ML17262B153
ML17262B153 | |
Person / Time | |
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Site: | Ginna ![]() |
Issue date: | 12/31/1992 |
From: | ROCHESTER GAS & ELECTRIC CORP. |
To: | |
Shared Package | |
ML17262B151 | List: |
References | |
PROC-921231, NUDOCS 9302020317 | |
Download: ML17262B153 (589) | |
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TITLE: DATE:
QUALITY APPENDIX B 12/3 1/92 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 1 of 4 TABLE OF CONTENTS Section 1.0 Inservice Inspection Program:
1.0 Introduction 1.1 General 1.2 Applicable Documents 1.3 Inspection Intervals 1.4 Classification of Components 1.5 Responsibility 1.6 Records 1.7 Examination Methods 1.8 Repair Requirements 1.9 Replacement Requirements 1.10 System Pressure Testing 2.0 CLASS 1 Program Plan 2.1 Basis for Preparation 2.2 Components Subject to Examination 2.3 Extent and Frequency of Examinations 2.4 Exemption Criteria 2.5 Examination of Reactor Coolant Pump Flywheels 2.6 Inservice Inspection Program Plan 3.0 CLASS 2 Program Plan 3.1 Basis for Preparation 3.2 Components Subject to Examination 3.3 Extent and Frequency of Examinations 3.4 Exemption Criteria 3.5 Inservice Inspection Program Plan 4.0 CLASS 3 Program Plan 4.1 Basis for Preparation 4.2 Components Subject to Examination 4.3 Extent and Frequency of Examinations 4 ' Exemption Criteria 4.5 Inservice Inspection Program Plan 5.0 CLASS 1, CLASS 2 6 CLASS 3 Component Supports 5.1 Basis for Preparation 5.2 Components Subject to Examination 5.3 Extent and Frequency of Examinations 5.4 Exemptions 5.5 Inservice Inspection Program Plan Section 2.0 Relief Requests:
1.0 Introduction 1.1 General 930202031? 930125 r PDR ADOCK 05000244 8 PDR
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QUALITY APPENDIX B 12/3 1/92 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 2 of 4 TABLE OF CONTENTS Section 3.0 PAID Boundary Drawings:
1.0 General P&ID Boundary Drawings Section 4.0 Code Tables:
1.0 General Table 1, Class 1 Table 2, Class 2 Table 3, Class 3 Table 4, Class 1, 2, and 3 IWF Component Supports Section 5.0 Line List:
1.0 General Line List Section 6.0 Allocation Tables:
1-.0 General ASME Code Required Class 1 Class 2 Class 3 Class 1, 2, and 3 Supports Augmented Required Reactor Coolant Pump Flywheel Program Reactor Vessel Augmented Program, Category B-A High Energy Program Snubber Program Section 7.0 UT Calibration Blocks:
1.0 General VT Calibration Blocks Section 8.0 High Energy Program:
1.0 General 2.0 Examination Requirements 3.0 Examination Methods 4.0 Frequency of Examination 5.0 Examination Evaluation 6.0 Repair and Testing Requirements
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QUALITY APPENDIX B 12/3 1/92 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 3 of 4 TABLE OF CONTENTS 7.0 Replacement and Testing Requirements 8.0 Scheduling 9.0 Reports and Records Table 1 High Energy Welds and Component Supports identification.
Section 9.0 Snubber Program:
1.0 General 2.0 Examination, Testing and Monitoring Requirements 3.0 Examination and Testing Methods 4.0 Examination and Testing Frequency 5.0 Examination, Testing and Monitoring Evaluation 6.0 Repair, Replacement and Modification Requirements 7.0 Scheduling 8.0 Reports and Records Section 10.0 Seismic Support Program:
1.0 General 2.0 Examination Requirements 3.0 Examination Methods 4.0 Frequency of Examination 5.0 Examination Evaluation 6.0 Repair and Testing Requirements 7.0 Replacement and Testing Requirements 8.0 Scheduling 9.0 Reports and Records Table 1 Seismic Support Identification Section 11.0 Additional Programs:
1.0 General 2.0 Steam Generator Tube Inspection Program 2.1 General 2.2 Examination Requirements 2.3 Examination Methods 2.4 Frequency of Examination 2.5 Examination Evaluation 2.6 Repair, Replacement and Testing Requirements 2.7 Scheduling 2.8 Reports and Records 3.0 Reactor Coolant Pump Flywheel Program 3.1 General 3.2 Examination Requirements 3.3 Examination Methods'
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QUALITY APPENDIX B 12/3 1/92 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 4 of 4 TABLE OF CONTENTS 3.4 Frequency of Examination 3.5 Examination Evaluation 3.6 Repair, Replacement and Testing Requirements 3.7 Scheduling 3.8 Reports and Records 4.0 Class 1 Bolting Program (IEB 82-02) 4.1 General 4.2 Examination Requirements 4.3 Examination Methods 4' Frequency of Examination 4.5 Examination Evaluation 4.6 Repair, Replacement and Testing Requirements 4.7 Scheduling 4.8 Reports and Records 5.0 Reactor Vessel Augmented Program, Category B-A 5.1 General 5.2 Examination Requirements 5.3 Examination Methods 5.4 Frequency of Examination 5.5 Examination Evaluation 5.6 Repair, Replacement and Testing Requirements 5.7 Scheduling 5.8 Reports and Records SUPPLEMENT 1 TO APPENDIX B "Inservice Inspection Program Plan" (Seperate Document controlled by Materials Engineering & Inspection Services)
SUPPLEMENT 2 TO APPENDIX B "The Repair, Replacement and Modification Program"
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PAGE QUALITY ASSURANCE MANUAL GINNA STATION Section 1 1 of 30 EFFECTIVE DATE:
December 31, 1992 ROCHESTER GAS 8 ELECTRIC CORPORATION SIGNATURE DATE TITLE: PREPARED BY APPENDIX B QUALITY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM REVIEW:
FOR THE 1990-1999 INTERVAL APPROVED Y: fg."Af-g
1.0 INTRODUCTION
General This Appendix "B" to the Quality Assurance Manual outlines the third interval inservice inspection examination (ISI) requirements for Class 1, Class 2, and Class 3 systems, and components for Rochester Gas & Electric Corporation's (RG&E) R.
E. Ginna Nuclear Power Plant (Ginna Station). The third inspection interval begins on January 1, 1990, as permitted by Paragraph IWA-2400 of ASME Code Section XI, the second interval concluded December 31, 1989.
1~ 1.1 This program is based on the requirements of the 1986 Edition of the American Society for Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, "Inservice Inspection of Nuclear Power Plant Components" as adopted by the Code of the Federal Regulations, 10 CFR Part 50, (Federal Register 53FR16051) May 5, 1988.
- l. 1.2 This program excludes the controls of the Enforcement Authority, and N-Stamp, in addition to IWE by of the ASME Section XI, since Regulation.
it is not endorsed
- 1. 1.3 Inservice Testing of pumps (IWP) and valves (IWV) is performed in accordance with Appendix C of the Ginna Station Quality Assurance Manual.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 2 of 30 1.2 Applicable Documents RG&E has adopted the following documents as the for the third inspection interval and is 'asis committed to satisfying their requirements.
Specific exceptions to the requirements of ASME Section XI are identified and located in the "Relief Requests", Section 2.0 of this document.
This program was developed in accordance with these documents:
1.2. 1 ASME Boiler and Pressure Vessel Code Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," 1986 Edition, no Addenda.
1.2.1. 1 ASME Boiler 6 Pressure Vessel Code,Section XI, 1986 Edition, no Addenda, Appendix IV.
1.2.2 U.S. Nuclear Regulatory Commission (USNRC)
Regulatory Guides:
a ~ 1.14, Rev. 1, "Reactor Coolant Pump Flywheel Integrity"
- b. 1.147, Latest Revision, "Inservice Inspection Code Case Acceptability ASME Section XI,,
Division l" c ~ 1.150, Rev. 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations"
- d. 1.83, Rev. 1, "Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes"
- e. 1.26, Rev. 3 "Quality Group Classifications and Standards for Water, Steam, and Radioactive Waste Containing Components of Nuclear Power Plants.
1.29, Rev. 3, "Seismic Design Classification".
g, 1.121, Rev. 0, "Bases for Plugging Degrated PWR Steam Generator Tubes".
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUALSTATION'ITLE INSERVICE INSPECTION PROGRAM PAGE GINNA FOR THE 1990-1999 INTERVAL 3 of 30 1.2.3 '- ASME Code Cases In accordance with 10CFR50.55a, Footnote 6, ASME Section XI Code Cases referenced in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability- ASME Section XI, Division l," may be incorporated into the Ginna Station ISI Program. Those Code Cases included in Regulatory Guide 1.147 that will be implemented at Ginna Station are identified in this section. Each Code Case is preceded by information on the applicable component/area, ASME requirements, and how the Code Case will be implemented.
,Paragraph 1.2.3.1 lists those code cases that have been technically reviewed and endorsed, with or without, conditions by the NRC in Regulatory Guide 1.147 and will be implemented during the 3rd ten year Inspection Interval.
Use of any subsequent NRC endorsed code cases that are identified in revisions to the Regulatory Guide 1.147 may be incorporated in the program and used during the 3rd ten year Inspection Interval.
1 ~ 2.3 ~ 1 USNRC Regulatory Guide 1.147 Approved ASME Section XI Code Cases Code Case No. Sect. XI References Com onent Area N 307-1 IWB-2500-1 Studs and Bolts with Heater Holes N-416 IWA-4400 Any repaired or replaced Class 2 piping component that cannot be isolated by valves or requires securing safety/
relief valves.
N-401 IWA-2233 Eddy Current Examinations N-402 IWA-2233 Eddy Current Calibration Standard Material
1 TITLE QUALITY APPENDIX Section 1 ASSURANCE GINNA NUCLEAR POWER PLANT B'.E-MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 4 of 30 N-427* Code Cases Code Cases in Inspection Plans N-437 IWA-5260 Use of digital readout and digital measurement devices for performing Pressure Tests.
N-446 IWA-2300 Recertification of Visual Examination Personnel.
N-460 IWA-2000/3000 Alternative examination coverage for Class 1 and Class 2 welds.
N-498 IWX-5000 Alternative rules for 10-year Hydrostatic Pressure Testing for Class 1 and 2 systems.
- As amended by USNRC Regulatory Guide 1.147, April, 1992.
1.2.4 R.E. Ginna Updated Final Safety Analysis Report (UFSAR):
Section 5.4 For Class 1 Section 6.6 For Class 2 and 3 1.2.5 Letter dated March 23, 1981, from Darrell G.
Eisenhut, Director, Division of Licensing, USNRC, regarding Technical Specification Revisions for Snubber Surveillance 1.2.5.1 First Addenda to ASME/ANSI OM-1987, Part 4 Published in 1988.
1.2.6 USAS B31.1.0-1967, "Power Piping" for High Energy Systems 1 '.7 R. E. Ginna, Technical Specification 4.2.
"Inservice Inspection".
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 5 of 30 1.2.8 R. E. Ginna Technical Specification 3.13 and 4.14 "Snubbers".
1.2.9 Letter dated February 16, 1989, to Mr. Carl Stahle, USNRC, PWR Project Directorate No. 1, regarding proposed changes to Snubber Technical Specification R. E. Ginna Nuclear Power Plant, Docket No. 50-244.
1.2.10 Rochester Gas and Electric Corporation Mechanical Engineering Specification, ME-256, Titled "Snubber Inspection and Test Program".
1.2.11 Electric Power Research Institute PWR Steam Generator Inspection Guidelines, Rev. 2.
1.2.12 Code of the Federal Regulations, 10CFR Part 50.
1.3 Inspection Intervals 1.3. 1 The inservice inspection intervals for Class 1 components started on January 1, 1970, with the second interval starting on January 1, 1980. The third inspection interval shall start on January 1/ 1990.
1.3.2 For Class 2 and Class 3, the first inspection interval started on May 1, 1973; the second on January 1, 1980; and the third on January 1, 1990.
1.3.3 The third inspection interval for Class 1, 2 and 3 is scheduled to end December 31, 1999. However, this date is subject to change as allowed by IWA-2400, which states that each inspection interval may be decreased or extended (but not cumulatively) by as much as one year. If R. E.
Ginna Nuclear Power Plant is out of service continuously for 6 months or more, the inspection interval and associated period during which the outage occurred may be extended for a period of time equivalent to the outage.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 6 of 30 1.3.4 A 10-year examination Program plan.-(Supplement. 1 to Appendix B), will describe the distribution of examinations for Class 1, Class 2, and Class 3 components in accordance with Inspection Program B, the IWB-2400, IWC-2400, IWD-2400 and IWF-2400 of Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components", 1986 Edition, no Addenda.
1.4 Classification of Components The Program Plan components and piping have been classified by RGGE for purposes of inservice inspection based on Section XI, Article IWA-1320, and definitions contained in 10CFR50.2.
1.5 Responsibility As specified in Paragraph IWA-1400 of ASME Section XI, RG6E bears the overall responsibility for implementation of an ISI program. Administrative Procedures, NDE Procedures, ISI Plans and Schedules are in place to control and implement these inservice inspection requirements.
1.6 Records Examination records and documentation of results provide the basis for evaluation and facilitate comparison with previous results and subsequent inspections. In accordance with Section XI, IWA-6000, these records will be maintained for the plant life.
1.6. 1 An Inservice Inspection Report shall be generated to document applicable Inservice Inspection and associated Repair, Replacement and Modification activities. ASME NIS-l and NIS-2 Forms shall be generated and included within the Inservice Inspection Report.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 7 of 30 1.7 Examination Methods Examination methods which will be used to satisfy Code examination requirements have been listed for nonexempt Class 1, Class 2, and Class 3 components, as applicable.
Provided in the following is a brief explanation of the examination methods which will be performed to satisfy the Code requirements.
Personnel performing nondestructive examinations will be qualified in accordance with written procedures prepared as required .by Paragraph IWA-2300 of Section XI. Methods of examination are also described in the applicable sections of this Inservice Inspection Program.
1.7. 1 Visual Examination Method Visual examinations (VT) will be performed in accordance with IWA-2210 of ASME Section XI.
IWA-2210 defines the three types of VT examinations as follows:
1.7.1.1 VT-1 examinations are conducted to determine the condition of the part, component or surface examined. The examination shall determine conditions such as cracks, wear, corrosion, erosion, or physical damage on the surfaces of the part or components. This type of examination may be performed by direct or remote methods as defined in IWA-2211.
1.7. 1.2 VT-2 examinations are conducted to detect leakage (or abnormal leakage) from pressure-retaining components during system pressure or functional tests as defined in IWA-2212.
1.7.1.3 VT-3 examinations are conducted to determine general mechanical and structural conditions of components and their supports such as the presence of loose parts, debris, or abnormal corrosion products, wear, erosion, corrosion, and the loss of integrity at bolted or welded connections. VT-3 Examinations are also conducted to determine conditions related to operability of mechanical and hydraulic snubbers, and spring devices.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 8 of 30 1.7.2 Surface Examination Method 1.7.2. 1 A surface examination is performed to detect the presence of discontinuities open to the surface of a material. Techniques for surface examination include either magnetic particle (MT) or liquid penetrant (PT) methods. Surface examinations will be conducted as defined in IWA-2220.
1.7.3 'Volumetric Examination Method 1.7.3.1 A volumetric examination is performed to detect the presence of discontinuities in the volume of a material. Such volumetric examinations include radiographic (RT), ultrasonic (UT), and eddy current (ET). Volumetric examinations will be conducted as defined in IWA-2230.
1.7.3.2 Radiography may be performed by utilizing either x-ray or gamma ray techniques.
1.7.3.3 The UT examinations may be performed by utilizing either manual or mechanized UT (Mech UT) techniques in accordance with Appendix I of Section XI and Regulatory Guide 1.150, Rev. 1 for the Reactor Vessel Examination only.
1.7.3.4 The ET Method will be utilized in the examination of heat exchanger tubing in accordance with Appendix IV of ASME Section XI and USNRC Regulatory Guide 1.83.
1.7.4 Alternative Examination Methods 1.7.4. 1 Alternative examination methods may be performed to those described in 1.7.1, 1.7.2 and 1.7.3 as allowed in IWA-2240. These may include such things as newly developed techniques, provided that these alternative methods are shown by practical demonstration to be equivalent or superior to those of the specific method to the satisfaction of the Level III NDE Examiner & Authorized Nuclear Inservice Inspector (ANII).
1.7.4.2 Examinations that detect flaws which require evaluation may be supplemented by other examination methods and techniques to determine the character of the flaw.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 9 of 30 1.7.5 Evaluation of Examination Results l.7.5. 1 The evaluation of nondestructive examination results shall be in accordance with Article IWA-3000 of Section XI. All reportable indications will be subject to comparison with previous data to aid in characterization and determination of origin.
Class 1 and Class 3 components containing relevant conditions will be considered acceptable for continued service providing an analytical evaluation performed demonstrates the component's acceptability and is subsequently examined in accordance with the requirements of IWB-3132.4, IWB-3142.4 and IWB-3144(b). Successive inspections for Class 1 shall comply with the requirements of IWB-2420(b) and (c). For Class 3, successive inspections shall comply with the requirements of IWC-2420(b) and (c).
Class 2 and High Energy Program components containing relevant conditions will be considered acceptable for continued service providing an evaluation performed demonstrates the component's acceptability and is subsequently examined in accordance with the requirements of IWC-3122.4, IWC-3132.3 and IWC-3134(b). For Class 2 and High Energy Program components, successive inspections shall comply with the requirements of IWC-2420(b) and (c) .
Class 1, Class 2, Class 3 and High Energy Program Supports containing relevant conditions will be considered acceptable for continued service providing an evaluation or test is performed that demonstrates the component's acceptability and is subsequently examined in accordance with the requirements of IWF-3122.4 and IWC-3134(b).
Successive inspections on these supports shall comply with the requirements of IWF-2420(b) and (c) .
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 10 of 30 1.8 Repair Requirements Repairs shall be performed to the requirements specified within Supplement 2 to Appendix B, The Repair, Replacement & Modification (RR&M) Program.
This RR&M Program shall apply to Class 1, 2 & 3 pressure boundary piping, components & supports; High Energy Pressure Boundary piping, components &
supports; Snubbers & identified Seismic Category I supports.
1.8.1 Performance of Repair Repairs shall be performed in accordance with the requirements of IWA/B/C/D/F-4000 of ASME Section XI, 1986 Edition or later Edition/Addenda that, is approved via 10CFR50.55a. Alternatively, repairs may be performed either to the requirements of the original Construction Code, which the component or system was fabricated to, or to later approved editions of the Construction Code, or later approved editions of ASME Section III along with any Code Cases approved via Regulatory Guide 1.147.
1.8.2 Examination/Test for Repairs 1.8.2. 1 Applicable examination requirements of the Construction Code shall be met. If the repair is performed on an existing weld that requires the complete removal of the original weld metal within that joint before rewelding, the following additional NDE requirements are required:
(a) Pressure retaining components greater than 2 inches in diameter shall require both surface and 1004 volumetric examinations to be performed on the new weld.
Or (b) Pressure retaining components 2 inches or less in diameter shall require a surface examination to be performed on the new weld.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL GINNA STATION INSERVICE INSPECTION PROGRAM FOR THE 1990-1999 INTERVAL PAGE ll of 30 "Examinations and testing of snubbers and supports that have been repaired shall be performed in accordance with 5.0 below and Section 9 of this program of this program.
The examination shall include the original examination method that detected the flaw, if applicable.
1.8.2.3 Applicable examination requirements of ASME section XI, 1986 Edition shall also be met to serve as a new PSI baseline for future ISI examinations, unless the examination performed under paragraph 1.8.2.1 was conducted under conditions and with equipment and techniques equivalent to those required by Section XI.
1.8.2.4 A hydrostatic test shall be performed in accordance with ASME Section XI, 1986 Edition, unless specifically exempted by Section XI Article IWA-4400, or the Alternate Rules of Incorporated Code Cases or Relief Requests, as applicable..
1..8.3 Surface flaws in Class 1, 2, or 3, bolts, studs, nuts and ligaments may be removed by mechanical means provided the removal of that flaw does not alter the, basic configuration of the item. Bolts, studs, and nuts that have flaws that cannot be removed by mechanical means shall be replaced or reported for evaluation as indicated by Section XI, Article IWA-3l00.
1.9 Replacement Requirements Replacement requirements, which includes modifications, are applicable to Class 1, 2 and 3 and high energy pressure retaining components and piping systems and their supports, unless specifically exempted by ASME Section XI, 1986 Edition, Article IWA-7400.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 12 Qf 30 1.9.1 Replacement Performance Replacements shall meet the requirements of ASME Section XI, 1986 Edition or a later Edition/Addenda approved via 10CFR50.55a. In addition, replacement items shall meet the requirements of the original Construction Code to which the original part was constructed. The R.
E. Ginna Nuclear Power Plant Replacement Program is defined within Supplement 2 to Appendix B.
1~ 9.1.1 A "Like for Like" replacement is an item that meets the original requirements of the design and procurement documents for the item being replaced and does not require reconciliation, reanalysis or changes to the item's design and technical requirements.
1.9. 1.2 An approved equivalent is a replacement that will result in a design or technical change to the original requirements based on reconciliation, re-analysis and/or testing per Paragraph 1.9.2.1.
1.9.2 Alternatively items used for replacement may meet all or portions of the requirements later additions of the Construction Code or Section III, when the construction code was not Section III. In order to use this alternate approach, the following additional requirements apply.
1.9.2. 1 Reconcile the requirements affecting the design, fabrication and examination of the replacement with the Design Analysis or Design Criteria or other methods of analysis that demonstrates the item is satisfactory for the specified design and operating conditions.
1.9.2.2 Mechanical interfaces, fits and tolerances that provide satisfactory performance are compatible with the system and component requirements.
1.9.2.3 Materials used are compatible with installation and system requirements.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA,NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 13 cf 30 1.9.2. 4 If item a replacement is because of a failure of the being replaced, a design evaluation or analysis shall consider the cause of the failure and its impact on other similar items, and the necessary actions to be taken to preclude recurrence.
1.9.2.5 When welding is to be performed as part of the replacement, the rules of Section IX shall be followed to satisfy the requirements of IWA-7320 and IWF-7000.
1.9.2.6 Items identified in IWA-7400 will be exempt from the requirements of the replacement program.
1.9.3 Examination/Test 1.9.3. 1 The replacement item shall be examined and pressure tested in accordance with the Construction Code or later Code provided the requirements of 1.9.1 above.
it meets Snubbers shall be examined and tested in accordance with Section 9 of this program.
1.9.3.2 Where the attachment of the non-pressure-retaining item is welded to a pressure boundary, the weld shall be examined in accordance with the requirements or 1.8.2.
1.9.3.3 Applicable examination requirements of ASME Section XI, 1986 Edition, shall also be met to serve as a new PSI baseline for future ISI examinations, unless the examination performed under 1.9.3.1 was conducted under conditions and with equipment and techniques equivalent to those required by ASME Section XI.
1.9.3.4 Replacements installed by mechanical methods shall be pressure tested at nominal operating pressure, or for Class 1 systems, the pressure associated with 100> rated reactor power.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION, FOR THE 1990-1999 INTERVAL 14 of 30 1.9.4 Reports and Records Reports and records to the extent required by the construction code and IWA 7520, as applicable for the replacement, shall be completed for all replacements.
- l. 10 System Pressure Testing 1.10.1 General Requirements 1.10.1.1 Pressure testing shall be conducted on all Class 1, 2, and 3 pressure retaining components in accordance with the requirements of Section XI Articles IWA-5000, IWB-5000, IWC-5000, IWD-5000; and Section XI Table IWB-2500-l Examination Category B-P, Table IWC-2500-1 Examination Category C-H, and Table IWD-2500-1 Examination Category D-A, D-B and D-C.
In addition, "High Energy" Main Steam and Feedwater Piping shall also be pressure tested on these non-class systems in accordance with the rules of IWC-5000.
1.10.1.2 Pressure tests are conducted from normal operating pressure, to a pressure up to 254 over design pressure. The degree of pressurization and the test. boundary depends upon the type of pressure test being performed. A visual examination (VT-2 method) is performed in concert with the pressure test on pressure retaining components under test pressure. Specific exceptions from achieving Section XI requirements are detailed in the Relief Request Section of this document.
- 1. 10. 2 Type of Pressure Tests The various types of pressure tests which are required during the inspection interval are described in the following:
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE XNSPECTXON PROGRAM PAGE GXNNA STATION FOR THE 1990-1999 INTERVAL 15 af 30 1.10.2.1 LEAKAGE PRESSURE TEST This test is performed subsequent to refueling outages. The boundary subject to test pressurization and the associated VT-2 examination during a leakage pressure test will extend to the pressure retaining components within the system boundary containing pressurized reactor coolant under the plant mode of normal reactor startup.
1.10.2.2 'FUNCTIONAL PRESSURE TEST This test is performed once each inspection period. The boundary subject to test pressurization and the associated VT-2 examination during a system functional pressure test will include only those pressure retaining components within the system boundary pressurized under the test. mode required during the performance or a periodic system/component surveillance test.
1.10.2.3 INSERVICE PRESSURE TEST This test is performed once each inspection period. The boundary subject to a test pressurization and the associated VT-2 examination during a system inservice pressure test will include only those pressure retaining components under operating pressure during normal system service.
l.l0.2.4 HYDROSTATIC PRESSURE TEST This test is performed once each inspection interval. The boundary subject to test pressurization and the associated VT-2 examination during a system hydrostatic 'pressure test includes all Class 1, 2, and 3 components and piping.
1.10.2.5 PNEUMATIC PRESSURE TEST This test is limited to Class 2 and 3 systems.
The boundary limits subject to test pressurization and the associated VT-2 examination during a system pneumatic pressure test are the same as a hydrostatic pressure test.
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QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 16 of 30 1.10.2.6 REPAIR/REPLACEMENT PRESSURE TESTS The boundary subject to test pressurization and the associated VT-2 examination is limited to the portion repaired or replaced, within the Class 1, 2 or 3 boundary. The specific type of pressure test is either a hydrostatic, pneumatic or a test at operating pressure such as the Leakage, Inservice or Functional. The specific type of test to be performed and the exemptions which apply to repair/replacement pressure testing are described in Supplement 2 to Appendix B, the Repair, Replacement & Modification Program.
1.10.3 Examination Requirements 1.10.3.1 During the conduct of pressure tests, certified VT-2 examination personnel, using RG&E approved NDE VT-2 examination procedure and the associated recording form, will examine the portions of piping under pressurization. The examination and test boundaries are depicted on controlled color-coded P&ID's. In some cases, the test boundary extends beyond the examination boundary due to valve location and/or check valve flow direction. In general, personnel will examine for evidence of leakage, inoperative leakage collection systems and evidence of corrosion.
1.10.3.2 Insulation removal during the VT-2 examination is not required, however, in accordance with IWA-5242 (a) "systems borated for the purpose of controlling reactivity" shall have insulation removed at bolted connections during conduct of the VT-2 examination. This requirement is only applicable to those VT-2 examinations performed during a hydrostatic pressure test,'since Leakage, Functional and Inservice tests are intended to be non-intrusive type tests. At Ginna, this requirement is considered to be applicable to borated lines only in the primary flow path of piping from the boric acid supply and CVCS Charging to the Reactor Vessel and return through CVCS Letdown, and is not applicable to branch lines connected to the primary flow path.
~ -e TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 17 of 30 1.10.4 Test Requirements 1.10.4.1 GENERAL The contained fluid in the system or fluid added to the system shall serve as the pressurizing or test medium. In steam systems either water or air may be used. Where air is used, the test procedures shall permit the detection and location of through wall leakages in components of the system tested. The temperature of the test medium will be that of the available source unless otherwise specified by the implementing test procedure/document. The test medium will be of a quality which is equal to or better than the system operating medium.
1.10.4.1.1 During conduct of hydrostatic tests, all entrained air will be vented from the system except in the following cases:
a~ Atmospheric Storage Tanks
- b. 0-15 psi Storage Tanks c ~ Class 1 systems where a mixture of steam, water and noncondensible gases are present. in a proportion typical of normal startup conditions.
- d. Normal Steam Systems 1.10.4.1.2 For. Leakage, Functional, Xnservice and Hydrostatic tests, the level of system pressure and temperature indicated or recorded by normal operating system instrumentation, or alternatively by test instrumentation is acceptable. For hydrostatic tests, the instrument requirements of IWA-5260 and applicable Code Case N-437 must be met.
1.10.4.2 LEAKAGE PRESSURE TEST REQUIREMENTS The system leakage pressure test shall be conducted at a pressure not less than nominal operating pressure associated with 100% Rated Reactor Power.
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 18 of 30 h
,The pressure and temperature will be attained at a rate in accordance with the heat-up limitations specified in the Ginna Technical Specifications for the component/piping system being tested.
- 1. 10.4.3 FUNCTIONAL PRESSURE TEST REQUIREMENTS The system functional pressure test is normally performed during performance or a Periodic Surveillance Test. The test is used to establish the test conditions associated with normal system operating pressure and temperature for performance of the system functional pressure test.
1.10.4.4 INSERVICE PRESSURE TEST REQUIREMENTS The operating pressure and temperature during normal system operation is used during performance of this test.
l.l0.4.5 HYDROSTATIC PRESSURE TEST REQUIREMENTS 1.10.4.5.1 General Code Case N-498 stipulations may be employed on Class 1 and 2 Required Systems in lieu of the 10 year Hydrostatic Test. Hydrostatic test pressure requirements vary among each of the classifications. The minimum test pressure will be maintained for the entire duration of the test, with the exception of tests performed at temperatures greater than 200'F, where the examination phase may be performed at a lower pressure corresponding to 200'F. The test pressure shall not exceed the maximum allowable test pressure of any component within the test boundary.
1.10.4.5.2 The hydrostatic test pressure, including static head, will not exceed 1064 of the specified te t pressure for the system anywhere within the test boundary.
- 1. 10. 4. 5. 2. 1 In cases where the highest and lowest elevations cannot be isolated, and the test pressure including static head, would exceed 106> of the specified test pressure at the lowest point, the test pressure will be reduced.
TITLE REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 19 of 30 1.10.4.5.2.2 In cases where the high and low elevations result in unreasonably small differences between the test pressure including static head, and 106% of the specified test pressure, the test pressure will be reduced to accomodate the precision of the test instrument.
1.10.4.5.2.3 The test pressure, as identified in sections 1.10.4.5.2.1 and 1.10.4.5.2.2, will be adjusted such that the test pressure, including static head, does not exceed 106> of the specified test pressure at the lowest point. The resulting pressure at the highest point will be considered acceptable.
- 1. 10. 4. 5. 3 The hydrostatic test boundary will end at the transition between system piping and instrumentation tubing shown on P&ID drawings.
When this transition does not occur at an isolation valve, the boundary will be extended to the first isolation valve after the transition.
- l. 10. 4. 6 The following sections list the hydrostatic test pressure requirements which will be met for each Code classification:
1.10.4.6.1 Class 1 The system hydrostatic pressure test is conducted at the pressure calculated from the following table based on the test temperature:
Test Tem erature De . F. Test Pressure 100 or less l. 10 Po 200 1- 08 Po 300 1. 06 Po 400 1.04 Po 500 or greater 1.02 Po "Po" is the nominal operating pressure cor'responding with 1004 rated reactor power.
Linear interpolation will be used at intermediate test temperatures. Technical Specification heat-up/cool-down limits will be observed.
N TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 20 01 30
- 1. 10.4. 6.2 Class 2, 3, and High Energy:
a0 The test pressure will be at least 1.10 times the system pressure for systems with a design temperature of 200 F or less, and at least 1.25 times the system pressure for systems with a design temperature above 200'F. The system pressure will be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure will be substituted for the system pressure.
- b. In the case of atmospheric storage tanks, the nominal hydrostatic pressure developed with the tank filled to its design capacity will acceptable as the test pressure.
'e c ~ For 0 to 15 psi storage tanks, the test.
pressure will be 1.1 times the design pressure of vapor or gas space above liquid level for which overpressure protection is provided by the relief valves. If relief valves are not "installed, the test pressure will be equal to 1.1 times the normal operating pressure.
- d. For the purpose of the test, open-ended portions of a suction or drain line from a storage tank extending to the first shutoff valve are considered as an extension of the storage tank.
- e. For open ended portions of discharge lines beyond the last shutoff valve in nonclosed systems, a test that demonstrates unimpaired flow will be performed in lieu of a system hydrostatic pressure test. Unimpaired flow for Class 2 is defined as an "open flow path" and for Class 3 as "adequate flow during system operation".
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 21 of 30 1.10.5 Test Implementation 1.10.5.1 All pressure testing is implemented using both the VT-2 examination procedure and the specific test procedure for the type of test and portion of system being tested.
- 1. 10.5.2 Applicable Required System Pressure Test Boundaries shall be confirmed prior to examination performance.
1.10.6 Scheduling When using the inspection plan, it should that during a period or refueling outage in which be noted a hydrostatic test is performed on a system or portion(s) of a system, the leakage test (Functional or Inservice) required for that period on the same system or portion of the system, may be deleted from that period or outage.
The hydrostatic test will satisfy the requirements for that leakage test.
2.0 CLASS 1 PROGRAM PLAN 2.1 Basis for Preparation 2.1.1 Preparation of the Class 1 ISI program plan was based on the requirements of Articles IWB-1000 and IWB-2000 of Section XI. These articles provide rules and guidelines for exemptions, inspection schedules, and examination requirements for Class 1 pressure retaining components and their integral attachments.
2.1 ~ 2 As allowed by 10CFR50.55a 2(ii), Class 1 Category B-J weld selection is based upon ASME Section XI
'74/Summer '75 Code and does not utilize stress level criteria.
2.2 Components Subject to Examination 2.2.1 Based on the requirements of Section XI, Class 1 nonexempt pressure-retaining components and their integral attachments will be subject to examination during the third inspection interval.
TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 22 of 30 2.3 Extent and Frequency of Examinations 2.3.1 Class 1 components, as listed- in Section 4, Table 1 shall be examined to the extent and frequency required in Table IWB-2500-1 and Figures IWB 2500 through IWB 2500-20 of Section XI.
2.4 Exemption Criteria 2.4. 1 In accordance with IWB-l220, certain Class 1 components are exempt from examination. The following criteria were applied to exempt components from surface and volumetric examinations in accordance with Section XI Exem tion Criteria Code Reference Piping or 1 inch nominal pipe IWB-1220(b) (1) size (NPS) and smaller, except for steam generator tubing Components and their connections ~
IWB-l220(b) (2) in piping of 1 inch NPS and smaller 2.5 Examination of Reactor Coolant Pump Flywheels The Reactor Coolant Pump Flywheels shall be examined as specified in Section 11 of this program. These examinations shall be scheduled in the Class 1 section of the ISI program plan.
2.6 Inservice Inspection Program Plan (Supplement 1 to Appendix B)
This plan provides the examination requirements for Class 1 components per the 1986 Edition of Section XI. These requirements shall be satisfied during the third inspection interval. The plan also shows the results of examinations performed in the previous two intervals.
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REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 23 of 30 A detailed description of the contents of the Class 1 Examination Plan can be found in the "Introduction". This immediately precedes the tables and the isometric and component drawings, in the plan. From this plan, an examination schedule is extracted for implementation of the examinations for each outage of the third interval.
3.0 CLASS 2 PROGRAM PLAN 3.1 Basis for Preparation 3.1.1 Preparation of the Class 2 ISI program plan is based on the requirements of Articles IWC-1000 and IWC-2000 of Section XI. These articles provide rules and guidelines for exemptions, inspection schedule, and examination requirements for Class 2 pressure retaining components and their integral attachments.
3.2 Components Subject to Examination 3.2.1 Based on the requirements of Section XI, Class 2, nonexempt pressure-retaining components and their integral attachments will be subject to examination during the third inspection interval.
3.3 Extent and Frequency of Examinations 3.3.1 Class 2 components, as listed in Section 4, Table 2, shall be examined to the extent and frequency required in Table IWC 2500-1 and Figures IWC 2500-1 through IWC 2500-13 of Section XI.
3.4 Exemption Criteria IWC-1220 of Section XI provides the exemption criteria for Class 2 components. The following criteria were used to exempt Class 2 components from surface and volumetric examinations in accordance with IWC-1220.
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 24 af 30 The following components (or parts ok components) of Residual Heat Removal (RHR), Emergency Core Cooling (ECC), and Containment Heat Removal (CHR),
systems (or portions of systems) are exempt from the volumetric and surface examination requirements of IWC-2500:
(a) vessels, piping, pumps, valves and other components 4 inches NPS and smaller in all systems except in high pressure safety injection systems of pressurized water reactor plants; (b) vessels, piping, pumps, valves, and other components l-l/2 inches NPS and smaller in high pressure safety injection systems of pressurized water reactor plants; (c) component connections 4 inches NPS and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size in all systems except in high pressure safety injection systems of pressurized water reactor plants; (d) component connections 1-1/2 inches NPS and smaller ( including nozzles, socket fittings and other connections) in vessels, piping, pumps, valves and other components of any size in high pressure safety injection systems of pressurized water reactor plants; (e) vessels, piping, pumps, valves, other components, and component connections of any size in statically pressurized, passive (i.e., no pumps) safety injection systems of pressurized water reactor plants; and (f) piping and other components of any size beyond the last shutoff valve in open- ended portions of systems that do not contain water during normal plant operating conditions.
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 25 of 30 l 3.4.2 The following components (or parts of components) of systems (or portions of systems) other than RHR, ECC and CHR systems are exempt from the volumetric and surface examination requirements of IWC-2500:
(a) vessels, piping, pumps, valves and other components 4 inches NPS and smaller; (b) component connections 4 inches NPS and smaller (including nozzles, socket fittings and other connections) in vessels, piping, pumps, valves and other components of any size; (c) vessels, piping, pumps, valves, other components and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200'F; and (d) piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions.
3.5 Inservice Inspection Program Plan (Supplement 1 to Appendix B)
This plan provides the examination requirements for Class 2 components per ASME Section 11, 1986 Edition, no Addenda. These requirements shall be satisfied during the third inspection interval.
The plan also shows the results of examinations performed during the previous two intervals.
A detailed description of the contents of the Class 2 Examination Plan can be found in the "Introduction". This immediately precedes the tables and the isometric and components drawings in the plan. From the plan an examination schedule is extracted for implementation of the examinations for each outage of the third interval.
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 26 af 30 4.0 CLASS 3 PROGRAM 4.1 Basis for Preparation 4.1.1 Preparation of the Class 3 ISI program was based on the requirements of Articles IWD-1000 and IWD-2000 of Section XI.. These articles provide rules and guidelines for exemptions, inspection schedules, and examination requirements for Class 3 pressure retaining components and their integral attachments.
4.2 Components Subject to Examination 4.2. 1 Based on the requirements of Section XI, Class 3 nonexempt pressure-retaining components integral attachments will be subject to examination during the third inspection interval:
4.2.1. 1 Other Class 3 systems are not subject to the examination or System Pressure Testing requirements of ASME Section XI because they do not meet the system function requirements of Examination Categories D-A, D-B and D-C, where:
D-A Systems in support of Reactor Shutdown Function D-B Systems in support of Emergency Core Cooling, Containment Heat Removal, Atmosphere Clean-up and Reactor Residual Heat Removal.
D-C Systems in support of Residual Heat Removal From Spent Fuel Storage Pool.
4.3 Extent and Frequency of Examinations 4.3.1 Class 3 components, as listed in Table 3 shall be examined to the extent and frequency required in Table IWD 2500-1 and Figure IWD 2500-1 of Section XI.
4.3.2 Integrally welded attachments shall be examined, utilizing the VT-3 method, once during the Interval on all Class 3 Component Supports. The associated Class 3 Categories have been grouped since the methods of examination and their Requirements are identical.
TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 27 of 30 4 4 Exemption Criteria 4.4.1 In accordance with IWD-l220, certain Class 3 components are exempt from. examination. The following exemption criteria was applied to Class 3 systems as specified in IWD-1220:
Section XI Exem tion Criteria Reference Integral attachments of IWD-1220.1 supports and restraints to components that are 4 inches NPS and smaller within the system boundaries of Examination Categories D-A, D-B, and D-C shall be exempt from the VT-3 examination, except for Auxiliary Feedwater. Exemption for the Auxiliary Feedwater System is 1 inch and less.
Integral attachments of IWD-1220.2 supports and restraints to components exceeding 4" nominal pipe size may be exempted from the visual examination VT-3 of Table IWD-2500-1 provided:
C (a) the components are located in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling; and (b) The components operate at a pressure of 275 psig or less and at a temperature or 200 degrees F or less.
JC TITLE REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 28 of 30 4.5 Inservice Inspection Program Plan (Supplement 1 to Appendix B)
This plan provides the examination requirements for Class 3 IWD components per the 1986 Edition of Section XI. These requirements shall be satisfied during the third inspection interval. The plan also shows the results of examinations performed during the previous two intervals.
A detailed description of the contents of the Class 3 Examination Plan can be found in the "Introduction". This immediately precedes the tables and the isometric and component drawings in the plan. From the plan an examination schedule is extracted for implementation of the examinations for each outage of the third interval.
5.0 CLASS 1~ CLASS 2q AND CLASS 3 COMPONENT SUPPORTS 5.1 Basis for Preparation 5.1.1 Preparation of the component support, ISI Program was based on the requirements of Articles IWF-1000 and IWF-2000 of Section XI. These articles provide rules and guidelines for exemptions, inspection schedules, and examination requirements for Class 1, Class 2, and Class 3 component supports. Inservice test requirements and VT-3 inspection requirements for snubbers shall be conducted in accordance with Section 9 of this program which implements the requirements of Article IWF-2000.
5.2 Component Supports Subject to Examination 5.2.1 Based on the requirements of Section XI, nonexempt component supports for the Class 1, Class 2 and Class 3 systems identified in Sections 2.2, 3.2 and 4.2 of this plan will be subject to examination during the third inspection interval.
The component supports requiring examination shall be as follows:
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TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 29 of 30 5.2.1.1 Plate and Shell-Type Supports Supports which are fabricated from plate and shell elements, such as vessel skirts and saddles, and are normally subjected to a biaxial stress.
5.2.1.2 Linear-'ype Supports Supports acting under essentially a single component or direct stress. Such elements may also be subjected to shear stress. Examples of such structural elements are: tension and compression stxuts; beams and columns subjected to bending; trusses; frames; arches; rings; and cables.
5.2.1.3 Component Standard Supports A support assembly consisting of one or more generally mass-produced units usually referred to as catalog items. Examples of such items are shown in Figure IWF-1210-1 of Section XI.
5.3 Extent and Frequency of Examination 5.3. 1 'Component supports selected for examination shall be those components required to be examined under the requirements of 2.3, 3.3 and 4.3, IWF-2500.
The inservice test requirements of Article IWF-5000 shall be satisfied by the requirements of Section 9 of this program.
5.3.1. 1 Class 3 component supports containing integrally welded attachments shall be examined, utilizing the VT-3 method, once during the Interval. The associated IWF categories have been grouped since the methods of examination and associated Requirements are identical.
5.3.2 In addition, snubbers shall be functionally tested at the frequency required by Section 9 of this program.
TITLE: REV.
QUALITY APPENDIX B Section 1 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 30 of 30 5.3.3 On piping systems where a piping seismic analysis boundary is beyond the safety class boundary, the component supports within the portion between those two boundaries shall be examined. The extent of these examinations shall be consistent with the examination requirements of that safety class system.
5.3.4 High Energy Piping Component Supports shall be examined to IWC-2500 Category C-C and to IWF-2500 in an Augmented Program.
5.4 Exemptions 5.4.1 ASME Section XI, 1986 Edition, no Addenda, does not contain defined exemption criteria for component supports.
5.4.2 Exemption criteria specified in IWB, IWC and IWD have been used.
5.5 Inservice Inspection Program Plan (Supplement 1 to Appendix B)
This plan provides that Class 1, Class 2 and Class 3 component supports shall be incorporated into the appropriate Class 1, Class 2 or Class 3 section of the examination plan. These requirements shall be satisfied during the third inspection interval. The plan also shows the results of examinations performed in the previous two intervals.
REV PAGE QUALITY ASSURANCE MANUAL Section 2 2 1 of 27 GINNA STATION EFFECTIVE DATE: December 31, 1992 ROCHESTER GAS 5 ELECTRIC CORPORATION SIGNATURE DATE TITLE: PREPARED BY rp-4-gg APPENDIX B QUAUTY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM REVIEW: lZ (P')g FOR THE 1990-1999 INTERVAL ISI RELIEF REQUESTS APPROVED B 1.0
Introduction:
General:
In accordance with 10CFR50.55a(g) (5) (iv), Rochester Gas Sr Electric has requested relief from those ASME Section XI requirements that have been determined impractical for certain areas. This section identifies each active and owner withdrawn Relief Request submitted to the Nuclear Regulatory Commission for their consideration and acceptance.
Table 1 provides information in a summary format for both active and withdrawn relief requests applicable to R. E.
Ginna Nuclear Power Plant.
Following Table 1, detailed Relief Requests are listed.
These provide information on the component for which relief is requested, ASME requirements, proposed alternate method, and other pertinent information, as needed.
Existing active relief requests can be withdrawn by the owner at any time. Additional relief requests will be su)mitted to the Nuclear Regulatory Commission, as appropriate.
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION 'OR THE 1990-1999 INTERVAL 2 of 27 ISI RELIEF RE UESTS TABIE 1
SUMMARY
OP RKIIFP Rdief ASME Requirement Reason for Proposed Request Sccdon XI lor Whkh Reliel Relief hkemsle Exandnathn
~n Number Reference Component is Requested Request 1525001 RPV Shell to FLange Volumeuk To perform all exanunstions Pafonn all auunhsthns assochted Cst. B.A WeM durhg two different assochted with the Shell-to wkh the Shell.to.Fhnge at or near pcnods Bangs during the sane . the end of the IntervaL pakxL IWB-2500.1 Cat. BD RPV Nozzle to Vessd WcMs Vohmeuie Examinadon durhg two different To paform all examhstions assochtcd wkh the Horde-to.
Perform all ~ns assoehtcd wkh the Nozzk.to Vessel weMs at or pc nods Vasel wcMs durhg the same near the end ol the htavaL pakxL 3 (') IWA.LC00 Authorhcd Inspecthn Use of huthorhed Agency'cw York State has Use L E. Ginna Ctuality Assurance Inspecdon Agency. not endorsed ASM E Codes Pmgram.
and docs not have an Authorized Inspecdon IWB-2500 Cat. BI I I Reactor Coohnc Pump Case Wekb and Volumcuie ~n case wekb and visual of of Agency.
Pump materhl and conliguradon.
Ifydrostatk test, surface and visual exams ol outside surfaces.
B.L2 Iiuccvals. Internsb.
IWB 2500-1 Chss I Valves Visual Exsmhstion of Excessive rsdhtion Exanune valve lnternsb when Cst. B.M.2 G recta than NPS 4 valve Intemab. expcuure and hbtorkal dbsssembkd lor mahtcnance.
rcihbiTay of valves.
IWD 2500.1 Rsdisx&e Waste Vbusl Examhsthn of Tank wBI be rcndacd Perform vbual cxsmlnsdon each Cat. DB HoMup Tank hydrostatk pressure. inoperative duiing tests. perkd at normal operathg prcssure.
IWC 5222(a) Vbusl Examinsuon at Pumps have maximum Pcrlonn Hydro Test and visual at 3C20 prig Hydmstadc prcssure hmk on the scab. 240 psig.
Pressure.
IWC.5222(a) Valves PCV 430 Visual Examination at Valve dhphragms cannoc Hddroscsde Test to Bex Connecthn and PCV 43 IG Hydosttk Prcssure. Test Pressure. operate dhphrsgm per valve test requirements and perform fnsecvke visual cxamhsdon once per perkxL IWG5222(a) Seconday Side Visual Exanunldon at RGB E adopted pressure Hydrostatic Test at ol stcam Hydrostatic Prcssure. diffcrenthl Bmitation of 800 1.10 dmes instead of generator and psig to prevent primary 1.25 Psv sctdng and assochtcd msh side tube sheec chdding perform visual steam piping. scpamthn. examination.
10 (') IWD.5223(a) Standby AuxBhry Visual Examhstion Pressure Rcducdon Bow Perform Inservke Pump RcciicuL at Hydrostatic Pressure. OriTice requires removal and Visual Exsminsuon Une fmm AOV bhnkofL System piphg does once pcf pcilod.
0710A, AOV 97 100 noc pmvMe bolauon to and their assocl. condensate supply tank ated downstream SigniTicant tank reduction Ihw oriyices. wouM be required for removal and b considacd hnpractieaL
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J TITLE:
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 3 of 27 ISI RELIEF RE UESTS TABLE I (Coul)
Relief ASME Rcquucment Reason for Proposed Request Section XI for Whch Relief Relief Ah crusts Nufsber Component Rcfercnce h Requested Request Examination IWD-5223(s) Bori ACM Filter and Vhual Eanunadon at Test Pressure rcqidfcd wiB Perform Inscfvke Visual Examhadon assoriated piphg llydtuststk Pressure. exceed Bffdts for safe workhg oflcc pcf pcnod between Valves 347, pressure on Bork AcM Fiber 348A and 349A. Houshg Fhnge Ca&etc.
12.1 IWD 5223(a) Air Qart for Visual Examinsdon at Air Qart Pressure Test wouM Perform Inservke Examination Diesel Generator Hydrostatk Pressure. require tefminathn prior to once per period and once each hclebng reschhg engine skM to quarter a pressure decay test Receiver Tanks preclude air to air start performed on air receiver.
and Piping. motor, leaving portion of piping untestable.
12.2 IWD-5223(s) Fuel Oil Transfer Visual Examination at Rcquhes hohthn of Deisel Perform system Functional Tcsdng Pumps and associated Hydrostatic Pressure. Oil Qorage and Dry Tank with assochtel Visual piphg to tennlnsh at i where no means of hohdon Examinathn once per period.
Oil Qorage Tank h pmvMcd st Transfer pump discharge piping with tanks
~n vented to atmosphere.
IWD-5223(s) Jacket Cooling Visual at COCBllg Watcf Expoluhfl Perform System Functional Testing Water System and )lydfenatic Pressure. Tank vented to atmosphere whh assochtcd Visual Kxandnathn sssochted piping requiring Isohtion with vart oilcc pcf pcfhd, to tenninah at plphg Involved willbc unable Water Expansion to PfCSSUflXC Tanks.
)3 IWD-5222(a) NOO.ISI classified Visual Examim&n at Requirements shouM be based Test in accordance with IOCFRSO syrtcms penetrating Hydroststk Pressure. on the containment system Appendbt J with the safety function primary containment. design not the assochted the line performs in accordance whh process system design Technical Specifications, sufveglance fcqldfemcflL requirements, In ad Mon, perform Inservice Visual Exsfnination on exposed porthns.
IWD 5223(a) Chss 3 Portion of the Visual Xxanunsdon st Hdtostsatk Tcsdng h Perform lnservke Vismd Examination scfvkc wstcf syrtcm Hydrostatic Pressure. hnpractkal due to system once per pcnoiL design that h openetdcd and emphying bunerily valves that werc not dcslgilcd to provide a leaktight boundary.
15 (') IWC.2500 Chss 2 Piping Piping WeMs ( At a mhumum terminal Surface and Volumetric Examinatklns Cat. GF I WeMS ( wall thkkncss do 3/8'ominal connecthn wekh of MentiTied on terndnsl connccthn weMs ol Bl GF-2 wall thickness 3/8'ondnsl not require surfs<<e and excmptel wekh per Items MentBied exempted wekh shall be Iteus CS.IO for piping > NPS4. volumetric examinations. CS.IO and C5.50 of Tabk pcffofllicd to dlc requirements of and C5.50 resp. IWG2500 shouM be IWC.2500-1.
cxsmineL 16 (') IWB.2500 Chss I & 2 Integral Volumcnie or surface Integral Anschments on Chss On Chss I & 2 Integral Attachments, Cat. B K.I Item Anschment on piping examination on Chss I I & 2 support attachments s surface exanunsdon shaB be B10.10 and IWC. spenTicsBy support integral Anaclunents is shoUld have a surface pcrionncd on support anachments to 2500, Cst. GC, Attachments. required on Base <<xsmination to Insure safety IWB and IWC 2500 rcquhements.
item C320 Attachment ) 5/8'. Chss and system integrity.
2 rcqullcs a surface examinadon on Base Anachments ) 3/4 .
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TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 4 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO. 1 DEFER RPV EXAMINATIONS TO END OF INTERVAL Components for Which Relief is Requested:
The component for which relief is requested is the Reactor Pressure Vessel (RPV) Shell-to-Flange Weld.
ASME Requirement from Which Relief is Requested:
Table IWB-2500-1, Examination Category B-A, requires that the RPV Shell-to-Flange weld be examined during the first and third periods in conjunction with the nozzle examinations, with at least 50 percent examined during the first period and the remainder by the end of the third period. The required Shell-to-Flange examination is impractical it if performed during the periods specified as can only be accomplished from the flange surface.
III. Proposed Alternate Method:
During the first two inspection intervals, 100 percent of the accessible length of the RPV welds including the Shell-to-Flange weld were examined at or near the end of the interval when the entire examination could be performed from both the flange surface and the vessel wall. This is a more practical approach in that the required examinations from both surfaces can be performed at the same time. During the third interval, 100 percent of the accessible length of all RPV welds including the shell-to-flange weld will be performed at or near the end of the interval when all the required examinations can be performed at the same time.
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TITLE:
QUALITY APPENDIX B Section 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE XNSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 5 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO ~ 2 Components for Which Relief is Requested:
The components for which relief is requested are the RPV Nozzle-to-Vessel welds and Nozzle Inside Radius Sections.
ASME Requirements for Which Relief is Requested:
Table IWB-2500-1, Examination Category B-D, Item B3.90, Nozzle-to-Vessel welds allows partial deferral. "Xf examinations are conducted from inside the component and the nozzle weld is examined by straight beam ultrasonic method from the nozzle bore, the remaining examinations required to be conducted from the shell inside diameter may be performed at or near the end of each interval."
Examination Category B-D, Item 3.100 Nozzle Inside Radius Sections, does not allow deferral to the end of the interval, and requires (footnote 2) 25 percent to 50 percent of the nozzles to be examined during the first period, with the remainder to be examined at the end of the interval.
Examinations from the nozzle bore and nozzle inside radius examinations can only be performed on two (outlets) of the six major nozzles without removal of the core barrel. The mechanized examination of the two accessible nozzle and inside radius sections is quite expensive, and the nozzle-to-vessel examination is only a partial examination from the nozzle bore. From a technical position considering the progress which is being made in ultrasonic examination equipment and techniques and for the correlation of data obtained from the bore with that obtained from the shell, it is highly desirable to perform both examinations at the same time.
III. Proposed Alternate Method:
Rochester Gas 6 Electric (RGGE) proposes to perform both nozzle-to-vessel examinations (from the nozzle bore and from the shell inside diameter) at or near the end of the interval. The nozzle inside radius examinations will also be performed at this time. This more practical approach will allow all the required examinations to be performed at the same time on all the nozzles and nozzle inside radii.
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 6 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 3 I ~ Examination Requirement for Which Relief is Requested:
The ASME Boiler and Pressure Vessel Code,Section XI, 1986 Edition, IWA-l400(f), requires an arrangement with an Authorized Inspection Agency to provide inspection services. In addition, the Code requires that certain administrative functions be performed by the "Enforcement Authority" and "Authorized Nuclear Inservice Inspector".
Proposed Alternative:
Ginna Station is located in the state of New York. This state has not endorsed ASME Codes and therefore does not provide administrative organization and controls such as "Enforcement Authority", "Authorized Nuclear Inservice Inspector" and "Reporting Systems". However,'inna Station's Quality Assurance Program does provide equivalent administrative control. Therefore, RG&E requests that Ginna's Station Quality Assurance Program be used in lieu of Code administrative functions.
Rochester Gas & Electric's program for the inservice inspection, governed by the R. E. Ginna Station Quality Assurance Manual, contains the requirements and responsibilities for implementation of the program and procedures. The procedures have been prepared and approved by the responsible organizations within Rochester Gas & Electric (e.g., Ginna Station, Engineering, Materials Engineering and Inspection Services, Electric Meter and Laboratory and Purchasing).
Approved procedures will be implemented to control the standards for examination evaluation. These procedures include the identifications of the organization performing the inspection, description of the method of inspection to be used, acceptance and rejection criteria, and requirements for providing evidence of completion and certification of the inspection activity.
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I'
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION"PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 7 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO. 3 (Con~t)
In addition, procedures are developed by Ginna Station to prescribe the disposition of nonconformances. The procedures implemented for the repairs, the retest procedures and the test results will be reviewed by the Plant Operating Review Committee. The members of this committee include technically qualified staff personnel.
Examination techniques have been established in accordance with written requirements and incorporated into written procedures. qualifications for nondestructive test personnel are in compliance with Regulatory Guide 1.58, "Qualification of'uclear Power Plant Inspection, Examination and Testing Personnel."
Records and reports of the inservice inspection will be developed and maintained by Rochester Gas and Electric and include .such items as examination plans and schedules, examination of results and corrective actions.
The functions of the authorized nuclear inservice inspector, namely their review and verification of inservice examinations, personnel qualification and
,equipment certification during the annual outages at Ginna Station will be performed by personnel of the Hartford Steam Boiler Inspection and Insurance Company. The qualifications of the inspectors, inspections specialists and inspection agency are in compliance with the Code.
~~
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E- GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 8 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 4 Components for Which Relief is Requested:
Each of the 27.5 inch diameter recirculation loops at R.E.
Ginna has a Class 1 Reactor Coolant Pump. The function of these two pumps is to provide forced circulation through the RPV core during normal reactor operation.
Code Requirement for Which Relief is Requested:
Table IWB-2500-1, Examination Categories B-L-l and B-L-2 require volumetric examination of casing welds and visual examination of internal pressure boundary surfaces of one pump case in each of the pump groups performing similar system functions each inspection interval. These examinations are impractical for the reactor coolant pumps at Ginna Station and relief is, therefore, requested.
A. Supporting Information 1 ~ The two reactor coolant pumps (RCP) for R.E.
Ginna are Westinghouse Model 93 pumps. Each pump casing is fabricated by welding four stainless steel (5A351 CF8) castings together.
Thus, there are 3 circumferential pressure-retaining welds that are to be volumetrically inspected in accordance with Category B-L-l.
- 2. The unsuitability of ultrasonic examination was demonstrated during the "A" reactor coolant pump examination in 1980. An attempt was made to determine the wall thicknesses using ultrasonic examination, the casing welds must be inspected using the miniature linear accelerator (MINAC).
3 ~ Radiographic examination using the MINAC was performed on the R. E. Ginna "A" RCP during the Spring 1981 refueling outage. In addition, the same type of examination has been performed at several other sites.
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TITLE REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTXON PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 9 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 4 (Con~ t)
This examination was performed by placing the MINAC inside the pump casing and placing the film on the outside of the pump. To perform the examination, the pump was completely disassembled.
Disassembly to this extent is far beyond any disassembly expected for this examination.
Also, insulation on the casing exterior was removed for film placement.
Additionally, the pump bowl must be dry for installation of the MINAC. Therefore, all fuel assemblies were removed from the reactor vessel and the vessel water level lowered to below the nozzles. Complete disassembly of the pump was also required to conduct the VT-l examination in accordance with Category B-L-2.
4 ~ No problems have been found with the welds at R.
E. Ginna or other sites. Additionally, no problems have been found during the Category B-L-2 visual examination. The visual examination was conducted at R. E. Ginna by using the video camera on the MINAC.
The whole body exposure to personnel during the Spring 1981 directly attributable to the RCP "A" examinations 93,067 millirem. This does not include the dose received during the complete core unload to get the plant in condition for the RCP disassembly.
- 5. The nuclear industry has been successfully applying leak-before-break concepts to primary loop and Class 1 auxiliary piping systems of commercial nuclear power plants. Currently, the analyses supporting such concepts comes under the review of the Nuclear Regulatory Commission by General Design Criteria-4 (GDC-4).
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TfTLE:
QUALITY APPENDIX B Section 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 10 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 4 (Con~t)
There are eight different models of RCP's in Westinghouse-type PWRs. Model 93 methodology used in the analyses is consistent with that recommended in NUREG 1061, Vol. 3 and GDC-4. A finite element stress analysis model for the Model 93 pump was developed.
The RCP casings are cast stainless steel. The chemistries of each heat of material used in the pumps were used to determine the fracture toughness. The phenomenon of thermal aging was addressed.
The program successfully demonstrates that leak-before-break analyses are applicable to all primary pump casings of all Westinghouse design PWRs for which the screening loads are reasonably applicable and the fracture toughness are known.
- 6. We believe that performing a volumetric examination of the Ginna Station Unit 1 RCP casing welds and a visual examination of the interior pressure retaining surface of one pump during the third lO-year inspection period does not provide an increase in safety and expected radiation exposure. The following items have been considere'd:
a ~ Visual examination (VT-2) of the exterior of all pumps during the hydrostatic pressure test required by Table IWB 2500-1 Category B-P.
- b. Perform a Visual examination (VT-l) of the external surfaces of the welds of one pump casing.
c ~ Perform a visual examination (VT-3) of the internal surfaces each time pump disassembly is required for maintenance.
~ J TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL GXNNA STATION XNSERVICE XNSPECTION PROGRAM FOR THE 1990-1999 XNTERVAL ISI RELIEF RE UESTS PAGE ll Of 27 RELIEF REQUEST NO. 4 (Con~t)
- d. Perform an evaluation to demonstrate the safety and serviceability of the pump casing. The evaluation will include:
Establishing material properties including fracture toughness values.
(ii) Performing a stress analysis of the structure.
(iii) Reviewing of the operating history of the structure.
(iv) Selection of locations for postulating flaws.
(v) Determination of a flaw size resulting in the detectable leak rate (vi) Establishing the stability of the selected flaw.
(vii) Demonstration that a postulated through-wall flaw which yields detectable leakage remains stable for all design loadings, with a margin of 2 on flaw size.
NOTE: Xn making this assessment, thermal aging embrittlement and any other processes which may degrade the properties of the pump casing during service will be considered.
1 4
TITLE:
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GXNNA STATXON FOR THE 1990-1999 INTERVAL 12 of 27 XSI RELIEF RE UESTS RELIEF REQUEST NO 5 Components for Which Relief is Requested:
Class 1 valves requiring valve body internal VT examination.
Size
~In. Valve No. M~FG T~e Line No.
10 842A Darling/Check 10A-SI2-1502-A 10 842B Darling/Check 10A-SI2-2501-B 10 867A Darling/Check 10A-SI2-2501-A 10 867B Darling/Check 10A-SI2-2501-B 10 700 Velan/Gate 10A-RC02501-A 10 701" Velan/Gate 10A-RCO-2501-A 10 720 Velan/Gate 10A-RCO-2501-B 10 721 Velan/Gate 10A-RCO-2501-B 853A Velan/Check 6A-RC-2501-A 853B Velan/Check 6A-RC-2501-B 852A Velan/Gate 6A-RC-2501-A 852B Velan/Gate 6A-RC-2501-B ASME Requirement for Which Relief is Requested Table IWB-2500-1, Examination Category B-M-2, requires an internal VT-3 examination on at least one valve within each group of valves that are of the same size, constructional design (such as globe, gate or check valves) and manufacturing method, that perform similar functions in the system. This relief request is based on the following points:
- 1. to complete the subject examination, unnecessary expenditures of man-hours and manrem are required with essentially no compensating increase in plant safety, and
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~ 4
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 13 of 27 ISI RELIEF REQUESTS RELIEF REQUEST NO. 5 (Conlt) 2 ~ the structural integrity afforded by valve casing material utilized will not significantly degrade over the lifetime of the valve.
Based on data compiled from a plant similar in age and design to Ginna Station, it is expected that approximately 100 manhours and 5 manrem exposure would be required to disassemble, inspect, and reassemble these vales.
Performing this visual examination under such adverse conditions, high doses rate (30-40 R/hr), and poor as-cast surface conditions, realistically provides little additional information as to the valve s casing integrity.
The valves material, a high-strength cast stainless steel (ASTM A35l-CF8), is widely used in the nuclear industry and has performed extremely well. The presence of some delta ferrite (typically 54 or more) substantially increases resistance to intergranular stress corrosion cracking. The delta ferrite also helps the material to resist pitting corrosion in chloride containing environments.
RG&E feels that, adequate safety margins are inherent in the basic valve design and that the public's health and safety will not be adversely affected by not performing a visual examination of the valve internal pressure boundary surfaces. Additionally, this visual examination adds little or no value to the overall safety of the plant and subjects plant personnel to unnecessary radiation exposure. Therefore, a request for relief from this requirement is sought.
III. Proposed Alternative Method:
As stated above, RG&E does not believe that the visual examination required each ten-year interval is warranted.
However, as standard maintenance practice dictates, when these valves are disassembled for maintenance purposes, a visual examination of the internals and internal pressure boundary surfaces will be performed, to the extent practical.
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- ~ I 4
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 14 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO 6 Component for Which Relief is Requested:
The radioactive waste hold-up tank in the waste disposal system provides a means of storing contaminated water that has been used in the operation of the nuclear power plant.
The waste disposal system and waste hold-up tank may be required to function in all modes of reactor operation including cold shutdown and refueling.
ASME Requirement for Which Relief is Requested:
Table IWD 2500-1, Examination Category D-B, Item No.
D2.10, requires VT-2 examination of the waste hold-up tank at hydrostatic testing levels (at least 1.10 system pressure) during each interval as well as VT-2 examinations at nominal operating pressure during each period.
The design of the waste disposal system is such that contaminated water is stored in the waste holdup tank until such time as the level of contamination is below the limits for discharge. At this time the holdup tanks may be reavailable for use by emptying the stored liquid.
Several important systems within the chemical volume and control system drain into the waste disposal system hold-up tanks. These are the volume and control tank drains, reactor coolant letdown system, reactor coolant drain tank discharge, and the demineralizer system drains.
If the tank was to be hydrostatically tested by filling it with water and pressurizing to 1.10 system pressure, the hold-up tank would be rendered useless. The plant would then be potentially put into an unsafe condition for any abnormal plant function and if startup occurred without a holdup tank being available.
TITLE:
QUALITY APPENDIX B Section 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 15 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 6 (Con~t)
Since this hold-up tank constantly stores liquid, any degradation of the tank material would show up prior to becoming a problem. RG&E believes that hydrostatically it testing the rad-waste hold-up tank puts Ginna's plant in an unsafe condition and therefore a request for relief from this requirement is sought.
III. Proposed Alternative Method:
A Visual VT-2 examination shall be performed once every period with the system at normal operating pressure to verify continued structural integrity.
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 16 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO 7 Components for which Relief is Requested:
CVCS, Three Charging Pumps and Discharge Piping to Discharge Isolation Valves.
ASME Requirements for which Relief is Requested:
IWC-5222(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200 F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. This corresponds to a test pressure of 3420 psig.
The charging pumps have a minimum hydrostatic test pressure limitation on the seals of 2400 psig, as specified by the pump manufacturer. As a result, the pumps and associated discharge piping to the first isolation valves cannot be tested to the required Code Test Pressure.
III. Proposed Alternate Method:
During the hydrostatic test and associated VT-2 examination, the charging pumps and associated discharge piping to the first isolation valves will be tested at a pressure of 2400 psig.
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TITLE REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT
'MANUAL XNSERVXCE INSPECTXON PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 17 of 27 ISX RELIEF RE UESTS RELIEF REQUEST NO ~ 8 I. Components for Which Relief is Requested:
RCS Overpressure Protection Nitrogen Accumulator System Valves PCV 430 and PCV 431C II. ASME Requirements for Which Relief is Requested:
IWC-5222(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200 F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. This corresponds to a test pressure of 137.5 psig.
The diaphragms in the operators of the subject valves are only designed to withstand a maximum pressure of 105 psig, and therefore cannot. be tested to the required Code test pressure.
III. Proposed Alternate Method:
The RCS overpressure nitrogen accumulator system will be tested to the Code requirements up to the flex connection to the valve operator. Operability of the diaphragm and operator is verified by valve testing requirements. In addition, an inservice pressure test at operating pressure will be performed once each inspection period on the piping, including the diaphragm.
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TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 18 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO ~ 9 Components for Which Relief is Requested:
Main Steam Secondary Side of Steam Generator and Downstream Piping to Class Boundary.
ASME Requirement from Which Relief is Requested:
IWC-5222(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200'F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. Since the design temperature of the Main Steam system is greater than 200 F, the test pressure is required to be 1.25 times Psv, or 1356 psig.
A pressure differential limitation of 800 psig between the primary and secondary side of the Steam Generator has been adopted. This was established early in plant life due to the experiences of some plants with primary side tube sheet cladding separation. To maintain this 800 psig differential, and the required pressure on the secondary side, the primary system must be heated up to a minimum of 160'F which would result in a problem with heat balance and a potential operational problem during implementation of the test procedure. The administrative controls necessary to assure a proper and safe test and the complexity required for the test procedure result in a situation that should be minimized.
In addition to the Section XI volumetric and surfaces examination requirements, the piping is part of the augmented inspection program since high energy break criteria.
it falls within the
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TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 19 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO. 9 (con~ t)
A letter was submitted to Dennis L. niemann. Chief Operating Reactor Branch 52, USNRC, Dated: November 8, 1979, requesting relief.
Subject:
System Pressure Test Restriction for Steam Generator and associated Feedwater and Main Steam piping, R.E. Ginna Nuclear Power Plant Nl, Docket No. 50-244.
III. Proposed Alternate Method:
Test the secondary side of the Steam Generator and associated Main Steam piping at a pressure of 1194 psig, which corresponds to 1.10 times the Psv setting.
These components are inside containment and any significant leakage would be detected by various leakage monitoring systems during plant operation.
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TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 20 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 10 Component for Which Relief is Requested:
Feedwater, Standby Auxiliary Pump recirculation line between AOV 9710A, AOV 9710B and their associated downstream flow orifices.
ASME Requirement from Which Relief is Requested:
IWD-5223(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design. Temperature of 200'F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200 F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be test:ed. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure Pd shall be substituted for Psv.
In order to hydrotest this piping to Section XI requirements, the pressure reducing flow orifices downstream of AOV 9710A & B would require removal and blank flanges inst:alled. System piping does not provide an isolation valve between the orifices and the Condensate Supply Tank. A significant reduction in tank level would be required to facilitate orifice removal, which is considered to be impractical.
II. Proposed Alternate Method:
The Class 3 portion of this piping shall be VT-2 examined at operational discharge pressure during functional testing which is performed once each period.
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TITLE REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 21 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 11 Component for Which Relief is Requested:
CVCS, Boric Acid Filter (CSFLBA) and all piping between valves 347, 348A and 349A.
ASME Requirement from Which Relief is Requested:
IWD-5223 (a) System Hydrostatic Test: The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200 F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure Pd shall be substituted for Psv.
The hydrostatic test pressure necessary to satisfy Section XI requirements will exceed the limits for the safe working pressure on the Boric Acid Filter housing flange gaskets.
II. Proposed Alternate Method:
The Boric Acid Filter and associated piping shall be VT-2 examined, at full operational pressure during inservice testing which shall be performed once each period.
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TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 22 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO ~ 12 I ~ Component for Which Relief is Requested; Emergency Diesel Generation:
Starting Air including receiver tanks and associated piping.
2 Fuel Oil Transfer pumps, suction and discharge including miscellaneous lines terminating at oil
~
storage tanks.
3 ~ Jacket Cooling Water system including miscellaneous line terminating at cooling water expansion tanks.
ASME Requirement from Which Relief is Requested:
IWD-5223(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200 F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200 F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure Pd shall be substituted for Psv.
Only portions of the piping associated with the components identified above are capable of being pressure tested.
The Air Start System pressure test would require termination prior to reaching the engine skid to preclude administrating air to the Air Start Motors. This would leave that portion of piping between the Air Start Motors to the first isolation, prior to reaching the engine skid, untestable.
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REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GXNNA STATXON FOR THE 1990-1999 INTERVAL 23 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO. 12 (Con~t)
The Diesel Fuel Oil Transfer system would require flange connection disassembly and the installation of blind flanges to isolate the Diesel Oil Storage Tank, which is vented to atmosphere, and the Day Tank where no means is provided to isolate the transfer pump discharge piping at a point close to the day tank. Additionally, the overflow piping from the day tank to the storage tank, which is identified as Class 3, has no isolation valves installed and is vented to the atmosphere.
The Jacket Cooling Water System would require isolating the Cooling Water Expansion Tank, due to vents to the atmosphere, which would include most of the piping subject to pressure testing. Due to the amount of piping within the class boundary which is unable to be pressurized, testing in accordance with Section XI requirements would not prove system integrity over and above the existing Surveillance Inservice and Functional Testing.
Proposed Alternate Method:
Inservice Testing shall be performed on the Air Start System at least once each period in accordance with the requirements of Section XX. Additionally, once each quarter a pressure decay test shall be performed on the air receiver to verify check valve operability in the reverse direction for the air receiver inlet check.
System Functional Testing shall be performed at least once each period on the Diesel Fuel Oil Transfer and Jacket Cooling Water Systems in accordance with the requirements of Section XI.
In addition to the testing discussed above, Technical Specifications 6.4.1 requires surveillance testing to be performed on a monthly basis. Such as, verifying operability of the fuel oil transfer pumps and verifying that the diesel starts from normal standby conditions.
II W
v b
1
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE XNSPECTION PROGRAM PAGE GXNNA STATION FOR THE 1990-1999 INTERVAL 24 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO ~ 13 Components for Which Relief is Requested:
Non-XSI classified systems, which do not carry radioactive gases or fluids, that contain line penetrating primary containment.
ASME Requirement from Which Relief is Requested:
XWC-5222(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200'F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested.
The safety function of these lines is to become part of the containment isolation system during periods when containment isolation is required. Therefore, the pressure testing requirements should be based on the containment system design not the associated process system design requirements.
III. Proposed Alternate Method:
Test these lines in accordance with 10CFR50 Appendix J, Reactor Containment Leakage Testing for Water Cooled Power Reactor commensurate with the safety function the line performs in accordance with Technical Specification, Surveillance requirements. Additionally, at least once each period exposed portions of the lines penetrating primary containment will be examined during normal system operation.
I lg
~ 4A 0
TITLE:
QUALITY APPENDIX B Section 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 25 of 27 ISI RELIEF RE UESTS RELIEF REQUEST NO. 14 Component for Which, Relief is Requested:
Service Water, All pressure retaining components within the Class 3 portion of the Service Water System.
ASME Requirement for Which Relief is Requested:
IWD-5223(a) System Hydrostatic Test; The system hydrostatic test pressure shall be at least 1.10 times the system pressure Psv for systems with Design Temperature of 200'F (93C) or less, and at least 1.25 times the system pressure Psv for systems with Design Temperature above 200'F (93C). The system pressure Psv shall be the lowest pressure setting among the number of safety or relief valves provided for overpressure protection within the boundary of the system to be tested. For systems (or portions of systems) not provided with safety or relief valves, the system design pressure Pd shall be substituted for Psv.
Rochester Gas and Electric believes that the hydrostatic test requirement for the service water system is impractical due to system design which dictates the use of an open-ended test. The portion of the system downstream of the heat exchanger is also open-ended and cannot be hydrostatically tested. The remaining section of the system is only isolatable by means of butterfly valves which were not designed to provide a leak-tight boundary.
With the system as such it would be impractical to expect the leakages other than at the valves could be detected.
The ample margin in cooling capacity inherently provided by system design does not dictate the need for an essentially leak-tight boundary. Since the system is in constant operation, its integrity is continually monitored. Thorough inspection of the system each period at the full operating pressure is adequate to detect any gross failures in the system without degrading system safety or availability.
III. Proposed Alternate Method:
Pressure retaining components within the operational boundary will receive an inservice test at operating pressure and an associated VT-2 examination each period during the interval.
% ~
TITLE REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 26 of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 15 Component for Which Relief is Requested:
Class 2, IWC 2500-1 Table Examination Category C-F-1 and C-F-2, Items C5.10 and C5.50.
ASME Requirements for Which Relief is Requested:
Category C-F-1 and C-F-2 (Items C5.10 and C5.50, respectively) for piping welds > 3/8 inches nominal wall thickness for piping > NPS4, a surface and volumetric examination is required on 1004 of each weld requiring examination at each inspection interval.
III. Proposed Alternate Method:
Rochester Gas and Electric believes as a minimum that the terminal connection welds of identified exempted welds (<
3/8" nominal wall) should be examined to the requirements of IWC 2500-1 Table, Category C-F-1 and C-F-2, Items C5.10 and C5.50 respectively. 'hese examinations are identified in the Class 2 Allocation Tables as Augmented Examinations and also are included under the category C-F-1 and C-F-2.
In the Program Plan Tables (Supplement 1 to Appendix B).
These are identified as C-F-1 or C-F-2 followed by These components are also noted in the instruction field.
I I~
lg
TITLE: REV.
QUALITY APPENDIX B Section 2 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 27 Of 27 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST N00 16 Components for Which Relief is Requested:
Class 1 and Class 2 Integral Attachments on Piping specifically to Support Attachments.
ASME Requirements for Which Relief is Requested:
For Class 1, Xntegral Attachments on piping as indicated in XWB-2500-1, Category B-K-l, Item B10.10, requires volumetric or surface examination be performed on Base Attachment Thickness > 5/8". For Class 2, Integral Attachments on piping as indicated in IWC-2500-1, Category C-C, Item C3. 20, requires a surface examination be performed on Base Attachments > 3/4".
It has been felt that support attachments to the pressure boundary such as gussets and stanchions should have a surface examination performed to insure the safety and integrity of the Class 1 and Class 2 Systems.
III. Proposed Alternate Method:
Surface examinations shall be performed on Integral Attachments on piping specifically support attachments once per interval in accordance with IWB-2500-1 and IWC-2500-1, B-K-l and C-C, respectively.
PAGE QUALITY ASSURANCE MANUAL P'ection 3 1 of 3 GINNA STATION EFFECTIVE DATE:
December 31, 1992 ROCHESTER GAS 8c ELECTRIC CORPORATION l~
SIGNATURE DATE TITLE PREPAREO BY X~JLSf IO I99 9g APPENDIX B QUAUTY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM FOR THE 1990-1999 INTERVAL APPROVED Y:
PAID BOUNDARY DRAWINGS
1.0 General
The PAID drawings included within this section identify drawings containing lines classified as ASME Class 1, 2, 3 and High Energy pressure boundary. A unique line identifier has been established for each class line and high energy pressure boundary line. The line identifier was used in the preparation of the "Line List" to identify the pressure boundary as well as document the line on the applicable PAID drawing.
The rules of ASME Section XI were applied to both class and high energy pressure boundaries as specified by IWB/C/D/F-1200. The color coded lines appearing on the applicable drawings identify those lines requiring Inservice Inspection and are not exempt under ASME Section XI for volumetric, surface and/or visual examinations.
Leakage examination boundaries and accompanying visual examinations for leakage is not addressed on these drawings.
Pressure boundaries that are not color coded reflect lines that are exempt from volumetric, surface and/or visual examinations.
The following color codes were applied to Class 1, 2, 3, and High Energy pressure boundaries; Class 1 Blue Class 2 & High Energy = Red Class 3 Green
TITLE REV.
QUALITY APPENDIX B Section 3 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 2 of 3 PAID BOUNDARY DRAWINGS The following list identifies the P&ID drawings that contain ASME Class 1, 2, 3, or High Energy pressure boundaries.
33013-1231 Rev. 18 33013-1231 Rev. 18 33013-1232 Rev. 7 33013-1236 Sheet 1 Rev. 3 33013-1236 Sheet 2 Rev. 5 33013-1236 Sheet 2 Rev. 5 33013-1237 Rev. 24 33013-1238 Rev. 8 33013-1239 Sheet 1 Rev. 6 33013-1239 Sheet 2 Rev. 6 33013-1245 Rev. 14 33013-1246 Sheet 1 Rev. 6 33013-1246 Sheet 2 Rev. 4 33013-1247 Rev. 17 33013-1248 Rev. 15 33013-1250 Sheet 1 Rev. 10 33013-1250 Sheet 2 Rev. 6 33013-1250 Sheet 3 Rev. 5 33013-1258 Rev. 12 33013-1260 Rev. 14 33013-1261 Rev. 19 33013-1262 Sheet 1 Rev. 7 33013-1262 Sheet 2 Rev. 3 33013-1263 Rev. 6 33013-1264 Rev. 12 33013-1265 Sheet 1 Rev. 5 33013-1265 Sheet 2 Rev. 4 33013-1266 Rev. 14 33013-1267 Rev. 10 33013-1268 Rev. 8 33013-1270 Sheet 1 Rev. 4 33013-1270 Sheet 2 Rev. 2 33013-1272 Sheet 1 Rev. 5 33013-1272 Sheet 2 Rev. 2 33013-1273 Sheet 1 Rev. 2 33013-1273 Sheet 2 Rev. 2 33013-1275 Sheet 1 Rev. 1 33013-1275 Sheet 2 Rev. 1 33013-1277 Sheet 1 Rev. 4 33013-1278 Sheet 1 Rev. 4 33013-1278 Sheet 2 Rev. 4 33013-1279 Rev. 7 33013-1863 Rev. 7
<<u
'll 1
TITLE: REV.
QUALITY APPENDIX B Section 3 2 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVXCE XNSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 3 of 3 PSrID BOUNDARY DRAWXNGS 33013-1865 Rev. 6 33013-1866 Rev. 9 33013-1870 Rev. 6 33013-1882 Rev. 5 33013-1886 Sheet 2 Rev. 3 33013-1887 Rev. 4 33013-1893 Rev. 8 33013-1908 Sheet 3 Rev. 2 33013-1915 Rev. 5 33013-1991 Rev. 4 33013-2275 Sheet 2 Rev. 1 33013-2278 Rev. 0 33013-2279 Sheet 1 Rev. 0
PAGE QUALITY ASSURANCE MANUAL Section.4 1 of 27 GINNA STATION EFFECTIVE DATE:
December 31, 1992 ROCHESTER GAS 5 ELECTRIC CORPORATION SIGNATURE DATE TITLE PREPARED BY APPENDIX B QUAUTY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM'OR REVIEW:
THE 1990-1999 INTERVAL CODE TABLES APPROVED BY:
1.0 General The following Code Tables were developed to address applicable ASME Section XI Requirements specifically to R.
E. Ginna Nuclear Power Plant. The Format of each Table is identical and a definition appears before the start of each text. ASME Section XI Code, 1986 Edition, no Addenda, and 10CFR50 specify that this program conforms to Articles IWB, C, D and F of ASME Section XI, Request and/or Code Case.
if not altered by a Relief The following list identified, the Code Table applicability.
Code Table Com onent Jurisdiction Class 1 Class 2 Class 3 Class 1, 2 and 3 IWF Component Supports-
W'I h
Table I item No.
Components aud Parts To be Exanuncd Exam!nation Methods Examination Requirements for Third Inspecdon Interval
~n Techrdque/Ertamfuadon Area Comments O
H The AqME Section XI Item No. Each type of <<mination Tbe NDE method TIds column provides inforznation regarding This column provides Informauon specific to cxamlnadon techniques, 0 and Category of the component area is listed in &is required to satisfy the number and/or percent of examinations examination areas, or comments.
are listed in these columns. column. Code requimnents requited to be performed for the inspection is listed In this Iruervah column.
H m
0 UI Q
g H W O o
~ LO H I 0 o M OX I
+ tCt O
> tf) Z
+ LEI H MHg 0 M
ID n
0
INSERVFX INSPECIION PROGRAM (BASS I mMPONENIS Item Examinsthn '
Components and Pans Rcq umnents for Examhation Tcchrdquc/Examhsrion No. Cattgory To be Examhed Tlurd Inspecaon Interval O
Area Comments W t3 REACfOR PRESSURE VESSEL K 8-A 0
81.10 Cucumfcrcntiaf and Volun>>ak 1004b of one circumfcrenthl weM to be Exsmhatlon of cbcumfereuriaf shefl weMs willbe performed with UT 81.11 Lony'tudinsl Shefl WCMs examined at autturrd disconthuity fn cbe tcchtdques. 'Ihere are no hny'telhsl weMs at IL E. Gbma.
81.12 behline region. The three other circumfercnthf weMs shsfl be pcriormed as a Auynentcd Program as speciT>>d by 10 CFR Part 50. Enunations may be performed at K or near the cnd of the hspccchn htctvaL 0 (/I 81.20 8-A Cucumfercnthl snd N/A Thcfc afc no meridionsl of c~cfcnthl head wclds at R. E. Chris.
81.21 Mcridionri Htad WeMs 81.22 R O 8120 8-A Shefl.to.Fhnge WCM Volumeak 100% of the weM to be cxsmheL At kast 5044 of the weM shsfl be cxamhed from the The shell.to.lhnge weM willbe cxsmhed hom the vessel seal surface and fnun the vessel waB Inside surhcc with UT. NOTEt Reliel OK~
flange hce by the end of che fust Inspecthn Request ia l. g 'LI3 K period and the remainder by the cnd of the third hapection pcrieL O (/l I
nw 8140 BA Head-to.phnge WeM Volurncak and 100% of the weld to be cxamheL The head.to.flange weM wfllbe examhed with UT and surhee + CI3O Surface examination techniques when the head h removeL >cf)
+ cf3 Z RIM 8-A K
BIBI Repair WeMs No repair weMs at R. E. Ghna.
~Kg 0 8350 B.D Nozde.to.Vessel WeMs and Volumetri 10096 of nozzks. At kasc 2544 but not sere The noule-to.vessel weMs and nozzle inside radius sections willbe 83.100 8-D Nozzle Inside Radius chan 5094 of the noules shall be examined by ecsmined whh UT.
Section the end of the first period and the rcmahder by 0>> end of the huevaL 84.10 8 E Paichl Penetration WCMs: Visual (Vl'2) At kast 2544 ol each group of welds of Examination willbe performed during the systen hydrostatic test or 84.11 Vessel Nouks comparabk size and function to be examhcd. ahcrnatively to Code Case NAPIL VT cxaminsrions wiflbe performed 84.12 Contml Rod Drive Nozzks in accordance whh IWA-5240.
84.13 Insaumentation Nozdes 85.10 ILF Prcssure. ctaining Volumeaic gi AB Nouk-to-safe end bua weMs NPS 4 or Exams msy be periormcd coinrident with the vessel nouk exams Disshnihr Metal WeMs Surface hrger. required by Category B.D.
I%20 B.F prcssure.Rctainisg No disshuhr metal weMs on RPV at lb E. Ginna.
8530 Disrimflar Mead WeMs 86.10 BG I Cksure Head Nuts >2 Surhce 1004b of nuts to be cxsmheL Examhation Nuts willbe examhcd with MT when removed for refueling.
Inches h Diameter msy be performed at or near che end of the hspection buetvsL
li
'I>>I " -8 ~ ~
INSERVKX INSPECllON PROGRAM CLASS 1 (Coca'd)
Item Examhsdon Comp@tents and Pars Examhadoa Examhathn Raluirements for Examination Tcchnhiue/Exanunadon No. Category To be Examhed Methods Tldcd Inspection Interval Area Comments REACTOR PRESSURE VESSEL cont'd H
0 B620 BG.I Qosure Head Studs, In Volmnetrk 100% of snub to be xamhecL Examhadon Qosure stud examhadons may be performed 'in place.'xamhadons Pkce, >2 Inches h may be performed at or near the end of d>> shouM be sehcdukd when stutb are mnovcd to reduce radhdon Diameter hspccdon htecvaL exposure and aBow the most thorough examhadon. Code Case N~-I sMI apply.
Qosure Head Buds, When Volumetric and I Kysb of studs to be examhed. Examhacion The sttxb willbe exanuned whh UT and MT. Code Case N~-I H
~ 0 RemovaL>>2 Inches in Smface may be performed ac or near the end of the shaB a pply.
(I)
Diameter inspecdon htccvaL Q
Threads h Fhnge Volumeuie 100% of tlueakd hoks to be Thc threads in Ilange willbe cxamhed fmm the Ihnge face with UT. H
>2 Inches in Diameter Emination may be performed at or near the ead of the nspection intervaL Er) O Prcssure.Rctahhg Qccure Visual (VT.I) AB washers and bushhgs to be examined OH~
0 Washers and Bushhgs upon stud removaL Examhation may be tg) H h
>>2 Inches Diameter periormed at or nem the end of d>> hspecdon > O(f) hcetvsL I B7.10 BG 2 Pressure.Retaining Bohng No pressure. retaining bohing <<2' diameter on RPV at L E. GhnL tf) O H s2 Inches in Muneter
+ tf) H g
td CRD Houshgs Bohs, Studs, and huts Visual (VT. I) AB bohr, studs, and nuts to be examhccL Pressure. retaining CRD housing bohing wBI be cxamhed when isassemblaL
~He B&.10 B-H Integral Attachments for No huegraBy weMed attachments oa RPV at IL E. Ghna that meet the d
RPV raiuremencs of Category B.H.
B13.10 B.N I Vessel Incerior Visual (VM) Accessible areas co be cxamhed during each hspecdon perioL B I%50 B-N.2 Interior Accachmencs Visual (VT-I) Accessibk attachment webb to be eaunbeL Withia Behline Region Examhadons may be completed at or near the end of the Inspeedoa hcetvaL (I)
BI%60 B-N-2 Visual (VM) Accessible cachmecx webb to be examhaL
(()
Examhadons may be compktcd at or near 0 the end of the hspection ItuetvaL po B13.70 B-NQ CoteSuppott Structure Visual (VT4) Wah core-support taructure removed, aB 0 accessible smfaccs to be examhaL Examinadons may be corn pleccd at or near the end of the inspecdon intecvaL BIC.IO BO Control Rod Drive Housing Volumetric or Webb h 10% of the peripheraI CRD housings The CRD housing webb wBI be examined with PT.
Webb Surface to be eramhaL ~docu may be performed at or near the end of the hspccdon htervaL
TaMe I INSHMCE INSPKCIION PROCRAM (2ASS 1 CDMPONENIS (Ocot'd)
Item Components and Parts Examinarion Exaadm&n Roidfrements for Exaudnation Technhlue/Euunination No. To be Examined Methods 'lhhd Inspection Iruetval Area Conunents O
b3 REACIQR PRESSURE VESSEL H 0
~ru nna'15.10 B.P All Pressure-training Visual (VT.2) All components to be examined during system VT examinations wBI be performed in accordance whh IWA.5240.
Boundaries for Vessel kaksge test. to be performed B15.11 B.P Components All Pressure.Retaining boundarie for Vessel Components Visual (VT-2)
In accordance with IWB4221 prior to plant startu p aher each refueling outage.
All components to be examined during system hydrosraric test or ahernatively to Code Case NAPL Examinations to be performed in accordance whh IWB.5221/5222 at or near VT eminatioas wiB be periormed in ~ wih IWA-5240.
0 H
(I)
Q the end of the aspccrion hterval 4H (x) O OH 0
I sE)
@ o(/) OX I
~tf) wa O ~ H
'Ll) Q
~
~ tD H 0 0
g0 (I)
(()
n 0
w I
TaMe I INSERVXX DISPEC I)N PROGRAM CXASS 1 COMPONKNIS (Coca'd)
Item Components and Parts
' for Req Irements Examination Technique/Exanunarion No. To be Exanuncd Third Inspect'on Interval Area Comments O
PRESSURIZER H 0
B2.10 B-B Chcumferentfd and Volurnctrie The upper and hwcr head-to.shell weMs to be Exauunation wiB be performed wih UT tccluuqucs.
B2.11 Longitudinal SheB-to.Head xaamfn<<L One foot of one Ionlp'tehaal weM B2.12 WeMs Intcrsccdng each headhoaheB weM to be examln<<L B220 B-B Cfrcumferentfd and Volumetric There are no p essurizcr bead weMs at R. E. H i&21 Meridhual Head WeMs B222 C irma.
0 U)
KLIIO BD Nozde-to.Vessel WcMs Q
No nozzle.to.vessel weMs. Hordes are huegraBly cast Mto the head L
of pressurizer at E. Ciima. gH XO B3.120 BD Nozde Inside Radius Sccrions Vofumetric All inside radius sections to be exandned. At least 251k but not more than 5016 of the nozdcs shall be exandncd by the cnd of the Examinarions wBI be performed with UT techniques.
or
~ tf) H ( s()
first pcriod and the remdudcr by the end of the nspcction hxcrvaL O U) OX I
U Heater Penctmtion WeMs Visual (VT-2) All p cssuizer heater enctrarion weMs shaB VT xandnarion wBI be performed h accordance with IWA.5240. +tf) O td P H be examined during system hdrostaric test or akernativcly to Code Case NA98 in accor- + 'Lf) H dance wkh IWB-5222 at or near the end of p the hupcction huctvaL 0 Notre.togafe End Surface and All butt welds to be examined. Exandnarions wBI be performed whh PT and UT techniques. d Dissinuhr Mead WeMs a4 Inches NPS Volumetric N 0 BL50 ILF NozdotoSafe End No nozzle-to.safe end dissimihu metal weMs on pressurizer at R. E.
DisshuBar Metal WeMs Cinna.
<4 Inches NPS I%60 B.F Nozde toSafe End Socket No disshnBar metal socket wekh on prcssurizer at L E. Ginna.
WcMs U)
(t) rcssilfe RctdnL1g Bokhig No bohfng on pcssurizer >2 inches h diameter at IL E. Chma. 0
>2 Inches in Diameter 0
B7M BC-2 Boks, Studs, and Nuts Visual (VT-I) All bohs, studs, and nuts to be examined. BoMng exanunations msy be performed ln place under tension or s2 Inches h Diameter when disassembhd or removed.
~ a%0 IM integrally Welded cssurizer support sMrt not required for examination Ior 3rd and 4th Attachments Intcrvds by Code.
I ~4 TaMe I INSERVICK INSPECIION PROGRAM I
CLASS CXtMPONENIS (Coat'd)
Item No.
~n Catcgorp Components and Pans To be Examined Xamlnarion Methods Examination Roluirements for Third Inspection Interval Examinsrion Tcchnhiue/Exatnlnarion Area Comments PRESSURIZER B1520 Conrd B.P Pressure-Retaining Boundarim for Vessel Components Visual (VT.2)
~
AU pressurizer components to be ~
during ay@em Icahage test. Examlnsrion to be pcriormcd in whh IWB-5221 VT examlnsrion wBI be performed in accordance whh IWBS24L H
0 BI521 B.P Prcssure Retaining ounfaries for Vessel Components Visual (VT.2) prior to plant startu p after each refuebng outage.
~
All pressurizer components to be examined during system h)drostarie test or ahernarivcly to Code Case NAPL Examlnarion to be per.
formed in whh IWB.5221/5222 at VT cxamlnarion will be periormcd in ~ whh IWA-5240.
0 H
(I) or near tbe cnd of tbe Inspecrion Interval XH (z) 0 o
~ tf) H O (I)
+Itf) ta I
O C' tf) Z MHg 0 4Q 0
I i QC SERVICX INSPKCllON PROGRAM C1ASS 1 0)MPONKNIS (Ooufd)
Item Xamnfnathn Components and Parts Examhathn Rcquhcments for Examhation Tcchdque/~n No. Guegory To be Examhed Thud Inspecrioa Intaval Area Cotnments O
th 0
B230 B.B Circumfercnrial and There are no stean generator hwcr head rircumfcrendaI or 8231 Meridiomd WcMs h the meridional weMs.
B232 Iiower Heal Tubeshcct.to Heal WeM Volumetri lhe tubesheet.to-head weM wBI be examineL Examinadon will be performed with UT techniques. H The <<xaminarion wBI be I'usted to one of 0>>
two steam generatots. 0 M B3.130 B.D Nozzlect>Vessel WeMs N/A These are no stean generator nozrie-to-vessel weMs. Nozzks are Q integraUy cast hto heah. H LH WO B3.1C0 B.D Nozzle InsMe Radius Volumctrie AB nozzle hsMe rahus sections will be Exanhations wBI be periotmcd with UT techniques. There are no OH~
Seaions cxanhed. At least 25qb but not more thm steam generator nozzle-to vessel wekh. H 50qb of the nozdes shall be exanhed by the 3 Cf) cnd of the fiat nspecrion period and the O M emafnder by the end of the Inspecthn I X
+COO ~ U Intervak
~'l) H B5.70 B-F NPS 4 or Larger Nozzle.to. Suriace and Examinations wBI be petfoaned whh PT and UT tcchdques at R. E. + 'Q HZ GInna Safe End DissimBar Mctrd Volumetric MH~ 0 WcMs 0
Less than NPSS, Noxricco. There are no steam generator weMs of tlds type.
Safe End DissimBar Meted WeMs Nozzle-toSafe End There are no steam generator welds of this type.
DissimBar Metal WeMs B6.90 BG-I Prcssure-Rcauning Bohng No bohhg on steam generatoa >2 inches in B6.100 >2 Inches in lÃanctcr diameter at Ih E. GInnL 85 110 M
B730 B4.2 Bohr, Studs, and Nuts Visual (VT.I) AB bohs, studs, and nuts willbe examinesL Examinarions may be performed la place under tension or when (t) s2 Inches h Diameter Tbe cxahation wBI be limhed to one of the rc\novcd 0 two aean generators.
BB30 B.H IntegraBy WcMed There are no steam generator huegraBy weMcd aaaehmcnts. 0 Attachmena IWB.2500 13.
B1530 B.p Pressure.Retaining Visual (Vl'2) AB steam generator components to be Boundaries exaudncd during system leakage test.
Exauunations to be periormcd h accordance whh IWB.5221 prior to plant stattup after each refuehng outage.
4
~ - 4 ~ ~ ~,
TaMe 1 I CSHIVICK QtSPECllON PIIOCRAM tlAiS I COMPONKNIS (Coca'd)
Itetn Examination Components and Parts Examination Requirements for Examination Technique/Examination Ntx Category To be Examined Third Inspection interval Area Comments A H
SIEAM CENERATORS Com'd 0
'g B1591 BP Visual (VT.2) All steam generator components to be VT xaminations willbe performed in accordance whh IWA-524L examined during system hydrosratk test or ahcrnativcly to Code Case NAPE Exanunadons to be performed in ccotdance whh IWtLS22I/5222 at or near the cnd of H t the inspectioa IntevaL m 0 U) 81620 Bt) Stcam Generator Tubing Volumetric The tubing In the hoc kg shk. Upend portkn. Examinatian requirements, xamfnadon method, and the extent and and optionally coM kg side wBI be examfnoL frequency of examinatkn shall be in accordance whh PLsnt Tcchnical SpceiTteatious. gH XO o~~
tf) H 'd a~<
+o0) $
I
+tf) A
> tf) I-3
+ tf) H 48 g
ahSS 1 COMPONENIS (Coot'd) kem Exambation Components and Parts Exambadon Examhsdon Rcq irements for No. Category To be Examined Methods qhftd hspeedon Intaval HEAT EXCHANGERS 8250 88 Head WeMs No meridional or eircmnfaendal heal weMs in heat cxehanger at R.
BXSI E. Cinna.
82.52 Tubeshect4o.Head WcMs Volumetric Each tubahcct to head weM wiB be Eam'nation wiB be paformed with UT techniques.
cxamfnoh I m
82.70 B.B longitudinal WeMs N/A There are no RHE bngbxBntd weMs at R. E. C boa 0 (I)
BX80 88 Tubcshat-toSheB WeMs Vol~ Each tubesheet to.sheB weM wiB be xaminadon willbe performed whh UT techniques.
cxambcd XH XA 83.150 83.160 B.D 8D Noxzle.to Vessel Welds and Nouk inside Radius Volumetric AB nosxk-to shell and noxrk bside radius sccdons wBI be examiners ht last 2546 but n wiB be performed with UT techniques.
OH~
Seenon not more than 50% of the noxtks shaB be tf) H examhcd by the end of the fkst bspecdon +4()
~ Q ( 'E) paiod and the remsiakr by the end of the Chloe. o(/) OX btavd. I
> tf) A 85100 SF Pressure.Retahung No heat exchanger dissimBar wcMs at lL E. C irma.
tf)
+QH Z 85 110 85.120 Dissbnilar Metal WeMs
~Hg 0 86.120 84.1 Pressure Retunlng BoMng N/A No heat exchatigcf bohbg at fL E.
86.130 >2 Inches in Dbmctcr 86.140 87AO SC.2 Bohr, Studs, and Nuts No heat exchanger bohing at R. E. Chna, s2 Inches in Diameter 88AO 8-H lntegraBy Wehkd Integrally weMcd anachments not requited for examination for 3rd Attarbnents and 4th btefvals by code.
815AO 8P Pressure.Retahung oundaries Visual (VT 2)
~ts AB RHE components wiB be examined duthg system k&age test. to be performed In a eordance whh IWB-5221 prbr VT enunations wiB be performed h aeeordance with IWA.5240 (/)
(D n
to phut startup after each rebelbg outage.
I 815AI B.P Visual (VT.2) AB RHE components will be cxambed during VT exambatbts wiB be performed b accordance with IWA-5240.
0 system hyrostade tat or alternatively to Code Case NA98. Exanunadots to be per-formed h accords'hh IWB-5221/S222 at or near the end of the bs pccdon htcrval
~-
t 4 f
TaMe I INSKRVKXINSPECIION PROGRAM I
CLASS COMPONKNIS (Coal'd)
Itan Components and Pans Exanunadon Xxamfnatbn Rcqufcmennts lor Examination Teeludqucf~a No. To be Examined Methods Th'ud Inspection interval Area Comments O
PIPING
'Ihe welh wiB be examined whh UT and PT.
0 B5.130 B-F DisshnBar Metal Welh Voiumetde and AB bun welh to be exanuneL a4 laches NPS Surface B5.140 B.F DissimBar Metal WeMs No issfmBar mecal wells <4 behes NPS at R. E. Glans.
<4 Inches NPS H
DhsfmBar Metal Socket No dissimBar metal socket weMs at L E. 6innL B5.150 B.F WeMs 0M B4.1 No pressure.retaitdng bohing >2 inches ht diatncter at L E. Ginna. Q B6.150 B6.160 PrcssureRetafning Bohfng
>2 laches in Diameter H
B6.170 Exj O BTM B4-2 Pressure Retaining Bohng No prcauze retaining bohing <2 inches on piping at L E. Glans. OH~
0 s2 laches in Dhmeter > cfk H B9.10 BJ Circumfacndal Pipe Wekh Volumetric and 2546 of the required ehcumfacntial butt oZ
~ cO fn OX
+ tfk O xa weMs to be exambeL See Note I at end of I IN. I I a4 Inches NPS Smface Tabk I lor sekction ahab. P H 89.12 Bd longitudinal Pipe WeMs Volumetrk and lo g'tudinal weMs chat adjob schcdukd Ihe piping wekh wiB be examined whh UT and PT. + cO H
<<4 heber NPS Surface ein:umlerenthl welds are to be examfneL d tTI One pipe diameter noc to exceed 12 Inches of 0 each bnghudinsl weM kngth required. R Cuemdaenthl WeMs Surface 2516 of the required drcumferendal butt 'Ihe piping weMs wiB be examined with PT.
<4 inches NPS weMs to be cxamfneL See Note I at cnd of Table I for sckction <<dtaia.
Longhudfnal Pipe Welh
<4 inches NPS No pressure-retaining bgitudinal pipe wekh <4 Inches NPS at R. E.
Glnna.
>If Bmnch Pipe Connecdon Volmnetrie and 2546 ol the required branch <<onneetbn joints The branch connecdon wekh wBI be examined whh UT and PT.
WeMs a4 Inches NPS Surface to be examined. Sce Note I at cnd of Table I for sekaion eritcrh. 0)
Branch Pipe Conneaion 25% ol the rcquhel branch conneedoa joints The branch onnectbn wekh will be examined whh PT. 8 Welh <4 inches NPS to be xarni'net See Note I at end of Table I 0 for selection aiteria. r.
Sockec Wekh 2546 of the requhed socket weMs to be The socket weMs willbe examined whh PT.
0 examineL See Note I at end ol Tabk I for selection criteria.
IntcgraBy WeMed lntegraBy welded attachments noc require! Ior examination for 3rd Anat hmaus and 4th intavals by code.
TaMe 1 INSKRVKXlKPKC1lON PIIOGRAM CKASS 1 COMPONENIS (Cont'd)
Item Components and Parts Examine don ExamlnatNn Reidfrements for xamfnation Tcchnhlue/Zxaminadon Nlx To be Exandned Methods 'Ihhd Inspecthut Interval Ama Comments O
PIPING Cont'd H 0
B15.50 BP All Pressure-Retaining Visual (VT 2) Allcomponents to be examined during system VT cxaminadon willbe performed h accordance whh IWA-524L for Piping kahage test. Examination to be performed in Components accordance with IWB 5221 prior to plant startup after each refueling outage.
=I BI %51 BP All Pressure-Retaining Visual (VT.2) Allcomponents to be examinM during system VT xammlnadon willbe performed h accordance with IWA-5240. r m
Components for Piping hdtostadc test or alternatively to Code Case NAPL Kxamlnadon to be performed h 0 (I) accordance whh IWB.5221/5222 at or near Q the cnd of the inspection Intcrvah H g H Er) O OH0) H
> O(f) (OX rE)
+ I ta Q 2s I
+tOO
>u)Z L tg)H 0
0 g0 (I)
(()
n ft 0
4 TaHe 1 INSERVKX INSPECIION PROGRAM (1ASS 1 COMPONENIS (Coat'd) CCC~
Item Components aad Pans Examination Rcquhements for Exammaaon Tcchmquc/Examfaatsoa No. To be Examined Third Inspectioa Interval Area Commelxs O
PUMPS H 0
B6.180 B4.1 Bohs aad Studs, >2 Inches Volumctrk All bohs and studs to be examined for items The bohing will be exaadned with UT. Code Cam N407.1 shall In Diameter associated whh oae component Ia a apply. The bohing may be examined In place under tension or when muhicomponcat group. 'sassernbkd or removeL Flange Surface for Bohfag Vins) (VT.I) Exaauaarion Ineieks I Inch annular surface VT xammaatkn wifibe performed when isassembkd.
>2 Inches Ia Diameter of flange amund each stud hole smfaee when When Coanecrion Is isassembkd isassembk4 0 (/)
auaO BG I Nuts, Bulling, and Visual (VT.I) All nuts, btuhiags, aad washers to be VT examination wfilbe performed when disassembluL Washers >2 Inches in examined when isasscmbkd on one Diameter component Ia a muhk>>mponeat group. td O No prcssure-rctahdng bo)riag <2 inches ht diameter oa pumps at Ib E.
OH~
Pressure.Retaining Bohing 0tf) H sQ Inches In Dklneter Ginna, R~+
0 (/)
BIL20 B.K.I Integrally Wchkd Integrafiy weued aaachments not rcquhcd for xammnarion for 3rd I
Aaachmeats and Cth Intervals by code.
+If)
~ O B12.10 B.l I Pump Caring Welds Volumetric Reactor coolant pump casing wehk on one ~If) Q pmnp to be cxamfnoL gtf) H 0
312'-I 2 Internal Surfaces of Pump Casings One reactor coolant pump to be examined when dhasscmbloL Pump casing hteraal surface wQI be examined with visual techniques when isassembktL Blu0 B.P Prcssure.Retaining Visual (VT.2) All compoaents to be examined during system X0 Boundarie for Pump kahage test. xaminarion to be performed in Components accordance whh IWB.5221 prior to plant sautup after cath refueling outage.
Pressure Retailuag Boundaries for Pump Components Visual (VT.2)
Surface and
~
All components to be examined during rystem hydrostaric test or ahernarivcly to Code Case NA98. Examfaarion to be periormed ln with iWS.5221/5222 at or near the end of the Inspecrioa Intervah Each reactor coolant pump fiywheel to be Vl'xaminations willbe pcriormed in accordance with IWA42C0.
Exanunations will be performed with MT and UT tcchrdqucs oa the 2 ft) n Volumcuie per Reg. Guide I.lc. Ifigh.stress active cola poaents as a minimum on an Auglncnted program V areas to be examined cash pcriod with full 0 cxaaunation at the end of IatetvaL
%7 ~ ~~
4
TaMe 1 INSKRVXXINSPECBON PROGI4W I
CIASS COMPONKNIS (Gxa'd)
Item Components and Parts Xxaminarion n Rcq drements for Examinarion Tcehrdque/Examine& n No. To be Exanuned Methods Third lnspecrion Interval Area Comments O
H 0
Et'0 BC-I Bolting>>2 Inches in N/A No valve bohfng >2 Inches in diameter at R. E. Ginna 86.220 Diameter 86230 Bohing s2 Inches in Visual (VI'I) Allbohs, studs, and nuts to be examined on 1 qhc boMng may be exanuned in place under tension or when Diameter component pef group. hisassembkd or removed.
H 81030 8 K. I Integrally Wekkd Imegrally weMed attachments not required for examinarion for 3rd Q (I)
Anachmcnts and 4th Intctvah by code. Q 81230 8 M. I Pressure.Retainbg Wc)dr No prcssure-retaining weMs in valve bodies <4 inches NPS at Ib E. R in Valve Bodies <4 Inches Ginna. WO NPS OH~
812.C0 8 M.l Pressure.Retaining Wehk Volumcak Ail weMs on one valve In each group of > cf) H
< tf) Z C r(3 in Valve Bodies trt Inches NPS valves that k of the sane consttucrion and simQar function to be cxamfneL o(/) AX
+ I ta X U I
81250 8-M-2 Imernal Smfaces of Valve Visual (VM) One valve in each group of valves that k of Internal smfaces of valve bodies to be examined whh VT4.
~ tD O ~ H Bodies on Valves the same co truction and simBar funcrion to tg b3
>4 Inches NPS be examined.
'El (El 815.70 8-P AB Prcssure-Retaining Vimal (VT.2) All components to be examined during system VT examlnarions will be performed In accordance whh IWB 52C0. 0 Boundaries for Valve kahsge test. Examinathn to be pcriormed in Corn ponents accordance whh IW8.5221 prior to plant stattup after each refueling outage.
~ns 815.71 8-P All Prcssure-Retaining undaries for Valve Components Visual (VT.2) AII components to be examined during system hdhrostatk test or ahernatively to Code Case NA98. Examinarion to be performed in accordance whh IW8.5221/5222 at or near VT wiU be peformed in accordance with IWA.52'/)
the end of the Inspection IntervaL (D
0 0
41 (1) ~ns shall be distnbutcd thtoughout the piping system to satisfy the extent required by itSME Section Xl 1974 Mtfon wkji hddenda though Summer 1975 as permktcd by 10CB50SSa(b)(2)(B).
Xamhathns reacted shaB be distributed ht such a manner to fncludet O
h3 (a) Terminal ends in each pipe or branch nm connected to vcsscb. H Terminal ends and joints in each pipe or branch run connected to other components.
0 (b)
(e) hB issimBar metal welds.
(d) hddkional piping wcMs so that the teal number of citcumfcrential bint weMs (or branch connection or socket wcMs) selected for examinatkut equals 259'f the chcumfercntial butt weMs (or branch connection or socket weMs) in the reactor coolant piping system. This total does not include weMs excluded by ItNB-1220. These a diYional wcMs may be located in one loop.
(2) ~n fs limited to those fntcgragy weMcd auachments that meet the foBowing condiYionst 0 V)
Q (a) the attachment is on the outsMe surface of the prcssure.rctsdnfng <<omponent; H
(b) the attachment provides component support as defined in NF-1110; Q H Exj O (e) the attachment base material design thickness fs 5/8 inch or greaterl and o
'4 H (d) the attachment weM joins the attachment either direcdy to the Nuface of the component or to an huegraUy cast or forged attachment to the component. U C sEj O (f) 0 Exj Examinations include the weMed attachments of piping rcquhcd to be exaudncd by Examhtation Category BJ and the weMcd aaachments to associated pumps and valves huegral to such piping. I
+ tf) A sgj Cl 0
R 'tl 8
0 0
M M K Item Components and Pats n Exandnation Rcq irements for Emination Tcehnviue/Examination O No. To be Examined Methods Third inspection Interval Area Comments Tbe ASME Section XI Item Ntx and Category of the component Eath type of agnination area is bsted in this The NDE method retiuired to satisfy This column provides information regarding the nmnber and/or percent of examinations Tbfs column provides information speei6e to examination technitiues, 0 examination areas, or conunents.
are listed h there columns. column. Code re irements required to be performed for the inspection fs fisted in this interval column.
m OM Q
H g H W O
>Vu) Z H
~oM I
~tgt O
+ lgt Q tgt H 0
CQ M
8 n
V 0
1 . ~
TaMe 2 (i) gK item Components and Parts Examination Rccptirements for Examlnatfon Tcchnklue/Examhation O No. To be Examined Tldrd Inspecthn Interval Area Comments H
PRESSURE VESSEIS 0
C1.10 GA Shell Circumferential 10096 of cath weM to be examined (applies WeMs only to wekb at gross ructural discontinuiYies). For mukfple vessels of shnBar design sbe and service, examinathns may be limited to one vcsseL H
m C130 GA Head Cueumferential WeMs Vohunetrie 100% of each head-to.sheB weM to be examhed. For inuhpk vesseb of shnBar 0M design size and seniice, examhations may be Q hmhed to one vesseL H g H C130 GA Tubeshcet.to-SheB WeM Volumctrk 10096 of cath weM to be exaaunoL For W O muldpk vesseb oi sinuhr design she and O rs service, examinations may be I'united to one 0 tf) H vesseL
~ o(/) (OX '0 C2.10 CB Nozzks in Vesseb xl/2. Surface AB nozzks at tcrmhal ends of plphg runs 'lhe weMs willbe examhed with MT or PT as a pplicabk. I C2.11 Inch Nominal Thickness which are sekctcd for <<xamhation under XU Nozzle-tohheB (or Head) Categories GF-1 and C F-2. 100% of each +NO WeM weM to be cxaminoL hlanways and hand Cd
+ tf) tf)
~
H holes excludetL For multiple vesseb of amBar design, size, and sctvke, xamhations may be limited to one vesseL
(/) H~0 M 0 (Ef C220 GB Nozzks Whhout Surface and AB nozzks to be selected at terminal ends of The weMs wBI be examhed whh UT and MTor PT as applieabk.
C221 Reinforcing Plate h Voluinetrie piping runs sekcttd for examhation under Verb > 1/2 Inch Categories C F.l and CF-2. 100% of each Nomhal lldckness Nozzle. weM to be examined. Manways and hand toSheB (or Head) WeM hoks excluded. For muhipk vesseb of
~
simBar design, sbe, and sctvke, examinatkns may be limited to one vesscL C222 CB Nozzle Inside Radius 1001k of each area to be examhed. The nozzle hside radius secrion wiB be cxamhed with UT.
Section and hand hoks are cxcludoL For muhpk vesseb of sinular design, sbe, and sctvke, M
8 emninarions may be limited to one vesseL 0 C230 GB Nozzks with Rehforeing See CLBI, C232, C232 r-Plate in Vesseb > I/2 heh 0 Nominal 'Ihkkness C231 GB Reinforcing Phte WcMs Surface All norzks at termhal ends of piping runs. The reinforcing phte webb willbe examhed with PT or MT as Surface to Nozzk gi Vessel For rnuhiple vest of shnBar design, size, applicable.
and sctvke, examharions may be limited to one vesseL
II TaMe 2 BCSERVXX BISPEc:IlON PROGRAM CLASS 2 COMPONENIS (Cont'd)
'n Components and Pans Examination Eauninadon Requirements for Examlnat>>n Technique/Eaunination Category To be Examined Methods Thhd Inspect'on Interval Area Counents O
PRESSURE VESSEIS Cont'd t3 H
C232 CB Nozzb-TogheB (or Head) N/h AB nozzbs at tclnunsI club of piping runx 1he Inside of vessel is not accessible, performing CL33 Instea/L 0
Webb When InsMe of For muhlple vesseb of shnBar design, size, Vessel is Accessible and service, xaminations may be Bmhcd to one vesseL CE33 CB N~ToShcB (or Head) Visual (VT-2) AB nozzles a! terminal cab of piping rurs. VT examinat>>ns willbe performed In ecoodance with IWB-524k H
Webb When InsMe of Vessel is I acccssibk and service, one vesseL
~ns For muhiple vesseb of simBar design, sbe, may be limited to The ~lllfcdcomponents to be examined during syacal kaklge test.
0 (I) m CLIO GC IntcgraBy WeMcd At. 1001b of cath weM to be cxaminccL Ihe weMed aaachments willbe examined whh MT or PT as tacluncnts Aaaclunents whose base material is 3/4 inch applicable. gH or greater to be sekctoL Where muhfple td O vessels are provided whh a number of simBar td aaaehmcnts, the aaachmcnts may be OH 0 tf) disaBnxed an>>ng the vcsseb. For mtddple H
vesscb of simBsr design and service, + o(/) OX examinauons may be limhcd to one vesseL
+ I I ta Q 2e C4.10 GD Bohhlg N/A No pressure.retaining vessel bohing >2 Inches in diameter at IL E. +tOO
+ ~
(EI tf)
>2 Inches in Diameter Ginna.
'LO H C7.10 GH Pressme Retaining Compo Vbual (Vl'-2) AB pressure.retaining hendarin for vesseb VT examinations willbe performed In accordance whh IWA.524th
~He 0
~tu to be examined during system pressure test.
to be performed In accordance with IWG5221 for each hapection petfoL g0 Pressure. Retaining Compo. Visual (VT.2)
~
AB prcssure-retaining boundancs for vesseb to be examined daring system hydrostadc tca or ahelnatively to Code Case NAPIL Exanunations to be performed In with IWG5221/S222 at or near the end of each hspcctioa Interval or during same bupection periods of each intervaL VT examinations will be performed in ccordace with IWA.5240.
(I) 8 n
Fso 0
I wary ~ I ww' ~ ' ~
~ I 1~ I' ~ I
~ I'
~ I'
~ I
~ ~ 0~
I~'
~ 0
~ ~
~
Table 2 SCEIRVICE INSPECIION PROGRAM CXhSS 2 a)MPONENIS (Coca'd)
Item No.
Components and Parts To be Examined
~n Requirements for
'Ihhd Inspection Ituctval Exaudnatbn Teetuuque/Eiraudnation Area Comments
~PtPtNG Nd CS c2 CF-I Pipe Branch onnections No bgittxUnal welds of this hcm number at R. E. Ghma.
M Austenitie Stainless Steel or High.ADoy Piping a2 NPS, Ionghudhud C550 C F-2 Pipbg WeMs ht Carbon or Surface and 100% of each eueumferenriat and 2.5t of 'the wekh willbe exanuned with UT and MT. H CRSI Low.hlby Steel a3/8. Volumetric cash bngitudinat weM requiYing xammnatb'on.
CR52 loch Nouunat Wall See Note 2 at cnd of Table 2 for selection 0 (I)
Thickness for Piping eritcria. In addiion, 10096 of the mdn stcam
>4 NPS, Cheumferenrial and uuun fecdwater weMs bested outside Q and longitudinal ~ ontainment and traversing safety ares shaB H be examined per Tcehnkal SpeciTieations. g hl 0 H
CSW CF2 Piping Wekh M Carbon or No eaten or bw~ steel nonexempt weMs in this category at L E.
CSs)t tow~ Steel > I/S.inch Ginna, oI C5.62 Nominal Wall Ttdckncss Otf) H for Piping a2 NPS and +(I)
> Q I xc NPS, ateumfcienriat O (I) OX and Longitudinal I
+ XU 70 C F-2 Socket WeMs M Cahon or No carbon or bw~ steel nonexempt weMs in this category at L E.
()t tf) O P H CS tf) Z Low.hlby Steel Gtnna. + tf) H Cx80 C F-2 Pipe Branch Co nectioiu in Carbon or Low.hlby ~a.
IIXBbof each cucumferenrial weld iequhfng See Note 2 at end of Table 2 (I) Hg 0El t))
Steel a2 NPS, for sebetbn eritcria. R 'E)
Cheumfcrentlst L E. Ghma.
g0 Pipe Branch Connections No bnghudinat weMs of ttds hem number at in Carbon or Low.ADoy Steel a2 NPS, Longitudinal Pressure.Retahung Compo Pressure.Rctahdng Compo-Visual (VT-2)
Visual (VT.2)
~n AB pressure. retaining to be performed in ~
undaries for piping to be examined iluring system pressure test.
with IWC.S221 for cath Inspection periaL AB pressure-retahung boundaries for piping to VT cxamtnatbru willbe performed In accordance with IWA-5240.
VT xaudnations wiB be performed In ~ with IWA.5240.
0 0
be examined during system hydrostatic test or aheinatively to Code Case NIPS. Exaudna-tion to be performed in accordance whh IWC.5221/5222 at or near the cnd of cath inspection interval or dming same hupeetion periods of each IntcrvaL
I Ta8e 2 NCSERVICE INSPECIION PROGRAM CXASS 2 COMPONKNIS (Coca'd)
Ran Components and Pats Examhathn Exauunauon Raiulfcments for Euuaim&n Technique/Examhathn No. To be Exanined Methods 'Iblfd Inspecrion Ifxaval Area Cnunents O PUMPS H 0
Cga GC Integrally WeMcd At- 10096 of each weM to be examhcfL Attachments whose base material tar hnents h 3/4 inch or greater to be scleacd. Sekcrion limhed to those components sekctcd under Examhation Categay GG. No pumps fnect tMs aherh at IL E. Ginna.
Prcssure.Rctahhg Bohing No pressure.rctahing pump bohing >2 Inches in diameter at IL E. H
>2 Inches h Dianeter Ginna.
m 0 (I)
C6.10 GG Prcssure Rctafnhg Was No prcaure.rctahing pump weMs at L E. Ginna. g Cd Q
h Pump Cashgs H Pfcssuft-Rctammg Compo.
Pressure. Retaining Compo Vhual (VT.2)
Vhual (VT-2)
Examhathn to be Peformed h ~
All pressure.retaining boundaria for pumps to be exanined dming prcssure test.
wkh IWC 5221 for each inspccrion pai>L All prcssure.rctahing boundaies for pumps VT examifmtions wQI be performed in accordance whh IWA.52ath VT cxamlnathns wBI be perfoaned in accordance whh IWA.524IL n
g Cd Otf) H Cd H
0 O (f) O I sT)
Cd I
to be examhed during system hyostatie tat or ahernarively to Code Case NAPIL + tf) A Examhation to be afoffncd in accordance whh IWG522I/5222 at or near thc end of + tf) H
<<ach hspeaion intaval or during same in. p speaion periods of each htavaL (I) H g0 gO 4A f'
n 0
ll 7
l7 l
V l
L I 0
TaMe 2 INSEWKX lÃSPECIION PROCRAbt (1ASS 2 CXiMPt)NENIS (Coca'd) item Components and Parts Exaaumuhn Rniuhemeats for Exanunadon Tcchnhluc/Examiaathn Ntx Tobe~ TIdtd I spectioa Interval Area Comments O IntcgrnUy WeMcd At- Surface l0095 of each weM to be exaadnoL 'Ihe weMs willbe examhed whh MTor PT as apptkabk.
0 tee bmcats Aaachments whose base material is B/4 inch or greater to be sckcteL Selectha limited to those components sckctcd under Examhatioa Category GC.
I C4AO GD Prcssure.Retaining Bohing No prcssure-retaining boMng >2 hches In dhmcter at IL E. Ginna. m
>2 Inches in DLtmctcr (I) g td Q
Prcssure Retaining WeMs No pressure.retaining welds in valve bodies at IL E. Chma.
hr Valve Bodies g H Visual (VT.2) AU pressure.retain)ng Ounddaries for valves m be examined during system pressure test.
VT ~ns wiU be perfoaned h accordce whh !WAS'(h OH' W 0 Examination to be performed ln accordance 'D H with IWG522I for each Inspccthn perhxL U tf)
VT cxaadnathns wUI be performed in accordance whh IWA.524'
<o(0 (OX M) E)
Pressure.Retaining Compo. Visual (VT-2) AU prcssure.rcttdning boundarks for valves to be cxaauned durhg system hydrostatic test or +I I
+(() ta ahcrnadvely to Code Case NA98. Exaauaa. t)) A 1() t3 tioa to be performed In accordance with IWG522I/5222 at or near the ead of each
~
~t() H hupcctioa interval or dming same hapecthn 0 0 periods of each huctvaL
(I) Thc weMs selected for cauninarion shaB Include 78%, but not kss than 28 wcMs, of aB austenhk stafnkss steel or hlghaBoy weMs not exempted by IWC.1220. (Some welds not exempted by this Case are not O rctiulrcd to be no destnxtivcly cxandned per Eaunfnation Category Gp.l. These weMs, however, shall be included In the toad weM count to whkh the 7$ % sampling rate is appBccL) 'lhe eaunfnarions dudl be disaibutcd as foUowst (a) the examinations shaB be distributed among the Qass 2 systems prorated, to the degree pmcticable, on the number of nonexempt austenitk stainkss steel or high~ wcMs in cath system (I.cv if a system 0 contains 3096 of the nonexempt wcMs, thea 30% of the no dcstrucrive exanunations requited by ExamInztion Category C F-I shouM be performed on that system)l (b) wftidn a system the examinations shaB be htrBnxed among terminal ends [see Note (3)) and tuuctural iscontinuiYies [see Note (4)) pmratcd, to the degree practkable, on the nmnber of nonexempt terminal ends and struttura) discontinuhks In that system; and (e) whhin each system, xaudnarions shaB be istrBnucd between Bne sizes pmratcd to the degree H ~
practkabk.'2)
[II The wcMs sekctcd for examfnatkn shall Include 75%, but not kss than 28 wcMs, of aB carbon or kw~ weMs not exempted by IWC 1220. (Some weMs not cxcmptedby this Case are not ruiufred to be 0 (I) ~
nondcstrucrively exaudncd per Examination Category CF.2. These weMs, however, shaB be Included in the total weM coum to which the 7.516 sampling rate is appBcd.) 'Ihe xauninations shaB be tBstribtucd as fo Bows:
gQ WHY (a) the exattdnatfons shaB be d'stributed among the Ckss 2 systems prorated, to the degree practkable, on the number of nonexempt carbon or the nonexempt weMs, thea 30lb of the nondestructive cxaminarions rctluired by Exauunation Category C F-2 shouM be performed on that system);
kw~ weMs in each system (ie., Ifa system contains 30)b of Ezj 0 Z OH tf) H 'g Q (b) witidn a system, the examinarions shaB be distributed among term!naI ends [see Note (3)) and structural discontinuiYks [see Note (4)) prorated, to the degree practicable, on the number of nonexempt terminal ends and structural discontinuhies in that system; and g
~ tf) R C 'Tj DV) O[zj (e) widun cash system, euuninarions shaB be distributed between line sizes pmratcd to the degree practkable. + HI ta V a
+CO Q H (d) ()nly those weMs showing rcpottabk prcsetvke uansvvcrse Indications need to be caunfned for transverse rcBcctots. ~ g L tf)H (3) Terminal ends are the extreauYies of piping runs that connect to suctures, components (such as vcsseh, pmnps, valves), or pipe anchors, each of wluch acts as a rigM restraint or provides at least two degrees of ansiYxed restraint to piping thermal expansion.
(I) H~o oem (0) Structural discontinuides include pipe weM joints to vessel nozzks, valve bodks, pump casings, pipe fuungs (such as efbows,tees, reducers, Ilangcs, ctc., conforming to ANSI 816.9), and pipe branch conncctkns R IT) [xj and parings.
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Item Components and Parts Examination Examfnanon Requirements for Examination Techtdque/Examfnadon No. To be Examined Methods Third Inspecdon Interval Area Comments H
The ASME Section XI Item No. Each type of examination The NDE method TIds column provides Information regarding Tlds colmnn prtwides Informatioa speeiTtc to examination techrdques, 0
aud Category of the component area is bstcd in this required to satisfy the nmnber and/or percent of examinations examination areas, of comments are lhted in these columns. column. Code req irements required to be performed for the lnspccnon is bsted in thiis Intcrvab column.
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Item Components and Parts 'n Eaination Rcqu'rements for O No To be Examined Methods Mrd Impaction Interval (BASS 3 COMPONENTS D1.10 DA Pressure Retaining Compo- Visual (VT 2) AB components to be exandned during system VT examfnatkns wiB be performed in accordance with IWA-5240.
D2.10 DB pressure or system hdrrostatk test as defined D3.10 DC by IWD.2500 I Tabk Descriptive.
Exam(nstkn to be performed In coordance whh IWD 5221 for each hupcctkn perkd and performed once each interval in accordance whh IWD-5223. 0 (I)
D120 DA integral Attachments of Vhue) (VM) AB rnlufred attachments to be examined The Integral attachnents wiB be xaamincsL Q Supports and Stsarnints, during each Inspcctkn Interval as defined by H through D1.60 Hydraulk Snubbers, IWD 2500 I Table Deter( ptkns. For mahple XH X
DL20 DB Spring, Constant load, and components In a system of simfiar design, O through Shock Absorbcrs functkn, and service, the hucgral attachment o w IXL60 of only one of the muhfpk components shaB 0 tf) H D320 DC be examfneL The integral attachmerus ~ + C s()
through sekctcd sMl correspond to those support Pa O (I) 0 X D3.60 components sekcted for examination In accor-dance with IWP-2510(b).
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~n Category Components and Parts To be Examined Examination Methods
' Reqfirements for Thhd inspection interval Examfnathn Techtdque/Examination Atra Comments 0
The ASME Section XI Item Ntx Each type of exination The NDE method This column provides information regarding This colmun provides information specfTic to examination techniques, a
and Category of the component area is hsted in this required to satisfy the number and/or percent of examinations cxammatfon areas, or cottunents.
are listed fn these cofunus. cohunn, Code requirements required to be petformed for the nspection is listed in this intetvaL column. H 0 M Q
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1 0
PIATE AND SHELL TIPE SUPPORTS LINEAR TVPE SUPPORTS AND COMPONENT STANDARD SUPPORTS F1.10 F.A Mechanical Connccrions to Visual (VTZ) Component supports to be selected for Functional tesung of snubber type support components shall be through Pressure.Retaining Compo. examination are the supports of the performed In accordance with Tcchnical Spceificarions and ln FIAO nents and BuBding nonexempt f2ass I, 2, and 3 components accordance with requirements of this program. =I F2.10 F-B ; WeM Connce- scheduled to be cxamfnecL Dtamfnatkut r m
through F2AO tions to BuBding Structure; WeM and Meehanieal boundaries cstabhshed in accordance with IWF-1300. Examinations may be performed 0 M F3.10 Connccrions at Interme- during normal system operation or plant out. Q through diate Joints in Mulri. ages.
fX50 connected Integral and g H Nonfntcgrai Supports; and W A Component Displacemcru Settings of GuMes and o~~
Stops, MisaBgnment of a+ cd H Supports, Assembly of Su-tfy O
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