ML20140H061

Forwards Summary of Results of re-examination of Three Indication on Core Spray Piping in Reactor Vessel at CNS Performed During 1997 Refueling Outage

ML20140H061
Person / Time
Site:	Cooper
Issue date:	05/07/1997
From:	Graham P NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS970088, NUDOCS 9705120315
Download: ML20140H061 (5)

Text

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NLS970088 May 7,1997 l

U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555-0001 l

Subject:

Inspection of Core Spray Spargers and Piping Cooper Nuclear Station, NRC Docket 50-298, DPR-46 Reference 1. Letter (No. NLS950228) to USNRC Document Control Desk from J. H. Mueller l (NPPD) dated November 22,1995, "IE Bulletin Response; Visual Inspection of Core Spray Spargers"

2. Letter (No. NLS950244) USNRC Document Control Desk from J. IL Mueller (NPPD) dated December 18,1995," Follow-up Information to IE Bulletin Response; Visual Inspection of Core Spray Spargers"

3. Letter to G. R. IIorn (NPPD) from J. R. IIall (USNRC) dated December 21,1995,

" Cooper Nuclear Station - Evaluation of Core Spray Piping Indications"(TAC No. M94097).

4. Letter (No. NLS960007) to USNRC Document Control Desk from J. IL Mueller (NPPD) dated March 29,1996, " Impact of Core Spray Line Crack Indications"

5. Letter (No. NLS960198) to USNRC Document Control Desk from P. D. Graham (NPPD) dated November 27,1996," Inspection of Core Spray Spargers and Piping" The purpose of this letter is to report the results of the re-examinations of three indications on the Core Spray piping in the Reactor Pressure Vessel at Cooper Nuclear Station (CNS) perfonned during the 1997 refueling outage. f lladstounil l In References 1 and 2, the Nebraska Public Power District (the District) reported the results of , O the Core Spray Sparger and Piping examinations performed during the 1995 refueling outage, provided an evaluation of three crack-like indications, and provided a supporting fracture mechanics evaluation. In Reference 3, the Commission provided their findings with respect to the evaluation of the indications and authorized operation for one additional cycle. In Reference 4, the District provided additional information regarding the potential impact of the postulated cracks in the Core Spray lines. In Reference 5, the District notified the Commission that future examinations of the Core Spray Spargers and Piping would be performed in accordance with the 13ailingWalgr Reactor Vessel Internals Project guidelines, BWRVIP-18. g ggggg 9705120315 970507 =e . -

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NLS970088

, May 7,1997 Page 2 of 3 The indications observed visually in 1995 were also examined ultrasonically at that time. Two of the indications were confirmed, and the surface lengths corresponded with the visual examination results. The third indication was not confirmed ultrasonically, but was conservatively assumed to be relevant.

Discussion The District has recently (during the 1997 refueling outage) completed visual re-examinations of the three indications previously identified using enhanced visual techniques in accordance with BWRVIP-18. Prior to the visual examinations, the surfaces were cleaned with a nylon bristle brush. The examinations did not identify any change in the length in two of the indications. The third indication could not be located after cleaning. This same indication could not be confirmed by ultrasonic examination during the 1995 outage.

The three indications were also re-examined ultrasonically during the 1997 refueling outage using bi-directional, focused, dual element 65" shear wave units. This technique satisfies the

BWRVIP-18 criteria and was approved by EPRI. A summary ofinspection results reported in 1995 and 1997, and maximum allowable flaw sizes from the analysis submitted in Reference 1, is attached.

Although proximity guidance contained in ASME Section XI would allow the indications on the A-loop #1 weld to be considered separately, the total length was evaluated considering the indications to be connected as was done in 1995. On this basis, the length was slightly greater than the length reported in 1995; however, still within the analyzed maximum flaw length considering two cycles of crack growth. The length of the indication on the A-loop #21 weld was slightly greater than the length reported in 1995. The length reported in 1997 for this weld l

was well within the analyzed maximum flaw length considering two cycles of crack growth. No l

fiaws were identified in the area of the B-loop #12 weld. We now believe that this third indication reported in 1995 is non-relevant.

The minor changes in the reported indication lengths may have resulted from either differences )

in UT examination methods, or from crack growth. If crack growth is indeed occurring, the rate is significantly less than the bounding rate assumed in the analysis submitted in Reference 1, and 1 the conclusions of this analysis remain valid. l The absense of significant growth in the indications, and the fracture mechanics evaluation submitted with References 1 and 2, demonstrate that flaw sizes at the end of the next fuel cycle i will not exceed the Code allowable flaw lengths. Based on the previous evaluations, the additional information submitted in Reference 4, and using the criteria in BWRVIP-18, there is sufficientjustification for at least one additional cycle of operation. The District will re-examine the two indication sites again during the 1998 refueling outage in accordance with the BWRVIP-18 criteria.

, NLS970088

, May 7,1997 Page 3 of 3 As governed per CNS procedures, the results of BWRVIP-18 inspections performed during the 1997 refueling, outage will be included in the ISI Summary Report.

If you have any questions regarding this submittal, please call.

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Sincerely, PbD-P. D. Graham Vice President-Nuclear

/dum Attachment I cc: Regional Administrator USNRC- Region IV Senior Project Manager USNRC-NRR Project Directorate l

Senior Resident Inspector USNRC NPG Distribution

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Summary ofInspection Results reported in 1995 and 1997 {gg and -Se s,82 Maximum Allowable Flaw Sizes -a a

Length Weld Reported In Length Measured Projected Length At Next Maximum Number 1995 (inches) In 1997 (inches) Cycle (inches)

• Allowable Flaw Size A1 8.9 9.1 10.3 11.8 A21 5.5 5.6 6.8 11.8 B12 1.5 NA NA 10.7

• Crack growth rate assumed to be 5 x 10 5 nches per hour, or 1.2 inches per 18 month operating cycle.

l ATTACHMENT 3 LIST OF NRC COMMITMENTS j

Correspondence No: NLS970088 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 The District will re-examine the two indication sites again during the 1998 refueling outage in accordance with the BWRVIP- * * "9 #9" 18 criteria.

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PROCEDURE NUMBER 0.42 l REVISION NUMBER 4 l PAGE 8 OF 9 l )

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