ML19257A722

LER 79-027/03L-0:on 791126,six Occurrences Identified Whereby Excessive Leakages Were Measured W/Respect to Primary Containment Isolation Valves During Local Leak Rate Testing.All Valves to Be Repaired & Tested

ML19257A722
Person / Time
Site:	Quad Cities
Issue date:	12/24/1979
From:	Vahrenwald J COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
Shared Package
ML19257A719	List: ML19257A718 ML19257A722
References
LER-79-027-03L-04, LER-79-27-3L-4, NUDOCS 8001070467
Download: ML19257A722 (4)

Text

NRC FORM'366 U. S. NUCLEAR REGULATORY COMMISSION (7 77)

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69 EVENT DATE 74 75 REPORT D ATE 80 EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 l While performina the Main Steam Isolation Valve local leak rate tast valves o 3 l A0 2-203-IA, A0 2-203-2A, A0 2-203-18, and A0 2-203-2B were found to have leakane i o 4 rates in excess of the 11.5 SCFH allowable limit as soecified by Technical SoeCificatlon o 5 4.7.A.2.i.c. All valves will be repaired and tested before Unit start-up. The 1 o e C and D steam line MSIV's had satisfactory leak test results. I o 7 l l

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7 8 9 SYSTEM CAUSE CAUSE C OY P. VA LV E CODE CCDE SU8COOE CCYPONENT CCDE SU8 CODE SUSCODE 7

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36 40 42 44 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS i o l Not all valves have been reoaired at this date. A supolemental report will be sent I i i l upon completion of all testinO and when all corrective actions reaardina these i

, 7 valves have been completed. l i 3 l 1 4 l l 7 8 9 80 STA S  % POWER OTHER STATUS ISCO RY DISCOVERY DESCRIPTION i s

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REPORT DATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 l While performing the main steam line drain local leak rate test. valve MO 2-220-2 was I lo a l found to have a leakage of 26.0 scfh. This is in excess of the 18.36 scfh allowable 1 o 4 I limit as Specified by Technical Specification 4.7.A.2.i. The MO 2-220-2 valve will be I ois repaired and tested before Unit Start-up. In-line isolation valve M0 2-220-1 had an l _

o e I acceptable leakage value. 1 o 7 l 0 8 l l 7 8 9 80 SYSTEM CAUSE CALSE COYP- VALVE CODE CCDE SLBCCOE COMPONENT CODE SUBCODE SUSCODE 7

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3J 34 35 36 37 40 43 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS 1 0 l The MO 2-770-7 valv, has -nr Loan ran,traa ,* este a,o. A sucolemantal rannrt will l 11 It I be sent upon comOletion of all testino and when all correctiva actions hava hean I i 7 completed. I t a l l 1 4 l l 7 8 9 80 ST S  % POWER OTH ER STATUS IS O RY DISCOVERY DESCRIPT'ON 1 s Hlh l 0 l 0 l 0 l@l NA l W @l Local Laak Rate Test l A TiVITY CONTENT RELEASED OF RELEASE AMOUNT OF ACTivlTY LOCATION OF RELEASE

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LICENSEE EVENT REPORT CONTROL BLOCK: l 1

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lh (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) 7 o i 8 9 lilLIoIAlDl2l@l0l0l0 -l0l0l0l-l0l0l0lhlhl1l1 l1 l 1lhl LICENSEE CODE 14 15 LICENSE NUMSER 25 26 LICENSE TYPE J0 l

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!o 1 2'RCI L}@l015 l 0l O l o l 2 l 61568 l@l 111121617 do 7 8 60 61 DOCKET NUYSER 69 EVENT DATE 74 75 REPORT DATE EVENT OFSCRIPTION AND PROBABLE CONSEQUENCES h o 2 l While performing the RCIC turbine steam sv2nly valve local leak rate test, valve I o 3 l M0 2-1301-16 was found to have a leakare of 76.7 scfh. This is in excess of the I o 4 18.36 scfh allowable limit as spec fied by Technical Specification 4.7.A.2.i. The l o s 1 M0 2-1301-16 valve will be repaired and tested before Unit Start-up. In-line valve l 0 6 l M0 2-1301-17 was found to have satisfactory leakage. l 0 7 l l 0 8 l l 7 9 g 80 SYSTEM CAUSE CAUSE COVP. VALvg COPE CODE SUBCOCE COMPONENT CODE SUSCODE SUSCCCE o 9 a

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43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h i o l The M0 2-1301-16 valve has not been recairad at th e darn A so-nin-o *,1 roenr+ l

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NRC FORM 366 U. S. NUCLEAR . .LATORY COMMISSION (7-77)

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7 8 60 DOCKET NUM8ER 68 69 EV ENT DATE 74 75 REPORT DATE EVENT DESCRIPTION AND PRO 8 ABLE CONSEQUENCES h a 2 l While oerforming the feedwater check valve local leak rara +mr- *ba leakaga m*a I o 3 through feedwater check valve 2-220-62B was found to be 405.8 SCFH. This is in 1 o 4 I excess of the 18.6 SCFH allowable limit as specified bv Technical Soecificat'on o s l 4.7.A.2.i. The 2-220-628 check valve will be repaired and tested before unit I o s I start up. In-line check valve 2-220-58B was found to have a leakage value wi thin 1 o 7 l limits. I o a I 7 8 9 30 SYSTEM CAUSE CAUSE COMP. VA LV E CODE CODE SLOCODE COMPONENT CODE SUSCODE SUSCOCE o 9 lC lH j@ l El@ l 8 l@ Vl Al Ll Vl El X[@ l C l@ 20W @

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41 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS h i o l The 2-220-628 valve has 7t been repaired at this date. A supolemental report will I i i l be sent upon completion of all testing and when all corrective actions have been 1

, , g completed. l t 3 1 1 4 l l 7 8 9 80 STA S  % POWE R OTHER STATUS ISCO RY OtSCOVERY DESCRIPTION i 5 Hj g l0 l0l0lgl NA l l Blgl Local Leak Rate Tesc l ACTIVITY CCf TENT RELEASED CF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELEASE i 6 NA l l NA I 7 3 9 l Z l @ l10Z !@l 11 44 45 80 8E RSC'.'.EL Ex POSUR ES NUMBE A TYPE OESCRIPT:ON 8 i 7 l 0 l 0 l 0 l@l Zl@l NA l PERSONNEL INr RIES NUYBE4 CESCRIPTICN i NA 7 d 9 0 l 0 l 0 l@l 't ^2 3; l

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NRC FORM 366 U. S. NUCLEAR REGULATORY COMMISSloN (7-77)

LICENSEE EVENT REPORT (PLEASE PRINT OR TYPE ALL REQUIREO INFORMATION)

CONTROL BLOCK: l 1

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REPORT DATE 80 EVENT DESCRIPTION AND PROBA8LE CONSEQUENCES h o 2 l While performing local leak rate testing, the containment purge volume bounded by l g 3 l valves A0 2-1601-21, 22, 55, and 56 was found to have a leak rate of 136.1 SCFH. l o 4  ; This is in excess of the 18.36 SCFH allowable limit for any one valve as specified l o 5 1 by Technical Specification 4.7.A.2.i. The containment purge valves will be l o 6 l repaired and tested before unit start up. l o 7 l l 0 3 l  !

7 8 9 80 SYSTEM CAUSE CAUSE COMP- VA LV E CODE CODE SUBCODE COMPONENT CODE SU8 CODE SOBCODE 7

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36 40 44 CAUSE CESCRIPTION AND COPRECTIVE ACTIONS h i

o l Valve 1601-21 is suspected to be leakina excessivelv. The valve has not been I i repaired at this date. A supplemental report will be sent uoan comoletion of all I

, 7 l testing and when all corrective actions have been completed. l 1

3! l l i 4 1 7 8 9 80 STA S  % POWER CTHER STATUS ISCOV RY DISCOVERY DESCR6PTION Local Leak Rate Test i 5 Hl@ l0l0l0l(l NA l 8l@l l AC TI VITY CO TENT AELEASED OF AELEASE AMOUP.1 OF ACTIVITY LOCATION OF RELEASE 7

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. akenwald PHONE 309-654-2241 ext. 195 f

NRC FOfiM'366 U. S. NUCLEAR REGULATORY COMMISSION (7 771 LICENSEE EVENT REPORT CONTROL BLOCK: l 3

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lh (PLEASE PRINT OR TYPE ALL REQUIRED INFORMATION) o i l l l L l Q l A l 0 l 2 l@l 0] 0l 0] -l 0 0l 0 l - j 0 l 0 l 0 l@l 4 l 1 l 1 l 1 l 1 l@l 14 UCENSE NUMSEN 25 29 LICENSE TYPE J0 l

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? 8 60 61 DOCKET NUYBER 66 69 EVENT DATE REPORT DATE EVENT DESCRIPTION AND PROBABLE CONSEQUENCES h o 2 l A local leak rate test of the drywell-suppression chamber vent exhaust valves [

o 3 l was performed in accordance with procedure QTS 100-30. The as found leak rate from l o 4 l the volume bounded by valves A0 2-1601-23, 24, 60, 61, 62, and 63 was found to be l o s i 27.0 SCFH. THis is in excess of the 18.36 SCFH allowable limit for any one l o 6 l valve as specified by T.S. 4.7.A.2.i. l 0 7 l l 0 8 l l 7 8 9 d SY S T E *.1 CAUSE CAUSE C OY P. VA LV E COCE CODE SUBOODE COMPONENT CODE SUSCODE SUSCOCE 7

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@ REPORT l 7l 9l l-l 0l 2l 7l l/ 013 l l30 Ll l_l l0 21 22 23 24 26 27 28 29 31 32 T K N A T ON ON PLANT ET HOURS 22 SB IT FO S. SUPPL E MAN ACTLRER l33B lgl34Z l g 3

[5Z lg l36 Zjg l 0l 0l 0l 31 l lNjg 41 lYlg [ Nlg Z[ZlZlZlg 40 42 43 44 47 CAUSE DESCRIPTION AND CORRECTIVE ACTIONS 1 o l Valve A0-2-1601-23 is suspected to be leaking excessively. A supOlement report I i i l will be sent upon comOletion of all testina and when all corrective actions have I i 7 been completed. l 5 3 I _

l i 4 I l 7 3 9 80 STA S  % POWER OTHER STATUS blSCO RY . DISCOVERY DESCRIPTION R H @ l 0 l 0 l 0 lgl NA l l B l@l*

Local Leak Rate Test l A TIVITY COTTENT RELEASEO OF RELE ASE AMOUNT OF ACTIVITY LOCATION OF RELEASE 6 NA NA 1

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