ML18012A438

LER 96-021-00:on 961107,inadequate post-maint Testing Following Repairs on Containment Isolation Valve 1SP-208 Identified.Caused by Misinterpretation of ASME Code. Procedures revised.W/961209 Ltr

ML18012A438
Person / Time
Site:	Harris
Issue date:	12/09/1996
From:	Donahue J, Verrilli M CAROLINA POWER & LIGHT CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
HNP-96-204, LER-96-021, LER-96-21, NUDOCS 9612160233
Download: ML18012A438 (7)

Text

CATEGORY j REGUZATO INFORMATION DISTRIBUTION aTEM (RIDS) 1

~ ACCESSION NBR:9612160233 DOC.DATE: 96/12/09 NOTARIZED: NO DOCKET FACIL:50-400 Shearon Harris Nuclear Power Plant, Unit 1, Carolina 05000400 AUTH. NAME AUTHOR AFFILIATION VERRILLI,M. Carolina Power & Light Co.

DONAHUE,J.W. Carolina Power a Light Co.

RECIP.NAME RECIPIENT AFFILIATION

SUBJECT:

LER 96-021-00:on 961107,mode 1 at 100% power, containment isolation valve 1SP-208 was declared operable. Caused by incorrect interpretation of ASME.Operations personnel will be trained a completed by 970330.W/961209 ltr.

DISTRIBUTION CODE: IE22T COPIES RECEIVED:LTR ENCL SIZE:

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

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NOTES:Application for permit renewal filed. 0500040$

RECIPIENT COPIES RECIPIENT COPIES ID CODE/NAME LTTR ENCL ID CODE/NAME LTTR ENCL PD2-1 PD 1 1 LE,N 1 1 INTERNAL: ACRS 1 1 PD AB 2 2 AEOD/SPD/RRAB 2 2 LE CE 1 1 NRR/DE/ECGB 1 1 NRR DE/EELB 1 1 NRR/DE/EMEB 1 1 NRR/DRCH/HHFB 1 1 NRR/DRCH/HICB 1 1 NRR/DRCH/HOLB 1 1 NRR/DRCH/HQMB 1 1 NRR/DRPM/PECB 1 1 NRR/DSSA/SPLB 1 1 NRR/DSSA/SRXB 1 D

RES/DET/EIB 1 1 RGN2 FILE 01 1 1 EXTERNAL: L ST LOBBY WARD 1 1 LITCO BRYCEgJ H 1 1 0

I NOAC MURPHY,G.A 1 1 NOAC POOREiW. 1 1 C NRC PDR 1 1 NUDOCS FULL TXT 1 1 NOTE TO ALL "RIDS" RECIPIENTS:

PLEASE HELP US TO REDUCE WASTE! CONTACT THE DOCUMENT CONTROL DESK, ROOM OWFN SD-5(EXT. 415-2083) TO ELIMINATE YOUR NAME FROM DISTRIBUTION LISTS FOR DOCUMENTS YOU DON'T NEED!

FULL TEXT CONVERSION REQUIRED TOTAL NUMBER OF COPIES REQUIRED: LTTR 26 ENCL 26 Aof

Carolina Power & Light Company Harris Nuclear Plant

, POBox165 New Hill NC 27562 DEC 9 1996 U.S. Nuclear Regulatory Commission Serial: HNP-96-204 ATTN: NRC Document Control Desk 10CFR50.73 Washington, DC 20555 SHEARON HARRIS NUCLEAR POWER PLANT UNIT 1 DOCKET NO. 50-400 LICENSE NO. NPF-63 LICENSEE EVENT REPORT 96-021-00 Sir or Madam:

In accordance with Title 10 to the Code of Federal Regulations, the enclosed Licensee Event Report is submitted. This report describes a condition involving inadequate post maintenance testing following repairs on a containment isolation valve.

Sincerely, J. W. Donahue Director of Site Operations Harris Plant MV Enclosure c: Mr. J. B. Brady (HNP Senior NRC Resident)

Mr. S. D. Ebneter (NRC Regional Administrator, Region II)

Mr. N. B. Le (NRC - NRR Project Manager) 96i2i60233 9bi209 05000400 PDR ADOCK 8 PDR

) QQ() U<~

State Road 1134 New Hill NC

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB No. 3150 0104 F405) EXPIRES 04/30/9B ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS MANDATORY INFORMATION COLLECTION REDUEST: 500 HRS. REPORTED LESSONS LEARNED ARE ANO FEO BACK TO UIOUSTRY.

LICENSEE EVENT REPORT (LER) INCORPORATED INTO THE UCENSING PROCESS FORWARD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORMATiON ANO RECORDS MANAGEMENT BRANCH IT.G F33L US. NUCLEAR REGULATORY COMMISSION, (See reverse for required number of WASHINGTON, OC 20555000). ANO TO THE PAPERWORK REDUCTION PROJECT l3I50.

0104), OFFICE OF MANAGEMENT ANO BUDGET, WASHINGTON, DC 20503.

digits/characters for each block)

OOCKET NUMBER IE) PAGE (3)

FACILITY NAME (I)

Harris Nuclear Plant Unit-1 50-400 1 OF 3 TITLE (4)

Inadequate post maintenance testing following repairs on containment isolation valve 1SP-208.

EVENT DATE (5) LER NUMBER (6) REPORT DATE (7) OTHER FACILITIES INVOLVED (B)

FACIUTY NAME OOCKET NUMBER SEOUENTIAL REVISION MONTH DAY YEAR NUMBER NUMBER MONTH DAY YEAR 05000 FACILITY NAME DOCKET NUMBER 7 96 96 021 0 12 9 96 05000 OPERATING THIS REPORT IS SUBMITTED PUR SUANTTO THE REQUIREMENTS OF 10 CFR 5: (Chock one or more) (11)

MODE (9) 20.2201(b) 20.2203(a)(2)(v) X 50 73(a)(2)(i) 50.73(a)(2)(vm) 20.2203(a) (1) 20.2203(a)(3) (i) 50.73(a)(2)(ii) 50.73(a) (2) (x)

POWER LEVEL (10) 20.2203(a)(2)(i) 20.2203(a)(3)(ii) 50.73(a) (2) (iii) 73.71 20.2203(a) (2) (ii) 20.2203(a) (4) 50.73(a) (2) (iv) OTHER 50.36(c) (1) 50.73(a) (2) (v) Specify in Abstract below 20.2203(a)(2) (iii) or in NRC Form 306A 20.2203(a)(2) (iv) 50.36(c)(2) 50.73(a) (2) (vii)

LICENSEE CONTACT FOR THIS LER (12)

NAME TELEPHONE NUMBER lindude Aiee Code)

Michael Verrilli Sr. Analyst - Licensing {919) 362-2303 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DES CRIBED IN THIS REPORT (13)

(E es REPORTABLE REPORTABLE COMPONENT MANUFACTURER CAUSE SYSTEM COMPONENT MANUFACTURER CAUSE SYSTEM TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14) MONTH DAY YEAR EXPECTED YES SUBMISSION (If yes, complete EXPECTED SUBMISSION DATE).

X No DATE (15)

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single. spaced typewrinen lines) (16)

On November 7, 1996, with the plant operating in mode 1 at 100% power, containment isolation valve 1SP-208 was declared operable following replacement of the reed switch position indication assembly. This replacement was necessary to resolve erratic valve position indication. Post-maintenance stroke time testing was performed on 1SP-208 in accordance with the applicable surveillance test procedure prior to declaring the valve operable. However, during subsequent documentation review, it was discovered that remote position indication testing should have also been performed to satisfy ASME Code Section XI and Technical Specification requirements. This valve is a fully enclosed Target Rock solenoid operated valve, which prevents external observation of stem movement.

Investigation identified additional previous instances where remote position indication tests were not performed following maintenance on valves of this type. These previous instances, as well as the current event, were caused by an incorrect interpretation of ASME Code Section XI post maintenance testing requirements. Corrective actions will include procedure revisions and training for appropriate personnel to ensure a clear understanding of required post maintenance testing.

U.S. NUCLEAR REGUUITORY COMMISSION NRC FORM 366A (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION OOCKET LER NUMBER (6) PAGE (3)

FACILITY NAME (I)

SEQUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris Nuclear I'lant ~ Unit N1 50400 2 OF 3 96 - 021 00 TEXT ril moro spssois rovvdod, vso oddrsinosl sopios ol NRC Form 3&W (11)

EVENT DESCRIPTION:

On November 7, 1996, with the plant operating in mode 1 at 100% power, post maintenance testing was completed for Post Accident Sampling System, Containment Isolation Valve (1SP-208, EIIS Code: JM) following replacement of the reed switch position indication assembly to correct erratic valve position indication. This valve is a fully enclosed Target Rock solenoid operated valve, which prevents external observation of stem movement. The post maintenance testing that was completed consisted of a stroke time verification utilizing control switch indication in accordance with the Sampling System ISI Valve Test - Quarterly Interval Operations Surveillance Test (OST-1038). Following this test, the valve was declared operable at 0213 hours0.00247 days <br />0.0592 hours <br />3.521825e-4 weeks <br />8.10465e-5 months <br />. During subsequent documentation review, Operations personnel questioned'the need for additional testing for 1SP-208. Later that morning, Engineering personnel determined that post maintenance testing should have included a remote position indication verification to satisfy ASME Code Section XI requirements prior to declaring the valve operable. Based on this determination, 1SP-208 was declared inoperable and deactivated at 1535 hours0.0178 days <br />0.426 hours <br />0.00254 weeks <br />5.840675e-4 months <br /> on November 7, 1996.

1SP-208 was closed prior to the reed switch replacement and remained closed throughout the above sequence except for during the stroke time test. To confirm actual position of 1SP-208, verify proper stem movement, and satisfy the remote position indication verification requirement, radiography was performed with the valve in the open and shut positions on November 15, 1996.

Investigation identified additional previous instances where remote position indication tests were not performed following maintenance on solenoid operated valves such as 1SP-208. In each of these cases subsequent local leak rate tests verified proper valve position.

CAUSE:

This event, as well as the additional testing-deficiencies identified during investigation, were caused by an incorrect interpretation of ASME Code Section XI post maintenance testing requirements. This incorrect interpretation allowed stroke time verifications to satisfy post maintenance testing requirements when the maintenance performed /id not affect the valve's seating characteristics. Associated procedures also failed to provide adequate guidance to ensure proper testing.

SAFETY SIGNIFICANCE:

There were no adverse safety consequences as a result of this event. During the time period that 1SP-208 was incorrectly considered to be operable, the valve was shut, which completes its containment isolation safety function.

Radiography performed on November 15, 1996 confirmed that 1SP-208 was shut. For the additional containment isolation valves that did not receive remote position verifications following maintenance, subsequent local leak rate tests have verified their ability to isolate flow.

This condition is being reported in accordance with 10CFR50.73.a.2.i as a violation of Technical Specification 4.0.5.

PREVIOUS SIMILAR EVENTS:

There have been no previous similar events reported to the NRC pertaining to inadequate ISI program post maintenance testing. A similar occurrence was documented on an internal condition report in March of 1990 which also involved returning a solenoid operated containment isolation valve to service without performing the required post maintenance testing. However, corrective actions for this condition were narrowly focused on PMTR assignment responsibilities, did not clarify ISI testing requirements and thus, did not prevent recurrence.

R RM I4. )

NRC FORM 366A U.S. NUCLEAR REGUIATORY COMMISSION (4 BSI LICENSEE EVENT BEPOBT (LEB)

TEXT CONTINUATION FAGIUTY NAME ul DOCKET (ER NUMBER (6l PAGE (3)

SEQUENTIAL REVISION YEAR NUMBER NUMBER Shearon Harris Nuclear Plant ~

Unit //1 50400 3 OF 3 96 - 021 - 00 TEXT fffmore spsceis reeofrerf, ose eo'dirioosi copies of fVRC Form 366(l (lil CORRECTIVE ACTIONS COMPLETED:

1. A self assessment was performed on the ISI program to ensure completeness in meeting ASME Section XI requirements for solenoid operated valves. This was completed on November 11, 1996 and verified that adequate testing has been performed after maintenance on solenoid operated valves.

2. Information regarding proper post maintenance testing requirements was placed in an Operations Night Order to ensure that oncoming shift personnel are aware of these requirements.

CORRECTIVE ACTIONS PLANNED:

1. Appropriate Engineering (ISI individuals) and Operations personnel will be trained to clarify post maintenance testing requirements. This training will include a review of this LER and will be completed by March 30, 1997.

The following procedures will be revised to ensure appropriate post maintenance testing:

~ PLP400, "Post Maintenance Testing Program" (completion target date 1/24/97)

~ EST-212, "Type C Local Leak Rate Tests" (completion target date 3/7/97)

~ OST-1062 "Sampling, Chemical Addition and Main Steam Drain Systems ISI Valve Test and Remote Position Indication Test - Refueling Interval" (completion target date 2/21/97)

NR M (4.