ML17329A415

LER 92-002-00:on 890124,containment Isolation Valves Not Repaired When ASME Section XI Leakrate Acceptance Criteria Exceeded.Caused by Procedure Requirements Not Being Included in Test Procedure.Packing replaced.W/920313 Ltr

ML17329A415
Person / Time
Site:	Cook
Issue date:	03/13/1992
From:	Blind A, Weber G INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LER-92-002-01, LER-92-2-1, NUDOCS 9203200062
Download: ML17329A415 (6)

Text

ACCELERATED' TRIBUTION DEMONS~TION SYSTEM REGULATO Y INFORMATION DISTRIBUTION SYSTEM (RIDS)

ACCESSION NBR:9203200062 DOC.DATE: 92/03/13 NOTARIZED: NO DOCKET ¹ FACIL:50-316 Donald C. Cook Nuclear Power Plant, Unit 2, Indiana M 05000316

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AUTH. NAME AUTHOR AFFILIATiON WEBERgG.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele BLIND,A.A. Indiana Michigan Power Co. (formerly Indiana & Michigan Ele RECIP.NAME RECiPIENT AFFILIATION SUB ECT: LER 92-002-00:on 890124,containment isolation valves not repaired when ASME section XI leakrate acceptance criteria exceeded. Caused by procedure requirements not being included in test procedure. Packing replaced.W/920313 ltr.

DISTRIBUTION CODE". IE22T COPIES RECEIVED:LTR t ENCL TITLE: 50.73/50.9 Licensee Event Report (LER), IncidentQ Rpt, etc.

SIZE:

NOTES:

RECIPIENT COPIES RECIPIENT COPIES ZD CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD3-1 LA 1 1 PD3-1 PD 1 1 STANG,J 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/PRPBll 2 2 NRR/DST/SELB 8D 1 1 NRR/DST/SICB8H3 1 1 NRR/DST SPLB8D1 1 1 NRR/DST/SRXB 8E 1 1 RE

'-- 02 1 1 RES/DS1R/EIB 1 1 ,RQN3 - ~

1 1 EXTERNAL: EG&G BRYCEiJ.H 3 3 L ST LOBBY WARD 1 1 NRC PDR 1 1 NSIC MURPHYiG.A 1 1 NSIC POOREgW ~ 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIPIBNIS:

PLEASE HELP US TO REDUCE WASTE( CONTACT THE DOCUMENT CONTROL DESK ROOM Pl-37 (EXT. 20079) TO ELIMINATEYOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 30 ENCL 30 g0 t

Indiana thichi'.:

Power Cornpar",

Cook Mrrciear Pin.',

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616 465 59ir IRQMNfi'tCHtGAM'OVd'M'nited March 13, 1992 States Nuclear Regulatory Commission Document Control Desk Rockville, Maryland 20852 Operating Licenses DPR-74 Docket No. 50-316 Document Control Manager:

In accordance with the criteria established by 10 CFR 50.59 entitled Licensee Event Re ort S stem the following report is being submitted:

k 92-002-00 Sincerely, A. A. Blind Plant Manager

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

A. B. Davf.s, Region E.,E. Fitzpatrick III P. 'A. Barrett B. F. Henderson R. F. Kroeger B. Walters Ft. Wayne NRC Resident Inspector T. Colburn - NRC J. G; Keppler M. R. Padgett G. Charnoff, Esq.

D. Hahn INPO S. J. Brewer/B. P. Lauzau B. A. Svensson 9203200062 920313 PDR ADOCK 05000316 8 PDR

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION (64)9) APPROVED OMB NOi 31500104 EXPIRES: 4(30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS INFORMATION COLLECTION REOUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS AND REPORTS MANAGE'MENT BRANCH IP 533), U.S. NUCLEAR REGULATORY COMMISSION. WASHINGTON, OC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2I PA 6 3 D. C. COOK NUCLEAR PLANT UNIT 2 0 5 0 0 0 3 1 6 1 OF 0 4

ONTAINMENT ISOLATION VALVES NOT REPAIRED WHEN ASHE SECTION XI LEAKRATE ACCEPTANCE CRITERIA EXCEEDED EVENT DATE (SI LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED IS)

MONTH DAY YEAR YEAR gyLSr SEOUENTIAL;.?gy REVIBIO I MONTH DAY YEAR FACILITYNAMES DOCKET )IUMBER(S)

NUMBER NUMBER 0 5 0 0 0 r

01 24 8 9 9 2 0 0 2 .00 03 13 9 2 0 5 0 0 0 OPERATING THI6 REPORT IS SUBMITTED PURSUANT T 0 THE REQUIREMENTS OF 10 CFR (I: (Cheek one or more ol the lollovninPl (11 MODE (9) 20.402(B) 73.71(BI 20AOS(cl 50,73(e) (2)(ivl 20,405( ~ l(1)(i) 50.36(el(1) 60.73(e ) (2) (v) 73.71(cl

"(I"0)' 0 0 20.405 I

( ~ ) ( ) (ii) 50.36(cl(2) 50,73(e) (2)(vill OTHER (Speclly in AottreCt Oelovl end ln Text, HRC /orm P 20.40S( ~ )(l)(BII 60.73( ~ )(2)(i) 60,73(e)(2)(viii)(AI 366A)

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20.405(e ) I I ) (iv) 60.73( ~ I (2) (ii) 50.73(el(2)(viiil(BI 6 20,405( ~ ) (1)(v) 60.73( ~ l(2)(iiil 50.73( ~ l(2)(xl LICENSEE CONTACT FOR THIS LER (12I NAME TELEPHONE NUMBER AREA CODE G. A. WEBER PLANT ENGINEERING SUPERINTENDENT 6 16 46 5-5 901 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

MANUF AC. REPORTABLE MANUFAC.

CAUSE SYSTELI COMPONENT TO NPRDS SYSTEM COMPONENT TURER TURER ISV DO 20 B ISV DO 20 SUPPLEMENTAL REPORT EXPECTED (14) MONTH OAY YEAR EXPECTED .

SUBMISSION DATE (15)

YES ill yn, COmplete EXPECTED S(iB4IISSIOht DATE/ X NO ABSTRACT (Limit tO le00 tpetet, le., epprOximete!y Alteen tipple IpeCe typeWritten lintel (15)

During a review of IST Valve Program Test Data, the ISI Engineer discovered that the Boron Injection Tank Outlet Valves, 2-ICM-250 and 2-ICM-251, had- been returned to service following the 1989 Unit 2 Steam Generator Replacement Outage with seat leakage in excess of the established ASME Section XI leakage limits. The valves remained in service until August, 1990 when the Unit was shut down for refueling. During the outage, the valves were retested. Valve 2-ICM-250 was found to meet the leakage limit, while 2-ICM-251 was found to significantly exceed the limit. Subsequently, stem packing was replaced in both" valves. Both valves were then retested and found to be within leakage limits.

Allowing the valves to be returned to service following the 1989 outage, while being outside Section XI limits, was due to a procedural deficiency affecting,Section XI valves which were tested under the Containment Local Leak Rate Test (LLRT) procedure. Until this condition was discovered, the procedure only included acceptance criteria for the LLRT, which, in certain circumstancesl can be less restrictive than Section XI limits. The Unit 2 LLRT procedure has

, been revised to include the Section XI limits; the Unit 1 procedure is scheduled for revision prior to its next use.

NRC F or m 366 (64)9)

NRC FORM 366A U.S. NUCLEAR REGULATORY COICMISSION (64)9) APPROVED 0MB NO. 31504104 EXPIRES: 4l30l92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (31500104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITYNAME (I) DOCKET NUMBER (2) I.E R NUMBE R (6) PAGE IS)

YEAR @(rI SCQVCNT<AL NVMSER Qgg REVISION

'A: NVMBER D. C. COOK NUCLEAR PLANT UNIT 2 0 5 0 0 o 3). 6 92' 002 0 0 0 2 OF 0 4 TEXT II!more Secre le reorrirecd6 ore eddiaionel ill(CForm 366A3) (12)

Conditions Prior to Occurrence Unit-2 in Mode 5 (Cold Shutdown)'escri tion of Event During a review of the IST Valve Program Test Data, the ZSI Engineer discovered that Valves 2-ICM-250 (EIIS/ISV-JM) and 2-ZCM-251'EIIS/ZSV-JM) had been returned to service following the Unit 2 1989 Steam Generator Replacement Outage, with seat leakage in excess of the ASME Section XZ permissible leakage value of 600 sccm, as identified in the Second 10-Year Interval Long-Term Inservice Examination and Testing Plan for Class 1, 2, and 3 Systems and Components for D.C. Cook Plant, Unit 2. Valves 2-ICM-250 and 2-ICM-251 are the Boron Injection Tank Outlet Valves. Valve 2-ICM-250 was returned to service with a leakrate of 806.8 sccm and 2-ICM-251 was returned to service with a leakrate of 704.7 sccm.

The valves were in service from January of 1989 until they were retested during the next Refueling Outage in August of 1990. The August, 1990 tests revealed that 2-ZCM-250 had a leakrate of 289.4- sccm and .2-ICM-251 had a leakrate of 9466.9 sccm. Following replacement of the valve packing, 2-ICM-250 was returned to service with a 119.3 sccm leakrate. Following the replacement of the valve stem packing and torque switch adjustment of the motor operator, 2-1CM-251 was returned to service with a leakrate of 163.5 sccm. Valves ICM-250 and 251 are double disc gate valves. The leakrate is measured by pressurizing the valve body between the seats. Any valve stem packing leakage is included in the leak rate measurement.

A review of the test data prior to the 1989 tests revealed that acceptable test results were obtained in June, 1988. Valve 2-ICM-250 had a leakrate of 219.0 sccm and 2-ICM-251 had a leakrate of 239.3 sccm. Between the 1988 and the 1989 tests, the unit was in a Steam Generator Replacement Outage.

The ISI Engineer was in the process of assuming the ZST Valve Program duties, for the valves in the Containment Local Leakrate (LLRT) Test Procedure and identified this discrepancy while entering the test data into the IST Valve Program. Credit was taken for the Section XI reviews as being part of the LLRT, without specific Section XZ acceptance criteria in the procedure. The LLRT Procedure is identified on-site as the Type B and C Leakrate Test (1 and 2 EHP 4030 STP.203) and fulfills the requirements of 10CFR Appendix-J and Technical Specifications Type B and C Testing.

NRC Form 366A (669)

NRC FORM 366A V.S. NUCLEAR REGULATORY COMMISSION (649) APPROVED OMB NO. 31500104 I EXP I A ES'( E/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT (LER) INFORMATION COLLECTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH IP 530). V.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555. AND TO THE PAPERWOAK REDUCTION PAO)ECT (3(50010E), OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (I) DOCKET NUMBER (2I LER NUMBER (6) PAGE I3)

YEAR SEOUENTIAL ..>Kg REVISION NUMSER NUMBER D.C. COOK NUCLEAR PLANT UNIT 2 0 5 0 0 0 3 1 6 9 2 0 0 2 0 0 0 3 OF 0 4 TEXT /// IIRYP gw EE i) /Eqv/IEd, vEE EddIS'one/ NRC FomI 366A3/ ((7)

Cause of Event At the time of this event, specific procedural requirements were not included in the appropriate test proceduze which would detect that the ASME Section XI Gui.deline and Maximum Permissible Leakrate Limits were not exceeded. The Type B and C Leakrate Test Procedures (1 and 2 THP 4030 STP.203) did not fully implement the requirement for analysis of leakage rates and the corrective action requirements of Paragraphs IWV-3426 and 3427(a). The B and C Test Engineer's primary concern was fulfillingTechnical Speci.fication and 10CFR50, Appendix-J requirements. An independent review of the test data was not performed for Section XI purposes. The Test Engi.neer was also required to perform the Section XI reviews for these particular valves.

The excessive leak rates for 2-ICM-250 and 251 is attributed to packing leakage. Following replacement of the packing, acceptable leakrates were obtained.

nal sis of Event Valves 2-ICM-250 and 2-ZCM-251 have two safety functions. On a safety injection signal the valves open to allow the charging pumps to inject make-up water to the Reactor Coolant System (RCS).. The valves may also be zequired to be closed to establish Containment Integrity. When used for Containment Isolation, the maximum seat leakage is determined by the Containment Local Leakrate Test (LLRT), which is conducted to satisfy 10CFR50 Appendi.x-J Type C Testi.ng. The leakrate for all penetrations and valves subject to the Type =B and C Leakrate Test is required to be less than 0.6 La. The total Type B and C leakrate was 0.076 La, well below the 0.6 La limi.t, despite the leakage contribution of these valves.

This event was determined to be reportable per the requirements of 10CFR50.73, Paragraph (a)(2)(i)B as a condition prohibited by the Plant's Technical Specifications. The condition did not represent a significant threat to public safety since these valves were capable of performing their safety functions.

Corrective Actions The valve stem packing was replaced on both 2-ICM-250 and 251.

The Type B and C Leakrate Test Procedures (1 and 2 THP 4030 STP.203) are being revised to ensuze that the ASME Section XI reviews are completed prior to entering Mode 4, when Containment integzity and is required by Technical Specifications. 2 THP 4030 STP.203 has been revised 1 THP 4030 STP.203 is scheduled to be revi.sed by June 30, 1992, prior to its next use.

The ZSZ Program Engineer now performs an independent review of the Type C Test Data for those valves in the IST Valve Program. This review ensures that the Section XZ maximum permissible criteria and leakrate trending activities are satisfactorily completed. This duty was previously performed by the B and C Test Engineer whose primary concezn was fulfilli.ng Techni.cal Speci,fication and 10CFR50, Appendix-J requi.rements.

NAC FRIIn 366A (6$ 9)

NRC FORM 366A US. NUCLEAR REGULATORY COIAMISSION (64)9) APPROVED OMS NO. 31500')04 EXP IR ES: 4/30/92 ESTIMATED BURDEN PER RESPONSE TO COMPLY WTH THIS LICENSEE EVENT REPORT ILER) INFORMATION COLLECTION REOUESTI 504) HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE RECORDS TEXT CONTINUATION AND REPORTS MANAGEMENT BRANCH (P4)30), U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON, DC 20555, AND TO THE PAPERWORK REDUCTION PROJECT (3(504)104). OFFICE OF MANAGEMENTAND BUDGET, WASHINGTON, DC 20503.

FACILITY NAME (11 DOCKET NUMBER (21 LER NUMBER (6) PAGE (3)

YEAR gC>h SEOUENTIAL h2R REVISION NUMSER NUMSSR D. C. COOK NUCLEAR PLANT UNIT 2 p 5 p p p 3 1 6 9 2 0 0 2 0 0 040FO 4 TEXTJ//mors spsss is rsr/uh/sC uss sCI/I'Corhs/iVRC Form 366A's/ (17)

Failed Com o ent Identification Component Name: Boron Injection Tank Train-A Outlet Containment Isolation Valve Component ZDE 2-ZCM-250 Manufacturer: Anchor/Darling Valve Co.

Model No.: S 350 WDD Component Name: Boron Injection Tank Train-B Outlet Containment Isolation Valve Component ID: 2-ZCM-251 Manufacturer: Anchor/Darling Valve Co.

Model No.s S 350 WDD Previous Similar Events This is the first event of this type.

NRC Form 356A (64)9)