ML18155A487

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 37381 June 4, 2018 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Watts Bar Nuclear Plant, Unit 2 Facility Operating License No. NPF-96 NRC Docket No. 50-39'1

Subject:

Response to Request for Additional lnformation Regarding Watts Bar Nuclear Plant (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 1 Steam 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 generator inspection report for the Watts Bar Nuclear Plant (WBN) Unit 2 Cycle 1 in accordance with the requirements of WBN Technical Specification (TS) 5.9.9, "Steam Generator Tube lnspection Report." ln Reference 2, the Nuclear Regulatory Commission (NRC) transmitted a request for additional information (RAl) regarding Reference 1 and requested TVA respond by June 4, 2018. Enclosure 1 to this letter provides the TVA response to the NRC RAI.

As noted in Enclosure 1, TVA discovered two errors in Reference 1. Further details of the errors are provided in Enclosure 1. Enclosure 2 to this letter contains the revised WBN Unit 2, Cycle 1 Steam Generator Tube lnspection Report. The errors in Reference 1 have been entered into the TVA corrective action program.

There are no new regulatory commitments in this submittal. Please direct any questions concerning this matter to Kim Hulvey, Site Licensing Manager, at423-365-7720.

U.S. Nuclear Regulatory Commission Page 2 June 4, 2018 Respectfull Paul Simmons Site Vice President Watts Bar Nuclear Plant

Enclosures:

1. Response to Request for Additional lnformation (RAl) Regarding the Watts Bar Nuclear Plant Unit 2 Cycle 1 Steam Generator Tube Inspection Report (EPr D-L-201 8-LRO-001 0)

2. Revised Watts Bar Nuclear Plant Unit 2 - Cycle 1 Steam Generator Tube Inspection Report cc (Enclosures):

NRC RegionalAdministrator - Region Il NRC Project Manager - Watts Bar Nuclear Plant, Unit 1 NRC Senior Resident lnspector - Watts Bar Nuclear Plant, Unit 1

Enclosure 1 Response to Request forAdditional lnformation (Ml) Regarding the Watts Bar Nuclear Plant 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 Bar Nuclear 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 fhe following addition al inform ation :

1. The technical specifications require that a condition monitoring assessmenf be performed, during each outage duing which fhe SG tubes are inspected or plugged, to confirm that the performance criteia are being met. The technical specifications a/so require the reporting of location, orientation (if linear), and measured sizes (if available) of seruice-induced indications. Please clarify the information provided in the report from 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. Please drscuss how signal injection and engineering assessment were used to develop the basrs for concluding that condition monitoring was confirmed.

b. Please describe the anomaly in tube support plate H01 of SG 3 (identified during inspection) and how it affected the eddy cunent inspection scope performed in the tubes near Row 47, Column 48.

c. Please confirm that all seruice-induced indications that were identified during the fall 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 were used 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 Ghent coils have typically been used to disposition PVN indications. This was the case for all PVN 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 large PVN indication that was not completely suppressed with the magnetically biased +Point probe. This PVN indication was plugged during WBN U2R1. The PVN indication was located in the U-bend region extending from tube support plate (TSP) H08+7.66 inches to TSP 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 with 1

+Point is a registered trademark of Zetec, lnc.

E1-1 of 5

Enclosure 1 the magnetically biased U-bend +POINT probe. Being in the U-bend region, the solid body Ghent probe could not be used to suppress the signal. Condition monitoring of this indication was accomplished by a combination of signal injection technology and engineering assessment. The potential degradation mechanisms within the area of the PVN were AVB wear and tubeto-tube contact wear. Axial outside diameter stress corrosion cracking (ODSCC) was identified as a potentialdegradation mechanism in the freespan while primary water stress conosion cracking (PWSCC) was identified as a potential degradation mechanism only for low row U-bends. Cracking was not a potential degradation mechanism for the PVN location in tube R24-C97.

AVB wear was observed during the current inspection and is considered an existing degradation mechanism that may be masked by the PVN in Tube R24-C97. The use of signal 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 the resultant signal shows evidence of the injected flaw. The actual PVN indication would then be reviewed for similar flaw evidence. The basis for acceptable condition monitoring is that the 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 the length of the PVN. The 4lo/oTW flaw is smaller than the condition monitoring limit of 660/o TW. The injected AVB wear was discernable at all injected locations, thereby demonstrating acceptable condition monitoring. This conclusion encompasses any tube volumetric degradation, which would occur due to tubetotube contact.

WBN Unit 2 operated only 0.74 effective full power years (EFPY) in its first cycle of operation. No indications of ODSCC or PWSCC were identified in any tube location during this SG inspection, including expansion transitions, low row U-bends, dents/dings, TSP intersections, or other locations of higher residual stress. Experience from other alloy 600 mitl annealed (MA) plants, including the originalWBN Unit 1 SGs, indicates that the onset of SCC occurs in regions of higher residua! stress prior to occurring in tube locations containing lower residual stress. Several plants with alloy 600MA tubing and more operating 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 at row 24. The original WBN Unit 1 SGs experienced SCC in the U-bend region but this was limited 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 tube length 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 a structurally significant crack could be discerned. A 60% TW axial inner diameter (lD) and a 60% TW axial outside diameter (OD) electric discharge machining (EDM) notch signalwas scaled in signal amplitude to simulate 60% TW axial !D and OD cracks. These simulated flaws were injected into the most severe region of the PVN. The resultant injected signals had little to marginal discernibility in terms of amplitude and phase change. There was a slight opening of the lissajous lobe at the area of the injected OD and lD flaw. There was no similar type of lissajous opening observed when the unaltered PVN signal was reviewed on a scan line basis. Additionally, there was no phase shift of the unaltered PVN signal throughout the indication when compared side-by-side to the baseline data. By comparison, a 4Oo/oTW axial lD EDM notch signal was injected into the most severe PVN E1-2 of 5

Enclosure 1 location and amplitude and phase changes were both readily detected. Although the results of the signal injection were marginal in terms of discerning structurally significant SCC, the results provided sufficient assurance that a significant flaw could be discerned under strict data review for lissajous lobe openings and a side-by-side comparison to the baseline 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 partial support plate (PSP) indication. The PSP would be designated to the diminished support plate signal. The PSP was present in the preservice inspection and scheduled for follow up monitoring at subsequent in-service inspections. Westinghouse performed a detailed review of the condition and potential causes. The review of this signal suggests that the response of the low frequency on the +Point probe is the result of installation of a patch plate in the carbon steel tube support plate. The basis for this determination is provided below:

o The TSPs are inspected in the manufacturing shop following the drilling of the tube supports. The TSPs are only deemed acceptable for installation if found to be free from surface defects.

o With no operating history on the WBN Unit 2 SGs at the time of the inspection, a crack would have had to be initiated either during installation of the tube or sometime during the extended layup period..

o The alloy 600MA tube material has a Rockwell hardness less than the carbon steel tube 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 the pre-service inspection.

o lf the cause were to have been systematic due to an unidentified SG layup process, it is presumable that further tubetotube support plate intersections would be affected by the 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 seam between the tube support plate and the patch plate installed while tubing the SG. The design drawings for the patch plate specifically indicate that care should be taken when welding the patch plate into place.

These observations support the conclusion that the PSP indication in SG3 is most likely the result of installation and welding of a patch plate in the carbon steel tube support plate and 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 all tubes surrounding tube location Row 47 Column 48 at H01 in SG3 which served as further confirmation 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 that degradation exists in the support plate at this location. A graphic of the PSP indication is provided in Figures 1 and 2.

E1-3 of 5

Enclosure 1 Figure 1 Eddy Current Testing (ECT) Graphic from Baseline 2010 Figure 2 ECT Graphic from U2R1 2017 E1-4 of 5

Enclosure 1 TVA Response to RAI 1.c.

TVA confirms that the service-induced indications that were identified during the WBN Unit 2 Fall2Q17 (U2R1) outage were included in the referenced letter. However, as a result 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 of identifying the tube at location SG 3 Row 40 Column 42 as not plugged when this tube was actually plugged prior to initial SG service.

. Table 2-2 of the referenced letter lists eight volumetric indications listed for SG 3 with a total 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. 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 Generator Tube lnspection Report," dated February 16,2018 (ML18047A370)

E1-5 of 5

Enclosure 2 Revised Watts Bar Nuclear Plant Unit 2 - Cycle 1 Steam Generator Tube lnspection Report

WgsrTxGHOUSn NoN-PnOpNIETARY CIaSS 3 SG.SGMP-I 7-35 May 2018 Revision I Watts Bar IJlRI 180 Day Steam Generiltor Titbe Inspection Report

WESTINGHOUSE NONI.PROPRIETARY CLASS 3 SG-SGMP.I7-35 Revision 1 Watts Bar U2Rl 180 Day Steam Generator Tube Inspection Report Prepared for:

Tennessee Valley Authority Author's Name: Signature I Date For Pages Jesse S. Baron *Electron icallv A p prsved All SG Management Programs Verifier's Name: Signature I Date For Pages Jivan G. Thakkar *EI ec troai caII v A n pro ved A11 SG Management Programs Manager's Name: Signature I Date For Pages David P. Lytle, Manager *EI ec troni call v A n a ro ved AII SG Management Programs Reviewer's Name: Signature I Date For Pages

' Jeremy W. Mayo All TVA SG Program Manager Reviewer's Name: Signature lDate For Pages Tammy C. Sears Ail TVA Watts Bar SG Program Owner Revie\ryer's Name: Signature lDate For Pages Daniel P. Folsom A11 TVA NDE Level III

• Electronically Approvd Records are Autbenticatd in the Elcctronic Document Managcmcnt System

@2018 Westinghouse Electric Company LLC AII Rights Reserved SG-SGMP- 17 -35 May 2018 Revision I Page 2 of 12

Record of Revisions Revision Date f)escription December 0a Preliminary draft for Tennessee Valley Authority review and comment.

20t7 January 0 Incorporated comments from Tennessee Valley Authority, final approved and issued.

2A18 Revised Table 2-2 volumetric indication count and Table 2-4 indication listing for May I SG3 tube Row 40 Column 42 to conectly show as plugged durit g the pre-service 201 8 inspection. Chanses are marked with bars on the rieht hand side.

SG-SGMP- 17 -35 May 2018 Revision I Page 3 of 12

Table of Contents 2.0 180 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...................... 10

f. The Number and Percentage of Tubes Plugged to Date and the Effective Plugging Percentage in each SG l1

g. The Results of Condition Monitoring, lncluding the Results of Tube Pulls and In-Situ Testing.................. I I List of Tables and Figures Figure 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 SG-SGMP- 17 -35 May 2018 Revision I Page 4 of 12

1.0 Introduction The first in-service inspections (ISI) of the Watts Bar Unit 2 (WBN2) steam generators (SGs) were performed during the fall 2017 refueling outage designated as U2Rl. The U2Rl inspection was performed afrer 0.74 effective full power years (EFPY) of plant operation. The inspections included eddy current testing of the SG tubing as well as primary side visual inspections, secondary side visual inspections and secondary side cleanings. This report documents the "Watts Bar U2Rl 180-Day Steam Generator Tube Inspection Report" as required by the WBN2 Technical Specifications. The steam generators at WBN2 are a Westinghouse Model D3 preheater-type design where the majority of the feedwater enters near the top of the tubesheet on the cold leg side and the tubing is made from mill annealed Alloy 600 (Alloy 600MA) matedal. Figure 1-l below provides the arrangement and location designation 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) Location HTS/CTS  : Hot Leg Top ofTubesheeVCold Leg Top ofTubesheet HTE/CTE: Hot Leg Tube End/Cold Leg Tube End SG-SGMP-17-35 May 2018 Revision I Page 5 of 12

2.0 180 Day Steam Generator Tube Inspection Report In accordance with WBN2 Technical Specification Section 5.7.2.72, "Steam Generator Program", and Technical Specification Section 5.9.9, "Steam Generator Tube lnspection Report", this report documents the scope and results of the Watts Bar U2RI SG inspections. There are seven specific reporting requirements associated with the Technical Specification. Each lettered reporting requirement listed below is followed with the associated information based on the inspections performed during U2Rl.

a. The Scope of Inspections Performed on each SG The inspection progftrm addressed the known degradation observed in the Watts Bar Unit 2 SGs during the pre-sewice inspection, potential in-service SG tube degradation mechanisms and included proactive examinations to address areas where no degradation is anticipatdd but monitoring is performed regardless. The inspections were performed with qualified non-destructive examination (NDE) techniques for each existing and potential mechanism. The defined 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 4 to 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 on the 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/or Array 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. This inspection included all CL peripheral tubes two (2) tubes deep.

. I00oA +POINT or Array probe inspection of DNTs and DNGs > 5 Volts in the HL straight lengths, 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 a manufacturing burnish mark (MBM).

. +POINT or Array probe inspection of tubes surrounding known locations of foreign objects from the pre-service inspection.

. +POINT or Anay probe inspection of all tubes within a two (2) tube pitch of the region surrounding 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 at this 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 CL side.

. Visual inspection in all SGs of channel head primary side HL and CL inclusive of the entire divider plate to channel head weld and all visible clad surfaces.

The Watts Bar U2Rl inspection included all tubes with prior indications of degradation. The table below summarizes the number and type of eddy current examinations performed during U2Rl excluding the special interest inspection scope.

SG-SGMP-17.35 May 2018 Revision I Page 6 of 12

Table 2-1: Watts Bar U2Rl Steam Generator Eddy Current Inspection Scope Eddy Current Exam Type sGl SG2 SG3 SG4 Full Length Bobbin 4,200 4,197 4,214 4,,204 CL R1-R4 Low Row Bobbin 45t 452 454 454 HL Rl-R4 Low Row Bobbin 451 4s0 4s4 454 U-Bend +Point Rl-R4 451 4s0 454 454 HL +Point Tubesheet 4,651 4,647 4,669 4,658 CL Straight Leg X-Probe C06 to CTS-2 inch 2,662 2,649 2,7ll 2,705 ln addition to the NDE and primary side inspections discussed, visual inspection was performed in all SGs in order to determine the deposit and foreign object removal effectiveness of the tubesheet 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, visual inspection was also performed of the SG upper internal components in SGl and SG4 during Watts Bar U2Rl.

b. Degradation Mechanisms Found Volumetric 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 with either TSPs or AVBs. Volumetric indications generated during tube manufacture and bundle assembly and initially detected during the pre-service inspections were also detected during U2Rl. 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 no indications of stress corrosion cracking were detected during U2R1.

Table 2-2: Nurnber of Indications Detected for Each Degradation Mechanism Degradation Mechanism sGl SG2 SG3 SG4 Total Volumetric Indications (Pre-Service) t2 3 7 15 37 Wear at Tube Support Plates 0 0 5 0 5 Wear at Anti-Vibration Bars 3 5 I 8 t7

c. Nondestructive Examination (NDE) Techniques Utilized for Each Degradation Mechanism Table 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 also listed which were available for diagnostic testing, resolution and confirmation of anomalous indications. All the examination technique specification sheets (ETSSs) used during U2Rl are from the electric power research institute (EPRI) database. In some cases a variable 'X' is used in the listing of techniques in Table 2-3 which is in reference to a series of ETSSs.

SG-SGMP-17-35 May 2018 Revision I Page 7 of 12

Table 2-3: NDE Techniques for Each Existing or Potential Degradation Mechanism Desradation Mechanism ETSS Detection Technique Volumetric Indications due to B: 27091.1 Tube Fabrication and Installation B:27491 .2 (Pre-Service)

B: 96041.1 Wear at

• Pt: I 0908.4 AVBs A: 1 7908.1 A: 17908.2 B:96442.1 A: 1 1956.1 Wear at A: I 1956.2 Tube Support Plates A: I 1956.3 A: 11956.4 B:27091.1 Wear due to +Pt: 21998.1 Foreign Objects A: 1790X.1 A: l79AX.3 Tube-to-Tube B: 13091.2 Contact Wear B: 96005.3 OD Pitting of the Tube Material -t-Pt: 21998.1 A:24998.1

• Pt: 20511.1 Ax Axial and Circumf'erential +Pt: ll1524 Cir PWSCC at the TTS A: 20501.1 Ax A: 20500.1 Cir

+Pt: 128424 Ax Axial and Circumferential +Pt: 128425 Ax ODSCC at the TTS +Pt: 21410.1 Clir A: 20400.1 Ax/Cir Axial ODSCC B: 12841I at Tube *'Pt: 128424 Support Plates A:20402.1 Axial and Circumferential +Pt: 96511.2 PWSCC in the or Low Row U-bends +Pt: 99997.1 B: 128411 or B:24013.1 ODSCC at B: I 0013.1 Tube Dents and Dings "t-Pt: 22401.1 A: 20400.1 A: 20403.1

+Pt: 128424 Ax SCC at Tube Bulges and +Pt: 128425 Ax Overexpansions +'Pt: 21410.1 Cir A: 20400.1 Ax/Cir B: 12841 3 Axial ODSCC in the Freespan A: 20403.l ODSCC at Dents and Dings +-Pt: 22401.1 Coincident with an MBM Gh: 20406.1 Anomalous Gh: 20507.1 Indications Gh: 20508.1 Gh: 20509.1 Acronym Definitions for Table 2-3

+Pt: +POIN]'Probe Gh: Ghent Probe A: Aray Probe MBM: Manufactutng Bumish Mark AVB: Anti-Vibration Bar OD: Outcr Dianreter Ax: Axial ODSCCI: Outer f)iameter Stress C)orrosion Cracking B: Bobbin Probe PWSCC: Prinrary Water Stress Clorrosi<ln Cracking Cir: Circrunfcrential SCC: Stress Clonosion Cracking ETSS: Eddy cun:cnt Techniquc Specitication Sheet TTS: Top of the Tubeslteet SG-SGMP-17.35 May 2018 Revision I Page 8 of 12

d. Location, Orientation (if Linear), and Measured Sizes (if Available) of Service Induced Indications 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 NDE technique and an indication of whether the tube has been plugged.

Table 2-4: Watts Bar U2RI Pre-Service Volumetric Indications - All SGs SG Row Col Location Inch I Indication %TW Plugged?

I 3 99 c14 0.08 VOL 17 No I l0 38 ct2 -2.5 VOL 23 No I l0 55 cl3 32.84 VOL l1 No I 14 l8 AVI 3.8 VOL l4 No I l4 98 cr4 t.69 VOL r8 No 1 30 73 cl3 28.64 VOL 5 No 1 3r 80 ct4 1.84 VOL 24 No I 32 65 c0r 2.4 VOL I No I 34 77 c14 1.52 VOL 6 No I $ ,,.i # $

.l::j ffi

:a : :i::::::lI::!

t$$ii::iii:iliiliiiii:i $,i*:,,.il.lir :f?.:.:.:::r::l iiiliiiffi 1 3s 48 H08 2s.26 VOL l8 No

l  ::i.r.j.$ff t?) ixr$:$:l:ill.:':':,:,:,:':,:,:i i :.Yes,- frf--S$t$icc ;

I 39 72 ct4 -2.7 4 VOL t7 No I 41 73 CL4 0.58 VOL 30 No 2 5 103 H01 0.62 VOL t4 No 2 21 l6 H03 33.19 VOL t1 No a

J l9 37 cl3 2.7 4 VOL l4 No J

a 21 4l c0l 3.88 VOL 20 No 3 22 43 c0r 1.46 VOL l5 No J

a 32 46 H01 4.89 VOL 32 No 3 38 42 H01 1.75 VOL 21 No 3 38 42 H0l t.t2 VOL 1l No a

J 38 58 H06 40.09 VOL 21 No

.iiis s 4 1 87 HOI 2.96 VOL 2t No 4 )

,l 106 clr 12.01 VOL 10 No 4 I 84 H04 21 .84 VOL 2A No 4 9 44 H07 30.94 VOL l0 No 4 2A 46 H07 22.43 VOL l0 No 4 21 31 H06 9.99 VOL l8 No 4 2l 3l H06 2.03 VOL t4 No 4 2t 3l H06 t7.06 VOL 7 No 4 Z)

^ta 77 c09 7.31 VOL l3 No 4 27 46 H04 22.89 VOL 20 No 4 28 95 H04 25,6 voL l6 No 4 29 100 cr3 36.1 VOL 30 No 4 30 14 c10 9.02 VOL 2A No 4 30 35 c10 8.8 VOL 23 No 4 38 84 H04 3.39 VOL l5 No SG-SGMP-17-35 May 2018 Revision I Page 9 of 12

Tabte 2-5: Watts Bar I.J2RI Anti-Vibration Bar Wear Indications - All SGs SG Row Col Location Inchl Indication %TW Plugged?

I 40 7q AY2 0.16 PCT l0 No I 36 91 AV3 0.1 8 PCT t4 No 1 24 108 AV4 0.1 8 PCT 15 No 2 24 8 AV2 0 PCT 15 No 2 35 85 AV3 0 PCT t4 No 2 22 8 AV3 0.19 PCT 1l No 2 30 23 AV4 0.12 PCT l1 No 2 23 6 AV4 0.16 PCT l3 No a

J 30 105 AV3 -0.09 PCT 10 No 4 42 65 AV3 0.16 PCT 1l No 1 44 40 AV3 0 PCT 8 No 4 42 4A AV3 0.21 PCT t4 No 4 42 36 AV3 -0.1 1 PCT r5 No 4 44 28 AV3 0.12 PCT l0 No 4 41 24 AV3 0.17 PCT 8 No 4 38 21 AV3 0.t2 PCT 14 No

,t 7')

4 JJ t4 AV3 0.04 PCT 9 No Table 2-6: Watts Bar U2R I Tube Support Plate Wear Indications - All SGs SG Row Col Location Inchl Indication %TW Plugged?

3 48 66 c06 -0.26 PCT 5 No 3 49 61 c06 -0.2r PCT 7 No a

J 49 60 c06 -0.r4 PCT l0 No 3 48 60 c06 -0.28 PCT ll No a

J 47 60 c06 -0.24 PCT t4 No

e. Number of Tubes Plugged During the Inspection Outage for Each Degradation Mechanism There were eight (8) tubes plugged during the Watts Bar U2R1 SG in-service inspection. Only one (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 a 46o/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 object which was unable to be retrieved. Table 2-7 below provides the numbers and percentages of tubes 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 full tube length was not able to be obtained on the first data collection attempt. However, a full test was later completed using altemate means which included data collection from the opposite leg and/or use of downsized eddy current probes. These tubes were plugged in order to eliminate the possibility of being unable to collect data along the full tube length and evaluate condition monitoring at future inspections.

One tube in SG3 had a permeability variation indication in the tube material fully contained within the U-bend region. The use of alternative eddy current probes, such as a Ghent probe, to cleaf, the permeability indication was not an option since the solid body Ghent probe is not capable 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 SG-SGMP-17-35 May 2018 Revision I Pagel0of12

assessment of potential degradation mechanisms. This tube was plugged for its potential to mask degradation at future inspections.

Four tubes were plugged due to a foreign object located at the hot leg top of tubesheet which was unable to be retrieved. The object was identified as a piece of weld slag which was rigidly contained in between a group of four tubes. Multiple retrieval attempts from all accessible angles and orientations were made to remove the object from the SG and none were successful. Three of the tubes surrounding the foreign object had possible loose part indications (PLPs) from the eddy current test program, although none of the four surrounding tubes had indications of wear. Al1 four tubes were stabilized with a cable stabilizer tmversing the tubesheet region and plugged.

Table 2-7: Number of Tubes Plugged for each Degradation Mechanism SGI SG2 SG3 I scn Total

.J Plugged Tubes prior to U2Rl 23 27

JfuJ rr3ffi72 Volumetric lndication frorn Pre-Service 0 I 0 0 I Restricted Tube 0 2 0 0 2 Permeability Variation 0 0 I 0 1 ForeiEr Ob ect Unable to be Retrieved 0 4 0 0 4 ri!rl

,l Total Pluggecl Following U2R1 23 34 7 16 80 Percentage Plugged Following U2Rl 0.49% 0.73% a)5% 0.34% a.$%

The Number and Percentage of Tubes Plugged to Date and the Effective Plugging Percentage in each SG Table 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-Situ Testing Condition Monitorine. Tube Pulls and In-Situ Testing A condition monitoring (CM) assessment was perfiormed as required by the Watts Bar Unlt 2 steam generator program. Volumetric tube wear was the only in-service degradation mechanism detected 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 wear had 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%TW which is less than a conservatively determined CM limit of 58%TW. These CM limits include uncertainties for material properties, NDE depth sizing, and the burst pressure relationship. Since the 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 much lower than the three times normal operating pressure differential associated with the CM limits SG-SGMP-17-35 May 201 8 Revision 1 Page I I of12

for structural integtty. Therefore, CM for accident-induced leakage integrity was also demonstrated since volumetric wear indications will leak and burst at essentially the same pressure. Operational leakage integrity was demonstrated by the absence of any detectable primary-to-secondary leakage during the operating interval prior to U2Rl. Since tube integrity was demonstrated analytically, in-situ pressure testing was not required nor performed during the U2Rl outage. No tube pulls were planned or performed during U2Rl.

Primarv and Secondary Side Visual Inspection Results Visual inspections were performed on both the primary and secondary sides during U2R1 in accordance with Westinghouse nuclear safety advisory letter NSAL-12-1. Primary side inspections included visual inspections of all previously installed tube plugs as well as the channel head bowl cladding and the divider plate. The installed tube plug inspections showed no conditions indicative of degradation. However, &e inspections of the channel head bowl cladding and the divider plate showed visually appaxent evidence of minor indications of degradation of the cladding in SGl located on the hot leg side just above the primary manway opening. 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 that acceptable margin exists for maintenance of structural integrity of the SG channel head base metal 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 removed as 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 the hot leg and cold leg annulus and center no-tube lane. A total of 25 foreign objects were removed from the top of the tubesheet region. Any foreign objects not able to be retrieved were charucteized and an analysis performed to demonstrate acceptability of continued operation without exceeding the tube integrity performance criteria. A limited top of tubesheet in-bundle visual inspection was also performed for the purpose of assessing and trending the level of hardened 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 limited amount 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.

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