ML18012A231

LER 96-006-00:on 960325,separation of Valve Stem & Disk for Main FW Isolation Valve Caused Valve to Become Inoperable. Caused by Mfg Defect.Valve Stem Machined & Procedures revised.W/960424 Ltr

ML18012A231
Person / Time
Site:	Harris
Issue date:	04/24/1996
From:	Rowell L CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-96-069, HNP-96-69, LER-96-006, LER-96-6, NUDOCS 9604300382
Download: ML18012A231 (7)

Text

CATEGORY 1 REGULATO INFORMATION DISTRIBUTION YSTEM (RIDS).ACCESSION NBR:9604300382 DOC.DATE: 96/04/24 NOTARIZED:

NO FACIL:50-400 Shearon Harris Nuclear Power.Plant, Unit 1, Carolina AUTH.NAME AUTHOR AFFILIATION ROWELLiL ST Carolina Power 6 Light Co.RECIP.NAME RECIPIENT AFFILIATION I DOCKET 05000400

SUBJECT:

LER 96-006-00:on 960325,separation of valve stem&di k f main F i W isolatIon valve caused valve to become inoperable.

g Is or Caused by mfg defect.Valve stem machined 6 procedures revised.W/960424 ltr.DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE: TITLE: 50.73/50.9 Licensee Event Report (LER)', Incident Rpt, etc.NOTES:Application for permit renewal filed.E 05000400 G RECIPIENT ID CODE/NAME PD2-1 PD INTERNAL: ACRS AEOD/SPD/RRAB NRR/DE/ECGB NRR/DE/EMEB NRR/DRCH/HICB NRR/DRCH/HQMB NRR/DSSA/SPLB RES/DSIR/EIB EXTERNAL: L ST LOBBY WARD NOAC MURPHY,G.A NRC PDR COPIES LTTR ENCL 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 1 RECIPIENT ID CODEtNAME LEgN/NP B FILE CENTER B NRR/DRCH/HHFB NRR/DRCH/HOLB NRR/DRPM/PECB NRR/DSSA/SRXB

'GN2 FILE 01 LITCO BRYCE,J H NOAC POORE,W.NUDOCS FULL TXT COPIES LTTR ENCL 1 1 2 2 1 1 1 1 1 1 1 1 1 1 1~1 1 1 2 2 1 1 1 1 p E N NOTE TO ALL"RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE!CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN 5D-5(EXT.415-2083)TO ELIMINATE YOUR NAME FROM DISTRIBUTION IISTS FOR DOCUMENTS YOU DON'T NEED!FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 26 ENCL 26 Carolina Power 8 Light Company Harris Nuclear Plant PO Box 165 New Hill NC 27562 U.S.Nuclear Regulatory Commission ATTN: NRC Document Control Desk ApR p 4 1996 Washington, DC 20555 Serial: HNP-96-069

'0CFR50.73 SHEARON, HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO.50-400 LICENSE NO.NPF-63 LICENSEE EVENT REPORT 96-006-00 Gentlemen:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted.

This report describes a separation of the valve stem and disk for a Main Feedwater Isolation Valve, resulting in a violation of the Technical Specification Action Statement.

Sincerely, LSR/lsr Enclosure~()Q$$3 c: Mr.J.B.Brady (NRC-HNP)Mr.S.D.Ebneter (NRC-RII)Mr.N.B.Le (NRC-PM/NRR)9tc04300382 960424 PDR ADOCK 05000400 S PDR State Road 113k New Hill NC NRC FORM 366 g.95)U.S.NUCLEAR REGULATORY COMMISSION LICENSEE EVENT REPORT (LER)(See reverse for required number of digits/characters for each block)APPROVED BY OMB No.3150-0104 EXPIRES 04/30/99 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THI MANDATORY INFORMATION COllECTION REOUESII SM HRS.REPORTED lESSONS LEARNED ARE INCORPORATED INTO THE UCENSUIG PROCESS AND FEO BACK TO INDUS'FRY.

FORWARD COMMENTS REGAROUIG BURDEN ESTIMATE TO THE INFORMATION AHO RECORDS MANAGEMENT BRANCH IT4(F33L US.HUClEAR REGU(ATORT COMMISSIOIL WASHU(GTON, OC 205550001, AND TO THE PAPERWORK REDUCTION PROJECT (3150.010(L Off ICE Of MANAGEMEN'r AND BUDGET, VIASHUIGTOH, OC 20503.FACIUTY NAME ll I Harris Nuclear Plant-Unit 1 DOCKET NUMBER (2I 50-400 PAGE (3)1 OF 3 TITLE (4I A separation of the valve stem and disk for Main Feedwater Isolation Valve 1FW-277 caused the valve to become inoperable and resulted in a violation of the Technical Specification Action Statement.

MONTH DAY YEAR EVENT DATE (5)LER NUMBER (6)SEQUENTIAL REVISION NUMBER NUMBER REPORT DATE (7)MONTH DAY YEAR FACILITY NAME OTHER FACILITIES INVOLVED (B)DOCKET NUMBER 05000 25 96 96-006-00 4 24 96 FACILITY NAME DOCKET NUMBER'5000 OPERATING MODE (9)POWER LEVEL (10)THIS REPORT IS SUBMITTED PUR 20.2201(b) 20.2203(a)(1) 20.2203(a)

(2)(i)20.2203(a)(2)(ii)20.2203(a)(2)(iii)20.2203(a)(2)(iv)

SUANT To THE REQUIREMENTS OF 10 CFR B: (Check one o 20.2203(a)

(2)(v)50.73(a)(2)(i) 50.73(a)(2)(ii)50.73(a)(2)(iii) 50.73(a)(2)(iv)20.2203(a)(3)(i)20.2203(a)(3)(ii) 20.2203(a)

(4)50.73(a)(2)(v) 50.73(a)(2)(vii) 50.36(c)(1) 50.36(c)(2) r moro)(11)50.73(a)(2)(viii)50.73(a)(2)(x) 73.71 OTHER Specify in Abstract below or in NRC Form 3BOA LICENSEE CONTACT FOR THIS LER Lewis S.Rowell-Project Engineer-Licensing/Regulatory Programs MPONENT FAILURE DES (12)TELEPHONE NUMBER Urcluda Area Coda((919)362-2287 CRIBED IN THIS REPORT (13)CAUSE SYSTEM COMPO NEIIT MANUFACTURER REPORTABI.E TD NPRDS CAUSE SYSTEM COL'IPONENT

'ANUFACTURER REPORTABLE TO NPRDS SJ ISV B350 SUPPLEMENTAL REPORT EXPECTED (14)YES (If yes, complete EXPECTED SUBMISSION DATE).No EXPECTED SUBMISSION DATE (15)MONTH DAY YEAR ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines)(16)On February 15, 1996 the plant was operating in Mode 1 with Reactor Power at 100%.At approximately 0108, the Metal Impact Monitoring System (MIMS)for the"B" Steam Generator pegged high and simultaneously a momentary low nitrogen pressure alarm was received on the"B" Steam Generator Feedwater Isolation Valve (1FW-277).

On March 4, 1996, 1FW-277 was partially stroked in accordance with surveillance test OST-1018.On March 15, 1996 a flow reduction trend was noted on Main Feedwater preheat line flow data.At this point, plant personnel recognized that the February 15, 1996 MIMS and low nitrogen pressure alarms may have been related to this trend in flow reduction.

Troubleshooting was commenced.

On March 22, 1996 the plant was shutdown and valve 1FW-277 was stroked with personnel near the valve to monitor for characteristic sounds of valve stroking.Because the characteristic sounds were not heard, the valve was disassembled on March 25, 1996.Inspection revealed that the valve stem had fractured.

Metallurgical analysis revealed that the crack had originated at the small diameter of the backseat to stem transition and propogated by low cycle fatigue until the load was sufficient to cause the final tension overload failure.Further investigation showed that a manufacturing defect had allowed the disk of this gate valve to be pulled hard into the bonnet when opened, causing unacceptable bending loads in the stem.The replacement valve stem was machined ch that the valve disk will not contact the bonnet when the valve.opens.Additionally, the Corrective intenance procedure for Main Feedwater Isolation Valves will be revised to include a verification prior final reassembly that the disk is not striking/contacting the bonnet.

~y~

NRC FORM 366A I4.951 LICENSEE EVENT REPORT (LERj TEXT CONTINUATION U.S.NUCLEAR REGULATORY COMMISSION FACILITY NAME 01 Harris Nuclear Plant~Unit 1 TEXT Pl eOrO OpOOO a rdrrOOOd, raO OddirrOOOl OOpOO Ol NR FOre (ITI OOCKET 50 400 LER NUMBER (6)YEAR SEOUENTIAL REVISION NUMBER NUIABER 96-006-00 PAGE IBI 2 OF 3 EVENT DESCRIPTION:

On February 15, 1996, the plant was operating in Mode 1 with Reactor Power at 100%.At approximately 0108, the Metal Impact Monitoring System (MIMS)for the"8" Steam Generator (SG)pegged high and simultaneously a momentary low nitrogen pressure alarm was received on the"8" SG Feedwater Isolation Valve (1FW-277)(EIIS Code SJ-ISV, Borg Warner, Valve Assembly-Gate, 16", 0 4350885-001, Hydraulic Operator).

On March 4, 1996 the valve was partially stroked in accordance with surveillance procedure OST-1018.On March 15, 1996, a flow reduction trend was noted on the"8" Main Feedwater line by the system engineer.At this point, plant personnel recognized that the February 15, 1996 MIMS and low nitrogen pressure alarms may have been related to this trend in flow reduction.

Troubleshooting was commenced.

This trend was confirmed by an increase in Preheater Bypass flow, increased opening of the Feedwater Regulating valve, and increased pressure upstream of the 1FW-277 valve.On March 22, 1996 the plant was shutdown for load sequencer inoperability (LER 96-002-003), and this valve was stroked with personnel near the valve listening for the characteristic loading up of the actuator hydraulic pump and the sound of the disk unwedging.

The absence of these characteristic sounds provided confirmation that the valve was not functioning properly.n March 25, 1996 the 1FW-277 valve was disassembled and inspection revealed that the valve stem had ractured.Metallurgical analysis revealed that the crack had originated at the small diameter of the backseat to stem transition and propogated by low cycle fatigue until the load was sufficient to cause the final tension overload failure.Unusual marking in the domed area inside the valve bonnet was noted.This indicated that the disk had firmly contacted the bonnet when opened instead of the stem backseat contacting the bonnet backseat area.This was confirmed by assembling the disk, a replacement stem, and the bonnet and noting the interference between the disk and bonnet.The bonnet backseat area inset was deeper than would be expected for the valve assembly.Based upon discussions with the valve manufacturer, it was determined that the actual inset of 0.406" was 0.181" deeper than the nominal inset dimension of 0.225".Thus, the stem had to travel approximately 3/16" farther than"nominal" to reach the backseat.The valve is designed to backseat when open.Valve 1FW-277 has an orientation of approximately 50 degrees from vertical which causes the disk to rest in the bottom of the lower guide body while stroking.This results in an offset between the disk centerline and the stem centerline of approximately I/4" and caused a bending moment on the lower stem, consistent with the crack initiation location.The loads, bending moment, and number of cycles on the valve stem are consistent with the low cycle fatigue symptoms observed on the failed stem.An engineering review determined that the valve would be incapable of performing its containment isolation function with the fractured stem.This condition had potentially existed since February 15, 1996.Therefore, the Technical Specification 3/4.6.3 Action Statement, requiring an inoperable valve to be restored to OPERABLE status or the affected penetration isolated within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, or that the plant be placed in COLD SHUTDOWN within the following 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />, was violated.This event is being reported per 10CFR50.73(a)(2)(i).

CAUSE: he root cause of the stem failure was a manufacturing defect which allowed the disk of this gate valve to be lied hard into the bonnet when opened, binding the disk to the stem and placing unanticipated bending loads in the stem.The effect of this failure was the inability of this valve to fully close and perform its design basis containment isolation function.

l$NRC FORM 366A l495)LICENSEE EVENT REPORT (LER)TEXT CONTINUATION US.NUCLEAR REGULATORY COMMISSION FACILITY NAME (I)Harris Nuclear Plant~Unit 1 DOCKET 50400 LER NUMBER I6)YEAR SEQUENTIAL NUMBER NUMBER 96-006-0.00 PAGE)3)3 OF 3 TEXT Pr pisrs sptss it ssFiisaf, siss stcam)itasl snpis pr)VRC Falsi 366iU I)T)SAFETY SIGNIFICANCE:

An engineering review was conducted of analyzed accident scenarios, assuming that the 1FW-277 valve would not be capable of performing its design basis containment isolation function.The review concluded that failure of the valve to close fully would not alter the conclusions of those accident scenarios since operator action or the system control logic would require closure of the Main Feedwater Regulating valve for feedwater and containment isolation.

Also, the safety-related check valve upstream of 1FW-277 would act as backup for containment isolation for those scenarios in which containment pressure is greater than upstream pressure.Additionally, based upon the probabilistic safety assessment, the unavailability of one or more main feedwater isolation valves is not safety-significant.

The potential consequences of valve failure are overfilling the affected steam generator and subsequent overcooling of the RCS.This could be mitigated by plant design features (tripping of the main feedwater pumps or automatic closure of the Feedwater regulating valves), or by operator intervention to control the main feedwater system.This event is judged to be nonsafety-significant.

The only other valves of this type at this facility are the other two MFIVs.Unlike 1FW-277, both of these valves have been disassembled since plant startup (1FW-159 in late 1994 and 1FW-217 in late 1995)with the bonnets inspected.

Personnel involved with the disassembly of those valves did not observe any contact marks as seen on 1FW-277.P PREVIOUS SIIVIILAR EVENTS: ere have been no similar events of component failures of Main Feedwater Isolation valves at the Harris Nuclear ant.CORRECTIVE ACTIONS COMPLETED:

The valve stem was replaced and the replacement stem"T Head" was machined such that the valve disk will not contact the bonnet when the valve is opened.CORRECTIVE ACTIONS PLANNED: The Corrective Maintenance procedure (CM-M0204) which provides instructions for disassembly, maintenance, inspection, and reassembly of the Main Feedwater Isolation Valves will be revised by August 31, 1996 to include verification prior to final valve reassembly that a valve disk does not strike/contact the bonnet.EIIS CODES: Main Feedwater Isolation Valve (MFIV)(SJ-ISV).