ML17263A523: Difference between revisions
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This RR&M Program shall apply to ASME Class 1, 2 & | This RR&M Program shall apply to ASME Class 1, 2 & | ||
3 pressure boundary piping, components & supports. | 3 pressure boundary piping, components & supports. | ||
1.8.1 Performance of Repair Repairs on ASME Class 1, 2 or 3 components 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 | 1.8.1 Performance of Repair Repairs on ASME Class 1, 2 or 3 components 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 th'e 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. 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: | 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: |
Revision as of 08:09, 10 November 2019
ML17263A523 | |
Person / Time | |
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Site: | Ginna ![]() |
Issue date: | 12/31/1993 |
From: | ROCHESTER GAS & ELECTRIC CORP. |
To: | |
Shared Package | |
ML17263A522 | List: |
References | |
PROC-931231, NUDOCS 9401250044 | |
Download: ML17263A523 (519) | |
Text
QUALITY DATE ASSURANCE APPENDIX B 12/31/93 MANUAL R.E. GINNA NUCLEAR POWER PLANT GINNA STATION INSERVICE INSPECTION PROGRAM PAGE FOR THE 1990-1999 INTERVAL 1 of 5 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 and Requirements 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 4 Exemption Criteria 4.5 Inservice Inspection Program Plan 5.0 Class 1, 2 & 3 Component Supports 5.1 Basis for Preparation 5.2 Components Subject to Examination 5.3 Extent and Frequency of Examinations 5.4 Exemption Criteria 5.5 Inservice Inspection Program Plan 9401250044 940111.
ADOCK 05000244.
PDR 8 'DR
QUALITY TITLE: DATE ASSURANCE APPENDIX B 12/31/93 MANUAL R.E. GINNA NUCLEAR POWER PLANT GINNA STATION INSERVICE INSPECTION PROGRAM PAGE FOR THE 1990-1999 INTERVAL 2 of 5 Section 2.0 Relief Requests:
1.0 Introduction 1.1 General Section 3.0 P&ID 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, & 3 IWF Component Supports Section 5.0 Line List:
1.0 General Line List Section 6.0 Allocation Tables:
1.0 General ASME Code Requirements Class 1 Class 2 Class 3 Class 1,2, & 3 Supports Augmented Reactor Coolant Pump Flywheel Program Reactor Vessel Augmented Program, Category B-A High Energy Program Snubber Program Seismic Support Program
QUALITY DATE ASSURANCE APPENDIX B 12/31/93 MANUAL R.E. GINNA NUCLEAR POWER PLANT GINNA STATION INSERVICE INSPECTION PROGRAM PAGE FOR THE 1990-1999 INTERVAL 3 of 5 Section 7.0 UT Calibration Blocks:
1.0 General UT 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 & Testing Requirements 7.0 Replacement & Testing Requirements 8.0 Scheduling 9.0 Reports & Records Table 1 High Energy Welds & Component Support Identification.
Section 9.0 Snubber Program:
1.0 General 2.0 Examination, Testing & Monitoring Requirements 3.0 Examination & Testing Methods 4.0 Examination & Testing Frequency 5.0 Examination, Testing & Monitoring Evaluation 6.0 Repair, Replacement & Modification Requirements 7.0 Scheduling 8.0 Reports & 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 & Testing Requirements 7.0 Replacement & Testing Requirements 7.0 Scheduling 8.0 Reports & Records
QUALITY DATE ASSURANCE APPENDIX B 12/31/93 MANUAL R.E. GINNA NUCLEAR POWER PLANT GINNA STATION INSERVICE INSPECTION PROGRAM PAGE FOR THE 1990-1999 INTERVAL 4 ot 5 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 & Testing Requirements 2.7 Scheduling 2.8 Reports & Records 3.0 Reactor Coolant Pump Flywheel Program 3.1 General 3.2 Examination Requirements 3.3 Examination Methods 3.4 Frequency of Examination 3.5 Examination Evaluation 3.6 Repair, Replacement & Testing Requirements 3.7 Scheduling 3.8 Reports & Records 4.0 Class 1 Bolting Program (IEB 82-02) 4.1 General 4.2 Examination Requirements 4.3 Examination Methods 4.4 Frequency of Examination 4.5 Examination Evaluation 4.6 Repair, Replacement & Testing Requirements 4.7 Scheduling 4.8 Reports & Records 5.0 Reactor Vessel Augmented Program, Category B-A 5.1 General'.2 Examination Requirements 5.3 Examination Methods 5.4 Frequency of Examination 5.5 Examination Evaluation 5.6 Repair, Replacement & Testing Requirements 5.7 Scheduling 5.8 Reports & Records
QUALITY TITLE: DATE ASSURANCE APPENDIX B 12/31/93 MANUAL R.E. GINNA NUCLEAR POWER PLANT GINNA STATION INSERVICE INSPECTION PROGRAM PAGE FOR THE 1990-1999 INTERVAL 5 of 5 SUPPLEMENT 1 TO APPENDIX B "Inservice Inspection Program Plan" (Separate Document controlled by Materials Engineering 6 Inspection Services.)
SUPPLEMENT 2 TO APPENDIX B "The Repair, Replacement and Modification Program"
PAGE QUALITY ASSURANCE MANUAL GINNA STATION Section 1 1 of 34 EFFECTIVE DATE:
December 31, 1993 ROCHESTER GAS 8E ELECTRIC CORPORATION SIGNATURE DATE TITLE APPENDIX B PREPARED BY:
QUALITY p +Dt g ~ ( q3 R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM REVIEW:
FOR THE 1990-1999 INTERVAL APPROVE BY:
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.
- 1. 1.2 This program excludes'he controls of the Enforcement Authority, and N-Stamp, in addition to IWE of ASME Section XI, since by'the 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVXCE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 2 of 34 1.2 Applicable Documents RG&E has adopted the following documents as the basis for the third inspection interval and is 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 & Pressure Vessel Code,Section XI, 1986 Edition, no Addenda, Appendix IV.
1 '.2 U.S. Nuclear Regulatory Commission Regulatory Guides:
(USNRC) a ~ 1.14, Rev. 1, "Reactor Coolant Pump Flywheel Integrity"
- b. 1.147, Latest Revision, "Xnservice Inspection Code Case Acceptability ASME Section XI, Division l" c ~ 1.150, Rev. 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Xnservice 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".
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 3 of 34
- h. 1.65, Rev..0, "Materials and Inspections for Reactor Vessel Closure Studs".
1.2.3 ASME Code Cases In accordance with 10CFR50.55a, Footnote 6, ASME Section XI Code Cases referenced in Regulatory Guide 1.147, "Xnservice Inspection Code Case Acceptability- ASME Section XX, Division 1," 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 1.147 may be incorporated in the program and Regulatory'uide 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
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 4 of 34 N-402 IWA-2233 Eddy Current Calibration Standard Material 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-481 B-L-1, Item B12.10 Alternative examination requirements for Cast Austenitic Pump Casings.
N-491 IWF-1000/2000/3000 Alternative rules for examination of Class 1, 2, 3, and MC Component Supports of Light Water Cooled Power Plants.
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
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TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVXCE INSPECTION PROGRAM PAGE GXNNA STATXON FOR THE 1990-1999 INTERVAL 5 of 34 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 1.2.7 R. E. Ginna, Technical Specification 4.2.
"Inservice Inspection".
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.
1.2.13 USNRC, NRC Region I Inspection Report 50-244/90-12, Section 3.4 regarding repair program "Service Induced" and "Code Rejectable" repairs.
1' 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVXCE XNSPECTION PROGRAM PAGE GXNNA STATION FOR THE 1990-1999 INTERVAL 6 ot 34 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 ' 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.
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 RG&E for purposes of inservice inspection based on Section XI, Article IWA-1320, and definitions contained in 10CFR50.2.
l.5 Responsibility As specified in Paragraph IWA-l400 of ASME Section XI, RG&E bears the overall responsibility for implementation of an XSX program. Administrative Procedures, NDE Procedures, ISX Plans and Schedules are 'in place to control and implement these Inservice Xnspection requirements.
TITLE REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 7 of 34 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 Xnspection and associated Repair, Replacement and Modification activities. ASME NXS-1 and NIS-2 Forms shall be generated and included within the Xnservice Inspection Report.
f 1.7 Examination Methods and Requirements 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:
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TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 8 of ~ 34 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're 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 a's 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.
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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 9 of 34 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.
l.7.5 Evaluation of Examination Results and Successive Inspections 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 service induced 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). Acceptance by Repair or Replacement can also be performed.
Successive inspections for Class 1 relevant service induced conditions shall comply with the
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATXON FOR THE 1990-1999 INTERVAL 10 of 34 requirements of IWB-2420(b) and (c). For Class 3 relevant, service induced conditions, successive inspections shall comply with the requirements of IWC-2420(b) and (c).
Class 2 components containing relevant service induced 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 XWC-3134(b). Acceptance by Repair or Replacement can also be performed. For Class 2 relevant service induced conditions, successive inspections shall comply with the requirements of XWC-2420(b) and (c).
Period 1 and u to the '9'4 Outa e '86 Code Class 1, Class 2, and Class 3 Supports containing relevant service induced conditions will be considered acceptable for continued service providing'n evaluation or test is performed that demonstrates the component's acceptability.
Acceptance by Repair or Replacement can also be performed. Successive inspections on supports containing relevant service induced conditions shall comply with the requirements of IWF-2420(b) and (c) .
Periods 2 Startin '94 Outa e and 3:
Class 1, Class 2, and Class 3 Supports whose visual examination does not reveal conditions as described in the acceptance standards (-3400 of Code Case N-491), shall be acceptable for service.
Visual examinations which detect relevant service induced conditions as described in the acceptance standards shall be unacceptable for service unless accepted or corrected by one of the following methods; Corrective measures (adjustment, 'repair, or replacement), and subsequent reexamination including the requirements of -2220(b) (Note:
Successive examination requirements of -2420 must be applied to supports containing relevant service induced conditions in which corrective measures were utilized for acceptance.)
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TITLE I REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 11 of 34 Evaluation 'r Test to the extent necessary to substantiate its integrity for its intended service (Note: Visual examinations that detect surface flaws that exceed -3400 criteria shall be supplemented by either surface or volumetric examinations) .
Note 1: Flaws or relevant conditions in which analytical evaluations are being performed to determine acceptance, may be considered acceptable for continued service during this evaluation process.
1.7.6 Additional Examinations 1.7.6.1 Inservice examinations performe'd that revealed service induced flaws that exceeds the acceptance standards shall have additional examinations performed.
1.7.6.2 Additional examinations shall be performed to the extent required by IWB-2430 whenever a service induced rejectable flaw is identified during the performance of ISI examinations in accordance with the requirements specified within IWB and IWD.
1.7.6.3 Additional examinations shall be performed to the extent. required by IWC-2430 whenever a service induced rejectable flaw is identified during the performance of ISI examinations in accordance with the requirements specified within IWC.
1.7.6.4 Additional examinations shall be performed, starting the '94 outage, to the extent required by
-2430 (Code Case N-491) whenever a service induced rejectable flaw is identified during the performance of ISX examinations in accordance with the requirements specified within Code Case N-491 and XWF.
TITLE:, REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 12 of '34 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 ASME Class 1, 2 &
3 pressure boundary piping, components & supports.
1.8.1 Performance of Repair Repairs on ASME Class 1, 2 or 3 components 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 th'e 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.
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.
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TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 13 of 34 1.8.2.2 The examination shall include the original examination method that detected the flaw, applicable.
if 1.8.2.3'pplicable 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-3100.
1.9 Replacement Requirements Replacement requirements, which includes modifications, are applicable to ASME Class 1, 2 and 3 pressure retaining components and piping systems and their supports, unless specifically exempted by ASME Section XI, 1986 Edition, Article IWA-7400.
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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 14 of 34 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.
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'e 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 C
the requirements of the replacement program.
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TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 15 of 34 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'ressure boundary, the weld shall be examined in accordance with the requirements of 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.
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.
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QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 16 of 34
- 1. 10 System Pressure Testing
- 1. 10. 1 General Requirements
- l. 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-2500-1 Examination Cat.egory 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 int:erval are described in the following:
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 st;artup.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E..GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 17 of 34 1.10.2.2 FUNCTIONAL PRESSURE TEST This test is performed once each inspection period. The boundary subject, to test pressurizat'.ion and the associated VT-2 examination during a system functional pressure t.est 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-' examination during a system inservice pressure test will include only those pressure retaining component'.s under operating pressure during normal system service.
1.10.2.4 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 hydrostat:ic 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 associat;ed VT-2 examination during a system pneumatic pressure test are the same as a hydrostatic pressure test.
TITLE:
QUALITY APPENDIX B Section 1 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVXCE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 18 of 34 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, Xnservice 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 6 Modification Program. Code Case N-498 does not apply to Repairs, Replacements or Modifications.
1.10.3 Examination Requirements 1.10.3.1 During the conduct of pressure tests, certified VT-2 examination personnel, using RGGE 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 PAID'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.
1.10.4 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 19 of 34 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, Inservice and Hydrostatic tests, the level of syst: em pressure and temperature indicated or recorded by normal operating system instrumentation, or alternatively by test instrumentation is acceptabl'e. 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.
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.
TITLE REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 20 of 34 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 106% of the specified test 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.
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 accommodate 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 21 of 34
- 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.
- 1. 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'he 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 1.10 P 200 1.08 P 300 1.06 P 400 1.04 P 500 or greater 1.02 P "P " is the nominal operating pressure corresponding with l00% rated reactor power.
Linear interpolation will be used at intermediate test temperatures. Technical Specification heat-up/cool-down limits will be observed.
1.10.4.6.2 Class 2, 3, and High Energy:
a ~ 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 syst: em pressure.
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QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 22 of 34
- b. In the case of atmospheric storage tanks, the nominal hydrostatic pressure developed with the tank filled to its design capacity will be acceptable as the test pressure.
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 f'irst 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".
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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 23 of 34 1.10.6 Scheduling When using the inspection plan, it should be noted that during a period or refueling outage in which 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-l000 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 (b)(2)(ii), the extent and frequency requirements for Class 1 Category B-J weld examinations may be based on the 1974 Edition of ASME Section XI with Addenda through Summer, 1975. This earlier Code does not. have any criteria established for the selection of specific welds to be examined, and therefore, stress level criteria, and terminal end criteria is not required for the selection process. Instead of utilizing the earlier Code rules for Category B-J welds (which has no selection guidance), the 1986 Code rules will be utilized to the extent, practical. In lieu of the stress level selection criteria Table IWB-2500-1, Note l(b), all accessible terminal end, welds (including terminal ends to vessel per Note l(a) of Table IWB-2500-1) shall be selected. Refer to Code Table 1 in Section 4 for selection criteria.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 24 of 34 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.
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-1220, 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 of 1 inch nominal pipe IWB-1220(b) (1) size (NPS) and smaller, except for steam generator tubing Components and their connections IWB-1220(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 Znservice 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 25 of 34 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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QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION -FOR THE 1990-1999 INTERVAL 26 of 34 3.4.1 The following components (or parts of 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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 27 of 34 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 requirement's 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.
3.4.3 In addition to the exemptions of IWC-1220, non-piping component size exemptions shall be as follows: Nonpiping components having a cumulative inlet and a cumulative outlet nominal cross section area, neither of which exceeds the nominal cross section area of the applicable exemption size, shall be exempted from the surface, volumetric, and visual VT-1 and VT-3 examination requirements of IWC-2500. The applicable exemption size shall be 4" NPS for all systems except the Class 2 High Pressure Safety Injection System, which shall be 1 1/2" NPS. This position is based on the interpretation of the clarification made by footnote 1 of IWC-1220 of the 1989 Addenda to ASME Section XI. For the purpose of applying this criteria, it is assumed that the intention of the Code is not to accumulate the cross-section areas of piping from different fluid systems entering a given component ( e.g., heat exchanger shell &
TITLE REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 28 of 34 tube sides), nor the non-process piping (e.g.,
vents, drains and instrumentation).
3.5 Inservice Inspection Program Plan (Supplement 1 to Appendix B)
This plan provides the examination requirements for Class 2 components per ASME Section XI, 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.
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.
~
~
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 29 of 34 4.2 Components Subject to Examination 4.2.1 Based on the requirements of Section XI, Class 3 nonexempt pressure-ret:aining 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'ore 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 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 30 of 34 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-l220:
Section XI Exem tion Criteria Reference Integral attachments of IWD-l220.1 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:
(a) the components are located in systems (or portions of systems) whose function is not rec{uired 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.
REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 31 of" 34 4.4.2 In addition to the exemptions of IWD-1220, non-piping component size exemptions shall be as follows: Nonpiping components having a cumulative inlet and a cumulative outlet nominal cross section area, neither of which exceeds the nominal cross section area of the applicable exemption size, shall be exempted from the surface, volumetric, and visual VT-1 and VT-3 examination requirements of IWD-2500. The applicable exemption size shall be 4" NPS for all systems except the Class 3 Auxiliary Feedwater System, which shall be 1" NPS. This position is based on the interpretation of the clarification made by footnote 1 of IWD-1220 of the 1989 Addenda to ASME Section XI. For the purpose of applying this criteria, it is assumed that the intention of the Code is not to accumulate the cross-section areas of piping from different fluid systems entering a given component ( e.g., heat exchanger shell &
tube sides), nor the non-process piping (e.g.,
vents, drains and instrumentation).
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.
t ll
TITLE:
QUALITY APPENDIX B Section 1 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 32 ot 34 5.0 CLASS 1q CLASS 2~ AND CLASS 3 COMPONENT SUPPORTS 5.1 Basis for Preparation The following paragraphs (5.1 through 5.4) describe the Inservice Inspection Program for Class 1, 2, and 3 component supports for the First and Second Periods, up to and including the first outage ('93). The remainder of the Second and Third Periods shall be in accordance with Code Case N-491 ( for IWF-1000, -2000 and -3000) and IWF of the '86 Section XI Code.
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.'hese 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 paragraphs 2.2, 3.2 and 4.2 of this section shall be subject to examination during the third inspection interval.
The component supports requiring examination shall be as follows:
5.2.l.l 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.
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QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 33 of 34 5.2.1.2 Linear-Type 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 struts; beams and columns subjected to
, bending; trusses; frames; arches; rings; and c'ables.
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.
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.
TITLE: REV.
QUALITY APPENDIX B Section 1 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 34 of 34 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 identifies and incorporates the appropriate component supports into the Class 1, 2, and 3 sections 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.
PAGE QUALITY ASSURANCE MANUAL GINNA STATION Section 2 1 of 39 EFFECTIVE DATE:
December 31, 1993 ROCHESTER GAS & ELECTRIC CORPORATION SIGNATURE DATE TITLE: PREPARED BY APPENDIX B QUALITY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM FOR THE 1990-1999 INTERVAL ISI RELIEF REQUESTS APPROVED GY 1.0
Introduction:
General:
In accordance with 10CFR50.55a(g) (5) (iv), Rochester Gas
& 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 submitted to the Nuclear Regulatory Commission, as appropriate.
c TITLEI REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 2 of 39 ISI RELIEF RE UESTS TABLE I
SUMMARY
OP RHJFP R Rebel ASME Requiscment Reason for Proposed Request Section XI for Which Relief Relief Akemate Number Refesence Component is Requested Request Examination IWB.2500.1 RPV ShcU to Bange Volumetric Examination To perform aU examinations Perform aU examinations associated Cat. B.A WeM during two dUferent assoriatcd with the SheU.to with the SheUao Range at or near pcAods Bange during the same the end of the interval.
pcskxL IWB.2500.1 RPV Nozxlc to-Vcsscl Volumetric Examinadon To perform aU examinations Perform aU cxaminatkins associated Cat. B-D WeMs during two different assocktcd with the Nosxk.to. with the Noxxk to.Vessel weMs at or pcAods. Vessel wcMs during the same near the end of the lntervaL pcsloL IWA.1400 huthorixed Inspection Use of 'huthorixed Agency'cw York State has Use R. E. Cinna Quality Assurance Inspection Agency. not endorsed ASME Codes Pmgram.
and does not have an Authorixcd Inspection Agency.
4 (') IWB.2500 I Reactor Coolant Pump Volumetric Examinatkn of Pump materiai and Hydsostatie test, surface and visual Cat. B.L I Case WeMs and ease weMs and visual of conBgumtion. exams of outside surfaces.
B-L2 Intervals. intemab.
IWB-2500 I Chss I Valves Visual Examination of Excessive radiation Examine valve Intcmak when Cat. B M.2 Creatcr than NPS 4. valw intemds. ex posuse and historical disassembled for maintcnanee.
reliability of valves.
IWD 2500-1 Rsdioanive Waste Visual Bueucadon at Tank wUI be rendered Perform visual examination each Cat. D.B HoMup Tank, hydsostade prcssure. inoperatiw during tests. period at noAnal opcradng prcssure.
IWC 5222(a) Charging Pumps Visual Examination at Pumps have maximum Perform Hydro Test and visual at 3420 psig Hydrostatic pressure limit on the scab. 2400 psig.
Pressure.
IWC 5222(a) Valves PCV 430 Visual Exasninadon at Valve diaphragms cannot Hydrostatic Test to Fkx Connection and PCV 431C. Hydrostatic Pressure. withstand Test Prcssure. operate diaphragm pcr valve test requirements and pcrfonn Insesvice visual examination once per period.
IWC 5222(a) Secondary Side Visual Examination at RCBE adopted pressure Hydrostatk Test at of steam Hydrostatic Pressure. differential limitadon of 800 1.10 times instead of generator and psig to prevent primary 1.25 Psv setting and assoriatcd main side tube sheet chdding perform visual steam piping. scpamtiolL cxammatfoiu 10 (') IWD-5223(a) Standby Auxiliary Visual Examination Pressure Redunkn Bow Perform Inservice Pump RceiscuL at llydsostatie Pressure. OriTice requires removal and Visual Examination Une from AOV biank4sff. System piping docs once per periosL 9710A, AOV 9710B not provMe lsolatkn to and thek assoei. condcmate supply rank.
atcd downstream SigniTicant tank reduction Row orifices. wouM bc required for removal and is comidered impractkai.
IWD.5223(a) Boric Acid Filter and Visual Examinatkn at Test Prcssure required will Perform Inservice Visual Examination assoriatcd piping Hydrostatic Pressure. exceed limits for safe working once per period.
between Valves 347, pressure on Boric Acid Fil(cr 348A and 349A. Housing Fhnge Cmkets.
(v) ~ WillIDRAWN
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 3 of 39 ISI RELIEF RE UESTS TAB!X I
SUMMARY
OF RFJJFP R URIS Relief ASME Requirement Reason lor Proposed Request Sectfon XI for Which Rriief Relief Alternate Number Reference k Requested Request Examination 12.1 IWD-5223(a) hir Stan for Visual Examinatkn at AirStart Pressure Test wouM Perform Inseivke Examination Diesel Generator Hydrostatic Pressure. fcqiilfc tclndnatlon pfiof to once per period and once each htciuding fcacMng engine skid to quarter a prcssure decay test Receiver Tanks preclude ak to air stan performed oii air rcccivcl; and Piping. motors, kavtng portion of pfping untestable.
IWD-5223(a) Fuel Oil Transfer Visual Examinarion at Requires hohtkn of Diesel Perform system Functional Tesring Pumps and assoriated Hydrostatic Pressure. Oil Storage and Dry Tank with assockted Visual plpuig to tert(dr(all at where no means of kola(ion Examination once per period.
Oil Storage Tank k provided at Transfer pump discharge piping with tanks vented to atmosphere.
IWD.5223(a) Jacket Cooling Visual Examination at Cooling Water Expansion Perform System Functional Testing Water System and Hydrostatic Prcssure. Tank vented to aunosphcre with associated Vkual Examination associated piping requiring isoktkn with vast once pcf pcf iud.
to terminals at piping Involved willbe unable Water Ex panskn to prcssurixc Tanks.
13 (111) Table IV C2500 Ckss 2 Piping Under Revision.
I, Category CH Pen((rath(3 containment vessel with balance of piping system outskle Section XI scope.
Ckss 3 Ponion of the Visual Examination at I lydrostatic Tesung is Perform Inservice Visual Examination servke water system Hydrostatic Pressure.. impra(tical due to system once per period.
design that is openwnded and empkying butterily valves that wcfc not dcsigncd to provide a kaktight boundary.
15 (') IWC.2500 Qass 2 Piping Piping WeMs c At a mhdmum terminal Surface and Volumetric Examinations Cat. C F-I WeMs c wall thickness do 3/8'ominal connection weMs of idcntilied on tcmunal connection weMS of
& C F-2 wall thkknem Men(ipcd exempted weMS shall be 3/8'ondnal no( require surface and exempted weMs pcr Items l(ems C5.10 for piping > NPSC. volumetric examinanons. C5.10 and C5.50 of Tabk performed to the requirements of and C5.50 resp. IWC2500 shouM be IWC.2500 1.
examinetk IWB-2500 Qass I & 2!ntcgral Volumetric or surface I ntegtul Attachments on Ckss On Ckss I & 2 Integral hnachmcnts, Cat. 8 K.l Item Attachment on piping examination on Qass I I & 2 suppon attachments a surface examination shall be B10.10 and IWC. speciTically support Integral Attachments is shoUM have a sUffacc performed on suppon attachments to 2500, Cat. CC, Attachments. required on Base examination to insure safety IWB and IWC-2500 requirements.
l(em C3.20 Attachment > 5/8. Qass and system integrity.
2 requires a sUrfacc examination on Base Attachments > 3/0'.
I9 IVC2500 Pulsation Dampener, Applicable Surface and/or Physical configuration not Perform Surface and/or Volumetric Category C B, Nossk WeM & Inside Volumetric Examinations. conducive for full Examination to Maximum extent Items Col & Radius. Examination coverage. practkaL C2.22
(') "
(1 ~ 1) ~
'4 N
'I 1
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE, R.E. GINNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 4 of 39 ISX RELIEF RE UESTS TABIX 1
SUMMARY
OP RHIKF mSlS Relief ASME Requirement Reason for Proposed Request Section XI for Which Relief Relief Alternate Number Reference Component is Requested Request Examination 23 (~) IWAA400 Chss 3 Valves: llydrostatie Testing of Code Impraetieality ASME Section III NDE with PT of V~
VMII Rephcement Activities. Root Pass. Perform a VT.2 leak Test. This Relief is a one-time'se VMI 3 only.
V~
V~
VA738 V.4739 VA760 24 (~) IWA-4400 Class 3 Vrdvcst Ifydrostatic Testing of Code Impractkality ASME Section III NDE with PT of VMI3 Re placement Activities. Root Pass. Pcrfonn a VT 2 leak V.4027 Test. This Relief is a one-time V 4020 usc only.
V~
(~) ~ ONE TIME USE ONLY.
t A
I
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 5 of 39 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.
I TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 6 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 2 DEFER RPV NOZZLE EXAMINATIONS TO END OF INTERVAL Components for Which Relief is'equested:
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. "If 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 same is highly desirable to perform both examinations at the time.
III. Proposed Alternate Method:
Rochester Gas & Electric (RG&E) 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.
It II't I
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT
'ANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE'990-1999 INTERVAL 7 of 39 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO ~ 3 U
USE OF AUTHORIZED INSPECTION AGENCY Examination Requirement for Which Relief is Requested:
The ASME Boiler and Pressure Vessel Code,Section XI, 1986 Edition, IWA-1400(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, Ginna 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.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 8 of 39 ISI RELIEF REQUESTS WITHDRAWN 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 Nuclear 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.
I t
t II t
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 9 of 39 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 4 RCP CASING WELDS VOLUMETRIC AND INTERNAL EXAMINATION 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 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 (SA351 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,'he 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!'CP during the Spring 1981 refueling outage. In addition, the same type of examination has been performed at several other sites.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 10 of 39 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).
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 11 of 39 ISI RELIEF RE UESTS 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 considered:
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.
co Perform a visual examination (VT-3) of the internal surfaces each time pump disassembly is required for maintenance.
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 12 of 39 ISI RELIEF RE UESTS WITHDRAWN 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 of the structure.
history (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: In making this assessment, thermal aging embrittlement and any other processes which may degrade the properties of the pump casing during service will be considered.
jV P
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 13 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 5 VISUAL INTERNAL EXAMINATION OF CLASS 1 VALVES Components for Which Relief is Requested:
Class 1 valves requiring valve body internal VT examination.
Size
~Xn. Valve No. MFG 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 Vela n/Check 6A-RC-2501-B 852A Velan/Gate 6A-RC-2501-A 852B Velan/Gate 6A-RC-2501-B II. 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:
to complete the subject examination, unnecessary expenditures of man-hours and manrem are required with essentially no compensating increase in plant safety, and
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 14 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 5 (Con't)
- 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:
A's 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.
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 15 of 39 ISI RELIEF REOUESTS WITHDRAWN RELIEF REQUEST NO. 6 HYDROSTATIC TESTING OF RADIOACTIVE WASTE HOLDUP TANK 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.
II. 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 unavailable 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.
f C
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 16 of 39 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. RGGE 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 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR, THE 1990-1999 INTERVAL 17 of 39 ISI RELIEF RE UESTS RELIEF REQUEST N0.7 REDUCED HYDROSTATIC PRESSURE TESTING OF CHARGING PUMPS DISCHARGE PIPING 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.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GXNNA STATXON FOR THE 1990-1999 INTERVAL 18 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 8 ALTERNATIVE TESTING OF RCS OVERPRESSURE PROTECTIONS NITROGEN ACCUMULATOR SYSTEM OF VALVES PCV 430 & PCV 431C X. Components for Which Relief is Requested:
RCS Overpressure Protection Nitrogen Accumulator System Valves PCV 430 and PCV 431C IX. 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.
l XIX. Proposed Alternate Method:
The RCS overpressure nitrogen accumulator syst: em 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.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 19 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO ~ 9 REDUCED HYDROSTATIC TEST OF THE STEAM GENERATOR SECONDARY SIDE 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 200oF (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
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL XNSERVICE INSPECTION PROGRAM PAGE GXNNA STATION FOR THE 1990-1999 INTERVAL 20 of 39 XSI RELIEF RE UESTS RELIEF REQUEST NO. 9 (con't)
A letter was submitted to Dennis L. Ziemann. Chief Operating Reactor Branch ¹2, 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 ¹1, 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 3 ASSURANCE R.E- GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 21 of 39 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 10 ALTERNATIVE HYDROSTATIC TESTING OF STANDBY AUXILIARY PUMP RECIRCULATION LINE I. Component for Which Relief is Requested:
Feedwater, Standby Auxiliary Pump recirculation line between AOV 9710A, AOV 9710B and their associated downstream flow orifices.
I. 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 syst: em 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.
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 installed. 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.
v a
TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 22 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 11 ALTERNATIVE HYDROSTATIC TESTING OF BORIC ACID FILTER &
ASSOCIATED PIPING I. 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.
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.
1 TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 23 of 39 ISI RELIEF RE UESTS RELZEF REQUEST NO 12 ALTERNATIVE HYDROSTATIC TESTING OF THE EMERGENCY DIESEL GENERATOR COMPONENTS & ASSOCIATED PIPING 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.
II. 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 syst: em 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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TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 24 of 39 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 XI. 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.
11 TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA 'NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 25 of 39 ISI RELIEF REOUESTS RELIEF REQUEST NO. 13 I
ALTERNATIVE PRESSURE TESTING OF CONTAINMENT PENETRATION PIPING Under Revision.
I' TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVXCE XNSPECTXON PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 26 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO 14 ALTERNATIVE HYDROSTATIC TESTING OF THE CLASS 3 SERVICE WATER SYSTEM 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.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 27 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 14 (Con't)
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 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL XNSERVICE XNSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 XNTERVAL 28 of 39 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO. 15 CLASS 2 PIPING < 3/8", AUGMENTED EXAMINATION Component for Which Relief is Requested:
Class 2, IWC 2500-1 Table Examination Category C-F-l and C-F-2, Items C5.l0 and C5.50.
ASME Requirements for Which Relief is Requested:
Category C-F-l 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 100% 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,XWC 2500-1 Table, Category C-F-l and C-F-2, Xtems C5.10 and C5.50 respectively. These examinations are identified in the Class 2 Allocation Tables as Augmented Examinations and also are included under the category C-F-l and C-F-2.
In the Program Plan Tables (Supplement 1 to Appendix B).
These are identified as C-F-l or C-F-2 followed by "-- ".
These components are also noted in the instruction field.
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TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GXNNA NUCLEAR POWER PLANT MANUAL INSERVICE XNSPECTXON PROGRAM PAGE GXNNA STATION FOR THE 1990-1999 INTERVAL 29 of 39 ISI RELIEF RE UESTS WITHDRAWN RELIEF REQUEST NO 16 CLASS 1 & 2 INTEGRAL ATTACHMENTS'UGMENTED EXAMINATIONS 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, Integral Attachments on piping as indicated in IWB-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, Xtem 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.
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TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 30 of 39 ISI RELIEF REOUESTS RELIEF REQUEST NO. 19 PULSATION DAMPENER'LTERNATIVE EXAMINATIONS FOR NOZZLE WELDS &
INSIDE RADIUS
'Components for Which Relief is Requested:
Charging System, Class 2 Pulse Dampener nozzle inside radius and nozzle welds examination volumes per Category C-B, Items C2. 21 and C2. 22.
ASME Requirements for Which Relief is Requested:
Within Table IWC-2500-1, Category C-B provides examination requirements for Class 2 pressure retaining nozzle welds in vessels. The Pulse Dampener contains three (3) nozzles, in line, located at. the bottom of the unit. These nozzles require both surface and volumetric examinations as specified by Item C2.21 for nozzle-to-shell welds. The same three (3) nozzles also require volumetric examination as identified by Item C2.22 for the nozzle inside radius sections. ASME Section XI requires "essentially 100> of the weld length" as specified within the Notes within the IWC-2500-1 tables. ASME Section XI Code .Case N-460 states that if the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in coverage is acceptable provided that the coverage is less than 10>.
R. E. Ginna Nuclear Power Plant was constructed to B31.1, 1955 edition. This code did not contain requirements to ensure that items be accessible for future examinations.
The Pulse Dampener was constructed and installed in the early 1970's, the construction code utilized did not provide provisions for accessibility for ISI NDE.
The Pulse Dampener contains three (3) nozzles that require examination under ASME Section XI, 1986 Edition, no addenda per Category C-B. These nozzles are identified as CF-N1, CF-N2 and CF-N3. The three (3) Nozzle-to-Shell welds require both surface and volumetric examinations as specified by Item C2.21. The same nozzles are also examined volumetrically for the Nozzle Inside Radius Section per Item C2.22. The design of the Pulse Dampener is not conducive for examination of the nozzles as identified within the above item numbers. The outboard
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TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE XNSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 31 of 39 ISI RELIEF RE UESTS RELIEF REQUEST NO. 19 (Con't) nozzle is identified as CF-N1. Between this nozzle and the middle nozzle (identified as CF-N2) is a support that covers the edge of one nozzle weld heat affected zone.
From the middle nozzle to the third nozzle (identified as CF-N3) is a space of only 7/8" from the edge of one nozzle weld heat affected zone to the edge of the other nozzle heat affected zone. Due to the identified interferences of the nozzles and vessel support, the associated volumetric and surface volume coverage acceptance criteria is below the stated value within Code Case N-460. The following table summarizes the coverage obtained using the indicated examination methods.
Nozzle Weld NDE Method of Covera e CF-N1 PT 664 UT 65%
CF-N2'T UT 664 65~o CF-N3 PT >90~o UT 80<
Since Section XI requires examination of this ASME component and since itCode is the only one of its type, style and function; the above identified surface and volume coverage should be acceptable in meeting code requirements.
XI. Proposed Alternate Method:
Rochester Gas and Electric (RG&E) proposes that the NDE surface and volume coverage identified within the table above be acceptable in fulfilling the code required examinations. The actual physical configuration of the component being examined is not conducive in obtaining the requirements specified within Code Case N-460.
In addition, RG&E proposes to perform a VT-2 visual examination on the entire Pulse Dampener during a leakage test and hydrostatic pressure test in accordance with IWA-5000 and Table IWC-2500-1, as applicable.
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TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 32 of 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO ~ 23 ALTERNATIVE CLASS 3 HYDROSTATIC TESTING FOR CODE REPLACEMENTS Components for Which Relief is Requested:
A. Service Water System, Rules for Hydrostatic Testing of Repairs, Replacements or Modifications to pressure retaining Class 3 components.
Installation of isolation valves a ~ V-4611
- b. V-4613 C. V-4626
- d. V-4669
- e. V-4738
- f. V-4739 g V-4760 B. Turbine Driven Auxiliary Feedwater System, Rules for Hydrostatic Testing of Repairs, Replacements or Modifications to pressure retaining Class 3 components.
Installation of valve V-4023 II. ASME Requirements for Which Relief is Requested:
Hydrostatic testing of Repairs, Replacements or Modifications on Class 3 systems is required by IWA-4400 which specifies that hydrostatic testing shall be performed to IWD-5223(a). The system hydrostatic test pressure shall be at least 1.10 times the system pressure P, for systems with Design Temperature of 200'F or less, and at least 1.25 times the system pressure P, for systems with Design Temperature above 200'F. The system pressure P, 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 P,.
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TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTXON PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 33 of 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 23 (Con't)
IXI. Basis:
The first activity to be performed is the replacement of seven (7) valves in the Class 3 Service Water System. The second activity is the replacement of a check valve in the Class 3 Turbine Driven Auxiliary Feedwater System.
Service Water S stem:
Seven (7) valves within the service water system are currently being replaced: valve numbers 4611, 4613, 4626, 4669, 4738, 4739, and 4760. The replacement of the valves is being performed to the 1986 edition of ASME B&PV Code, Section III.
An evaluation to establish the work area test boundaries for the valve replacements was performed as discussed below (refer to drawings 33013-1250 sheets 1-3):
V-4626 The replacement valve is a gate valve. Both the upstream and downstream isolations would require the use of butterfly type isolation valves, thus making an elevated (hydrostatic) pressure test impractical. (see Relief Request g14)
V-4611/ V-4669/ V-4738/ V-4739/ V-4760/ V-4613:
The first five replacement valves are gate valves and serve as loop cross-connects. V-4613 is a butterfly type design and serves as a loop isolation valve. In order to isolate the selected areas for replacement, freeze seals would be required to maintain some systems supplied by the Service Water System operational. The use of freeze seals as isolation boundaries is not feasible. In order to establish test boundaries, complete service water loops would be required to be isolated, thus creating an operational hardship resulting in a reduction in plant safety due to rendering certain critical operational and safety related equipment unavailable.
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TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 34 of 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 23 (Con't)
Turbine Driven Auxiliar Feedwater S stem:
The work activity being performed is the replacement of check valve 4023 in the recirculation test line. The Code used for this work is the 1989 edition of ANSI B31.1. Once the valve is installed, the pipe section cannot be isolated to perform the elevated (hydrostatic) pressure test.
Referring to drawing 33013-1237, boundaries for the work area would be the pump suction valves (V-4024 & 4098), the pump discharge valve (V-3996) and downstream flow control valve V-4291. Both V-3996 and V-4291 are not designed for isolation applications, and would require the installation of test blocks. Installation of these blocks was found to be impractical.
Part of the work area would incorporate piping having a design pressure of 150 psi, with a required test pressure of 2304 psi, therefore, an unacceptable condition would exist. Installation of a test block at the discharge of the pump was reviewed and found to be impractical because disassembly of the pump would be required.
IV. Proposed Alternate Method:
Alternative Pressure Test Requirements for Welded Repairs or Installation of Replacement Items by Welding:
Class 3:
Using ANSI B31.1 and ASME B&PV Code Case N-416 as a basis for relief, Rochester Gas & Electric considers the following alternative requirements to be acceptable:
NDE shall be performed in accordance with the methods and acceptance criteria of Subsection ND of the 1986 Edition of Section III. Additional NDE also is performed by dye penetrant testing of the root pass weld for the service water valves. For V-4023, a surface examination will be performed.
TITLB REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 35 of 39 1l ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 23 (Con't)
- 2. Prior to, or immediately upon return to service, a VT-2 visual examination shall be performed in conjunction with an inservice or functional leakage test, using the 1986 Edition of Section XI, in accordance with IWA-5000, at nominal operating pressure and temperature.
This code required test pressure and the nominal system pressure at which a leak test will be performed as proposed are shown in the table below:
Code Required Test Pressure to SWS Valves Test Pressure be Performedf'~
V-4611 165 psig 75 psig (Nom.)
V-4613 165 psig 75 psig (Nom.)
V-4626 165 psig 75 psig (Nom.)
V-4669 165 psig 75 psig (Nom.)
V-4738 165 psig 75 psig (Nom.)
V-4739 165 psig 75 psig (Nom.)
V-4760 165 psig 75 psig (Nom.)
Test pressure reflects system operating pressure of nominal 75 psig.
The test pressure (and nominal operating pressure) is considered adequate to detect potential leakage after performing the replacement.
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TITLE REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 36 of 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 23 (Con't)
Code Required Test Pressure to AFW Valve Test Pressure be Performed V-4023 2304 psig 1100 psig (Nom.)
(2) Test pressure reflects system operating pressure of nominal 1100 psig.
- 3. Use of this Relief Request will be documented on the NIS-2 Form for the Replacement.
N TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVXCE INSPECTXON PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 37 0( 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO 24 ALTERNATIVE CLASS 3 HYDROSTATIC TESTING FOR CODE REPLACEMENTS Components for Which Relief is Requested:
Service Water System, Rules for Hydrostatic Testing of Repairs, Replacements or Modifications to pressure retaining Class 3 components.
Installation of isolation valves a ~ V-4663
- b. V-4013 c ~ V-4027
- d. V-4028 ASME Requirements for Which Relief is Requested:
Hydrostatic testing of Repairs, Replacements or Modifications on Class 3 systems is required by IWA-4400 which specifies that hydrostatic testing shall be performed to XWD-5223(a). The system hydrostatic test pressure shall be at least 1.10 times the system pressure P, for systems with Design Temperature of 200'F or less, and at least 1.25 times the syst: em pressure P, for systems with Design Temperature above 200'F. The system pressure P, 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 P y.
TITLE: REV.
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 38 of 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 24 (Con't)
III. Basis:
This request for relief is for a "one-time" use for the replacement of four (4) Class 3 service water valves. The four valves being replaced are valve numbers 4663, 4013, 4027, and 4028. The replacement represents a previously unscheduled activity attributable to defects encountered during a refurbishment program. The replacement of the valves is being performed to the 1986 edition of ASME B&PV Code,Section III.
An evaluation to establish the work area test boundaries for the valve replacements was performed as discussed below (refer to drawings 33013-1250 sheet 3 for V-4663 and 33013-1237 for V-4013, V-4027, and V-4028)
All replacement valves are gate valves. Isolation boundaries for the valves would require the use of butterfly valves, making an elevated (hydrostatic) pressure test impractical. (see Relief Request g 14)
IV. Proposed Alternate Method:
Alternative Pressure Test Requirements for Welded Repairs or Installation of Replacement Items by Welding:
Class 3:
Using ANSI B31.1 and ASME B&PV Code Case N-416 as a basis for relief, Rochester Gas & Electric considers the following alternative requirements to be acceptable:
NDE shall be performed in accordance with the methods and acceptance criteria of Subsection ND of the 1986 Edition of Section III. Additional NDE also is performed by dye penetrant testing of the root pass weld.
TITLE:
QUALITY APPENDIX B Section 2 3 ASSURANCE R.E. GINNA NUCLEAR POWER PLANT MANUAL INSERVICE INSPECTION PROGRAM PAGE GINNA STATION FOR THE 1990-1999 INTERVAL 39 ot 39 ISI RELIEF RE UESTS ONE-TIME USE ONLY RELIEF REQUEST NO. 24 (Con't)
- 2. Prior to, or immediately upon return to service, a VT-2 visual examination shall be performed in conjunction with an inservice or functional leakage test, using the 1986 Edition of Section XI, in accordance with IWA-5000, at nominal operating pressure and temperature.
This code required test pressure and the nominal system pressure at which a leak test will be performed as proposed are shown in the table below:
Code Required Test Pressure to SWS Valves Test Pressure be Performed V-4663 165 psig 75 psig (Nom.)
V-4013 165 psig 75 psig (Nom.)
V-4027 165 psig 75 psig (Nom.)
V-4028 165 psig 75 psig (Nom.)
Test pressure reflects system operating pressure of nominal 75 psig.
The test pressure (and nominal operating pressure) is considered adequate to detect potential leakage after performing the replacement.
- 3. Use of this Relief Request will be documented on the NIS-2 Form for the Replacement.
PAGE QUALITY ASSURANCE MANUAL GINNA STATION Section 3 1 of 3 EFFECTIVE DATE:
December 31, 1993 ROCHESTER GAS 5 ELECTRIC CORPORATION SIGNATURE DATE TITLE PREPARED BY:
1~ "1 f3 APPENDIX B QUALITY R. E. GINNA NUCLEAR POWER PLANT ASSURANCE INSERVICE INSPECTION PROGRAM I 8-z89 FOR THE 1990-1999 INTERVAL P&ID BOUNDARY DRAWINGS APPROVED Y: l~ C aha g-.y3
1.0 General
The P&ID 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 P&ID 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 P&ID BOUNDARY DRAWINGS The following list identifies the P&ID drawings that contain ASME Class 1, 2, 3, or High Energy pressure boundaries.
Drawin Number Revision 33013-1231 Rev. 22 33013-1232 Rev. 9 33013-1236 Sheet Rev. 8 33013-1236 Sheet Rev. 6 33013-1237 Rev. 27 33013-1238 Rev. 9 33013-1239 Sheet, Rev. 11 33013-1239 Sheet Rev. 9 33013-1245 Rev. 18 33013-1246 Sheet Rev. 7 33013-1246 Sheet Rev. 6 33013-1247 Rev. 20 33013-1248 Rev. 17 33013-1250 Sheet 1 Rev. 13 33013-1250 Sheet 2 Rev. 14 33013-1250 Sheet 3 Rev. 9 33013-1258 Rev. 14 33013-1260, Rev. 15 33013-1261 Rev. 21 33013-1262 Sheet Rev. 10 33013-1262 Sheet Rev. 5
(*) 33013-1263 Rev. 7 33013-1264 Rev. 14 33013-1265 Sheet Rev. 6 33013-1265 Sheet Rev. 5 33013-1266 Rev. 16 33013-1270 Sheet 1 Rev. 5 33013-1272 Sheet 1 Rev. 6 33013-1272 Sheet 2 Rev. 5
(*) 33013-1275 Sh eet 1 Rev. 2 33013-1275 Sheet 2 Rev. 2 33013-1277 Sheet 1 Rev. 6
(*) 33013-1278 Sh eet 1 Rev. 8
(*) 33013-1278 Sh eet 2 Rev. 9 33013-1279 Rev. 9 33013-1863 Rev. 11 33013-1865 Rev. 8 33013-1866 Rev. 13 33013-1870 Rev. 8
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 3 of 3 P&ID BOUNDARY DRAWINGS Drawin Number Revision 33013-1882 Rev. 9 33013-1886 Sheet 2 Rev. 5 33013-1887 Rev. 5
(*) 33013-1893 Rev. 10 33013-1908 Sheet 3 Rev. 3 33013-1915 Rev. 9 33013-1991 Rev. 8
(*) 33013-2278 Rev. 0 33013-2279 Sheet 1 Rev. 0 Note: (*) = Identifies P&IDs containing ASME Class piping NPS 1" and below. These identified drawings are not included within this section.
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