ML20101L367

Submit Addl Info Re Potential Impact of Postulated Cracks in Core Spray Lines for Plant

ML20101L367
Person / Time
Site:	Cooper
Issue date:	03/29/1996
From:	Mueller J NEBRASKA PUBLIC POWER DISTRICT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
NLS960007, TAC-M94097, NUDOCS 9604030431
Download: ML20101L367 (4)

Text

e COOPER NUCLEAR STATION h h P.O. BOX 98. BROWNVILLE. NEBRASKA 68321 4: ' Nebraska Public Power District

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NLS960007 March 29,1996 l

l U.S. Nuclear Regulatory Commiss on Attention: Document Control Desk Washington, D.C. 20555-0001

Subject:

Impact of Core Spray Line Crack Indications Cooper Nuclear Station, NRC Docket No. 50-298, License No. DPR-46

References:

1. Letter (No. NLS950244) to USNRC Document Control Desk from J.11.

Mueller (NPPD) dated December 18,1995: Follow-up Information to IE Bulletin 80-13 Response; Visual Inspection of Core Spray Spargers

2. Letter to G. R. Horn (NPPD) from J. R. IIall (US NRC) dated December 21,1995: Cooper Nuclear Station - Evaluation of Core Spray Piping Indications (TAC No. M94097)

Gentlemen:

The purpose of this letter is to submit to the Nuclear Regulatory Commission (NRC) additional (

information regarding the potential impact of postulated cracks in the Core Spray lines for Cooper Nuclear Station (CNS). The Nebraska Public Power District (District) committed, by letter dated December 18,1995 (Reference 1) to provide this information as a follow-up to the NRC's Safety Evaluation Report (Reference 2). The District committed in Reference 1 to forward this information to the NRC by February 8,1996. However, as discussed with and agreed to by the NRC CNS Senior Project Manager, submittal of this letter by the end of March 1996 was acceptable in order to address minor technical issues.

The District has completed a review of the efTect of potential leakage through crack indications which have been observed in both the "A" and "B" core spray lines. The review assumes a through wall crack with a length equal to the maximum predicted cycle 17 growth and a width of 10 mils. The review also considered the unrelated leakage from the T-box vent holes associated with the core spray piping / core shroud penetrations. There is one vent hole for each T-box and one T-box for each core spray line. The total leakage flow rate for the two T-box vent holes combined is conservativel.< :stimated to be 20.2 gpm.

For the flaws in lines "A" and "B," the total poteuii! kakage rate has been calculated to be 40.2 gpm. For line "A," the calculated potential leakage rate consists of 20.8 gpm for sleeve weld #1 and 13.8 gpm for sleeve weld #21. For line "B " the calculated potential leakage rate consists of 5.6 gpm from T-box weld #12. With the addition of the vent hole flows (20.2 gpm), the total leakage rate for both loops sums up to 60.4 gpm. This translates to less than 1.5% of total core spray flow rate.

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NLS960007 March 29,1996 Page 2 of 3 A change of 10% core spray flow during the worst case Design Basis Analysis (DBA) Loss of Coolant Accident (LOCA), which consists of an 80% recirculation line break, is conservatively estimated to yield a 20 F change in Peak Clad Temperature (PCT). A change of 1.5% core spray flow is conservatively estimated to yield an increase of 3 F PCT during the DBA LOCA For CNS, the thermal-mechanical Maximum Average Planer Linear Heat Generation Rates (MAPLIIGR) bound the DBA LOCA. The design PCT for the most limiting DBA LOCA is 2,200 F LOCA at an MAPLIIGR of approximately 13.8 kilowatts per foot (kW/ft). For Cycle 17, the limiting value MAPLilGR is 12.93 kW/ft, which conservatively translates to a PCT value of approximately 2,150 F. When adding the additional 3 F to the PCT range for the current cycle, there is considerable margin between the resultant PCT and the design PCT. Therefore, the evaluation indicates that the postulated leakage would not adversely affect core spray performance and that such leakage would not affect plant operating limits.

The District has perfomied an analysis to evaluate the safety concerns associated with a postulated piping failure which results in loose piping parts falling into the vessel annulus region.

The analysis evaluated four major concerns associated with loose parts: The potential for fuel bundle flow blockage and consequent fuel damage, the potential for fretting wear of fuel cladding, the potential for interference with control rod operation, and the potential for corrosion or chemical reaction with other reactor materials.

The evaluation results indicate that a single 360 degree (full circumferential) through-wall crack l would not result in loose pieces and the sparger would remain attached to the shroud. Further, in the unlikely event that multiple failures were to occur and a piece was to break loose, the l evaluation concludes that the loose part would result in no significant fuel bundle flow blockage, no potential for interference with control rod operation and no potential for corrosion or adverse chemical reaction with other reactor materials. As for the potential for fretting wear of fuel l cladding, possible fuel rod leaks would be detected by the off-gas system so that appropriate action can be taken to maintain the off-site radiation release within acceptable limits. Any such cladding damage would only be an operational concern and not a safety concern to the magnitude of the other three concerns identified above.

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

! Sincerely, j i  !

[ s -

John IL Mueller j Site Manager j

/ dam Attachment

.- NLS960007 March 29,1996 Pagie 3 of 3 1 o

cc:. Senior Project Manager USNRC - NRR Project Directorate IV-1 l Senior Resident Inspector l USNRC - Cooper Nuclear Station Regional Administrator  :

USNRC - Region IV NPG Distribution i

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LIST OF NRC COMMITMENTS ATTACHMENT 3 Correspondence No; NLS960007 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 i r

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l PROCEDURE IM4BER 0.42 l FEVISION iM4BER 1 l PAGE 9 OF 11 l