ML17325B182

LER 89-005-00:on 890221,containment Type B & C Leak Rate Exceeded Limiting Condition for Operation Using Min Pathway Calculation Method,Causing Excessive Leakage in Several Valves.Valves Repaired & retested.W/890323 Ltr

ML17325B182
Person / Time
Site:	Cook
Issue date:	03/23/1989
From:	Postlewait T, Will Smith INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-89-005, LER-89-5, NUDOCS 8904030243
Download: ML17325B182 (7)

Text

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ACCESSION NBR:8904030243 DOC.DATE: 89/03/23 NOTARIZED: NO DOCKET FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana & 05000316 AUTH. NAME AUTHOR AFFILIATION POSTLEWAIT,T.K. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele SMITH,W.G. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 89-005-00:on 890221,containment type B&C leak rate exceeds LCO value due to excessive valve leakage.

W/8 DISTRIBUTION CODE: IE22D COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73 Licensee Event Report (LER), Incident Rpt, etc.

NOTES RECIPIENT COPIES RECIPIENT COPIES A.

ID CODE/NAME LTTR ENCL ID CODE/NAME L'TTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 D

STANG,J 1 1 INTERNAL: ACRS MICHELSON 1 1 ACRS MOELLER 2 2 D ACRS WYLIE 1 1 AEOD/DOA 1 AEOD/DS P/TPAB 1 1 AEOD/ROAB/DS P 2 2 DEDRO 1 1 IRM/DCTS/DAB 1 1 NRR/DEST/ADE 8H 1 1 NRR/DEST/ADS 7E 1 0 NRR/DEST/CEB 8H 1 1 NRR/DEST/ESB 8D 1 1 NRR/DEST/ICSB 7 1 1 NRR/DE ST/MEB 9H 1 1 NRR/DEST/MTB 9H 1 1 NRR/DEST/PSB 8D 1 1 NRR/DEST/RSB 8E 1 1 NRR/DEST/SGB 8D 1 1 NRR/DLPQ/HFB 10 1 1 NRR/DLPQ/QAB 10 1 1 NRR/DOEA/EAB 11 1 1 NRR/DREP/RPB 10 2 2 NRR DRIS SIB 9A 1 1 NUDOCS-ABSTRACT 1 1 R FILE 1 1 RES/DSIR/EIB 1 1 fK~Y B 1 1 RGN3 FILE 01 1 1 R

EXTERNAL: EG&G WILLIAMS,S 4 4 FORD BLDG HOY,A 1 1 H ST LOBBY WARD 1 1 LPDR 1 1 NRC PDR 1 1 NSIC MAYS,G 1 1 I

NSIC MURPHYiG A ~ 1 1 D

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TOTAL NUMBER OF COPIES REQUIRED: LTTR 44 ENCL 43

NRC Form 388 U.S. NUCLEAR REGULATORY COMMISSION (9PLTI APPROVED OMB NO. 3)604104 LICENSEE EVENT REPORT (LER) EXPIRES: 8/31/88 FACILITY NAME (I) DOCKET NUMBER (2) PA E 3 D. C. Cook Plant, Unit 2 o 5 o o o316 soF04 Containment Type BKC Leak Rate Exceeds L. C. 0. Value Due To Excessive Valve Leakage EVENT DATE IS) LFR NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (8)

MONTH DAY YEAR YEAR Ore SEOVENTIAL NVMSER I ;. REV6)ON NVMSER MONTH DAY YEAR FACILITYNAMES DOCKET NUMBER(S) 0 5 0 0 0 0 2 2 1 8 9 8 9 0 0 5 0 0 0 3 238 9 0 5 0 0 0 THIS REPORT IS SUBMITTED PURSUANT T0 THE REOVIREMENTS OF 10 CFR (): /Check onr or morr of tnr /olloWlnp/ (11 OPERATING MODE (9) 20.402(ir) 20.40S(e) 60.73(el(2)(ivl 73.71()r)

POWER 20.406( ~ )(1)(ll 60.38(c) (1) 60.73(el(2((vl 73.71(el LEYEL 0 0 0 (10) 20.405( ~ l(1)(ill 50.38 I e) (21 50.73(e)(2) (vii) OTHER ISprclfy In AOI)rect

/rr/ovrenrfln Text, HRC Form 20,406(e l (I)(ill) x 60.73(e) (2)(i) 60,73(e) (2) (viil)(A) 366AI 20A06( ~ ) (10)vl 50.73(e) (2) (ii) 60.73( ~ l(2)(viSI(B) 20.405(e)(1)(v) 50,73(el(2)(lill 60.73( ~ ) (2) (x I LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER T. K. Postlewait- AREA CODE Technical Engineering Superintendent 616 4 65 - 5901 COMPLETE ONE LINE FOA EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAVSE SYSTEM COMPONENT MANUFAC.

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TURER REPORTABLE TO NPADS I%I!M B BD I SV I2 07 Y  %%I4 LWN>

B BD I SV C4 18 SVP)'LEMENTAL REPORT EXPECTED (1 ~ ) MONTH DAY YEAR EXPECTED SUBMISSION DATE I)S)

YES Ilf yrL complete EXPECTED SIIBMISSION DATE/

ABSTRACT ILlmlt to /AX) rpecrt, I r.. rpproximrtrly /i/rem tinplr Iprce typevvrlnrn linrr/ (18)

In performing the Type BSC Contair)ment Penetration Leak Rate Tests and Contair)ment Integrated Leak Rate Test, during the 1988/1989 outage, the B and C "Penalty" applied to the total leakage was arrived at using the "minimum" pathway calculation method. On February 21, 1989, the and C leak rate was recalculated using the "maximum" pathway

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methodology.

B This resulted in an as-found leak rate of 0.61 La, which exceeds the Technical Specification 3.6.1.2.b limit of 0.60 La.

The excessive leakage of several valves subject to the Type B&C test is the cause for exceeding Technical Specification 3.6.1.2.b limit of 0.60 La. Valves that exhibited excessive leakage were repaired and retested.

The total as-left B and C leakage was calculated to be 0.076 La, which is far below the 0.60 La limit required by Technical Specification 3.6.1.2.b.

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r.vm ososs ss m TCXT (/ ssmrm ~ Is ayvA4 ~ IlNShmsss'ANC Asms ~'ll1171 o s o o o31 6 8 9 005 << 0 2 0 4 Conditions Prior to Occurrence Unit 2 in h'tode 5 (Cold Shutdown) during Steam Generator Replacement Outage.

Descri tion of Event In performing the Type B&C Containment Penetration Leak Rate Tests and Containment Integrated Leak Rate Test, during the 1988/1989 outage, the B and C "Penalty" applied to the total leakage was arrived "minimum" pathway calculation method. On February 21, at using the 1989, the Type B and C leak rate was recalculated using the "maximum" pathway methodology.

This resulted in an as-found leak rate of 0.61 La, which exceeds the Technical Specification 3.6.1.2.b limit of 0.60 La.

Cause of Event The excessive leakage of several valves subject to is the cause for exceeding Technical Specification the Type B and C test 3.6.1.2.b limit,'f 0.60 la. The major contributors to the excessive leak rate. were valves 2-SF-160 (EIIS:ISV/BD) and 2-VCR-205 (EIIS:ISV/BD) . 2-SF-160 is the containment isolation valve from the Reactor Coolant Drain Tank to the Refueling h'ater Purification Pumps. 2-VCR-205 is the containment Isolation Valve for Upper Containment, Purge Air. The leak rates of valves 2-SF-160 and 2-VCR-205 represented Supply 89 percent of the reported B and C leakage.

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t&4ll UCENSEE EVENT REPORT (LER) TEXT CONTlNUATlON AffhOVlO OQN NO. 1I ~04 KXPNCS: NISI /%

fACILITYNAllt Ill OOCKtT HNOEh QI Lth NUM44h IO fAO4 Q 0(4IJCNTML D. C. Cook Plant, Unit 2 YIAh NUM II o oooo 316 89 005 0 0 0 3 0 0 TTXTIy~4INN4~~~WeansmaWIITI Anal sis of Event The containment is designed to have two valves in series for each penetration that is tested during the Appendix J Type B&C Test Program.

During the Type C Test, there was not a case in which both of the in-series isolation valves for a given. penetration exhibited excessive leakage. The Containment Integrated Leak Rate was below 0.75 La (The limit specified in Technical Specification 3.6.1.2) -and takes into account the type B and C leakage rates. Therefore, the Containment would have performed its design function and this event is not of safety significance.

In conclusion, containment integrity was not caapranised and this event did not represent a significant hazard to public health and safety.

Corrective Actions 2-VCR-205 was leaking at seat area near the stem. The clampseal bolts were tightened, eliminating the leakage.

2-SF-160 was repaired by replacinq the diaphragm.

Valves 2-VCR-205 and 2-SF-160 were the major contributors for exceeding the BSC allowable leak rate limit.

Other valves that exhibited excessive leakage were repaired and retested.

The total as-left B and C leakage was calculated to be 0.076 La, which is far below the 0.60 La limit required by Technical Specification 3.6.1.2.b.

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NIIC Fono IQEEI

~ LICENSEE EVENT REPORT ILERI TEXT CONTINUATION UA. NUCLEAN IIECULATOIIYCOOMIIEOOH AFFllOVEO OME NO. $ 1~104 EXFIAES: EIEI/N FACILITY NAME III OOCKET NUINEII IEI LEII NUMEEII IEI ~ ACE IEI D. C. Cook Plant, Unit 2 YEAN ~ EECUENTIAL @5 rruu o >> rrvu ISIOrr Eo TEXT 8'ore Nrore N eeeroferE rroe ~ NIIC hvm 454'ol IITI o s o o o 3 1 6 89 005 00 04oF 0 4 Failed C nent Identification Plant I.D. No: 2-VCR-205 (EIIS: ISV/BD)

'anufacturer: Clow Corporation (C418)

Canponent Name: Upper Containment Purge Supply Containment Isolation Valve Model No: S Plant I.D. No: 2<<SF-160 (EIIS: ISV/BD)

Manufacturer: ITT Grinnell (I207)

Canponent Name: Reactor Coolant Drain Tank Isolation to Refueling Water Purification Pump Model No: MKF-5-SS Previous Similar Events Previous Licensee event reports subnitted for excessive type B&C Leak Rate Test results include:

050-315/79-34 050-316/79-20 050-315/81-11 050-316/79-53 050-315/81-25 050-316/81-18 050-315/82-58 050-316/83-16 050-315/83-72 050-316/84-05 050-315/85-17 050-316/86-09 050-315/87-12 NIIC FIIIIu OooA IE4EI

Indiana Michigan Power Company Cook t4uclear

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P.O. Box 458 Bridgman. Ml 49106 616 465 5901 GNARL INDIANA NICHI POWER March 23, 1989 United States Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 Operating License DPR-74 Docket No. 50-316 Document Control Manager:

In accordance with the criteria established by 10 CFR 50.73 entitled Licensee Event Re ortin S stem, the following report is being submitted:

89-005-00 Sincerely, W. G.

g f'

lpDl Smith, i8-Jr.

Plant Manager WGS:clw Attachment cc: D. H. Williams, Jr.

A. B. Davis, Region III M. P. Alexich P. A.'arrett J. E. Borggren R. F. Eroeger NRC Resident Inspector Wayne Scott, NRC R. C. Callen G. Charnoff, Esq.

Dottie Sherman, ANI Library D. Hahn INPO PNSRC A. A. Blind S. J. Brewer/B. P. Lauzau