ML17276A348

R.E. Ginna, 2017 Steam Generator Tube Inspection Report

ML17276A348
Person / Time
Site:	Ginna
Issue date:	09/26/2017
From:	Wilson D F Exelon Generation Co, Ginna
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML17276A348 (15)

Text

4 1 Exelon Generation R.E. Ginna Nuclear Power Plant 1503 Lake Rd. Ontario, NY 14519 26, 2017 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20505-0001

Subject:

R.E. Ginna Nuclear Power Plant Facility Operating License No. DPR-18 NRC Docket No. 50-244 2017 Steam Generator Tube Inspection Report www.exeloncorp.corn Exelon Generation Company, LLC (Exelon) completed an' examination of the tubing in the R.E. Ginna Nuclear Power Plant (Ginna) steam generators during the end of Cycle 39 Refueling Outage. Ginna Technical Specification 5.6.7 requires that a report of the inspection results be submitted within 180 days after reactor coolant temperature exceeds 200°F. Attachment 1 contains the "Steam Generator Tube Inspection Report, End of Cycle 39 Refueling Outage, May 2017," which documents the results of the examinations.

There are no regulatory commitments contained in this letter. Should you have any questions regarding this submittal, please contact Kyle Garnish at 585-771-5321 . Respectfully, David F. Wilson Director, Site Engineering R.E. Ginna Nuclear Power Plant, LLC DW/ef

Attachment:

Steam Generator Tube Inspection Report, End of Cycle 39 Refueling Outage, May 2017 cc: NRC Regional Administrator, Region 1 NRC Project Manager, Ginna NRC Resident Inspector, Ginna Document Control Desk

• September 26, 2017 Page 2 bee: W. B. Carsky P. M. Swift K. G. Garnish D. P. Ferraro M.A. Slaby T. R. Loomis V. V. Gallimore

Attachment Steam Generator Tube Inspection Report, End of Cycle 39 Refueling Outage, May 2017

)i '=:::--Exelon Generation R. E. Ginna Nuclear Power Plant STEAM GENERATOR TUBE INSPECTION REPORT END OF CYCLE 39 REFUELING OUTAGE MAY 2017 1503 Lake Road Ontario, N.Y. 14519 Steam Generator Tube Inspection Report EOC 39 Refueling Outage TABLE OF CONTENTS Page 2of12 *

1.0 INTRODUCTION

............................................................................................................

3 2.0

SUMMARY

.....................................................................................................................

3 3.0 REPORT ........................................................................................................................

4 3.1 Scope oflnspections Performed on each Steam Generator (TS 5.6.7.a.)

...............

4 3.1.1 Primary Side Base Scope ......................

..............................................................

4 3.1.2 Primary Side Visual Inspection Scope ................................................................

6 3.1.3 Secondary Side Inspection Scope .......................................................................

6 3.2 Active Degradation Mechanisms Found (TS 5.6.7.b.)

...............................................

7 3.3 Nondestructive Examination Techniques Utilized for each Degradation Mechanism (TS 5.6.7.c.)

..................................................................*............................

7 3.4 Location, Orientation (if Linear), and Measured Sizes (if Available) of Service-Induced Indications (TS 5.6.7.d.)

................................................................................

7 3.5 Number of Tubes Plugged During the Inspection Outage for each Active Degradation Mechanism (TS 5.6.7.e.)

..........................................................................

8 3.6 Total Number and Percentage of Tube Plugs to-Date (TS 5.6.7.f.)

..........................

9 3.7 The Results of Condition Monitoring, including the Results of Tube Pulls and In-Situ Testing (TS 5.6. 7 .g.) ..............................................................................................

9 3.7.1 Lattice Grid Wear .............................................................. , ..................................

10 3.7.2 Foreign Object Wear ...........................................................................................

10 3.7.3 Leakage lntegrity

.................................................................................................

11 4.0 ACRONYMS ................................................................................................................

12

5.0 REFERENCES

.............................................................................................................

12 Steam Generator Tube Inspection Report EOC 39 Refueling Outage

1.0 INTRODUCTION

Page 3of12 The R.E. Ginna Nuclear Power Plant (Ginna) design has two (2) re-circulating design steam generators (SG) designed and fabricated by Babcock and Wilcox (BWI) of Cambridge, Ontario, Canada. The nomenclature used for fabrication and subsequent service inspections is SG-A and SG-B. Each BWI steam generator was designed to contain 4 765 tubes. One tube in each steam generator was removed from service during fabrication by means of a shop welded lnconel 690 plug. SG-A contains 4764 open tubes, and SG-B had four (4) tubes plugged at End of Cycle (EOC 32) due to a loose part that brings the total to 4760 open tubes. The tubing material is thermally treated lnconel 690 having a nominal outer diameter (OD) of 0.750 inch and a nominal wall thickness of 0.043 inch. The nominal thickness of the tube sheet is 25.25 inches, with a full depth hydraulic expansion of all the tubes into the tube sheet material.

The tubes are supported in the straight section by eight 410 stainless steel lattice grid supports which are comprised of high, medium, and low bars. The tubes are supported in the U-bend by ten 410 stainless steel fan bar I collector bar assemblies. Ginna Technical Specifications (TS) 5.5.8.d provides the requirements for SG inspection frequencies

[5.1 ]. The TS requires that 100% of the tubes are to be inspected at sequential periods of 144, 108, 72, and, thereafter, 60 effective full power months (EFPM). Additionally, inspect 50% of the tubes by the refueling outage nearest the midpoint of the period and the remaining 50% by the refueling outage nearest the end of the period. At the beginning of the Ginna EOC 39 refueling outage, Ginna was at 18.16 EFPY (217.95 EFPM). Ginna has operated for a total of 2.871 EFPY or 34.452 EFPM since the last Steam Generator inspection in May 2014. Therefore, Ginna was at 73.95 EFPM within the second 108 EFPM inspection period. In accordance with the Ginna TS 3.4.17, "Steam Generator (SG) Tube Integrity," Ginna TS 5.5.8, "Steam Generator (SG) Program," and American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code Section XI, 2004 Edition [5.2], IWB 2500-1, Examination category B-Q, item B16.20, SG eddy current examinations were performed during the Ginna EOC 35 refueling outage. All inspections were completed in accordance with Ginna TS 5.5.8 and EPRI "Pressurized Water Steam Generator Examination Guidelines" [5.4]. 2.0

SUMMARY

The EOC 39 2017 refueling outage (RFO) was the eighth in-service inspection of the replacement Ginna SGs. The inspection was formally completed on May 5, 2017. Ginna entered Mode 4 on May 13, 2017. A degradation assessment was performed prior to the EOC 39 inspection to assure qualified inspection techniques would be used to detect any existing and potential damage mechanisms.

The modes of tube degradation detected during the EOC 39 RFO were secondary side foreign object wear, and minor tube to lattice grid wear.

Steam Generator Tube Inspection Report EOC 39 Refueling Outage Page 4of12 Foreign objects were also detected on the tubesheet secondary face of both "A" and "B" steam generators (SG). A process was established to prioritize the removal of foreign objects, with the highest priority given to the foreign objects that posed a higher risk to tube integrity.

Each foreign object that was detected and left in-service was evaluated in accordance with EPRI Steam Generator Integrity Assessment Guidelines, Revision 3, with Ginna SG plant-specific thermal performance inputs. The basis for leaving each foreign object in-service is the disposition of each foreign object evaluation.

Tube manufacturing anomalies were sampled with no detectable degradation and no detectable changes since the original SG manufacturing baseline.

Denting at the top of tubesheet on the cold leg side was first observed during the 2008 inspection in both SGs (2 C/L locations in SG A, 80 C/L locations in SG B). In the 2011 inspection, denting at the top of tubesheet was observed to a greater extent (4 C/L locations in SG A, 236 C/L locations in SG B, 1 H/L location in SG B). The 75% full length 2017 RFO eddy-current bobbin-coil examination plan incorporated potential extent of condition for denting at the tubesheet secondary face. This scope was the same for both the "A" and "B" SG. Starting with the 2014 inspection and continuing through the 2017 inspection, the top of tubesheet denting is no longer active. A total of 250 tubesheet dents were reported in 2014 and a total of 251 dents were reported in 2017. A comparison of the 2017 inspection results with those from 2014 indicates that there continues to be no active progression of the denting. Rotating probe ECT was employed to determine if any tube degradation existed as a consequence of the denting. No tube degradation was identified in any of the tubes tested. The denting is suspected to be a consequence of the build-up of hard sludge near the center of the tube bundle. Denting is not considered a tube degradation mechanism.

Considering the SCC resistance of the tubing and that the denting is no longer active, this condition represents a negligible increase in cracking risk. 3.0 REPORT 3.1 Scope of Inspections Performed on each Steam Generator (TS 5.6.7.a.)

3.1.1 Primary Side Base Scope Eddy current examinations included full-length bobbin inspections in 7156 of 9524 service tubes as well as array probe examinations of tubes near the periphery and tube lane. Additional supplemental array and +PointŽ examinations were performed on locations of special interest (dents, previous eddy current PLPs, etc.). Table 1 below shows the quantity and type of examinations performed.

Steam Generator Tube Inspection Report EOC 39 Refueling Outage Table 1 -Primary Side Inspection Scope SGA Number of Installed Tubes 4765 Number of Tubes in Service Prior to EOC39 4764 Number of Tubes Inspected FIL w/Bobbin Probe 3581 Previously Plugged Tubes 1 Number of Obstructed Tubes 0 Bobbin Exam Full Length 2975 H/L Candy Cane 605 H/L Straight 1 C/L Straight 606 Array Coil Exam H/L Tubesheet (01 HTEH) 1364 H/L PLP Bounding 93 C/L Tubesheet (01 CTEC) 1364 C/L PLP Bounding 52 Proximity Tubes (08HOBC) 13 Rotating Coil Exam H/L OXP 28 H/L Tubesheet 0 Historical WAR Indications 3 Ubend Special Interest 1 H/L Straight Special Interest 1 C/L Straight Special Interest 4 Total Inspections 7110 SGB 4765 4760 3575 5 0 2969 606 0 606 1495 35 1398 0 12 0 1 4 0 3 247 7376 Note: One tube in SG-A was restricted with a 0.61 O" diameter bobbin probe in the U-bend. This tube was tested along its full length with a combination bobbin coil and rotating coil examinations.

Page 5of12 Total 9530 9524 7156 6 0 5944 1211 1 1212 2859 128 2762 52 25 28 1 7 1 4 251 14486 Steam Generator Tube Inspection Report EOC 39 Refueling Outage 3.1.2 Primary Side Visual Inspection Scope Page 6of12 The primary side channel head (hot and cold leg) of both steam generators was visually inspected using a remote operated camera in accordance with Ginna inspection procedures.

The channel head general area and cladding was inspected for the following:

through holes or breaches that would expose carbon steel base material under the cladding, rust colored discoloration or stains visible on cladding surface and channel head cladding degradation such as cracks or significant deformation.

The tubesheet, tube ends, and tube plugs were inspected for the following:

cracking, degradation, water leakage, boron deposits, tube sheet or tube end deformation.

No degradation was observed in any of these areas in either steam generator.

The divider plate was visually inspected from both hot and cold legs using a remote camera specifically looking for the following:

cracks on the divider plate surface, surface deformation, foreign material that may mask any degradation and any other degradation.

Special attention was made when inspecting the weld deposit seat bar, divider plate weld, 6 inch divider plate corner windows and the divider plate weld heat affected zone. No degradation was observed in any of these areas in either steam generator.

3.1.3 Secondary Side Inspection Scope Secondary side inspections were performed with a variety of remote tooling. For each steam generator, a visual inspection (top of tubesheet) was performed after sludge lancing including:

• 100% of the annulus to 5 tubes deep

• 100% of the no-tube lane to 5 tubes deep

• Slowdown and drain holes

• Shroud supports

• Inspection of tube support structures (1st support only)

• In-bundle inspection of previously identified foreign objects as directed by BWXT Engineering

• In-bundle inspection of ECT detected potential loose parts (PLP) as directed by BWXT Engineering A steam drum and upper internals inspection was performed on both steam generators.

The steam drum and upper internals inspection included:

I

• Upper Internals Visual Inspection o Secondary moisture separators, structural welds, etc. o Secondary moisture separator base plates o Steam outlet venturi

• Laser scanning of all secondary separator base plates

• Ultrasonic inspection of selected secondary separator base plates Steam Generator Tube Inspection Report EOC 39 Refueling Outage 3.2 Active Degradation Mechanisms Found (TS 5.6.7.b.)

Page 7of12 The only detected tube degradation was volumetric resulting from foreign object wear and tube to lattice grid support wear. There were a total of 7 wear locations in 6 tubes, all of these exist from previous outages. These were sized with the techniques shown below in Table 2. All 85 of the secondary separator base plates were inspected in both steam generators.

The inspections included both visual inspections and laser profilometry.

These

• inspections were performed in response to base plate degradation that was previously observed in the EOC37 outage. The most significant degradation was at separator

1. 46 in SG A which had degradation as deep as 51 % through the plate. This was the only separator that showed degradation deeper than 50% through the plate. Eighteen other separators (seven in SG A and eleven in SG B) showed moderate degradation between 30% and 50% through. the plate. The remaining separators showed minor degradation of less than 30% through the plate. The degradation is caused by flow accelerated corrosion.

These results are consistent with.results from other utilities with BWXT steam generators.

3.3 Nondestructive Examination Techniques Utilized for each Degradation Mechanism (TS 5.6.7.c.)

See Table 2. 3.4 Location, Orientation (if Linear), and Measured Sizes (if Available) of Service-Induced Indications (TS 5.6.7.d.)

See Table 2.

Steam Generator Tube Inspection Report EOC 39 Refueling Outage Page 8of12 Table 2 -Lattice Grid and Foreign Object Wear Indications Maximum Prior Prior Delta SG Row Col Location ETSS Depth Outage Outage Cause %TW Delta Current Depth Depth Growth EFPY Outage (Original) (Resized) 10%TW Lattice A 1 25 05H -1.76 96910.1 10%TW (96910.1)

NA Grid Wear 0 2.871 6%TW 6%TW Lattice B 2 20 06C-1.62 96910.1 (96910.1)

NA Grid Wear 0 2.871 B 78 24 01H + 0.97 96910.1 8%TW 25%TW 8%TW Lattice 0 2.871 (21998.1)

(96910.1)

Grid Wear B 78 24 01H + 1.26 96910.1 7%TW 21%TW 6%TW Lattice 1 2.871 (21998.1)

(96910.1)

Grid Wear 32%TW 37%TW Foreign A 91 51 05H + 0.35 27901.1 37%TW (21998.1)

(27901.1)

Object 0 2.871 Wear 21%TW 21%TW Foreign A 53 85 03H -2.07 27901.1 21%TW (27901.2)

(27901.1)

Object 0 2.871 Wear 19%TW 25%TW Foreign B 2 78 02H -1.84 27901.1 25%TW (21998.1)

(27901.1)

Object 0 2.871 Wear 3.5 Number of Tubes Plugged During the Inspection Outage for each Active Degradation Mechanism (TS 5.6.7.e.)

There were no tubes that required removal from service by plugging during the Ginna 2017 RFO examination.

%TW Growth per EFPY 0 0 0 . 0.3 0 0 0 Steam Generator Tube Inspection Report EOC 39 Refueling Outage 3.6 Total Number and Percentage of Tube Plugs to-Date (TS 5.6.7.f.)

See Table 3 below. Table 3 -Plugged Tubes SG-A SG-8 Tubes plugged to date 1 5 Tubes Installed 4765 4765 % of tubes pluaaed to date 0.02% 0.10% Page 9of12 The tube plugging performed to-date included 1 tube in each SG during pre-service examinations.

The additional 4 tubes plugged in the SG-8 were from a foreign object that was not able to be removed during the 2005 RFO in-service examination.

3.7 The Results of Condition Monitoring, including the Results of Tube Pulls and In-Situ Testing (TS 5.6. 7 .g.) A condition monitoring assessment was performed for each in-service degradation mechanism detected during the EOC 39 2017 RFO SG examination.

The condition monitoring assessment was performed in accordance with Ginna TS 5.5.8.a, NEl-97-06

[5.3], EPRI Steam Generator Integrity Assessment Guidelines, Revision 3 [5.5], and the EPRI Steam Generator Degradation Specific Management Flaw Handbook, Revision 2 * [5.6]. For each identified degradation mechanism, the as-found condition was compared to the appropriate performance criteria for tube integrity, accident induced leakage and operational leakage as defined in TS 5.5.8.b. For each damage mechanism a tube structural limit was determined to ensure that SG tube integrity would be maintained over the full range of operating conditions and design basis accidents.

This included retaining a safety factor of 3.0 against burst under normal steady-state full power operation primary-to-secondary pressure differential and a safety factor of 1.4 against burst under the limiting design basis accident pressure differential.

The as-found condition of each degradation mechanism found during the EOC 39 RFO was shown to meet the appropriate limiting structural integrity performance parameter with a probability of 0.95 at 50% confidence level, including consideration of relevant uncertainties.

There were no tube pulls or in-situ pressure testing performed during the EOC 39 2017 RFO. The following sections provide a summary of the condition monitoring assessment for each damage mechanism.

Steam Generator Tube Inspection Report EOC 39 Refue l ing Outage Page 10 of 12 3. 7 .1 Lattice Grid Wear 100 90 80 70 60 !': Q. ., 0 50 c 0 .., 40 30 20 JO 0 0 EPRI Examination Technique Specification Sheet (ETSS) 96910.1 were used for depth sizing of lattice grid wear. Based on the sizing parameters for this technique , a Condition Monitoring (CM) curve (Figure 4) was generated from data documented in the Degradation Assessment (DA). As shown below , all indications lie well below the CM curve. Hence , structural integrity is demonstrated for all the lattice grid wear indications. Figure 4: CM Eval u a tion for Lattice Grid We a r (ETSS 96910.1) I l t I I ' ........... I I -. I

• I 0.5 1.5 2.5 3 3.5 Degradation Length (in) -cM Limit

• t...ittice Grid Weon SGA

• Lattice Grid Wedr SG 8 3.7.2 Foreign Object W ear All of the foreign object wear indications were located at or near lattice grid supports. These flaws were confirmed to be foreign object wear based on review of the +PointŽ results and the fact that the flaws were not coincident with the contact points between the tubes and lattice supports.

In all cases, the foreign object that caused the w ear was confirmed to not be present based on review of +PointŽ inspection results of the affected and bounding tubes. ETSS 27901.1 was used for depth sizing of foreign object wear. Based on the sizing parameters for ETSS 27901.1 and the axial thinning model , the CM curve shown in Steam Generator Tube Inspection Report EOC 39 Refueling Outage Page 11of12 100 90 80 70 t ., .,, so c 0 'i .., 40 ., 0 30 20 10 0 0 Figure 5 was generated from data documented in the DA [5.7]. As shown , all indications lie well below the CM curve. Hence , structural integrity is demonstrated for all the foreign object wear indications. Figure 5: CM Evaluation for Foreign Object Wear (ETSS 27901.1) ---' I l """"" I I -------* I I

• j 1 I 0.5 1.5 2.5 3.5 Degr*da t ion Length (in) -cM Limit

• Foreisn Object Wear SG A

• Foreign Object Wear SG B 3.7.3 Leakage Integrity Per Reference

[5.5], for volumetric flaws with axial extents ::::0.25" , the onsets of popthrough leakage and burst are coincident.

All wear indications detected had lengths ::::0.25". Therefore , since structural integrity was satisfied at the 36P value of 4383 psid , accident and operational leakage integrity is also satisfied at all lower pressure differentials.

Steam Generator Tube Inspection Report EOC 39 Refueling Outage 4.0 ACRONYMS ASME C/L CM DA EFPM EFPY EOC EPRI ETSS FOSAR H/L NEI PLP SG TS TW American Society of Mechanical Engineers Cold Leg Condition Monitoring Degradation Assessment Effective Full Power Months Effective Full Power Years End of Cycle Electrical Power and Research Institute Examination Technique Specification Sheet Foreign Object Search and Retrieval Hot Leg Nuclear Energy Institute Potential Loose Part Steam Generator Technical Specification Through-Wall

5.0 REFERENCES

[5.1] Ginna Technical Specifications

[5.2] ASME Section XI, 2004 Edition Page 12of12 [5.3] Steam Generator Program Guidelines, Nuclear Energy Institute, NEI 97-06, Rev. 3, January 2011 [5.4] EPRI, "Pressurized Water Reactor Steam Generator Examination Guidelines:

Revision 7,"1013706, October 2007 [5.5] EPRI SGMP, "SG Integrity Assessment Guidelines Rev. 3," 1019038, November 2009 [5.6] EPRI Steam Generator Degradation Specific Management Flaw Handbook, Revision 2", 3002005426, October 2015 . [5.7] R.E. Ginna Station EOC 39 Steam Generator Condition Monitoring and Operational Assessment (CMOA) 0192-AST-101215 Rev. 000