ML18155A487

Watts Bar, Unit 2 - Response to Request for Additional Information Regarding Cycle 1 Steam Generator Tube Inspection Report

ML18155A487
Person / Time
Site:	Watts Bar
Issue date:	06/04/2018
From:	Simmons P Tennessee Valley Authority
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
EPID L-2018-LRO-0010
Download: ML18155A487 (20)

Text

Tennessee Valley Authority, Post Office Box 2000, Spring City, Tennessee 37381June 4, 2018ATTN: Document Control DeskU.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001

Subject:

Watts Bar Nuclear Plant, Unit 2Facility Operating License No. NPF-96NRC Docket No. 50-39'1Response to Request for Additional lnformation Regarding Watts Bar NuclearPlant (WBN) Unit 2 Cycle 1 Steam Generator Tube lnspection Report(EPr D-L-201 8-LRO-00 1 0)References 1 ,TVA letter to NRC, "Watts Bar Nuclear Plant WBN) Unit 2 - Cycle 1Steam Generator Tube lnspection Report," dated February 16,2018(M1180474370)

2. NRC Electronic Mail to TVA, "Request for Additional lnformation Re:Unit 2 Cycle 1 Steam Generator Tube lnspection Report(EPID L-2018-LRO-0010)," dated May 3, 2018 (ML18123A394) ln Reference 1, Tennessee Valley Authority (rvA) submitted the 180-day steam generatorinspection report for the Watts Bar Nuclear Plant (WBN) Unit 2 Cycle 1 in accordance withthe requirements of WBN Technical Specification (TS) 5.9.9, "Steam Generator Tubelnspection Report." ln Reference 2, the Nuclear Regulatory Commission (NRC)transmitted a request for additional information (RAl) regarding Reference 1 and requestedTVA respond by June 4, 2018. Enclosure 1 to this letter provides the TVA response to theNRC RAI.As noted in Enclosure 1, TVA discovered two errors in Reference

1. Further details of theerrors are provided in Enclosure

1. Enclosure 2 to this letter contains the revisedWBN Unit 2, Cycle 1 Steam Generator Tube lnspection Report. The errors in Reference 1have been entered into the TVA corrective action program.There are no new regulatory commitments in this submittal.

Please direct any questionsconcerning this matter to Kim Hulvey, Site Licensing Manager, at423-365-7720.

U.S. Nuclear Regulatory Commission Page 2June 4, 2018Respectfull Paul SimmonsSite Vice PresidentWatts Bar Nuclear Plant

Enclosures:

cc (Enclosures):

NRC RegionalAdministrator - Region IlNRC Project Manager - Watts Bar Nuclear Plant, Unit 1NRC Senior Resident lnspector - Watts Bar Nuclear Plant, Unit 11.2.Response to Request for Additional lnformation (RAl) Regarding the Watts BarNuclear Plant Unit 2 Cycle 1 Steam Generator Tube Inspection Report(EPr D-L-201 8-LRO-001 0)Revised Watts Bar Nuclear Plant Unit 2 - Cycle 1 Steam Generator Tube Inspection Report Enclosure 1Response to Request forAdditional lnformation (Ml) Regarding the Watts Bar NuclearPlant Unit 2Cycle 1 Steam GeneratorTube Inspection Report (EPID-L-2018-LRO-0010)

Nuclear Reoulatorv Commission (NRC) RA!"By letter dated February 16, 2018 (Agencywide Documents Access and Management Sysfem Accession No. ML18047A370),Iennessee Valley Authortty, (the licensee)submitted the results of the steam generator (SG) inspections performed at Wafts BarNuclear Plant, Unit 2. These inspections were pefformed during the first refueling outage(RFO 1).ln orderto complete its review, the U.S. Nuclear Regulatory Commission staff requesfs fhefollowing addition al inform ation :1. The technical specifications require that a condition monitoring assessmenf beperformed, during each outage duing which fhe SG tubes are inspected or plugged, toconfirm that the performance criteia are being met. The technical specifications a/sorequire the reporting of location, orientation (if linear), and measured sizes (if available) of seruice-induced indications.

Please clarify the information provided in the reportfrom RFO 1 regarding the following:

a. Section 2.0, "180 Day Steam GeneratorTube lnspection Repoft," sub-section e.,Number of Tubes Plugged During the lnspection Outage for Each Degradation Mechanism," dlscusses a permeability variation found in a tube rn SG 3. Pleasedrscuss how signal injection and engineering assessment were used to develop thebasrs for concluding that condition monitoring was confirmed.

b. Please describe the anomaly in tube support plate H01 of SG 3 (identified duringinspection) and how it affected the eddy cunent inspection scope performed in thetubes near Row 47, Column 48.c. Please confirm that all seruice-induced indications that were identified during thefall 2017 outage were included in the February 16, 2018, repoft."TVA Response to RAI 1.a.The following information describes how signal injection and engineering assessment wereused to develop the basis for concluding that condition monitoring was confirmed.

Permeability variation (PVN) indications are noise interferences that could compromise detection of flaw signals. lnspections with magnetically biased +Point@1 probe and Ghentcoils have typically been used to disposition PVN indications.

This was the case for allPVN indications during the Watts Bar Nuclear Plant WBN) Unit 2, first refueling outage(U2R1) with the exception of one indication.

Tube R24-C97 in SG 3 contained a largePVN indication that was not completely suppressed with the magnetically biased +Pointprobe. This PVN indication was plugged during WBN U2R1. The PVN indication waslocated in the U-bend region extending from tube support plate (TSP) H08+7.66 inches toTSP H08+22.75 inches and encompassed an anti-vibration bar(AVB) support location.The maximum bobbin coilvoltage was2.2l volts and was reduced to only 1.22 volts with1 +Point is a registered trademark of Zetec, lnc.E1-1 of 5 Enclosure 1the magnetically biased U-bend +POINT probe. Being in the U-bend region, the solid bodyGhent probe could not be used to suppress the signal. Condition monitoring of thisindication was accomplished by a combination of signal injection technology andengineering assessment.

The potential degradation mechanisms within the area of thePVN were AVB wear and tubeto-tube contact wear. Axial outside diameter stresscorrosion cracking (ODSCC) was identified as a potentialdegradation mechanism in thefreespan while primary water stress conosion cracking (PWSCC) was identified as apotential degradation mechanism only for low row U-bends. Cracking was not a potentialdegradation mechanism for the PVN location in tube R24-C97.AVB wear was observed during the current inspection and is considered an existingdegradation mechanism that may be masked by the PVN in Tube R24-C97. The use ofsignal injection technology was also used to confirm condition monitoring requirements have been satisfied.

Signal injection uses the Westinghouse Data Union Software (DUS)to inject a structurally significant size flaw into the PVN signal and verifying that theresultant signal shows evidence of the injected flaw. The actual PVN indication would thenbe reviewed for similar flaw evidence.

The basis for acceptable condition monitoring is thatthe resultant injected flaw and PVN signal show evidence of a structurally significant flaw;no similar evidence is discernable in the actual PVN signals. ln this case, a 40 percent (%)through-wall (TW) AVB wear flaw was injected into the PVN signa! at all locations along thelength of the PVN. The 4lo/oTW flaw is smaller than the condition monitoring limit of 660/oTW. The injected AVB wear was discernable at all injected locations, therebydemonstrating acceptable condition monitoring.

This conclusion encompasses any tubevolumetric degradation, which would occur due to tubetotube contact.WBN Unit 2 operated only 0.74 effective full power years (EFPY) in its first cycle ofoperation.

No indications of ODSCC or PWSCC were identified in any tube location duringthis SG inspection, including expansion transitions, low row U-bends, dents/dings, TSPintersections, or other locations of higher residual stress. Experience from other alloy600 mitl annealed (MA) plants, including the originalWBN Unit 1 SGs, indicates that theonset of SCC occurs in regions of higher residua! stress prior to occurring in tube locationscontaining lower residual stress. Several plants with alloy 600MA tubing and moreoperating life have experienced SCC in the U-bend region up to tube row 13. However, no' plant has detected SCC in higher tube rows such asthe WBN Unit 2 PVN affected tube atrow 24. The original WBN Unit 1 SGs experienced SCC in the U-bend region but this waslimited to tubes in row four or less and was detected after five full cycles of operation.

Therefore, it is highly unlikely that PWSCC or ODSCC has initiated within the limited tubelength encompassed by the PVN after only 0.74 EFPY.Although SCC within the PVN affected area of SG3 tube R24-C97 is highly improbable, Westinghouse's eddy current signal injection technology was used to determine if astructurally significant crack could be discerned.

A 60% TW axial inner diameter (lD) and a60% TW axial outside diameter (OD) electric discharge machining (EDM) notch signalwasscaled in signal amplitude to simulate 60% TW axial !D and OD cracks. These simulatedflaws were injected into the most severe region of the PVN. The resultant injected signalshad little to marginal discernibility in terms of amplitude and phase change. There was aslight opening of the lissajous lobe at the area of the injected OD and lD flaw. There wasno similar type of lissajous opening observed when the unaltered PVN signal was reviewedon a scan line basis. Additionally, there was no phase shift of the unaltered PVN signalthroughout the indication when compared side-by-side to the baseline data. Bycomparison, a 4Oo/oTW axial lD EDM notch signal was injected into the most severe PVNE1-2 of 5 Enclosure 1location and amplitude and phase changes were both readily detected.

Although theresults of the signal injection were marginal in terms of discerning structurally significant SCC, the results provided sufficient assurance that a significant flaw could be discernedunder strict data review for lissajous lobe openings and a side-by-side comparison to thebaseline data as is typically performed for PVN signals.TVA Response to RAI 1.b.The anomaly identified in tube support plate H01 of SG 3 was categorized as a partialsupport plate (PSP) indication.

The PSP would be designated to the diminished supportplate signal. The PSP was present in the preservice inspection and scheduled for followup monitoring at subsequent in-service inspections.

Westinghouse performed a detailedreview of the condition and potential causes. The review of this signal suggests that theresponse of the low frequency on the +Point probe is the result of installation of a patchplate in the carbon steel tube support plate. The basis for this determination is providedbelow:o The TSPs are inspected in the manufacturing shop following the drilling of the tubesupports.

The TSPs are only deemed acceptable for installation if found to be freefrom surface defects.o With no operating history on the WBN Unit 2 SGs at the time of the inspection, a crackwould have had to be initiated either during installation of the tube or sometime duringthe extended layup period..o The alloy 600MA tube material has a Rockwell hardness less than the carbon steeltube support plate. Any damage to the support plate would have caused significantly more degradation to the tube and to the support. Only a lowJevel manufacturing burnish mark was detected in the U-bend region on the affected tube during thepre-service inspection.

o lf the cause were to have been systematic due to an unidentified SG layup process, itis presumable that further tubetotube support plate intersections would be affected bythe mechanism.

No other such indications were detected during the 100% bobbin pre-service inspection or during this inspection.

o The tube location in SG3 Row 47 Column 48 at H01 aligns with the weld seambetween the tube support plate and the patch plate installed while tubing the SG. Thedesign drawings for the patch plate specifically indicate that care should be taken whenwelding the patch plate into place.These observations support the conclusion that the PSP indication in SG3 is most likelythe result of installation and welding of a patch plate in the carbon steel tube support plateand did not affect the eddy current inspection scope performed in the tubes near Row 47,Column 48. The inspection scope for WBN U2R1 included a +Point probe inspection of alltubes surrounding tube location Row 47 Column 48 at H01 in SG3 which served as furtherconfirmation of this conclusion and to monitor the condition.

A 100o/o bobbin inspection capable of detecting these types of indications was also performed in al! SGs.The PSP indication did not change from preservice to U2R1. There is no indication thatdegradation exists in the support plate at this location.

A graphic of the PSP indication isprovided in Figures 1 and 2.E1-3 of 5 Enclosure 1Figure 1Eddy Current Testing (ECT) Graphic from Baseline 2010Figure 2ECT Graphic from U2R1 2017E1-4 of 5 Enclosure 1TVA Response to RAI 1.c.TVA confirms that the service-induced indications that were identified during the WBNUnit 2 Fall2Q17 (U2R1) outage were included in the referenced letter. However, as aresult of developing this RAl response, TVA identified the following errors in the referenced letter. The errors existed in Table 2-2 and Table 2-4 of the referenced letter as a result ofidentifying the tube at location SG 3 Row 40 Column 42 as not plugged when this tube wasactually plugged prior to initial SG service.. Table 2-2 of the referenced letter lists eight volumetric indications listed for SG 3 with atotal of 38 indications for all of the SGs. The correct value is seven volumetric indications with a total of 37 indications.

o Table 2-4 of the referenced letter did not list tube SG 3 Row 40 Column 42 as plugged.As noted above, this tube was plugged during preservice inspections in 2010.Therefore, the status of tube SG 3 Row 40 Column 42has been changed to plugged.Enclosure 2 contains a revision to the referenced letter correcting the above errors.TVA letter to NRC, "Watts Bar Nuclear Plant WBN) Unit 2 - Cycle 1 Steam GeneratorTube lnspection Report," dated February 16,2018 (ML18047A370)

E1-5 of 5 Enclosure 2Revised Watts Bar Nuclear Plant Unit 2 - Cycle 1 Steam Generator Tube lnspection Report WgsrTxGHOUSn NoN-PnOpNIETARY CIaSS 3SG.SGMP-I 7-35Revision IMay 2018Watts Bar IJlRI180 Day Steam Generiltor Titbe Inspection Report WESTINGHOUSE NONI.PROPRIETARY CLASS 3SG-SGMP.I7-35 Revision 1Watts Bar U2Rl180 Day Steam GeneratorTube Inspection ReportPrepared for:Tennessee Valley AuthorityAuthor's Name:Jesse S. BaronSG Management ProgramsVerifier's Name:Jivan G. ThakkarSG Management ProgramsManager's Name:David P. Lytle, ManagerSG Management ProgramsReviewer's Name:' Jeremy W. MayoTVA SG Program ManagerReviewer's Name:Tammy C. SearsTVA Watts Bar SG Program OwnerRevie\ryer's Name:Daniel P. FolsomTVA NDE Level IIISignature I Date*Electron icallv A p prsvedFor PagesAllSignature I Date*EI ec troai caII v A n pro vedFor PagesA11Signature I Date*EI ec troni call v A n a ro vedFor PagesAIISignature I DateFor PagesAllSignature lDateFor PagesAilSignature lDateFor PagesA11* Electronically Approvd Records are Autbenticatd in the Elcctronic Document Managcmcnt System@2018 Westinghouse Electric Company LLCAII Rights ReservedSG-SGMP- 17 -35Revision IMay 2018Page 2 of 12 Record of RevisionsRevisionDatef)escription 0aDecember20t7Preliminary draft for Tennessee Valley Authority review and comment.0January2A18Incorporated comments from Tennessee Valley Authority, final approved and issued.IMay201 8Revised Table 2-2 volumetric indication count and Table 2-4 indication listing forSG3 tube Row 40 Column 42 to conectly show as plugged durit g the pre-service inspection.

Chanses are marked with bars on the rieht hand side.SG-SGMP- 17 -35Revision IMay 2018Page 3 of 12 Table of Contents2.0 180 Day Steam Generator Tube Inspection Report .........6

a. The Scope of Inspections Performed on each SG.............. ..................6

b. Degradation Mechanisrns Found.... ..............7

c. Nondestructive Examination (NDE) Techniques Utilized for Each Degradation Mechanism........................7

d. Location, Orientation (if Linear), and Measured Sizes (if Available) of Service Induced lndications ...........9

e. Number of Tubes Plugged During the Inspection Outage for Each Degradation Mechanism......................

10f. The Number and Percentage of Tubes Plugged to Date and the Effective Plugging Percentage in each SG l1g. The Results of Condition Monitoring, lncluding the Results of Tube Pulls and In-Situ Testing..................

I IList of Tables and FiguresFigure l-l: Tube Support Arrangement for Watts Bar Unit 2 Model D3 Steam Generators ..........5 Table 2-l: Watts Bar U2Rl Steam Generator Eddy Current Inspection Scope.......... .....................7 Table2-2:

Number of lndications Detected for Each Degradation Mechanism. .........7 Table 2-3: NDE Techniques for Each Existing or Potential Degradation Mechanism. ...................8 TableZ-4:

Watts BarU2Rl Volumetric Indications-All SGs....... .........9 Table 2-5: Watts Bar U2Rl Anti-Vibration Bar Wear Indications - All SGs.... ........l0 Table2-6:

Watts Bar U2Rl Tube Support Plate Wear Indications - All SGs.... ........10 Table}-7:

NumberofTubesPluggedforEachDegradationMechanism. ................11 May 2018Page 4 of 12SG-SGMP- 17 -35Revision I

1.0 Introduction

The first in-service inspections (ISI) of the Watts Bar Unit 2 (WBN2) steam generators (SGs) wereperformed during the fall 2017 refueling outage designated as U2Rl. The U2Rl inspection wasperformed afrer 0.74 effective full power years (EFPY) of plant operation.

The inspections included eddycurrent testing of the SG tubing as well as primary side visual inspections, secondary side visualinspections and secondary side cleanings.

This report documents the "Watts Bar U2Rl 180-Day SteamGenerator Tube Inspection Report" as required by the WBN2 Technical Specifications.

The steamgenerators at WBN2 are a Westinghouse Model D3 preheater-type design where the majority of thefeedwater enters near the top of the tubesheet on the cold leg side and the tubing is made from millannealed Alloy 600 (Alloy 600MA) matedal. Figure 1-l below provides the arrangement and locationdesignation of the tube support structures for the WBN2 SGs.Figure 1-1: Tube Support Arrangement for Watts Bar Unit 2 Model D3 Steam Generators Notes: HICIAV: Hot Leg SupportiCold Leg Support/Anti-Vibration Bar (AVB) LocationHTS/CTS : Hot Leg Top ofTubesheeVCold Leg Top ofTubesheet HTE/CTE: Hot Leg Tube End/Cold Leg Tube EndSG-SGMP-17-35 Revision IMay 2018Page 5 of 12 2.0 180 Day Steam Generator Tube Inspection ReportIn accordance with WBN2 Technical Specification Section 5.7.2.72, "Steam Generator Program", andTechnical Specification Section 5.9.9, "Steam Generator Tube lnspection Report", this report documentsthe scope and results of the Watts Bar U2RI SG inspections.

There are seven specific reportingrequirements associated with the Technical Specification.

Each lettered reporting requirement listedbelow is followed with the associated information based on the inspections performed during U2Rl.a. The Scope of Inspections Performed on each SGThe inspection progftrm addressed the known degradation observed in the Watts Bar Unit 2 SGsduring the pre-sewice inspection, potential in-service SG tube degradation mechanisms andincluded proactive examinations to address areas where no degradation is anticipatdd butmonitoring is performed regardless.

The inspections were performed with qualified non-destructive examination (NDE) techniques for each existing and potential mechanism.

Thedefined scope that was implemented in all four SGs included:. l00yo bobbin inspection of all open tubes in all four SGs full length and tube Rows I through 4to the top fube support plate from both the hot leg (HL) and cold leg (CL) sides.. 100% +POINT probe inspection of tube Rows I through 4 from the top tube support plate onthe HL side to the top tube support plate on the CL side.. +POINT probe 'Special Interest' inspections of rube locations with non-resolved bobbin and/orArray probe signals.. l00o +POINT probe inspection of the hot leg top of tubesheet region from HTS+21-2 inches.. SOyo Combination bobbin and Array probe inspection from C06 to CTS-2 inches. Thisinspection included all CL peripheral tubes two (2) tubes deep.. I00oA +POINT or Array probe inspection of DNTs and DNGs > 5 Volts in the HL straightlengths, U-bends and the top tube support plate (TSP) on the CL side. 20yo +POINT or Array probe inspection of all DNTs and DNGs > 2 Volts. IAA0A +POINT probe inspection of any DNT or DNG signal located within L0 inch or less of amanufacturing burnish mark (MBM).. +POINT or Array probe inspection of tubes surrounding known locations of foreign objectsfrom the pre-service inspection.. +POINT or Anay probe inspection of all tubes within a two (2) tube pitch of the regionsurrounding any foreign object wear or possible loose part (PLP) locations.. +POINT probe inspection of SG3 tube Row 47 Column 48 at HOl and all tubes within one (l)tube of this location at the same elevation.

An anomaly in the support plate was identified atthis location during the pre-service inspection.. +POINT probe inspection of bobbin tube-to-tube proximity (PRO or PRX) signals >1.25 Volt.. 100% visual inspection of all installed tube plugs from the primary side on both the HL and CLside.. Visual inspection in all SGs of channel head primary side HL and CL inclusive of the entiredivider plate to channel head weld and all visible clad surfaces.The Watts Bar U2Rl inspection included all tubes with prior indications of degradation.

Thetable below summarizes the number and type of eddy current examinations performed duringU2Rl excluding the special interest inspection scope.SG-SGMP-17.35 Revision IMay 2018Page 6 of 12 Table 2-1: Watts Bar U2Rl Steam Generator Eddy Current Inspection ScopeEddy Current Exam TypesGlSG2SG3SG4Full Length Bobbin4,2004,1974,2144,,204CL R1-R4 Low Row Bobbin45t452454454HL Rl-R4 Low Row Bobbin4514s04s4454U-Bend +Point Rl-R44514s0454454HL +Point Tubesheet4,6514,6474,6694,658CL Straight Leg X-Probe C06 to CTS-2 inch2,6622,6492,7ll2,705ln addition to the NDE and primary side inspections discussed, visual inspection was performedin all SGs in order to determine the deposit and foreign object removal effectiveness of thetubesheet cleaning process applied. This was followed by a foreign object search and retrieval(FOSAR) inspection performed at the top of the tubesheet in all four SGs. Finally, visualinspection was also performed of the SG upper internal components in SGl and SG4 duringWatts Bar U2Rl.b. Degradation Mechanisms FoundVolumetric tube weal was the only degradation mechanism detected during the U2Rl inspection.

All of the in-service volumetric wear indications detected were located at tube intersections witheither TSPs or AVBs. Volumetric indications generated during tube manufacture and bundleassembly and initially detected during the pre-service inspections were also detected duringU2Rl. These indications are not considered an active or ongoing in-service degradation mechanism but are listed for completeness.

Table 2-2 below shows the number of indications reported for each degradation mechanism during the U2Rl inspections.

It is notable that noindications of stress corrosion cracking were detected during U2R1.Table 2-2: Nurnber of Indications Detected for Each Degradation MechanismDegradation MechanismsGlSG2SG3SG4TotalVolumetric Indications (Pre-Service) t2371537Wear at Tube Support Plates00505Wear at Anti-Vibration Bars35I8t7c. Nondestructive Examination (NDE) Techniques Utilized for Each Degradation MechanismTable 2-3 provides the NDE techniques that were used for the detection of each degradation mechanism considered as existing or potential for the U2Rl inspection.

NDE techniques are alsolisted which were available for diagnostic testing, resolution and confirmation of anomalousindications.

All the examination technique specification sheets (ETSSs) used during U2Rl arefrom the electric power research institute (EPRI) database.

In some cases a variable 'X' is usedin the listing of techniques in Table 2-3 which is in reference to a series of ETSSs.SG-SGMP-17-35 Revision IMay 2018Page 7 of 12 Table 2-3: NDE Techniques for Each Existing or Potential Degradation MechanismDesradation MechanismETSS Detection TechniqueVolumetric Indications due toTube Fabrication and Installation (Pre-Service)

B: 27091.1B:27491 .2Wear atAVBsB: 96041.1* Pt: I 0908.4A: 1 7908.1A: 17908.2Wear atTube Support PlatesB:96442.1A: 1 1956.1A: I 1956.2A: I 1956.3A: 11956.4Wear due toForeign ObjectsB:27091.1+Pt: 21998.1A: 1790X.1A: l79AX.3Tube-to-Tube Contact WearB: 13091.2OD Pitting of the Tube MaterialB: 96005.3-t-Pt: 21998.1A:24998.1Axial and Circumf'erential PWSCC at the TTS*Pt: 20511.1 Ax+Pt: ll1524 CirA: 20501.1 AxA: 20500.1 CirAxial and Circumferential ODSCC at the TTS+Pt: 128424 Ax+Pt: 128425 Ax+Pt: 21410.1 ClirA: 20400.1 Ax/CirAxial ODSCCat TubeSupport PlatesB: 12841I*'Pt: 128424A:20402.1Axial and Circumferential PWSCC in theLow Row U-bends+Pt: 96511.2or+Pt: 99997.1ODSCC atTube Dents and DingsB: 128411orB:24013.1B: I 0013.1"t-Pt: 22401.1A: 20400.1A: 20403.1SCC at Tube Bulges andOverexpansions

+Pt: 128424 Ax+Pt: 128425 Ax+'Pt: 21410.1 CirA: 20400.1 Ax/CirAxial ODSCC in the FreespanB: 12841 3A: 20403.lODSCC at Dents and DingsCoincident with an MBM+-Pt: 22401.1AnomalousIndications Gh: 20406.1Gh: 20507.1Gh: 20508.1Gh: 20509.1Acronym Definitions for Table 2-3+Pt: +POIN]'Probe Gh: Ghent ProbeA: Aray Probe MBM: Manufactutng Bumish MarkAVB: Anti-Vibration Bar OD: Outcr DianreterAx: Axial ODSCCI: Outer f)iameter Stress C)orrosion CrackingB: Bobbin Probe PWSCC: Prinrary Water Stress Clorrosi<ln CrackingCir: Circrunfcrential SCC: Stress Clonosion CrackingETSS: Eddy cun:cnt Techniquc Specitication Sheet TTS: Top of the Tubeslteet SG-SGMP-17.35 Revision IMay 2018Page 8 of 12

d. Location, Orientation (if Linear), and Measured Sizes (if Available) of Service InducedIndications Table 2-4 through Table 2-6 provide a listing of all pre-service and service-induced indications reported during the U2Rl inspection including the estimated depths from the associated NDEtechnique and an indication of whether the tube has been plugged.Table 2-4: Watts Bar U2RI Pre-Service Volumetric Indications - All SGsSGRowColLocationInch IIndication

%TWPlugged?I399c140.08VOL17NoIl038ct2-2.5VOL23NoIl055cl332.84VOLl1NoI14l8AVI3.8VOLl4NoIl498cr4t.69VOLr8No13073cl328.64VOL5No13r80ct41.84VOL24NoI3265c0r2.4VOLINoI3477c141.52VOL6No.l::j : :a : :i::::::lI::!

t$$ii::iii:iliiliiiii:i

$,i*:,,.il.lir

f?.:.:.:::r::l iiiliiiffi ffiI $ ,,.i # $13s48H082s.26VOLl8No:l::i.r.j.$ff t?)ixr$:$:l:ill.:':':,:,:,:':,:,:i i :.Yes,- frf--S$t$icc

I3972ct4-2.7 4VOLt7NoI4173CL40.58VOL30No25103H010.62VOLt4No221l6H0333.19VOLt1NoaJl937cl32.7 4VOLl4NoaJ214lc0l3.88VOL20No32243c0r1.46VOLl5NoaJ3246H014.89VOL32No33842H011.75VOL21No33842H0lt.t2VOL1lNoaJ3858H0640.09VOL21No.iiiss4187HOI2.96VOL2tNo4,l)106clr12.01VOL10No4I84H0421 .84VOL2ANo4944H0730.94VOLl0No42A46H0722.43VOLl0No42131H069.99VOLl8No42l3lH062.03VOLt4No42t3lH06t7.06VOL7No4^taZ)77c097.31VOLl3No42746H0422.89VOL20No42895H0425,6voLl6No429100cr336.1VOL30No43014c109.02VOL2ANo43035c108.8VOL23No43884H043.39VOLl5NoSG-SGMP-17-35 Revision IMay 2018Page 9 of 12 SG-SGMP-17-35 Revision ITabte 2-5

Watts Bar I.J2RI Anti-Vibration Bar Wear Indications - All SGsSGRowColLocationInchlIndication

%TWPlugged?I407qAY20.16PCTl0NoI3691AV30.1 8PCTt4No124108AV40.1 8PCT15No2248AV20PCT15No23585AV30PCTt4No2228AV30.19PCT1lNo23023AV40.12PCTl1No2236AV40.16PCTl3NoaJ30105AV3-0.09PCT10No44265AV30.16PCT1lNo14440AV30PCT8No4424AAV30.21PCTt4No44236AV3-0.1 1PCTr5No44428AV30.12PCTl0No44124AV30.17PCT8No43821AV30.t2PCT14No4,t 7')JJt4AV30.04PCT9NoTable 2-6: Watts Bar U2R I Tube Support Plate Wear Indications - All SGsSGRowColLocationInchlIndication

%TWPlugged?34866c06-0.26PCT5No34961c06-0.2rPCT7NoaJ4960c06-0.r4PCTl0No34860c06-0.28PCTllNoaJ4760c06-0.24PCTt4Noe. Number of Tubes Plugged During the Inspection Outage for Each Degradation MechanismThere were eight (8) tubes plugged during the Watts Bar U2R1 SG in-service inspection.

Onlyone (l) tube was required to be plugged in accordance with the plant Technical Specification requirements due to having a measured depth of 40% through-wall (TW) or greater. This was a46o/oTW volumetric indication in SG2 at tube location Row 5 Column 110 as listed in TabLe 2-4.The remainders of tubes were plugged for preventative measures including two (2) restricted tubes, one (1) for an indication of permeability variation and four (4) due to a foreign objectwhich was unable to be retrieved.

Table 2-7 below provides the numbers and percentages oftubes plugged following U2R1 and the subsequent sections elaborate on the plugging basis.Regarding the restricted tube locations in SG2 which were plugged, a complete test of the fulltube length was not able to be obtained on the first data collection attempt. However, a full testwas later completed using altemate means which included data collection from the opposite legand/or use of downsized eddy current probes. These tubes were plugged in order to eliminate thepossibility of being unable to collect data along the full tube length and evaluate conditionmonitoring at future inspections.

One tube in SG3 had a permeability variation indication in the tube material fully containedwithin the U-bend region. The use of alternative eddy current probes, such as a Ghent probe, tocleaf, the permeability indication was not an option since the solid body Ghent probe is notcapable of traversing the U-bend region. Condition monitoring was subsequently demonstrated through the use of a combination of simulated eddy current flaw signal injection and engineering May 2018Pagel0of12 SG-SGMP-17-35 Revision 1assessment of potential degradation mechanisms.

This tube was plugged for its potential to maskdegradation at future inspections.

Four tubes were plugged due to a foreign object located at the hot leg top of tubesheet which wasunable to be retrieved.

The object was identified as a piece of weld slag which was rigidlycontained in between a group of four tubes. Multiple retrieval attempts from all accessible anglesand orientations were made to remove the object from the SG and none were successful.

Three ofthe tubes surrounding the foreign object had possible loose part indications (PLPs) from the eddycurrent test program, although none of the four surrounding tubes had indications of wear. Al1four tubes were stabilized with a cable stabilizer tmversing the tubesheet region and plugged.Table 2-7: Number of Tubes Plugged for each Degradation MechanismSGISG2SG3I scn .JTotalPlugged Tubes prior to U2Rl2327;JfuJ72rr3ffiVolumetric lndication frorn Pre-Service 0I00IRestricted Tube02002Permeability Variation00I01ForeiEr Obect Unable to be Retrieved04004,lri!rlTotal Pluggecl Following U2R1233471680Percentage Plugged Following U2Rl0.49%0.73%a)5%0.34%a.$%The Number and Percentage of Tubes Plugged to Date and the Effective PluggingPercentage in each SGTable 2-7 in the previous section provides the nurnber and percentage of tubes plugged to date.g. The Results of Condition Monitoring, Including the Results of Tube Pulls and In-SituTestingCondition Monitorine.

Tube Pulls and In-Situ TestingA condition monitoring (CM) assessment was perfiormed as required by the Watts Bar Unlt 2steam generator program. Volumetric tube wear was the only in-service degradation mechanismdetected during the Watts Bar U2Rl inspection.

All of the in-service volumetric wear indications detected were located at tube intersections with either AVBs or TSPs. Volumetric indications generated during tube manufacture and assembly and initially detected during the pre-service inspections were also detected during U2Rl.The deepest indication of AVB wear had an estimated depth of l5%TW which is significantly less than a conservatively determined CM limil of 66yoTW. The deepest indication of TSP wearhad an estirnated depth of l4%TW which is significantly less than a conservatively determined CM limit of 64%TW. The largest volumetric indication from the pre-service measured 46%TWwhich is less than a conservatively determined CM limit of 58%TW. These CM limits includeuncertainties for material properties, NDE depth sizing, and the burst pressure relationship.

Sincethe deepest flaws have an estimated depth less than the associated CM limits, the structural integrity performance criterion was met for the operating interval preceding U2Rl.The limiting pressure differential associated with accident induced leakage integrity is muchlower than the three times normal operating pressure differential associated with the CM limitsMay 201 8Page I I of12 for structural integtty.

Therefore, CM for accident-induced leakage integrity was alsodemonstrated since volumetric wear indications will leak and burst at essentially the samepressure.

Operational leakage integrity was demonstrated by the absence of any detectable primary-to-secondary leakage during the operating interval prior to U2Rl. Since tube integritywas demonstrated analytically, in-situ pressure testing was not required nor performed during theU2Rl outage. No tube pulls were planned or performed during U2Rl.Primarv and Secondary Side Visual Inspection ResultsVisual inspections were performed on both the primary and secondary sides during U2R1 inaccordance with Westinghouse nuclear safety advisory letter NSAL-12-1.

Primary sideinspections included visual inspections of all previously installed tube plugs as well as thechannel head bowl cladding and the divider plate. The installed tube plug inspections showed noconditions indicative of degradation.

However, &e inspections of the channel head bowlcladding and the divider plate showed visually appaxent evidence of minor indications ofdegradation of the cladding in SGl located on the hot leg side just above the primary manwayopening. A site condition report (CR) was initiated to document the condition and an associated engineering evaluation was performed.

The conclusion of the engineering evaluation was thatacceptable margin exists for maintenance of structural integrity of the SG channel head basemetal for at least six cycles of operation.

Prior to the secondary side foreign object search and retrieval (FOSAR) inspections, sludge,scale, foreign objects, and other deposit accumulations at the top of the tubesheet were removedas part of the top of tubesheet sludge lancing process. The secondary side FOSAR inspections performed in all four SGs included visual examination of tube bundle periphery tubes from thehot leg and cold leg annulus and center no-tube lane. A total of 25 foreign objects were removedfrom the top of the tubesheet region. Any foreign objects not able to be retrieved werecharucteized and an analysis performed to demonstrate acceptability of continued operationwithout exceeding the tube integrity performance criteria.

A limited top of tubesheet in-bundlevisual inspection was also performed for the purpose of assessing and trending the level ofhardened deposit buildup in the kidney region. Finally, there were no structurally significant anomalies observed during inspection of the upper internals of SGI and SG4. Only a limitedamount of foreign material was observed and retrieved during the upper internals inspections.

Therefore, no potential for the upper internal components to have an effect on SG tube integrity.

SG-SGMP-17-35 Revision 1May 2018Page 12 of 12