ML19327B328

LER 89-007-00:on 890926,four Safety Relief Valves Had Exhibited Drift in Mechanical Lift Setpoints in Excess of 3% Tolerance Specified by Inservice Testing Requirements.Root Cause Being Investigated.Valves refurbished.W/891023 Ltr

ML19327B328
Person / Time
Site:	Hatch
Issue date:	10/23/1989
From:	Hairston W, Tipps S GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-797, LER-89-007-03, LER-89-7-3, NUDOCS 8910300186
Download: ML19327B328 (7)

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. Geof Da Poe Company h"... e~~

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• 333 Pe.1 mort Avenue -

Arttsa, Georg.a 303D8 be;Aone 4D4 (20 3195

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40 trwness Center Prtway Post Offee 00: 1?95 Dirmingham Apatiama 35?01-loteptiore 205 N.B 14,01 t*c s.o c',,rn e4 x:rr r y:. trim s- W. G. Holrotori, all

$ Seno Voe President Nxtew oweons HL-797 L 0355V

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october 23, 1989 i

l' U.S. Nuclear Regulatory Commission

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~ ATTN: Document Control Desk j Hashington, D.C. 20555 i.

PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 l ..

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LICENSEE EVENT REPORT SAFETY RELIEF VALVES WITH PH13-8Mo PILOT yALYE DISCS EXPERIENCE SETPOINT DRIFT Gentlemen:

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

This event occurred at Plant Hatch - Unit 2.

Sincerely, W.1b h H. G. Hairston, III l

l SHR/ct

Enclosure:

LER 50-366/1989-007 c: (See next page.)

G910300186 891023 PDR S

ADOCK 05000366 FDC N}'

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GeorgiaIbwer b U.S. Nucleir Regulatory Commission october 23- 1989 Page Two '  !

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

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GO-NORMS U.S. Nuclear Regy111ory Commission. Washington. D.C. I Mr. L. P. Crocker, Licensing Project Manager - Hatch U.S. Nuclear Reculatory Commission. Reaion II ,

Mr. S. D. Ebneter, Regional Administrator Mr. J. E. Henning, Senior Resident Inspector - Hatch r

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PLANT HATCH, UNIT 2 o Is I o Io I o D l 616 1 lofl 015 l 16148 6 1 l SAFETY RELIEF VALVES WITH PH13-8Mo PILOT VALVE DISCS EXPERIENCE SETPOINT DRIFT tytWT DAT8 (El 489WUMDth11 htPOtf Daf t IPs OTM1 h 9 ACILif tll flWVOLV801.)

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~A.4 n a , osi us.. a AktA (QD4 Stevee D. Tipps, Manager Nuclear Safety and Compliance Hatch 9 i 1,2 3 6,7 i i 78 i Sl 1 i COMPLtit DN8 tlht FOle R ACH COMPOh8WT f AlLUtt Dt.Chl.tD IN THIS hlPO8T 1131 COU$5 SYltt M COM*0NINT "M 'C' To$,'n'[I' CAvst sittiv COM*0% TNT "#7gf 0 Tp",'n'jj'

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An t; &c f ,u-, ,= ... . . . ., , , . .,. ,, ,...,,.. - n . , t On 9/26/89, at approximately 1200 EDT, Unit 2 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 i notification of the results of off-site testing of pressure vessel safety I relief valves (SRVs). Of the eleven SRVs, four had exhibited drif t in the mechanical lif t setpoints in excess of the + 3% tolerance specified by in-

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service testing (IST) requirements. This voluntary report is being submitted due to the potential industry interest in this event since three of the four referenced SRVs had pilot valve discs of PH13-8Mo which is the

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L new disc material being tested by the Boiling Water Reactor Owners' Group (BWROG) to reduce setpoint drift. The experienced setpoint drift was well within the analytical limits existing for reactor vessel overpressure protection. '

The root cause of the event is being investigated in cooperation with the BWROG effort. The experienced setpoint drift in this event is inconsistent with previous industry data showing average setpoint drif t of about 50% less with this new disc material, j Corrective actions for this event thus far included refurbishing the [

l valves and continuing the root cause investigation.

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! UCENSEE EVENT REPORT lLER) TEXT CONTINUAT60N amovio on wo m-om u . tomia swa emm isami m . nonii wounin u, ,,,,,,,,,,,, ,,,,,,

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~ PLANT HATCH, UNIT 2 Twi n-. we o 16 lc t o l o l 3 l 6l 6 8l9 -

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@0 q2 or 0 15 wc e am.wi PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System codes are identified in the text i as (E!!S Code XX).

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SUMMARY

OF EVENT

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On 9/26/89, at approximately 1200 EDT, Unit 2 was in the Refuel mode at l 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 l l safety relief valves (SRVs E!!S Code RV). Of the eleven SRVs, four had l exhibited drift in the mechanical lif t setpoints in excess of the + 3%

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tolerance specified by in-service testing (IST) requirements of thii i American Society of Mechanical Engineers (ASME)Section XI, IWY-3512. [

This voluntary report is being submitted due to the potential industry interest in this. event since three of the four referenced SRVs had pilot i valve discs of PH13-8Mo which is the new disc material being tested by the Boiling Water Reactor Owners' Group (BWR03) to reduce setpoint i drift. The experienced setpoint drif t was well within the analytical '

limits existing for reactor vessel overpressure protection. ,

The root chuse of the event is being investigated in cooperation with the BWROG effort. The experienced setpoint drift in this event is  :

inconsistent with previous industr/ data showing average setpoint drift  !

of about 50% less with this new disc material. l Corrective actions for this event thus far included refurbishing the  !

valves and continuing the root cause investigation.  !

DESCRIPTION OF EVENT On 9/09/89, as part of ongoing Unit 2 refueling outage activities, the  :

SRVs were removed from the main steam lines and sent to in cff-site  !

contract test laboratory for the purpose of conducting in-service testing (IST) in accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, IWV-3512.

On 9/26/89, by approximately 1200 EDT, plant engineering personnel had -

been notified of the test results for all the SRVs. Of the eleven SRVs,  !

four had exhibited drif t in the mechanical lift setpoints in excess of '

the + 3% tolerance specified in Section XI. The following is a tabuTation of test results for the eleven SRVs.  ;

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,  : Plant Pilot Nameplate Initial  % Naneplate .  : f

Hatch Cartridge Set Press. Lift Press. Actuation  :  :

MPL S/N (psig) (psig) Pressure  :

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2B21-F013A* 31 5 1100 1077 -

2.14  :

.: 2B21-F013B 31 2 1090 1199 + 2.66  :

2B21-F0130* 308 1090 1129 + 3.58  :

2B21-F013D 1001 1100 1115 + 1.36  :  :

2B21-F013E 303 1110 1135 + 2.25  :  !

2B21-F013F 31 0 1090 1103 + 1.19  : l

' 2021-F013G 314 1090 1150 + 5.50  :

2B21 -F013H* 306 1110 1227 + 10. 54  : '

2821-F013K* 302 1100 '1201 + 9.18  : i

2B21 -F013L* 307 1110 1137 + 2.43  :

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2B21-F013M 301 1100 1118 + 1.64  :

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• Indicates valve discs were made of PH13 8Mo steel. The remainder  :

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were made of Ste111te-6.  :

r While the setpoint drif t demonstrated by the four valves (2821-F013C, G, '

H, K) has been determined to be not reportable under the requirements of i 10 CFR 50.73, this event is of potential interest to the industry in '

-view of ongoing efforts by the BWROG to verify the performance of a new ,

SRV pilot valve disc material, PH13-8Mo. The BWROG has identified '

PH13-8Mo as a disc material which has 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 are installing PH13-8Mo discs in up to 50% of their SRVs. This will facilitate the gathering of in-service data to compare the performance of the new  :

material with the existing Ste111te-6 discs exposed to the same l

^ environment.

Plant Hatch has been participating in the BWROG program and three of the ,

four valves which demonstrated excessive setpoint drift had PH13-BMo '

discs. Prior to this point the in-service data collected has indicated '

that SRVs with PH13-8Mo discs display an average setpoint drift of about ,

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50% less than those with Stellite-6 discs that were exposed to the same environment. Therefore, the two SRVs with PH13-8Mo discs with drift ,

magnitudes of +9.18 and +10.54 are of particular interest due to their inconsistency with previous data. "

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ms w . ,m e <~no mu,mn CAUSE OF THE EVENT The root cause of the event has not yet been conclusively determined.

Since the two SRVs which experienced the highest drift, +9.18 and +10.54 percent, had PH13-8Mo discs, the assistance of the BWROG and General Electric has been requested in the root cause investigation. Pertinent data has already been provided to Ger.eral Electric and the BWROG has a meeting planned in November,1989, to review this recent Plant Hatch SRV data as well as new in-service data from other BWRs which have installed PH13-8Mo discs in some of their SRVs.

1 REPORTABILITY ANALYSIS AND SAFETY ASSESSMENT This report is being submitted voluntarily because the event may have some bearing on the ongoing effortr, of the BWROG to address the issue of SRV setpoint 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 in the main steam lines between the reactor pressure vessel and the main steam isolation valves (MSIVS E!!S Code ISV). The SRVs are manufactured by Target Rock Company in compliance with the requirements of ASME Section III (1968 with Winter t 1968 addenda), Paragraph N911.4(a)(1) for pilot operated valves. There are three sets of valves; four valves are designed to open at 1090 psig, l four at 1100 psig, and tnree designed to open at 1110 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 1250 psig, or a maximum of 1375 psig.

In this event, four of the eleven SRVs had setpoint drifts in excess of the +3% tolerance specified in ASME Section XI, with the two maximum setpFint drif t magnitudes being +10.54 and +9.18. However, a plant

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specific analysis has been performed for Georgia Power Company by General Electric which demonstrates that Plant Hatch has sufficient margin for overprest.ure 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 pressurizatior, event, the MSIV closure-flux scram event. If it

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was conservatively assumed that all eleven SRVs opened at a lif t l pressure +9% above the stated nameplate pressure, the resulting pressure i

transient would be limited to approximately 1300 psig, which is less than the design limit of 1375 psig. Since the total combined setpoint l drift experienced in the event addressed in this report was l significantly less than the uniform +9% assumed in the referenced analysis, it is concluded that the limiting pressure transient occurring in conjunction with the measured SRV setpoint drif t would not have resulted in exceeding the 1375 psig limit.  !

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Based on the above information, it is concluded that this event had nn  !

l adverse impact on nuclear plant safety. The analysis is conservative in  !

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that it assumes worst case initial conditions, and is therefore  !

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applicable to all power levels. '

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i I i l CORRECTIVE ACTIONS ,

Corrective actions for this event include:

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1. refurbishing the SRVs to bring lift pressures within a +1% '

tolerance. .

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2. conducting a root cause investigation in cooperation with the BWROG i activity concerning SRV setpoint drift. ' The conclusions of this -

root cause investigation and any resulting further corrective  ;

actions will be reported in a revision to this LER.  !

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ADDITIONAL INFORMATION .

1. Previous Similar Events: j

No similar events where SRVs with pilot discs of PH13-8Mo exceeded '

the +3% tolerance occurred.

2. Affected Components Identification:

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Master Parts List Number: 1821-F013C , G, H. K Manufacturer: Target Rock Company [

Root Cause Code: X  ;

Model Number: 7567F EIIS Componcnt Code: RV

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Type: Two Stage Safety Relief Yalve Manufacturer Code: T020  !

EIIS System Code: JE l

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Reportable to NPRDS: Yes I

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3. Other Affected Equipment:

ll No systems other than the Unit 2 Safety Relief Valves were affected by this event.

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