ML20247C560

Special Rept:On 890224,during Local Leak Rate Surveillance Test,Overall Max Pathway Containment Leakage Failed to Meet Acceptance Criteria & Containment Isolation Valves Exhibited Excessive Leakage.Caused by Improper Valve Stroking

ML20247C560
Person / Time
Site:	Byron
Issue date:	05/18/1989
From:	Chrzanowski R COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
References
NUDOCS 8905240500
Download: ML20247C560 (6)

Text

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a. j 7" Commonwealth Edison -

• i j , / 72 West Adams Street, Chicago, Illinois _-

' ' ' ' ~

.i Address RIply to: Post Office Box 767

\v ' Chicago, Ilknois 60690 - 0767 May 18, 1989 Dr. Thomas E..Murley, Director

' Office of Nuclear Reactor _ Regulation

' U.. S. Nuclear Regulatory Commission Washington, DC. 20555'

Subject:

Byron Station Unit 2 ~

Failed Local Leak Rate Test Report NRC Docket No. 50-45ji Reference (a):- May 3, 1989, letter'from D. R. Muller to T. - J. Korach

'(b): Jahuary:18,.1989, letter from J. A. Silady to T. E. Murley

Dear Dr. Murley:

This letter provides the. report for the Byron Unit 2 local leak rate test .

l (LLRT). 'The specific LLRT failure involved three valves in separate lines that contributed tu exceeding the. acceptance criteria of 0.60 La. Upon' reviewing

, reference (a), since this failure did not consist of'two containment isolation valves ~1n series that were found to leak and exceed the Technical Specification allowable limits, and this failure did not identify a single cause that resulted in a group of components becoming inoperable, an LER was not required.

Commonwealth Edison is submitting the attached special report as documented in reference (b), to ensure that the staff is notified of the LLRT failure, even though an ILRT was not performed.

Please contact this office should any additional information be required.

l Very truly yours, l' ,

R. A. Chrzanowski Nuclear Licensing Administrator

'8905240500 890518 1, PDR ADOCK 05000455 S PDC 0147T 1 cc: Regional Administrator - RIII psg,Z L. N. Olshan - NRR F. A. Maura - RIII [

Byron Resident Inspector (

M..C. Parker - IDNS

,, DEVIATION REPORT DVR NO.

06 - 02 - 89 - 029 STA UNIT YEAR NO. Form Rev 2.0

_ FART 1 l TITLE OF DEVIATION OCCURRED 02/24/89 1400 FAILED LOCAL LEAK RATE TESTS DATE TIME SYSTEM AFFECTED PLANT STATUS AT TIME OF EVENT TESTING MODE 5 POWER (%) 0% WORK REQUES NO. l X l l l PC DESCRIPTION OF EVENT During the execution of Local Leak Rate Tests, the following valves had excessive leakage:

2RY8026, 2 SIB 968, 2500020. All were repaired and retested.

POTENTIALLY SIGNIFICANT EVENT PER NSD DIRECTIVE A-07 YES g g lX l .NO 10CTR50.72 NRC RED PH0HE l l NOTIFICATION MADE l l l X l TIME RESPONSIBLE SUPERVISOR DATE PART2l OPERATING ENGINEER'S C0tttENTS None.

N')N REPORTABLE EVENT l X l NOTIFICATION g g 30 DAY REPORTABLE /10CFR REGION III DATE TIME 5 DAY REPORT PER 10CFR21 l l NSD DATE TIME g ANNUAL /SPECIAL REPORT REQUIRED CECO CORPORATE NOTIFICATION MADE IF AB9VE NOTIFICATION IS PER 10CFR21 R. # ,

,, , TELECOPY CECO CORPORATE OFFICER DATE TIME PRELIMINARY REPORT COMPLETED AND REVIEWED J. W. Schrock 02/27/89 OPERATINL ENGINEER DATE /

INVESTIGATION REPORT 3. RESOLUTION g ,

/ j ACCEPTED BY STATION REVIEW N yUA W v. 4 -4' ~I[ y

/ oM+ t/, leS, -

RESOLUTION APPROVED AND AUTHORIZED FOR DISTRIBUTION ~s _4A N ./ / J STATIONMANMi(T DdTE/

86-5176 (Form 15-52-1) 11-20-85 DOCUMENT ID (0278R/0034R)

o. _ _ _ .

. OEVIATION INVESTIGATION REPORT (DIR)

Facility Name PAGE

_Syron Nuclear Power Station 1 10 0 4 Title F6ILED LOCAL LEAK RATE TESTS EVENT DATE DIR NUPEER REPORT DATE p

/ SEQUENTIAL // REVISION

//j/

f

/j/j MODE MONTH DAY YEAR STA UNIT YEAR NUteER NU M MONTH DAY YEAR 5 POWER

_Al 2 21 4 81 9 1 01.6 01 2 81 9 -

0 l 21 9 --

010 l l l l l0 CONTACT FOR THIS DIR NAE TELEPHONE NUPEER AREA CODE A. Javorik. Assistant Tech Staf f Supervisor Ext. 2206 8l115 2l314l-l5l414l1 COMPLETE ONE LINE FOR EACH COMPONEN A LURE DESCRIBED IN THIS REPORT CAUSE l SYSTEM COMPONENT MANUFAC- REPORTABLE CAUSE SYSTEM COMPONENT MANUFAC- REPORTABLE 6 TURER TO NPRDS TURER TO NPRDS X WlI l Il Sl V El 31 51 0 Y X Al B l Il St V Cl61315 Y X Bl0 l Il Sl V KlOl815 Y l l l l l SUPPLEMENTAL REPORT EXPECTED MONTH DAY YEAR p

SUBMISSION l YES (if ves. complete EXPECTED SUBMISSION DATE) Xl NO TEXT Energy Industry Identification System (E!IS) codes are identified in the text as (XX)

A. PLANT CONDITIONS PRIOR TO EVENT:

, Event Date/ Time 02/24/89 / 1400 Unit 1 MODE 1 - Power Operation Rx Power 99% RCS (AB) Temperature / Pressure Normal Operatino Unit 2 MODE 5 - Cold Shutdown Rx Power 0% RCS (AB) Temperature / Pressure 100'F / 0 PSIG B. DESCRIPTION OF EVENT:

During the Unit 2 first Refueling Outage (B2R01) excution of the Primary Containment (PC) (BD) local Leak Rate Test Surveillance, it was determined that the overall maximum pathway containment leakage failed to meet the acceptance criteria of 0.60 La (277.76 SCFH). The surveillance was started on 01/09/89 and ended on 02/24/89, Three containment isolation valves exhibited excessive leakage that contributed to the overall failure. The valves involved and the dates of failure discovery are: Steam Generator Blowdown Isolation Valve 2SD002D (1/28/89), Pressurizer Relief Tank (PRT) to the Auto Gas Analyzer Isolation Valve 2RY8026 (2/4/89), and Safety Injection Accumulator Nitrogen Supply Check Valve 25I8968 (2/16/B9). Nuclear Work Requests were ger.erated and corrective maintenance was nerfonned to reduce leakage rates. Because the Unit was in Modes 5 and 6, the Limiting Condition for Opera on Action Requirement was not applicable. All components were repaired and retested prior to the unit entering Mode 4, when containment integrity is required.

(0278R/0034R)

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. DEVIATION INVESTIGATION REPORT TEXT CONTINUATION

___ Form Rev 2.0 FACILITY HAHE DIR NUMBER PAGE SEQUENTIAL REVISION STA UNIT YEAR NUMBER NUPEER Byron Nuclear Power Station 01 6 01 2 81 9 -

0l2l9 0 10 2 0F 014 TEXT Energy Industry Identification System (E!IS) codes are identified in the text as (XX)

C. CAUSE OF EVENT:

The Steam Generator Biowdown Isolation Valve failed due to improper valve stroke. The root cause of the improper stroke is indeterminate.

The PRT to the Auto Gas Analyzer Isolation Valve had a diaphragm which had degraded through normal usage.

The Nitr. gen Supply Check Valve failed due to rubber debris lodged in the valve internals. The source of the debris could not be determined.

D. SAFETY ANALYSIS:

The safety consequences of the Steam Generator Blowdown Isolation Valve leakage is minimized by the IB classification (seismic and safety related) design of the steam generators and associated piping inside containment. In a design basis Loss of Coolant Accident, the steam generators and associated piping provide the containment isolation function. The isolation valve is not challenged by post accident containment pressure. The isolation valves receive an auto closure signal during a Phase A containment isolation in order to maintain Steam Generator inventory, and thus do not perfonn a containment isolation function.

The other failed isolation valves (2RY8026, 2SI80968) have redundant automatic containment isolation valves in series with them. These redundant valves passed the Local Leak Rate Testing and provided acceptable containment isolation.

E. CORRECTIVE ACTIONS:

Valve 2SD002D had its valve stroke adjusted to ensure proper seating. The diaphragm and 0-ring for 2RY8026 were replaced. The valve internals for check valve 2S18968 were cleaned and the seating surfaces were lapped. All components were retested with acceptable results. The valves will continue to be tested at j the frequency specified by 10CFR50 Appendix J and the Technical Specifications.

A Technical Specification change request is being submitted to exempt the Steam Generator Blowdown Containment Isolation valves, 2SD002A-H and 2SD005A-D, from Type C Testing. This exemption is based on the evaluation given in the first paragraph of the safety analysis above. The same evaluation is already j applied to the feedwater and Main Steam Isolation Valves. Action Item Record 88-0267 tracks this corrective action.

i (0278R/0034R) m - - - - - - - - --- _ - -

h

. . DEVIATION INVESTIGATION REPORT TEXT CONTINUATION Form Rev'2.0

-FACILITY NANE. DIR NUMBER PAGE SEQUENTIAL REVISION STA UNIT YEAR NUMBER ' NUMBER 8vron Nuclear Power Station 01 6 01 2 81 9 --

012l9 -

0 10 3 0F 0l4

. TEXT. . Energy Industry ! identification System (EIIS)' codes are identified in the text as [XX].

'F. BECURRING QENTS SEARCH AND ANALYSIS:

a )' '{yENT SEARCH IDIR. LER)

DIR NUMBER ~ IjJJ,g -

6-1-87-061 Local Leak Rate Test Failure.

6-2-87-085 Local Leak Rate Test failure.

6-2-87-118 Local Leak Rate Test Failure.

6-1-88-205 Local Leak Rate Test failure.

Valve 2RY8026 had previously failed Local Leak Rate Testing (DIR 6-2-87-085)..

y .. b ) INDUSTRY SEARCH (OPEX's NPRDS)

An NPRDS search was not performed due to the generic nature of the failures.

'c).. 18fR'

' All of the afinAed valves had work histories that include repacking, stroke adjustment, and operator repair. 2RY8026 had a stroke adjustment to repair the previous LLRT failure. 2518968 was repaired during pre-operational testing due to a failed LLRT. None of the maintenance history is of.

'an exceptional nature, d) ANALYSIS No additional information can be obtained from the recurring events search, j  ; G. CONPONENT FAILURE DATA:

I

't%NufACTURER NDENCLATURE Borg-Warner 1500 lb, 2 inch Valve Assy 851-C-0148/072 -

. Cope-Vulcan 3/8 ASNE Valve Class 1500 3/8TA78RL Kerotest 1 Series 600# Y-Type Check Valve IC66 H. OTHER RELATED DOCUENTS:

None.

(0278R/0034R)

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l .' .t. ,

DEVIATION INVESTIGATION REPORT TEXT CONTINUATION-Fonn Rev 2.0 FACILITY NAME DIR NUPEER PAGE j, SEQUENTIAL REVISION l, 1 STA UNIT YEAR NUPBER NUPBER l Byron Nuclear Power Station 01 6 01 2 81 9 -

012'l9 -

0 l0 4 0F O'l 4

-TEXT- Energy Industry Identification System (EIIS) cedes are identified in the text as [XX)

I. EFFECTIVENESS REVIEW:

. None Scheduled.

J. -ADDITIONAL DATA:

a)- Affected Technical Specification: 3.6.1.C, 4.6.1.2 b) Procedures: .BVS 6.1.2.d-1.1 through .26 c) Equipment Involved: See Component Failure Data

d) Other

Local Leak Rate Test Failures. Type B, Type C Failures

'(0278R/0034R)

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