Text

Steam Generator Eddy Current Test Report, 2015 Refueling Outage
	ML15328A491
Person / Time
Site:	Fort Calhoun
Issue date:	11/24/2015
From:	Cortopassi L; Omaha Public Power District
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	LIC-15-0136
	Download: ML15328A491 (11)
	v • d • e

Omaha Public Power District November 24, 2015 LIC-15-0136 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Fort Calhoun Station (FCS), Unit 1 Renewed Facility Operating License No. DPR-40 NRC Docket No. 50-285

Subject:

Fort Calhoun Station (FCS), Unit No. 1, Steam Generator Eddy Current Test Report-2015 Refueling Outage Pursuant to Technical Specification 3.17(3), attached is the FCS Steam Generator Eddy Current Test Report summarizing testing performed during the spring 2015 Refueling Outage.

If you have any questions or require additional information, please contact Mr. Bill Hansher at (402) 533-6894. No commitments to the NRC are made in this letter.

Louis P. Cortopassi Site Vice President and CNO LPC/T JH/mle

Enclosure:

Fort Calhoun Station Steam Generator Eddy Current Test Report, 2015 Refueling Outage c:

M. L. Dapas, NRC Regional Administrator, Region IV C. F. Lyon, NRC Senior Project Manager S. M. Schneider, NRC Senior Resident Inspector

LIC-15-0136 Enclosure FORT CALHOUN STATION STEAM GENERATOR EDDY CURRENT TEST REPORT 2015 REFUELING OUTAGE

Page 1 TABLE OF CONTENTS

1.0 Introduction........................................................................................................................ 2 2.0 The Scope of the inspections performed on each SG (3.17(3)a)......................... 3 3.0 Active Degradation Mechanisms Found (3.17(3)b).................................................. 4 4.0 Nondestructive examination techniques utilized for each degradation mechanism (3.17(3)c)....................................................................................................... 5 4.1 Table 1 - NDE Techniques Utilized for Identified Degradation...................... 5 5.0 Location, orientation (if linear), and measured sizes (if available) of service induced indications (3.17(3)d)....................................................................................... 5 5.1 Broached (Trefoil) Tube Support Plate (TSP) Wear............................................... 5 Table 2 - Summary of TSP Wear Indications............................................................. 5 6.0 Number of tubes plugged during the inspection outage for each active degradation mechanism (3.17(3)e)............................................................................... 8 7.0 Total number and percentage of tubes plugged to date and the effective plugging percentage in each steam generator (3.17(3)f and 3.17(3)h)............... 8 Table 3 - Tube Plugging Summary.............................................................................. 8 8.0 The results of condition monitoring, including results of tube pulls and in-situ testing (3.17(3)g)....................................................................................................... 8 8.1 Broached (Trefoil) Tube Support Plate (TSP) Wear............................................... 8 8.2 Operational Leakage Criterion and Validation of Previous Operational Assessment (OA).......................................................................................................... 9

Page 2 FORT CALHOUN STATION STEAM GENERATOR EDDY CURRENT TEST REPORT SPRING 2015 REFUELING OUTAGE 1.0 Introduction This report summarizes steam generator eddy current test results obtained during the Fort Calhoun Station (FCS) Unit No. 1 2015 Refueling Outage (RFO) hereafter referred to as FCR27.

Eddy current examinations in the steam generator tubing at FCS were performed during April and May of 2015. 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 requirements.

The inspection conducted in FCR27 was the second inspection of the steam generators after replacement of the original steam generators in 2006. Pursuant to FCS Technical Specifications, 62% of the tubes in each steam generator were inspected. Minor tube support plate wear was detected, however, no tubes were required to be repaired as a result of this inspection.

The condition monitoring assessment was performed to determine that NEI 97-06 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.

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 visual inspections and sludge lancing in both steam generators. The visual inspections included steam drum inspections, upper bundle and fifth tube support plate inspections, and top of tube sheet foreign object search and retrieval (FOSAR) and in-bundle inspections.

Description of FCS Steam Generators FCS Unit No. 1 is a two-loop Combustion Engineering design nuclear steam supply system (NSSS). The two FCS steam generators (SG-A and SG-B) were replaced during the 2006 RFO.

The FCS replacement steam generators are re-circulating type, designed and manufactured by Mitsubishi Heavy Industries and are designated as model MHI-49TT-1. The tubing material is seamless nickel-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 each steam generator for a total of 5,200 tubes. In the upper bundle the tubes are restrained from out of plane movement by a series of 10 anti-vibration bars (AVB). 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 AVBs is series 405 stainless steel. The operating temperature (Thot) is 593F.

Technical Specification (TS) 5.23 provides the requirements for SG inspection frequencies and requires periodic tube inspections be performed. TS 5.23 requires that 100% of the tubes be inspected at sequential periods of 144, 108, 72, and thereafter 60 effective full power months (EFPM). FCS replacement SGs are still in their 1st (144-month) Inspection period. This

Page 3 examination meets the requirement to inspect 50% of the tubes by the refueling outage nearest the mid-point of the period.

During the FCS spring 2015 refueling outage (FCR27) both steam generators (SG-A and SG-B) were inspected in accordance with FCS TS 5.23. This was the second in-service inspection of the replacement steam generators (RSG). Due to the length of the 2011 RFO when FCS was shut down due to Missouri River flooding and entry into Inspection Manual Chapter 0350, the FCS RSGs have only operated for approximately 6 years and accumulated 5.22 EFPY by FCR27.

Below summarizes the results of the inspection in accordance with the 180-Day reporting requirements of TS 3.17(3) Bold wording restates the TS requirement, followed by the required FCR27 information.

A report shall be submitted within 180 days after exceeding 210°F reactor coolant temperature (Tcold) following completion of an inspection performed in accordance with the Technical Specification 3.17(3), Steam Generator (SG) Program.

The report shall include:

2.0 The Scope of the inspections performed on each SG (3.17(3)a)

The Inspection Plan was developed from the Fort Calhoun Station Degradation Assessment (DA) for FCR27. There were no pre-existing degradation mechanisms in the Fort Calhoun Steam Generators. The potential degradation mechanisms include:

Foreign Object Wear Tube Wear at TSP intersections Tube Wear at AVB Intersections (Periphery Tubes and Inner Bundle Tube above Row 15)

Tube Wear at Retainer Bar Intersections Tube-to-tube Wear Reactor coolant temperature (Tcold) exceeded 210°F on May 29, 2015. As required by the EPRI Examination Guidelines Rev. 7, the inspection program addressed potential degradation mechanisms. The following outlines the FCR27 initial inspection plan, as justified in the Degradation Assessment, applicable to FCS steam generators SG-A and SG-B:

Bobbin Coil Inspection The Bobbin Base Scope consisted of a full-length inspection of periphery tubes and a full length inspection of a 50% sample of the inner bundle tubes. The periphery tube inspection consisted of a minimum of 3 tubes deep along the diagonal and 2 tubes deep in any row. The inner bundle tubes consisted of all tubes in alternating columns plus all tubes in rows 91 and higher, which are not part of the inner bundle tubes.

Rotating Coil Inspection U-Bend

+Point' probe inspection of the U-bend region from the upper TSP hot leg (HL) to upper TSP cold leg (CL) of six tubes in Rows 1 and 2 tubes that are on the periphery (R1C1, R1C3, R1C125, R1C127, R2C2, and R2C126).

Page 4 Rotating Coil Inspection - Straight Section

+Point' probe inspection of the outer three tubes of periphery, within three inches of the Top of the Tubesheet (TTS +/-3 inches), HL and CL.

Rotating Coil Inspection - Special Interest

+Point' probe inspection of all previously identified dents and dings 5 volts.

+Point' probe inspection of all prior and FCR27 "I-code" and/or non-quantifiable indications as determined by bobbin inspection or any previously reported signal that has changed.

+Point' probe inspection of new potential loose parts (PLPs) and any in the eddy current database as identified by previous eddy current inspections.

+Point' probe inspection of all foreign objects not retrieved in FCR25.

+Point' probe inspection of a minimum two tube locations surrounding any newly identified PLP.

+Point' probe inspection of a minimum two tube locations surrounding newly identified foreign objects that are classified as Priority 1 or Priority 2.

MRPC (Motorized Rotating Pancake Coil) inspection of all tube-to-tube (TTW) indications detected by bobbin coil.

MRPC inspection of all reportable proximity (PRO) indications.

MRPC inspection of all MBM (Manufacturing Burnish Mark) bobbin indications that have increased by 0.5 volts.

MRPC inspection of all Channel 6 calls that are paired with another Channel 6 call in an adjacent tube and at a common elevation.

Other Primary Side Inspections o Video Inspection of both installed plugs (in SG-B, the HL and CL of Tube B-R57C71). The plugs were installed during manufacturing prior to SG-B being place in service.

o Video Inspection of hot and cold leg channel heads looking for thinning or missing cladding and associated wastage using the recommended inspections of the SG channel head, including the entire divider plate to channel head weld and all visible clad surfaces.

Secondary Side Inspections The secondary side visual inspections in both SGs included: top of tubesheet annulus, no tube lane and in-bundle (after sludge lancing), steam drum, moisture separators, feedring, upper bundle, uppermost (5th) tube support, retainer bars and support structures.

3.0 Active Degradation Mechanisms Found (3.17(3)b)

One degradation mechanism was confirmed to be present in the FCS SGs; broach (Trefoil) tube support plate wear. No other degradation mechanisms, including AVB wear, retainer bar wear, loose part wear or tube-to-tube wear, were detected. TSP wear had not been previously detected at FCS, however, it has been found on most replacement SGs so TSP wear was not unexpected based on the FCS degradation assessment.

The primary side visual inspection of the cladding, previously installed plugs, and divider plate found no degradation.

Page 5 The secondary side visual inspections at the top of tubesheet annulus, no tube lane and in-bundle, steam drum, moisture separators, feedring, upper bundle, uppermost (5th) tube support, retainer bars and support structures found no evidence of degradation in either SG.

4.0 Nondestructive examination techniques utilized for each degradation mechanism (3.17(3)c)

Table 1 below identifies nondestructive examination (NDE) techniques utilized for each identified active degradation mechanism.

4.1 Table 1 - NDE Techniques Utilized for Identified Degradation Degradation Mechanism Inspection Type EPRI ETSS Broached Tube Support Wear Bobbin (Detection) 96004.1 (Rev. 13)

MRPC +Point' (Sizing) 96910.1 (Rev. 10) 5.0 Location, orientation (if linear), and measured sizes (if available) of service induced indications (3.17(3)d) 5.1 Broached (Trefoil) Tube Support Plate (TSP) Wear A total of 19 tubes had indication of wear related to TSPs reported during the FCS outage (10 in SG-A and 9 in SG-B). Five (5) tubes had wear at more than one TSP. All of the TSP wear locations were inspected with +Point' to confirm that the location and depth of the indications were consistent with tube support wear and not some other damage mechanism such as foreign object wear. The +Point' examination also determined that nearly all of the wear locations involved more than one of the 3 trefoil lands so all sizing was performed using +Point'.

This was the first occurrence of tube support wear at FCS. However, this is not an unexpected occurrence since tube support wear has already been observed at most other replacement SGs.

Table 2 provides a listing and sizes of these indications.

Table 2 - Summary of TSP Wear Indications SG Row Col TSP TSP Land Distance from TSP Centerline (inch)

Max.

Depth

%TW Max.

Length (inches)

Max.

Width1 (Inch)

Max.

Width (Degrees)

A 1

29 04C 1

-0.54 9

1.25 0.47 72 2

-0.52 8

1.15 0.41 63 3

0.4 9

0.86 0.39 59 A

1 37 03C 1

-0.32 5

1.31 0.44 67 2

0.53 12 1.31 0.4 61

Page 6 SG Row Col TSP TSP Land Distance from TSP Centerline (inch)

Max.

Depth

%TW Max.

Length (inches)

Max.

Width1 (Inch)

Max.

Width (Degrees)

A 1

37 04C 1

-0.53 2

0.58 0.42 64 2

-0.49 8

1.12 0.48 74 3

0.35 10 1.09 0.48 74 A

1 121 04C 1

-0.75 5

0.64 0.42 64 2

-0.64 8

1.34 0.47 72 3

0.44 14 0.89 0.48 74 A

1 127 04C 1

-0.59 16 1.34 0.42 64 2

-0.51 9

1.25 0.42 64 3

0.44 6

0.9 0.39 59 A

2 16 04C 1

-0.55 3

0.29 0.44 67 2

0.48 5

0.37 0.49 76 A

2 38 04C 1

-0.44 4

1.06 0.44 67 A

2 100 03C 1

-0.4 3

1.15 0.44 67 2

0 5

1.07 0.39 59 A

2 100 04C 1

0.03 7

1.09 0.39 59 2

0.17 9

1.04 0.4 61 A

59 115 04C 1

0.43 8

0.7 0.38 58 A

77 99 05C 1

-0.35 9

1.29 0.4 61 A

102 80 03C 1

-0.69 3

0.51 0.47 72 2

0.49 10 0.8 0.47 72 A

102 80 04C 1

-0.64 7

0.61 0.48 74 2

0.51 12 0.81 0.43 66 B

1 5

04C 1

-0.51 3

0.41 0.4 61

Page 7 SG Row Col TSP TSP Land Distance from TSP Centerline (inch)

Max.

Depth

%TW Max.

Length (inches)

Max.

Width1 (Inch)

Max.

Width (Degrees) 2

-0.5 4

0.93 0.37 56 3

0.4 12 0.93 0.37 56 B

1 15 04C 1

-0.51 8

0.85 0.38 58 2

0.42 13 0.76 0.33 50 B

1 69 03C 1

-0.47 2

0.58 0.25 39 2

0.41 8

1.22 0.35 53 3

0.41 8

1.19 0.27 42 B

1 69 04C 1

-0.47 4

1.11 0.26 40 2

0.36 8

1.16 0.29 45 3

0.4 5

1.19 0.28 43 B

1 123 03C 1

-0.48 6

0.57 0.44 67 2

0.37 5

0.82 0.33 50 3

0.39 7

0.46 0.33 50 B

2 12 04C 1

-0.49 7

1.06 0.31 47 2

-0.42 8

0.92 0.31 47 3

0.4 7

0.44 0.28 43 B

2 26 03C 1

-0.44 3

0.5 0.27 42 2

0.39 3

1.13 0.31 47 3

0.49 2

0.3 0.31 47 B

2 38 04C 1

0.36 9

1.14 0.31 47 2

0.37 1

0.94 0.25 39 B

2 124 03C 1

-0.36 1

0.31 0.41 63 2

0.47 6

0.89 0.39 59

Page 8 SG Row Col TSP TSP Land Distance from TSP Centerline (inch)

Max.

Depth

%TW Max.

Length (inches)

Max.

Width1 (Inch)

Max.

Width (Degrees) 3 0.49 7

0.47 0.44 67 B

4 36 03C 1

0.34 7

1.19 0.32 48 B

4 36 04C 1

-0.49 4

1.11 0.27 42 2

0.45 6

1.14 0.26 40 Note 1: Actual width of trefoil land is 0.18 inch (nominal) 6.0 Number of tubes plugged during the inspection outage for each active degradation mechanism (3.17(3)e)

Zero (0) tubes were plugged during the FCR27 outage.

7.0 Total number and percentage of tubes plugged to date and the effective plugging percentage in each steam generator (3.17(3)f and 3.17(3)h)

Table 3 provides the post-FCR27 outage tube plugging status of the FCS SGs.

Table 3 - Tube Plugging Summary SG Tubes Installed Tubes plugged to-Date (plugging percentage*)

SG-A 5200 0 (0.000%)

SG-B 5200 1 (0.019%)

Total 10,400 1 (0.010%)

There are no sleeves installed in the FCS steam generators, therefore the effective plugging percentage is the same as stated in Table 3 above.

8.0 The results of condition monitoring, including results of tube pulls and in-situ testing (3.17(3)g) 8.1 Broached (Trefoil) Tube Support Plate (TSP) Wear Based on the inspection data in comparison to the limits in the Degradation Assessment, structural integrity requirements have been met at the FCR27 inspection. The largest TSP wear depth found was 16% through-wall (TW) compared to the condition monitoring limit of 42%TW for a 1.4 inch long uniform wear flaw. Regarding TSP wear locations, satisfaction of structural integrity implies satisfaction of leakage integrity at accident conditions since steam line break accident loading condition does not result in ligament tearing or pop-through for volumetric flaws with an axial extent greater than 0.25 inch. For TSP volumetric axial thinning greater than 0.25

Page 9 inch in length, pop through and tube burst are coincident. Therefore, condition monitoring has been satisfied for degradation associated with TSP wear indications at the Fort Calhoun FCR27 inspection. All reported degradation falls below the applicable condition monitoring limit and therefore satisfies the Technical Specification structural performance criteria. Therefore, no tube-pulls or in-situ pressure testing were required.

8.2 Operational Leakage Criterion and Validation of Previous Operational Assessment (OA)

The operational leakage criterion was also satisfied by the absence of any measureable primary to secondary leakage since the previous inspection.

The results of the 2015 inspection and the condition monitoring assessment confirm that the 2011 Operational Assessment was appropriately bounding.