LIC-08-0120, Steam Generator Eddy Current Test Report - 2008 Refueling Outage

Steam Generator Eddy Current Test Report - 2008 Refueling Outage

ML083440629
Person / Time
Site:	Fort Calhoun
Issue date:	12/09/2008
From:	Clemens R Omaha Public Power District
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
LIC-08-0120
Download: ML083440629 (12)

Text

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-Im e Omaha Public Power District 444 South 16th Street Mall Omaha NE 68102-2247 December 9,2008 LIC-08-0120 U. S. Nuclear Regulatory Comrr~ission ATTN: Docun~entControl Desk Washington, DC 20555

Reference:

Docket No. 50-285

SUBJECT:

Fort Calhoun Station (FCS), Unit No. 1, Steam Generator Eddy Current Test Report 2008 Refueling Outage Pursuant to Technical Specification 3.1 7(3), attached is the FCS Steam Generator Eddy Current Test Report surr~niarizingtesting performed during the Spring 2008 Refueling Outage.

If you have any questions or require additional information, please contactTeddy Hutchinson at (402) 533-7389. IVo commitments to the NRC are made in this letter.

Division Manager IVuclear Engineering

Attachment:

Fort Calhoun Station Steam Generator Eddy Current Test Report, 2008 Refueling Outage Employment with Equal Opportunity

Attachment LIC-08-0120 Page 1 FORT CALHOUN STATION STEAM GENERATOR EDDY CURRENT TEST REPORT 2008 REFUELING OUTAGE

Attachment LIC-08-0120 Page 2 FORT CALHOUN STATION STEAM GENERATOR EDDY CURRENT TEST REPORT 2008 REFUELING OUTAGE Introduction

-This report summarizes steam generator eddy current test results obtained during the Fort Callioun Station (FCS) 2008 Refueling Outage (RFO). The Omaha Public Power District (OPPD) submitted test result summaries to the NRC for the two previous eddy current inspections in the following documents:

0 Fort Calhoun Station (FCS) Steam Generator Eddy Current Test Report -

2003 Refueling Outage, dated March 26, 2004 (LIC=04-0040)

Fort Calhoun Station (FCS) Steam Generator Eddy Current Test Report -

2005 Refueling Outage, dated September 26, 2005 (LIC-05-0118)

Westinghouse Nuclear Service Division performed eddy current examinations of the steam generator tubing at Omaha Public Power District's Fort Calhoun Station during April and May of 2008. The purpose of the examination was to assess the condition of the steam generators, to identify tubes requiring repair and to provide the information necessary to fulfill plant Technical Specification (TS) requirements. Entry into MODE 4 was achieved on June 14, 2008.

-The inspection conducted during the 2008 RFO was the first inspection conducted since replacement of the steam generators in 2006. Pursuant to the Fort Calhoun Station Technical Specifications, 100% of the tubes in each steam generator were inspected. No tube degradation was detected and no tubes were required to be repaired as a result of this inspection.

The condition monitoring assessment was performed to determine that Nuclear Energy Institute (IVEI) 97-06 Revision 2, Steam Generator Program Guidelines, performance criteria were satisfied at the time of ,the inspection. -The operational assessment was performed to determine that NEI 97-06 performance criteria will be satisfied until the next planned steam generator inspection.

Because no degradation was detected, condition monitoring is projected to be satisfied for three cycles of operation of 1.5 effective full power years (EFPY) each for both steam generators. Thus the operational assessment criteria are satisfied for at least three cycles of operation.

Secondary side maintenance consisted of sludge lancing in both steam generators. Top of tube sheet foreign object search and retrieval (FOSAR) and in-bundle inspections were co~iductedas well.

Attachment LIC-08-0120 Page 3 Description of FCS Steam Generators FCS is a two-loop Combustion Engineering design nuclear steam supply system (NSSS). The FCS steam generators were replaced during the 2006 RFO. The FCS replacement steam generators are re-circulating type, designed and manufactured by Mitsubishi Heavy Industries (IWHI) and are designated as model MHI-49TT-1. The tubing material is searr~lessnickel-chromium-iron Alloy 690 thermally treated (UNS N06690) with an outside diameter of 0.750 inches and a nominal wall thickness of 0.043 inches. There are 127 columns and 104 rows installed in the steam generator for a total of 5,200 tubes. There are 12 stay rods spaced throughout the tube bundle. The tubing is installed in a tri-angular configuration with a nominal pitch of 1.0 inch center to center. The tubing as installed has a single uniform u-bend with a minimum bend radius of 3.5 inches in row 1. All of the u-bends in rows 1 through 14 were stress relieved after bending by a second thermal treatment. In the upper bundle the tubes are restrained From out of plane movement by a series of 10 anti-vibration bars. The vertical section of the tubing is supported by 5 horizontal support plates which are of broached design with a three point contact on the tube. Each broached support is nominally 1.38 inches thick and is chamfered at both the top and bottom edges. The material for all of the support structures including the anti vibration bars (AVB) is series 405 stainless steel. The tube to tubesheet joint is comprised of a mechanical hard-roll at the primary face of the tubesheet and a hydraulic expansion of the remainder of tlie 25.79 inch thickness of the tubesheet. The nominal diametral tube expansion is 0.009 inches. The tubes are seal welded at the primary face of the tubesheet. The material for tubesheet itself is low alloy carbon steel and is clad with inconel on the primary face. The operating temperature (That) is 593'F.

At the time of the 2008 RFO, the FCS steam generators had operated for approximately 18 months and accumulated 1.34 EFPY. One (1 ) tube in steam generator RC-26 was plugged at the WlHl fabrication facility during manufacturing due to a manufacturing anomaly.

The total number and percent of tubes plugged at the start of the 2008 RFO are:

RC-2A has 0 tubes plugged, which is 0.000% of the total number of tubes.

RC-26 has 1 tube plugged, which is 0.019% of the total number of tubes.

Scope of Examination The Inspection Plan was developed from the FCS Degradation Assessment (DA) for the 2008 RFO. There are no existing degradation mechanisms in the FCS steam generators. The potential degradation mechanisms include:

O anti vibration bar (AVB) wear O

tube support plate (TSP) wear

Attachment I-IC-08-0120 Page 4 a

tube-to-tube fretting or proximity (PRO) wear a loose part wear A tube end to tube end bobbin coil inspection was performed on 100% of the tubes in the FCS steam generators. The examination program included multi-frequency bobbin probe testing for indications of degradation from wear at the AVB's, TSP, or because of tube-to-tube proximity or loose parts wear. The tests were conducted from both the hot and cold leg. A majority of the testing were full length exams, but the lower rows (rows 1-4) were tested from the hot leg, tube end hot to the top support on the hot leg side (48 inches per second) and from the cold leg, tube entry cold leg to the top support on the hot leg side (24 inches per second).

The special interest inspection program (informational exams) was + PointTM probe testing of locations such as all manufacturing burnish marks (MBM)',

bulges (BLG) and geometric distortions (GEO) from the pre-service inspection (PSI) were part of the base examination. All dents (DNT)* and dings (DNG)~2 0.5 volts from the PSI were also inspected. Additional + PointTMexams were performed for further evaluation of indications detected with bobbin coil tests.

Table 1 summarizes the steam generator eddy current examinations performed at FCS during the 2008 RFO.

Table 1 Steam Generator RC-2A The initial inspection program consisted of the following:

Bobbin examination: 100% of accessible tubes full length (5200 tubes) 2871 Hot leg (H/L) bobbin program tested at 48 inches per second 2329 Cold leg (C/L) bobbin program tested at 48 inches per second 252 Cold leg bobbin program tested at 24 inches per second

+ PointTMexamination special interest inspection programs:

13 Special lnterest historical indications H/L inspections in 13 tubes 5 Special Interest H/L inspections in 5 tubes 7 Special Interest historical indications C/L inspections in 5 tubes 1

Attributable to the manufacturing process.

2 Service induced dent (not present in baseline examination).

3 Ding - Baseline dent. The threshold for reporting a ding was a 5501140 differential mix P1 response of 1.0 volts for new dings (not initially identified in the baseline examination). Historical dings (called in baseline examination) were tracked at 0.5 volts.

Attachment I-IC-08-0120 Page 5 Steam Generator RC-2B The initial inspection program consisted of the following:

Bobbin examination: 100% of acqessible tubes full length (5199 tubes) 3002 Hot leg bobbin program tested at 48 inches per second 2197 Cold leg bobbin program tested at 48 inches per second 273 Cold leg,bobbin program tested at 24 inches per second

+ PointTMexamination of special interest inspection programs:

8 Special Interest historical indications H/L inspections in 8 tubes 2 Special Interest HIL inspections in 2 tubes 13 Special lnterest historical indications C/L inspections in 10 tubes 1 Special Interest CIL inspections in 1 tube The results of the inspection are discussed later in this report in the lnspection Results section.

All examinations were in compliance with the Steam Generator Program, Technical Specification Task Force (TSTF)-449 and the FCS Technical Specifications. The FTC inspection program, as conducted, meets or exceeds the recommendations contained in the Electric Power Research Institute (EPRI)

Pressurized Water Reactor (PWR) Steam Generator Guidelines, Revision 6. All data acquisition personnel were trained and qualified in the eddy current method and were certified to a minimum of Level I.

Personnel from Westinghouse, NDE, Young Technical Services (YTS),Tecnatom, HRlD and MoreTech performed the eddy current analysis.

Primary analysis was performed by Tecnatom, HRlD and Moretech data analysts. All primary analysis was done manually. NDE and YTS performed independent secondary analysis utilizing an automated data-screening program (ADS) for bobbin data and manual analysis for + PointTMdata. Representatives of Westinghouse, Wiltec and YTS performed resolution analysis. MoreTech provided independent Qualified Data Analysts (QDA) oversight for the analysis process. All analysis personnel assigned to the OPPD Fort Calhoun Station examination were Qualified Data Analysts in accordance with Appendix G of the EPRl PWR Steam Generator Examination Guidelines Revision 6. In addition, all data analysis personnel were qualified by a written and proficiency examination utilizing the EPRl site specific performance demonstration software program.

The Westinghouse ST Max data management system provided primary data management of the eddy current examination results. The ST Max system also provided electronic test plans to acquisition. The EddynetTM lnspection

Attachment LIC-08-0120 Page 6 Management System (EIMS) ZetecTMsoftware was used for secondary data management. The outputs of the Westinghouse ST Max data management system and the EIMS data management system were compared upon completion of the examination to ensure data accuracy.

Eddy current data was transmitted from the acquisition stations to the analysis stations via a local area network (LAIV) and stored on local hard disks. The data was subsequently transmitted via a wide area network (WAN) to two independent teams of analysts. Primary and secondary analysis was done in the Westinghouse remote data room (REDAC) located at the Waltz-Mill site in Madison, PA. Lead analysts (eddy current Level Ill), resolved discrepancies between the two sets of evaluation results. The removal of potential degradation indications from the database received the concurrence of two Lead Analyst personnel representing the primary and secondary analysis teams. All analysis was in concurrence with the Westinghouse procedure MRS-SSP-2229-CFTCI rev 00. All analysis was performed using Westinghouse Anser software rev 299 on a Hewlett Packard (HP)-UX operating system installed on HP UNlX workstations. The raw data along with the evaluated results from these analyses were then transferred to optical disks for permanent storage.

In addition to the regular analysis the outer peripheral tubes were re-analyzed in a separate analysis group. This re-analysis was done by specifically screening the low frequencies to identify loose parts. This re-analysis was done by the secondary analyst.

During the course of the examination, feedback was provided to all analysis personnel by utilizing the Analysis Performance Tracking Software (APTS). All analysts were required to review all of their missed calls and 20% of their overcalls.

In addition to OPPD's Steam Generator Program Manager, OPPD was represented by an Eddy Current Level Ill, from MoreTech, who performed as lndependent QDA. She was not part of the Primary, Secondary, or Resolution analysis teams. The independent QDA provided an overview function for Omaha Public Power District in accordance with Westinghouse procedure MRS-SSP-2229-CFTCI rev 00. Among other duties the independent QDA randomly reviewed the data to ensure that the data quality was acceptable. The lndependent QDA also checked to ensure the resolution process was properly performed and that field calls were properly reported.

Inspection Equipment Westinghouse Electric Company performed the nondestructive examination (NDE) of the steam generator (SG) tubes. Westingh,ouse PEGASYS manipulators with dual guide tube fixtures were used to perform testing from the

Attachment I-IC-08-0120 Page 7 hot leg and cold leg plenums in both steam generators. Corestar Ornni-200 multi-frequency eddy current testerstprobe pushers were used to energize the test probes on all test stations. HP workstations controlled the PEGASYS fixture, and the Omni-200 testertprobe pushers. The software package that was used is documented in MRS-DFD-2217-SR Rev 2. The raw eddy current data was stored on hard disks attached to the acquisition workstations; this data was then stored on the data server computers. Final data storage was on optical disks.

Inspection Techniques The Fort Calhoun Station Steam Generator 2008 RFO Degradation Assessment Report (DA) number SG-CDME-08-04 rev 1 identifies the following EPRl Appendix H Qualified ECT Techniques (ETSS) as applicable for Fort Calhoun Station for potential damage mechanisms or informational exams with the bobbin or MRPC +Point.

Table 2 Potential Damage Mechanisms ETSS # Probe Application 96004.1 r l 1 Bobbin - Wear at Tube Supports 96004.1 r l 1 Bobbin Wear at Anti-Vibration Bars and Diagonal Straps 27091.2 r0 Bobbin Detection and Depth Sizing of Loose Part Volumetric Indications, Freespan area (loose part not present) 96910.1 rO Plus Point Wear, Plus Point, 300t100 kHz mix TSP 27901 . I to 27907.1 all Plus Point Detection and sizing of various orientation of r0 indications caused by loose parts (loose part not present)

Informational exams ETSS # Probe Application 96010.1 r7 Bobbin Small volume MBM in free span of tubing 24013.1 r2 Bobbin Outer Diameter Stress Corrosion Cracking in Freespan Dings less than 5 volts 22401.1 r4 Plus Point Detection of axial ODSCC at dented support structures 22484.1 r4 Plus Point Detection and length sizing of circumferential ODSCC in parent tubing with dented support structures.

9651 1.2 r16 Plus Point Detection of Axial and Circumferential Primary Water Stress Corrosion Cracking in low row U-Bend Regions.

Attachment LIC-08-0120 Page 8 The test frequencies used for the bobbin probe were 550, 280, 140 and 40 kHz differential and absolute. The test frequencies used for the 1 Coil NlRPC U-Bend

+Point Midrange probe were 400,300, 150 and 35 kHz.

The bobbin probe calibration standards used for the examination were manufactured to conform to the requirements of Section V of the 1998 to 2000 addenda of the ASME Boiler and Pressure Vessel Code. Tests of the calibration standards and the as-built drawings of each standard were used to set spans, voltages and phase rotations of each bobbin test, as well as settirrg calibration curves for sizing of tube wear.

MRPC standards were manufactured with various EDM notches to permit the set up of the spans and rotations to achieve the best possible detection. The dimensions and tolerances of the notches are in accordance with the recommendations provided in the EPRl PWR Steam Generator Guidelines, Revision 6.

Inspection Results There are currently no degradation mechanisms in the Fort Calhoun steam generators. The results for the potential degradation mechanisms and non damage mechanisms are as follows:

Potential Degradation Mechanisms:

o Mecliarrical Wear lndications at the Anti Vibration Bar: There were 0 tubes in SG RC-2A and 0 tubes in SG RC-2B, with indications of wear at the anti vibration bars.

o Mechanical Wear lndications at Support Structures: There were 0 tubes in SG RC-2A and 0 tubes in SG RC-2B, with indications of wear at support str~~ctl-lres.

o Mecharlical Wear lndications due to Tube-to-Tube Fretting or Proximity (PRO) Wear. There were 0 tubes in SG RC-2A and 0 tubes in SG RC-2B, with indications of wear from tube proximity.

No indication of tube-to-tube proximity was detected this outage.

o Mechanical Wear lndications due to Loose Parts: There were 0 tubes in SG RC-2A and 0 tubes in SG RC-2B, with indications of loose part wear. No loose parts were detected during this outage from either ECT testing or secondary side visuals.

Attachment LIC-08-0120 Page 9 Informational Exams:

o Manufacturing Burnish Marks (MBM): Manufacturing Burnish Marks (MBM) are conditions where shallow volumetric scuffs introduced into the tubes exterior during the final tube polishing.

There were two MBM indications reported during the baseline.

Using the analysis procedure flow charts for freespan indications these indications were not reportable. Both of the historical iVlBMs were tested with the plus point coil. R 56 C12 showed a volumetric indication and was called an MBM, testing of R 36 C 68 showed no indication.

If an indication had changed based on the phase or voltage requirements or was a new indication, an absolute drift indication (ADI) or distorted freespan indication (DFI) was called and the indication was examined with a plus point cod. There were two (2)

DFI indications identified in two (2) different tubes in SG RC-2A and one (1) AD1 and one (1) DFI indications identified in two (2) different tubes in SG RC-2B. All the AD1 and DFI indications were inspected with a plus point coil. The results of the tests were that no defect was found (NDF).

o Ding with indication (DDI) is a condition where a possible flaw signal forms within a ding signal. There was one (1) DDI indication reported in SG RC-2A. The DDI signal was inspected with a MRPC plus point coil. The results of the tests were no defect was found (NDF). There were no DDI signals reported in SG RC-2B .

o Ding (DNG) is a baseline indication where the tubing inside diameter is less than normal. The degradation asses'sment specified that all dings 0.5 volts or greater that were reported in the baseline exarrrination would be inspected with a plus point coil.

There were 15 dings in 13 tubes in SG RC-2A and 17 dings in 14 tubes in SG RC-2B. All the dings in SG RC-2A and SG RC-2B were inspected with a MRPC probe with a plus point coil. The results of the tests were that no defect was found (NDF).

o Dent (DNT) is a service induced dent where the tubing inside diameter is less than normal. Dent indications were reported if the P I mix channel was greater than 1.0 volt and there was no dent indication found in the baseline data. There were three (3) dentslding identified in three (3) different tubes in SG RC-2A and one (1) new dent identified in SG RC-2B. All the dents in SG RC-2A and SG RC-2B were inspected with a MRPC probe with a plus

Attachment LIC-08-0120 Page 10 point coil. The results of the tests were that no defect was found (MDF). ,

o Bulge (BLG) is a condition where the tubing outside diameter is greater than normal. There were two (2) bulges identified in two (2) different tubes in SG RC-2B. The location of these bulges were near the 5th AVB support. All the BLG indications were inspected with a MRPC probe with a plus point coil. The results of the tests were that no defect was found (IVDF).

o Geometric Distortion (GEO) is an abnormal geometry change, usually indicative of a lift-off signal caused by a local inside diameter change from tubing manufacturing. There were five Geometric Distortion identified with the MRPC probe using a p l ~ ~ s point coil in SG RC-2A. The location of these GEOs were near the hot leg 5th TSP. There were no GEO signals reported in SG RC-2B.

o All special interest was inspected with the single coil plus point coil.

The sample density was 30 samples circumferential and 30 samples axial.

Repairable Indications:

o Steam Generator RC-2A, there were no tubes plugged.

o Steam Generator RC-2B, there were,no tubes plugged.

The total number and percent of tubes plugged at the completion of this outage:

o RC-2A has 0 tubes plugged, which is 0.000% of the total number of tubes.

o RC-2B has 1 tube plugged, which is 0.019% of the total number of tubes.

Tube Plug Inspections:

o There are zero (0) plugged tubes in Steam Generator RC-2A.

o There is one (1) plugged tube in Steam Generator RC-2B located at R71 C57. This tube was plugged during manufacturing. A visual inspection was performed on the tube plug and it was found to be in excellent condition.

Inspection Plan Expansions:

o No inspection plan expansions were performed as part of the 2008 steam generator inspections.

Attachment 1-IC-08-0120 Page 11 Secondary Side Maintenance and inspections:

o Approximately five (5) pounds of sludge was removed from each steam generator.

o Top of tube sheet FOSAR and in-bundle inspection found zero (0) loose parts in each steam generator.