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Category:Fuel Cycle Reload Report
MONTHYEARML24267A0402024-09-19019 September 2024 Cycle 27 Core Operating Limits Report Revision 0 ML24109A0272024-04-16016 April 2024 Cycle 27 Core Operating Limits Report Revision 0 ML23103A1952023-04-13013 April 2023 Cycle 26 Core Operating Limits Report Revision 0 ML22332A4632022-11-28028 November 2022 Cycle 26 Core Operating Limits Report ML21319A3412021-11-12012 November 2021 Cycle 25 Core Operating Limits Report, Revision 0 ML21138A8722021-05-18018 May 2021 Cycle 25 Core Operating Limits Report, Revision 0 ML20121A0092020-04-29029 April 2020 Cycle 24 Core Operating Limits Report Revisions 0, 1 and 2 ML19345D2522019-12-0909 December 2019 Cycle 24 Core Operating Limits Report, Revisions 1 and 2 ML19301A4712019-10-24024 October 2019 Cycle 24 Core Operating Limits Report Revision 0 ML18319A1902018-11-14014 November 2018 Cycle 23 Core Operating Limits Report, Revision 0 ML18128A1782018-05-0808 May 2018 Cycle 23 Core Operating Limits Report, Revision 0 ML17152A2722017-06-0101 June 2017 Cycle 22 Core Operating Limits Report, Revision 0 ML16357A5562016-12-21021 December 2016 Cycle 22 Core Operating Limits Report, Revision 0 ML15364A0102015-12-23023 December 2015 Cycle 21 Core Operating Limits Report ML15328A0522015-11-16016 November 2015 Cycle 21 and Unit 2 Cycle 20 Core Operating Limits Reports, Revision No. 1 ML15154A5122015-05-27027 May 2015 Cycle 21 Core Operating Limits Report, Revision 0 ML14181A0382014-06-25025 June 2014 Cycle 20 Core Operating Limits Report ML13346A4272013-12-0404 December 2013 Cycle 20 Core Operating Limits Report ML13010A3862013-01-0808 January 2013 Cycle 19 Core Operating Limits Report ML12104A2682012-04-10010 April 2012 Cycle 19, Core Operating Limits Report ML0934909642009-12-14014 December 2009 Cycle 17 Core Operating Limits Report ML0912402462009-04-30030 April 2009 Cycle 17 Core Operating Limits Report (COLR) ML0912106992009-04-27027 April 2009 Submittal of Cycle 16 Core Operating Limits Report (Colr), Revision 1 ML0902603172008-11-21021 November 2008 Cycle 15 - 180-Day - Steam Generator Inspection Report ML0815803012008-06-0303 June 2008 Unit 2 Cycle 16 Core Operating Limits Report (COLR) ML0812901852008-04-23023 April 2008 Cycle 15 (U1C15) - 180-Day - Steam Generator (SG) Inspection Report ML0810803372008-04-14014 April 2008 Technical Specification Change - 08-01 Revision of Core Operating Limits Report References for Realistic Large Break Loss of Coolant Accident Methodology ML0732000922007-11-14014 November 2007 Cycle 16 Core Operating Limits Report (COLR) Revision ML0715606012007-06-0404 June 2007 Cycle 15 Core Operating Limits Report, Revision 1 ML0704401472007-01-31031 January 2007 Cycle 15 Core Operating Limits Report (COLR) Revision 1 ML0636204062006-12-20020 December 2006 Cycle 15 Core Operating Limits Report (COLR) Revision ML0612404542006-05-0404 May 2006 Cycle 14 Core Operating Limits Report (COLR) Revision ML0514408132005-05-23023 May 2005 Unit 2 Cycle 14 Core Operating Limits Report (COLR) ML0432900602004-11-23023 November 2004 Cycle 14 Core Operating Limits Report ML0335107792003-12-0909 December 2003 Cycle 13 Core Operating Limits Report (COLR) ML0315507682003-06-0303 June 2003 Unit 1 Cycle 13 Core Operating Limits Report (COLR) ML0212904362002-05-0808 May 2002 Unit 2 Cycle 12 Core Operating Limits Report (COLR) 2024-09-19
[Table view] Category:Letter
MONTHYEARML24304A8492024-10-31031 October 2024 December 2024 Requalification Inspection Notification Letter IR 05000327/20250102024-10-29029 October 2024 Notification of Sequoyah, Units 1 and 2 - Comprehensive Engineering Team Inspection - U.S. Nuclear Regulatory Commission Inspection Report 05000327/2025010 and 05000328/2025010 ML24298A1172024-10-24024 October 2024 Cycle 26, 180-Day Steam Generator Tube Inspection Report CNL-24-074, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-10-23023 October 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions 05000327/LER-2024-001, Reactor Trip Due to a Turbine Trip2024-10-17017 October 2024 Reactor Trip Due to a Turbine Trip ML24282B0412024-10-15015 October 2024 Request for Withholding Information from Public Disclosure for Sequoyah Nuclear Plant, Units 1 and 2 and Watts Bar Nuclear Plant, Units 1 and 2 ML24260A1682024-10-0404 October 2024 Regulatory Audit Summary Related to Request to Add and Revise Notes Related to Technical Specification Table 3.3.2-1, Function 5 ML24284A1072024-09-26026 September 2024 Affidavit for Request for Withholding Information from Public Disclosure for Sequoyah Nuclear Plant, Units 1 and 2, and Watts Bar Nuclear Plant, Units 1 and 2 05000328/LER-2024-001, Reactor Trip Due to an Electrical Trouble Turbine Trip2024-09-25025 September 2024 Reactor Trip Due to an Electrical Trouble Turbine Trip CNL-24-060, Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description2024-09-24024 September 2024 Supplement to Request for Approval of the Tennessee Valley Authority Nuclear Quality Assurance Program Description CNL-24-047, Decommitment of Flood Mode Mitigation Improvement Systems2024-09-24024 September 2024 Decommitment of Flood Mode Mitigation Improvement Systems ML24262A0602024-09-23023 September 2024 Summary of August 19, 2024, Meeting with Tennessee Valley Authority Regarding a Proposed Supplement to the Tennessee Valley Authority Nuclear Quality Assurance Plan ML24267A0402024-09-19019 September 2024 Cycle 27 Core Operating Limits Report Revision 0 CNL-24-065, Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-09-18018 September 2024 Tennessee Valley Authority – Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24185A1742024-09-18018 September 2024 Cover Letter - Issuance of Exemption Related to Non-Destructive Examination Compliance Regarding Sequoyah Nuclear Plant Independent Spent Fuel Storage Installation ML24253A0152024-09-0808 September 2024 Emergency Plan Implementing Procedure Revisions ML24247A2212024-08-29029 August 2024 Notification of Deviation from Pressurized Water Reactor Owners Group (PWROG) Letter OG-21-160, NEI 03-08 Needed Guidance: PWR Lower Radial Support Clevis Insert X-750 Bolt Inspection Requirements, September 1, 2021 ML24247A1802024-08-28028 August 2024 Application to Revise the Fuel Handling Accident Analysis, to Delete Technical Specification 3.9.4, Containment Penetrations, and to Modify Technical Specification 3.3.6, Containment Ventilation Isolation Instrumentation for Sequoyah Nuclea IR 05000327/20240052024-08-26026 August 2024 Updated Inspection Plan for Sequoyah Nuclear Plant, Units 1 and 2 - Report 05000327/2024005 and 05000328/2024005 ML24239A3972024-08-23023 August 2024 Rssc Wire & Cable LLC Dba Marmon - Part 21 Final Notification - 57243-EN 57243 CNL-24-061, Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08),2024-08-19019 August 2024 Supplement to Application to Revise Function 5 of Technical Specification Table 3.3.2-1, ‘Engineered Safety Feature Actuation System Instrumentation,’ for the Sequoyah and Watts Bar (SQN-TS-23-02 and WBN-TS-23-08), IR 05000327/20240022024-07-31031 July 2024 Integrated Inspection Report 05000327/2024002 and 05000328/2024002 ML24211A0572024-07-29029 July 2024 Submittal of Emergency Plan Implementing Procedure Revision ML24211A0542024-07-29029 July 2024 Operator License Examination Report ML24211A0412024-07-26026 July 2024 Unit 1 Cycle 26 Refueling Outage - 90-Day Inservice Inspection Summary Report ML24199A0012024-07-22022 July 2024 Clarification and Correction to Exemption from Requirement of 10 CFR 37.11(c)(2) ML24172A1342024-07-15015 July 2024 Exemptions from 10 CFR 37.11(C)(2) (EPID L-2023-LLE-0024) - Letter ML24191A4652024-07-0909 July 2024 Response to NRC Regulatory Issue Summary 2024-01, Preparation and Scheduling of Operator Licensing Examinations ML24177A0282024-06-25025 June 2024 Emergency Plan Implementing Procedure Revisions ML24176A0222024-06-24024 June 2024 Retraction of Interim Report of a Deviation or Failure to Comply – Transducer Model 8005N ML24089A1152024-06-21021 June 2024 Transmittal Letter, Environmental Assessments and Findings of No Significant Impact Related to Exemption Requests from 10 CFR 37.11(c)(2) ML24145A0852024-05-30030 May 2024 1B-B Diesel Generator Failure - Final Significance Determination Letter ML24145A1052024-05-29029 May 2024 301 Exam Approval Letter ML24134A1762024-05-13013 May 2024 Submittal of 2023 Annual Radiological Environmental Operating Report CNL-24-040, Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-05-0808 May 2024 Tennessee Valley Authority - Central Emergency Control Center Emergency Plan Implementing Procedure Revisions ML24128A0352024-05-0707 May 2024 Providing Supplemental Information to Apparent Violation ML24120A0582024-04-26026 April 2024 10 CFR 50.46 Annual Report for Sequoyah Nuclear Plant Units 1 and 2 ML24116A2612024-04-25025 April 2024 Interim Report of a Deviation or Failure to Comply - Transducer Model 8005N ML24114A0482024-04-23023 April 2024 Annual Radioactive Effluent Release Report for 2023 Monitoring Period CNL-24-037, Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 422024-04-22022 April 2024 Clinch River, Sequoyah, Units 1 and 2, Watts Bar, Unit 1 and 2, Nuclear Quality Assurance Plan, TVA-NQA-PLN89-A, Revision 42 ML24144A2362024-04-20020 April 2024 Discharge Monitoring Report (Dmr), March 2024 ML24144A2322024-04-20020 April 2024 Tennessee Multi-Sector Permit (Tmsp), 2024 Annual Discharge Monitoring Report for Outfalls SW-3, SW-3, and SW-9 ML24089A0882024-04-18018 April 2024 – Exemption from Select Requirements of 10 CFR Part 73; Security Notifications, Reports, and Recordkeeping and Suspicious Activity Reporting ML24102A1212024-04-18018 April 2024 Summary of Conference Call with Tennessee Valley Authority Regarding Sequoyah Nuclear Plant, Unit 1 Spring 2024 Steam Generator Tube Inspections CNL-24-024, Hydrologic Engineering Center River Analysis System Project Milestone Status Update2024-04-17017 April 2024 Hydrologic Engineering Center River Analysis System Project Milestone Status Update CNL-24-033, Central Emergency Control Center Emergency Plan Implementing Procedure Revisions2024-04-17017 April 2024 Central Emergency Control Center Emergency Plan Implementing Procedure Revisions IR 05000327/20240012024-04-17017 April 2024 Integrated Inspection Report 05000327/2024001 and 05000328/2024001 ML24109A0272024-04-16016 April 2024 Cycle 27 Core Operating Limits Report Revision 0 CNL-23-006, Application to Modify Technical Specifications 3.8.1, AC Sources – Operating, and 3.8.2, AC Sources – Shutdown, for Sequoyah Nuclear Plant (SQN-TSC-22-03)2024-04-15015 April 2024 Application to Modify Technical Specifications 3.8.1, AC Sources – Operating, and 3.8.2, AC Sources – Shutdown, for Sequoyah Nuclear Plant (SQN-TSC-22-03) ML24106A0502024-04-12012 April 2024 Discharge Monitoring Report (Dmr), February 2024 2024-09-08
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Tennessee Valley Authority, Post Office Box 2000, Sodciy-Daisy, Tennessee 37384-2000 April 23, 2008 U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555 Gentlemen:
In the Matter of ) Docket No. 50-327 Tennessee Valley Authority (TVA) )
SEQUOYAH NUCLEAR PLANT (SQN) - UNIT 1 CYCLE 15 (U1C15) - 180-DAY -
STEAM GENERATOR (SG) INSPECTION REPORT In accordance with the requirements of SQN Unit 1 Technical Specification 6.8.4.k, Steam Generator Program and Technical Specification 6.9.1.16, Steam Generator Inspection report, TVA is submitting the 180-day report that includes the results of inservice inspections performed on Unit 1 SGs during the U1C15 refueling outage.
There are no commitments contained in this letter.
Please direct questions concerning this issue to me at (423) 843-7170 or Russell R. Thompson at (423) 843-6672.
Sincerely, James D. Smith Manager, Site Licensing and Industry Affairs Enclosure cc: See page 2 Prinjec on recycled paper
U.S. Nuclear Regulatory Commission Page 2 April 23, 2008 cc (Enclosure):
Mr. Thomas H. Boyce, Senior Project Manager U.S. Nuclear Regulatory Commission Mail Stop 08G-9a One White Flint North 11555 Rockville Pike Rockville, Maryland 20852-2739
Tennessee Valley Authority Sequoyah Nuclear Plant Unit 1 Cycle 15 Refueling Outage October 2007 180 Day Steam Generator Inspection Report Prepared by:
Verified by:
Approved by: />t*<< ^/^MMA ?>'>lzfL Zf/of
SQN Unit 1 Cycle 15 180-Day Report In accordance with Technical Specification 6.8.4.k, Steam Generator Program and Technical Specification 6.9.1.16, Steam Generator Tube Inspection Report, this report documents the results of the Unit 1 Cycle 15 (U1C15) steam generator (SG) tube inservice inspection. The following list is the required scope of the 180-day report.
- 1. The scope of inspections performed on each SG.
EDDY CURRENT EXAM TYPE (Exams) RSG 1 RSG 2 RSG 3 RSG 4 Total Full Length Bobbin (.610 coil) 2477 2481 2479 2480 9917 Full Length Bobbin (.590 coil) 42 42 42 42 168 Partial Length Bobbin (.570 coil HTE-VS3) 123 122 123 123 491 Expansion Full Length Bobbin (.610 coil) 479 286 116 0 881 Predetermined Diagnostic Plus Point 2 3 29 1 35 Diagnostic/PID Plus Point 0 1 1 1 3 Total Exams Completed 3123 2935 2790 2647 11495 Total Tubes Examined 3121 2933 2775 2645 11474 In accordance with the U1C15 Degradation Assessment, expansion sample examinations were performed in SGs 1, 2, and 3. No expansion sample was required in SG 4.
- 2. Active Degradation Mechanisms found.
INDICATIONS (Tubes) RSG 1 RSG 2 RSG 3 RSG 4 Total U-Bend Support Wear 32 2 11 2 47 Total 32 2 11 2 47
- 3. Nondestructive examination technique utilized for each degradation mechanism.
Bobbin coil examinations were utilized for the detection of U-bend Support Wear.
- 4. Location, orientation (linear), and measured sizes (if available) of service induced indications.
Refer to Table 1
SQN Unit 1 Cycle 15 180-Day Report
- 5. Number of tubes plugged during the inspection outage for each active degradation mechanism.
PLUGGING STATUS RSG 1 RSG 2 RSG 3 RSG 4 TOTAL Previously Plugged Tubes 14 6 6 5 31 Tubes Plugged by Damage Mechanism U-Bend Support Wear 0 0 0 0 0 Total Plugged Cycle 15 0 0 0 0 0 Total Tubes Plugged 14 6 6 5 31 Plugged Tube Percentage 0.28% 0.12% 0.12% 0.10% 0.16%
- 6. Total number and percentage of tubes plugged to date.
Refer to the response to Number 5 above. The U-bend Support Wear indications were tapered along the tube axially.
- 7. The results of condition monitoring, including the results of tubes pulls and in-situ testing.
No tube pulls or in-situ pressure tests were performed.
U-BEND SUPPORT WEAR Structural Integrity The SQN U1C15 degradation assessment predicted, based on past indications and growth rate data, that 25 tubes could be preventively plugged. A total of 51 indications in 47 tubes were detected. None exceeded the plugging limit of 40 percent. The U-bend Support Wear was assumed to have grown from 0 percent through wall over the last 2 fuel cycles. Using these growth rates, a repair limit of 30 percent was calculated which would ensure that a safety factor of 3 would be maintained (with a 95 percent probability) until the next planned inspection during the U1C18 refueling outage. Condition monitoring assumed the axial length of the U-bend Support Wear indication to be the width of the U-bend Support (i.e., 2 inches).
All indications were a fraction of this length. The limiting indications were 16 percent maximum depth. The limiting tubes were in SG1, Row 76, Col 100 at VS2-0.97, Row 85, Col 67 at VS4+0.91, and Row 87, Col 73 at VS2+1.28. The EPRI, Steam Generator Degradation Specific Management Flaw Handbook, Section 5.3.3, Axial Thinning with Limited Circumferential Extent, equations were utilized to calculate the burst pressure. The calculated 95th percentile lower limit burst pressure was 6958 psig (3P is 4114 psig). The largest amplitude U-bend Support Wear indication was 0.58 volts, and therefore did not exceed the in-situ pressure testing leakage screening criteria of 7.9 volts. All U-bend Support Wear indications met condition monitoring performance criteria. Zero indications exceeded the 40 percent plugging limit and zero indications exceeded the 30 percent repair limit.
SQN Unit 1 Cycle 15 180-Day Report Leakage Integrity During the previous two fuel cycles, the Unit 1 operational primary to secondary leakage was below detection. The majority of the indications detected during U1C15 were less than 0.5 volts which in accordance with EPRI Steam Generator In-Situ Pressure Test Guidelines Revision 2 (TR-1007904), dated August 2003, Section B.3.4 is a quick screen for structural and leakage integrity, and therefore provides a 95 percent probability that none of the indications would leak should a postulated main steam line break have occurred. The greatest through wall was 16 percent.
The associated 95th percentile lower limit burst pressure was 6958 psig. The equations utilized for burst pressure determine the ligament tearing pressure or the pressure at which a partial through-wall indication goes through-wall (i.e., pop-through). If an indication has a 95 percent probability of not going through wall at 6958 psig, then it should not go through wall at the Main Steam Line Break pressure of 2405 psig. This provides additional confidence that no leakage would occur should a postulated main steam line break occur.
- 8. The effective plugging percentage of all plugging in each SG.
Refer to the response to Number 5 above.
SQN Unit 1 Cycle 15 180 Day Report Table 1 List of Indications The indications below were characterized as U-bend Support Wear. They were detected by bobbin coil. The associated maximum depth and voltage are included. The plugging limit at SQN Unit 1 is 40 percent through wall (TW).
SG ROW COL LOCATION %TW VOLTS CHARACTERIZATION RESOLUTION In service <30% Repair 1 68 76 VS2-.26 7 0.21 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 75 73 VS3+.80 9 0.28 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 78 78 VS2+.91 8 0.23 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 83 77 VS2+.91 10 0.32 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 87 73 VS2+1.28 16 0.58 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 87 73 VS3+1.00 15 0.51 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 87 75 VS2+1.14 13 0.4 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 87 81 VS4-1.11 15 0.53 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 88 70 VS4-1.11 13 0.39 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 90 70 VS4-1.00 13 0.37 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 92 70 VS2-.91 9 0.24 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 92 70 VS4+1.00 15 0.45 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 95 77 VS4-1.11 9 0.27 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 96 76 VS3+1.28 11 0.35 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 55 115 VS3-.91 15 0.47 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 61 85 VS2-1.09 8 0.21 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 67 87 VS3-.92 6 0.19 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 76 90 VS2+.86 9 0.26 U-BEND SUPPORT WEAR Limit In service <30% Repair 4 68 64 VS3-.83 10 0.37 U-BEND SUPPORT WEAR Limit In service <30% Repair 4 97 55 VS4-.83 8 0.29 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 69 59 VS3+.77 10 0.31 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 70 66 VS4-1.03 15 0.53 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 74 48 VS3-.88 6 0.17 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 83 45 VS4-.91 6 0.15 U-BEND SUPPORT WEAR Limit In service <30% Repair 2 20 80 VS3-.06 14 0.45 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 83 49 VS2+.77 7 0.2 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 84 64 VS4+1.03 11 0.31 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 85 65 VS4+.80 11 0.33 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 85 67 VS4+.91 16 0.54 U-BEND SUPPORT WEAR Limit 1 88 52 VS4-.91 8 0.22 U-BEND SUPPORT WEAR In service <30% Repair 1
SQN Unit 1 Cycle 15 180 Day Report Table 1 List of Indications SG ROW COL LOCATION %TW VOLTS CHARACTERIZATION RESOLUTION Limit In service <30% Repair 1 92 52 VS3+.83 6 0.17 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 92 68 VS4+.62 14 0.41 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 93 65 VS2+.85 9 0.25 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 95 63 VS3-.63 13 0.42 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 95 63 VS4-1.14 11 0.33 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 98 66 VS3+.82 8 0.24 U-BEND SUPPORT WEAR Limit In service <30% Repair 2 92 42 VS2-.74 8 0.23 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 62 112 VS3+.77 13 0.41 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 66 110 VS3+.80 11 0.34 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 76 100 VS2-.97 16 0.56 U-BEND SUPPORT WEAR Limit In service <30% Repair 1 92 88 VS4-.85 11 0.33 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 73 75 VS3+.74 7 0.23 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 79 79 VS2-.86 7 0.23 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 79 79 VS3+.86 8 0.27 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 80 72 VS4-.83 9 0.31 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 85 69 VS4+.83 10 0.35 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 87 63 VS2+1.07 9 0.28 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 87 65 VS3+.95 12 0.4 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 89 71 VS4+1.01 8 0.26 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 97 71 VS3+.74 8 0.26 U-BEND SUPPORT WEAR Limit In service <30% Repair 3 98 56 VS4-1.06 12 0.39 U-BEND SUPPORT WEAR Limit