W3P87-1700, Responds to Violation 8634-01 Noted in Insp Rept 50-382/86-34.Corrective Actions:Local Leak Rate Test Configuration Modified to Provide Greater Flow Measuring Capability

Responds to Violation 8634-01 Noted in Insp Rept 50-382/86-34.Corrective Actions:Local Leak Rate Test Configuration Modified to Provide Greater Flow Measuring Capability

ML20215G208
Person / Time
Site:	Waterford
Issue date:	06/18/1987
From:	Cook K LOUISIANA POWER & LIGHT CO.
To:	NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
W3P87-1700, NUDOCS 8706230161
Download: ML20215G208 (3)

Text

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• P. O. BOX 60340 LOUISI POWER AN A

& LiG HT/ 317 NEWBARONNE STREET ORLEANS, LOUISIANA 70160 *

(504) 595-3100

$iSikiks svSNM June 18, 1987 W3P87-1700 A4.05 QA U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555

Subject:

Waterford 3 SES Docket No. 50-382 License No. NPF-38 NRC Inspection Report 86-34 Attached is the Louisiana Power and Light Company (LP&L) response to Violation No. 8634-01 identified in Inspection Report No. 86-34.

If you have any questions on the response, please contact G.E. Wuller, Operational Licensing, at (504) 464-3499.

Very tr y yours,

,hk '

K.W. [ook Nuclear Safety and Regulatory Affairs Manager KWC:PTM:pmb Attachment cc: R.D. Martin, NRC Region IV J.A. Calvo, NRC-NRR J.H. Wilson, NRC-NRR NRC Resident Inspectors Office l E.L. Blake l W.M. Stevenson l

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\ l 8706230161 B7061B 2 PDR ADOCK 0500 h a l 8'AN EQUAL OPPORTUNITY EMPLOYER" j

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Attachment to W3P87-1700 Sheet 1 of 2 LP&L Response to Violation No. 8634-01 VIOLATION NO. 8634-01 Failure to Quantify Containment Isolation Valve Leakage The requirements of. Appendix J to 10CFR Part 50.54(o), are invoked as a condition of the license for all water cooled power reactors. Appendix J requires that a continuous running sum of local leak rates for penetrations and isolation valves subjected to this testing not exceed 0.6 times the maximum allowable. leakage rate at the calculated peak containment internal  !

pressure related to the design basis loss of coolant accident. '

l Contrary to the above, it was found during the week of December 8-11, 1986 that the local leak rates for six valves required to be tested had recorded as-found leak results of "off-scale" with no recorded value, so that the l continuous running sum of local leak rates was not known and the required comparison with the Appendix J criterion could not be made for the as-found condition.

This is a Severity Level IV violation.

RESPONSE

(1) Reason for the Violation The requirements of Appendix J Section III.C.3 to 10CFR Part 50 as invoked through Part 50.54(o) states in part: "The combined leakage rate for all penetrations and valves subject to Type E and C tests shall be less than 0.60 La." At the time the local leak rate testing was being performed the plant was in Mode 5.

As the result of local leak rate testing, six valves had as-found leak l rates which exceeded the capability of the tett equipment. The data as recorded was identified as "off scale" with no numerical value identified. During this inspection the NRC inspector noted that because of the recorded values of "off scale" he was unable to ascertain if the criterion of 0.60 La for leakage was satisfied. j l

(2) Corrective Action That Has Been Taken Prior to performing maintenance on the six valves the local leak rate test configuration was modified to provide a greater flow measuring capability. Of the six valves identified as having leak rates of "off scale", numerical values were identified for 5 of these valves. The sixth valve, a 24 inch containment vacuum check valve (CVR-102) could .

not be pressurized with the modified test unit, thus no leak rate {

could be established.

Corrective action was initiated by maintenance on all six valves and the appropriate retests were conducted. Acceptable leak rates were identified for these valves prior to returning to Mode 4. In accordance with the Appendix J criterion the required comparison for j

Attachment to W3P87-1700-Sheet 2 of 2 the as-found condition could not be made for CVR-102. For the purpose of this test it was assumed that the leakage for CVR-102 would result in exceeding the 0.60 La limit. As per AppenA ac J Section V.B.3 requirements this valve will be reported in a separate accompanying summary report'to the " Reactor Containment Building Integrated Leak 'l Rate Test" report.

L The ability to indicate and quantify leakage rates has been increased to account for valves with larger leak rates. A continuous running sum total leak rate has been established to be able to, at any time, identify present leakage as compared to the 0.60 La limit.

It is conceivable in the future that a gross leakage may exceed the capability of our test unit. In this event, alternative test methods or evaluations will be conducted to determine whether the 0.60 La criteria for Type C testing is met and a summary report will be prepared in accordance with Appendix J Section V.B.3.

(3) Corrective Action To Be Taken l

Surveillance Procedure (PE-5-002) " Local Leak Rate Test (LLRT)" will be modified to maintain a continuous. running sum of leak rates that fall within the 0.60 La criteria. Failed valves will be reworked and retested to ensure the total leakage is less than 0.60 La prior to entering a mode where the requirement exists.

(4) Date When Full Compliance Will Be Achieved Changes to the surveillance Procedure " Local Leak Rate Test (LLRT) are expected to be complete by July. 15, 1987, a