ML17329A700

LER 90-015-02:on 901102,discovered Containment Type B & C Leakage Exceeded L.C.O. Value Due to Degradation of Isolation Valve Seating Surfaces.Seat Was Lapped & Disc Was replaced.W/921207 Ltr

ML17329A700
Person / Time
Site:	Cook
Issue date:	12/07/1992
From:	Blind A, Weber G INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-90-015, LER-90-15, NUDOCS 9212150109
Download: ML17329A700 (6)

Text

A.CCELERA TED DOCUiVKNTDISTRIBUTlOb SYS'1'Eb'1 REGULATORONFORMATZON DISTRIBUTION %STEM (RZDS)

ACCESSION NBR:9212150109 DOC.DATE: 92/12/07 NOTARIZED: NO DOCKET FACIL:50-315 Donald C. Cook Nuclear Power Plant, Unit 1, Indiana M 05000315 AUTH. NAME AUTHOR AFFILIATION WEBERFG.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele BLINDFA.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECIPIENT AFFILIATION 1

SUBJECT:

LER 90-015-02:on 901102,discovered containment Type B & C leakage exceeded L.C.O. value due to degradation of isolation valve seating surfaces. Seat was lapped & disc was replaced.W/921207 ltr.

DISTRIBUTION CODE: ZE22T COPIES RECEIVED:LTR ENCL SIZE:

TITLE: 50.73/50.9 Licensee Event Report (LER), Incident Rpt, etc.

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 DEANFW 1 1 INTERNAL: ACNW 2 2 AEOD/DOA 1 1 AEOD/DSP/TPAB 1 1 AEOD/ROAB/DSP 2 2 NRR/DET/EMEB 7E 1 1 NRR/DLPQ/LHFB10 1 1 NRR/DLPQ/LPEB10 1 1 NRR/DOEA/OEAB 1 1 NRR/DREP/PRPB11 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 /@SAN. LB8Dl 1 1 NRR/DST/SRXB 8E 1 1 G 02 1 1 RES/DSIR/EIB 1 1 RGN3 FILE 01 1 1 EXTERNAL: EG&G BRYCEFJ.H 2 2 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHYFG.A 1 1 NSIC POORE,W. 1 1 NUDOCS FULL TXT 1 1 YOTE TO ALL RlDS REC!!'IEiTS:

PL AS H L> US TO R DUCE 4IAS I P CONTACT THE DOCUME>NT CONTROL DES" ROOli! P!-S7 (EXT. 50'-2055) TO EL!ii!!i%ATE YOUR iAN!E FROX! D!S:

R!BU !Oi LISTS FOR DOCU4!EATS YOU DOi'7 FEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 29 ENCL 29

Indiana Michigan Power Company Cook Nuclear Plant One Cook Place Bridgman, Ml 49106 616 465 5901 INDIANA MICHI6iAN PQWER December 7, 1992 United States Nuclear Regulatory Commission Document. Control Desk Rockville, Maryland 20852 Operating Licenses DPR-58 Docket No. 50-315 Document Control Manager:

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

90-015-02 Sincerely, p.ma4 A. A. Blind Plant Manager

/sb Attachment c; D. H. Williams, Jr.

A. B. Davis, Region III E. E. Fitzpatrick P. A. Barrett R. F. Kroeger B. Walters Ft. Wayne NRC Resident Inspector W. M. Dean NRC J. G. Keppler M. R. Padgett G. Charnoff, Esq.

D. Hahn INPO S. J. Brewer B. A. Svensson 9212150109 921207 PDR ADOCK 05000315 S PDR

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kRC FORM $ 56 (665)

I'ACILITYNAME Ill MONTH

'OWER

'5%ILk~m kAME 0 2 9 OPERATING MODE 15)

LEYEL 0 0 0 9 0 5

0 LICENSEE EVENT REPORT (LER)

D. C. COOK NUCLEAR PLANT UNIT 1 CONTAINMENT TYPE B AND C LEAKAGE EXCEEDS ISOLATION VALVE SEATING SURFACES EVENT DATK (Sl DAY YKAR 20A02(5)

LKR NUMSER (61 0

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20AOS(el(1)(5) 20AOS( ~ l(I)IS)I 20AOS 4) (I II Irl 20.iOS le) Ill(r) 1 G. A. WEBER PLANT ENGINEERING SUPERINTENDENT 5 0 2 X

1 U.S. NUCLEAR REGULATORY COMMNSION SKOUKNTIAL g~x IIKVIJION MONTH NVMOER NVMOSR SOW(el RKPORT DATE 207 92 OAY THIS REPORT IS SUSMITTEO PUINUANT TO THE RLOUHIKMKNTSOF 10 CFR 20.i05 I cl 50.$ 5 (c) 12) 60.724) (2) III 60.7$ (el(2) (5) 50.724)(2) I III)

L.C.O.

ltl YEAR LICENSEE CONTACT FOR THIS LER (12)

(J:

APPROVFD OMS NO. $ )504)De EXPIRES: i/$0/52 ESTIMATED SURDEN PER RESPONSE TO COMI'LY WTH THIS INFORMATION COLLECTION REOUEST: 60.0 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGEMENT BRANCH IP6$ 0), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT ($ 150010il. OFFICE OF MANAGEMENTAND SUDGET. WASHINGTON. DC 2050$ .

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COMPLETE ONE LINE FOR EACH COMPONENT FAILURE OKSCRI ~ ED IN THIS REPORT (1$ )

~I AREA CODE 0

OTHER FACILITIES INVOLVED (5)

FACILITYNAMES 616 465 -

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This updated report is being submitted to modify a Corrective Action activity and rephrase commitments as completed activities.

With the Reactor Coolant System in Mode 5 (Cold Shutdown), the measured leakage, using the maximum pathway methodology, for the Type B and C Leak Rate Tests on Containment penetrations was 6.74 La. In addition, there were two penetrations that had leak rates that could not be quantified. This exceeded the L.C.O. value (0.60 La) of Technical Specification 3.6.1.2.b.

The measured leakage for the Weld Channel Pressurization System Valve Enclosure Manway for 1-ICM-305 (EIIS1TK/BD) was 94 percent of the measured Type B and C leak rate. 1-CS-442-1 and 1-CS-442-3 (EIISIISV/CB) are Containment, Isolation Valves for the seal water injection lines to Reactor Coolant Pump numbers ll and 13 and had unquantifiable leak rates.

deficiencies would not have resulted in any additional leakage from The three Containment, due to their function in an accident. There was no leakage into, the ICM-305 valve enclosure, therefore, The final Type B and C leak rate was 0.112 La.

it would not have been a contributor.

NRC Form $ 66 16J)9)

NRC FORM SSSA U.S. NUCLEAR REGULATORY COMMISSION APPAOVEO OMS NO. 2150010i (6411) 5 XPIR ES: 9/20/92 ES TED BUADEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVE REPORT (LER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IPS20), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. OC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT )2150010J). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20502.

FACILITY NAME III DOCKET NVMSER 12) LER NUMBER (5) PAGE IS)

YEAR SEQUENTIAL FEVISION NUM 5 II NUM EN D. C. COOK NUCLEAR PLANT UNIT 1 0 5 0 0 0 0 02 02 OF O 3 g 5 9 p 1 5 TEXT /////XNP APMP /1 /Pqvira/, wr Oda5aW A'RC F/rm 20549/ I)7)

This updated report is being submitted to modify a Corrective Action activity and rephrase commitments as completed activities.

Conditions Prior to Occurrence:

Unit-1 in Mode 5 (Cold Shutdown)

Descri tion of Event:

With the Reactor Coolant System in Mode 5 (Cold Shutdown), the measured leakage, using the maximum pathway methodology, for the Type B and C Leak Rate Tests on Containment penetrations was 6.74 La. In addition, there were two penetrations that had leak rates that could not be quantified. This exceeded ~

the L.C.O. value (0.60 La) of Technical. Specification 3.6.1.2.b.

The measured leak rate for the Weld Channel Pressurization System Valve Enclosure Manway for 1-ICM-305 (EIIS:TK/BD) was 94 percent of the measured Type B and C leak rate. Check, valves 1-CS-442-1 and 1-CS-442-3 (EIIS:ISV/CB),

are Containment Isolation Valves for the seal water injection lines to Reactor Coolant Pump numbers 11 and 13, respectively. These valves had leak rates that could not be quantified.

Cause:

The excessive leak rate of 1-ICM-305 valve enclosure manway cover was caused by a portion (approximately 12 inches) of the 0-ring being out of its channel.

This manway cover was last installed on June 10, 1989. On June 11, 1989 the valve enclosure was tested and had a leak rate of 1500 sccm. When the manway cover was removed, a portion of the 0-ring fell off. The corresponding portion of the 0-ring groove was full of grease that is used to lubricate the 0-ring. It is believed that a portion of the 0-ring was pulled out of its channel when the manway cover was aligned for bolting. The grease used on the O-ring, when it was installed, must have provided a sufficient seal to allow the valve enclosure to pass the June ll, 1989 Leak Rate Test, but degraded since then.

The excessive leak rates for 1-CS-442-1 and 1-CS-442-,3 are attributed to pieces of'neoprene found in the valves. A small piece was found in 1-CS-442-1 and a piece approximately 0.19 inches in diameter was found in 1-CS-442-3.

These check valves are downstream of the seal water filters. The neoprene pieces found in 1-CS-442-1 and 1-CS-442-3 are believed to have come from the seal water filters and either broke off during filter replacement, or were trapped by the filter and fell off during removal. This is the first time we

,have experienced such problems.

NRC Form SSSA 1549)

NRC FORM 3SSA U.S. NUCLEAR REGULATORY COMMISSION APPROVED OMS NO. 31500(OO (SJ)9)

EKPIRES: o/30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REOUESTI 500 HRS. FORWARD COMMENTS REGARDING SURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT SRANCH (F430). U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON. DC 20555, AND TO 1HE PAPERWORK REDUCTION PROJECT (3150010(). OFFICE OF MANAGEMENTAND SUDGET,WASHINGTON, DC 20503.

FACILITY NAME (I) DOCKET NUMSER (2) LER NUMSER (5) PAGE C))

YE A II Il SEQVENTIAL ~ala IIEVIS ION NVM 5 II ..II) NVM D. C. COOK NUCLEAR PLANT UNIT 1 0 5 0 0 0 3 I 5 9 0 0 1 5 0 2 O3oFO TEXT lllmore g>>oe lr ooII)klaf, ooo ~ JJSrmol HAC Form 30rJI SJ (12)

Corrective Actions:

The 1-ZCM-305 valve enclosure manway cover 0-ring was replaced. A retest indicated no leakage (000 sccm). No further action is planned since the 0-ring installation problem was an isolated event and not an indication of a generic problem or practice.

Valve 1-CS-442-1 had the small piece of neoprene removed. The seat was lapped and the disc was replaced. The as-left leak rate was 40 sccm. Valve 1-CS-442-3 had the piece of neoprene removed. .No additional repair actions were necessary. The as-left leak rate was 65 sccm. The Maintenance procedure used for replacement of the Seal Water Injection Filters was reviewed ensuring that adequate steps are in place to prevent the introduction of material to the system during seal replacement. In addition, the filter seal degradation was reviewed with the manufacturer. The filter seal failure is considered to I be an isolated case since the manufacturing process for the filter did not change the filter composition or configuration.

Other Containment Isolation Valves that exhibited leak rates in excess of the guideline acceptance criteria were repaired and retested to ensure the leak rates were within allowable limits., The final as-left Type B and C leak rate was 0.112 La.

~Anal sls:

The ZCM-305 valve enclosure is outside of Containment and provides a Containment barrier for any leakage from ICM-305, the Containment Sump Recirculation Isolation Valve. This valve enclosure is not part of the Containment Zntegrated Leak Rate Test boundary. Therefore, an addi.tional leak into the enclosure would have to have developed for radioactivity to escape out of Containment. No other leaks were discovered in this valve enclosure during the as-found testing or during the as-left testing. Therefore, if Containment had been pressurized and radioactivity had been released into the Containment atmosphere, the radioactivity would have been contained within Containment. The public health and safety was never threatened by this condition.

Valves 1-CS-442-1 and 1-CS-442-3 are located in the Reactor coolant Pump Seal Water Injection System. Both of these valves were found wi.th foreign material in the seat area and were unable to fully close. The leakage from these valves could not be quantified. These check valves would not be required to isolate Containment during an accident scenario as the Reactor Coolant. Pump Seal Water Injection System remains in service following a desi.gn basi.s accident, preventing a flow path from Containment to the outside atmosphere.

Zf a postulated break in either of these lines occurred, Containment integrity would be maintained by the two check valves located in each line inside Containment, and downstream of 1-CS-442-1 and 1-CS-442-3.

This event has been determined to be reportable under 10 CFR 50s73 (a)(2)(i)(C).

Based on the above, however, it has also been determined that this condition did not create a significant sa'fety concern. The as-found condition of the Containment would not have put the plant in an unsafe condition.

NRC Form 35SA (589)

l(ROFORM 388A (LS. NUCLEAR REGULATORY COMMISSION (889)

LICENSEE EVENT REPORT (LER)

O APPROVED OMB NO. 31504104 EXPIRES; e(30I92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUESTI 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION 'AND REPORTS MANAGEMENT BRANCH (P 530). U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO 1HE PAPERWORK REDUCTION PROSECT (31500)08). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503, FACILITY NAME ()) DOCKET NUMBER (2) LER NUMBER (8) PAGE (3)

YEAR;:(@ SEOUENTIAL 48VISIOrr HUM EA rI v M ee II COOK NUCLEAR PLANT UNIT 1 D. C. 0 5 0 0 0 3 g 5 9 0 OIlS 0 204 0F0 TEXT N more opec>> le requeed, vee odd(9mel htRC Frvm 388A'el I(2)

Failed Com onent Identifications Component nameS 1-ICM-305 Valve Enclosure Plant I.D. No.s 1-TK-84 (EIISSTK/BD)

Manufacturer: Unknown Model No.s Unknown Component name: Reactor Coolant Pump No. 11 Seal Water Injection Containment Isolation Valve Plant I.D. No. 3 1-CS.-442-1 (EIISSISV/CB)

Manufacturer: Conval, Inc.

Model No.s 12C2 Component name: Reactor Coolant Pump No. 13 Seal Water Injection Containment Isolation Valve Plant I.D. No.s 1-CS-442-3 (EIISSISV/CB)

Manufacturers Conval, Inc.

Model No.s 12C2 Previous Similar Events:

Previous Licensee Event Reports submitted for excessive type BGC Leak Rate Test results include:

050-315/79-34 050-316/79-20 050-315/81-11 050-316/79-53 050-315/81-25 050-316/81-18 050-315/82-58 050-316/83-16 050-315/83-72 050-316/84-05 050-315/85-17 050-316/86-09 050-315/87-12 050-316/89-05 050-315/89-04 050-316/90-07 NRC Form 388A (849)