ML20042E684

LER 90-005-00:on 900329,safety Relief Valves Experienced Setpoint Drift in Excess of Tolerance.Caused by corrosion- Induced Bonding of Surface Between Pilot Valve Disc & Seat. Valves refurbished.W/900424 Ltr

ML20042E684
Person / Time
Site:	Hatch
Issue date:	04/27/1990
From:	Hairston W, Tipps S GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-1058, LER-90-005-01, LER-90-5-1, NUDOCS 9004300267
Download: ML20042E684 (7)

Text

i Georga Power Company :

33J Pemont Annue t'

w 1< - Manta. Georg.a 30308 '

.. TeWohons 404 526 3195 -

Maihng Arkirow A0 ineness Center Pad way Post 0%ce Boy 1295 Birmengham. Alat;ama 35201 Tdephono 205 868 5581

' tre wom m t Ar!T s<:+m W. G. liairston, til Senor Vce Presicient -

Nuciear Opera',om HL-1058 000386 April 24, 1990 U.S. Nuclear Regulatory Commission ATTN: ' Document Control Desk Washington, D.C. 20555 t

PLANT HATCH - UNIT 1 NRC DOCKETS 50-321 OPERATING LICENSE DPR-57 LICENSEE EVENT REPORT SAFETY RELIEF VALVES EXPERIENCE'SETP0 INT QRIFT DVE TO CORROSION INDUCED BONDING Gentlemen:

Georgia Power Company is submitting the-enclosed -voluntary Licensee Event Report (LER) due to the potential . industry interest in the event.

.This event occurred at Plant Hatch - Unit 1.

Sincerely, gM.)/k -W W. G. Hairston, III-SWR /ct

Enclosure:

LER 50-321/1990-005

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IGeorgiaPower d L U.S. Nuclear. Regulatory Commission-l: April 24, 1990'-

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L c: Georaia Power Comoany, Mr. H. C. Nix, General Manager - Nuclear. Plant-Mr. J. D. Heidt, Manager Engineering and Licensing.- Hatch G0-NORMS

'U.S. Nucle'ar Reaulatory Commission. Washinaton.'U.C.

Mr. L. P. Crocker,1 Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaion II Mr. S. D. Ebneter,'. Regional Administrator _

Senior Resident Inspector _- Hatch -

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Fem att U.S. NUCLEM LEQULATORY COMMIS$10N APPR:VED OMB C 31540104 LICENSEE EVENT REPORT (LER) '" P'a' 5 ' '8'

F ACILITY hAME til DOCKET NUMGE R Ul PAGE i3' PLANT HATCH, UNil 1 o l5 l0 lo l0l3 12 l1 1 lorl 0 l5 flTLE tes l SAFETY RELIEF VALVES EXPERIENCE SETPOINT DRIFT DUE TO CORROSION INDUCED BONDING EVENT DATE 181 LlR NUMBER 46) REPORT DATE 171 OTHER F ACILITIES INVOLVED ist

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• DOCKET NUMetRis MONT H DAY YEAR YE All -

g (' ' y,[ MONTH DAY YEAR F ACILIT V N AMtt Ol510I0101 I I 0 l3 2l9 9 0 9l0 0l0l5 0 l0 0l4 2l7 9l0 o isio io i oi i j OPE R ATING TH18 REPORT IS SUBMiff ED PUR$UANT TO THE REQUIREMENTS OF 10 CFR $ ICheck one e, sno,e e,,he ,onowmg/ (11) 5 n .0 m n .Ost.> .0nuuH > nmi POvyt R 20 406teH1Hil 50 38I.1til 60.73teH2Hvl 73.71tel

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NAME TELEPHONE NUMBER ARE A CODE i j

Steven B. Tipps, Manager Nuclear Safety and Compliance, Hatch 9 i l2 3,6 7 i i 78 1 51 i i i i COMPLETE ONE LINE FOR E ACH COMPONENT F AILURE DESCRIBE 0 IN THis REPORf (131 CAUSE SYSTEM COMPONENT "'" O A [

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On 3/29/90 at approximately 0800 EST, Unit 1 was in the Refuel mode at an approximate j power level of 0 MWt (approximately 0% of rated thermal power). At that time plant engineering personnel received written notification of the results of off-site _

testing of pressure vessel safety relief valves (SRVs, Ells Code RV). Of the eleven SRVs, six exhibited drif t in the mechanical lift setpoints in excess of the + 3% i tolerance specified by in-service testing (IST) requirements. This voluntary report I is being submitted due to the potential industry interest in this event in view of l the ongoing efforts of the Boiling Water Reactor Owners' Group (BWR0G) to reduce  !

setpoint drift. The experienced setpoint drift was well within the analytical limits "

existing for reactor vessel over-pressure protection, i The root cause of the event is corrosion-induced bonding of the surface between the pilot valve disc and seat. The experienced setpoint drift in this event is consistent with current industry data demonstrating that both PH13-8Mo discs and stellite discs can occasionally form corrosion bonds with the stellite seat resulting ,

in setpoint drift.

Corrective actions for this event include refurbishing the valves and continuing to  ;

participate in the BWROG effort to develop a new corrective action plan to resolve the SRV setpoint drif t issue.

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saceptv asaast tis DOChlt NUMD4 R (2) Lf R WUMDlh iSI P&Of (31 vs . "tu;r a vm PLANT HATCH, UNIT 1 o is lo lo lo 13l2l1 910 --

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010 01 2 0F 01 5 PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System codes are identified in the text as (EIIS Code XX).

SUMRARY OF EVENT On 3/29/90 at approximately 0800 EST, Unit I was in the Refuel mode at an approximate power level of 0 MWt (approximately 0% of rated thermal power). At that time plant engineering personnel received written notification of the results of off-site testing of pressure vessel safety relief valves (SRVs, EIIS Code RV). Of the eleven SRVs, six exhibited drift in the mechanical lift setpoints in excess of the + 3% tolerance specified by in-service testing (IST) requirements. This voluntary report is being submitted due to the potential industry interest in this event in view of the ongoing efforts of the Boiling Water Reactor Owners' Group (BWROG) to reduce setpoint drif t. The experienced setpoint drift was well~ within the analytical limits existing for reactor vessel over-pressure protection.

The root cause of the event is corrosion-induced bonding of the surface between the pilot valve disc and seat. The experienced setpoint drift in this event is consistent with current industry data demonstrating that both PH13-8Mo discs and stellite discs can occasionally form corrosion bonds with the stellite seat resulting in setpoint drif t.

Corrective actions for this event include refurbishing the valves and continuing to participate in the BWROG effort to develop a new corrective action plan to resolve the SRV setpoint drif t issue.

l I DESCRIPTION OF EVENT On 2/26/90, as part of ongoing Unit I refueling outage activities, the SRVs were  ;

removed from the main steam lines and sent to an off-site contract test laboratory ]

for the purpose of conducting in-service testin American Society of Mechanical Engineers (ASFE)g (IST)

Boiler and in accordance Pressure Yessel Code, with the Section XI, IWV-3512. On 3/29/90, by approximately 0800 EST, plant engineering {

personnel had been notified of the test results for all the SRVs. Of the eleven -

SRVs, six exhibited drif t in the mechanical lift setpoints in excess of the + 3%

tolerance specified in Section XI. The following is a tabulation of test results for the eleven SRVs.

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LICENSEE EVENT REPORT (LER) TEXT CONTINUATION ***aoap ove =o voo-om wm e r .,

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PLANT HATCH, UNIT 1 o p ;o ;o io ; 3l2 l1 9; O __ 0j0l5 .- 0;0 0;3 or 0l 5

Plant Pilot Nameplate Initial  % Nameplate :

Hatch Cartridge Set Press. Lift Press. Actuation :

MPL S/N (psig) (psig) Pressure  :

1B21-F013A 1009 1080 1069 - 1.02  :

1821-F013B 1188 1100 1242 + 12.90  :

1B21-F013C 1003 1100 1171 + 6.45  :

1B21 -F013D* 1189 1090 1089 -

0.08  :

1B21-F013E* 1002 1080 1114 + 3.15  :

1B21-F013F 1007 1090 1102 + 1.10  :

1021 -F013G 1011 1080 1102 + 2.03  :

lB21 -F013H 1006 1090 1100 + 0.84  :

1 B21 -F013J

• 1186 110C 1142 + 3.82  :

1821 -F013K 31 6 1090 1130 + 3.67  :

1021 -F013L* 31 3 1080 1114 + 3.15  :

• Indicates valve discs were made of PH13-Ul10 steel. The remainder were  :

made of Ste111te-6.  :

While the setpoint drif t demonstrated by the six valves (1821-F013B, C, E, J K L) has been determined to be not reportable under the requirements of 10 CFR 50.73, this event is of potential interest to the industry in view of ongoing efforts by the BWROG to address the issue of SRV setpoint drift by eliminating corrosion induced bonding as a contributor.

The BWROG had identified PH13-8Mo as a disc material which had the potential to be less susceptible to forming an adherent corrosion (oxide) bond to the Ste111te-6 seat. This corrosion at the SRV pilot seat-disc interface is one of the causes of SRV setpoint drift. In cooperation with the BWROG study, several BWRs with Target Rock 2-stage SRVs, including Plant Hatch, had installed PH13-BMo discs in up to 50%

of their SRV pilot valves. This facilitated the gathering of in-service data to compare the performance of the new material with the existing Ste111te-6 discs exposed to the same environment.

Early in-service performance of PH13-8Mo appeared to indicate a marked improvement over the stellite discs. However, following a review of the latest in-service data as of November,1989, the BWROG reached the conclusion that the PH13-8Mo discs were not providing the improved setpoint drift performance originally expected. The data indicated that the performance of PH13-8Mo is not significantly different than that of stellite; both materials can occasionally form corrosion bonds which result in significant setpoint drift.

The excessive setpoint drif t demonstrated by the six valves is consistent with the '

previous in-service data reviewed by the BWROG. In this particular case, three of these six valves had PH13-BMo discs; however, the two valves with the highest drift i magnitudes at +12.90 and +6.45 had stellite discs, g ,o. .

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PLANT HATCH, UNIT 1 o p l o l o l o l 3 l2 l1 9l0 -

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The root cause of the event is corrosion-induced bonding of the pilot valve disc and ,

seat. Georgia Power Company is continuing to participate in the BWROG efforts to resolve the SRV setpoint drift issue. .

REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT ,

The report is being submitted voluntarily because the event may h0ve some bearing on the ongoing efforts of the BWROG to address the issue of SRV setpcint drift.

The purpose of the SRVs is to provide over-pressure protection for the reactor pressure vessel and associated reactor coolant system piping. There are a : total of eleven SRVs located on the main steam lines between the reactor pressure vessel and the main -

steamline isolation valves (MSIVs EIIS Code ISV). The SRVs are manufactured by Target Rock Company in compliance with the requirements of ASME Section III (1968 with Winter 1968 addenda), Paragraph N911.4(a)(1) for pilot operated valves. Tnere are three sets of valves; four valves are designed to open at 1080 psig, four at 1090 psig, and three at 1100 psig. The size of the valves coupled with the designated lift pressures is intended to limit a vessel pressure transient to +110% of the reactor vessel' design pressure of l?50 psig, or a maximum of 1375 psig.

In this event, six of the eleven SRVs had setpoint drifts in excess of the +3% ,

tolerance specified in ASME Section XI, with the two maximum setpoint drift magnitudes being + 12.90% and + 6.45%. However, a plant specific analysis has been performed for '

Georgia Power Company by General Electric which demonstrates that Plait Hatch has sufficient margin for over-pressure protection and can tolerate up to a maximum 200 psi '

drift. Specifically, the analysis evaluated the peak vessel pressure at various setpoint drifts up to 200 psi for the plant's most limiting pressurization event, the

! MSIV closure-flux scram event. If it was conservatively assumed that all eleven SRVs

! opened at a lift pressure +9% above the stated nameplate pressure, the resulting '

) pressure transient would be limited to approximately 1300 psig, which is less than the i

l design limit of 1375 psig. Since the total combined setpoint drift experienc0d in the _;

event addressed in this report was significantly less than the uniform +9% assemed in i

the referenced analysis, it is concluded that the limiting pressure transient occurring in conjunction with the measured SRV setpoint drift would not have restalted in exceeding the 1375 psig limit.  :

Based on the above information, it is concluded that this event had no adverse impact  !

on nuclear plant safety. The analysis is conservative in that it assumes worst case initial conditions, and is therefore applicable to all power levels.

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PLANT HATCH, UNIT 1 _

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0l0 0l 5 or 0l 5 nnoa, ,. - , auacw maenn CORRECTIVE ACTIONS ,

Corrective actions for this event include: [

1. Refurbishing the SRVs to bring lift pressures within a + 1% tolerance.

2. Continuing to participate in the BWROG efforts to resolve the SRV setpoint drift issue to determine a new corrective action plan for Plant Hatch.

ADDITIONAL INFOR!% TION i

1. Previous Similar Events:

i An event was reported in LER 50-366/1989-007 dated 10/23/89, in which SRVs with PH13-8Mo pilot valve discs experienced setpoint drift in excess of + 3%.

2. Affected Components Identification: ,

Master Parts List Number: 1821 -F013 B , C , E , J , K , L Manufacturer: Target Rock Company Root Cause Code: B '

Model Number: 7567F EIIS Component Code: RV Type: Two Stage Safety Relief Yalve ,

Manufacturer Code: T020 EIIS System Code: JE -

Reportable to NPRDS: Yes

3. Other Affected Equipment:

No systems other than the Unit 1 Safety Relief Valves were affected by this event. '

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