Text

Third Inservice Inspection Interval
	ML18018B052
Person / Time
Site:	Nine Mile Point
Issue date:	09/27/1999
From:	; Niagara Mohawk Power Corp
To:	; Office of Nuclear Reactor Regulation
References
	Download: ML18018B052 (532)
	v • d • e

V NIAGARA 4 MOHAWK Niagara Mohawk Power Corporation 300 Erie Boulevard West Syracuse, New York 13252 Nine Mile Point Nuclear Power Station THIRD INSERVICE INSPECTION INTERVAL INSERV!CE INSPECTION PROGRAM PLAN Prepared For Nine Mile Point Nuclear Power Station P.O. Box 63 Lycoming, New York 13093 Commercial Service Date: December 26, 1969 NRC Docket Number: 50-220 Document Number: NMP1-ISI-003 Revision Number: 0 Date: September 27, 1999 Prepared by:

NMP1 Sl Program Manager Approved by: ~r lfP'lenn R. Perkins - Supervisor, ASME Section XI Programs Approved by:

Willi Ye r a

/ H'1 er, E gineering Services Reviewed by:

x-z~ H Au orized Nuclear In rvice ector ocrr ]3sg lilt<)

()n PIC rl8 Srrqo l'Ib

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 El Y NIAGARA lIkl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE Of CONTENTS Table of Contents .. I-2 Record of Revision . i-5 Abbreviations . i-6 Glossary of Terms i-9 Abstract ..... i-12 Section 1.0 introduction 1-1 thru 1-11 Section 2.0 ASME Code Class 1 Systems/Components . 2-1 thru 2-16 Section 3.0 ASME Code Class 2 Systems/Components . 3-1 thru 3-10 Section 4.0 ASME Code Class 3 Systems/Components . 4-1 thru 4-4 C

Section 5.0 ASME Code Class 1, 2, and 3 Component Supports 5-1 thru 5-6 Section 6.0 Augmented Examinations .. ~....... 6-1 thru 6-35 Section 7.0 Requests For Relief 7-1 thru 7-4 Section 8.0 Acceptance Criteria .. 8-1 thru 8-12 Section 9.0 ASME Repairs and Replacements 9-1 thru 9-9 Section 10.0 Records .. ... 10-1 thru 10-14 LIST Of FIGURES Figure 6-3 Feedwater Nozzle Zones ................................... 6-19 Figure 10-1 Form NIS-1 Owner's Data Report for lnservice Inspection .......... 10-6 Figure 10-2 Form NIS-2 Owner's Report for Repairs or Replacements .............. 10-8 Figure 10-3 OAR-1 Owner's Activity Report . . 10-10 Figure 10-4 Form NIS-2A Repair/Replacement Plan Certification Record .. . 10-14 File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN LIST OF TABLES Table 1-2 NMP1 Inservice inspection Periods 1-4 Table 1-6 Inspection Program B 1-10 Table 2-1 NMP1 Valve Groupings . 2-13 Table 5-1 Category F-A Selection Process . 5-4 Table 5-2 Snubber Visual Inspection Frequency 5-6 Table 6-1 IGSCC Category A Weld Selections . 6-7 Table 6-2 IGSCC Category D Weld Selections 6-8 Table 6-3 IGSCC Category F Weld Selections .. 6-12 Table 6-4 IGSCC Category G Weld Selections 6-12 Table 6-5 IGSCC Category S Weld Selections . 6-13 Table 6-6 IGSCC Examination Requirements .. 6-14 Table 6-7 Routine Inspection Intetvals Refueling Cycles ...... 6-17 Table 6-8 Feedwater Nozzle/Sparger Inspection recommendations .. .. 6-17 Table 6-9 Augmented Feedwater Nozzle Examination ...... 6-18 Table 6-10 In-Vessel Augmented Examinations . .. 6-20 Table 6-11 Instrument Nozzles .. 6-22 Table 6-12 Instrument Nozzles .. 6-23 Table 6-13 Instrument Nozzles .. 6-23 Table 6-14 IRM Incore Dry Tube Assembly .. 6-24 Table 6-15 SIL-455 Selected Welds ... 6-25 Table 6-16 SIL-474 Steam Dryer .. 6-26 Table 6-17 RPV Head Cladding .. .. 6-27 File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN LIST OF TABLES - Continued Table 6-18 Upper Core Grid . 6-27 Table 6-19 SIL-588 Upper Core Grid 6-28 Table 6-20 SIL-459 Pump Shafts .. 6-29 Table 6-21 Core Spray Sparger . 6-29 Table 6-22 IfB 96-03 Suction Strainers . 6-30 Table 6-23 CRDH Upper J Weld and Roll Area Listing . 6-31 Table 6-24 Tee Base material . . 6-34 Table 8-1 Class 1 Acceptance Standards . 8-11 Table 8-2 Class 2 Acceptance Standards . 8-11 Table 8-3 Class 3 Acceptance Standards . 8-12 Table 8-4 Class 1, 2, 3 Component Supports Acceptance Standards .. 8-12 Table 8-5 Class MC Acceptance Standards 8-12 APPENDICES Appendix A Class 1 Summary Tables 1 thru 33 Appendix B Class 2 Summary Tables 34 thru 46 Appendix C Class 3 Summary Tables . . 47thru48 Appendix D Class 1, 2 and 3 Component Support Summary Tables ... 49thru58 Appendix E Code Boundary Diagram Listing E-1 thru E-3 Appendix F Relief Requests ISI-1 thru ISI-12 F-1 thru F-4 File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station N MP1-ISI-003 Unit 1 El V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION REV. PROGRAM DATE OF REVISION AFFECTED REASON FOR No. SECTIONS PAGES CHANGE Section 0 September 27, 1999 page i-1 thru i-11 Initial Issue Section 1 September 27, 1999 page 1-1 thru 1-11 Initial Issue Section 2 September 27, 1999 page 2-1 thru 2-16 Initial Issue Section 3 September 27, 1999 page 3-1 thru 3-10 Initial Issue Section 4 September 27, 1999 page 4-1 thru 4-4 Initial Issue Section 5 September 27, 1999 page 5-1 thru 5-6 Initial Issue Section 6 September 27, 1999 page 6-1 thru 6-34 Initial Issue Section 7 September 27, 1999 page 7-1 thru 7-4 Initial Issue Section 8 September 27, 1999 page 8-1 thru 8-12 Initial Issue 9 September 27, 1999 page 9-1 thru 9-9 Initial Issue

'ection Section 10 September 27, 1999 page 10-1 thru 10-14 Initial Issue Appendix A September 27, 1999 page A-1 thru A-3 Initial Issue October 26, 1999 Table 1 thru 33 0 Appendix B September 27, 1999 page B-1 thru B-3 Initial Issue October 26, 1999 Table 34 thru 46 Appendix C September 27, 1999 page C-1 thru C3 Initial Issue October 26, 1999 Table 47 thru 48 0 Appendix D September 27, 1999 page D-1 thru D-3 Initial Issue October 26, 1999 Table 49 thru 58 Appendix E September 27, 1999 page E-1 thru E-3 Initial Issue Appendix F September 27, 1999 page F-1 thru F-3 Initial Issue RR 1 thru RR 12 Fiie: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 4V NIAGARA MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ABBREVIATIONS Listed below are the abbreviations utilized in this document:

ANII Authorized Nuclear Inservice Inspector ANSI American Nuclear Standard Institute ASME American Society of Mechanical Engineers B&PV Boiler & Pressure Vessel Code BC Branch Connection BWR Boiling Water Reactor BWROG Boiling Water Reactor Owner's Group CFR Code of Federal Regulations CRC Corrosion Resistant Cladding CRD Contr'ol Rod Drive System CRA Code Required Surface Area CRS Core Spray System CT Condensate Transfer CTN-SP Containment Spray System CRV Code Required Volume CU Reactor Water Clean-Up System ECS Emergency Cooling System DPI Drywell Inerting CAD and Purge System FSAR Final Safety Analysis Report FWS Feedwater System FPS Spent Fuel Pool Filtering and Cooling System GE General Electric File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V N1AGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ABBREVIATIONS(Continued)

GL Generic Letter HPCI High Pressure Coolant Injection System IEB Inspection and Enforcement Bulletin (USNRC)

IEN Inspection and Enforcement Notice (USNRC)

IHSI Induction Heat Stress Improvement ISI Inservice Inspection IVVI In-Vessel Visual Inspections LPS Liquid Poison System MSS Main Steam System Magnetic Particle Examination N/A Not Applicable NBVI Nuclear Boiler Vessel Instrumentation NDE Nondestructive Examination NMP1 Nine Mile Point Nuclear Power Station Unit1 NPS Nominal Pipe Size NMPC Niagara Mohawk Power Corporation NSSS Nuclear Steam Supply System NWT Nominal Wall Thickness OD Outside Diameter PAID Piping and Instrumentation Diagram PT Liquid Penetrant Examination RBCLC Reactor Building Closed Loop Cooling System Regulatory Guide (USNRC)

File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ABBREVIATIONS(Continued)

RHR Residual Heat Removal System RHSI Resistant Heat Stress Improvement RICSIL Rapid Information Communication Services Information Letter RR Reactor Recirculation System RG Regulatory Guide (NRC)

RPV Reactor Pressure Vessel RWC Reactor Water Cleanup System RXVI Reactor Vessel Instrumentation SD Structural Discontinuity SDC Shutdown Cooling Water System SIL Services Information Letter SRP Standard Review Plan (USNRC)

SWS Service Water System SURF Surface Examination Sl Stress Improvement T.S. Technical Specifications TE Terminal End UFSAR Updated Final Safety Analysis Report UT Ultrasonic Examination USNRC United States Nuclear Regulatory Commission VOL Volumetric Examination Visual Examination (suffix number denotes type of exam, (VT-1, VT-2, VT-3))

WinISI Computerized Inservice Inspection Data Base Management Software File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI Y NIAGARA H 4 MOHAWK THIRD INSERVICE INSPECT(ON INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN GLOSSARY OF TERMS ASSESS - to determine by evaluation of data compared with previously obtained data such as operating data or design specifications.

AUTHORIZED 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 INSPECTOR - an employee of an authorized Inspection Agency who has been qualified in accordance with NCA-5000 of Section III.

COMPONENT - an item in a nuclear power plant such as a vessel, pump, valve or piping system.

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.

CONSTANT LOAD TYPE SUPPORT - spring type support that produces a relatively constant supporting force throughout a specified deflection CORE SUPPORT STRUCTURES - those structures or parts of structures that are designed to provide direct support or restraint of the core (fuel and blanket assemblies) within the reactor pressure vessel 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.

Defect - a flaw (imperfection or unintentional discontinuity) of such size, shape, orientation, location, or properties as to be reject able DISCONTINUITY - a lack of continuity or cohesion; an interruption in the normal physical structure of material or a product 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.

File: NMP1PPO.WPD

0 Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit1 V NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN GLOSSARY OF TERMS - Continued EXAMlNATIONCATEGORY - a grouping of items to be examined or tested.

FLAW - an imperfection or unintentional discontinuity that is detectable by nondestructive examination HANGER - an item that carries the weight of components or piping from above with the supporting members being mainly in tension IMPERFECTION - a condition of being imperfect; a departure of a quality characteristic from its intended condition INDICATION- the response or evidence from the application of a nondestructive examination INSERVICE EXAMINATION- the process of visual, surface, or volumetric examination performed in accordance with the rules and requirements of Section XI.

INSERVICE INSPECTION - methods and actions for assuring the structural and pressure-retaining integrity of safety-related nuclear power plant components in accordance with the rules of Section XI.

INSPECTION - verification of the performance of examinations and tests by an Inspector.

INSPECTION PROGRAM- the plan and schedule for performing examinations or tests.Section XI Inservice Inspection Program Plan -A term used to address the programmatic requirements as required by 10 CFR 50.55a and USNRC Guidelines, dated 1981.

Inservice Inspection Plan and Schedule - A term used to address the ASME Section XI, Appendix F, Inspection Plan.

Long-Term Ten-Year Inspection Plan - A term used to address the entire data base printout (WinISI), that includes all items subject to examination.

Inservice Inspection Period Plan - A term used to address those items scheduled for an inspection period (Period 1, 2 or 3 as applicable), as required by Inspection Program "B".

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 Xl.

ITEM - a material, part, appurtenance, piping sub-assembly, component or component support.

File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN GLOSSARY OF TERMS - Continued MAINTENANCE - routine servicing or work undertaken to correct, adjust or prevent an abnormal or unsatisfactory condition.

NONDESTRUCTIVE EXAMINATION- an examination by the visual, surface or volumetric method.

OPERATIONAL READINESS - The ability of a component or system to perform its intended function when required.

OPEN ENDED - a condition of piping or lines that permits free discharge to atmospheric or containment atmosphere OWNER - the organization legally responsible for the operation, maintenance, safety and power generation of the nuclear power plant.

RELEVANT CONDITION - a condition observed during a visual examination that requires supplemental examination, corrective measure, repair, replacement, or analytical evaluation RECORDABLE INDICATION- an indication which equals or exceeds the recording criteria REGULATORY AUTHORITY- a federal government 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 components; (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, valve bodies, pump casings, pipe fittings (such as tees, elbows, reducers, flanges, etc. conforming to ANSI B16.9) and pipe branch connections and fittings.

TERMINALENDS - the extremities of piping runs that connect structures, components, or pipe anchors, each of which acts as a rigid restraint or provides at least 2 degrees of restraint to piping thermal expansion.

VARIABLE SPRING TYPE SUPPORT - a spring type support providing a variable supporting force throughout a specified deflection 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.

Fiie: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA V MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ABSTRACT This document describes the Updated Inservice Inspection Program for the Third Ten-Year Inservice Inspection Interval for the Nine Mile Point Nuclear Power Station, Unit 1.

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, 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 of 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 Nine Mile Point Nuclear Power Station, Unit 1, Third Inservice Inspection Interval Program Plan and Schedule is the 1989 Edition, with no Addenda of Section XI, hereafter referred to as the Code.

File: NMP1PPO.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-1 of 1-11 TABLE OF CONTENTS SECTION 1 - INTRODUCTION Table of Contents 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 Code of Federal Regulations 1-5 ASME Code Editions and Addenda . 1-5 USNRC Regulatory Guides 1-5 NMP1 Specific Documents 1-5 USNRC NUREGS/SRP's 1-6 USNRC Bulletins ............... 1-6 USNRC Generic Letters 1-6 USNRC Informational Notices ..... 1-6 ASME Code Cases . 1-7 1.4 Applicable Code Editions and Addenda 1-7 1.4.1 Third Inspection Interval . 1-7 1.4.2 Subsequent Code Edition and Addenda 1-7 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 1-8 1.5.6 Optional Construction 1-9 1.5.7 Piping Penetrating Containment . 1-9 1.5.8 Classification Diagrams 1-9 File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y N1AGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN Pa e 1-2 of 1-11 TABLE OF CONTENTS (Continued)

SECTION 1 - INTRODUCTION 1.6 Inspection Program B 1-9 1.6.1 Class 1, 2 and 3 Components 1-9 1.6.2 Component Supports 1-10 1.7 Development of Inspection Program Plan 1-10 1.8 Substitute Examinations 1.9 Exclusions/Exceptions 1.9.1 Examination Category B-J . 1-11 1.9.2 IWE Inspection Program 1-11 1.9.3 Additional Programs . 1-11 LIST OF TABLES TABLE 1-2 NMP1 Inservice Inspection Periods 1-4 TABLE 1-6 Inspection Program B 1-10 File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NlAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-3 of 1-11

,'";;REV!BIO'M,:'.'

September 27, 1999 Entire Updated Inservice Inspection Program Plan Document for the 3" Ten Year Inservice Inspection Interval File; NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 N U MOHAWK Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN Pa e 1Q of 1-11

1.0 INTRODUCTION

This document details the basis and plans for the Inservice Inspection Program for the Third Ten-Year Inservice Inspection Interval for components, welds, supports, bolting, pump casings, valve bodies, and reactor pressure vessel internals for the Nine Mile Point Nuclear Power Station, Unit 1.

Niagara Mohawk Power Corporation (NMPC), is the Owner of Record.

The Commercial Service Date for the Nine Mile Point Nuclear Power Station, Unit 1 is December 26, 1969.

1.1 Inspection Interval The Third Inservice Inspection Interval becomes effective on December 26, 1999 and is scheduled to end on December 25, 2009.

1.2 Inspection Periods The Third Inservice Inspection Interval is divided into three successive inspection periods as determined by calendar years of plant service within the inspection interval. Identified below are the period dates for the third inspection interval as defined by Inspection Program II B II 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 NMP1's plant outages.

TABLE 1-2 NMP1 INSERVICE INSPECTION PERIODS INSPECTION PERIOD START PERIOD END REFUEL REFUEL PERIODS DATES DATES OUTAGE OUTAGE YEAR December 26, 1999 December 25, 2002 RFO-16 2001 December 25, 2002 December 26, 2006 RFO-17 2003 RFO-18 2005 December 26, 2006 December 25, 2009 RFO-19 2007

The plant operating license is currently scheduled to expire on August 22, 2009.

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 developed after giving due consideration to the following documents and subject to the limitations and modiTications listed in 10 CFR 50.55a(b), and to the extent practical within the limitations of design, geometry and materials of construction. Specific areas within this document where these document are used File: NMP IPP1.WP0

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-5 of 1-11 in the preparation of this program are addressed within the area affected.

Code of Federal Regulations 10 CFR 50.55(a) Code of Federal Regulations; Federal Register, Volume 64, Number 183, dated September 22, 1999, amendment to the regulation.

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 Addenda, "Rules for Inservice Inspection of Nuclear Power Plant Components",

Subsections IWE and IWL.

USNRC Regulatory Guides The following list of Regulatory Guides are applicable to the Nine Mile Point Nuclear Power Station 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.

1.65 Materials and Inspections for Reactor Vessel Closure Studs, dated October 1973, Regulatory Position C.4.b.

1.84 Design and Fabrication Code Case Acceptability ASME Section III, Division 1, Latest Revision.

1.85 Material Code Case Acceptability ASME Section III, Division 1, Latest Revision.

1.147 Inservice Inspection Code Case Acceptability ASME Section XI, Division 1, Revision 12..

NMP1 Specific Documents Nine Mile Point Unit 1 Updated Final Safety Analysis Report, Sections 1, 5, 7, 9, 12 and 16.

Nine Mile Point Unit 1 Technical Specifications, USNRC Docket number 50-220, Sections 3.6 and 4.6.

File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-IS 1403 Uni) 1 LI Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-6 of 1-11 USNRC NUREGS/SRP's USNRC NUREG 0313, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2.

USNRC NUREG 0619, BWR Feedwater Nozzle and CRD Return Lines.

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 Generic 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.

81-11 Feedwater and Control Rod Drive Nozzle Cracking, 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.

94-03 Intergranular Stress Corrosion (IGSCC) Cracking of Core Shroud in Boiling Water Reactors, July 25, 1994.

98-05 Boiling Water Reactor Licensees Use of the BWRVIP 05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, December 10, 1998.

Boiling Water Reactor Vessel Inspection Program (BWRVIP) References or Commitments Requirements of the complete BWRVIP implementation criteria for the Third Ten Year Inservice Inspection Interval is currently under development. Upon completion of this process the applicable information will be incorporated within ISI Program. General Information pertaining to BWRVIPs implementation is contained in the ISI Program NMP1-ISI-003, Section 6.0 Titled Augmented Examinations.

USNRC Informational Notices 89-79 Degraded Coatings and Corrosion of Steel Containment Vessels, December 1 1989.

~

File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-7 of 1-11 ASME Code Cases Code Cases approved through Regulatory Guide 1.147 may be proposed for revision to the inspection plan. Specific Code Cases used in the preparation of this document are identified below.

N-416-1 Alternative Pressure Test Requirements for Welded Repairs or Installation of Replacement Items by Welding, Class 1, 2 and 3,Section XI, Division 1. RG 1.147,Rev. 12 N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds,Section XI, Division 1. RG 1.147, Rev. 12 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.

RG 1.147, Rev. 12 N-509 Alternative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded Attachments,Section XI, Division 1. RG 1.147, Rev. 12, Subject to conditions.

N-524 Alternative Examination Requirements for Longitudinal Welds in Class 1 and 2 Piping Section XI, Division 1. RG 1.147, Rev. 12.

N-526 Alternative Requirements for Successive Inspection of Class 1 and 2 Vessels,Section XI, Division 1. Subject to Request for Relief ISI-4.

N-532 Alternative Requirements to Repair and Replacement Documents Requirements and Inservice Summary Report Preparation and Submission As Required by IWA-4000 and IWA-5000,Section XI, Division 1. Subject to Request for Relief ISI-8.

N-573 Transfer of Procedure Qualification Records Between Owners,Section XI, Division 1. Subject to Request for Relief ISI-10.

Note: See Appendix D "Code Cases" of the 10 Year Inspection Plan for a complete listing and copy of applicable Code Cases..

1.4 Applicable 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 system pressure testing for the Third Inservice Inspection Interval are based on the rules set forth in the 1989 Edition of Section XI, that was endorsed twelve months prior to the start of the Third Inspection Interval.

File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-8 of -11 1.4.2 Subsequent Code Editions and Addenda As permitted by 10 CFR 50.55a(g)(4)(iv), the NMPC 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 all related requirements to the respective Editions and Addenda are met. NMPC intends to 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 systems.

1.5 System Quality Group Classifications System safety classifications, 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 Nine Mile Point Nuclear Power Station, Unit 1, Updated Final Safety Analysis Report (UFSAR).

Water, steam and radioactive containing components (other than turbines and condensers) are designated Quality Group A, B, 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 NMP1 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 NMP1 FSAR, and apply to those components of the Reactor Coolant System not classified as Quality Group A, (ASME Code Class 1), and that are safety-related.

1.5.3 Quality Group C (ASME Code Class 3)

Quality Group C system boundaries were developed based on Regulatory Guide 1.26 and the NMP1 FSAR, and apply to those components that are not classified as Quality Group A or B, (ASME 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, and as such are not included within this document. Exception, Reactor Water Cleanup System welds located outside the containment isolation valves. See Section 6 Augmented Examinations.

File: NMP1PPi.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V'IAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 N U MOHAWK Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN Pa e 1-9 of 1-11 1.5.5 Application Application of the rules of Section XI are governed 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 higher than the minimum class established in the component 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.

1.5.7 Piping Penetrating Containment The portions of piping that 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 Nine Mile Point Nuclear Power Station, are designated on various ASME Section XI Boundary Diagrams (P8I D). These designations identify the system class breaks by color coding.

The rules of IWB, IWC and IWD were applied to these drawings in order to determine those components/systems subject to examination/test. 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 ASME Section XI Boundary diagrams (P8i Diagrams), to this program. Copies of these diagrams are available through the NMPC drawing control system.

1.6 Inspection Program B The Nine Mile Point Nuclear Power Station inspection intervals comply with IWA-2432, Inspection Program B. With the exceptions of the examinations identified in 1.6.1, the required examinations in each examination category shall be completed in accordance with Table 1-6.

1.6.1 Class 1, 2 and 3 Components The required examinations in each Examination Category shall be completed during each inservice inspection interval, in accordance with IWB-2412-1, IWC-2412-1, and IWD-2412-1, with the following exceptions:

File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 T NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-10 of 1-11 (1) Examination Categories B-N-1, B-P, B-Q, C-H, and the system pressure test requirements of D-A, D-8 and D-C; (2) the percentage required by Note 2 of Examination Category 8-D; (3) the examinations that may be deferred to the end of an inspection interval, as specified in Table IW8-2500-1.

(4) If there are less than three items to be examined in an Examination 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.

(5) Within various Code Categories the total number of items scheduled for examination exceeds Inspection Program "8" requirements. Adjustments to those Code Categories which exceed the allowable Program "8" percentages, may be reduced to meet Program "8" requirements.

1.6.2 Component Supports The required examinations shall be completed in accordance with the inspection schedule established for the components under IWB, IWC, and IWD.

TABL'E 1-6 INSPECTION PROGRAM:B Inspection Inspection Period Minimum Maximum Interval Calendar Years of Examination Examination Plant Service Within Completed, % Credited, %

the Interval Note. 1 2 NMP1 3'6. 16% 34%

Inservice Inspection 50% 67%

Interval 10 1PP 100%

Note: 1 Exce t as noted in Table IW8-2500-1, 81.30.

Note: (1) The examination of shell<oAange welds may be performed during the first and third inspection periods in conjunction with the nozzle examinations of examination category B-D. At least 50% of shell-to-flange welds, shall be examined by the end of the First Inspection Period, and the remainder by the end of the Third Inspection Interval.

(2) At kit 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 last inspection period, of the respective inspection interval. (See paragraph 1.6.1 (2)).

1.7 Development of Inspection Program Plan Sections 2 through 6 detail the narrative description of the Nine Mile Point Unit 1 Third lnsefvice Inspection Program Plan basis for Quality Group A, 8 or C, (ASME Code Class 1, File: NMPIPPI.WPD

Nine Mile Point Nuclear Power Station NMP1-IS I-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Pa e 1-1 of 1-11 2 and 3), (including their Supports and Augmented Examinations), of components and/or systems subject to examination/test.

1.8 Substitute Examinations NMPC 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 access, 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 required, 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.

1.9 Exclusions/Exceptions This paragraph defines the exclusions/exceptions, NMPC has taken due to the unit being Docketed prior to June 1978.

1.9.1 Examination Category B-J All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the following limits under loads associated with specific seismic events and operational conditions:

(1) Primary plus secondary stress intensity range of 2.4Sm for ferritic and stainless steel (2) cumulative usage factor U of 0.4 1.9.2 IWE Inspection Program The Inservice Inspection Program for Class MC, Subsection IWE and IWL is not addressed within this program, with the exceptions of Repairs and Replacements as defined in Section 9.

1.9.3 Additional Programs In addition to the above items, the following Programs are outside the scope of this document. They are addressed in separate documents.

Snubber Examination and Testing Program Repair and Replacement Program Inservice Pump and Valve Test Program Metal Containment Examination Program System Pressure Test Program Containment Pressure Test Program File: NMP1PP1.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN P 2-1 f 2-1 6 TABLE OF CONTENTS SECTION 2 - CLASS 1 SYSTEMS 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 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 -15 2.2.14 Category B-0 2 -16 2.2.15 Category B-P 2 -16 2.2.16 Category B-Q 2 -16 2.2.17 Successive Inspections . 2 -16 File: NMP1PP2.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NIAGARA HU MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-

-'. -.*RBCORDOF.REVISION

REVISION "DATE;,; ".:,,

"AFF,ECTED REASON;FOR REVISION',

'. 'No." '~'",~PAGES':

0 September 27, 1999 'ntire Updated Inservice Inspection Program Plan Document for the 3"'en Year Inservice Inspection Interval 0

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAI Rey. P Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN 2.0 CLASS 1 SYSTEMS/COMPONENTS The ASME Code Class 1 system boundaries subject to examination and testing were developed based upon the requirements of 10 CFR 50.2(v) and Nine Mile Point Unit 1 (NMP1), Final Safety Analysis Report (FSAR). The ASME Code Class 1 components and systems (including their supports) subject to examination and testing are described in detail 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 pressure 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 from on-site emergency power.

(b) 1. Piping of 1" nominal pipe size and smaller; and

2. Components and their connections in piping of 1" nominal pipe size and smaller.

(c) Reactor vessel head connections and associated piping, 2" nominal pipe size and smaller, made inaccessible by control rod drive 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 Retaining Welds in Reactor Vessel All examinations are performed from the inside and/or outside surface using manual/automated inspection equipment, (as applicable) and volumetric examination techniques.

Noe NMPC submitted and received authorization to utilize an "Alternative for Examination of Reactor Pressure Vessel Shell Welds, (TAC No. MA4383). As authorized by the USNRC the use of PDI qualified personnel and procedures results in a more sensitive examination and will provide added assurance for flaw detection and sizing, and is an acceptable alternative to the requirements of the 1989 Edition of Section XI Code and Regulatory Guide 1.150. The error band for flaw sizing has been established within the limits of ASME Section XI, Appendix VIII.

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 El V NIAGARA l1 U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. 0 Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN B B 2- e Scope of Examination - includes volumetric examination of essentially 100% of all longitudinal and circumferential shell weld lengths. (does not include shell to flange weld). Pursuant to USNRC Generic Letter 98-05 and BWRVIP-05, NMPC applied for and was granted relief from performing examinations of the circumferential weld till the end of the NMP1 operating license. (August 22, 2009)

(4) B1.11 Circumferential shell welds, none required (12) B1.12 Longitudinal shell welds, (12) required Subject to Request for Relief: ISI-1 Granted under USNRC TAC No. MA4768, and ISI-4.

B121 B Scope of Examination - includes volumetric examination of essentially 100% of accessible length of all circumferential and meridional head welds.

(2) B1.21 Circumferential head weids, (2) required (14) B1.22 Meridional head welds, (14) required B B Scope of Examination - includes volumetric examination of essentially 100% of accessible length of all circumferential and meridional head welds.

(1) B1.21 Circumferential head welds, (1) required (8) B1.22 Meridional head welds, (8) required Subject to Request for Relief: ISI-2 B

Scope of Examination - Volumetric examination of 100% of the shell to flange weld.

(1) B1.30 Circumferential shell to flange weld, (1) required Subject to Request for Relief: ISI-2 and ISIS.

Haia2: If partial examinations are conducted from the flange face, the remaining volumetric examinations required to be conducted from the vessel wall may be performed at or near the end of each inspection interval.

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 File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NlAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-Period, and the remainder by the end of the Third Inspection Period.

B Scope of Examination - includes volumetric and surface examination of essentially 100% of the reactor vessel head to flange weld length.

(1) B1.40 Circumferential head to flange weld, (1) required, 1/3 Each Period Subject to Request for Relief: ISI-2 Not applicable to Nine Mile Point Nuclear Power Station 2.2.2 Category B-B, Pressure Retaining welds in vessels other than Reactor Vessels.

This Examination Category is not applicable to Nine Mile Point Nuclear Power Station.

2.2.3 Category B-D, Full Penetration Welds of Nozzle in Vessels (Program B)

Scope of Examination - Volumetric examination of 100% of all nozzles with full penetration welds to vessel shell (or head) and integrally cast nozzles.

(40) B3.90 RPV Nozzle Welds, (40) required (40) B3.100 RPV Nozzle Inner Radius, (40) required Subject to Request for Relief: ISI-3 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, Generic Letter 81-11, ILE Bulletin 80-13, revision 1, Supplement 1 and GE-NE 523-A71-0594 are addressed in Section 6, Augmented Examinations.

Were practical Code credit will be taken during these Augmented examination.

Pr B Not applicable to Nine Mile Point Nuclear Power Station.

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 V NIAGARA kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. Q INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Not applicable to Nine Mile Point Nuclear Power Station.

Not applicable to Nine Mile Point Nuclear Power Station.

2.2.4 Category B-E, Pressure Retaining Partial Penetration Welds in Vessels B - V v N No Scope of Examination - Visual VT-2 examination on 25% of all partial penetration welds each interval. The examinations are performed during the System Pressure Testing. These examinations and tests are addressed in the Nine Mile Point Unit 1 Inservice Pressure Test Program, Document NMP1-PT-003..

(0) B4.11 Vessel nozzles, not applicable to NMP1 (129) B4.12 Control rod drive nozzles, (32) required (64) B4.13 Instrumentation nozzles, (16) required 2.2.5 Category B-F, Pressure Retaining Dissimilar Metal Welds Scope of Examination - Volumetric and surface or surface examinations are required of all dissimilar metal safe end welds in each loop and connecting branch of the Reactor Coolant System. For the reactor vessel nozzle dissimilar metal safe end welds, the examination may be performed coincident with the vessel nozzle examinations required by Examination Category B-D.

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, Supplement 1. See Section 6.0 of this Program for details. Completed examinations shall be used to satisfy the percentage requirements of Inspection Program "B" and NUREG 0313, as applicable.

(33) B5.10 NPS 4" or Larger Nozzle-to-Safe end butt welds, (33) req'd (4) B5.20 Less than NPS 4" Nozzle-to-Safe end butt welds,(4) req'd (N/A) B5.30 Nozzle-to-Safe end socket welds, Not applicable to NMP1 Augmented examination requirements of NUREG-0313, Rev. 2, NRC Generic Letter 88-01, Sup. 1, BWRVIP-06, 27, and 49, SIL 455, Rev. 1, Sup. 1, SIL 571, and RICSIL 072 are discussed in Section 6 of this document.

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NlAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. Q INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN z 8 Not applicable to Nine Mile Point Nuclear Power Station.

o 0 Not applicable to Nine Mile Point Nuclear Power Station.

8 2 Not applicable to Nine Mile Point Nuclear Power Station.

-D 4 II Scope of Examination - Volumetric and surface or surface examinations are required of all dissimilar metal safe end welds in each loop and connecting branch of the Reactor Coolant System.

Examination Category B-F welds are scheduled and examined as part of the IGSCC Augmented Inspection Program. The extent and frequency of examinations

're in accordance with NUREG 0313, Revision 2 and Generic Letter GL 88-01, Supplement 1. See Section 6.0 of this Program for details. Completed examinations shall be used to satisfy both Inspection Program "B" and NUREG 0313 requirements.

(N/A) 85.130 ~ 4" NPS dissimilar butt welds, Not applicable to NMP1 (N/A) 85.140 < 4" NPS dissimilar butt welds, Not applicable to NMP1 (N/A) 85.150 dissimilar socket welds, Not applicable to NMP1 2.2.6 Category 8-G Pressure Retaining Bolting, Greater Than 2 in. In Diameter ea 8 8 .4 Scope of Examination - Examination includes all bolts, studs, nuts, bushings, and 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-B, B-J, B-L-2, and B-M-2.

Examinations consist of visual exams of reactor vessel closure head nuts, volumetric exams of RPV studs in place, and surface exams of RPV studs when removed. Regulatory Guide 1.65, regulatory position C.4.b will be invoked by NMPC to examine a representative sample of a minimum of (12) studs on a geometric distribution of (3)studs within each 90 degree segment. A volumetric examination of the threads in the base material of the reactor vessel flange will be conducted only when the connections are disassembled. A Visual (VT-1)

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H llMOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN P e 2- of 2-1 examination of the RV washers and bushings may be examined in place, when the connection is disassembled or when the bolting is removed. Flange surfaces, when the connection is disassembled, include 1 inch annular surface of flange surrounding each stud.

(64) B6.10 Nuts, (64) required, 1/3 each period (64) B6.20 Studs in Place, (64) required, 1/3 each period (64) B6.30 Studs, when removed, (12) min. required, (4) each period (64) B6.40 Threads in Ligaments, (64) required, 1/3 each period (128) B6.50 Washers, (128) required, 1/3 each period (64) B6.50 Bushings, (64) required, 1/3 each period Subject to Request for Relief: ISI-11 Pr B 70 B6 Not applicable to Nine Mile Point Nuclear Power Station.

B690 B61 B Not applicable to Nine Mile Point Nuclear Power Station.

m B612 B 3 B Not applicable to Nine Mile Point Nuclear Power Station.

pi I B 16 B Not applicable to Nine Mile Point Nuclear Power Station.

B B6 200 Scope of Examination - All bolts, studs, nuts, bushings, and flange surfaces.

Examinations applicable to five (5) Reactor Recirculation Pumps 32-185, 32-186, 32-1 87, 32-1 88 and 32-1 89.

~No 5. 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 examined only when the connections are disassembled. Bushings may be examined in place.

Flange surface requires 1 inch annular surface of flange surrounding each stud hole.

(80) B6.180 Studs, 16 studs per pump, (16) one pump required (5) B6.190 Flange surfaces, (1) per pump, one pump required (240) B6.200 Nuts, Bushings, and Washers, (48) per pump, one pump File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. P INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-required Scope of Examination - All bolts, studs, nuts, bushings, and flange surfaces.

Examinations applicable to Feedwater, Core Spray and Shutdown Cooling systems and are limited to valve selected under Category B-M-2.

(5) B6.210 Valve Bolts, (3) required (5) B6.220 Valve Flanges, (3) required (9) B6.230 Valve Nuts, Bushings, washers, (5) required 2.2.7 Category B-G-2, Pressure Retaining Bolting, 2 in. And Less in Diameter B B B 7 B Scope of Examination - Visual VT-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.

(18) B7.10 Reactor Pressure Vessel, (18) required (N/A) B7.20 Pressurizer, Not applicable to NMP1.

(N/A) B7.30 Steam Generator, Not applicable to NMP1.

(N/A) B7.40 Heat Exchanger, Not applicable to NMP1.

(12) B7.50 Piping Flange Bolting, (12) required (80) B7.60 Pump, (5) pumps 16 cap screws per pump, one pump required (75) B7.70 Valves, (33) required (129) B7.80 CRD Housing, when disassembled Haia%. Augmented examination requirements of SIL 419 and SIL 483 Revision 2 are discussed in Section 6 of this program.

2.2.8 Category B-H, Integral Attachments for Vessels 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, "Alternative Rules for the Selection and Examination of Class 1, 2 and 3 Integrally Welded AttachmentsSection XI, Division 1". Code Examination Categories and Item Numbers are as denoted in the Code Case.

See Examination Category B-K File: NMP1PP2.WPD

Nine Mlle Point Nuclear Power Station NMP1-Isl-003 Unit 1 T NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. Q INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-B Not applicable to Nine Mile Point Nuclear Power Station.

B Not applicable to Nine Mile Point Nuclear Power Station.

2.2.9 Category B-J, Pressure Retaining Welds in Piping B B B Scope of Examination-All dissimilarmetal 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.

HaIIJI Table IWB-2500-1, Examination Categoly B-J, Footnote (1)(b)(1) and (2) are not applicable to NMP1, with the exception of the Reactor Recirculation system, due to the unit being docketed prior to June 1978.

All augmented Main Steam and Feedwater System welds, to the extent practical shall be used to satisfy the percentage requirements of Inspection Program "B" and the augmented requirements of NUREG 0313, Generic Letter 88-01, Supplement 1, I8 E Bulletin 80-1 3, SIL 289 and INPO SER 5-85 shall also be used for satisfying the percentage requirements of Inspection Program "8", to the extent practical. See Section 6.0 Augmented Examinations of this Program for details.

All longitudinal pipe welds intersecting any of the selected circumferential welds will also be examined. As an alternate to Table IWB-2500-1, 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:

(A) When only a surface examination is required, examination of longitudinal 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 of longitudinal piping welds is not required beyond those portions of the 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, examination requirements shall be met for both transverse and parallel flaws at the intersection of the welds and for that length of File: NMP1PP2.WPD

e Nine Mile Point Nuclear Power Station N MP1-ISI-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2- 1 f 2-1 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 and parallel flaws within the entire examination volume of intersecting circumferential welds.

(345) B9.11 Circumferential welds, (86) required (288) B9.12 Longitudinal welds, No minimum required (164) B9.21 Circumferential welds, (41) required (None) B9.22 Longitudinal welds, Not applicable to NMP1 (9) B9.31 Branch Conn. NPS 4" or Larger, (2) required (26) B9.32 Branch Conn. Less than NPS 4", (7) required (69) B9.40 Socket welds, (17) required (613) Nonexempt Welds subject to examination, (153) required, (1 54) scheduled (288) Long. Welds subject to examination, (46) selected See Appendix A Tables for selection details.

2.2.10 Category B-K, Integral Attachments to Piping, Pumps 8 Valves

'B B1 2 B um V achmen Scope of Examination - Volumetric or Surface examination to include essentially 100% of the length of the attachment weld at each integrally welded attachment subject to examination. Integral attachments selected for examination will consist of attachments whose base material design thickness is equal to 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 and valve integral to such piping.

Examirlations 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". Examination Category and Item Numbers shall be as defined within the Code Case.

~No e 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.

(6) B10.10 Vessel Integral Attachments, (1) required (155) B10.20 Piping Integral Attachments, (16) required (0) B10.30 Pump Integral Attachments, Not applicable to NMP1 (8) B10.40 Valve Integral Attachments, (1) required Fiie: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA Il4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2- 2 f 2-1 Subject to Request for Relief: ISI-5 2.2.11 Category B-L-1, Pressure Retaining Welds in Pump Casings, B-L-2, Pump Casings

'B $

Scope of Examination -100% volumetric examination of all welds in one Pump.

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 Nine Mile Point Nuclear Power Station Unit 1. The five (5)

Reactor Recirculation Pumps do not have casing welds.

B Scope of Examination - Visual examination of the interior surfaces of one of the five (5) Reactor Recirculation Pumps when disassembled for maintenance. Pump to be identified when pump is disassembled.

(5 ) B12.20 Recirc. Pumps (1) Pump Required 2.2.12 Category B-M-1, Pressure Retaining Welds in Valve Bodies, B-M-2, Valve Bodies 2 Valv B Not applicable to Nine Mile Point Nuclear Power Station Unit 1 . Valves less than NPS 4 do not have any valve body welds.

B24 V v Bo P "o La Scope of Examination - Volumetric examination to include essentially 100% of weld length. Examinations are limited to at least one valve within each group of valves that are the same size, constructional design, and manufacturing method, and perform similar functions in the system.

(6) Valve Body Welds, (1) required 8 V v Bod I 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.

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-Isl-003 Unit 1 El V NIAGARA H Ll MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-The Table below list all valves grouped by system, size, and type subject to examination under B-M-2.

.Tabie:,2-1 "'...

VALVE;GROUPINGS .

GRP SYS " SIZE', TYPE' '.G"f " ~

'-M-2 " '-M-1""';;;","',",;""',SELECT

'O..

"?;.:, VALVE

'D,:

V01 01.0 24.0'LOBE N/A 01-01-VB 01-01-VBY N/A 1 Valve 01-02-VB 01-02-V BY among a 01-03-VB 01-03-VBY group of 01-04-VB 01-04-VBY valves V02 01.0 6.0'ATE N/A 01-07-VB 01-07-VBY N/A 1 Valve 01-08-VB 01-08-VBY among a 01-09-VB 01-09-VBY group of 01-10-VB 01-10-VBY valves 01-11-VB 01-11-V BY 01-12-VB 01-12-VBY V03 01.0 6.0'elief N/A 01-102-A-SV8 01-102-A-SVBY 01-102-A-WD-001 1 Valve 01-102-B-SVB 01-102-B-SVBY 01-102-B-WD-001 among a Ot-102-C-SVB 01-102-C-SVBY 01-102-C-WD-001 group of 01-102-D-SVB 01-102-D-SVBY 01-102-D-WD-001 valves 01-102-E-SVB 01-102-E-SVBY 01-102-E-WD-001 01-102-F-SVB 01-102-F-SVBY 01-102-F-WD-001 V04 31.0 18.0'heck 31-01R-VB 31-01R-VBY N/A 1 Valve 31-02R-VB 31-02R-V BY among a group of vatvfts VOS 31.0 18.0'ate N/A 31-07-VB 31-07-VBY 1 Valve 31-08-VB 31-08-VBY among a group of valves V06 32.0 28.0'ate 32-380-VB 32480-VBY 1 Valve 32-381-VB 32.381-V BY among a 32-382-VB 32482-VBY group of 32-383-VB 32-383-VBY valves 32484-VB 32-384-VBY 32-375-VB 32-375-VBY 32-376-VB 32476-VBY 32-377-VB 32-377-VBY 32-378-VB 32-378-VBY 32479-VB 32-379-VBY V07 33.0 6.0'ate N/A 33-01R-VB 33-01R-VBY N/A 1 Valve 33-02R-VB 33-02R-V BY among a group of valves V08 38.0 14.0 Gate N/A 38-01-VB 38-01-VBY 1 Valve 38-02-VB 38-02-VBY among a 38-13-VB 38-13-VBY group of valves File: NMP1PP2.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. P INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-

, '"',Tabfe2-1 VALVEr GROUPINGS GRP SYS'." "SlZE TYPE 'BG '-';, '8-8-2 SELECT

'O. lD VALVE '.

V09 39.0 10.0'lobe N/A 39-05-VB 39-05-VBY N/A 1 Valve 39-06-VB 39-06-VBY among a group of valves V10 38.0 14.0'heck 38-12-VB N/A 38-12-VBY 1 Valve among a group of valves V11 40.0 12.0'heck 40-03-VB N/A 40.03-VBY 1 Valve 40-13-VB 40-13-VBY among a group of valves V12 39.0 N/A 39-03-VB 39-03-VBY N/A 1 Valve 39-04-VB 39-04-VBY among a group of 10.0'3.0 valves V13 Check N/A 33-03-VB 12.0'heck 33-04-VB 33.03-VBY 33.04-VBY N/A 1 Valve among a 6.0'0.0 group of valves V14 Gate 40-01-VB 40-01-VBY N/A 1 Valve 40-02-VB 40-02-VBY among a 40-09-VB 40-09-VBY group of 40-10-VB 40-10-VBY valves 40-11-VB 40-11-VBY 40-12-VB 40-1 2-VBY V15 39.0 N/A 39-01R-VB 39-01R-V BY N/A 1 Valve 39.02R-VB 39-02R-VBY among a 10.0'0.0 group of valves V16 6.0'ate Gate N/A 40-05-VB 40-06-VB 40-05-VBY 40-06-VBY N/A 1 Valve among a 6.0'0.0 group of valves V17 Safety N/A CH-576-12A-B V-BK-01-1 19A 1 Valve Relief CH-576-12B-B V-BK-01-119B among a CH-576-12C-B V-BK-01-1 19C group of CH.576-12D-B V-BK-01-119 D valves CH-576-12F-B V-BK-01-119F CH.576-12G-B V.BK-01-119G CH-576-12H.B V-BK-01-119H CH-576-12J-B V-BK-01-119J File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 Y NIAGARA kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. Q INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-

'i'Table'i2>>I .':.i'i'I. I VALVE.GROUPINGS GRP Na.

BYS

'StZE

lo...','," '

'TYPE VALVE,,

'B-G 1 B-'6-2 .'B.M.1:" SELECT V18 39.0 10.0'ate N/A 39-07R-VB 39-07R-VBY 1 Valve 39-08R-VB 39-08R-VBY among a 39-09R-VB 39-09R-VBY group of 39-10R-VB 39-10R-VBY valves 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, Removable Core Support Structures.

Augmented IVVI examinations are addressed in Section 6 of this Program.

0 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.

(3) Accessible Areas, once each period

'B 0 Reactor Vessel (BWR)

Scope of Examination - Visual VT-1 examination of accessible welds of interior attachments within the Beltline region (once per interval).

(5) Interior Attachments, (5) required c e Scope of Examination - Visual VT-3 examination of accessible welds of interior attachments beyond the Beltline region (once per interval).

(N/A) Interior Attachments, Not applicable to NMP1 0 c u Scope of Examination - Visual VT-3 examination of accessible surfaces of core support structures (once per interval).

(49) Accessible surfaces, (49) required B 37 I Filo: NMP1PP2.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rey. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 2-Reactor Vessel (PWR)

Not applicable to Nine Mile Point Nuclear Power Station Unit 1.

Augmented examination requirements are addressed in Section 6 of this Program.

2.2.14 Category B-O, Pressure Retaining Welds ln Control Rod Housings

'B W D Scope of Examination - Volumetric or surface examination of 10% of the peripheral CRD housings.

(129) CRD Housings, (32) Peripheral CRD Housings, (3) required (10%)

Subject to Request for Relief: ISI-6 2.2.15 Category B-P, All Pressure Retaining Components Scope of Examination - System pressure tests are conducted on All Class 1 systems and components in accordance with the Nine Mile Point Unit 1 System Pressure Testing Program, Document NMP1-PT-003.

2.2.16 Category B-Q, Steam Generator Tubing Not applicable to Nine Mile Point Nuclear Power Station Unit 1.

2.2.17 Successive Inspections The sequence of component examinations established during the first inspection interval was repeated during the third inspection intelval, to the extent practical.

In accordance with IWB-2420(b), several welds examined during the second inspection interval were evaluated in accordance with IWB-3142.4, and were determined by analysis to qualify as acceptable for continued service. The areas containing these indications shall require reexamination during the third inspection interval. Applicable welds are uniquely identified within the inservice inspection plan Tables.

File: NMP1PP2.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 3 - CLASS 2 SYSTEMS Table of Contents .. 3-1 Record of Revision 3-2 3.0 CLASS 2 SYSTEMS/COMPONENTS .. 3-3 3.1 ASME Code Exemptions 3-3 3.1.1 IWC-1221 3-3 3.1.2 IWC-1222 3-3 3.1.3 IWC-1 230 . 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 3-7 3.2.5 Category C-F-1 3-7 3.2.6 Category C-F-2 3-8 3.2.7 Category C-6 .. 3-9 3.2.8 Category C-H .. 3-10 3.3 Successive Inspections 3-10 File: NMP1PP3.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit1 Y NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

"";y4: ,.>>>>',:"'.;";;;,',";"P";; .'>>,.;.; ':;RECORD'.OF.:REVISION '",i:."':l

,REVISION;"-, K>>.,',", ':"'"DATE

.:-",": REASON FOR'REVISION~;-;, :-

0 September 27, 1999 Entire Updated Inservice Inspection Program Plan Document for the 3" Ten Year Inservice Inspection Interval ~

File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 3.0 CLASS 2 SYSTEMS/COMPONENTS The Class 2 System Boundaries were developed based upon the requirements of Regulatory Guide 1.26 and the NMP1 FSAR.

The Class 2 components and systems (including supports) subject to examination and testing are described in detail below:

3.1 ASME Code Exem'ptions 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 CHR 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 pressurized water reactor plants.

, (b) Vessels, piping, pumps, valves, and other components NPS 1 ~/e and smaller in high pressure safety injection systems of pressurized water reactor plants.

(c) Component connections NPS 4 and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size in all systems except high pressure safety injection systems of pressurized water reactor plants.

(d) Component connections NPS 1 y2 and smaller (including nozzies, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size in high pressure safety injection systems of pressurized water reactor plants.

(e) Vessels, piping, pumps, valves, other components, and component connections of any size in statically pressurized, passive (i.e., no pumps) safety injection systems of pressurized water reactor plants.

(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.

3.1.2 IWC-1222- Components within systems (or portions of systems) other than RHR, ECC and CHR Systems (a) Vessels, piping, pumps, valves, and other components NPS 4 and smaller.

File: NMP1PP3.WPD

Nine Mlle Point Nuclear Power Station N M P1-IS I-003 Unit 1 hl Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN (b) Component connections NPS 4 and smaller (including nozzles, socket fittings, and other connections) in vessels, piping, pumps, valves, and other components of any size.

(c) Vessels, piping, pumps, valves, other components, and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200 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 described in detail below:

3.2.1 Category C-A, Pressure Retaining Welds in Pressure Vessels Scope of Examination: 100% of all welds at gross structural discontinuities only.

The examinations are limited to one vessel among a group of vessels.

Not applicable to Nine Mile Point Nuclear Power Station Unit 1 ie 1 Scope of examination: 100% of head-to-shell welds, (limited to,one vessel of multiple vessels)..

Not applicable to Nine Mile Point Nuclear Power Station Unit 1 o Wid Scope of examination: 100% of Tubesheet to shell welds (limited to one vessel of multiple vessels). Components applicable to this examination category are the (4)

Emergency Condenser Heat Exchanger' (111, 112, 121, and 122), and (4)

Reactor Containment Spray Heat Exchanger's (111, 112, 121 and 122).

File: NMP1PP3.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI V NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 C

INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN (16) C1.30 Shell Circ. Welds, (4) welds required 3.2.2 Category C-B, Pressure Retaining Nozzle Welds In Vessels oz V o a Components applicable to (4) Reactor Containment Spray Heat Exchanger's (111, 112, 121, and 122).

.(8) C2.11 Nozzle to Shell Welds, (2) required I 22 - ozzl o ifo Ves

~ches Components applicable to (4) Emergency Condenser Heat Exchanger's (111, 112, 121, and 122).

e 221- zz Il Scope of Examination - All nozzles at terminal ends of piping runs (limited to one vessel of multiple vessels). Includes only those piping runs selected for

. examination under Examination Category C-F.

(8) C2.21 Nozzle to Shell or Head Welds, (2) required I C2 22- No adi e o Scope of Examination - All nozzles at terminal ends of piping runs (limited to one vessel of multiple vessels).

Not applicable to Nine Mile Point Nuclear Power Station Unit 1

~o~e The Emergency Condenser Heat Exchanger Nozzles do not have inner radius sections.

I C2 -R pc e ed o ozzl dV s Scope of Examination - All nozzles at terminal ends of piping runs (limited to one vessel of multiple vessels).

Not applicable to Nine Mile Point Nuclear Power Station.

2 2 No o II Wel i o Ve essibl '

Not applicable to Nine Mile Point Nuclear Power Station.

C2 - ozzie o o e W o V File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - Visual VT-2 of tell tale hole in reinforcing plates (limited to one vessel of multiple vessels). Examination performed in accordance with system pressure test program.

Not applicable to Nine Mile Point Nuclear Power Station.

3.2.3 Category C-C, Integral Attachments for Vessels, Piping, Pumps & Valves 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 Attachments Section Xl, Division 1".Examination Category and Item Numbers shall be as defined in the Code Case.

Haia 2. 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.

le u esses e r I d chmen

. Scope of Examination - 100% of the length of the attachment weld of only one integrally welded attachment of only one of the multiple vessels selected.

(8) C3.10 Integral Welded Attachments, (1) required I r Scope of Examination - 100% of the length of the attachment weld of 10% of the welded attachments 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.

(1031) Integrally welded attachments, (103) required

-Pu raIl W I 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 Nine Mile Point Nuclear Power Station.

em -Va ves In e rail Welde c e 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).

File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NlAGARA M U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Not applicable to Nine Mile Point Nuclear Power Station.

3.2.4 Category C-D, Pressure Retaining Bolting > 2" in Diameter 10Pr V Scope of Examination - 100% bolts and studs at each bolted connection of components required to be inspected. The examination of bolting for vessels may be performed on one vessel in a group of vessels.

Excluded per IWC-2500-1 size <2.0" Dia.

ems 4 OC4 & 4 Not applicable to Nine Mile Point Nuclear Power Station.

3.2.5 Category C-F-1, Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping m 1 2 40 C54 &C 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 < NPS 4 apply to PWR high pressure safety injection systems in accordance with the exemption criteria of I WC-1220.

(2) The welds selected for examination 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 pipe to pipe welds, not exempted by IWC-1220, and are not required to be nondestructively examined per Examination Categoiy C-F-1). These welds, however, were included in the total weld count to which the 7.5% sampling rate was applied). The total welds selectedfor examination is based on adding non-exempt circumferential welds to excluded circumferential welds and multiplying by 7.5%.

(72) C5.11 Circ. Welds, (28) required (0) C5.41 Circ. Welds, Not applicable to NMP1 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 non-exempt welds, then 30% of File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 (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.

(4) ~ The welds selected for examination shall be reexamined during subsequent inspections over the service lifetime of the piping component.

3.2.6 Category C-F-2, Pressure Retaining Welds in Carbon or Low Alloy Steel Piping 82 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 < NPS 4 apply to PWR high pressure safety injection systems in accordance with the exemption criteria of IWC-1 220.

(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%

sampling rate is applied).

(721) C5.51, Circ. Welds, (16) C5.52, Long. Welds, (0) C5.61, Circ. Welds, (0) C5.62, Long welds, (0) C5.70, Socket Welds, (38) C5.81, Circ. Welds, (0) C5.82, Long Welds, Non-Exempt Welds (759), Includes Excluded Welds (759) x 7.5% = (56.9) or (57) minimum required, (66) scheduled File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN The examinations shall be distributed as follows:

(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 required by Examination Category C-F-2 should be performed on that system);

(b) Within a system, the examination 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.

(4) The welds selected for examination shall be reexamined during subsequent inspection intervals over the service lifetime of the piping component.

(5) Only those welds showing reportable preservice transverse indications need to be examined for transverse reflectors.

Refer to Appendix B for a complete summary of welds selected and distributed per each system for examination.

3.2.7 Category C-G, Pressure Retaining Welds in Pumps and Valves s C6108 C6 Scope of Examination - 100% of welds in all components in each piping run examined under Examination Category C-F. This Category is applicable to (4)

Reactor Containment Spray Pumps (111, 112, 121, and 122), and (4) Reactor Core Spray Pumps (111, 112, 121 and 122). In the case of multiple pumps and valves of similar design, size, function, and service, the examination of only one pump and one valve among each group of multiple pumps and valves is required. The examination may be performed from either the inside or outside surface of the component.

(80) C6.10 Pump Casing Welds, (20) required (0) C6.20 Valve Casing Welds, Not applicable to NMP1.

File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Subject to Request for Relief ISI-7 3.2.8 Category C-H, All Pressure Retaining Components 0 C72 0 7 0 Scope of Examination - System pressure tests are conducted on all Class 2 systems and components in accordance with the Nine Mile Point Unit 1 System Pressure Testing Program, Document NMP1-PT-003.

3.3 SUCCESSIVE INSPECTIONS The sequence of component examinations established during the first inspection interval will be repeated during the third inspection intervai, to the extent practical.

File: NMP1PP3.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 4- CLASS 3 SYSTEMS Table of Contents 4-1 Record of Revision 4-2 4.0 CLASS 3 SYSTEMS/COMPONENTS 4-3 4.1 ASME Code Exemptions 4-3 4.1.1 IWD-1220.1 4-3 4.1.2 IWD-1220.2 4-3 4.2 Component/System Examination Development 4-3 4.2.1 Category D-A 4-3 4.2.2 System Pressure Tests - Class 3 4-4 File: NMP1PP4.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H V NIAGARA H U MOHAWK THIRD INSERVICEINSPECTIONINTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

,'; .RECORD OF REVISION REVISION;; " -,::,-.-;, DATE'; ".,': AFFECTED'

"~-"'i'!PAGES REASON',,FOR;REVISION 0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year Inservice Inspection Interval File: NMP1PP4.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unlt1 Ll V NIAGARA 1% U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 NMP1 FSAR. The Class 3 components and 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 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 Summary 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 1: Examination Categories and Examination Item Numbers for Class 3 Integrally Welded Attachments are defined in accordance with the ASME Code Case N-509 classification criteria.

4.2.1 Category D-A Integral Attachments for Class 3 Vessels, Piping, Pumps ssur V Scope of Examination - Perform Visual (VT-1) examination of 100% of the weld length of all integrally welded attachment required each interval. Applicable to Emergency Condenser Heat Exchanger's (111, 112, 121 and 122), Reactor Building Closed Loop Cooling Heat Exchanger's (3), Shutdown Cooling Water Heat File: NMP1PP4.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA I% U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Exchanger's (11, 12, and 13), Spent Fuel Pool Cooling Heat Exchanger's (11, and 12).

D1.10 (28) Integral Attachments, (9) required e D Scope of Examination - Perform Visual (VT-1) examination of 100% of the weld length each inspection interval (459) D1.20 Integral Attachments, (46) required.

D I W Scope of Examination - Perform Visual (VT-1) examination of 100% of the weld length of all required integrally welded attachments each inspection interval.

(0) D1.30 Integral Attachments, not applicable to NMP1.

D -V v W de Scope of Examination - Perform Visual (VT-1) examination of 100% of the weld length on all integrally welded attachments required for the inspection interval.

(0) D1.40 Integral Attachments, not applicable to NMP1 Note 2: 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 of integrally welded attachments for each item in each Code Class per interval will be examined.

Note 3: Integral Attachment to Shock Absorbers and snubbers are included as part of the population for which the 10% sample is taken.

4.2.2 System Pressure Tests - Class 3 The pressure retaining components within the boundary of each system specified for Examination Categories D-A, D-B and D-C are pressure tested and visually examined (VT-2), for leakage in accordance with the Nine Mile Point Nuclear Power Station Unit 1 System Pressure Test Program, document NMP1-PT-003.

File: NMP1PP4.WP 0

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unlt1 hl V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 5 - COMPONENT SUPPORTS Table of Contents . ~ ~ ~ ~ ~ ~ ~ ~ ~ ~ 5 1 Record of Revision . 5-2 5.0 CLASS 1, 2 AND 3 COMPONENT SUPPORTS - IWF 5-3 5.1 Supports Exempt From Examination 5-3 5.2 Support Examination Development 5-3 5.2.1 Class 1 Component Supports 5-3 5.2.2 Class 2 Component Supports 5-3 I

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 Fi!e: NMP1PP-5.WPD

Nine Mile Point Nuclear Power Station NMP1-IS I-003 Unit 1 El Y NIAGARA 1% U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD'OF REVISION

REVISION DATE '. REASON FOR REVISION: -""+;, '-;.-':,"

,NO,"

AFFECTED'AGES September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year Inservice Inspection Interval File: NMP1PP-5.WPD

Nirie Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 5.0 CLASS 1, 2 AND 3 COMPONENT SUPPORTS - IWF Component supports selected for examination shall be the supports of those components that are required to be examined under IWB, IWC and IWD. NMPC will conduct examinations in accordance with alternate examination requirements of 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", as noted below:

As an alternate to the Class 1, 2 and 3 Component Support Requirements of Table IWF-2500-1, NMPC 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'. NMPC 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 examination are identified in Appendix D.

5.2.2 Class 2 Component Supports Class 2 component 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.

The Class 1, 2 and 3 Summary Tables satisfy the requirements of IWA-2420 (a) (1) through (6) respectively.

File: NMP1PP-S.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 Type, to the extent practical '.

Table 5-1 CATEGORY F<<A SELECTION PROCESS Exam . ASME Code Applicable System Type of Supports Item No. Class F1.10 Code Class 1 44.2 Control Rod Drive a. Rod Hangers 39.0 Emergency Condenser b. Anchors 31.0 Feedwater C. Springs 42.1 Liquid Poison d. Stanchion s F1.20 Code Class 2 01,02,03.0 Main Steam e. Sway Braces 33.0 Reactor Clean Up f. Dead Weight 40.0 Reactor CoreSpray g. Snubbers F1.30 Code Class 3 32.0 Reactor Recirculation 38.0 Reactor Shutdown Cooling 80.0 Reactor Containment Spray 54.0 Spent Fuel Pool Cooling F1.40 Other than piping 72.0 Service Water 70.0 Reactor Building Close Loop Cooling 5.3.1 Examination Category F-A Supports Scope of Examination - Visual VT-3 examination of 25% of all non-exempt Class 1 Supports.

(190) F1.10 Supports, (48) required AII 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: NMP1PP-S.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - Visual VT-3 examination of 15% of all non-exempt Class 2 Supports.

(433) F1.20) Supports, (65) required 3

Scope of Examination - Visual VT-3 examination of 10% of all non-exempt Class 3 Supports (503) F1.30 Supports, (50) required Scope of Examination - Visual VT-3 examination of 100% of all non-exempt Supports, other than piping supports. This item is applicable to the Emergency Condenser Heat Exchanger's (111, 112, 121, and 122); Emergency Service Water Pump (11 and 12);

Reactor Building Closed Loop Cooling Heat Exchanger, Pump and MU Tank; Reactor Containment Spray Heat Exchanger's (111, 112, 121 and 122); Reactor Containment Spray and Raw Water Pumps (111, 112, 121 and 122); Reactor Core Spray Pumps 111,121 and 122) and Reactor Core Spray Topping Pumps (111, 112, 121 and 122); Reactor Recirculation Pumps (11, 12, 13, 14 and 15); Reactor Vessel Supports; Shutdown Cooling Water Heat Exchanger, (11, 12, and 13); Spent Fuel Pool Cooling Filter (11 and 12); Spent Fuel Pool Cooling Heat Exchanger and Pumps (11 and 12) and the Spent Fuel Pool Cooling Surge Tank. For multiple components, only one of the multiple components are required to be examined.

(89) F1.40 Supports, (26) required Note 1: The recording of Hot or Cold positions will be performed in conjunction with the VT-3 examination .

5A Snubber Examination and Performance Testing Program The following section provides a description of Nine Mile Point Unit 1 Snubber Program for Examination and Performance Testing of Dynamic Restraints (Snubbers).

The Snubber Program is currently defined in plant technical specification, amendment 142, section 4.6.4, and applies to the visual inspection and periodic testing requirements for shock suppressors (snubbers), by assuring the snubbers perform their intended function.

Exclusions from the inspection program are those snubbers that are installed on non safety-related systems and then only if their failure or failure of the system on which they are installed, would have no adverse effect on the safety-related system.

Scope of Examination - Snubbers are categorized into two types (mechanical and hydraulic), and File: NMP1PP-S.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA a UmoHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN classified as accessible or inaccessible. Although snubbers are listed within the ISI Program as part of the total population for which the percentage of Code Case N-491-1 were taken, no snubbers are scheduled and no Code credit is being taken for snubber examinations within this program.

Snubbers inspections are performed by the plant maintenance organization in accordance with Plant Technical Specification and are visually (VT-3) inspected in accordance with the following schedule.

Table 5-2 VISUAL INSPECTION FREQUENCY Number of snubbers found inoperable Next required inspection interval during inspection or-during inspection...,.

interval Refueling Period 12 months *25%

6 months *25%

3,4 124 days ~ 25%

5,6,7 62 days *25%

8 or more 31 days ~ 25%

The required inspection interval shall not be lengthened more than one step at a time.

Scope of Testing - At least once each refueling cycle, 10% of the total population of each type (mechanical or hydraulic), (accessible or inaccessible) of snubbers in the plant shall be functionally tested. Testing requirements shall be in accordance with Plant Technical Specification requirements.

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 NMP1 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 ISI-9.

File: NMP1PP-5.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 6 - AUGMENTED EXAMINATIONS Table of Contents 6-1 Record of Revision 6-4 6.0 AUGMENTED EXAMINATIONS 6-5 6.1. Regulatory Documents 6-5 6.1.1 USNRC Regulatory Guide 1.65 Closure Head Studs . ~............ 6.5 6.1.2 USNRCRegulatoryGuide1.150Ultrasonic Examination ~ . ~... ~... ~ ~....... 6-5 6.1.3 Generic Letter 88-01 IGSCC Cracking in Piping .. 6-5 6.1.4 Generic Letter 94-03 IGSCC Cracking of Core Shrouds, BWRVIP 07 6-14 6.1.5 Generic Letter 98-05 Reactor Vessel Shell Welds BWRVIP 05 6-15 6.1.6 NUREG 0619 BWR Feedwater Nozzle and CRD Return Line Nozzle Cracking ... 6-15 6.1.7 NUREG 0803, GL 81-34, 86-01 BWR Scram System Pipe Break .. 6-20 6.2 Industry Documents 6-20 6.2.1 In-Vessel Visual Examinations 6-20 6.2.2 SIL 571 Instrument Nozzle Safe End Cracking BWRVIP 06, 27, 49 6-22 6.2.3 SIL409, RICSIL073 Incore Dry Tube Cracks .. ..

~ 6-24 6.2.4 SIL 419 CRD Hydraulic Control Valves . 6-24 6.2.5 SIL 433 Shroud Head Bolt Cracks 6-24 6.2.6 SIL455, RICSIL072AIloy182 Nozzle Weldments 6-25 6.2.7 SIL 474 Steam Dryer Drain Channel Cracking ~ 6-26 6.2.8 SIL 483 CRD Cap Screw . 6-27 6.2.9 SIL 539 RPV Head Clad Cracking 6-27 File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Table of Contents - Continued 6.2.10 SIL 554 Top Guide Cracking 6-27 6.2.11 SIL 588 Top Guide and Core Plate Cracking . 6-28 6.2.12 SIL 459 Recirculation Pump Shaft Cracking . 6-28 6.2.13 IE Bulletin 80-13, RICSIL 073 Core Spray Sparger . 6-29 6.2.14 IE Bulletin 96-03 Emergency Core Cooling Suction Strainers . 6-30 6.3 Additional Commitments 6-31 6.3.1 NCTS 503783-01 CRD Housings . 6-31 6.3.2 INPO SER 5-85 Mixing Points 6-34 List of Tables 6.1 IGSCC Category A Welds Selected . 6-7 6.2. IGSCC Category D Welds Selected 6-8 6.3 IGSCC Category E Welds Selected 6.11 6.4 IGSCC Category F Welds Selected 6.12 6.5 IGSCC Category G Weids Selected ~ . 6-12 6.6 IGSCC Category S Welds Selected .. 6-13 6.7 IGSCC Examination Requirements 6-14 6.8 Routine Inspection Intervals Refueling Cycles . ~ .. ~ 6-17 6.9 Feedwater Nozzle/Sparger Inspection Recommendations 6-17 6.10 Augmented Feedwater Nozzle Examinations .. 6-18 6.11 In-Vessel Augmented Examinations (BWRVIP's) 6-20 6.12 Instrument Nozzies 6-22 6.13 Instrument Nozzles 6-23 6.14 Instrument Nozzles 6-23 6.15 IRM Incore Dry Tube Assembly . 6-24 File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unlt1 V NIAGARA M U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Table of Contents - Continued 6.16 SIL 455 Selected Welds 6-25 6.17 SIL 474 Steam Dryer 6-27 6.18 RPV Head Cladding . 6-27 6.19 Upper Core Grid 6-28 6.20 SIL 588 Upper Core Grid 6-28 6.21 SIL 459 Pump Shaft 6-29 6.22 Core Spray Sparger 6-30 6.23 IEB 96-03 Suction Strainers 6-30 6.24 CRDH Upper J Welds . 6-31 6.25 Tee Base Material 6-34 List of Figures 6.3 Feedwater Nozzle Examination Zones . 6-19 File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD QF REVISION REVISION REASON FOR REVISION No.

September 27, 1999 Entire Document Updated lnservice Inspection Program Plan for the 3" Ten Year Inservice Inspection Interval File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 6.0 AUGMENTED EXAMINATIONS This section of the Third Inservice Inspection Program provides a summary description of additional requirements identified in regulatory, industry or other documents in addition to the ASME Code. NMPC has reviewed these additional documents for consideration and were determined to be applicable have incorporated them within the Third Inservice Inspection Plan Tables. NMPC plans on utilizing where applicable, Augmented Examinations to satisfy the requirements of ASME Section XI.

6.1 REGULATORY DOCUMENTS 6.1.1 USNRC Regulatory Guide 1.65 Materials and Inspections for Reactor Vessel Closure Studs.

This regulatory guide defines materials and testing procedures acceptable to the regulatory staff for implementing material and examination criteria for reactor vessel closure head studs for light-water-cooled reactors. Currently the ASME Code,Section XI requires a visual examination to be performed on nuts; a volumetric examination to be performed on studs, in place; and both a volumetric and a surface examination when the stud is removed. At NMP1 the reactor pressure vessel studs are not removed.

Scope of Examination - The sixty-four (64) closure head studs receive an ultrasonic examination of each stud, in-place. As required by the Code, approximately twenty-two (22), (1/3 of the 64 studs) will be examined in-place each period.

Regulatory Position C.4.b will be invoked to examine by either a magnetic particle or liquid penetrant method a representative sample of a minimum of twelve (12) studs and on a reasonable geometric distribution of three (3) studs per each 90-degree quadrant. Additionally, NMP1 will remove 12 studs per interval, essentially 4 per period, with a distribution of 3 per 90-degree quadrant. These studs receive both a volumetric and a surface examination.

(64) Closure Head Studs, (12) required, (4) per Period 6.1.2 USNRC Regulatory Guide 1.150- UT of RPV Welds during PSI and ISI This regulatory guide identifies the acceptable methods of performing ultrasonic examination of Reactor Pressure Vessel welds. NMP1 deviates from the guide by performing examinations of the RPV shell welds in compliance with the spirit of PDI and Section XI, Appendix Vill.

Scope of Examination - Excluded from the Third Inservice Inspection Interval by Relief Request ISI-1.

6.1.3 Generic Letter 88-01, Augmented IGSCC Examinations The Nine Mile Point Technical Specifications, Section 4.6.F.3, requires NMPC 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, "Intergranular 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 Pressure Boundary Piping."

Fiie: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V N1AGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Generic Letter 88-01, USNRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, presents the USNRC staff positions on materials, processes, water chemistry, weld overlay reinforcement, partial replacement, stress improvement of cracked weldments, clamping devices, crack characterization and repair criteria, inspection methods and personnel, inspection schedules, sample expansion, leak detection, and reporting requirements. The technical bases for these positions are detailed in NUREG-0313, Rev. 2, "Technical Report on Material Selection and process Guidelines for BWR Coolant Pressure Boundary Piping."

Generic Letter 88-01, Supplement ¹1 - In its first supplement to GL 88-01, issued February 2, 1992, the USNRC provided several acceptable alternative staff positions to those originally in the Generic Letter.

NMP1 has elected to use two of these alternative staff positions. One of these positions allows sample expansion for Category D welds to be limited to the piping system where the crack was found. NMP1 has elected to examine 50% of Category D welds, by system loop, each cycle and will use this relaxation of sample expansion criteria should cracking be found. The second position is described in Examination Category S below.

NMP1 Background - In 1981 NMP1 identified IGSCC in large bore austenitic stainless steel piping and replaced the piping with low carbon austenitic stainless steel. NMPC considers the replacement material to be IGSCC resistant material, Category A, as defined in GL 88-01.

Examinations performed under the scope of GL 88-01 (and this program plan) are in compliance with 1989 Edition of ASME Section XI. In addition, as required by NMP1 Technical Specification 4.2.6a.2, "the Inservice Inspection Program for piping identified in USNRC Generic Letter 88-01 shall be performed in accordance with the staff positions on schedule, methods, personnel and sample expansion identified in GL 88-01." Where practical examinations performed on IGSCC welds shall also be used to satisfy Section XI Program "B" compliance.

GL 88-01 divides piping welds into 7 categories lettered A through G, of which four are applicable to NMP1.

They are categories A, D, F and G. A summary of these categories as well as their applicability to NMP1 is shown in applicable Table identified below.

IGSCC Category A Weldments - Identifies welds which are fabricated from resistant materials.

Category A weldments are those welds with no known cracks, that have a low probability of incurring IGSCC problems, because they are made entirely of IGSCC resistant material or have been solution heat treated after welding. Augmented examinations required by GL 88-01 are identified in the Third Inservice Inspection Plan and Schedule Tables.

Scope of Examination - IGSCC Category A welds are examined 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 with at least 12% in 6 years. There are one hundred forty-one (141) Category "A" welds at NMP1, one hundred thirty-one (131) are ASME Class 1 and ten (10) welds are ASME Class 2.

(141) Category A Welds, (35) required File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl T NlAGARA H kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

,. :-', "';:188CC";Category "A"'Welds S~lected.-. '= ':

Component Component Period Component -

','Component Period Identifloafion Description Select'ed Identification'..;.Description Selected, ',;

32-WD-022 P-V 2~o PERIOD 32-WD-125 N1D-SE 3" PERIOD 32-WD-035 V-P 3" PERIOD 32-WD-128A P-V 2" PERIOD 32-WD-040 E-P 2"o PERIOD 32-WD-167 N1E-SE 1s PERIOD 32-WD-041 P-N2A 2" PERIOD 32-WD-171BR V-R 3" PERIOD 32-WD-042 N2A-SE 2" PERIOD 33-WD-015 P-V INACC. Each Outage 32-WD-061 E-P 3" PERIOD 33-WD-035 V-P INACC Each Outage 32-WD-082 N2B SE-N 3" PERIOD 39-WD-002 N5A-SE 1sT PERIOD 32-WD-110 P-E o 3 PERIOD 39 WD 003 NSA SE-P 1 PERIOD 32-WD-121 P-N2C SE 1 PERIOD 39-WD-006 P-P 1sT PERIOD 32-WD-122 N2C SE-N 1~ PERIOD 39-WD-008 P-P 1s PERIOD 32-WD-164 N2D SE-N 2" PERIOD 39-WD-009 P-V INACC Each Outage 32-WD-208 N2E SE-N 3" PERIOD 39-WD-097 P-V INACC Each Outage 32-WD-002 N1A-SE 2" PERIOD 39-WD-211C1 P-E 2" PERIOD 32-WD-045 N1B-SE 3"o PERIOD 39-WD-037C1'-E 2" PERIOD 32-WD-085 N1C-SE 1~ PERIOD 39-WD-041C1* P-HX 3 o PERIOD 32-WD-092 P-E 1~ PERIOD 39-WD-100" R-P 1~ PERIOD 39-WD-121C1* P-HX 1sT PERIOD 39-WD-117C1'-E 3" PERIOD 32-WD-062 P-V 3" PERIOD 32-WD-101 P-P 1sT PERIOD 32-WD-119 P-E 1~ PERIOD 32-WD-120 E-P 1 PERIOD 32-WD-157 V-P 3"o PERIOD Class 2 welds IGSCC Category B Weldments - Identifies welds which are fabricated from non-resistant material Category B weldments are those welds made of resistant materials, but have had an Sl performed either before service or within two years of operation.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unlt1 Y NIAGARA 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - There are no welds in this category at the Nine Mile Point Nuclear Power Station.

IGSCC Category C Weldments - Identifies welds which are fabricated from non-resistant materials.

Categoly C weldments are those welds not made of resistant materials, and have been given an Sl process after more than two years of operation.

Scope of Examination - There are no welds in this category at the Nine Mile Point Nuclear Power Station.

IGSCC Category D Weldments - non-resistant materials: no stress improvement Category D weldments are those welds not made with resistant materials, and have not been given an Sl treatment, but have been examined and found to be free of cracks. Included in this category are all bimetallic nozzle weldments made with non-resistant material and 182 inconel weld butter.

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. Welds classified as category D are examined in accordance with GL 88-10, Table 1, as modified by alternative staff position ¹4 for sample expansion, as contained in Supplement 1 to GL 88-01. All 142 Category D welds are ultrasonically examined every other refueling outage, with sample expansion limited to the piping system loop where cracking was found.

(142) Category D Welds, (284) required Tabie 6-2 IGSCC Category "D" Selected WeIds Component Identification Component Component Identification Component Descrfption Description 33-W D-004 P-P 40-WD-38A N6A SE-SE 33-WD-046 P-P 40-WD-039 N6A N-SE 33-WD-049 P-E 40-W D-041 V-P 37-WD-003 F-R 40-W D-043 BC 38-WD-001 V-P 40-W D-043A P-N 38-WD-002 P-E 40-WD-044 P-E 38-WD-003 E-E 40-WD-045 E-P 38-WD-004 E-P 40-WD-047 P-V 38-WD-005 P-E 40-WD-048 P-E 38-W D-006 E-P 40-WD-049 E-V 38-WD-089 P-E 40-W D-051 P-E File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

" ':.'":'-:':-.7j-:,'.,:::;;;-;-',".",.:-';>>'" " ':: ';""::":-"IGSCC'Cat gory.'"'Selectecf Vfeids ",:,-",.'".'-.',-",,'~,"-,.~;:,.";:

Comporient:Identification Co'mpo riant:.:;." Component, identification Compcinent Description,:: ,Description 38-WD-090 E-P '0-WD-052 E-P 38-WD-091 P-P 40-W D-053 P-E 38-WD-092 P-P 40-WD-054 E-P 38-WD-093 P-E 40-WD-055 P-E 38-WD-094 E-V 40-WD-056 E-P 39-WD-195 P-E 40-WD-058 P-T 39-WD-196 E-P 40-WD-059 T-V 39-WD-197 P-E 40-WD-060 V-P 39-WD-198 E-P 40-WD-061 P-E 39-WD-1 99 P-E 40-WD-062 E-P 39-WD-200 E-P 40-WD-063 P-E 39-WD-201 P-E 40-WD-064 E-P 39-WD-202 E-P 40-WD-065 P-E 39-WD-203R P-E 40-WD-066 E-P 39-WD-204A P-P 40-WD-067 P-T 39-WD-204R E-P 40-WD-068 T-P 39-WD-226B P-P 40-WD-069 P-V 39-WD-227 P-E 40-WD-070 V-P 39-WD-228 P-E 40-WD-072 E-P 39-WD-229 P-E 40-WD-073 P-T 39-W D-230 E-P 40-WD-074 T-P 39-WD-231 P-P 40-WD-075 P-P 39-WD-232 P-E 40-WD-076 P-R 39-W D-233 E-P 40-WD-077 R-P 39-WD-233A P-E 40-WD-079 N6B SE-E File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NIAGARA H U MOHAWK THIRD INSERVICEINSPECTIONINTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

...- Tabte,.s,'-,,2"~.:; .

'.;;:;,:,';::,",,,.;;:,lGSCC;Categorjj .".0,"...,':;Sefectecf.,:,Nelds .: ...,:,;,.,:j;;;:,;;~,",',;;,,

'Component'Iclentif fcation'"'," "Component ~

" ':.Coinponent fdentiftcation:. rCornponent" '"'"

, Descrtptlon';:;: ~ ;;;,;;:,':,~,',,;;, <<Description ': "

39-WD-234A P-E 40-WD-079A N6B SE-SE 39-WD-235R E-P 40-WD-080 N6B N-SE 39-WD-090 N5B-SE 39-WD-014 P-P 40-WD-001 V-P 39-WD-016 P-P 40-WD-003 BC 39-WD-017 P-P 40-WD-004 P-E 39-WD-020 P-P 40-WD-005 E-P 39-WD-021 P-P 40-WD-006 P-P 39-WD-022 T-P 40-W D-007 P-V 39-WD-024 T-P 40-WD-008 P-E 39-WD-025 E-P 40-WD-009 E-V 39-WD-026 P-P 40-WD-011 P-E 39-WD-027 P-P 40-WD-012 E-P 39-WD-030 E-P 40-WD-013 P-P 39-WD-034 P-P 40-WD-014 P-P 39-WD-035 P-P 40-WD-015 P-E 39-WD-038 E-P 40-WD-016 E-P 39-WD-102 P-P 40-W D-017 P-E 39-WD-104 P-P 40-WD-018 E-P 39-WD-105 P-P 40-WD-020 P-T 39-WD-108 P-P 40-WD-021 T-P 39-WD-109 P-T 40-WD-022 P-E 39-WD-110 T-P 40-WD-023 E-P 39-WD-112 T-E 40-WD-024 P-V 39-WD-113 E-P 40-WD-025 V-E 39-WD-114 P-P File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI V NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

,,IGSCC,Category",9" Selected'eids ...:.:,',",

Com sonant Identification Component Component Identfffcatfon Component

'?

Description oA,> Oescriptfon 40-WD-026 E-P 39-WD-115 P-E 40-W D-027 P-T 39-WD-118 E-P 40-WD-028 T-V 39-WD-122 P-E 40-WD-029 V-P 39-WD-123 P-P 40-WD-030 P-T 39-WD-124 P-E 40-WD-031 T-P 39-WD-125 E-P 40-WD-032 P-E 39-WD-128 E-P 40-W D-033 E-E 39-WD-129 P-P 40-WD-034 P-E 40-WD-036 R-P 40-WD-035 E-R 40-WD-038 N6A SE-E IGSCC Category E Weldments - All welds included ln this category are weld overlays.

Category E weldments are those welds with known cracks that have been reinforced by an acceptable weld overlay or have been mitigated by an Sl treatment welding.

Scope of Examination - should be inspected once every two refueling outages after repair. Approximately 50% shall be inspected during the first refueling outage and subsequent outages.

(1) Category E welds, (1) required once every two refueling outages TABLES@ "

I GSCC:,CATEGORY "E" SELECTEO WELDS.".>>",

'OMPONENT: REMARKS"'.':: ~-':::"

2 COMPONENT, DESCRIPTION IDENTIFICATION 33-FW-22 PIPE TO SS NOZZLE Outside CIV, 10% each refueling, weld overlay, once every 2 RFO's IGSCC Category F Weldments - Cracked, inadequate or no repair Category F weldments are those welds with known cracks that have been approved by analysis for limited additional service without repair.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NlAGARA H kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - inspected each refueling outages.

(5) Category F welds, (5) required each refueling outage.

,, " IGSCC Categor'y",F",Selected'Vfeldi '"

Weld No,.'..," Component'identification ..' ", Re-.eximlnatlon'Frequency':,"';,,;

32-WD-046 Nozzle N1B Safe End-Elbow Examine each refueling outage 32-WD-050 Valve 32-376-Pipe Per IWB-2420(b) each period 32-WD-086 Nozzle N1C Safe End-Elbow Examine each refueling outage 32-WD-126 Nozzle N1D Safe End-Elbow Examine each refueling outage 32-WD-168 Nozzle N1E Safe End-Elbow Examine each refueling outage IGSCC Category G Weldments - Non-resistant and not inspected by UT Category G weldments are those welds not made of resistant materials, have not been given an Sl treatment.

Scope of Examination - Welds classified as Category G (excluding RWCU see GL 88-01 Supplement ¹1) are examined in accordance with GL 88-01. Twenty-seven (27) RWCU piping welds are included in this category, two (2) of which are inside penetrations, one (1) examined every 2 refueling outages and twenty-four (24) are part of the Level 2 expanded sample An additional twelve (12) welds, seven (7) of which are

~

inside penetrations and five (5) can not be examined due to configuration, receive a visual examination for evidence of leakage at each refueling outage.

(39) Category G welds, (15) required

.,'.;,Table6<<8

~ ' ';~lGSCC'Category "6" Selected~'.WelCh:.', ",'

Component ...- Coinpon'e'nt~.'"" :,Com'pon eat Examin'ation":

!Exa'mtnatio'n'.,:.'omp'oiient::.:Requirements,'.".;,,

.':,Identlffcation ,:;, ',Descriptton"P'~, IdentltIc'ation,',', ~D'es'crlption,,,,Requtremerits",;.;

33-WD-014 P-P Inacc. VT-2 38-WD-087 V-P Inacc. VT-2 each outage each outage 33-WD-036 P-E Inacc. VT-2 38-WD-088 P-P Inacc. VT-2 each outage each outage 33-WD-22R P-N 2No and 3Ro 39-WD-1 94 V-P VT-2 each Period outage 38-FW-007 P-P Inacc. VT-2 39-WD-194A P-P Inacc. VT-2 each outage each outage File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 al V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

>Co'mpone'nt:-:,:.,', ':Component-.;'; Examfnition,'. Compoiient -:...: Component Examination

'.Identification:-".- - 'Description ',: 'Requirements;. :.Identification - DescrIption Requirements 38-FW-008 P-V Inacc. VT-2 39-WD-226 V-P VT-2 each each outage outage 39-WD-226A p-p Inacc. VT-2 39-09R-WD- V-EW VT-2 each each outage 001 outage 39-10R-WD- V-E VT-2 each 40-WD-010A V-P Inacc. VT-2 001 outage each outage 40-WD-050A V-P VT-2 each outage IGSCC Category S Weldments - outboard of Cl's Category S weldments are those welds that are located on the Reactor Water Clean Up system outboard of the Containment Isolation Valve.

Scope of Examination - The Reactor Water Cleanup System outside of the containment isolation valves has very high levels of radiation. Therefore until actions associated with GL 89-10 on motor operated valves are completed by licensees, (ref. file code M96-0007 and M96-0016) the USNRC staff has determined that the inspection of the subject piping on a sampling basis of at least 10% of the weld population may be performed each refueling outage. The population of RWCU piping that lies outboard of containment isolation valves is 32. Based on the 10% sample requirement, NMP1 will examine three (3), (10% of 32) welds each outage.

+'eld (6) IGSCC Categoly S welds, (3) required each refueling outage.

Number '-

'Table 6A IGSCC iCategory"S" Setected:

Welds'...

'omponent'Description- 'xamlnatiori Frequency 33-FW-35 Elbow-Nozzle Each Refueling Outage 33-FW-36 Nozzle-Elbow Each Refueling Outage 33-FW-37 Elbow-Nozzle Each Refueling Outage Inspection Schedule - The extent and frequency of inspection for various weldment categories are detailed in Table 6-6 below.

File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 EI Y NlAGARA HU MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

.TABLE 6-7-;: ".;, -",:-";:"-.

. ;::;,::-',:;" lGSCC EXAMINATfONREQUlREMENTS,

'GSCC "'.':.'XA'MINATION, EXTENT OF.,EXAMINATION, REMARKS CATEGORY: ,'>>REQUIREMENTS A 25% Every 10 Year Interval At least 12% in 6 years 50% Every 10 Year Interval At least 25% in 6 years All Within Two Refueling Cycles At least 50% in 6 years after the Post-Sl Inspection, and All Every 10 Years thereafter D All Every Two Refueling Outage 50% each refueling outage All Every Two Refueling Outages 50% each refueling outage All Every Refueling Outage G All Next refueling Outage All due to inaccessibility and high radiation S Each refueling Outage RWCU outboard of CIV's 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.

6.1.4 Generic Letter 94-03, "Intergranular Stress Corrosion Cracking of Core Shrouds in BWRs" and BWRVIP-07, "EPRI Report TR-105747 Guidelines for Reinspection of Core Shrouds.

In September 1994, the USNRC issued GL 94-03, which required that all BWR licensees inspect their core shroud. In response to GL 94-03 NMP1 elected to pre-emptively repair the core shroud in lieu of inspection.

This repair installed stabilizer assemblies that replaced the load carrying capability of the shroud circumferential welds H1 through H7. By letter dated March 31, 1995, the USNRC requested NMP1 to submit reinspection plans for the core shroud and its repair assemblies. Reinspection plans were submitted in December 1998 which were consistent with the guidelines provided in BWRVIP-07, EPRI Report TR-105747, "Guidelines for Reinspection of BWR Core Shrouds.

In a letter dated March 24, 1999, the USNRC transmitted an SER which accepted the NMP1 inspection plans submitted in the December 30, 1998 NMPC letter. These plans were accepted for RFO-15 only.

Shroud vertical welds were inspected during RFO-14 and RFO-15. During RFO-15 shroud vertical welds V9 and V10 were repaired. Future inspections of the core shroud and its repair assemblies will be performed in accordance with the inspection guidelines documented in BWRVIP-07 and BWRVIP-63, EPRI Report TR-File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 113170, "Shroud Vertical Weld Inspection and Evaluation Guidelines".

Scope of Examination - As stated in the USNRC SER, NMPC commitment was to submit the scope of the examinations for RFO-16 to the USNRC, three ( 3) months prior to the start of the outage.

6.1.5 Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP - 05 Report to Request Relief From Augmented Examination Requirements on Reactor Vessel Shell Welds.

In NMPC letter NMP1L 1391, dated December 10, 1998, relief was requested pursuant to GL 98-05.

Niagara Mohawk Power Corporation (NMPC) requested relief from the inservice inspection requirements of 10CFR50.55 (g) for volumetric examination of circumferential reactor pressure vessel (RPV) welds (ASME Code Section XI, Table IWB-2500-1, Category B-A, Item 1.11, Circumferential welds). This relief request also includes an alternative to the required inspections for RPV shell welds specified in 10CFR50.55a(g)(6)(ii)(A)(2).

Additionally, NMPC requested approval of an alternative to the examination requirements specified in 10CFR50.55a(g) for volumetric examination of longitudinal RPV shell welds and the shell-to-flange weld (ASME Code Section XI, Table IWB-2500-1, Category B-A, Item 1.12, and Item 1.30). NMPC proposes to perform an automated inspection of certain RPV welds using personnel and procedures qualified to the Performance Demonstration Initiative, (PDI). The use of these inspection procedures is a alternative to 10CFR50.55a(b)(2).

NMPC has incorporated the information in BWRVIP-05 into these alternative requirements and addressed the USNRC positions in the USNRC's July 28, 1998 safety evaluation report. See NMPC letter NMP1L 1391 for specific details.

Scope of Examination - NMP1 is received a final Safety Evaluation Report from the USNRC. See Examination Category B-A, Section 2 of this program plan for additional information.

6.1.6 NUREG 0619 BWR Feedwater and Control Rod Drive Return Line (CRDRL) Nozzle Cracking, Generic Letter 81-11 and GE NE-523-A71-0594 USNRC Generic Letter 81-11 forwarded NUREG 0619 and corrected a footnote in the guide. This regulatory guide describes the technical issues associated with the discovery of cracking in feedwater nozzles and control rod drive return line nozzles. The NUREG also describes technical studies and analysis performed by the General Electric Company and the USNRC staff, the staff's technical positions based on these studies, and the staff's requirements for licensee implementation of the technical positions.

NMP1 initial response, dated December 29, 1980, committed to a program of periodic ultrasonic examinations every other refueling outage, visual inspections of the Feedwater spargers every fourth refueling outage, and liquid penetrant examinations every sixth refueling outage or every 90 startup/shutdown cycles, which ever occurs first. Subsequent responses amended the examination methods and included the use of the General Electric Nuclear Energy (GE-NE) GERIS 2000 automated System (UT) in place of performing dye penetrant examinations. The USNRC, in their letter dated February 5, 1999, accepted the examination program submitted in the NMPC latest submittal, dated September 4, 1998. This submittal identifies the NMP1 commitments for the third inservice inspection interval, and summarized as follows:

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 hI V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN The USNRC, in a letter dated June 5, 1998, forwarded the safety evaluation that accepted, with modifications, the Boiling Water Reactor Owners Group (BWROG) report GE-NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements." Subsequently, in the September 4, 1998 NMPC letter, NMPC revised its commitment for NMP1's feedwater (FW) and control rod drive return line (CRDRL) nozzle inspections. The revised commitment for the FW nozzle inspections would be in accordance with the BWROG report GE-NE-523-A71-0594, dated October 1, 1995, subject to the modifications in the related USNRC safety evaluation report dated June 5, 1998.

Also in the September 4, 1998 letter NMPC revised its commitment for CRDRL nozzle inspections by replacing periodic liquid penetrant (PT) with less frequent ultrasonic testing (UT). The examination techniques will be in accordance with the requirements of ASME Code using the GE GERIS-2000 system to perform contact pulse-echo UT examinations. The GERIS-2000 system inspection is performed with sensitivity for detecting flaws that is more sensitive than ASME Code requirements. These revised inspections were accepted by the USNRC in the February 5,1999 letter.

F w The following is the list of modifications in the USNRC safety evaluation report, Section 5.0, which modify the requirements in the BWROG report:

1. The UT techniques should have the ability to reliably detect axially oriented flaws from a depth equal to 0.25 inches for each of the Zones 1 through 3 and axially and radially oriented flaws in the area of the nozzle-to-safe end welds located in Zone 5 (Figure 1). The nozzle-to-safe end butt weld in Zone 5 is required to be inspected according to paragraph IWB 2500-1 of the ASME Code.

The PT may be eliminated from FW nozzle examinations, provided that the UT techniques satisfy the requirements of the 1986 or later approved editions of ASME Code or the objectives of Appendix Vill. UT techniques that do not satisfy the 1986 or later approved editions of ASME Code or the objectives of Appendix Vill shall follow the PT frequency shown in Table 1.

3. The automated UT (gated peak threshold recording) multiplication factors shall be those shown for manual UT in Table 2, Method 1. Automated UT (gated peak threshold recording) techniques qualified according to the objectives of Appendix Vill may use multiplication factors in Table 2, Methods 2 or 3.

4. The automated UT (no threshold recording) multiplication factors in Table 2, Method 3 are adequate, provided that the UT techniques are qualified according to the objectives of Appendix Vill.

5. The inspection of Zone 3 shall be at the same frequency as Zones 1 and 2, except that licensees using triple sleeve with double piston ring design sparger may follow the proposed inspection frequency for Zone 3, but not less than one inspection every ASME Code interval.

The fracture mechanics analysis shall be recalculated using the more recent fatigue curves in the ASME Code that address environmental effects. The examination requirements and frequency for the third inspection interval will be determined based on the results of this recalculation.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hI T NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN (4) Visual inspection of flow holes and welds in sparger arms and sparger tees. These requirements are the same as those specified in NUREG-0619.

Scope of Examination - The zones to be examined by the ultrasonic techniques shall be the regions as shown for identification purposes in Figure 6-9.

Table 6- 10 AUGMENTED FEEDWATER NOZZLE EXAMINATION Feedwater Nozzle identification Zone No Examination Method Extent and Frequency Nozzle N4-A Automated UT Techniques 3" Periods 31-WD-030-IR 1 and 2 31-W D-030 3 Nozzle N4-B Automated UT Techniques 3" Periods 31-WD-021-IR 1 and 2 31-W D-021 3 Nozzle N4-C Automated UT Techniques 3" Periods 31-WD-051-IR 1 and 2 31-WD-051 3 Nozzle N4-D Automated UT Techniques 3Re Periods 31-W D-060-IR 1 and 2 31-WD-060 3 RV-05-I Feedwater Sparger N/A Visual VT-3 Examination Every 4'. Refueling Assemblies and Feedwater Outage, commencing at Nozzles RFO-8, (1981),

Scheduled 3" Period CRD Nozzle N9 N/A Automated UT Techniques 3" Period 44.1-WD-018-IR File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Vcsscl Nozzle M

A t%

Safe-End 3 28 2A Figure 6-9 FEEDWATER NOZZLE ZONES Filo: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 6.1.7 NUREG-0803/ Generic Letter 81-34 and 86 BWR Scram System Pipe Break Generic Letter 81-34 transmitted NUREG-0803 to all BWR licensees.

NUREG-0803, "Generic SER Regarding Integrity of BWR Scram System Piping", addresses the need for improvement in procedures, periodic inservice inspection and surveillance for the scram discharge volume (SDV) system. These guidelines were developed to address the consequences of a postulated leakage crack in the SDV piping and the resulting large leakage (up to 550 gpm) downstream of the system isolation valves.

Generic Letter 86-01, "Safety Concerns associated with Pipe Breaks in the BWR Scram System",

addressed the staff's position based on information provided in BWROG and General Electric Company supplied generic information (NEDO-22209, BWROG-8420) and staff generic analysis of the SDV piping system integrity. The staff has concluded that SDV piping satisfies BTP MEB 3-1, position B.2.C (1). A through wall leak need not be postulated. Also BWROG emergency procedure guidelines and visual verification of the SDV integrity provide sufficient measures to verify the detecting and mitigating the consequences of leakage.

Scope of Examinations - Based on the above information NMP1 will perform the examinations and tests required by the 1989 Edition of Section XI for Class 2 systems. No additional augmented examination is required.

6.2 Industry Documents 6.2.1 In-Vessel Visual Examinations (BWRVIP)

In addition to the ASME Code requirements, this section identifies those additional examination activities that NMPC has evaluated and determined to be applicable to the Third Inspection Interval. Augmented examinations may be recommended by regulatory documents, NSSS recommendations, industry experience, and/or good engineering practice.

TABLE 6-.11,, ','.;:';;::,;,',,",",;;,,:,;;:.~~;;,~;."

,',:,:,: '.":,':"..IN VESSEL AUGMENTED EXAMINATIQNS,=',::.",js":,,",,",'...'."K.';.;.,;,j:.;.,;

Compoiient (1} Idet1tlfication",.Exanj Inspe,etio'rt;..,!"'.:,~<!,"-, 'Exter'it,&iid'Prequ'e'ncy,;.

',Method 'equirements Top Guide Visual GE RICSIL 071 Periodically as required ( Note 4)

VT-3 BWRVIP-26 EPRI TR-1072851 RICSIL 059 SIL 554 Steam Dryer Assembly Visual BWROG 91 Periodically inspect to assess GE SIL 474 component condition; UT inspect as necessary to characterize cracking; (Note 3)

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NIAGARA a VMOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE 6,- 11

., IN(VESSEL,AUGMENTED'EXAMINATlONS ',

Exam ...;. ;-.IrispectIon: Extent and'Frequency

'omponent,(4}'Identification

',.Requirements .

Method"'isual Core Shroud Assembly, BWRVIP-07 Examine Shroud Tie Rod Shroud Repair Components VT-3 (EPRI TR-105747) Assemblies per details in NER-1M-and Shroud Supports UT BWRVIP-01, 053; Shroud Vertical welds, shroud EVT-1 BWRVIP-38, support welds H8 8 H9, and shroud BWRVIP-63 Component Welds (Note 4)

Core Spray and Core Spray Visual NUREG/CR 4523 Examine CS Piping and CS Sparger Sparger BWRVIP-18 Assembly and piping component EPRI TR-1 06740 condition (Note 4),

Moisture Separator/Shroud Visual BWROG 91 Periodically inspect to assess Heads Assembly GE SIL 433 component condition (Note 3)

Good Engineering Practice Shroud Head Bolts Visual Sil-433 Periodically as required (Note 3)

UT Feedwater Sparger Nozzle Visual Good Engineering Visually inspect each fourth Practice refueling outage, Examine the nozzles on the replacement FW Sparger on a periodic basis (Note 3)

IRM/SRM Dry Tubes (2) Visual BWRVIP-47 Examine on a periodic basis to SIL 409 Rev. 1 assess the conditions of the tubes RICSIL 073 (Note 3,4)

Control Rod Drive Stub Visual BWRVIP-47 When Accessible (Note 3,4)

Tube/Housing Penetrations Good Engineering Practice In-Core Visual Good Engineering When Accessible (Note 3)

Instrumentation/Housing Practice Penetrations Core Delta Pressure/Standby Visual BWRVIP-27 When Accessible (Note3,4)

Liquid Poison Control Good Engineering Practice Core Plate Visual GE RICSIL 071 Periodically as required (Note 4)

BWRVIP-25 EPRI TR-107284 SIL 588 Fife: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI>>003 Unit 1 LI V NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

..."..., "'"':::.':;:"':".:.-. .. TABLE 6 -"I'.I'; '.::,,: ; "

,;.;:,.,;. ':.,;;,';...;, .'iN;VESSEL"'AuGMENTED EXAMINATIONS;;.,'..

. Component:(1) identification>>".>>' :."Exam Inspection:; '-'.;-:-.;:"., Extent and Frequency, Method RequIremerit>>s Control Rod Drive Return Line Visual NUREG 0619 Inspect each outage until final PT is Nozzle performed. NMPC Letter Notes: (1) NMPC has not determined whether reexamination of all these areas, (as defined above) is warranted during the Third Inservice Inspection Interval.

(2) IRM/SRM Dry Tubes have all been replaced in 1986 with three (3) different type designs and two manufacturers with Type 347 material.

(3) These inspections are not commitments and are subject to change as necessary to support good engineering practices.

(4) These inspections will be based on approved BWRVIP guidelines for the identified components.

6.2.2 SIL 571-instrument Nozzle Safe End Cracking and BWRVIPs -06 (TR-1507), VIP-27 (TR-107286),

VIP-49 (TR-108695)

SIL 571, Instrument Nozzle Safe End Cracking, identified a leak in a reactor vessel water level instrument nozzle safe end at a GE BWR/4 plant and documents the results of investigations and examinations that were performed. Subsequently the issue has received substantial industry exposure including investigations and reviews as part of the BWR Vessel and Internals Project (BWRVIP) work.

NMPC has evaluated this concern in Deviation/Event Report ff 1-93-2209. The following examinations were established for the instrument nozzles and the Standby Liquid Control (SLC) nozzle.

Perform a VT-2 examination after vessel flood up to assess for the presence of leakage at the start of each refuel outage and, Follow up with a VT-2 examination at the end of each refuel outage during the system pressure test.

The NMP1 design includes eleven (11) nozzles with stainless steel safe ends, ten (10) instrument nozzles and one (1) standby liquid control nozzle. These nozzles are identified below.

4 Table 6-'12.

',Instrument Nozzies - Stainless Steel Safe Ends

'Safe;End 'levation '-

Azimuth 'escription

.'-'Material de rees'8 N12 SA336-F8 266'1" Standb Li uid Control N13A SA336-F8 306'0" 57 Instr. Level Protection N13B SA336-F8 295'1" 62 Instr. Level Protection N14A SA336-F8 305'0" Instr. Level Control Ran e N14B SA336-F8 295'1" 82 Instr. Level Control Ran e N15A SA336-F8 305'0" 236 Instr. Level Control Ran e N15B SA336-F8 295'1" 236 Instr. Level Control Ran e N16A SA336-F8 306'0" 244 Instr. Level Protection N16B SA336-F8 295'1" 244 Instr. Level Protection N17A SA336-F8 311'2" 250 Instr. Wide Ran e Level File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El T NIAGARA a &MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

-::.>:Table 6-12 .

Instrument Noiztes - Stainless Steel Sate: Ends

'Nozzle'ff "";:.-'. Safe:End ', -,. Elev'ation:

=-

".';"',;';::,'",.::Mate'r'Ial"," "".:.'. ~~"."" ." "."

Azimuth',~.'":;"'

de'"rees, ',

Descrlptlon::-;

N178 SA336-F8 295'1" 252 Instr. Wide Ran e Level All of the reactor instrument nozzle safe ends and the standby liquid control safe end are fabricated of stainless steel material SA-336 F8 Three (3) safe end to nozzle welds are scheduled for examination as follows:

Nozzle':.:.'

, t N12 N13A

'.'-.';-'-"";: -"'~.i ".Instrument

';)804 14

'."'.'."Exam 42.1-WD-034 36-WD-003 ID::--:,,':"-,

Table 6-13>,',

IItozzles - Code Examina'lions

'Code Category/item "~:;:~~;;...:Exam 8-F/85.20 8-F/85.20

~

Surface Surface Exam Schedule 3'f 1 eriod lcd N16A 8C 36-W D-924 8-F/85.20 Surface 3'riod Only the code required exams are scheduled for the above welds. The examination is performed using liquid penetrant and includes the adjacent h" portion of the base material on each side of the weldment.

Should the examination reveal indications, Engineering shall determine the necessity for expansion.

Additionally the remaining nozzles receive a VT-2 examination each refuel outage during system pressure test.

Standby Liquid Control Nozzle Due to the importance of the SLC nozzle the BWRVIP-027 Guidelines recommend a volumetric examination be performed on the nozzle to safe end weld and the safe end extension. NMPC has evaluated this recommendation in Deviation/Event Report ff 1-97-0378. The following supplemental examination has been incorporated in this plan.

""Table,6-14"'->> ~-: '"'- ""':

.": Instrument Noiztes"- Stan'db "LI uid'Cohtrol ISO :"; 'Exam, ID . ',", 'Code,",."'".'-'.:.,',.;,"Examination'. Examination

-Cate o /Item':-~ Method ~

" Schedule N12 42.1-W D-034 8-F/85.20 Volumetric 3'4 Period The volumetric examination includes the weld between the nozzle (Alloy 600 material) and the safe end (Stainless Steel) and the accessible length of the safe end base metal. These examinations will be performed using ASME Section XI methodology and criteria to the extent practical. If indications are located, flaw sizing shall be performed in accordance with Section XI requirements and reported to Engineering. Since the RPV instrument nozzles are fabricated in the same manner, Engineering will also determine the necessity to expand the scope to the instrument nozzles.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 T NlAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 6.2.3 SIL 409 and RICSIL 073, Incore Dry Tube Cracks SIL 409 and SIL 409, Rev. 1, Incore Dry Tube Cracks found in incore dry tubes at three operating BWR's and to alert BWR owners to the need for inspections. Recommended action was to inspect the dry tubs during the next refueling outage. NMPC performed visual examinations on 12 dry tubes in 1984 using an underwater TV camera. The dry tubes were cracked but still serviceable. The dry tubes were replaced in 1986 (RFO-10). The new dry tubes are three different type design and two manufacturers, Kraftwork Union (KWU) and General Electric (GE). There are two new style KWU design, eight new style GE design, and two old style GE design. The new KWU design is made of a higher strength material, (347 stainless steel). The new style GE design has a higher purity, better quality stainless in the area where the cracking had occurred. Both designs are crevice free.

In 1988, both of the old style GE and one each of the new designs were examined. No indications were found. These examinations were repeated during RFO-06 and no indications were found. RICSIL 073, dated May 12, 1995, presented additional information on recent cracking in incore dry tubes.

Scope of Examinations - Conduct visual examination of two old style GE design, every other even numbered outage.

Table 6-'I5 IRM Incore DryTube Assembly Component Identification Component Description Outage Selection SIL409-IDTC1245 IRM Incore Dry Tube Assembly RFO-16 RFO-18 SIL409-IDTC3645 IRM Incore Dry Tube Assembly RFO-16 RFO-18 6.2.4 SIL 419, CRD Hydraulic Control Unit Isolation Valves In accordance with the recommendation of SIL 419, the CRD 101 and 102 valve wedges are examined with liquid penetrant when their companion CRD is removed for maintenance, to a maximum of five (5)

CRD's per refueling.

Scope of Examination - No exams required, all valves replaced and/or upgraded.

6.2.5 SIL 433, Shroud Head Bolt Cracks SIL 433 and SIL 433, Supplement 1, identified cracking of head bolts in four BWR/4's and one BWR/3.

Recommendations were ultrasonic examinations of all BWR/2-5's head bolts the next time the reactor vessel head is removed and the shroud head and separator assembly is removed to the equipment storage pool. Additional information was presented in SIL 433 Supplement 1, dated September 15, 1993 that revised the. recommendations. The cause of cracking was confirmed to be crevice corrosion IGSCC and recommended the entire length of the head bolts be examined.

During RFO-1 0, NMP1 examined all 36 bolts and no indications were identified. These examinations were repeated during RFO-11 and no indications were identified. Examinations will be repeated until an evaluation supports their deletion. See NMPC letter NMP-36220.

File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - Examinations are to be identified later.

6.2.6 SIL 455 and RICSIL 072, Recommendation for Additional ISI of Alloy 182 Nozzle Weldments SIL 455 identified IGSCC type indications in one of ten-recirculation inlet safe- end to nozzle Inconel 182 weldments. One of the axial components had extended about one quarter inch into the alloy steel material and has been repaired by temper bead weld overlay. The SIL recommended BWR owners perform additional ultrasonic examinations, during planned inservice examinations, of safe-end to nozzle weldments if cracks are found in alloy 182. Subsequently Revision 1, and Revision 1 Supplement 1 has been issued. These revisions recommend owners review their design for the for Recirculation inlet and outlet nozzles, high pressure core spray, low pressure core spray and low pressure coolant injection nozzles, Jet pump instrumentation nozzles, and control rod drive hydraulic return nozzles. For those designs where alloy 182 extends back into the nozzle bore, UT examinations should be performed in this extended area. Also GE recommends that a 45-degree and a 60 degree refracted longitudinal wave be used for crack detection and sizing. AT NMP1 34 nozzles utilize designs in which Inconel weld butter extends back into the nozzle bore. See NMPC letter MNP 36220.

Additional information was presented in RICSIL 072, dated January 10, 1995.

Table 6-16 .. '::,:,

SIL466 Selected Neids-...

"Compon'ant:fderttifIcatl~oit'" ".'"-',~ '-",'" 'Coiiip'orIeiit Disci)ptfon"""'"-," Exam" Wh art"..

Method Selected RV-WD-01 1 Nozzle N7A - Flange UT 1e'eriod RV-WD-013 Nozzle N7B - Flange UT 3" Period RV-WD-015 Nozzle N7C - Flange UT 3~ Period RV-WD-017 Nozzle N8 D - Flange UT 3" Period RV-WD-019 Nozzle N7E - Flange UT 1" Period RV-WD-021 Nozzle N7F - Flange UT 3" Period RV-WD-023 Nozzle N7G - Flange UT 3"e Period RV-WD-025 Nozzle N7H - Flange UT 3Ro Period RV-WD-027 Nozzle N7J - Flange UT Period RV-WD-029 Nozzle N7K - Flange UT 1~ Period RV-WD-031 Nozzle N7M - Flange UT 1eT Period RV-WD-033 Nozzle N7N - Flange UT 3 Period RV-WD-035 Nozzle N7P - Flange UT 3"e Period File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA V MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Table 6-16 SINGS Selected %aids Component Identification Component Description Exam When Method Selected RV-WD-037 Nozzle N7R - Flange UT 3"0 Period RV-WD-039 Nozzle N7S - Flange UT 3"e Period RV-WD-041 Nozzle N7T- Flange UT 1sT Period RV-WD-043 Nozzle N7U - Flange UT 1 Peripd 32-WD-042 N2A Nozzle - Safe End UT 2"e Period 32-WD-082 N2B Safe End - Nozzle UT 3"e Period 32-WD-122 N2C Safe End - Nozzle UT 1s'eriod 32-WD-164 N2D Safe End - Nozzle UT 2" Peripd 32-W D-208 N2E Safe End - Nozzle UT 3"e Perjpd 32-WD-002 N1A Nozzle - Safe End UT 2" Period 32-WD-045 N1B Nozzle - Safe End UT 3Ro Period 32-WD-085 N1C Nozzle - Safe End UT pel jpd 32-WD-125 N1D Nozzle - Safe End UT 3" Period 32-WD-167 N1E Nozzle - Safe End UT 1 Period 36-WD-1074 N7L Closure Head UT 3RD Perjpd 37-WD-002 N8 Nozzle - Flange UT 2'e Period 39-WD-002 N5A Nozzle - Safe End UT 1 Peripd 39-WD-090 N5B Nozzle - Safe End UT 3" Period 40-WD-039 N6A Nozzle - Safe End UT 1" Period 40-WD-080 N6B Nozzle - Safe End UT 1 Period 44.1-WD-017 N9 Safe End - Nozzle UT 1s'eriod 6.2.7 SIL 474, Steam Dryer Drain Channel Cracking Scope of Examination - As recommended by SIL 474, Steam Dryer Drain Channel Cracking, NMP1 will visually examine steam dryer drain channels each period.

File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unlt1 El Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

. ~~~X+,,~~c

', 'Table,6'-".'l7;;:."',:."....

SIL 474 Stear'n,Dryer'.

a'Component'Identification;>'..

Component 0'esc'rlption Exam Method

'.- ':.Frequenc j RV-01-I Steam Dryer Drain VT-1 Once every period 6.2.8 SIL 483, CRD Cap Screw Crack Indications SIL 483, CRD Cap Screw Indications, identify circumferential cracking and corrosion pitting in the shank directly below the cap screw head. SIL-483. Rev.2 recommends visual examination of all removed cap screws for crack indications in the shank -to-head transition region whenever cap screws are removed for routine maintenance.

This SIL resulted from examinations performed by NMP1. An analysis of these examination results was issued by GE in March 17, 1989. This SIL relaxes previous NMP1 commitments.

Scope of Examinations - An LP examination will only be performed on suspect cap screws slated for reuse. No examinations are scheduled pursuant to this plan.

6.2.9 SIL 539, RPV Head Clad Cracking SIL 539, RPV Head Clad Cracking, identifies cracking in the stainless steel cladding of the RPV top head. The cracking was identified by visual examination that revealed rust streaks on the cladding surface. Visual examinations will be per Code with particular attention given to back clad areas of the flange to dome weld. If indications are found, a confirmatory liquid penetrant examination or enhanced ultrasonic testing is performed. Also see USNRC Information notice No. 90-29.

Scope of Examination - No examinations are planned during this interval.

Table;6.,:,~1'8 '2

,: 'RPV Head;Cladding.

Component Identification'-.',"-=.',,- 'omponent .

Exam

,Description, ':,-;;-',-'.';:., Method,,

RV-16-I Vessel Cladding VT-3 No Exams Planned 6.2.10 SIL 554, Top Guide Cracking SIL 554, Top Guide Cracking, identified a through wall crack approximately 1 and ~A inches long in a top guide of a BWR/2. This SIL recommends owners of BWR/2, 3, 4 and 5 plants, with top guide fluence levels above 1x10" neutrons per square centimeter, perform visual inspections of the top guide at grid locations where fuel and blade guides have been removed. Whenever cracking is found, perform an ultrasonic examination of the beam top guide intersection. NMP1 performed the recommended visual examinations in RFO-12 and no cracking was found.

File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit1 al T NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN Scope of Examination - Examine per BWRVIP-26 recommendations.

', ',Tabfe'6-'19,

,, ~ Upper-,'Cori.Grid ';,

Compor'merit Ideritlfication . 'omponent'.:,;; ., Exam Frequency,

.: Descrlptfori" '"'"::,=:: Method RV-13-I Upper Core Grid Later Per BWRVIP-26 6.2.11 SIL 588, Top Guide and Core Plate Cracking SIL 588, Top Guide and Core Plate Cracking, identified that inspections showed significant cracking in the core shroud, top guide and core plate rims. The items were manufactured from type 347 stainless steel, which is a niobium stabilized austenitic stainless steel. GE considers welded 347 SS to have susceptibility to IGSCC. The SIL recommends a visual examination at the next refueling outage.

NMP1 has documented the evaluation of this SIL in Deviation Report/Event Report DER ¹ 1-95-0436.

Scope of Examinations - Visual examinations will be performed as part of the Section XI examinations, Category B-N-2, item B13.40. Augmented examinations will be performed per BWRVIP-26.

. Table 6-,20.

SIL-'SSS Upper, Core',Grid MQp Comporient.fde'ritlffcatiort,::. 'Component,;-::;;. -:,',', 'Exam ',.: ., Frequency',:,',';,

""-'" 'Desc lptl n':""." """'""'

RV-13-I Upper Core Grid Later Per BWRVIP-26 6.2.12 SIL 459, Recirculation Pump Shaft Cracking SIL 459, supplement 2, was issued to inform GE BWR owners of shaft cracking in recirculation pumps manufactured by Byron Jackson (BJ) and by Bingham. BJ manufactured the recirculation pumps at NMP1. Supplement 2, of the SIL provided additional information form that presented in SIL 459, Rev. 0 and SIL 459, Supplement 1. The combined information is summarized below. There is a SIL 459, Supplement 3; however, it is not applicable to NMP1.

SIL 459 was issued in December 1987 and identified the first instance of thermal fatigue cracking in BWR reactor recirculation pump shafts. The recommended action was to perform NDE on the shafts when the pump is disassembled. For additional information see NMPC letter NMP 31096, QA91-U1-080, NMP 34194 and NMP 45971.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA V MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN In SIL 459, Supplement 1, there was no change in the recommendation to perform NDE on the pump shafts. For additional information see NMPC letter QA91-U1-281. This letter indicates that 3 of the 5 shafts have been examined and the remaining 2 will be examined.

In SIL 459, Supplement 2, provided additional information about pumps with heat exchangers and a discussion on shaft vibration. There was no change in the recommendation to perform NDE on the pump shafts when disassembled.

Scope of Examinations - NDE examination of the pump shafts when disassembled for maintenance has been included in this inspection interval.

TaMe 6-21 SIL-459 Pump Shafts Component Identlficatlon Component Exam Frequency Descrfptlon Method 32-187-Shaft Pump 11 Shaft When disassembled 32-188-Shaft Pump 12 Shaft VT-1 When disassembled 32-189-Shaft Pump 13 Shaft VT-1 When disassembled 32-1 90-Shaft Pump 14-Shaft VT-1 When disassembled 32-1 91-Shaft Pump 15-Shaft VT-1 When disassembled 6.2.13 IE Bulletin 80-13 SIL 289 and RISIL 073, Cracking in Core Spray Spargers IE Bulletin 80-13, Cracking in Core Spray Spargers, identified that Oyster Creek and Pilgrim had found cracks in their core spray spargers. Both Oyster Creek and Pilgrim had performed their examinations in accordance with the recommendations in GE SIL 289. Subsequently examinations were performed by NMP1 on four (4) core spray spargers and associated internal piping segment (the section of piping between the inlet nozzle and the vessel shroud) using a remote underwater television camera.

Indications identified on Loop A Core Spray Sparger pipe near Spray Nozzle 23A during RFO-13 were evaluated and submitted to the USNRC on April 21,1995. The indications were evaluated by fracture mechanics and the sparger will not be prevented from performing its function. No repairs are necessary.

NMP1 will continue to perform examinations at each refueling outage and any propagation of indications will be evaluated.

SIL 289, Rev 1, Supplement 1, Revision 1, issued March 15, 1989 identified additional cracking in the core spray sparger. Subsequently RISIL 074, dated November 1, 1995 was issued with additional information.

Scope of Examination - Examinations will be in accordance with BWRVIP-18.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA kl MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

';:,:-'..:;...;;Table'6 <<',22':~~.'"~~';;;"-::: '.

pray, Sparger. ..

Comp orient Compoiient ".'"

Exam, Frequency...Comments

IdentNcatlon Description"."-', Method' RV-06-I Core Spray VT-3 Each Outage Examine per BWRVIP-18.

Sparger RV-07-I Core Spray VT-1 Each Outage Examine Per BWRVIP-18 Lines 6.2.14 IE Bulletin 96-03, Potential Plugging of Emergency Core Cooling Suction Strainers By Debris in Boiling Water Reactors In NMPC letter NMP1L 1151, dated 11/4/98, NMP1 committed to performing an augmented visual examination of the core spray suction strainers in accordance with the recommendations of IE Bulletin 96-03.

Scope of Examination - Perform visual examinations each refueling outage.

.: -; ".i;::-:,",';,";,~P, IEB.96-03.Suction Stralriers

-'Component,. '.':: Exa'rn-,'"

Com'ponent-ldentlflcatlon,'-:

Frequency

.",-: 'Descrlption "- '-:." .Method; 80-09 Ctmt. Spray VT-3 Each outage 80-30 Ctmt. Spray VT-3 Each outage 80-10 Ctmt. Spray VT-3 Each outage 80-39 Ctmt. Spray VT-3 Each outage 81-05 Core Spray VT-3 Each outage 81-06 Core Spray VT-3 Each outage 81-24 Core Spray VT-3 Each outage 81-25 Core Spray VT-3 Each outage File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 El V NIAGARA a U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 6.3 Additional Commitments 6.3.1 NCTS Commitment No. 503783-01, Augmented Examinations of CRD Housing In 1984 NMP1 identified a degraded material condition to Control Rod Drive (CRD) stub tubes. The condition was initially evaluated by the USNRC in June 1984 and that the Roll Repair method used was acceptable. This method limited leakage from the penetrations and assured safe operation of the facility. The US NRC's safety evaluation was supplemented by USNRC Safety Evaluation dated March 25, 1987. Subsequently USNRC correspondence dated August 9,1993 requested that NMP1 develop an augmented inspection program for periodic examination of the critical areas of the CRD housings.

Specifically the stub tube upper J welds and the previously rolled repaired regions. The housings that have not been repaired are included in the program.

Scope of Examination - The augmented program includes ultrasonic examination of the upper-J weld and the roll-repaired areas of the CRD housings prior to and after each roll repair. The UT exam includes the upper-J weld plus a minimum of 1" above and below the upper-J weld to account for the heat affected zone (HAZ).

The UT exam of the rolled area of the housing shall include the volume of base material associated with the roll; and to account for the rolled to unrolled transition areas, will extend at least 1" above and below the roll band.

Each refueling outage a minimum of two (2) previously roll-repaired CRD housings will require a UT exam and only, if previously roll-repaired CRD housings are made available through normal drive maintenance. The housings will be selected from those that are disassembled for maintenance.

Engineering Design Drawing D6340-100-163 shall be updated as required.

In addition bolting will receive a VT-1 on housings that are disassembled. All housings and CRD penetrations will receive a VT-2 exam for evidience of leakage during the system pressure test conducted each refueling outage and the mid-cycle shutdown when drywell is de-inerted.

, '~;-

: ,",;..:.CRDH:Upper"=;-J:,Weld'and Roll"Area'Ltstin j "::;:.

',Component:-.;"::::.:;. ..:;;-":g;":.,';,;;:, Com jonent,-:=':. ':,.'.",'; Exam, Method Frequency:,.:. -,

,Identlflcation'::,',::,"":,':."'." :."'~";-,:,,Qescriptfon':; -- "~",.".

DRV-44-0219 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-0631(NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-1015 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NlAGARA a UMOHAwK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

'"'- -': 4"-- 'abie6'-24"" """':""">>"'-"'-

Component ""

Identification

':.;",-.'-CRDH Upper; J Vfefd and RoII.Area,LIstiiig;

'Compotierif Description, Exam Meth'od

'requency DRV-44-1027 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-1411 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-1803 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-1819 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-2251 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-2647 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3007 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3047 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3051 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3407 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3415 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unlt1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

'.- '::. '"':-'-"':-"= -:.-:;-::.-;;"-': -'"'CRDH Upper - J Nefd'and:;Ro'II,'Area'Listfn'9

,Compon'ent Exam':Meth'od FfeqUeACQ, Ideittfffcstlon,'"-'",,-',',;"'" ":,.';: Description DRV-44-3419 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3431 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3439 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRY-44-3451 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3811 (NC02) CRDH Upper-J Weld UT When disassembled for I and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3831 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-4227 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-4239 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRY-44-4627 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRY-44-4639 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-5019 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage File: NMP1PP6.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

. CRDH Upjii'.;;"O',Vteidand Riff Area Usting '--;.: ";.'::;.: >>" 7, Com'ponent"-: Component ',"=', "'.'*"' Exam Method Fre Identiffcatlori 'escrlptiara ju'en'RV-44-4219 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-5023 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-4231 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-3403 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-0223 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-0235 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage DRV-44-5035 (NC02) CRDH Upper-J Weld UT When disassembled for and Roll Area Bolting - VT-1 maintenance (2) DR Housings each Outage NCTS503783-1 AIISRDH CRDH Exterior Bottom Head I Rx VT-2 All CRD Housings, CRD Pens., Leakage required during each RFO and Mid-cycle shutdown when dry well De-inerted 6.3.2 INPO Recommendation SER 5-85, Thermal Fatigue Cracking at Mixing Points In accordance with the recommendations of the Institute of Nuclear Power Operations (INPO) Significant Event Report (SER) 5-85 the mixing points of systems that experience fluctuations in temperature which could lead to thermal fatigue cracking should be examined each refueling outage. The Reactor Recirculation System has mixing points for the Emergency Cooling and Shutdown Cooling Systems. An examination of these three areas shall be performed during each refueling outage.

Scope of Examination - The base material of the tee between welds 32-WD-004, 32-WD-004A, and 32-WD-005. The examination shall cover 100% of the base material or 12" from the centerline, Fife: NMP1PP6.WPD

0 Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN whichever is less.

The base material of the tee between welds 32-WD-170, 32-WD-171, and 32-WD-172. The examination shall cover 100% of the base material or 12" from the centerline, whichever is less.

The base material of the tee between welds 32-WD-203, 32-WD-204, and 32-WD-205. The examination shall cover 100% of the base material or 12" from the centerline, whichever is less.

Table'8:-,,25 Tee Base Mate rial...

Component-'-";,.

identifi~atio~'.-':Tee'ocation ISO'o.-'uehng "Re '

Outage 2"e'Perio'd, 3~ Period 14.0" Discharge 15D Yes Yes Yes Yes

'2-WD-204-MT 32-WD-004A-MT 12.0" Suction 11S Yes Yes Yes Yes 32-WD-171-MT 28.0" Suction 15S Yes Yes Yes Yes If flaws deeper than those allowed by ASME Section XI, IWB 3514.3 are found, the weld must be evaluated in accordance with the crack evaluation criteria identified Attachment 2 of USNRC Generic Letter 84-11.

File: NMP1PP6.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 7- RELIEF REQUESTS 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 7.3 Relief Request Content .. 7-3 File: NMP1PPT.WPD

Nine Mile Point Nuclear Power Station N MP1-IS!-003 Unit 1 El Y NIAGARA IIU MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

" 4::. ~ "> '>>>>>>~~: "f~<.<<" ~

'ECORD OF'REVISION.j ':0>> ~ ..'"~'AWE.i "" ~ "i "" "',~g@jP%'"

REVISION', DATE '~""' 'FFECTED ; '""',.:.-'::,:.'-",';.':REASON;FOR R EVISION "

'- PAGES' No.':.,-'~

0 September 27, 1999 Entire Updated Inservice Inspection Program Plan Document for the 3" Ten Year Inservice Inspection Interval File: NMP1PP7.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V N(AGARA a 4 woHAwK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 Relief 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 and the status of each Request for Relief submitted to the USNRC as part of this program.

Note: Examination volume or surface area that cannot be examined due to interference by another component or part geometry, a reduction 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 Case N-460, will be documented in the form of a Relief Request per 10 CFR 50.55a (g)(4)(iv).

In cases where parts of the required examination areas cannot by effectively examined because of a combination of component design or current inspection technique limitations, NMPC 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; D. Basis for Relief - describes justification to support the reason relief is being requested; E. Alternative Examination - describes examination(s) or tests that NMPC proposes File: NMP1PP7.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 4V NIAGARA MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN to use in lieu of the current requirements; Implementation Schedule - denotes the interval, period, and/or outage (whichever is applicable), that NMPC proposes to implement the relief; G. Attachments to the Relief - 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 may be added to each Relief.

File: NMP1PP7.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA A@MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 8- ACCEPTANCE STANDARDS Table of Contents 8-1 Record of Revision 8-3 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 .. 8-4 8.4.1 Class 1 Components 8-5 8.4.2 . Class 2 Components 8-5 8.4.3 Class MC Components 8-5 8.5 Acceptance by Engineering Evaluation . 8-5 8.6 Acceptance by Correction 8-6 8.7 Acceptance by Supplemental Examination 8-7 8.8 Acceptance Criteria in Course of Preparation . 8-7 8.9 Additional Examinations . 8-7 8.9.1 Class 1 8-7 8.9.2 Class 2 8-8 8.9.3 Class 3 8-8 8.9.4 Component Supports 8-9 8.9.5 Class MC 8-9 8.9.6 IGSCC Sample Expansion 8-10 File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS - Continued 8.10 Defects Found Outside Section XI Examination 8-10 LIST OF TABLES Table 8-1 Class1 Acceptance Standards 8-11 Table 8-2 Class 2 Acceptance Standards 8-11 Table 8-3 Class 3 Acceptance Standards 8-12 Table 8-4 Class 1, 2, 3 Component Support Acceptance Standards . 8-12 Table 8-5 Class MC Acceptance Standards .... 8-12 File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION REVISION DATE AFFECTED REASON FOR REVISION No. PAGES September 27, 1999 Entire Updated Inservice Inspection Program Plan Document for the 3" Ten Year Inservice Inspection Interval File: NMP1PPB.WPD

0 Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA N V MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 criteria 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 with indications/conditions exceeding the acceptance criteria above shall be in accordance with the 8.2 through 8.6.

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 criteria in 8.1. Repairs are further addressed in Section 10.

Note: The additional examination requirements of IWB-2430, IWC-2430, IWE-2430 or IWF-2430, (as applicable) shall be performed for service induced defects/conditions, 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 alternative to repair requirements of 8.2, the component, or the portion of the component containing the defect/condition shall be replaced. Replacements are further addressed in Section 9.

8.4 Acceptance by Analytical Evaluation Components whose volumetric or surface examination reveals defects that exceed the acceptance criteria of Tables 8-1 through 8-5 are acceptable for continued service without defect removal, repair or replacement if an analytical evaluation meets the acceptance criteria of IWB-3600 or IWC-3600 as applicable, or for Class MC meets the engineering evaluation criteria 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.

Fife: NMP1PPS.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI V NIAGARA 1% U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 acceptance standards and the component is analyzed as acceptable for continued service, the areas containing such defects shall be reexamined during the next three (3) inspection periods of Inspection Program B (IWB-2412-1). 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 seivice-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 during the next inspection period of Inspection Plan B (IWB-2412-1). Provided the defects remain essentially unchanged for the next inspection period, the component examination schedule 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-2412-1). 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.

8.5 Acceptance by Engineering Evaluation Examinations that reveal indications exceeding the acceptance criteria identified in Tables 8-1 through 8-5 will be submitted to Nuclear Engineering for evaluation and disposition:

A. Indications found to be acceptable by the materials and welding criteria specified in the Construction Code and/or Section III Edition applicable to the construction of the component shall be acceptable for continued service.

B. Indications determined to be acceptable by the NMPC Design and/or Manufacturer's Specifications shall be acceptable for continued service.

C. Indications believed to be surface anomalies (e.g., fabrication marks, scratches, surface abrasion, material roughness or other conditions are acceptable for File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NIAGARA a U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN continued service provided the indication is removed by light flapping and/or grinding (surface preparation), and the material removed does not violate the design 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.

Components whose examination results reveal flaws or areas of degradation that exceed the acceptance criteria 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 structural integrity of the containment.

When supplemental examinations of 8.7 are required, if either the thickness of the base metal is reducedby 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.

Nuclear Engineering evaluation and/or disposition may include the need for corrective measures, repairs, analytical evaluation, or replacement, as appropriate.

8.6 Acceptance by Correction Component supports whose examinations reveal conditions described in IWF-341 0(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 mechanicai 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 may be analyzed and/or tested to the extent necessary to substantiate its integrity for its intended service.

File: NMP1PPB.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 4Y NlAGARA MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 Criteria in Course of Preparation If acceptance criteria for a particular component, examination category, or examination method are not specified, defects that exceed the acceptance criteria for materials and welds specified in the Construction Code and/or Section III Edition applicable to the construction of the component shall be evaluated to determine disposition.

8.9 Additional Examinations 8.9.1 Class 1 The additional examination requirements identified in IWB-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:

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.

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.

File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP't-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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.

8.9.2 Class 2 The additional examination requirements identified in IWC-2430 shall be performed for service induced detects/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. 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 detects/conditions, the examinations shall be further extended to include remaining number of similar components or areas within the same examination category.

c. 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.

8.9.3 Class 3 There are no additional examination requirements identified in IWD-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 exists, additional examinations shall include the following:

An additional number of components 'or areas, within the same examination item number, system, and line, and will include the following:

The next component or area, upstream and downstream of the 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.

File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NIAGARA A@MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECT(ON Date: September 27, 1999 PROGRAM PLAN Additional examinations will be performed before the end of the outage.

8.9.4 Component Supports The additional examination requirements identified in IWF-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 component supports immediately adjacent to the initially identified support with the defect/condition.

c. Additional supports equal in number and similar in type, design, and function to those initially examined.

d. When these additional examinations reveal defect/conditions, the remaining supports within the item number, system or line 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.

8.9.5 Class MC The additional examination requirements identified in IWE-2430 shall be performed for any one inspection that reveals flaws or areas of degradation as follows:

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.

b. When additional flaws or areas of degradation that exceed 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), NMPC 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 File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-Isl-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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; B. An evaluation of each area, and the results of the evaluation, and; C. A description of necessary corrective actions.

8.9.6 IGSCC Sample Expansion 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 welds in that category should be inspected, approximately equal in number to the original sample. This additional sample should be similar in distribution (according to pipe size, system, and location) to the original 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.

Significant crack growth for overlayed welds is defined as crack extension to deeper than 75% of the original wall thickness, or for cracks originally deeper than 75% of the pipe wall, evidence of crack growth into the effective weld overlay.

b. Significant crack growth for Sl 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).

C. Category D weld expansions are limited to piping systems where cracking was identified.

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 8-1 through 8-4, as applicable.

File: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

'-'::."';" '>A ': " " '-; "-"'TABLES-1'-'L'ASS 1 'ACCEPTANCE'STANDARDS-' .."- """

""," 'CATEGORY,'-.":".::. -.:~ "~."- " " '-':,"'. '"i'8 "

EXAMINATION;j;:~r~";;;,>'-':--CQMPONENTORPART;:,EXAMINED.:,-=-'--;=-':-;"" ';:",:;

'. '"--" ",""'l".;"'i.-

ACCEPTANCE,

-:S1ANDARD

'k'-A Welds in Reactor Vessels IWB-3510 B-B Welds in Other Vessels IWB-3510 B-D Vessel Nozzle Welds IWB-3512 B.E Partial Penetration welds in Vessels IWB-3522 B-F, B-J Dissimilar and Similar Metal Welds in Pi in IW8-3514 B-G-1 Boltin >2'dia. IWB-3515/351 7 B-G-2 Boltin <2'dia. IW8-3517 B-H. B-K-1 Inte ral Attachments for Pi in, Valves, Pum s 8 Vessels IW8-3516 B-L-1. B-M-1 Welds in Pum s 8 Valves IWB-3518 B-L-2. B-M-2 Pum Casin s & Valve Bodies IWB-3519 B-N-1 B-N-2 Interior Surfaces 8 Internal Components of Reactor Vessels I W8-3520.1 B-N-3 IW8-3520.2 B-0 Control Rod Drive Housin Welds IW84523 B-P Pressure Retainin Bounda IWB-3522 B-Q Steam Generator Tubin IWB.3521 TABLE 8 CLASS 2 ACCEPTANCE STANDARDS EXAMINATION ACCEPTANCE CATEGORY COMPONENT OR PART EXAMINED STANDARD C-A Welds in Pressure Vessels I WC-3510 C-B Nozzle Welds in Vessels IWC-3511 C-C Inte ral Attachments for Vessels. Pi in, Pum s and Valves IWC-3512 C-D Boltin IWC-3513 C-F-1,C.F-2 Welds in Pi in IWC-3514 C-G Welds inPum sand Valves IWC-3515 C-H Pressure Retainin Com onents IWC-3516 Fife: NMP1PPB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unlt1 H Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN VAEILEse- CLASS 3 ACCE TANCE STANDARDS EXAMINATION COMPONENT OR PART EXAMINED ACCEPTANCE CATEGORY STANDARD D-A Integral Attachments (VT-1) Mlg. Code 8 Applicable nd fd ASME Section XI Acceptance Standard in course of preparation.

TABLE 8< - COMPONENT SUPPORT ACCEPTANCE STANDARDS EXAMINATION COMPONENTOR PART EXAMINED ACCEPTANCE

"'ATEGORY'-A STANDARD Supports IWF-3410 TABLE'SW"-".CL'ASS'C ACCEPTANCE STANDARDS ' '

EXAMINATION COMPQNENTOR PART EXAMINED ACCFPTANCE CATEGORY STANDARD E-A Containment Surface IWE-3510 E-B Pressure Retainin Welds IWE-3511 E-C Containment Surfaces requiring Augmented Examinations IWE-3512 E-D Seals, gaskets, and moisture barriers IWE-3513 E-F Pressure retaining dissimilar metal wefds IWE-3514 E-G Pressure retaining bolting IWE-3515 E-P All ressureretafnin corn nents A endix J Fife: NMP1PPS.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 EI Y NIAGARA H I!MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 9- REPAIRS, REPLACEMENTS, MODIFICATIONS Table of Contents 9-1 Record of Revision 9-3 9.0 REPAIRS, REPLACEMENTS AND MODIFICATIONS 9-4 9.1 Repairs 9-4 9.1.1 Exemptions 9-4 9.1.2 Repair Operations 9-5 9.1.3 Pressure Testing 9-5 9.1.4, Baseline Examinations 9-5 9.2 Replacements 9-5 9.2.1 Replacement Operations .. 9-5 9.2.2 Engineering Approval .. 9-6 9.2.3 Pressure Testing 9-6 9.2.4 Preservice Examinations 9-7 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 9-7 9.3.3 IWE Class MC Components Examination Requirements .. 9-8 9.3.4 IWE Class MC Components Examination and Pressure Test Requirements . 9-8 9.3.5 IWE Class MC Components Examination Qualifications 9-8 File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS (Continued) 9.4 Modifications 9-8 9.5 Evaluation 9-8 9.6 Access 9-9 9.7 Construction Codes 9-9 9.8 Authorized Nuclear Inservice Inspector 9-9 9.9 Implementations 9-9 File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NlAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION REVISION AFFECTED REASON FOR REVISION No. PAGES September 23, 1999 Entire Updated Inservice Inspection Program Plan for the 3" Ten Document Year Inservice Inspection Interval File: NMP1PP9.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA

&MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 9.0 REPAIRS, REPLACEMENTS AND MODIFICATIONS Scope This section establishes the program by which the Nine Mile Point Nuclear Power Station will define the managerial and administrative controls over the implementation and completion of repairs, replacement (modifications) and maintenance of items that require subsequent inservice examinations 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," hereafter referred to as the Code, 1989 Edition, no Addenda, for the Repair, Replacement (modification), and Installation of Replacement Activities at the Nine Mile Point Nuclear Power Station. The repairs and replacements for components which are within the provisions of ASME Boiler and Pressure Vessel Code,Section XI, 1992 Edition with the 1992 Addenda of Subsection IWE, "Requirements for Class MC Metallic Liners of Class CC Components of Light-Water Cooled Power Plants," incorporated by reference in 10 CFR 50.55a, will be controlled in accordance with this program.

The repairs and replacements for which these provisions apply are restricted to those performed on systems and components classified Class 1, 2, 3, and Class MC pressure retaining components and their integral attachments.

ASME Section XI Repairs and Replacements shall be conducted in accordance with the Niagara Mohawk Power Corporation implementing Repair/Replacement procedures.

9.1 Repairs Repairs for which these provisions apply are restricted to those performed on systems and components classified Quality Group A, B or C, (Class 1, 2, 3) and Class MC pressure retaining components and their integral attachments. Repairs shall be performed in accordance with NMPC's NMP1 Design Specification and the original Construction Code of the component or system.

However, as allowed by paragraph IWA-4120, later Editions and Addenda of the Construction Code or of Section III, 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 Editions and Addenda of Section XI at the time of the planned repair have been incorporated by reference in amended regulations of the regulatory authority 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 are exempt 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 NMPC Quality Assurance Program.

File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H T NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 the complete repair cycle and shall include 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 procedures and postweld heat treatment, and the non-destructive examination program to be used after the repair.

(d) Evaluation as described in 9.5.

(e) The repair programs shall be subject to review by the enforcement and regulatory authorities having jurisdiction at the plant site.

9.1.3 Pressure Testing After repairs by welding on the pressure retaining boundary, a pressure test shall be performed in accordance with the requirements of the Inservice Pressure Testing Program, Document NMP1-PT-003.

9.1.4 Baseline Examinations When required by Section Xl, the repaired area shall be reexamined to establish a new preservice record. The examination shall include the method that detected the flaw.

9.2 Replacements Replacements are performed using approved procedures or instructions in accordance with IWA-7000.

All procedures for the installation of renewal, spare, and replacement parts shall be in accordance with IWA-4100. Alternatively, subsequent Editions and Addenda of Section XI may be used for replacement provided these Editions and Addenda are acceptable to the enforcement and regulatory authorities having jurisdiction at the site.

9.2.1 Replacement Operations The program for replacements shall include the following:

(a) The applicable Construction Code to which the original item was constructed File: NMP1PP9.WPD

Nirie Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN (b) The existing design requirements (if the original item 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 alternative of (c) below is adopted).

(c) Alternatively, an item to be used for replacement may meet all or portions of the requirements of later editions of the Construction Code or Section III, when the Construction Code was not Section III, 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.

(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 Pressure testing shall be performed on replacements in accordance with the Inservice Pressure Testing Program, Document NMP1-PT-003.

File: NMP1PP9.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 9.2.4 Preservice Examinations Prior to the system's return to service, a preseivice inspection shall be made in accordance with IWB-2200, IWC-2200, IWD-2200, IWF-2200, IWE-2200.

Post-Work testing and pressure testing will be performed as delineated in applicable Plant Procedure(s).

9.3 Repair/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 Nine Mile Point ¹clear Power Station, as the NMP1 Plant uses a steel primary containment.

9.3.1 IWE Exempt Components The following components (or parts of components) are exempted from the examination 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 inaccessible 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:

(a) The Primaly Containment System at NMP1, which is defined as a General Electric Mark 1 Pressure Suppression Containment.

File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NlAGARA IiU MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 9.3.3 IWE Class MC Components Examination Requirements (a) Examination requirements 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 Examinations 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 Inservice Pressure testing Program, Document NMP1-PT-003.

9.3.4 IWE Class MC Examination and Pressure Test Requirements Examination and Pressure Test requirements shall be performed in accordance with the requirements defined in IWE-2500, Examination and Pressure Test Requirements, IWE-5200, System Test Requirements, applicable Code Cases that are approved for use and accepted for implementation within the ISI Repair/Replacement Program, or approved USNRC 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 s as applicable.

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 acceptance and evaluation of IWB, IWC, IWD, IWF, and IWE components/systems.

File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA a @MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 9.6 Access Accessibility for inservice inspection was considered during the design of the reactor vessel and insulation to ensure adequate working space and access 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 considered by Nuclear Engineering as part of the processing of design or 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 with 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.

Welding activities shall be performed in accordance with NMPC Weld Control Manual.

9.8 Authorized Nuclear Inservice Inspector The services of an Authorized Nuclear Inservice Inspector (ANII) shall be used when making all repairs/replacements. The repair planner shall be made available for review by the ANII for all welded repairs/replacements. The ANII shall determine what hold points, if any, are required to monitor the repair/replacement activity. NMPC shall notify the ANII prior to starting the repair, replacement or modification and keep the 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.

File: NMP1PP9.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 4Y MOHAWK NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS SECTION 10- RECORDS 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 10.4 Summary Report Submittal 10-4 10.5 Reporting Requirements for IGSCC 10-4 10.6 Reporting Requirements for Class MC 10-4 10.7 Reporting Requirements for NUREG 0619 10-4 LIST OF FIGURES 10- 1 NIS-1 Owners'ata Report for Inservice Inspections 10-6 10 - 2 NIS-2 Owners'eport for Repairs and Replacements . 10-8 10- 3 OAR-1 Owner's Activity Report . 10-10 Table 1-Abstract of Examination and Tests .. 10-11 Table 2-Items with Flaws or Relevant Conditions that Require Evaluation for Continued Service 10-12 Table 3-Abstract of Repairs, Replacements or Corrective Measures Required for Continued Service 10 -13 10- 4 NIS-2A Repair/Replacement Plan Certification Record 10- 14 File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD QF, REVISION REVISION 'AFFECTED REASON FOR'RFVISION No. -. "<<", PAGES 0 September 27, 1999 Entire Updated Inservice Inspection program Plan for the 3"0 Ten Document Year Inservice Inspection Interval File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V'lAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 10.0 RECORDS This section provides the requirements for the preparation and submittal of Inservice Inspection records and reports as required by IWA-6000.

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 Summaly 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 alternate to the requirements of IWA-6000, NMPC has submitted a Request for Relief to implement 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,Section XI, Division 1".

\

Each Summary Report will contain the following:

a. Refueling outage number (when applicable).

b. Owner's Data Report for Inservice Inspections, Form NIS-1, Figure 11-1 or Form OAR-1, Figure 10-3.

c. Owner's Data Report for Repairs or Replacements Form NIS-2 or Form NIS-2A, Figure 11-2 and Figure 11-3.

Subject to Request for Relief: ISI-S 10.3 Cover Sheet Each Summary Report will have a cover sheet providing the following information:

a. Date of document completion

b. Name and address of Owner Name and address of generating plant Name and number designation of the plant Commercial service date for the unit File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 10.4 Summary Report Submittal Ninety (90) days following the units return to service, NMPC shall forward a Summary Report of the Inservice Inspection activity to the United States Nuclear Regulatory Commission in accordance with IWA-6220, or the requirements specified in Code Case N-532, upon USNRC approval.

10.5 Reporting Requirements for IGSCC If any cracks are identified that do not meet the criteria for continued operation without evaluation given in Section Xl 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)(x), NMP1 shall provide the following in the Inservice Inspection Summary Report required by IWA-6000:

(1) A description of the type and estimated extent 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 necessaly corrective action.

10.7 Reporting Requirements for NUREG 0619 Within 6 months of completing an outage at which these examinations were performed, NMP1 must submit a detailed report to the NRC Regional Director, IE, with copies to the Director, IE, and Director, NRR; discussing the inspections performed, including:

The number of startup/shutdown cycles since the previous inspection, and the total number of cycles. This includes cycles accumulated during the initial startup and testing of the plant.

2. A Summary of methods used and results of previous inspections, including maximum crack depth, number of cracks found, and number of startup/shutdown cycles between such inspections.

A description of any additional system changes or changes in operating procedures that will affect Feedwater flow or temperature and that should be considered in predicting future cracking tendencies based on past history.

A detailed discussion of the inspection results, including a complete description of cracking locations, dimensions, and a crack profile. The USNRC, if available, requests drawings and photographs.

Information regarding the results of leakage monitoring. However, the USNRC staff must be File: NMP1PP10.WP D

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y N(AGARA a U woHAwK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN informed immediately if on-line leakage monitoring, during operations, discloses any leakage from welded spargers, or leakage on the order of 0.3 gpm through single-sleeve/single-piston-ring spargers or triple sleeve spargers.

Information regarding all UT crack-like indications and any subsequent PT indications. Information regarding UT techniques should be as precise and as extensive as possible in order that it may be of benefit in future inspections.

File: NMP1PP10.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 EI Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-1 FORM NIS-1 OWNERS'ATA REPORT FOR INSERVICE INSPECTIONS As re uired b the Provisions of the ASME Code Rules

1. Owner: W (Name and Address of Owner)

2. Plant: Nin i r inP Yr 1 (Name and Address of Plant)

3. Pinot Unit:~ 4. Owner Oertlficete of Authorlzetlon (tf required)

5. Commercial Service Date:~~QQQQ 6. National Board Number for Unit: ~~

7. Components Inspected:

r ompnonnent',or,

'rtentuan'n'Ced'-.'-;.'"'anufactui'er'or'. Manufactu'rer'o'r', State or'Pro'vfn'c'e Nartional Boar'd

.. 'nstaller Serial No. No.

c'nstaller No.'

Note: Supplemental sheets in form of lists, sketches, or draviings may be used, provided (1) size is 8 th 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.

File: NMP1PP10.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-1 (Continued)

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 concerning status of work required for inspection Plan.

14. Abstract of Results of Examinations and Tests.

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 Authorization NO. (If applicable) Expiration Date Date 19 Signed by Owner CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and/or the State or Province of and employed by have inspected the components described in this OWNERS'ata Report during the period to , and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in the Owners'ata 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, concerning the examinations, and 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.

Commission Number Inspector's Signature National Board, State, Province, and Endorsements Date: ~ 19 File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1<<ISI-003 Unit 1 Y NIAGARA N U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-2 FORM NIS-2 OWNERS'EPORT FOR REPAIRS OR REPLACEMENTS As Re uiredb the Provisions of the ASME Code Section Xi

1. Owner: i n Date:

Name B I v Y 1 Sheet of Address

2. Plant: Unit:

Name (Repair Organization P.O. No., Job No., etc.)

3. Work Performed By: Type Code Symbol Stamp:

Name Authorization No.:

Expiration Date:

e Address

4. Identification of System:

5. (a) Applicable Construction Code: Edition Addenda Code Case (b) Applicable Edition of Section XI Utilized for Repair or Replacement 19

6. Identification ofcom onentsRe alrodor Re lacedandRe IacementCom onents trams ol ttsmoof Msnutacturer ttatlonat othor Year Replaced ASMB Component atsnufacturor Serial kumber Board tdontlflcstlon Built ttepalred Code tto. Stamped ne lscement ear No

7. Description of Work:

t

8. Test Conducted: HydrostatIc [ ] Pneumatic [ ] Nominal Operating Pressure [ ]

OTHER [ ] Pressure: Psl Test Temp: Degree F NOte: Supplemental sheets in form of lists, sketches, or drawings may be used, provided (1) size is 8 tra x11 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.

File: NMP1PP10.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-2 (CONTINUED)

NIS-2 REPORT CONTINUED

9. Remarks:

Appficabfe Manufacturer's Data Reports to be attached CERTIFICATE 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'esignee, Title CERTIFICATE OF INSERVICE INSPECTION to, I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State or Province of employed by of have inspected the components described in this Owners'eport during period and state that to the best of my knowledge and belief, the Owner has performed examinations and taken corrective measures described in this Owners'eport in accordance with the requirements of the ASME Code,Section XI.

By signing this certification neither the Inspector nor his employer makes any warranty, expressed or and implied, concerning the examinations and corrective measures described in this Owners'eport.

Furthermore, 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 DATE: ,19 File: NMP1PP10.WPD

Nine Mile Pofnt Nuclear Power Station NMP1<<ISI-003 Unit 1 hl T NIAGARA l1 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-3 FORM OAR-1 OWNER'S ACTIVITYREPORT As required by the provisions of the ASME Code Case N-S32 Page of Report Number Owner

~

(Name and Address of Owner)

Plant I rP (Name and Address of Plant}

Plant Unit Commercial Service Date Refueling Outage Number Current Inspection Interval (1st, 2nd, 3rd, 4th, Other)

Current Inspection Period (1st, 2nd, 3rd)

Edition and Addenda of Section XI applicable to the Inspection Plan Date and Revision of inspection Pian n and Addenda of ASME Section XI applicable to Repairs and Replacements, if different than spection Plan CERTIFICATE OF CONFORMANCE I certify that the statements made in this Owner's Activity Report are correct, and that the examinations, tests, repairs, replacements, evaluations and corrective measures represented by this report conform to the requirements of Section XI.

Certificate of Authorization No. Expiration Date (If applicable)

Signed Date (Owner's Representative and Title)

CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State or Prownce of and employed by of have inspected the items described in this Owner's Activity Report, during the period 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 requirements of Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or implied, concerning the examinations, 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 damage or a loss of any kind arising from or connected with this inspection.

Commissions Inspector's Signature National Board, State, Province, and Endorsements Date 19 File: NMP t PP10.WP0

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 T NIAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-3 (Continued)

TABLE 1 ABSTRACT OF EXAMINATIONAND TESTS As required by the provisions of the ASME Code Case N-532 Page of TOTAL TOTAL TOTAL TOTAL EXAMINATIONS CODE EXAMINATIONS EXAMINATIONS EXAMINATIONS CREDITED (%) To EXAMINATION REOUIRED FOR REQUIRED FOR THIS CREDITED (%) FOR THE DATE FOR THE CATEGORY INTERVAL PERIOD PERIOD INTERVAL REMARKS NOT APPUCABLE TO NINE MILE POINT NUCLEAR POWER STATION UNIT I 37 BG I NOT APPLICABLE TO NINE MILE POINT NUCLEAR POWER S TATION UNIT 1. THIS CATEGORY HAS BEE NCOM BINE 0 WITH CATEGORY B.K PER COO E CASE N.509 BJ CODE CASE N-509 APPUES NOT APPLICABLE TO NINE MILE POINT NUCLEAR POWER STATION UNIT I BMI B M.2 17 WHEN DISSEMBLED B.N.2 76 B.N-3 NOT APPUCABLE T0 NINE MILE POINT NUCLEAR Po WER STATION UNIT I C.C 105 CODE CASE N.509 APPUES CF I 62 C-F-2 C.G DA 487 CODE CASE N 509 APPUES CODE CASE N 491 ~ I APPLIES File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-3 (ContinLted)

TABLE 2 ITEMS WITH FLAWS OR RELEVANT CONDITIONS THAT REQUIRE EVALUATIONFOR 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 EXAMINATIONOR TEST ES OR NO B-J B9.11 32-WD-050 Valve to Pi e IWB-3600 Yes B-J B9.11 32-WD-126 N1D Noz. SE to Elbow IWB-3600 Yes B-J B9.11 32-WD-168 N1E Noz. SE to Elbow IWB-3600 Yes B-J B9.11 . 32-WD-046 N1B Noz. SE to Elbow IWB-3600 Yes B-J B9.11 32-WD-086 N1C Noz. SE to Elbow IWB-3600 Yes B-A B1.12 RV-WD-140 Lower Vessel Lon Weld IWB-3600 Yes B-A B1.30 RV-WD-099 Shell to Flan e Weld IWB-3600 Yes File: NMP1PP10.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NlAGARA H U MOHAWK THIRDINSERVICEINSPECTIONINTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN 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 REPLACEMEI4T CODE OR CORRECTIVE ITEM OF EXAMINATIONOR TEST DATE PLAN CLASS MEASURE DESCRIPTION WORK (YES OR NO) COMPLETE NUMBER File: NMP1PP10.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NlAGARA a UMOHAwK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN FIGURE 10-4 FORM NIS-2A REPAIR/REPLACEMENT PLAN CERTIFICATION RECORD As required by tho provisions of tho ASMS Codo Case N-532 Pago of asaaaaaaaaa saaaaasass ssssssssaaaaaassasaaassassas aaaaaaaasassasassssaaaaasaaasaasassassasaaaaasaasaaassaaasaasasaa aaasaaaa OWNER'S CERTIFICATE OF COMPLIANCE I certify that the represented by Repair/Replacement repair or replacement Plan Number conforms to tho requirements of Section XI.

Type Code Symbol Stamp Certificate of Authorization No. Expiration Date Signed Date 19 Owner or Owner's designee, Title CERTIFICATE OF INSERVICE INSPECTION I, the undersigned, holding a valid commission issued by the National Board of Boiler and Pressure Yesset Inspectors and tho State or Province of and employed by of have inspected tho components described in Repair/Replacement Plan No. during the period to

~ and state that to tho best of my knowledge and belief, the Owner has performed all the activities described fn tho Repair/Replacement Plan in accordance with the requirements of Section XI.

By signing this certificate neither the Inspector nor his employer makes any warranty, expressed or imptied, concerning theactivities described in the Repair/Replacement Plan. Furthermore, neithertho Inspector norhis employershallbo liable inanymannerforany 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 Date 19 File: NMP1PP10.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 hI V NIAGARA a lfMOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX A - CLASS 1

SUMMARY

TABLES Table of Contents A-1 Record of Revision A-2 ASME CODE CLASS 1 SECTION XI

SUMMARY

TABLES A-3 Examination Category B-A Examination Category B-D 2-5 Examination Category B-E 6-6 Examination Category B-F 7-9 Examination Category B-G-1 10- 14 Examination Category B-G-2 15- 17 Examination Category B-J 18- 21 Examination Category 8-K . 22- 24 Examination Category B-L-1, B-L-2 25- 26 Examination Category B-M-1, B-M-2 27- 29 Examination Category B-N-1, B-N-2. B-N-3 30- 32 Examination Category B-0 33- 33 File: APPENDIXA.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA H U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

,':'::.",':" .." ":;;:":::.':,"'".'2.'-.":;; "",",;... "RECORD'OF REVISION"';:,'.-" .';:.:..'.; "'".:,',:.:-",;:;

,-- .,~-.: -DATE-.:-: <<;-<< AFFECTED ,REASON'FOR;REVISION "

'EVISION>

0 September 27, 1999 Entire Updated Inservice Inspection Program Plan Document for the 3" Ten Year Inservice Inspection Interval File: APPENDIXA.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NlAGARA U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ASME CODE CLASS 1

SUMMARY

TABLES File: APPENDIXA.WPD

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 1 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING WELDS IN REACTOR VESSEL ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM (J OF NO. SCHEDULED/COHPLETED iTEN (( ITEM DESCRIPTION METHOD DESCRI('TlvN COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS Bl.ll CIRCUMFERENTIAL SHEL( VOLUMETRIC Reactor Vessel 4 0 0 0 0 0 0 0 All Welds, 100%

WELDS Welds 0.0% 0.0% p 0(( weld length, Deferral permissible Bl . 12 LONGITUDINAL SHELL WELDS VOLUMETRIC Reactor Vessel 12 12 0 0 0 0 12 0 All Welds, 100%

Welds 0. 0'1 p pg lpp pg weld length, deferral permissible.

Bl . 21 CIRCUMFERENTIAL HEAD VOLUMETRIC Reactor Vessel 3 3 0 0 1 0 2 0 Access. length of WELDS Welds p 0$ 33 3Q lpp pg All welds, lppi weld length, Deferral permissible.

/

HER1DIONAL HEAD WELDS VOLUHFTRIC R~a"t ~ t V seel ~~ 6 0 2 G 14 0 Accessible length of All weids, )00%

weld length, Deferral permissible.

Bl . 30 SHELL-TO-FLANGE WELD VOLUMETRIC Reactor Vessel 2 2 1 0 0 0 1 0 1008 of weld Welds 50. 0% 5p pg lpp pi length, Partial Deferral Permissable See footnote (3) and (4) ~

Bl.40 HEAD-TO-FLANGE WELD VOLUMETRIC Reactor Vessel 1 3 1 0 1 0 1 0 100% of weld SURFACE 33 3$ 66 6g lpp pg length, Partial Deferral Permissable see footnote 3.

Bl.b! REPAIR WELDS>>BELTL(NE REGION

ATE lvRY TvrAL: 4 ( (. 8 v 4 0 30 0 19.0i 8.5'4 1G0.0%

NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 2 DATE: 10/26/99 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A FULL PENETRATION WELDS OF NOZZLES IN VESSELS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM ll OF NO. SCHEDULED/COMPLETED ITEM DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COMMENTS ITEM ii B3.100 REACTOR VESSEL-NOZZLE VOLUMETRIC Control Rod 1 1 1 0 0 0 0 0 All Nozzles, 25ii INSIDE RADIUS SECTION Drive pp pii lpp p, lpp pq to 50% 1st period, Remainder by end of Interval VOLUHETRIC Emergency 2 2 1 0 0 0 1 0 All Nozzles, 258 Condenser Supply 50.0% 50.0% 100.0% to 50% 1st period, Remainder by end of Interval VvLUME1 ii1 i.'e vows < .: ."ys a vni u 0 0 All Nozzles, 25i

0. 01 p pii Ipp.pii to Spn 1st period, Remainder by end of Interval VOLUHETRIC Main Steam 2 2 1 0 0 0 1 0 All Nozzles, 250 System 50.0% 50.0% 100.05 to 50~ 1st period Remainder by end of Interval VOLUMETRIC React.ur Core 1 0 1 0 0 0 All Nozzles, 25%

Spray 50.0'h 100.0% 100.0% to 50~ 1st period Remainder by end of Interval VOLUHETRIC Reactor Head 0 0 1 0 0 0 All Nozzles, 25%

Vent p pa Ipp.pg Ipp.pg to 50% 1st periodi Remainder by end of Interval VOLUHETRI( Rea-tit 0 0 i 0 All Nozzles, 251 instrumentation 0.0% p.pli lpp.pa to 50% 1st period, Remainder by end of Interval VOLUHETRIC Reactor 5 5 1 0 2 0 2 0 All Nozzles, 25%

Recirculation 20.0ii 60.0li 100.0% to 50% 1st period, Discharge Remainder by end of Interval VOLUMETRIC Reactor 5 5 2 0 1 0 2 0 All Nozzles, 25%

Reci r~u't lair 6p pq lpp pii to 50% 1st period, Suction Remainder by end

0 DATE: 10/26/99 tkINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 3 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A FUIL PENETRATION WELDS OF NOZZLES IN VESSELS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEH tt OF NO. SCHEDULED/COMPLETED ITEM (t ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COHMENTS B3.100 of Interval VOLUHETRIC Reactor Vessel 17 17 5 0 5 0 7 0 All Nozzles, 258 Nozzles 29 4$ 58 8() lpp p(t to 50() 1st period/

Remainder by end of Interval ITEM TOTAL: 40 40 12 0 10 0 18 0 30.00 55.0(t 100.0t B3. I!0 PRESSURIZER-NOZZLE-TO-VE N/A SSEL WELDS PRESSURIZER-NOZZLE INSIDE RADIUS SECTION B3.130 STEAN GENERATORS

( PRIMARY SIDE)-NOZZLE-TO-VESSEL WELDS 40 STEAH GENERATORS (PRIHARY SIDE)-NOZZLE INSIDE RADIUS SECTION B3.90 REACTOR VOLUMETRIC Control Rod 1 1 1 0 0 0 0 0 All Nozzles, 25%

VESSEL-NOZZLE-TO-VESSEL Drive lpp p(t lpp p(t lpp 0() to 50% 1st period WELDS Remainder by end of Interval VOLUMETRIC Emergency 2 2 1 0 0 0 1 0 All Nozzles, 254 Condenser Supply 50.0t 5p p(t lpp p(t to 50% 1st Period, Remainder by end of Interval 0 4 0 All Nozzles, 25%

to 50% 1st period, Remainder by end o! Interval VOLUMETRIC Hain Steam 1 0 0 0 1 0 All Nozzles, 25'4 System 5p pg 5p 0() lpp 0() to 50% 1st period/

Remainder by end of Interval VOLUMETRIC Reactor Core 2 2 1 0 1 0 0 0 All Nozzles, 25%

Spray 50.0() 100.0% 100.0% to 50~ 1st Per"od Remainder by end

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 4 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUHMARY TABLE A FULL PENETRATION WELDS OF NOZZLES IN VESSELS ASME NO. OF COHPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM t) ITEN DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B3.90 of Interval VI>,. 'HETRi ' ~ "i < t lt~ >>c I 0 0 I 0 0 0 All Nozzles, 25%

Vent 0 0(t 100.0% 100.0% to 50% 1st Period, Remainder by end of Interval VOLUMETRIC Reactor 1 1 0 0 0 0 1 0 All Nozzles, 25(t Instrumentation 0.0% 0.0% 100.0() to 505 1st period, Remainder by end of Interval VOLUHETRIC Reactor 5 5 I 0 2 0 2 0 All Nozzles, 25%

Recirculation 20.0(t 60.0% 100.0% to SOS 1st period/

Discharge Remainder by end of Interval VOLUMETRIC Reactor 5 5 2 0 I 0 2 0 All Nozzles, 25%

Recirculation 4p pg 6p 0() lpp p(t to 50% 1st period/

Suction Remainder by end of Interval VOLUHETR/i: R<<,s"t. ". V<<ss<<. :7 5 0 5 0 t 0 All Nozzles, 25%

100 0(i tO 50(t 1st period, Remainder by end of Interval ITEM TOTAL: 40 40 12 0 10 0 18 0 30.0% 55.0% 100.0(i B3. 150 HEAT EXCHANGERS (PRIHARY N/A SIDE)-NOZZLE-TO-VESSEL WELDS r

83. 160 HEAT EXCHANGERS (PRIMARY N/A SIDE) -NOZZLE INSIDE RADIUS SECTION B3. 10 REACTOR VESSEL- N/A tiOZZLE-TO-VESSEL WELDS B3.20 REACTOR VESSEL-NOZZLE ti/A INSIDE RADIUS SECTION PRESSURIZER-NOZZLE-TO-VE SSEL WELDS Bv.4u i'RESSURIZER-NOZZLE NIA INSIDE RADIUS SECTIVit

0 DATE: 10/26/99 NIHE MZLE POINT NUCLEAR PLAHT UNIT 1 PAGE: 5 REVISIOH: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION'89 CLASS 1 SECTIOH XZ

SUMMARY

TABLE A FULL PENETRATION WELDS OF NOZZLES ZN VESSELS (INSPECTION ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM I ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B3.50 STEAM GENERATORS N/A (PRIMARY SIDE)-NOZZLE-TO-VCSSEL WELDS B3.60 STEAM GENERATORS H/A (PRIMARY SIDE)-NOZZLE INSIDE RADIUS SECTION B3.70 HEAT EXCHANGERS (PRIMARY SIDE)-NOZZLE-TO-VESSEL NELDS B3.80 HCAT CXCHANGCRS (PRIMARY N/A SIDE)-NOZZLE INSIDE RADIUS SECTIOH CATEGORY TOTAL: 80 80 24 0 20 0 36 0 30.0(j 55.0% 100.0(j

DATE: 10/"6/99 II.'NE H.'LF. r'OIIIT liU::.EAR I'LAN>'NIT PAGE: 6 REVISION: QQ I:ISERVI ..! I!ISIE'" Io!! P:.AN i>lR '."IIE TH:RD INTERVAI. CODE EDITION: E89

'.l>SS: Se: "OiIi~ SUHMARY TABLE A PRESSURE RETAINING PARTIAL PENETRATION WELDS IN VESSELS ASHE NO. OF COMPONENTS SEC XI EXAM SYSTEH I OF NO. SCHEDULED/COMPLETED ITEH 8 ITEH DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COMMENTS B4.10 PARTIAL PENETRATION I4ELDS B4.11 VESSEL NOZZLES N/A B4.12 CONTROL ROD DRIVE VISUAL Reactor Vessel 1 1 0 0 0 0 1 0 258 of nozzles, NOZZLES Nozzles 0.0% 0.0% 100.0% Deferral permissible B4. 13 INSTRUMENTATION NOZZLES VISUAL Reactor Vessel 1 1 0 0 0 0 1 0 25% of nozzles, Nozzles Q.QI, Q.QII 100.0II permissible

~~

'I ~ I'hI;SSI!8,1 "ER-IIEATER ~

!I I'ENETRATION WELDS CATEGORY TOTAL: 2 2 0 0 0 0 2 0 0.01 0.0% 100.00

DATE: 10/26/99 NINE HILE POIttT NUCLEAR PLANT UNIT 1 PAGE:

REVISION: 00 It)SERVICE INSPECTION PLA!$ FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTIOtt XI

SUMMARY

TABLE A PRESSURE RETAINING DISSIMILAR METAL WELDS ASHE NO. OF COHPONENTS SEC XI EXAM SYSTEH OF NO. SCHEDULED/COMPLETED ITEM ITEN DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS REACTOR VOLUHETRIC Emergency 2 2 1 0 0 0 1 0 All welds May VESSEL-NOZZLE-TO-SAFE SURFACE Condenser Supply coincide with 0 0% 50 0% 100 pl END BUTT WELDS NPS 4 or Category B-D LARGER examinations VOLUMETRIC Reactor Core 2 2 1 0 1 0 0 0 All welds May SURFACE Spray coincide with 5p.pg Ipp.pnt lpp pg Category 8-D examinations VOLUHETRIC Reactor Head 0 0 1 0 0 0 All welds Hay SURFACE Vent O.pit 100.0't 100.0%

Category B-D examinations VOLUMETRIC Re~cV>r 1 1 v v u v 1 0 All welds Hay SURFA L 1nst i will") L.i< n-uI coincide with

~

0.0% 100.0%

Category 8-D examinations VOLUMETRIC Reactor 5 5 1 0 2 0 2 0 All welds May SURFACE Recirculation 2p pg 6p ptt lpp 01, coincide with Discharge Category B-D examinations VOLUMETRIC Reactor 5 5 2 0 1 0 2 0 All welds May SURFACE Recirculation qp pg 6p 01, lpp ptt coincide with Suction Category B-D examinations VOLUMETRIC Reactor Vessel 17 17 6 0 5 0 6 0 All welds May SURFACE Nozzles coincide with 35 2$ 64 7g lpp 0~

Category B-D examinations L ~ I, J'L ~

33.a'~ 63.6t 100.0%

B'..10v HEAT N/A EXCHANGERS-NOZZLE-T<)-SAF E END BUTT WELDS, NPS or LARGER B5.110 HEAT N/A EXCHANGERS-NOZZLE-TO-SAF E END BUTT WELDS LESS

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UtJIT 1 PAGE: 8 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING DISSIMILAR METAL WELDS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM tt OF NO. SCHEDULED/COMPLETED ITEM li ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B5.110 THAN NPS 4 B5. 120 HEAT EXCHANGERS-ttOZZLE-TO-SAF E END SOCKET WELDS B5.130 PIPING-DISSIMILAR METAL BUTT WELDS, NPS 4 LARGER B5. 140 PI PING-DISSIMILAR METAL BUTT WELDS, LESS THAtt NPS 4 B5. 150 PI PING-DISSIMILAR METAL SOCKET WELDS B5.20 REACTOR SURFACE Control Rod 1 1 1 0 0 0 0 0 All welds May VESSEL-NOZZLE-TO-SAFE Drive lpp p% lpp p% lpp p% coincide with END BUTT WELDS, LESS Category B-D THAN NPS 4 examinations SURFACE Liquid Poison 1 1 1 0 0 0 0 0 All welds May lpp p% lpp p% lpp p% coincide with Category B-D examinations SURFACE Reactor 0 0 2 0 All welds May t ~ ~ ~

0 ~ 0% 100 p% coincide with Category B-D examinations ITEM TOTAL: 4 4 2 0 0 0 2 0 50.0% 50.0% 100.0%

B5.30 REACTOR N/A VESSEL-NOZZLE-TO-SAFE END SOCKET WELDS B5.40 PRESSURIZER-NOZZLE-TO-SA N/A FE END BUTT WELDS, NPS 4 or LARGER B5.50 PRESSURIZER-NOZZLE-TO-SA N/A FE END BUTT WELDS LESS THAN NPS 4 PRESSURIZER-NOZZLE-TO-SA FE END SOCKET WELDS

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 9 REVI S!OH: 00 IHSERVICE 1HSPECT)OH PLAH fOR THE THIRD INTERVAL CODE EDITION: E89 CLASS ! Sf.('T!OH i'I

SUMMARY

TABLE A PRESSURE RETAINING DISSIMILAR METAL WELDS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM I OF NO. SCHEDULED/COMPLETED ITEM 5 ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER CCMMENTS B5.70 STEAM N/A GENERATOR-NOZZLE-TO-SAFE END BUTT WELDS, NPS 4 or LARGER B5.80 STEAM N/A GENERATOR-NOZZLE-TO-SAFE END BUTT WELDS, LESS THAN NPS 4 B5. 9v STEAM N/A GENERATOR-NOZZLE-TO-SAi'E END SOCWET WELDS CATEGORY TOTAL: 3 l 37 13 0 10 0 14 0 35.1'1 62.1t 100.0S

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UHIT 1 PAGE: 10 REVISIOH: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN ASME NO. Of COMPONENTS SEC XI EXAM SYSTEM Of ttp. SCHEDULED/COMPLETED ITEM I ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B6.10 REACTOR VESSEL-CLOSURE VISUAL Closure Head 64 64 22 0 21 0 21 0 All nuts HEAD NUTS Nuts 34.3% 67.1% 100.0t B6.100 STEAM GENERATORS-FLANGE N/A SURFACE, WHEN CONNECTION DISASSEMBLED B6.110 STEAM GENERATORS-NUTS, BUSHINGS'ND WASHERS B6. 120 HEAT EXCHANGERS"BOLTS N/A AND STUDS B6. 130 HEAT EXCHAHGERS-FLANGE N/A SURFACE, WHEN COHNECTION DISASSEMBLED Ke 14>> HEAT EXCttANGERS-NUTS, BUSHINGS'ttD WASHERS PI PING-BOLTS AND ST'."'."

PIPING-FLANGE SURfACE, tt<A WHEN CONNECTIOH DISASSEMBLED

86. 170 PI PING-NUTS, BUSHINGS, AND WASHERS B6. 180 PUMPS-BOLTS AND STUDS VOLUMETRIC Reactor 0 0 1 0 0 0 All bolts 4 studs Recirculation p pg lpp ptt lpp ptt Limit to comp Pump 11 sched. by Cat.

B-L-2 VOLUMETRIC Reactor 1 0 0 0 0 0 0 0 All bolts 4 studs Recirculation O.OS 0.0% p pg Limit to comp Pump 12 sched. by Cat.

B-L-2 VOLUMETRIC Reactor 1 0 0 0 0 0 0 0 All bolts 4 studs

>>.Os v.p>> 0 pg Limit to comp p<>ml sched. by Cat.

B-L-2 VOLUMETRIC React.> 0' 0 0 0 0 0 All bolts 4 studs Recirculation O.OS 0.0% p pg Limit to comp Pump 14 sched. by Cat.

B-L-2

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 11 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUHHARY TABLE A PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM lf OF NO. SCHEDULED/COHPLETED ITEM rr ITEM DESCRIPTION HETHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VOLUMETRIC Reactor 0 0 0 0 0 0 0 All bolts a studs Recirculation 0.00 o.orr p pg Limit to comp Pump 15 sched. by Cat.

8-L-2 ITEM TOTAL: 5 1 0 0 1 0 0 0 0.0% 100.0% 100.0%

B6.190 PUMPS-FLANGE SURFACE, VISUAL Reactor 1 1 0 0 1 0 0 0 All flange surf.,

WHEN CONNECTION Recirculation 0 pq lpp pq lpp prr Limit to comp DISASSEMBLED Pump 11 sched. by Cat.

B-L-2 VISUAL Reactor 1 0 0 0 0 0 0 0 All flange surf.,

Recirculation 0. OS 0. 0% p pg Limit to comp Pump 12 sched. by Cat.

B-L-2 VISUAL Reactor 1 0 0 0 0 0 0 0 All flange surf.,

Recirculation 0.08 O.OS p pg Limit to comp Pump 13 sched. by Cat, B-L-2 VISUAL Reactor 1 0 0 0 0 0 0 0 All flange surf.,

Recirculation 0. 0'1 0. Orr p pg Limit to comp Pump 14 sched. by Cat.

B-L-2 VISUAL Reactor 1 0 0 0 ~

0 0 0 0 All flange surf.,

Recirculation 0. 0't 0. 0'rr p pg Limit to comp Pump 15 sched. by Cat.

B-L-2 ITEM TOTAL: 5 1 0 0 1 0 0 0 0.0% 100 '% 100.0%

B6.20 REACTOR VESSEI-CLOSURE VOLUMETRIC Closure Head 64 64 22 0 21 0 21 0 All studs STUDS, IN PLACE SURFACE Studs 34.3% 67.1% 100.0%

Ba <<PV PUMPS "NUTS, BUSHINGS, VISUAL Reactor. 0 0 3 0 0 0 All nuts, bush.

AND WASHERS Re~!rrculati.r py lpp prt lpp pg Mash Limit to comp Pump 11 sched. by Cat.

B-L-2

0 DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 12 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUMHARY TABLE A PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN ASME NO. OF CCHPONENTS SEC XI EXAM SYSTEH OF NO. SCHEDULED/COMPLETED ITEN I ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VISUAL Reactor 3 0 0 0 0 0 0 0 All nuts, bush. a Recirculation 0.0% 0.0% p 0% wash Limit. to comp Pump 12 sched. by Cat.

B-L-2 VISUAL Reactor 3 0 0 0 0 0 0 0 All nuts, bush. a Recirculation 0. 0% 0. 0% p p% wash Limit to comp Pump 13 sched. by Cat.

B-L-2 VISUAL Reactor u 0 0 0 0 0 All nuts, bush. S Recirculation 0. 0'%. 0'% p p% wash Limit to comp Pump 14 sched. by Cat.

B-L-2 VISUAL Reactor 3 0 0 0 0 0 0 0 All nuts, bush. a Recirculation 0.0% 0.0% p p% wash Limit to comp Pump 15 sched. by Cat.

B-L-2 ITEM TOTAL: 15 3 0 0 3 0 0 0 0.0% 100.0% 100.0%

B6.210 VALVES-BOLTS AND STUDS VOLUHETRIC Feedwater System 2 1 0 0 1 0 0 0 All bolts and p p% lpp p% lpp p% studs, Limited to components selected under Category B-M-2, VOLUHETR1C Re."~i i 1 1 v 0 0 0 0 All bolts and

>>>u 0% lpp p% studs, Limited to components selected under Category B-M-2, VOLUHETRIC Reactor Shutdown 1 1 0 0 0 0 1 0 All bolts and Cooling 0.0% p p% lpp p% studs, Limited to components selected under Category B-H-2, ITEM TOTAL: 5 3 1 0 1 0 1 0 33.3% 66.6% 100.0%

DATE: lp/26/99 NINE MZLE POINT NUCLEAR PLANT UNIT 1 PAGE: 13 REVISION: 00 ZNSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUHHARY TABLE A PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN ASHE NO. OF COHPONENTS SEC XI EXAM SYSTEM OF WO. SCHEDULED/COMPLETED ITEH ITEM DESCRIPTION HETHOD DESCRIPTION COMP RE@ 1ST PER 2ND PER 3RD PER COMMENTS B6.220 VALVES-FLANGE SURFACE, VISUAL Feedwater System 2 1 0 0 1 0 0 0 All flange NHEN CONNECTION 0.0% 100.08 100.08 DISASSEMBLED to valve selected under Category B-M-2 VZSUAL Reactor Core 2 1 1 0 0 ' 0 0 All flange Spray lpp pg lpp pq lpp pII surfaces, Limited to valve selected under Category B-H-2 VISUAL Reactor Shutdown 1 1 0 0 0 0 1 0 All flange Cooling 0.0% p 0II lpp 0II surfaces, Limited to valve selected under Category 8-H-2 ITCH TOTAL: 5 3 1 0 1 0 1 0 33.31 66.6% 100.0%

B6.230 VALVES-NUTS, BUSHINGS, VISUAL Feedwater System 1 0 0 1 0 0 0 All nuts, AND NASHERS p pq lpp 0II lpp bushings, washers, pg Limited to valve selected under B-M-2 VISUAL Reactor Core 6 3 3 0 0 0 0 0 All nuts, Spray Ipp. 0II lpp pI, lpp plI bushings, washers, Limited to valve selected under B-M-2 VISUAL Reactor Shutdown I 1 0 0 0 0 1 0 All nuts, Cooling 0.0% p pg lpp pg bushings, washers, Limited to valve selected under B-H-2 ITEM TOTAL: 9 5 3 0 1 0 1 0 60.0't 80.08 100.0%

REACTOR VESSEL-CLOSURE SURFACE Gl osure Head 1 3 1 0 1 0 1 0 12 studs, Deferral STUDS, WHEN REMOVED Studs y3 y$ 66 6'pp 01 Permissable,

1. 26, C. 2.b.,

RG a

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 14 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES ZN ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM I OF NO. SCHEDULED/COMPLETED ITEM i ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B6.30 minimum of 4 per period B6.40 REACTOR VESSEL-THREADS VOLUMETRIC Reactor Vessel 64 64 22 0 21 0 21 0 All threads in IN FLANGE Flange 34 3$ 67 1$ lpp pg flange, B6.50 REACTOR VESSEL-CLOSURE VISUAL Closure Head 128 128 44 0 42 0 42 0 All washers 4 WASHERS, BUSHINGS Washers 34 3tj 67 Itt Ipp pg Bushings VISUAL Reactor Vessel 64 64 22 0 21 0 21 0 All washers 4 Flange 34.3tt 67.1tt 100.0tt Bushings ITEM TOTAL: 192 192 66 0 63 0 63 0 34.3tt 67.18 100.0tt PRESSURIZER-BOLTS AND STUDS PRESSURIZER-FLANGE SURFACE, WHEN CONNECTION DISASSEMBLED PRESSURIZER-t<UTS, tlat A BUSHINGS, AND WASHER BB. 00 STEAM GENERATORS-BOLTS tilA AND STUDS CATEGORY TOTAL: 429 403 138 0 135 0 130 0 34.2% 67.7'0 100.0tt

DATE: 10/26/99 NINE MILE POlNT NUCLEAR PLANT UNIT 1 PAGE: 15 REVISIONr 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

Th BLE A 2 - PRESSURE RETAINING BOLTING, 2 INCHES AND LESS IN ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM II OF NO. SCHEDULED/COMPLETED ITEM ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS 87.10 REACTOR VESSEL-BOLTS, VISUAL Reactor Vessel 18 18 5 0 5 0 8 0 hll bolts, studs STUDS, AND NUTS Nozzles 27.7rt 55.5% 100.0%

87.20 PRESSURIZER-BOLTS'TUDS, AND NUTS 87.30 STEAM GENERATORS-BOLTS, STUDS, AND NUTS HEAT EXCHANGERS-BOLTS, N/h STUDS, AND NUTS sr 5r P! PING-BOLTS, STUDS, A"'" O'SUA'UTS Hen" 0 0 0 0 0 hll bolts, studs

> Pe.u't luu v'0 100.0'4 and nuts, Limited to 8-J VISUAL Reactor 5 5 2 0 2 0 1 0 All bolts, studs Recirculation 4p pg Sp.pg Ipp.pg and nuts, Limited Discharge to 8-J VISUAL Reactor 5 5 0 0 3 0 2 0 All bolts, studs Recirculation O.pt 60.00 100.0% and nuts, Limited Suction to 8-J ITEM TOTAL: 12 12 4 0 5 0 3 0 33.3rr 75.0t 100.08 87.6U PUMPS-BOLTS, STUDS, AND VISUAL Reactor 1 1 0 0 1 0 0 0 hll bolts, studs NUTS Recirculation p prr lpp prr lpp prr and nuts, Limited Pump 11 to 8-L-2 ViSUAL n~ cl%'r,"A, 0 n v 0 0 u 0 All bolts, stud's Rec fl r'oui sr. r '> 0 0~ p pq 0 pg and nuts, Limited Pu lap to 8-L-2 VISUAL Reactor 1 0 0 0 0 0 0 0 All bolts, studs Recirculation p pg p pg p pq and nuts, Limited Pump 13 to 8-L-2 VISUAL Reactor 1 0 0 0 0 0 0 0 All bolts, studs Recirculation pg p pg p pq and nuts, Limited Pump 14 to 8-L-2

DATE: 10/26/99 ttlNE NILE PC>! ttT tlUCLEAR PL%'<T UttIT 1 PAGE: 16 REVISION: 00 INSERVICE INSPECTtr>tt Pltut FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECT10tt XI

SUMMARY

TABLE A PRESSURE RETAINING BOI TING, 2 INCHES AND LESS IN ASME NO. OF COHPONENTS SEC XI EXAM SYSTEM tt OF'O. SCHEDULED/COMPLETED ITEM tt ITEH DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS Bt.60 Reactor All bolts, studs Recirculation and nuts, Limited Pump 15 to 8-L-2 ITEM TOTAL: 5 1 0 0 1 0 0 0 O.OF> 100.0% 100.0%

B7 70 VALVES BOLTS> STUDS> AND VISUAL Control Rod 3 3 1 0 0 0 2 0 All bolts, studs NUTS Drive 33 3$ 33 3>1 lpp pF> and nuts, Limited to B-M-2 VISUAL Emergency 6 3 2 0 1 0 0 0 All bolts, studs Condenser Return 66.6t 100.0% Ipp.pir and nuts, Limited to B-M-2 VISUAL E>r>v> ge>>.y v 0 0 0 0 0 0 All bolts, studs or>censer Supply p ptt 0 pi p pg and nuts, Limited to B-M-2 VISUAL Feedwater System 2 1 1 0 0 0 0 0 All bolts, studs 100. 0% 100. ptt Ipp. 0>t and nuts, Limited to B-H-2 VISUAL l.iquid Poison 2 2 1 0 1 0 0 0 All bolts, studs 5p ptt 1pp pt> 1pp ptt and nuts, Limited to B"H-2 VISUAL Hain Steam 17 4 0 0 4 0 0 0 All bolts, studs System p.prt lpp pg lpp pg and nuts, Limited to B-M-2 VISUAL Reactor Clean Up 4 1 0 0 1 0 0 0 All bolts, studs p.pg 100.0>t Ipp.pq and nuts, Lirrited to B-H-2 VISUAL kea" r r ~ ""> <: l. 5 "

0 I 0 2 0 All bolts, studs Spray 0.0't Gp.pt 100.0% and nuts, Limited to 8-H-2 VISUAL Reactor Drain 4 0 0 2 0 2 0 All bolts, studs p.ptt 5p ptt lpp ptt and nuts, Limited to 8-M-2

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 17 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING BOLTING, 2 ZNCHES AND LESS IN ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM ii ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VISUAL Reactor Head 3 3 3 0 0 0 0 0 All bolts, studs Vent lpp pi lpp pq lpp pq and nuts, Limited to B-M-2 VISUAL Reactor 10 5 1 0 2 0 2 0 All bolts, studs Recirculation 20 Oh 60 OS 100.0% and nuts, Limited Discharge to B-M-2 VISUAL Reactor 5 1 0 0 1 0 0 0 All bolts, studs Recirculation p pq lpp pg lpp pg and nuts, Limited Suction to B-M-2 VISUAL Reactor Shutdown 3 1 0 0 0 0 1 0 All bolts, studs Cooling p pg p pt, lpp pg and nuts, Limited to B-M-2 ITEM TOTAL: 75 33 11 0 13 0 9 0 33.3% 72.7% 100.0S B7 8(>> CRD HOUSINGS BOLTS'TUDS, VISUAL Control Rod 136 4 1 0 2 0 1 0 All bolts, studs AND NUTS Drive Housings 25 0$ 75 pg lpp pii and nuts, when CRD disassembled

'ATEUt>R': T( .'AL: .'46 6t'1 0 6 0 21 0 3v. 8'a 69. 1'4 100. 0%

DATE: 10/26/99 NINE HILE PO1NT ttUCLEAR PLANT UNIT 1 PAGE: 18 REVISION: 00 INSERVICE IttSPECTIOtt PLAN FOR THE THIRD IttTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUHMARY TABLE A PRESSURE RETAINING WELDS IN PIPING ASHE NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM ITEM DESCRIPTION METHOD DESCRIPTION CCHP REQ 1ST PER 2ND PER 3RD PER COMMENTS B9.11 CIRCUMFERENTIAL PIPE VOLUMETRIC Emergency 32 0 0 0 0 0 0 0 At least 25tt of WELDS, NPS 4 or LARGER Condenser Return O.pit 0.0% O.ptt the welds VOLUHETRIC Emergency 16 5 5 0 0 0 0 0 At least 25tt of SURFACE,~,ndensor Supply Ipp.pi Ipp.pi 'pp.pg the welds VOLUMETRIC Feedwater System 50 12 0 0 1 0 11 0 At least 25% of SURFACE the welds p pg 8 3tt lpp ptt VOLUMETRIC Main Steam 34 11 2 0 1 0 8 0 At least 25'I of SURFACE System 18.1% 27.21 100.00 VOLUMETRIC Reactor Clean Up 30 11 0 0 3 0 8 0 At least 25% of SURFACE the welds O.ptt 27.24 100.05 VOLUMETRIC Reactor Core 70 6 3 0 2 0 1 0 At least 25% of SURFACE Spray the welds 50.0% 83.3$ 100.pent VOLUHETRIC Reactor 43 17 5 0 4 0 8 0 At least 25% of Recirculation 29.4'2.9tt 100.0lt the welds Discharge VOLUMETRIC Reactor 54 19 4 0 8 0 7 0 At least 258 of SURFACE t<ecl t rule l n 1 pp ptt the welds tuf I ~ <<g VOLUt4ETRIC Reactor Shutdown 16 3 1 0 1 0 1 0 At least 254 of Cooling 33. 3'6. 6% 100. 0% the welds ITCH TOTALt 345 84 20 0 20 0 44 0 23.81 47.6% 100.0l B9. 12 LONGITUDINAL PIPE WELDS, VOLUHETRIC Emergency 47 0 0 0 0 0 0 0 One pipe diameter, NPS 4 or LARGER Condenser Return 0.0% O.OS p ptt or 12 inches of each weld selected under Categories B-F and B-J, Code Case N-524 VOLUMETRIC Reactor Core 55 0 0 0 0 0 0 0 One pipe diameter SURFACE Spr ay or 12 inches of O.ptt p ptt p ptt each weld selected

DATE: 10/26/99 ttlttF. Hl I.E I'Ot ttT t!U .LEAF PLAttT Uttli I PAGE: 19 REVI 8 IvH: 00 ! ttSERVICF, IttSI'E -,TIARA! I".hl':-"i)k '!'HF THIRD It'T RVAL CODE EDITION: E89 CI>>ASS I SE- lr)ll Xl SUMHARY TABLE A PRESSURE RETAINING WELDS IN PIPING ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM tt OF NO. SCHEDULED/COMPLETED ITEH tt ITEM DESCRIPTION HETHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS B9.12 under Categories B-F and B-J, Code Case N-524 VOLUMETRIC Reactor 78 18 5 0 5 0 8 0 One pipe diameter Recirculation 27.7% 55.5% 100.0% or 12 inches of Discharge each weld selected under Categories B-F and B-J, Code Case N-524 VOLUHETRIC Reactor 83 26 4 0 13 0 9 0 One pipe diameter SURFACE Recirculation 15 3% 65 3% lpp p% or 12 inches of Suction each weld selected under Categories B-F and B-J, Code Case N-524 VOLUHETR!C Reaorut Shutdewn 5 " 0 0 " 0 0 0 One pipe diameter Cool I ng 0.0% 100.0% 100.0% or 12 inches of each weld selected under Categories B-F and B-J, Code Case N-524 ITEH TOTAL: 288 46 9 0 20 0 17 0 19.5% 63.0% 100.0%

B9. 21 CIRCUMFERENTIAL PIPE SURFACE Control Rod 20 8 7 0 0 0 1 0 At least 25% of

!4ELDS>> LESS THAN NPS 4 Drive 7.5% 87,5% lpQ.Q% the welds SURFACE Liquid Poison 20 5 4 0 0 0 1 0 At least 25% of 80.0% 80.0% 100.0%

SURFACE Hain Steam 14 4 2 0 1 0 1 0 At least 25% of System 5Q 0% 75. p% I pp p% the welds

.::!HFA 'E t\ ~ '>>>> '. I ~ s 3 0 5 0 At least 25% of i Q. 0% IOP. O%

sURFAcE Reactor Drain 12 1 0 0 1 0 0 0 At least 25% of p p% lpp p% lpp p% the welds

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 20 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING NELDS IN PIPING ASME NO. OF'OMPONENTS SEC XI EXAM SYSTEM ll OF NO. SCHEDULED/COMPLETED ITEM ll ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS SURFACE Reactor Head 11 4 4 0 0 0 0 0 At least 25% of Vent 100.0% 100.0% 100.0%

SURFACE Reactor 5 " 0 0 0 0 2 0 At least 25% cf Instrumentation p.p% p p% lpp.p% the welds SURFACE Reactor 1 0 0 0 0 0 0 0 At least 25% of Instrumentstion p 0% p p% p p% the welds SURFACE Reactor 30 5 0 0 0 0 5 0 At least 25% of Recirculation 0.0% 0.0% 100.0% the welds Discharge SURFACE Reactor 2 1 0 0 0 0 1 0 ht least 25% of Recirculation p p% p p% lpp p% the welds Suction

!TEM TOTAL: 164 l4 23 0 5 0 16 0 52.2% 63.6% 100.0%

B9. '" LONGITUDINAL PIPE WELDS, N/A LESS THAN NPS 4 B9. 3) BRANCH CONNECTION WELDS, VOLUMETRIC Main Steam 6 2 2 0 0 0 0 0 At least 25% of NPS 4 or LARGER SURF'ACE System 100 0% 100 0% 100 0% the welds VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 ht least 25% of SURF'ACE Spray the welds 0 0% 0 0% 0 0%

VOLUMETRIC Reactor 1 0 0 1 0 0 0 At least 25% of SURFACE Recirculation p p% lpp p% lpp p% the welds Suction ITEM TOTAL: 9 3 2 0 1 0 0 0 66.6% 100.0% 100.0%

B4.4 'RANCH CONNECTION WELDS, SURFACE Main SL,eaa( 1 0 0 1 0 0 0 At least, 25%

LESS THAN NPS 4 Sysl ~ 'ml 0.0% 100.0% 100.0%

SURFACE React.ov CLean Up 1 1 1 0 v 0 0 0 ht least 25%

100.0% 100.0% 100.0%

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 21 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUYAARY TABLE A PRESSURE RETAINING WELDS IN PIPING ASHE NO. OF'CHPONENTS SEC XI EXAH SYSTEH Of'O. SCHEDULED/CCHPLETED ITEH F ITEH DESCRIPTION HETHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHHENTS SURFACE Reactor Core 4 1 0 0 1 0 0 0 At least 25%

Spray 0.0% 100.0% 100.0%

SURF'ACE Reactor 10 2 1 0 1 0 0 0 At least 25%

Recirculation 50.0% 100.0% 100.0%

Dischar9e

UKFACF. ke~ct 0 0 ! 0 1 0 At least 25%

keel rcuiat ioit 0. 0% 50. 0% 100.

Suction 0'%TEH TOTAL: 26 7 2 0 4 0 1 0 28.5% 85.7% 100.0%

B9.40 SOCKET NELDS SURFACE Hain Steam 9 1 0 0 1 0 0 0 At least 25% of System p p% lpp p% lpp p% the welds SURFACE Reactor Drain 10 5 0 0 2 0 3 0 At least 25% of p p% 4p 0% lpp p% the welds SURFACE Reactor Head 14 4 4 0 0 0 0 0 At least 25% of Vent lpp p% lpp p% lpp p% the welds SURFACE Reactor 6 1 0 0 0 0 1 0 At least 25% of Instrumentation p p% lpp p% the welds

.Nt lk FA '.r. kt' s ~ t t~ 0' u 5 0 At least 25% of "t I y ~ ~ we 1 ds tDi sabha r:t-ITEH TOTAL: 69 20 4 0 7 0 9 0 20.0% 55.0% 100.0%

CATEGORY TOTAL: 901 204 60 0 57 0 87 0 29.4% 57.3% 100.0%

0 DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 22 REVISION: 00 IHSERVICE INSPECTIOH PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A INTEGRAL ATTACHMENTS FOR PZPZNG, PUMPS, AND VALVES ASME NO. OF COMPONENTS SEC Xl EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM N ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2HD PER 3RD PER COMMEHTS Blp. 20 PIPING-ZNTEGRALLY WELDED SURFACE Control Rod 1 0 0 0 0 0 0 0 10% of all Welded ATTACHMENTS Drive 0.0% 0.0% p pg attachments of piping required under Category B-J, Code Case N-509 SURFACE Emergency 10 2 2 0 0 0 0 0 10% of all Welded Condenser Return lpp pq lpp pg lpp pg attachments of piping required under Category B-J, Code Case N-509 SURFACE Feedwater System 18 4 0 0 0 0 4 0 10'1 of all Welded

0. 0% p pi lpp pt attachments of piping required under Category B-J, Code Case H-509 SURFACE Main Steam 16 4 0 0 4 0 0 0 10% of all Welded System p pg lpp pq lpp pg attachments of piping required under Category B-J, Code Case N-509 SURFACE Reactor Clean Up 23 2 0 0 2 0 0 0 108 of all Welded 0 pg lpp pg lpp pq attachments of piping required under Category B-J, Code Case H-509 SURFACE Reactor Core 34 4 0 0 0 0 4 0 10% of all Welded Spray 0. 0'8 p pg lpp pg attachments of piping required under Category B-J, Code Case N"509 SURFACE Reactor Drain 0 0 0 0 0 0 0 10% of all Welded p pg p pg p pg attachments of

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 23 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION Xl

SUMMARY

TABLE A INTEGRAL ATTACHMENTS FOR PIPING, PUMPS, AND VALVES ASME NO. OF COMPONENTS SiU '::i EXAM S I STEM OF NO. SCHEDULED/COMPLETED ITEM 0 ITEM DESCRI PTION METHOD:lES: Ri PTION C(44i REQ 1ST PER ND PER 3RD PER COMMENTS B10.20 piping required under Category B-J, Code Case N-509 SURFACE Reactor 24 4 4 0 0 0 0 0 10% of all Welded Recirculation 100.0% 100.0% 100.0% attachments of Discharge piping required under Category B-J, Code Case N-509 SURFACE Reactor 20 4 0 0 0 0 4 0 10% of all Welded Recirculation 0.0% p 0% lpp p% attachments of Suction piping required under Category B-J, Code Case N-509 SURFACE fee~i t.'. hu orwn 5 0 0 0 0 0 0 0 10% of all Welded Coi~ ~ ~ lq p p% attachments of piping required under Category B-J, Code Case N"509 ITEM TOTAL: 155 24 6 0 6 0 12 0 25.0% 50.0% 100.0%

B10.30 PUMPS-INTEGRALLY WELDED N/A ATTACHMENTS Bl 0. 40 VALVES-INTEGRALLY WELDED SURFACE Control Rod 0 0 1 0 0 0 10% of all Welded ATTACHMENTS Drive 0.0% 100.0% 100.0% attachments of valves required under Category B-J, Code Case N-509 SURFACE Liquid Poison 1 0 0 0 0 0 0 0 10% of all Welded p p% p p% p p% attachments of valves requi ed under Category B-J, Code Case N-509

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 24 REVISION: 00 INSERVICE INSPECTION PLAN fOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A INTEGRAL ATTACHMENTS FOR PIPING, PUMPS, AND VALVES ASHE NO. OF COMPONENTS SEC XI EXAM SYSTEM II OF NO. SCHEDULED/COMPLETED ITEM ll ITEM DESCRIPTION HETHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COMMENTS Blp. 40 10% of all Welded attachments of valves required under Category B-J, Code Case N-509 SURfACE Reactor Core 2 0 0 0 0 0 0 0 10% of all Welded Spray 0.0% 0.0'll p pg attachments of valves required under Category B-J, Code Case N-509 SURFACE Reactor Shutdown 2 1 0 0 0 0 1 0 lpll of all Welded Cooling 0.0% p pg lpp pg attachments of valves required under Category B-J, Code Case N-509

'ITEH TOTAL: 8 3 0 0 1 0 2 0 0.0% 33.3ll 100.0'1 Blp. 10 PRESSURE SURFACE Reactor Vessel 6 3 1 0 1 0 1 0 100% of the length VESSEL-INTEGRALLY WELDED Supp'ts 33 3% 66 6% 100 0% of the welds to ATTACHHENTS the vessel. Code Case N-509 CATEGORY TOTAL: 169 30 7 0 8 0 15 0 23.3% 50.08 100.0%

DATE: 1v/ "6/99 tt1NF. MILL Pvttt't'UCLEAR PLANT UttIT 1 PAGE: 25 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTIOlt XI

SUMMARY

TABLE A PRESSURE RETAINING NELDS IN PUMP CASING ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM 8 OF NO. SCHEDULED/COMPLETED ITEM I ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B12. 10 PUMPS-PUMP CASING WELDS N/A CATEGORY TOTAL: 0 0 0 0 0 0 0 0 0.0% 0.Otal O.OS

DATE: 10/26/99 NINE NILE POINT NUCLEAR PLANT UNIT 1 PAGE: 26 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A 2 - PUMP CASINOS ASME NO. OF COMPONENTS SEC XI EXAM >!.'YSTEM I OF'O. SCHEDULED/COMPLETED

'. TFM tt ITEM DESCRI PTIOtt NETHr DESCR!PTIOtt Crgt['EQ 1ST PER 2ND PER 3RD PER COMMENTS Bl . 20 PUMPS-PUMP CASINGS VISUAL tK4 t 1 0. 0 1 0 0 0 One pump, when

>'cula Cion h.'leet p.ptt Ipp.ptt lpp ptt disassembled for Pur<

t p ptt disassembled for Pump 13 ma int. VISUAL Reactor. 1 0 0 0 0 0 0 0 One pump, when Recirculation p.ptt p pg p.pt> disassembled for Pump 14 ma int. VISUAL Reactor 1 0 0 0 0 0 0 0 One pump, when Recirculation p pg p ptt p ptt disassembled for Pump 15 maint. IFt t'r>T/>I . 5 ) 0 0 1 0 0 0

0. u't 100. 0>> 100. 0%

CATEGORY TOTAL: 5 1 ' 0 1 0 0 0 0.0% 0.0% 100.0% DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 27 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAINING WELDS ZN VALVE BODIES ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM II OF NO. SCHEDULED/COMPLETED ITEM II ITEM DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS B12.30 VALVES-VALVE BODY WELDS, N/A tootle LESS THAN NPS 4 BI2.40 VALVES-VALVE BODY WFI,DS, VOLUMFTRI< Hain Sr~am 6 1 0 0 1 0 0 0 Limited NPS 4 or LARGER Sr/s r. errr 0.0'1 100.0'00.0II of valves CATEGORY TOTAL: 6 1 0 0 1 0 0 0 O.OII 0.0% 100.0II

DATE: )0/ 6/99 PAGE: 28 REVISION: 00 iffsERvfcE:Nsf'F 'T:tsf:"jvf =.>R .!lF '."!if!if" flfTERvAI. CODE EDITION: E89 CLASS 1 SECTION Xf

SUMMARY

TABLE A VALVE BODIES ASME NO. OF COHPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM ll ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B12. 50 VALVES-VALVE BODIES VISUAL Emergency 6 3 2 0 1 0 0 0 One valve each EXCEEDING 4 INCHES Condenser Return 66.6tf 100.0% 100.0% group of valves if NOMINAL PIPE SIZE diss. for mnt.

reasons Defer Permis VISUAL Emergency 4 1 0 0 0 0 1 0 One valve each Condenser Supply 0.0% 0.0% 100.0% group of valves !.f diss. for mnt.

reasons Defer Permis V!SUAl. I n 1 0 0 0 One valve each 5u.pa lpp.ui 100.0% group of valves if diss. for mnt.

reasons Defer Permis VISUAL Hain Steam 16 3 0 0 3 0 0 0 One valve each System 0.0% 100.0!f 100.0% group of valves if diss. for mnt.

reasons Defer Permis VISUAL Reactor Clean Up 4 2 0 0 2 0 0 0 One valve each p ptl lpp pg lpp ptl group of valves if diss. for mnt.

reasons Defer Permis VISUAL Reactor Core 10 2 2 0 0 0 0 0 One valve each Spray lpp pi lpp ptf lpp pg group of valves if diss. for mnt.

reasons Defer Permis VISUAL Reactor 5 0 0 0 0 0 0 0 One valve each Recirculation 0. Off 0. 0% p ptl group of va)ves if Discharge diss. for mnt.

reasons Defer Permis VISUAL Reactor 5 1 0 0 1 0 0 0 One valve each Recirculation p pg lpp pi lpp ptf group of valves if

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT Ut!IT 1 PAGE: 29 R'=V:.t! UN: pt IIISERVICE INSPECTIOII PLAN FOR THE THIRD INTERVAL CODE EDITIOtl: E89 i!ASS ! SFcTIr>!I !I

SUMMARY

TABLE A VALVE BODIES ASME NO. OF COHPONENTS SEC XI EXAM SYSTEH OF'O. SCHEDULED/COMPLETED ITEM ll ITEM DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COMMENTS B12.50 Suction diss. for mnt.

reasons Defer Permis VISUAL Reactor Shutdown 4 2 0 0 0 0 2 0 One valve each Cooling 0.0% p 01 lpp 01 group of valves if diss. for mnt.

reasons Defer Permis VISUAL Reactor Vessel 8 1 1 0 0 0 0 0 One valve each Nor@les Ipp.p!t Ipp.pl! Ipp.p!t group of. valves if diss. for mnt.

reasons Defer Permis

TEM TiiTAI.: 66 I> 6 0 8 0 3 0 35.25 8 .3'L 100.0'1 i'ATEGORY TUTAL: 66 17 6 0 8 0 3 0 35.2% 82.3!t 100.0S

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 30 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI SUMHARY TABLE A INTERIOR OP REACTOR VESSEL ASME NO. OF COHPONENTS SEC XI EXAM SYSTEH II OF NO. SCHEDULED/COMPLETED ITEM ITEM DESCR1PTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B13 10 F REACTOR VESSEL-VESSEL VISUAL Reactor Vessel 3 6 2 0 2 0 2 0 First refuel INTERIOR Interior 33.3% lpp pg outage then once each inspection period

'A:: 4.>AY TOTAL: 6 0 0 2 0 33.3% 0.0'1 100.0'h

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 31 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A 2 INTEGRALLY WELDED CORE SUPPORT STRUCTURE ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM 8 ITEM DESCRIPTION METHOD DESCRIPTION COMP RE{} 1ST PER 2ND PER 3RD PER COMMENTS B13.40 REACTOR VESSEL VOLUMETRIC Reactor Vessel 49 61 35 0 12 0 14 0 Accessible (BWR)-CORE SUPPORT Interior pg lpp pg surfaces STRUCTURE B13.20 REACTOR VESSEL VISUAL Reactor Vessel 5 5 1 0 2 0 2 0 All welds Deferral (BWR)-INTERIOR Interior 20.0% 60.0% 100.0% Permissable ATTACHMENTS WITHIN BELTl,INE REGION B)3.30 REACTOR VESSEL (BWR) -INTERIOR ATTACHMENTS BEYOND BELTLINE REGION B13.50 REACTOR VESSEL N/A (PWR)-INTERIOR ATTACHMENTS WITHIN BELTLINE REGION REACTOR VESSEL N/A (PWR)-INTERIOR ATTACHMENTS BEYOND BELTLINE REGION CATEGORY TOTAL: 54 66 36 0 14 0 16 0 54.5'1 75.7S 100.0%

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 32 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECT I Off:;".

SUMMARY

TABLE A 3 - REHOVABLE CORE SUPPORT STRUCTURES ASHE NO. OF COMPOHENTS SEC Y.I EYAH SYSTEM OF'O. SCHEDULED/COMPLETED ITEM 8 ITEH DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS BI3.70 REACTOR VESSEL (PWR)-CORE SUPPORT STRUCTURE CATEGORY TOTAL: 0 0 0 0 0 0 0 0 O.OS 0.0% 0.0t

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 33 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A

- PRESSURE RETAINING WZLDS IN CONTROL ROD HOUSINGS ASME NO. OF CCMPONENTS SEC Xl EXAM SYSTEM tl OF NO. SCHEDULED/COMPLETED ITEM li ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS B14. 10 REACTOR VESSEL-WELDS IN SURFACE Control Rod 8 8 2 0 3 0 3 0 108 peripheral CRD CONTROL ROD DRIVE Drive Housings 25 pg 62 5g lpp pg Housings, Deferral HOUSINGS Permissable CATEGORY TOTAL: 8 8 2 0 3 0 3 0 25.08 0.0% 100.0%

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 EI Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev.

H U MOHAWK 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX B - CLASS 2

SUMMARY

TABLES Table of Contents B-1 Record of Revision B-2 ASME Code Class 2 Section XI Summary Tables B-3 Examination Category C-A 34- 35 Examination Category C-B 36-37 Examination Category C-C 38- 39 Examination Category C-D 40- 40 Examination Category C-F-1 .. 41-41 Examination Category C-F-2 42- 45 Examination Category C-G . 46- 46 File: APPENDIXB.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION'""

REVISION; ,;:.",'..';."',GATE";;." .' AFFECTED >@REASON FOR REVISION.

PAGES 0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3"'en Year Inservice Inspection Interval File: APPENDIXB.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unlt1 El Y NIAGARA H 4 MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ASME CODE CLASS 2

SUMMARY

TABLES File: APPENDIXB.WPD

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 34 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION XI

SUMMARY

TABLE A PRESSURE RETAZNZNG WELDS ZN PRESSURE VESSELS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM I OF NO. SCHEDULED/COMPLETED ITEM 5 ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS Cl.lp SHELL CIRCUMFERENTIAL WELDS C1.20 HEAD CIRCUMFERENTIAL N/A WELDS C1.30 TUBESHEET-TO-SHELL WELDS VOLUMETRIC Emergency 2 0 0 0 0 0 0 0 Limited to one Condenser Heat p pg p pg p pg vessel in a group Exchanger 111 of vessels u 0 0 0 Limited to one p pg vessel in a group Ex.'n~r>g>>: ! 1. of vessels VOLUMETRIC Emergency 2 0 0 0 0 0 0 0 Limited to one Condenser Heat p pg p pg p pg vessel in a group Exchanger 121 of vessels VOLUMETRIC Emergency 2 2 1 0 1 0 0 0 Limited to one Condenser Heat 5p pg lpp pg lpp pg vessel i.n a group Exchanger 122 of vessels VOLUMETRIC Reactor 2 0 0 0 0 0 0 0 Limited to one Containment p pg p pg p pg vessel in a group Spray Heat of vessels Exchanger ill VOLUMBTRIC Reactor 2 0 0 0 0 0 0 0 Limited to one asr> air>>>>er>t u 0~ p p~ 0 pg vessel in a group Sf, '>'>" w(

~ of vessels VOLUMETRIC Reactor v 0 0 0 0 0 0 Limited to one Containment p pt, p pg p pg vessel in a group Spray Heat of vessels Exchanger 121 VOLUMETRIC Reactor 2 2 0 0 0 0 2 0 Limited to one Containment p p>1 p pa lpp p>1 vessel in a group Spray Heat of vessels Exchanger 122 ITEM TOTAL: 16 4 1 0 1 0 2 0 25.00 50.0% 100.0%

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UN1T 1 PAGE: 35 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI SUMHARY TABLE B PRESSURE RETAINING WELDS IN PRESSURE VESSELS ASME NO. OF COHPONENTS SEC XI EXAM SYSTEM I OF NO. SCHEDULED/COMPLETED ITEN fi ITEM DESCRIPTION HETHOD DESCRI PTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS CATEGORY TOTAL: 16 4 1 0 1 0 2 0 25.0% O.OS 100.08

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 36 REVISIONr 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE B PRESSURE RETAINING NOZZLE %tELDS IN VESSELS ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM tt OF NO. SCHEDULED/COMPLETED ITEM s ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS NOZZLE-TO-SHELL (OR <

tR FACE Rr>~ 1" ~

0 0 0 0 0 0 0 All nozzles at HEAD) WELD <~ 1/2 Lli. Con%at lrmvlrk 0.0% 0.0% p 0% Terminal Ends of NOMINAL THICKNESS Spray Heat piping runs Ex twanger 111 selected under C-F SURFACE Reactor 2 0 0 0 0 0 0 0 All nozzles at Containment 0.0% 0.0% p p% Terminal Ends of Spray Heat piping runs Exchanger 112 selected under C-F SURFACE Reactor 2 0 0 0 0 0 0 0 A)1 nozzles at Containment 0.0% 0.0% p p% Terminal Ends of Spray Heat piping runs Exchanger 121 selected under C-F SURFACE Reactor 2 2 0 0 0 0 2 0 All nozzles at Containment 0.0% p p% lpp 0% Terminal Ends of Spray Heat piping runs Exchanger 122 selected under C-F

.'"Et) TrrTAL: 0 0 0 0 2 0 u.0% v.0% 100.0%

I

~ r ttOZZLE-TO-SHELL tOR VOl UMETRI< t'ier c 0 0 0 0 All nozzles at ftEAD) WELD > 1/2 Itt. Coridens r Heat 0.0% p p% p p% terminal ends of NOMINAL THICKNESS Exchanger 111 piping runs WITHOUT REINFORCING selected under PLATE Category C-F, Limited to one vessel VOLUMETRIC Emergency 2 0 0 0 0 0 0 0 All nozzles at SURFACE Condenser Heat 0.0% 0.0% 0.0% terminal ends of Exchanger 112 piping runs selected under Category C-F, Limited to one vessel VOLUMETRIC Emergency 2 0 0 0 0 0 0 0 All nozzles at Condenser Heat P. 0% 0. 0% p p% terminal ends of u(a><u( "

piping runs selected under Category C-F, Limited to one

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLAttT Ut)IT 1 PAGE: 37 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE B ASME SEC XI C2.21

'XAM PRESSURE RETAINING NOZZLE NELDS ITEM ff. ITEM DESCRIPTIOW IN VESSELS METHOD SYSTEM DESCRIPTION VOLUMETRIC Emergency i OF COMP 2

NO.

REQ 2

NO. OF COMPONENTS SCHEDULED/COMPLETED 1ST PER 2ND PER 3RD PER 1 0' 0 0 0 COMMENTS vessel All nozzles at SURFACE Condenser Heat Sp pg lpp pg Ipp.pg terminal ends of Exchanger 122 piping runs selected under Category C-F, Limited to one vessel ITEM TOTAL: 8 2 1 0 1 0 0 0 Su Ob

~ 100 8 100 'f) ttOZZLE IttSIDE RADIUS W/A SECTION > 1/2 IW.

NOMINAL THICKNESS WITHOUT REINFORCING PLATE REINFORCING PLATE WELDS N/A TO NOZZLE AND VESSEL )

1/2 IN. NOMIHAL THICKNESS C2.32 NOZZLE-TO-SHELL fOR HEAD) WELDS WHEN INSIDE OF VESSEL IS ACCESSIBLE

'i 1/2 IN. NOMINAL THICKNESS C" 33

~ NOZZLE-TO-SHELL (OR W/*

HEAD) WELDS ttHEN INSIDE OF HELD IS INACCESSIBLE 1/2 IN. NOM1NAL THICKNESS CA EGURY TOTAL: 16 4 1 0 1 0 2 0 25.0t 0.01 100.0ft

DATE: 10/26/99 ll>Ne M~ >>. <<..;>>1> in> ~

> I.'AR t>l~tt~ PAGE: 38 R. VIS!OH: 00 !NSERVICE !HSI EC'I'lr>N il~N FOR TNE T!!IRD !tlTERVAL CODE EDITION: E89 C!rAcc .'E T!Otl XI

SUMMARY

TABLE 8 INTEGRAL ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND ASME HO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM tl ITEM DESCRIPTION METHOD DESCRIPTION CCHP REQ 1ST PER 2ND PER 3RD PER COMMENTS C3. 10 PRESSURE SURFACE Reactor 2 0 0 0 0 0 0 0 10% of All VESSELS-INTEGRALLY Containment p pit p pit p pit attachments, Code WELDED ATTACHMENTS Spray Heat Case N-509 Exchanger 111 SURFACE Reactor 2 0 0 0 0 0 0 0 10% of All Containment 0.0% O.OI> O.OI> attachments, Code Spray Heat Case H-509 Exchanger 112 SURFACE React, or 0 0 0 0 0 0 0 10% of All C >;>< a!r>me<<r 0 0~ p pq p pit attachments, Code 3 p I >y I!i-'v ( Case H-509 SURFACE Re~ct,~> 1 0 0 0 0 1 0 10% of All Containment p pq p OI> lpp 0!t attachments, Code Spray Heat Case H-509 Exchanger 122 ITEM TOTAL: 8 1 0 0 0 0 1 0 O.pit 0.0% 100.0%

C3.20 PIPING-INTEGRALLY WELDED SURFACE Control Rod 24 0 0 0 0 0 0 0 10!t of All ATTACHMENTS Drive 0.0% O.pent p pg attachments, Limited to those Category C-F and C-G, Code Case H-509 SURFACE Emergency 61 8 0 0 0 0 8 0 10% of All Condenser Return 0. OS p pg lpp pg attachments>

Limited to those Category C-F and C-G, Code Case N-509 SURFACE Emergency 18 4 4 0 0 0 0 0 10% of All Condenser Supply 100.0% 100.0% Ipp.pit attachments, Limited to those Category C-F and C-G, Code Case N-509

DATE: 10/26/99 NINE HILE POINT ttUCLEAR PLANT UNIT 1 PAGE: 39 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE ED1TION: E89 CLASS 2 SECTIOtt XI SUHMARY TABLE B INTEGRAL ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM tt OF NO. SCHEDULED/COMPLETED ITEN tj ITEM DESCRI PTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS SURFACE Main Steam 15 4 4 0 0 0 0 0 10% of All lpp pt lpp pi lpp ptt attachments, Limited to those Category C-F and C-G, Code Case N-509 SURFACE Hain Steam 56 8 0 0 8 0 0 0 lptt of All System 0 0% 100 Ott 100 Ott attachments, Limited to those Category C-F and C-G, Code Case N-509 SURFACE Reactor 526 54 9 0 25 0 20 0 10% of All Containment 62 9t lpp pg attachments, Spray Limited to those Category C-F and C-G, Code Case N"509 SURFACE Reactor Core 331 26 13 0 2 0 11 0 105 of All Spray 5p.nit 57.6$ lpp ptt attachments, Limited to those Category C-F and C-G, Code Case N-509 ITEM TOTAL: 1031 104 30 0 35 0 39 0 28.8t 62.5% 100.0tt C3.30 PUMPS-INTEGRALLY WELDED N/A ATTACHMENTS C3.40 VALVES-INTEGRALLY WELDED ATTACHMENTS CATEGORY TOTAL'039 105 30 0 35 0 40 0 28.5'4 61.9't 100.0t

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 40 REVI S I ON: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI SUHMARY TABLE B PRESSURE RETAINING BOLTING GREATER THAN 2 INCHES IN ASME NO. OF CCHPONENTS SEC XI EXAM SYSTEM lt OF NO. SCHEDULED/COMPLETED ITEM ir ITEH DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C4. 10 PRESSURE VESSELS-BOLTS N/A AND STUDS C4.20 PIPING-BOLTS AND STUDS C4.au PUMPS-BOLTS AND STUDS Nxh C4 ~ 40 VALVES-BOLTS AND STUDS N/n CATEGORY TOTAL: 0 0 0 0 0 0 0 0 0.0% O.OS O.OS

DATE: 10/26/99 tllHE MI LE POItlT tlVCLEAR PIAIIT UNIT 1 PAGE: 41 RFV.'SIOHt 00 ltlSERVICE ltlSlECTI:.>I: PI.AII 8'>R THE Ttl)RD ItlTERVAL CODE EDITION: E89 TAB' B PRESSURE RETAINING WELDS ZN AUSTENITIC STAINLESS STEEL ASME NO. Of'OMPONENTS SEC KI EXAM SYSTEM tt OF HO. SCHEDULED/COMPLETED ITEM tl ITEM DESCRIPTIOH METHOD DESCRIPTIOH COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C5.12 LONGITUDINAL WELDS > 3/8 VOLUMETRIC Emergency 54 15 7 0 3 0 5 0 2.5t at IN. HOMINAL WALL Condenser Return lpp 0% intersecting Circ.

THICKNESS FOR PIPING > Welds selected for NPS 4 examination, Code Case H-524 VOLUMETRIC Emergency 67 22 4 0 11 0 7 0 2.5t at SURFACE Condenser Supply intersecting Circ.

lpp p%

Welds selected for examination, Code Case N-524 ITEM TOTAL: 121 37 11 0 14 0 12 0 29.7% 67.5% 100.0%

': KCUMFEKEtlTIAL VtEL-1/b 1N. NOM1HAL WALL THICKNESS FOR PIPING NPS 2 AND < NPS 4 LONGITUDINAL WELDS > 1/5 IH. NOMINAL WALL THICKNESS FOR PIPZHG >

NPS 2 AND < NPS 4 C5.30 SOCKET WELDS H/A C5.41 CIRCUMFERENTIAL PIPE N/A BRANCH CONNECTIONS OF BRANCH PIPING > NPS 2 C5.42 LONGITUDINAL PIPE BRANCH N/A CONNECTIONS OF BRANCH PIPING > NPS 2 CIRCUMFERENTIAL PIPE VOLUMETRIC Emergency 31 8 4 0 2 0 2 0 7.5% but not less WELDS > 3/8 IN. NOMINAL SURFACE Condenser Return than 28 welds 5p p% 75 p% lpp p%

WALL THICKNESS FOR PIPING >NPS 4 VOLUMETRIC Fm. r ge>>~y tu 4 0 7 0 6 0 /.5% but not less SUKf'ACF. ('~ >denqi. >uppt y

~

5% 64 7% lpp p% than 28 welds ITEM TOTAL: 72 25 8 0 9 0 8 0 32.0% 68.0% 100.0%

CATEGORY TOTAL: 193 62 19 0 23 0 20 0 30.6% 67 '% 100.0%

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 42 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE B 2 - PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL ASME NO. OF COMPONENTS SEC XI EXAH SYSTEM fi OF NO. SCHEDULED/COMPLETED ITEM 4 ITEN DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS C5. 53 CIRCUHFERENTIAL WELDS VOLUHETRIC Control Rod 59 6 2 0 2 0 2 0 7.58 but not less

>3/8" NOMINAL WALL SURFACE Driye 33 3% 66 6% 100 0% than 28 welds THICKNESS FOR PIPING

>NPS 4 VOLUMETRIC Hain Steam 30 3 1 0 1 0 1 0 7.5% but not less SURFACE than 28 welds 33.3% 66.6% 100.0%

VOLUMETRIC Main Steam 49 6 2 0 2 0 2 0 7.5% but not less SURFACE System than 28 welds

33. 3% 66. 6% 100. 0%

VOLUMETRIC Reactor 342 27 7 0 8 0 12 0 7. 5% but not less SURFACE r,i, ~

a i nmg. ~

25. 9>> 55. 5~ lpp pg than 28 welds voLUMETRIc Reactor 0 0 0 0 0 0 0 7.5% but not less SURFACE Coniafnment than 28 welds 0 0%

~ 0.0% 0.0g Spray Pump 111 VOLUHETRIC Reactor 0 0 0 0 0 0 0 7.5% but not less Containment p pg p pg p pg than 28 welds Spray Pump 112 VOLUMETRIC Reactor 2 0 0 0 0 0 0 0 7.5% but not less SURFACE Containment than 28 welds p pg p pg p pg Spray Pump 121 VOLUHETRIC Re~ tn! 0 0 0 0 0 0 0 7.5% but not less SURFA' 0 pg p pg than 28 welds VOLUMETRIC Reactor Cor. 225 21 8 0 7 0 6 0 7.5% but not less SURFACE Spray than 28 welds 38 pg 71 4g lpp pg

0 DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 4 3 REVISION: 00 INSERVICE INSPECTION PLAN fOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI SUYMARY TABLE B PRESSURE RETAINING WELDS ZN CARBON OR LOW ALLOY STEEL ASME NO ~ Of COMPONENTS SEC XI EXAM SYSTEM I OF NO. SCHEDULED/COMPLETED ITEM 5 ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMHENTS CS. 51 VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 7.5S but not less SURFACE Spray Pump 111 p pS p pS p pS than 28 welds VOLUMETRIC Re~or~a Cor~ 0 0 0 0 0 0 0 7.5S but not less rl'llL'I, 'F ai.-r,ly i urnj than 28 welds p OS p.pS p pS VOLUHETRIC Reactor Core 2 0 0 0 0 0 0 0 7.5S but not less SURFACE Spray pump 121 p.pS p.pS p.pS tharr 28 welds VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 7.5S but not less SURFACE Spray pump than 28 welds p pS p pS p OS ITEH TOTAL: 721 63 20 0 20 0 23 0 31.7S 63. 4S 100. OS C5.52 LONGITUDINAL WELDS >3/8" VOLUMETRIC Reactor 2 0 0 0 0 0 0 0 2.5t at NOHINAL WALL THICKNESS SURFACE Containment O.OS O.OS p pS intersecting Circ.

FOR PIPING >NPS 4 Spr*y Pump l 1] welds, Code Case N-524 VOLVHETirl", Re-~ " .: u v 0 0 v 0 2.5t at Containment O.OS O.OS p.pS intersecting Circ.

Spray Pump 112 welds, Code Case N-524 VOLUMETRIC Reactor 2 0 0 0 0 0 0 0 2 St at Containment O.OS 0.0'S p OS intersecting Circ.

Spray Pump 121 welds, Code Case N-524 VOLUMETRIC Reactor 2 0 0 0 0 0 0 0 2.5t at Containment. O.OS O.OS p pS intersecting Circ.

Spray Pump 122 welds, Code Case N-524

DATE: 10/26/99 NittE MILE POINT tfUCLEAR PLANT UNIT 1 PAGE: 44 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTI Oft Xl

SUMMARY

TABLE B PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM Of'O. SCHEDULED/COMPLETED ITEM tt ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 2.5t at SURFACE Spray pump O.pft O.pt p pft intersecting Circ.

welds, Code Case N-524 VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 2.5t at SURFACE Spray pump 0. Oft 0. 0% p,pft intersecting Circ.

welds, Code Case N-524 VOLUMETRIC React>r C~ re 0 0 0 0 0 0 0 2.5t at SURFACE Spray pump lpl O.pft 0.0% p pft intersecting Circ.

welds, Code Case N-524 VOLUMETRIC Reactor Core 2 0 0 0 0 0 0 0 2.5t at SURFACE Sptay pumk" 0.04 0. Pf! p pft intersecting Circ.

welds, Code Case N-524 ITEM TOTAL: 16 0 0 0 0 0 0 0 0.0% 0.0% 0.0%

C5.61 CIRCUMFERENTIAL WELDS > N/A 1/5" NOMINAL HALL THICKNESS FOR PIPING >

NPS 2 AND < NPS 4 C5. 62 LONGITUDINAL WELDS > N/A

'1/5" tfOMINAL WALL TH f CKNL'SS FOR 1". f 1 1t(1 ttPS 'Nb ~ NPS C5.70 SOCKET WELDS NIA

'5.81 CIRCUMFERENTIAL PIPE SURf'ACE tfain Steam 16 1 0 0 1 0 0 0 7.5L, But not less BRANCH CONNECTIONS System p pg lpp pft lpp pft than 28 welds PIPING > NPS 2 Of'RANCH SURFACE Reactor 6 1 1 0 0 0 0 0 7.5L, But not less Containment lpp pft lpp pq lpp pft than 28 welds Spk ay SURFACE Reactor 2 0 0 0 0 0 0 0 7.5%, But not less Containment p pft p pg p pg than 28 welds Spray Pump 111

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 45 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE B PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL ASME NO. OF COMPONENTS SEC /1 E.'iAM SYSTEM ff OF NO. SCHEDULED/COMPLETED ITEM I ITEM DESCRIPTION METHOD DESCklP'f'lON COMf'EO 1ST PER 2ND PER 3RD PER COMMENTS SURF'ACE Reactor 0 0 0 0 0 0 0 7.5%, But not less Containment p.p% p.p% p.p% than 28 welds Spray Pump 112 SURFACE Reactor 2 1 0 0 0 0 1 0 7.5%, But not less Containment p p% p p% lpp p% than 28 welds Spray Pump 121 SURFACE Reactor 2 0 0 0 0 0 0 0 7.5%, But not less Containment p p% p p% p p% than 28 we)ds Spray Pump 122 SURFACE Reactor Core 2 0 0 0 0 0 0 0 7.5%, But not less Spray Pump 111 p p% p p% p p% than 28 welds SURFACE Reactor Core 0 0 0 0 0 0 0 7.5%, But not less Spray Pump 11." 0.0% p.p% p.p% than 28 welds SURFACE Reactor Core 2 0 0 0 0 0 0 0 7.5%, But not less Spray Pump 1 "1 0.0% 0.0% p.p% than 28 welds SURFACE Reactor Core 2 0 0 0 0 0 0 0 7.5%, But not less Spray Pump 122 p p% p p% p p% than 28 welds ITEM TOTAL: 38 3 1 0 1 0 1 0 33.3% 66.6% 100.0%

C5.82 LONGITUDINAL PIPE BRANCH N/A CONNECTION OF BRANCH PIPING ) NPS 2 CATEGORY TOTAL: 775 66 21 0 21 0 24 0 31.8'%3.6% 100.0%

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 46 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION t E89 CLASS 2 SECTION XI

SUMMARY

TABLE B PRESSURE RETAINING WELDS ZN PUMPS AND VALVES ASME NO. OF COMPONENTS SEC XI EXAM SYSTEH OF NO. SCHEDULED/COMPLETED ITEN N ITEM DESCRIPTION HETHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COHMENTS C6.10 PUHP CASING WELDS SURFACE Reactor 10 0 0 0 0 0 0 0 Only one pump Containment 0 Og 0 Og 0 Og among each group Spray Pump 111 of pump SURFACE Reactor 10 0 0 0 0 0 0 0 Only one pump Containment 0.0tt 0 Og 0 Og among each group Spray Pump 112 of pump SURFACE Rea.tor 10 5 1 0 4 0 0 0 Only one pump Conc ~ > r aietit 0 01 100 0% 100 Oi among each group Spray Pump 1 "1 of pump SURFACE Reactor 10 0 0 0 0 0 0 0 Only one pump Containment 0 Ott 0 Og 0 Ott among each group Spray Pump 122 of pump SURFACE Reactor Core 10 0 0 0 0 0 0 0 Only one pump Spray Pump 111 0 Og 0 Og 0 Og among each group of pump SURFACF. Ri a Spray "t ~: ~" r" Pump 11

'<< 0 0 0 Oi 0 0 0 Og 0 0 0 Ott 0 Only one pump among each group of pump SURFACE Reactor Core 10 4 0 0 0 0 4 0 Only one pump Spray Pump 121 0 Ot, 0 Ott 100 Og among each group of pump SURFACE Re ct.or Core 10 0 0 0 0 0 0 0 Only one pump Spray Puaip 1 "2 O.Ott 0.0tt 0.0% among each group of pump ZTEM TOTAL: 80 9 1 0 4 0 4 0 11.1% 55.5% 100.0%

C6.20 VALVE CASING WELDS CATEGORY TOTAL: 80 9 1 0 4 0 4 0 11.11 O.OS 100.08.

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 LI T NlAGARA THIRD INSERVICE INSPECT(ON INTERVAL H U MOHAWK Rev: 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX C - CLASS 3

SUMMARY

TABLES Table of Contents C-1 Record of Revision C-2 ASME Code Class 3 Section XI Summary Tables . C-3 Examination Category D-A 47-48 File: APPENDIXC.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 EI Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION REVISION

'ATE AFFECTED REASON FOR REVISION PAGES I" 0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year Inservice Inspection Interval Fiie: APPENDIXC.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ASME CODE CLASS 3

SUMMARY

TABLES File: APPENDIXC.WPD

0 DATE< lp/26/99 NlNE Ml l E P<,<lNT NUCLEAR PLANT PAGE: g7 INSPECTION PLAN FOR THE THIRD INTERVAL UNIT'NSERVICE REVISION: 00 CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE B INTEGRAL ATTACHMENTS FOR CLASS 3 VESSELS < PI PING< PUMPS <

ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM II OF NO. SCHEDULED/COMPLETED ITEM ll ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMFNTS Dl.lp PRESSURE VESSEL VISUAL Emergency 3 0 0 0 0 0 0 0 1008 of weld INTEGRALLY WELDED Condenser Heat O.pl O.pll p pll length, Code Case ATTACHMENTS Exchanger 111 N-509, 104 required VISUAL Emergency 3 0 0 0 0 0 0 0 1008 of weld Condenser Heat O.OS O.pl< O.pg length, Code Case Exchanger 112 N-509, 108 required V'UAl. E<<<':"g<<'f< "/ 0 0 V 0 0 0 0 1004 of weld u ~ <<, << I<, (,<<< length, Code Case H-509, 104 required VISUAL Emergency 3 3 0 0 0 0 3 0 100<i of weld Condenser Heat 0. Ol< p pg lpp pa length, Code Case Exchanger 122 N-509, 10%

required VISUAL Reactor Building 6 2 2 0 0 0 0 0 1008 of weld Closed Loop lpp pll lpp pl, lpp p<l length, Code Case Cooling Heat N-509, 10%

Exchange required VISUAL Shutdown Cooling 2 2 0 0 0 0 2 0 100% of weld Water Heat 0.0% p p<l lpp pll length, Code Case Exchanger 11 H-509, 10%

required VISUAL Sl<<<< ~iw;< ~' . i<<I p p 0 0 100% of weld

<,>. 0% u. 01 0 pg length, Code Case N-509, 10%

required VISUAL Shutdown Cooling 2 0 0 0 0 0 0 0 100% of weld Water Heat 0. O'L 0. Ol< p pl< length, Code Case Exchanger 13 N-509, 10%

required VISUAL Spent Fuel Pool 2 0 0 0 0 0 0 0 100% of weld Coo)ing Heat p pg p p<l p pll length, Code Case Exchanger 11 N-509, 10%

DATE: 10/26/99 HINE MILE POIHT NUCLEAR PLANT UNIT 1 PAGE: 48 REVISIOH: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABI E C INTEGRAL ATTACHMENTS FOR CLASS 3 VESSELS, PIPING, PUMPS, ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM j> ITEM DESCRIPTIOH METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS Dl. 10 required VISUAL Spent Fuel Pool 2 0 0 0 0 2 0 lpptt of weld Ccnling Heat 0. 0% p ptt lpp pg length, Code Case Exchanger 12 N-509, 1 ptt required ITEM TOTAL: 28 9 2 0 0 0 7 0 22.2'2.2S 100.0't D1.20 PIPING INTEGRALLY WELDED VISUAL Emergency 29 2 0 0 0 0 2 0 lpptt of weld ATTACHMENTS Service Water 0.0% p ptj lpp pg length, 10$ of all integral attachments, Code Case N-509 VISUAL Reactor Buildir>g 53 7 0 0 2 0 5 0 100% of weld Closed Loop O.ptt 28.5% 100.0% length, 10% of all Cooling integral attachments, Code Case N-509 VISUAL Reactor "19 15 15 0 0 0 0 0 100% of weld C'r a j nm>et>r 100.0% 100.0% 100.0% length, 10% of all Spray integral attachments, Code Case N-509 VISUAL Spent Fuel Pool 158 20 0 0 16 0 4 0 100% of weld Cooling p.ptt Bp pjt lpp pg length, lptj of all integral attachments, Code Case N-509 ITEM TOTAL: 459 44 15 0 18 0 11 0 s4.0% 75.0t> lpp.ptj Dl.3u PUMPS INTEGRALLY WELDED NtA ATTACHMENTS Dl. 40 VALVES INTEGRALLY WELDED ATTACHMENTS CATEGORY TOTAL: 487 53 17 0 18 0 18 0 32.0tj 66.0t 100.0'1

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev.

U MOHAWK 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX D - CLASS 1, 2, 3 COMPONENT SUPPORT

SUMMARY

TABLES Table of Contents D-1 Record of Revision . D-2 ASME Code Class 1, 2, 3 Section XI Component Support Summary Tables .. D-3 Examination Category F-A 49-58 ASME Code Class 1 Supports 49- 51 ASME Code Class 2 Supports 51 -52 ASME Code Class 3 Supports 52- 53 Other Than Piping Supports . 53- 58 File: APPENDIXD.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL H U MOHAWK Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN RECORD OF REVISION REVISION AFFECTED REASON FOR REVISION PAGES .

0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year lnservice Inspection Interval File: APPENDIXD.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 T NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN ASME CODE CLASS 1, 2, 3 COMPONENT SUPPORT

SUMMARY

TABLES File: APPENDIXD.WPD

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 49 REVISION: 00 INSERVICE INSPECTION PLAN fOR THE THIRD INTERVAL CODE EDITZON: E89 CLASS 2 SECTION XI SUHMARY TABLE C Class I Piping Supports ASHE NO. OF COMPONENTS SEC XZ EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEH I ITEM DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COHMENTS Fl.lp 25% of Class 1 VISUAL Control Rod 1 0 0 0 1 0 Categorized to Drive 50.0i 50.0'i 100.08 identify suPPort types by component function A, B, C, etc, Code Case N-491-1 VISUAL Emergency 15 3 1 0 0 0 2 0 Categorized to Condenser Return 33 3$ 33 3Q lpp pi identify support types by component function A, B, C, etc, Code Case N-4 91-1 VISUAL Feedwater System 18 5 1 0 2 0 2 0 Categorized to 2p pq 60 pg lpp pg identify support types by component function A, B, C, etc, Code Case N-4 91" 1 VISUAL Liquid Poison 7 1 0 1 0 0 0 Categorized to 50.0% 100.0% 100.0% identify support types by component function A, B, C, etc, Code Case N-4 91-1 VISUAL Hain Steam 16 1 0 1 0 2 0 Caregorized to System 25 pi 5p pi lpp pt i ent y upport types by component function A, B, C, etc, Code Case N-491-1 VISUAL Reactor Clean Up 18 2 1 0 0 0 1 0 Categorized to 5p pg 5p pi Ipp.pq identify support types by component function A, B, C, etc, Code Case N-4 91-1 VISUAL Reactor Core 6 2 0 2 0 2 0 Categorized to Spray lpp pi identify support

DATE: 10/26/99 N 1 WE Mi LL'VltlT NUCLEAR Pt&flT UNIT 1 PAGE: 50 REVISIOH: 00 INSERVICE INSPEC 1Otl PLAtl FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECL'IOtl il

SUMMARY

TABLE D

- Class 1 Piping Supports ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEM tt ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS Fl. 10 types by component function A, B, C, etc, Code Case N-4 91-1 VISUAL Reactor Drain 1 0 0 0 1 0 Categorized to 50.0tt 50.0$ 100.0% identify suPP rt types by component function A, B, C, etc, Code Case N-491-1 VISUAL Reactor Head 10 2 1 0 1 0 0 0 Categorized to Vent ipp ptt iden'tify suppor't types by component function A, B, C, etc, Code Case H-491-1 VISUAL Reactor 4 1 0 0 0 0 1 0 Cat:egorized to Instrumentation 0.0% p.ptt lpp pg identify support types by component function A, B, C, etc, Code Case H-491-1 VISUAL Reactor 30 8 2 0 0 2 0 Categorized to Recircuiation 25. 0'1 01 lpp pi identify support.

Discharge types by component function A, B, C, etc, Code Case N-491-1 VISUAL Reactor 25 8 2 0 2 0 4 0 Categorized to Recirculation 25.ptt 5p pg ipp pit identify support Suction types by component function A, B, C, etc, Code Case tl-491-1 VISUAL Reactor Shutdown 8 1 0 0 0 0 1 0 Categorized to Cooling 0.0% p,pg lpp pt, identify support types by component function A, B, C, etc, Code Case H-4 91-1

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UtlIT 1 PAGE: 51 REVISION: 00 INSERVICE INSPECTION PLAtl FOR THE THIRD ItlTERVAL CODE EDITION: E89

'tJd:: t SE /Iuti Xi

SUMMARY

TABLE D Class 1 Piping Supports ASME NO. OF COMPONENTS SEC XI EXAM STSTEM ft OF NO. SCHEDULED/COMPLETED ITEM tt ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS ITEM TOTAL: 190 46 14 0 13 0 19 0 30.4tt 58.6tt 100.0%

F1.20 15% of Class 2 Supports VISUAL Control Rod 32 4 3 0 0 0 1 0 Categorized to Drive 75 ~ 0% 75. 0tt 100. Ott identify support types by component function A, B, C, etc.Code Case N-491-1 VISUAL Emergency 31 3 0 0 1 0 2 0 Categorized to Condenser Return 0 01 33 3tt 100 0$ identify support types by component function A, B, C, etc.Code Case N" 4 91 "1 VISUAL Emergency 32 3 3 0 0 0 0 0 Categorized to Condenser Supply 100.0'1 100.0tt 100.Otal identify suPPort types by component function A, B, C, etc.Code Case N-491-1 VISUAL Main Steam 13 2 2 0 0 0 0 0 Categorized to 100 0tt 100 Otal 100 0tt identify support types by component function A, B, C, etc.Code Case N-491-1 VISUAL Main S>~am 34 3 0 < 3 0 0 0 Categorized to S'/s~ em 100 0< 100 0tt identify support types by component function A, B, C, etc.Code Case N-4 91-1 VISUAL Reactor 178 28 14 0 9 0 5 0 Categorized to Containment 50.0% 82.1tt 100.0S identify support

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 52 REVISION: 00 INSERVICE INSPECTION PLAN fOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 3 SECTION XI SUMHARY TABLE F Class 2 Piping Supports ASHE NO. Of COMPONENTS SEC XI EXAM SYSTEH ll OF NO. SCHEDULED/COMPLETED ITEH 5 ITEH DESCRIPTION METHOD DESCRIPTION COHP REQ 1ST PER 2ND PER 3RD PER COMMENTS Fl . 20 Spray types by component function A, B, C, etc.Code Case N-4 91-1 VISUAL R '.i "', i Core

~ 7 0 6 0 9 0 Categorized to Spkey 31 8t 59 Ot 100 Ot iden ti fY support types by component function A, B, C, etc.Code Case N-491-1 ITEM TOTAL: 433 65 29 0 19 0 17 0 44.6t 73.8t 100.0t F1.30 10t of Class 3 Supports VISUAL Emergency 57 6 3 0 1 0 2 0 Categorized to Service Water 50.0t 66.6t 100.0t identify types by component support function A, B, C, etc.Code Case N-4 91-1 VISUAL Reactor Building 214 19 0 0 4 0 15 0 Categorized to Closed Loop 0 Ot 21 Ot 100 Ot identify types by Cool in p component support function A, B, C, etc.Code Case N-4 91-1 VISUAL Reactor 140 15 10 0 5 0 0 0 Categorized to Containment 66.6t 100.0t 100.0t iden<<fy types bY Spray component support function A, B, C, etc.Code Case N-4 91-1 VISUAL Spent Fuel Pool 2 0 0 0 0 0 0 0 Categorized to Cooling 0. 0't 0. Ot 0 Ot identify types by component support function A, B, C, etc.Code Case N-491-1

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 53 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 3 SECTIOtt XI

SUMMARY

TABLE F

- Class 3 Piping Supports ASME HO. OF COMPONENTS SEC XI EXAM SYSTEM OF HO. SCHEDULED/COMPLETED ITEM tt ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS Fl . 30 Categorized to identify types by component support function A, B, C, etc.Code Case H-491-1 VISUAL Spent Fuel Pool 5 1 0 0 0 0 1 0 Chtegorized to coil ing System 0.0% 0 0% 100 0% identify types by component support function A, B, C, etc.Code Case H-4 91-1 ITEM TOTAL: 503 51 14 0 17 0 20 0 2t.4% 60.t% 100.0%

Fl. 40 100% of the supports, VISUAL Emergency 1 0 0 0 0 0 0 0 Code Case H-491-1 For multiple components, Condenser Heat 0.0% 0.0% 0.0%

only one of multiple Exchanger 111 components required VISUAL Emergency 1 0 0 0 0 0 0 0 Code Case N-491-1 Condenser Heat 0.0% "

0.0% 0.0%

Exchanger 112 VISUAL Emergency 1 0 0 0 0 0 0 0 Code Case H4911 Condens r t!.AL 0.0% 0.0%

0.0'%lSUAL Emergency 1 1 v 0 0 0 1 0 Code Case H-491-1 Condenser Heat 0.0% 0.0% 100.0%

Exchanger 122 VISUAL Emergency 1 1 0 0 0 0 1 0 Code Case tt-491-1 Service Water 0.0% 0.0% 100.0%

Pump ll VISUAL Emergency 1 ~ 0 0 0 0 0 0 0 Code Case tt-491-1 Service Water 0.0% 0.0% 0.0%

Pump 12

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 54 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 3 SECTION yI SUMHARY TABLE F

- Supports Other Than Piping Supports ASME NO. OF COMPONENTS SEC XI EXAM SYSTEM jj OF NO. SCHEDULED/COMPLETED ITEM jj ITEM DESCRIPTION HETHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VISUAL Hain Steam 1 1 0 0 0 0 1 0 Code Case N-491-1 O.OL 0.0jj 100.0't VISUAL React:or Building 3 1 1 0 0 0 0 0 Code Case N-491-1.

Closed Loop 100.05 100.0'00.0tj Cooling Heat Exchange VISUAL Reactor Building 1 1 0 0 0 0 1 0 Code Case N-491-1 Closed Loop Cooling HU Tank 0 'jj O.OS 100.0jj VISUA!, RP qt t.+ I R< j 1 d j nc j n n 0 0 1 0 Code Case N-491-1 0.0e 0.0% 100.0%

VISUAL Reactor 1 0 0 0 0 0 0 0 Code Case N-491-1 Containment 0.0% 0.08 0.0'j Spray Heat Exchanger 111 VISUAL Reactor 0 0 0 0 0 0 0 Code Case N-491-1 Containment 0.0jj 0.0% 0.0'5 Spray Heat Exchanger 112 VISUAL Reactor 1 0 0 0 0 0 0 0 Code Case N-491-1 Containment O.OS 0.0'.0%

Spray Heat Exchanger 121 VISUAL Reactor 1 1 1 0 0 0 0 0 Code Case N-491-1 ContdinlfLenl 100.0'4 100.0% 100.04 VISUAL React. or 1 0 0 0 0 0 0 0 Code Case N-491-1 Containment 0. 0jj 0. 0% 0. 0%

Spray Pump 111

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 55 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITIOHt E89 CLASS 2 SECTIOH XI SUMHARY TABLE E Supports Other Than Piping Supports ASHE NO. OF COMPONENTS SEC XI EXAH SYSTEM fi OF HO. SCHEDULED/COMPLETED ITEM h ITEM DESCRI PT!OH HET!ivv DESCRIPTIOlr COMP REQ IST PER 2ND PER 3RD PER COHMENTS Fl . 4v VISUAL Reactor I 0 0 0 0 0 0 0 Code Case N-491-1 Containment 0.0% 0.0% 0.0%

Spray Pump 112 VISUAL Reactor I 1 1 0 0 0 0 0 Code Case N-491-1 Containment 100.0% 100.0% 100.0%

Spray Pump 121 VISUAL Reactor 1 0 0 0 0 0 0 0 Code Case H-491-1 Containment 0.0% 0.0% 0.0%

Spray Pump 122 VISUAL Reactor I I 0 0 I 0 0 0 Code Case H-491-1 Containment 0.0% 100.0% 100.0%

SP . <<'/ kaw Wet er t'irrfe VISUAL keact.ul'onta

! u 0 0 0 0 0 0 Code Case N-491-1 inrrient, 0.0% 0.0% 0.0%

Spray Raw Water Pump 112 VISUAL Reactor 1 0 0 0 0 0 0 0 Code Case N-491-1 Containment 0.0% 0.0% 0.0%

Spray Raw Water Pump 121 VISUAL Reactor I 0 0 0 0 0 0 0 Code Case H-491-1 Containment 0.0'%.0% 0.0%

Spray Raw Water Pump 122 VISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case H-491-1 Spray Pump I! I D. 0% O. 0'%. 0%

VISUAL ke 9 ".t or Ct t ! r'r u 0 0 0 0 0 Code Case H-491-1 Purr'1" t'pray 0.0% 0.0% 0.0%

DATE: 10/26/99 NINE HILE POINT NUCLEAR PLANT UNIT 1 PAGE: 56 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 2 SECTION XI

SUMMARY

TABLE E

- Supports Other Than Piping Supports ASHE NO. OF'OMPONENTS SEC XI EXAM SYSTEH OF NO. SCHEDULED/COMPLETED ITEH 8 ITEM DESCRIPTION METHOD DE SCR I PTI ON COMP REQ 1ST PER 2ND PER 3RD PER COMHENTS Fl. 40 VISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case N-491-1 Spray Pump 121 0.0% 0.0%

0.0'%ISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case N-491-1 Spray Pump 122 0.0% 0.0% 0.0%

VISUAL Reactor Core 1 1 1 0 0 0 0 0 Code Case N-491-1 Spray Topping 100. 0% 100. 0% 100. 0%

Pump 111 VISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case N-491-1 Spray Topping 0.0% 0.0% 0.0%

Pump 112 VISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case N-491-1 Spray Topping 0.0% 0.0% 0.0%

Pump 121 VISUAL Reactor Core 1 0 0 0 0 0 0 0 Code Case N-491-1 Spr~y Trpplng 0.0% 0.0% 0.0%

VISUAL Reactor 7 1 0 0 1 0 0 0 Code Case N-491-1 Recirculation 0.0% 100.0'%00.0%

Pump 11 VISUAL Reactor 7 0 0 0 0 0 0 0 Code Case N-491-1 Recirculation Pump 12

0. 0% 0. 0'%. 0%

DATE: !u/26/99 WINE MILE POINT NUCLEAR PLANT UtkIT 1 PAGE: 57 REVISIO!f: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 1 SECTION VI SUMHARY TABLE D Supports Other Than Piping Supports ASME NO. OF CCHPONENTS SEC /I EXAM SYSTEM fk OF WO. SCHEDULED/COHPLETED ITEN E ITEM DESCRIPTION METHOD DESCRIPTION COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS E'1. 40 Reactor Recirculation Pump 13 VISUAL Reactor 7 0 0 0 0 0 0 0 Code Case N-491-1 Recirculation 0.0% 0.0% 0.0%

Pump 14 VISUAL Reactor 7 1 0 0 1 0 0 0 Code Case N 491 1 Recirculation 0. 0% 100. 0% 100. 0%

k'Umf VISUAL Re<<t.oc Vessel 9 9 2 0 2 0 5 0 Code Case N-491-1 Supports 22.2% 44.4% 100.0%

VISUAL Shutdown Cnnl ing 1 1 0 0 0 0 1 0 Code Case N-491-1 Water Heat 0.0% 0.0% 100.0%

Exchanger 11 VISUAL Shutdown Cooling 1 0 0 0 0 0 0 0 Code Case N-491-1 Water Heat 0.0% 0.0% 0.0%

Exchanger 12 VISUAL Skiutdown Cool i ng 1 0 0 0 0 0 0 0 Code Case W-491-1 Water Hear 0.0% 0.0% 0.0%

VISUAL Spent Fuel Pool 1 1 0 0 1 0 0 0 Code Case W-491-1 Cooling E'liter 0.0% 100.0% 100.0%

ll VISUAL Spent Fuel Pool 1 0 0 0 0 0 0 0 Code Case N-491-1 Cooling E'ilter 0.0% 0.0% 0.0%

l~

DATE: 10/26/99 NINE MILE POINT NUCLEAR PLANT UNIT 1 PAGE: 58 REVISION: 00 INSERVICE INSPECTION PLAN FOR THE THIRD INTERVAL CODE EDITION: E89 CLASS 3 SECTION XI

SUMMARY

TABLE F Supports Other Than Piping Supports ASME NO. OF COMPONENTS SFC XI EXAM SYSTEM OF NO. SCHEDULED/COMPLETED ITEMi >>> ITEM DESCRIPTION METHOD DESCR I PTI ON COMP REQ 1ST PER 2ND PER 3RD PER COMMENTS VI 8>.'A> :> i.:,> F>>nl Pool 0 0 0 0 0 0 0 Code Case N-491-1

> )1>: Heat 0.0% 0.0% 0.0%

x 'ha>>g"r 1 1 VISUAl Spent Fuel Pool 1 1 0 0 1 0 0 0 Code Case N-491-1 Cooling Heat 0.0% 100.0% 100.05 Exchanger 12 VISUAL Spent Fuel Pool 1 1 0 0 0 0 1 0 Code Case N-491-1 Cooling Pump 11 0.0% O.OS 100.0%

VISUAL Spent Fuel Pool 1 0 0 0 0 0 0 0 Code Case N-491-1 Cooling Pump 12 0.0% 0.0% 0.0%

V:. UAL 1 0 0 0 Code Case N-491-1

] >>'0. 0>> 100. 0$

ITEM TOTAL: 89 26 6 0 8 0 12 0 23.0% 53.88 100.0%

CATEGORY TOTAL: 1215 188 63 0 57 0 68 0 33.50 63.8% 100.0%

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 hl V NlAGARA THIRD INSERVICE INSPECTION INTERVAL H U MOHAWK Rev. 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX E - CODE BOUNDARY DIAGRAMS Table of Contents Record of Revision E-2 ASME Code Boundary Classification Diagram Listing . E-3 Fiie: APPENDIXE.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NAGARA THIRD INSERVICE INSPECTION INTERVAL Rev.

NU MOHAWK 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN

~.:~AX~~"','9 " '~""-.~,REC@AD.'OF;RBVf8ION" -.~+,,;mme%-.Sinai~;~.:;-,'='-:.".."."",-'";;

>',ZPR~~~4~~

.";REVISlON4 TREASONFOR;REVISION::;",,' "-...'

PAGES +%~@+ > 4'AP c 0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year Inservice Inspection Interval File: APPENDIXE.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit1 hl V NIAGARA a U MOHAWK THIRD INSERVICE INSPECTION INTERVAL Rev.

Date:

0 September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN CODE BOUNDARY CLASSIFICATION BOUNDARY DIAGRAMS The NMP1 Code Boundary Classification Diagrams identifying Quality Group A, B and C (ASME Code Class 1, 2 and 3) systems are described below.

CODE BOUNDARY DIAGRAM LISTING Boundary Diagram Number System Title F-63002-C Main Steam and HP Turbine F-63003-C Condensate Flow F-63005-C High Pressure FW Flow F-63006-C Drywell and Torus Isolation Valves F-63007-C Reactor Core Spray F-63008-C Spent Fuel Storage Pool Filter & Cooling F-63009-C Reactor Cleanup System F-63011-C Instrument Air F-63012-C Reactor Containment Spray F-63013-C Reactor Building Heat and Cooling F-63014-C Drywell & Torus Inert Gas & Cooling F-63015-C Reactor vessel Instrumentation F-63016-C Control Rod Drive F-63017-C Emergency Cooling System F-63018-C Reactor Shutdown Cooling F-63019-C Reactor Liquid Poison System F-63020-C Reactor Recirculation Loops F-63021-C Turbine Building Heating & Cooling F-63022-C Service Water, Closed Loop Cooling F-63026-C Diesel Generator Air, Water, Oil & Fuel F-63027-C Service Water F-63035-C Resin Transfer Regeneration F-63036-C Sealing Water File: APPENDIXE.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 El V NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev.

H @MOHAWK 0 INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN CODE BOUNDARY DIAGRAM LISTING Boundary Diagram Number System Title F-63041-C Sampling F-63045-C Waste Disposal F-63046-C Air Conditioning F-63047-C Heating, Ventilating 8 Air Conditioning F-63048-C Condensate Transfer File: APPENDIXE.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN TABLE OF CONTENTS APPENDIX F - RELIEF REQUESTS Table of Contents F-1 Record of Revision F-2 Requests for Relief Summary. F-3 ISI-1 1-1 of 1-8 ISI-2 2-1 of 2-8 ISI-3 3-1 of 3-7 ISI-4 4-1 of 4-4 ISI-5 5-1 of 5-5 ISI-6 6-1 of 6-3 ISI-7 7-1 of 7-4 ISI-8 8-1 of 8-6 ISI-9 9-1 of 9-3 ISI-10 10-1 of 10-3 ISI-11 11-1 of 11-1 ISI-12 12-1 of 12-3 File: APPENDIXF.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 El Y NIAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK Date: September 27, 1999 INSERVICE INSPECTION PROGRAM PLAN

',".'" "..I<""-".RECORD'OFIREVISION;..":,";.:.'i';"'",~".""':-""""'~"';-'".'",." '".":."'".",,",;""

. 'REVISION >> ,'<'" ','DATE '" AFFECTED.",;.:".".:i:,"; '-REASON'FOR REVISION;, '.;

PAGESl..","--"':.".':.-"'

0 September 27, 1999 Entire Document Updated Inservice Inspection Program Plan for the 3" Ten Year Inseivice Inspection Interval File: APPENDIXF.WPD

Nine Mlle Point Nuclear Power Station NMP1-ISI-003 Unit 1 V NlAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN CLASS 1, 2 AND 3 RELIEF REQUEST

SUMMARY

Relief (Exam,',,'item:

Su'm'mary.'of,R'equest ""',

..Proposed Alternatives- Relief

'Req. System'.or Exam. ", for Relief t ."<

Request

', No. 'ompo'nent;.,

, Cat;g'.,-,-," ;No;,-':, Status, ISI-1 Reactor Pressure 8-A 81.11 Relief is requested for NMPC proposes to Granted Vessel 81.12 permanent relief from perform examination as TAC examining RPV Circ. defined in the Relief MA4383 Welds Request ISI-2 Reactor Pressure 8-A 81.21 Relief is requested Perform exams to the Vessel 81.22 from the performing extent practical 81.30 100% of weld length 81.40 ISI-3 Reactor Pressure 8-D 83.90 Request Relief from Perform exams to the Vessel 83.100 performing 100% of extent practical CRV ISI-4 Reactor Pressure 8-A 81.30 Relief is requested NMPC proposes to Vessel 81.12 from IWB-2420(b) utilize ASME Code reexamination Case N-526, requirements "Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels" ISI-5 Reactor Pressure 8-K 810.10 Relief is requested NMPC proposes to vessels from performing 100% perform additional exam of weld length per Relief ISI-6 Reactor Pressure 8-0 814.10 Relief is requested NMPC proposes to vessel from performing 100% perform additional exam of CRA per Relief Request ISI-7 Reactor Containment C-G C6.10 Relief is requested Perform examinations Spray & Reactor Core from performing to the extent practical Spray Pumps examination of 100% from the outside welds on one pump surface, and from the among group of pumps inside when disassembled File: APPENDIXF.WPD

Nine Mile Point Nuclear Power Station NMP1-ISI-003 Unit 1 H Y NlAGARA THIRD INSERVICE INSPECTION INTERVAL Rev. 0 H U MOHAWK INSERVICE INSPECTION Date: September 27, 1999 PROGRAM PLAN CLASS 1, 2 AND 3 RELIEF REQUEST

SUMMARY

": i,"'::.':;: ".:..':-',

Relief Req.'o. SysteIn cia, Comp'onent,:,;;

,Exa'm'~'at.,

';.Sx'a'm~",".-'

-'for Relief,. ',,

Summarjj of Request':;.;: Proposed'Alternatives

';:;~'-; ,

Retief.:...':-.

Reque'st.

Status ISI-8 ISI Summary Report N/A N/A Relief is requested NMPC proposes to IWA-6000 from Article IWA-6000 utilize 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" Sl-9 ASME Code Class 1, 2 N/A N/A Relief is requested NMPC proposes to and 3 Snubbers from Section 2.3.2.2 utilize USNRC Generic and 2.3.2.3 of OM- Letter 90-09 for 1988, Part4 and intervals and sampling Article IWF rates, perform exams in accordance with Plant Tech Specs ISI-10 Article IWA-4000 N/A N/A Relief is requested NMPC proposes to from the requirements utilize ASME Code of Article IWA-4000, Case N-573, "Transfer IWA-4400 of Procedure Qualification Records between Owners" ISI-11 Reactor Vessel Closure B-G-1 B6.10 Relief is requested NMPC proposes to Head Nuts from performing 100% utilize the Visual VT-1 surface examination of examination criteria of 64 RVCH Nuts the 1989 Addenda ISI-12 Quality Group A, ASME N/A N/A Relief is requested NMPC proposes to Code Class 1 from performing full utilize Vol and Surface Augmented volumetric and surface exams to the extent Examinations examination of 21 practical and VT-2 each nonconforming service outage for evidence of sensitive piping welds leakage File: APPENDIXF.WPD

NINE MILE POINT UNIT)

THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 ONENT IDE 0 System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Reactor Pressure Vessel Shell Welds E TION E ASME Section XI, Table IWB-2500-1, Examination Category B-A, "Pressure Retaining Welds in Reactor Vessel, Examination Item Number B1.10, "Shell Welds"

,Circumferential Nefds Axial Vfetds";'-'..:-.."..

RVWD-100 RVWD-130 RVWD-131 RVWD-101 RVWD-132 RVWD-133 RVWD-137 RVWD-134 RVWD-135 RVWD-138 RVWD-139 RVWD-140 RVWD-141 RVWD-142 RVWD-143 RVWD-144 10CFR50.55a(g)(6)(ii)(A)(2) states that all licensees shall augment their reactor vessel inservice inspection program by implementing the examination requirements for Reactor Pressure Vessel shell welds specified in Code Item No. B1.10 of Examination Category B-A, Pressure Retaining Welds in Reactor Vessel, Table IWB-2500-1 of Subsection IWB of the 1989 Edition of Section XI, Division I, of the ASME Boiler and Prcssure Vessel Code,and subject to the conditions speci fied in 10CFR50.55a(g)(6)(ii)(A)(3) and (4). As stated in 10CFR50.55a, for thc purposes of this augmented examination, essentially 100 percent as used in Table IWB-2500-1 means more than 90 percent of the required examination volume for each weld. Additionally, IOCFR50.55a(g)(6)(ii)(A)(5) requires licensees that are unable to completely satisfy the augmented Reactor Vessel shell weld examination requirement to submit information to the United States Nuclear Regulatory Commission (USNRC) to support thc determination, and propose an alternative to the extent necessary as to provide an acceptable level of quality and safety.

RELIEF E UE ED Pursuant to USNRC Generic Letter 98-05, "Boiling Water Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds."

NMPC requests permanent relief from the inservice inspection requirements of 10CFR50.55a(g) for volumetric examination of circumferential reactor pressure vessel (RPV) welds (ASME Section XI, Table IWB-2500-1, Examination Category B-A, Examination Item Number B1.11.

D.

ELIMINATIONOF REACTOR PRESSURE VESSEL CIRCUMFERENTIALWELDS FROM INSPECTION On November 10, 1998, the USNRC issued Generic Letter 98-05, "Boiling Water Reactor Licensees Use Of The BWRVIP-05 Report To Request Relief From Augmented Examination Requirements on Reactor.

Pressure Vessel Circumferential Shell Welds" The Letter stated that BWR licensees may seek File: RRISI1.WPD IS I 1-1 of IS I 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-0 permanent relief from performing examinations of the RPV circumferential shell welds for the duration of the original operating license This determination was supported by USNRC staffs Safety Evaluation Report to the BWRVIP, dated July 28, .'1998 which would require, on a plant specific basis, licensees to demonstrate that: (1) at the end of a units license the RPV circumferential welds will continue to satisfy the limiting conditional failure probability, and (2) licensees have implemented operator training and established procedures that limit the frequency of beyond design basis low temperature over-pressure events (LTOP) to the limits specified in usNRC Safety Evaluation Report. The plant specific evaluation against both criteria is provided below.

Demonstrate at the end of license, the circumferential welds willsatisfy the limiting conditional failure probability for circumferential welds ln the USNRC staffs July 28,1998, safety evaluation to the BWRVIP-05 report.

Table 1 illustrates that Nine Mile Point Unit 1(NMP1) has conservatism in comparison to the USNRC Final Evaluation of BWRVIP-05 Limiting Plant Specific Analysis. The chemistry factor, adjustment of the reference temperature (c RT,>>),and mean RT>>, are calculated consistent with the guidelines of USNRC Regulatory Guide 1.99, Rev. 2. The data presented for NMP1 in the BWRVIP response to the USNRC Request For Additional Information (RAI) on BWRVIP-05 is also shown in Table 1.The fluence value on Table 1 bounds the highest fluence beltline circumferential weld. The maximum Cu% and Ni%

variability from the most current data available is also bounded.

TABLE 1 PARAMETER NMP1 COMPARATIVE USNRC LIMITINGPLANT DESCRIPTION PARAMETERS AT32 EFPY SPECIFIC ANALYSIS (BOUNDING PARAMETERS AT 32 EFPY CIRCUMFERENTIAL1ljbj'ELD) SE TABLE 2.$ 4 SE "VIP" SE eCEOGe Fluence, 2.21 x10ts 2 P x1018 2.P x10is RT n/cm'nitial

'F -50 Chemistr Factor 112 151.7 172.2 CU7% 0.22* 0.13 0.183 Ni% 0.20* 0.71 0.704 ART 'F 66.5 86.4 98.1 Mean RT 'F 16.5 86.4 98.1 Notes: SE = USNRC Safety Evaluation, entitled, 'Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC NO. M93925), "dated July 28, 1998 The CU% and Ni% bounds the maximum GL 92-01 NMP1 weld chemistry variability as documented in NMPC's September 4, 1998 RAI response to TAC No. MA1200.

As shown above, the impact of irradiation results in lower plant-specific mean RT >> for the NMP1 circumferential weld material as compared to that for any of the Staff's plant-specific analyses which were performed for the CE fabricated RPV's with the highest adjusted reference temperatures. Comparison of the NMP1 specific data and the data used in the USNRC Final Safety Evaluation indicates the difference is the combined effects of the Ni% and CU% on the Chemistry Factor, which is by itself bounded by the USNRC Independent Assessment, and the initial RTQ>> Therefore, the limiting plant-specific conditional probability of failure P(FIE), determined by the Staff, bounds the NMP1 case through the projected end of license.

Thus the BWRVIP specific results relative to NMP1 as presented in BWRVIP-05 and subsequent RAI responses are consistent with those in the USNRC Independent Assessment. Both analyses conclude that File: RRISI1.WPD ISI 1-2 of ISI 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 the failure probability associated with the circumferential welds is extremely small, and that it is orders of magnitude less than that for axial welds. Therefore, the NMP1 circumferential weld satisfies, at the end of license, the limiting conditional failure probability for circumferential welds in the USNRC Staff's July 28, 1998, Safety Evaluation.

Demonstrate that licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in NRC Safety Evaluation Report to the BWRVIP, dated July 2S, 199B The USNRC staff indicated that the potential for,and consequences of, non-design basis events not discussed in the BWRVIP-05 report should be addressed. In particular, the USNRC stated that non-design, low temperature over-pressure transients (LTOP) should be considered. The USNRC further went on to describe several types of events that could be precursors to an LTOP. The BWRVIP provided a response to this issue concluding that Condensate and Control Rod Drive (CRD) pumps could cause such a condition leading to an LTOP event. This was summarized in the USNRC Safety Evaluation of BWRVIP-05.

NMPC has in place procedures which monitor and control reactor pressure, temperature, and water inventory during all aspects of cold shutdown minimizing the likelihood of an RPV LTOP event. Additionally, these procedures are reinforced through NMPC's reactor operator training program.

The RPV Leakage and Hydrostatic pressure test procedures used at NMP1, have sufficient procedural guidance to prevent LTOP. The leakage test is performed at the conclusion of each refueling outage, while the hydrostatic test is performed once every ten years. These pressure tests are infrequently-performed, complex tasks, and the test procedures are controlled as Special Plant Evolutions. As such, a requirement is included in the procedures for an extensive pre-job briefing to be conducted with all essential personnel including Operations management. The briefing details the anticipated testing evolution with special emphasis on conservative decision making, plant safety awareness, lessons learned from similar in-house or industry operating experiences, the importance of open communications and finally the process in which the test would be aborted if plant systems responded in an adverse manner. Vessel pressure and temperature are required to be monitored throughout the tests to ensure compliance with the plant Technical Specification pressure-temperature cuwe. Also, the procedures require the designation of a "Principal Test Engineer" for the duration of the test who is a single point of accountability, responsible for the coordination of testing from initiation to closure, and maintaining operations and plant management cognizant of the test status.

With regard to inadvertent','ystem injection resulting in an LTOP condition, NMP1 high pressure make-up system, (I.E., THE High Pressure Coolant Injection (HPCI)) as well as the normal Feedwater system are interconnected. The portion of the system for HPCI operation is comprised of two (2) motor driven condensate pumps, feedwater booster and feedwater pumps. HPCI is a mode of operation of the Condensate and Feedwater systems rather than an independent, stand alone system. As such, the HPCI system contains only ILC components as its own dedicated equipment. HPCI initiation is prompted by the Reactor Protection System under the following conditions: (1) a turbine trip, or (2) low reactor water level.

During shutdown of the unit, the associated booster and feedwater pumps in the system are secured in accordance with operating procedures. Equipment malfunction or inappropriate operational action would be necessary to cause inadvertent system operation.

During normal cold shutdown conditions, with the RPV head installed, RPV level and pressure are controlled with the CRD System, Condensate Feedwater System, and Reactor Water Cleanup (RWCU) systems using a "feed and bleed" process. The RPV is not taken solid during these times, and plant procedures require opening of the head vent valves after the reactor has been depressurized to approximately 15 psig.

The Liquid Poison System is another high pressure water source to the RPV, however, there are no means of automatic system activation. System injection requires an operator to manually reposition a key-locked control switch to start the system from the Control Room. The system may also be operated from a remote local test station.. The only injection path to the RPV is through two explosive actuated injection valves that File: RRISI1.WPD ISI 1-3 of ISI 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 are interlocked with the key-locked switch in the Control Room. Local testing of the pumps uses demineralized water from a test tank and is a closed test loop. The injection rate for each pump is approximately 30 gpm, which would give the operator sufficient time to control reactor pressure.

Procedural controls are in place to respond to an unexplained rise in reactor pressure which could result from a spurious activation of an injection system. Actions specified include determination and isolation of the pressure source, verification of reactor head vents and/or MSIVs open and, as necessary, relieving reactor pressure using available plant equipment (e.g., electromagnetic relief valves, reactor water cleanup and reactor bottom drain).

During normal cold shutdown conditions, reactor water level, temperature are maintained within established ranges in accordance with operating procedures. The Operations manual governing Control Room activities requires that the Control Room operators frequently monitor for indications and alarms to detect problems and abnormalities as early as possible. An Operations procedure also requires that the control room supervisor be notified immediately of any change or abnormality in plant indications and controls.

Furthermore, reactor water level and temperature operating bands and changes thereto are established under the direction of the Station Shift Supervisor. Therefore, any deviations in reactor -water level or temperature from a specified band will be identified and corrected. Finally, plant conditions and on-going activities are discussed during each shift turnover. This ensures that on-coming operators are cognizant of activities that could adversely affect reactor level, pressure, or temperature.

Plant specific procedures have been developed to provide operator guidance regarding compliance with the plant Technical Specifications and RPV pressure-temperature curve limits. Additionally, operators receive training on RPV brittle fracture and the relation of these pressure-temperature curve limits.

During plant outages, NMP1 work control processes ensure that the outage schedule and changes to the schedule receive a thorough shutdown risk assessment review to ensure defense-in-depth is maintained.

Work is coordinated through the Work Control Center which provides an additional level of Operations oversight. In the Control Room, the Station Shift Supervisor is required, by procedure, to maintain cognizance of any activity that could potentially affect reactor safety during refueling outages. Expected plant responses and contingency actions to address unexpected conditions, that may be encountered, are required to be evaluated as stated in the administrative controls for risk management and management of outages.

As discussed above, NMPC has implemented procedural controls and training to minimize the probality of an LTOP event. Accordingly, the above information and the supporting technical documentation contained in the BWRVIP-05 report and USNRC Safety Evaluation provide a basis for excluding RPV circumferential welds from the augmented examination requirements of 10CFR50.55a(g) and ASME Section XI.

2. WELD ACCESSIBILITY NMP1, a BWR/2, has an RPV that was designed and fabricated to the rules of ASME Sections I and Vill, including Nuclear Code Case 1270N and 1273M. Additionally, General Electric's Specification for design and fabrication included additional requirements for materials and inspection that were similar to ASME Section III. Early vintage plants of this type were designed, fabricated and erected prior to the examination and inspection requirements of ASME Section XI. Specific ultrasonic (UT) examination criteria was not required by ASME I, III, or Vill for preservice inspection of the vessel and not factored into the plant design, hence external access to the RPV axial shell welds is constrained due to inadequate clearances between the bioshield wall and vessel insulation.

The NMP1 examination plan requires examination of 100% of all accessible regions of the RPV axial welds.

The ability to inspect 100% of the axial welds will be limited, in some cases, due to physical constraints of the RPV internal vessel design and arrangement of internal components. An internal vessel accessibility study of the RPV was performed by General Electric to determine the inspectability of the RPV axial shell welds and to obtain clearance measurements for the GERIS-2000. Several internal vessel components will File: RRISI1.WPD IS I 1-4 of IS I 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 limit a 100% ID UT examination including interference from the Feedwater Sparger, Specimen Brackets, Vibration Brackets, the Shroud Support Baffle Plate, and Shroud Repair Tie Rod Assembly. Even with these limitations, the overall projected percentage of effective weld examination coverage in the beltline region is approximately 92%. Tables 2 and 3 provide an illustration of the anticipated examination coverage of the axial welds. Included in Table 3 is a column identifying the specific limitation precluding essentially 100%

of the axial shell welds. A drawing also provides the location of welds in relation to the reactor pressure vessel.

NMPC concluded that permanent deferral of the examination of the RPV circumferential shell welds for the life of the operating license and the reduced examination coverage of the axial welds is justified and presents an acceptable level of quality and safety to satisfy the requirements of 10CFR50.55a(a)(3)(i),

10CFR50.55a(a)(3)(ii) and 10CFR50.55a(g)(6)(ii)(A)(5).

The proposed examinations are an alternative to the augmented examinations required for RPV shell welds specified in 10CFR50.55a(g)(6)(ii)(A)(2), and an alternative to the inservice inspection requirements of the ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition (Table IWB-2500-1, Examination Category B-A, Item No. B1.10).

The proposed alternative for examination of the RPV shell welds includes performing an examination, from the internal ID surface, of only the RPV axial shell welds (Item No. B1.12) and approximately 2-3 percent of the intersecting circumferential RPV shell weld (Item No. B1.11) to the maximum extent possible. The examination of the remaining accessible portions of the RPV circumferential shell welds will be permanently deferred for the life of the original unit license.

NMPC also proposes, as part of this request for relief, to perform an automated inspection of the RPV axial shell welds using personnel and procedures qualified to the Performance Demonstration Initiative, (PDI). The examinations will be performed using the General Electric Remote Inspection System (GERIS-2000). The GERIS

-2000 system and procedures were demonstrated and qualified to the satisfaction of PDI and in accordance with ASME Section XI, 1992 Edition with the 1993 Addenda, Appendix Vill.

The alternative program identified in Attachment 1 provides assurance of structural integrity and, therefore, an acceptable level of quality and safety is assured.

F. IMPLEMENTATIONSCHEDULE Pursuant to 10 CFR 50.55a(a)(3)(i) the request for relief was authorized and is effective from the date of the USNRC Safety Evaluation (4/7/99) until the expiration of the operating license (8/22/09), under Tac number MA4383, dated April 7, 1999 File: RRISI1.WPD ISI 1-5 of ISI 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 TABLE:2.':,".', ":; ',;;:,~r~'.'",,:,:...

. "PROJECTED,EXAMINATlON:COVERAGE OF, RPV.BELTt;INE REGION'AXIAL'WEI.'DS:,;-":;;:,',:;;,:;

','Weld

)

Number ID, Weld'Lengthfn'Beftllrie',.", ProJected ID'-.Examfnatlon Length /o,Of Weld Length In Reglon:,",--:,".'" ,ln Beltllne Region (fn) "Beltllne to be Ex)amfned

'"'"""'ln)';;;;;,;;,.

'VWD-139 128 128 100%

RVWD-1 40 128 128 100o/

RVWD-141 128 91 71'/

RVWD-142 40 40 100%

RVWD-143 40 40 100o/

RVWD-144 40 40 100%

TOTAL 504 467 92 6o/o TABLE,3, EXAWIINATION;COVERAGEOF ALL REACTOR;VESSEL AXIALWEL'DS.: 'ROJECTED Weld Number ID., ', 'Tote f Weld:: ',"-""';. ':;, V>Pr'oJected,lD": ';,";; lo Of Total Weld" Ca'us'e.'of, L'imitation 2

",;L'ength:, ';,';;Examlnatlon Total; ""<~Length to.be ".'>-",:.-,.'.

(See'Notes)

'-,'"::","'(in)'32

-;::: ',""""':!=',Examined

.'::.,:.';"Length:;";:;.':,-;:(In)

F:.

RVWD-130 132 100%

RVWD-1 3'f 132 132 100'/

RVWD-132 132 132 100o/

RVWD-133 133 76 57% FWS RVWD-134 133 76 FWS,VB,SB RVWD-135 133 80 60% FWS RVWD-139 132 132 100 RVWD-140 132 132 100 RVWD-141 132 91 68% SRTRA RVWD-142 133 101 76% SSBP RVWD-143 133 101 76% SSBP RVWD-1 44 81* 45 63% SSBP TOTAL 1538 1230 79 9%

NOTES: FWS - Feedwater Sparger, VB - Vibration Bracket, SB - Specimen Bracket, SRTRA - Shroud Repair Tie Rod Assembly, SSBP - Shroud Support Baffle Plate Weld RVWD-144 is rcduccd due to the intersection with a Recirculation Nozzle File: RRISI1.WPD ISI 1-6 of ISI 1-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 ATTACHMENT'I PROPOSED EXAMINATIONALTERNATIVE NMPC intends to inspect the RPV axial shell welds using personnel and procedures qualified to the Performance Demonstration Initiative, (PDI). The automated ultrasonic examinations will be performed using the General Electric Remote Inspection System (GERIS-2000). The GERIS -2000 system and procedures were demonstrated and qualified to the ASME Section XI, 1992 Edition with the 1993 Addenda, Appendix Vill.

The examination procedure uses echo-dynamic motion and tip diffraction characteristics of flaws for detection and sizing in lieu of ASME Code amplitude based techniques. All accessible weld examination volumes will be interrogated by the same straight and angle beam search units required by ASME Section V, Article 4 and an additional 70 degree refracted longitudinal search unit will be employed to ensure adequate investigation of the RPV axial weld clad base metal interface A comparison between the ASME Section V, Article 4 ultrasonic methods and procedures developed to satisfy the PDI can best be described as a comparison between a prescriptive, compliance procedure (ASME Section V) and a demonstrative results driven procedure (PDI). A typical ASME Section V procedure derives examination sensitivity to detect and size flaws based on the amplitude of a known reflector in a calibration standard. This method provides a means for standardization during examination and was easily specified in applicable documents controlling the process. This ultrasonic method has however, since been recognized as potentially providing inaccurate results for the application.

The PDI process, rather than just specifying the means by which an examination will be performed, specifies the results of the examination. Simply stated, a group of inspection specimens containing actual cracks and imbedded flaws are provided for demonstration of a procedure. These flaws are atypical of those that may be encountered in an insitu inspection of RPV shell welds and are characteristic to the flaw acceptance criteria contained in ASME Section XI, thereby validating the examination through performance. Compliance procedures would have great difficultyeven detecting these type of flaws during a PDI process since the reflective amplitudes are very low or even discernible. The difference being the PDI process requires detection and measurement of tip diffracted signals whereas the compliance process relies on larger specular reflectors.

USNRC Regulatory Guide 1.150, AUltrasonic Testing of Reactor Vessel wclds During Preservice and Inservice Examinatione was issued by thc Staff in 1981 as a means to initiate a change to ultrasonic procedures to be results based versus compliant. The regulatory guide contains concepts for flaw detection and sizing but falls short in providing the means by which to perform thc demonstrations. As such, this allows for different interpretations of thc intended requirements. Based solely on cost, it would be prohibitive for an individual licensee to demonstrate ultrasonic examination procedures as has been done with the PDI process.

The regulatory guide does not provide for thc number of flawcd spccimcns, blind tests, or mandate an cxpccted level of performance as does PDI. The regulatory guide only requires an estimate of expcctcd capability The use of PDI qualified personnel and procedures results in a more scnsitivc examination and will provide added assurance for fiaw dctcction and sizing, meeting or exceeding thc current requircmcnts of the 1989 Edition of the ASME Section XI Code and USNRC Regulatory Guide 1.150. The error band for flaw sizing has been established within the limits of ASME Section XI, Appendix VIII..

The above information and supporting technical documentation contained in the BWRVIP-05 rcport and NRC Safety Evaluation provides a basis for excluding the performance of RPV circumferential welds from the augmented examination requircmcnts of IOCFR50.55a(g)(ii)(A)(2) and ASME Section XI and that the RPV axial welds being volumetrically examined using PDI and qualiTied personnel and procedures. Niagara Mohawk firmly belicvcs permanent deferral for examination of the RPV circumferential shell welds for the life of the original operating license is justified and presents an acceptable level of quality and safety to satisfy the requirements of IOCFR50.55a(a)(3)(i), IOCFR50.55a(a)(3)(ii) and 10CFR50.55a(g)(6)(ii)(A)(5).

File: RRISI1.WPD IS I 1-7 of IS I 1-8

tl 0$ 00 Vtfill. ILAICC N 0 I)S 0 ~ V CL OSLO RVVD 000

~ Oo Oo Jg OIO' tL SIOO'L, Q'IO CuNID O

~ IS SO I Z

~SZ Z I- ~

0'l ITI

0) C'IOIILLaLOOI -.>>'.LIOLO ~

~

ITI 0$ 1000'L IIO'O IPCRLID

~ O-m ITI I 0 RVVD IOI SI SII0'l ITI Z ITI CI 1$ 1 IuP IO ACIIVI I ill 0C~O 0)

V)

CO C

C V

C

~8 0'OP SOIL'l ITI Ill Z NOH

+ +C 9 ltllllltOILIER S 0 Z RVVR.ISI, ~ IOS ISO'o 10 1 001 RSOIP CL 200 Z

SIIL'l

~ OIIISI IO ACIIVC lull ITI

~(

Qo Qo ISO ~0 luPPINI vll0 0'l I

IOL RVVO IN IL 00'CSICI neo IICVLIIue

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 ENT DE System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Pressure Retaining Welds in Reactor Vessels ME E T RE IRE E Section XI, Table IWB-2500-1, Examination Category B-A requires:

C'ode'Item'No." Component ID;:, Examination Description B1.21 Circumferential Head Welds Accessible length of all welds B1.22 Meridional Head Welds Accessible length of all welds B1.30 Shell to Flange Weld 100% of weld length B1.40 Head to Flange Weld 100% of weld length C. ELIEF RE E TED Relief is requested from performing 100% volumetric examination of the Code Required Volume (CRV) for those components identified, on Table 1 attached.

D. BA ELE NMP1, a BWR/2, has a Reactor Pressure Vessel that was designed and fabricated to the rules of ASME Sections I and Vill, including Nuclear Code Case 1270N and 1273N. Early vintage plants of this type were designed, fabricated and erected prior to examination requirements of ASME Section XI. Specific ultrasonic (UT) examination criteria was not required by ASME I, III, or Vill for preservice inspection of the vessel and was not factored into

. plant design.

The NMP1 Reactor Pressure Vessel design precludes essentially 100% examination of the weld lengths due to the following:

Closure Head The Closure Head Dollar Plate Weld RV-WD-002, limits essentially 100% examination of weld length due to the physical location of six (6) nozzles and the close proximity of a steel platform. See attached sketch.

The Closure Head Meridional Welds (8 each) RV-WD-003, 004, 005, 006, 007, 008, 009 and 010, limits essentially 100% examination of the weld length due to the physical location of eighteen (18) nozzles and insulation lugs.

File RRISI2.WPD ISI 2-1 of ISI 2-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 The Closure Head to Flange Weld RV-WD-001, limits essentially 100% examination of the weld length due to the configuration of the weld being a one sided exam. See attached sketch Bottom Head All Bottom Head circumferential welds two (2)RV-WD-147,RV-WD-162 and Bottom Head Meridional Welds fourteen (14) RV-WD-148, 149, 150, 151, 152, 153, 154, 155, 156, 157, 158, 159, 160 and 161 are inaccessible due to one hundred twenty-nine (129) Control Rod Drive nozzles and sixty-four (64) In-core Flux Nozzles.

Reactor Vessel Shell to Flange Weld The Reactor Vessel Shell to Flange Weld RV-WD-099 limits 100% examination of the weld length due to Guide Rods located at the 0 and 180 degree position and the Main Steam Nozzle Plug hoses.

In addition to the above external access to the reactor pressure vessel bottom head welds is constrained due to inadequate clearances between the bio-shield wall and vessel insulation.

RV-WD-099(A), (B), (C) and (D), Reactor Pressure Vessel Shell to Flange Weld from the flange side, was divided into four (4), equal 90 degree segments during the First Inservice Inspection Interval and the remainder of the weld was examined from the vessel inside surface at the end of the interval. During the preparation of the Second Inspection Interval the same division process was included in the inspection plan in order to stay consistent with the First Interval. NMP1 performed the shell to flange weld in the same sequence as conducted in the first interval with the exception of segment D, which was performed from the shell side. With the completion of refueling outage 15, weld RV-WD-099 will have been examined to the extent practical.

Compliance with the ASME Section XI examination requirements would require a redesign of the Reactor Pressure vessel, which would provide an undue hardship on NMPC without a compensating increase in the quality and safety of the unit.

E. L E VE XA A No alternate examinations of the Closure Head Dollar Plate Weld RV-WD-002. Examine to the extent practical.

No alternate examinations of the Closure Head Meridional Welds (8 each) RV-WD-003, 004, 005, 006, 007, 008, 009 and 010. Examine to the extent practical.

No alternate examinations of the Closure Head to Flange Weld RV-WD-001. Examine to the extent practical 1/3 each period.

No alternate examinations of the Bottom Head circumferential welds two (2)RV-WD-147,RV-WD-1 62 and Bottom Head Meridional Welds fourteen (14) RV-WD-148, 149, 150, 151, 152, 153, 154, 155, 156, 157, 158, 159, 160 and 161, as they are inaccessible.

File RRISI2.WPD ISI 2-2 of ISI 2-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 No alternate examinations of the Reactor Vessel Shell to Flange Weld RV-WD-099. Examine to the extent practical.

The extent of examination performed on the Reactor Pressure Vessel will assure an acceptable level of quality and safety.

F. I LE E ED LE Third Inservice Inspection Interval File RRISI2.WPD ISI 2-3 of ISI 2-8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 omponent ID Component Estimated . Description of Limitation Description o/o of CRV Achieved RV-WD-001 Head to Flange Weld 67% Obstructed by twelve nozzles, eighteen insulation lugs, three lifting lugs and the configuration of head to flange weld.

RV-WD-002 Closure Head Dollar 68% Obstructed by six (6) nozzles and the close Plate Circ. Weld proximity Steel Platform.

RV-WD-004 CH Merd. Weld 70'/ Obstructed by Nozzle N7C and N7P RV-WD-010 CH Merd. Weld 80o/ Obstructed by Nozzle N7M, N7N RV-WD-099 Shell to Flange Weld 83 3% Obstructed by Guide Rod, MS Nozzle Plug hoses RV-WD-147 BH Dollar Plate Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-148 BH Merd. Weld 00/ Inaccessible due to CRD Nozzles and In Core Flux Monitors V-WD-149 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-150 BH Merd. Weld 0 Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-151 BH Merd. Weld 0 Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-152 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-153 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-154 BH Merd. Weld 00/ Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-155 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-156 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors V-WD-157 BH Merd. Weld 0 Inaccessible due to CRD Nozzles and In Core Flux Monitors File RRISI2.WPD ISI 2-4 of ISI 2-8

0 NINE MlLE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 ornponent ID 'Component Estimated DesdriptIon of LimitatIon

'Description %of CRY Achieved

RV-WD-158 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and ln Core Flux Monitors RV-WD-159 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-159 BH Merd. Weld 00/ Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-160 BH Merd. Weld 00/ Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-161 BH Merd. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors RV-WD-162 BH Circ. Weld 0% Inaccessible due to CRD Nozzles and In Core Flux Monitors File RRISI2.WPD ISI 2-5 of ISI 2-8

0 NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 T 81 < T1

' Rl Q4 i ~ 1 42 4~

41 42 RT

$8 4

ij) 0/

6 6' Q1 44 42 Ul e+e ee 9'e e e.ee,~

21 'M+O e S .e e.e 77 '8089 4'e888P 72 ee'ee .~ceo.e 2~

21 n

24 21 22

)4 Td U

6 86888+8e 9e e,-e 6 eSe 4 888 S2 9C'3 17 14 +0 e p e 8 0,8 84 U3 00 11 Rd i4884 ~a 4+4'86$

04 07 4 9~6 $ .8 +4 i 01

4. ON I ~

I I

A I sr 0

I Ud I,

00:04 Td 14 ia aa I

Oa '12 10)2 24}aa)S sa 02 Oa 10 14 14 22 2 ~ 20 24.2442 ia 40 5

2,

~

r

~ ~

1 CVRLR!

r I

~ ~

\ J

'I ~,

File RRISI2.WPD lsl 2 - 6 OF IS I 2- 8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 00 A

1 44 File RRISI2.WPD IS I 2 - 7 OF ISI 2- 8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-2 0'

t, ~~~~P <~

~I

~M~ A (NlA)

(N 7 H)~ ~ g &8(N7 0)

K p (N1U)

( H'F K ) '. g are'

( '

WE (N14

("'4) (~VS) 00'ile RRISI2.WPD ISI 2 - 8 OF IS I 2- 8

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-3 COMPONEN DE F CAT N System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Full Penetration Welds of Nozzles in Vessels - Inspection Program 8 B. ESE ON E E Section XI, Table IWB-2500-1, Examination Category 8-D requires:

,Code Item: No...':. !Co'mponent ID::::.:':, ' Examin'ation'Desciiption.,",",',

83.90 Nozzle to Vessel Welds AII Nozzles, Exam Volume per IWB-2500-7 83.100 Nozzle Inside Radius Section All Nozles, Exam volume per IWB-2500-7 EL EF E UE D Relief is requested from performing 100% volumetric examination of the Code Required Volume (CRV) for those components identified in Table 1 attached.

D EL NMP1, a BWR/2, has an Reactor Pressure Vessel that was designed and fabricated to the rules of ASME Sections I and Vill, including Nuclear Code Case 1270N and 1273N. Early vintage plants of this type were designed, fabricated and erected prior to examination requirements of ASME Section XI. Specific ultrasonic (UT) examination criteria was not required by ASME I, III, or Vill for preservice inspection of the vessel and was not factored into plant design.

The NMP1 Reactor Pressure Vessel design of the nozzles precludes essentially 100% examination of the Code required volume due to the following conditions:

1. Nozzle locations (close proximity to each other), doesn't allow enough scan distance between nozzles to interrogate the entire Code Required volume.

Nozzle configurations and shell tapers do not provide parallel surface, therefore providing areas of non scanning.

lifting lugs and insulation lugs limit the scan distances required to interrogate portions of Code volume.

limited access for examination personnel between the reactor pressure vessel and the biological shield limits the maximum search unit scanning distance for each nozzle.

Compliance with the ASME Section XI examination requirements would require a redesign of the Reactor Pressure vessel, which would provide an undue hardship on NMPC without a compensating increase in the quality and safety of the unit.

File RRISI3.WPD ISI 3-1 of ISI 3-7

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-3 LE N VEE No alternate examinations proposed. Examine each nozzle to vessel weld and inner radius section to the extent practical.

The extent of examination performed on the Reactor Pressure Vessel Nozzles will assure an acceptable level of quality and safety.

L E Third Inservice Inspection Interval File RRISI3.WPD ISI 3-2 of ISI 3-7

NINE MlLE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-3 omponent ID Component Description Estimated Description of Limitation

'/o of CRV Achieved 01-WD-001 N3A Nozzle to Vessel 87% Adjacent nozzle, Bio-shield wall 01-WD-001-IR N3A Noz. Inner Radius 54 Adjacent nozzle, Bio-shield wall 01-WD-033 N3B Nozzle to Vessel 91% Manual Adjacent Nozzle, Bio-shield wall Geris 80.1%

01-WD-033-IR N3B Noz. Inner Radius 790/ Adjacent nozzle, Bio-shield wall 31-WD-021 N4B Nozzle to Vessel 78% Adjacent nozzle, Bio-shield wall 31-WD-021-IR N4B Noz. Inner Radius 47 Adjacent Nozzle, Bio-shield wall 31-WD-030 N4A Nozzle-Vessel 42.1% Adjacent Nozzle, Bio-shield wall.

31-WD-030-IR N4A Noz. Inner Radius 82% Adjacent Nozzle, Bio-shield wall.

31-WD-051 N4C Nozzle to Vessel 84 Adjacent Nozzle, Bio-shield wall.

31-WD-051-IR N4C Noz. Inner Radius 86% Adjacent Nozzle, Bio-shield wall.

31-W D-060 N4D Nozzle to Vessel 85% Bio-shield wall.

1-WD-060-IR N4D Noz. Inner Radius 59% Bio-shield wall.

32-WD-001 N1A Nozzle-Vessel 33.8% Lug, Adjacent Nozzle, Bio-shield Wall 32-WD-001-IR N1A Noz. Inner Radius 73% Adjacent Nozzle, Bio-shield wall.

32-WD-043 N2A Nozzle to Vessel 75% Bottom Head Taper of Shell Thickness 32-WD-043-IR N2A Noz. Inner Radius 82% Bottom Head Taper of Shell Thickness 32-WD-044 N1B Nozzle to Vessel 67 9'/ Bio-shield wall.

32-WD-044-IR N1B Noz. Inner Radius 73% Bio-shield wall.

32-WD-083 N2B Nozzle to Vessel 75.4% Bottom Head Taper of Shell Thickness 32-WD-083-IR N2BNoz. Inner Radius 82% Bottom Head Taper of Shell Thickness 32-WD-084 N1C Nozzle to Vessel 67.8% Bottom Head Taper of Shell Thickness 32-WD-084-IR N1C Noz. Inner Radius 73% Lug, Adjacent Nozzle, Bio-shield Wall 32-WD-123 N2C Nozzle to Vessel 75 4% Bottom Head Taper of Shell Thickness 32-WD-123-IR N2C Noz. Inner Radius 82% Bottom Head Taper of Shell Thickness 32-WD-124 N1D Nozzle to Vessel 68% Bio-shield Wall 32-WD-124-IR N1D Noz. Inner Radius 73% Bio-shield Wall

-WD-165 N2D Nozzle to Vessel 75 4% Bottom Head Taper of Shell Thickness 32-W D-1 65-IR N2D Noz. Inner Radius 82% Bottom Head Taper of Shell Thickness File RRISI3.WPD ISI 3-3 of ISI 3-7

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-3 omponent ID Component Description Estimated Description of Limitation ofo of CRV Achieved 32-WD-166 N1E Nozzle to Vessel 33.8% Lug, Thermocouple, Bio-shield wall 32-WD-166-IR N1E Noz. Inner Radius 73% Lug, Adjacent Nozzle, Bio-shield Wall 32-WD-209 N2E Nozzle to Vessel 74 5% Bottom Head Taper of Shell Thickness 32-WD-209-IR N2E Noz. Inner Radius 82% Bottom Head Taper of Shell Thickness 39-WD-001 N5A Nozzle to Vessel 65% Adjacent Nozzle, Bio-shield Wall 39-WD-001-IR N5A Noz. Inner Radius 55 Lug, Adjacent Nozzle, Bio-shield Wall 39-WD-089 N5B Nozzle to Vessel 52.5% Adjacent Nozzle, Bio-shield Wall 39-WD-089-IR N5B Noz. Inner Radius 79 3% Adjacent Nozzle, Bio-shield Wall 40-WD-040 N6A Nozzle to Vessel 65.8% Adjacent Nozzle, Bio-shield Wall 40-WD-040-IR N6A Noz. Inner Radius 89 5% Adjacent Nozzle, Bio-shield Wall 40-WD-081 N6B Nozzle to Vessel 57.3% Adjacent Nozzle, Bio-shield Wall 40-WD-081-IR N6B Noz. Inner Radius 78.1 Adjacent Nozzle, Bio-shield Wall 4.1-W D-018 N9 Nozzle to Ve I 46.2/0 io-s ie wa 44.1-WD-018-IR N9 Noz. Inner Radius 48% Bio-shield Wall File RRISI3.WPD ISI 3-4 of ISI 3-7

MIERIALTSAI 05 GTL II HATER I ALT SA336 NOZZLf fxAH IO N4A 31 10030 31 TETOZS N40 3l IETOZ I 15 I" HiC 31 IETOZ0 31IET051 31161050 X

VI7 Nio 31 IET060 3116055 Zl O

Z HOTf5e II THIS IXTANING IDENTIFIES COH%tCNIS SLIJECT ~SZ ill IO INSERTICE INSPECIITxl ExAIRHATIlxiS AS REOVIREO 01 ASIA SECTION xl AIET GEFTHEO IH ~ +ill Z THE IAPC I NSERrl CE INSPECTION PROGRAH PLAN. p ~

CO E43 ZI THIS ORAIRHG IS IHIEIOEO fOR USE IH CONJIPTCIIOH Nl IH 'INE IHSERTICE INSPfCITGH m~g EZI pRGGRAH pLAH TFAT.II IS INIEIETEO To PRGTIOE OTAT A GEHEITAL REPRESENTAII OH OF IHE fll I 0 31 IEXT30 IOENIIFlfO CGPPONENIIEGUIPFEHT lUCl IS NOT ITl 3l le0ZT 10 Of USEO FOR ANT OTHER PMPOSE.

31 141051

'll TET060 M

0) 14 ~ ' SECIION xl CLASS I. ()

RfFI Ill Ill Z 6e'REFI 3I IETOZS 31 IETRZ0 31 IET050 31 TC055 REFEREHCEST MO+ l COP ORAT@NO CE Z31 ~ 566 Zg \

SECTION xl OGLAOaRr OlaQQW F'63006 C, STLI TELO HAP IFTANING F45103.C SIEEI 5 AID Z0 cn<K2q I REF I m

NOTE> IRIS GRAVING VAS CREAIEO USING COHPUIE R AIRE O GRAF I INC OO HOI HARE rixiuaL CHANGES

(

I IUlÃN HIAUARA NONANC PONXR LXTRPGR*TIOH FEEOWATER NOZZLE ASSEI%LY NINE HILf POINT MTCLEAR STATION IPRT I CIXO~HT Of IAIL GRAVING FfEOvAIER NOZZLE ASSEs0Lr INSEIITI C I HSPECI I ON PROGRAH

~ ~ IO H' I IOC

~ ISI WZ 004

~ I s~

+

3 SSSIERI ALs 314eS NOZZLE fYAH IO 32sO881 32se882 Hl 8 32K%I ~ 32leBIS 32se88< 32se885

~ ~ 32seI 24 32Y0125 32sel 66 32IOI 61 x BVI IEREO INCDCL skIIE Ss U

HAIERI ALS I ~ 182 Il IRIS IOENIIFlfS COHSICNIS R

+ MZ ORASslNG SSAIJECI fII IO INSERYICE INStECIION YAs8HAIIONS AS 32SES88I 32ie8s Y REOVIRfO BY ASSf SECIIOH GESTURAL ASE) OEFIHEO IN 1HE INC INSERYICE INSPECIIIRI PROGRAH PLAIL ill 2 32sO88S 32sEN 3 sOI66 2i j

/ 21 IRIS ORASslNG IS IHIftOEO FOR USE IN I

~ lll 0)

CD IRfF>

12s' REF I s s COQUHCIIOH PROGR4H OH.Y A SSI YH YHE ISSSERYI Cf INSPECII OH tLAH OSA.Y.I I IS INIflOEO 10 PROYIOE REPRESEHIAIIOH OF IHE IOENIIFIEO C~NISEOVI PSf NI ASO IS M)I

~nK Ill I IO BE USEO FIFI ANr Orson R PURPOSE. 37 E31 O 3S SECIIOH 11 CLASS I. ITI Z 32YO882 C) m

/ 32sO8 s5 C~O ITI lll 2 32sO885

0) / 32sOI 25 f REF RENCESs 1 32sOI 61 V)O+

CD

~

COH'R4lslHG SELO HSP ORASSI CE 231 SSG

~

SECIIOH Xl BOVSNARY OIAGRASA F45183 56$

C SHEE F

I 63886'C.

2 ANO 28 SH. I

++C "sSSF sssss S OZ ~

s 4l ~~

X HOIEs IRIS ORAsslHG M4S CREAIEO USING COHPUIER AIRE 0 GRAF IISsG ITI OO HOI HAkE SSASASSN EHANGf 5 C

I RECIRCUL4TION OUTLET NOZZLE 4SSEIRLY ISIAOARA HOHAWS tOSSSR LYSRPORAYIOH RISE HILE POIHI SASCLEAR SIAIIOH still I C~HI Of IAIL ORASSING

RECIRCSAAIIOH OVILEI NOE?Lf INSERYI C INSPECI I ON PROGRAH

~ ~ sk %s sskk

~ I SI.HOI ~ 881

37 2

Ct)

O HATEAIALt SAIOS GIL II HATERI ALT SA336 NOllLf EXAH IO

~

~ I ICOSI ~ I IDAA2

~

IS'2 Al IDA)3 ~ I ICO) I Ul7 z Zl NOT ESI O II

-Jl-THIS ORAVIIC IOENllflfS CDPOtCHTS SIXI)ECI f

10 IHSERTICE INStECIIOH XWRNATIOHS AS REOUI NEO dr ASIA SECTION Xl WD OFF THEO IN THE tPPC I NSEATI CE INStECTION tROGRAH PLWL gSZ z

0) 21 THIS IXIAVING IS INTEICXO FTXI USE IN mmz COTLIUtCIION VI TH 'IHE INSEATICE INSPECIIOI I >+m PROGRAH tL4H OtC,T I I IS INTEIDEO IO PROTIOE Q

Ct)

C)

Rl TDAOI

~ 2TDO)3 tt t

t OK.T A CEIERAL REPAESENTAIIIXI OF IHE IOENIIFIEO COttOtAHT/EOUIttfNI WD IS 10 SE IISEO fIXI WIT OIHEII PURPOSE. NOT ~O-m Ill I 21 ~

)t I. zl m AEFI SECTION Xl CLASS ITI Z~

0)

Et)

I Rffl I

~ I l0002 REFEREHCESt COTE ORAVltC CE 231 S66 Cl M C 'll-Ill m p

Z

~ 2ICO)I SECIIOH Xl OOUtOART OIAGWW F ~ 63666.C, TELO tftt ORAIrlHG F4SIA) C Stf El 2 AtD 26 SIL I 0)OH I

M oz z

NOIEt IRIS ORAVIIC HAS CAE4IEO USING Ill COHPUIEA 41OEO GRAF I ING OO HOT HAXE HAHUAL CHANGES I

I Affl IMIRN NIAGARA HOHANS PORXR CORPORATION STEAH OUTLET - TUBINE NOZZLE ASSEMBLY NINE Hllf POINT M%LEAR S141IOH IDIOT I COttOtA'HI OETAIL ORAHIIC SIE4H OUTLET TIXISINE tDllLE ASST INSERT I C INSPECTIPI PAOGA4H sa Ja a +wtititt s c NOtO

~ Altl W He I tAT

<< ISI.NOI OA)

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISIS D

System: Reactor Pressure Vessel Class: Quality Group A, (ASME Code Class 1)

Component

Description:

Pressure Retaining Welds In Reactor Vessel Section XI, Table IWB-2500-1, Examination Category B-A requires:

. Cocfe,'Item w,'arts",Examined.-'=',N~';~~~,";;,-..'F-.;--:,.; ,'.Ex'am,'Re jilrements",-; >Extent"artd:Frequeiicy;::.

B1.12 Reactor Vessel Longitudinal IWB-2500-2 Includes essentially 100% of Shell Weld weld length B1.30 Reactor Vessel Shell to Flange IW8-2500-4 Includes essentially 100% of Weld weld length Pursuant to IWB-2420(b) If flaw indications or relevant conditions are evaluated in accordance with IWB-3132.4, and the component qualifies as acceptable for continued service, the area containing such flaw indications shall be reexamined during the next three inspection periods listed in the schedules of the inspection programs of IWB-2410.

Relief is requested from paragraph IWB-2420(b), from performing reexaminations of the flaw indications during the next three inspection periods.

D.

During the automated ultrasonic examinations of the Reactor Pressure Vessel shell welds, several sub-surface indications were observed that exceeded the acceptance criteria of IWB-3000 on welds RV-WD-099 and RV-WD-140; Weld RV-WD-099 identified six (6) unacceptable flaws, located in the region of the weld fusion lines and were attributed to lack of fusion and thin film slag deposits left from the fabrication process.

Weld RV-WD-140 identified two (2) unacceptable flaws, located in the region of the weld fusion lines and were attributed to lack of fusion and thin film slag deposits left from the fabrication process.

Review of the construction radiographs (RT) provided a correlation with the ultrasonic indications.

An analytical evaluation was performed in accordance with IWB-3600 and the welds were found to be acceptable for continued service. These evaluations took into consideration flaw growth that is unlikely to occur with fabrication related flaws. The flaws were found to be acceptable for continued service until the intended end of plant life.

ASME Code Case N-526, Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels,Section XI, Division 1, (attached), provides an alternate to the reexamination required by IWB-2420(b).

File RRISI4.WPD ISI 4-1 of ISI4-4

NINE MILE POINT UNIT 1,,

THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-4 Compliance with the ASME Section XI reexamination requirements would provide an undue hardship on NMPC, without a compensating increase in the quality and safety of the unit.

No alternate examinations proposed. Reexamine the flaws along with the Code required examinations for welds RV-WD-140 and RV-WD-099 as currently scheduled in the ISI Program.

The extent of examination performed on the Reactor Pressure Vessel will assure an acceptable level of quality and safety.

Third Inservice Inspection Interval File RRISI4.WPD ISI 4-2 of ISI 4-4

33 CO D

O S U<<%1 ~s s ITS' s I<<'10'<<0'L 000 0 VCSSCL CLANGOR RVVD.ODD

,. Qo Qo tI0' Cl. SIT ~ IO SUICLD 0'L.

C SO<<0'l.

~ k

~

@ jfTTTTTTTT Ool 11 S<<IL'N'F ST<<0'L IN'F SHtllJO O 'I ~ S1I RVvD IOI 0'l SSS III IF ACIIVC I Utl DllILIK RIGIOU CL RVVD'101 8790'l IDS 4<<S ZI8 S/10'L oc AcIIVC Ilail

'.,Oo 0o Oo 8 Oo IS20'oIIIDI SHROUD SUPPIN I VII 0 Cl RVVD ISD

<<000 IDIO'l 00'CSSIL Ilao CICVAIILH

XI U

z

~V)2 mmz r- >+m m ~

~O

~mt m Figure 1 Flaw Location Plots (Typical) in Cross Section, RV-WD-099 0 mm2 MOW g)OZ II tlat

)g 4

2 M~ OIS / OI ~

m XI t

Figure 2 Flaw Location Plots in Cross Section, RV-WD-140

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-5 System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Integral Attachments For Class 1 Vessels, Piping, Pumps and Valves N-Code Case N-509, Table 2500-1, Examination Category B-K requires:

Code:Item,'; ;Parts",'Examtnedg:;:,'-',,'.,';.~'..: '~.', Exam Requirements; .Exterit and Frequ'ency No.,',; ".

B10.10 Reactor Vessel Integrally Welded IWB-2500-13, 14 100% of required areas of Attachments and 15 each welded attachment Note: Regulatory Guide 1.147, Revision 12, dated May 1999, is acceptable to the USNRC provided that 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 should be examined.

C. E TED Relief is requested from performing 100% of the length of the attachment weld at each attachment subject to examination.

D. B LIEF NMP1, a BWR/2, has an Reactor Pressure Vessel that was designed and fabricated to the rules of ASME Sections I and Vill, including Nuclear Code Case 1270N and 1273N. Early vintage plants of this type were designed, fabricated and erected prior to examination requirements of ASME Section XI.

Of the (6) integral attachments subject to examination, four (4) are the earthquake stabilizer brackets and two (2) are. the reactor pressure vessel support skirt integral attachment, which is broken down into inside surface and outside surface.

EARTHQUAKE STABILIZER BRACKET ATTACHMENTS The four (4) alloy steel Reactor Vessel earthquake stabilizer attachments brackets are located at 22.5, 112.5, 202.5 and 292.5 degree axis points around the outer circumference of the vessel approximately eleven (11) feet and eight (8) inches below the vessel flange. See attached drawing for locations. Access for examination purposes only allows a maximum of 50% of the attachment weld length to be examined on all four (4).

REACTOR VESSEL SUPPORT SKIRT The Reactor Vessel support skirt is divided within the examination plan as two (2) separate items, these being the inside surface of the attachment weld and the outside surface of the attachment weld. Access to the support skirt is limited to the outside surface of the attachment weld only. The inside surface is inaccessible.

File RRISI5.WPD ISI 5-1 OF ISIS-5

Qi NINE MILE POINT UNIT 1,,

THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-5 EXAMINATIONREQUIREMENTS The Code Case requires that a surface examination be performed in accordance with Figures IWB-2500-13 and IWB-2500-15.

The earthquake stabilizer bracket (Figure IWB-2500-15) attachments require the weld plus 0.50" on each side of the weld and essentially 100% of the weld length to be examined by the surface (Magnetic Particle or Liquid Penetrant Method).

The reactor vessel skirt weld (Figure IWB-2500-13) attachment requires the weld plus 0.50" on each side of the weld and essentially 100% of the weld length to be examined by the surface (Magnetic Particle or Liquid Penetrant Method).

The use of the ultrasonic examination method in lieu of the surface exam is not appropriate due to the access provision would be the same as that for the surface examination. In addition the ultrasonic examination of the outside surface of the vessel skirt from one side would be inappropriate due to the design and geometry of the skirt being non parallel surfaces on the forging knuckle. The additional areas achieved would be negligible.

The 10% sample requirements for the six (6) Code Item Number B10.10 integral attachments would require as a minimum one (1) integral attachment required to be examined over the interval.

Compliance with the ASME Code Case examination requirements would require a redesign of the Reactor Pressure vessel integral attachments, which would provide an undue hardship on NMPC without a compensating increase in the quality and safety of the unit.

NMPC proposes to perform the following examinations:

Schedule two of the four Earthquake stabilizer brackets for surface examination to the extent practical. The anticipated Code Required Area to be achieved will be 50% on each integral attachment, which would be equivalent of completing essentially one bracket.

In addition to the stabilizer attachment, NMPC proposes to perform to the extent practical a surface examination on the outside surface of the RPV Support skirt only.

The extent of examination performed on the Reactor Pressure Vessel Integral Attachment will assure an acceptable level of quality and safety.

Third Inservice Inspectlbn Interval File RRISI5.WPD ISI 5-2OF ISI 5-5

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-5

)

Component fde'ntiffca ;Comp'onent Descriptton :.Pe'rcent;(%);of. Selection

'RA'to. be, ."..'...;;

-achieved: ;..'"

RV-SB1-IA-371/372 Earth Quake Stabilizer Integral Attachment 50% Selected RV-SB2-IA-373/374 Earth Quake Stabilizer Integral Attachment 50% Selected RV-SB3-IA-375/376 Earth Quake Stabilizer Integral Attachment 50% Not Selected RV-SB4-IA-377/378 Earth Quake Stabilizer Integral Attachment 50% Not Selected RV-WD-356-ID Support Skirt Integral Attachment 50 Not Selected RV-WD-356-OD Support Skirt Integral Attachment 50% Selected File RRISI5.WPD IS I 5 - 3 OF IS I5-5

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST IS[-5 File RRISI5.WPD IS I 5 - 4 OF IS I 5 - 5

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-5

/p~

]

~Pg@

Wc~o

~SCAN 0 I I g.v Tlute D

/

64Ck' CQubKUdÃm>f~

18

/

I

/

~ 0 5

5

@lÃ~z

%>i W u File RRISIS.WPD ISI 5-5OF ISI 5-5

NINE MILE POINT UNIT I THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-6 System: Reactor Pressure Vessel Class: Quality Group A, ASME Code Class 1 Component

Description:

Pressure Retaining Welds In Control Rod Drive Housing Section XI, Table IWB-2500-1, Examination Category B-O requires:

Code Item Parts Examined Exam Requirements Extent and Frequency No, B14.10 Reactor Vessel Welds in CRD IW8-2500-18 10% of the Peripheral CRD Housing Housings required Pursuant to 10 CFR 50.55a(g)(6)(i) NMPC requests relief from performing 100% of the Code Required Volume of 10% of the peripheral CRD Housing welds length as defined in Figure IWB-2500-18.

BA I R E NMP1, a BWR/2, has an Reactor Pressure Vessel that was designed and fabricated to the rules of ASME Sections I and Vill, including Nuclear Code Case 1270N and 1273N. Early vintage plants of this type were designed, fabricated and erected prior to examination requirements of ASME Section XI.

There are one hundred twenty-nine (129) Control Rod Drive Housings located on the bottom head. Thirty-two (32) are peripheral CRD Housing for which 10% or 3 are required to be examined during the interval.

A sector of approximately 180 degrees of each CRD peripheral housing circumference is obstructed by the adjacent CRD housings and their hydraulic lines. See attached drawing.

NMPC proposes to perform surface examinations on six (6) of the peripheral control rod drive housing in lieu of the 3 required. The adrfltional 3 housing examinations will result in the same weld length being examined, thereby meeting the intent of the Code requirement.

This approach was preciously granted per USNRC Safety Evaluation, TAC No. M83099, dated April 6, 1994.

The extent of examination performed on the Control Rod Drive Housings will assure an acceptable level of quality and safety.

F. E T D File RRISI6.WPD ISI 6-1 of ISI6-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-6

'Comp'on'ent'; ';.=".:,'.:; ~~pporlent'",Desciiption',:";; :Pete'eiit;(%)-of.,';-'-.-'; '.Sefectiori: ';,'."-;",~;:.;;.;:;";.:;;-'.";,,,"'::.

.IdentificatIo'n::.'." ;:CRA'to:be",.'..'--',,'-,

j';=-'-.-'V-CRD-S1

'achieved-"" ,;;-",., i

'elected CRD Housing Weld 50%

RV-CRD-S3 CRD Housing Weld 50% Selected RV-CRD-R1 CRD Housing Weld 5p Selected RV-CRD-R5 CRD Housing Weld 5p Selected RV-CRD-T3 CRD Housing Weld 5p Selected RV-CRD-T7 CRD Housing Weld 50% Selected RV-CRD-U2 CRD Housing Weld 50% Not Selected RV-CRD- U6 CRD Housing Weld 50% Not Selected File RRISI6.WPD ISI 6-2 of ISI6-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-6 7 Sl + 71 Ad 4 i~

44 AT I 0 40d 88 08 "e 8P 14 47 A2 44 i1 e~e e e e.e e,~

2~

M~O e t4j SB 84O eeeee~ @cove 21 24 24 2t 77

~~'&4~

e e'e e.sp. ie e o.e T2 S2 9C 24 2l 22

'I

~

Td e 'eeeee+eeeee e e e 6 ~,e-e 8s e ~ T3 11 14 +o e o e e e,e ee e,o' V3 14 11 Ad

'e e e j,e e.e ~ ~e 04 04 OT + +~6 0. + i ~

01 4$ 0~ N 4 !>

> eii A

I Ud Td Qu Quan 00:01 OI I

is ~'ebs S1 telo L 1I es 02 0 ~ 10 s1 s ~ 2! 2 ~ 40 41 se14 1e IO I

l File RRISI6.WPD ISI 6-3 of ISI6-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-7 System: Reactor Containment Spray Class: Quality Group B, ASME Code Class 2 Component

Description:

Pressure Retaining Welds in Pumps E 0 MNA ON E T ASME Section XI, Table IWB-2500-1, Examination Category C-G, "Pressure Retaining Welds in Pumps, Examination Item Number C6.10, "Pump Casing Welds", requires a surface examination of 100% welds in all components in each piping run examined under Examination Category C-F.

In case of multiple pumps of similar design, function, and service in a system, the examination of only one pump among each group of multiple pumps is required.

The examination may be performed from either the inside or outside surface of the component.

Systems affected' '" --',::,; -": ""'i!"=:,'-.,'-. ':;-P'ump., '.

.Weldh Affected,",'::, ". Reason:Affected :

Affected:

80.0 Reactor Containment Spray 121 80'-03-WD-009 Embedded in Concrete 80-03-WD-012 Embedded in Concrete 80-03-WD-014 Embedded in Concrete 80-03-WD-010 When disassembled 80-03-WD-011 When disassembled 81.0 Reactor Core Spray 121 81-03-WD-009 Embedded in Concrete 81-03-WD-012 Embedded in Concrete 81-03-WD-014 Embedded in Concrete 81-03-WD-010 When disassembled 81-03-WD-011 When disassembled NMPC request relief from ASME Section XI, Table IWB-2500-1, Examination Category C-G, "Pressure Retaining Welds in Pumps, Examination Item Number C6.10, "Pump Casing Welds", of performing surface examinations of 100% v0elds, of only one pump among each group of multiple pumps in a system.

D. B I 0 Reactor Containment Spray Pumps, Figure ISI-PUMP-002(attached), provided a typical drawing of the pump 80-03, 80-04, 80-23 and 80-24. This drawing identifies sixteen (16) welds on each pump subject to examination.

Of the 16 welds ten (10) are subject to Examination Category C-G surface examinations. Of the ten (10) welds subject to surface examination, three (3) are embedded in concrete, and two (2) can only be examined when the pump is disassembled.

Reactor Core Spray Pumps, Figure ISI-PUMP-003 (attached), provided a typical drawing of the pump 81-03, 81-04, 81-23 and 81-24. This drawing identifies sixteen (16) welds on each pump subject to examination. Of the 16 welds ten (10) are subject to Examination Category C-G surface examinations. Of the ten (10) welds subject to surface examination, three (3) are embedded in concrete, and two (2) can only be examined when the pump File: RRISI7.WPD ISI 7-1 of ISI 7-4

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-7 is disassemblqd.

All the pumps in each system provide the same limitations as previously discussed above, therefore, the five (5) welds in question on the selected pump can not be substituted for five (5) welds on another pump.

NMPC considered selecting an alternate five (5) welds on another pump within each system to substitute for those welds that are inaccessible. This consideration was dismissed as the information provided would not provide meaningful information relating to the inaccessible welds.

Two (2) of the five (5) inaccessible welds on each pump are accessible when the pump is disassembled, welds 80-03-WD-010, 011 and 81-03-WD-010, 011. Table IWB-2500-1, footnote (2) allows the examination to be performed either from the inside or outside surface of the pump. Therefore these welds would be required to be examined when and if the pump is disassembled.

The three (3) weld on each pump that are embedded within concrete are inaccessible from the outside surface, but even if the pump was disassembled would provide some accessibility problems from the inside surface.

NMPC feels that the welds imbedded in the concrete would provide an greater acceptable level of safety over and above the limited surface examinations required by the Examination Category C-G.

NMPC proposes to the extent practical, and only when the pump is disassembled for maintenance, repair and or replacement to perform the surface examinations on welds 80-WD-03-010, 011 and 81-03-WD-010, 011,as required by Examination Category C-G.

On welds 80-03-WD-009,012, 014 and 81-03-WD-009, 012, 014, NMPC proposes to the extent practical and only when disassembled for maintenance, repair and or replacement to perform a Visual examination of the interior surface of the pump casing embedded in concrete.

The examination performed on accessible welds, coupled with the proposed examinations and the system pressure test will provide an acceptable level of quality and safety.

F.

Third Inservice Inspection Interval G. ACH E Drawing ISI-PUMP-003 Reactor Core Spray Pump Drawing Drawing ISI-PUMP-002 Reactor Containment Spray Pump Drawing File: RRISI7.WPD ISI 7-2 of ISI 7-4

37 33 0)

O CFPP IO EXAH IO PH'N 0) 80'0) ID Ml 80 0) ID.M2 M ~ ) ID M)

~

M.O) NO NA 80 0) ID M5 MO) IDM6 80 0) ID.M6 80 N 0)

0) r0 MT ID.NO x N.O).ID.OI6 N 03 ID NI NO) IDNS N 0) ID 809 M. 0) "ID OI ~

~

xU N 0).ID. ~ I I M 0) ID. ~ I2 z M 0) ID F 13 g)SZ M 0).ID NT N.O).ID ~ IS N 0) ID OIA mmz SUCTION.

0 ~

0 I SCHAAGE M M

0) ID ~ IS 03 ID ~ I6 r ><m EA)

M. Ol. ID. NI f XAHS TTPICAL FCXI PH'.O<, M-23 8 N.)i ~oKml C) f > m 0)

M ~ ) ID 08) M. 03. ID. M2 NOT Ss ll THIS GRAVING IDENT IF lfS C~NTS IO INSERT ICE INSPECI IGN EKANINATIDNS AS SUBJECT mz~

OSp REQUIRED Or ASIA SECT IOI Xl AID OEF INEO IN IHE IOPC INSfrlff INSPECT IOI PLAIL C 0 N 0)

IMPER ID ~ l3 CASE ~ LIA M 0) ID ON 2l THIS GRAVING IS lvlflDEO FOR INSERr ICE INSPE C I IOI PLAN IXX.r.

USE IN THE mmz M 0+

AEFEAE NCf Ss M 0) ID. ~ VTXITHINGIOH I 5HI.

CIAO GRAVING rV

) I l. ISDEN.I 7)TO I<i)I, MpZ z"

~

CASE.LVI I'LOIEA M.03 ID ~I ~

SECT IIXI XI, CLASS 2 SECT TDH Xl SQXDART OTAGIUVL F 63812 C SNEEI 2 N 0)"KO ~ I I XEIDNI HVANL Nlv)ISOOPIJV'NI

~ VELD FHP TXIAVING F ~ 45103 C SI8EIS l5, 15A. I58, 15C IDIEI THIS GRAVING VAS CAEATEO USING fll CDHPUIEA AIDEO GRAF IING 00 NOI HAKE HANUAI. CHANGES gl C

M. 0). VO. NS Null l NIAGARA NOHAVX PORN COAPOAATNWI HIIC HILf POINT IDCLE4A STATION lHll r f8 I ICAL lcf IN)PEcl IDH tut'N)EAr PRDGAAH 88 0) ID-OI2 ~ 4>>I>> IA N>> N>>4 a ISI PIAP 882

EO 33 33 CO O

COH'O EIAH IO PH" 61 03

~ SI 0) ID Nl Sl ~ ) ID N2 ee 41 03 ID Nl 61 03 IO Ni 61.03 IO N5 61.0') IO NC

~

61 ~

0) ID NC Sl ~ 03 IO.NT Sl 0) VO NS x 4 I . 0). IO ~ I C 41 0) IO NS 61-0). IO. NS 4 I . 0). ID. NT Sl 0) ID ~ I ~ U Sl 4) ID ~ II Sl 0)'IO 012 z Sl 0) IO Sl) g)SZ Ill g 41 0) ID. NT

~ ~

SI.S).IO.415 Sl 0)'IO' I~

~

r. XI 4

SUC I IOI 0 I SCHARGE 61 61 ~

0) lD 015 03 lO SIC

<m e

EEAHS TIP ICAL FOR -g 0 K

~ All tIP 41 N, Sl 23 6 Sl Cl 61-03 IO.NI Sl . 03. ID. N2 NO IfSe CO Sl ~ ) IO Nl ll 'IHIS ORAIIING IOEHllfIES COHVICHIS SUS)'ECI

~ I ~ 03 IO ~ IlLVI IO INSERV lff f INSPECT IIXI EAHIHATIONS AS AEIEIIAEO Sr ASIA SECI ION x I IVO OEF IIRO IN THE IAPC INSEV ICE IeeltECI IGN PLAIL C m lo 0-2 I IfetER CASE 41 0).IO NS 21 THIS ORAVIID IS INIEIOEO FOR USE IH INSERr ICE INSPE CI IGN PLAN Oex.r.

THE MOW I

$ 0z REF EAEICESe

- VORIHINGTI)I 15err.)lt <S Sl I LOVE4 Sl IO 01 ~

CASE 'LIA SI ~ Sl. IO. ~ I ~

CIWP I)IAVING VV.I~ r21 OEH.1 2314 SECT IOI XI, CLASS 2 Z =I Sl-0)-ID Sll

- SECT IOI xl SauoAAr DIAGRAM F ~ C)NT.C SIff1 I VEIExxl IHIENL Nl V)1546PIPPNI VELD lrve IXVIVIIOf <SIS) C SeffTS 13, 13A, 136. 13C z

IDlfe THIS GRAVING VAS CAEAIEO USING m COIWUIER AIOEO ORAf I ING 00 NOT HAKE HANUAL CHANGES C

Sl ~ 03 IO.NS D NU~ I HIAOARA HOHAWE tOWER COAPORATIOH NINE HILE POIN'I IAXLEAR STAIR)I IPIIT c<<NL<<eel e VERI ICAL PItP INSEAV IC INStfCI IOH tAOGAAH

<<e a <<~<<<<eeeere ~ ~

61 03 IO ~ 12 ~ << II<<A

<< ISI tert N)

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-8 Article IWA-4000 Repair Procedures Article IWA-6000 Records and Reports Article IWA-7000 Replacement B. E I E IWA-4800 The records required by IWA-6000 shall be completed for all repairs.

IWA-7520(8) Completed Owner's Report for Repairs or Replacements, Form NIS-2 IWA-6210(c) The Owner shall prepare inservice inspection summary report for Class 1 and 2 pressure retaining components and their supports.

IWA-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.

IWA-6220(d) Each summary report shall contain the following:

(2) Owner's Report for Inservice Inspection, Form NIS-1 (3) Owner's Report for Repair or Replacement, Form NIS-2 IWA-6230 Within 90 days of the completion of the Inservlce inspection conducted during each refueling outage, the Owner shall file ISI Summary Reports with the enforcement and regulatory authorities.

C. RELIEF REQUESTED:

Pursuant to 10 CFR 50.55a(a)(3)(i), NMPC requests Relief from the following:

1. Preparation of the Owner's Report for Inservice Inspection, Form NIS-1

2. Preparation of the Owner's Report for Repair or Replacement, Form NIS-2

3. Submittal of the summary report within 90 days following completion of the inservlce inspection conducted during each refueling outage.

D. BASIS FOR RELIEF:

NMPC 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 NMP1.

The summary report does not furnish evidence of compliance with the ASME Boiler and Pressure Vessel Code,Section XI, Inspection Program B, percentage requirements as mandated by IWB-2412, IWC-2412, and IWD-2412.

Class 3 components are excluded from the summary report Submittal.

File: RRISI8.WPD ISI 8-1 of ISI 8-6

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-8 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 NMPC without an increase in assurance of Code compliance.

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,Section XI Division 1", has not been published in Regulatory Guide 1.147, dated May 1999 "Inservice Inspection Code Case Acceptability ASME Section XI, Division 1", however, the USNRC staff has approved it's use at other nuclear stations.

The information provided in the documentation pertaining to the use of Code Case N-532, can be used in the same manner to assess the safety implications of Code activities performed during the outage. A review using the information as prescribed by the Code Case will, therefore, provide the same or improved level of quality and safety as reviews that may be conducted using the Code reporting requirements.

E.

As an alternate to the requirements of IWA-4800, IWA-6000, and IWA-7528(8), NMPC will implement ASME Code Case N-532, "Alternative Requirements to Repair and Replacement Documentation Requirements and Insetvice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000', Division 1",

(Note: 1 - ASME 1992 Edition Section XI).

IMPLEMENTATIONSCHEDULE:

The Alternate requirements of ASME Code Case N-532 will be incorporated into NMPC Inservice Inspection Program during the 3rd Ten-Year Interval.

G. ATTACHMENTSTO THE RELIEF:

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 N

1 File: RRISI8.WPD ISI 8-2 of ISI 8-6

CASE N-532 CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: December 12, 1994 She Numeric Index for expiretion end eny reeffirmetion detes.

Case N-532 (d) Form NIS-2A shall be presented to the Inspector Alternative Requirements to Repair and for certification.

Rephcement Documentation Requirements and (e) The completed Form NIS-2A shall be maintained Inservice Summtry Report Preparation and by the Owner.

Submission as Required by IWA~ and (0 The Owner shall maintain an index of Repair/

IWA~t Replacement Plans in accordance with 1WA-6340. The Section XI, Division 1 index shall identify the identification number required by (b) above the inspection interval and period during Inquiry: What alternatives may be used to the re- which each repair or replacement was completed.

quirements of IWAA910(d) and IWA-6210(e) for completion of Form NIS-2 following repair or replace- 2.0 OWNER'S ACTIVITY REPORT ment, and IWA-6210(c) and (d), IWA-6220, PREPARATION AND SUBMITTAL IWA-6230(b), (c), and (d), and IWA-6240(b) for prep-aration and submittal of thc inseivice sumniary rcport An OWNER'S ACTIVE REPORT FORM OAR-1 and Fortn NIS-1? shall be prepared and certified upon completion of each refueling outage. Each Form OAR-1 prepared during an inspection period shall be submitted following the Reply: It is the opinion of the Committee that as end of the inspection period. Each Form OAR-1 shall an alternative to the requirements of IWAA910(d),

contain the following:

IWA-6210(c), (d), and (c), IWA4220, IWA-6230(b).

(a) Abstract of applicable examinations and tests (c), and (d), and IWA-6240(b), the following provi-with the informafiion and format of Table l.

sions may be usecL This Case shall be utilized at least (b) A listing of item(s) with flaws or relevant condi-until the cnd of the inspection period in which it was tions that required evaluation to determine acceptability invoked.

for continued service, whether or not the flaw or relevant condition was discovered during a scheduled 1.0 CERTIFICATION OF THE REPAIR OR examination or test. Me listing shall provide thc infor-REPLACEMENT mation in the format of Table 2.

(c) Abstract for repairs, replacements and corrective (o) The Owner's Repair/Replacement Program shall measures performed, which were required due to an item identify use of this Case. containing a flaw or relevant condition that exceeded (b) A Repair/Replacement Plan shall be prepared in IWB-3000, IWC-3000, IWD-3000, IWE-3000, accordance with IWA%140', and shall bc given a IWF-3000, or .IWI 3000 acceptance criteria; even unique identification number. though the discovery of the flaw or relevant condition (c) Upon completioq~f all required activities associ- that necessitated the repair, replacement or corrective ated with the Repair/Replacement Plan, the Owner shall measure, may not have resulted from an examination prepare a REPAIIVREPLACEMENTCERTIFICATION or test required by this Division. If acceptance criteria, RECORD, FORM NIS-2A. for a particular item is not specified in this Division, the provisions of IWA-3100(b) shall be used to determine which repairs, replacements, and concctive measures are

'll references to to the l992 Edition.

IWA~ and IWA4000 used in this Case refer required to be included in the abstract. The abstract shall provide the information in the format of Table 3.

829

CASE (continued)

N-532 CASES OF ASME BOILER AND PRESSURE VESSEL CODE FOAM NIS-2A REPAIR/REPLACEMENT CERTIFICATION RECORD OWNER'S CERllRCATE OF CONFORMANCE I certify that the represent by Repair/Replacement ieooII or weeonwnr Plan number conforms to the requirements of Section XI.

Type Code Symbol Stamp Certificate of Authoritation No. Expiration Date Signed Oate Owner or Owed ~ Oooenee. toro CERTIRCATE OF INSERVICE INSPECTION I, the undersipned. holdlnp ~ valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors and the State or Province of ~ nd employed by of have inspected the items described in Repair/Replacement Plan num.

ber during the period to ~ nd state that to the best of my knowledge and belief, the Owner,has performed all the activities described in the Repair/Replacement Plan in accordance with the requirements of Section XI.

By signinp this certificate neither the Inspector nor his employer makes any warranty, expressed or implied. concerning the activities described in the Repair/Replacement Plan. Furthermore, 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 ln ooeoro r' Xrnaervro rreoonol soonL stere. orownoo. end treoenw ~ ie Date lhle form lkoot20) msy be obtained from the order ooot AsME, 22 taw Drive. sox 2200, Foirllold, IIJ 01007 2200.

830

CASE (continuedj N-532 CASES OF ASME BOILER AND PRESSURE VESSEL CODE FOAM OAR-1 OWNERrS ACTIVITY REPORT Report Number Owner (Nerne end Adorned OI Owner)

Plant

'd INerne end Addreee OI Ainu Unit No. Commercial service data Refueling outage no.

eddeeeelol Current inspection interval I IN. 2nd, snx edr. edrerl Current inspection period Ila,mrs a dl Edition and Addenda of Section XI applicable t the inspectin plan Date and revision of Inspection plan Edition and Addenda ot Section XI applicable to repairs and replacements, 'han e inspection plan I certify that the statements made In this Owner's Rap re correct. and that the examinations, tests, repairs. replacements, evaluations.

and corrective measures represented by this report n h requirements ot Section XI.

Certificate of Authorltatlon No. Expiration Date ed ~

Sign Data Owner or Owner e I. the undersigned. holding a valid commission issued by the National Board of Boiler and Pressure Vessel Inspectors add the State or Province of and employed by of haveinspactad the items described in this OwneVs Acdvity Report. during the period to and state that to the bast ot my knowledge snd belief, the Owner has pertormed all activities represented by this report ln accordance with the requirements of Section XI.

By signing this certifioata neither the Inspector nor his employer makes any warranty, expressed or implied. concerning tha examinations. tests.

repairs. replacements, evaluations and corrective maaures described ibis repon. Furthermore, neither tha inspector nor his employer shall be liable in any manner for sny personal injury or property damage or a loss of any kind arising from the connected with this inspection.

Commissions Neeenu aderrx sINK dwerrnee. end t ee ~ nrerru Data f

This form I f00127) mev be obtained from the Order OeotAS ME, 22 Ldw Drtvd. Box 2200, eirlteld. HJ 07007 2XO.

831

CASE (continued)

N-532 CASES OF ASME BOILER AND PRESSURE VESSEL CODE TABLE 1 ABSTRACT OF EXAMINATIONSAND TESTS Total Total Total Total Examinations Examinations 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 (IWA-3300) Examination or Test (Yes or No)

TABLE 3 ABSTRACT OF REPAIRS, REPLACEMENTS, OR CORRECTIVE MEASURES REQUIRED FOR CONTINUED SERVICE Flaw or Relevant Condition Found Repair, . During Scheduled Replacement, Section XI Repair/

Code or Corrective Item Description Examination or Date Replacement Class Measure Description of Work Test (Yes/No) Completed Plan Number 832

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-9 IWF-5000 Class: Quality Group A, B and C, (ASME Code Class1,2 and 3)

Identification of Components: Mechanical and Hydraulic Snubbers Systems: All B.

ASME BBPV Code, Section Xi, Article IWF-5000 1989 Edition invokes the snubber examination requirements of Standard OMa-1988, Part 4, Section 2.3.2.2 which states 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.

ED'elief is requested from performance of visual inspections 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 OMa-1988 Section 2.3.2.3 is also requested.

D. B ELIEF:

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 number of unacceptable snubbers found during the previous visual inspection, irrespective of the size of the snubber population. The 18-month inspection interval is incompatible with current operating cycle lengths of 24 months. Due to the large number of snubbers in use at the Nine Mile Point 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 no commensurate increase in quality or safety. As concluded by the USNRC staff in Generic Letter 90-09, the proposed alternative inspection maintains the same confidence level in snubber operability. The proposed alternative is compatible with the current 24-month operating cycle and generally will allow inspections to be performed during plant outages, 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.

In addition to the ASME Code,Section XI requirements, Shock Suppressors (Snubbers) surveillance requirements are addressed in Plant Technical Specifications 3.6.4/4.6.4. The requirements of the Technical Specifications snubber visual inspections and testing provides the necessary assurance for snubber operability and visual examination requirements to fulfillthe ASME Code,Section XI requirements without duplicating the inspections.

The proposed alternative inspection conforms with USNRC Generic Letter 90-09.

FILE: RRISI9.WPD ISI 9-1 of ISI 9-3

NINE MILE POINT UNIT 1 ...

THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-9 Examinations of snubbers will be performed at intervals and sampling rates in accordance with the requirements specified in Generic Letter 90-09, "Alternative 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 criteria:

Population Column Column Column Category Interval A'xtended Interval 8'epeat Interval C'educe 80 100 150 200 300 12 25 The next visual inspection 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 intewal. 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 made and documented before any inspection 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, 8, 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 8 but greater than the number in Column A, the next inspection interval shall be the same as the previous Intervyl.

If the number of unacceptable snubbers is equal to or greater than the number 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 8, 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 8 to the difference in the numbers in Columns 8 and C.

The standard 25% extension on surveillance intervals is applicable to any examination interval determined in accordance with this alternative.

All inservice inspection (VT-3 examinations) of snubbers shall be performed per the requirements of the Nine Mile Point Unit 1 Technical Specifications.

FILE: RRISI9.WPD ISI 9-2 of ISI 9-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-9 F.

The proposed alternative inspection will be implemented during the third 10-year inspection interval.

G.

None FILE: RRISI9.WPD ISI 9-3 of ISI 9-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-10 D

Class: All Identification of System: All B. DE E E 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 Program in accordance with IWA-4120.

C.

Pursuant to 10 CFR 50.55a(a)(3)(ii), NMPC requests Relief from the requirements of ASME Section XI, Article IWA-4000, IWA-4400.

D. B 0 ELI 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 qualify all welding procedures even though they have met the qualification process at another facility. Code Case N-573 recognizes and addresses this fact and proposes an alternative which maintains an acceptable level of quality and safety.

E.

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.

1. NMPC will perform a technical review of the supplying Owner's PQR

2. The supplying Owner will state in writing that the PQR 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.

NMPC will generate a NMPC WPS using the variables established in the supplied PQR(s).

NMPC PQR's may supplement these or other Owner supplied PQR's.

V

4. The WPS will be approved and signed by NMPC.

5. The WPS will be demonstrated successfully by NMPC by completing a welder performance qualification test using the parameters of the NMPC WPS.

6. NMPC will not transfer the supplied PQR to any other Owner.

7. NMPC will document the use of this Code Case on the appropriate NIS-2 form.

File: RRISI10.WPD ISI 10-1 of ISI 10-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-1 0 The Alternate requirements of ASME Code Case N-573 will be incorporated into the NMPC Inservice Inspection Program during the 3rd Ten-Year Interva, or until Code Case N-573 is approved for general use by reference in Regulatory Guide 1.147. After that time, NMPC will follow the conditions, if any specified in the regulatory guide.

ASME Code Case N-573, Transfer of Procedure Qualification Records Between Owners,Section XI, Division 1 Pressure Test of Containment Penetration Piping, Section Xl, Division 1.

File: RRISI10.WPD ISI 10-2 of ISI 10-3

CASE N-573 CASES OF ASME BOlLER AND PRESSURE VESSEL CODE Approvet Date: March 1Z, 1997 See Numerical Index for expiration and any reaffirmatfon dates.

Case N-$ 73 Assurance Program that satisfies the requirements of Transfer of Procedure QuaHfication Records IWA-1400.

Between Owners (c) The Owner accepting the completed PQR shall Section XI, Division I accept responsibility for obtaining any additional sup-porting information needed for WPS development.

Inquiryt What alternatives to the welding and braz- (d) The Owner accepting the completed PQR shall ing procedure qualification requirements of IWAP000 document, on each resulting WPS, the parameters appli-cable to ~elding. Each WPS shall be supported by all may be used?

necessary PQR's.

(e) The Owner accepting the completed PQR shall Reply: It is the opinion of the Committee that as accept responsibility for the PQR. Acceptance shall be an alternative to the welding and brazing procedure documented by the Owner's approval of each WPS qualification requirements of IWAA000, a procedure that references the PQR.

qualification record (PQR) qualified by one Owner may be used by another Owner. When this alternative g The Owner accepting the completed PQR shall demonstrate technical competence in application of the is used, the following requirements shall be met: received PQR by completing a perfonnance qualification (a) The Owner that performed the procedure qualifi- test using the parameters of a resulting WPS.

cation test shall certify, by signing the PQR, that testing (g) The Owner may accept and use a PQR only was performed in accordance with Section IX. when it is received directly Rom the Owner that certified (b) The Owner that performed the procedure qualifi- the PQR, cation test shall certify, in writing, that the procedure (h) Use of this Case shall be shown on the NIS-2 qualification was conducted in accordance with a Quality form documenting welding or brazing.

1145 SUPP. 8 NC

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-11 System: 00.0 Reactor Pressure Vessel Class: Quality Group A (ASME Code Class 1)

Component

Description:

Reactor Pressure Vessel Closure Head Nuts ASME Section XI, Table IWB-2500-1, Examination Category B-J, "Pressure Retaining Bolting, Greater Than 2 Inch In Diameter, Examination Item Number B6.10, "Reactor Vessel Closure Head Nuts ", requires 100%

surface examination.

C. REL Pursuant to 10 CFR 50.55a (a)(3)(i), NMPC requests relief from performing a 100% surface (magnetic particle) examination of the sixty-four (64) Reactor Pressure Vessel Closure Head Nuts as required by Table IWB-2500-1.

Due to the closure head nuts distinct size and geometric configuration, surface (magnetic particle) examination methods as required by IWB-2500-1 (89 Edition), added considerable costs associated with removal, preparation (both post and pre-cleaning), and examination time with little or no compensating increase in the quality and safety of the plant.

The 1989 Edition of Section XI does not provide acceptance criteria for the mandated surface examination of Table IWB-2500-1.

ASME Section XI subcommittee recognized this minimal increase in quality by mandating a surface examination over a visual examination, The 1989 Addenda, Table IWB-2500-1 was changed by requiring a Visual (VT-1) examination of the Reactor Vessel Closure Head Nuts, which also referenced acceptance criteria for VT-1 examination of bolting greater than 2 inches.

Both the visual and magnetic particle examination address the examination on the surface of the component.

The additional subsurface depth of the magnetic particle examination over. the visual examination of the surface does not provide a substantial increase in the level of quality and safety, NMPC proposes to utilize the Visual VT-1 examination requirements and acceptance criteria of the 1989 Addenda of Section XI for Reactor Vessel Closure Head Nuts, in lieu of the surface examination requirements of the 1989 Edition with no acceptance criteria.

The extent of examination performed will provide an acceptable level of quality and safety.

IMPLEMENTATIONSCHEDULE Third Inservice Inspection Interval File: RRISI11.WPD ISI 11-1 OF ISI 11-1

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-12 System: Various Class: Augmented Quality Group A, (ASME Code Class 1)

Component

Description:

Nonconforming Service Sensitive Piping Welds.

B. UD D E

NUREG 0313, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping (Reference 11), requires augmented volumetric and surface examination of nonconforming service-sensitive piping welds.

E E E Relief is requested from performing full volumetric and surface examination of nonconforming service sensitive piping welds. Relief is requested for twenty-one (21) of the augmented piping welds.

ELIE The welds listed on the attached Table were not fully examined by volumetric and/or surface examination methods during the first and second 10-year interval due to limitations of design, geometry, and material of construction.

The dendritic weld structure of the stainless steel material can result in both sound redirection and attenuation phenomena which limit ultrasonic interrogation. Thus, such welds necessitate examination from both sides in order to be fully examined. In particular, non-parallel surfaces and product form of the material and product form of the material of valves preclude meaningful ultrasonic examination from the valve side.

Four (4) stainless steel welds continue to be limited by configuration, two (2) by permanent attachment to the piping and fifteen (15) by containment penetrations. The percentage of weld required area (WRA) and Weld Required Volume (WRV) that was completely examined is tabulated with the nature of the obstruction on the attached Table.

Per NUREG-0313, the Core Spray System (40) piping is defined as nonconforming service sensitive; the extent and frequency of examination is 100% of those welds every outage. Other system welds that had been selected for this augmented examination program were also examined each outage and thus had been more frequently inspected than required by NUREG-0313.

Perform ultrasonic and surface examinations to the extent practical.

Perform a Visual (VT-2) examination of the inaccessible IGSCC Category welds each refueling outage for evidence of leakage per NMPC submittal dated July 28, 1988 and September 4, 1990 commitment.

The examinations as proposed, together with the other pressure tests (as applicable) provide an acceptable level of assurance of nonconforming service sensitive piping weld integrity.

Filo: RRISI12.WPD ISI12-1 of ISI12-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-12 Pursuant to 10 CFR 50.55a(g)(6)(i), this request for relief was granted per USNRC Safety Evaluation, dated April 6, 1994, TAC No. M83099, for the second inspection interval.

Approval of this request for relief for the second inspection interval also included a submittal dated July 28, 1988 and September 4, 1990, that committed to performing a visual examination of the inaccessible IGSCC Category welds each refueling outage for evidence of leakage.

FIIQ: RRISI12.WPD ISI12-2 of ISI12-3

NINE MILE POINT UNIT 1 THIRD INSERVICE INSPECTION INTERVAL RELIEF REQUEST ISI-12 OMPONENT COMPONENT DESCRIPTION EXAMINATION EXTENT LIMITATION IDENTIFICATION METHOD EXAMINED 40-WD-050.A VALVE40-12 TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X-14 40-WD-010A VALVE40-02 TO PIPE UT/PT P INACCESSIBLE INSIDE PENETRATION X-13A 40-WD.005 PIPE TO ELBOW UT/PT WRV 58% PERMANENT HANGER OBSTRUCTION WRA 86%

40.WD.006 PIPE TO PIPE UT/PT WRV 82% PERMANENT HANGER INTERFERENCE 40.WD-011 ELBOW TO PIPE UT/PT WRV 31% INACCESSIBLE INSIDE PENETRATION X-14 WRA 25%

40.WD.051 PIPE TO ELBOW UT/PT WRV 50% INACCESSIBLE INSIDE PENETRATION X-14 37-WD 003 REDUCER TO FLANGE UT/PT 0% FITTING CONFIGURATION 39.09R-WD.001 VALVE39.09R TO PENETRATION UT/PT P CONFIGURATION 39-10R-WD.001 VALVE39-10R TO PENETRATION UT/PT 0% CONFIGURATION 39-WD-194 VALVE39.05 TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X-SB 39.WD-194A PIPE TO PIPE UT/PT P INACCESSIBLE INSIDE PENETRATION X-SB 39-WD.226 VALVE39-06 TO PIPE UT/PT 0% CONFIGURATION 39-WD-226A PIPE TO PIPE UT/PT P% INACCESSIBLE INSIDE PENETRATION X-5A 38-WD.007 PIPE TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X.8 38-WD-087 VALVE38-12 TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X-7 38-WD 088 PIPE TO PIPE UT/PT P INACCESSIBLE INSIDE PENETRATION X-7 33.WD 014 PIPE TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X-9 33-WD.036 PIPE TO ELBOW UT/PT P% INACCESSIBLE INSIDE PENETRATION X-154 33-WD.035 VALVE33-03 TO PIPE UT/PT 0% INACCESSIBLE INSIDE PENETRATION X-154 33-W D.015 VALVE33-04 TO PIPE UT/PT P INACCESSIBLE INSIDE PENETRATION X.9 Filo: RRISI12.WPD ISI 12- 3 of ISI 12- 3

ML18018B052

Text

1.0 INTRODUCTION

1.0 INTRODUCTION

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Description:

Navigation menu

Search