ML20116G638

Forwards Response to RAI Re Rev of Third 10-yr Interval Inservice Insp Program.Request for Relief RI-23, Withdrawn

ML20116G638
Person / Time
Site:	Cooper
Issue date:	08/05/1996
From:	Graham P NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS960131, TAC-M94000, NUDOCS 9608080236
Download: ML20116G638 (5)

Text

I l COOPER NUCLEAR STATON P.O. BOX 98. BROWNVILLE NEBRASKA 68321 h- \ Nebraska Public Power District "' h

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NLS960131 l

August 5,1996 i i

U.S. Nuclear Regulatory Commission l Attention: Document Control Desk l Washington, D.C. 20555-0301 l

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Subject:

Response to Request for Additional Information Regarding Revision 1 to the

! Third 10-Year Interval Insenice Inspection Program (TAC No. M94000) and l Withdrawal of Request for Relief RI-23 l Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46 l

References:

1) Letter (No. NLS950157) to USNRC from J. H. Mueller (NPPD), dated October 18,1995, " Third Ten-Year Interval Inservice Inspection Program"

2) Letter (No. NLS960050) to USNRC from J. H. Mueller (NPPD), dated l April 11,1996, " Response to Request for Additional Information and 1

Submittal of Revision I to the Third Ten-Year Interval Inservice Inspection Program"

3) Letter to G. R. Horn (NPPD) from D. L. Wigginton (USNRC), dated June 3,1996, ' Cooper Nuclear Station - Request for Additional Information Regarding Third 10-Year Interval Insenice Inspection Program (TAC No.

M94000)"

Gentlemen:

In Reference 1, the Nebraska Public Power District (District) submitted Revision 0 of the Third Ten-Year Interval Inservice Inspection Program. In Reference 2, the District submitted its response to a request for additional information and Revision 1 of tl.e Third Ten-Year Interval Inservice Inspection Program. In Reference 3, the NRC forwarded a request for additional l

information (RAI) regarding this revised program. In response to the RAI, the District is !g providing, in an attachment to this letter, its response to each of the individual NRC questions. '

This letter also withdraws Request for Relief RI-23. .

If you have any questions or require any additional information, please contact me.

l Sincerely, Y

J Philip D. Graham Site Manager 9608000236 960805 DR ADOCK0500g8 "Pdwurful2 2deln

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U.S: Nuclear Regulatory Commission August 5,1996 -

Page 2 of 2

/dnm Attachments l l

i j cc: Senior Project Manager USNRC - NRR Project Directorate IV-1 1

l i l Senior Resident Inspector

! USNRC - Cooper Nuclear Station Regional Administrator USNRC - Region IV l NPG Distribution t

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l Attachment to NLS960131 j Page1 of2 l

i Nebraska Public Power District's l Response to Request for Additional Information l Regarding Revision 1

< to the 1 i

Third 10-Year Interval Inservice Inspection Program f NRC Request 1.

In Request for Relief RI-21, Rev. O, NPPD included Item B3.100, Nozzle Inner Rauds, in the component identification. However, in the basis, inner radius sections are not included. Is this request specific to nozzle-to-shell welds only or should it also include inner radius, sections? Ifinner radius examinations are part of this request for relief, I provide the percent of coverage obtained for these examination areas.

l l Some of the percentages of coverage for the nozzle-to-shell weld appear low. Describe the technique being implemented to perform the examinations and what efforts have been made to maximize coverage.

l District Response I In Request for Relief RI-21, Rev. O, Code Item B3.100, Nozzle Inner Radius, applies to the Inner Radius of the CRD Return Nozzle, N9. It does not apply to any of the other nozzles listed. Only 70% of the CRD Return Line Nozzle Inner Radius was examined as indicated in the relief request.

Typical nozzle ultrasonic examinations included 0 ,45 , and 60* scans. Supplemental 70* scans were used in zone 1 of the inner radius. Supplemental manual exams were used where possible to increase coverage when automated examinations were performed. The percent examined in Table RR-21-1 represents the total composite coverage. The extent t

of coverage for individual scans is documented in the examination record. Examination l results were independently reviewed by a UT Level III examiner and by the Authorized Nuclear Inservice Inspector.  ;

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! NRC Request 2.

For Request for Relief RI-23, NPPD has requested relief from Code examination requirements associated with ihree RCIC welds in the Torus. Describe the function of the subject system relative to RHR, ECC, and CHR. Provide the basis for the code i classification of the subject welds on the strainer, when the subject welds are within the i torus. l

Attachment to NLS960131 l

! Page 2 of 2  ;

i District Response Request for Relief RI-23 applies to three submerged welds on the alternate suction supply piping for the Reactor Core Isolation Cooling system (RCIC). The RCIC system functions to provide core cooling water if the reactor becomes isolated from the main

, condenser. Although the RCIC system does not provide a RHR, ECC, or CHR function,

CNS has conservatively classified the RCIC suction piping as ASME Class 2 for inservice ,

i, inspection. Exemption IWC-1222(c) is applicable to this section of piping. Therefore,  !

CNS withdraws Request for Relief RI-23.

NRC Request 3.

For Request for Relief PR-02, NPPD requested relief from the Code requirements for the i system leakage test during startup. In the response to the request for additional l

information, the licensee stated that it proposes to perform the test prior to startup with '

the first isolation valve, as applicable, closed to facilitate the test. Verify that proposed leakage test will be performed following the reclasing of system (s) following a refueling

outage. Describe how the proposed alternative will provide the same level of quality and l safety as the Code required pressure test.

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District Response j Request for Relief PR-02 describes pressure test boundaries that are essentially equivalent I to the Code required pressure test boundaries. The test frequency and pressure meet  ;

Code requirements. The pressure test is performed at the end of the refueling outage afler - .

the reactor vessel is reassembled. The only difference in the test is that certain valves that l would normally be open for reactor startup, such as the Main Steam Isolation Valves and the Feedwater Check Valves, will be closed for the pressure test. It would be

! impracticable to perform a pressure test of the reactor vessel with these valves open. The  !

j steam lines would have to be flooded up to the turbine stop valves. Temporary supports 3 would have to be added for the additional weight of the water. Mechanicaljumpers would ,

l have to be installed around the Feedwater Check Valves. Request for Relief PR-02 l

1 clarifies that the systems are in their normalline up to the extent practicable. The VT-2 '

examination will extend to the outermost valves in these systems, as required by the Code.

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l e LIST OF NRC COMMITMENTS l ATTACHMENT 3 l Correspondence No: NLS960131 The following table identifies those actions committed to by the District in this document. Any other actions discussed in the submittal represent intended or planned actions by the District. They are described to the NRC for the NRC's information and are not regulatory commitments. Please notify the Licensing Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None N/A J

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l PROCEDURE NUMBER O.42 l REVISION NUMBER 1.1 l PAGE 9 OF 11 l