ML20196F533

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Forwards Examination Summary Rept for FW Nozzle Examinations Performed During 1998 Fall Refueling Outage at Cooper Nuclear Station.Rept Covers Past Examinations Up to & Including 1998 Fall Refueling Outage
ML20196F533
Person / Time
Site: Cooper Entergy icon.png
Issue date: 06/22/1999
From: Swailes J
NEBRASKA PUBLIC POWER DISTRICT
To:
NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS990062, NUDOCS 9906290233
Download: ML20196F533 (7)
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l Nebraska Public Power District Nebraska's Energy Leader N!.S990062 June 22,1999 U.S. Nuclear Regulatory Commission Attention: Documer't Control Desk Washington, D.C. 20555-0001 Gentlemen:

Subject:

Feedwater Nozzle and Control Rod Drive Return Line Report Cooper Nuclear Station, NRC Dock-t No. 50-298, License No. DPR-46

References:

1.

NUREG-0619, published November 1980, "BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking" 2.

Letter to G. R. Horn (NPPD) from R. B. Eevan (USNRC), dated February 13,1992, " Cooper Nuclear Station - Staff Acceptance of Fracture Mechanics Evaluation of Flaw Indications (TAC No. M82258)"

3.

Letter (No. NLS960095) to USNRC Document Control Desk from John H. Mueller (NPPD) dated June 28,1996, "Feedwater Nozzle and Control Rod Drive Return Line Report" In accordance wkh Reference 1 (Paragraph 4.4.3.1 (2), Reporting), the Nebraska Public Power District (District) is submitting to the Nuclear Regulatory Commission (NRC) an examination summary report for the feedwater nozzle examinations performed during the 1998 Fall refueling

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outage at Cooper Nuclear Station (CNS). The attached report covers past examinations up to

/

and including the 1998 Fall refueling outage. Based on the results of the examinations performed in the 1998 Fall refueling outage and plant performance, the following information is provided:

1.

No reportable indications were identified by ultrasonic examinations of the feedwater nozzle bore and inner radius areas.

g"O 2.

No changes to systems, procedures, or plant operation that would change the feedwater temperature or flow, since the last report, have been implemented.

3 No feedwater io nozzle bypass leakage in excess of the 0.3 gallons per minute reporting threshold, established in Reference 2, has been identified since the submittal of

- Reference 3.

Cooper Nudear Station m

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PO. Box 98/ BrownvnHe, NE 68321D098

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l Page 2 of 2 Should you have any questions concerning this matter, please contact me.

I Sincerely, 6.

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J Jo 'a ilos V e Pr t ofNuclear Energy 1

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Regional Administrator USNRC - Region IV Senior Project Manager d

USNRC - NRR Project Directorate IV-1 i

Senior Resident Inspector USNRC NPG Distribution 1

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.' Attachment to NLS990062 Page1of4 NUREG-0619 FEEDWATER NOZZLE EXAMINATION

SUMMARY

REPORT 1

m s Attachment to NLS990062 Page 2 of 4 -

Cooper Nuclear Station NRC Docket No. 50-298 License No. DPR-46 History Boiling Water Reactor (BWR) feedwater nozzle and control rod drive (CRD) return line nozzle cracking was identified at operating plants in the mid-to-late 1970's and in 1980. Accordingly, the NRC issued NUREG-0619 to:

1.

Describe technicalissues.

)

2.

Describe technical studies and analyses performed by General Electric Company and the

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- NRC staff.

1 Previde the NRC staff positions based on these studies.

4.-

Describe the NRC staff requirements for the licensee and applicants implementation of

]

the technical positions.

The NRC required that within 45 days of their letter (dated Nov. 13,1980), the licensee confirm implementation of the technical positions. In accordance with NRC technical positions, CNS

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performed the following:

l.

As a method to eliminate the potential for CRD nozzle cracking, CNS performed a CRD makeup water test to determine if the CRD return line was required for reactor pressure vessel (RPV) makeup flow. During the Fall 1977 refueling outage, the CRD system was tested for proper operation with the return line valved out of service. Based on satisfactory results from that test, CNS chose to perform a RPV nozzle design modification to eliminate the potential for thermal cycle cracking by cutting and capping the CRD retum line at the reactor vessel nozzle. Dye penetrant inspection of the CRD nozzle and surrounding vessel wall was performed at this time and revealed no crack indications.

'2.

In conjunction with CRD nozzle modification, CNS verified that the alternative CRD flow path was capable of providing adequate flow for RPV makeup. A CRD System return flow capability test was performed in accordance with paragraph 8.2(4) of NUREG-0619 (Ref. USAR Section 5.5.2). Special Test Procedure 81-1 was performed during the 1981 refueling and maintenance outage and demonstrated that the flow capacity was acceptable.

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3.

The introduction of reactor water cleanup (RWCU) flow to the feedwater flow has i minimal effect on feedwater temperature reduction. Therefore, the RWCU is not a

' contributing factor to feedwater nozzle thermal cycling. As a result, rerouting of the 1

RWCU system is not applicable to CNS.

4.

During the Spring 1980 refueling outage, the interference-fit feedwater spargers were removed, the nozzle cladding was removed, and the nozzles were inspected. Triple-

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.' Attachment to NLS990062 Page 3 of 4 Cooper Nuclear Station NRC Docket No. 50-298 License No. DPR-46 thermal sleeve, double piston ring spargers were installed. Engineering reviews of this design have shown it to be the most effective in preventing thermal sleeve bypass leakage because the design provides multiple leakage barriers. With the thermal sleeve design change, thermal sleeve bypass leakage has been minimized (does not exceed 0.3 gallons per minute). With the minimization of bypass leakage, rapid thermal cycling that causes thermal expansion / contraction stresses and the potential for crack initiation in the bore and radius regions is prevented.

5.

There were no changes made to the feedwater control system or other system changes that would affect feedwater flow temperature.

6.

During the Fall 1991 refueling outage, automated ultrasonic examination and thermal l.

sleeve bypass leakage monitoring was implemented in lieu of the nozzle dye penetrant (PT) examination required by NUREG-0619. NUREG-0619 originally required a PT l

examination be performed on a periodic basis of the feedwater bore and inner radius l

areas. As an alternative to the PT examination, CNS requested and obtained relief from l

the NRC to implement an ultrasonic examination technique as an attemative.

(

Reference:

USNRC letter to G. R. Hom, dated October 2,1991, " Review of l

NPPD's Request Regarding Feedwater Nozzle Examination Methods," TAC No. 79612).

j As shown on the attached table, CNS has performed periodic ultrasonic and visual examinations L

to ensure no crack initiation is occurring. Based on results through the 1998 Fall refueling outage, it is concluded that the new design is proven effective in minimizing thermal sleeve bypass leakage, preventing the potential for thermal cycle related crack initiation.

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  • .l'4
  • ATTACHMENT 3 LIST OF NRC COMMITMENTS l
l. -

Correspondence No: NLS990062

~The following table identifies those actions committed to by the District in this document. Any other actions discussed ir the submittal represent intended or

l. planned actions by the District.

They are deecribed to the NRC for the NRC's

-information and are not regulatory commitments.

Please notify the NL&S Manager at Cooper Nuclear Station of any questions regarding this document or any associated regulatory commitments.

COMMITTED DATE COMMITMENT OR OUTAGE None-N/A l

PROCEDURE' NUMBER 0.42 l

REVISION NUMBER 6 l

PAGE 9 op 13

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