ML20012F505

LER 89-027-00:on 891005,determined That Leakage Rate for Containment Isolation Check Valve in Excess of Leakage Rate Allowed Per Tech Specs.Caused by Loose Weld Slag in Valve Seat Area.Valve Cleaned & reassembled.W/900405 Ltr

ML20012F505
Person / Time
Site:	Arkansas Nuclear
Issue date:	04/05/1990
From:	Ewing E, Millar D ARKANSAS POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
2CAN049008, 2CAN49008, LER-89-027-01, LER-89-27-1, NUDOCS 9004160035
Download: ML20012F505 (4)

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[ April 5,1990 2CAN049008 U. S. Nuclear Regulatory Commission

! Document Control Desk l.

Mail Station P1-137 Washington, D. C. 20555

SUBJECT:

Arkansas Nuclear One - Unit 2 L Docket No. 50-368 License No. NPF-6 Licensee Event Report No. 50-368/89-027-00 Gentlemen:

In accordance with 10CFR50.73(a)(2)(1)(B), attached is the subject report concerning loose weld slag found in a containment isolation check valve which rendered the valve inoperable.

Very truly yours, E. C. Ewing General Manager, Technical Support and Assessment ECE/DM/sgw .

Attachment l cc: . Regional Administrator Region IV i U. S. Nuclear Regulatory Commission l 611 Ryan Plaza Drive, Suite 1000  ;

Arlington, TX 76011  ;

INPO Records Center  !

Suite 1500 1100 Circle 75 Parkway i Atlanta, GA 30339-3064 9004160030 891003 q f/Z*

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• Fsrm 1062.01A NRC Fors 366 U.S. Nuclear Regulatory Commission ,

(9 83) Approved OMB No. 3150-0104  ;

Expirest 4/30/92 L1CEN5EE EV[NT REPOR1 (L E R) ,

IACILITY NAME (1) Arkansas Nuclear One, Unit Two IDOCKET NUSER (2) IPAGE (3) 10151010101 31 61 Bl110Fl013  !

TITLE (4) Loose Weld Slag Found in Containment Isolatio.e Check Valve Rendering the Valve Inoperable

! EVEN" DATE (b) JR Ml55IR (6) REPORT DATE (7) OTHER F ACILITIES INVOLVED (8)

I lioquentiell I Revision i 11 Month Day Year' Year Number Number Monthi Day Year Facility Names Docket Nunber(s) ,

t 0 5 0 0 0 >

11 0 01 f, 8 9 81 9P! 01 21 71 -I 01 0 01 41 01 51 91 0 0 5 0 0 0 OPERA"ING TH'5 REPORT 15 5UBMITTED PURSUANT 'O THE REQUIREMENTS OF 10 CFR 5:  ;

WDE (9) 15 (Check one or more of the followinn) (11)  !

POWERl l_ 20.402(b) l_ 20.405(c) 4 1_ I b0.73(a)(2)(iv) l_l 73.71(b) '

LEVEll  : _I 20.405(a)(1)(1) _l 50.73(a)(2)(v) l_l 73.71(c)

(10) 101010 20.405(a)(1)(ii) l_ll' 50.36(c)(1)50.36(c)(2) _

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_ 20.405(a)(1)(iii) l3150.73(a)(2)(1) _ 50.73(a)(2)(viii)(A) 1 , Abstract below and in Text, NRC Fom i

_ 20.405(a)(1)(iv) _l50.73(a)(2)(11) l

_I 50.73(a)(2)(viii)(B)l l

l 20.405(a)(1)(v) 1 50.73 a)(2?(iii) '

50.73(a)(2)(x) I' 366A)

LICEN5EE CON"ACT FOR THIS LER (12) .

Nast l Telephone Number l Area l Dana Millar, Nuclear Safety and Licensing Specialist Code 1 510111916141-13111010 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (L3) 11 l l l Reportable, 1 I l l lReportablel Cause System Component knuf acturerl to NPRDS lCause Svstes conoonent Manufacturer to NPRDS l 1 l l l l l l l I i 1 i l I i 1 1

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I I l 1 l 1 l l I I l i SUPPLEMEN" REl' ORT EXPECTI D (A4) l EXPECTED. Month Day Year l $UBMISSION l l~l Yes (If yes, conolete Expected Submission Date) til No l DATE (15) I l' 1 I I AB5 TRACT (Limit to 1400 spaces, i.e., approximately fifteen single-space typewritten lines) (L6)

On October 5,1989, while performing local leak rate testing (LLRT) during a refueling outage (2R7), it was determined that the leakage rate for a containment isolation check valve (25A-69) in the Service '

Air (SA) System was in excess of the leakage rate allowed by Technical Specifications. The valve was declared inoperable and a maintenance job order issued to repair or replace the valve as necessary.

The maximum pathway leakage rate is considered when calculating total leakage for Type '8' and Type 'C' valves. The cause of the excessive leakage through 25A-69 was detemined to be loose weld slag found in the valve seat area. It could not be determined how the weld slag was introduced into the valve, although it could have been introduced during a maintenance outage in May 1989. The valve was reassembled and a leakage rate test satisfactorily performed. The penetration is isolated by two valves 25A-68, a normally closed, locked closed, manual isolation valve and 25A-69. The result of the last LLRT performed on 25A 68 indicated zero leakage. With no leakage through 25A-68, it is reasonable to assume that the actual leakage through the penetration would be minimal and, therefore, there was no safety concern.

On March 6,1990, after a detailed reevaluation of this condition and discussions with engineering personnel, this condition was determined to be reportable pursuant to 10CFR50.73(a)(2)(1)(B).

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i Form 1062.01B U.S. Nuclear Regulatory Commission NRC Form 3666 (9 83) Approved OMB No. 3150-0104 Expires 4/30/92 LICENS(( (V[NT REPORT (L(R) T[XT CONTINUATION  ;

FAtlLITY NAME (1) lDOCkLT NUMBLR (2) l ,[R NLMBIR (6) l PAGE (3) l l l lSequentiall l Revision)

Arkansas Nuclear One, Unit Two I l Year Number Number l m 1015101010I 31 6l BF 81 9 --

01 21 7 --

01 Ol01210Fl013 i I

T[XT (If more space is required, use additional NRC Form 366A's) (17)

A. Plant Status

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At the time of occurrence of this event, Arkansas Nuclear One, Unit Two (ANO 2) was in Mode 5 (Cold Shutdown). Reactor Coolant System (RCS) (AB) pressure was at atmospheric and RC$ temperature at approximately 120 degrees Fahrenheit. The seventh refueling outage (2R7) was in progress (September 25, 1989 to November 20,1989). i B. [ vent Description On October 5,1989, while performing the Local Leak Rate Test (LLRT) on a Service Air (SA) system 7 containment isolation check valve (25A 69), it was identified that 2SA 69 leaked excessively.

Additional testing was performed on 25A-69 to better quantify the leakage rate. The results of the additional testing indicated that the leakage rate was approximately 21264 absolute cubic centimeters per minute (aces).

Technical Specification 3.6.1.2.b requires that the combined leakage rate for penetrations and valves subject to Type 'B' and Type 'C' tests be less than or equal to 0.6 L t s ANO 2). As found leakage tests for Type 'B' and Type 'C' are calculated ustRg(20990 acem the maximum for pathway  !

method. The maximum pathway leakage rate for the penetration (2P43), of which 25A 69 isolates one side, exceeded the total-leakage rate allowed by Technical Specifications. The containment isolation valve (25A-69) was declared inoperable and a maintenance job order issued to open, clean, inspect, repair or replace the valve, as necessary.

C. Root Cause The cause of the excessive leakage rate for 25A 69 was a large quantity of loose weld slag which was found in the valve. It cannot conclusively be determined how the weld slag was introetced into the valve, although it could have been introduced during a maintenance outage in May 1989.  ;

(This is the only time the SA penetration, a normally closed penetration, had been opened since the previous refueling outage (2RG) at which time a satisfactory LLRT was performed.) A probable cause of the introduction of the weld slag into the SA system was inadequate cleanliness controls associated with the maintenance activities on the SA system. /

D. Corrective Actions The valve was cleaned and reassembled, and an as-lef t LLRT was performed resulting in a satisf actory leakage rate of approximately 54.8 accm. The condition found did not indicate any generic problems with the particular valve type, site or application. A review of previous maintenance activities associated with the valve was performed with no indication of how the weld slag could have been introduced into the valve. Since it could not be concluded how the weld slag was introduced into the $A system ANO management decided that until the next refueling outage (2R8), an LLRT would be performed on 25A 69 prior to plant heat-up if the valve is opened to supply SA to containment During 2R8 scheduled to begin in March 1991, a flush of the SA system will be performed to clean out foreign material.

Additionally, a review of the cleanliness controls associated with maintenance activities performed on the SA system has indicated that the controls were inadequate. An administrative procedure which addresses cleanliness control for maintenance activities on fluid systems will be revised to include maintenance activities on air systems. This procedure will be revised by March 1991 consistent with the 2R8 outage schedule.

E. Safety Significance The only SA penetration into conta'inment has two valves, 25A 69 and 25A-68. Both valves are leak tested. 25A 68 is a manual, normally locked closed (with strict administrative controls) containment isolation valve. The leakage rate on 25A-68 was verified to be acceptable, with zero leakage.

The use of the maximum pathway leakage rate is a conservative method of calculating leakaDe through individual penetrations. $1nce there was zero leakage measured through 2$A-68, it is reasonable to assume that the actual leakage through the penetration would be minimal. Although it may appe6r that the actual leak rate f rom containment exceeded the allowable Technical Specification values due to the method of calculation (i.e., maximum pathway), Technical Specification allowable values were not exceeded. Therefore, ANO has concluded that this condition is not safety significant.

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. Fsrs 1062.01B 1 NRC Form 366A U.S. Nuclear Regulatory Commission (9-83) Approved DMB No. 3150-0104 Empires: 4/30/92 LICENSEE EVENT REPOR1 (LER) TEXT CDNTINUATION FACILITY NAME (1) (DOCKET NUMBER (2) l .[R NUMBER (6) l PAGE (3) l l l i sequentiell lRevisioni  ;

~ Arkansas Nuclear One, Unit Two l l Yeer. Number Number l 10151010101 31 61 BI FT 9 --

01 ri 7 --

oi ol01310r101~3 TEXT (If more space is required, use additional NRC Form 366A's) (17)

F. Basis for Reportability  ;

based upon a detailed investigation of the use of the SA system, it was identified that 5A was aligned to the containment building during a maintenance outage in May 1989, which required the opening of 25A-69. The weld slag could have beer introduced into the valve at this time, rendering the valve inoperable. Technical Specification 3.0.4 requires that the Limiting Conditions for Operation be satisfied prior to changing operational modes. Since the maintenance outage in May 1989 operational mode changes have occurred without knowing 25A 69 was inoperable. This condition was, therefore, determined to be reportable pursuant to 10CFR50.73(a)(2)(1)(B), operation prohibited by Technical Specifications on March 6, 1990.

G. Additional Information There have been no previously reported events in which foreign material rendered a containment isolation valve inoperable.

' Energy Industry Identification System (E!!$) codes are identified in the text as (XX).

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