05000324/LER-1992-010

LER 92-010-01:on 921207,penetration Leakage Exceeded TS Allowable Limit Due to actuator-to-valve Alignment & disk-to-seat Alignment Problems.Line Bored Valve to Achieve concentricity.W/930812 Ltr

ML20056D626
Person / Time
Site:	Brunswick
Issue date:	08/12/1993
From:	Jonathan Brown, Jones T CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-010, LER-92-10, NUDOCS 9308170204
Download: ML20056D626 (5)

LER-2092-010,

Event date:
Report date:
3242092010R00 - NRC Website

text

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Carolina Power a Light Company .

teemmeswomammaawnwa  ;

Brunswick Nuclear Plant f P. O. Box 10429  ;

Southport, N.C. 28461-0429- -j August 12, 1993  :

i FILE: B09-13510C 1 0CFR50.73 i I

SERIAL: BSEP-93-0129 i

U.S. Nuclear Regulatory Commission ATTN: Document Control Desk .:

Wasl.ington, D. C-. 20555 BRUNSWICK NUCLEAR' PLANT UNIT 2 f DOCKET'NO. 50-324 >

LICENSE NO. DRP-62 l' SUPPLEMENTAL LICENSEE EVENT REPORT 2-92-010-01 Gentlemen:

i In accordance with Title 10 of the Code of Federal Regulations, the enclosed l Supplemental Licensee Event Report is submitted. The original report fulfilled  :

the requirement for a written report within thirty (30) days of a reportable  !

occurrence and was submitted in accordance with the format set forth in NUREG-1022, Septenber 1983.

Very truly yours, )

'. M. Brown, Plant Manager - Unit 1 8runswick Nut: lear Plant TMJ/

Enclosure cc: Mr. S. D. Ebneter Mr. P. D. Milano BSEP NRC Resident Office i

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l 170037 93081702o4 930812

. DR ADOCK 05000324 /)

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1 NRC FORM 366 U.S. NUCLEAR REGULAVORY COMMISSION APPROt/ED OMB NO. 3150-0104

. (5/92f EXPIRES: 5/31/95 ESTIMATED BURDEN PER HISPONSE TO COMPty WITH THIS d INF0FtM ATION COLL.E CTlDN RE QUE ST: 60.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMutmS REGARoma BURDEN ESTum ,0 THE ,NFORuATiON AND FIECORDS MANAGEMENT BRANCH [MNBB 7714L U.S. NUCLEAR REGUL ATORY COMMISSION, W ASHINGTON, DC 205$6 0D01. AND TO THC F APERWORK REDUCTION PROJECT (3150 c104L OFFICE OF '

MANAGEMINT AND BUDOET, WASHINGTON, DC 20503.

F ACIUTV NAME (1) DOCEET NVMEiER (2) P AGE (3)

Brunswick Steam Electric Plant, Unit 2 '

05000324 1 of 4 TITLE (4l PENETRATION LEAKAGE IN EXCESS OF TECHNICAL SPECIFICATION ALLOWABLE LIMIT DURING LOCAL l LEAK RATE TEST 1hu EVENT DATE (5) LER NUMBER (6) REPORT D ATE (7) OTHER FACILITIES INVOLVED (8) l MONTH DAY VEAR YEAR MONTH DAY YEAR NUVDER NUM bER 05000 12 07 92 92 - 010 - 01 OB 12 93 FACW NAME DOCF ET NVVBER 05000 ,

OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR 8: (Check one or rnore of the followingM11) 4 MODE (9) 20.402(bl 20 405tcl 50.731aH2Hivl 73.71(b) )

20.405 taH1 % 50.36(cH1) X 50.73taH2Hvi 73.71 tc)

LEVEL (10) 000 20 405(aH11tM 50 36(cH2) 50.731aH2)fvii) OTHER

0 405taH1Hiid so 73(aH2Ha 50.73(aH2HviiiHA) (Specify in Abstract 1 20 405taH1Hrv) 50 73(aH2Hi0 50.73(aH2HvnilfB) and Tert) 20 405taH1Hv1 EO 73faH2Hsd j 50 73talt2Hz)  !

I LICENSEE CONTACT FOR THIS LER (12)

NAME TELEFHONE NUMBIR Theresa M. Jones, Regulatory Compliance Specialist (919) 457-2039 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

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CAUSE SYSTEM C OMPONI NT M A NUf ACTURER CAUSE SYSTEM C OMPONE NT M ANUF ACTURE R X SB ISV R344 Y SUPPLEMENTAL REPORT EXPECTED (14) M NTH DAY YtAR EXPECTED SUBMISSION E DATE (15) nf yes r,smoieve E ME CTf D SUBM!sDON CATE)

ABSTRACT (Umst to 1400 spaces, i e. appmximately fifteen single space typewntten knes) (10)

On December 7, 1992, the Unit 2 reactor was shutdown in day 230 of an outage. Local leak rate testing 'LLRT) of tha min steam isolation valves (MSIVs) had been performed. The results of testing indicated that MSIV leakage on main steam line iMSL) D was 54 standard cubic feet per hour (s ef h) , exceeding the technical specification limit of 11.5 scfh.

l MSLs A, B, and C tested satisfactory. It was determined that the Inboard MSIV leakage was caused by an a ct uator- t o- valve alignment problem. The actuator guide tube feet faces (which bolt to the bonnet; were found to have excessive run-out resulting in actuator-to-valve stem misalignment of 0.290 inches. Leakage of the outboard MSIV was caused by a disk-to-seat alignment problem. Inspections revealed that the seat bore and body bore were not concentric and that disk-to-guide clearances were excessive. Bonnet r egister-to-valve clearances were also found Lv be ej.cessi w. The inboard MOIV actuatcr guide tube feet faces were machined flat reducing actuator-to-valve stem misalignment to less than half (0.125) of the as-found readings, The outboard and inhoard MSIV were line-bored to obtain concentricity. On January 27, 1993, post-maintenance testing was performed twice (for repeatability) with satisf actory results (i.e., MSL D leakage measured 4.013 scfh in both tests). Troubleshooting ef f orts were applied in determining the cause of leakage since this was a repeat event for MSL D. The safety significance of this event is considered minimal. Previous similar events involving MSIVs not meeting the technical specification leakage requirements are re;)orted in LERs 1-B8-025 and 2-91-019.

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i NRC FORM 366A U. S. NUCLEAR REGULATORY COMMISSloN APPROVED OMB NO. 3150-0104 (5/92) EXPIRES: 5/31/95 E STIM ATED BURDE N PER RE SPONSE TO COMPLY WITH THIS j INFORMATION COLLECTION REQUEST: . 50.0 HRS. FORWARD UCENSEE EVENT REPORT (LER) COMMEuTS nEcARoiNo Bv4 DEN ESTIMATE To THE iNronMATiON ANo

. TEXT CONTINUATION RECORDS M AN AGEME NT BRANCH IMNBB 7714L U.S. NUCLE AR REGULATORY COMMISSION, W ASHINGTON. DC206500001. AND TO THE PAPERWORK REDUCTION PROJECT (3160-0104), OfTICE OF M AhAGEMf NT AND BUDGET. WASHINGTON, DC 205C3. )

l FACILfTY NAME {11 DOCKET NUMBER (2) LER NUMBER 16) PAGE (3)

Sf QUENTIAL REVISION B swick Steam Electric Plant """"" ""

05000324 2 of 4 92 - 010 - 1 TEXT I!f more space as requireri, use additionalNRC Form 366Nsl (17>

TITLE PENETRATION LEAKAGE IN EXCESS OF TECHNICAL SPECIFICATION ALLOWAELE LIMIT DURING LOCAL LEAK RATE TESTING SUPPLEMENTAL INFORMATION INITIAL OONDITIONS ,

On December 07, 1992, the Unit 2 reactor was shutdown in day 230 of the 1992 outage resulting from structural concerns with the emergency diesel generator building walls.

Local leak rate testing (LLRT) of the main steam isolation valves (MSIVs) had been i performed to satisfy technical specification primary containment leakage rate surveillance l requirements. Technical specifications require that the primary containment leakage rate be limited to less than or equal to 11.5 standard cubic feet per hour (scfh) for any one- i main steam line (MSL) isolation valve when tested at 25 psig.

The MSIVs are Rockwell Model 1612, 24", Y-type globe valves. l F

EVENT NARRATIVE The LLRT is accomplished by draining the MSL, closing the inboard and outboard MSIVs, pressurizing the space between them to 25 psig with air while providing a vent path for any leakage, and measuring the leakage rate. In the event of a leakage rate equal to or greater than I1.5 scfh, the test procedure directs that the associated MSL be filled with ,

water between the reactor vessel and the inboard MSIV to seal any leakage past the inboard l MSIV. The space between the MSiVs is again air pressurized and the leakage rate resulting '

from leakage past the outboard MSIV is measured. This method of testing indicates the total MSL leakage, the leakage associated with the outhcard MSIV and, by substraction, the leakage associated with the inboard MSIV.

On November 29, 1992, leakage testing of MSL D commenced. Testing was initially performed by closing the inboard and outboard MSIVs and attempting to pressurize the space between them; however, test pressure was not obtainable. Subsequent testing, af ter stroking the outboard MSIV, was accomplished by raising reactor vessel level and filling the MSL unti]

a static water head of 25 psig was reached at the inboard MSIV. Leakage past the outboard MSIV was found to be 105 scfh. A retest was performed, af ter draining the water from MSL D and a leakage rate of 159 scfh was obtained. This indicates an inboard MSIV leakage rate Of M ccfh. Therefore. leakege through W L D inhnard and outboard MSIVs was limited to 54 scfh. The inboard and outboard MSIVs were repaired after determining the cause of 1 % r.a ge and, on January 27, 1993, post-maintenance testing was performed twice (for repeatability) with satisfactory results (i .e. , MSL D leakage measured 4.013 scfh in both tests). Troubleshooting efforts were applied in determining the cause of leakage since this was a repeat event for MSL D.

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NRC EORM 366A U. S. NUCLEAR REGULATORY COMMISSION APPROVED OMB NO. 3150-0104

, (5/92) EXPIRES: 5/31/95 ISTIM ATED BifRDE N PER RE SPONSE TO COMPLY wtTH THl5 (NFORMATION COLLECTION RE QUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COuurNTs REcARn No euRoEN tsTiu*TE TO TwE iNrORu ATiON ANo

. TEXT CONTINUATION RECORos uANacEusNT eRANCH ruses 7714), u.s. NoCLE AR REGOLATORY COMMISS80N. WASHINGTON, DC20555-000), AND TO THE FAPERWORK REDUCTION PROJECT (3150-C104L OFFICE OF MANAGEMENT AND SUDGET, WASHINGTON, DC 20503.

1 FACluTY NAME (1) DOCKET NUMBER (2) tER NUME.ER 16) PAGE (3)  !

tiEQUENTIAL REVISION 4 Brunswick Steam Electric Plant "" "#"'"

05000324 3 of-4 Unit 2 92 -- 010 - 1

• TEXT Uf mwe space is required, use additwroalNRC form 366A'st (17) i CAUSE OF EVENT J

MSL D INBOARD MSIV Leakage of the inboard MSIV was caused by an actuator-to-valve alignment problem. This ,

was confirmed during initial cause and ef fect troubleshooting when a satisf actory leakage >

rate was obtained utilizing temporary rigging to correct the misalignment. Thrust measurements were taken and found to be satisfactory with and without actuator to valve misalignment. The actuator guide tube feet faces (which bolt to the bonnet) were found to have run-out readings as much as 0.060 inches on one foot and 0.040 inches on another, {

resulting in a actuater-to-valve stem misalignment of 0.290 inches. Further evidencing j an actuator-to-valve alignment problem were a scored, or galled, valve stem and bonnet l spacer ring. Possibly contributing to the leakage was a disk-to-seat alignment problem i caused by the body bore, seat joint, and seat bore not being concentric. i l

MSL D OUTBOARD MSIV Leakage of the outboard MSIV was caused by a disk-to-seat alignment prcblems. Inspections revealed that the seat bore and body bore were not concentric and that disk-to-guide l clearances were excessive. Bonnet register-to-valve bore clearances were also found to j be excessive. d i

CORRECTIVE ACTICNS MSL D INBOARD MSIV The irboard MSIV actuator guide tube feet Iaces were machined flat reducing actuator-to-valve stem misalignment to less than half (0.125 inches) of the as-found readings.

Additionally, the valve was line bored to achieve concentricity. This was done as a conservative measure even thour.1 the vendor field service representative did not feel valve bore run-out was excessive . Because the valve bore was slightly oversized after line boring, a new oversized disk was required to be installed to maintain proper disk-to-bore and disk-to-guide rib clearances. A new stem and spacer ring was installed.

MSL D OUTBOARD MSIV The outboard MSIV was line-bored to obtain concentricity between the seat bore and body bcre. The disk was oversized to achieve manufacturer recommended clearances between the disk outside diameter and body guides, and weld build-up to the bonnet register was performed to maintain proper clearancca with the new valve bore dimensions.

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l' NRC EORM 366A (3/92)

U. S. NUCLEAR REGULATORY COMMISSION APPROVED OMB No. 3150 0104 ,

l ,, EXPIRES: 5/31/95 i

} ES*lMATED BURDEN PER RESPONSE TO COMPLY WIT H THIS  ;

I INFD*1M ATION COLLECTION REQUESTS SD.0 HRS. FORWARD i

LICENSEE EVENT REPORT (LER) CoMMtN1S nloARa,No ,UnoEN Es1,M ATE To rst ,N,onM ATioN ANo >

. TEXT CONTINUATION RECORDS MANAGEMENT BRANCH (MNBB 7714L U.S. NUCLEAR REGULATORY COMMIS$10N. WASHINGTON, DC 20555-D001, AND 10 T HE PAPERWORK Rf DUCTION PROJE CT 0150-01041, OFFICE OF MANAGEMENT AND BUDGET, WASHINGTON. DC 205C3.

FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (6) PAGE (3) f l

SE QUENTIAL REVISION i l

Brunswick Steam Electric Plant "*"'" * * *'"

Unit 2 05000324 4 of 4 1 92 - 010 - 1 j

TEXT (If rnore space is required, use additional n/RC Forrn 36&A 's! l17) -

i BUTY ASSESSMENT  ;

Primary containment integrity ensures that the release of radioactive materials from the  ;

containment atmosphere will be restricted to those leakage paths and associated leak rates  !

assumed in the accident analyses. This restriction, in conjunction with the leakage rate l limitations, will limit the site boundary radiation doses to within the litnits of 10CFR100 during accident conditions. Technical specifications require that primary containment  !

leakage rates be limited to less than or equal to 11.5 scfh for any one MSL isolation valve I

tested at 25 psig. '

The safety significance of this event is considered minimal. Although the technical ,

specification limits were exceeded, a MSL leakage rate of 54 scfh would have not resulted t in the 10CFR100 off-site dose limits being exceeded during a design basis accident. This  ;

assessment is based upon dose calculations recently completed for Brunswick Units 1 and  !

2 by General Electric (GE letter OG-91-1063-09, dated Decenter 17, 1991) , using the Boiling t Water Reactor Owners Group (Bh70G) methodology at 100 scfh per steam line.

PREVIOUS SIMILAR EVENTS Past f ailures of MSIVs to meet the TS required 11.5 scfh leakage requirements were reported in LERs 1-88-025 and 2-91-019. LER 1-88-025 involved the outboard MSIV on two separate  !

l MSLs. On November 12, 1991, LER 2-91-019 reported the failure of both MSLs C and D. The }

l associated MSIVs were repaired and the unit was. returned to service in January,1992 until l April 21, 1992, when it was shutdown for an outage.

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ETIS COMPONENT IDENTIFICATION l

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l System / Component ETIS Cod 2 l

l MSIv Ss/ISV l I

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