ML20199G566
| ML20199G566 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 01/06/1998 |
| From: | Anderson E, Penny R, Alexandra Smith POWER AUTHORITY OF THE STATE OF NEW YORK (NEW YORK |
| To: | |
| Shared Package | |
| ML20199G519 | List:
|
| References | |
| [[::JAF-ISI-0002|JAF-ISI-0002]], JAF-ISI-2, NUDOCS 9802040298 | |
| Download: ML20199G566 (273) | |
Text
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l JAMES A. FITZPATRICK NUCLEiR POWER I*IANT i
l THIRD INSERVICE INSPECTION INTERVAL
- INSERVICE INSPECTION PROGRAM -
1 i
COMMERCIAL SERVICE DATE: JULY 28,1975 t DOCUMENT No. [[::JAF-ISI-0002|JAF-ISI-0002]] l
l l
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- NewYorkPbwer 4 Authority i
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, NEW YORK POWER AUTHORITY 123 MAIN STREET WIIITE PLAINS, NEW YORK 10601 i
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e3 V 3 h@@""" Eg3NTROLLED New York Power Authority 123 Main Street White Plains, New York 10601 THlRD INSERVICE INSPECTION INTERVAL INSERVICE INSPECTION PROGRAM Prepared Fot James A. Fitzpatrick Nuclear Power Plant P.O. Box 41 Lycoming, New York 13093 Commercial Service Date: July 28,1975 NRC Docket Number: 50-333 Document Number: [[::JAF-ISI-0002|JAF-ISI-0002]] l
Revision Number: 0 Date: January 6,1998 Prepared by Wd old Edw ,
rd L. -ISI Consultant Reviewed by: -
. <4c #a Aturo DTmith -ISl Technical Specialist Approved by: !
R6bert y- ' ector Engineering Programs Reviewed by: -
_m. p Authorized Nuclear Inservice inspector
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.m New York Power Authority 123 Main Street White Plains, New York 10601 THIRD INSERVICE INSPECTION INTERVAL INSERVICE INSPECTION PROGRAM Prepared For
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. .Y James A. Fitzpatrick Nuclear Power Plant P.O. Box 41 Lycoming, New York 13093 Commercial Service Date: July 28,1975 NRC Docket Number: 50-333 Document Number. [[::JAF-ISI-0002|JAF-ISI-0002]] Revision Number: 0 Date: January 6,1998 Prepared by h///9 oM Edw rd L. rs - ISI Consultant Reviewed by: . ne ma Aturo DTmith -ISI ochnical Specialist Approved by: !
R6beit .y- ' ector Engineering Programs Reviewed by:
Authorized Nuclear Inservice inspector
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Y JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] i
NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 12 of 1 11 TABLE OF CONTENTS i
Table of Contents . . . .. . . 1-2 Record of Revision . . . . 1-4 Abbreviations . . . .. . . . .... . 1-5 Glossary of Terms . . . . . . 1-8 Abstract . . . . 1-11 Section 1.0 Introduction . . . 1 1 thru 1-11 Section 2.0 ASME Code Class 1 Systems / Components . 2-1 thru 2-16 in i Section 3.0 ASME Code Class 2 Systems / Components . 3-1 thru 3-11 Section 4.0 ASME Code Class 3 Systems / Components . 4-1 thru 413 ,
Section 5.0 ASME Code Class 1,2, and 3 Component Supports . 5-1 thru 5-5 Section 6.0 Augmented Examinations 6-1 thru 6-12 Section 7.0 Requests Fr.r Relief . 7-1 thru 7-4 Section 8.0 Acceptance Critena 8-1 thru 8-13 Section 9.0 ASME Repairs and Replacements . . 9-1 thru 9-9 Section 10 0 Records .. .. . 10-1 thru 10-13 LIST OF FIGURES Figure 6-5 Feedwater Nozzle Zones . . .6-9 Figure 10-1 Form NIS-1 Owner's Data Report for Inservice Inspection . . 10-5 Figure 10-2 Form NIS-2 Owner's Report for Repairs or Replacements . . 10-7 Figure 10-3 OAR-1 Owner's Activity Report . 10-9
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k Figure 10-4 Form NIS-2A Repair / Replacement Plan Certification Record . 10-13 File: SECT 00.TXT.E1
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1 JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT l
- 5 NewWrkPower
& Authority --
G THIRD INSERVICE INSPECTION INTERVAL INSERVICE INSPECTION Rev. O Date: January 6,1998 ,
PROGRAM l Page 13 of 1 11 l l
1 LIST OF TABLES j Table 1-2 JAF Inservice inspection Periods . . 1-4 I 1
Table 16 Inspection Program B . . . 1-10 Table 6-1 Applicable RPV Augmented Welds . . . . .. 6-3 Table 6-2 IGSCC Examination Requirements . . . . 6-7 Table 6-3 Main Steam and Feedwater Augmented Examinations .6-8 Table 6-4 Augment 6a Feedwater Nozzle Examinatic? . .. .6-9 Table 6-5 In-Vessel Augmented Examinations . 6-10 Table 64 Core Spray Augmented Weld Examinations . . . . 6-12 Table 8-1 Class 1 Acceptance Standards . 8-12 Table 8 2 Class 2 Acceptance Standards . . 8-12 Table 8-3 Class 3 Acceptance Standards . . B-13 Table 8-4 Class 1,2,3 Component Supports Acceptance Standards 8-13 Table 8-5 Class MC Acceptance Standards . . 8-13 APPENDICES Appendix A Class 1 Summary Tables . . A-1 thru A-37 Appendix B Class 2 Summary Tables . . B-1 thru B-29 Appendix C Class 3 Summary Tables . . . C-1 thru C-13 Appendix D Class 1,2 and 3 Support Summary Tables D-1 thru D-10 Appendix E Augmented Examination Summary Tables . . . E-1 thru E-6 Appendix F Relief Requests . . .. . . F-1 thru F-55 t]
~J Appendix G Code Boundary Classification Diagrams . G-1 thru G-2 File: CECT00.TXT.E1
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JAMES A. FITZPATRICK JAF ISI-0002 i NUCLEAR POWER PLANT tv THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 l PROGRAM
,Page 1-4 of I-11 I
RECORD OF REVISION
! REVISION); ,
' [ DATED $ yAFNTE.6bl fREASUA FdhEVISIUN); <
, l No. ! '
iM :PAGESW 0 January 6,1998 Entire Updated Inservice !n@ection Program Plan Docum9nt for the 3* Ten Year Inservice Inspection Interval 1
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File: SECT 00.TXT.E1
JAMES A.FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT nntYh THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page I-5 of I-11 l ABBREVIATIONS Listed below are the abbreviations utiltzed in this document:
ANil . Authorized Nuclear Inservice Inspector _
ANSI American Nuclear Standard Institute ASME American Society of Mechanical Engineers B&PV Boiler & Pressure Vessel Code BWROG Boiling Water Reactor Owner's Group CFR Code of Federal Regulations CRC Corrosion Resistant Cladding CRD Control Rod Drive System CRS Code Required Surface CRV Code Required Volume CS Core SF ay System DPI Drywellinerting CAD and Purge System FSAR Final Safety Analysis Report FW Feedwater System GE General Electric GL Generic Letter HPCI High Pressure Coolant injection System IE Inspection and Enforcement IHSI Induction Heat Strest Improvement
/~S. ISI Inservice Inspection
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v File: SECTdo.TXT,El
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE IPSPECTION Date: January 6,1998 PROGRAM Page i-6 of i-11 ABBREVIATIONS (Continued)
IST Inservice Testing IWI in-Vessel Visual Inspections JAFNPP James A. Fitzpatrick Nuclear Power Plant MS Main Steam System N/A Not Applicable NBVI Nuclear Boiler Vessel Instrumentation NDE Nondestructive Examination NPS Nominal Pipe Size 1
( NYPA New York Power Authority P&lD Piping and Instrumentation Diagram
! REF Refueling Outage and Year l
Reactor Water Recirculation System RC l RCIC Reactor Core Isolation Coolant S) stem I
i R.G. Regulatory Guide l
RHR Residual Heat Removal System RHSI Resistant Heat Stress improvement RPV Reactor Pressure Vessel RWC Reactor Water Cleanup System SURF Surface Examination l
l St Stress improvement T.S. Technical Specifications l f3 l () UT Ultrasonic Examination i USNRC United States Nuclear Regulatory Commission l
File: SECT 00.TXT.E1
l i
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] 1 NUCLEAR POWER PLANT i
- > NewWrkPower 4# Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM
, Page 17 of I 11 ABBREVIATIONS (Continued)
VOL Volumetric Examination
'VOL Ultrasonic Refracted Longitudinal Examination
. Ultrasonic 0 degree Examination Only VT Visual Examination (suffix number denotes type of exam, (VT-1, VT-2, VT-3))
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Fb: SECT 00.TXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > New%rkPower
& Authority . l THIRD INSERVICE ..sPECTION INTERVAL Rev. O I INSERVICE INSPECTION Date: January 6,1998 PROGRAM !
Page I8 of I-11 i GLOSSARY OF TERMS 1
ASSESS - to determine by evaluation of data compared with previously obtained data such as vrating data or design specifications.
AUTHORIZEb INSPECTION AGENCY - an organization that is empowered by an enforcement authority to provide inspection personnel and services as required by Section XI.
AUTHORIZED NUCLEAR INSERVICE INSPECTOR a person who is employed and has been qualified by an authorized inspection Agency to verify that examinations, tests and repairs (that do not include welding or brazing) are performed in accordance with the rules and requirements of Section XI.
AUTHORIZED NUCLEAR INSPFCTOR - an employee of an authorized Inspection Agency who has been qualified in accordance with NCA-5000 of Section Ill.
COMPONENT - an item in a nuclear power plant such as a vessel, pump, valve or piping system.
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/~N COMPONENT SUPPORT - a metal support designed to transmit loads from a component to the load-carrying building or foundation structure. Component supports include piping supports and encompass those structural elements relied upon to either support the weight or provide structural stability to components.
CONSTRUCTION - an all-inclusive term comprising materials, design, fabrication, examination, testing, inspection and certification required in the manufacturer and installation of items.
CONSTRUCTION CODE - the body of technical requirements that governed the construction of the item.
ENFORCEMENT AUTHORITY - a regional or local governing body, such as a State or Municipality of the United States or a Province of Canada, empowered to enact and enforce Boiler and Pressure Vessel Code legislation.
ENGINEERING EVALUATION - an evaluation of indications that exceed allowable acceptance standards to determine if the margins required by the Design Specification and the Construction Code are maintained.
EVALUATION - the process of determining the significance of examination or of test results, including the comparison of examination or test results with applicable acceptance criteria or previous results.
EXAMINATION CATEGORY - a grouping of items to be examined or tested.
INSERVICE EXAMINATION - the process of visual, surface, or volumetric examination performed in
]O accordance with the rules and requirements of Section XI.
Fde: SECT 00.TXT Ei
I l
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT 1 I
- > NewWrkPower er Authortty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERV!CE INSPECTION Date: January 6,1998 PROGRAM Page i-9 of I-11 GLOSSARY OF TERMS (Continued)
INSERVICE INSPECTION - methods and actions for assuring the structural and pressure-retaining integnty of safety-related nuclear power plant components in accordance with the rules of Section XI.
INSERVICE TEST - a test to determine the operational readiness of a component or system.
INSPECTION - verification of the performance of examinations and tests by an It'spector.
INSPECTION PROGRAM- the plan and schedule for performing examinations or tests.
INSPECTOR - an Authorized Nuclear Inservice Inspector, except for those instances where so designated as Authorized Nuclear Inspector.
INSPECTION INTERVAL - a duration of time,10-years.
INSPECTION PERIOD - a duration of time within an inspection interval, as determined by Plant Technical Specifications and/or Inspection Program B of Section XI.
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Q ITEM - a material, part, appurtenance, piping sub-assembly, component or component support.
MAINTENANCE - routine servicing or work undertaken to correct, adjust or prev'ent an abnormal or unsatisfactory condition.
NONDESTRUCTIVE EXAMINATION - an examination by the visual, surface or volumetric method.
OPERATIONAL READINESS - The abihty of a component or system to perform its intended function when required.
OWNER - the organization legally responsible for the operation, maintenance, safety and power generation of the nuclear power plant.
REGULATORY AUTHORITY - a federal govemment agency, such as the United States Nuclear Regulatory Commission, that is empowered to issue and enforce regulations affecting the design, construction, and operation of nuclear power plants.
SUBSEQUENT PERIOD - is the next following period, even if it is in the following interval.
SUPPORT - (1) an item used to position components, resist gravity, resist dynamic loading, or maintain equilibrium of componentE (2) an item that carries the weight of a component or piping from below with the supporting members being mainly in compression.
STRUCTURAL DISCONTINUITY - As used in this program: includes pipe weld joints to vessel nozzles, fm valve bodies, pump casings, pipe fittings (puch as tees, elbows, reducers, flanges, etc.
) conforming to ANSI B16.9) ant! pipe branch connections and fittings.
Fde: SECT 00.TXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > NewYorkPower 4# Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page I 10 of I 11 I
GLOSSARY OF TERMS (Continued)
TERMINAL ENDS the extremities of piping runs that connect structures, components, or pipe anchors, each I of which acts as a rigid restraint or provides at least 2 degrees of restraint to piping thermal l expansion.
TEST - a procedure to obtain ir. formation through measurement or observation to determine the operational readiness of a component or system while under controlled conditions.
VERIFY - to determine that a particular action has been performed in accordance with the rules and requirements of Section XI either by witnessing the action or by reviewing records.
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File: SECT 00 TXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > NewWrkPower
& Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 1 - 11 of I-11 ABSTRACT This document describes the Updated inservice Inspection Prog am for the Third Ten-Year Inservice inspection Interval for the James A. Fitzpatrick Nuclear Power Plant. -
This document defines the basis for those pressure retaining components and/or systems (including their supports), which are classified Quality Group A, B, and C, (ASME Code Class 1 Class 2, and Class 3), and subject to examination and testing, as set forth in the applicable Edition of the ASME Boiler and Pressure Vessel Code,Section XI, to the extent practical within the limitations of design, geometry and materials o' construction of the components pursuant to Title 10. Part 50. Section .55a (b)(2) of the Code of Federal Regulations.
The ASME Boiler and Pressure Vessel Code, Edition applicable to the James A. Fitzpatrick Nuclear Power Plant's Third inservice Inspection Interval Program Plan and Schedule is the 1989 Edition, with no Addenda of Section XI, hereafter referred to as the Code.
,O in addition to the 1989 Edition of Section XI, and as mandated by the Federal Register, Volume 61, Number V 154, dated August 8,1996, the applicable Edition and Addenda for Subsection IVE, " Requirements for Class MC and Metallic Liners of Class CC components of Light-Water Cooled Power Plants", applicable to James A. Fitzpatrick Nuclear Power Plant is the 1992 Edition with the 1992 Addenda of Section XI.
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Fde: SECT 00.TXT.E1
JAMES A.FITZPATRICK JAFISl0002 NUCLEAR POWER PLANT e
- > WrkPower THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 FROGRAM Paae 1-1 of 1 11 l
TABLE OF CONTENTS SECTION 1 Table of Contents . .. .. . . 1-1 Record of Revision . . . . . . . . 1-3
1.0 INTRODUCTION
. . , . . . . . 1-4 1.1 Inspection interval . . . . 1-4 1.2 Inspection Periods . . . 1-4 1.3 Applicable Documents . . 1-4 O
Code of Federal Regulations . 1-5 ASME Code Editions and Addenda . . .15 USNRC Regulatory Guides . 1-5 JAFNPP Specific Documents . . . . 1-5 USNRC NUREGS/SRP's . . . 1-6
, USNRC Bulletins . . . . . 1-6 l USNRC Generic Letters .. 1-6 USNRC Informauonal Notices . . . 1-6 ASME Code Cases , . 1-6 l
1.4 Applicable Code Editions and Addenda 1-7
, 1.4.1 Third inspection Interval . . .. .. . . 1-7 l 1.4.2 Subsequent Code Edition and Addenda . . 1-7 l
1.5 System Quality Group Classifications . . . 1-8 1.5.1 Quality Group A . . .. . 1-8 1.5.2 Quality Group B . ... . . . 1-8 1.5.3 Quality Group C , . . . . 1-8 1.5.4 Quality Group D . . . 1-8 1.5.5 Application . . . .. .. . . '- 1 -o 1.5.6 Optional Construction . . . 1-8 1.5.7 Piping Penetrating Containment . . 1-9 l
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1.5.8 Classification Diagrams . 1-9 File: SECT 01.TXT-E1
1 JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1908 PROGRAM Pam 12 of 1-11 TABLE OF CONTcNTS (Continued) l l
l 1.6 Inspection Program B . . . 1-9
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1.6.1 Class 1,2 and 3 Components . 19 1.6.2 Component Supports , ~. 1-10 1.7 Development of Inspection Program Plan . . . . 1-10 1.8 Substitute Examinations . .. . ... . 1-11 1.9 Exclusions / Exceptions .. . . . ... 1-11 1.9.1 Containment Penetrations . . . 1 11 1.9.2 Control Rod Drive System - Class 2 . . 1-11 1.9.3 Emergency Diesel 6: erator- Class 3A 1-11 rh
( LIST OF TABLES TABLE 1-2 JAFNPP Inservice inspection Periods . . . , 1-4 TABLE 1-6 Insps:; tion Program B . . . . 1-10
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JA' ES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT p> gWrkPwver THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICS INSPECTION Date: January 6,1998 PROGRAM Paae 1-3 of 1 11 RECORD OF REVISION
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< REVISION / / DATES '
, { EFFECTEDI REdSON FOR REVISION?
- 'Ndi '
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,PAGES4 0 January 6,1998 Entire Updated Inservice Inspection Program Plan Document for the 3'8 Ten Year Inservice Inspection interval v
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ig rk: SECT 01.TXT Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- gWrkPower THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,g, 9,4 gg 3,99 l
1.0 INTRODUCTION
This document details the basis and plans for the Inservice inspection Program for the Third Ten- I Year Inservice Inspection Interval for the James A. Fitzpatrick Nuclear Power Plant. 1 l
The Construction permit for James A. Fitzpatrick Nuclear Power Plant was issued in May 1970. New l York Power Authority (NYPA), hereafter referred to as the Authority, is the Owner of Record.
The Operating License for James A. Fitzpatrick Nuclear Power Plant was issued on October 17, 1974.
The Ce mmercial Service Date for James A. Fitzpatrick Nuclear Power Plant is July 26,1975.
1.1 Inspection Interval The Third Inservice Inspection Interval t'ecomes effective on September 28,1997 and is scheduled to end on September 27,2006.
Note: The third inspection interval is nine (9) years in duraticq, due to NYPA extension of q the second inspection interval by one (1) yt:
V 1.2 Inspection Periods The Third Inservice inspection interval is divided into three successive inspection periods c.is determined by calendar years of plant service within the inspection interval, Identified below are the perioo dates for the third inspection interval as defined by inspection Program "B".
In accordance with IWB-2412(b) the inspection period specified below may be decreased or extended by as much as 1 year to enable inspection to coincide with JAF's plant outages.
TABLE 1-2 JAF INSERVICE INSPECTION PERIODS INSPECTION PERIOD START DATES PERIOD END DATES PERIODS 1 September 28,1997 September 27,2000 2 September 27,2000 September 28,2004 3 September 28,2004 September 27,2006
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1.3 Applicable Documents -
The Third Inservice inspection Program for Quality Group A, B, and C (ASME Code Class 1,2 and 3), systems and components (including their supports) was dcveloped after giving due consideration to the following documents and subject to the limitations end modifbations File: SECT 01.TXT-E1
JAMES A. FITZPATRICK JAF ISI-C002 NUCLEAR POWER PLANT
- > gPower THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pm 15 of 1-11
==4 listeJ in 10 CFR 50.55a(b), and to the extent practical within the limitations of design, geometry and materials of construction.
Code of Federal Regulations 10 CFR 50.55(a) Code of Federal Requiations; Federal Register, Volume 61, Number 154, dated AugustB,1996, amendment to the regulation to incorporate by reference the 1992 Edition with the 1992 Addenda of Subsection IWE and IWL of Section XI, Division 1, of the ASME Code.
ASME Code Editions and Addenda ASME Boiler and Pressure Vessel Code, Sections V,1989 Edition, Nondestructive Examination" ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition, " Rules for Inservice inspection of Nuclear Power Plant Components" ASME Boiler and Pressure Vessel Code,Section XI,1992 Edition through the 1992
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s Addenda, " Rules for Inservice inspection of Nuclear Power Plant Components
- Subsections IWE and IWL
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- USNRC Regulatory Guides The following list of Regulatory Guides are applicable to the James A. Fitzpatrick Nuclear Power Plant Third Inservice inspection Program:
1.26 Quality Group Classifications and Standards for Water-Steam, and Radioactive-Waste-Containing Components of Nuclear Power Plants, Revision 2, June 1975.
1.84 Design and Fabrication Code Case Acceptability ASME Section lit, Division i 1, Latest Revision.
1.85 Material Code Case Acceptability ASME Section lil, Division 1, Latest Revision.
1,147 Inservice Inspection Code Case Acceptability ASME Section XI, Division 1, Revision 11 and Draft Revision 12..
JAFNPP Specific Documents James A. Fitzpatrick Updated Final Safety Analysis Report, Sections 1,5,7,9,12 and 16.
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James A. Fitzpatrick Technical Specifications, USNRC Docket number 50-333, Sections 3.6 and 4.6.
File. SECT 01.TXT E1 is i-, i r
JAMES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT
- 5 NewWrkPower THIRD INSERVICE INSPECTION INTERVAL Rev, O l
INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,y, y ,, q ,9 9 USNRC NUREGS/SRP's !
l USNRC NUREG 0313, Technical Report on Material Selection and Processing l Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2.
USNRC NUREG 0619, BWR Feedwater Nozzle and CRD Return Lines.
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- USNRC Bulletins 82-03 Stress Corrosion Cracking in Thick-Wall, Large-Diameter, Stainless Steel, Recirculation System Piping at BWR Plants, Revision 1, October 28,1982.
USNRC Genetic Letters 88-01 USNRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, January 25,1988.
88-01 NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping, Supplement 1, February 4,1992.
90-05 Guidance for Performing Temporary Non-Code Repairs to ASME (ISI) Code Class 1,2 and 3 Piping and Components, June 15,1990.
90-09 Alternative Requirements for Snubber Visual inspection Intervals and Corrective Actions, December 11,1990.
USNRC Informational Notices 89-79 Degraded Coatings and Corrosion of Steel Containment Vessels,
( December 1,1989.
ASME Code Cases Code Cases approved through Regulatory Guide 1.147 may be proposed for revision to the inspection plan.
N-355 Calibration Block for Angle Beam Ultrasonic Examination of Large Fittings in accordance with Appandix ill-3410,Section XI, Division 1.
N-416-1 Altemative Pressure Test Requirements for Welded Repairs or j Installation of Replacement items by Welding, Class 1,2 and 3,Section XI, Division 1.
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N-460 Altemative Examination Coverage for Class 1 and Class 2 Welds, Q,o Section XI, Division 1.
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Fue: SECT 01.TXT E1
JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM ,p,q, 17 of 1 11 N-491 1 Alternative Rules for Examination of Class 1, 2, 3 and MC Component Supports of Light Water Cooled Power Plants,Section XI, Division 1.
N-498-1 Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1 and Class 2 Systems,Section XI, Division 1.
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N-509 Alternative Rules for the Selection and Examination of Class 1,2 and 3 Integrally Welded Attachments,Section XI, Division 1.
N-522 Pressure Testing of Containment Penetration Piping Section XI, Division 1.
N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI, Division 1.
N-532 Altemative Requirements to Repair and Replacement Document
- Requirements and Inservice Summary Report Preparation and Submission As Required by IWA--4000 and IWA-5000,Section XI, p Division 1.
Q N-573 Transfer of Procedure Qualification Records Between Owners,Section XI, Division 1.
Note: See Appendix D
- Code Cases" of the 10 Year inspection Plan.
1.4 Appilcable Code Editions and Addenda 1.4.1 Third inspection Interval Pursuant to Title 10, part 50, Section 55a(g)(4), of the Code of Federal Regulations, the Inservice inspection requirements applicable to nondestructive examination and sy3 tem pressure testing for the Third inservice Inspection Interval are based on the rJIes set forth in the 1989 Edition of Section XI, that was endomed twelve months prior to the start of the Third Inspection Interval.
1.4.2 Subsequent Code Editions and Addenda As permitted by 10 CFR 50.55a(g)(4)(iv), the Authority may elect to meet the requirements set forth in subsequent Editions and Addenda of Section XI that are incorporated by reference into 10 CFR 50.55a(b)(2), subject to the applicable limitations and modification and subject to USNRC approval.
Portions of Editions and Addenda may also be used provided that a!! related
! , requirements to the respective Editions and Addenda are met. NYPA intends to i
(,k- ,) continually evaluate and apply, as appropriate, changes in adopted Code Editions and Addenda which provide the continuing assurance of the quality and safety of pressure retaining components and systeins.
File: SECT 01.TXT E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- 5 Power .._
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: Janrary 6,1998 PROGRAM Paae 1-8 of 1-11 l l
1.5 System Quality Group Classifications j System safety c'assifications, design and fabrication requirements meet the intent of 10 CFR 50.2v and Regulatory Guide 1.26, to the extent practical within the limitations of design, geometry and materials of construction of the components, as identified within the James A.
Fitzpatrick Nuclear Power Plant Final Safety Analysis Report (FSAR).
~ Water, steam and radioactive contair'ing components (other than~ turbines and condensers) are designated Quality Group A. G, or C , (ASME Code Class 1,2 or 3), and that are safety-related.
1.5.1 Quality Group A (ASME Code Class 1)
Quality Group A system boundaries were developed based on 10 CFR 50.2(v), and the JAF FSAR, and apply to the reactor coolant pressure boundary components.
The Reactor Coolant system includes a single cycle, forced circulation, General Electric Boiling Water Reactor.
1.5.2 Quality Group B (ASME Code Class 2)
Quality Group B system boundaries were developed based on Regulatory Guide 1.26 and the JAF FSAR, and apply to %ose components of the Reactot Coolant System not classified as Quality Group A, (ASME Code Class 1), and that are safety-related.
1.5.3 Qua!!ty Group C (ASME Code Class 3)
Quality Group C system boundaries were developed based on Regulatory Guide 1.26 and the JAF FSAR, and apply to those components that are not classifieo as Quality Group A or B, (ASM2 Code Class 1 or 2), and that are safety-related.
1.5.4 Quality Group D (Non-Nuclear Safety Related)
Quality Group D applies to those components not related to nuclear safety.
1.5.5 Application Application of the rules of Section XI are govemed by the group classification criteria as defined above. The rules of IWB, IWC, and IWD were applied to those systems whose components are classified Quality Group A, B, or C, (ASME Class 1,2 or 3).
1.5.6 Optional Construction of an Component Optional construction of a component within a system boundary to a classification (Ly higher than the minimum class established in the comt :nent Design Specification (either upgrading from Class 2 to Class 1 or from Class 3 to Class 2) shall not affect the overall system classification by which the applicable rules of Section XI are determined. _ ,
mym sur File: SECT 01lTXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT i THIRD INSER\!CE INSPECTION INTERVAL Rev. O F tiRVICE INSPECTION Date: January 6,1998 Page 19 of 1-11 1.5.7 Piping Penetrating Containment Tbs portions of piping that a penetrate the containment vessel which are required to be constructed to Class 1 or 2 rules for piping and which may differ from the classification of the balance of the piping system, need not affect the overall system classification that determines the applicable rules of Division.
1.5.8 Classification Diagrams The system Quality Group A, B and C,(ASME Code Class 1,2 and 3) classification interfaces between components of different quality groups applicable to James A.
Fitzpatrick Nuclear Power Plant, are designated on various ISI Flow Diagrams (ISI.
FM(s)). These designations identify the system class breaks by a numeric identifier, either a 1,2 or 3.
The rules of IWB, IWC and IWD were applied to these drawings in order to determine those components / systems subject to examination /ter* Components subject to surface, volumetric and visual examination are listed in the Ten-Year Inservice inspection plan tables.
Appendix G provides a list of the applicable flow diagrarra (ISI-FM(s), to this program. Copies of these diagrams are avuilable through the drawing control system.
1.6 Inspection Program B The James A Fitzpatrick Nuclear Power Plant inspection intervals comply with IWA-2432, inspection Program B. With the exceptions of the examinations identified in 1.6.1, the required examinations in eacn examination category shall be completed in accordance with Table 1-6.
1.6.1 Class 1,2 and 3 Components The requead examinations in each Examination Category shall be completed during each inservice inspection interval, in accordance with IWB-2412-1, lWC-2412-1, and IWD-2412-1, with the following exceptions:
(1)- Examination Categories B-N-1, B-P, B-O, C-H, and the system pressure test requirements of D A, D-B and D-C; (2) the percenq required by Note 2 of Examination Cateoory B-D; (3) the examinations that may be deferred to the end of an inspection interval, as specified in Table IWB-2500-1.
If there are less than three items to be examined in an Examination (4)
Category, the items may be examined in any two periods, or in any one period if there is only one item, in lieu of the percentage requirements of Table 1-6 below.
File: SECT 01.TXT-E1
l JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] j NUCLEAR POWER PLANT '
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: sanuary 6,1998 l
PROGRAM Paoe 1 10 of 1-11 (5) Within various Code Categories the total number of items scheduled for examination exceeds inspection Prcgram 'B' requirements. Adjustments to those Code Categories which exceed the allowable Program "B" percentages, may be reduced to meet Program *B" requirements.
1.6.2 Component Supports
~
The required examinations shall be completed in accofdance with the inspection schedule established for the components under lWB, IWC, and IWD.
TABLE 1-6 INSPECTION PROGRAM B Inspection inspection Period Minimum Maximum Interval Calendar Yeart of Examination Examination Plant Service Within Completed, % Credited, %
the Interval [Notc (1) (2)]
CN JAF 3* 23 16 34 O Inservice 27 50 67 Inspection Interval 30 100 100 Note: (1) Except as noted in Table IWB-2500-1, B1.30.
I Note: (1) The examination of shell-to-flange welds may be performed dunng the first and third inspection penods in conjunction with the nozzle examinations of examination ct'egory B-D.
At least 50% of shell-to-flange weids, shall be examined by the end of the First inspection j Penod, and the remainder by the end of the Third Inspection Interval.
(2) At least 25% but not more than 50% (credited) of the nozzles shall be examined by the end of the 4t inspection period, and the remainder by the end of the last inspection penod, of the respective inspection interval. (See paragraph 1.6.1 (2)).
1.7 Development of Inspection Program Plan Sections 2 through 6 detaii the narrative description of the James A. Fitzpatrick Third Inservice Inspection Program Plan basis for Quality Group A, B or C, (ASME Code Class 1, 2 and 3), (including their Supports and Augmented Examinations), of components and/or systems subject to examination / test.
Quality Group A, B and C, (Class 1,2 and 3) systems subject to examination and testing include the following:
, 02-2 RC Reactor Water Recirculation System (m)' 03 CRD Control Rod Drive System 10 RHR Resuual Heat Remova! System 12 RWC Reactor Water Clean-Up System i
13 RCIC 03 actor Core Isolation Coolant System File. SECT 01.TXT E1
JAMES A. FITZPATR!CK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT ter ity THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,,, q,$ $ ,, q ,q , j l
14 CS Core Spray System 15 RBCLCS Reactor Building Closed Leap Cooling System ,
19 FPCS Fuel Pool Cooling System l 23 HPCI High Pressure Coolant injection System 29 MS Main Steam System i 34 FW Feedwater System 46 SWS Service Water System
' 66 RBV&CS Reactor Buile.ng Vent and Cooling Sysfem 70 CRV&CS Control Room and Relay Room Vent and Cooling System 1.8 Substitute Examinations The Authority may substitute items scheduled in the Inspection Plan for others not previously scheduled when the original selection was part of the additional piping welds.
This substitution may be done due to such conditions as limited physical ace w, high radiation levels, etc. Such changes will be noted in the Summary Report submittal as required by IWA-6000 of the applicable Code Edition. Specific examinations that are sequired, and can not be completed within the period / interval will be identified within the Summary Report, and as applicable, may be the subject of a request for relief.
p 1.9 Exclusions / Exceptions This paragraph defines the exclusions / exceptions, NYPA has taken due to the u.7it being Docketed prior to June 1978.
1.9.1 Containment Penetrations A. For containment penetration configurations, where there are no inside valves, the following will apply:
- 1. The rules of Section XI shall appi, to the first valve outside the drywell through the penetration to the first weld inside the drywell.
- 2. Repairs / replacements shall be in accordance with NYPA design specification and the original conit uction code.
- 3. Pressure tests shall be in accordance with the Appendix J testing Program.
1.9.2 Class 2 Control Rod Drive System All pressure tests shall be performed during Reactor startup and/or shutdown during a refueling and/or maintenance outage.
7 1.93 Emergency Diesel Generator (EDG)
)
'd NYPA has optionally upgraded portions of the EDG system to an augmented class 3 category (3A) for the purposes of pressure testing only.
File: SECT 01.TXT-E1
JAMES A. FITZPATRICK l [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT E THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 2-1 of 2.'O TABLE OF CONTENTS SECTION 2 Table of Contents . .. .. 2 -1 Record of Revision . . ., , . 2 -2 2.0 CLASS 1 SYSTEMS / COMPONENTS . 2 -3 2.1 ASME Code Exemptions . . . . . . . 2 -3 2.2 Component / Piping Examination Development .. . . 2 -3 2.2.1 Category B-A . . . . . 2 -3 2.2.2 Category B-B . . . 2 -5 O
V 2.2.3 Category B D .. . 2 -5 2.2.4 Category B-E . . . 2 -6 2.2.5 Category B-F 2 -6 2.2.6 Category B-G-1 2 -7 2.2.7 Category B-G-2 . 2 -9 2.2.8 Category B-H 2 -9 2.2.9 Category B-J . 2 -10 2.2.10 Category B-K 1 . . 2 -11 2.2.11 Category B-L-1 and B-L-2 2 -12 2.2.12 Category B-M-1 and B-M-2 2 -12 2.2.13 Category B-N-1, B-N-2, and B-N-3 .. 2-13 -
2.2.14 Category B-0 . . . . . 2 -14 2.2.15 Category B-P 2 -14 2.2.16 Category B-Q . . . 2 -16 File: SECT 02.TXT-U1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT l
- > gYorkPower THIRD INSERVICE INSPECTION INTERVAL Rev. O l
INSERVICE INSPECTION Date: January 6,1998 l PROGRAM Paco 2-2 of 2 18 g RECORD OF REVISION
. . . . , s .. .
.m .
, w y,
^ REVISION . .,TQDATE[- ,
!AFFECTEDf '. REASON FOR REVISION lJ x-g 9 <
- . PAGES * % < ~-
0 January 6,1998 Entire Updated inservice inspection Program Plan Document for the 3* Ten Year Inservice inspection Interval D,
%.)
l l
l l
l l
I i
l l
l m
s File: SECT 02.TXT E1
JAMES A. FITZPATRICK JAF ISI 0002 )
NUCLEAR POWER PLANT 4 WakPower tv 8 THIRD INSERVICE INSPECTION fNTERVAL Rev. O Date: January 6,1998 INSERVICE INSPECTION PROGRAM Paae 23 of 2 16 2.0 CLASS 1 SYSTEMSiCOMPONENTS The ASME Code Class 1 system boundaries subject to examination and testing were developed based upon the requirementt of 10 CFR 50.2(v) and James A. Fitzpatnck (JAF), Final Safety y Analysis Report (FSAR). The ASME Code Clats 1 components and systems (including thir supports) subject to examination and testing are desenbed in det9il below:
2.1 ASME Code Exemptions IWB 1220 - The following components (or parts of components) are exempted from the volumetric and surface examination requirements of IWB 2500:
(a) Components that are connected to the Reactor Coolant System and part of the reactor coolant prNsure boundary and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the Reactor Coolant System, under normal plant operating conditions, is within the capacity of makeup systems which are operable frota on site emergency power.
(O Note: NYPA has determined through evaluation that Class 1 components, piping, and associated valves, vessols, (including their supports), that are three (3) inches NPS and smaller are exempt from the volumetric and surface examinations. Sufficient normal makeup capacity using on site power exists for 6 aqual to or less than a 3.0 inch pipe break.
(b) 1. Piping of 1" nominal pipe size and smaller; and
- 2. Componer.ts and their connections in piping of 1" nominal pipe size P so smaller, (c) Reactor vessel head co.inections and associated piping,2" nominal pipe size and smaller, made inaccessible by control rod dnve penetrations.
2.2 Component / Piping Examination Development A narrative discussion of Class 1 components subject to examination and testing are described in detail below; 2.2.1 Category B A, Pressure Retalning Welds le Reactor Vessel All examinations are performed from the inside and/or outside surface using manual / automated inspection equipment, (as applicable) and volumetric examination te:hniques.
Note: NYPA has not .brmally adopted Regulatory Guide (R G.) 1.150, Revision 1, Appendix A, but
(~'s does perform all RPV weld exams and evaluations in accordance with the volumetric V examination technique and requirements outhned in the regulatory gdde. The ultrasonic techniques of R.G.1.150 are performed in add *on to those detailed in ASME Section XI.
Fe SECT 02.TXT Et
JAMES A.FITZPATRICK JAF lSI 0002 NUCLEAR POWER PLANT p
y h THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 24 of 2 16 11 ems B1.11.B1.12 ShellWelds Scope of Examination - includes essentially 100% of all longitudinal and circumferential shell weld lengths. (does not include shell to flange weld).
(4) 81.11 Circumferential shell welds (12) B1.12 Longitudinal shell welds Note: Pursuant to 10 CFR 50.55a(a)(g)(6)(ii)(A)(3)(ii), the deferred augmented examinations of the reactor vessel shell welds specified in item B1.10 of Examination Category B-A
- Pressure Ret ining Welds in Reactor Vesse, in Table IWB 2500-1 of Subsection lWB of Section XI, Division 1, of the ASME Boiler and Pressore Vessel Code, as mandated by 10 CFR 50.55a(a)(g)(6)(ii)(A), will be used as a substitute for the reactor vessel shell weld examinations scheduled for inspection during the JAF 380 Inservice inspection Interval. Second Interval deferral of augmented examinations to the 3"D inspection interval was approved by the USNRC, TAC No.
M87158, dated 03/08/94.
Items B1.21. B1.22 Bqttom Head Welds Scope of Examination - includes essentially 100% of accessible length of all circumferential and meridional head welds.
(2) B1.21 Circumferential head welds (14) B1.22 Meridional head welds items B1.21. B1.22 - Too Head Welds Scope of F.xamination includes essentially 100% of accessible length of all circumferential and meridional head welds.
(1) 81.21 Circumferential head welds (8) B1.22 Meridional head welds item B1.30 Shell-to-Flanae Weld Scope of Examination 100% of the shell to flange weld.
(1) B1.30 Circumferential shell to flange weld Note: If partial examina%ns are conducted from the flange face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or
("]
'J near the end of each inspection interval.
Flie: SECT 02.TXT.Et
JAMES A. FITZPATRICK JAF ISI 0002 NUCLEAR PGWER PLANT 4# ty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 DROGRAM Pace 2-5 of 2 16 The examination may be performed during the first and third inspection periods in conjunction with the nozzle examinations of Examination Category B-D (Program B). At least 50% of the weld shall be examined by the end of the First inspection Period, and the remainder by the end of the Third Inspection Period.
item B1.40 Head to FlanaqJygid .
Scope of Examination includes essentially 100% of the head to flange weld length.
(1) B1.40 Circumferential head to flai go weld item B1.51 - Renalr Welds.1 Beltline Rjsign)
- No bae metal weld repairs were performed during the 1" or 2"D inspection intervals at the James A. Fitzpatrick Nuclear Power Plant.
2.2.2 Category B-8, Pressure Retaining welds in vessels other than Reactor Vessels.
O V This Examination Category is not applicable to James A. /itzpatrick Nuclear Power Phnt.
2.2.3 Category B-D, Full Penetration Welds of Nozzle in Vessels (Program B)
Heactor Vencls: ltems B3.90. B3.100 - Nozzle to Vessel Welds and Nozzle inside Radius Ses11gn Scope of Examination - 100% of all nozzles with full penetration welds to vessel shell (or head) and integrally cast nozzles.
(28) B3.90 RPV Nozzle Welds, (27) Selected (28) B3.100 RPV Nozzle inner Radius, (27) Selected Note: At least 25% but not more than 50% (credited) of the nozzles shall be examined by the end of the First inspection Period, and the remainder by the end of the inspection interval.
Augmented Examinations of the Feedwater Nozzle, in accordance with USNRC NUREG 0619, are addressed in Section 6, Augmented Examinations.
! Subject to Relief Request 7
- hessurizer
- ltems B3.110. B3.12Q.
I O
) - Not applicable to James A. Fitzpatrick Nuclear Power Plant.
l File SECT 02.TXT-Et
JAMES A. FlTZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 2-6 of 2 16 Steam Generators: llems B3.130. B3J.3 4
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Heat Exchanger: ltems 83.150. B3.160 .
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
2.2.4 Category B.E, Pressure Retalning Partial Penetration Welds in Vessels items B4.11. B4.12. B4.13 - Venel Nr>zzles. Control Rod Drive. Nozzles.
Instrumentation Norzin Scope of Examination - Visual VT 2 examination on 25% of all partial penetration welds each interval. The examinations are performed during the System Hydrostatic Pressure Testing. These examinations and tests are part of the Inservice Pressure G Test Program. See paragraph 2.2.15 (b) for hydrostatic pressure testing
/j
's requirements.
(2) B4.11 Vessel nozzles, (1) selected (137) B4.12 Control rod drive nozzles (34) selected (49) B4.13 instrumentation nozzles, (12) selected Subject to Relief Request 3 2.2.5 Category B F, Pressure Retalning Dissimilar Metal Welds Reactor _Yessel: ltemplido. B5.20. B5.30 Scope of Examination - Examinations are required of all safe end welds in each loop and connecting branch of the Reactor Coolant System. For the reactor vessel nozzle safe ends, the examination may be performed coincident with the vessel nozzle examinations required by Examination Category B-D.
All Examination Category B-F welds are scheduled and examined as part of the IGSCC Augmented Inspection Program. The extent and frequency of the examinations are in accordance with NUREG 0313, Revision 2 and Generic Letter GL 88-01. See Section 6.0 of this Program for details. Comnleted examinations shall be used to satisfy the requirements of both inspection Program *B* and NUREG 0313.
en -
(16) B5.10 NPS 4" or Larger Nozzle-to-Safe end butt welds
{) -
( 0) B5.20 Less than NPS 4" Nozzle-to-Safe end butt welds 35.30 Nozzle-to-Safe end socket welds
( 0) i File. SECT 02.TXT-E1
JAMES A. FITZPATRICK JAF lSI-0002 i NUCLEAR POWER PLANT l THIRD INSERVICE INSPECTION INTERVAL Rev. O n INSERVICE INSPECTION Date: January 6,1998 PROGRAM i hae 27 of 2 16 _.
~q Pressurizer items B5.40. B5.50. 85.60
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Steam Generators ltems B5.70. 85.80. 85.90 -
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Heat Exch1Dger: Items B5.100. B5.110. B5.129
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Piolna: ltems B5.130. B5.140. B5.150 Dissimilar Metal Butt Welds > or = 4" NPS. < 4" NPS. S.peket Weld Scope of Examination Examinations are required of all safe end welds in each loop and connecting branch of the Reactor Coolant System.
O Q' All Examination Category B-F welds are scheduled and examined as part of the IGSCC Augmented Inspection Program The extent and frequency of examinations are in accordanco with NUREG 0313, Revision 2 and Generic Letter GL 88-01. See Section 6.0 of this Program for details. Completed examinations shall be used to satisfy both inspection Program 'B' and NUREG 0313 requirements (9) B5.130 a 4" NPS dissimilar butt welds (0) B5.140 < 4" NPS dissimilar butt welds (0) B5.150 No dissimilar socket welds 2.2.6 Category B-G 1 - Pressure Retalning Bolting, Greater Than 2 in,in Diameter Reactor Vessel: Items B6.10. B6.20. B6.30. 86.40. B6.50 Scope of Examination - Examination includes all bolts, studs, nuts, bushings, threads in flange stud holes. Bolting may be examined in place under tension, when the connection is disassembled, or when the bolting is removed. For heat exchangers, piping, pumps, and valves, examinations are limited to components selected for examination under Examination Categories B 8, B J, B-L 2, and B M-2.
Bushings and threads in base material of flanges are required to be examined only when the connections are disassembled. Bushings may be inspected in place.
Flange surfaces, when the connection is disassembled, include 1 inch annular Q surface of flange surrounding each stud.
NI FC: SECT 02.TXT Ei
\
JAMES A.FITZPATRICK JAF lSI-0002 j NUCLEAR POWER PLANT
')
THIRD IfdSERVICE INSPECTION INTERVAL Rev. O yh.
?
INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 2-8 of 2 16 i
(52) B6.10 Nuts,1/3 each period (52) B6.20 Studs in Place,1/3 each period (4) B6.30 Studs, when removed (52) B6.40 Threads in Ligaments,1/3 each period (104) B6.50 Washers, bushings,1/3 each period Note: In addition to item BS.40 listed above, NYPA committed to inspect 22 threads in the RPV flange during the 1" Period of the 35 Interval.
Pressurizer: litIDM60. B6.70. B6.80
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Sjesm Generators: Items B6.90. B6.100. B6.110 Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Heat Exchanaer: Items B6.120. B6.130. B6.140 O
V - Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Ploina: liems B6.150. B6.160. B6.11.Q
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Pumos: Items B6.180. B6.190. B6.200 Scope of Examination All bolts, studs, nuts, bushings, and flange surfaces.
l Examinations applicable to Reactor Recirculation Pumps 02-2P-1 A and 02 2P-18.
Note: Pump bolting is limited to the pump selected under Examination Category B-L 2.
Bolting may be examined in place under tension, when the connection is
- disassembled, or when the bolting is removed.
Bushings and threads in base material of flanges are required to be examin6d only when the connections are disassembled. Bushings may be examined in place.
Flange surface requires 1 inch annutar surface of flange surrounding each stud hole.
l -
(32) 86.180 Studs,16 studs per pump, one pump required (32) B6.190 Flange surfaces,16 per pump, one pump required (32) 86.200 Nuts, Bushings. and Washers,16 per pump, one pump p required O
I File: SECT 02.TXT Ei I
JAMES A. FITZPATRICK JAFISI0002 i NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 I.
PROGRAM Paae 29 of 2 16 3 Yelves: ltems B6.210. B6.220. B6.23D i I
- Not applicable to James A. Fitzpatrick Nuclear Power Plant. I 2.2.7 Category B G 2, Pressure Retaining Botting,2 in. And Less in Diameter ltems: B7.10. B7.20. B7.30. 87.40. B7.50. B7.60. B7.70. B7.80 Scope of Examination VisualVT 1 examination each interval of all bolts, studs, and nuts. Examinations are limited to components selected for examination under Examination Category B B, B J, B L 2, and B-M 2.
(3) B7.10 Reactor Pressure Vessel (0) B7.20 Pressurizer, not applicable to JAF (0) B7.30 Steam Generator, not applicable to JAF (0) B7.40 Heat Exchangers, not applicable to JAF (3) B7.50 Piping Flange Botting (2) selected (32) B7.60 Pump,2 pumps 16 cap screws per pump, one pump required l
(60) 87.70 Valves, (18) required
( - (137) B7.80 CRD Housing, when disassembled 2.2.8 Category B H, Integral Attachments for Vessels BeactorVessel:ltem B8.10 Inteurally Welded Attachment Scope of Examination - Examination includes essentially 100% of the length of the attachment weld at each attachment subject to examination. Examinations limited to the Reactor Pressure Vessel skirt weld and stabilizers.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509,'Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1"
- (5) Integrally welded attachments, (5) selected Subject to Relief Roquest 4 Pressurizer: Item B8.20 -Intearally Welded Attachment Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Steam Gent.tstor: Item B8.30 -Intearally Welded Attachment
)
(,/ -
Not appl; cable to James A. Fitzpatrick Nuclear Power Plant.
F0e: SECT 02.TXT Et
l I
JA~iES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT
- > WrkPower tv 9 THIRD INSERVICE INSPECTION INTERVAL Rev. O Date: January 6,1998 INSERVICE INSPECTION l PROGRAM Pane 2 10 of 2 16 L ua.
11 cat Exchanaer: ltem BBdD_drif carally Welded Attachment
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
2.2.9 Category B J, Pressure Ret .Ining Welds in Piping jlems: B9.11. B9.12. 89.21. B9.22. 89.31. B9.32. 89.40 Scope of Examination - All dissimilar metal pipe welds, terminal ends, plus an additional number of piping welds so that 25% of all non-exempt circumferential and branch connection pipe welds are examined.
Note: Table IWB-2500-1, Examination Category B-J. Footnote (1)(b)(1) and (2) are not applicable to JAF, due to the unit being docketed prior to June 1978.
All augmented Main Steam and Feedwater welds, to the extent practical shall be used to satisfy the requirements of both inspection Program *B" and the augmented requirements. See Section 6.0 of this Program for details.
O i \
V Alliongitudinal pipe welds intersecting any of the selected circumferential welds will also be examined. As an attemate to Table IWB-25001, ASME Code Case N 524,
" Alternative Examination Requirements for Longitudinal Welds in Class 1 Piping Section XI, Division 1", shall be used as defined below:
i
- i l (A) When only a surface examination is required, examination of longitudinal l piping welds is not required beyond those portions of the welds within the
! examination boundaries of intersecting circumferential welds.
(B) When both surface and volumetric examinations are required, examination i of longitudinal piping welds is not required beyond those portions of the l welds within the examination boundaries of intersecting circumferential welds providing the following requirements are met.
(1) Where longitudinal welds are specified and locations are known, l examination requirements shall be met for both transverse and l parallel flaws at the intersection of the welds and for that length of l longitudinal weld within the circumferential weld examination volume; (2) Where longitudinal welds are specified but locations are unknown, or the existence of longitudinal welds is uncertain, the examination requirements shall be met for both transverse aad parallel flaws within the entire examination volume of intersecting circumferential
/~T welds.
FO: SECT 02.TXT-Et
JAMES A.. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 2-11 of 2 16
- (444) 89.11 Circumferential welds, (130) scheduled
- (158) 89.12 Longitudinal welds, (46) scheduled (23) 89.21 Circumferential welds, (1) scheduled (0) 89.22 Longitudinal welds (38) 89.31 Branch Conn. NPS 4" or Larger, (11) scheduled (98) 89.32 Branch Conn. Less than NPS 4", All exempt per IWB-1220(a)
(0) B9.40 Socket welds, not applicable to JAF
- (505) Cire. Welds subject to examination, (126) required. (142) scheduled,28%
(158) Long. Welds subject to examination, (46) scheduled,29%
See Appendix A Tables for details.
Subject to Relief Request 5 2.2.10 Category B-K 1, Integral Attachments to Piping Pumps & Valves items: B10.10. B10.20 Ploing and Pumo intearal Attachments V
Scope of Examination Allintegrally welded attachments with a design thickness of 5/8 inch and greater. The examinations include only the welded attachments to piping required to be examined under Examination Category B-J and the welded attachments to pumps associated with this piping. The multi-component concept is Dol applicable to this examination category.
Examinations will be performed in accordance with the attemate requirements of Code Case N-509, " Alternative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1".
Note: In addition to those conditions specified in the Code Case, a minimum 10% sample of integral!y welded attachments for each item in each Code Class per interval will be examined.
- (121) B10.10 Integral Attachments, (12) required, (15) scheduled (18) B10.20 Integral Attachments, (1) required, (1) scheduled item: B10.30 Valve Intearal Attitchments
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Fae: SECT 02.TXT-Ei
I JAMES A. FITZPATRICK JAF.lSI-Ou02 NUCLEAR POWER PLANT h ty THIRD INSERVICE INSPECTION INTERVAL Rev. 0 l
INSERVICE INSPECTION Date: January 6,1998 )
PROGRAM Paoe 2-12 of 2 16 2.2.11 Category B L 1, Pressure Retalning Welds in Pump Casings, B-L 2, Pump Casings
((em: B12.10 Pumo Casing Welds-Scope of Examination -100% volumetric examination of.all welds in one of the two Reactor Recirculation Coolant Pumps. The pump selected shall be based on pump disassembly for maintenance under B L 2 or end of inspection interval, whichever comes first.
- Not applicable to James A. Fitzpatnck Nuclear Power Plant. Reactor Recirculation Pumps do not have casing welds.
Item: B12.20 Pump _ Casing Scope of Examination - Visual examination of interior surfaces of one of the two (2)
Reactor Recirculation Pumps when disassembled for maintenance. Pump to be identified when pump is disassembled.
(3 V -
(2) B12.20 Recire. Pumps (1) Pump Required 2.2.12 Category B-M-1, Pressure Retalning Welds in Valve Bodies, B.M 2, Valve Bodies litmt: B12.30. B12.40 Valve Body Welds
- Not applicable to James A. Fitzpatrick Nuclear Power Plant. Valves do not have any valve body welds.
Item: B12.50 Valve Bodv Interior Scope of Examination - Visual VT 3 examination of at least one valve in a group of valves that are the same size, constructional design (such as globe, gate, or check valves), and manufacturing method, and that perform similar functions in the system (such as containment isolation and system over pressure protection). Examinations are performed once per interval when disassembled for maintenance or repair.
Valves to be identified when valve is disassembled.
Group 1, Sys 02 2 (11) 4" Relief Valves, i required
- Group 2, Sys 02 2 (4) 28" Gate Valves,1 required
- Group 3, Sys 10 (2) 24" Globe Valves,1 required Group 4, Sys 10 (4) 24" Gate Valves,1 required i
- Group 5, Sys 10 (2) 24" Check Valves, i required
- Group 6, Sys 10 (3) 20" Gate Valves,1 required l
p)g C - Group 7, Sys 12 (3) 6" Gate Valves,1 required l
Group 8 Sys 12 (1) 4" Gate Valve,1 required Fde: SECT 02.TXT-Ei
JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM P e 2 13 of 2 16
- Group 9, Sys 13 (1) 4" Gate Valve,1 required
- Group 10, Sys 13 (1) 4" Check Valve,1 required
- Group 11, Sys 14 (6) 10" Gate Valves,1 required
- Group 12, Sys 14 (2) 10" Check Valves,1 required
- Group 13, Sys 23 (1) 14" Gate Valves,1 required
- Group 14, Sys 23 (1) 14" Check Valve,1 required
- Group 15, Sys 23 (2) 10" Gate Valves,1 required
- Group 10, Sys 29 (8) 24" Globe Valves,1 required
- Group 17, Sys 34 (2) 18" Gate valves,1 required
- Grosp 18, Sys 34 (4) 18" Check Valves,1 required 2.2.13 Category B-N 1, Interior of Reactor Vessel, B N 2, Integrally Welded Core Support Structures and Interior attachments to Reactor Vessels, B-N 3, Reraovable Core Support Structures.
Note: Augmented IWI examinations are addressed in Section 6 of this Program.
Item: B13.10 V111tel InterigI A
d Scope of Examination - Visual VT 3 examination of accessible areas (areas above and below the reactor core made accessible for examination by removal of components during normal refueling), once each inspection period.
(1) Accessible Areas items: B13.20 Interior Attachments Within Beltline Region Reactor Vessel (BWR)
Scope of Examination Visual VT 1 examination of accessible welds of interior attachments within the Beltline region (once per interval).
(23) Interior Attachments item: B13.30 Interior Attachments - Beyond Beltline Region Scope of Examination - Visual VT-3 examination of accessible welds of interior attachments beyond the Belt!ine region (once per interval).
(60) Interior Attachments item: B13.40 Core Support Structuig
(]j r Scope of Examination - Visual VT-3 examination of accessible surfaces of core support structures (once per interval).
File SECT 02.TXT-E1
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT A fjewWrkPower tv Authority G THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pam 2 14 of 2 16 ltem: B13.50. B13.60. B13.70 RPV (PWR's)
Reactor Vessel (PWR)
- Not apphcable to James A. Fitzpatrick Nuclear Power Plant.
Note: Augmented examination requirements are addressed in Section 6 of this Program.
2.2.14 Category B 0, Pressure Retalning Welds in Control Rod Housings hem: B14.10 Welda in CRD Housinan Scope of Examination -Volumetric or surface examination of 10% of the peripheral CRD housings.
- (137) CRD Housings, (28) Peripheral CRD Housings. (3) required (10%)
O V 2.2.15 Cetegory B P, All Pressure Retaining Components hgms: B15.10. B15.11. B15.20. B15.21. B15.30. B15.31. B15.50. B15.51. B15.60.
B15.61. B15.70. B15.71 Scope of Examination - System pressure tests are conducted on Class i systems and components as follows:
(a) A System Leakaoe Test, IWA 5211(a)-is conducted prior to plant startup following each reactor refueling outage.
The pressure retaining boundary subject to the leakage test corresponds to the Reactor Coolant System boundary, as established with all valves aligned as required by approved plant operating procedures for startup and normal reactor operation. The VT-2 examination boundary extends to include the second closed valve at the boundary extremity, which may be a check valve opposing Reactor Coolant System pressure. The test is conducted at system operating temperature and pressure.
(b) A System Hydrostatic Test: will be performed in accordance with alternate examination techniques of Code Case N-4981
- Alternative Rules for 10 Year Hydrostatic Pressure Testing for Class 1 Systems,Section XI, Division 1", as noted below:
(]
U' As an alternative to the 10-Year hydrostatic pressure test required by Table IWB-2500-1, Category B-P, NYPA will perform the following:
Fi:1 SECT 01TXT Et
JAMES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT
- > WrkPower i ev '
G THIRD INSERVICE INSI /CTION INTERVAL Rev. 0 l l
INSERVICE INSPECTION Date: January 6,1998 i PROGRAM Pace 2 15 of 2 16
~
(1) A system leakage test (IWB 5221) shall be conducted at or near the end of each inspection interval, prior to reactor startup.
)
(2) The boundary subject to test pressurization during the system leakage test shall extend to all Class 1 pressure retaining components within the system boundary.
(3) Prior to performirq VT-2 visual examination, the system shall be pressurized to nominal operating pressure for at least four hours for insulated systems and ten minutes for non-insulated systems. The system shall be maintained at nominal operating pressure during the performance of the VT 2 visual examination.
Note: (PWR issue, included for completeness only), Per IWA 5242, for those systems borated for the purpose of controlling reacuvity, insulation shall be removed from pressure retaining bolted connections for visual examination VT-2. The VT 2 examination will be performed in accordance with Code Case N 533,
- Alternative Requirements for VT 2 Visual Examination of Class 1 insulated Pressure Retaining Bolted Connections,Section XI, Division 1", as follows:
f
(
(a) A system pressure test and VT 2 visual examination shall be performed each refuehng outage without removal of insulation.
(b) Each refueling outage the insulation shall be removed from bolted connection, and a VT-2 visual examination for evidence of leakage shall be performed and evaluated. The connection is not required to be pressurized.
l (4) Test temperatures and pressures shall not exceed hmiting conditions for hydrostatic test curve as contained in the JAF plant Technical Specifications.
(5) The VT 2 visual examination shallinclude all components within the boundary identified in (2) above.
All ASME Section XI Pressure Testing requirements are controlled by site procedures, The system boundaries subject to system leakage and system hydrostatic tests are shown on the ISI Flow Diagrams (See Appendix G). Details of the system pressure test program are defined in Inservice Inspection System Pressure Tesf Plari.
l (6) Test instrumentation requirements of IWA-5260 are not applicable.
/ \
() Subject to Relief Requests 3 and 10 File: SECT 02.TXT Et
l l
l JAMES A. FITZPATRICK JAF ISI-0002 1 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 216 of 2 16 2.2.16 Category B-Q, Steam Generator Tubing item: B16.10. Steam GentIBtor Tubina in Stralatt Tube Deslan
- Not applicable to James A Fitzpatrick Nuclear Power Plant.
O O
File SECT 02.TXT Ei
_ _ _ _ _ ~_ __ . . _ . . . _ _ .__. _ _ _ _ _ _ _ . . . _ _ _
JAMES A. FITZPATRICK JAF lSI 0002 4 NUCLEAR POWER PLANT
- > Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998
, PROGRAM Page 31 of 3 11 TABLE OF CONTENTS SECTION 3 4
Tablo of Contents . ...... . ... ....... .. .. ........ .. .. ... ..... . ...... 31 Record of Revision . . . . . . . . . ... .. ... .. .. . ... .. . .. .....32 3.0 CLASS 2 SYSTEMS / COMPONENTS . .. .. . ... . . . . . . . . 33 3.1 ASME Code Exemptions . . . . . . . . . . .. . ... .. . ... 3-3 3.1.1 IWC-1221 . ...... .. . . .. . ... . 3-3 3.1.2 IWC 1222 . . ...... . . .... ... ..... .. .. ........ 3-4 3.1.3 IWC 1230 . . . . . . .. . .. . . . .. . 3-4 3.2 Component / Piping Examination Development . .. .. . .. . . 3-4 3.2.1 Category C A .. . . . . .. . 3-4 3.2.2 Category C B . ... . . . 3-5 3.2.3 Category C-C . . .. . . .. 3-6 3.2.4 Category C D . .. .. .. .. . . . ... . . . . . .. 37 3.2.5 Category C-F-1 . ... . ... .. . ... .. .. . .. ........ 3-8 3.2.6 Category C-F 2 . . . . . . . . . . . . . . . . . . . .. . . . . ...... .. . 39 3.2.7 Category C-G . , .... ... . . . .. ... ........... .... 3-10 3.2.8 Category C-H . . . . . . . . . . . ..... . ... ... .. . . . .. ... 3-10 O '
V Fes: SECT 03.TXT-Et
iam JAMES A. FITZPATRICK JAF IS14002 NUCLEAR POWER PLANT 4# y THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 32 of 3 11 RECORD OF REVISION REVISION - IDhTE- AFFECTED1 : REASON FOR REVISION No. PAGESL 0 January 6,1998 Entire Updated inservice Ins'pection Program Plan Document for the 3* Ten Year Inservice Inspection Interval i
V L)1 File: SECT 03.TXT Et
JAMES A. FITZPATRICK JAF ISI-0022 NUCLEAR POWER PLANT
- > ljlowWrkPower
& wherW THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 33 of 3 11 3.0 CLASS 2 SYSTEMS / COMPONENTS The Class 2 System Boundaries were Jeveloped based upon the requirements of Regulatory Guide 1.26 and the JAF FSAR.
The Class 2 components and systems (including supports) subject to examination and testing are descobed in detail below: -
3.1 ASME Code Exemptions IWC 1220 The following components (or parts of components) are exempted from the volumetric and surface examination requirements of IWC 2500; 3.1.1 IWC 1221 Components within RHR, ECC and Charging Systems (or portions of systems).
(a) Vessels, piping, pumps, valves, and other components NPS 4 and smaller in all systems except high pressure safety injection systems of pres 8unzed water reactor plants.
U (b) Vessels, piping, pumps, valves, and other componerits NPS 1 % and smalbr in high pressure safety injection systems of pressurized water reactor plants.
(c) Component connections NPS 4 and smaller (including nozzles, socket fittings, and other co"ections) in vessels, piping, pumps, vah/es, and other components of any size in all systems except high pressure safaty injection systems of pressurized water reactor plants.
(d) Component connections NPS 1 % 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 pressunzed water reactor plants.
(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.
O Fde: SECT 03 TXT-Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT 4#
1 THIRD INSERVICE INSPECTION INTERVAL Rev. O v INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 34 of 3 11 3.1.2 IWC 1222 Components within systems (or portions of syster es) other than RHR, ECC and Charging Systems (a) Vessels, piping, pumps, valves, and other components NPS 4 and smaller.
(b) Component connections NPS 4 and smaller (including nozzles, socket frttings, 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 degrees F.
(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.
3.1.3 IWC.1230 Concrete Encased Components Piping support members and piping support components that are encased in concrete shall be exempted from the examination requirements of IWC 2500.
3.2 Component / Piping Examination Development Class 2 components subject to examination are identified in Appendix B. The Class 2 Summary Tables satisfy the requirements of IWA-2420 (a) (1) through (6) respectively.
A narrative discussion of Class 2 components subject to examination and testing are desenbed in detail below:
3.2.1 Category C-A, Pressure Retaining Welds in Pressure Vessels item C1.10 ShellCircumf2tentialWelds Scope of Examination: 100% of all welds at gross structural discontinuities only. The examinations are limited to one vessel among a group of vessels. Components applicable to this examination category are the two (2) RHR Heat Exchangers 10E-2A and 10E 2B.
(4) C1.10 Shell Cire. Welds, (2) welds selected O
File: SECT 03.TXT Et ,
JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT tv THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Dcte: January 6,1998 PROGRAM Page 35 of 3 11 item C1.20 - Ht11.Chrdmferentia8 Welds Scope of examination: 100% of head-to-shell welds, (limited to one vessel of multiple vessels). Components applicable are the (2) RHR Heat Exchangers (10E-2A and 10E-28) and tha (2) Control Rod Drive Scram Tanks (03TK-1 A and 03TK-18).
(6) C1.20 Head Cire. Welds, (3) welds selected llem C1.30 Tubesheet to SheliWelds Scope of examination: 100% of Tubesheet to shell welds (limited to one vessel of multiple vessels).
(0) C1.30 Not applicable to JAF, located on the Class 3 side.
3.2.2 Category C 8, Pressure Retaining Nozzle Welds in Vessels item C2.10 and C2.11 Nozzles in Veitels < % in. Nominal Thickness i \
V Components applicable to (2) RHR Heat Exchangers (10E 2A and 10E-28) and (2)
Coatrol Rod Drive Scram Tanks (03TK-1 A and 03TK 1B).
(0) C2.11 Not applicable to JAF llem C2.20 Nozzles Without Reinforcina Plate in Vengis > % in. Nominal Thickness Components applicable to (2) RHR Heat Exchangers (10E 2A and 10E-2B) and (2)
Control Rod Drive Scram Tanks (03TK-1 A and 03TK 1B).
Item C2,21 - Nozzle to Shell or Head Welda Scope of Examination - All nozzles at terminal ends of piping runs (limite ' *.o one vessel of mutiple vessels). Includes only those piping runs selected for exe.nination under Examination Category C-F.
(6) C2.21 Nozzle to Shell or Head Welds, (3) required Ifem C2.22 Nozzle Inside Radius Section Scope of Examination - All nozzles at terminal ends of piping runs (limited to one vessel of multiple vessels).
C2.22 Nozzle Inside Radius Section, (1) required
{}
V (2)
Note: The RHR Heat Exchanger Nozzles do not have inner radius sections.
File: SECT 03.TXT.E1
l Imusume i JAMES A. FITZPATRICK JAF ISIM02 l NUCLEAR POWER PLANT
- > Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 l i
PROGRAM jPage 36 of 3 11
)
ltem C2.30. C2.31. C2.32. & C2.3
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
ltem_Q2.31 - Reinforcina Plate Welds tgE.gzzle and Vessel Scope of Examination All nozzles at terminal ends of piping runs (limited to one vessel of multiple vessels).
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
Item C2.32 Nozzle to Shell for Heitdl Welds When inside of VesstLis Accessible > % inch
- Not applicab'e t3 James A. Fitzpatrick Nuclear Power Plant.
[tgrn C2.33 - Nozzle to Shell_ wr Head) Welds When inside of Vessel is inaccessibli Scope of Examination Visual VT-2 of tell tale hole in reinforcing plates (limited to l]
V one vessel of multiple vessels), Examination performed in accordance with system pressuie test program.
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
3.2.3 Category C C, integral Attachments for Vessels, Piping, Pumps & Valves Examinations will be performed in accordance with the attemate requirements of Code Case N 509, *Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" Note: in addition to those conditions specified in the Code Case: A minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
Subject to Relief Request 4 ltem C3.10 - Pressure Vessels. Intearally Welded Attachments Scope of Examination - 100% of the length cf the attachment weld of only one integrally welded attachment of only one of the multiple vessels selected.
(24) C3.10 Integral Welded Attachments, (2) selected
,rx
,v )
Fate: SECT 031 AT-E1
I JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] l
NUCLEAR POWER PLANT )
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 37 of 3 11 llem C3.20 Pipina. Integrally Welded Attachments Scope of Examination - 100% of the length of tne attachment weld of 10% of the welded att:chments associated with the component supports selected for examination of components examined under C F 1 and C-F 2. Multiple component concept is not applicable.
- (210) Integrally welded attachments, (25) scheduled item C3.30 - Pumns. Intearally Welded Attachments Scope of Examination 100% of required areas of each welded attachment (limited to attachments of components examined per C-F and C-G).
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
(~T ltem C3.40 - VallY1gJniggrally Welded Attachments b Scope of Examination 100% of required areas of each welded attachment (limited to attachments of those components required to be examined under Examination Categories C-F and C-G).
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
3.2.4 Category C D, Pressure Retaining Bolting > 2"in Diameter
[tems C4.10 Pressure Vess.gj Scope of Examination - 100% bolts and studs at each bolted connection of components required to be intpected. The examination of bolting for vessels may be performed on one vessel in a group of vessels.
(144) RHR Ht. Exch. Bolting, Excluded per IWC-2500-1 size <2.0" Dia.
Items C4.20. C4.30. & C4.40
- Not applicable to James A. Fitzpatrick Nuclear Power Plant.
l l
I l
l t
rM l \ h 1 </
Fue: SECT 03.TXT Et
JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT G h tty THIRD INSEFVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 38 of 3 11 3.2.5 Category C F 1, Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping items C5.10. C5.11. C5.12. C5.20. C5.21. C5.22. C5.30. C5.40. C5.41 & C5.42
~
Welds are selected for examination as defined below. Refer to Appendix B for a summary detail of welds selected for examination.
(1) Requirements for examination of welds in piping 5 NPS 4 ap#v to PWR high pressure safety injection systems in accordance with the exemption criteria of IWC-1220.
f2) The welds selected for examinstion shall include 7.5%, but not less than 28 welds, of all austenitic stainless steel of high alloy welds not exempted by IWC 1220. (The Category Total includes (0) pipe to pipe welds, not exempted by IWC 1220, and are not required to be nondestructively examined per Examination Category C-F-1). These welds, however, were included in ue total weld count to which the 7.5% sampling rate was
{
x apphed). The total welds selected for examination is based on adding non-exempt circumferential welds to exempt circumferential welds and multiplying by 7.5%.
- (12) C5.11 Cire. Welds, Excluded IWC-25001, (3) selected (0) C5 41 Cire. Welds, Excluded IWC-2500-1 (12) x 7.5% = 0.9 or (1) min., (3) scheduled, see note for selection criteria The examinations shall be distributed as follows:
(a) The examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of non-exempt austenitic stainless steel or high alloy welds in each system (i.e., if a system contains 30% of the neal-exempt welds, then 30% of the nondestructive examinations required by Examination Category C F 1 should be performed on that system);
(b) Within a system, the examinations shall be distributed among terminal ends and structural discontinuities [See Note (3)) below prorated, to the degree practicable, on the number of non-exempt terminal ends and structural discontinuities in that system; and
(")
(j (c) Within each system, examinations shall be distributed between line sizes prorated to the degree practicable.
Fde: SECT 03.TXT Ei
I
=
JAMES A. FITZPATRICK JAF.lSI-0002 NUCLEAR POWER PLANT
- > YorkPower G & rhy THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM 1 Page 39 of 3 11 (3) Structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc., conforming to ANSI 016.9), and pipe branch connections and fittings.
(4) The welds selected for examination shall be reexamined dering subsequent inspections over the service lifetime of the piping component.
Note: Since all welds within this Examination Category are excluded from examination, no examinfjons are required. The (3) welds selected are for monitoring potential pump discharge vibration.
3.2.6 Category C-F 2, Pressure Retaining Welds in Carbon or Low Alloy Steel Piping items C5.$0. C5.61 & C5.62
. Not applicable to James A. Fitzpatrick Nuclear Power Plant.
O
'd items C5.50. C5.51. C5.52. C5.70. C5.80. C5.81 & C5.8.2 Welds are selected as defined below. Refer to Appendix B for a complete summary of welds selected for examination.
(1) Requirements for examination of welds in piping 5 NPS 4 apply to PWR high pressure safety injection systems in accordance with the exemption criteria of IWC-1220.
(2) The welds selected for examination shall include 7.5%, but not less than 28 welds, of all carbon and low alloy steel welds not exempted by IWC 1220.
(Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Category C-F-2. These welds, however, shall be included in the total weld count to which the 7.5%
sampline rate is applied).
Non-Exempt Welds (752)
Excluded Welds (329)
(1081) x 7.5% = (81.0) or (81) minimum required, 90 scheduled The examinations shall be distributed as follows:
p
()
Flie: SECT 03.TXT Ei
I l
I
. LAMES A. FITZPATRICK JAF ISI 0002 NUCLEAR POWER PLANT
- D fjewWrkPower S 4# nuthortty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 3 10 of 3 11 (a) The examination shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of non-exempt carbon and low alloy steel welds in each system (i.e., if a system contains 30% of the non-exempt welds, then 30% of the nondestructive examinations iequired.by Examination Category C-F 2 should be performed on that system);
(b) Within a system, the examinatinn shall be distributed among terminal ends and structural discontinuities [See Note (3) below) prorated, to the degree practicable, on the number of non-exempt terminal ends and structural discontinuities in that system; and (c) Within each system, examinations shall be distributed between line sizes prorated to the degree practicable.
(3) Structural discontinuities include pipe weld joints to vessel nozzles, valve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges,
(' etc., conforming to ANSI B16.9), and pipe branch connections and fittings.
C Tha welds selected for examination shall be reexamined during subsequent (4) inspection intervalt over the service lifetime of the piping component.
(5) Only those welde showing repertable preservice transverse indications need to be examined for transvere reflectors.
3.2.7 Category C G, Pressure Retainir.g Welds in Pumps and Valves items C6.10 & C6.20
- Not applicable to James A Fitzpatrick Nuclear Power Plant. There are no pressure retaining welds in pumps and valves.
3.2.8 Category C H, All Pressure Retaining Components llems C7.10. C7.20. C7.30. C7.40. C7.50. C7.60. C7.70 & C7.80 Scope of Examination - The pressure retair,ing components within the Class 2 system boundaries are subjected to system pressure tests in accordance with IWC-5210 and visually examined (VT 2) per IWA-5240. The tests are conducted as follows:
]
O Fite SECT 03 TXT E1
JAMES A. FITZPATRICK JAF ISI-0002 i NUCLEAR POWER PLANT S h ty THIRD INSERVICE INSPECTION INTERVAL Rev. O j INSERVICE INSPECTION Date: January 6,1998 PROGRAM l Page 3 - 11 of 3 - 11 (a) System Functional Test lWA-5211(b)- For those systems or portions of systems not required to operate during normal reactor operation, Dut for j which periodic system cr component functional tests are performed as required by the Plant Te.:hi al Specifications and/or the Pump and Valve (IST) program, a VT 2 examination is performedat least once during each period during the, system or component functional test. The boundary subject to prersurization during a System Functional Test includes only those pressure retaining components within the system boundary pressurized under the test mode required during the performance of the periodic system (or componant) functional test. Nominal operating pressure of the system functional test is acceptable as the system test pressure.
(b) A System Hydrostatic Test. will be performed in accordance with alternate examination techniques of Code Case N-4981,'Altemative Rules for 10 Year Hydrostatic Pressure Testing for Class 2 Systems,Section XI, Division 1", as noted below:
A As an alternative to the 10-year Hydrostatic Pressure Test required by Table U IWC-25001, Category C-H, JAF will perform the following:
(1) A system pressure test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of inspection Program B.
(2) All ASME Section XI Pressure Testing requirements are controlled by site procedures. The boundary subject to test pressurization during the system pressure test shall extend to all Class 2 components included in those portions of systems required to operate or support the safety system function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required.
(3) Prior to performing VT 2 visual examination, the systern shall be pressurized to nominal operating pressure for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for non-insulated systems.
The system shall be maintained at nominal operating pressure during the performance of the VT 2 visual examination.
(4) The VT 2 visual examination shall include all components within the boundary identified in (2) above.
,o (5) Test instrumentation requirements of IWA-5260 are not applicable.
i, 1 Subject to Relief Requests 3 and 10 De SECT 03 TXT-E1
JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT
& THIRD INSERVIC,E INSPF.CTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 41 of 4 13 TABLE OF CONTENTS ,
SECTION 4 Table of Contents . ...... .... ... .. .. .. . . . 4-1 Record of Revision ... . ... .. . .... . .. . . . .. 42 4.0 CLASS 3 SYSTEMS / COMPONENTS . . .. . .. , .43 4.1 ASME Code Exemptions . . . . .. .. . . . .. . .. . .. 4-3 4.1.1 IWD-1220.1 . .. .. ... .. .. . ..... . ... . . .. . 4-3 4.1.2 IWD 1220.2 .. . .... . . . .. . . .. . 43 4.2 Component / System Examination Development . . . .. ... .. 4-3 4.2.1 Category D-A .. ... . . . . . . 4-4 l 4.2.2 Category D-B . . . . 4-6 1
4.2.3 Category D-C .. 49 4.2.4 System Pressure Tests - Class 3 . . .. . .. .. 4 12 l
Fae. SECTOLTXT.E1 l
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JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT ter thor ty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 42 of 4 13 RECORD OF REVISION 4 REVISION . .DATE W: i' iAFFECTEDi : REASON FOR REVislON : jh;( < s
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iPAGES' r 0 January 6,1998 Entire Updated Inservice inspection Program Plan Document for the 3* Ten Year Inservice Inspection interval r
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\_J Fh: SECT 04.TXT-Ei
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > gVorkPower THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 43 of 4 -13 4.0 CLASS 3 SYSTEMS / COMPONENTS The Class 3 system boundaries subject to examination and testing were developed based upon the requirements of Regulatory Guide 1.26, and the JAF FSAR. The Class 3. components anJ systems subject to examination and testing are described in detail below:
4.1 ASME Code Exemptions Employed 4.1.1 IWD 1220.1 Integral attachments of supports and restraints to components that are 4" nominal pipe size and smaller within the system boundaries of Examination Categories D-A, D B and D-C of Table IWD-2500-1 shall be exempt from the visual examination VT-3, except for the Auxiliary Feedwater System.
4.1.2 IWD 1220.2 O
V Integral attachments of supports and restraints to components exceeding 4" nominal pipe size may be exempted provided:
(a) The components are located in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling; and (b) The components operate at a pressure of 275 psig or less and at a temperature of 200"(93'C), or less.
4.2 Component / System Examination Development Class 3 components subject to examination are identified in Appendix C. The Class 3 Summar/ Tables satisfy the requirements of IWA-2420 (a) (1) through (6) respectively.
A narrative discussion of Class 3 components subject to examination and testing are described in detail below:
Note: Examination Categories and Examination item Numbers for Class 3 Integrally Welded Attachments are defined in accordance with the ASME Code Section XI, 1989 Edition, Article IWD, Table IWD-25001 in lieu of Code Case N 509 classification criteria.
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. Fde. SECTD4.TXT-E1 l
JAMES A. FIT 7. PATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT m
ter
- THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 44 of 4-13 4.2.1 C ategory D-A Systems in Support of Reactor Shutdown Function item D1.10 - Pressure 3etalnina Components
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Scope of Excminatic:: Perform System inservice Test and Visual (VT 2) examination on pressure retaining components within the system boundary containing operating pressure during system operation each inspection period.
Scope of Examination - Perform a System Hydrostatic Pressure Test and Visual (VT-
- 2) examination on pressure retaining components at or near the end of each inspection interval.
Paragraph 4.2.4, System Pressure Test, Closs 3 of this Section provides the attemate testing requirements of Code Case N-498-1, that JAF will implement during the third inspection interval.
O Subject to Relief Requests 3 and 10 Item D1.20 -Intearal Attachments to Component Supports and Restraints Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval (9) D1.20 Integral Attachments, (1) selected.
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the attemate requirements of Code Case N-509, "Altemative Rules for the Selection and Exemination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the Code Case: a minimum 10% sample of integrally welded attachments for each item in each Code Class per interva! will be examined.
Subject to Relief Request 4 Item D1.30 -Integral Attachments to Mechanical and Hydraulic Snubbers Scope of Examination - Perform Visual (VT-3) examination on all integrally welded
- f. S attachments each inspection interval.
- (0) D1.30 Integral Attachments, Not applicable to JAF. I l
Frier SECT 04.TXT-E1
l JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O g INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 45 of 4-13 Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the' alternate requirements of Code Case N-509,"Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the Code Case, a minimun.10% sample of integrally welded attachmeats for each item in each Code Class per interval will be examined.
ligm D1.40 -Intearal Attachments to Sprina Tvoe Sypports Scope of Examination - Perform Visual (VF-3) examination on all integrally welded attachments each inspection interval.
- (0) D1.40 Integral Attachments, Not applicable to JAF Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509,"Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the C'de Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
Item D1.50 -Intearal Attachments to Conpf ant Load Tvoe Suonorts Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
- (0) D1.50 Integral Attachments, Not applicable to JAF Note: In the case of multiple components within a system of similar design, function and service, the integral atfachment of only one of the multiple components shall be examined.
A
( ) Examinations will be performed in accordance with the alternate requirements of Code Case N-509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1",
Fde: SECT 04.TXT-E1
JAMES A. FITZPATRICK JAF lGI-0002 NUCLEAa POWER PLANT 4# ty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM t
Page 46 of 4 -13 In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
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ltem D1.60 -Intearal Attachment to Shock Absorbers Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
- (0) D1.60 Integral Attachments, Not applicable to JAF.
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only or.e of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of p Code Case N-509," Alternative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" V
In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
4.2.2 Category D-B Systems in Support of Emergency Core Cooling, Containment Heat Romoval, Atmosphtsre Cleanup, and Reactor Residual Heat Removal item D2.10 - Pressure Retainina Comoonents Scope of Examination - Perform a System Functional Test and Visual (VI'-2) examination on pressure retaining components within the system boundary at nominal operating pressure each inspection period.
Scope of Examination - Perform a System Hydrostatic Pressure Test and Visual (VT-
- 2) examination on pressure retaining components at or near the end of each inspection interval.
Paragraph 4.2.4, System Pressure Test, Class 3 of this Section provides the attemate testing requirements of Code Case N-498-1, tnat JAF will implement during the inspection interval.
Subject to Relief Requests 3 and 10 l}
V File: SECT 04,TXT E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT tv THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4-7 of 4-13 ltem D2.20-Intearal Attachments to Component Suonorts and RestraLott Scope of Exarnination - Perform Visual (VT-3) examination on all integrally weided attachments each inspection interval, ,
(119) D2.20 Integral Attachments (14) selected,12%
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be er<amined.
Examinations will be performed in accordance with the attemate requirements of Code Case N-509,'Altemative Rules for the Selection and Examinatbn of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" in addition to those conditions specified in the Code Case, a minimum 10% sample
(~3 of integrally welded attachments for each item in each Code Class per interval will V be examined.
Subject to Relief Request 4 Item D2.30 -Integral Attachments to Mechanical and Hydraulic Snubbers Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
(1) D2.30 Integral Attachment, selection per T.S. Section 3.6.1 Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance witn the altemate requirements of Code Case N-509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded Attachments,Section XI, Division 1" In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachmentc for each item in each Code Class per interval will be examined.
Subject to Relief Request 4 O
'j File: SECT 04.TXT-Ei
JAMES A. FITZPATRICK [[::JAF-lSI-0002|JAF-lSI-0002]] NUCLEAR POWER PLANT tsf rity THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 48 of 4-13 ftem D2.40 -Intearal Attachments to Sprina Tvoe Supports Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
- (3) D2.40 Integral Attachments, (1) selected 33%
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the attemate requirements of Code Case N-509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integtally Welded AttachmentsSection XI, Division 1" O in addition to those conditions specified in the Code Case, a minimum 10% sample
( of integrally welded attachments for each item in each Cr 'e Cl ass per interval will be examined.
Subject to Relief Request 4 Item D2.50 -Intearal Attachments to Constant Load Tvoe Supports Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
(0) D2.50 Integral Attachments, Not applicable to JAF Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509, *Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Weied AttachmentsSection XI, Division 1" in addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
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Fde: SECT 04.TXT E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT tv THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4-9 of 4-13 Item D2.60 -Integral Attachment to Shock Absorbers Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval. _
- (0) D2.60 Integral Attachments, Not applicable to JAF Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509,' Alternative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" in addition to those conditions specified in the Code Case, a minimum 10% sample n
g of integrally welded attachments foi each item in each Code Class per interval will be examined.
4.2.3 Category D-C Systems in Support of Residual Heat Removal from Spent Fuel Storage Pool item D3.10 - Pressure Retalnina Comconents Scope of Examination - Perform a System inservice Test and Visual (VT-2) examination on pressure retaining components within the system boundary containing operating pressure during system operation each inspection period.
Scope of Exaniination - Perform a System Hydrostatic Pressure Test and Visual (VT-
- 2) examination on pressure retaining components at or near the end of each inspection interval.
Paragraph 4.2.4, System Pressure Test, Class 3 of this Section provides the attemate testing requirements of Code Case N 498-1, that JAF will implement during the inspection interval.
Subject to Relief Requests 3 and 10 A
File: SECT 04.TXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT ty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4 - 10 of 4 13
((gm D3.20 -Integral Attachments to Com9onent Supports and Restraitas Scope of Examination - Perform Visual (VT 3) examination on all integrally Nelded attachments each inspection interval.
(16) D3.20 Integral Attachments, (2) selected,12%
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1",
n In addition to those conditions specified in the Code Case, a minimum 10% sample (j of integrally welded attachments for each item in each Code Class per interval will be examined.
Subject to Relief Request 4 Item D3.30 -Inteoral Attachments to Mechanical and Hydraulic Snubbers Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
(0) D3.30 Integral Attachments, Not applicable to JAF Note: in the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N 509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
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YI File: SECT 04.TXT E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]]
, NUCLEAR POWER PLANT O C WW'** THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4 - 11 of 4-13 ltem D3.40 -Intearal Attachments to Sprina Tvoe Suocorts Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval. ,
- (1) D3.40 Integral Attachment, (1) selected,100%
Note: In the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the altemate requirements of Code Case N-509,"Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" in addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will (q) be examined.
Subject to Relief Request 4 Item D3.50 -Intearal Attachments to Constant Load Type Suooorts Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
- (0) D3.50 Integral Attachments, Not applicable to JAF.
Note: In the case of multiple components within a system of similar design, function ar.d service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance vvith the altemate requirements of Code Case N-509,"Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
Item D3.60 -Intearal Attachment to Shock Absorbers (g)
Scope of Examination - Perform Visual (VT-3) examination on all integrally welded attachments each inspection interval.
File: SECT 04.TXT-Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER Pl. ANT 4# Rev.
THIRD INSERVICE INSPECTION INTERVAL O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4 - 12 of 4 13
- (0) D3.60 Integral Attachments, Not applicable to JAF.
Note: in the case of multiple components within a system of similar design, function and service, the integral attachment of only one of the multiple components shall be examined.
Examinations will be performed in accordance with the alternate requirements of Code Case N-509, "Altemative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1" In addition to those conditions specified in the Code Case, a minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined.
4.2.4 System Pressure Tests - Clau 3 r-~S The pressure retaining components within the boundary of each system specified for Examination Categories D-A,0-8 and D-C are pressure tested and visually
() examined (VT-2), for leakage during the following tests:
(a) Svstem inservice Test, IWA-5211 (c) - For systems required to operate during normal plant operation, a VT-2 examination is conducted at least once each period while the system is in operaNon and at operating pressure.
The boundary subject to test pressunzation during a System Inservice Test extends to those pressure retaining components under operating pressures during normal system operation.
or (b) System Functional Test, IWA-5211(b)- For those systems or portions of systems not required to operate during normal reactor operation, but for which periodic system or component functional tests are performed as required by the Plant Technical Specifications and/or the Pump and Valve (IST) program, a Vl-2 examination is performed at least once durina each period during the system or component functional test. The boundary subject to pressurization during a Svstem Functional Test includes only those pressure retaining components within the system boundary pressurized under the test mode required during the performance of the periodic system (or component) functional test. Nominal operating oressure of the System Functional Test is acceptable as the system test pressure.
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(c) System Hydrostatic Test. will be performed in accordance with alternate examination techniques of Code Case N-498-1, "Altemative Rules for 10-File: SECT 04.TXT-E1
JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT 4sr ity THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 4 - 13 of 4 13 Year Hydrostatic Pressure Testing for Class 3 Systems,Section XI, Division 1" as noted below:
As an attemate to the 10-year Hydrostatic Pressure~
Test required by Table IWD-2500-1, Category D-A, D-8, and D-C, JAF wi!! perform the following:
(1) A system pressure test shall be conducted at or near the end of each inspection interval or during the same inspection period of each inspection interval of Inspection Program B.
(2) The boundary subject to test pressurization during the system pressure test shall extend to all Class 3 components included in those portions of systems required to operate or support the safety system function up to and including the first normally closed valve, including a safety or relief valve, or valve capable of automatic closure when the safety function is required.
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- j (3) Prior to performing the VT-2 visual examination, the system shall be pressurized to nominal operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated systems and 10 minutes for non-insulated systems. The system shall be maintained at nominal operating pressure during the performance of the VT-2 visual examination.
(4) The VT-2 visual examination shall include 6ll components within the boundary identified in (2) above.
(5) Test instrumentation requirements of IWA-5260 are not applicable.
Subject to Relief Requests 3 and 10 All ASME Section XI Pressure Testing requirements are controlled by site procedures. The boundaries subject to system pressure tests (Functional, inservice and Hydrostatic), are shown on ISI Flow Diagrams. (See Appendix G).
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File: SECT 04.TXT-Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > lyewYorkPower 4# Authortty 9 THIRD INSERVICE INSPECTION INTERVAL INSERVICE INSPECTION Rev. O Date: January 6,1998 PROGRAM Page 5-1 of 5-5 1 TABLE OF CONTENTS
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SECTION 5 Table of Contents . . 5-1 Record of Revision . . . 5-2 5.0 CLASS 1,2 AND 3 COMPONENT SUPPORTS -lWF . . 5-3 5.1 Supports Exempt From Examination . . . . . 5-3 5.2 Support Examination Development . 5-3 O
V 5.2.1 Class 1 Component Supports . . 5-3
- 5.2.2 Class 2 Component Supports . .. . 5-3 5.2.3 Class 3 Component Supports . 5-3 5.3 Narrative Discussion .. . . . 5-4 5.3.1 Examination Category F-A Supports , . . 5-4 5.4 Snubbers . . . . . . 5-5
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Fit SECT 105.TXT Ei
l JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > WrkPower
& rity G THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 52 of S-5 RECORD OF REVISION 1REVISIOND > OdTEi !NFFECTED$ 4
/REAS N FOR REVISION l* ' $d F
' NOlJ bc - .PAGES ! eg
< :_y 0 January 6,1998 Entire Document Updated Inservice inspection Program Plan for the 3" Ten Year Inservice Inspection Interval 1
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l FC: SECT 105.TXT E1
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > VorkPower O 4#
Q THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998
- PROGRAM Page 5-3 of 55 5.0 CLASS 1, 2 AND 3 COMPONENT SUPPORTS lWF Compone,nt supports selected for examination shall be the supports of those components that are required to be examined under IWB, IWC and IWD. JAF will conduct examinations in accordance with attemate examination requirements of Code Case N 491-1,"Altemative Rules for Examination of Class 1,2, 3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1", as noted below; As an alternate to the Class 1,2 and 3 Component Support Requirements of Table IWF-2500-1, JAF will perform the following
Class 1, 2 and 3 supports receive a Visual (VT-3) examination to determine their general mechanical and structural condition, and when required, conditions relating to their operability. The supports subject to
! examination have been selected in accordance with Code Case N-491-1. (Refer to Appendix D for Detail Tables).
5.1 Supports Exempt From Examination Exemptions are as stated in IWB-1220, IWC-1220 and IWD-1220, (Sections 2, 3 and 4 of this Program, respectively),
- a. In addition, portions of supports that are inaccessible by being encased in concrete, buried underground, or encapsulated by guard pipe are also exempt from the examination requirements.
- b. NYPA has determined that a support that does not fully meet the definition of a component support, as defined within ASME Section XI, Article IWA-9000, Glossary definition for Component Support, is exempt for examination. Pipe whip restraints, insulation lugs, or unused pipe supports, which do not provide structural stability or support the weight of the pipe, are exempt.
5.2 _ Support Examination Development 5.2.1 Class 1 Component Supports Class 1 component supports subject to exar"ination are identified in Appendix D.
5.2.2 Class 2 Component Supports Class 2 comronent supports subject to examination are identified in Appendix D.
5.2.3 Class 3 Component Supports Class 3 component supports subject to examination are identified in Appendix D.
Fik SECT 105.TXT-Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT e 4#
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE 6WPECTION Date: January 6,1998 PROGRAM Page 5-4 of 5-5 The Class 1,2 and 3 Summary Tables satisfy the requirements of IWA-2420 (a) (1) through (6) respectively.
5.3 Narrative Discussion A narrative discussion of Class 1,2 and 3 component supports subject to examination are described in detail below:
In order to assure that a representative sample of supports within each Code Class is examined.
(Code Examination Category F-A, Examination item Numbers F1.10 Class 1, F1.20 Class 2, F1.30 Class 3, and F1.40 other than piping), selection was based on Class, System and Size, to the extent practical '
5.3.1 Examination Category F A Supports ltem F1.10 - Class 1 Ploino Suonorts iO Scope of Examination -Visual VT-3 examination of 25% of all non-exempt Class 1 Supports.
(85) F1.10 Supports, (21) required, (28) scheduled, 32%
Item F1.20 - Class 2 Ploina Suncorts Scope of Examination - Visual VT-3 examination of 15% of all non-exempt Class 2 Supports.
(313) F1.20) Supports,46.9 (47) required, (51) scheduled,16%
' tem F1.30 - Class 3 Ploina Supports Scope of Examination -Visual VT-3 examination of 10% of all non-exempt Class 3 Supports
- (329) F1.30 Supports, (33) required, (41) scheduled,12%
Item F1.40 - Suncorts Other than Pioina Sucoorts (Class 1. 2. 3 and MC)
Scope of Examination -Visual VT-3 examination of 100% of all non-exempt Supports.
- (36) F1.40 Supports, (18) required, (18) scheduled, 50%
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' I All component supports subject to examination have been classified (a, b, c, d, etc.), to the extent practical. As these supports could be classified by one or more of the suffixes for the same support. only one suffix was selected. These classifications are identified in the 10-year inspection Tables.
File: SECT 105.TXT E1
xw? -
JAMES A. FITZPATRICK [[::JAF-ISI-OF|JAF-ISI-OF]].'
NUCLEAR POWER PLANT
- > NewYorkPower 4# Authority G THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 55 of 5-5 Note: The recording of Hot or Cold positions will be performed in conjunction with the VT-3 examination .
5.4 Snubber Examination and Performance Testing Program The following section provides a description of James A. FitzPatrick's Snubber Program for Examination and Performance Testing of Dynamic Restraints (Snubbers).
The Snubber Program has been revised in accordance with the ISI Program update requirements to institute ASME Section XI,1989 Edition no Addenda, utilizing OMa 1988 Part 4.These changes are based on guidance in NUREG-1433, NEDO-31466, criteria of 10CFR 50.36, and are consistent with USNRC and industry efforts to simplify Technical Specifications. NYPA has submitted a Technical Specification Amendment (letter, JPN 96-051, dated November 26,1996, James A. FitzPatrick Nuclear Power Plant Docket No. 50-333, Proposed Changes to the Technical Specifications to Relocate Requirements for Snubber to Plant Controlled Documents (JPTS-96-001).
c The Presevice Examination Requirements detailed in OMa-1988 Part 4, Para. 2.2 Thermal Movement
( Examination have been evaluated and considered preoperational construction requirements not
\ applicable to the JAF Snubber Program.
Snubber Program compliance for general and specific requirements along with exemptions from OMa-1988 Part 4 will be defined within applicable plant procedures. Relief from specific requirements within OMa-1988 Part 4, are located in Appendix F, Relief Request # 11.
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LY Fb: SECT 105.TXT-Ei l
l JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT 4# ty THIRD INSERVICE INSPECTIOtt INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-1 of 6 12 TABLE OF CONTENTS SECTION 6 -
Table of Contents . . .. . , 6-1 Record of Revision . . .. . . . . . .. . 6-2 6.0 AUGMENTED EXAMINATIONS .. . .. . . ., . 6-3 6.1. Augmented Reactor Pressure Vessel Examinations .. . .. .. . 6-3 6.1.1 Reactor Pressure Vessel and Closure Head . . . . .. . .. 6.3 6.1.2 Applicable Welds Affected . . .. . . . 6-3
( 6.2. Augmented IGSCC Examinations . 6-4 6.2.1 Categonzation Process . 6-4 6.2.2 Inspection Schedule . , .. . . . . . . 6-6 6.2.3 Sample Expansion . . . 6-7 6.3 Main Steam and Feedwater Augmented Examinations . . . 6-7 6.4 NUREG 0619 BWR Feedwater Nozzle and CRD Return Line Nozzle Cracking 6-8 6.5 In-VesselVisual Augmented Examinations . . , 6-10 6.6 Core Spray Augmented Examinations . . . 6-12 List of Tables 6.1 Applicable Welds Affected . . .. .. .. ... . . . . .. . 6-3 6.2. IGSCC Exam Requirements . .. . . . .. . .. 6-7 6.3 Main Steam and Feedwater Augmented Examinations . . . 6.8 6.4 Augmented Feedwater Nozzle Examinations . . . . .. . . 6-9 6.5. In-Vessel Augmented Examinations . .. . .. .. . 6-10 6.6 Core Spray Augmented Examination Welds . . . . 6-12 im l List of Figures x.s) 6.6 Feedwater Nozzle Examination Zones . . .. . 6-9 File: SECT 06.TXT-E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- 1> Power THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM '
Page 6-2 of 6-12 RECORD OF REVISION iREVISIONj D
- DATE ? Oc (:AFFECTED5 @ NEhSON FORLREVISIONi ~
~' ' '
iNoi 5 :ydjz , ,
+
cLPAGES!: ' eg *i
- E, O January 6,1998 Entire Document Updated inservice inspection Program Plan for the 3" Ten Year Inservice inspection Interval C
O File: SECT 06.TXT-Ei
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > Power
. THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 63 of 6-12 6.0 AUGMENTED EXAMINATIONS The following section provides a detailed description of the James A. Fitzpatrick Inservice Inspection Program Basis for Augmented Examination of additional components / systems. NYPA plans on utilizing where applicable, Augmented Examination results to satisfy the requirements of ASME Section XI. See Appendix E for Summary Tables of Augmented Examinations.
Class 1 Augmented Examinations 6.1 Augmented Reactor Pressure Vessel Examinations 6.1.1 Reactor Vessel (including the Closure Head)
- a. Regulatory Guide 1.150 Examinations are performed in accordance with the ASME Code,Section XI and the additional requirements of Regulatory Guide 1.150, Rev.1.
- b. New Regulation 10 CFR 50.55a(g)(6)(ii)(A)2 The change in the regulation affected the Second Ten-Year Inservice Inspection interval.
NYPA received approval from the USNRC to defer the essentially 100% volumetric examination untill the first refueling outage, of the first inspection period, of the third inspection interval.
As required by 10 CFR 50.55a(a)(g)(6)(ii)(A), NYPA shall augment the reactor vessel examination by implementing once, as part of the inservice inspection interval in effect on September 8,1992, the examination requirements for reactor vessel shell welds specified in Examination item B1.10 of Examination Category B-A, " Pressure Retaining welds in Reactor Vessel."in Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI.
6.1.2 Applicable Welds Affected TABLE 6 - 1 APPLICABLE WELDS AFFECTED Exam Examination item Description Number of Welds Remarks / Comments item No. Affected B1.11 Circumferential Shell Welds 4 100% of the Weld Length B1.12 Longitudinal Shell Welds 12 100% of the Weld Length O
Nj 2
Deferral approved by the USNRC, TAC M87158, dated 03/08/94.
File: SECT 06.TXT E1
JAMES A.FITZPATRICK [[::JAF-lSI-0002|JAF-lSI-0002]] NUCLEAR POWER PLANT
- > NewWrkPower tv Authority Th ID INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-4 of 6-12 All augmented examinations will be performed during the First Outage, of the First Period, of the Tni-d Inspection Interval. Credit will be taken for these examinations to satisfy the Section XI 1989 Edition requirements for the Third Inspection Interval.
6.2 Augmented IGSCC Examinations The James A. Fitzpatrick Technical Specifications, Section 4.6.F.3, requires NYPA to implement an augmented inspection program for those welds designated as IGSCC susceptible. The requirements for an augmented IGSCC inspection program are mandated by Generic Letter GL 88-01, GL 88-01 Supplement 1,
- lntergranular Stress Corrosion Cracking in BWR Austenitic Stainless Steel Piping" and NUREG-0313 Revision 2, ' Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressuro Boundary Piping."
6.2.1 Categorization Process
- 1. IGSCC Category A Weldments V Category F Sidments are those welds with no known cracks, that have a low probability of incurring lbCC problems, because they are made entirely of IGSCC resistant material or have been solution heat treated after welding. NYPA has further defined those welds in Category A by using the following suffixes:
Scope of Examination - IGSCC Category A welds are inspected in accordance with a schedule similar to that called out for in Section XI. A 25% sample of welds shall be examined during this inspection interval.
Cateaorv A : Identifies welds which are fabricated from resistant materials.
(23) Category A Welds, (6) required Q1gqory A-1 : Identifies lonaitudinal seam welds (163) Category A-1 Welds, (41) required.
Note: This sample includes only those longitudinal seam welds that intersect a circumferential weld.
Cateaorv A* -
ldentifics sween-o-let welds that have been solution annealed.
(8) Category A* Welds, (2) required bh L._,I Fue: SECT 06.TXT-Ei
x JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > VDrkPower THIRD INSERVICE INSPEC TION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page f-5 of 6-12
- 2. IGSCC Category B Weldme':ts Category B weldments are thow welds made of resistant materia s, but h:We had an SI performed edher befcro service or within two years of operation.
There are no welds in this category at James A. Fitzpatrick Nuclear Power Plant.
- 3. IGSCC Category C Weldments Category C weldnents are those welds not made of resistant naterials, and have been given an Si process after more than two years of operation. NYPA has further defined those welds in Category C by using the following suffixes:
Scope of Examination - All welds shall be laspected once dunng this inspection interval.
Cateaorv C-2 - Identifies welds alven an Sl process after more than two ve1[1 of operatiqn,
'd -
(57) Category C-2 Welds, (57) required Cateaorv C* - Identifies welds treated with a Resistance Heatina Stress Imorovement (RHSI) orocess after more than two years of operation.
(2) Category C* Welds, (2) required Cateaorv C-3 - jdentifies welds alven an Si orocess after more than two years of operation and havina service stress over 1.0 SM. Reference NUREG 0313. Rev. 2. Section 4.5.
(3) Category C-3 Welds, (3) required.
Note: During Refueling Outages 1988 and 1990,100% of all NUREG 0313, Category C welds were inspected. Based on clarification by the USNRC regarding Extent and Frequency requirements, this category falls under an Augmented 10-Year inspection Interval, which began in 1992 and will be completed during the scheduled refueling outage in 2002.
- 4. IGSCC Category D Weldments Category D weldments are those welds not made with resistant materials, and have not been given an Si treatment, but have been examined and found to be free of cracks.
? ' tm i
-) included in this category are all bimetallic nozzle weldments made with non-resistant material and 182 inconel weld butter.
File: SECT 06.TXT Ei
JAMES A. FITZPATRICK [[::JAF-lSI-0002|JAF-lSI-0002]] NUCLEAR POWER PLANT 4 NewYorkPower
& Authority THIRD INSERVirE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-6 of 6-12 Scope of Examination - All welds shall be examined at least every two refueling outages.
Approximately 50% of all Category D welds shall be examined each refueling outage.
(28) Category D We!ds, (28) required, (14) each outage S. IGSCC Category E Weldments Categxy E weldments are those welds with known cracks that have been reinforced by an acceptable weld overlay or have been mitigated by an Si treatment welding. NYPA has further defined those welds in Category E by using the following suffixes:
Scope of Examination - should be inspected once every two refueling outages after repair.
Appro.vimately 50% shall be inspected during the first refueling outage and subsequent outages.
Catecorv E : All welds included in this cateaorv are weld overlavs.
O Category E Welds, (23) required, (11) each outage.
(23)
Cateoorv E' - Welds deslanated for an increased inspection freauencv.
(2) Category E' Welds, (2) required. (1) each outage Note: There are two welds in the E* Category (12-02-2-13 and 12-02-2-66).
- 6. IGSCC Category F Weldments.
l Category F weldments are those welds with known cracks that have been approved by analysis for limited additional service without repair.
There are no welds in this Category at James A. Fitzpatrick Nuclear Power Plant.
- 7. IGSCC Category G Weldments Category G weldments are those welds not made of resistant materials, have not been given an Si treatment.
There are no welds in this Category at James A. Fitzpatrick Nuclear Power Plant.
l 6.2.2 Inspection Schedule i
! The extent and frequency of inspection for various weldment categories are detailed below.
U File: SECT 06.TXT-E1
1 JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-7 of 6-12
@ .._ ... _ _.ITABLE[6y2Q . .. f . . . ,
[,
- m:; ilG8CC EXAMINATION REQUIREMENTSj ,,
flGSCdi ., . }EXhMINATION( ,' EXTENT-OFj og 8REMhRES$
iCATEGORY) ? REQUIREMENTS?
-- k EX/J8INATIONp: ' si A 25% Every 10 Year interval At least 12% in 6 years B 50% Every 10 Year Interval At least 25% in 6 years C All Within Two Refueling At least 50% in 6 years Cycles after the Post-Si inspection, and All Every 10 Years thereafter D All Every Two Refueling 50% each refueling outage Outage E All Every Two Refueling 50% each refueling outage (Vr') Outages F All Every Refueling Outage G All Next refueling Outage 6.2.3 Sample Expansion if one or more cracked welds in IGSCC Categories A B, or C, are found by a sample inspection during the Third Ten-Year Interval, an additional sample of welds should be inspected. Specific expansion requirements for IGSCC welds are defined in Section 8.0.
6.3 Main Steam and Feedwater Augmented Examinations An Augmented inservice Examination program as required by NYPA Technical Specifications, Section 4.6.F.2 is required for those high stressed circumferential pipe joints located in the Main Steam and Feedwater systems in lines larger than 4.0 in. in diameter, where no restraint against pipe whip is provided.
Scope of Examination - 100% inspection of these welds each inspection interval.
i
- g
.Y File: SECT 06.TXT-Ei
! l JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT 4# Authority -
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-8 of 6 12
.__. 1, jable 6s 3i & , . . .
LMAIN STEAM and FEEDWATER AUGMENTED EXAMINATIONSh _o Main' Steam 4 Main Stem'mi Feoowaten .^
FeedwaterY MSK 30314 N MSK 30325 D MSK-30332 LMSK-3034S 24 29-541 24-29-588 12-34-368 12-34-402 24-29-557 24-29-584 12-34-369 12 34-403 24-29-552 24-29-589 12-34-370 12-34-404 24 29-553 24-29-605 12-34-371 12 34-405 24 29-559 24-29-630 12-34-375 12-34 4 09 24 29-572 12-34-373 12-34-406 O 12 34-372 12-34-407 18-34-374 18-34-408 18-34-382 18-34-416 18-34-389 16-34-423 18-J4-391 18-34-425 Note: In addressing the augmented examination requirements of those portions of systems, pipe-to-pipe welds and longitudinal seams are required to be examined. Additionally, the Code boundary is extended past the Code Class boundary (MSIV) to the first restraint. The welds and supports thus affected are specifically identified through Examination Notes in the Main Steam and Feedwater Examination Tables located in the Ten-Year Inspection Plan.
6.4 NUREG 0619 BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking NYPA committed to a program of periodic examinations of the (4) Feedwater Nozzles /Feedwater Sparger Assemblies in accordance with the requirements of NUREG 0619, "BWR Feedwater Nozzle and Control Rod Drive Retum Line Nozzle Cracking". In compliance with this NUREG, NYPA has removed the cladding from the feedwater nozzles and cut and capped the CRD Retum Line. In addition, NYPA has submitted a Request for Relief number 7 from performing Liquid Penetrant (PT) examination of the feedwater nozzle bore and nozzle blend radius. The Visual Examination of the Control Rod Drive Retum Line Nozzle shall continue to be performed in accordance with NYPA Letter, JPN-83-64, dated 07/07/83.
%/
File: SECT 06.TXT-Et
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- 5 Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 69 of 6-12 au Scope of Examination - The zones to be eramined by the ultrasonic techniques shall be the regions as shown for identification purposes in Figure 6-5.
' ~
~
.. ^'l . . JTable8l4i 9& ; , $ ; yf^$f :l l t' AUGMENTED F.EEDWATER Nf1778 F EXAMINATION): M . ;g 9 FeedwktelNozzlSS !Nozzlebldeb ExamiriatidrIMethod[: * $itent snd Frequelncp; ,>
llys I,
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'IdentifH::ation ? s
%k N4-A 12.0" Automated UT Techniques At or near the end of interval N4-B 12.0" Automated UT Techniques At or near the end of interval N4-C 12.0" Automated UT Techniques At or near the end of interval N4-D 12.0" Automated UT Techniques At or near the end of interval Feedwater Sparger 12.0" Visual VT-3 Examination Every 4*. Refuehng Outage O
V Assemblies and Feedwater Nozzles yessei-l//
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Nozzle #
x ;
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r Safe-End
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3 jN sf ,
i i j 3 1 2B l 2A i
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Figure 6-S File: SECT 06.TXT-E1
JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT 4#
THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6 10 of 6-12 0.5 in-Vessel Visual Examinations in addition to the ASME Code requirements, this section identifies those additional examination activities that NYPA has evaluated and determined to be applicable to the Third inspection interval. Augmented examinations may be recommended by regulato:y documents, NSSS reco7nmendations, industry experience, and/or good engineering practice.
b ' Nb wg _
^3s . _ . h _$jTdBLE M 5p
. c ~ DIN-VES$ELLAUGMENTED EXAMINATIONSt #
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Component l >
EMinMAd (OishectionLRssENemst5ts[ Mxnn.tlaMFrequensp} <
"m'ation i Illiothod? ' WW s F
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i, Top Guide Visual VT-3 GE RICSIL 071 Periodically as required BWRVIP-26 (11)
EPRI TR-1072851 Steam Dryer Assembly Visual GEJAF 86 Periodically inspect to U BWROG 91 assess component GE Sll 474 condition; UT inspect as necessary to characterize cracking;(10)
Core Shroud Assembly Visual VT-3 BWRVIP-07 Examine Shroud Tie Rod and Supports (EPRI TR-105747) Assembly; EVT-1 Examins Nut Engagement Examine Component Welds (11)
Jet Pump Beams (2) Visual NUREG/CR 30f 2, inspect each inspection USNRC Bulletin 80-07 Interval as required (11),
GE SIL 330 Beams were replaced with BWRVIP-41 a resistant material.
EPRI TR-1087281 Core Spray and Core Visual NUREGICR 4523 Examine CS Piping and CS Spray Sparger (3) BWRVIP-18 Sparger Assembly and EPRI TR-106740 piping component condition (11), Commitment Letter JPN-97-013 Moisture Visual GEJAF 86 Periodically inspect to
/N Separator / Shroud BWROG 91 assess component
() Heads Assembly GE Sll 433 condition (10)
Good Engineering Practice File: SECT 06.TXT-E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-11 of 6 12 2
m.. "i_ 1
. UThBLE6NU' 4 M# ; <
,[m n# 1NhVEUSEUhUGMENTED ENAMINAilONST 4
Componentzu Exa$ir$dids hispscilon R$40Uementsi Extendsl FYequencyT Method 1 i' G
identifisatidn7 Shroud Head Bolts (4) Visual Replacement Bolts Periodically as required (10)
Jet Pump Assembly Visual GE SIL 420 Examine Jet Pump GE SIL 465 assemblies, diffuser to shelf GE SIL 574 welds, sensing lines and BWRVIP-41 brackets, etc., on a periodic EPRI TR-108728 basis to assess component condition. (11)
Feedwater Sparger Visual Good Engineering Practice Visually inspect each fourth (3 Nozzle (5) refueling outage, Examine the nozzles on the
() replacement FW Sparger on a periodic basis (10)
IRM/SRM Dry Tubes (6) Visual industry Recommendations Examine on a periodic basis to assess the conditions of the tubes (10)
Control Rod Drive Stub Visual Good Engineering Practice When Accessible (10)
Tube / Housing Penetrations in-Core Visual Good Engineering Practice When Accessible (10)
Instrumentation / Housing Penetrations Core Delta Visual Good Engineering Practice When Accessible (10)
Pressure / Standby Liquid Control Core Plate Visual GE RICSIL 071 Periodically as required BWRVIP-25 (11)
EPRI TR-107284 i
Control Rod Drive Visual NUREG 0619 Inspect each outage until Retum Line Nozzle final PT is performed.
) NYPA Letter JPN-83-64 i \ ,/
i File: SECT 06.TXT Ei l
I JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POW 6R PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 6-12 of 6-12 Notes: ,
(1) NYPA has not determined whether reexamination of all these areas, (as defined above) is warranted during the Third interval.
The Jet Pump beams were replaced per Modification DI-92-077.
(2) ' Core Spray Sparger *B" clamp was repaired per Modification F1-18199.
(3)
(4) Bolts were replaced. Type 347, per Modification F1-86-122. (1)
(5) Feedwater Sparger assemblies replaced per Modification F1-78-036.
(6) IRM/SRM Dry Tubes have all been replaced with Type 348 material per Modification F1-85-099.
(10) These inspections are not commitments and are t.ibject to change as necessary to support good engineering practices.
(11) These inspections will be cased on approved BWRVIP guidelines for the identified components, 6.6 Core Spray Augmented Examinations Augmented examinations identified below are welis that are dedicated to monitoring the pump discharge piping for vibration.
r (w Scope of Examination - Perform volumetric and surface examination of selected welds as identified below:
- _ .y . f. ' , J , [ J _3 ,lj l , q _ , Nf N' _ g
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, M' ^ ;, M i wg; CORE 8 PRAY AUGMENTED, WELD INSPECTIONSha, j s '*
L60P'hSMSKdE2i[.M . , s 100(BYMSK202k % Me M M 12-14-750 inaccessible 12-14-851 Inaccessible 8-14-779 8-14-878 8-14-780 8-14-879 12-14-724 10-14-884A 12-14-734 12-14-823 12-14-834 Eh
! I V
{
l Flie: SECT 06.TXT-E1
I JAMES A. FITZPATRICr( JAF-IS!-0002
- > I)rvyYorkPower NUCLEAR : OWER PLANT
& mnority
- THIRD INSERVICE INSPECTION INTERVAL Rev. O i INSERVICE INSPECTION Date
- January 6,1998 l PROGRAM Page 7-1 of 7-4 1
TABud OF CONTENTS SECTION 7 Table of Contents ... . . . . . 7-1 Record of Revision . .. ... .. . .. . 7-2 7.0 RELIEF REQUESTS .. . 7-3 7.1 Second Inspection Interval . .. .. 7-3
- 7.2 Third Inspection Interval . 7-3 5
7.3 Relief Request Content . 7-3
Pawer NUCLEAR POWER PLANT p.
(g THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 72 of 7-4 RECORD OF REVISION DATE: 0' , ~ REASON FOR REVISION:
- REVISDN , 'AFFECTED
_ Noi
.... c PAGES:- :Q1:;-
,,, ; g:
~
O January 6,1998 Entire Updated inservice inspection Program Plan Document for the 3'* Ten Year Inservice inspection Interval
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File: SECT 07.TXT41
l t
! JAMES A. FITZPATRICK JAF lSI-0002
- 5 Ptmer NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 73 of 7-4 7.0 RELIEF REQUESTS 7.1 Second Inspection Interval During the Second Ten Year Inservice Inspection Interval, there were cases where component configuration and/or interference prevented the code required volume or surface area from being examined in it's entirety. In each case where such limitations were encountered, the details were documented on a Request for Relief and submitted to the USNRC as required by 10 CFR 50.55a.
7.2 Third inspection interval A detailed review of the previously submitted Requests for Pelief was performed, and based on that review, Requests for Relief on items which remain applicable for the Third inservice inspection interval are included in Appendix F of this program. Appendix F includes a listing s
and the status of each Request for Relief submitted to the USNRC as part of this program.
i Note: Examination volume or surface area that cannot be examined due to interference by another component or part geometry, a reductiori in examination coverage on any weld will be considered acceptable provided the reduction in coverage for that weld is less than 10%.
Subject of Code Case N-460. Examination volume or surface area interference that does not meet the coverage requirements of Code Ccse N-460, will be documented in the form of a Relief Request per 10 CFR 50.55a (g)(4)(iv).
In cases where p:,tts of the required examination areas cannot by effectively examined because of a combination of component design or current inspection technique limitations, NYPA will continue to evaluate the development of new or improved examination techniques with the intent of applying these techniques where a practical improvement on the examination can be achieved.
7.3 Relief Request Content Each Request for Relief will contain the following information:
A. Component identification - describes the Code Class, Code Examination Category (if applicable) and a brief description; B. Examination Requirement describes the Code item Number (s) and the examination requirements; C. Relief Requested provides a description of the relief from the requirements of the Code that cannot be complied with; Fit SECT 07.TXT Ei
JAMES A. FITZPATRICK JAF ISl 0002
- 5 power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 7-4 of 7-4 D. Basis for Relief - describes justification to support the reason relief is being requested; E. Alternative Examination - describes examination (s) or tests that NYPA proposes to use in Heu of the current requirements; F. Implementation Schsdule denotes the interval, period, and/or outage (whichever is applicable), that NYPA proposes to implement the relief, G. Attachments to the Reli'ef identify all Figures Tables, Sketches, Photographs, etc.,
attached to the Request for Relief. All attachments should be referenced within the applicable text.
Note: Following receipt of the USNRC Safety Evaluation, a USNRC Response Section is added to each Relief.
f sg l
I
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V File: SE.CT07.TXT E'1 l
l
1 JAMES A. FITZPATRICK JAF ISI-0002 l
NUCLEAR POWER PLANT 6 #5 W Uterty
%wkPower THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 81 of 8 13 TABLE OF CONTENTS SECTION 8 Table of Contents , . . . . .. . . . .+. . . 8-1 Record of Revision . . . . . . . 83 8.0 ACCEPTANCE CRITERIA . .. . . . . . .... . 8-4 8.1 Acceptance by Examination . . .. . .. . .. 8-4 8.2 Acceptance by Repair .... . . .. . .. . 8-4 8.3 Acceptance by Replacement . ... . . . . .. . 8-4 8.4 Acceptance by Analytical Evaluation . .. ... 85 8.4.1 Class 1 Components . . .. . . 8-5 8.4.2 Class 2 Components . . . .. . . 85 8.4.3 Class MC Components . .. . . . . 8-5 8.5 Acceptance by Engineering Evaluation .. 86 8.6 Acceptance by Correction . . . . 8-7 8.7 Acceptance by Supplemental Examination , . . . 8-7 8.8 Acceptance Criteria in Course of Preparation .. . .. 8-7 8.9 Additional Examinations . . ... . . ....... .. . 8-8 8.9.1 Class i ... ... .. . . . . .. . . . . . .... 8-8 8.9.2 Class 2 . . , . . . . . . . . ... .. ... ., ... .. . .. 8-8 8.9.3 Class 3 . ... . . ..... . .... .. . . .. ... 89 8.9.4 Component Supports . . . . . 3-9 8.9.5 Class MC ,..., . ... .. .. . . .. 8 -10 v
8.9.6 IGSCC Sample Expansion . . 8 -10 File. SECTOS TXT-Ei
JAMES A. FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT
& hority THIRD INSERVICE INSPECTION INTERVAL Rev. 0 ,
INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 82 of 8-13 l
l TABLE OF CONTENTS Continued 8.10 Defects Found Outside Section XI Examination . . . . . ... .. . . 8 11 LIST OF TABLES Table 8-1 Class 1 Acceptance Standards . .. . . . 8 - 12 Table B2 Class 2 Acceptance Standards . .. . . .. . . . . 8 12 Table 8-3 Class 3 Acceptance Standards . .... .. ... .. .. 8 -13 Table 8-4 Class 1,2,3 Component Support Acceptance Standards . ... 8 - 13 Table 8-5 Class MC Acceptance Standards . . . . . .. . 8-13 (3
%)
File: SECT 08.TXT-Et
I JAMES A.FITZPATRICK JAF.lSI-0002 NUCLEAR POWER PLANT S #> liewYwkPower
& Author #if THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998
^ Pace 83 of 8 13 RECORD OF REVISION
' REVl810N: .DATE: l_ AFFECTED ' . REASON FOR REVISION Noi. , ' PAGES <;
O January 6,1998 Entire Updated Inservice Inspection Program Plan for Document the 3* Ten Year Inservice Inspection interval ID V
+
f
\m)
File: SECT 08.TXT-Ei
JAMES A.FITZPATRICK JAFISI0002 NUCLEAR POWER PLANT h hortty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 84 of 8 13 8.0 ACCEPTANCE CRITERIA Indications detected by inservice examinations shall be compared against the acceptance criteria of Section XI as defined in Tables 8-1 through 8-4.
8.1 Acceptance by Examination Components whose examination either confirms the absence of flaws / conditions or reveals indications that do not exceed the acceptance enteria identified in Tables 8-1 through 8-5 shall be acceptable for continued service. Verified changes of flaws / conditions from prior examinations shall be recorded in accordance with Section 10 of this program.
Acceptance of components for continued service w;th indications / conditions exceeding the acceptance criteria above shall be in accordance with the 8.2 through 8.6.
/"'N i 4 V 8.2 Acceptance by Repair Components whose volumetric or surface examination reveals defects / conditions that exceed the acceptance criteria of Tables 8-1 through 8-5 shall be unacceptable for continued service until removed by mechanical methods or until the component is repaired to the extent necessary to meet the acceptance enteria in 8.1. Repairs are further addressed in Section 10.
Note: The additional examination requirements of IWB 2430, lWC 2430, IWE-2430 or IWF-2430, (as applicable) shall be performed for service induced defects /condrtiono, ,
and/or those construction or manufacturing defects determined by Nuclear Engineering to be detrimental to the quality or safety of the component / system.
8.3 Acceptance by Replacement As an attemative to repair requirements of 8.2, the component, or the portion of the component containing the defecticondition shall be replaced. Replacements are further addressed in Section 10.
p File: SECT 08.TXT Ei
JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT h hority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pace 85 of 8 13 8,4 Acceptance by Analytical Evaluation Components whose volumetric or surface examination reveals defects that exceed the acceptance cnteria of Tables 81 through 8-5 are acceptable for continued service without defect removal, repair or replacement if an analytical evaluation meets the acceptance enteria of IWB-3600 or IWC-3600 as applicable, or for Class MC meets the engineering evaluation enteria of 8.5.
Where the acceptance criteria of IWB 3600 or IWC-3600 are satisfied, the area containing the defect shall be subsequently reexamined in accordance with 8.4.1,8.4.2 or 8 4.3.
Note: Reexamination shall be accomplished only on service induced defects / conditions.
8.4.1 Class 1 Components Pursuant to the Section XI Code, sub-article IWB 2420 (b), in the case, where examinations reveal the presence of service-induced defects that exceed the O acceptance standards and the component is analyzed as acceptable for continued h service, the areas containing such defects shall be reexamination during the next three (3) inspection periods of Inspection Program B (IWB 24121). Provided the defects remain essentially unchanged for three successive inspection periods, the component examination schedule will revert to the original schedule of successive inspections.
8.4.2 Class 2 Components Pursuant to the Section XI Code, sub-article IWC-2420 (b), in the case, where examinations reveal the presence of service-induced defects that exceed the acceptance standards and the component is analyzed as acceptable for continued service, the areas containing such defects shall be reexamined dunng the next inspection period of Inspection Plan B (lWB-2412-1) Provided the defects remain essentially unchanged for the next inspection period, the component examination senedule will revert to the original schedule of successive inspections.
8.4.3 Class MC Components When component examination results require evaluation of flaws, areas of degradation, or repair in accordance with IWE 3000, and the component is found to be acceptable for continued service, the areas containing such flaws, degradation, or repairs shall be reexamined during the next inspection period listed in the schedule of Inspection Program B (IWE-24121). When the reexamination required by IWE 2412(b) reveals that the flaws, areas of degradation, or repairs remain
,/] essentially unchanged for three consecutive inspection periods, the areas containing
(/ such flaws, degradation, or repairs no longer require augmented examination in accordance with Table IWE-2500-1, Examination Category E C.
File; SECTOS.TXT-Ei l
JAMES A.FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT h THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paao 8-6 of 8 13 8.5 Acceptance by Engineering Evaluation Examinations that revealindications exceeding the acceptance criteria identified in Tables 81 through 8-5 will be submitted to Nuclear Engineering for evaluation and disposition:
A- Indications found to be acceptable by the materials and welding enteria specified in the Construction Code and/or Section 111 Edition applicable to the construction of the component shall be acceptable for continued service.
B. Indications determined to be acceptable by the NYPA Design and/or Manufacturer's Specifications shall be acceptable for continued service.
C. Indications believed to be surface anomalies (o g., fabrication marks, scratches, surface abrasion, material roughness or other conditions are acceptable for continued service provided the indication is removed by light flapping and/or grinding O (surface preparation), and the material removed does not violate the design V minimum wall thickness.
D. If the evaluations conducted on a component support demonstrates that the support was functional for its intended safety function, additional exams are not required.
E. Components whose examination results reveal flaws or areas of degradation that exceed the acceptance enteria of Table 8-5 are acceptable for continued service without defect removal, repair or replacement if an engineering evaluation indicates that the flaw or area of degradation is nonstructural in nature or has no effect on the structuralintegrity of the containment.
F. When supplemental examinations of 8,7 are required, if either the thickness of the base metal is reduced by no more than 10% of the nominal plate thickness or the reduced thickness can be shown by analysis to satisfy the requirements of the Design Specifications, the component shall be acceptable by evaluation.
G. Where the flaw or area of degradation are accepted by engineering evaluation, the area containing the flaw or degradation shall be reexamined in accordance with 8.4.3, O Nuclear Engineering evaluation and/or disposition may include the need for corrective
\j measures, repairs, analytical evaluation, or replacement, as appropriate.
Fue: SECT 08 TXT-Et
JAMES A.FITZPATRICK JAF.lSI-0002 NUCLEAR POWER PLANT
- > lilewWrkPower W Authony G THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,g, g,7 og g,3 3 8.6 Acceptance by Correction Component supports whose examinations reveal conditions described in IWF 3410(a) shall
, be unacceptable for continued service until such conditions are corrected by one or more of the following: ,
Adjustment and reexamination for conditions such as:
- 1. Detached or loosened mechanical connections;
- 2. Improper hot or cold positions of spring supports and constant load supports;
- 3. Misalignment of supports; or
- 4. Improper displacement settings of guides and stops.
A component support or portion of a component support which is unacceptable per Table 8-4, for continued service ma'y be analyzed and/or tested to the extent necessary to substantiate O, its integrity for its intended service.
8.7 Acceptance by Supplemental Examination Volumetric, visual, or surface examinations that detect indications requiring evaluation may be supplemented by other examination methods and techniques to determine the character of the indication / condition.
Components containing indications and/or relevant conditions shall be acceptable for continued service if the results of supplemental examinations meet the acceptance requirements of 8.1.
Examinations that detect flaws or evidence of degradation that requires evaluation in accordance with the requirements of 8.5 may be supplemented by other examination methods and techniques (IWA 2240) to determine the character of the flaw (i e., size, shape, and orientation) or degradation. Visual examinations that detect surface flaws or areas that are suspect shall be supplemented by either surface or volumetric examination.
8.8 Acceptance Criteris in Course of Preparation if acceptance criteria for a particular component, examination category, or examination method are not specified, defects that exceed the acceptance enteria for matenals and welds specified in the Construction Code and/or Section ll1 Edition applicable to the construction of the component shall be evaluated to detennine disposition.
p File: SECT 08 TXT-Ei
JAMES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT 49 THIRD INSERVICE INSPECTION INTERVAL Rey, O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pane 8-8 of 8-13 8.9 Additional Exarninations 8.9.1 Class 1 The additional examination requirements identified in IWD 2430 shall be performed for service induced defects / condition, and/or those construction or manufacturing defects determined by Nuclear Engineering to be detrimental to the quality or safety of the component / system only. When this situation exists, additional examinations shall include the following:
a, The remaining welds, areas, or parts within the same Code item Number for;
- 1. The existing period
- 2. The next subsequent inspection period, even if the period falls within the next interval.
O O b. If the examinations for that inspection item are not scheduled in the subsequent period, the most immediate period containing scheduled examinations shall be taken as the subsequent period.
- c. If the additional examinations reveal service induced defects / conditions, the examinations shall be further extended to include all welds, areas, or parts of similar design, size, and function.
Additional examinations will be performed before the end of the outage.
8.9.2 Class 2 The additional examination requirements identified in IWC-2430 shall be performed for service induced defects / condition, and/or those construction or manufacturing defects determined by Nuclear Engineering to be detrimental to the quality or safety of the component / system only. When this situation exists, additional examinations shallinclude the following:
- a. An additional number of components or areas, within the same examination category, approximately equal to the number of components or areas examined initially.
- b. If the additional examinations reveal service induced defects / conditions, the examinations shall be further extended to include remaining number of
('N similar components or areas within the same examination category.
File: SECT 08TXT E1
JAMES A.FITZPATRICK JAF ISl4002 NUCLEAR POWER PLANT 9 tif tertty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 89 of 8 13
- c. Additional examinations of welds, areas, or parts may be limited to welds, areas, or parts of similar design, size, and functiori Additional examinations will be performed before the end of the outage.
8.9,3 Class 3 There are no additional examination requirements identified in LWD-2000, therefore additional examinations shall be performed for service induced defects / condition, and/or those construction or manufacturing defects determined by Nuclear Engineering to be detrimental to the quality or safety of the component / system only.
When this situation exit,ts, additional examinations shallinclude the following:
- a. An additional number of components or areas, within the same examination item number, system, and line, and will include the following:
O 1. The next cornponent or area. Upstream and downstream of the b initial defect or condition.
- 2. If the additional examinations reveal service induced defects / conditions, the examinations shall be further extended to include remaining number of similar components or areas within the same item number, system or line.
- b. Additional examinations of welds, areas, or parts may be limited to welds, areas, or parts of similar design, size, and function.
Additional examinations will be performed before the end of the outage.
8.9.4 Component Supports The additional examination requirements identified in lWF 2430 shall be performed for service induced defects / condition, and/or those construction or manufacturing defects determined by Nuclear Engineering to be detrimental to the quality or safety of the component / system only. When this situation exists, additional examinations shallinclude the following:
, a. The component supports immediately adjacent to the initially identified l Support with the defect / condition,
- c. Additional supports equal in number and similar in type, design, and function (7 to those initially examined.
I Y Flie: SECT 08.TXT E1
JAMES A. FITZPATRICK JAF-ISI 0002 NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 FROGRAM Pane 8 10 of 8 13
- d. When these addrtional examinations reveal defect / conditions, the remaining supports within the item number, system or hne shall be examined.
- e. Additional examinations of supports may be limited to supports within the same system or line of the same type, design, and function.
Additional examinations will be performed before the end of the outage.
0.9.5 Class MC The addrtional examination requirements identified in IWE 2430 shall be performed for any one inspection that reveals flaws or areas of degradation as follows:
- a. Examinations performed during any one inspection that reveal flaws or areas of degradation exceeding the acceptance standards of Table IWE-3410-1 shall be extended to include an additional number of examinations within the same category approximately equal to the initial number of examinations during the inspection.
O b. When additional flaws or areas of degradation that 6>ceed the acceptance standards of Table 8-5 are revealed, all remaining examinations within the same category shall be performed to the extent specified in Table IWE-2500-1 for the inspection interval.
Note: Additional examinations will be performed before the end of the outage.
Per 10CFR 50.55a(b)(x), NYPA shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. The evaluations shall include the following information:
A. A description of the type and estimated extent of degradation, and the conditions that led to the degradation; D. An evaluation of each area, and the results of the evaluation, and; C. A description of necessary corrective actions.
8.9.G IGSCC Sample Expansion A- if one or more cracked welds in IGSCC Categories A, B. or C, are found by a sample inspection during the 10 year interval, an additional sample of the
[_
L welds in that category should be inspected, approximately equal in number to the original sample, This additional sample should be similar in distr:bution (according to pipe size, system, and location) to the original Fue: SECT 08.TXT Et
I l
JAMES A. FITZPATRICK JAF ISI-0002 l NUCLEAu 90WER PLANT G. THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 8 11 of 8 13 sample, unless it is determined that there is a technical reason to select a different distribution.
If any cracked welds are found in this second sample, all of the welds in that IGSCC Category should be inspected.
B. If significant crack growth, or additional cracks are found during the inspection of one or more IGSCC Category E welds, all other Category E welds should be examined.
- a. Significant crack growth for overlayed welds is defined as crack extension to deeper than 75% of the original wall thickness, or for cracks originals/ deeper than 75% of the pipe wall, evir.sence of crack growth into the effective weld overlay.
- b. Significant crack growth for Si mitigated Category E welds is defined as growth to a length or depth exceeding the criteria for Sl mitigation. (10% of circumference or 30% in depth),
io\
8.10 Defects Found Outside Section XI Examination Defects / conditions that are found outside the course of a Section XI examination, shall be compared against the acceptance standards of Tables 81 through 8-4, as applicable.
p C/
File: SECT 08 TXT Ei
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] NUCLEAR POWER PLANT
- > l]lowYurkPower e' Authorty THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 8 12 of 8 13 TABLE 81. CLASS 1 ACCEPTANCE STANDARDS EXAMINATION ? CC4APor$ENT OR PART f'.XAMINED ' ACCEPTANCE' CATEGORY- ,
- STANDARD" B-A Welds in Reactor Vessels IWB 3510 B-B Welds in Other Vessels IWB 3510 BD Vessel Nonle Welds IWB 3512 B-E Partial Penetration welds in Vessels IWB 3522 B-F. B J Dissimilar and Gimilar MetalWelds in Piping IWB 3514 B G-1 Bolting > 2" dia fWB 3515/3517 B-G.2 Bolting 5 2"dia IWB 3517 B-H. B K 1 Integral Attachments for Piping. Vatves. Pumps & Vessels IWB 3516 A
( B-L 1, B-M-1 Welds in Pumps & Valvos IWB-3518 B-L 2 B-M-2 Pump Casings & Vatve Bodies IWB 3519 B N-1 B-N 2 Intenor Surfaces & Intemal Components of Reactor Vessels IWB-3520 B-N 3 BO Control Rod Drive Housing Welds IWB 3523 B-P Pressure Retaining Boundary IWB 3522 BO Steam Generator Tubing IWB3521 TABLE 8 2 CLASS 2 ACCEPTANCE STANDARDS
^ ^
COMPONENT OR PART EXAMINED
^ ^"
CATEGORY STANDARD C-A Welds in Pressure Vessels iWC 3510 C-B Nonle Welds in Vessels r#C-3511 C-C Integral Attachments for Vessels, Piping. Pumos and Valves IWC-3512 C-D Bolting IWC-3513 C F-1.C.F-2 Welds in Piping IWC 3514 C-G Welds in Pumps and Valves IWC 3515 l [] C-H Pressure Retaining Components r#C-3516 k
File: SECT 08 TXT Ei
JAMES A. FITZPATRICK JAFISI0002 NUCLEAR POWER PLANT tv fy THIRD INSERVICE INSPECTION INTERVAL Rev. 0
, INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,,, g,3 3 ,, g,3 3 TABLE 8 3. CLASS 3 ACCEPTANCE STANDARDS EXAMINATION ., ; COMPONENT OR PANT EXAMINED
. ACCEPTANCE CATEGORY l ' STANDARD ^
D-A Pressure Doundary (VT 2) *No Leakage ' Mfg Code integral Attachments NT 3) & Applicable Standards D-D Pressure Boundary (VT 2) *No Leakage Gifg Code Integral Attachments (VT 3) & Applicable Standards D-C Pressure Doundary (VT 2) IWA 5250
- Mfg Code &
Integral Attachments NT 3) Apphcable Standards ASME Section XI Acceptance Standard in course of preparation.
^ Defined as no leakage from the pressure boundary, in excess of that which is expected and m determlried to be within acceptable limits, i.e., at valve packing, bolted connections, pump seals, etc.
TABLE 84 - COMPONENT SUPPORT ACCEPTANCE STANDARDS
- fJXAMINATION / ' j COMPONENT OR PART EXAMINED L ;h 0 ACCEPTANCEI ,
- CATEGORY " STANDARD -'
FA Supports IWF.3410 TABLE 8 5. CLASS MC ACCEPTANCE STANDARDS dXAMINATION ' D' s
' COMPOkhMT OR PART EXAMINED I LACdEPTANOE)
T:
- CATWOORY ? ' - STANDARD" E-A Containment Surface IWE-3510 ED Pressure Retaining Welds IWE 3511 E-C Containment Surfaces requinng Augmented Examinations fWE 3512 E-D Seals, gaskets, and moisture barriers IWE-3513 EF Pressure retaining dissimilar metal welds IWE-3514 7 EG Pressure retaining botting IWE 3515
/
(,) E-P All pressure retaining components Appendix J fW SECT 08 TXT-E1
JAMES A. FITZPATRICK JAF lSI 0002 NUCLEAR POWER PLANT t
- > New York Power !
G 4# Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 9 - 1 of 99 TABLE OF CONTENTS SECTION 9 Table of Contents . . . .. . .. . . .. . 91 Record of Revision , . ,. . .. ... .. . ... .. .. .... .. .. 93 9.0 REPAIRS, REPLACEMENTS AND MODIFICATIONS .. .. . . . .... 94 9.1 Repairs . . . . .. . ... .. . . 94 9.1.1 Exemptions . .. . ...... .. . .. .. . . 94 9.1.2 Repair Operations . . ... .. ...... . . . ... 9-5' 9.1.3 Pressure Testing .. . ... .. . .. . .. . . .. . . ... .. 95 9.1.4 Baseline Examinations . .. . . ...... . . . .. . 9-5 9.2 Replacements . .... .. . . , . . 95 9.2.1 Replacement Operations . .. 95 9.2.2 Engineering Approval . ,,. . ,, , . . .. . . . 9-6 9.2.3 Pressure Testing .. . . . 9-6 9.2.4 Preservice Examinations .. ,,.. , ... . . . ,,.. , 97 9.3 Repair / Replacement Activities for IWE Class MC Components . . .. . 9-7 9.3.1 IWE Exempt Components . . .. .. , , .. .... . 9-7 9.3.2 IWE Class MC Components Operations . . . . .. . .,,. . 97 9.3.3 IWE Class MC Components Examination Requirements . . .. . 9-8 9.3.4 IWE Class MC Components Examination and Pressure Test Requirernents ... 9-8
[ 9.3.5 IWE Class MC Components Examination Qualifications . . . 98
\_
Fde SECT 09 TXT-Ei
__m.__
JAMES A. FITZPATP.lCK JAFISl0002 NUCLEAR POWER PLANT
- > New York Power 4# Authority E THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Page 92ef 99 TABLE OF CONTENTS (Continued) 9.4 Modifications . .. .. . . . . ... . .. . ,, 9-8 9.5 - Evaluation . .. .... .... . . . . .-. . .. . . 98 9.0 Access. . . ... . . . . . . . . . . . .. . ..... . 99 9.7 Construction Codes . . . .. . . . . . . . . . . 99 9.8 Authorized NuclearInserviceinspector . . . . . . . . .. 9-9 9.9 Implementations . .. . . . . . . . . . . . .... 99 O
File: SECT 09.TXT Ei 1
JAMES A.FITZPATRICK JAFISI0002 NUCLEAR POWER PLANT
- > New York Power g Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,g, g,3 g 9,g RECORD OF REVISION REVISION > DATE AFFECTED) : REASON FOR REvlslON .
No, , PAGES 40 0 January 6,1998 Entire Updated inservice Inspection Program Plan for the 3* Ten Year Document inservice inspection Interval O
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( s l
l L)
! I File SECT 09 TXT Et i
l l
~#_ _
JACES A.FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT New York Power g #>#
4 Autherity THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,g, , , 4 ,, 9,9 9.0 REPAIRS, REPLACEMENTS AND MODIFICATIONS 1
Scope This section establishes the program by which the James A. Fi2 Patrick Nuclear Power Plant will define the managenal and administrative controls over the implementation and completion of repairs, replacement (modifications) and maintenance of items that require subsequent inservice exam l nations or tests.
This section implements the requirements of the ASME Boiler and Pressure Vessel Code,Section XI,
- Rules for Inservice inspection of Nuclear Power Plant Components,' hereafwr referred to as the Code,1989 Edition, no Addenda, for the Repair, Replacement (modification), and Installation of Replacement Activities at the James A FitzPatrick Nuclear Power Plant. The repairs and replacements for components which are within the provisions of ASME Boller and Pressure Vessel Code,Section XI,1992 Edition with the 1992 Addenda of Subsection IWE,' Requirements for Class MC Metallic Liners of Clm CC Components of Light Water Cooled Power Plants,' incorporated by reference in 10 CFR 50.55a. will be controlled in accordance with this program.
p The repairs and replacements for which these previsions apply are restricted to those performed on systems
(-) and components classified Class 1,2,3, and Class MC pressure retaining comporents and their integral attachments.
ASME Section XI Repairs and Replacements shall be conducted in accordance with the James A. FitzPatrick implementing Repair / Replacement procedure.
9.1 Repairs Repairs for which these provisions apply are restrirted to those performed on systems and components classified Quality Group A, B or C, (Clas51,2, 3) and Class MC pressure retaining components and their integral attachments. Repairs shel be performed in accordance with NYPA's JAF Design Specification and the original Construction Code of the component or syr. tem. However, as allowed by paragraph IWA-4120, later Editions and Addenda of the Construction Code or of Section ill, either in their entirety or portions thereof, and Code Cases may be used. The later editions and Addenda of Section XI, either in their entirety or portions thereof, may be used for the repair program, provided these Edr! ions and Addenda of Section XI at the time of the planned repair have been incorporated by reference in amended regulations of the regulatory authonty having jurisdiction at the plant site.
9.1.1 Exemptions The repair of piping, valves and fittings, nominal pipe size (1) one inch and less ar' ;x6mpt from NDE and pressure testing, but shall comply with all other rules of this section. These repairs shall be made in accordance with the applicable plant procedure for repair / replacement and meet the requirements of the JAF Ouality Assurance Program.
Fue: SECT 09 TXT Ei
JAMES A.FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT
- > New York Power THIRD INSERVICE INSPECTION INTERVAL Rev. 0 l lNSERVICE INSPECTION Date: January 6,1998 PROGRAM Pa 9 5 of 99 m
9.1.2 Repair Operations Code repairs are performed in accordance with approved procedures or instructions that meet the requirements of IWA-4000.
Repair operations shall be performed in accordance with a program delineating requirements of tha complete repair cycle and shallinclude the following:
(a) The NDE method that revealed the flaw and the description of the flaw.
(b) The flaw removal method, method of measurement of the cavity created by removing the flaw and dimensional requirements for reference points during and after the repair.
(c) Weld procedu7s and postweld heat treatment, and the non-destructive examination program to be used after the repair, q (d) Evaluatien as described in P.5.
O (e) The repair programs shall be subject to review by the enforcement and regulatory authorities having jurisdiction at the plant site.
9.1.3 Press:.te Testing After repairs by welding on the pressure retaining boundary, a pressure test shall be performed in accordance with the requirements of IWA-4700/5000.
9.1.4 Baseline Examinations When required by Section XI, the repaired area shall be reexamined to establish a new preservice record. The examination shallinclude the method that detected the flaw, j 9.2 Replacements
! Replacements are performed using approved procedures or instructions in accordance with IWA-7(.00.
All procedures for the installation of renewal, spare, and replacement parts sha'l be in accordance l with IWA-4100. Altematively, subsequent Editions and Addenda of Section XI may be used for l replacement provided these Editions ana Addenda are acceptable to the enforcement and regulatory
( authorities having jurisdiction at the site.
l l 9.2.1 Replacement Operations n
( ) The program for replacements shall include the following:
(a) The applicable C9nstruction Code to which the otiginal item was constructed Fde- SECT 09 TXT Ei
JAMES A.FITZPATRICK JAF ISI-0002 NUCLEAR POWER PLANT
- > New York Power THIRD INSERVICE INSPECTION INT ERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM P 96M 99 (b) The existing design requirements (if the originalitem was constructed without Code requirements, the item to be used for replacement shall be in accordance with the design, fabrication, and examination requirements for the original item unless the atternative of C below is adopted).
(c) Altematively, an item to be used for replacement may meet all or portions of the requirements of later editions of the Construction Code or Section lil, when the Construction Code was not Sect,on ill, provided that the following requirements are met.
(1) The requirements affecting the design, fabrication, and examination of the item to be used for replacement are reconciled with the Owner's through the Stress Analysis Report, Design Report, or other suitable method that demonstrates the item is satisfactory for the specified design and operating conditions.
(d) A description of the work to be performed.
/~'N Q (e) The Code Edition, Addenda and Code Cases applicable to materials, design manufacture, and installation.
(f) Any special requirements pertaining to materials, welding, heat treatment, and nondestructive examination requirements.
(g) Mechanical interfaces, fits, and tolerances that provide satisfactory performance are compatible with system and component requirements.
(h) Materials are compatible with installation and system requirements.
(I) The test and acceptance criteria to be used to verify the acceptability of the replacement.
(J) The documentation required by IWA-7500.
(k) The application of the ASME NA Code Symbol Stamp is neither required nor prohibited for the installation of a item to be used for replacement.
9.2.2 Engineering Approval Replacements that involve substitution of materials, dimensional changes, process changes, deviations to specifications or changes to design codes require engineering approval.
9.2.3 Pressure Testing
.(p; Pressure testing shall be performed on replacements in accordance with IWA-7000 and by the requirements as defined in IWA-4100.
File- SECT 09 TXT El j
JAMES A. FITZPATRICK JAF ISI 0002 NUCLEAR POWER PLANT
- > New York Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 Page 9 7 of 99 9.2.4 Preservice Examinations Prior to the system's retum to service, a preservice inspection shall be made in accordance with IWB 2200, IWC 2200, LWD 2200, IWF 2200, IWE 2200.
Post Work testing and pressure testing will be performed as delineated in applicable Plant Procedure (s).
9.3 Repalr/ Replacement Activities for IWE Class MC Components The USNRC amended 10 CFR 50.55a, by reference the 1992 Edition with the 1992 Addenda of Section XI, to incorporate subsection IWE
- Requirements for Class MC and Metallic Liners of Class CC Components of Light-Water Cooled Power Plants *, and Subsection IWL Requirements for Class CC Concrete Components of Light-Water Cooled Power Plants *.
Note: Subject IWL is not applicable to James A. Fitzpatrick Nuclear Power Plant , as the JAF Plant g uses a steel primary containment.
9.3.1 IWE Exempt Components The following components (or parts of components) are exempted from the examint. tion requirements of IWE-2000.
(a) Vessels, parts, and appurtenances that are outside the boundaries of the containment as defined in the Design Specifications; (b) Embedded or inaccessible portions of containment vessels, parts, and appurtenances that met the requirements of the original Construction Code; (c) Portions of containment vessels, parts, and appurtenances that become embedded or inacc3ssible as a result of vessel repair or replacement if the conditions of IWE-1232 and IWE-5220 are met; (d) Piping, pumps, and valves that are part of the containment system, or which penetrate or are attached to the containment vessel. These components shall be examined in accordance with the rules of IWB or IWC, as appropriate to the classification defined by the Design Specifications.
9.3.2 IWE Class MC Components Operations The program for Class MC components and their integral attachments for Repair / Replacements shall include the following:
(V5 (a) The Primary Containment System at JAF, which is defined as a General Electric Mark 1 Pressure Suppression Containment.
File: SECT 09 TXT-Et
JAMES A. FITZPATRICK JAF lSI 0002 NUCLEAR POWER PLANT
- > New York Power THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM p , , 3 ,, ,,,
9.3.3 IWE Class MC Components Examination Requirements (a) Examination requirerrants shall apply to Class MC pressure retaining components and their integral attachments. These examinations shall apply to surface areas, including welds and base metal.
Note: Pursuant to 10 CFR50 55a(b)(2)(x)(C) examination of pressure retaining welds and pressure retaining dissimilar metal welds are optional.
(b) Preservice Exarninations shall be performed in accordance with the requirements defined in IWE 2200, Preservice Examination.
(c) Visual examinations performed during the conduct of a system pressure test shall be in accordance with IWE 5240.
9.3.4 IWE Class MC Examination and Pressure Test Requirements Examination and Pressure Test requirements shall be performed in accordance with the i requirements defined in IWE-2500, Examination and Pressure Test Requirements, IWE-C' 5200 System Test Requirements, applicable Code Cases that are approved for use and accepted foi implementation wrthin the ISl Repair / Replacement Program, or approved NRC Relief Requests for the component and/or part.
9.3.5 IWE Class MC Components Examination Qualifications Examination qualifications shall meet those requirements of IWA-2300, applicable Code Cases that are approved for use and accepted for implementation within the ISI Repair / Replacement Program, or USNRC approved Relief Request submitted under NYPA Letter JPN 97-031, dated 10/06/1997..
9.4 Modifications The performance of modifications is controlled in accordance with applicable plant procedures.
9.5 Evaluation When the repair, replacement or modification is required because of failure of a part or component pressure boundary, an evaluation shall be done to ensure that the replacement is suitable and the repair procedure selected is suitable. The cause of failure shall be evaluated in accordance with the Code. Engineering evaluations shall be used to document conditional use of equipment ("use as is*)
or equipment substitutions when it is impractical to restore the equipment to the original design configuration by modification, repair, or direct replacement.
! [ Note: Refer to Section 8 of this program for specific criteria for the acevance and evaluation of V; IWB, lWC, IWD, IWF, and IWE components / systems.
i l
l File; SECT 09.TXT-Ei
JAMES A. FITZPATRICK JAF lSI-0002 NUCLEAR POWER PLANT
- > New York Power THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6,1998 PROGRAM p,g, , , , , , ,,9
=
9.6 Access Accessibility for inservice inspection was considered dunng the design of the reactor vessel and insulation to ensure adequate working space and eccess for inspection. The selection of the components and locations to be inspected meet the intent of the ASME Boiler and Pressure Vessel Code.Section XI,
- Inservice Inspection of Nuclear Reactor Coolant Systems', dated January 1,1970.
Adequate access and clearances for examination and tests shall be cor sidered by Nuclear Engineering as part of the processing of design er arrangement changes of system components in accordance with applicable Nuclear Engineering Procedures / Instructions.
9,7 Construction Codes ,
The procurement, design, fabrication and installation Components, parts. and piping shall be in accordance wit:1 the requirements of the FSAR and design specifications. Later Editions and Addenda of the Construction Code or of Section Ill, either in their entirety or portions thereof, and Code Cases may be used.
p)
( Welding activities shall be performed in accordance with Weld Control Manual.
xJ 9.8 Authorized Nuclear Inservice Inspector The services of an Authorized Nuclear inservice inspector (ANil) shall be used when making all repairs / replacements. The repair planner shall be made available for review by the ANil for all welded repairs / replacements. The ANil shall determine what hold points, if any, are required to monitor the repair / replacement activity. NYPA shali notify the ANil prior to starting the repair, replacement or modification and keep tl.e inspector informed of the progress of the work so that necessary inspections may be performed.
9.9 Implementation All ASME Section XI Class 1,2 and 3 Repairs and Replacements are controlled by site procedures.
3 t b
~% _)
File: SECT 09 TXT-Ei
JAMES A.FITZPATRICK JAF lSI-0002 Power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 10-1 of 10 13 .
TABLE OF CONTENTS SECTION 10 Table of Contents . . . . . . .. . . . 10 1 Record of Revision . .. . . . .. . . 10-2 10.0 RECORDS . . ... . . . . 10-3 10.1 General . . . . . . . 10 3 10.2 Inservice inspection Summary Report .. .. . . .. .. 10-3 10.3 Cover Sheet 10-3
( .. . . . . .. .
10.. Summary Report Subm.ttal . .. . . . . 10-4 10.5 Reporting Requirements for IGSCC . . ... . .. . 10-4 10.6 Reporting Requirements for Class MC . . . .. . . . 10 4 LIST OF FIGURES 10-1 NIS 1 Owners' Data Report for Inservice inspections . . .. . 10-5 10-2 NIS-2 Owners' Report for Repairs and Replacements . . . .. . 10-7 10-3 OAR-1 Owner's Activity Report . . . .. . . .. .. . . 10-9 Table 1-Abstract of Examination and Tests . . . .. , . 10-10 Table 2-Items with Flaws or Relevant Conditions that Require Evaluation for Continued Service . . . . . . . .. . . . . .. 10 - 11 Teble 3-Abstract of Repairs, R3 placements or Corrective Measures
,-m Required for Continued Service .. . . .. 10 -12 i !
'd'~ 10-4 NIS-2A Repair / Replacement Plan Certification Record . . .. 10-13 Fila' SECT 10 TXT-Ei
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]]
/> YorkPower NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 RO N M Paae 10 2 of 10 13 RECORD OF REVISION REVISION DATEl AFFECTED L . REASON FOR REVISION i .
. No. - PA.
~ . GES e
' ..( ;g -
0 January 6,1998 Entire Updated Inservice Inspection program Plan for the 3* Ten Year Document inservice inspection Interval l
i l
l r
f r%
Ej l l .
File SECT 10.TXT-Ei
JAMES A. FITZPATRICK JAF lSI-0032 g Power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTlON INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 10-3 of 10 13 10.0 RECORDS This section provides the requirements for the preparation and submittal of Inservice Ins'>ection records and reports as required by IWA4000.
10.1 General -
Examinations, tests, replacements, and repair records are prepared in accordance with the requirements of the ASME Boiler and Pressure Vessel Code,Section XI.
10.2 Inservice Inspection Summary Report An Inservice inspection Summary Report will be prepared at the completion of each inspection conducted during a refueling outage, Examinations, tests, replacements, and repairs conducted since the preceding summary report shall be included.
Note: As a afternate to the requirements of IWA-6000, NYPA has subr .itted a Request for Relief to implement Code (g
'y Case N-532,
- Alternative Requirements to Repair and Replacement Documentation Requirements and inservice Summary Report Preparation and Submission as Required
- v IWA-4000 and IWA-6000,Section XI, Division 1" Each Summary Report will cor tain the following:
- a. Refueling outage number (when applicable).
- b. Owner's Data Report for inservice inspections, Form NIS-1, Figure 11-1.
- c. Owner's Data Report for Repairs or Replacements Form NIS-2 or Form NIS-2A, Figure 112 and Figure 113.
10.3 Cover Sheet Each Summary Report will have a cover sheet providing the following information:
- a. Date of docL ent completion
- b. Name and address of Owner
- c. Name ond address of generating plant
- d. Name and number designation of the plant
- e. Commercial service date for the unit (v)
Foo SECT 10.TXT E1 l
JAMES A.FITZPATRICK JAFlSI0002
- > Power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTIGN INTERVt Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pace 10-4 of 10-13 10A Summary Report Submittal 4
g Ninety (90) days following the units retum to service, NYPA shall forwa' .1 Sumr..ary Report of the inservice 1 Inspection activity to the United States Nuclear Regulatory Commission . . accordance with IWA-6220, or the requirements specified lii Code Case N-532, upon USNRC approval.
10.5 Reporting Re M.rements forIGSCC if any cracks are identified that do not meet the criteria for conbiued operation without evaluation given in Section XI of the Code, USNRC approval of flaw evaluation and/or repairs in accordance with IWB-3000 and IWA-4000 is required before resumption of operation.
10.6 Reporting Requirements for Class MC Per 10 CFR 50.55a(b)(vi, JAF shall provide the following in the inservice inspection Summary Report required by IWA-6000:
(1) A description ot~the type and estimated extekt of degradation, and the condition that led to the degradation; (2) An evaluation of each area, and the results of the evaluation, and; (3) A description of necessary corrective action.
t O
Fde: SECT 10.TXT E1
JAMES A. FITZPATRICK [[::JAF-ISl-0002|JAF-ISl-0002]] Power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O L
INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pane 10 5 of 10-13 FIGURE 101 FORM NIS 1 OWNERS' DATA REPORT FC P. INSER /105 INSPECTIONS As required by the Provisions of the Ashy C4e Rules
- 1. Owner: New York Power Authority.123 Main Street. Whitr, Plains. New York 10601 (Name and Address of Owner) -
- 2. Plant: James A. Fitzoatrick Nuclear Power Plant. P.O Box 41. Lvcomina. New York 1309L
,l (Name and Address of Plant)
- 3. Plant Unit: N/A 4. Owner Certificate of Authorization (if required) N/A
- 5. Commercial Service Date : 07/28/1975 6. National Board Number for Unit: N/A
- 7. Components inspected:
k n ., - ~.: < ~ . . ~x..-
Component orj '3, . Manufactures off RANufacturer 4 State or Provin; >ce' NationalBoasd Apputtonance: 'C ' : Installer? s - Instalter Serial Nd.S No? A , ' - Noc '
Note: Supplemental sheets in form of lists, sketches, or drawings may be used, provided (t) size is 3 % x 11 in. (2) information in items 1 through 6 on this report is included on each sheet, and (3) each sheet is numbered and the number of sheets is recorded at the top of this form.
Fi"; SECT 14.TXT-E1
JAMES A. FITZPATRICK JAF ISI-0002 power NUCLEAR POWER PLANT S THIRD INSERVICE INSPECTION INTERVAL INSERVICE INSPECTION Rev.
Date: January 6,1998 O
PROGRAM Pace 10-6 of 10 13
,~
FIGURE 101 (Conlued)
NIS 1 Owner's Data Report FORM NIS-1 (Back)
- 8. Examination Dates: to
- 9. Inspection Period Identification:
- 10. Inspection Interval Identification:
- 11. Applicable Edition of Section XI:
- 12. Date/ Revision of Inspection Plan:
- 13. Abstract of Examinations and Tests. Include a list of examinations and tests and a statement conceming status of work required for inspection Plan, p 14. Abstract of Results of Examinations and Tests.
V)
I
- 15. Abstract of Corrective Measures.
We certify that a) the statements made in this report are correct b) the examinations and tests meet the inspection Plan as required by the ASME Code,Section XI, and c) corrective measures taken conform to the rules of the ASME Code,Section XI.
Certificate of Authonzation NO (if applicable) Expiration Date Date 19 Signed by Owner l
CERTIFICATE OF INSERVICE INSPECTION 1, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vesselinspectors and/or the State or Province of and employed by have inspected the components desenbed in this OWNERS' Data Report dunng the penod to . and state that to the best of my knowledge and behef, the Owrier has performed examinations and taken corrective measures described in the Owners' Data Report in accordance with the requirements of the ASME Code,Section XI. By signing this certificate, neither the inspector nor his employer makes any warranty, expressed or implied, conceming the examinations, and neither the inspector nor his employer shall be liable in any manner for any personalinjury or property damage or loss of any kind ansing from or connected with this inspection.
Commission Number inspector's Signature National Board, State Province, and Endorsements Date. .ig l O i
I File: SECT 10.TXT.E1
JAMES A. FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] A WrkPower NUCLEAR POWER PLANT 4 18' rity THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 10-7 of 10-13 FIGURE 10 2 FORM NIS 2 OWNERS' REPORT FOR REPAIRS OR REP!.ACEMENTS As Required by the Provisions of the ASME Code Section XI
- 1. owner: New York Power Authority Date:
- Name j.23 Main Street. White Plains. New York 10601 Sheet of Address
- 2. Plant: James A. Fitzcatrick Nuclear Power Plant Unit: N/A Name P.O Box 41. Lvcomina. New York 13093 (Repair Organization P.O. No., Job No., ate.)
- 3. Work Performed Dy: Type Code Symbol Stamp-Name Authorization No.:
O Expiration Date:
Address
(
- 4. Identification of System:
- 5. (a) Applicable Construction Code: 19 Edition _ Addenda Code Case (b) Applicable Edition of Section XI Utilized for Repair or Replacement 19
- 6. Identification of Components Repaired or Replaced and Replacement Componenta kame of Name of Mar.atacturer National other Year Replaced ASME Component Manufacturer sertal Number Board identsftcation Built Repaired Code No. or stamped Replacement (Yes a No)
- 7. Description of Work: , , , _ .
- 8. Test Conducted: Hydrostatic [ ] Pneumatic [ ] Nominal Operating Pressure [ ]
, OTHER [ ] Pressure: Psi TestTemp: Degree F l
(,) Note: Supplemental sheets in form of lists, sketches. or drqwings may be used, provided (1) size is 8 % x 11 in. (2) information in items 1 through 6 on this report is included on tsch sheet and (3) each sheet is numbered and the number of sheets is recorded at the top of this form.
File: SECT 10.TXT E1
l JAMES A.FITZPATRICK [[::JAF-lSI-0002|JAF-lSI-0002]] ;
M NewWrkPower NUCLEAR POWER PLANT !
W Mhority THIRD INSERVICE INSPECTION INTERVAL Rev. 0 l
lNSERVICE INSPECTION Date: January 6,1998 l PROGRAM Pace 10-8 of 10 13 FIGURE 10 2 (CONTINUED)
NIS 2 REPORT CONTINUED
- 9. Remarks:
Apphcable Manufacturers Data Reports to be attached ,
CERT!FICAT5 OF COMPLIANCE We certify that the statements made in the report are correct and this conforms to the rules of the ASME Code Section XI. (Repair / Replacement)
Type Code Symbol Stamp: N/A
,<~,
) Certificate of Authorization No.: Expiration Date:
Signed Date .19__
Owner or Owners' Designee, Title CERTIFICATE OF INSERVICE INSPECTION 1, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure VesselInspectors and the State or Province of and employed by of have inspected the components described in this Owners' Report during period to
. and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in this Owners' Report in accordance with the requirel., ants of the ASME Code,Section XI.
By signing this certification neither the inspector nor his employer makes any warranty, expressed or implied, conceming the examinations and corrective measures described in this Owners' Report.
Furtherrnore, neither the inspector nor his employer shall be liable in any manner for any personal injury or property damage or loss of any kind arising from or connected with this inspection.
Commissions inspector's Signature National Board, State, Province and Endorsements
! l J DATE: .19 FA SECT 10,TXT-E1
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] e NewVorkPower NUCLEAR POWER PLANT tv Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 10-9 of 10 13 FIGURE 10-3 FORM OAR-1 OWNER'S ACTIVITY REPORT As required by the provisions of the ASME Code Case N-532 Page of nons==========================================================s====================================3==
R: port Number Owner New York Power Authority - 123 Main Street. White Plavaa. New York 10601 (Name and Address of Owner)
Plint James A. FitzPatrick Nuclear Power Plant. P. O. Box 41. Lvcomina. New York 13093 (Name and Address of Plant)
Plint Unit N/A Commercial Service Date 7/28/75 Fsefueling Outage Number Curr:nt inspection Interval (1st,2nd, 3rd,4th, Other)
Curr:nt inspection Period (1st,2nd, 3rd)
Edition and Addenda of Section XI applicable to the Inspection Plan 1989 Edition. No Addenda D t3 and Revision of inspection Plan
(\
( iition and Addenda of ASME Section XI applicable to Repairs and Replacements, if different than N.Ja inspection Plan CERTIFICATE OF CONFORMANCE l
l l c:rtify that the statements made in this Owners Activity Report are correct, and that the examinations, tests, repairs, r: placements, evaluations and corrective nieasures represented by this report conform to the requirements of Section XI.
Cartificate of Authorization No. Expiration Date (if applicable) l Signed Date 1 (Owners Representativt and Title) l CERTIFICATE OF INSERVICE INSPECTION l the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the Stite or Province of_ and employed by of have inspected the it;ms described in this Owners Activity Report, dunng the penod to , and state that to the best of my knowledge and belief, the Owner has performed all activities represented by this report in accordance with the r;quirements of Section XI.
l By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, conceming the cx;minations, tests, repairs, replacements, evaluations and corrective measures described in this report. Furthermore, neither the Inspector nor his employer shall be liable in any manner for any personal injury or property rJamage or a loss of any kind arising from or connected with this inspection.
l k Commissions Inspectors Signature National Board, State, Province, and Endorsements D;te 19 i
File: SECT 10.TXT-E1
JAMES A.FITZPATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] b er NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Paae 10-10 of 10-13 FIGURE 10-3 (Continued)
TABLE 1 ABSTRACT OF EXAMINATION AND TESTS As required by the provisions of the ASME Code Case N-532 Page of ss=================================================================================================
TOTAL TO1AL TOTAL TOTAL EXAMINATIONS CODE EMMINATONS EXAMINATIONS ERAMINATIONS CREDffED(%) TO EXAMINATION REQUIRED FOR REQUIRED FOR THl3 CREDITED (%) FOR THE DATE FOR THE CATEGORY IMTERVAL PERIOD PERIOD INTE RVAL REMARKS s.A se so s-F EG1 bG2 N. S-H EJ SK-1 er.i EL-2 En1 SE2 WHEN DISSEMBLED EN-1 842 843 cA C4 cc C41 C44 0+
,r~5 o.a 1
x.s o.c 4 F.A File: SECT 10.TXT-E1 ;
l
JAMES A. FIT PATRICK [[::JAF-ISI-0002|JAF-ISI-0002]] M NewWrkPowcr NUCLEAR POWER PLANT tv Authority THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Date: January 6,1998 PROGRAM Pace 1011 of 10-13 FIGURE 10-3 (Continued)
TABLE 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRE EVALUATION FOR CONTINUED SERVICE As required by the provisions of the ASME Code Case N 532 Page of
=============================================================================== =====
FLAW OR RELEVANT FLAW CONDITION EXAMINATION ITEM ITEM CHARACTERIZATION FOUND DURING SCHEDULED CATEGORY NUMBER DESCRIPTION (IWA-3300) SECTION XI EXAM! NATION OR 1EST (YES OR NO)
File. SECT 10.TXT E1
l l
JAMES A.FITZPATRICK JAF.lSI-0002 l Power NUCLEAR POWER PLANT THIRD INSERVICE INSPECTION INTERVAL Rev. O INSERVICE INSPECTION Da .s: January 6,1998 PROGRAM Paae 1012 of 10-13 FIGURE 10-3 (Continued)
TABLE 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE As required by the provisions Of the ASME Code Case N 532 Page of FLOW OR RELEVANT CONDITION FOUND REPAIR DURING SCHEDULED REPAIR REPLACEMENT DESCRIPTION SECTION XI REPLACEMENT CODE OR CORRECTNE ITEM OF EXAMINATION OR TEST DATE PLAN CLASS MEASUPE DESCRIPTION WORK (YES OR NO) COMPLETE NUMBER l
l L l
l File: SECT 10.TXT-Ei
JAMES A.FITZPATRICK JAF ISI 0002 M NewYorkPower NUCLEAR POWER PLANT
& Authority THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: January 6, s998 PROGRAM Paae 10-13 of 10 13 FIGURE 10-4 FORM NIS-2A REPAIR / REPLACEMENT PLAN CERTIFICATION RECORD As required by the provisions of the ASME Code Case N-532 Page of
- m.
OWNER'S CERTIFICATE OF COMPLIANCE 1 certify that the represented by Repair / Replacement repair or replacement Pian Number conforms to the requirements nf Section XI.
Type Code Symbol Stamp Certificate of A6thorization No. Expiration Date Signed Date .1 g l G Owner or Owners designee, Title l
i CERTIFICATE OF INSERVICE INSPECTION l 1, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State or
! Province of and employed by of j have inspected the components described in Repair / Replacement Plan l No. during the penod to , and State that to the best of my knowledge l cnd belief, the Owner has performed all the activities desenbed in the Repair / Replacement Plan in accordance with the requirements of l Section XI.
l Dy signing this certificate neither the inspector not his employer makes any warranty, expressed or implied, conceming j the activities described in the Repair / Replacement Plan Furthermoie, neither the Inspector not his employer shall be liable in any l manner for any personalinjury or property damage or loss of any kind arising from or connected with this inspection.
l t
Commissions inspectora Signature National Board, State, Province, and Endorsements l
l Date 19
["N (I
l File: SECT 10.TXT.E1
,, ,, n M
y g P weer James A. Fitzpatrick Nuclear Power Plant jar-tSi-0002 APPENDIX A - Program Summary Tables Revision: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret 5 Req Remarks /Commertts item Method Component. lD No. Items Sch'd 1' 2* 3" EXAMINATtON CATEGORY 8-A, PRESSURE RETAINING WEI.DS IN REACTOR VESSEL B1.11 Circ. SheM Welds Vol 01 RPV 02V-1 3036 4 4 4 0 0 A8 Welds,100% of we4d length.
Augmented Examination per 10 CFR SO SSa(gX6Xii)(A)(3)(ii)
B1.12 Long. Sheu Welds Vol 01 RPV 02V-1 3036 12 12 12 0 0 A8 Welds,100 of weld length.
Augmented Examination per 10 CFR SO SSa(gX6Xii)(A)(3)(ii) 8121 Circ HeadWelds Vol 01RPV 07/-1 3036 3 3 0 2 1 Accessible length of aR welds, Deferral to end of interval permissible B1.22 Mendional Head Welds Vol 01 RPV 02V-1 3036 22 22 0 11 11 Accessble length of as wekis, Deferral to end of interval permissible B1.30 SheH-to-Flange Weld Vol 01 RPV 02V-1 3036 1 1 % 0 % 100% of weld lengm 50% IST Period,50% 3". Penod **
B1.40 Head-to-Flange Wold Vol/ Surf 01 RPV 02V-1 3036 1 1 0 1 f 100% of weld length B1.51 Bettline Region Repair Welds Vol 01 RPV 02V-1 3036 0 0 0 0 0 AR repair welds, no base metal weid repairs during the 1* and 2'd inspection interva4 at JAF' Total Examination Category: 43 43 16 % 14 12 %
38 % 70% 100 %
Notes:
(1) Material (base rnetal) weld repairs where repair depth exceeds 10% nominal of the vessel wall. Bettime region extends for the length of the vessel thermai shsid, or in the absence of a thermat shield, the effective length of the reactor fuel elements. If the location of the repair is not positivety and accurately known, then the individual sheE plate, forging, or sher course contaming the repair shau be included.
(2) Includes essantialty 100% of the weld leng'A (3) If partial examinations are conducted from the flange face, the remaining volumetric examination equired to be conducted from the vessel wa!! may be >fwM st or near the end of each inspection interval.
(4) The examination of shen-to-flange welds may be W fwmd during the first and third inspection periods in conjundion with the nozzle examinations of Exam. Cat. B-D (Program B). At least 50% of shen-to-flange welds shall be examined by the end of the first inspection period, and the remainder by the end of the third inspection period.
File: APP-AE1.WPD Appendix A-1 C5 A - 38
Wi O,
James A. Fitzpatrick Nuclear Power Plant JAF-!SHm02 g gamer APPENDIX A - Program Summary Tables Revision: 0 l January s.199s ASME CODE CLASS 1 SYSTEMS AND COMPONENTS i
i ISO No. No. INSPECTION PERIODS Rel Excm item Description Exam System Line or Req Remarks / Comments Component. ID No- Items Sch'd 1s' 2* 3*
item Method Examination Category: B-D, FULL PENETRA flON WELDS OF NOZZLES IN VESSELS -INSPECTION PROGRAM B 3 1 1 (2) 5 75*, (1) 3.812* Nozzles Vol 01 RPV 01 Closure H=ad 3036 3 1 03.90 Nozzle-to-VesselWelds 2 2 1 1 0 Vol 01 RPV 02-2. 28" RC 3036 3036 10 10 5 1 4 02-2.12" RC 02-3, 4" JP 3036 2 2 1 0 1 Vol 01 RPV Enerrpt IWB-1220(a) 03, 3" CRD 3036 1 0 0 0 0 Vol 01 RPV 14,10" CS 3036 2 2 0 1 1 Vol 01 RPV 29,24" MS 3036 4 4 2 1 1 Vol 01 RPV RR-7 NUREG 0619 34,12" FW 3036 4 4 0 0 4 Vol 01 RPV 23 27 10 5 12 Total Examination item 55% 100 %
37%
3 3 1 1 1 (2) 5.75", (1) 3.812" Nezzles 01 RPV 01 Closure Head 3036 B3.100 NozzM inside Radius Section Vol 2 2 1 1 0 Vol 01 RPV 02-2, 28" RC 3036 3036 10 10 5 1 4 02-2,12" RC 02-3, 4" JP 3036 2 2 1 0 1 Vol 01 RPV Exempt lWB-1220(a) 03,3" CRD 3036 1 0 0 0 0 Vol 01 RPV l
01 RPV 14,10" CS 3036 2 2 0 1 1 Vol l
01 RPV 29, 24" MS 3036 4 4 2 1 1 Vol 0 4 RR-7 NUREG 0619 01 RPV 34,12" FW 3036 4 4 0 j
Vol 28 27 10 5 12 Total Examination item f
37 % 55% 100 %
i 20 10 24 Total Examination Category 56 54 37% 55% 100 %
Notes:
Inc; odes nozzles with fun penetratico welds to vessel shes (or head) and integrany cast nozzles, but excludes manway and hand holds either wek$ed to or integraNy cast in vesset.
-(1) At least 25% but not more than 50% (credited) of the nozzles shan be examined by the end of the first inspection oeriod, and the remainder by the end of the inspection intervat.
(2)
( (3) inspections may be partia!!y deferred under the fonowing conditiens; if examinations are conducted from inside the w..ve.ent and the nozzle we8d is examw'ed by straight b method from the nozzle boro, the remaining exsminations required to be conducted from the sheB inside diameter may be performed at or near the end of the inspedRm interval.
The examination volume shan apply to the applicable figure shown in Figs. IW13-2500-7(a) through (d).
(4)
File: APP AE1.WPD Appendix A-2 of A-38 .
1
O O O p> gemmer James A. Fitzpatrick Nuclear Power Plant JAF41@02
, APPENDIX A - Program Summary Tables Roi. ion: 0.
ASME CODE CLASS 1 SYSTEMS AND COMPONENTS Janvery s.1sse Exarn item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. lO No. Items Sch'd 1st 2" , 3"' Req Remarks / Comments Examination Category: B-E, PRESSURE RETAINING PARTIAL PENETRATION WtELDS IN VESSELS B4.11 Partial Penetration Welds VT-2 01 RPV Vessel Nozzles 3036 2 1 0 0 1 RR-3 25% of nozz'es, CC N4?S-1 B4.12 Partial Penetration Welds VT-2 01 RPV Control Rod Drive 3036 137 34 0 0 34 RR-3 25% of nozzles.CC N-4981 B4.13 Partial Penetration Welds VT-2 01 RPV inst. Nozzles 3036 49 12 0 0 12 RR-3 25% of rezz'es, CC N-498-1 Total Examination Category 188 47 0 0 47 0% 0% 100 %
. Notes: Examinations to be performed during the conduct of a si stem hydrostatic test per IWB-5222.
t I
File: APP-AE1.WPD Appendix A -3 of A - 38
g gPower James A. FitzpatriCk Nuclear Power Plant JAF-:St-0002 APPENDIX A - Program Summary Tables Revis:en- 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret Item Method Component ID No- Items Sch'd 15' 2* 3" Req Remarks / Comments ,
Examination Category: B-F, PRESSURE RETAINING DtSSIMILAR METAL WELDS B5.10 NPS 4 or Larger Nozzle-to-Safe End Butt Welds VoVSurf 02-2 RC 28 0"-WH-GE-1 A 3001 1 1 1 0 0 AN welds, NUREG 0313 28.0"-WH4E-1B 3002 1 1 0 1 0 An we!ds, NUREG 0313 Total 28.0" 2 2 1 1 0 50 % 100 % 100 %
Vol/ Surf 02-2 RC 12 0"-WH-GE-4A 3001 1 1 1 0 0 AR welds, NUREG 0313 12.0"-WH-GE-5A 3001 1 1 1 0 0 AR welds, NUREG 0313 12.0"-WH4E-6A E01 1 1 0 1 0 AR welds, NUREG 0313 12.0"-WH-GE-7A 3001 1 1 0 0 1 All welds, NUREG 0313 12.0"-WH-GE-8A 3001 1 1 1 0 0 AH welds, NUREG 0313 12.0"-WH4E-48 3002 1 1 0 0 1 All welds, NUREG 0313 12.0"-WH-GE-5B 3002 1 1 0 0 1 A3 welds, NUREG 0313 12.0"-WH-GE-6B 3002 1 1 1 0 0 A5 welds, NUREG 0313 12.0"-WH-GE-78 3002 1 1 0 0 1 AM weids, NUREG 0313 12.0 -WH-GE-88 3002 1 1 1 0 0 AN welds, NUREG 0313 Total 12.0" 10 10 5 1 4 50*/. 60 % 100 %
Vot/ Surf 02-3 NBVI NBA - 1 1 0 1 0 N88 - 1 1 1 0 0 Total RC System 14 14 7 3 4 50% 66 % , 100 %
Vol/ Surf 14 CS 10 0"-WES-1504-5A 3022 1 1 0 1 0 AH welds, NUREG 0313 10.0 -W23-902-5B 3023 1 1 0 0 1 AN welds, NUREG 0313 Total CS System 2 2 0 1 1 0% 50 % 100 %
Total Examination item 16 16 7 4 5 43% 68% 100 %
B5.20 Less than NPS 4 Nozzle-to-Safe End Welds Surf N/A N/A 0 0 0 0 0 AN welds, not applicable to JAF File: APP-AE1.WPD Appendix A-4 of A - 38
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g wJ % \vl g g Povver James A. FitzpatriCk Nuclear Power Plant [[::JAF-iSi-0002|JAF-iSi-0002]] APPENDIX A - Program Summary Tables Revision: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1996 Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 15' 2* 3* Req Remarks / Comments 85-.30 Nozzle-to-Safe End Socket Welds Surf N/A N/A 0 0 0 0 0 As wetos, not applicable to JAF B5.130 NPS 4 or Larger Dissimilar Metal Butt Welds Vot/ Surf 10 RHR 24*-W20-902-14A 3013 3 3 1 1 1 AR welds, NUREG 0313 24"-W20-902-148 3013 3 3 1 1 1 AM wetds, NUREG 0313 Vol/Sud 10 RHR 20"-20-1504-42 3011 1 1 0 0 1 AH welds, NUREG 0313 Total RHR System 7 7 2 2 3 28 % $7% 100 %
i VotSurf 14 CS 10"-WES-1504-5A 3022 1 1 0 1 0 AM welds, NUREG 0313 10"-W23-902-58 3023 1 1 0 1 0 AN welds, NUREG 0313 Total CS System 2 2 0 2 0 0% 100 % 100 %
Total Examination item 9 9 2 4 3 22% 66 % 100 %
B5.140 f.ess than NPS 4 Dissimilar Metal Butt Welds Surf N/A N/A 0 0 0 0 0 Not appreable to JAF B5.150 Dissimitar Metal Socket Welds Surf IFA N/A 0 0 0 0 0 Not appr:caole to JAF Total Examination Category 25 25 9 8 8 36% 68% - 100 %
Notes:
(1) Examinations are required of each safe end in each loop and connecting branch of the reactor coolant system.
(2) For the reactor vessel nozzle safe ends, the examinatbns may be performed coincident with the vessel nozzle examinations required by Examination Category B-D.
(3) includes dissimilar metal welds between combinations of.
(A) carbon or low alloy stects to high alloy steels (B) carbon or low alloy steels to high nickel allows (C) high alloy steet- to high nickel afloys Fife: APP-AE1.WPD Appendix A -5 of A - 38
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- > gRmer James A. FitzpatriCk Nuclear Power Plant [[::JAF-tSI-0002|JAF-tSI-0002]] APPENDIX A - Program Summary Tables Revissorv 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 8 1' 2* , 3* Req Remarks / Comments Examination Category: B-G-1, PRESSURE RETAINING BOLTING GREATER THAN 2 in. IN DtAMETER B6.10 Closure Head Nuts Surf 01 RPV Closure Head 3036 52 52 17 17 18 AB nuts B620 Closure Studs,in place Vol 01 RPV Closure Heed 3036 52 52 17 17 IS A8 studs B6.30 Closure Studs Vol/ Surf 01 RPV Closure Head 4 4 0 0 4 At or near end of interval When romoved B6.43 Threads in Flango Voi 01 RPV Ctosure Head 3036 52 52 17 17 18 AN Threads in Flange B6.50 Closure Washers, Bushings VT-1 01 RPV Closure Head 3036 104 104 34 34 36 All washers and bushings B6.180 Pump Botts sc.d Studs Vol 02-2 RC 02-2-P-1 A 3001 16 16 0 0 0 AR bolts and ctuds, Limited Vol 02-2 Rt; 02-2-P-1 B 3002 16 0 0 0 0 to one pump in a group of pumps Limited to pump selected under Cat.
B-t-2 B6.190 Flange Surface, when connection disassembled VT-1 02-2 RC 02-2-P-1A 3001 16 16 0 0 0 Only when disassembled VT-1 02-2 RC 02-2-P-1 B 3002 16 0 0 0 0 Only when disassembled 86 200 Nuts Bushings and Washers VT-1 02-2 RC 02-2-P-1 A 3001 16 16 0 0 0 Limi'ed to one pump in a group VT-1 02-2 RC 02-2-P-1B 3002 16 0 0 0 0 c' pumps, Limsted to pump selected under B-t-2 Total Examination Category 360 312 85 85 94 27 % 54 % t 100 %
Notes:
(1) Botting may be examined:
(A) in place unrfer tension; (B) when the rnnection is disassembled, (C) when the txsng is removed (2) Bushings and threads in base materia! of 11anges are required to be examined only when the connections ere disassembled. Bushings may be inspected in place.
(3) For heat exchangers, piping, pumps, and vatves, examinatons are Irnited to cccm as selected for exanwnstion under Examination Categories B-B. B-J. B-L-2, and B4A-2.
(4) Examination incfudes t in. Annular surface of flange surrounding each stud.
g (5) Deferral of inspecton is permissible except when the detected leakage of borated water requires a visual VT-1 in accordance with IWA-5250(a)(2).
(6) The 48 additional item are nct required, limited to one pump in a group of pumps, and Limited to pump seleded urder B-t-2 (7) In addition to Exam item B6 40. NYPA cor.1mitted to examine 22 threads in the RPV Flange during the 1" period of t 'e 3 interval missed during the 2"" interval.
File: APP-AE1.WPD Appendix A -6 of A - 38
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James A. Fitzpatrick Nuclear Power Plant .!AF4SM002 APPENDIX A - Program Summary Tables Revision: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6.1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rel item Method Component ID No. Items Sch'd is' 2'* , 3"> Req Remarks / Comments Examination Category: B-G-2. PRESSURE RETAINING BOLTING,2 in. AND LESS IN D!AMETER B7.10 Reactor Vessel Botts, Studs and Nuts VT-1 01 RPV Closure Head 3036 3 3 1 1 1 Total Examination item 3 3 1 1 1 33% 66% 100 %
B7.50 Piping Bolts, Studs and Nuts VT-1 01 RPV Vent Line Flange 3036 1 0 0 0 0 Exempt fWB-1220(c)
VT-1 02-2 RC 6*LP-A Decon Flange 3001 1 1 1 0 0 Limited to w.. ,ts selected 6* LP-B Decon Flange 3002 1 1 0 1 0 under B4 Total Examination item 3 2 1 1 0 50 % 100 % 100 %
B7.60 Pumps Bolts Studs and Nuts VT-1 02-2 RC 02-2-P-1 A 3001 16 16 0 0 0 Cap Screws, Limited to pump selected VT-1 02-2 RC 02-2-P-1B 3002 16 0 0 0 0 under B-L-2 ; Cap Screws Total Examination item 32 16 0 0 0 0% 0% 0%
B7.70 Valve Bolts Studs et Nuts Limited to vatves selected under B-M-2 VT-1 01 RPV 4* 02RV-71A 3031 1 0 0 0 0 Safety Relef Valves 4* 02RV-71B 3031 1 0 0 0 0 Safety Relief Valves 4* 02RV-71C 3031 1 0 0 0 0 Safety Refef Valves 4" 02RV-71D 3031 1 0 0 0 t 0 Safety Refef Valves 4* 02RV-71E 3031 1 0 0 0 0 Safety Relief Valves 4* 02RV-71F 3032 1 0 0 0 0 Safety Relef Valves 4* 02RV-71G 3031 1 0 0 0 0 Safety Rettef Valves 4* 02RV-71H 3032 1 0 0 0 0 Safety Relief Vatves 4* 02RV-71J 3032 1 0 0 0 0 Safety Refef Valves 4* 02RV-71K 3031 1 0 0 0 0 Safety Reref Valves 4* 02RV-71L 3032 1 0 0 0 0 Safety Relief Vatves Total RPV Valves 11 1 0 0 0 File: APP-AE1.WPD Appendix A -7 of A - 38
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- > geweer James A. Fitzpatrick Nuclear Power Plant jar-isi-0002 APPENDIX A - Program Summary Tables Reviskwr 0
- ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 l
l Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 8 1' 2* . 3* Req Renaark*JComments 87.70 VT-1 02-2 RC 28"MOV-53A 3001 1 0 0 0 0 Gate Vatve 28" MOV-43A 3002 1 0 0 0 0 Gate Vatve 28' MOV-43B 3002 1 0 0 0 0 Gate Valve 28* MOV-438 3001 1 0 0 0 0 Gate valve Total RC Valves 4 1 0 0 0 VT-1 10 RHR 24" RHR-81 A 3013 1 (0 0 0 0 Globe Vafve 24" RHR-81B 3013 1 D 0 0 0 Globe Vake 24" AOV48A 3013 1 0 0 0 0 Check Valve 24" AOV-68B 3013 1 0 0 0 0 Check Vatve 24" MOV-25A 3013 1 0 0 0 0 Gate Valve 24" MOV-25B 3013 1 0 0 0 0 Gate Valve 24" MOV-27A 3013 1 0 0 0 0 Gate Valve 24" MOV-27B 3013 1 0 0 0 0 Gate van' 20" MOV-17 3011 1 0 0 0 0 Gate Valve 20" MOV-18 3011 1 0 0 0 0 Gate Vaive 20" RHR-88 3011 1 0 0 0 0 Gate Valve Total RHR Valves 11 3 0 0 0 VT-1 12 RWC 6" MOV-15 3018 1 0 0 0 0 G ete Vabe 6" MOV-18 3018 1 0 0 0 0 G te Valve 6* RWC-46 3018 1 0 0 0 0 Gate Vatve 4" MOV-69 3018 1 0 0 0 0 Gate Valve Total RWC Valves 4 2 0 0 0 VT-1 13 RCIC 4" MOV-21 3019 1 0 0 0 r0 Gate Vatve 4* RCI-22 3019 1 0 0 0 0 Check valve 3" MOV-16 3020 1 0 0 0 0 Gate Valve. Exempt IWB-1220(a) 3" MOV-15 3020 1 0 0 0 0 Gate Va've. Exempt IWB-1220(a)
Total RCIC Valves 4 2 0 0 0 File: APP-AE1.WPD Appendix A-8 of A- 38
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- > gyukPwwer James A. Fitzpatrick Nuclear Power Plant jar-rSI-0002 APPENDIX A - Program Summary Tables neviason: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret 5 Req Remarks / Comments Item Method Component ID No. Items Sch'd 1' 2* 3*
B7.7D VT-1 14 CS 10" MOV-11 A 3022 1 0 0 0 0 Gate Valve 10" MOV-118 3023 1 0 0 0 0 Gate Vane 10" MOV-12A 3022 1 0 0 0 0 Gate Vane 10" MOV-128 3023 1 0 0 0 0 Ga'e Valve 10" AOV-13A 3022 1 0 0 0 0 Check Vane 10" AOV-138 3023 1 0 0 0 0 Check Vane 10" CSP-14A 3022 1 0 0 0 0 Gate Vafve 10" CSP-14B 3023 1 0 0 0 0 Gate Vafve Total CS Valves 8 2 0 0 0 VT-1 23 HDCI 14"MOV-19 3026 1 0 0 0 0 Gate Vafve 14" HPI-18 3024 1 0 0 0 0 Check Valve 10" MOV-15 3024 1 0 0 0 0 Ga*e Valve 10" MOV-16 302S 1 0 0 0 0 Gate Vane Total HPCI Valves 4 3 0 0 0 VT-1 29 MS 24* AOV-80A 3031 1 0 0 0 0 24" AOV-808 3031 1 0 0 0 0 Globe Valve 24" AOV-80C 3032 1 0 0 0 0 Globe Valve 24" AOV-80D 3032 1 0 0 0 0 Globe Vatve 24" AOV-86A 3031 1 0 0 0 0 Globe Vane l
24" AOV-868 3031 1 0 0 0 0 Globe Vane 24" AOV-86C 3032 1 0 0 0 0 Globe Valve 24" AOV-860 3032 1 0 0 0 0 Globe Vafves Total MS Valves 8 1 0 0 0
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VT-1 34 FW 18" FWS-28A 3033 1 0 0 0 0 Check Va!ve 18" FWS-28B 3034 1 0 0 0 0 Check vane 18" FWS-29A 3033 1 0 0 0 0 Gate Vane 18" FWS-29B 3034 1 0 0 0 0 Gate Vane 18" NRV-111 A 303't 1 0 0 0 0 Check vatve 18" NRV-1118 3034 1 0 0 0 0 Check valve Total FW Valves 6 2 0 0 0 Total Examination item 60 18 0 0 0 Limited to valves selected 0% 0% 0% under B-M-2. to be identifed later l
File: APP-AE1.WPD Appendix A -9 of A - 38
O C O J4F a o002 James A. Fitzpatrick Nuclear Power Plant
-#> geswer APPENDIX A -Program Summary Tables novi. son: 0 January s,1ns ASME CODE CLASS 1 SYSTEMS AND COMPONENTS l
ISO No. No. INSPECTION PERIODS Ret item Description Exam System Une or Exam Component. ID No. Items Sch'd 1" 2* , 3'" Req Remarks / Comments item Method B7.80 CRD Housings BoRs, Studs and Nuts 0 0 0 0 8 boRs per assembty, Note 1 01 RPV 02V-1 CRD Pens. 3036 137 VT-1 137 0 0 0 0 Total CRD System 0% 0% 0%
232 36 1 1 0 The additional 34 Items are Totaf Examination Category under Category B-M-2 0% 0% 100 %
Notes:
(1) Botting may be examined:
(A) in place under tension; (B) when the connection is disassembled; (C) when the boring is removed (2)
For heat exchangers, piping, pumps, and valves, examinations are limited to -vv.mi,"s selected for examination under Examination Categones B-8, B-J B-L-2. and B M (3) CRD bo! ting examinaed when disassembled k
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i File: APP-AE1.WPD Appendix A-10 of A-38
v v U e gamer James A. Fitzpatrick Nuclear Power Plant APPENDIX A - Program Summary Tables [[::JAF-tSI-0002|JAF-tSI-0002]] Revissort o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rei 8 Req Remarks / Comments Item Method Component. ID No. Items Sch'd 1' 2" , 3*
Examination Category: B-H, INTEGRAL ATTACHMENTS FOR VESSELS 88.10 Reactor Vessel integna!!y Welded Attachments Vol/ Surf 01 RPV RPV Support Skirt 3036 1 1 1/2 1/2 0 RR-4 CC N-509 01 RPV RPV Vessel Stabitzee 3036 4 4 0 2 2 RR-4 CC N-509 Total Examination Category 5 5 1/2 21/2 2 1% $0% 100 %
Notes:
(1) Weld buildup on nozzfes that is in compression under normal conditions and provides only -vvis-d support is excluded from examination.
Examination is limited to those integrally welded attachments that meet the following conditions:
(A) the attachment is on the outcide surface of the pressure retaining comporent; (B) the attachment provides component support as defined in NF-1110, fC) the attachment base material design thickness is 5/8 trt or gmater; and (D) the attachment weki joins the attachment either directly to the surface of the vessel or to an integrafty cast or forged attachment to the vetsel.
(2) The extent of the examination includes essentia#y 100% of the length of the attachment weld at each attachment sutW to examination.
(3) In case of multiple vessets of similar design, size, and service, the examination is limited to the attachmW welds of one vesset.
(4) For the configuration shown in Fig. IWB-2500-14, a vetumetric examinaten of vosume A-B-C-D from one side (B-C) of the circumferential weld may be performed in lieu of the surface examinaten.
(5) Examinations will be performed in accordanes with the attemate requirements of Code Case N-509, *A!!ernative Rules for the Selection and Examination of Class 1,2 arx: 3 Integrally Welded Attachments *,Section XI Division 1.Only one vessel in a group of vessels is required.
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Film APP-AE1.WPD Appendix A -11 of A - 38
- > gPower James A. Fitzpatrick Nuclear Power Plant jar-is -0002 APPENDIX A - Program Summary Tables Revi. ion: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item IWethod Component. lD No. Iterrs Sch'd 1st 2* 3* Req Remarks / Comments
~ Examination Category: 8,J. PRESSURE RETAINING WELDS IN P1 PING 89.10 NPS 4 or t.arger Vol/ Surf 25% of circum'erential webs or Branch Cv.-Gis 89.11 Ciretunferentlaf Webs Vol/ Surf 01 RPVCH 5.75" Dia.. Inst Nozzles 3036 2 1 1 0 0 3.812" Dia.. Vent Nozzle 3036 1 0 0 0 0 Total RV 3 1 1 0 0 100 % 100 % 100 %
Vol/ Surf 02-2 RC 28*-WH-GE-1 A 3001 12 4 1 1 2 NUREG 0313 28"-Wi+GE-2A 3001 7 2 1 1 0
' 28*-WH-GE-1 B 3002 13 4 1 1 2 28"-WH-GE-28 3002 7 2 0 0 2 Total 28.0" 39 12 3 3 6 25% 50 % 100 %
Vot/ Surf 02-2 RC 22 -WH4E-3A 3001 4 1 1 0 0 NUREG 0313 22"-WH-GE-3B 3002 4 1 0 1 0 Total 22.0 8 2 1 1 0 50 % 100 % 100 %
Vol/ Surf 02-2 RC 12"-WH-GE-4A '4001 3 1 1 0 0 NUREG 0313 12"-WH-GE-5A 3001 3 1 1 0 0 12 -WH-GE-6A 3001 4 1 0 1 0 12"-WH-GE-7A 3001 3 1 0 0 1 12"-WH-GE-8A 3001 3 1 1 0 t 0 12"-WH-GE-43 3002 3 1 0 0 1 12"-WH-GE-5B 3002 3 1 0 0 1 12 -WH-GE-6B 3002 4 1 1 0 0 12"-WH-GE-7B 3002 3 1 0 0 1 12"-WH-GE-8B 3002 3 1 1 0 0 Total 12.0" 32 10 5 1 4 50 % C0% 100 %
File: APP-AE1.WPD Appendix A-12 of A-38
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Q m grener James A. Fitzpatrick Nuclear Power Plant APPENDIX A - Program Summary Tables jar-:SI-0002 Revision: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January s.19ss Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rel 8 Req Remarks / Comments item Method Component. lD No. Items Sch'd 1' 2" , 3*
89.11 Vot/ Surf 02-3 NBVI N8A/N8R - 8 2 1 0 1 Total JPl 8 2 1 0 1 50 % 50 % 100 %
Total RC System 87 26 10 5 11 38% 57 % 100 %
Vot/ Surf 10 RHR 24*-W20-902-14A 3013 14 5 2 2 1 24*-W20-902-14B 3013 15 4 1 3 0 Total 24.0" 29 9 3 5 1 33 % 88% 100 %
Vol/ Surf 10 RHR 20"-W20-1504-42 3011 16 4 2 1 1 Total 20.0" 60 % 75% 100 %
Total RHR System 45 13 5 $ 2 38 % 84 % 100 %
Vol/ Surf 12 RWC 6*-WR-1504-73 3018 1 0 0 0 0 6"-WR-902A-1 3018 19 6 2 2 2 4"-WD-902A-14 3018 10 3 1 1 1 Total RWC System 30 9 3 3 3 33% 66 % 100 %
Vol/ Surf 13 RCIC 4*-W22-902-4 2019 24 6 2 2 2 4"W22-902A-4A 3019 2 0 0 0 r0 Total RCIC System 26 6 2 2 2 33% 62% 100 %
Vol/ Surf 14 CS 10*-W23-902-5A 3022 25 6 2 2 2 10"-W23-902-58 3023 20 5 1 2 2 Total CS System 45 11 3 4 4 27 % 63% 100 %
File: APP-AE1.WPD Appendix A-13 of A- 38 I
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U s James A. Fitzpatrick Nuclear Power Plant JAF ast-Oo02 I
ggh APPENDIX A - Program Summary Tables Rwi um: o Janwry s,199s ASME CODE CLASS 1 SYSTEMS AND COMPONENTS Line or ISO No. No. INSPECTION PERIODS Ret Exam item Description Exam System Component. ID NO. Items Sch'd 1" 2**
- 3* Req Remarks! Comments item Method 8 3 1 1 1 VoVSurf 23 HPCI 14*-W25-902A-3A 3026 89.11 4 2 10-SHP-902-19A 3024 17 1 1 Total HPCI System 25 7 2 2 3 28 % 57 % 100 %
3031 20 5 5 0 0 TS Secten 4 6F VoVSurf 29 MS 24"-SHP-902-1 A 24*-SHP-902-1B 3031 25 7 0 7 0 24"-SHP-902-1C 3032 23 7 0 0 7 24"-SHP-902-1D 3032 18 5 2 2 1 6*-SHP-902 3031/32 22 6 2 2 2 Total MS System 108 30 9 11 10 30 % 66 % 100 %
3033 20 5 0 0 5 TS Section 4.6F VoVSurf 34 FW 18"WFP-902A-4A 18*-WFP-902F-48 3034 15 4 2 2 0 12*-WFP-902A-5A 3033 8 2 0 0 2 12"-WFP-902A-5B 3034 8 2 0 0 2 12"-WFP-902A-5C 3033 12 7 2 4 1 12"-WFP-902A-5D 3034 12 7 0 2 5 l
Total FW System 75 27 4 8 15 14% 44 % 100 %
Total Examination item 444 130 39 41 50 30 % 61 % 100 %
NPS 4 or Larger Long. Welds VoVSurf 02-2 RC 28"-WH-GE 1 A 3001 22 7 1 3 ?3 RR-5 NUREG 0313 B912 2 0 RR-5 28"-WH-GE-2A 3001 8 1 1 28"-WH-GE-1B 3002 18 7 2 1 4 RR-5 i
28"-WH-GE-28 3002 5 1 0 0 1 RR-5 Total 28.0" 53 17 4 5 8 No percentage required 22"-WH-GE-3A 3001 10 1 1 0 0 RR-5 NUREG 0313 VoVSurf 02-2 RC 22"-WH-GE-38 3002 5 2 0 2 0 RR-5 20 3 1 2 0 No percentage required Total 22.0" File: APP-AE1.WPD Appendix A-14 of A-38
, p (O} (\ (Jo) g gPbwer James A. Fitzpatrick Nuclear Power Plant jar-tSi-0002 APPENDIX A - Program Summary Tables Revision: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS hnuary 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION FERIODS Rel item Method Component. lD No. Items Sch'd 1" 2* . 3* Req Remarks / Comments EW.12 Vol/ Surf 02-2 RC 12"-WH-GE-4A 3001 6 1 1 0 0 RR-5 NUREG 0313 12"-WH-GE-5A 3001 6 2 1 1 0 RR-5 12"-WH-GE-6A 3001 5 1 0 1 0 RR-5 12"-WH-GE-7A 3001 6 1 0 0 1 RR-5 12"-WH-GE-8A 3001 6 1 1 0 0 RR-5 12"-WH-GE-48 3002 6 1 0 0 1 RR-5 12"-WH-GE-5B 3002 6 1 0 0 1 RR-5 12" WH-GE-68 3002 5 1 1 0 0 RR-5 12"-WH-GE-7B 3002 6 1 0 0 1 RR-5 12"-WH-GE-89 3002 6 1 1 0 0 RR-5 Total 12.0** 58 11 5 2 4 No percentage required Total RC System 131 31 10 9 12 Percentages not required, and 32% 61 % 100 % not included in Exam Category Vol/ Surf 13 RCIC 4"-W22-902-4 3019 4 0 0 0 0 RR-5 No percentages required Vol/ Surf 14 CS 10"-WES-1504-5A 3022 23 15 5 6 4 RR-5 No percentages required 20 % 100 % 100 %
Total Examination Item 158 46 15 15 16 32% ES% 100 % Percentages not required, listed for inforr"ation purposes only B9.21 Less Than NPS 4 Circ. Welds Surf 03 CRD 3 0" Line 3003 1 1 0 0 Exempt (WB-1220(a) 13 RCIC 3*-SHP-902-17A 3020 22 0 0 0 0 Exempt IWB-1220(a) t Total Examination item 23 1 1 0 0 100 % 100 % 100 %
B9.22 Less Than NPS 4 Long. Welds N/A N/A N/A 0 0 0 0 0 Not arplicat>le to JAF File: APP-AE1.WPD Appendix A-15 of A- 38
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- > geweer James A. Fitzpatrick Nuclear Power Plant JAF IS8-0002 APPENDIX A - Program Summary Tables nev+. ion: 0 ASME COCE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam ' Item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret
. Item Method Component ID No. Items Sch'd is' 2" + 3"" Req Remarks /Comrr,ents B9_31 Branch Pipe Conn. Welds NPS 4 or Larger Vo8/ Surf 02-2 RC 28"-WH-GE-1 A 3001 1 0 0 0 0 Selection and frequency per 28"-WH-GE-2A 3001 4 2 1 1 0 NUREG 0313 28*-WH-GE-1B 3002 1 0 0 0 0 28"-WH-GE-2B 3002 4 1 0 0 1 Total 28.0" 10 3 1 1 1 33% 66 % 100 %
89.31 Vot/ Surf 02-2 RC 22" WH-GE-3A 3001 4 1 0 1 0 Selection and frequency per 22"-WH-GE-3B 3002 4 1 0 1 O NUREG 0313 Total 22.0" 8 2 0 2 0 0% 100 % 100 %
Vot/ Surf 02-2 RC 12"-WH-GE-4A 3001 1 1 1 0 0 Selection and frequency per 12"-WH-GE-5A 3001 1 1 0 1 0 NUREG 0313 12"-WH-GE-7A 3001 1 0 0 0 0 12"-WH-GE-8A 3001 1 0 0 0 0 Selection and frequency pw 12"-WH-GE-48 3002 1 0 0 0 0 NUREG 0313 12"-WH-GE-5B 3002 1 0 0 0 0 12"-WH-GE-78 3002 1 0 0 0 0 12*-WH-GE-8B 3002 1 0 0 0 0 Total 12.0" 8 2 1 1 0 50 % 100 % 100 %
TotalitC System 26 7 2 3
- 2 28% 71 % 100 %
Vol/ Surf 29MS 24"-SHP-902-1 A 3031 3 1 1 0 0 24"-SHP-902-1B 3031 2 1 0 1 0 24"-SHP-902-1C 3032 4 1 0 0 1 24*-SHP-902-1 D 3032 3 1 1 0 0 Total MS System 12 4 2 1 1 50 % 75% 10G%
Total Examination item 38 11 4 5 ?
36 % 81 % 100 %
File: APP-AE1.WPD Appendix A-16 of A-38
O O O g gPbwer James A. Fitzpatrick Nur,le d'ower Plant [[::JAF-ISI-0002|JAF-ISI-0002]] APPENDIX A - Program Summary Tables Revision: 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam Item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. lD No. Items Sch'd is' 2" . 3* Req Remarks / Comments 89.32 Branch Conn. Less Than NPS 4 Welds Surf 02-2 RC 28"-WH-GE-1A 3001 4 0 0 0 0 Exempt IWB-1220(a) 28"-WH-GE-1 B 3002 6 0 0 0 0 28"-WH-GE-2A 3001 1 0 0 0 0 28"-WH-GE-28 3002 3 0 0 0 0 Total 28.0" 14 0 0 0 0 Exempt fWB-12Ma)
Surf 02-2 RC 12"-WH-GE-4A 3001 1 0 0 0 0 Exempt IWB-1220(a) 12"-WH-GE-SA 3001 1 0 0 0 0 12"-WH-GE-6A 3001 1 0 0 0 0 12"-WH-GE-7A 3001 1 0 0 0 0 12"-WH-GE-8A 3001 1 0 0 0 0 12"-WH-GE-48 3002 1 0 0 0 0 12"-WH-GE-58 3002 1 0 0 0 0 12 -WH-GE4B 3002 1 0 0 0 0 12"-WH-GE-7B 3002 1 0 0 0 0 12 -WH-GE-8B 3002 1 0 0 0 0 Total 12.0" 10 0 0 0 0 Exempt fWB-1220(a)
Total RC System 24 0 0 0 0 Exempt IWB-1220(a)
Surf 10 RHR 24"-W20-902-14A 3013 3 0 0 0 0 Exempt IVs-1220(a) 24"-W20-902-14B 3013 4 0 0 0 0 Surf 10 RHR 20"-W20-150442 3011 3 0 0 0 0 Exempt IWB-1220(a)
Total RHR System 10 0 0 0 3 Exempt IWB-1220(a)
Surf 12 RWC 6"-WR-902A-1 4 0 0 0 0 Exempt IWB-1220(a)
Total RWC System 4 0 0 0 0 Exempt IWB 1220(a, Surf 13 RCIC 4"-W22-902-4 3019 2 0 0 0 0 Exempt IWB-1220(a) 3"-SHP-902-17A 3020 9 0 0 0 0 Exempt IWB-1220(a)
Total RCIC System 11 0 0 0 0 Exempt IWE-1220(a)
File: APP-AE1.WPD Appendix A-17 of A-38
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V U V g grower James A. Fitzpatrick Nuclear Power Plant APPENDIX A - Program Summary Tables aAr-aSi-0002 nevision- 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO N o. No. INSPECTION PERIODS Rel item Method Component. ID No. Items Sch'd 1st 2* . 3'" Req Remarks / Comments B9 32 Surf 14 CS 10"-WES-1504-5A 3022 5 0 0 0 0 Exempt hvB-1220(a) 10*-W23-902-58 3723 5 0 0 0 0 Exempt IWB-1220(a)
Total CS System 10 0 0 0 0 ExeM IWB-1220(a)
Sud 23 HPCI 14 -W25-902A-3A 3026 2 0 0 0 0 Exempt WsB-1220(a) 10"-SHP-902-19A 3024 3 0 0 0 0 Exempt WsB-1220(a)
Total HPCI System 5 0 0 0 0 Fxempt IWB-1220(a)
Surf 29 MS 24*-SHP-902-1 A 3031 8 0 0 0 0 Exempt IWB-1220(a) 24*-SHP-902-1B 3031 7 0 0 0 0 Exempt fWB-1220(a) 24*-SHP-902-1C 3032 6 0 0 0 0 Exempt UdB-1220(a) 24*-SHP-902-1D 3032 7 0 0 0 0 Exempt IWB-1220(a)
Total fps System 28 0 0 0 0 Exempt fWB-1220(a)
Surf 34 FW 18 -WFP-902A-4A 3034 3 0 0 0 0 Exempt Ins-1220(a) 18"-WFP-902A-4B 3034 3 0 0 0 0 Exempt h48-1220(a)
Total FW Sys.em 6 0 0 0 0 Total Examination item 98 0 0 0 0 Exempt fWB-1220(a), not included in the count for the Category 8940 Socket Welds Suff N/A N/A N/A 0 0 0 0 0 Not applicable to JAF Total Examination Category 505 142 44 46 52 30% 63% t 100 %
File: APP-AE:.WPD Appendix A-18 of A-38
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~4 PPENDIX A - Program Summary Tables ne% on. o ASWiE CODE CLASS 1 SYSTEMS AND COMPONENTS January s.tssa Exam Item Description Exam System Une or ISO No, No- INSPECTION PERJODS Ret Item tilethod Compe ent.ID No. Items Sch'd 1 ** 2* 3** Req Remarks / Comments Notes:
(1) Examinstons shas include the folknnrwy (A) AN temwial ends in each pipe or tranch rua con arcted to vesse's.
(B) AN temunal ends and jords in each pee or tranch run coh to other & w sts where the shess ; eve 4 exceed ether or the fonownng Imts urwh k> ads assty:ra+ed with spec Tc seismic events and operatenal conddens' (1) prriary plus secondary stress intensity range of 2.4Sm fa ferrde steel and austanitic steel (2) cumutstrve usage factor U of 0 4 (C) A5 damnaar metal welds be* ween comboaters of.
(1) carbon or low alloy steels to hgh alloy steels (2) e et low afloy steels to hgh mckel alloys (3) high aBoy steels to righ nickel anoys ,
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(D) Addrenal pipeg welds to that the total arreer of crcum8erettal butt weMs (o- branch w.Me, or socatet welds) seleded for exammaten eques 26% of the cm.".. .. ; butt welds (or branch connediori or socket welds) in the reactor coolant p'pwg system. This total does not include weMs escluded by na.B-1220. These
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adf tional weMs may be locatad iri one loop (one loop is defined for both PWR and BWR plants in the 1977 Editen)
(2) The initia!!y selected weMs shall be reexamroed during each inspecten interval.
(3) includes essent:aBy 100% of weld lengttt (4) The exammation includes si least a pea 4*rameter length but not more than 12 et Of each long*vdrial weid intersectmg the catw. A m 'A weids requrred to be exammed >y Exammaten Ca+egories B-F and B-J.
(5) For walds in carbon or low alloy steefs. only those welds shorving reportable preservce transver=e indcatens need to be enarnmed for L s,we.w re c eders.
t File: APP-AE1M'PD Appendix A-19 of A-38
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. APPENDIX A - Program Summary Tables Revissori- 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS Jem,ery e.1ssa Exam item Description Exam System Une or ISO No. :lo. INSPECTION PERIODS Ret item Method Component. ID No- Items Sch'd 1st 2* . 3* Req Remarks / Comments Examination Category: 84-1. INTEGRAI. ATTACHMENTS FOR PIPING, PUMPS, AND VALVES t 10.10 Poing integ a#y Welded AW. e.
Surf C2-2 RC 28*-WHGE-1 A 3001 1 0 0 0 0 RR-4 CC 4509, fan.10%
28"-WM-GE-1B 3002 3 0 0 0 0 RR-4 28" W4GE-2A 3001 3 1 0 1 O RR-4 28"-W4GE-2B 3002 3 0 0 0 0 RR 4 Total 28.0" to 1 0 1 0 0% 100 % 100 %
Surf 02-2 RC 22"-WRGE-3A 3001 4 1 0 0 1 RR4 CC N-509 Mai.10%
22"-WH-GE-3B 3002 4 0 0 0 0 RR4 Total 22.0" 8 1 0 0 1 0% 0% 100 %
Total RC System 18 2 0 1 1 0% 50 % 100 %
Surf 10 RHR 24'-W20-902-14A 3013 4 0 0 0 0 ' !'R-4 CC N-509. Mni 10%
24" W20-902-14B 3013 5 1 0 1 0 RR-4 20 -W20-1504-42 3011 5 1 1 0 0 RR-4 Total RHR System 14 2 1 1 0 33% 100 % 100 %
Surf 12 RWC 6*-WR-902A-1 ~3018 2 1 0 1 0 RR4 CC 4509 Mm.10%
4"-WD-902A-14 3018 4 0 0 0 t O RR-4 Total RWC System 6 1 0 1 0 0% 100 % 100 %
Surf 13 RCIC 4*-W22-902-4 3019 18 2 0 1 1 RRA CC 4509. Me 10%
13 RCIC 3"-SHP-902-17A 3320 2 0 0 0 0 Exernpt IWB-1220(s)
Total RCIC System 20 2 0 1 1 0% 50 % 100 %
File: APP-AE1.WPD Appendix A-20 of A-38
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4,g h James A. Fitzpatrick Nuclear Power P! ant aAr. Sum 0-APPENDIX A - Program Summary Tables nevision: e ASEE CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No- No- INSPECTION PERIODS Ret hem Method Component ID No- Items Sch'd 8 1' 2'* , 3* Req Remarks / Comments _
B10.10 Surf 14 CS 10*-W23-902-5A 3022 4 1 1 0 0 RR4 CC N-509 Mct 10%
10"-W23-902-5B 3023 5 0 0 0 0 RR-4 Total CS System 9 1 1 0 0 100 % 100 % 100 %
Surf 23 HPCI 14*-W25-902A-3A M26 1 1 0 0 1 RR4 CC N-509 Me 10%
10"-SHP-902-19A 3024 3 1 1 0 0 RR-4 Total HPCI System 4 2 1 0 1 50 % 50 % 100 %
Surf 29 MS 24*-SHP-902-1A 3031 8 0 0 0 0 RR-4 24*-SHP-902-1B 3031 9 0 0 0 0 RR-4
~24*-SHP-902-1C 3032 9 0 0 0 0 RR-4
- 2 T-SMP-902-1D 3032 8 3 1 1 1 RR4 Total MS System 34 3 1 1 1 33 % 65 % '100 %
Surf 34 FW 18" YEP-902A-4A 3033 2 1 0 0 1 RR-4 18*-VEP-902F 48 3034 2 0 0 0 0 RR-4 12*-WTP-902A-5A 3033 1 O O O O RR-4 12"-VEP-902A-58 3034 1 0 0 0 0 RR-4 12"-WTP-902A-SC 3033 5 1 1 0 0 RR-4 12*-WTP-902A-SD 3034 5 0 0 0 0 RR-4 Total FW System 16 2 1 0 1 50 % 50 % t 100 %
Total Examination item 121 15 5 5 5 33 % 66 % 100 %
File: APP-AE1.WPD Appendix A-21 of A - 38
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m/ v v g g Psever James A. Fitzpatrick Nuclear Power Plant aAr-mac2 APPENDIX A - Program Summary Tables R siori. O ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No- No. IMOPECTION PERf00S Ret item Method Component. ID No. I: ems Sch'd 1" 2"* 3 Req RemarWm C1020 Pump in*egra5y WeMed Attachments 02-2 RC 02-2-P-1A 9 1 0 0 1 RR-4 CC N-509, Met _1C%
02-2 RC 02-2 0-1B 9 0 0 0- 0 RR-4 Total Exsmination item 18 1 0 0 1 0% 0% 100 % 1 selected, multiple purnpst Total Examination Category 133 16 5 5 6 31 % $2% 100 %
No*es:
(1) 2xammation is Irnited to those integraHy welded attachments that meet the fonoweg condhs.
(A) the attachment is on the outsee sur' ace of the pressure retaining we ww 4 (B) the attachment prowdes w.we-a support as defined in NF-1110 (C) the attachment base matenal design thickness is 5/8 irt Or greater, and (D) the attachment weM joms the .--W . a ether dwect!y to the surface of the w+e e 4 or to an intagra!!y cast or forged c e ,G.ma to the wm.a.
(2) The extent of the exammation includes essentiacy 100% of the length of the attachment weld at eacn attachment subrect to erammaten.
(3) Examinations include the weMed aN a. of piping requeed te be exammed by Exammation Category B-J and weMed attachments to assoca'ed purms and vsW Integral to such piping-(^) For the configuration shown in Fig. IWB-2500-14, a volumetnc examination of volume A-B-C-D from one see (B-C) of the aw.a.m J ; weld may be s,:,b.4 in lieu of the surface exammatert.
The extent and frequency of exammations for Inspechon Program B is only epsA.W to the 1". and the 2'*. Inspeden intervais (5) Examinatens wiR be perfonred in accordance wth the attemaM requirements of Code Case N-509, "Altemabwe Rules for the Seieden and Exammaton of Class 1,2 and 3 Integracy WeMed Attadiments". Se : ten XI. Drvisen 1.
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File: APP-AE1.WPD Appendix A-22 of A-38
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- D NeurysmPouver James A. Fitzpatrick Nuclear Power Plant JAF-IS8 4MIC2 W AuthorW APPENDIX A - Program Summary Tables Reviss erv 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 1" 2* 3"' Req Remarks.C,.a . s .^u.
E=.J. L. Ca'egory: B61. PRESSURE RETAINWG VVELDS IN PU1WP CASmCS; B&2, PUIWP CASMGS C12.10 Purm Caseg Welds Vol 02-2 RC C2-2-P-1A 0 0 0 0 0 r Net acc ocab4e to JAF C2-2 RC 02-2-P-t B C 0 0 0 0 Not .sv to .lAF C12.20 Pump Cawg intemar Surface VT-3 C2-2 RC C2-2-P-1A 3001 1 1 0 0 t Ordy w'*n emm- ibi one -
C2-2 RC 02 2-P-1B 3002 1 0 0 0 0 pump in g oup of pervs Total Exe--J..Je,i Category 2 1 0 0 1 0% 0% 100 %
Notes:
(1) Ex., 4Lv > are Imted to at least one pump in each gaaup of purms sm - -y ss,itar funciens irt the systmt. e g. recreuisteg coeurit .w.
(2) Exame.aten is required only at=n a psym or wahre is dsassemeied for <- ee e, reper. or ve*urme*ne examstatm. Exarn!na+en of the intemmi pressu e boundary shas bew ", . W to the erM pracbcable Examesaten is requeed ordy once danng the bssv Iw interval (3) Purm A selected for mustraten purposes on*y. to prevede the number of purnes recured The actual semen wa be based upon cuaw 65.sw < e4 dureg the intenrat (c) includes essercally 100% of the weld length.
(5) Sunce a y surface examinatens may be " W ui treenor and/or er*enor tur* aces to assrst in &6 -y the loca*en of PJws detect =d by 4 xL,; ene-- e^ws [see r.*/B-35181(d)).
File: APP-AE1%'PD Appendix A-23 of A-38
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James A. FitzpatriCk Nuclear Power Plant JAF-tS862 APPENDIX A -Program Summary Tables R.vaiorr 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PElk!ODS Rei ST Item Method Component ID No. Items Sch'd 1 2'" 3* Reg Remarks / Comments B12.50 VT-3 to rJ4R 24 0" Check 24 AOV-68A 3013 1 0 0 0 0 When desassen@ed, one valve 24" AOV488 3013 1 0 0 0 0 in a gap of watves VT-3 10 RHR 20 0" Ga'e Vanes 20" RHR-88 3011 1 0 0 0 0 When desassen@ed. one wake 20" MOV-17 3011 1 0 0 0 0 m a group of vanes 20" MOV-18 3011 1 0 0 0 0 Total RHR System 11 3 0 0 0 VT-3 12 RWC 6"MOV-15 Gate 3018 1 0 0 0 0 % desassen@ed. one 6" MOV-18 Gate 3018 1 0 0 0 0 va:ve in a group of valves 6" RWCAS Gate 3018 1 0 0 0 0 VT-3 12 RWC 4* MOV-69 Gate 3018 1 0 0 0 0 Total RWC System 4 1 0 0 0 VT-3 13 RCIC 4* MOV-21 Gate 3019 1 0 0 0 0 When drsasse r@ad, one VT-3 13 RCIC 4* RCI-22 Ct.eck 3019 1 0 0 0 0 valve in a g oup of valves VT-3 13 RCIC 3" MOV-16 Gate 3020 1 0 0 0 0 Exc!uded IWB-2500-t size 3" MOV-15 Gate 3020 1 0 0 0 0 Excluded IWB-2500-1 size Total RC:C System 4 2 0 0 0 VT-3 14 CS to 0" Gate Vanes 10" MOV-11A 3022 1 0 0 0 ' O When dess M. one 10" MOV-118 3023 1 0 0 0 0 valve in a g oup of vanes 10" MOV-12A 3022 1 0 0 0 0 10" MOV-128 3023 1 0 0 0 0 10" CSP-14A 3022 1 0 0 0 0 10" CSP-148 3023 1 0 0 0 0 VT-3 14 CS 10 0" Check 10* AOV-13A 3022 1 0 0 0 0 When disassemtwed, one 10" AOV-s3B 3023 1 0 0 0 0 vsNe in a group of valves Total CS System 8 2 0 0 0 File: APP-AE1.WPD Appendix A -25 of A - 38
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& Author #r APPENDIX A - Program Summary Tables nevisioer 0 ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January s,19ss Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret 5 39 Req Remarks / Comments item Method Component. lD No. Items Sch'd 1' 2*
C12.50 VT-3 23 HPCI 14 0" Gate Vafves 14* MOV-19 302S 1 0 0 0 0 When dmswiA,2. one vatwe in a gmup of wakes VT-3 23 HPCI 14 0* Chem Vatwes 14* HP5-18 3026 1 0 0 0 0 When osassenCed, one vane in 3 gmup of waltes VT-3 23 HPCI 10 0* Gate Vanes 10* MOV-15 3024 1 0 0 0 0 When disassemtded, one vane 10" MOV-16 3024 1 0 0 0 0 m a group of vanes Total HPCI System 4 4 0 0 0 VT-3 29 MS 24 0* Globe Valves 24" AOV-80A 3031 1 0 0 0 0 On4 when d%sse.40, one 24* AOV-808 3031 1 0 0 0 0 valve M a gream of wakes 24" AOV40C 3032 1 0 0 0 0 24* AOV-800 3032 1 0 0 0 0 24* AOV-86A 3031 1 0 0 0 0 24* AOV-868 3031 1 0 0 0 0 24* AOV-86C 3032 1 0 0 0 0 24* AOV-860 3032 1 0 0 0 0 Total MS System 8 2 0 0 S VT-3 34 FW 18 0* Gate Va5v es I 18* FWS-29A 3033 1 0 0 0 0 Only when d=sassemtded.
18* FWS-298 3034 1 0 0 0 '
O Lwroted to one vane in a group of watwes VT-3 34 FW 18.0* Check VaNet 18* FWS-28A 3033 1 0 0 0 0 Only when d'sassemt9ed 18* FWS-288 3034 1 0 0 0 0 Lim 4ed to one vane in a group 18* NRV-111 A 3033 1 0 0 0 0 of vanes 18' NRV-1118 3034 1 0 0 0 0 Total FW System 6 2 0 0 0 Total Examination Category 60 18 0 0 0 Vahres To be identfried when disassembled File: APP-AE1.WPD Appendix A -26 of A-38
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, item Method Component. !D No. Items Sch*d 187 2* 3"' Reg Remarksh,ments Notes:
(2) Examw ste is requred ordy when a pump or wat<e es d sassembed for u.,am o,w. repar. or volumetnc exammabort Exammatrori cf the reenal pressure boundary shes be W L m 4 to the extent prachcable. Exammabon is rectired once dunng the especten h%
(3) Examirrates are Imted to at least one vahre withe each group of vatves that are of the same se. constructonal desvi (such as globe. gate, or check vam), and marWacturrg method.
and that perform simAar funchons in the system (such as w a- a isolatm and system over pressure protecten).
(5) Supplementary surface exammate may be pmLu e on intence arui'er evterior surfaces to assrst in JJ. ...-,.g the locaten of flaws detedad by volumetnc ex . 4.v.re [see IWB-5181(d:-}
t Fi;e: APP-AE1.WPD Appendix A-27 cl A-38
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p> mewverkPuner James A. Fitzpatrick Nuclear Power Plant jar.:S 4oc2
& Author #r APPENDIX A - Prograrr Summary Tables Revisio.t o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PEhUODS Ret item Method Component. ID No. lien , Sch'd 1'5 2*' 3"D Req Remarks / Comments i
Examination Category: B-N-1. INTERIOR OF REACTOR VESSEL; B-N-2, INTEGRALLY WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACH 1pENTS TO REACTOR VESSEL C13.10 Resctor Vessef Interior VT-3 RPV 01 3036 1 3 1 1 1 Access @e areas B1320 Reador Vessel (BWR) Inte cr At:achments wt*m Bettiene Regen VT-1 RPV 01 3036 23 23 0 0 23 Accessbe we ses C13.30 Intenor Attachments beyond Beltime Regiyt VT-3 RPV 01 3036 EO 60 0 0 60 Accessible welds B13.40 Core Support Structure VT-3 RPV 01 3036 75 75 0 0 75 AccessNe surfaces Total Examination Category 159 161 1 1 159 Notes:
(1) Areas to be examened SNtil include the spaces above and baiow the reactor core that are made accessNe for exame tatxrt by removal uf G,..w .G dtong nor'-tal re'umang outages.
(2) The structure shall be removed from the reactor vessel for exammaton.
(3) At 1st refueleg outage, and subsequent re8ueling outages at approximately 3 year ritervah (4) Refer to Note (1), Exammaten Category B-A, for derwuten of belttee regon.
7 File: APP-AE1.WPD Appendix A -28 of A - 38
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Exarnination Category: B-0. PRESSURE RETAINING WELDS IN CONTROL ROD HOUSINGS C14.10 Reador VesselWekts in CRD Housmg Vot'Swf 01 RPV 02V-1 3036 25 3 0 0 3 10% of penpheral CRD Houssys 0% 0% 100 % ,
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File: APP-AE1.WPD Appendix A-29 of A-38
,, c O G g g essser James A. Fitzpatrick Nuclear Power Plant JAF-1584002 APPENDIX A - Program Summary Tables Revision: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January s.199s Exam ' Item Description Er,m System Une or ISO No. No. INSPECTION PERIODS Ret 5
Item 16thod Comporient ID No. Items Sch'd 1' 2* 3* Reg RemarksComments Examination Category: B-P, ALL PRESSURE RETAINING COMPONENTS C15.10 Reador VessM System Leakage Test VT-2 01 RPV 02V 1 3036 1 3 1 1 1 RR-10 Each r**u=3ing ot4 age B1511 Reador vesset System Hydrostate Pressure Test VT-2 01 RPV 02V-1 3036 1 1 0 0 1 RR-3/10 Once per entenral C15 40 Heat Excharger System Leakage Test VT-2 N/A WA WA 0 0 0 0 0 Not apphcable to JAF B15 41 Heat Exchanger System Hydrostatic Pressure Test VT-2 N/A N/A WA 0 0 0 0 0 Not appicable to JAF C1550 Pipmg System Leakage Test VT-2 02-2 RC 28"-M-GE-1 A 3001 1 3 1 1 1 RR-10 Each Refuaing Ottaga 28" WH-GE-2A 3001 1 3 1 1 1 RR-10 23*-N-GE-1B 3002 1 3 1 1 1 RR-10 28 -VA4-GE-28 3002 1 3 1 1 1 RR-10 22"-WH-GE-3A 3001 1 3 1 1 1 RR-10 22" WH-GE-38 3002 1 3 1 1 1 RR 10 12*-WH4E-4A 3001 1 3 1 1 1 RR-10 17-WH-GE-5A 3001 1 3 1 1 1 RR-10 12"-YMGE-6A 3001 1 3 1 1 1 P't-t o 12" WH-GE-7A 3001 1 3 1 1 1 RR-10 12"-WH-GE-8A 3001 1 3 1 1 1 RR-10 1T-WH-GE-48 3002 1 3 1 1 ' 1 RR-10 1T-WH-GE-58 3002 1 3 1 1 1 RR-10 17-WH-GE4B 3002 1 3 1 1 1 RR-10 1T-WH-GE-78 3002 1 3 1 1 1 RR-10 17-WH-GE-8B 3002 1 3 1 1 1 RR-10 4*-WH-GE-9A 3001 1 3 1 1 1 RR-10 4"-W%GE-98 3002 1 3 1 1 1 RR-10 2" W%1504-22A 3001 1 3 1 1 1 RR-10 2 -WH-1504-22B 3002 1 3 1 1 1 RR-10 File: APP-AE1.WPD Appendix A-30 of A-38
V V C g g a m er James A. Fitzpatrick Nuclear Power Plant jar-ass.coc2 APPENDIX A - Program Summary Tables Reion- o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Une or ISO No. No- INSPECTION PER!OOS Ret Item Method Component. ID No. Items Sch'd 1" 2* 3* eteq Remarks.8 Comments C15 50 VT-2 10 RHR 24 -W20-902-14A 3013 1 3 1 1 1 RR-10 Each Refuesing Outrpt 24~-W20-902-148 3013 1 3 1 1 1 RR-10 20~-W20-1504-42 3011 1 3 1 1 1 RR-10 4*-W20-902-36 3012 1 3 1 1 1 RR-10 4 4Y20-902-43 3012 1 3 1 1 1 RR-10 12 RWC 6*-WR-1504-73 3018 1 3 1 1 1 RR-10 6 -WR-902Ar1 3018 1 3 1 1 RR-10 13 RCIC 6" Secton 3019 1 3 1 1 1 RR-10 4"-W22-902-4 3019 1 3 1 1 1 RR-10 4 -W22-902A-4A 3019 1 3 1 1 1 RR-10 3 -SHP-902-17A 3020 1 3 1 1 1 RR-10 VT-2 14 CS 10*-W23-902-5A 3022 1 3 1 1 1 RR-to 10"-VES-1504-5A 3022 1 3 1 1 1 RR-10 10"-W23-902-5B 3023 1 3 1 1 1 RR-10 VT-2 23 HPCI 14*-W25-902A-3A 3026 1 3 1 1 1 RR-10 10 -SHP-9rJ2A-19A 3024 1 3 1 1 1 RR-10 VT-2 29 MS 24"-SdP-902-1A 3031 1 3 1 1 1 RR-10 24*-SHP-902-1B 3031 1 3 1 1 1 RR-10 24*-SHP-902-1C 3032 1 3 1 1 1 RR-10 24 -SHP-902-1D 3032 1 3 1 1 1 RR-10 6'-SHP-902 3031/32 1 3 1 1 1 RR-10 t
VT2 34 FW 18VEP-902A4A 3033 1 3 1 1 1 RR-10 18"-VEP-902F-48 3034 1 3 1 1 1 RR-10 12"4*EP-902A-5A 3033 1 3 1 1 1 RR-10 12"-VEP-902A-5B 3034 1 3 ? 1 1 RR-10 12"-WFP-902A-SC 3033 1 3 1 1 1 RR-10 12"-VEP-902A-50 3034 1 3 1 1 1 RR-10 8 Secten 3033 1 3 1 1 1 RR-10 To*al Examination Item 56 168 55 56 56 File: APP-AE1.WPD Appendix A-31 of A-38
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& Amiher#r APPENDIX A - Program Summary Tables Revision: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No- No- INSPECTION PERIODS Ret item Method Component ID No- Items Sch'd 157 2* 3* Reg Remarks /Co..... d.
C15 61 Pipeg System Hydrostabc Test VT-2 02-2 RC 28"-VMGE-1 A 3001 1 1 0 0 1 RR-3/10 Once per Interval 28"4WA-2A 3001 1 1 0 0 1 RR-3/10 28*-WH4E-1R 3002 1 1 0 0 1 RR-3/10 28*-WH-GE-2 3002 1 1 0 0 1 RR-3/10 22"-VM-GE-3. 3001 1 1 0 0 1 RR-3/10 27-WH-GE-38 3002 1 1 0 0 1 RR-3/1C 12*-WH4E-4A 3001 1 1 0 0 1 RRGto 12*-WH-GE-SA 3001 1 1 0 0 1 RR-3/10 12*-VM-GE4A 3001 1 1 0 0 1 RR-3/10 12*-WH-GE-7A 3001 1 1 0 0 1 RR-3/10 12"4%%GE-8A 3001 1 1 0 0 1 RR-3/10 12"-WM-GE-48 3002 1 1 0 0 1 RR-3/10 12"-VM-GE-SB 3002 1 1 0 0 1 RR-3/10 12"-WH-GE-6B 3002 1 1 0 0 1 RR-3/10 12" WH4E-78 3002 1 1 0 0 1 RR-3/10 12" YMGE-8B 3002 1 1 0 0 1 RR-3/10 4"-WH-GE-9A 3001 1 1 0 0 1 RR-3/10 4*4TH-GE-98 3002 1 1 0 0 1 RR-3/10 2*4*M-1504-22A 3001 1 1 0 0 1 RR-3/10 2"-WH-1504-228 3001 1 1 0 0 1 hR-110 VT-2 10 RHR 24*4*J20-902-14A 3013 1 1 0 0 1 RR-3/10 24*-W20-902-148 3013 1 1 0 0 1 RR-3t10 20"-W20-1 SO4-42 3011 1 1 0 0 1 RR-3/10 4*-W20-902-36 3012 1 1 0 0 1 RR-3/10 4 -W20-902-43 3012 1 1 0 0 1 RR-3/10 VT-2 12 RWC 6*-WR-1 SO4-73 3018 1 1 0 0 1 RR-3/10 6*-WR-902A-1 3018 1 1 0 0 1 RR-3/10 4"-WD-902A-14 3018 1 1 0 0 1 RR-3/10 VT-2 13 RCIC 6" Sectet 3019 1 1 0 0 1 RR-3/10 4*-W22-902-4 3019 1 1 0 0 1 RR-3/10 4*-W"'2-902A-4A 3019 1 1 0 0 1 RR-3/10 3*-D "-302-17A 3020 1 1 0 0 1 RR-3/10 File: APP-AE1.WPD Appendix A-32 of A- 38
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& AufherW APPENDIX A - Program Summary Tables Re wie,c 0 ASME CODE CLASS 1 SYJTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No- No- INSPECTION PEhtlODS Ret ite m Methoo Component. lO No- Ite ns Sch'd 187 2* 3* Req Remarks! Comments C15.51 VT-2 14 CS 10"-W23-902-5A 3022 1 1 0 0 1 RR-3/10 10"-W23-1504-5A 3022 1 1 0 0 1 RR-3/10 10*-W23-902-58 3023 1 1 0 0 1 RR-3/10 VT-2 23 HPCt 14~-W25-902A-3A 3026 1 1 0 0 1 RR-3/10 10"-SHP-902-19A 3024 1 1 0 0 1 RR-3/10 VT-2 29 MS 24*-SHP-902-1 A 3031 1 1 0 0 1 RR-3I10 24 -SHP-902-1B 3031 1 1 0 0 1 RR-3/10 24'-SHP-902-1C 3032 1 1 0 0 1 RR-3710 24~-SHP-902-1D 3032 1 1 0 0 1 RR-3/10 6"-SHP-902 3031/32 1 1 0 0 1 RR-3f10 VT-2 34 FW 18"V5P-902A-4A 3033 1 1 0 0 1 RR-3/10 16'-WFP-902F-48 3034 1 1 3 0 1 RR-3/10 12"-VEP-902A-5A 3033 1 1 0 0 1 RR-3/10 12*-WFP-902A-5B 3034 1 1 0 0 1 RR-3/10 12"-WFP-902A-5C 3033 1 1 0 0 1 RR-3/10 12"-WFP-902A-5D 3034 1 1 0 0 1 RR-3/10 8* Section 3033 1 1 0 0 1 RR-3/10 Total Examination Item 56 56 0 0 56 B15 60 Pump System Leakage Test VT-2 02-2 RC 02-2-P-1A 3001 1 3 1 1 1 RR-40 Each reftAng outage VT-2 02-2 RC 02-2P-1B 3002 1 3 1 1 1 RR-10 t
Total Examination item 2 6 2. 2 2 C15 61 Pump System Hydrostatic Pressure Test VT-2 02-2 RC 02-2 P-1 A 3001 1 1 0 0 1 RR-3/10 Once per intemal VT-2 02-2 RC 02-2-P-1B 3002 1 1 0 0 1 RR-3/10 Total Examination item 2 2 0 0 2 File: APP-AE1.WPD Appendix A-33 of A-38
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g g Po = James A. Fitzpatrick Nuclear Power Plant .A n s m c2 APPENDIX A - Program Summary Tables Re.+ on- o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1998 Exam item Liescription Exam System Une or ISO No. No. INSPECT 1086 PERIODS Ret item Method Co.iiw it. lD No. Items Sch'd 1" 2* 3* Reg Remarkstomments C15 70 Vahre System Leakage test VT-2 01 RPV 02RV-71A 3036 1 3 1 1 1 RR-10 Each refuehng c>cge C2RV-71B 3036 1 3 1 1 1 RR-10 C2RV-71C 3036 1 3 1 1 1 RR-10 C2RV-71D 3036 1 3 1 1 1 RR-10 02RV-71E 3036 1 3 1 1 1 RR-10 02RV-71F 3036 1 3 1 1 1 RR-10 02RV-71G 3036 1 3 1 1 1 RR-10 02RV-71H 3036 1 3 1 1 1 RR-10 02RV-71J 3036 1 3 1 1 1 RR-10 02RV-71K 3036 1 3 1 1 1 RR 02RV-711. 3036 1 3 1 1 1 RR-10 C2-2 RC 28" MOV-53A 3001 1 3 1 1 1 RR-10 23 MOV-43A 3001 1 3 1 1 1 RR-10 28" MOV438 3002 1 3 1 1 1 RR-10 23" MOV 43B 3002 1 3 1 1 1 RR-10 10 RHR 24" RHR-81 A 3013 1 3 1 1 1 RR-10 24* RHR418 3011 1 3 1 1 1 RR-10 20* RHR-BS 3011 1 3 1 1 1 RR-10 24* AOV-68A 3013 1 3 1 1 1 RR-10 24 AOV-688 3013 1 3 1 1 1 RR-10 24 MOV-25A 3013 1 3 1 1 1 RR-10 24 MOV-258 3013 1 3 1 1 1 RR-10 24" MOV-27A 3013 1 3 1 1 1 RR-10 24* MOV-27B 3013 1 3 1 1 1 RR-10 20" MOV-17 3011 1 3 1 1 1 RR-10 20" MOV-18 3011 1 3 1 1 1 RR-10 12 RWC 6" MOV-15 3018 1 3 1 1 1 RR-10 6 MOV-18 3018 1 3 1 1 1 RR-10 6* RWC-46 3018 1 3 1 1 1 RR-10 4 MOV-69 3018 1 3 1 1 1 RR-10 13 RCIC 4* MOV-21 3019 1 3 1 1 1 RR-10 4' RCI-22 3019 1 3 1 1 1 RR-10 3* MOV-16 3020 1 3 1 1 1 RR-10 3" MOV-15 3020 1 3 1 1 1 RR-10 File: AFP-AE1.WPD Appendix A-34 of A-38
O O jar-esteoc2 Jemes A. Fitzpatrick Nuclear Power Plant g genser APPENDIX A - Program Summary Tables Revi.so e o January s.1sse ASME CODE CLASS 1 SYSTEMS AND COMPONENTS ISO No- No- INSPECTION FE t!ODS Ret Exam item Description Exam System Une or Req Remarks / Comments Method Compv.ent. ID No. Items Sch'd 1" 2* 3*
item 3022 1 3 1 1 1 RR-10 VT-2 14CS 10" MOV-1 tA C1510 3023 1 3 1 1 1 RR-10 10" MOV-118 3022 1 3 1 1 1 RR-10 10" MOV-12A 3023 3 1 1 1 RR-10 10" MOV-128 1 3022 1 3 1 1 1 RR-10 10* AOV-13A 3023 1 3 1 1 1 RR-10 10" AOV-138 3022 1 3 1 1 1 RR-10 10" CSP-14A 3023 3 1 1 1 RR-10 to* CSP-148 1 3024 1 3 1 1 1 RR-10 23 HPCI 14 MOV-10 3026 1 3 1 1 1 RR-10 14* HPI-18 3024 1 3 1 1 1 RR-10 10" MOV-15 302S 1 3 1 1 1 RR-10 10" MOV-16 3031 1 3 1 1 1 RR-10 29 MS 24* AOV-80A 3031 1 3 1 1 1 RR-10 24* AOV-808 .
3032 1 2 1 1 1 RR-10 24* AOV-80C 3032 3 1 1 1 RR-10 2r AOV-800 1 3031 1 3 1 1 1 RR-10 24* ADV-86A 3031 3 1 1 1 RR-10 24 AOV-868 1 3032 1 3 1 1 1 RR-10 24* AOV-86C 3032 3 1 1 1 RR-10 24* AOV-860 1 3033 1 3 1 1 1 RR-10 34 FW 18" FWS-28A 16" FWS-288 3034 1 3 1 1 1 RR-10 18" FWS-29A 3033 1 3 1 1 1 RR-10 18" FWS-298 3034 1 3 1 1 ' 1 RR-10 18 NRV-111A 3033 1 3 1 1 1 RR-10 18" NRV-1118 3034 1 3 1 1 1 RR-10 Total Examination item 49 147 49 49 49 f
File: APP-AE1.WPD Appetidix A-35 of A-38
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& AmhorW APPENDIX A - Program Summary Tables Reve wm: O ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January s.19ss Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rei item Method Component ID No- Items Sch'd 157 2'* 3* Reg Remarks / Comments C15.71 Vaive System Hydmtatic Pressure Test VT-2 01 RPV 02RV-71A 3036 1 1 0 0 1 RR-3/10 Once per interval 02RV-718 3036 1 1 0 0 1 RR-3/10 02RV-71C 3036 1 1 0 0 1 RR-3/10 02RV-71D 3036 1 1 0 0 1 RR-3/10 02RV-71E 3036 1 1 0 0 1 RR-3/10 02RV-71F 3036 1 1 0 0 1 RR-3/10 02RV-71G 3036 1 1 0 0 t RR-3/10 02RV-71H 3036 1 1 0 0 1 RR-3/10 02RV-71J 3036 1 1 0 0 1 RR-3/10 02RV-71K 3036 1 1 0 0 1 RR 3/10 02RV-71L 3036 1 1 0 0 1 RR-3/10 02-2 RC 28" MOV-53A 3001 1 1 0 0 1 RR-3/10 28* MOV-43A 3001 1 1 0 0 1 RR-3/10 28" MOV438 3002 1 1 0 0 1 RR-3/10 28* MOV-438 3002 1 1 0 0 1 RR310 10 RHR 24' RHR-81 A 3013 1 1 0 0 1 RR-3/10 24 RHR-8tB 3013 1 1 0 0 1 RR-3/10 20" RHR-88 3011 1 1 0 0 1 RR-3/10 24" AOV-68A 3013 1 1 0 0 1 RR-3/10 24" AOV489 3013 1 1 0 0 1 RR-3/10 24* MOV-25A 3013 1 1 0 0 1 RR-3/10 24" MOV-258 3013 1 1 0 0 1 RR-3/10 24* MOV-27A 3013 1 1 0 0 1 RR-3/10 24" MOV-27B 3013 1 1 0 0 ? 1 RR-3/10 20" MOV-17 3011 1 1 0 0 1 RR-3/10 20" MOV-1 F 3011 1 1 0 0 1 RR-3/10 12 RWC 6" MOV-15 3018 1 1 0 0 1 RR-3/10 6* MOV-18 3018 1 1 0 0 1 RR-3/10 6" RWC 46 3018 1 1 O O 1 RR-3/10 4 MOV-69 3018 1 1 0 0 1 RR-3/10 13 RCIC 4" MOV-21 3019 1 1 0 0 1 RR-3/10 4" RO1-22 3019 1 1 0 0 1 RR-3/10 3* MOV-16 3020 1 1 0 0 1 RR-3/10 3" MOV-15 3020 1 1 0 0 1 RR-3/10 File: APP-AE1.WPD Appendix A-36 of A-38
O O O g gPower James A. FitzpatriCk Nuclear Power Plant aar-rsm0c2 APPENDIX A - Program Summary Tables R.vi on: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS .ranuary s.1ns Exam item Description Exam System Line or ISO No. No- INSPECTION PERIODS Rei Item Method Component. ID No. Items Sch'd 1" 2* 3"' Reg RemarkstComments C15 71 VT-2 14 CS 10" MOV-11 A 3022 1 1 0 0 1 RR-3/10 Once per Interval 10" MOV-11B 3023 1 1 0 0 1 RR-3/10 10" MOV-12A 3022 1 1 0 0 1 RR-3/10 10" MOV-128 3023 1 1 0 0 1 RR-3/10 10* AOV-13A 3022 1 1 0 0 1 RR-3/10 10" AOV-138 3023 1 1 0 0 t FA-3/10 10" CSP-14A 3022 1 1 0 0 t etR-3/10 10" CSP-148 3023 1 1 0 0 1 RR-3/10 23 HPCI i4 MOV-19 3024 1 1 0 0 1 RR-3/10 14 HPl-18 3026 1 1 0 0 1 RR-3/10 10" MOV-15 3024 1 1 0 0 1 RR-3/10 10" MOV-16 302S 1 1 0 0 1 RR-3/10 29 MS 24" AOV-80A 3031 1 1 0 0 1 RR-3/10 24* AOV-808 3031 1 1 0 0 1 RR-3/10 24" ADV-BOC 3032 1 1 0 0 1 RR-3/10 24* AOV-800 3032 1 1 0 0 t RR-3/10 24* AOV-86A 3031 1 1 0 0 1 RR-3/10 24* AOV-868 3031 1 1 0 0 1 RR-3/10 24* AOV-86C 3032 1 1 0 0 1 RR-3!10 24* AOV-860 3032 1 1 0 0 1 RR-3/10 34 FW 18" FWS-28A 3033 1 1 0 0 1 RR-3/10 18* FWS-288 3034 1 1 0 0 1 RR-3/10 18* FWS-29A 3033 1 1 0 0 1 RR-3/10 18* FWS-298 3034 1 1 0 0 1 RR-3/10 18" NRV-111 A 3033 1 1 0 0 1 RR-3/10
'8" NRV-1118 3034 1 1 0 0 1 RR-3/10 Total Examination fram 49 49 0 0 49 Total Examination Category 216 432 108 108 216 File: APP-AE1.WPD Appendix A-37 of A-38
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- > g asiver James A. Fitzpatrick Nuclear Power Plant JAns.4oc2 APPENDIX A - Program Summary Tables nevissori: o ASME CODE CLASS 1 SYSTEMS AND COMPONENTS January 6,1996 Exam item Description Exam System 1.ine or ISO No. No. INSPECTION PER'ODS Ret Item Method Component. ID No. Items Sch'd 1** 2* 3"* Reg Remarks / Comments Notes:
(1) The presstre retaining bounda y during the System Leakage Test shasw mw4 to the reactor coolart system bounda y. wth at va4ves in the normal posden, w% is requeed for normat reactor operation startup. The VT-2 examinatort shaE, however, extend to and include the second closed wahre at the boundary entremty (7) The pressure retaining boundary during the System Hydrostatic Test shat include sE Class 1 w1we.G withm the system bounda y (3) System pressure tests of the reador coolant system shas be conducted in anse wdh fWA-5000. System pressure tests for repawed, replaced, or atte'ed w w m..^ shot be govemed by IWA-5214(c).
(4) Vesual examination of fWA-5240.
(5) The System Leakage Tes1(rWB-5221) shas be conducted enor to plant startup fonowrg each reactor refuehng o Aage.
(6) The System Hydrostate Test (fWB-5222) shat be conducted at or cear the end of each '..w. m interval.
(7) A System Hydrostatic Test (PWB-5222) arx' the accompanyng VT-2 erammaton are acceptab4 in Eeu of the System Leakage Test (rv'JB-5221) and VT-2 examinaten.
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File: APP-AE1.WPD Appendix A-38 of A-38 ;
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v v v 48k gPausw James A. Fitzpatrick Nuclear Power Plant JAFruxn 2 APPENDIX B - Program Summary Tables Revissori: o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Jam-y s, tese Exam Item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Asethod Component. lD No. Items Sch'd 1 2* 3 Req Remarks / Comments Examination Category- C-A. PPISSURE RETAINING WELDS IN PRESSURE VESSELS C1.10 Shen Cacum'e enbal Weus Vol 10 RHR 10E-2A 3037 2 2 0 1 1 One vesselin a g oup of Vol 10 RHR 10E-28 3037 2 0 0 0 0 Vessels Total Examination stem 4 2 0 1 1 0% 50 % 100 %
C1.20 Head Cacum'erential Wees Vol 03CRD 03TK-1A 3003 2 2 1 1 0 One vesset in a group of Vol 03 CRD 03TK-1B 3003 2 0 0 0 0 Vesse ss Vol 10 RHR 102-2A 3037 1 1 0 0 1 One vesselin a group of Vol 'RHR 10E-2B 3037 1 0 0 0 0 Vessets Total Examination item 6 3 1 1 1 33 % $6% 100 %
C1,30 Tubesheet-to-Shes Weld Vol Not arphcable JAF Total Examination Category to 5 1 2 2 20 % 60 % 100 %
Notes:
(1) includes essentiaWy 100% of the weld length.
(2) Gross structural discontwuty is defined in NB-3213.2. Examples are junctions between sheffs o' dMerent thickness, cylindncal she8Mo-torncai shea junctens, shes (or headMange weUs, and head-t@B welds.
(3) in the case of muniple vessets of sanitar design, size, and senrece (such as steam generators, heat exchanges), the requwed ex a,Je 4 may be limited to one vesse! or distributad among the vessels.
(4) The vessel areas selected for the incal examenation shas be reexar med over the service Irfetrne of the -vv. + -
File: JAPP-BE1.WPD Appendix B - 1 of 8 - 29
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Exam item Description Exam System Une or ISO No. No. INSPECTTON I'ERIODS Ret 8 Req Item Method Component. ID No- Items Sch'd 1' 2* 3* Remarks! Comments Exammation CMegory:C-8. PRESSURE RETAINING NO2ZLE WELDS IN VESSELS C2.10 Nozzles in Vessets <1/2 b. Nommal Thckness C2.11 Nozzia-txShea (or Head) Weld Surf N/A WA NA 0 0 0 0 0 Ah nozz%s at te mr.at ends of pceg runs. One vessel in a grote of Vesses.
Not .ah to JAF C2.20 Nozzfes Wthout Remforong Plate m Vesse6s > % in. Nomnal Thoness C2.21 Nozzle-toShe5 (or Head) Weld AE nozzws at temealends of pceg runs VoVSurf 03CRD 03TK-1A 3003 1 1 1 0 0 Ore vessel m a g oup of VoVSu f 03 CRD 03TK-1B 3003 1 0 0 0 0 Vessets VoVSorf 10 RHR 10E-2A 3037 2 2 0 1 1 One vesset in a group of VoVSurf 10 RHR 10E-28 3037 2 0 0 0 0 Vessets Total Ex... ..euen item 6 3 1 1 1 33% 66 % 100 %
C2_22 Nozzie Inside Radrus Secton A3 eczzies at termmal ends of pceng runt Vol 03CRD 03TK-1 A 3003 1 1 1 0 0 One vesse4 m a group of Ved 03CRD 03TK-1B 3003 1 0 0 0 0 Vesseis Total Examination Item 2 1 1 0 0 100 % 100 % 100 %
C2.30 Nodes WGemforong P' ate in Vessets > % in. Nor wnal Thuness
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C2.31 Rer8crong Plate #ields to Nczzle and Vessel Su:1 AN nozzles at temunal ends of comg runs 0 0 0 0 0 Not we e to JAF C2.32 Nor'is &> MB (or leead) Welds When Insde of Vessel is Accessbe voi AE nozzWs at termmat ends of pceg runs 0 0 0 0 0 Not AA to JAF File: JAPP-BE1.VFD Appendix B - 2 of B - 29
V' N v g grower James A. Fitzpatrick Nuclear Power Plant APPENDIX B - Program Summary Tables 2Ar-is:40c2 nevisson- o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s. ine Exam item Description Exam System Line or ISO No. No. INSPECTION PER10CS Ret Item Method Cw6,, ,r.er,t. ID No. Items Sch'd 15T 2* 3* Reg Remarks / Comments C2.33 Neule-to-Shes (or Head) Welds When inside of Vessel is inaccessble VT-2 AR noules at temunal ends of pping runs.
The tentale hob sha8 be examened for evidence cf leakage O O O O O Not applicab4 to JAF Total Examination Category 5 4 2 1 1 50 % 75% 100 %
Notes:
(1) Includes nozzles welded to or integrapy cast in vessels that connect to piping runs (manways and handholds are exduded)
(2) include unty those piping runs selected for examination under Exammation Category C F.
(3) The nozzles selected initialty for examination shaE be reexamined over the service lifetime of the w:vv.m.t (4) in she cese of multiple vessets of similar desgn, size, and sennce (such as steam generators, heat exchangers), the required ex... k, = may be Irruted to one vessel or dist ibuted amor g the vessels.
(S) The tentab hole in remforcing plate shaE be examsned for evidence of leakage while vesse s is t, e,ww w the system pressure test (1WC-5221 or IWC-5222) as rW by Exammaton Category C-H.
7 File:JAPP-BE1.WPD Appendix B - 3 of B - 29
AY g g Poseer James A. Fitzpatrick Nuclear Power Plant JArass-0002 APPENDIX B - Program Summary Tables Rem.on: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s.1m Exam item Description Exam System Une or ISO No. No- INSPECTION PERIODS Ret item Method Component. lD No- items Sch'd 5 1' 2* 3* Reg Remarks / Comments Examination Category: C4. INTEGRAL ATTACHMENTS FOR VESSEL 3. PIPING. PUMPS AND VALVES C3.10 Pressure Vessel Integrasy Weided AW- a. 100% cf requwed areas of each welded attactv ent Surf 03 CRD 03TK-1A 3003 6 1 0 0 3 RR-4 One vesselin a gw of Surf 03 CRD 03TK-1B 3003 6 0 0 0 0 RR-4 Vesseh, CC N-509 Total CRD System 12 1 0 0 1 0% 0% 100 %
Surf 10 RHR 10E-2A 3037 6 1 1 0 0 RRA One vesselin a group of 10E2B 3037 6 0 0 0 0 RR-4 vessels, CC N-509 Total RHR System 12 1 1 0 0 100 % 100 % 100 %
Total Examination item 24 2 1 0 1 50 % 50 % 100 %
C3.20 Piping integra#y Welded Attachrnents 100% of requeed areas of each welded attactwrwrt 10 RHR 24~-W20-152-3A 3010 2 0 0 0 0 RR-4 CC N-509 24*-W20152-38 3009 2 0 0 0 0 RR-4 24*-W20-302-11 A 3006 4 1 0 1 0 RR-4 24"-W20-302-17 3006 1 0 0 0 0 RR-4 24"-W20-302-118 3006 2 0 0 0 0 RR-4 TOTAL 24.0* 11 1 0 1 0 0% 100 % 100 %
C3 20 Surf to RHR 2Cr-W20-152-2A 3010 1 0 0 0 0 RR-4 CC N-509 20"-W20-152-28 3009 9 1 1 0 0 RR-4 20"-W20-152-2C 3010 11 1 0 1 0 RR4 20~-W20-152-2D 3009 1 0 0 0 0 RR-4 20"-W20-302-8A 3005 3 0 0 0 0 RR-4 20"-W20-302-8B 3004 4 2 1 0 1 RR-4 20"-W20-302-17 3006 1 0 0 0 0 RR-4 TOTAL 20.0* 30 4 2 1 1 50 % 75 % 100 %
Fi!e: JAPP-BE1.WPO Appendix B -4 of B -29
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.: i j \ / a g geswer James A. FitzpatriCk Nuclear Power Plant 2Ar asi-0002 APPENDIX B - Program Summary Tables Revisieru o ASME CODE CLASS 2 SYSLARS AND COMPONENTS J.nuary s. ms Exam item Deseription Exam System Une or ISO No- No- INSPECT ON PERIODS Rel item Method Component ID No- Items Sch'd 1" 2* 3* Req Remarks; Comments C320 Surf 10 RHR 16"-W20-320-5A 3017 2 0 0 0 0 RR-4 CC N-509 16"-W20-152-5B 3016 2 1 0 0 1 RR-4 16"-W20-302-7A 3005 2 1 0 0 1 RR-4 16"-W20-302-7B 3004 1 0 0 0 0 RR-4 Excluded ff&2500-1(1Xc) 16 -W20-302-7C 3005 1 0 0 0 0 RR4 16"-W20-302-7D 3004 2 0 0 0 0 RR-4 16"-W20-302-9B 3004/15 4 0 0 0 0 RR-4 16~-W20-320-10A 3005 4 0 0 0 0 RR-4 16~-W20-320-108 3004 3 1 0 1 0 RR-4 16"-W20-302-15A 3007/17 3 0 0 0 0 RR-4 16"-W20-302-158 3008/16 11 1 1 0 0 RR-4 16*-W20-302-34 3014 4 1 1 0 0 RR-4 16"-WS-302-368 3004 1 0 0 0 0 RR-4 16*-WS-302-56A 3005 2 0 0 0 0 RR4 TOTAL 16.0" 42 S 2 1 2 40 % 60 % 100 %
12*-W20-302-13A 3007 4 1 1 0 0 RR-4 CC N-509 12"-W20-302-13B 3008 7 2 0 1 1 RR-4 TO'. AL 12.0" 11 3 1 1 1 33 % 66 % 100 %
10"-W20-302-12A 3007 4 0 0 0 0 Excluded FWC-2500-1 10"-W20-302-128 3008 3 0 0 0 0 Excluded fWC-2500-1 TOTAL 10.0" 7 0 0 0 0 0% 0% 0%
v C320 Serf 10 RHR 8*-SHP-902-32A 3014 5 2 0 1 I RR-4 CC N-509 8" SHP-902-32B 3n15 5 1 1 0 0 RR-4 8"-W20-302-38 3006 12 0 0 0 0 RR-4 Exdudedh h N 1 8*-W20-152-39 3009 8 0 0 0 0 RR-4 Excfuded hh%1, size TOTAL 8.0" 30 3 1 1 1 33 % 66 % 100 %
6"-SLP-302-34A 3014 3 0 0 0 0 Excluded IWC-2500-1 sire 6~-SLP-302-348 3015 4 0 0 0 0 Excluded IWC-2500-1 size 6~-W20-152-44B 3016 1 0 0 0 0 Excluded IWC-2500-1 size 6"-W20-152-44A 3017 1 0 0 0 0 Excluded IVC 2500-1 size File: JAPP-BE1.WPD Appendix B - 5 of B -29
- > gPnser James A. Fitzpatrick Nuclear Power Plant jar- S -0002 APPENDIX B - Program Summary Tables Re 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS a.nuny 6.1,9s Exam item Description Exam System Line or ISO No. No- INSPECTION PERIODS Ret item Method Component. RD No- ttems Sch'd 5 1' 2* 3* Reg Remarks / Comments TOTAL 6.0" 9 0 0 0 0 0% 0% 0%
Tots' PNG System 140 16 6 5 5 37 % 68 % 100 %
Surf 13 RCIC 8' 61P-152-22 3019 1 0 0 0 0 EnduoJd IWC-2500-1 size Surf 13 RCfC 6 -W22-152-16 3020 1 0 0 0 0 Exduded IWC-2500-1 see Total RCIC System 2 0 0 0 0 0% 0% 0%
Surf 14 CS 16" W23-152-1 A 3021 4 0 0 0 0 RR-4 CC N-509 16"-W23-152-18 3021 5 1 0 0 1 RR-4 12"-W23-302-4A 3022 6 0 0 0 0 RR4 12"-W23-302-48 3023 7 1 0 1 0 RR-4 10" W23-302-8A 3022 3 0 0 0 0 RR4 Enduded fvC,-2500-1 see 10" W23-302-88 3073 2 0 0 0 0 RR-4 Erduded IV&2500-1 size 10"-W23-163-9A 3022 1 0 0 0 0 Enduded IWC-25001-1 see 10"-W23-163-98 3023 1 0 0 0 0 Enduded IWC-2500-1 size TOTAL CS System 29 2 0 1 1 25% 50 % 100 %
Surf 15 RBCLC 6"-WCL-151-29A 103F 2 0 0 0 0 RR-4 Exduded IWC-2500-1. see Total RBCLC System 2 0 0 0 0 0% 0% 0%
r C3.20 Surf 23 HPCI 24 -SLP-152-26 3027 1 0 0 0 0 RR-4 Enduded fWC-2500-1 Me 20"-SLP-152-25 3027 5 2 1 1 0 RR-4 CC N-509 16 -WCP-152-1 3028 2 1 1 0 0 RR-4 16"-W25-152-7 3028 3 2 0 1 1 RR-4 14 -W25-902-3 3025 7 1 0 1 0 RR-4 10"-W25 902-4 3026 1 0 0 0 0 RR-4 Esduded fV&2500-1 size 10"-SHP-902-19 3024 8 0 0 0 1 RR-4 Enduded IWC-2500-1 size TOTAL HPCI System 27 7 2 3 2 33% 66 % 100 %
File: JAPP-BE1.WPD Appendix B -6 of B -29
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LJ J g geweer James A. Fitzpatrick Nuclear Power Plant APPENDIX B - Program Summary Tables 24F-tSur;02 Revisiert 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s.1998 Exam item Description Exam System Une or ISO No. No. INSPECTION PERfODS Rei Item Method Component. ID No. Items Sch'd 15' 2* 3* Reg Remarks / Comments C3.20 Surf 27 CP 24*-N-152-18A 3079 1 0 0 0 0 Endudedh h N 1see 2C"-N-152-16 3030 2 0 0 0 0 Excluded hE'500-1 sire 20*-N-152A-20 3029 2 0 0 0 0 Excluded hvC-2500-1 see 18*-N-152A-6 3029 1 0 0 0 0 Exduded IWC-2500-1 size 14*-N-152A-9 3029 1 0 0 0 0 Excluded hvC-2500-1 size Tota! CP System 7 0 0 0 0 0% 0% 0%
Surf 29 MS 24*-SHP-902-101 A 3031 2 1 1 0 0 RR-4 24*-SHP-902-101B 3031 2 0 0 0 0 RR-4 24*-SHP-902-101C 3032 2 0 0 0 0 RR-4 24*-SMP-902-101D 3032 2 0 0 0 0 RR-4 TOTAL MS System 8 1 1 0 0 100 % 100 % 100 %
Total Examination item 210 25 8 9 8 33 % 66 % 100 %
C3.30 Pump Integra9y Welded Attachmonts 100% of requwed areas of each welded attactw,ent Surf 0 0 0 0 0 Not apphcable to JAF C3 40 Vahre Integratty Welded Attachments 100% of sequwed areas of each welded attachment Surf 0 0 0 G 0 Not applicable to JAF TOTAL EXAMINATION CATEGORY 234 27 9 9 9 33 % 66 % 100 %
Notes:
r (1) The examination is Emited to those etegrally welded attachtrents that meet the followeg windawis:
(a) The attachment is on the outside surface of the pressure retaining www as; (b) the attachment provides component support as defined m NF-1110.
(c) the attachment base material design thickness is 3/4 in. Or greater, and (d) the attacher.ent weld joins the attachment either directfy to the surface of the ww 64 of to an integraWy cast or forged attachment to the wyv.e,1 (2) in case of multiple vessets of similar design and sennce, the required exammatens may be conducted on only one vessel. Where multiple vessets are prowded with a r umber of simitar attachments, the examinaton of the attachment may be distributed among tne vessals.
(3) The areas selected for the initial examination shall be reexarnined over the service Efetime of the wwwi
(') Limited to attachments of those wi ~,e.-as requred to be examined under Examination Categones C-F and C-G-(5) Selection included only those items not exempted by fWC-1220. or excluoed from enamination per fWC-2500-1 Tables.
Examinations will be pmhi.md in accordance witn the attemate requirements of Code Case N-509 %'temative Rules for the Selecton and Examination cf Class 1. 2 and 3 Integrapy Weided (7)
Attachments". Secton XI. Division 1.Only one vessel in a group of vessets is required.
File: JAPP-BE1.WPD Appendix B -7 of 8 -29
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- > g Peveer James A. Fitzpatrick Nuclear Power Plant JAF N 2 APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6,1m Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret Item Method Component. ID No. Items Sch'd 1 87 2* 3" Reg Remarks / Comments Examination Category: C-0. PRESSURE RETAINING BOLTING GREATER THAN 2 in. IN DtAMETER C4.10 Pressure Vessel BoRs and Studs Vol 100% bots and studs at each bored w-,vde, of wr w.v as requwed to be inspected Vol 10 RHR 10E.2A 3037 72 0 0 0 0 One vesselin a group of Vol 10 RHR 10E-2B 3037 72 0 0 0 0 Vesse4. Excluded thN 1., <2.0" Diameter C4.20 Piping Bons and Studs Vol 100% bots and studs at each bored w-Gi of w.w. cots regurred to be inspected 0 0 0 0 0 Not apphcaele to JAF C4.30 . Pumps BoRs and Studs Vol 100% bots and studs at eadi bored w-G, of w.w,vots requred to be inspected 0 0 0 0 0 C4 40 Vatves Bons and Studs Vol 100% bots and studs at each boned connectkm of wiw vi.ie requ red to be inspected O O O O O Not apphcable to JAF Total Examination Category 144 0 0 0 0 0% 0% 0%
Notes:
(1) The examination may be perfermed on botmg in place under load or upon drsassembly of the connectert (2) The examination of boning for vessels, pumps, and vatves may be conducted on one vessel, one ptwnp one vane among a group of vessels, pumps, and valvas in each system required to be examined and which are similar in design, size, functon, and service. In adddion, where the one w,weat to be exammed contains a group of bored connectens of similar design and size (such as flange connections, manway covers), it e examination may be conducted on one bored cormecten among tte group. t (3) The examination of flange botmg in piping systems required to % exammed may be fwnited to the flange connectens in pipe runs seected for examination under Exammaten Category C-F.
(4) The areas selected for initial examination shall be reexamined over the service lifetrne of the w.w ent.
File: JAPP-BE1.WPD Appendix B - 8 of B - 29
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g g Po w James A Fitzpatrick Nuclear Power P: ant APPENDIX B - Program Summary Tables jar-tSim02 Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6,19M Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rel item Method Component. ID No. Items Sch'd 1 2* 3* Req Remarks / Comments Examination Category: C-F-1. PRESSURE RETAINING YELDS IN AUSTENmC STA'NLESS STEEL OR HIGti ALLOY PIP 1NG C3.10 Piping Weids > 3/8 in. Nominal Wall Thickness for Piping > NPS 4 Not applicable to JAF C5.11 circumferential Weld StWNot 14 CS 10"-W23-163-9A 3022 6 1 1 0 0 1C0% of each wet requiring 10" W23-163-9B 3023 6 2 0 1 1 Exammation. Exempt fWC-2500-1 Total Examination item 12 3 1 1 1 33% 66 % 100 %
C5.12 LongitudinalWeld SurfNoi 0 0 0 0 0 2.St - at the intersectrng creumferential weld C5.20 Piping Welds > 1/5 in NominalWatt Thickness for Piping > NPS 2 and < NPS 4 C5.21 Circumferential Weld SurfNo! 0 0 0 0 0 100% of each weld requiring examination C5.22 Longitudinal Weld SurfNoi 0 0 0 0 0 2.St - at the intersecting circu... beg 1 weld C5.30 Socket Welds Su:1 0 0 0 0 0 100% of each weld requong examination CSXO Pipe Branch Connections of Branch Piping > NPS 2 C5.41 CircumferentialWeld Surf N/A N/A N/A 0 0 0 0 0 100% of each wald requiring. not applicable to JAF C5.42 Longitudinal Weld Surf 0 0 0 0 0' 2.5t - at the intersectmg cwcumferentist weld Total Examination Category 12 3 1 1 1 33 % 66 % 100 %
File: JAPP-BE1.WPD Appendix B - 9 of B - 29
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- > gemer James A. Fitzpatrick Nuclear Power Plant [[::JAF-tSI-0002|JAF-tSI-0002]]
, APPENDIX B - Program Summary Tables nevissorr o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Jano.ry s,im Exam item Descriptior Exam System Line or ISO No. No. INSPECTION PERIODS Rel
- Item FSthod Component. lO No. Items Sch'd 187 2* 3'* Req Remarks /Commente Notes:
Requirements for examinatic.. of welds in piping s NPS 4 apply to PWR high pressure safety inreden systems in accordance with the exempton criteria d IWC-1220. I (1)
(2) The welds selected for examination sha9 include 7.5%, but not less than 28 wekts, of aR austenttic stainless steet of high ahoy welds not exempted ty IWC-1220. (Scme welds not exerrpted by IWC-1220 are not required to be ,westrucuvely examined per Examination Category C-F-1. These nids, however, sha3 be included in the total weld count to which the 7.5% samphng rate is applied.) The examinations shaR be distributed as follows:
(a) the examination shaR be distributed among the Class 2 systems prorated, to the deg ee practicable, on the number of nonexempt w ee udive examinations required by Examination Category C-F-1 should be imJ M on that system);
(b) within a system, the exa ninstions shaN be distributed among termmal ends and stmetural discontinuites (See Note (3)) prorated, to thw degree practicable, on the number of nonexempt terminal ends and structural discontinuities in that system; and (c) within the system, examinations shau ba 6strbuted between line sizes prorated ta the degree practicable.
(3) Struc 4.31 discontinuities include pipe weld joints to vessel nozzles, vahre bodies, pump casings pipe f?'gs (such as elbows, tees, reducers, flanges, etc. Jwna s to ANSI B16.9), and pipe tranch connections ard fittings.
(4) The wekts selected for examination sha9 be reexamined during subsequent inspection intervats over the service lifetime of the piping m.wnt.
t File JAPP-BE1.WPD Appendix B - 10 of 8 - 29
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$ gRn*er James A. Fitzpatrick Nuclear Power Plant APPENDIX B - Prograin Summary Tables [[::JAF-ISI-0002|JAF-ISI-0002]] R vision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January e.19es Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. !D No. Ite m s Sch'd 187 2* 3" Req Remarks / Comments Examination Category: C.F-2, PRESSURE RETAINING WELD
- 1N CARBON OR LOW ALLOY STEEL PtPING C150 Piping Welds > 3/8 in. Nominal Was TNckness fw Piping > NPS 4 C5.51 CircumferentialWeld 100% of each weld requeng examination Vol/ Surf 03 CRD 10"-WR-901 3039 15 2 2 0 0 10"-WR-901 3040 13 1 1 0 0 8"-WR-901 3039 10 2 2 0 0 8"-WR-901 3040 10 0 0 0 0 Total CRD System 48 5 5 0 0 100 % 100 % 100 %
Vot/ Surf 10 RHR 24*-W2G 152-3A 3010 9 1 0 0 1 24"-W20-152-3B 5009 10 1 1 0 0 24"-W20-302-11 A 3003 18 3 0 2 1 24*-W20-302-17 3006 5 1 1 0 0 24*-W20-302-118 3006 15 3 1 1 1 Total 24.0" RHR 57 9 3 3 3 42% 71 % 1C3%
20"-W20-152-2A 3010 16 1 1 0 0 20"-W20-152-2B 3009 16 2 0 0 2 20"-W20-152-2C 3010/11 22 2 1 1 0 20"-W20-152-2D 3009 10 1 0 1 0 20*-W20-1524A 3010 1 0 0 0 0 20"-W20-152-CB 3009 9 1 0 0 1 20"-W20-152-6C 3010 9 1 1 0 0?
20"-W20-152-6D 3009 9 1 0 1 0 20"-W20-302-8A 3005 6 1 0 1 0 20"-W20-302-88 3004 6 0 0 0 0 20" W20-302-17 3006 13 1 0 0 1 20"-SLP-302-33A 3014 3 0 0 0 0 20 -SLP-302-33B 3015 3 1 1 0 0 Total 20.0" RHR 123 12 4 4 4 33 % 66 % 100%
File: JAPP-BE1 WPD Appendix B - 11 of B - 29
N-g gPower James A. Fitzpatrick Nuclear Power Plant JAF-tS!M2 APPENDIX B - Program Summary Tables Revi.ioru O ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Janu.ry s.1 ,es Exam item Description . Eram System Line or ISO No. No- INSPECTION PERIODS Ret item Method Component. ID No- Items Sch'd 1st 2* 3*" Req Remarks / Comments C5.51 16"-W20-320-5A 3017 7 0 0 0 0 16"-W20-152-5B 3016 10 1 1 0 0 16*-W20-302-7A 3005 9 1 0 0 1 16* W20-302-7B 3004 10 1 0 1 0 16"-W20-302-7C 3005 10 1 0 0 1 16*-W20-302-7D 3004 9 1 1 0 0 16*-W20-302-9A 3005 1 0 0 0 0 16"-W20-302-98 3004/15 16 2 1 0 1 16"-W20-302-15A 3007/17 12 2 2 0 , 0 16"-W20-302-15B 3008/16 11 0 0 0 0 16*-W20-302-34 3014 12 1 0 0 1 16*-WS-302-368 3004 11 1 1 0 0 16*-WS-302-56A 3005 10 1 0 1 0 16*-WS-302-10B 3004 16 1 0 0 1 16*-W20-302-10A 3005 15 1 0 1 0 Total 16.0" RHR 159 13 6 3 5 46% 76% 100 %
14*-W23-152-98 3015 2 1 0 0 1 Total 14.0" RHR 2 1 0 0 1 0% 0% 100 %
12"-W20-302-13A 3007 22 2 1 1 0 12"4V20-302-138 3008 19 1 0 0 1 Total 12.0" RHR 41 3 1 1 1 33% 66% 100 %
7 10"4V20-302-12A 3007 28 0 0 0 0 Excluded IWC-2500-1 size 10 -W20-302-128 3008 30 0 0 0 0 Excluded IWC-2500-1 size Total 10.0" RHR 58 0 0 0 0 0% 0% 0%
C5.51 8*-SHP-902-32A 3014 31 3 1 1 1 8"-SHP-902-32B 3015 33 6 1 4 1 8"-SLP-302-37B 3015 4 0 0 0 0 8"-W20-302-38 3006 34 0 0 0 0 Exduded M.'C-2500-1 size 8*-W20-152-39 3009 22 0 0 0 0 Excluded IWC-2500-1 size File: JAPP-BE1.WPD Appendix B - 12 of B -29
O LJ Q) p> grower James A. Fitzpatrick Nuclear Power Plant JAF-:S 40o2 APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS .l.nu.ry 6,1eos Exam ' Item Desc-iption Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component ID No. Items Sch'd is' 2* 3* Req Remarks / Comments Total 8.0" RHR 124 9 2 5 2 12% 77% 100 %
C5.51 Vol/ Surf 10 RHR 6*-W20-302-16A 3017 4 0 0 0 0 Erduded fWC-2500-1 size 6"-W20-302-16B 3016 2 0 0 0 0 Exduded fWC-2500-1 size 6"-SLP-302-34A 3014 20 0 0 0 0 Erduded IWC-2500-1 size 6"-SLP-302-34B 3015 19 0 0 0 0 Exduded fWC-2500-1 stre 6"-W20-152-44A 3017 12 0 0 0 0 Exduded fWC-2500-1 size 6"-W20-152-44B 3016 12 0 0 0 0 Enduded IVC 2500-1 size Total 6.0" RHR 69 0 0 0 0 0% 0% 0%
Total RHR System 63J 48 16 16 16 33% 66 % 100 %
Vol/ Surf 13 RCIC 8"-SLP-152-22 3019 13 0 0 0 0 Exduded IWC-2500-1, size 6"-W22-152-16 3020 1 1 1 0 0 Total RCIC System 14 1 1 0 0 100 % 100 % 100 %
Vol/ Surf 15 RBCLC 6"-WCL-151-29A 103F 7 0 0 0 0 Exduded IWC-25001, size Total RBCLC System 7 0 0 0 0 0% 0% 0%
C5.51 Vol/ Surf 14 CS 16"-W23-152 ' A 3021 9 0 0 0 0 16"-W23-152 B 3021 9 0 0 0 0 16"-W23-152-3A 3021 4 1 0 1 0' 16*-W23-152-38 3021 4 1 1 0 0 12"-W23-152-2A 3021 1 0 0 0 0 12"-W23-152-2B 3021 1 0 0 0 0 12"-W23-302-4A 3022 30 3 1 1 1 12"-W23-302-4B 3023 24 2 0 1 1 10"-W23-302-8A 3022 2 1 1 0 0 Exduded IVC 2500-1, sizs 10 -W23-302-80 3023 2 1 0 0 1 Exduded fWC-2500-1, size Total CS System 86 9 3 3 3 33 % 66 % 100 %
File: JAPP-BE1.WPD Appendix B - 13 of B - 29
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n U U, U e l}lsw wrkpswer James A. Fitzpatrick Nuclear Power Plant jar 4SI-0002
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APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s,199s -
Exam item Description Exam System Une or ISO No. No, INSPECTION PERIODS Ret item Method Component ID No- Items Sch'd 8 1' 2* 3* Req Remarks / Comments C5.51 VoVSurf 23 HPCI 24*-SLP-152-26 3027 7 1 0 0 1 20*-SLP-152-25 3027 17 1 0 0 1 16"-WCP-152-1 3028 23 2 0 1 1 16"-W25-152-7 3028 12 3 0 1 2 14*-W25-902-3 3025 25 3 1 1 1 10"-W25-902-4 3026 8 1 0 1 0 10*-SHP-902-19 3024 27 3 1 1 1 10"-SHP-902-34 3024 8 1 1 0 0 Total HPCI System 127 15 3 5 7 20% 53 % 100 %
VoVSurf 27 CP 30 -N-152-13A 3030 10 0 0 0 0 Excluded IWC-2500-1, she 30"-N-152-13B 3030 10 0 0 0 0 Enduded !WC 2500-1, size 30*-N-152-13C 3030 10 0 0 0 0 Enduded IWC-2500-1. size 30"-N-152-13D 3030 10 0 0 0 0 Excluded IWC-2500-1, size 33*-N-152-13E 3030 10 0 0 0 0 Exduded IWC-2500-1, size 24"-N-152A-8 3029 3 0 0 0 0 Excluded IWC-2500-1, size 24"-N-152A-18A 3029 6 0 0 0 0 Exduded iWC-2500-1 size 24"-N-152A-1 BB 3029 3 0 0 0 0 Exc!uded lWC-2500-1, size 20"-N-152A-7 3029 5 0 0 0 0 Excluded fWC-2500-1, sim 20*-N-152A-16 3030 9 0 0 0 0 Excluded fWC-2500-1, size 20"-N-152A-20 3029 12 0 0 0 0 Exduded IWC-2500-1, size 18*-N-152A-6 3029 5 0 0 0 0 Excluded IWC-2500-1, size 18"-N-152A-26A 3029 4 0 0 0 0 Excluded IVC-2500-1, size 18"-N-152A-268 3029 4 0 0 0 0 Excluded IWC-2500-1, size 14*-N-15A-9 3029 14 0 0 0 0 Exduded IWC-2500-1, size Total CP System 115 0 0 0 0 t
Vol/ Surf 29 MS 24*-SHP-902-101 A 3031 1 0 0 0 0 24*-SHP-902-101 B 3031 1 0 0 0 0 24*-SHP-902-101C 3032 1 0 0 0 0 24"-SHP-902-101D 3032 1 0 0 0 0 Total MS System 4 0 0 0 0 0% 0% 0%
Total Examination item 1034 78 28 25 25 35% 67% 100 %
File: JAPP-BE1.WPD Appendix B - 14 of B - 29 L __ __ ._
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- ,D gPower James A. Fitzpatrick Nuclear Power Plant JAFas:4002 APPENDIX B - Program Summary Tables Revi. ion: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6 m8 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret 5
item - Method Component. ID No. Items Sch'd 1' 2* 3* Req Remarks / Comments C5.52 Longitudinal WeM VoVSurf 0 0 0 0 0 2.51- at the intersechng swJwwaial weM. Not appreable to JAF C5.60 Piping WeMs > 1/5 in. NominalWa5 Thickness for Piping > NPS 2 and < NPS 4 Does not apply to BWR's C5.61 CircumferentialWeM Vol/ Surf 03 CRD 3.0" Line 3003 5 0 0 0 0 100% of each weM requong exammadon 10 RHR 4*- W20-902-36 3012 4 0 0 0 0 Does not apply to BNs 4*-W20-902-35 3012 9 0 0 0 0 Does not app!y to BWR's Total Examination item 18 0 0 0 0 Does not apply to BWR*s C5.62 Longitudinal Wed Vol/ Surf 0 0 0 0 0 2.5t - at the intersecting crci cJmm.^o 4 weM Does not apply to BWR's C5.70 Socket Welds Surf 0 0 0 0 0 100% of each weld requinng examination, Not applicate to JAF C5.80 Pipe Branch Connections of Branch Piping > NPS 2 C5.81 CircumferentialWeld Surf 100% of each weM requmng exarmnation 10-RHR 24"-W20-302-11 A 3006 1 1 0 1 0 24*-W20-302-11 B 3006 2 0 0 0 0 Total 24.0'* RHR 3' 1 0 1 0 0% 100 % 100 %
20*-W20-302-8A 3005 4 1 0 0 1 20"-W20-302-8B 3004 4 2 1 1 0f '
20*-W20-302-17 3006 1 0 0 0 0 Total 20.0" RHR 9 3 1 1 1 33% 66 % 100 %
16"-W20-320-5A 3017 2 2 0 1 1 16"-W20-152-5B 3016 2 1 0 0 1 16"-W20-302-15A 3007/17 2 1 1 0 0 16*-W20-302-158 3008/16 1 0 0 0 Total 16.0" RHR 7 4 1 1 2 25% 50 % 100 %
File: JAPP-BE1.WPD Appendix B - 15 of 8 - 29
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A NewVukPtmer James A. FitzpatriCk Nuclear Power Flant [[::JAF-4SI-0002|JAF-4SI-0002]] W Authority APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6.19ee Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rel item Method Component. lD No- Items Sch'd ,1 5 2* 3" Req Remarks / Comments C5.81 Surf 10 RHR 10"-W20-302-128 3008 2 ( 0 0 0 Excluded IWC-2500-1 size Total 10.0" RHR 2 0 0 0 0 0% 0% 0%
8"-W20-302-38 3006 1 0 0 0 0 Excluded rWC-2500-1 size Total 8.0" RHR 1 0 0 0 0 0% 0% 0%
6"-W20-152-44A 3017 1 0 0 0 0 Excluded IWC-2500-1 size Total 6.0" RHR 1 0 0 0 0 0% 0% 0%
Total RHR System 23 8 2 3 3 33% 66 % 100%
Surf 14 CS 16"-W23-152-1 A 3021 1 0 0 0 0 16"-W23-152-1B 3021 1 0 0 0 0 12"-W23-302-48 3023 1 1 0 1 0 10"-W23-163-8A 3022 1 1 1 0 0 Total CS System 4 2 1 1 0 50 % 100 % 100%
Surf 23 HPCI 20"At!'-152-25 3027 1 1 0 0 1 10"-SHP-902-19 3024 1 1 0 1 0 Total HPCI System 2 2 0 1 1?
0% 50 % 100%
Total Examination item 39 12 3 5 4 25% 66% 100 %
C5 82 LongitudinalWeld Surf 0 0 0 0 0 2.5t - at the intersecting circumferentet weld Total Examination Category 1081 90 31 30 29 34 % 67% 100 %
Fila: JAPP-BE1.WPD Appendix B - 16 of B - 29
- /x ,m o U U U g g revner James A. Fitzpatrick Nuclear Power Plant JAF-!S14002 l
APPENDIX B - Program Summary Tables nevision: o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s,1em i
Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component ID No. Items Sch'd 1'8 2* 3'" Req Remrrks/ Comments Notes:
(1) Requirements for examination of welds in piping s NPS 4 apply to PWR high pressure safety irjection systems in accordance with the exemption crteria of hNC-1223.
(2) The webs selected for exardnation shas include 7.5%, but not less than 28 welds, of aR austenite stainless steel c8 high a!!oy wees not exempted by fWC-1220.
(Some welds not exempted by IWC-1220 are not required to be nondestructivety exammed per Examination Category C-F-1. These welds, however, shall be included in the total we'd count to which the 7.5% sampling rate is appned )The examinations shan be distributed as fonows:
(a) the examination shaR be distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt nondestrudive examinations required by Examination Category C-F-1 should be peiL +j on that system);
(b) within a system, the examinations shs3 be distributed among termmal ends and structural discontinuities [See Note (3)] prorated, to the degree predicable, on the number of nonexempt terminal ends and structural discontinuities in that system; and (c) within 'he systerg examinations shaR be distributed between line sizes prorated to the degree pradicable.
(3) Structural discontinuities include pipe weld joints to vessel nozzles, va!ve bodies, pump casings, pipe fittings (such as elbows, tees, reducers, flanges, etc.,
conformrng to ANSI B16.9), and pipe branch connections and fittings.
The welds selected for examination shan be reexammed during subsequent inspection interva!s over the sennce lifetime of the piping www-,t.
(4)
(5) Onfy those welds showing reportable preservice transverse indicatsns need to be examined for transverse reflectors.
t File: JAPP-BE1.WPD Appendix B - 17 of B -29
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4, grewer James A. FitzpatriCk Nuclear Power Plant APPENDIX B - Program Summary Tables -
JAF41-0002 nevi ion: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s me Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'J 5 1' 2" 3* Req Remarks / Comments Examination Category: C-G. PRESSURE RETAIN!NG WELDS IN PUMPS AND VALVES C6.10 Pump Casing WeMs Surf 0 0 0 0 0 wees in as w.w m.e in each peng run exammed tmder Catagory C.F Nd app 6 cab 8e to JAF C620 Vahe Body WeMs Surf 0 0 0 0 0 100% welds in all ,
w.w m.G in each piping run examined under Category C-F, NM apphcable to JAF Total Examination Category 0 0 0 0 0 0% 0% 0%
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File: JAPP-BE1.WPD Appendix B - 18 of B - 29
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James A. Fitzpatrick Nuclear Power Plant [[::JAF-tSI-0002|JAF-tSI-0002]] g gPower APPENDIX B - Program Summary Tables Revision: o January 6.1998 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Exam System Line or ISO No. No. INSPECTION PERIODS Rel Ex m item Description 5 2" 3" Rec Remarks /Cvmments Method Component. ID No. Items Sch'd 1' item Examination Category: C-H, ALL PRESSURE RETAINING COMPONENTS Pressure Vessels Pressure retaining boundary. Each inspection Period. IWC-5221 Test C7.10 Pressure Retaining Compor'ents VT-2 RR-10 VT-2 03 CRD 03TK-1A 3038 1 3 1 1 1 3038 1 3 1 1 1 RR-10 VT-2 03 CRD 03TK-1B 3037 1 3 1 1 1 RR-10 VT-2 10 RHR 10E-2A 3037 3 1 1 1 RR-10 VT-2 10 RHR 10E-28 1 4 12 4 4 4 Total Examination item Pressure retainhg boundary, each inspection Intervci, IWC-5222 Test C7.20 Pressure Retaining Components VT-2 RR-3/10 03TK-1A 3038 1 1 0 0 1 VT-2 03 CRD 3038 1 1 0 0 1 RR-3/10 VT-2 03 CRD 03TK-1B 3037 1 1 0 0 1 RR-3/10 VT-2 10 RHR 10E-2A 3037 1 1 0 0 1 RR-3/10 VT-2 10 kHR 10E-2B 4 4 0 0 4 Total Examination item Piping Pressure retaining boumlary Each inspecbon Period, IWC-5221 Test.
C7.30 Pressure Retaining Components VT-2 RR-10 VT-2 C3CRD 10"-WR-901 3039/40 1 3 1 1 1 3039/40 1 3 1 1 1 RR-10 8*-WR-901 3010 1 3 1 1 1 RR-10 VT-2 10 R HR 24"-W20-152-3A 3009 3 1 1 RR-10 24"-W20-152-38 1 3006 3 1 1 1 RR-10 24" W20-30. * '* 1 3006 3 1 1 1' RR-10 24*-W20-302 './ 1 20"W20-152-2A 3010 1 3 1 1 1 etR-lu 3009 3 1 1 1 RR-10 l
' 20"-W20-152-28 1 3010 1 3 1 1 1 RR-10 20"-W20-152-2C 3009 1 3 1 1 1 RR-10 4 20"-W20-152-2D 3010 1 3 1 1 1 RR-10 20"-W20-152-6A 3009 3 1 1 1 RR-10 20"-W20-152-6B 1 3010 1 3 1 1 1 RR-10 20"-W20-152-6C 3009 1 3 1 1 1 RR-10 20"-W20-152-6D 3005 3 1 1 1 RR-10 20"-W20-30A8A 1 3004 3 1 1 1 RR-10 20"-W20 *J2-8B 1 Appendix B - 19 of B -29 Fife: JAPP-BE1.WPD
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g g P m er James A. Fitzpatrick Nuclear Power Plant Jc-is -0002 APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6,1998 Exam . Item Description Exam System IIne or ISO No. No- INSPECTION PERIODS Rel Method Component ID Items 5 item No. Seh'd 1' 2* 3* Req Remarks / Comments C7.30 VT-2 10 RHR 20"-W20-302-17 3006 1 3 1 1 1 RR-10 20"SLP-302-33A 3014 1 3 1 1 1 RR-10 20"-SLP-302-33B 3015 1 3 1 1 1 RR-10 16"-W20-320-5A 3017 1 3 1 1 1' RR-10 16*-W20-152-5B 3016 1 3 1 1 1 RR-10 16*-W20-302-7A 3005 1 3 1 1 1 RR-10 16*-W20-302-78 3004 1 3 1 1
- RR-10 16"-W20-302-7C 3005 1 3 1 1 1 RR-10 16"-W20-302-7D 3004 1 3 1 1 1 RR-10 16 -W20-302-9A 3005 1 3 1 1 1 RR-10 16*-W20-302-98 3004 1 3 1 1 1 RR-10 16*-W20-302-15A 3007 1 3 1 1 1 RR-10 16"-W20-302-158 3008 1 3 1 1 1 RR-10 16*-W20-302-34 3014 1 3 1 1 1 RR-10 16"-W20-302-36B 3004 1 3 1 1 1 RR-10 16*-WS-302-56A 3005 1 3 1 1 1 RR-10 14*-W23-152-9A 3017 1 3 1 1 1 RR-10 14"-W23-152-98 3016 1 3 1 1 1 RR-10 12"-W20-302-13A 3007 1 3 1 1 1 RR-10 12"-W20-302-138 3008 1 3 1 1 1 RR-10 10"-W20-302-12A 3007 1 3 1 1 1 RR-10 10*-W20-302-128 3008 1 3 1 1 1 RR-10 8*-SHP-902-32A 3014 1 3 1 1 1 RR-10 8"-SHP-902-32B 3015 1 3 1 1 1 RR-10 8*-SLP-302-37B 3015 1 3 1 1 1 RR-10 8"-W20-302-38 300G 1 3 1 1 1 RW-10 8*-W20-152-59 3008 1 3 1 1 1 RR-10 6"-W20-302-16A 3017 1 3 1 1 1 RR-10 6*-W20-302-16B 3016 1 3 1 1 1 RR-10 6"-SLP-302-34A 3014 1 3 1 1 1' RR-10 6"-SLP-302-34B 3015 1 3 1 1 1 RR-10 6"-W20-152-44A 3017 1 3 1 1 1 RR-10 6"-W20-152-44B 3016 1 3 1 1 1 RR-10 4"-W20-302-19A 3017 1 3 1 1 1 RR-10 4*-W20-152-408 3016 1 3 1 1 1 RR-10 4"-SLP-302-86A 3014 1 3 1 1 1 RR-10 4*-SLP-302-86B 3015 1 3 1 1 1 RR-10 VT-2 13 RCIC */-SLP-152-22 3019 1 3 1 1 1 RR-10 6*W22-152-16 3020 1 3 1 1 1 RR-10 File: JAPP-BE1.VPI Appendix B - 20 of B - 29
,.m (V) . uJ (n) (m'uj) g gPower James A. Fitzpatrick Nuclear Power Plant [[::JAF-ISt-0002|JAF-ISt-0002]] APPENDIX B - Program Summary Tables Revi.rori- o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Jmuery 6, m8 Exam item Description Exam System Line or ISO No. No, INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 15' 2* 3" Req Remarks / Comments C7.30 VT-2 14 CS 16"-W23-152-1 A 3021 1 3 i 1 1- RR-10 16"-W23-152-1B 3021 1 3 1 1 1 RR-10 16"-W23-152-3A 3021 1 3 1 1 1 RR-10 16"-W23-152-38 3021 1 3 1 1 1 RR-10 12"-W23-152-2A 3021 1 3 1 1 1 RR-10 12"-W23-152-2B 3021 1 3 1 1 1 RR-10 12"-W23-302-4A 3022 1 3 1 1 1 RR-10 12"-W23-302-4 B 3023 1 3 1 1 1 RR-10 10"-W23-302-8A 3022 1 3 1 1 1 RR-10 10"-W23-302-8B 3023 1 3 1 1 1 RR-10 10"-W23-163-98 3023 1 3 1 1 1 RR-10 VT-2 23 HPCI 24 -St.P-152-26 3027 1 3 1 1 1 RR-10 20"-Str-152-25 3027 1 3 1 1 1 RR-10 16"-WCP-152-1 3028 1 3 1 1 1 RR-10 16"-W25-152-7 3028 1 3 1 1 1 RR-to 14*-W25-902-3 3025 1 3 1 1 1 RR-10 10"-W25-902-4 3026 1 3 1 1 1 RR-10 10"-SHP-902-19 M24 1 3 4 1 1 RR-10 10"-SHP-902-34 3024 1 3 1 1 1 RR-10 C7.30 VT-2 27 CP 30" N-152-13A 3030 1 3 1 1 1 RR-10 30 -N-152-13B 3030 1 3 1 1 1 RR-10 30 -N-152-13C 3030 1 3 1 1 1 RR-10 r 30"-N-152-13D 3030 1 3 1 1 1 RR-10 30"-N-152-13E 3030 1 3 1 1 1 RR-10 24*-N-152A-8 3029 1 3 1 1 1 RR-10 24"-N-152A-18A 3029 1 3 1 1 1 RR-10 24"-N-152A-18B 3029 1 3 1 1 1 RR-10 20"-N-152A-7 3029 1 3 1 1 1? RR-10 20 -N-152A-16 3030 1 3 1 1 1 RR-10 20"-N-152A-20 3029 1 3 1 1 1 RR-10 18"-N-152A4 3029 1 3 1 1 1 RR-10 18"-N-152A-26A 3029 1 3 1 1 1 RR-10 18*-N-152A-26B 3029 1 3 1 1 1 RR-10 14"-N-15A-9 3029 1 3 1 1 1 RR-10 VT-2 29 MS 24"-SHP-902-101 A 3031 1 3 1 1 1 RR-10 24"-SHP-902-101 B 3031 1 3 1 1 1 RR-10 24"-SHP-902-101C 3032 1 3 1 1 1 e '-10 24"-SHP-902-101D 3032 1 3 1 1 1 RR-10 Fi!e: JAPP-BE1.WPD Appendix B - 21 of B - 29
/n o pb O b ggh James A. Fitzpatrick Nuclear Power Plant jar-tsi-0002 APPENDIX B - Program Summary Tables Revision: o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January s.199s Exam item Description Exam System Line or ISO No. No. IMSPECTION PERIODS Ret item Method Component ID No. Items Sch'd f s' 2* 3'* Req Remarks / Comments Total Examination item 95 285 95 5 95 C7.40 Pressure Retaining Components VT-2 Pressure retaining boundary Each inspection interval, IWC-5222 Test VT-2 03 CRD 10"-WR-901 3039/40 1 1 0 0 1 1R-3/10 VT-2 03 CRD 8"-WR-901 3039/40 1 1 0 0 1 RR-3/10 VT-2 10 RHR 24"-W20-152-3A 3010 1 1 0 0 1 RR-3/10 24"-W20-152-38 3009 1 1 0 0 1 RR-3/10 24"-W20-302-11 A 3006 1 1 0 0 1 RR-3/10 24*-W20-302-17 3006 1 1 0 0 1 RR-3/10 20"-W20-152-2A 3010 1 1 0 0 1 RR-3/10 20"-W20-152-28 3009 1 1 0 0 1 RR-3/10 20"-W20-152-2C 3010 1 1 0 0 1 RR-3/10 20 -W20-152-2D 3009 1 1 0 0 1 RR-3/10 20"-W20-1524A 3010 1 1 0 0 1 RR-3/10 20"-W20-1524B 3009 1 1 0 0 1 RR-3/10 20"-W201524C 3010 1 1 0 0 1 RR-3/10 20"-W20-1524D 3009 1 1 0 0 1 RR-3/10 20"-W20-302-8A 3005 1 1 0 0 1 RR-3/10 20"-W20-302-8B 3004 1 1 0 0 1 RR-3/10 20"-W20-302-17 3006 1 1 0 0 1 RR-3/10 20"-SLP-302-33A 3014 1 1 0 0 1 RR-3/10 20"-SLP-302-33B 3015 1 1 0 0 1 RR-3/10 16*-W20-320-5A 3017 1 1 0 0 1 RR-3/10 16"-W20-152-5B 3016 1 1 0 0 1 RR-3/10 16"-W20-302-7A 3005 1 1 0 0 1 RR-3/10 16"-W20-302-78 3004 1 1 0 0 1 RR-3/10 16"-W20-302-7C 3005 1 1 0 0 1 RR-3/10 16"-W20-302-7D 3004 1 1 0 0 1 RR-3/10 16"-W20-302-9A 3005 1 1 0 0 1' RR-3/10 16"-W20-302-9B 3004 1 1 0 0 1 RR-3/10 16"-W20-302-15A 3007 1 1 0 0 1 RR-3/10 16"-W20-302-15B 3008 1 1 0 0 1 RR-3/10 16"-W20-302-34 3014 1 1 0 0 1 RR-3/10 16*-W20-302-36B 3004 1 1 0 0 1 RR-3/10 16*-WS-302-56A 3005 1 1 0 0 1 RR-3/10 14*-10-W23-152-9A 3017 1 1 0 0 1 RR-3/10 14*-W23-152-98 3016 1 1 0 0 1 RR-3/10 12"-W20-302-13A 3007 1 1 0 0 1 RR-3/10 File: JAPP-BE1.WPD Appendix B - 22 of B - 29
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L - V V e grower James A. Fitzpatrick Nuclear Power Plant APPENDIX B - Program Summary Tables [[::JAF-4SI-0002|JAF-4SI-0002]] Revi. ion: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS a nuny s.1ese Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd is' 2" 3* Req Remarks! Comments C7.40 VT-2 10 RHR 12"-W20-302-13B 3008 1 1 0 0 1 RR-3/10 10"-W20-320-12A 3007 1 1 0 0 1 RR-3/10 10"-W20-302-12B 3008 1 1 0 0 1 RR-3/10 8"-SHP-902-32A 3014 1 1 0 0 1 RR-3/10 8"-SHP-902-32B 3015 1 1 0 0 1 RR-3/10 8"-SLP-302-37B 3015 1 1 0 0 1 RR 3/10 8"-W20-302-38 3006 1 1 0 0 1 RR-3/10 8"-W20-152-59 3009 1 1 0 0 1 RR-3/10 t -W20-302-16A 3017 1 1 0 0 1 RR-3/10 s '-W20-302-168 3016 1 1 0 0 1 RR-3/10 6*-SLP-302-34A 3014 1 1 0 0 1 RR-3/10 6"-SLP-302-34B 3015 1 1 0 0 1 RR-3/1G 6"-W20-152-44A 3017 1 1 0 0 1 RR-3/10 6"-W20-152-44B 3016 1 1 0 0 1 RR-3/10 4 -W20-302-19A 3017 1 1 0 0 1 RR-3/10 4"-W20-152-408 3016 1 1 0 0 1 RR-3/10 4"-SLP-302-86A 3014 1 1 0 0 1 RR-3/10 4*-SLP-302-86B 3015 1 1 0 0 1 RR-3/10 VT-2 13 RCIC 8"-SLP-152-22 3019 1 1 0 0 1 RR-3/10 6"W22-152-16 3020 1 1 0 0 1 RR-3/10 i VT-2 14 CS 16"-W23-152-1 A 3021 1 1 0 0 1 RR-3/10 i 16"-W23-152-1B 3021 1 1 0 0 1 RR-3/10 l 16"-W23-152-3A 3021 1 1 0 0 1 RR-3/10 i 16"-W23-152-3B 3021 1 1 0 0 1 RR-3/10 l 12"-W23-152-2A 3021 1 1 0 0 1 RR-3/10
! 12"-W23-152-2B 3021 1 1 0 0 1 RR-3/10 12"-W23-302-4A 3022 1 1 0 0 1 RR-3/10 l 12"-W23-302-4B 3023 1 1 0 0 1F RR-3/10 10"-W23-302-8A 3022 1 1 0 0 1 RR-3/10 10"-W23-302-88 3023 1 1 0 0 1 RR-3/10 10"-W23-163-98 3023 1 1 0 0 1 RR-3/10 VT-2 23 HPCI 24"-SLP-152-26 3027 1 1 0 0 1 RR-3/10 20"-SLP-152-25 3027 1 1 0 0 1 RR-3/10 16"-WCP-152-1 3028 1 1 0 0 1 RR-3/10 16"-W25-152-7 3028 1 1 0 0 1 RR-3/10 14"-W25-902-3 3025 1 1 0 0 1 RR-3/10 10"-W25-902-4 3026 1 1 0 0 1 RR-3/10 10"-SHP-902-19 3024 1 1 0 0 1 RR-3/10 10"-SHP-902-34 3024 1 1 0 0 1 RR-3/10 l
Fra: JAPP-BE1.WPD Appendix B - 23 of B - 29 l
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g grower James A. Fitzpatrick Nuclear Power Plant APPENDIX B - Program Summary Tables jar-asi-0002 Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Jaraary s.1eos Exam item Description Exam System Line or ISO No- No. INSPECTION PERIODS Rel item Method Component ID No. Items Sth'd 1s' 2* 3* Req Remarks / Comments VT 27 CP 30*-N-152-13A 3030 1 1 0 0 1 RR-3/10 30"-N-152-13B 3030 1 1 0 0 1 RR "J10 30*-N-152-13C 3030 1 1 0 0 1 RR-3/10 30"-N-152-13D 3030 1 1 0 0 1 RR-3/10 30"-N-152-13E 3030 1 1 0 0 1 RR-3/10 24"-N-152A-8 3029 1 1 0 0 1 RR-3/10 24"-N-152A-1 BA 3029 1 1 0 0 1 RR-3/10 24"-N-152A-188 3029 1 1 0 0 1 RR-3/10 20"-N-152A-7 3029 1 1 0 0 1 RR-3/10 20"-N-152A-16 3030 1 1 0 0 1 RR 3/10 20"-N-152A-20 3029 1 1 0 0 1 RR-3/10 18*-N-152A-6 3029 1 1 0 0 1 RR-3/10 18"-N-152A-26A 3029 1 1 0 0 1 RR-3/10 18"-N-152A-26B 3029 1 1 0 0 1 RR-3/10 14"-N-15A-9 3029 1 1 0 0 1 RR-3/10 VT 29 MS 24"-SHP-902-101 A 3031 1 1 0 0 1 RR-3/10 24*-SHP-902-101B 3031 1 1 0 0 1 RR-3/10 24"-SHP-902-101C 3032 1 1 0 0 1 RR-3/10 24*-SHP-902-101D 3032 1 1 0 0 1 RR-3/10 Total Examination item 95 95 0 0 95 Pumps C7.50 Pressure Retaining Components VT-2 Pressure retaineg boundary, each inspectron Period, IWC-5221 Test VT-2 14 CS 14P-1A 3021 1 3 1 1 1 RR-10 14P-1B 3023 1 3 1 1 1 RR-10 VT-2 23 HPCI 23P-1A 3025 1 3 1 1 1' RR-10 23P-1B 3028 1 3 1 1 1 RR-10 Total Examination item 4 12 4 4 4 C7.60 Pressure Retaining Components VT-2 Pressure retaining boundary, each inspection Interval, IWC-5222 Test VT-2 14 CS 14P-1A 3021 1 1 0 0 1 RR-3/10 14P-1B 3023 1 1 0 0 1 RR-3/10 VT-2 23 HPCI 23P-1A 3025 1 1 0 0 1 RR-3/10 23P-1B 3028 1 1 0 0 1 RR-3/10 Total Examination item 4 4 0 0 4 File: JAPP-BE1.WPD Appendix B -24 of B -29
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APPENDIX B - Program Summary Tables R e ion: o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS a.nu.ry s. im Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret 5 Req Rem rks/ Comments Item Method Component. ID No. Items Sch'd 1' 2* 3*
Vanes C7.70 Pressure Retaining Components VT-2 Pressure retaining boundary. Each inspection Period, nW 5221 Test Vr-2 to RHR RHR-45B 3004 1 3 1 1 1 RR-10 RHR-42D 3004 1 3 1 1 1 RR-10 MOV-658 3004 1 3 1 1 1 RR-10 MOV-668 3004 1 3 1 1 1 RR-10 MOV-128 3004 1 3 1 1 1 RR-10 RHR-1778 3004 1 3 1 1 1 RR-10 '.
MOV-1498 3004 1 3 1 1 1 RR-10 RHR-42C 3005 1 3 1 1 1 RR-10 RHR-49C 3005 1 3 1 1 1 RR-10 RHR-42A 3005 1 3 1 1 1 RR-10 MOV-66A 2'Y15 1 3 1 1 1 RR-10 MOV-12A 3005 1 3 1 1 1 RR-10 RHR-177A 3005 1 3 1 1 1 RR-10 MOV-20 3006 1 3 1 1 1 RR-10 RHR-9 3006 1 3 1 1 1 RR-10 FPC-51 3006 1 3 1 1 1 RR-10 MOV-39A 3007 1 3 1 1 1 RR-10 MOV-26A 3007 1 3 1 1 1 RR-10 MOV '11 A 3007 1 3 1 1 1 RR-10 MOV-31B 3008 1 3 1 1 1 RR-10 MOV-268 3008 1 3 1 1 1 RR-10 MOV-398 3008 1 3 1 1 1 RR-10 MOV 158 3009 1 3 1 1 1 RR-10 MOV-151B 3003 1 3 1 1 1 RR-10 MOV-1383 3009 1 3 1 1 1 RR-10 C7.70 VT-2 10 RHR MOV-13D 3009 1 3 1 1 1 RR-1; MOV-15D 3009 1 3 1 1 1' RR-10 FPC-34 3009 1 3 1 1 1 RR-10 MOV-15C 3010 1 3 1 1 1 RR-10 MOV-13C 3010 1 3 1 1 1 RR-10 MOV-15A 3010 1 3 1 1 1 RR-10 MOV-13A 3010 1 3 1 1 1 RR-10 MOV-151A 3010 1 3 1 1 1 RR-10 MOV-17 3011 1 3 1 1 1 RR-10 MOV-65A 3014 1 3 1 1 1 RR-10 MOV-70A 3014 1 3 1 1 1 RR-10 POV-69A 3014 1 3 1 1 1 RR-10 SV-74A 3014 1 3 1 1 1 RR-10 MOV-708 3015 1 3 1 1 1 RR-10
- File: JAPP-BE1.WPD Appendix B-25 of B -29
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g g esauer Jar es A. Fitzpatrick Nuclear Power Plant [[::JAF-tSI-0002|JAF-tSI-0002]] APPENDIX B - Program Summary Tables Revision: o ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Janu=7 s.1,os Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No, items Sch'd 1s' 2'* 3'" Reg Remarks / Comments PCV-699 3015 1 3 1 1 1 RR-10 SV-748 3015 1 3 1 1 1 RR-10 MOV-39B 3016 1 3 1 1 1 RR-10 MOV-34B 3016 1 3 1 1 1 RR-10 MOV-38B 3016 1 3 1 1 1 RR-10 MOV-39A 3017 1 3 1 1 1 RR-10 MOV-38A 3017 1 3 1 1 1 RR-10 MOV-34A 3017 1 3 1 1 1 RR-10 Total RHR Valves 47 141 48 48 48 C7.70 VT-2 13 RCIC RCI-5 3019 1 3 1 1 1 RR-10 RCIC-3 3019 1 3 1 1 1 RR-10 RCIC-4 3019 1 3 1 1 1 RR-10 MOV-41 3020 1 3 1 1 1 RR-10 Total RCIC Valves 4 12 4 4 4 C7.70 VT-2 14 CS MOV-7A 3021 1 3 1 1 1 RR-10 MOV-7B 3021 1 3 1 1 1 RR-10 CSP-8A 3021 1 3 1 1 1 RR-10 CSP-8B 3021 1 3 1 1 1 RR-10 CSP-10A 302" 1 3 1 1 1 RR-10 CSP-10B 30".3 1 3 1 1 1 RR-10 MOV-26A 3022 1 3 1 1 1 RR-10 MOV-26B 3023 1 3 1 1 1 RR-10 Total CS Valves 8 24 8 8 8 f
VT-2 23 HPCI MOV-14 3024 1 3 1 1 1 RR-10 MOV-19 3025 1 3 1 1 1 RR-10 MCV-20 3025 1 3 1 1 1 RR-10 MOV-21 3026 1 3 1 1 1 RR-10 HPI-65 3027 1 3 1 1 1 RR-10 HPI-12 3027 1 3 1 1 1 RR-10 HPI-11 3027 1 3 1 1 1 RR-10 MOV-98 3028 1 3 1 1 1 RR-10 HPI-61 3028 1 3 1 1 1 RR-10 MOV-57 3028 1 3 1 1 1 RR-10 MOV-17 3028 1 3 1 1 1 RR-10 HP1-32 3028 1 3 1 1 1 RR-10 File: JAPP-BE1.WPD Appendix B - 26 of B - 29
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' g gPtmer James A. Fitzpatrick Nuclear Power Plant . jar-ass-Ooo2 APPENDIX B - Program Summary Tables Revisiorr. O ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Jano.ry s.199s Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. lO No. Items Sch'd 1 57 2* 3* Req Remarks / Comments Total HPCI Valves 12 36 12 12 12 C7.80 Pressure Retaining Components VT-2 Pressure retaining boundary, Each inspeKien Intervat. IWC-5222 test VT-2 10 RHR RHR45B 3004 1 1 0 0 1 RR-3/10 RHR42D 3004 1 1 0 0 1 RR-3/10 MOV-658 3004 1 1 0 0 1 RR-3/10 MOV-66B 3004 1 1 0 0 1 RR-3/10 MOV-128 3004 1 1 0 0 1 RR-3/10 RHR-177B 3004 1 1 0 0 1 RR-3/10 MOV-149B 3004 1 1 0 0 1 RR-3/10 RHR42C 3005 1 1 0 0 1 RR-3/10 RHR-49C 3005 1 1 0 0 1 RR-3/10 RHR42A 3005 1 1 0 0 1 RR-3/10 MOV-6GA 3005 1 1 0 0 1 RR-3/10 MOV-12A 3005 1 1 0 0 1 RR-3/10 C7.80 VT-2 10 RHR RHR-177A 3005 1 1 0 0 1 RR-3/10 MOV-20 3006 1 1 0 0 1 RR-3/10 RHR-9 3006 1 1 0 0 1 RR-3/?O FPC-51 3006 1 1 0 0 1 RR-3/10 MOV-39A 3007 1 1 0 0 1 RR-3/10 MOV-26A 3007 1 1 0 0 1 RR-3/10 MOV-31A 3007 1 1 0 0 1 RR-3/10 MOV-318 3008 1 1 0 0 1 RR-3/10 MOV-268 3008 1 1 0 0 1 RR-3/10 MOV-39B 3008 1 1 0 0 1 RR-3/10 MOV-15B 3009 1 1 0 0 1 RR-3/10 MOV-151B 3009 1 1 0 0 1 RR-3/10 MOV-138 3009 1 1 0 0 1 RR-3/10 MOV-13D 3009 1 1 0 0 1t RR-3/10 MOV-15D 3009 1 1 0 0 1 RR-3/10 FPC-34 3009 1 1 0 0 1 RR-3/10 MOV-15C 3010 1 1 0 0 1 RR-3/10 MOV-13C 3010 1 1 0 0 1 RR-3/10 MOV-15A 3010 1 1 0 0 1 RR-3/10 MOV-13A 3010 1 1 0 0 1 RR-3/10 MOV-151A 3010 1 1 0 0 1 RR-3/10 MOV-17 3011 1 1 0 0 1 RR-3/10 MOV-65A 3014 1 1 0 0 1 RR-3/10 MOV-70A 3014 1 1 0 0 1 RR-3/10 PCV-69A 3014 1 1 0 0 1 RR-3/10 SV-74A 3014 1 1 0 0 1 RR-3/10 File: JAPP-BE1 WPD Appendix B - 27 of 8 - 29
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\ v' s M gPower James A. FitzpatriCk Nuclear Power Plant jar-ass-0002 APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS Janoary s.19,s Exam wem Description Exam System Une or ISO No. No. INSPECTION PERIODS Rel item Method Component ID No- Items Sch'd 5 1' 2* 3" Req Remarks / Comments MOV-708 3015 1 1 0 0 1 RR-3/10 PCV-69B 3015 1 1 0 0 1 RR-3/10 SV-748 3015 1 1 0 0 1 RR-3/10 MOV-398 3016 1 1 0 0 1 RR-3/10 MOV-348 3016 1 1 0 0 1 RR-3/10 MOV-388 3016 1 1 0 0 1 RR-3/10 MOV-39A 3017 1 1 0 0 1 RR-3/10 MOV-38A 3017 1 1 0 0 1 RR-3/10 f.40V-34A 3017 1 1 0 0 1 RR-3/10 Total RHR Valves 46 46 0 0 46 C7.80 VT-2 13 RC8C RC1-5 3019 1 1 0 0 1 RR-3/10 RCIC- 3019 1 1 0 0 1 RR-3/10 RCIC-4 3019 1 1 0 0 1 RR-3/10 MOV-41 3020 1 1 0 0 1 RR-3/10 Total RCIC Valves 4 4 0 0 4 VT-2 14 CS MOV-7A 3021 1 1 0 0 1 RR-3/10 MOV-78 3021 1 1 0 0 1 RR-3/10 CSP-8A 3021 1 1 0 0 1 RR-3/10 CSP-8B 3021 1 1 0 0 1 RR-3/10 CSP-10A 3022 1 1 0 0 1 RR-3/10 CSP-108 3023 1 1 0 0 1 RR-3/10 MOV-2SA 3022 1 1 0 0 1 RR-3/10 MOV-2SB 3023 1 1 0 0 1 RR-3/10 I
Total C3 Valves 8 8 0 0 8' VT-2 23 HPCI MOV-14 3024 1 1 0 0 1 RR-3/10 MOV-19 3025 1 1 0 0 1 RR-3/10 MOV-20 3025 1 1 0 0 1 RR-3/10 MOV-21 3026 1 1 0 0 1 RR-3/10 HPl-65 3027 1 1 0 0 1 RR-3/10 HPI-12 3027 1 1 0 0 1 RR-3/10 HPI-11 3027 1 1 0 0 1 RR-3/10 MOV-98 3028 1 1 0 0 1 RR-3/10 14P8-61 3028 1 1 0 0 1 RR-3/10 MOV-57 3028 1 1 0 0 1 RR-3/10 tIOV-17 3028 1 1 0 0 1 RR-3/10 File: JAPP-BE1.WPD Appendix B -28 of B -29
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h 4 g genser James A. Fitzpatrick Nuclear Power Plant JAF 4SI-0002 APPENDIX B - Program Summary Tables Revision: 0 ASME CODE CLASS 2 SYSTEMS AND COMPONENTS January 6.1998 Ex8m > ltem Der- iption Exam System Line or ISO No. No. INSPECTION PEI-CDS Ret item Method Component. lD No. Items Sch'd 1st 2* 3"D Req Remarks / Comments HPI-32 3028 1 1 0 0 1 RR-3/10 Total HPCI Valves 12 12 0 0 12 Total Examination Category 310 628 157 157 314 Notes: .
(1) Other than open end',4 ponions of systems.
(2) System Pressure tests c!IWA-5000 and IWC-5000.
(3) Visual examination of IWA-5240.
(4) No components within the pressure retaining boundary [as defined by Note (7) are exempt or excluded from the examination requirements, except as speofed in fWA-5214(c) for repairs and replacements.
(5) The system hydrostatic test (fWC-5222) shaf! be conducted at or near the end of each inspection interval or during the same inspection of each inspection interval of Inspection Program B.
(6) Where portions of a system are subject to system pressure tests associated with two different system functions, the VT-2 examination need only be peikin.cd dunng the test conducted at the higher of the test pressures of the tespective system function.
(7) The pressure retaining boundary includes only those portions of the system required to operate or support the safety system function up to and inch.*dmg the frst normaWy closed valve (including a safety or refief valve) or valve capaDie of automatic closure when the safety function is required.
(8) A system hydrostatic test (tWC-5222) and accompa tying VT-2 examination are acceptable in lieu of the system pressure test (fWC-5221) and VT-2 examination.
r File: JAPP-BE1.WPD Appendix B-29 of 8 -29
, ,m a' neww kPener James A. Fitzpatrick Nuclear Power Plant jar. S!-0002 e Authertly APPENDIX C - Program Summary Tables Revision: 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rei Item Method Component. ID No. Items Sch'd 1" 2" . 3*8 Req Remarks / Comments Examination Category: D-A, SYSTEMS IN SUPPORT OF REACTOR SHtJTDOWN FtJNCTION D1.10 Pressure Retaining Cornponents VT-2 Pressure Retaining Boundary Each Inspecbon Period VT-2 13 RCIC 6*-WCP-152-2 102A1 1 3 1 1 1 RR-10 Note 1,4 and 5 6"-WCD-152-2 1 3 1 1 1 RR-10 6*-W22-152-16 1 3 1 1 1 RR-10 8"-SLP-152-22 102B1 1 3 1 1 1 RR-10 8 -SLP-152-22 167C1 1 3 1 1 1 RR-10 Total RCIC 5 15 5 5 5 VT-2 Pressure Retaining Boundary Each inspecten intenral VT-2 13 RCIC 6"-WCP-152-2 102A1 1 1 0 0 1 RR-3/10 Note 1.4 and 5 6"-WCD-152-2 1 1 0 0 1 RR-3/10 6*-W22-152-16 1 1 0 0 1 RR-3/10 8"-SLP-152-22 102B1 1 1 0 0 1 RR-3/10 8"-SLP-152-22 167C1 1 1 0 0 1 RR-3/10 Total RCIC 5 5 0 0 $
D1.20 Integral Attachment-Component Supports and Restraints VT-3 Integral Attachment Note 3 VT-3 13 RCtC 6"-WCP-152-2 102A1 1 1 0 1 0 RR-4 CC N-509.10%
6"-WCD-152-2 1 0 0 0 0 RR-4 CC N-509 6"-W22-152-16 5 0 0 0 0 RR-4 CC N-509 8"-SLP-152-22 167C1 2 0 0 0 0 RR-4 CC N-509 Total RCIC 9 1 0 1 F0 0% 100 % 100 %
D1.30 Integral Attachment-Mechanical and Hydraulic Snubbers VT-3 Integral Attachment 0 0 0 0 0 Note 3 and 6. JAF Technical Specifcations D140 Intigtal Attachment-Spring Type Support VT-3 Integral Attachment 0 0 0 0 0 Note 3. Not apphcable to JAF '
D1.50 Integral Attachment-Constant Load Type Support VT-3 integral Attach mmt 0 0 0 0 0 Note 3. Not appfcable to JAF File: APP-CE1.WPD Appendix C - 1 of C - 13 L
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g gRmer James A. Fitzpatrick Nuclear Power Plant APPENDIX C - Program Summary Tables [[::JAF-tSI-0002|JAF-tSI-0002]] nevision: 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System tJne or ido No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 15T 2* 3" Req Remarks / Comments D1.60 Integral Attachment-Shock Absorbers VT-3 Integral Attachment 0 0 0 0 0 Note 3 and 6.JAF Techmcal Specifications Total Examination Category 9 1 0 1 0 Does not include D1.10 G% 100 % 100 %
Notes:
(1) The system boundary extends up to and including the first norma!!y closed valve or valve capable of automaSc closure as required to perform the safety-related system function.
(2) The system hydrostatic test sha!! be conducted at or near the end of each inspection interval or dunng the same inspection period of each inspection interval for inspection Program B.
(3) In the case of multiple components within a system of similar design, function, and service, the ir:4egraf attachment of only one of the multiple w..ve ieas shall be examined. The integral attachment selected foi examination shall correspond to those cei.vv e-4 supports selected by IWF-2510(b).
(4) There are no exemptions or exclusions from these requirements except as specifed in IWA-5214(c).
(5) A system hydrostatic test (IWD-5223) and accompanying VT-2 examination are acceptable in lieu of the system pressure test (IWD-5221) and VT-2 examination.
(6) Snubbers are examined and tested in accordance with JAF Plant Technical Speerfecations Section 3.6.1.
7 File: APP-CE1.WPD Appendix C - 2 of C - 13
D G wr g gPmer James A. Fitzpatrick Nuclear Power Plant APPENDIX C - Program Summary Tables aAr- Ss 0002 Revision: 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System t ine or ISO No. No. INSPECTION PERIODS Rel item Method Component. ID No. Items Sch'd is' 2* , 3* Req Remarks / Comments ..
Examination Category: D-8, SYSTEMS IN SUPPORT OF EMERGENCY CORE COOLING, CONTAINMENT HEAT REMOVAL, ATMOSPHERE CLEANUP, AND REACTOR RESIDUAL HEAT REMOVAL .
D210 Pressure Retaining Components VT-2 Pressure Retaining Boundary Each inspection Period Notes 1,4 and 5 VT-2 10 RHR 16*-WS-151-298 166A1 1 3 1 1 1 RR-10 16*-WS-151-298 166B1 1 3 1 1 1 RR-10 16"-WS-151-MA 166C1 1 3 1 1 1 RR-10 12*-WS-151 7D 137A1 1 3 1 1 1 RR-10 12"-WS-151-27B 1 3 1 1 1 RR-10 16"-WS-151-28D 1 3 1 1 1 RR-10 16"-WS-151-288 1 3 1 1 1 RR-10 16*-WS *51-29B 13781 1 3 1 1 1 RR-10 16"-WS-151-29B 137C1 1 3 1 1 1 RR-10 16*-WS-151-568 1 3 1 1 1 RR-10 12* 'NS-151-27A 137D1 1 3 1 1 1 RR-10 12"-WS-151-27C 1 3 1 1 1 RR-10 16"-WS-151-28A 1 3 1 1 1 RR-10 16*-WS-151-28C 1 3 1 1 1 RR-10 16*-WS-151-29A 1 3 1 1 1 RR-10 16*-WS-151-29A 137E1 1 3 1 1 1 RR-10 16*-WS-151-29A 137F1 1 3 1 1 1 RR-10 16"-WS-151-56A 1 3 1 1 1 RR-10 16"-WS-151-29A 1 3 1 1 1 RR-10 16*-WS-151-30A 137H1 1 3 1 1 1 RR-10 16*-WS-151-30B 1 3 1 1 1 RR-10 ToutRHR 21 63 21 21 21
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VT-2 15 RBCLC 8"-WES-151-100 103A1 1 3 1 1 1 RR-10 8"-WES-151-99 10381 1 3 1 1 1 RR-10 6"-WES-151-104 1 3 1 1 1 RR-10 6"-WCL-151-103 103C1 1 3 1 1 1 RR-10 6*-WCL-151-29 1 3 1 1 1 RR-10 6"-WCL-151-61 103C2 1 3 1 1 1 RR-10 8"-WCL *51-29 1 3 1 1 1 RR-10 6"-WES-151-103 1 3 1 1 1 RR-10 6*-WES-151 104 103C3 1 3 1 1 1 RR-10 6"-WCL-151-29A 103F 1 3 1 1 1 RR-10 6"-WES-151-103 1 3 1 1 1 RR-10 File: APP-CE1.WPD Appendix C - 3 of C - 13
7 V V V g grower James A. Fitzpatrick Nuclear Power Plant APPENDIX C - Program Summary Tables
.lAF4S14002 R vi. ion: o ASME CODE CLASS 3 SYSTEMS AND COMPONENT S January 6,1998 Exam item Description - Exam System Line or ISO No- No. INSPECTION PERIODS Ret Item Method Component. ID No- Items Sch'd is' 2* . 3'* Req Remarks / Comments 6"-VES-151-104 103P 1 3 1 1 1 RR-10 8*-WES-151-99 103C8 1 3 1 1 1 RR-10 8"-VES-151-100 1 3 1 1 1 RR-10 6"-WCL-151-61 1 3 1 1 1 RR-10 6"-WCL-151-24 1 3 1 1 1 RR-10 6*-WCL-151-96 1 3 1 1 1 RR-10 6"-WES-151-104 1 3 1 1 1 RR-10 6*-WES-151-103 1 3 1 1 1 RR-10 6"-WCL-151 72 1 3 1 1 1 RR 10 6"-WCL-151-61 1 3 1 1 1 RR-10 Total RBCLC 21 63 21 21 21 VT-2 46 SWS 12"-WES-151-2A 137K1 1 3 1 1 1 RR-10 10"-WES-151-37 1 3 1 1 1 RR-10 12"-VES-151-2B 1 3 1 1 1 RR-10 10"-WES-151-38 1 3 1 1 1 RR-10 12"-WES-151-9 137N1 1 3 1 1 1 RR-10 8"-VES-151-10 1 3 1 1 1 ER-10 8"-WES-151-4B 137P1 1 3 1 1 1 RR-10 t 8"-WES-151-4A 1 3 1 1 1 RR-10 8"-WES-151-3B 1 3 1 1 1 FsR-10 8 -WES-151-3A 1 3 1 1 1 RR-10 8"-WES-151-4A 137Q1 1 3 1 1 1 RR-10 6*-WES-151-5A 1 3 1 1 1 RR-10 6*-WES-151-5B 1 3 1 1 1 RR-10 8"-WES-151-48 1 3 1 1 1 RR-10 6*-WES-151-6A 1 't 1 1 1 RR-10 6*-WES-151-6B 1 3 1 1 7 1 RR-10 10"-WES-151-38 166Et 1 3 1 1 1 RR-10 ,
8"-WES-151-100 1 3 1 1 1 RR-10 10"-WES-151-37 166F1 1 3 1 1 1 RR-10 8"-WES-151-99 1 3 1 1 1 RR-10 Total SWS 20 60 20 20 20 VT-2 66 RBV&C 6"-WES-151-69 1359 1 3 1 1 1 RR-10 6"-WES-151-69 1361 1 3 1 1 1 RR-10 6"-WS-151-2 1363 1 3 1 1 1 RR-10 File: APP-CE1.WPD Appendix C - 4 of C - 13
- 1> NewyerkPsever James A. Fitzpatrick Nuclear Power Plant [[::JAF-ISI-0002|JAF-ISI-0002]] 4# Autherty APPENDIX C - Program Summary Tables Revenion- 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System IIne or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No- Items Sch'd 1ST 2* . 3* Req Remarks / Comments Total RBV&C 3 9 3 3 3 FS 70 CRVSC 6"-WS-151-2 1474/75 1 3 1 1 1 RR-10 6"-WS-151-35 1475 1 3 1 1 1 RR-10 6" WS-151-1 1 3 1 1 1 RR-10 6"-WS-151-1 1476 1 3 1 1 1 RR-10 6"-WES-151-41 1456 1 3 1 1 1 RR-10 6*-WES-151-40 1 3 1 1 1 RR-10 6"-WES-151-41 1457 1 3 1 1 1 RR-10 6* WES-151-40 1 3 1 1 1 RR-10 6"-WES-151-41 1458 1 3 1 1 1 RR-10 6*-WES-151-40 1 3 1 1 1 RR-10 6"-WES-15141 1492 1 3 1 1 1 RR-10 6"-WES-151-40 1 3 1 1 1 RR-10 Total CRV&C 12 36 12 12 12 VT-2 Pressure Retaining Boundary Each inspection Interval Notes 1,4 and 5 VT-2 10 RHR 16"-WS-151 29B 166A1 1 1 0 0 1 RR-3/10 16"-WS-151-299 166B1 1 1 0 0 1 RR-3/10 16"-WS-151-29A 166C1 1 1 0 0 1 RR-3/10 12"-WS-151-27D 137A1 1 1 0 0 1 RR-3/10 12"-WS-151-27B 1 1 0 0 1 RR-3/10 16"-WS-151-28D 1 1 0 0 1 RR-3/10 16"-WS-151-288 1 1 0 0 1 RR-3/10 16"-WS-151-298 137:31 1 1 0 0 1 RR-3/10 16"-WS-151-29B 137C1 1 1 0 0 1 RR-3/10 16"-WS-151568 ! 1 0 0 1 RR-3/10 12"-WS-151-27A 137D1 1 1 0 0 t 1 RR-3/10 12"-WS-151-27C 1 1 0 0 1 RR-3/10 16"-W3-151-28A 1 1 0 0 1 RR-3/10 16*-WS-151-28C 1 1 0 0 1 RR-3/10 16*-WS-151-29A 1 1 0 0 1 RR-3/10 16"-WS-151-29A 137E1 1 1 0 0 1 RR-3/10 16*-WS-151-29A 137F1 1 1 0 0 1 RR-3110 16"-WS-151-56A 1 1 0 0 1 RR-3/10 16*-WS-151-29A 1 1 0 0 1 RR-3/10 16"-WS-151-30A 137H1 1 1 0 0 1 RR-3/10 16"-WS-151-308 1 1 0 0 1 RR-3/10 Total RHR 21 21 0 0 21 File: APP-CE1.WPD Qoendix C - 5 of C - 13
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- > gPower James A. Fitzpatrick Nuclear Power Plant wes-0002 APPENDIX C - Program Sunimary Tables R i.iore o ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January s,19es Exam item Descr ption Exam System Line or ISO Mo- No. INSPECTION PERIODS Ret item Method Component. ID No- Items Sch'd 15T 2* . 3* Req Remarks /Cosments VT-2 15 RBCLC 8*-WES-151-100 103A1 1 1 0 0 1 RR-3/10 8*-WES-151-99 10381 1 1 0 0 1 RR-3/10 6*-WES-151-104 1 1 0 0 1 RRar10 6*-WCL-151-103 103C1 1 1 0 0 1 RR-3/10 6 -WCL-151-29 1 1 0 0 1 RR-3/10 6*-WCL-151-61 103C2 1 1 0 0 1 RR-3/10 8" WCL-151-29 1 1 0 0 1 RR-3/10 6*-VES-151-103 1 1 0 0 1 RR-3/10 6*-WES-151-104 103C3 1 1 0 0 1 RR-3/10 6*-WCL-151-29A 103F 1 1 0 0 1 RR-3/10 6*-WES-151-103 1 1 0 0 1 RR-3/10 6*-VES-151-104 103P 1 1 0 0 1 RR-3/10 8*-WES-151-99 103C8 1 1 0 0 1 RR-3/10 8*-WES-151-100 1 1 0 0 1 RR-3/10 6*-WCL-151-61 1 1 0 0 1 RR-3/10 6*-WCL-151-24 1 1 0 0 1 RR-3/10 6 -WCL-151-96 1 1 0 0 1 RR-3/10 6*-WES-151-104 1 1 0 0 1 RR-3/10 6*-WES-151-103 1 1 0 0 1 RR-3/10 6* WC:.-151-72 1 1 0 0 1 RR-3/10 6*-WCL-151-61 1 1 0 0 1 RR-3/10 Total RBCLC 21 21 0 0 21 VT-2 46 SWS 12*-WES-1512A 137K1 1 1 0 0 1 RR-3/10 10 -WES-151-37 1 1 0 0 1 RR-3/10 12 -VES-151-2B 1 1 0 0 1 RR-3/10 10*-VES-15138 1 1 0 0 1 RR-3/10 12 -WES-151-9 137N1 1 1 0 C t 1 RR-3/10 8*-WES-151-10 1 1 0 0 1 RR-3/10 8*-WES-151-4B 137P1 1 1 0 0 1 RR-3/10 8*-WES-151-4A 1 1 0 0 1 RR-3/10 8*-WES-151-38 1 1 0 0 1 RR-3/10 8 -WES-151-3A 1 1 0 0 1 RR-3/10 8*-VES-151-4A 13701 1 1 0 0 1 RR-3/10 6*-WES-151-5A 1 1 0 0 1 RR-3/10 6*-VES-151-53 1 1 0 0 1 RR-3/10 8*-WES-151-4B 1 1 0 0 1 RR-3/10 6*-WES-151-6A 1 1 0 0 1 RR-3/10 6*-WES-151-6B 1 1 0 0 1 RR-3/10 File: APP-CE1.WPD Appendix C - 6 of C - 13
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g grower James A. Fitzpatrick Nuclear Power Plant APPENDIX C - Program Summary Tables JAF-IS8 0002 Revision: 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO Ho. No. INSPECTION PERIODS Ret item Method Component ID No. Items Sch'd 15' 2* , 3" Req Remarks / Comments 10"-WES-151-38 166E1 1 1 0 0 1 PR-3/10 B"-WES-151-100 1 1 0 0 1 RR-3/10 10*-WES-151-37 166F1 1 1 0 0 1 RR-3/10 B"-WES-151-99 1 1 0 0 1 RR-3/10 Total SWS 20 20 0 0 20 VT-2 66 RBV&C 6"-VES-151-69 1359 1 1 0 0 1 RR-3/10 6"-VES-151-69 1361 1 1 0 0 1 RR-3/10 6"-WS-151-2 1363 1 1 0 0 1 RR-3/10 Total RBV&C 3 3 0 0 3 VT-2 70 CRV&C 6"-WS-151-2 1474/75 1 1 0 0 1 RR-3/10 6*-WS-151-35 1A75 1 1 0 0 1 RR-3/10 6*-WS-151-1 1 1 0 0 1 RR-3/10 6"-WS-151-1 1476 1 1 0 0 1 RR-3/10 6*-WES-151-41 1456 1 1 0 0 1 RR-3/10 6"-WES-151-40 1 1 0 0 1 RR-3/10 6"-WES-151-41 1457 1 1 0 0 1 RR-3/10 6"-VES-151-40 1 1 0 0 1 RR-3/10 6"-VES-151-41 1458 1 1 0 0 1 RR-3/10 6*-WES-151-40 1 1 0 0 1 RR-3/10 6*-WES-151-41 1492 1 1 0 0 1 RR-3/10 6" WES-151-40 1 1 0 0 1 RR-3/10 Total CRVSC 12 12 0 0 12 Total Examination item 154 308 77 77 t 154 File: APP-CE1.WPD Appendix C - 7 of C - 13
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APPENDIX C - Program Summary Tables Reve . ion: O r' 7, ODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Descrip* ion Exam System Une or ISO No. No- INSPECTION PERIODS Ret item Method Cus.pc.wat ID No- Items Sch'd 15' 2* 3* Reg Remarks / Comments D120 Integraf Atta@ ment-Cor conent Supports ared Restraints VT-3 Integral Atta@rnent Note 3 VT-3 10 RHR 16 -WS-151-298 166A1 12 2 2 0 0 RR4 CC N-509 1G'-WS-151-29B ' 166B1 5 1 0 1 0 RR-4 CC N-509 1G"-WS 151-29A 166C1 2 1 0 0 1 RR-4 CC 4509 3
16*-WS-151-2SD 137At 1 0 0 0 0 RR-4 CC N-509 4
16"-WS-151-298 1 0 0 0 0 RR-4 CC 4-509 [
16*-WS-151-29B 137B1 1 0 0 0 0 RR-4 CC N-509 16"-WS-151-29B 137C1 1 0 0 0 0 RR-4 CC N-509 16"-WS-151-568 2 0 0 0 0 RR-4 CC N-509 ,
16"-WS-151-28A 137D1 1 0 0 0 0 RR-4 CC N-SO9 16" WS-151-28C 1 1 0 1 0 RR 4 CC N-509 16*-WS-151-29A 3 0 0 0 0 RR-4 CC 4509 16*-WS-151-29A 137E1 3 0 0 0 0 RR-4 CC N-509 16"-WS-151-29A 137F1 1 1 0 0 1 RR-4 CC N-509 16" WS-151-56A 1 0 0 0 0 RR-4 CC N-509 16 -WS-151-29A 2 0 J 0 0 RR-4 CC 4509 ,
16~-WS-151-30A 137H1 1 0 0 0 0 RR-4 CC N-SC'-
TotJ RHR P 6 2 2 2 33 % 66 % 100 %
VT-3 15 RBCLC 8"-WES-151-100 103A1 3 1 0 0 1 RR-4 CC 4509 S* YES-151-99 10381 2 0 0 0 '
RR-4 CC N-509 j 6"-WES-151-104 1 0 0 0 0 RR4 CC N#,,09 6*-WCL-151-29 10YA 1 0 0 0 0 RR-4 CC 4509 6'-WCL-151-61 103C2 1 0 0 0 0 RR4 CC 4509 6"-WCL -151-29 1 0 0 0 0 FR-4 CC N-509 6 -WES-151-103 8 1 1 0 t 0 RR4 CC N509 6 -VES-151-104 103C3 1 0 0 0 0 RR4 CC N-509 6*-WCL-151-29A 103F 1 0 0 0 0 RR-4 CC N-509 6" WES-151-103 1 0 0 0 0 RR4 CC 4509 6 -VES-151-104 103P 1 0 0 0 0 RR-4 CC 4509 8*-VES-151-99 103C8 1 0 0 0 0 RR4 CC N-509 8 -WES-151-103 1 0 0 0 0 RR-4 CC 4509 6~-WCL-151-61 1 0 0 0 0 RR-4 CC N-509 6*-WCL-151-24 1 0 0 0 0 RR-4 CC 4509 6 -WCL-151-96 1 0 0 0 0 RR4 CC 4509 6 -VES-151-104 5 1 0 1 0 RR-4 CC 4509 6~-WES-151-103 6 1 0 1 0 RR4 CC 4509 6~-WCL-151-72 2 0 0 0 0 RR-4 CT,4 509 1
File: APP-CE1.WFD Appendix C - 8 of C - 13
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- D gh James A. FitzpatriCk Nuclear Power Plant jar Sum 02 APPENDIX C - Program Summary Tables em o ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January s.199s Exam item Description Exam System Line or ISO No- No. INSPECTION PERIODS Ret item Method Component. lD No. Items Sch'd 1" 2* 3* Req Remarks / Comments 6 -WCL-151-61 3 1 0 0 1 RR-4 CC N-509 Total RBCLC 42 4 1 1 2 25% 50 % 100 %
VT-3 46 SWS 12 -WES-151-2A 137K1 2 1 1 0 0 RR4 CC 45C9 12"-WES-151-2B 2 0 0 0 0 RR-4 CC N-509 12'-WES-151-9 137N1 1 0 0 0 0 RR-4 CC N-SC9 8"-WES-151-10 4 1 0 1 0 RR-4 CC N-509 8"-VES-151-48 137P1 3 0 0 0 0 RR-4 CC N-509 8"-WES-151-4A 1 0 0 0 0 RR-4 CC 4509 8"-VES-151-3B 2 0 0 0 0 RR4 CC N-5C9 8"-VES-151-3A 2 0 0 0 0 RR-4 CC N-509 8"-WES-151-4A 137Q1 1 0 (' O O RR-4 CC N-509 8"-WES-15148 1 0 C 0 0 RR-4 CC N-509 10 -WES-151-38 156Et 6 1 0 0 1 RR 4 CC N-509 8" WES-151-100 2 0 0 0 0 RR-4 CC N-509 10"-WES-151-37 166F1 1 0 0 0 0 RR4 CC N-509 8"-VES-151-99 1 0 0 0 0 RR-4 CC N-509 Total SWS 29 3 1 1 1 ;
33 % 66 % 100 %
VI-3 66 RSV&C 6 -VES-151-69 1359 1 0 0 0 0 RR-4 CC N-509 6*-VES-151-69 1361
- O O O O 2R4 CC 4509 Tctal RBV&C 2 0 0 0 0 0% 3% 0%
7 VT-3 70 CRV&C 6*-WS-151-2 1474/75 3 1 1 0 0 RR-4 CC 4509 6" WS-151-35 1475 1 0 0 0 0 RR-4 CC N-509 6"-WS-151-1 2 0 0 0 0 RR4 CC N-509 6"-WS-151-1 1476 1 0 0 0 0 RR-4 CC N 509 6"-WES-151-41 1456 1 0 0 0 0 RR4 CC 4509 6*-WES-151-40 1 0 0 0 0 RR4 CC 4509 6"-WES-151-41 1458 1 0 0 0 0 RR4 CC 4509 6 -WES-15140 1 0 0 0 0 RR4 CC 4509 Total CRV&C 11 1 1 0 0 100 % 100 % 100 %
File: APP-CE1.WPD Appendix C - 9 cf C - 13
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m k.) g 'd b g g Posser James A. Fitzpatrick Nuclear Power Plant .!AFM4002 APPENDIX C - Program Summary Tables Re 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam Itera Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Itet'is Sch'd 5 1' 2"* , 3* Rea Remarks / Comments Total Ezr-nination item 119 14 5 4 5 05% 64 % 100 %
D2.30 Integraf AN.-Wde-A.; arx!;'y: tragic Snubbe s VT-3 Integrat Attachment Note 3 VT-3 10 RHR 16*-WS-151-30A 137H1 1 0 0 0 0 RR-11 TS Se 3 61 Total Examination item 1 0 0 0 0 0% 0% 0%
D2.40 Integ al Attachment-Spnng Type Supper' VT-3 Integ si Attachment Note 3 VT-3 toRHR 16*-WS-151-298 137C1 1 0 0 0 0 RR4 CC N-509 16" WS-151-29A 137E1 1 0 0 "J 0 RR-2 CC N-509 Total RMR 2 0 0 J 0 0% 0% 0%
VT-3 15 RBCLC 6*-VES-151-103 103C2 1 0 0 1 RR-4 CC N-509 Total RBCLC 1 0 0 1 0% 100 % 100 %
Total Examination item 3 1 0 0 1 0% 0% 100 %
D150 Integral Attachment 4cnstant Load Type Support VT-3 Integral Attachment 0 0 0 G t 0 tkAe 3 D160 Integral Attachment-Shock Abscrbers VT-3 in*egeal AttacW 0 0 0 0 0 RR-11 TS Secten 3 61. Note 3 Total Ex&n-An t;v . Category 123 15 5 4 6 Does not include D2.10 33 % ES% 100 %
fee: APP-CE1.WPD Appendix C - 10 of C - 13
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) ( (v) e g Pavver James A. Fitzpatrick Nuclear Power Plant w .s occ2 APPENDIX C - Program Summary Tables Revision- o ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January s.1 sos Exam item Description Exam System Un3 or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 15' 2 . F F.sq Remarks /Co... ..u ^e Notes:
(1) The system boundary extends up to and includeg the Frst nom.aPy closed vatve or wahre capable of automa*c desire as reW to perform the safetyela'ed system functen (2) The system hydrostate test shat be conducted at or near the end of each %.I.v.. rtenrai or dunng the same inspecten pened of each inspecten interval for
- u@. Program B.
(3) in the case cf muttiple w.vv.en's within a system of smlar design, functen, and service, the inte7al .;,Ja. 4 of only one of the muriple ww a shal be exammed. The integraf attachme it semed for exammation shat w.w4v.4 to those = w m a supports selected by nW-25t0(b).
(4) There are no e-v^e or exdusens f'om these requirements except as speo5ed in IWA-5214(c).
(5) A system hyWestatic test (NVD-5223) and w.v...,4 VT-2 examinaten are acceptable in lieu of the system p essure test (1WD-5221) and VT-2 exar,mahon (6) Snubbers are examined and tested in ec.w e,- with Plant Techmcal Swee e Sedon 3 6.1.
t File: APP-CE1.WPD Appendh 11 of C - 13
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g gPhaw James A. FitzpatriCk Nuclear Power Plant [[::JAF-tS3-0002|JAF-tS3-0002]] APPENDIX C - Program Summany Tables Re o ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Line or ISO No- No- INSPEt," TION PERIODS Rei Method 5 Req Remarks / Comments stem Component ID No. Items Sch'd 1' 2* 3*
Examination Category: D-C. SYSTEMS IN SUPPORT OF REstDUAL HEAT REMOVAL FROM SPENT FUEL STORAGE POOL D3.10 Pressure Retaining Components VT-2 Pressure Retairung Boundary Each ;%e Period Note 1. 4. and 5 VT-2 19 FPC 6*-WD-153-14A 11381 1 3 1 1 1 RR-10 8"-WD-153-66 113E1 1 3 1 1 1 RR-10 8"-WD-151-67 1 3 1 1 1 RR-10 8"-W19-151-17 1 3 1 1 1 RR-10 10"-W19-151-1B 113D1 1 3 1 1 1 RR-10 8"-WD-153-13A 113F1 1 3 1 1 1 RR-10 6"-WD 153-14A 1 3 1 1 RR-10 6"-WD-;63-148 1 3 1 1 1 RR-10 6"-WD-153-14B 113C1 1 3 1 1 1 RR-10 Total FPC 9 27 9 9 ,* 9 VT-2 Prassure Retamng Boundary Each ;,.we, Intervat Note 1. 2,4 and 5 VT-2 19 FPC 6"-WD-153-14A 11381 1 1 0 0 1 RR-3/10 8"-WD-153-66 113E1 1 1 0 0 1 RR-3/10 8"-WD-151-67 1 1 0 0 1 RR-3/10 8"-W19-151-17 1 1 0 0 1 RR-3/10 10"-W19-151-1B 113D1 1 1 0 0 1 RR-3/10 8"-WD-153-13A 113F1 1 1 0 0 1 RR-3/10 6"-WD-153-14A 1 1 0 0 1 RR-3/10 6"-WD-153-148 1 1 0 0 1 RR-3/10 6"-WD-153-148 113C1 1 1 0 0 1 RR-3/10 Total FPC 9 9 0 0 t 9 Total Examination item 180 36 9 9 18 D120 Integral Attachment-Ce.wm,a Supp&ts and Restraints VT-3 Integral Attachrnent Note 3 VT-3 19 FPC 6"-WD-153-14A 11381 4 1 0 1 0 RR-4 CC N-509 8"-WD-153-66 113E1 1 0 0 0 0 RR-4 CC P?-509 8"-WD-151-67 2 0 0 0 0 RR-4 CC N-509 10"-W19-151-18 11301 2 0 0 0 0 RR-4 CC N-509 8"-WD-153-13A 113F1 1 0 0 0 0 RR-4 CC N-509 6"-WD-153-14A 4 1 1 0 0 RR-4 CC N-509 6"-WD-153-148 2 0 0 0 0 RR-4 CC N-509 Fife: APP-CE1.WPD Appendix C - 12 of C - 13
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- > gPower James A. Fitzpatrick Nuclear Power Plant JAF-ts:-Oo02 APPENDIX C - Program Summary Tables nevisert 0 ASME CODE CLASS 3 SYSTEMS AND COMPONENTS January 6,1998 Exam item Description Exam System Une or ISO No. No- INSPECTION PERIODS Ret item Method Component. ID No- Items Sch'd 157 2* , 3* Rec Remarks / Comments Total Examination item 16 2 1 1 , 0 50 % 100 % 100 %
D3.30 Integra! Attachment IAechanical and Hydraulc Snubbers VT-3 Integral Attachment 0 0 0 0 0 TS Section 3 6.1, Note 3 D3.40 Integral Attachment-Spreg Type Support VT-3 Integral AttachrreM Note 3 VT-3 19 FPC 8 -WD-151-67 113E1 1 1 0 0 1 RR-4 CC N-509 Total Examination item 1 1 0 5 1 0% 0% 100 %
D3.50 Integral Attachment-Constant Load Type Support VT-3 leeoral Attachment 0 0 0 0 0 Note 3 D3 60 Integral Attachment-Shock Absorbers VT-3 Integral Attachraant 0 0 0 0 0 Note 3 Total Examination Category 17 3 1 1 1 33 % 66 % 100 %
Notes:
(1) The ::ystem boundary extends up to and inchrding the fwst norma 5y closed valve or vatve capable of automate closure as requeed to perform the safety-related syste m funcien.
The systm hydrostate test shall be conducted at or near the end of each inspecten interval or dunng the same ;W.w. period cf each kW,w. intarval for (2) inspection Prygram B. r (3) In the case of rnuftiple wi ~.,v.4e within a syt iem of svnitar design, funcbon, and service, the integral eM ,+-d of only one of the multple cw.vwede shat be examined. The integral attachrnect selected for examination shall w.csgw 4 to those w vw,= 4 supports selected by rWF-2510(b).
(4) There are no exemptons or exclusions from these requrements except as specified in IWA-5214(c).
(5) A syster.: hydrostatic test (1WD-5223) and accompanying VT-2 examinaten are acceptable in lieu cf the system pressure test (1WD-5221) and VT-2 examinaten.
(6) Snubbers are examirN and tested in accordance wrth Plant Techn6 cal SW T el.wi. Secten 3 61.
File: APP-CE1.WPD Appendix C - .s of C - 13
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- > g essser James A. Fitzpatrick Nuclear Power Plant aAr-4Sim02 APPENDIX D - Program Summary Tables Revision: o ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January s,1sse Exam item Description Exam System Une or ISO No- No. INSPECTION PERIODS Ret item n'athod Component. lD No. Items Sch'd 5 1' 2* , 3* Req Remarks / Comments Examination Category: F-A, SUPPORTS F1.10 Class 1 Piping Supports Visual VT-3 25% of Class 1 VT-3 02-2 RC 28-W14-GE-2A 3001 1 1 0 1 0 RR-2 CC N-491-1 28-WH-GE-1A 3001 2 0 0 0 0 RR-2 CC N491-1 28-WH-GE-2B 3002 1 0 0 0 0 RR-2 CC N-491-1 28-W14-GE-19 3002 1 0 0 0 0 RR-2 CC N-491-1 22-WH-GE-3A 3001 2 1 0 0 1 RR-2 CC N-491-1 22-WH-GE-38 3002 2 1 1 0 0 RR-2 CC N491-1 Total RC 9 3 1 1 1 33 % 66 % 100 %
VT-3 toRHR 24-W20-902-14A 3013 2 1 0 0 1 RR-2 CC N 491-1 24-W20-902-14B 3013 3 1 0 1 0 RR-2 CC N-491 1 20-W20-1504-42 3011 3 2 1 1 0 RR-2 CC N-491-1 Total RHR 8 4 1 2 1 25% 75% 100 %
VT-3 12 RWC 6-WR-902A-1 3018 3 1 0 1 0 RR-2 CC N-491 1 4-WD-902A-14 3018 4 1 1 0 0 RR-2 CC N-491-1 Total RWC 7 2 1 1 0 50 % 100 % 100 %
VT-3 13 RCIC 4-W22-902-4 3019 12 3 1 1 1 RR-2 CC N-491-1 3-SHP-902-17A 3020 4 0 0 0 0 RR-2 CC N-491-1. Exempt IWB-1220(a) 7 Total RCIC 16 3 1 1 1 33% 66 % 1M%
VT-3 14 CS 10-W23-902-5A 3022 2 1 1 0 0 RR-2 CC N-491-1 10-W23-902-58 3023 1 1 0 0 1 RR-2 CC N-491-1 Total CS 3 2 1 0 1 25% 25% 100 %
File: JAPP-DE1.WPD Appendix D - 1 of D -9
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- > gh James A. FitzpatriCk Nuclear Power Plant APPENDIX D - Program Summary Tables JAF-:si-0002 Revision: 0 ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January s,199s Exam Item Description Esam System Une or ISO . No. No- INSPECTION PERIODS Rel item Method Component. ID No. Items Sch'd is' 2* 3* Req Remarks! Comments VT-3 23 HPCI 14-W25-902-3A 3026 1 1 0 0 1 RR-2 CC N-491-1 1(MiHP-902-19A 3024 2 1 1 0 0 RR-2 CC N-491 1 Total HPCI 3 2 1 0 1 25% 25 % 100 %
VT-3 29 MS 24-SHP-902-1A 3031 4 1 1 0 0 RR-2 CC N-491-1 21-SHP-902-1B 3031 5 2 0 2 0 RR-2 CC N-491-1 24-SHP-902-1C 3032 6 1 0 0 1 RR-2 CC N-491-1 24-SHP-902-1D 3032 6 3 1 1 1 RR-2 CC N-491-1 Total MS 21 7 2 3 2 28 % 71 % 100 %
VT-3 34 FW 18-WFP-902A-4A 3033 4 2 0 0 2 RR-2 CC N-491-1 18-WFP-902A-48 3034 4 1 0 1 0 RR-2 CC N491-1 12-WFP-902A-SC 3033 4 1 1 0 0 RR-2 CC N-491-1 12-WFP-902A-5A 3033 1 0 0 0 0 RR-2 CC N-491-1 12-WFP-902A-5D 3034 4 1 0 0 1 R".-2 CC N-491-1 12-WFP-902A-5B 3034 1 0 0 0 0 RR-2 CC N-491-1 Total FW 18 5 1 1 3 20 % 40 % 100%
Total Class 1 85 28 9 9 10 32 % 63 % 100 %
t File: JAPP DE1.WPD Appendix D - 2 of D -9
e, gPever James A. Fitzpatrick Nuclear Power Plant APPENDIX D - Program Summary Tables JArass-0002 Revision: o ASME CODE CLASS 1,2 AND 3 COMFONENT SUPPORTS January s,1ess Exam item Description Exam System Line or ISO No. No- INSPECTION PERIODS Ret Item Method Component ID No. Items Sch'd is' 2* 3"D Req Remarks / Comments F120 Class i Mpsg Supports Vsual VT-3 15% of Class 2 VT-3 03 CRD 10-WR-901 3039 4 1 1 0 0 htR-2 CC N-491-1 10WR-901 3040 3 1 1 0 0 RR-2 CC N491-1 B-WR-901 3029 12 2 0 0 2 RR-2 CC N491 1 8-WR-901 3040 15 2 0 2 0 RR-2 CC N491-1 Total CRD 34 6 2 2 2 33 % 66 % 100%
VT-3 10 RHR 24-W20-302-11 A 3006 4 2 0 1 1 RR-2 CC N-491-1 24-W20-302-118 3006 3 2 1 0 1 RR-2 CC N-491-1 244V20-302-17 3006 2 0 0 0 0 RR-2 CC N491-1 24-W20-152-38 3009 3 0 0 0 J RR-2 CC '-401-1 24-W20-152-3A 3010 2 1 1 0 0 RR-2 CC h 191-1 20-W20-302-8B 3004 2 1 0 0 1 RR-2 CC N491-1 20-W20-302-8A 3005 2 1 0 0 1 RR-2 CC N-491-1 20-W20-302-17 3006 3 0 0 0 0 RR-2 CC N491-1 20-W20-152-28 3008/10 8 3 2 1 0 RR-2 CC N491-1 20-W20-152-2D 3009 1 0 0 0 0 RR-2 CC N-491-1 20-W20-152-2C 3010/11 11 6 2 3 1 RR-2 CC N491-1 20-W20-152-2A 3010 1 0 0 0 0 RR-2 CC N491 1 20-W20-1524A 3010 1 0 0 0 0 RR-2 CC N491-1 20-W20-152-6B 3009 1 0 0 0 0 RR-2 CC '4491 1 16-W20-302-78 3004 1 0 0 0 0 RR-2 CC N-491-1 16-W20-302-7D 3004 1 0 0 0 t0 RR-2 CC N-491-1 16-W20-320-108 3004 2 1 0 1 0 RR-2 CC N491-1 16-WS-302-36B 3004 2 0 0 0 0 RR-2 CC N491-1 16-W20-02-7C 3005 1 0 0 0 0 RR-2 CC N491-1 16-W20-3f;2-7A 3005 1 0 0 0 0 RR-2 C. N491 1 16-W20-302-10A 3005 2 0 0 0 0 RR-2 CC N-491 1 16-WS-302-56A 3005 3 0 0 0 0 RR-2 CC N491-1 16-W20-302-15A 3007 3 1 0 1 0 RR-2 CC N491-1 16-W20-302-158 3008 4 1 0 0 1 RR-2 CC P491-1 16-W20-302-34 3014 2 1 1 0 0 RR-2 C 491-1 16-W20-302-9B 3015 4 1 0 1 0 RR-2 CC N-491-1 16-W20-152-58 3015 2 0 0 0 0 RR-2 CC N-491 1 16-W20-152-5A 3015 2 0 0 0 0 RR-2 CC N-491-1 File: JAPP-DE1.WPD Appendix D-3 of D -9
g g g V U p > g P m er James A. FitzpatriCk Nuclear Power Plant JAF-IS62 APPENDIX D - Program Summary Tables Revision: 0 ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January 6,1998 Exam item Description Exam Sys'em 1.ine or ISO No- No- INSPECTION PERIODS Rel item Method Corr >onent ID No- Items Sch'd 1 ** 2* 3* Reg Remarks / Comments 12-W20-302-13A 3007 10 1 1 0 0 RR-2 CC N491-1 12-W2042-13B 3008 11 3 0 1 2 RR4 CC N491-1 10-W20-302-12A 3007 3 0 0 0 0 Excuded fWC-2500-1 10-W20-302-128 3008 2 0 0 0 0 RR-2 CC N-491-1. Excluded IWC-2500-1 8-W20-302-38 3006 22 0 0 0 0 Excktded IWC-2500-1 8-W20-152-39 3009 12 0 0 0 0 RR-2 CC N-491-1. Excluded IWC-2500-1 8-SHP-902-32A 3014 13 3 1 1 1 RR-2 CC N-491-1 8-SHP-902-32B 3014 8 1 1 0 0 RR-2 CC N-491-1 6-SLP-302-34A 3014 6 0 0 0 0 Excluded IWC-2500-1 6-SLP-302-348 3015 6 0 0 0 0 RM-2 CC N-491-1. Excluded IWC-2500-1 Total RHR 171 29 10 10 9 34 % 68 % 100 %
VT-3 13 RCIC 8-SLP-152-22 3019 4 0 0 0 0 Excluded reVC-2500-1 Total RCIC 4 0 0 0 0 0% 0% 0%
VT-3 14 CS 16-W23-152-1 A 3021 4 1 1 0 0 RR-2 CC N-491-1 16-W23 iS2-1B 3021 4 0 0 0 0 RR-2 CC N491-1 12-WZ3-302-4A 3022 7 2 0 1 1 RR-2 CC N491-1 12-W2-302-4B 3023 10 2 0 1 1 RR-2 CC N-491-1 Total CS 25 5 1 2 , 2 20 % 60% 100 %
VT-3 1L SCLC 6 -WCL-151-29A 103F 3 0 0 0 0 Excluded IWC-1500-1 Total RBCLC 3 0 0 0 0 0% 0% 0%
VT-3 23 HPCI 10-SLP-152-25 3027 6 2 1 1 0 RR-2 CC N-491-1
.6-W25-152-7 3028 2 1 0 1 0 RR-2 CC N-491 1 16-WCP-152-1 3028 8 0 0 0 0 RR-2 CC N 491-1 14-W25-902-3 3025 16 3 1 1 1 RR-2 CC N-491-1 10-W25-902-4 3025 4 1 1 0 0 RR-2 CC N-491-1 File: JAPP-DE1.WPD Appendix D - 4 of D -9
,7 p p V V V e gigwer James A. Fitzpatrick Nuclear Power Plant [[::JAF-IS8-0002|JAF-IS8-0002]] APPENDIX D - Program Summary Tables Revision o ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January 6,1998 Exam item Description Exam System Line or ISO No- No. INSPECTION PERIODS Ret itera Method Component. ID No. Items Sch'd 1s' 2" . 3* Reg Rmiarkst > omments 10-SHP-902-19 3024 10 2 0 G 2 RR-2 CC N 491-1 10-SHP-902-34 3024 1 0 0 0 0 RR-2 CC N-491-1 Total HPCI 47 9 3 3 3 33 % 66 % 100 %
VT-3 27 CPS 30"-N-152-13A 3030 1 0 0 0 0 Enduded fWC-2500-1 30"-N-152-138 3030 1 0 0 0 0 Endudedt h W 1 30" N-152-13C 3030 1 0 0 0 0 Eh MC-2500-1 30"-N-152-13D 3030 1 0 0 0 0 Exduded thM 1 30"#152-13E 3030 1 0 0 0 0 Exduded W.h W 1 24*#152A-18A 3029 3 0 0 0 0 Exekndd WJC-2500-1 20~-n-152A-16 3030 2 0 0 0 0 Erduded IWC-2500-1 20"-N-152A-20 3029 2 0 0 0 0 Exduced IWC-2500-1 18~-N-152A-26A 3029 1 0 0 0 0 Exduded rWC-2500-1 14"-N-152A-9 3029 4 0 0 0 0 Exduded IWC-2500-1 Total CPS 17 0 0 0 0 VT-3 29 MS 24-SHP-902-101 A 3 1 1 0 0 RR-2 CC N-491-1 24-SHP-902-101B 3 0 0 0 0 RR-2 CC N-491-1 24-SHP-902-101C 3 1 0 0 1 RR-2 CC N-491-1 24-SHP-902-101D 3 0 0 0 0 RR-2 CC N-491-1 Total MS 12 2 1 0 1 50 % 50 % 100 %
Total Class 2 313 51 17 17 17 33 % 66% t 100 %
File: JAPP-DE1.WPD Appendix D - 5 of D -9
v s) J
- > geener James A. FitzpatriCk Nuclear Power F ant Jar- Si-0002 APPENDIX D - Program Summary Tables Revision- 0 ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January 6,1998 Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Component ID No- Items Sch'd 1st ge 3* Reg hmarks! Comments F1.30 Class 3 Pping Supports Vrsual VT-3 10% of Class 3 VT-3 10 RHR 16-WS-151-29A 166C1 28 3 0 1 2 RR-2 CC N491-1 16-WS-151-298 1668s 39 7 5 0 2 RR-2 CC N491-1 12"-WS-151-27D 137A1 1 1
- O O RR-2 CC N491-1 12"-WS-151-27B 1 0 0 0 0 RR-2 CC N491-1 16-WS-151-280 1 0 0 0 0 RR-2 CC N491 1 16-WS-151-288 1 0 0 0 0 RR-2 CC N491-1 16-WS-151-568 2 1 0 0 1 RR-2 CC N491
- 12-WS-151-27A 137D1 1 0 0 0 0 RR-2 CC N491-1 12-WS-151-27C 1 0 0 0 0 RR-2 CC N491-1 16-WS-151-28A 1 0 0 0 0 RR-2 CC N491-1 16-WS-151-28C 1 1 0 1 0 RR-2 CC N-491-1 16-WS-151-56A 1 0 0 0 0 RR-2 CC N491 1 Total 10 RHR 78 13 6 2 5 46% 61 % 100 %
VT-3 13 RCIC 8-SLP-152-22 102B1 2 0 0 0 0 RR-2 CC N491-1 6-WCD-152-2 102A1 3 2 0 2 0 RR-2 CC N491-1 6-W22-152-16 102A1 5 0 0 0 0 RR-2 CC N-491-1 Total RCIC 10 2 0 2 0 0% 100 % 100 %
VT-3 15 RBCLC 8-WES-151-100 103C8 5 1 0 0 1 RR-2 CC N491-1 8-WES-151-99 10381 4 1 0 1 0 RR-2 CC N-491-1 6-VES-151-104 103C8 12 1 0 0 e i RR 2 CC N491-1 6-WCL-151-103 103C1 9 2 1 0 1 RR-2 CC N491-1 6-WCL-151-29 3 0 0 0 0 RR-2 CC N491-1 6-WCL-151-61 103C2 12 1 0 0 1 RR-2 CC N-491-1 6-WCL-151-29A 103F 1 0 0 0 0 RR-2 CC N-491-1 6-WCL-151-24 1 0 0 0 0 RR-2 CC N491-1 6-WCL-151-96 1 0 0 0 0 RR-2 CC N-491-1 6-WCL-151-72 3 0 0 0 0 RR-2 CC N491-1 Total RBCLC 51 6 1 1 4 16% 33 % 100 %
File: JAPP-dei.WPD Appendix D -6 of D -9
V d V p> mewyukPower James A. FitzpatriCk Nuclear Power Plant [[::JAF-Si-c002|JAF-Si-c002]]
- # AufherNy APPENDIX D - Program Summary Tables Revision: 0 ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January 6,1998 Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret Item Method Component ID No. Items Sch'd is' 2* , 3" Req Remarts/ Comments VT-3 19 FPC 6-WD-153-14A 113B1 14 2 1 1 0 RR-2 CC N491-1 6-WD-153-148 113F1 8 1 1 0 0 RR-2 CC N491-1 10-W19-151-1B 11301 7 0 0 0 0 RR-2 CC N491-1 8-W19-151-17 1 0 0 0 0 RR-2 CC N491-1 8-WD-153-13A 113F1 1 0 0 0 0 RR-2 CC N491-1 8-WD-151-67 113Et 7 1 0 0 1 RR-2 CC N-491-1 Total FPC 38 4 2 1 1 50 % 75% 100 %
VT-3 46 SWS B-WES-151-4B 137P1 3 0 0 0 0 RR-2 CC N-491-1 8-WES-151-4A 127P1 4 0 0 0 0 RR-2 CC N491-1 12-WES-151-28 4 0 0 0 0 RR-2 CC N-491-1 10-WES-151-38 25 2 1 0 1 RR-2 CC N-491-1 12-WES-151-9 137N1 3 0 0 0 0 RR-2 CC N491-1 8-WES-151-10 6 1 0 1 0 RR-2 CC N491-1 8-WES-151-3A - 4 0 0 0 0 RR-2 CC N491-1 6-WES-151-5A 2 0 0 0 0 RR-2 CC N491-1 C-WES-151-58 2 1 0 0 1 RR-2 CO N491-1 6-WES-151-6A 2 0 0 0 0 RR-2 GC N491 1 6-VES-151-6B 2 0 0 0 0 RR-2 CC N491-1 8-WES-151-100 2 0 0 0 0 RR-2 CC N491-1 10-WES-151-37 166F1 20 2 0 1 1 RR-2 CC N491-1 8-WES-151-99 3 0 0 0 0 RR-2 CC N491-1 12-WES-151-2A 137K1 3 2 1 1 0 RR-2 CC N491-1 Total SWS 85 8 2 3 3 25% 62% e 100*/.
VT-3 66 RBVSC 6-WES-151-69 1361 21 2 1 0 1 RR-2 CC N-491-1 6-WS-151-2 1363 4 0 0 0 0 RR-2 CC N-491-1 ToutRBV&C 25 2 1 0 1 50 % 50 % 100 %
File: JAPP-DEt.WPD Appendix D - 7 of D -9
O O O Agh James A. Fitzpatrick Nuclear Power Plant jar-asim2 APPENDIX D - Program Summary Tables Revision: o ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January s 1ssa Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Re4 item Method Component. ID No. Items Sch'd 1" 2* 3* Req Remarks / Comments VT-3 70 CRV&C 6-YES-151-41 1456 16 3 1 0 2 RR-2 CC N491 1 i 6-WES-151-40 1457 12 2 0 0 2 RR-2 CC N-491-1 6-WS-151-35 1 0 0 0 0 RR-2 CC N-491-1 6-WS-151-1 5 0 0 0 0 RR-2 CC NJ91-1
- .-WS-151-2 1474 8 1 1 O O RR-2 CC N 191-1 Total CRV&C 42 6 2 0 4 16% 16 % 100 %
Total Class 3 329 41 14 9 18 34 % 56 % 100 %
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t File: JAPP-DE1.WPD Appendix D - 8 of D -9
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L) J g g Pavuor James A Fitzpatrick Nuclear Power Plant jar-ist-0002 APPENDIX D - Program Summary Tables Revisers: o ASME CODE CLASS 1,2 AND 3 COMPONENT SUPPORTS January 6.1998 Exam item Description Exam System Line or ISO No- No. INSPECTION PERIODS Ret item Method Component. lO No. Items Sch'd 1s' 2* T* Req Remarks / Comments F1.40 Supports Other than Peng Supports (Class 1. 2. 3. ard MC) Vrsual VT-3 100% of the supports VT-3 02-2 RCP 02-2-P-1A 3001 3 3 3 0 0 *1xcluded per fWT-2510(a)
VT-3 02-2 RCP 02-2-P-1B 3002 3 0 0 0 0 Emduded per WE-2510(a)
VT-3 03 CRD 03TK-1A 3033 4 4 3 0 1 VT-3 03 CRD 03TK-1B 3038 4 0 0 0 0 VT-3 10 RHR 10E-2A 3037 6 6 2 2 2 VT-3 10 RHR 10E-28 3037 6 0 0 0 0 VT-3 10 RHR 10P-38 3004 1 1 0 1 0 VT-3 10 RHR 10P-3D 3004 1 0 0 0 0 VT-3 10 RHR 10P-3C 3004 1 0 0 0 0 VT-3 10 RHR 10P-3A 3004 1 0 0 0 0 VT-3 13 RCIC 13TU-2 167C1 1 1 1 0 0 VT-3 13 RCIC 13P1 102A1 1 1 0 1 O VT-3 14 CS 14P-1 A 3022 1 1 0 0 1 VT-3 14 CS 14P-1B 3023 1 0 0 0 0 VT-3 23 HPCI 23TU-2 3026 1 1 0 0 1 Exduded per WE-2510(a)
VT-3 23 HPCI 23P-1M 3025 1 0 0 0 0 Total Others 36 18 9 4 5
!9% 72 % 100 %
Total Component Supports T63 138 49 39 r 50 35% 63 % 100 %
No'es:
(1) Item numbers shal be categorized to identh support types ty w.vu,e.1 suport funcbon (e g, A = supports such as one-GMee; rod hangers; B = s@
such as muitdredronal restraints; and C = supports that aBow thermal movement, such as spnngs).
(2) The total percentage sample sha8 be w,v.ind of supports from each system (e g . Mac Steam Feedwater. or RHR), where the indnndual sarm8e sizes are v.vvv.uv.el to the total number of nonexermt supports of each type and funcbon wethe each systent
^
(3) For muttpe m.v .ea5e other than peng within a system of similar dessn function, and service, the supports cf only one of the rnuttpe w.wsa. are required to be exammed.
(4; To the extent practical, the sa:N supports selected for examinaten dunng the first inspecbon interval shall be examined du ing each .m.w..;++ inspecbort interval, File: JADP-DE1.WPD Appendix D - 9 of D -9
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V d d Mgh y James A. Fitzpatrick Nuclear Power Plant APPENDIX E - Program Summary Tables [[::JAF-4St-CCO2|JAF-4St-CCO2]] Re
- ion: o AUGh*.ENTED IGSCC SYSTEMS AND COMPONENTS Janumey s.1sse Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Rei item Method Component. ID No. Items Sch'd 15' 2* 3* Req Remarks / Comments Examination Category: B-F. PRESSURE RETAINING DISSIM11.AR METAL WELDS IGSCC Category A B5.10 fFS 4* or Larger Nozzle-to-Safe End Butt Wekfs VoVSurf C2-2 RC 10.0" WA 1 1 1 0 0 NUREG 0313 Piping B5.130 HPS 4* cr Larger Dissimilar Metal Butt Wekts VoVSurf 1 A CS 20.0" WA 1 1 1 0 0 NUREG 0313 14 CS 10.0" WA 2 2 1 1 O NUREG 0313 Totals 4 4 3 1 0 Notes:
(1) Ex .-a Gs. 5 are requwed of each safe end in each loop and w e.w branch of the reactor coolant systent
, (2) Examinations wf!I be used to satisfy both Section XI and NUREG 0313 reqtarement t
l File: APP-EE1.WPD Appendix E - 1 of E -6
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CJ' V) g g%nw James A. Fitzpatrick Nuclear Power Plant jar-tSi-ooo2 APPENDIX E - Program Summary Tables nevision- o AUGMENTED IGSCC SYSTEMS AND COMPONENTS January s.1sse Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Cwiipsm,t. ID No. Items Sch'd 1'8 2'* 3* Reg Remarks / Comments Exsmination Category: B-F. PRESSURE RETAIMNG DiSS!MILAR METAL WELDS IGSCC Category D B5.10 NPS 4" or Larger Nozzle-t Safe End Butt Webs As weHs Every two outages VoFSurf 01RPV 28 0* N/A 2 6 2 2 2 NUREG 0313 01RPV 12.0" MA 10 30 10 10 10 NUREG 0313 01RPV 10.0" WA 1 3 1 1 1 NUREG 0313 01RPV 7 MA 2 6 2 2 2 NUREG 0313 BS.20 Less than NPS 4* Nozzle-to-Safe End Butt WeHs As welds. Every two outag-s Surf 01RPV 3.0 MA 1 3 1 1 1 NUREG 0313 Piping 25.130 NPS 4* or Larger Dessimilar Metal Butt WeHs A3 weids. Every two outages VoVSurf 10 RHR 24.0" MA 6 18 6 6 6 NUREG 031::
Totals 22 66 22 22 22 Notes:
(1) Exammations are required of each safe end in each loop and cormecimg tranch of the reactor cociaM system.
(2) For those welds requmng both a volumetne and surface examination. the surface examinat~e n is only requwed once per intervat 7
File: APP-EE1.WPD Appendix E-2 of E -6
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e gPawer James A. FitzpatriCk Nuclear Power Plant . AF.4SM202 APPENDIX E - Program Summary Tables Revi. ion- o AUGMENTED IGSCC SYSTEMS AND COMPONENTS January s.199s Exam item Description Exam System Une or ISO No. No. ;NSPECTION PERIODS Ret 5 3*
item Method Component. lO No- Items Sch'd 1' 2* Reg Remar % . .e . .t.
Examination Category 84 PRESSURE RETATMNG WELDS IN P1ptNG IGSCC Category A B9.11 NPS 4* or Larger Chw Amie.; Weds 25%.100% of wed length Vol/ Surf 14 CS 100" N/A 11 3 1 1 1 NUREG C3t3 B9.12 LongitudinalWees 12.0 irt of each long. iew 4 cire. wee Vol/Strf 02-2 RC 28 0" N/A 60 15 5 5 5 NYPA CAT. A-1 02-2 RC 22X NAl 20 5 2 2 1 NYPA CAT. A-1 02-2 RC 12.0" MA 60 15 5 5 5 NYPA CAT. A-1 14 CS 10 0" MA 23 6 2 2 2 NYPA CAT. A-1 B9.31 Branch Pipe weMs NPS 4* or La~ger Surival 100% of wed length Vol/ Surf 022RC 22F MA 8 2 1 1 0 NYPA CAT. A* NUREG 0313 G2-2 RC 4.0" MA 8 2 1 1 0 NUREG 0313 Totals 190 48 17 17 14 Notes:
Exea.-.Aiao sha4 kdude 25% of as cawerci >a weMs. (27 x 25% = 6 75 (mewnum of 7 crc. welds to be exarneed),ams irtersectng 6414-4 weMs.
(1)
(2) The volumetne examinahons shas sats'y both the ASME Sectiori XI and NUREG 0313 requeernents.
(3) The surface exams wiR satsfy ASME Secbon X1 requrements..
t File: APP-EE1.WPD Appendix E - 3 of E -6
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- > g Pesser James A. Fitzpatrick Nuclear Power Plant jar-e n oc2 APPENDIX E - Program Summary Tables Revision- o AUGMENTED IGSCC SYSTEMS AND COMPONENTS January s.199e Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret item Method Cor- ~. at. ID No. Items Sch'd 5 1' 2* 3* Req Remarks /Co.. ma^s Ex ..:. A catego y B-J. PRESSURE RETA!NtNG '* ELDS IN PIP!NG GSCC Category C AI weeds every to years E9.11 NPS 4* cr Larger C %h aa Waids 100% of weid leng$
Vol/ Surf C2-2 RC 28 0* KA 2 2 1 1 0 NYPA CAT. C* NUREG 0313 C2-2 RC 28 0' KA 26 2S 9 9 8 NYPA CAT. C-2 NUREG G313 C2-2 RC 22F M4 7 7 3 2 2 NYPA CAT. C-2 NUREG C313 02-2 RC 12 0* 4A 19 19 7 6 6 NYPA CAT. C-2 NUREG C313 C2-2 RC 12F RA 1 1 1 0 G NYPA CAT. C 3 NUREG C313 89 31 Brarch P5e Cu,wA,:, NPS 4* or Larger 100% of weu length Vol/ Surf C2-2 RC 12 0* WA 5 5 2 2 1 NYPA CAT. C-2 NUREG C313 02-2 RC 12F KA 2 2 1 1 0 NYPA CAT. C-3 NUREG 0:'13 02-2 RC 4 C" NA 2 2 1 1 0 NYPA CAT. C-2 NUREG C313 Totals 64 64 25 22 17 t
File: APP-EE1.Yv"PD Appendix E -4 of E -6
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a gewverkessmer James A. Fitzpatrick Nuclear Power Plant JAF@M2
- '-~' APPENDIX E - Program Summary Tables ne ision: o AUGMENTED IGSCC SYSTEMS AND COMPONENTS Januny s,1ses Exam item Description Exam System Line or ISO No. No. INSPECTION PERIODS Ret item Method Component. ID No. Items Sch'd 8 1' 2* 3'" Rea Remarks / Comments Examination Category- B4, PRESSURE RETAINING YELDS IN PtPING IGSCC Category D B9.11 NPS 4 or Larger Ch.J eJaiWelds 100% cf wets lengm Vol/ Surf 01RPV 12 0" MA 4 12 4 4 4 NUREG 0313
- 01RPV 8.0" MA 1 3 1 1 1 NUREG 0313 02-2 RC 40" MA 1 3 1 1 1 NUREG 0313 Totals 6 18 6 6 6 Notes
(1) Exarmnatens shas include ALL weeds every two refuehng outages.
(2) The volumetne exammatens wL..d shan be used to satisfy both the ASME Code Secten XI and NUREG 0313 critena (3) Surface ex ia Lo shas be performed on!y once, and only on 25% af t% mends. (6 x .25 =1.5 or (mimmum of 2 welds).
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File: APP-EE1.WPD Appendix E - 5 of E -6
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{O I g gPower James A. Fitzpatrick Nuclear Power Plant APPENDIX E - Program Summary Tables [[::JAF-ISt-0002|JAF-ISt-0002]] Re won. o AUGMENTED IGSCC SYSTEMS AND COMPONENTS January s.1sss Exam item Description Exam System Une or ISO No. No. INSPECTION PERIODS Ret ST item Method Component. ID No. Items Sch'd 1 2* 3 Req Remarks / Comments Examinadon Category: B J. PRESSURE RETAINING WELDS IN P1 PING IGSCC Category E 89.11 Circumferential Welds ALL welds,100% of weM length.
Every two refueling outages Vol/ Surf 02-2 RC 28.0'* N/A 8 24 8 8 8 NUREG 0313 02-2 RC 22.0* N/A 1 1 1 1 1 NYPA CAT. E* NUREG 031J 02-2 RC 22.0" N/A 4 12 4 4 4 NUREG 0313 02-2 RC 12.0" N/A 1 3 1 1 1 NYPA CAT. E' NUREG 0313 02-2 RC 12.0" N/A 9 27 9 9 9 NUREG 0313 02-2 RC 4.0" N/A 1 3 1 1 1 N'JREG 0313 B9.31 NPS 4* or Larger All welds,100% of weld ength Every two refueling outages Vol/ Surf 02-2 RC 12.0" N/A 1 3 1 1 1 fCREG 0313 Totals 25 73 25 25 15 Notes:
(1) Examinations shall include ALL welds, every two refuertng outages.
(2) The volumetric exs.a 1.v.is shaft be used to satisfy both the ASME Section XI and NUREG 0313 iw.m.--n (3) The surface examinations shall be W.L 4 only once, and on only 25% of the welds. (25 x 25% = 6,25Xminimum of 6 welds) t File: APP-EE1.WPD Appendix E - 6 of E -6
G James A. Fitzpatrick Nuclear Power Plant e
[[::JAF-tsi-0002|JAF-tsi-0002]]
- > WskPower APPENDIX F-Program Summary Tables Revision: 0
- CLASS 1,2 AND 3 RELIEF REQUEST
SUMMARY
Jan. fy 6,1998 MEMT^,
Q' @'
g3 [M4 j Emesnissoc ion Susentaryof flogesetfor} Proposed Alternettves fleNet- n
' assefnequests syssess occomponset _ s- menemo4 ~
nemet.' - -
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- ~
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category < ' 7- ss : -
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1 Welded Repairs on Class B-P, C-H. Various Relief is requested from the JAF proposes to utdtze ASME 1,2 and 3 systems D-A, D-B, Hydrostatic Pressure Test Code Case N-416-1, "Altematwe and D-C required by IWA-4000 Pressure Test Requirement for Welded Repairs or instattation of Replacement items by Welding
- 2 Class 1,2,3 and MC F-A Various Reliefis requested from the JAF proposes to utilize ASME Component Supports requirements of Subsection Code Case N491-1,-Altemative IWF for determining Rules for Examinaten of Class component supports 1,2,3 and MC Component subject to examination. Supports -
3 Hydrostatic Testing of B-P, C-H, Various Relief is requested from the JAF proposes to utilize ASME Class 1,2 and 3 systems D-A, D-B and requirements of System Code Case N-498-1,"Altemative D-C Hydrosta'ic Testing in Rules for 10-Year System accordance with IWB,IWC Hydrostatic Testing for Class 1, and IWD 2 and 3 Systems" 4 Selection and Examination B-H, B-K-1, Various Reliefis requested from JAF proposes to utilize ASME of Class 1,2, and 3 C-C, D-A, the sclection and Code Case N-509,"Altemative Integra!!y Welded D-8, and D-C examination ofintegrally Rules for the Selection and Attachments welded attachments in Examination of Class 1,2 and 3 accordance with IWB, IWC Intprally Weided Attachments
- and IWD 5 Class 1 and 2 Longitudinal B-J 89.12 Reliefis requested from JAF proposes to utilize ASME Piping Welds B9.22 surface and volumetric Code Case N-524. -Altemative C-F-1 C 5.12 examination of longitudinal Examination Requirements for C-F-2 L.52 piping welds LongitudinalWelds in Class 1 and Class 2 Piping
- FILE APPF. TAB-E1 Appendix F - 1 of F - 55
7 James A Fitzpatrick Nuclear Power Plant [[::JAF-ts8-0002|JAF-ts8-0002]] APPENDIX F - Program Summary Tables Revision: 0
- > Pouver CLASS 1,2 AND 3 RELIEF REQUEST SUIMIARY January e,199s i
^'
y $i 7 M - 1*
~
s: - SameneryetRequestk Proposed Annessouves s -
Setat Q 4 yc ,
amenres;@ nauet , wN y - ^ e Anguest 8 noget nequest synesesLorcomponenb Enandnamen ~ e-5: - ~ , .. a =.
- w ,
N/A N/A Rertefis requested from the JAF proposes to utilize ASME 6 N/A requirements of IWA-4000, Code Case N-532. -Altemative and IWA-6000 Rt.asirements to Repair and Replacement Documentation Requtrements and ine.:nnce Summary Report F'.eparation and Submission" N/A N/A Reliefis requested from Perform UT examinations every 7 NUREG 0619 performing ultrasonic third refueling outage.
Feedwater Nozzles examinations every other refueling outage.
Control Rod Drive System C-H Various Retefis requested from the Examinations / Tests wi!! be 8
requirements of Table IWC- performed in the 3"D. Period, Class 2 2500-1 during a scheduled shutdown E-F E7.10 Relief is requested IWA- Perform the requirements of the 9 Primary Containment 2000,1992 Edition of 1989 Edition of Seebon XI Structure Section XI Class 1,2 and 3 System N/A N/A Reliefis requested from Compfy with IWA-5250(a)(2),
10 Pressure Tests IWA-5250(a)(2).1989 1992 Edition of Section XI Edition of Section XI Class 1,2 and 3 Snubbers N/A N/A Reliefis requested from Proposes to utilize USNRC 11 Section 2.3.2.2 and 2.3.2.3 Generic Letter 90-09 for of OM-1988, Part 4. intervais and sampling rates All systems / Class 2 C-H Various ASME Section XI Code Case N-522 12 Examination Catep.ny C-H A!I systems / Class 1,2,3 N/A N/A ASME Sechon XI, IWA- Code Case N-573 13 4000 and IWA-4400 Appendix F - 2 of F - 55 FILE APPF. TAB-E1
i JAMES A.FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 1 A. COMPONENT IDENTIFICATION:
Class 1,2, and 3 Systems >
B. EXAMINATION REQUIREMENTS:
ASME Section XI Requirements: ,
Article IWA4000, paragraph IWA-4110(b) and paragraph IWA-4700, Pressure Test, contain the system hydrostatic test requirements for welded repairs and the attach)g of items to be used for replacement (as defined in IWA-7110) to the system where such attachment is by welding.
C. RELIEF REQUESTED:
Relief is requested from the following:
Pursuant to the provisions of 10 CFR 50.55a(a)(3) relief is requested from the requirements specifed in IWA-4000, paragraph IWA 4110(b) and paragraph IWA-4700, Pressure Test of the ASME Boller and Pressure Vessel Code,Section XI,1989 Edition for welded repairs and the attaching of items to be used for replacement (as defined in IWA 7110) to the system where such attachment is by welding.
I D. BASIS FOR RELIEF:
The need for relief from the hydrostatic test requirement for weld repairs and welded pipe replacements is sometimes unexpected as a result of system conditions requiring immediate repair, or repairs required prior to startup as identifed during outage related inspections. As a consequence, immediale communication with NRC for testing relief on a case-by-case basis is necessary to avoid exceeding limiting conditions for operation or startup delays. This places an unnecessary be den on the Authority's JAF Plant and NRC resources.
Draft RG 1.147, Rev.12 dated May 1997, includes Code Case N-4161, Altemate Pressure Test Requirements for Welded Repairs or installation of Replacement items by Welding, Class 1,2, and 3 Section XI, Division 1", this code case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved it'; use at JAF durin" '+ie 2nd ISI interval and other nuclear stations.
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(j' FILE.APPF.RR Ei Appendix F-3 of F 55
JAMES A. FITZPATRICK ,
THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 1 >
j The Authonty is also taking exception to the USNRC's imposed conditions in addition to those conditions specifed in Code Case N 4161, Additional surface examinations should be performed ,
on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class 3 '
components when the surface examination method is used in accordance with Section ill The additional examinations, specifically requiring examination of the root pass layer of butt and socket welds is neither beneficial nor warranted for the following reasons:
- 1. given thinness of the root pass layer, leaving it in place awaiting inspection exposes it to unnecessary stresses, 2.- - materials requiring elevated preheat will need to be cooled to allow inspection, introducing additional unnecessary stresses,
- 3. grinding, if required to pass the NDE, may further thin and weaken the root pass layer,
- 4. the root pass layer is essentially remelted by the succeeding pass, minimizing any perceived benefit of the NDE results.
The system pressure test required by Code Case N-4161 subjects the completed weld to heat up, thermal growth, and vibration. The hydrostatic pressure testing, eliminated by this Code Case, is a carefully controlled process with slow pressurization, brief hold times, and s',ow depressurization, The system pressure testing subjects the weld to a more robust and challenging environment than that experienced during hydrostatic pressure testing.
O FILE:APPF.RR-Ei Appendix F-4 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 1 E. ALTERNATIVE EXAMINATIONS OR TESTS: l The afternative examination requirements will be implemented as defined by ASME Section XI Code Case N-4161, Alternative Pressure Test Requirements for Welded Repairs or installation of Replacement items by Welding, Class 1,2, and 3 Section XI Division , with the following exceptions:
(1) Currently in Draft RG 1.147 Rev.12, dated May 1997, imposes conditions in addition to those conditions specified in the Code Case, Additional surface examinations should be performed ;
on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class
- 3 compone
- 1 when the surface examination method is used in accordance with Section 111 The Authority does not intent to perform a surface examination on the root pass layer of butt and socket welds of the pressure retaining boundary of Class 3.
(2) Code Case N-4161 requires NDE to be performed in accordance with the methods and r acceptance criteria of the applicable Subsection of the 1992 Edition of Section Ill. Currently the volumetric examination is designated as radiography.
The Authority intents to utilize, radiograph or ultrasonics as the volumetric method, PT or MT as the surface method, the techniques and Acceptance Standards of these examinations will be in accordance with 1989 Edition of Section XI or the 1992 Edition of Section Ill. Visual examination will be conducted in accordance with the 1989 or 1992 Edition of Section XI.
F. IMPLEMENTATION SCHEDULE:
The Attemate Examination requirements of ASME Code Case N4161 will be incorporated into JAF inservice Inspection Program during the 3rd Ten Year Interval.
G. ATTACHMENTS TO THE RELIEF:
ASM5 Code Case N-416-1, Attemate Pressure Test Requirement for Welded Repairs or installation of Replacement items by Section XI, Division 1.
H. USNRC RESPONSE FILE:APPF.RR Ei Appendix F-5 of F 55
. - . .- . _ _ . . . . _ _ - - - . ~ . - . . . - . _ .
_ . ~ . . . . _ . - - . . - _ _.
MSE N-416-1
/'" CASES OF ASME BOILER AND PRESSURE YESSELCODE
, t Approval Date: February 15.1994 See Nurneric Index for enpiration and any reaffittnerson dates. .
l Case N-416-1 placement items by welding, a system leakage test may Alternatise IYessure Test Requirement for Welded be used provided the following requirements are met.
Res.t.!rs or hstallation of Replacement items by (a) NDE shall be performed in acmrdance with the ;
Welding, Class I,2 and 3 methods and acceptance criteria of the applicable Sub-Section XI, Division I section of the 1992 Edition of Section III.
(b) Prior to or immediately upon retum to service, a visual examination (VT 2) shall be performed in con-Inquiry: What altemative pressure test may be per* junction with a system leakage test, using tha !992 Edi-formed in lieu of the hydrostatic pressure test required tion of Section XI, in accordance with para. IWA 5000, by para. IWA-4000 for wcided repairs or installation of at nominal operating pressure and temperature.
replacement items by welding? (c) Use eithis Case shall be documented on an NIS-2 Form.
Reply: It is the opinion of the Committee that in lieu If the previous version of this case were used to defer of performing the hydrostatic prenure test required by a Class 2 hydrostatic test, the deferred test may be clim-para. IWA-4000 for welded repairt installation of re- inated when the requirements of this revision are met.
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769
1 l
JAMES A, FITZPATRICK !
THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO,2 O
b ;
A, COMPONENT IDENTIFICATION:
Class: 1,2,3 and MC Identification of System: All i Description of Components: Component Supports of Light-Water Cooled Power Plants B. EXAMINA'.*lON REQUIREMENTS:
ASME Section XI Requirements:
Subsection IWF, Requirements for Class 1,2,3 ar, s MC Component Supports of Light-Water Cooled Power Plants. Tne 1989 Edition of ASME SecF Xi requires component supports selected for examination be the supports of those components L .at nre required to be examined under IWB, IWC, IWD, and IWE during the first inspection interval. Also for multiple components within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined.
Code Category ite m . Exam Extent of Examination Number Method FA F1.10 VT? IWF-1300 / IWF-2510 eq F-A F1.20 W3 IWF-1300 / IWF 2510 F-A F1.30 VT 3 IWF-1300 / IWF 2610 F-A F1.40 VT3 IWF-1300 / IWF 2510 F-A F1.50 VT 3 IWF 1300 / IWF-2510 F-A F1.60 VT 3 IWF 1300 / IWF-2510 F-A F1.70 VT3 IWF-1300 / IWF-2510 C, RELIEF REQUESTED:
Pursuant to the provisions of 10 CFR 50.L5a(a)(3), relief is requested from the requirements specified in the ASME Boiler and Pressure Vessel Code, Cection XI,1989 Edition, Subsection lWF, Requirements for Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants .
D. BASIS FOR RELIEF:
At JAF during the 2nd Ten Year interval, component supports were selected for examination per ASME Code Case N-491, Attemative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1", which is approved for use in Regulatory Guide 1.147, Rev 11. During the 3rd Ten Year interval JAF intends to implement ASME Code Case N-491 1, - Attemative Rules for Examination Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1". A detailed review and comparison of
-O. Code Cases N-491 and N-491 1 reveals the only change is Paragraph 1220, Snubber inspection Ll Requirements which specifies, The inservice inspection requirements for snubbers shall be in accordance with the Section XI Edition and Addenda specified in the Owner's Inservice inspection
- Program, when previously Code Case N-491 stated, inservice inspection requirements for snubbers
. FILE:APPF.RR-E1 Appendix F-7 of F-55
i JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 2 O7- shall be in accordance with the requirements of IWF 5000". This change ooes not effect the selection or examination rulet specified in Code Cass N-491 1. The snubber inspection program at JAF wi'l be conducted in accordance with Section XI Edition vpecified for the JAF Inservice Inspection Program along with ASME Code Cases and/or relief requests. ASME Code Case N-491-1 will be implemented at JAF for selection and examination of Class 1,2,3 and MC Component Supports during the 3rd Ten-Year Interval.
Draft RG 1.147 dated May 1997, includes Code Case N 491 1, Alternative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants.Section XI, Division 1", this Code Case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved its use at other nuclear stations. _
E, ALTERNATE EXAMINATION REQUIREMENTS:
The following attemative examination requirements will be implemented as defined by ASME Section XI Code Case N-491-1, Alternative Rules for Examination of Class 1,2,3 and MC Component Supports of Light-Water Cooled Power Plants,Section XI, Division 1" F. IMPLEMENTATION SCHEDULE:
The Attemate Examination requirements of ASME Code Case N-491-1 will be incorporated into JAF Inservice inspection Program during the 3rd Ten-Year Interval.
G. A TTACHMENTS TO THE RELIEF:
AWE Code Case N491-1, Alternative Rules for Examination of Class 1,2,3 and MC Ccmponent Supports of Light-Water Cooled Power Plants,Section XI, Division 1" H. llSNRC RESPONSE C :
f FILE:APPF.RR-Ei Appendix F-8 of F-55
CASE N-491-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE
, O Approval Date: April 30,1993 See NumericalIndex for expiration and any reaffsmation dates.
Case N 4911 1000 SCOPE AND RESPONSIBILITY A!.ernative Rules for Examination of Class 1,2,3, and MC Component Supports of Light Water 1100 SCOPE Cooled Power PlantsSection XI, Division 1 This Case provides alternative rules for insenice nspection of Class 1,2,3, and MC component sup. 3
_ ports. _
1200 COMPONENT SUPPORTS SUIlJECT
/nquhy: What alternative examination require-ments to those stated in Section XI, Division 1. Sub- TO EXAMINATION AND TEST section IWF may be used when determining the com- 1210 Examination Requirements ponent supports subject to examination and establishing requirements for component supports. Th mis gima M q m &
following:
(a) piping supports; (b) supports other than piping supports.
Reply: It is the opinion of the Committee that the 1220 Snubber inspection Requirements following alternative rules may be used for determin.
V} ing component supports subject to examination and The inservice inspection requirements for snub-for establishing examination requirements for Class bers shall be in accordance with the Section XI Edi-1,2,3, and MC component supports under Subsec- tion and Addenda specified in the Owner's Inservice tion IWF,Section XI, Division 1. Inspection Program.
TAllLE OF CONTENTS 1000 SCOPE AND RESPONSIBILITY -3000 STANDARDS FOR EXAMINATION
-1100 Scope EVALUATIONS 1200 Component Supports Subject to Exam- -3100 Evaluation of Examination Results ination and Test 3110 Presenice Examination 1210 Examination Requirements 3111 General
-1220 Snubber Inspection Requirements -3112 Acceptance
-1230 Supports Exempt from Examination 3120 Inservice Examinations
-1300 Support Examination Boundaries -3121 General 2000 EXAMINATION AND INSPECTION 3122 Acceptance
-2100 Scope 3200 Supplemental Examinations 2200 Preservice Examination 3400 Acceptance Standards
-2210 Initial Examination M10 Acceptance Standards - Component 2220 ' Adiustment, Repair, and Replacement Support Structural Integrity
-2400 Inspection Schedule
-2410 Inspection Program 2420 Successive Inspections lQ)
C 2430 Additional Examinations TAllLES 2500 Examination Requirements -2410-1 Inspection Program A
-2510 Supports Selected for Examination 2410-2 Inspection Program B 2520 Method of Examination -2500-1 Examination Categories
. 913
l CA0E (2onthued)
N-491-1 y CASES OF ASME BOILER AND PRESSURE VESSEL CODE 1230 Supports Exempt from Examination adjusted in accordance with 3000, repaired, or re.
Component supports exempt from the examination Pl aced.
requiremen:s of 2000 are those connected to com- (b) For systems that operate at a temperature ponents and items exempted from examination under greater than 200*F during normal plant operation, IWD 1220, IWC 1220, IWD-1220, and IWE-1220. In the Owner shall perform an additanal presenice ex-addition, portions of supports that are inaccessible amination on the affected component supports dur-by being encased in concrete, buried underground, ing r i 11 wing the subsequent system heatup and or encapsulated by guard pipe are also exempt from e idown cycle unless determmed unnecessary by the examination requirements of 200C evaluation. This examination shall be performed dur-ing operation or at the next refueling outage.
1300 SUPPORT EXAMINATION IlOUNDARIES Support examination boundaries shall be in ac- 2400 INSPECTION SCl!EDULE cordance with IWF 1300.
2410 Inspection Program (a) Inservice examinations shall be performed 2M EXAa11NAT10N AND either during normal system operation or plant out-INSPECTION ages.
2100 SCOPE (b) The required examinations shall be comoleted n accordance with the inspection schedule provided The requirements of this Case apply to the ex. in Table -2410-1 or Table 2410-2 amination and inspection of component supports, b (c) The inspection period specified in (b) above (V) not to the insenice test requirements of IWF 500' may be decreased or extended by as much as ona year to enable an inspection to coincide with a plant outage, within the limitations of IWA 2400.
2200 PRESERVICE EXAMINATION (d) Following completion of Program A after 40 2210 Initial Examination y ars, successive if Pection intervals shall follow the 10 year inspection interval of Program B.
(a) All examinations listed in Table 25001 shall be performed completely, once, as a presmice ex-amination. These presenice examinations - tall be 2420 Successive Inspections extended to include 100% of all supports not ex- (a) The - luence of component support exami-empted by -1230.
nations estat thed during the first inspection inter-(b) Examinations for systems that operate at a val shall be ~peated during each successive inspec-temperature greater than 200*F during normal plant tion interval, to the extent practical.
operation shall be performed during or following in. (b) When a component support must be subjected itial system heatup and cooldowi. Other examina- to corrective measures in accordance with -3000, that tions may be performed prior to initial system heatup support shall be reexamined during the next inspec-and cooldown. tion period listed in the inspection schedules of the inspection programs of -2410.
2220 Adjustment, Repair, and Replacement (c) When additional corrective measures are not
, required during the next inspection period as a result (a) Prior to return of the system to senice, the of the examinations required by (b) above, the in-applicable examinations listed in Table -2500-1 shall spection schedule may revert to the requirements of b, performed on component suoports that have been (a) above.
C\
U 914 l
. -. .. _ - .. . . . - . ~.
CASE (continued)
N-491-1 p)
L CASES OF ASME BOILER AND PRESSURE VESSEL CODE 2430 Additional Examinations (a) When component supports must be subjected to corrective measures in accordance with 3000, the component supports immediately adjacent to those
. TABLE 24101 for which corrective action is required shall be ex-INSPECTION PROGRAM A amined. Also, the examinations shall be extended to include additional supports within the system, equ.J I"*'ll'" P"d- M'al*"* M**'*"*
in number and of ti.e same type and function as those rva P S Co pl ,% C t ,
scheduled for examination during the inspection pe-nod.
. 1st 3 too too (b) When corrective measures in accordance with 3000 are required as a result of the additional ex.
2nd 7 33 67 aminations, the remaining component supports with-l' 200 l" in the system of the same type and function as in (a) 3,, 33 33 3a above shall be examined.
17 40 So (c)(1) When corrective measures in accordance 20 66 75 with 3000 are required as a result of the additional 23 too too examinations in (b) above, examinations shall be ex-4th 27 8 16 tended to include a4. .sonexempt support $ potentially 30 25 34 subject to the same failure modes that required cor-33 So 67 rective measures in ACCordance with (a) and (b)
- 37 75 loo above.
O d
80 100 --
(2) These additional examinations shall include nonexempt component supports in other systems when support failures requiring corrective measures indicate non system related support failure modes.-
(d) When corrective measures are required by (c) above, the Owner shall examine those exempt com-
. TABLE 2410-2 INSPECTION PROGRAM B Ponent supports that could be affected by the same observed failure modes and could affect nonexempt Inspection Period, components.
Calendar Years of Plant Service Minimum Maximum Inspection Within the Examinations Examinations 2500 EXAMINATION REQUIREMEhTS Interval Interval Completed. % Credited, %
The following shall be examined in accordance 1st 3 16 34 with Table 2500-1.
7 So 67 (4) mechanical connections to pressure retaining to 100 100 components and building structure; successive 3 16 3. (b) weld connections to building structure; 7 50 67 (c) weld and mechanical Connections at interme-10 100 100 diate joints in multiconnected integral and noninte-gral supports; (d) clearances of guides and stops, alignment of supports, and assemuly of support items; (e) hot or cold settings of sprir,g supports and con-stant load supports; p (f) accessible sliding surfaces.
915
gr,.
/ \ ,
Lj) _= O kD 6
TABLE -2500-1 EXAMINATION CATEGORIES EXAMINATION CATEGORY F-4, SUPPORTS itene Support Type Examination Requirements / Extmenation Acceptance zo-h s
Estent cf Examination No-~ Examined ib No. Method Standard (See -2500) Frequency of Examination
- Dm (D p A O3 F1.10 Class 1 Piping Sepperts IWF-13001 Visual, VT-3 -3410 a
25% of Cisss ?'
F1.20 Class 2 Piping Supports Each inspection interval ==A h
C IWF 1300-1 Visual, VT-3 -3*10 15% of Class 2' Each inspection laterval 8 F1JO Class 3 Piping Supports IWF-13001 Visual, VT-3 A -3410 O
10% of Class 3* Each ineeettien interval F1.40 Supports Other than IWF-1300-1 visual, VT 3 -3410 100% of the supports
- Each inspection intervat Piping Suppsets (Class 1,2,3, and MC) h i a
NOTES: %
(1) Item numbers shall be categortred to identify support ypes by component support function (e.g, A = supports such as onewfirectional rod N-hangers; 8 = supports such as nrattidirectional restraints; tnd C = supports that allow thermal movement, such as springst g
l'8 (2) The total percentage sample shall be comprised of supports from each system (e.g., Main Steam, Feedwater, or RHR), where the !ndnridual sample stres are proportional to the total nurnber of nonexempt supports of each type and function within each svstem. $
pl (3) For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple
, cormponents are required fa be examined. g '
g (4) To the extent practical, the same supports selected for examination durtag the first inspectio : Interval shall be examined during each successive Inspection Interval.
I I
O E
l l
t
CASE (continued) g
! N-491-1
(]
G' CASES OF ASME BOILER AND PRESSURE VESSEL CODE 2510 Supports Selected for Examination (2) improper hot or cold settings of spring sup-Component and piping supports shall be examined Ports and constant load supports; in accordance with Table 25001. Component sup-l#l *I5*III "**"* I '"PP '**; '
ports to be examined shall be the supports of those H) Improper displacement settings of guides and stops.
components that are required to be examined under (b) repalr m, accordance with IWA-4000 and reex-IWD 2500, IWC-2500, IWD-2500, and IWE 2500 by ,
volumetric, surface, or visual (VT 1 or VT 3) exam- aminati n in ace rgance with 2200; (c) replacement in accordance with IWA 7000 and ination methods. Piping supports to be examined reexaminati n in accordance with 2200.
shall be the supports of piping not exempted under IWB-1220, IWC 1220, IWD 1220, and IWE 1220. 3112.3 Acceptance-by Evaluation or Test. As an alternative to the requirement of 3112.2, a compo-nent support that is unacceptable for service may be 2520 Method of Examination analyzed or tested to the extent necessary to sub-The methods of examination shall comply with stantiate its integrity for its intended service. Records those in Table 25001. Alternative methods of ex- and reports shall meet the requirements of IWA-amination meeting the requirements of IWA-2240 6000.
may be used.
3000 STANDARDS FOR 3120 inservice Examinations EXAMINATION EVALUATIONS 3121 General. Inservice nondestructive examina-O O
3100 EVALUATION OF EXAMINATION tions performed during or at the end of successive inspect.on intervals to meet the requirements of Ta-RESULTS ble 2500-1 and conducted in recordance with the 3110 Preservice Examinations procedures of IWA-2200 shall be evaluated by com-3111 General. The preservice examinations per- Paring the results of examinations with the accept-formed to meet the requirements of 2200. hall be ance standards specified in 3400.
evaluated by comparing the examination results with 3122 Acceptance acceptance standards specified in 3400.
3122.1 Acceptance by Examination. Component 3112 Acceptance supports whose examinations do not reveal condi-31121 Acceptance by Examination. Component tions described in -3410(a) shall be acceptable for supports whose examinations do not reveal condi- continued service. Verified changes or conditions tions described in -3410(a) shall be a ' table for from prior examinations shall be recorded in accord-service. ance with IWA 6220.
-3112.2 Acceptance by Correction. Component 3122.2 Acceptance by Correction. Component supports whose examinations reveal conditions de- supports whose examinations reveal conditions de-scribed in -3410(a) shall be unacceptable for serice scribed in -3410(a) shall be unacceptable for conti-until such conditions are corrected by one or more nued service until such conditions are corrected by of the following: one or more of the following:
(a) adjus' ment and reexamination in accordance (a) adjustment and reexamination in accordance with 2200 for conditions such as with 2200 for conditions such as (1) detached or loosened mechanical connec- (1) detached or loosened mechanical connec-tions; tions; lD U
917
CASE (c ntinu:d)
N-491-1 CASES OF ASME BOILER AND PRESSURE VESSEL CODE V
(2) improper hot or cold settings of spring sup- 3400 ACCEI"TANCE STANDARDS ports and constant load supports; (3) misalignment of supports; or 3410 Acceptance Standards - Component (4) improper displacement settings of guides 3 3 g 3, and stops. '(a) Component support conditions which are un-(b) repair in accordance with IWA-4000 and reex- acceptable for continued service shall include the fol-amination in accordance with 2200; lowing:
, (c) replacement in accordance with IWA 7000 arid (1) deformations or structural degradations of reexamination in accordance with 2200. fasteners, springs, clamps, or other support items; 3122.3 Acceptance by Evaluat;on or Test. As an .
(2) missing, detached, or loosened support alternative io the requirement of 3122.2, a compo-nent support or portion of a component support
'"h) are strikes, weld splatter, paint, scoring which is unacceptable for continued service may be roughness, or general corrosion on close tolerance analyzed and/or tested to the extent necessary to sub- machined or sliding surfaces; stantiate its integrity for its intended service. Records (4) improper hot or cold settings of spring sup-reports shall meet the requirements of IWA- ports and constant load supports; (5) m salignment of supports; (6) improper clearances of guides and stops.
(b) Except as defined in (a) above, the following are examples of non-relevant conditions:
3200 SUPPLEMENTAL EXAMINATIONS (1) fabrication marks (e.g., from punching, lay-out, bending, rolling, and machmmg);
Examinations that detect conditions that require (2) chipped or discolored paint; evaluation in accordance with the requirements of (3) weld splatter on other than close tolerance 3100 may be supplemented by other examination machined or sliding surfaces; methods and techniques (IWA 2000) to determine (4) scratches and surface abrasion marks;
' the character of the fiaw (i.e., size, shape, and ori- (5) roughness or general corrosion which does entation). Visual examinations that detect surface not reduce the load bearing capacity of the support; flaws that exceed 3400 criteria shall be supplement. (6) general conditions acceptable by the mate-ed by either surface or volumetric examinations. rial, Design, or Construction Specifications.
f
\
b 918
JAMES A. FITZPATRICK THIRD INSERVICE INSPECb DN INTERVAL g RELIEF REQUEST NO. 3 O
A. COMPONENT IDENTIFICATION:
Class: 1,2 and 3 Identification of System: All Description of Components: RPV, Pressure Vessels, Piping, Pumps, Valves, and Pressure Retaining Components B, EXAMINATION REQUIREMENTS:
ASME Section XI Requirements:
Table'fWB-2500-1, Category B-P (for Class 1), Table IWC-2500-1, Category C-H (for Class 2), Table IWD-2500-1, Category D-A, D-8, and D-C (for Class 3) contain the requirements for system hydrostatic 1nd leakage testing. The ASME Code requires system hydrostatic testing once per 10-year interval at or near the end of the interval.
1 Code Category ite m . ' Exam Extent of ExaminationL Number Method B-P B15.11 VT-2 One test per Interval B-P B15.51 VT-2 One test per Interval
!ql B-P B15.61 VT-2 One test per Interval N
B-P B15.71 VT-2 One test per Interval C-H C7.20 VT-2 Pressure Retaining Boundary C-H C7.40 VT-2 Pressure Retaining Boundary C-H C7.60 VT-2 Pressure Retaining Boundary C-H C7.80 VT 2 Pressure Retaining Boundary D-A D1.10 VT-2 Pressure Retaining Boundary /
Each inspection interval D-B D2.10 VT-2 Pressure Retaining Boundary /
Each inspection Interval D-C D3.10 VT-2 Pressure Retaining Boundary /
Each inspection Interval C. RELIEF REQUESTED:
l Pursuant to the provisions of 10 CFR 50.55a(a)(3), relief is requested from the requirements specified l f'3 - in Table IWB-2500-1, Subsections IWA-5000 and IWB-5222; Table IWC-2500-1, Subsection IWC-l Q1 5222; and Table IWD-25001, Subsection IWD-2500-1 and Section IWD-5223 of the ASME Boiler and
)
Pressure Vessel Code,Section XI,1989 Edition for hydrostatic testing of Class 1, 2 and 3 '
components and systems.
)
1 FILE:APPF.RR-E1 - Appendix F-15 of F-55 j i
l
, JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 3 iv)
D. BASIS FOR RELIEF:
The hydrostatic test requirement results in unusual difficulties without a compensating increase in the level of quality and safety. The difficulties are associated with the installation of blank flanges to isolate the tested portion from connecting systems, the removal of check valve inte nals, and when necessary the erection of temporary supports to allow the use of water as the hydrostatic fluid.
Additional activities included in this effort are erection and removaf of scaffolding, the removal and replacement of insulation, the removal and restoration to service (along with retesting) of electrical components, the setup and removal of the testing equipment, and the retum of the system to its normal configuration.
Draft Regulatory Guide 1.147 dated May 1997, includes Code Case N-498-1, Alternative Rules for 10-Year Hydrostatic Testing for Class 1,2, and 3 Systems,Section XI Division 1", this Code Case has not been published in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved its use at other nuclear stations.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The following altemative examination requirements will be implemented as defined by ASME Section XI Code Case N-498-1, Altemative Rules for 10-Tear System Hydrostatic Testing for Class 1,2, and 3 Systems,Section XI, Division 1" F. IMPLEMENTATION SCHEDULE:
[
V The Altemate lixamination requirements of ASME Code Case N-498-1 will be incorporated into JAF Inservice inspection Program during the 3rd Ten-Year Interval.
G. A'ITACHMENTS TO THE RELIEF:
ASME Code Case N498-1, Alternative Rules for 10-Year System Hydrostatic Testing for Class 1, 2 and 3 Systems,Section XI, Division 1" H. USNRC RESPONSE NYPA was granted approval to implement this Code Case during the Second inservice inspection interval, per USNRC TAC Number M90801, dated 02/13/95.
O]
FILE.APPF.RR-E1 Appendix F-16 of F-55
CASE N-498-1 CASES OF ASME BOII.ER AND PRESSURE VLi5 ' ODE
/O l
Approval Date: May 11,1994 L/ See NurneticalIndex for expiration and any reaffirmation dates.
Case N-498-1 (2) The boundary subject to test pressurization Alternative Rules for 10-Year System Hydrostatic during the system pressure test shall extend to all Testing for Class 1,2, and 3 Systems Class 2 components included in those portions of sys-Section XI, Division I tems required to operate or support the safety system function up to and including the first normally closed Inquuy What alternative rules may be used in lieu valve, including a safety or relief valve, or valve ca-of those required by Section XI, Division 1, Table pable of automatic closure when the safety function IWB-2500-1, Category B-P, Table IWC-2500-1, Cat. is required.
egory C-H, and Table IWD.2500-1, Categories D-A, (3) Prior to perfoiming the VT 2 visual exami-D-B, and D.C, as applicable, for the 10-year system nation, the system shall be pressurized to nominal hydrostatic test? operating pressure for a minimum of 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for in-sulated systems and 10 minutes for noninsula'-f sys-tems. The system shall be maintained at nominal op-erating pressure during performance of the VT-2 Reply: visual examination.
(a) It is the opinion of the Committee that as an (4) The VT 2 visual examination shall include alternative to the 10-year system hydrostatic test re- all components within the boundary identified in quired by Table IWB-2500-1, Category B-P, the fol. (b)(2) above, lowing rules shall be used. (5) Test instrumentation requirements of IWA-(1) A system leakage test (IWB 5221) shall be 5260 are not applicable.
conducted at or near the end of each inspection in- (c) It is the opinion of the Committee that, as an (p) terval, prior to reactor startup. alternative to the 10-year system hydrostatic test re-(2) The boundary subject to test pressurization quired by Table IWD-2500-1, Categories D A, D-B, during the system leakage test shall extend to all or D-C (D-B for the 1989 Edition with the 1991 and Class 1 pressure retaining components within the sys- subsequent Addenda), as applicable, the following tem boundary, rules shall be used.
(3) Prior to performing the VT 2 visual exami- (1) A system pressure test shall be conducted at nation, the system shall be pressurized to nominal or near the end of each inspection interval or during operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated the same inspection period of each inspection inter-systems and 10 minutes for noninsulated systems. val of Inspection Program B.
The system shall be maintained at nominal operating (2) The boundary subject to test pressurization pressure during performance of the VT-2 visual ex- during the system pressure test shall extend to all amination. Class 3 components included in those portions of sys-(4) Test temperatures and pressures shall not tems required to operate or support the safety system exceed limiting conditions for the hydrostatic test function up to and including the first normally closed curve as contained in the plant Technical Specifica- valve, including a safety or relief valve, or valve ca-tions. pable of automatic closure when the safety function (5) The VT-2 visual examination shall include is reqGred.
all components within the boundary identified in (3) Prior to performing the VT 2 visual exami.
(a)(2) above. nation, the system shall be pressurized to ncminal (6) Test instrumentation requirements of IWA. operating pressure for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated 5260 are not applicable. systems and 10 minutes for noninsulated systems.
(b) It is the opinion of the Committee that, as an The system shall be maintained at nominal operating alternative to the 10-year system hydrostatic test re- pressure during performance of the VT-2 visual ex-quired by Table IWC-2500-1, Category C-H, the fol- amination, em lowing rules shall be used. (4) The VT-2 visual examination shall include
) (1) A system pressure test shall be conducted at all components within the boundary identified in or near the end of each inspection interval or during (c)(2) above.
the same inspection period of each inspection inter. (5) Test instrumentation requirements of IWA-va! of Inspection Program B. 5260 are not apphcable.
943
l i'
JAMES A.FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 4 A. COMPONENT IDENTIFICATION:
Class: 1,2, and 3 Identification of System: All Description of Components: Integrally Welded Attachments of vessels, piping, pumps, and valves B. EXAMINATION REQUIREMENTS:
ASME Section XI Requirements: 1989 Edition, no Addenda Table'lWB-2500-1, Category B-H & B-K-1 (for Class 1); Table IWC-2500-1, Ca:--l C-C (for Class 2 ) and Table IWD-2500-1, D-A, D-8, and D-C (for Class 3) contain the reqairemen.s for selection and examination of integrally welded attachments.
Code. Item Exam Method ' Extent of Examination Category Number B-H B8.10 Volumetric or 3rd Interval no examinations surface required (inspection Program B)
Note: 1, 2,4 B-K-1 B10.10 Volumetric or 3rd Interval no examinations iO surface required (Inspection Program B)
Note: 1,2,3,4 B-K-1 B10.20 Volumetric or 3rd hterval no examinations surface required (insnection Program B) Note: 1.2,3,4 B-K-1 B10.30 Volumetric or 3rd Interval no examinations surface required (Inspection Program B) Note: 1,2,3,4 C-C C3.10 Surface IWC Table-2500-1 Note: 1,2,3
!C-C C3.20 Surface IWC Table-2500-1 Note: 1,3.4 C-C C3.30 Surface IWC Table-2500-1 Note: 1,3,4 C-C C3.40 Surface IWC Table-2500-1 Note: 1.3.4 D-A D1.20 Visual. VT-3 IWD Table-2500-1 Note: 3 D-A D1.30 Visual, VT-3 IWD Table-2500-1 Note: 3 D-A D1.40 Visual, VT-3 IWD Table-2560-1 Note: 3 _
O FILE:APPF.RR-E1 Appendix F-18 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 4 B. EXAMINATION REQUIREMENTS: "Cantinued
.5 Code item Exam Method - Extentof Examination s Category - Number D-A D1.50 Visual, VT-3 IWD Table-2500-1 Note: 3 D-A D1.60 Visual, VT-3 IWD Table-2500-1 Note: 3 DB D2.20 Visual, VT-3 IWD Table-2500-1 Note: 3 D-B D2.30 Visual, VT-3 IWD Table-2500-1 Note: 3 D-B D2.40 Esual, VT-3 IWD Table-2500-1 Note: 3 D-B D2.50 Visual, VT 3 IWD Table-2500-1 Note: 3, D-B D2.60 Visual, VT-3 IWD Table-2500-1 Note: 3 D-C _
D3.20 Visual, VT-3 IWD Table-2500-1 Note: 3 D-C D3.30 Visual, VT-3 IWD Table .tS00-1 Note: 3 a
O\ D-C D3.40 Visual, VT-3 IWD Table-2500-1 Note: 4 D-C D3.50 Visual, VT-3 IWD Table-2500-1 Note: 3 D-C D3.60 Visual, VT-3 IWD Table-2500-1 Note: 3 FILEAPPF.RR-E1 Appendix F-19 of F-55
. - -- . - .- -- . ~ . .-.
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 4 C. RELIEF REQUESTED:
, Pursuant to the provisions of 10 CFR 50.55a(a)(3) relief is requested from the requirements specified in Table IWB-2500-1, and applicable portions of Subsection IWA-2000; Table IWC 2500-1, and applicable portions of Subsection IWC-2000, and Table IWD-2500-1,and applicable portions of Subsection LWD-2000 of the ASME boiler and Pressure Vessel Code,Section XI,1989 Edition for selection and examination of Class 1,2, and 3 integrally welded attachments.
D. BASIS FOR RELIEF:
At JAF, for the 3rd Ten-Year Interval component supports shall be selected for examination in accordance with ASME Code Case N-491-1, which has also been requested as a relief, (Reference Relief Request 2). Application of Code Case N-509 defines alternative examination requirements that may be applied to ASME Code Class 1,2, and 3 integrally welded attachments. The extent of examination as uaied in Note 5 of Table 2500-1, Examination Categories B-H, B-K, C-C, and DA of Code Case N 509 requires examination of integral attachments associated with component supports selected for examination under the 1989 ASME Section XI, paragraph IWF-2510. Code Case N-509 a has been issued by the American Society of Mechanical Engineers and has been included in the 1995 Addenda of Section XI.
l Industry experience in the United States has also shown that ASME Code integral attachment welds have not experience degradation that would warrant continued examination to the extent required by the 1989 Edition of ASME Section XI. To date, no significant loading conditions or known material O degradation mechanisms have become evident that specifically relate to integral attachment welds in nuclear power plant piping. Should a service induced defect be detected in these welds, ASME Code Case N-509, specifies examination expansion enteria to ensure degradation in other attachment welds would be detected. Therefore, the health and safety of the public will continue to be maintained while implementing the altemative examination requirements of Code Case N-509. The 1989 Edrtion of Section XI inspection requirement for Class 2 and 3 integrally welded attachments result in unusual difficulties without a compensating increase in the level of quality and safety. The difficulties and cost associated with the increased radiation exposure. Additional, activities included in this effort are i erection and removal of scaffolding, the removal and replacement of insulation.
Draft RG 1.147 dated May 1997, includes Code Case N 509, Alternate Rules for the Selection and Examination of Class 1,2, and integrally Welded AttachmentsSection XI, Division 1, this code case has not been published in Regulatory Guide 1.147, Inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved it's use at other nuclear stations.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The following attemative examination requirements will be implemented as defined by ASME Section XI Code Case N-509, Alternative Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded AttachmentsSection XI, Division 1. In addition to those conditions specified in Code Case N-509: A minimum 10% sample of integrally welded attachments for each item in each Code Class per interval will be examined , with the following exception:
1). Examiration Category and item No.(s) for Class 3 Integrally Welded Attachments are defined in accorder.ce with the ASME Section XI,1989 Edition, Article IWD, Table IWD-25001.
O FILE:APPF.RR-Ei Appendix F-20 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 4 F. IMPLEMENTATION SCHEDULE:
The Altemate Examination requirements of ASME Coda Case N 509, including a minimum 10%
sample of integrally welded attachments for each item in each Code Class, will be incorporated into
- JAF Inservice Inspection Program during the 3rd Ten-Year interval
~
G. ATTACHMENTS TO THE RELIEF:
ASME Code Case N-509, Alternate Rules for the Selection and Examination of Class 1,2, and 3 Integrally Welded Attachments,Section XI, Division 1.
' ~
H. USNRC RESPONSE a
O V
4 l
1 ry b
FILE:APPF.RR-E1 Appendix F-21 of F 55 s
- . - - - - .. - . . _ . . . . . . . . - - . - . - ~ _- - . . _ - .
CASE N '509 p_ CASES OF ASME BOILER AND PRESStJRE VESSEL CODE O ).
Approval Date: November 26,1992 See Numeric Index for expiration L and any reaffirmation dates.
Case N 509 1.1 Exemption Criteria Alternative Rules for the Selection and Examis,alon (a) The exemption criteria provided in IWB 1220, of Class 1,2, and 3 Integrally Welded Attachments IWC-1220, and IWD 1220 may be applied to ClassSection XI, Division 1 1,2, and 3 components respectively, with integrally welded attachments, required to be examined in ac-Inquiry: What alternative requirements to those of cordance with Table 2500-1.
IWB, IWC, and IWD rnay be used to select and (b) Class 1,2, and 3 integrally welded attachment .
examine integrally welded attachments? examinations performed as a result of component support deformation cannot be credited under the
. requirements of IWB 2411 or IWB 2412, IWC-2411 Reply: It is the opinion of the Committee that the or IWC 2412, and 1WD 2411 or IWD-2412, respec-following rules may be used to select and examine tively, integrally welded attachments:
(a) This Case is limited to Examination Categories B-H, B K 1, C-C, D A, D B, and D C. 1.2 Inspection Schedule (b) Class 1,2, and .} component supports shall be Class 1,2, or 3 integrally welded attachments se-selected for examination m accordance with IWF of the 1989 Edition with the 1990 Addenda, lected for examination by sample selection criteria in accordance with Table 2500-1, Examinction Cate.
(c) Except for the selection of component supports for examination, all references to Section XI within gories B-K, C-C, and D-A, shall meet the require-v ments of IWB-2411 or IWB-2412, IWC-2411 or
- this Case shall be from the edition and addenda spec-IWC-2412, or IWD 2411 or IWD-2412, repectively.
ified in the Owner's Inservice Inspection Program.
1.3 Additional and Successive Examinations
/s) Class 1,2, and 3 additional and successive ex- 1 1.0 SCOPE arnination requirements of IWB-2430 and IWB-2420 These requirements apply to examination and for Class 1, IWC 2430 and IWC 2420 for Class 2 and sample selection of Class 1,2, and 3 integrally welded 3 as applicable, shall be applied to integrally welded attachments of vessels, piping, pumps, and valves attachments whose examinations reveal flaws or rel.
listed in Table 2500-1 as follows: evant conditions that exceed the acceptance stan-(a) Table 2500-1, Examination Category B K shall dards of IWB-3000, IWC-3000, and IWD 3000, re.
be used for Class 1 integrally welded attachments in spectively.
Examination Categories B H and B-K 1 of IWB. (b) When integrally welded attachments are ex-(b) Tabic 2500-1, Examination Category C-C shall amined as a result of identified component support be used for Class 2 integrally welded attachments in deformation and the results of these examinations Examination Category C-C of IWC. exceed the applicable acceptance standards listed (c) Table 2500-1, Examination Category D A shall above, additional or successive examinations sall be be used for Class 3 integrally welded attachments in performed when determined necessary based on an Examination Categories D A, D B, and D C of IWD. evaluation by the Owner, u
985
o o '
- z. ,g ,
TABLE 2500-1 UI m :
EXAMINATION CATEGORIES O ^8
-- EXAMINATION CATEGORY B-K, INTEGRAL ATTACHMENTS FOR CLASS 1 VESSELS, PIPING, PUMPS, AND val.VES il 3- ,
Examination c item No. Parts Examined' Requarements/ . Exawination Acceptance Extent of Frequency of g
Fig. No. Method Standard ;
Exammation Examination' ;
B10.10 Fra Vessels IWB-2500-13, Surface' IWB-3516 100% of required areas of each Integrally Welded -14, and -15 Each identified occurrence and welded attachment each inspection interva *
- Attachments g g
B10.20 Piping Integra!!y Welded IWB-2500-13, Surface IWB 3516 100% of required aac of each Each identified occurrence and 5
-14, and 15 welded attachmer* each impection interwar e
M Attachments N I 810.30 Pumps IWB-2500-13, E Surface IWB 3516 100% of required areas of each Integrally Welded -14, and -15 Each identified occurrence and es welded attachrnett each inspection interval' 2 Attachments g 810.40 Valves IWB-2500-13, S'wface IWD-3516 100% of required aeems cf each
<> Integra!!y Weided Ecch identified occurrence and
-14, and -15 welded attachmen* each impection interval
- Attachtrents NOTES:
(1) Examination is limited to those integrally welded attachments that meet the foHow eg conditions-ta) the attachment is on the outside surface of the pressure sctaining component, (b) the attachment provides component support as defined in NF 1110; and f"
' (c) the attachment meld joins the attachment either directly to the surface of the component or to art integrally cast or forged attachment to the corvponent. g ,
A (2) The extent of the exantination includes essentially 100% of the tength of the attachment wetd et each attachment subject to esamination. 3 (3) Selected samples of integrally welded attachments shall be erarnined each inspection interval. [-;
(4) in the case of multiple vtvels of signitar design, function and ervice, only one integrally welded attachment of only one of the multiple vessels shall be selected for exa n
(5) beFor piping, purgs, and valves, a sarmle of 10% of the welded a *.achrnents associated with the component soports selected for examinatione examined. , -
h sa under t"3 the 1990 Add '
M) inservice Examination is orrequired inspection, testing. whenever component support member defonnation (e.g , brokan, bent, or faulted out parts) is identified during operation, refuentg ma; ,
r (7)examination For the configuration shown in Fig IWB-2500-14, a volumetric examination of volume A-B-C-0 from side (8-C) of the circumferential welds of surfaces A-D and B-C. urfacemay be performed in li h
e a e
V ( k TABLE 2500-1 (CONT'D)
- EXAMINATION CATEGORIES EXAMINATION CATEGORY C-C, INTEGRAL ATTACHMENTS FOR CLASS 2 VESSELS, PIPING, PtfMPS, AND VALVES Examination Item Requirements / Examination Acceptance Extent of Frespeency of N o. PMs Examined' Fig. No. Method Standard Examination ** Eaaminttion*
C3.10 Pressure ifessels IbC-2500-5 Integrally Weided Attachments Surface IWC-3512 100% of rerluired areas of each welded attachment Each identified occurrence and each inspettien interval * {
m C3.20 Piping IWC-2500-5 Surface Integrally Welded IWC-3512 100% of required areas of each Each identified occurrence and g welded attachment each inspection interval' K Attachments "
C3.30 Pumps 8
IWC 2500 5 Surface IWC-3512 100% of required areas of each Each identified occurrence and Integrally Welded welded attachment each inspection interval
- h 30 Attachments C3.40 Valves IWC-2500-5 Swface IWC-3512 100% of required areas et each Each identified occurrence and %
Integrally Welded welded attachrnent each inspection intervef' Q NOTES:
(1) Esamination is limited to those integrally welded attactrnents that meet the following conditions:
(a) the attachment is on the outside surface of the pressure retainrng comp < 1ent; h m
(b) the attachment provides component support as defined in NF-1110; and (c) The attachment weld joins the attachment either directly to the surface of the cormonent or to ain integrally cast or forged attachment to the component.
(2) The extent of the examination includes essentially 100% of the length of the attachrnent weld at each attactenent subject to examination.
n (3) Selected samples of integrally welded attachments shall be examined each inspection interval Q m
(4) In the case of multiple vessels of similar design, function and service, only one integrally welded attachment of only one of the multiple vesse$$ shall be selec.ed e for examination.
(5) For piping, pumps, and valves, a sample of 10% of the welded attachments associated with the component supports selected for examination under the 1990 Addenda, IWF-2510 shall be examined.
(6) Examination is required whenever component support member deformation (e.g, broken, bent, or pulled out parts) es identified during operation, refueling, maintenance, examination, inservice impettion, or testir4 0 '
3>
(t) .
m n
za8 On 5 oa m-
f .
GJ G
- z. .g TABLE 2500-1 (CONT'D) U1 m EXAMINATION CATEGORIES O$
g
- EXAMINATION CATEGORY D.A. INTEGRAL ATTACHMENTS F8R CeASS 3 VESSELS, PfPING, PtfMPS, ANO VALVE 5 Examination k
3 Item - Requirements / j Examinaties Acceptance Estent of No. Parts Erassined' Fig. No. Frequency of g Method Standard Eaamination*3 Enamination" D1.10 Pressuee Vessels IWD-2500-1 Visual, VI I IWD-3000 100% of required areas of each Integrally Welded Each identifed occurrtace and Attachments welded atta.hment each inspection interval D1.20 Piping IWD-2500-1 Visual, VT-1 IWD-3000 100% of required areas of each M Integrally Welded Each identifed occurrence and "
welded attachment Attachments each inspection inte val k g
D130 Pumps IWD-2500-1 &
Visual, VT-1 IWD-3000 100% of required areas of each Each identified occurrence and M Integrally Welded Attachments welded attachment each inspection interval O p *
, D1.40 Valves 1% D-2500-1 DO Visual, VT 1 IWD-3000 100% of required areas of each g Integraffy WeideJ Each identified occurrence and y Attachments welded attachment each inspection interva! C3 y
NOTES:
a m
(1) Examination is limited to those integrally welded attachrnents that eneet the following conditions:
(a) the attachment is on the outside surface of the pressure retaining component; h M
th) the attachment provides component support as defined in NF-1110; and (c) the attachment weld joins the attachment either directly to the surface of the component or to an integrally cast or forged e'tachment to the cormonent (2) The extent of the examiriation inclades essentiary 100% of the length of the attachment weld at each attachment s.hiect to examination. ,
M (3) Selected samples of integrally welded attachments shall be examined each inspection interval. Att integrally welded attactments a selected for examination sh ll b e prbjact to corrosion, as determined by the Ovmer, such as the integrally welded attachments of the Service Water or Emergency Service Water systems. In othe case of mul function and service, the integrally welded attachments of only one of the multiple vessels shatt be selected for examination. For integrally weld valves a 10% sample shall be selected for exami:.ation. This percentage sample shatt be proportional to the total number of noneuempt integr'
@ ally welded att piping, pumps, and valves, located within each system subject to these examinations.
j (4) inservice Examination is required inspection, or testing.whenever component support member deformation (e.g., broken, bent, or putted out parts) is identified during operation refuelin g, maintenance, examhuttion, l
i
( JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 5 10 V
A. COMPONENT IDENTIFICATION:
Class: 1 and 2 Identification of Components: LongitudinalWelds Description of Components: Intersecting LongitudinalWelds of Selected Circumferential Welds B. EXAMINATION REQJIREMENTS:
Examination Examination Examination Description Category item Number BJ B9.12 Nominal Pipe Size 4.0 inch or Greater B9.22 Nominal Pipe Size < 4.0 inch C-F-1 C5.12 Longitudinal Welds > 3/8 inch Nominal Wall Thickness for Piping > 4.0 inch diameter.
C-F-2 C5.52 Longitudinal Welds 2 3/8 inch Nominal Wall h
Y Thickness for Piping > 4.0 inch diameter C. RELIEF REQUESTED:
Relief is requested from the additional ASME Code surface examination requirements for the following:
- 1. Class 1 to include at least a pipe-diameter length but no more than 12 inches of each longitudinal weld intersecting the circumferential welds required to be examined by Examination Categories B-F and B-J.
- 2. Clas- 2 to :nclude 2.51 at the intersection circumferential weld required to be examined under Examination Categories C-F-1 and C-F-2.
D. BASIS FOR RELIEF:
- 1. Longitudinal welds are fabricated during original manufacturing under controlled si conditions, which oroduce higher quality and more uniform residual stress patterns.
- 2. Longitudinal piping welds undergo heat treatment in the shop, which enhances the material properties of the weld arid reduces the residual stresses created by welding.
- 3. Results of previous weld inspections throughout the industry indicate that longitudinal welds have not been a safety concern, nor has there been any evidence of longitudinal weld defech (c)
'J compromising safety at nuclear power plants.
FILE:APPF.RR-E1 Appendix F-26 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 5 O
V
- 4. Longitudinal welds have not been shown to be susceptible to any particular degradation mechanism.
- 5. The only areas of a longitudinal weld which may be considered suspect are the ends of the weld where it is adjacent to the field fabricated circumferential weld. These areas fall within the volumetric examination boundaries of the adjacent circumferential weld.
- 6. The man-rem exposure and cost associated with the inspection of longitudinal welds is
, dependent on the time it would take *o remove / reinstall insulation and laterferences, locate the weld, prepare the weld fm examination and perform the examination.
- 7. Based on the above arguments, there is little, if any, technical benefit to performing inservice inspections on longitudinal piping welds. In addition, there are substantial radiation exposure and cost considerations associated with these inspections.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The following alternative examination requirements will be implemented as defined by ASME Section XI Code Case N-524, Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping.Section XI, Division 1".
F. IMPLEMENTATION SCHEDULE:
(v~}
The Attemate Examination requirements of ASME Code Case N 524 will be incorporated into JAF Inservice inspection Program during the 3rd Ten-Year Interval.
G. ATTACHMENTS TO THE RELIEF:
ASME Code Case N-524, Attemative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping,Section XI, Division 1" H. USNRC RESPONSE
,r--
- t L-FILE:APPF.RR-E1 Appendix F-27 of F-55
( CASE N-524 CASES OF ASME BOILER AND PRESSURE VESSEL CODE x
Approval Date: August 9,1993 See NumericalIndex for expiration and any restfirmation dates.
Case N 524 Alternative Examination Requirements for longitudinal Welds in Class 1 and 2 Piping Section XI, Division 1 Inquiry: What alternative requirements may be ap-plied to the surface and volumetric examination of-longitudinal piping welds specified in Table IWB-2500-1, Examination Category B-J, Table IWC-2500-1, Examination Lategories C-F-1 and C-F-2 (Exam-ination Category C-F prior to Winter 1983 Adden-da), and Table IWC-2520, Examination Category C.
G (1974 Edition, Summer 1975 Addenda)?
Reply: It is the opinion of the Committee that the following shall apply:
(a) When only a surface examination is required, examination of longitudinal piping welds is not re-quired beyond those portions of the welds within the
[m]
E examination boundaries of intersecting circumfer-ential welds.
(b) When both surface and volumetric examina-tions are required, examination of longitudinal piping welds is not required beyond those portions of the welds within the examination boundaries of inter-secting circumferential welds provided the fol'owing requirements are met.
(1) Where longitudinal welds are specified and locations are known, examination requirements shi.ll be met for both transverse and parallel flaws at the intersection of the welds and for that length of lon-gitudinal weld within the circumferential weld ex-amination volume; (2) Where longitudinal wehls are specified but locations are unknown, or the existence of longitu-dinal welds is uncertain, the examination require-ments shall be met for both transverse and parallel flaws within the entire examination volume of inter-secting circumferential welds.
1035
1 JAMES A. FITZPATRICK l THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 6 A. ARTICLE IDENTIFICATION:
Article IWA-4000 Repair Procedures Article IWA-6000 Records and Reports Article IWA 7000 Replacement B. ARTICLE REQUIREMENTS:
IWA-4800 The records required by IWA-6000 shall be completed for all repairs.
IWA-7520(8) Completed Owners Report for Repairs or Replacements, Form NIS-2 lWA-6210(c) The Owner shall prepare inservice inspection summary report for Class 1 and 2 pressure retaining components and their supports.
lWA-6220(c) Inservice Inspection summary reports shall be required at the completion of each inspection conducted during a refueling outage. Examinations, tests, replacements, and repairs conducted since the preceding summary report shall be included.
J IWA-6220(d) Each summary report shall contain the following:
(2) Owners Report for Inservice inspection, Form NIS-1 (3) Owners Report for Repair or Replacement, Form NIS-2 IWA-6230 Withia 90 days of tne completion of the inservice inspection conducted during each refueling outage, the Owner shall file ISI Summary Reports with the enforcement and regulatory authorities.
C. RELIEF REQUESTED:
Relief is requested from the following:
- 1. Preparation of the Owners Report for Inservice Inspection, Form NIS-1
- 2. Preparation of the Owners Report for Repair or Replacement, Form NIS-2.
- 3. Submittal of the summary report within 90 days following completion of the inservice inspection conducted during each refueling outage.
D.' BASIS FOR RELIEF:
JAF feels that the summary report required by IWA-6000 does not contain the information necessary to assure compliance with Code requirements, and therefore does not provide a compensation
- increase in the quality and/or safety at JAF.
O- The summary report does not fumish evidence of compliance with the ASME Boiler and Pressure V Vessel Code, Section Xl, inspection Program B, percentage requirements as mandated by IWB-2412, IWC-2412, and IWD-2412.
- FILE:APPF.RR-E1 Appendix F-29 of F-55
l JAMES A. FITZPATRICK
,- THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST No. 6 Class 3 components are excluded from the summary report Submittal.
Both a Final Report and Summary Report must be prepared, reviewed and approved in order to comply with Sub-articles IWA-6220 and IWA-6310 respectively.
The preparation, review, approval and certification of each record and report, within the time frame of 90 days following completion of each refueling outage, increases substantially the costs associated with inservice inspection activities, and puts an unreasonable time constraint on JAF without an increase in assurance of Code compliance.
~ ~
Draft RG 1.147 dated May 1997, includes Code Case N 532, Attemative Requirements to Repair and Replacement Documentation requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000',Section XI Division , this Code Case has not been published in Regulatory Guide 1.147, inservice inspection Code Case Acceptability ASME Section XI, Division 1", however, the NRC staff has approved it's use at other nuclear stations.
2 E. ALTERNATIVE EXAMINATIONS OR TESTS:
As an alternate to the requirements of IWA-4800, lWA-6000, and IWA-7528(8), JAF will implement ASME Code Case N-532, Alternative Requirements to Repair and Replacement Documentation Requirements and inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000', Division 1", (Note: 1 - ASME 1992 Edition Section XI).
I O F. IMPLEMENTATION SCHEDULE:
The Alternate Examination requirements of ASME Cowe Case N-532 will be incorporated into JAF Inservice Inspection Program during the 3rd Ten-Year interval.
G. ATTACHMENTS TO THE RELIEF:
ASME Code Case i+532, Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000, Division 1".
H. USNRC RESPONSE 4
w
\ /
, FILE:APPF.RR-E1 Appendix F-30 of F-55
CASE I
N-532 p CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: December 12,1994 See Numeric Index for expiration and any reaffirmation dates.
~
Case N 532 (e) The completed l'orm NIS 2A shall be main-Alternative Requirements to Repair and tained by the Owner.
Replacement Documentation Requirements and (f) The Owner shall maintain an index of Repair /
Inservice Summary Report Preparation and Replacement Plans in accordance with IWA-6340.
Submission as Required by IWA 4000 and IWA- The index shall identify the identification number 60005 required by (b) above and the inspection interval and Section XI, Division 1 period during which each repair or replacement was completed.
Inquiry: What alternatives may be used to the re-quirements of IWA-4910(d) and IWA 6210(e) for 2.0 OWNER'S ACTITTIT REPORT com?l etion of Form NIS 2 following repair or re. PREPARATION AND SUDMITTAL place: ment, and IWA-6210(c) and (d), IWA 6220, An OWNER'S ACTIVITY REPORT FORM p)
C IWA-6230(b), (c), and (d), and IWA-6240(b) for preparation and submittal of the inservice summary OAR 1 shall be prepared and certified upon com-pletion of each refueling outage. Each Form OAR-report and Form NIS-17 1 prepand during an inspection period shall be sub-mitted following the end of the inspection period.
Reph: It is the opinion of the Committee that as Each Form OAR 1 shall contain the following:
an alternative to the requirements of IWA-4910(d), (a) Abstract of applicable examinations and tests IWA-6210(c), (d), and (c), IWA-6220, IWA-6230(b), with the information and format of Table 1.
(c), and (d), and IWA-6240(b), the following provi- (b) A listing of item (s) with flaws or relevant con-sions may be used.This Case shall be utilized at least ditions that required evaluation to determine ac-until the end of the inspection period in which it was ceptability for continued senice, whether or not the invoked. flaw or relevant condition was discovered during a scheduled examination or test. The listing shall pro-1.9 CERTIFICATION OF THE REPAIR OR vide the information in the format of Table 2.
REPLACEMENT (c) Abstract for repairs, replacements and correc-(a) The Owner's Repair / Replacement Program tive measures perforrned, which were required due shallidentify use of this Case, to an item containing a flaw or relevant condition (b) A Repair / Replacement Plan shall be prepared that exceeded IWB-3000, IWC-3000, IWD-3000, in accordance with IWA 41405, and shall be given a IWE 3000, IWF-3000, or IWL-3000 acceptance cri-unique identification number, teria; even though the discovery of the flaw or rele-(c) Upon completion of all required activities as- vant condition that necessitated the repair, replace-sociated with the Repair / Replacement Plan, the ment or corrective measure, may not have resulted Owner shall prepare a REPAIR / REPLACEMENT from an examinttion or test required by this Division.
CERTIFICATION RECORD, FORM NIS-2A. If acceptance criteria for a particular item is not (d) Form NIS-2A shall be presented to the specified in this Division, the provisions of IWA-Inspector for certification. 3100(b) shall be used to determine which repairs,
) replacements, and corrective measures are required V 'AH references to lWA 4000 and IWA-6000 used in this Case refer to be included in the abstract. The abstract shall pro-to the 1992 Edition. vide the information in the format of Table 3.
1061
. ...-~ , - ~ - . .- . - . . _ _ - -- .. = . _ . -- -. . - -
CASE (continued)
N-532 CASES OF ASME BOILER AND PRESSURE VESSF1 CODE i
V
, FORM NIS 2A REPAIR / REPLACEMENT CERTIFICATION RECORD OWNER'S CERTIFICATE OF CONFORMANCE I certify that the represent by Repair / Replacement repart or replacement .
Plan number conforms to the requirements of Section XI.
Type Code symbol Stamp 4
CertifMte of Authorization No. Expiration Date j Signed Date Owner or Owner s Des.0aee. Titie s
CERTIFICATE OF INSERVICE INSPECTION
- l. the undersigned, holdin0 a valid commission issued by the National Board of Boiler and Pressure Vesselinspectors nd the State or Province of and emplo,Jd by of have inspected the items desenbed in RepairMeolacement Plan number dunng the period to and state that to the best of my knowledge and belief, the Owner has performed all the activities described in the Repair / Replacement Plan irl accordance with the requirements of SMion XI.
By signing this certificate neither the inspector not his employer makes any warranty, expressed or imphed, concerning the activities described in the Repair / Replacement Plan. Furthermore, neither the inspector nor his employer shall be liable in any manner for any personalinjury or property damage or loss of any kind arising from or connected with this inspection.
Commissions inspector's Segnature National Roard. State. Provmce. and Endorsements Date This form (E00126) may be obtained from the Order Dept., ASME,22 Law Drive. Box 2300. Fairfield, NJ 07007 2300.
m a i
%)_
1062
- - -. - - -- -. - - ~. - . . . - . ..-.-.- _- - .- - - - . .
I
- CASE (continued) i -
N-532 CASES OF ASME BOILER AND PRESSURE VESSEL CODE s
FORM OAR 1 OWNER'S ACTIVITY PFPORT 1
1 neo.n ueber owner es.~ e~ u m e, ow ori e,a,e no one u.e e. ,ie ums No co-o,s.ai .e~ ice dare aeu v e*ng o. age no.
w .e.ii..a.
coneat mspair n ime~ai ii.a x emori eu, rent so-iion penod --
itst. 2nd. M Edetion and Adoends of Section xi apphcatWe to the espection paan Date and reviseon of mspecteon plan
. E stion and Addenda o' %ct<on XI apphcable to repairs and replacements,if different than the mspection plan 4
i CERTIFICATE OF CONFORMANCE I certh that the statements maos en tNs owner's Activity menort are correct. and that the examinations, tests repairs.
replacements evaluations, sad corrective measures represented by tNs report cordorm to the requirements of Section XI.
l Certificate of Autheriranon No. Expiration Date is opeiicaneen 4
j' Signed Date Owner se Owner s Det.eaes, fee CERTIFICATE OF INSERVICE INSPECTION
- 1. ttw unders$ned, holding a vahd commission issued by the National Board of Boiler and Pressure Vessel inspectors and
! the State or Provmce of and employed by of have insoected the items cesenbed m trus Owner's Actmtv Peport. durmg the pened to and state tttat to the best of my knowtersgo and behef. the Owner has performed all actmties represented by trws report in accordance with the
, .reausrements of Section XI.
- By segrung t'u8 certificate neither the inspector nor Ns emotover makes any warranty, expressed or imphed, concernmg the exammations, tests, repairs. reodocements, evaluations and correctwo measures cescnoed tNs report. FurtF.Jrmore. neither the inspector nor his employev shall be kaote m any manner toe any personal ensury or property damage or a loss of any kind ansmg
- from or connected with this msoection.-
Commiss ons respecier s s gaame Nare.w Seard. Swe. ,roomse, see Eneernements Date TNs form (E00127) may be obtaened from the Order Dece.. ASML 22 Law Dnve. Son 2300. Fedeld. NJ 07007 2300.
1063
- - CASE (continued)
. N-532 -
-p)
CASES OF ASME BOILTR AND PRESSURE VESSEL CODE ,
NJ' 1
4 TABLE 1 ABSTRACT OF EXAMINATIONS AND TESTS Total
_ Total Total Total Examinations Examin 3tions Examinations Examinations Credited (%) To Examination Required for Credited for Credited (%) Date for The Category The Interval This Period For The Period Interval Remarks TABLE 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRED EVALUATION FOR CONTINUED SERVICE Flaw Flaw or Relevant Condition Found Examination item item Characterization During Scheduled Section XI Category - Number Description O (IWA-3300) Examination or Test (Yes or No)
TABLE 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE i
Flaw cr Relevant Condition Found Repeir, During Scheduled Replacement,Section XI Repa '
Code or Corrective item Description Examination or Date Replacement Class Measure Description of Work Test (Yes/No) Complete Plan Number
~O
. D 4
m
JAMES A. FITZPATRICK _
THIRD INSERVICE INSPECTION INTERVAL n_ RELIEF REQUEST NO. 7 A. COMPONENT IDENTIFICATION:
Class: 1 identification of System: Feedwater Nozzles B. EXAMINATION REQUIREMENTS:
- 1. NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking ,
~
- a. As required by NUREG 0619, Para. 4.3.2.2 inspection lntervals The routine inspection intervals for representative feedwater nozzle and sparger configurations given in Table 2 reflect the NRC staff's present estimate of the effectivenees of these sparger types in preventing cracks in feedwater nozzles. The inspection intervals apply to all plants of each configuration but may be revised as experience accumulates.-
- b. As required by NUREG-0619 Para. 4.3.2.3, UT Inspection and Subsequent PT of
. Recordable Indication At scheduled refueling outages for which a UT inspection of feedwater nozzles is A called for in Table 2, perform an extemal UT examination of all feedwater nozzle safe ends, bores, and inside blend radii. If indications are found in the safeend, evaluate per Section XI of the ASME Code. If recordable indications (defined in ASME Section V, Article 4, Paragraph T-441.8) are interpreted to be cracks in any nozzle, proceed with the sparger removal, PT of the nozzle bore and the nozzle blend radius, and repair. An acceptable PT, whether required by the finding of a UT indication or by the routine inspection schedule given in Table 2, includes removal of a sparger from one nozzle (see exception below) followed by flapper wheel grinding and examining, by PT, both the nozzle of the removed sparger and the accessible portions of the other nozzles. If any cracks are detected, remove all spargers and completely examine all nozzles, and remove all nozzle cracks.
- 2. ASME Section XI,1989 Edition, Article IWB-2000, Table IWB-2500-1, Examination Category B-D, item No. B3.20.-
- a. As required by ASME Section XI,1989 Edition, Article IWB-2000, Table 2500-1, Examination Category B-D, item No. B3.20.
The Feedwater Nozzles inside Radius Section requires inspection once every interval with partial deferral permbsible during the Third and Fourth intervals under conditions specified in Table IWB-2500-1. Examination Category B-D, Note (3).
L}
FILE:APPF.RR-Ei ' Appendix F-35 of F-55
-+ + - ,
JAMES A. FITZPATRICK THIRD INSERVICE WSPECTION INTERVAL RELIEF REQUEST NO. 7 -
. O C. RELIEF REQUESTED:
- 1. Relief is requested from the ultrasonic examination testing frequency requirements specified in NUREG-0619. Table 2, Routine inspection Intervals and the elimination of the routine liquid penetrant examination testing. The Authority is also requesting the elimination of the FW Nczzle Leakage Monttoring System (LMS) currently in place at FitzPatrick.
- 2. Relief is also requested from the requirements of ASME Section XI,1989 Edition, Article IWB 2000. Table IWB 25001, Examination Category B-D, item No. B3.20, for Feedwater inside Radius examinations.
~
D. BASIS FOR RELIEF: t The technical bases for the proposed changes are: (1) the effectiveness of the past and current UT examination for detecting FW nozzle cracks; (2) the effectiveness of the improved thermal sleeve design in minimizing thn bypass leakage based on the results of the Leakage Monitoring System over the past two operating cycles; (3) the absence of reportable FW sozzle indications during previous nondestructive examinations; (4) the results of a fracture mech &nics analysis of the FW nozzle; (5) the normal fuli power FWinlet temperature of 420 F for the FitzPatrick plant is significantly higher than FWinlet temperatures at many other plants. This minimizes the thermat fatigue on the nozzles due to both normal FW temperature changes and thermal sleeve bypass leakage, and minimizes the potential for FW nozzle cracking.
The FitzPatrick plant conformed to NUREG-0619 by: (1) removing the stainless steel cladding from i the FW nozzles; (2) installing triple thermal sleeve, double Aston-ring seal spargers; (3) cutting and
- cepping the contro! rod drive retum line; (4) changing the in'; mal valve trim in the low flow feedwater control valve; and (5) implementing an augmented inspection program. The Authority demonstrated to the satisfaction of the NRC staff (Reference 2) that rerouting the Reactor Water Clernup System return flow to all FW lines, and installing new low flow FW controllers, was not necessary.
, 1. UT Examination Techniaues
, Since NUREG4819 was issued, significant improvements '1 UT technology have been made. One of the major developments is computer modeling which when applied to nozzle examinations can optimize the UT techniques (coverage, exam angles, and scan paths) ,
used in the designated areas of interest. Modeling has been, and is currently being used for the development of numerous UT examination techniques and to verify the adequacy of existing technigres and examinations. The Performance Demonstration initiative (PDI),
which is in the process adopted by the industry for ihe qualrfication of personnel, procedures, and equipment has incorporated computer modeling into various aspects of the qualification process to validate examination coverage of complex geometries. The Authority has had two separate computer modeling evaluations performed on the FW nozzico at Fitzpatrick, the first by GE to optimize the manual examination techniques applied during RO12 and the second by EPRI that performed a comparison between the EP81 <senerated FW nozzle computer model and that of the examination procedures used o_ ng the 1990 FW nozzle inspection (GE automated). The results confirmed that UT techniques and procedures, (automated and manual used during 1990 and 1996 outages respectively), employed during examinations of the FW nozzles are optimal.
L FILE:APPF.RR Ei Appendix F-36 of F 55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 7 LJ The automated UT techniques, used at Fitzpatrick during the 1990 UT examination of the FW nozzles, are capable of detecting small (0.25 inch deep) fatigue cracks, as demonstrated in a 1991 qualification test (Reference 6). The qualification tests, as proven on Electromagnetic Discharge Machining (EDM) notches, confirmed that the techniques used during the 1990 Fitzpatrick inspection are capable of detecting a 0.25 inch deep flaw in the blend radius and bore regions of the FW nozzles. A similar demonstration was conducted at GE's Huntersville Facility, attended by an ANil representative (Arkwright), NYPA Level lil, NYPA ISI Program Manager and GE representatives. The demonstration entailed using the procedure developed from the computer mMel and scanning a FW mock-up block with EDM notches of varying depths located throughout Zones 1,2, and 3 as depicted in Attachment
. 1 Figure 1. The manual technique used was able to detect all of_the simulated defects with a minimum signal to noise ratio of 21
- 2. Leakace Monitorinu Sv.!detI)
A Leakage Monitoring System (LMS), that monitors for feedwater leahge past the sparger thermal sleeve rsals, was installed on all four nozzles at Ptzpatrick in the spring of 1992. The leakage data covers a period of three operating cycles, and confirms acceptable bypass leakage levels , without any increasing trends, for all four FW nozzles. Thermal sleeves at other plants with leakage monitoring systems also exhibit insignificant leakage (Refereace 6). This confirms the effectlyeness of the improved thermal sleeve design in minimizing bypass leakage. The nozzle temperature data from the A, B, and D nozzles confirm the absence of leakage. The data from the C nozzle correspond to a leakage rate from previous
,,)
(J i.
studies for the FW nozzles. Based on the indicated leakage 0.75 g.p.m. from the C nozzle the resulting contnbution to 40 year cumulative fatigue usage remains at about 0.28 (Reference 7) which le significently less than the acceptance limit of 1.0..
- 3. In!ir ection Results Nondestructive examinations of the Fitzpatrick FW nozzles performed in accordance with NUREG 0619, have not revealed any reportable indication to date. This includes the UT examinations performed in 1982,1985,1990,1995 and 1996; and VT examinations performed in 1985,1987,1990, and 1994.
- 4. FW Nozzle Fracture Mechanics Analysis A fracture mechanics ar.alysis of the FitzPatrick FW nozzle concluded that stress cycling from conservative temperature and flow profiles, when added to those resulting from other crack growth phenomena, such as startup and shutdc.cn cycles, do not result in the growth of an initial 0.25 inch crack to greater than one inch during the remaining life of the plant f 0.47 inches for the worst case), consistent with the criteria of Generic Letter 81-11.
The analysis (Reference 6) conservatively assumed failure of the first seal on the triple thermal sleeve sparger, and remains valid for the Fitzpatrick FW nozzle. The results of this analysis were previously approved by the NRC (Reference 2).
g3
(
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FILE:APPF.RR-E1 Appendix F-37 of F-55
l JAMES A. FITZPATRICK l THIRD INSERVICE INSPECTION INTERVAL l REUEF REQUEST NO. 7 l O
Conclusion:
l S.
Since NUREG-0619 was issued, significant improvements in UT technology have been i made. UT techniques used by the Authority to inspect the FW nozzles are capable of detecting small (0.25 inch deep) fatigue cracks. Given that PT examination requirements issued in 1980 (NUREG-0619) were based on limited confidence in the UT technology available at that time, and that areas of concern can be adequately examined using current UT techniques, a routine PT examination requirement for ths FW nozzle adds no assurance of nozzle integnty beyond that provided by the UT, and is therefore unnecessary, If a crack is indicated by T, a PT examination will not aid in determining the depth of the flaw, and will
_ only serve to confirm the existence of the crack. The current UT ermination prograrn for the FW nozzles, which utilizes an optimtzed inspection technique, along with the VT ex&mination schedule, is effective for assuring the integrity of the FW nozzles. The LMS, in service for three operating cycles, confirms the effectiveness of the improved FW sparger thermal sleeve design and the absence of unacceptable fatigue usage due to rapid thermal cycling.
Further, the higher FW inlet temperature minimizes the potential for thermal cycle fatigue usage even in the event of bypass leakage. For these reasons, and considering the absence of any anomalies associated with the present FW nozzle configuration, an Irmse in the UT examination interval, he elimination of the required PT examination * . LMS will not compromise plant safety.
Eliminating the routine PT examination of the FW nozzle will avoid exposing many plant workers to substantiallevels of radiation. The PT examination is coupled with a replacement of the sparger therma! sleeve seals. The dose survey for the FW sparger/ thermal sleeve modification in 1978 was 415 person-rem. The potential radiation exposure during a PT examination /sparger replacement task is expected to be 50 to 200 person-rem, depending on the extent of the difficulties that may be experienced during this activity. The PT examination /:parger thermal sleeve seal replacement program represents outage critical path time estimated to be 10 - 14 days. This estimate assumes no unanticipated difficulties associated with the sparger removal / replacement task. The potential uncertainties associated with the sparger removal / replacement would have an unknown impact on future refuel outages and their schedules. Once ths spargers have been removed, chances that a sparger can be refurbished and reused is romote, considering the high radiation doses associated with the radioactive sparger mateial. This would require the fabrication of new spargers and their installation at an estimated cost of $2 $3 million. This cost, and the power replacement cost resulting from an extended outage, represents a substantial expense to be absorbed by the Authority.
O FILE:APPF.RR-Ei Appendix F-38 of F-55 !
JAMES A.FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 7 -
E. ALTERNATIVE EXAMINATIONS OR TESTS:
- 1. The Authority will perform ultrasonic examination in the areas identified in Attachment 1 Figure 1, on FW nonles at Fitzpatrick once every ten years and perform viscal inspection of the FW nonles in accordance with NUREG-0619, Table 2.
- 2. The Authority shall schedule, perform and credit NUREG-0619 ultrasonic examinations of I Feedwater Nonles in lieu of the required ASME Section XI Nonle inside Railus examinations. !
l
- 3. . The Visual Examination for the Control Rod Drive Return Line Nonle shall continue to be l performed in accordance with NYPA Letter, JPN 8344, dated 07/07/83.
F. IMPLEMENTATION FCHEDULE:
Based on the Augmented FW nonle inspection schedule and in accordance with this relief request, ultrasonic examina' son of the FW nonles will be conducted in 2006. Visual examination of the FW nonles and Control Rod Drive Line noule are scheduled in accordance with NUREG-0619, Table 2.
G. ATTACHMENTS TO THE RELIEF:
Attachment 1 (Figure 1) I O \
H. UENRC RESPONSE d
. [D v
FILE:APPF.RR E1 Appendix F-39 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL
- RELIEF REQUEST NO. 7 ;
O References' l
- 1. NRC NUREG-0019, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, November 1980.
- 2. NRC letter, H.l. Abelson to J.C. Brons, dated July 21,1986, Feedwater Nozzle Cracking in Bolling Water Reactors.
- 3. NYPA letter, J.C. Brons to NRC (JPN-88-010), dated March 25,1988, NUREG-0610, Feadwater Nozzle inspections.
4 4. NRC letter H.l. Abelson to J.C. Brons (TAC 67829), dated September 13,1988, Relief from Augmented Inspection of Feedwater Nozzle /Sparger. ,
- 6. GE report NEDC 32019P (Class Ill), inspection and Monitoring of Feedwater Nozzle, dated May 1992.
- 6. GE report NEDE 30799P, James A. Fitzpatrick Nuclear Power Station Feedwater Nozzle Fracture Mechar.ics Analysis to Show Compliance with NUREG-0619. December 1984.
- 7. Structuralintegrity Report: SIR 94-021, Rev.0, Evaluation of Feedwater Bypass Leakage Monitoring System Results for J.A. Fitzpatrick N.P.P.
O 1
FILE:APPF.RR Ei Appendix F-40 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 7 Attachment 1 I
i Floure 1 Vessel-- Q l-I Norrie #
! /
- Safe End j
,/ y jx y y 3 2D 2A NUREG-0619 EXAMINATION AREA 1, 2A, 'B, 3 FILE:APPF.RR-Ei Appendix F 41 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF RE(iJEST NO. 8
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A. COMPONENT !DENTIFICATION:
Class: 2 Identification of System: 03 - Control Rod Drive System B. EXAMINATION REQUIREMENTS:
ASME Section XI Requirements:
As required by Article IWC 2000, Table IWC 2500-1, Code Category Q H, Components shall be examined and pressure tested as specified in Table IWC-25001. Table IWC-2500-1, Code Category C H, requires the pressure retaining boundary be inspected each inspection period. Relief should be from the 10 year inservice / hydro test and Code Case N-4981 that the inservice test will be done dunng a normally scheduled shutdown from plant operation. The VT 2 wili be conducted during plant shutdowa when the CRD system is normally in operation. All components will be tested, but this does not meet the requirerNnt of an inservice test.
C. REL!EF REQUESTED:
Relief is requested from the following:
Relief is requested from the requirements specified in IWC-2000 Table IWC 25001, Code Category
(
C-H, All Pressure Retaining Components of the ASME Boller and Pressure Vessel Code,Section XI,1989 Edition for inservice precsure testing of the CRD System each inspection period. Also Code Case N-4981 for the third period test. An inspection will be done dur;ng plant shutdown when the CRD system is in operation.
D. BASIS FOR RELIEF:
Portions of the CRD system are not normally in operation except during plant startup or shutdown and thus are not pressurized nor do these portions meet the requirements for an inservice test. Inspection during a plant shut down from power operation ensures all components and piping are in operation.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The piping, valves and components will be inspected during a scheduled plant shut down from power operation during the third inspection period. This will ensure all piping, valves and components shall be tested. Inspection will be via a VT 2 examination. -
F. IMPLEMENTATION SCHEDULE:
The requirements as specified in this relief request will be incorporated into JAF Inservice inspection Program during the 3rd Ten-Year Interval.
G. ATTACHMENTS TO THE RELIEF:
None
'(q) H. USNRC RESPONSE FILEAPPF.RR E1 Appendix F-42 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 9 O
v A. COMPO!JENT IDENTIFICATlON:
Class: MC (Metallic Constrainments)
Identification of System: Primary Containment Structure B. EXAMINATION REQUIREMENTS:
The ASME Section XI 1992 Edition, Article IWE requires that repairs and replacements to the containment be made in accordance with the 1992 Uode Edition requirements. Additionally, the definitions of visual examination methods VT 1 and VT 3, and visual examiner and authorized inspection qualifications are required to comply with Article IWA in the 1992 Edition.
C. RELIEF REQUESTED:
Relief is requested from the requirements of ASME Section XI 1992, Edition Article IWA 2000, as evoked by 10 CFR 50.55a (b)(2)(6).
D. BASIS FOR REllEF:
The current ISI Program for Class 1,2, and 3 components and their supports is based on the e ASME Section XI,1989 Edition without Addenda. This program defines the VT 1 and VT 3 Iq s examinations and includes all factors necessary to qualify the examiners and assure the 4 examinations are performed at the highe'.t quakty levels. The imposition of evoking ASME Section XI,1992 Edition Code requirements for containment would necessitate a separate administrative program including procedures, certification tests and record keeping requirements, which would merely parallel the existing program with no benefit to quality. The existing program is used on the reactor and other Class 1 components, and continues to prove effective.
Maintaining two separate programs for the same activities is a significant burden that has no benefit to quality or safety. Similarly, the repair and replacement program for Class 1,2, and 3 components is based on the 1989 Edition and has proven effective. Use of the 1992 Code for containment would require a separate set of procedures with 1992 references and parallel record keeping. This is a significant burdon with no compensating increase in quality or safety.
Implementing procedures for all applicable portions of Article IWE will be in accordance with the Authority's Quality Assurance Program, which meets 10 CFR 50 Appendix B.
E. AL.TERNATIVE EXAMINATIONS OR TESTS:
The ISI Program at JAF will meet the requirements specified in ASME Section XI 1989 Edition without Addenda, Article IWA 2000, r\ '
V FILE:APPF.RR E1 Appendix F-43 of F 55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO. 9 F. IMPLEMENTATION SCHEDULE:
The requirements as specified in this relief request will be incorporated into JAF Inservice Inspection Program for the duratic.'t of the 3rd Ten Year interva!.
G. ATTACHMENTS TO THE RELIEF:
None H. USNRC RESPONSE NYPE submitted this request for relief under Let6er JPN-97-031, dated 10/06/97 for the Third Inservice Inspectioriinterval.
O l
l l
l l
l FILE:APPF.RR-Ei Appendir F-44 of F-55
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_ _ - -- . . . - .- _- . _ _ = - _ .
JAMES A. FITZPATRICK l THIRD INSERVICE INSPECTION INTERVAL l RELIEF REQUEST NO.10
(
V .
l A. ARTICLE IDENTIFICATION:
IWA 5000 Section IWA-5250(a)(2) ,
l Class: 1,2, and 3 '
System: All B, EXAMINATlON REQUIREMENTS:
IWA 5000, Section IWA-5250(a)(1), (a) The source of leakages detected during the conduct of a system pressure test shall be located and evaluated by the Owner for corrective measures as follows:
(2) if leakage occurs at a bolted connection, the botting shall be removed, VT-3 visually examined for corrosion, and evaluated in accordance with IWA 3100; C. RELIEF REQUESTED:
Relief is requested from the requirements of IWA 5250(a)(2) If leakage occurs at a bolted connection, the bolting shall be removed, VT 3 visually examined for corrosion, and evaluated in accordance with IWA 3100.
D, BASl3 FOR RELIEF:
d Later Codes and Addenda (ASME Section XI,1992 Edition) have acknowledged this requirement as a significant burden that has no benefit to quality or safety and have modified IWA-5250(a)(2) as follows:
if leakage occurs at a bolted connection on othei than a gaseous system, one of the bolts shall be removod. VT 3 examined, and evaluated in accordance w!th IWA 3100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall oe removed, VT 3 examined, and evaluated in accordance with IWA-3100" The degradation of botting due to leakage from borated Jystems not an occurrance that is considered a problem at Bolling Water Reactors (BWRs). Bolting degradation due to wastage Pressure testing at JAF will be conducted in accordance with Relief Request # 3, which evokes the requirements of Code Case N4981. Code Case N4931 requires that the visual examination (VT 2) be performed at nominal operating pressure but states that temperature shall not exceed limiting conditions for the hydrostatic test curve as contained in the plant Technical Specification.. JAF expects a certain amount of leakage during pressure testing from bolted connections (valves, flanges) during RCS pressure testing primarily due to the lower testing temperature. The majority of leakage identhied during testing is from packing leaks but a small percentage is attributed to the CRD Flanges and other pressure retaining bolted connections. Usually this leakage is arrested as the plant heats up or additional torquing is performed to stop the leakage, in those cases where leakage is not arrested based on the above actions an evaluation is performed and, when necessary, corrective measures taken.
L)
. Fil.E:APPF.RR-Ei Appendix F 45 of F 55
--- - = - _ _ _ - - - = . - . _ . - - - - - . - - - - _ - . _ _ - - . _ _ _ _
l l
JAMES A. FITZPATRICK l THIRD INSERVICE INSPECTION INTERVAL l RELIEF REQUEST NO.10 0 Removal and inspection of t,utting without indication of botting degradation, can result in a system or portions of a system being placed in a mode of inoperation, limited condition of operation (LCO), or place the plant in a shutdown condition. This constitutes a significant level of risk and burden on the plant without a compensating benfit to quality or safety.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The ISI Program at JAF will spectfy conducting visual (VT 3) examination in-place on Ltited connections when leakage occurs. Evidence of degradation to the bolted connection or bolting shall require the bolt closest to the source of the leakage to be removed visually examined (VT 3) and evaluated to IWA-3100. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA 3100.
F. IMPLEMENTATION SCHEDULE:
The requirements as specified in this relief request will be incorporated into JAF inservice Inspection Program during the 3rd Ten Year Interval.
G. ATTACHMENTS TO THE RELIEF:
None H, USNRC RESPONSE 3
t i
O FILE:APPF.RR Ei Appendix F 46 of F-55
~. ______ __ _ _ __ ._ _____ _. _ _ . _
JAMES A FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST No.11 b(3 .
A. ARTICLE IDENTIFICATION: IWF 5000 Class: 1,2 and 3 Identification of Components: Snubbers Systems: All B. EXAMINATION REQUIR2MENTS:
ASME B&PV Code,Section XI, Article IWF 50001989 Edition invokes the snubber examination requirements of Standard OMa-1988, Part 4, Section 2.3.2.2 which state _s that examinations shall be conducted at 18-month intervals and specifies schedule changes if unacceptable snubbers are revealed. Section 2.3.2.3 of Standard OMa-1988 requires that subsequent examinations for any given failure group not be lengthened more than one increment at a time.
C. RELIEF REQUESTED:
Relief is requested from performance of visual inspec'Jons of snubbers at 18-month intervals, and the associated schedule changes if unacceptable snubbers are revealed, as required by IWF 5000 which invokes Standard OMa 1988 Part 4, Section 2.3.2.2. Relief from the Subsequent Examination Schedule Adjustment of oms 1988 Section 212.3 is also requested.
D. BASIS FOR RELIEF:
The 18-month snubber visual inspection schedule as it appears in Standard OMa 1988, Part 4, Section 2,3.2.2 assumes that refueling intervals will not exceed 18 months, and is based only on the riumber of ut, acceptable snubbers found during the previous visualinspection,;rrespective of the size of the snubber population. The 18-month inspection intervalis incompatible with current operating evcle lengths of 24 months. Due to the large number of snubbers in use at the Fitzpatrick plant, the OMa schedule and snubber selection method is excessively restrictive and resource intensive.
Performance of these inspections during power operation, as would be necessary under the OMa 18-month inspection interval, would result in expenditures of significant resources and would subject plant personnel to unnecessary radiological exposure with r.o commensurate increase in quality or safety. As concluded by the NRC staff in Generic Letter 90-09, the proposed attemative inspection maintains the same confidence levelin snubber operability. The proposed alternative !s compatible with the current 24-month operating cycle and generally will allow inspections to be performed during plant Utages, thereby reducing radiological exposure of plant personnel.
Relief from Section 2.3.2.3, Subsequent Examination Schedule Adjustment is also requested since the schedule adjustment specified in this Section of the standard is based on the examination intervals of Section 2.3.2.2. of OMa-1988.
,The proposed alternative inspection conforr.s with NRC Generic Letter 90-09 and has been previously approved for use at the Fitzpatrick Nuclear Power Plant by the NRC as t.icense Amendment 180 to the Fitzpatrick Operating License on April 1'),1992.
Application to the Commission for amendment of the Technical Epecifications to conform to the revised ISI Snubber Fiogram was submitted in NYPA Letter JPN-96-05, dated 11/26/96, JAFNNP (7 Docket 50-333, Prc, nosed Changes To The Technical Specifications To Relocate Requirements For V. Snubbers To Plant Controlled Documents ,(JPTS-96-001) 4 FILE:APPF.RR-E1 Appendix F-47 of F-55
t JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL m RELIEF REQUEST NO.11 V
E. ALTERNATIVE EXAMINATIONS OR TESTS:
Examinations of snubbers will be performed at intervals and samphng rates in accordance with the requirements specified in Generic letter 90-09, Attemative Requirements for Snubber Inspection Intervals and Corrective Actions , December 11,1990. This proposed alternative is based upon the number of unacceptable snubbers found during the previous inspection, the total population or category size for each snubber type, and the previous interval Specifically, the visual inspection interval will be determined based upon the following cnteria:
Population Column A> Column B' Column C' Category Extended Interval Repeat Interval Reduce interval 1 0 0 1 80 0 0 2 100 0 1 4 150 0 3 8 200 2 5 13 300 5 12 25 A
The next visualinspection interval for the population of a snubber category shall be determined based upon the previous inspection interval and the number of unacceptable snubbers found during that interval. Snubbers may be categorized, based on their accessability during power operation, as accessible or inaccessible. These categories may be examined separately or jointly. This decision shall be mrado and documented before any inspecbon and used as the basis upon which to determine the next inspection interval for that category, Interpolation between population or category sizes and the number of unacceptable snubbers is permissible. The next lower integer for the value or limit for Columns A, B, C shall be used if that integer includes a fractional value of unacceptable snubbers as determined by interpolation.
If the number of unacceptable snubbers is equal to or less than the number in Column A, the next inspection interval may be twice the previous interval but not greater than 48 months.
If the number of unacceptable snubbers is equal to or less than the number in Column B but greater than the number in Column A, the next inspection Interval shall be the same as the previous interval.
If the number of unacceptable snubbers is equal to or greater than the nt, nber in Column C, the next inspection interval shall be two-thirds of the previous interval. However, if the number of unacceptable snubbers is less than the number in Column C but greater than the number in Column B, the next interval shall be reduced by a factor that is one-third of the ratio of the difference between the number of unacceptable snubbers found during the previous interval and the number in Column B to the difference in the numbers in Columns B and C.
/m N
i FILE:APPF.RR-Ei Appendix F-48 of F 55 l 1
JAMES A FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO.11 The standard 25% extension on surveillance intervals is applicable to any examination interval i determined in accordance with this alternative.
F. IMPLEMENTATION SCHEDULE: ;
The proposed alternative inspection was implemented in 1992 as a result of Amendment 180 to the Fitzpatrick Operating License on April 13,1992 and will be continued in the third 10-year inspection interval.
G. ATTACHMENTS TO THE RELIEF:
~
NRC' Generic Letter 90-09, Alternative Requirements for Snubber lnspection Intervals and Corrective Actions , December 11,1990.
H. USNRC RESPONSE i
G o
()
FILE:APPF.RR E1 Appendix F-49 of F 55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL I RELIEF REQUEST NO.12 I o
)
A. COMPONENT IDENTIFICATION: I l
Class: 2 Identification of System: All Description of Components: Examination Category C-H / Pressure Testing of Coniainment Penetration Piping B. EXAMINATION REQUIREMEN1S:
Table IWC 2500-1, Category 0 H contains the requirements for system Pressure test (IWC 5220) for Class 2 pressure retaining components. The ASME Code requires a system leakage test each inspection period.
C. RELIEF REQUESTED: ,
Pursuant to the provisions of 10 CFR 50.55a(a)(3), relief is requested from the requiremenis specified in Table IWC 25001, Examination Category C-H, of the ASME Boiler and Pressure Vessel Code,Section XI,1989 Edition for pressure testing of piping that penetrates a containment vessel, when the piping and Isolation valves that are part of the containment system are Class 2 but, the balance of the piping system is outside the scope of Section XI.
D. BASIS FOR RELIEF:
The leakage testing requirement results in unusual difficulties without a compensating increase in the O level of quality and safety.
ASME Code Case N-572 recognizes and addresses this fact and proposes an attemative which maintains an acceptable level of quality ara safety.
Draft RG 1.147 dated May 1997, includes Code Case N 522, Pressure Testing of Containment Penetration Piping,Section XI, Division, this Code Case has not been published in Regulatory Guide 1.147, inservice Inspection Code Case Acceptability ASME Section XI, Division 1" Penetrations at JAF are currently tested in accordance with the Appendix J hsting Program.
E. ALTERNATIVE EXAMINATIONS OR TESTS:
The following alternative examination requirements will be implemented as defined by ASME Section XI Code Case N-522, Pressure Testing of Containment Penetration Piping,Section XI, Division 1" In addition to the requirements specified in Codo Case N-522, the following conditions shall be met:
The test should be conducted at the peak calculated containment pressure and the test procedure should permit the detection and location of through-wall leakage in Containment isolation Valves (CIVs) and pipe segments between the CIVs.
,m .
FILE:APPF.RR E1 Appendix F-50 of F-55
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I JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST NO.12 F. IMPLEMENTATION SCHEDULE:
The Alternate Examination requirements of ASME Code Case N 522 will be incorporated into JAF Inservice inspection Program during the 3rd Ten Year Interval.
G. ATTACHMENTS TO THE RELIEF:
ASME Code Case N-522, Pressure Test of Containment Penetration Piping,Section XI, Division 1".
H. USNRC RESPONSE 4
0
.Lf3)
FILE:APPF.RR-Ei Appendix F-51 of F 55 e l
.__ .. . . _ . _ . - ~ . - - - - - - . _ - - - - . . . - - - -- . . _ . .
CASE l N-522 CASES OF ASME DOILEN AND PRES $tlRE VESSEL CODE Approval Date: December 9,1993 See NumericalIndex tot espiration and any tesfhrmetson dates, ;
Csse N 522 4
Pressure Testing of Containment Penetration i
Piping i Section XI. Division I
~
Inquiry What alternative to the rules of Table IWCf 25001, Category C.H may be used for pressure testing ,
piping that penetrates a containment vessel, when the piping and isolation valves that are part of the contain-ment system are Class 2 but the balance of the piping system is optside the scope of Section XI?
Reply: It is the opinion of the Committee that 10 CFR 50, Appendix J, may be used as an alternative to t-#
rules in Table IWC 2500,1, Category C II, for pressure testing piping that penetrates a containment vessel, when the piping and isolation valves that are part of the con.
tainment system are Class 2 but the balance of the pip-
' -,, ing system is outside the scope of Section XI?
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O 1029
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l l
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL g RELIEF REQUEST NO.13 b
A, COMPONENT IDENTIFICATION:
Clats: All Identification of System: All B. CODE REQUIREMENTS:
Article IWA-4000 welding and brazing procedure qualification requirements.
(a) All welding shall be performed in accordance with Welding Procedures Specifications that have been qualified by the Owner or repair organization in accordance with the requirements of the codes specified in the Repair P ogram in accordance with IWA 4120.
C. RELIEF REQUESTED:
Relief is requested from the requirements of ASME Sec%n XI, Articia IWA-4000, IWA-4400.
BASIS FOR RELIEF:
The basis for this relief is to implement ASME Code Case N-573, which eliminates the redundancy currently required by the Code for each organization to independently quality all welding procedures even though they have met the qualification process at another facility. Code f- Case N 573 recognizes and addresses this fact and promses an alternative which maintains an acceptable level of quality and safety.
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E. ALTERNATIVE EXAMINATIONS OR TESTS:
The following alternative testing requirements will be implemented as defined by ASME Section XI Code Case N-573, Transfer of Procedure Qualification Records Between Owners,Section XI, Division 1.
- 2. The supplying Owner will state in writing that the POR was performed under an acceptable Nuclear Quality Assurance program that meets ASME Section XI, IWA-1400 and that it was performed in accordance with ASME Section IX.
NYPA POR's may supplement these or other Owner supplied PQR's.
- 4. The WPS will be approved and signed by NYPA.
- 5. The WPS will be demonstrated successfully by NYPA by completing a welder performance qualification test using the parameters of the NYPA WPS.
- 6. NYPA will not transfer the supplied POR to any other Owner.
- 7. NYPA will document the use of this Code Case on the appropriate NIS-2 form.
V FILE:APPF.RR-Ei Appendix F 53 of F-55
JAMES A. FITZPATRICK THIRD INSERVICE INSPECTION INTERVAL REL!EF REQUEST NO.13 F. IMPLEMENTATION SCHEDULE:
The Alternatt, Testing requirements of ASME Code C sse N 573 will be incorporated into the JAF Inservice Inspection Program during the 3rd Ten Year Interval.
G. ATTACHMENTS TO THE RELIEF:
ASME Code Case N 573, Transfer orocedure Qual #ation Records Between Owners,Section XI, Division 1 Pressure Test of Containment Penetration Piping,Section XI, Division 1.
H. USNRC RESPONSE 7
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,- a l FILE:APPF.RR Ei Appendix F-54 of F 55
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M A N-573 1
t' Ares 00P AshtE 80tIAS AND PRENUEE M. CODE
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Appsevel Deer: Merah it 1997 l
See NumencolInerou & enrotion and any reeMnnation afstes. {
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! Case N.373 Assurance Progr'am that satis 6es the require:nents of Transfer ed Procedure Queliaceties Records IWA 1400.
Between Owoore (c) The Owner accepting the completed PQR shall l
, Sessies XI, Diviales I accept responsibility for obtaining any additional sup.
parting informauon needed for WPS development.
Inquiry What altetandves to the welding and bra. (d) The Owner accepdag the completed PQR shall lag procedure quali6 cation requirements of IWA 4000 document, on each resulting WPS, the parameters appil-may be used? cable to welding. Each WPS shall be supported by all necessary PQR's.
Aeply It is the opinion of the Committee that as (a) h Ownw accepdng the canpleW M M an ahernative to the welding and brazing procedure accept re8Ponsibility for the PQR. Acceptance shall be
, quali6 cation requirements of IWA-4000, a procedure
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- pq i
quah6 cation record (PQR) quali6ed by one Owner may be used by raother Owner. When this ahernative (f) The Owner accepting the completed PQR shall demonstrate technical competence in applicadon of the is used, the following requirements shall be met:
received PQR by completing a performance quali6 cation (a) The Owner that perfwmed the procedure qualifi-test using the parameters of a resulting WPS.
canon test shall certify, by signing the PQR, that testing i O)
% was perfanned in accordance with Section IX.
(g) The Owner may accept and use a PQR only when it is received duectly from the Owner that certined (b) The Owner that performed the procedure qualin- the PQR. >
cation test shall certify, in writing, that the procedure (h) Use of this Case shall be shown on the NIS 2 quali6 cation was conducted in accordance with a Quality form documenting welding or brazing, e
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.b James A. Fitzpatrick Nuclear Power Plant NYPA4AF-68-0002 g g inservice Inspection Program Summary Tables Revisiors: 0 y Date: January 22,1998 INSERVICE INSPECTION BOUNDARY CLASSIFICATION DIAGRAMS DIAGRAM NUMBER SYSTEM IDENTIFICATION REVISION DIAGRAM TITLE NUMBER NUMBER ISI-FB-10H 66 6 Flow Diagram Reactor Building Service Water Cooling ISI-FB-35E 70 8 Control Room Area-Service and Ch.11ed Water ISI-FM-15A 15 6 Reactor Build!ng Cooing Water ISI-FM-15B 15 7 Reactor Building Cooling Water ISI-FM-17A 20 4 Radwaste System ISI-FM-188 27 4 Containment Purge /C. oyste iSI-FM-18C 27 5 Pass Cooling Water Supply ISI-FM-18D 27 3 Containment Hydrogen & Oxygen Sampling System ISI-FM-19A 19 9 Fuel Pool Coohng (FPC)
ISI-FM ~"'A 14 7 Core Spray (CS)
ISI-FM-21A 11 6 Standby Liquid Control ISI-FM-22A 13 8 Reactor Core Isolation Cochng (RCIC)
ISI-FM-24A 12 6 ReactorWaterCleanup RWC)
ISI-FM-20A 10 9 Residual Heat Removal , RHR)
ISI-FM-20B 10 7 Residual Heat Removal (RHR)
ISI-FM-25A 23 9 High Pressure Coolant Injectron (HPCI)
ISI-FM-26A 03 8 Control Rod Dnve (CRD)
FILE: APP-GE1 WPo Appendix G - 1 of G - 2
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N) Q (I James A. Fitzpatrick Nuclear Power Plant NYPA4F4SMcC2 g y Inservice Inspection Program Summary Tables Revisiort. O y ,_L;g- Date: January 22.199s INSERVICE INSPECTION BOUNDARY CLASSIFICATION DIAGRAMS DIAGRAM NUMBER SYSTEM IDENTIFICATION REVISION DIAGRAM TITLE NUMBER NUMBER ISI-FM-278 02-2 8 ReactorWater Recirculabon (RC)
ISl-FM-29A 29 8 Main Steam (MS)
ISI-FM-39C 39 4 Instrument Air ISI-FM-46A 46 9 Service Water (SW)
ISI-FM468 46 & 15 9 Emergency Service Water (F.SWi ISI-FM-47A 02-3 7 Nuclear Boiler Vessel Inst umentabor. (NBVil ISI-FM-49A 16-1 5 Drywa!!/ Torus Leak Rate Arm' -
ISI-FM-93A 93 7 Ernergency Diesel Generator Fuel Otl and Combuston Air Systems ISI-FM-93C 93 2 Emergency Diesel Generator and Lubricabng Systems IS!-FM-94A 93 3 Emergency Diese! Generator Air Start-up Unes FILE: APP-GE1 wm Appendix G-2 of G-2
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