ML14314A033

Submittal Special Report (Spr) 2014-006

ML14314A033
Person / Time
Site:	Summer
Issue date:	11/05/2014
From:	Gatlin T South Carolina Electric & Gas Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
SPR 2014-006
Download: ML14314A033 (12)

Text

Thomas D. Gatlin Vice President,Nuclear Operations 803.345.4342 November 5, 2014 A SCANA COMPANY Document Control Desk U. S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

VIRGIL C. SUMMER NUCLEAR STATION (VCSNS), UNIT 1 DOCKET NO. 50-395 OPERATING LICENSE NO. NPF-12 SPECIAL REPORT (SPR) 2014-006

Reference:

Letter from S. A. Byrne (VCSNS) to Document Control Desk (NRC),

"SPECIAL REPORT (SPR 2000-005)," dated November 8, 2000

[ML003769321]

South Carolina Electric & Gas Company (SCE&G) hereby submits the Steam Generator Tube Inspection Report pursuant to the requirements of Technical Specifications 6.9.1.12. This report summarizes the examination conducted during the 2014 Spring Refueling (RF-21) and the state of the three steam generators installed at V. C. Summer Nuclear Station (VCSNS),

Unit 1.

In preparation for the RF-21 inspection, an error was identified. The referenced letter incorrectly identified the location of two of the three tubes that were plugged in 'A" Steam Generator during RF-12. This error has been corrected in Table 9 of the attached Steam Generator Tube Inspection Report. The error has been entered in the station's corrective action program. No tube plugging was required during RF-21.

Should you have any questions, please call Mr. Bruce Thompson at (803) 931-5042.

Very truly yours, Thomas D. Gatlin WLT/TDG/ts Attachment c: K. B. Marsh W. M. Cherry INPO Records Center S. A. Byrne V. M. McCree NSRC J. B. Archie S. A. Williams RTS (LTD 322)

N. S. Carns NRC Resident Inspector File (818.08)

J. H. Hamilton K. M. Sutton PRSF (RC-14-0171)

J. W. Williams Virgil C.Summer Station

• Post Office Box 88 *Jenkinsville, SC. 29065

• F (803) 941-9776

Document Control Desk Attachment LTD 322 RC-14-0171 Page 1 of 11 Steam Generator Tube Inspection Report VC Summer Nuclear Station Jenkinsville SC 29065 RF-21 April 2014

1.0 Background

Commissioned in January 1984, VC Summer originally had three Westinghouse Model D3 steam generators. The original steam generators were replaced with Westinghouse Delta-75 steam generators in the fall of 1994.

Each"of the replacement steam, generators contains 6307 thermally treated alloy 690 tubes with an outside diameter, of 0.6875 inches and a nominal wall thickness of 0.040 inches. Stress relief was performed during fabrication on the U-bends of the first 17 rows of tubing. The straight sections of the tubing are supported by nine tube support plates made of 1.125 inch thick, SA-240, Type 405 stainless steel. The tube support plates (TSP) have trefoil, broached holes which reduce dryout and the collection of impurities at the tube-to-support plate intersections. The u-bend sections of the tubing are supported by four sets of anti-vibration bars (AVB) made of SA-240, Type 405 stainless steel. A flow distribution baffle (FDB) plate is located between the tubesheet and the lowest support plate. The flow distribution baffle increases the secondary side flow velocity near the tubesheet, thus reducing the accumulation of sludge at the top of the tubesheet.

The D-75 steam generators contain a sludge collector above each primary separator which is designed to reduce the accumulation of sludge at the tubesheet. The tube-to-tubesheet joints are formed by a full depth hydraulic expansion and a weld to the primary cladding on the tubesheet.

2.0 Scope of Inspections Performed on Each Steam Generator (SG)

The RF-21 inspection program, as required by the EPRI PWR SG Examination Guidelines Revision 7, addressed the existing and potential degradation mechanisms for V. C. Summer. The defined scope for each steam generator included the following:

1) 55% (plus tubes with prior indications) full length bobbin inspection

2) U-bend +Point inspection of tubes that a bobbin probe would not pass through

3) 100% +Point inspection of peripheral tubes and at the tube lane (two tubes closest to the boundary in both cases) in the hot leg (HL) and cold leg (CL) from 3 inches above to 3 inches below Top of the Tubesheet (TTS)

4) 20% +Point inspection of non-peripheral tubes in hot leg from 3 inches above to 3 inches below TTS

Document Control Desk Attachment LTD 322 RC-14-0171 Page 2 of 11

5) 50% +Point inspection of all dents/dings > 5 volts

6) +Point inspection of legacy loose parts

7) Special interest RPC (freespan signals without historical resolution, bobbin I-code indications, etc.)

8) 100% tube plug video inspection in HL and CL (8 tubes total)

9) Video scan of channel head bowl in HL and CL as recommended by Westinghouse Nuclear Safety Advisory Letter (NSAL) 12-1 10)TTS secondary side video inspection including Foreign Object Search & Retrieval (FOSAR) 1 1)Upper bundle video inspection in 'A' SG 12)Steam drum inspection

.Tables 1 and 2 provide the RF-21 total eddy current examination scope.

Table 1: Base Scope. Exam --

Base Scope Exam Type SG A SG B SG C Cold Leg Full Length Bobbin 2889 2895 2885 Cold Leg Low Row Straight Section 619 545 322 Hot Leg Full Length Bobbin 0 80 298 Hot Leg Candy Cane 619 545 322 Cold Leg RPC Tubesheet Exams (Periphery and Intejior) 680 679 678 Hot Leg RPC Tubesheet Exams (Periphery and Interior) 1832 1893 1882 TOTAL 6639 6637 6387 Table 2: +Point Diagnostic Exams for Bobbin Indications (Special Interest Additional Scope)

Exam Type SG A SG B SG C Hot Leg Bobbin Dent/Ding > 5V 0 2 0 Hot Leg Bobbin I-Code Indications 6 1 4 Hot Leg PLPNOL and Box In 18 14 0 Hot Leg Mag Bias Special Interest 0 1 2 Cold Leg Bobbin Dent/Ding > 5V 34 47 31 Cold Leg Bobbin I-Code Indications 5 13 8 Cold Leg PLPNOL and Box In 0 0 0 Cold Leg New Bobbin Wear 10 20 13 Cold Leg Mag Bias Special Interest 0 0 0 U-bend Bobbin I-Code Indications 0 0 2 Hot Leg U-Bend Dent/Ding > 5V 2 2 0 Cold Leg U-Bend Dent/Ding > 5V 1 1 0 TOTAL 76 101 60

Document Control Desk Attachment.'

LTD 322 RC-14-0171 Page 3 of 11 3.0 Degradation Mechanisms Found Existing degradation mechanisms are those mechanical or corrosive processes previously and/or currently observed in a Steam Generator. Based on this definition of existing degradation mechanisms and prior inspection results, there are three existing degradation mechanisms within the VC Summer replacement steam generators.

3.1 Tube Wear at AVBs Wear related to tube interaction with AVB supports results from flow-induced vibration in the upper bundle. This mechanical process is related to the tightness of the upper bundle assembly as expressed in the distribution of tube-to-AVB gaps. Indications of AVB wear in-the replacement steam generators were first reported during RF-12. Very few AVB wear indications were reported in subsequent inspections during RF-1 5 and RF-1 8. No new AVB wear indications were reported in RF-21. The deepest wear indication during RF-21 was 10% through-wall depth (TWD). To date, no tubes have been plugged for this mechanism.

3.2 Tube Wear at TSPs and FDB Flow-induced vibration that causes tube-wear at TSPs and the FDB is governed primarily by secondary side thermal hydraulic characteristics and the geometry (sizes of the tube-to-support gaps, TSP spacing, etc.). Tube wear at TSPs was first reported in the replacement steam generators during RF-1 5. The deepest TSP wear reported during RF-21 was, 29% TW. One indication of tube wear at FDB was-reported during RF-18. The FDB wear reported during RF-21 was 16% TW. Both wear mechanisms are located and sized using identical eddy current testing techniques. With TSP wear being significantly deeper than FDB wear, TSP wear bounds the condition monitoring analysis results for FDB wear. To date, no tubes have been plugged for this mechanism.

3.3 Tube Wear Due to Foreign Objects Foreign objects may enter the SG through the feedwater stream and may cause tube wear. Foreign object wear was detected for the first time during RF-21 inspections.

During the bobbin probe base scope inspection program, an indication was noted at the top of the hot leg baffle plate in two tubes. When these locations were tested with a rotating coil probe, small indications indicative of wear were noted above the baffle plate. It was surmised that the wear was caused by a transient loose part. The tube at row 32, column 89 was sized at 7% TWD and the tube at row 35, column 92 was sized at 10% TWD. A review of the eddy current data from surrounding tubes was conducted with no additional indications discovered. Due to the lack of similar indications reported during the inspection and the shallow nature of the indications, examination expansion was deemed unnecessary. No loose parts associated with this wear were found in the

Document Control Desk Attachment LTD 322 RC-14-0171 Page 4 of 11 vicinity of these indications during FOSAR activities so it is believed that the loose part no longer remains in the SG. No tubes were plugged for this mechanism.

4.0 Nondestructive Examination Techniques for Each Degradation Mechanism Inspection programs and Examination Technique Specification Sheets (ETSS) identified in the Degradation Assessment for each degradation mechanism are identified in Table 3.

Table 3: SG Tube Degradation Mechanisms and Associated NDE Techniques Degradation Inspection Program Detection Sizing Mechanisms  % Sample Technique ETSS Technique

__ _ _ I____ETSS #

Exidting Mechanisms _

Wear at AVBs 55% Bobbin 96004.1 96004.1 Wear at TSPs and 55% Bobbin, 96004.1 96004.1 FDB +Point Special Interest 96910.1 96910.1 Potential Mechanisms ___

Wear due to Foreign 55% Bobbin, 27091.2 Objects +Point Special Interest, TTS +Point of 21998.1 21998.1 peripheral tubes in Alternate Alternate

- HL and CL 2790X.3 2 2790X.3 2 Diagnostic Inspection Inspection of Row 55% Bobbin 96004.1 96004.1 1-2 U-bends ODSCC at HL TTS 20% + Point HL TTS 128424 128431 Expansion 128425 128432 Transition. 21410.1 21410.1 ODSCC at 55% Bobbin, 10013.1 Dings/Dents 50% +Point dents/dings > 5V 24013.1 128413 128424 128431 128425 128432 21410.1 21410.1 S22841.3 Note 1: Prior to RF-21, wear due to foreign objects was classified as a potential degradation mechanism.

Note 2: The applicable ETSS are numbered 2790X.3 where X is a variable between 1 and 7. Techniques and corresponding uncertainty used for sizing of foreign object wear is dependent on indication geometry.

Document Control Desk Attachment LTD 322 RC-14-0171 Page 5 of 11 5.0 Location, Orientation (if Linear), and Measured Sizes (if Available) of Service Induced Indications Tables 4-8 provide the location and measured sizes of service induced indications. The majority of the indications were from tube wear by support structures. Two indications of wear due to a migratory loose part were detected in SG C.

Table 4: AVB Wear SG Row Col Supp Inch Depth A 19 140 AV7 0.43 10 A 26 139 AV2 -0.75 7 A 26 139 AV7 0.75 9 C 26 3 AV2 -0.75 9 Table 5: SG A TSP Wear Row Col Supp Inch Depth 17 2 05C 0.31 13 1.4... 10 06C -0.57 15 1 10 05C 0.00 20 1 10 04C 0.00 13 67 24 07C 0.41 12 91 28 07C -0.70 11 1 60 03C -0.43 8 1 60 03C 0.47 16 115 62 09H ,:0.45 12 2 77 08H -0.43 26 2 77 07C 0.00 29 4 81 07C 0.52 10 4 99 05C 0.36 14 4 103 05C -0.56 8 4 103 05C 0.45 10 77 122 07C 0.50 12 Table 6: SG B TSP Wear Row Col Supp Inch Depth 1 2 06C -0.63 15 1 2 06C 0.43 15 1 2 05C -0.68 7 1 2 05C 0.38 14 34 9 06C 0.43 11 2 17 07H 0.49 11 1 28 07H -0.55 14

Document Control Desk Attachment LTD 322 RC-14-0171 Page 6 of 11 Row Col Supp Inch Depth 1 28 07H 0.49 11 2 31 06C 0.58 12 2 31 05C 0.48 17 99 44 09C 0.43 10 94 45 09C 0.43 8 3 46 08C -0.53 14 3 46 07C -0.49 19 108 47 07H -0.61 11 108 47 07H 0.41 11 3 54 06C 0.56 13 2 61 04C -0.59 14 115 74 06H 0.43 12 6 75 07C -0.54 13 114 75 04H 0.49 6 3 84 .07C 0.47 16 5 84 07C 0.52 11 5 88 08C -0.57 12 5 88 08C 0.48 12 5 88 07C '0.54 18 5 88 07C 0.48 11 5 88 05C -0.47 10 5 90 08C 0.50 13 5 90 07C 0.57 13 99 106 07C -0.61 12 2 113 07C 0.47 15 2 113 06C 0.50 16 5 116 07C 0.43 14 39 120 08C 0.36 8 10 121 08C - -0.52 10 10 121 08C 0.58 7 54 121 07C -0.61 13 20 139 03H -0.58 11 1 140 06C -0.58 10 1 140 06C 0.41 14 Table 7: SG C TSP Wear and FDB Wear Row Col Supp Inch Depth 1 4 06C 0.38 11 1 8 07C 0.45 14 52 19 08C 0.47 10 25 36 07C 0.47 12 5 52 06C -0.56 15 7 54 06C -0.52 15

Document Control Desk Attachment LTD 322 RC-14-0171 Page 7 of 11 Row Col Supp Inch Depth 104 55 08C 0.38 25 110 73 08C 0.33 20 1 82 08H 0.53 14 114 83 09H 0.39 8 34 91 BPH 0.51 16 1 94 06H 0.59 21 6 107 05C -0.45 9 1 112 06C -0.52 19 1 112 06C 0.48 13 83 116 07C -0.55 13 30 137 08C 0.43 12 Table 8: SG C Foreign Object Wear Row Col . Supp Inch Depth 32 89 BPH 0.42 7 35 92 BPH 0.50 10 6.0 Tubes Plugged during the Inspection Outage -11 No tube plugging was required during RF-21.

7.0 Number and Percentage of Tubes Plugged to Date A total of eight (8) tubes have been removed from service by plugging in the VC Summer Steam Generators as shown in Table 9. The percentage of tubes plugged to date is less than 0.07% for each SG as shown in Table 10.

Table 9: List of all Plugged Tubes through RF-21 SG Outage Year Row Col Ind Plugging Attribute A7 RF-12 2000 26 25 NTE Anomalous tube-end expansion AK' RF-12 2000 26 31 NTE Anomalous tube-end expansion A RF-12 2000 94 51 NTE Anomalous tube-end expansion B Factory 1994 76 123 Factory-installed weld plug C Factory 1994 104 41 Factory-installed weld plug C RF-12 2000 57 96 NTE Anomalous tube-end expansion C Factory 1994 108 97 Factory-installed weld plug C RF-12 12000 99 100 NTE Anomalous tube-end expansion Note 1: Report RC-00-0345, dated 11/8/2000, incorrectly identified that the tubes at Row 25 Column 26 and Row 25 Column 31 were plugged in "A" Steam Generator during RF-12.

Document Control Desk Attachment LTD 322 RC-14-0171 Page 8 of 11 Table 10: SG Tube Plugging Percent ISG A SG B SG C Total Tubes 6307 6307 6307 Tubes Plugged 3 1 4 Effective Plugging 0.048% 0.016% 0.063%

Percentage 8.0 Results of Condition Monitoring, Including Tube Pulls and In-Situ Testing Condition monitoring results for each existing degradation mechanism is outlined in the sections that follow. Based on inspection data, no tube pulls or in-situ testing was required in RF Condition monitoring concluded that SG performance criteria for leakage and structural integrity were satisfied for the preceding three'cycle operating interval prior to RF-21.

8.1 Tube Wear at AVBs The maximum AVB wear indication from RF-21 was 10% TWD. This observed wear is well within the predicted wear from the operating assessment performed in RF-1 8 (approximated to be 21% TWD).

ETSS 96004.1, Rev. 13, was utilized for sizing AVB wear. The ETSS correlation for true depth to indicated depth is:

T = 0.98*1 + 2.89, where T is predicted depth and I is the indicated depth from EC.

From this correlation, the predicted depth of the maximum indicated wear (10%) is 12.69%.

The standard deviation for the correlation, a = 4.19. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645. Therefore, at 95% probability, the sizing uncertainty is 6.89% (1.645x4.19).

The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%

probability is:

Ut = 1.12

• 6.89% = 7.72%

Document Control Desk Attachment LTD 322 RC-14-0171 Page 9 of 11 The potential maximum depth of the indication at a confidence level > 95/50 is:

Maximum Predicted Indication 12.69% TWD NDE Sizing Uncertainty (95/50 CL) 7.72% TWD Total (>95/50) 20.41% TWD Structural Limit for AVB Wear 72% TWD Since the maximum observed AVB wear indication of 20.41 % TWD at 95% probability and 50% confidence is less than the structural limit of 72% TWD (defined in the RF-21 Degradation Assessment), the requirements for condition monitoring are met for AVB wear.

8.2 Tube Wear at TSPs and FDB The maximum TSP and FDB wear indication from RF-21 was 29% TWD and 16%

TWD, respectively. This observed wear is well within the predicted wear from the operating assessment performed at RF-18 (approximated to be 40% TWD and 24%

TWD, respectively). ..

ETSS 96004.1, Rev. 13, was utilized for sizing TSP and FDB wear. The ETSS correlationh for true depth to indicated depth is:

T = 0.98*1 + 2.89, where T is predicted depth and I is the indicated depth from EC.

From this correlation, the predicted depth of the maximum indicatedwear (29%) is 31.31%.

The standard deviation for the correlation, a = 4.19. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645. Therefore, at 95% probability, the sizing uncertainty is 6.89% (1.645 x 4.19).

The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%

probability is:

Ut = 1.12

• 6.89% = 7.72%

Document Control Desk Attachment LTD 322 RC-14-0171 Page 10 of 11 The potential maximum depth of the indication at a confidence level > 95/50 is:

Maximum Predicted Indication 31.31% TWD NDE Sizing Uncertainty (95/50 CL) 7.72% TWD Total (>95/50) 39.03% TWD Structural Limit for TSP and FDB Wear 66% TWD Since the maximum observed TSP and FDB wear indication of 39.03% TWD at 95%

probability and 50% confidence is less than the structural limit of 66% TWD (defined in the RF-21 Degradation Assessment), the requirements for condition monitoring are met for TSP and FDB wear.

8.3 Tube Wear Due to Foreign Objects The maximum foreign object wear indication from RF-21 was 10% TWD.

ETSS 27903.3, Rev. 1, was utilized for sizing foreign object wear. The ETSS correlation for true depth to indicated depth is: .

T =1.21*1 + 26.3; .

where T is predicted depth and I is the indicated depth from EC.

From this correlation, the predicted depth of the maximum indicated wear (10%) is 38.4%. The standard deviation for the correlation, a = 4.99. The multiplier on the standard deviation to achieve a 95% probability for a normal distribution is 1.645.

Therefore, at 95% probability, the sizing uncertainty is 8.21% (1.645x4.99).

The EPRI Steam Generator Integrity Assessment Guidelines specifies a factor of 1.12 to adjust for analyst uncertainty; thus, the total uncertainty, Ut, for sizing at 95%

probability is:

Ut = 1.12

• 8.21% = 9.19%

The potential maximum depth of the indication at a confidence level > 95/50 is:

Maximum Predicted Indication 38.4% TWD NDE Sizing Uncertainty (95/50 CL) 9.19% TWD Total (>95/50) 47.6% TWD Structural Limit for Foreign Object Wear 65% TWD

Document Control Desk Attachment -

LTD 322 RC-14-0171 Page 11 of 11 Since the maximum observed foreign object wear indication of 47.6% TWD at 95%

probability and 50% confidence is less than the structural limit of 65% TWD, the requirements for condition monitoring are met for foreign object wear.

9.0 References 9.1 Westinghouse Document SG-SGMP-14-1, V.C. Summer 1 RF21 (April 2014)

Steam Generator Degradation Assessment, Rev. 2, April 2014.

9.2 Westinghouse Document SG-SGMP-14-15, V.C. Summer Unit 1 Steam Generator Cycle 21 Condition Monitoring and Cycles 22, 23, and 24 Operational Assessment, Rev. 0, May 2014.

9.3 V.C. Summer Ul R21 SG Eddy Current Inspection Report, April 2014.

'.' J.