ML20024G740

LER 91-008-00:on 910327,main Steam Isolation Valve Local Leak Rate Test Failed Due to Normal Equipment Wear Resulting in Degradation of Valve Seating Surfaces.Valves Repaired & retested.W/910424 Ltr

ML20024G740
Person / Time
Site:	Hatch
Issue date:	04/24/1991
From:	Beckham J, Tipps S GEORGIA POWER CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HL-1596, LER-91-008-01, NUDOCS 9104290165
Download: ML20024G740 (6)

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HL-1596 001503 April 24, 1991 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 PLANT HATCH - UNIT 2 NRC DOCKET 50-366 OPERATING LICENSE NPF-5 LICENSEE EVENT REPORT MAIN STEAM ISOLATION VALVE LOCAL LEAK RATE TEST FAILURES Gentlemen:

In accordance with the requirements of 10 CFR 50.73(a)(2)(v), Georgia Power Company is submitting the enclosed Licensee Event Report (LER) concerning a condition that could have prevented an ESF from fully performing its safety function. This event occurred at Plant Hatch -

Unit 2.

4 Sincerely,

/ p -&$ y J. T. Beckham, Jr.

SWR /ct

Enclosure:

LER 50-3G6/1991 008 cc: (See next page.)

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s 0, J GeorgiaPower d U.S. Nuclear Regulatory Commission April 24, 1991 Page Two cc: Georaia Power Comoany Mr. H. L. Sumner, General Manager - Nuclear Plant Mr. J. D. Heidt, Manager Engineering and Licensing - Hatch NORMS-U.S. Nuclear Reaulatory Commission. Washinaton. D.C.

Mr. K. Jabbour, Licensing Project Manager - Hatch U.S. Nuclear Reaulatory Commission. Reaton 11 Mr. S. D. Ebneter, Regional Administrator Mr. L. D. Wert, Senior Resident Inspector - Hatch i

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rjn JM U.5.1%AALAR MLUil.AlWi WPmbhlJI hTA U b siiA LICENSEE EVENT REPORT (LER)

FALIL4II hAME (4) UUiktl h VM blii (2) FN f31 PLANT liA'IDI, INIT 2 05000366 L Or l4 IITLE (4)

MAIN STEAM ISOLATION VALVE LOCAL IIAK PATE TEST FAllllRES EsthI DATE (5) LER huMBER (6) REPORT DAIL (7) OTHER FACILITIES lhVOLbED (B)

M0hTH DAY YEAR YEAR SEQ hum KEV M0hTH DAY YEAR .FACILIII hAME5 DOCAEI h0MbER(5) 05000 03 27 91 91 008 00 04 24 91 05000 0?tRAflhG N5 m)RL 15 W W W NRMM M M WRNH M 10 UR U1)

MODE (9) 5 20.40r(b) ~ 20.405(c) 50.73(a)(r)(tv) ~ 73.71(b) 20.405(a)(1)(t) 50.36(c)(1) T 50.73(a)(!)(v) 73.71(c)

PgR 000 -

20.405(a)(1)(11) [ 50.36(c)(2) [ 50.73(a)(2)(vit) ~_ OTHER (Specify in

_ 20.40$(a)(1)(t11) _ 50.73(a)(2)(t) _ 50.73(a)(2)(vtit)(A) Abstract below) 20.405(a)(1)(ty) ~ 50.73(a)(2)(tt) ~ 50.73(a)(2)(vitt)(B) 20.405(a)(1)(v) 50.73(a)(2)(itt) $0.73(a)(?)(x)

LICE 45EE C0hTACT FOR THI5 LER (li)

NAME IELEPn0hE huMbEk LREA CODE STEVEN B. TIPPS, MANAGE NUCLEAR SA}YrY AND (XHPLIANCE, ilA*Iul 912 367 7851 COMPLETE OhE LlhE FOR [ACh FAILURE DESCRIBED lh IHl5 REPORI (13)

CAUSE SYSTEM COMPONENT MANUFAC- P RT CAUSE SYSTEM' COMPONENT MAhUFAC-7g hPROS 7URER groRT p

X SB ISV R340 YES SUPPLEMEhTAL REPORI EXPECTED (14D M0hin DAV VIAR

~""]YES(Ifyes,completeEXPECTEDSUSMISSIONDATE) ~E] NO DAN (n.ON)

AB5IRACT l16)

On 1/27/91, at approximately 1200 CST, Unit 2 was in the Refuel mode with the reactor vessel flooded. Fuel removal was in progress with fuel partially removed from the core. At that time, Local Leak Rate Testing (LLRT) of Main Steam Line Isolation valves (MSIVs, EIIS Code SB) 2B21-F022B and 2B21-F028B was completed, confirming that one pair of MSIV's in the same main steam line (MSL) leakad in excess of the Technical Specifications leakage limit. The other 6

-MSIVa passed their LLRT, meeting the Technical Specifications requirement. The leakage-of 92 SCFil through the one MSL.was well within the capacity of the MSIV Leakage Control System to handle in the unlikely event of a loss of Coolant Accident with the as found MSIV leakage rates.

Thel root cause of this event is attributed to normal equipment wear resulting in

slight degradation of the valve seating surfaces.

Corrective actions for this event include repairing and re testing the valves.

l

grpMA u.b. NEW RMAIC MmMWh he lo M LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAMI (1) DOCKET huMB(R (!) LIR huMBER (5) PAGE (3)

YEAR SEQ hum lil V PIAVT liATQi, UNIT 2 05000366 91 008 00 2 0F 4 un PIANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System codes are identified in the text as (EIIS Code XX).

DESCRIPTION OF Tile EVENT The Local Leak Rate Test (LIAT) for Main Steam Line Isolation Valves (MSIVs, EIIS Code SB) was begun on Unit 2 on 3/23/91 in accordance with surveillance procedure 42SV TET-001 2S, " PRIMARY CONTAINMENT PERIODIC TYPE B AND TYPE C LEAKACE TESTS." Of the eight MSIVs. inboard and outboard MSIVs 2B21.F022B and  !

2B21-F0288, respectively, were the only MSIVs which did not meet the Technical I Specifications limit on MSIV leakage of 11.5 standard cubic feet per hour  !

(SCril) . These two MSIVs are located on the same Main Steam Line (MSL).

1 Specifically, MSIV LLRT testing is performed in the following manner per procedure. In the first stage, the reactor vessel is not yet flooded for defueling activities. The cavity between the two MSIVs is pressurized with air to approximately 28.8 psig per Unit 2 Technical Specifications section 3.6.1.2.c, and then the leakage is measured. This is a measure of the total leakage from the cavity, and the proportion of the leakage from each valve exiting the cavity is not yet known. In the second stage of the test, the reactor vessel and MSL leading to the valves is flooded with water sufficient to provide pressure against the inhoard MSIV approximately equal to the test pressure. With the MSL thus flooded, the cavity between the MSIVs is again pressurized with air to 28.8 psig, and the leakage is measured. Flooding the MSL essentially seals off the inboard MSIV, limiting leakage to that of the outboard MSIV only. Thus the second measurement yields the sum of the leakage for the outboard valve and the MSIV Leakage Control System (MSIV LCS, EIIS Code BD) valve for that pair of MSIVs. The difference between the first measurement and the second is defined as the leakage of the inboard valve.

For inboard and outboard MSIVs 2B21 F022B and 2B21 F028B, in the first phase of their LLRT, the leakage was approximately 262 SCFM. The reactor was then flooded, and the second stage of testing on these valves was completed on 3/27/91. At that time, the leakage was measured as approximately 92 SCFH.

Since, in this case, the total leakage through the MSL cannot be greater than that of the two outboard valves, the total leakage throu6h the MSL was at that time estimated at approximately 92 SCFH. Thus, it was confirmed that Icakage in excess of the maximum allowable existed on this MSIV pair. Inspection of the MSIV's following disassembly revealed no disrupted metal or visible damage to the poppet or seat.

yn wA u.5.nuwkwuw um m wegg w LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION l rACIL11Y NAME (1) 00cKti NUMetR (2) LtR NUMBER (5) PAGE (3)

YEAR 50Q huh EEV piNir llAToi, UNIT 2 05000366 91 008 00 3 M 4 ILAT CAUSE OF Tile EVENT The root cause of this event is attributed to normal equipment wear resulting in slight degradation of the valve seating surfaces.

REpORTABILITY ANALYSIS AND SAFETY ASSESSMENT This event is reportable per 10 CPR 50.73(a)(2)(v) because the leakage rates of two MSIVs located in the same MSL exceeded the leakage rate limit of 11.5 SCFil per MSIV stated in Unit 2 Technical Specifications section 3.6.1.2.c.

The surveillance testing performed to measure primary containment leakage rates is in accordance with the requirements of 10 CFR 50, Appendix J (Types A, B, and C testing) and is performed on a frequency specified by the Unit 2 Technical Specifications. This testing allows timely detection of valve degradation so that maintenance and repairs can be performed as necessary to restore leakage rates to within their Technical Specifications limits. The limitations on containment leakage rates ensure that the total amount of leakage during a potential accident will not exceed the value assumed in the Final Safety Analysis Report (FSAR) for Hatch Unit 2. The maintenance of primary contaitunent integrity in conjunction with these leakage rate limitations will maintain the potential site boundary radiation doses below the limits of 10 CPR 100 during accident conditions.

In this event MSIVs 2B21 F022B and 2B21 F028B exceeded their leakage limit of 11.5 SCFH per valve. Both of these valves are located in the same MSL, and the total leakage through this MSL was approximately 92 SCFil. No other MSIVs had leakage rates in excess of the Technical Specifications limits. It is not possible to determine exactly when the MSIVs exceeded their leakage requirements during the surveillance interval. The MSIVs were within allowable limits at the beginning of the surveillance interval and such degradation is often a function of time and use.

The large volume of the main condenser and main steam lines could provide for sufficient holdup and plate out of any fission products contained in the HSIV leakage to minimize the contribution of MSIV leakage to offsite radiation doses during potential accident conditions. Unit 2 currently has an MSIV Leakage Control System (MSIV-LCS). This system can handle up to 100 SCFil leakage per MSL by directing the leakage into an area of the plant served by the Standby Cas Treatment System (SCTS, EIIS Code Bil) for processing prior to release to the atmosphere. The MSIV LCS would assure that offsite dose due to MSIV leakage does not contribute to exceeding 10 CFR 100 limits in the unlikely event of a Loss-of-Coolant Accident (LOCA) with the as found MSIV leakage rates.

Based on the above analysis, it is concluded that this event had no adverse impact on nuclear safety. This analysis is applicable to all power levels.

l gge mA u.a. maw muuuum mma mg g' cu LICENSEE EVENT REPORT (LER)

TEXT CONTINUATILN FACILITT NAME (1)- 00CKli NUMBER (2) LER huMBER (5) PAGE (3)

YEAR 5[Q hVM R[V PIRfr HATCH, INIT 2 05000366 91 008 00 4 0F 4 un CORRECTIVE ACTIONS The poppet of the outboard valve, 2B21 F028B, was replaced with a new poppet from stock. Replacement was necessary because some of the threads in the poppet assembly became galled during disassembly and not because of damage.to seating surfaces. The in body seat of this valve was also ground to provide a better seating surface. Repairs on the inboard valve have been completed and both valves have been successfully leak rate tested. ,

ADDITIONAL INFORMATION

1. Other Systems Affected: No other systems were affected by this event.

2. Previous Similar Events: There have been no evenca during the past two

, years in which two MSIVs in the same MSb failed LiRT.

3. Failed Components Identification:

Master Parts List Number: 2821 F0228 and 2B21 F028B Manufacturer: Rockwell Manufacturing Corporation Model Number:- 1612JMMNTY Type: Air Operated Valve Manufacturer Code: R340 EIIS System Code: SB Reportable to NPRDS: Yes Root Cause Code: -X EIIS Component Code; ISV

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