ML20129G374

Forwards Byron Unit-1 SG Interim Plugging Criteria 90 Day Rept for EOC 7 Insp

ML20129G374
Person / Time
Site:	Byron
Issue date:	09/09/1996
From:	Wozniak D COMMONWEALTH EDISON CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
Shared Package
ML20129G378	List: ML20129G374 ML20129G388
References
BYRON-96-5146, NUDOCS 9610070173
Download: ML20129G374 (4)

Text

Ormnvinwealth life.un Company ll> ron Genrrating Station .

4 e50 North German Church Road It> ron,11. 61010 979 4 Tel 815 23 M 411 September 9,1996 LTR: BYRON-96-5146 FILE: 3.11.0320 4

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention: Document Control Desk

SUBJECT:

Byron Station Unit 1 Steam Generator Interim Plugging Criteria 90 Day Report for the End-of-Cycle 7 Inspection Byron Nuclear Power Station Facility Operating License NPF-37 NRC Docket No. 50-454

References:

1. November 9,1995, Letter from M.D. Lynch (NRR) to D.L. Farrar (Comed) Issuing Amendment No. 77 to Facility Operating Licenses NPF-37 and 66, Docket Nos. STN 50-454 and STN 50-455.

2. NRC Generic Letter 95-05," Voltage-Based Criteria for the Repair of Westinghouse Steam Generator Tubes Affected by Outside Diameter Stress Corrosion Cracking," dated August 3,1995.

3. September 20,1995, Letter from H. Pontious (Comed) to NRC regarding the September 13,1995, Teleconference between Comed and NRC concerning the increase in Interim Plugging Criteria.

4. March 19,1996, Letter from D. Saccomando to NRC regarding Comed implementation of 3.0 volt Interim Plugging Criteria Probe Wear Criteria. ,

in Reference 1, NRC approved a license amendment for Byron Station to implement a voltage-based Interim Plugging Criteria (IPC) for Unit 1 through Cycle 8 for Outer (q"u 4

Diameter Stress Corrosion Cracking (ODSCC)in steam generator tubing. NRC Generic  ;\ )

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, , g c,4tter 95-05 (Reference 2) served as cuidance for this amendment. NRC Generic Letter ]

9610070173 960909 )

PDR ADOCK 05000454  ;

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95-05 requires that the final results of the steam generator inspection and tubing integrity evaluation be submitted to the Staff within 90 days ofplant restart following a steam generator inspection that implemented a voltage-based repair criteria. Reference 3 requires a complete report to be submitted to the Staff within 90 days of plant restan detailing the results of steam generator internal visual inspections performed to support the increased IPC. In addition, Reference 4 requires the IPC 90 day report that is submitting to the Staff address the implementation of the alternate probe wear criteria used during the eddy current inspection.

Pursuant to these reponing requirements, Commonwealth Edison (Comed) is submitting the enclosed reports concerning the end-of-cycle 7 refuel outage steam generator IPC inspections and tube integrity evaluation. Attachment A to this letter contains the final report of the steam generator internal visual inspections performed to verify integrity of components necessary to support 3.0 volt IPC. Attachment B to this letter contains the results of the steam generator IPC eddy current inspection, results of the tube integrity evaluation, and assessment of the alternate probe wear criteria implementation.

Please address any questions regarding this matter to his. hiarcia Lesniak, Byron Station Nuclear Licensing Administrator at Downers Grove 708/663-6484.

Sincerely, l

David B. Wozniak Site Engineering hianager Byron Nuclear Power Station DBW/JS/cb Attachments cc: A. B. Beach, Regional Administrator - RIII l S. Burgess, Senior Resident Inspector - Byron l

G. Dick, Byron Project hianager - NRR i Oflice of Nuclear Facility Safety - IDNS p:\sec\byritr\96-5146. doc 4

e ATTACllMENT A ,

I Tube Support Plate Integrity Verifications Byron Unit 1 Cycle 7 Refuel Outage (B1R07) i l

l The structural integrity of steam generator internal components that are important to the bases of 3.0 volt IPC were inspected during the Byron Unit I mid-cycle inspection in the Fall of 1995 in accordance with the "SG Structural Integrity Plan in Support of Braidwood-l and Byron-13.0 volt IPC"(Inspection Plan). Additionally, Byron committed to perform additional structural load path inspections during BIR07, as described in a September 8,1995, letter to the Staff from K.L. Graesser (Byron Letter 95-0308). The additional inspections included visual inspection of the vertical support bar welds below the Paw distribution bafile, verification of the tube bundle wrapper alignment, enhanced eddy current of tubes near the anti-rotational device, and eddy current verification of tube support plate presence.

A visual inspection of the vertical support bar welds (24) beneath the flow distribution bafile was performed in each steam generator. The inspection was performed following completion of sludge lancing operations. Proper lighting and resolution was verified to meet ASME VT-1 requirements. Degradation of the welds was not found in any location.

Tube bundle wrapper alignment in all four steam generators was verified through each of the four sludge lance inspection ports located 90 degrees apart. Each wrapper was visually verified to be aligned at the four inspection portsjust above the tubesheet. The ability to install the sludge lancing equipment also ensured no misalignment between the steam generator shell and the tube bundle wrapper.

Enhanced eddy current examinations were performed in the areas of the three anti-rotation devices in each SG using the EPRI developed technique. The focus of this inspection was to verify the integrity of the tube suppor: plate. The enhanced technique involved acquiring data with a bobbin coil probe at a reduced pull speed of 12 inches per second or less. Anomalies, if found, were to be compared to defect signals from laboratory rapport plates fabricated and tested by EPRI. Fifty (50) intersections were inspected at e- 5 anti-rotational device. Due to SG symmetry,75 tubes were inspected to encompass the 50 intersections per anti-rotation device. Data was collected for the entire tube and each support plate was evaluated. No anomalies indicative of degradation were detected in any SG.

The presence of each tube support plate was verified for all SG tubes. This was performed as part of the normal eddy current analysis of each tube.

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, J ATTACllMENT B Steam Generator Tube Integrity Evaluation Westinghouse Report SG-96-08-005 Byron Unit-1 End-of-Cycle ?B Interim Plugging Criteria Report August 1996 i

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