ML101670584

Inservice Inspection Examination Plan, Rev. 4, Fourth Interval, May 1, 2003 Through May 31, 2012

ML101670584
Person / Time
Site:	Monticello
Issue date:	06/10/2010
From:	Blomgren T Northern States Power Co, Xcel Energy
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
QF-0212, Rev 4
Download: ML101670584 (799)

Text

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NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN XcelEnergy XCEL ENERGY, INC. NSP-MINNESOTA 414 NICOLLET MALL MINNEAPOLIS, MN 55401 MONTICELLO NUCLEAR GENERATING PLANT 2807 WEST HIGHWAY 75 MONTICELLO, MINNESOTA 55362 INSERVICE INSPECTION EXAMINATION PLAN REVISION 4 FOURTH INTERVAL MAY 1, 2003 THROUGH MAY 31, 2012 Reviews and Approvals: PCR-01 125788 Approval Date: 6-5-2010

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN RECORD OF REVISIONS Pa2ie Rev.* Review and Approval (PCR-01125788) ..................................................... 4 i- x i .................................................................................................. . . .. . . 4 1.1-1 ........................................................................................................ ...2 1.2-1 through 1.2-7 ................................................................................. ..4 1.3-1 a nd 1.3-5 ....................................... ............................................... . . . 4 1.4-1 throug h 1.4 -7 ................................................................................. .. 3 1.5-1 through 1.5-246 ............................................................................... 3 1.6-1 throug h 1.6-5 .................................................................................. . . 22 1.7-1 through 1.7-7 ................................................................................. .. 2 Inspection Schedule (Page 1 to 370) ............................................................. 2 Note: The revision numbers reflect section-specific revisions. Drawings in Section 1.6 retain their own drawing specific revision number. However, the governing revision of all sections is the ISI Plan Revision listed on the Review and Approval page. 0 ii

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN RECORD OF REVISIONS (cont'd) Summary of Changes, Plan Revision 1 Title Page Modified Interval start date to May 1, 2003 and noted Revision 1 Page i Noted Rev. 1 for affected sections Page ii Added page for Summary of Changes, Revision 1 Page 1.2-1 4th Ten Year Interval

• Updated revised 3rd Interval extension dates
• Added discussion regarding overlap of 3rd and 4th Intervals
• Changed 4 th Interval start dates Component Selection
• Added requirement for examination of re-used CRD Bolting Page 1.2-2 Code Edition Summary
• Added requirement for examination of re-used CRD Bolting
• Removed reference to NF (Supports). NF not applicable, supports are examined per Subsection IWF
• Modified Appendix VIII Section to reflect latest modification of Appendix VIII implementation per 10CFR50.55a and remove references to specific Supplements and implementation dates Page 1.2-3 Examination Personnel / Procedures

              " clarified description of examination personnel and procedure requirements.

• Clarified to reflect additional use of Mandatory Appendix VIII requirements for UT personnel and procedures as modified by 1 OCFR50.55a dated September 26, 2002, except where relief has been granted.

              " Removed reference to Appendix VIII - Supplements.

Page 1.3-1 Removed reference to unpublished Reg. Guide 1.147 Rev. 13 (Draft Reg. Guide 1091) Section 1.4 Modified entire section to remove references to Code Cases listed as approved or conditionally approved in unpublished Draft Reg. Guide 1091, but not found in published Reg. Guide 1.147, Rev. 12. iii

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 2 Title Page Modified revision number and revised titles/names Page i Changed Revision Number-and pages for affected sections Pages iii & iv Added pages for Summary of Changes, Revision 2 Page 1.2-1

Background:

0 Format / spacing changes 4th Ten Year Interval

• Format / spacing changes
• Changed number of scheduled outages from Six to Five
• Removed reference to maintenance outages Component Selection 0 Format / spacing changes Page 1.2-2 Code Edition Summary
• Added provisions for implementation of approved ISI Relief Request #7 to use 2001 Edition of Section XI for Repair/Replacement activities and associated Pressure Testing. NRC exception noted. (see Corrective Action Program OTH020219)

Background for Plan / Schedule Development

• changed intent of scheduling from "subject to allowing meaningful accumulation of service time for new components" to "to the extent practical Page 1.2-3 ISI Plan Overall Description
• Added RI-ISI to the description of how components are listed in the Plan and Schedule a Capitalized Item Number Page 1.2-4 ISI Plan Overall Description (cont'd) a Added "Rev. B-A" to TR-112657 Page 1.3-1 Source Documents
• Added 1995 Edition, 1995 Addenda of Section XI
• Added 2001 Edition, No Addenda of Section XI iv

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 2 (cont'd) Page 1.3-2 Source Documents (cont'd) a Added NRC SER for Relief Request #7 Page 1.5-1 Relief Requests 0 Added Relief Request No.7 Page 1.5-35 Relief Request No.2

• Added Clarification in title regarding italicized text Page 1.5-48 Relief Request No.3
• added Reference 11, NRC SER Title for Relief Request No.3
• updated Status as approved Page 1.5-53 Relief Request No.5
• updated Status as approved and listed NRC SER Title for Relief Request No.5 Page 1.5-56 Relief Request No.6
• added Reference 4, NRC SER Title for Relief Request No.6
• updated Status as approved Page 1.5-57 Relief Request No.7 a added Relief Request No. 7 including NRC exceptions V

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 3 Title Page Modified revision number and revised titles/names Page i Changed revision number and pages for affected sections and added note discussing applicability of revision numbers Pages v - viii Added pages for Summary of Changes, Revision 3 Page 1.2-1 4 th Ten-Year Interval

• clarification of 3 rd Interval extension and overlap of 3 rd and 4 th Intervals Component Selection
• Added reference to ASME Section XI Code Case N-598 applied to per-period percentage requirements Page 1.2-2 Code Edition Summary
• Class 1 (Quality Group A), clarified description
• Class 1 CRD Bolting, removed reference to augmented program GE SIL. No. 483R2 for examination of B7.80 items.

10CFR50.55a requires use of 1995 Edition for reused CRD bolting.

• Class 2 (Quality Group B), clarified description
• Appendix VIII - Mandatory, updated to include subsequent publications of 10CFR50.55a
• Repair/Replacement..., added further restrictions invoked by 10CFR50.55a effective November 1, 2004 Page 1.2-3 Examination Personnel / Procedures 0 updated to include subsequent publications of 10CFR50.55a Page 1.3-1 Source Documents
• Changed 1995 Edition, 1995 Addenda of Section XI to 1995 Edition, No Addenda

              " Updated 10CFR50.55a to current reference in Federal Register

• Updated Reg. Guide 1.147 to Rev.13
• Added EWI-09.04.00
• Deleted letters regarding 3 rd Interval Relief Requests (RR) 8 and 13
• Added Jan. '03 letter regarding further extension of 3 rd Interval
• Added Dec. '02 letter requesting NRC review of 4th Interval relief requests included with ISI Plan submittal vi

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 3 (cont'd) Page 1.3-2 Source Documents (cont'd)

• Added letter for ISI Plan, Rev.2
• Deleted letter regarding 3 rd Interval RR 14
• Added cross-reference to 3 rd Interval RR 12 (4 th Interval RR 2)
• Added NRC SER for RR 3 and 6
• Added NRC SER for RR 5
• Added date to NRC SER for RR 7
• Added NRC SER for RR 8
• Added NRC SER for RR 9 (Fleet RR)
• Added NRC SER for RR 10 (Fleet RR)
• Added NMC Letter to NRC for RR 11 (Fleet RR)

Page 1,4-1 Section XI Code Cases thru 1.4-6

• Added reference to Rev.13 of Reg. Guide 1.147
• Added discussion that currently applied Code Cases are underlined Completely updated list for Code Cases that are applied to the ISI Plan. Also included Code Cases that would be beneficial if needed on a case by case basis, but have not been implemented for use to date.

Page 1.5-1 Relief Requests thru 1.5-2

• Added RR 8, 9, 10 and approval dates
• Added RR 11 submittal (pending approval)
• Added approval dates to RR 3, 4, 5, 6, 7
• Clarification for notes regarding RR 1 and 2 Page 1.5-50 Relief Request No.4 thru 1.5-54
• Replaced original submittal with updated final submittal and reference to NRC approval Page 1.5-77 Relief Request No.8 thru 1.5-82
• Added final submittal and reference to NRC approval Page 1.5-83 Relief Request No.9 thru 1.5-107 a Added final submittal and reference to NRC approval vii

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 3 (cont'd) Page 1.5-108 Relief Request No.10 thru 1.5-113

              . Added final submittal and reference to NRC approval Page 1.5-114 Relief Request No.11 thru 1.5-118
              . Added final submittal Page 1.6-1     Boundary and Isometric Drawing Index thru 1.6-4
              & Updated to show current revision / status ISI Boundary Drawings and ISl Class 1, 2, and 3 Isometric Drawings

• Updated multiple drawings for editorial corrections, piping configuration changes, and elimination of systems Page 1.7-1 Inspection Plan and Table thru 1.7-4
• Periods, updated period start / end dates
• Scheduled Outages, updated with current outage schedule
• Scheduled Outages, added Code Case N-598 for per-period percentage requirements
• Note 3, added "to the extent practical"
• Note 4, removed reference to GE SIL 483 and updated with requirements from 10CFR50.55a
• Note 6, added with exam requirements for Code Cat. B-G-2, Item B6.180
• Note 7 p, added further definition
• Note 7 I, added further definition
• Pressure Testing Note A, added details re: restructuring of the Plan and scheduling of Class 1 System Pressure Tests.
• Pressure Testing Note B, added details re:restructuring of the Plan and scheduling of Class 2 System Pressure Tests.
• Pressure Testing Note D, updated from Code Case N-488-1 to N-498-4
• Pressure Testing Note E, previously reserved, applied for Code Case N-522
• Pressure Testing Note F, added details regarding restructuring of the Plan and scheduling of Class 2 System Pressure Tests.
• NDE Note AA, clarified exam is performed near the end of the Interval.

viii

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 3 (cont'd) 4 th Interval Plan, Component Examination Schedule " Updated to reflect current status of examination status

• Updated to reflect inactivation of eliminated systems per Design Change 02Q195 (RPV Head Spray Removal) and 03Q145 (CGCS System Removal)

" Updated to include integrally welded attachments (B-K, C-C, or D-A) which were previously examined under the IWF category, or were exempted from examination in the 3 rd ISI Interval. (see CA 022418, CA 020622, CAP 029634) " Updated schedule to reflect incorporation of Code Case N-598

• Updated schedule for proportional examination of support by system and function (SA 021615, CAP 036901, CA 023866)

" Updated schedule for applied Code Cases (SA 021615, CAP 036901, CA 023866) " Updated schedule to optimize examination requirements for "multiple" Class 1 and 2 components ( (SA 021615, CAP 036901, CA 023866) " Updated schedule for items permitting deferral to end of interval (SA 021615, CAP 036901, CA 023866) " Updated schedule with other miscellaneous changes from SA 021615 and minor self-identified discrepancies. ix

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 4 Title Page Modified revision number and company information Page i Changed revision number and pages for affected sections Pages x - xi Added pages for Summary of Changes, Revision 4 Page 1.2-1

Background:

• Updated company information
• Added references to the Snubber and BWRVIP Programs Component Selection
• Deleted reference to 10CFR50.55a(b)(2)(xi) that was deleted from the regulation
• Corrected editorial error - deleted incorrect reference to 1995 Addenda for applicable to Item B7.80 Page 1.2-2 Code Edition Summary

              " Class 1 (Quality Group A), deleted reference to 10CFR50.55a(b)(2)(xi) that was deleted from the regulation
              " Class MC (Metal Containment), updated code of record to 2001 Edition with 2003 Addenda

• Appendix VIII - updated to include 2001 Edition referenced in subsequent publications of 10CFR50.55a Page 1.2-4 ISI Plan Overall Description
• Added reference to non-Code exams for welds on drawing NC-ISI-51 Page 1.2-4 Added new section, License Renewal Aging Management Plans and Commitments

              " Listed applicable Program Basis Documents /Aging Management Plans and applicable commitments

• Added new subsection for Augmented Program to examine Class MC supports per Subsection IWF requirements

              " Added new subsection for Augmented Program to examine Class 1 small bore piping < NPS 2 and < NPS 4 Page 1.3-1     Source Documents

• Deleted 1992 Edition, 1992 Addenda of Section XI
• Added 2001 Edition with 2003 Addenda for IWE x

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Summary of Changes, Plan Revision 4 (cont'd) Page 1.3-3, -5 Source Documents

• Added reference NRC Commitment M97025A
• Added reference to NUREG-1865 for License Renewal
• Added reference to RI-ISI first period update
• Added source documents related to relief requests 12 - 19
• Added references to applicable License Renewal PBD/AMPs and NRC Commitments Page 1.4-1, -7 Code Cases

              " Updated list for Code Cases published in RG 1.147, Rev.15 used, or potential for use at MNGP.

• Updated list for Code Cases that have been applied, or are proposed to be applied, via Relief Requests.

Page 1.5-1, -2 Relief Requests

• Updated list with proposed, submitted, and approved dates, as applicable.
• Updated list with Relief Requests 12 - 19.

Page 1.5-118 through 1.5-246

• Added Relief Requests 12 - 19
• Updated list with Relief Requests 12 - 19.

Section 1.6 ISI Boundary/Isometric Drawings

• Updated several drawings for applicable system modifications and editorial corrections
• Added new drawings for Class MC Supports Section 1.7 Inspection Plan and Schedule Tables

              " Added table to describe the Item Numbers applied to Risk-Informed ISI components (Category R-A)
              " Added new components to schedule table for Class MC Supports
              " Updated schedule to current exam and scheduling status xi

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN TABLE OF CONTENTS Page R ecord of R evi ion.......................................... ns ......................................... . Table of Contents ............................................................... 1.1-1 Intro d uctio n ................................................................................................ 1.2 -1 S ource Docum ents ..................................................................................... 1.3-1 S ection X I C ode C ases .............................................................................. 1.4-1 R eq uests fo r R e lief .................................................................................... 1.5-1 ISI Boundary/Isometric Drawings .............................................................. 1.6-1 Inspection Plan and Schedule Tables (Pages 1 to 332) ............................ 1.7-1 1.1-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN INTRODUCTION

Background:

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code (hereafter referred to as ASME Section XI, Section XI, or the Code), Section XI Inservice Inspection (ISI) Program is prepared and maintained by the Northern States Power Company - Minnesota (NSPM). The Inservice Testing Program (IST) is maintained separately from this program and is submitted under separate cover. The Containment Inspection Program, as allowed by 10CFR55a(g)(6)(ii)(B), and the Repair/Replacement Program are maintained separately from this program, and, although they are not submitted, they are available at the plant site for audit and review. The Snubber Program and Boiling Water Reactor Internals Project (BWRVIP) Program are also maintained separately from this plan. 4th Ten-Year Interval: The Monticello 4th Ten-Year Inservice Inspection Interval is slightly less than 120 months due to an extension of the 3rd Interval (Letters to the NRC in May 2002 and January 2003 providing notification of 3rd Interval extension initially through March 8, 2003 (M2002057) and subsequently through May 31, 2003 (L-MT 004). The 4th Interval overlapped the 3rd Interval as permitted by IWA-2430(d)(1),(2),(3), and (4) The 4th Interval start date is May 1, 2003 and end date is May 31, 2012. Five refueling outages are currently scheduled in this time frame. Component Selection: With the exception of Class 1 and 2 piping welds, components within the examination plan were selected and scheduled using criteria in the 1995 Edition of ASME Section XI with the 1996 Addenda (Inspection Program B and Code Case N-598) and 10CFR50.55a(g)(6)(ii)(A), except where relief has been requested. Per 10CFR50.55a(b)(2)(xxi)(B) reused CRD Bolting must meet examination requirements for Table IWB-2500-1, Category B-G-2, Item B7.80 of ASME Section Xl 1995 Edition. Selection of Class 1 and Class 2 piping welds in ASME Categories B-F, B-J, C-F-1 and C-F-2 are based on EPRI Topical Report 112657 Rev. B-A. "Revised Risk Informed Inservice Inspection Evaluation Procedure." The Risk Informed Class 1 and Class 2 application was also conducted in a manner consistent with ASME Code Case N-578 "Risk Informed Requirements for Class 1, 2, and 3 Piping, Method B." The use of the RI-ISI program was approved for use on July 27, 2002. (reference TAC MB3819 and Relief Request #1 for 4th ISI Interval) 1.2-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN INTRODUCTION (cont'd) Code Edition Summary: The code editions implemented in the ISI Program can be summarized as follows: Class 1 (Quality Group A) 1995 Edition with 1996 Addenda, Risk-Informed Program for Class 1 Piping Category B-F and B-J (Relief #1), Class 1 CRD Bolting (B7.80) 10CFR50.55a(b)(2)(xxi)(B) specifies 1995 Edition for examination requirements of reused CRD Bolting Class 2 (Quality Group B) 1995 Edition with 1996 Addenda, Risk-Informed Program for Class 2 Piping Category C-F-1 and C-F-2 (Relief #1) Class 3 (Quality Group C) 1995 Edition with 1996 Addenda MC (Metal Containment) 2001 Edition with 2003 Addenda, Subsection IWE Appendix VIII - Mandatory 1995 Edition with 1996 Addenda through 2001 Edition as modified by 10CFR50.55a dated September 26, 2002 and subsequently published Final Rules Repair / Replacement and 2001 Edition with No Addenda per ISI Relief associated Pressure Test Request No.7. NRC exception: must use IWA-4540(c) of the 1998 edition in lieu of the 2001 Edition requirement. Additional restrictions for the 2001 Edition include prohibited use of IWA-4340 (10CFR50.55a(b)(2)(xxv)) and IWA-2220 (10CFR50.55a(b)(2(xxii)) 1.2-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN INTRODUCTION (cont'd) Background for Plan/Schedule Development: The examination plan and schedule was developed from ASME Code requirements, Risk-Informed Methodology, individual component examination history and plant scheduling needs such as optimizing insulation removal and scaffolding needs. During the 2nd Interval, a substantial number of component replacements and alterations were made (e.g. the recirculation piping replacement). The intent of the 4th Interval scheduling was to be consistent with the 2nd and 3rd Interval, to the extent practical. For Class 1 (category B-F and B-J) and Class 2 Category C-F-1 and C-F-2) Piping Welds examined per the RI-ISI Plan, there may be little schedule correlation with previous ISI Intervals. Examination Personnel / Procedures: Inservice Inspection examination procedures and personnel certifications meet the requirements specified in the 1995 Edition of ASME Section Xl with the 1996 Addenda. Additionally, UT personnel and procedures meet the requirements of Mandatory Appendix VIII as modified by 10CFR50.55a dated September 26, 2002 and subsequently published Final Rules, except where relief has been granted. Reporting of Associated Section Xl Programs: The Section Xl Repair and Replacement Program, System Pressure Tests and Snubber Functional Tests are administered under separate program documents. Although these programs are administered separately, the activities required by the Repair and Replacement Program, System Pressure Tests and Snubber Functional Tests are reported in the "Inservice Inspection Summary Report" following each refueling outage. ISI Plan Overall

Description:

The ASME Section X1 Inservice Inspection Program is comprised of six parts: Introduction, Source Documents, Requests for Relief, ISI Boundary Drawings, ISI Isometric Drawings, and a table containing the Inservice Inspection Examination Plan and Schedule. The ISI Boundary Drawings outline Quality Group Classifications, (A, B and C). The ISI Isometric Drawings delineate ASME Section XI components or items that are included in the examination program. The Inservice Inspection Examination Plan and Schedule lists the ASME Section XI components by Isometric Drawing Number, System, Code or RI-ISI Category, Code or RI-ISI Item, Component Description and Required Examination. The Examination Plan and Schedule identify the ASME Section XI Item Number listed in Tables IWB-2500-1, IWC-2500-1, IWD-2500-1 and Subsection IWF, and Item Number for Risk Informed Tables as identified in EPRI TR-112657, thus identifying the examination method. The examination schedule lists the 1.2-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN INTRODUCTION (cont'd) anticipated period and outage for the examination of a given component. The examination schedule is intended to be flexible to allow for deviations in outage length and outage work scope. Therefore, the schedule may be changed, as allowed by the Code, without further notification. Examination distribution was developed in accordance with IWA-2432, Inspection Program B. The examination plan and schedule also contains certain non-code items to be examined, or examinations beyond Section Xl Code requirements. These augmented items include licensee-initiated examinations on NC-7879-6/ITank, NC-ISI-51/W-11, W-12,W-13, and NC-ISI-37/W-1, W-2, W-3, W-4, W-12, W-12A shown in the plan and schedule. These items will be examined to the extent practical in accordance with the Section XI Code, 1995 Edition with 1996 Addenda, not the RI-ISI Program. Relief requests will not be submitted for these non-code exams if Section XI Code requirements cannot be met. Non-code exams are also subject to change without prior notification to the NRC. The Monticello Plant was built prior to the implementation of Section XI Access Requirements. As a result, some components that require examination may not be completely accessible. Welds selected for examination under the Risk Informed Program were selected base on risk ranking, radiation area, and weld accessibility as allowed by EPRI TR-1 12657 Rev. B-A. LICENSE RENEWAL AGING MANAGEMENT PLANS AND COMMITMENTS This document supports the implementation of the following Renewed License Aging Management Programs and Commitments:

• PBD/AMP-004, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austentitic Stainless Steel (CASS) Program
• PBD/AMP-022, Primary Containment In-Service Inspection Program
• PBD/AMP-024, ASME Section XI, Subsection IWF
• PBD/AMP-033, ASME Section XI, Inservice Inspection, Subsection IWB, IWC, and IWD
• PBD/AMP-034, Reactor Head Closure Studs
• PBD/AMP-035, BWR Vessel IDAttachment Welds Program
• PBD/AMP-036, BWR Feedwater Nozzle
• PBD/AMP-037, BWR Control Rod Drive Return Nozzle
• PBD/AMP-038, BWR Stress Corrosion Cracking Program
• PBD/AMP-039, BWR Penetrations Program
• NRC Commitments M05008A, M05009A, M0501OA, M05011 A, M05020A, M05021A, and M05022A 1.2-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TI INTERVAL EXAMINATION PLAN CLASS MC SUPPORTS As required by License Renewal Commitment M0501 1A, Class MC Supports will be examined per the requirements of Subsection IWF. The following Class MC Supports are included in an augmented program:

• Torus/ Ring Header Seismic Restraints

         "   Drywell Male and Female Stabilizers

• Shield Stabilizers
• Torus Columns
• Torus Saddles
• Vent System Supports

         "   Downcomer Bracing SMALL BORE CLASS 1 PIPING As required by License Renewal to manage aging effects, examination of Small Bore Piping (Ref. USAR Appendix K, Section K2.1.2) has been added as an augmented program to the ISI Plan. The weld population includes W-1 through W-7 on ISI drawing ISI-786A and W-32 through W-34 on ISI drawing ISI-74215A.

• Augmented volumetric examinations of welds are performed on Class 1 stainless steel small bore piping butt welds > NPS 2 to <

NPS 4. The exams are performed in support of License Renewal and SHALL be performed through the Renewed License period of extended operation.

         " The base scope of approximately 10% of the population will be examined during each ISI interval.

• Examination personnel SHALL be certified to ASME Section Xl, Appendix VIII as modified by 10CFR50.55a.

         " The weld volume applicable to Cat. B-F or B-J will be used for the examination. Welds will be examined to the extent practical. If limitations are encountered that do not permit coverage of essentially 100%, a 10CFR50.55a request (relief request) is not required.

• Welds will be evaluated in accordance with IWB-3000 requirements applicable to Cat. B-F or B-J.

1.2-5

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN FEEDWATER NOZZLES As required by License Renewal Commitments M05020A, M05021A, and M05022A to manage aging effects, Reactor Vessel Feedwater Nozzle examinations will be performed in accordance with the requirements of General Electric Report NE-523-A71-0594-A, Rev.1 (Ref. USAR Appendix K, Section K2.1.8). The report requires volumetric examinations of specified zones in accordance with ASME Section XI Appendix VIII (as modified by 10CFR50.55a). RI-ISI Periodic Update Summary As a condition of NRC approval to implement Risk-Informed ISI (RI-ISI), Monticello made a commitment that 'Risk ranking of piping segments will be reviewed and adjusted on an ASME period basis.' (reference 4 th Interval ISI Relief Request -1 and Commitment M01003A). To meet the commitment, MNGP obtained the Electric Power Research Institute (EPRI) to perform this review in 2007 for activities applicable to Period 1 of the 4 th Interval. The purpose of the review was to determine if the methodology and conclusions applied to the original program implementation are still valid, or to determine if conditions have changed to a degree significant enough to warrant changes to the continued implementation of the RI-ISI program. This review compared the parameters originally used to generate the RI-ISI evaluation against the same or similar parameters as they were at the conclusion of Period 1. Examples of pertinent parameters:

1. Original scope of application compared to current scope,

2. Original PRA model used for consequence evaluation compared to current PRA model.

3. Original key inputs, assumptions, and references from the degradation mechanism (DM) evaluation reviewed to validate continued applicability,

4. Original DM evaluation and failure potential compared to actual service history and examination results, along with industry operating experience (OE) since implementation of RI-ISI to validate continued applicability,

5. Plant design changes (modifications and alterations) implemented since the original RI-ISI evaluation reviewed for possible impact on component RI-ISI classification that would necessitate programmatic changes,

6. Original risk-ranking reviewed using information based on results of 1-5 above, to validate continued applicability, 1.2-6

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

7. Original element selection, including meeting established sampling percentages, reviewed using information from 1-6 above to validate continued conformance with requirements.

The following review items from the report are summarized as follows:

1. The overall scope of the RI-ISI application has not changed: Class 1 and 2, Category B-F, B-J, C-F-i, and C-F-2 piping welds only.

2. The current PRA model is significantly more detailed than the original model, particularly the internal flooding study. However, review of the new PRA information indicates that the original consequence rankings are not impacted by the changes to updated PRA inputs. As a result, no changes to the program are warranted.

3. Service history, examination results, and system characteristics for Period 1 were reviewed to validate the basis of the failure potential and DM evaluations. Two conditions were identified that impacted the pressure boundary:

      " Piping downstream of MO-2008 and MO-2009, examined as part of the site's FAC Program, was confirmed to be susceptible to Erosion-Cavitation. Therefore, the original threshold for E-C susceptibility of
          <100 operational hours/year can no longer be applied to these segments. The threshold change actually has no impact on the program since the segments are already included in, and monitored in accordance with, the FAC Program.

• CRD insert and withdrawal piping crack indications were identified in the 1998 and 2000 outages, prior to implementing RI-ISI. Followup examinations conducted in 2003 did not identify any new issues.

Therefore, there is no impact on the program Review of other inputs into the DM susceptibility was conducted. System design requirements (other than those mentioned in 2 above), operating characteristics, the strategic water chemistry plan implementation, and insulation requirements (RG1.36) have not changed, and there is no new industry OE that was not bounded by the original evaluation. Therefore no program changes are warranted. Enhanced criteria for determining a location's susceptibility to crevice corrosion (C-C) was applied to welds/areas located on three sets of reactor nozzle piping with tuning fork style safe ends: reactor recirculation inlet nozzles (10 welds), core spray inlet nozzles (2 welds), and feedwater inlet nozzles (8 welds). These were originally classified as susceptible to C-C, however the enhanced criteria determined that a mechanical crevice 1.2-7

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN is not present, therefore they are not considered susceptible to C-C. This affected the risk ranking of the welds and their DM susceptibility, moving them from R-A (High) to R-A (Medium), and from DM=C-C to DM=None. Although the risk ranking and DM susceptibility changed, they remained selected for examination to ensure Class 1 weld inspection population continued to meet the requirements of EPRI TR-1 12657 Section 3.6.4.2 (not substantially below 10% of the population).

4. The original risk-ranking results are based on consequence assessment and DM evaluation. Except for the minor changes that had already been made to the program for the Rx Head Spray and CGC system removal and the application of the enhanced C-C criteria to the nozzle safe end locations, there have been no other impacts to the risk-ranking.

Therefore, no further changes are required.

5. Design changes impacted program components, i.e. Reactor (Rx) Head Spray and Combustible Gas Control (CGC) system removals, and replacement of piping downstream of MO-2008, but only effected the population of welds, and had no other impacts on the RI-ISI application.

These changes were reviewed through the site ASME Section Xl Repair/Replacement Program and the RI-ISI Plan was updated accordingly. No additional changes to the program are warranted.

6. The original element selection process ensured that the appropriate percentages were selected for each risk ranking group:

• 25% of R-A (High)
• 10% of R-A (Medium)
• R-A (Low) does not require examination
• Class 1 weld population: if the above criteria do not result in 10% of the Class I weld population being selected, the resulting percentage is not substantially below 10%.

The review concluded that percentage requirements are being met, and no further changes are required. Conclusion of Period 1 Update Review: The Monticello RI-ISI application is a Class 1 and 2 only applications. As such, it is typical that there are little to no changes required of the program as a result of the update process. The EPRI review confirmed this assumption as there were some, but not significant, changes to the RI-ISI program. All necessary changes have been captured through the site's processes and have been incorporated into the RI-ISI Program. Therefore, no further changes are required. 1.2-8

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Source Documents: The following referenced source documents described and listed below are basis documents used and applicable to the Monticello 4th Interval ISI Plan. ASME BPV Code Section XI, 1995 Edition with 1996 Addenda ASME BPV Code Section XI, 1995 Edition with No Addenda ASME BPV Code Section XI, 2001 Edition with No Addenda ASME BPV Code Section XI, 2001 Edition with 2003 Addenda, Subsection IWE 10CFR50.55a, Industry Codes and Standards (69FR58804) 10CFR-50.55a(g)(6)(ii)(A)(64FR51370) ASME Section XI, 1995 Edition with 1996 Addenda, Appendix VIII Supplements 10CFR-50.55a(g)(6)(ii)(A)(66FR16391) ASME Section Xl, 1995 Edition with 1996 Addenda, Appendix VIII Supplement 4 Length Sizing Correction Regulatory Guide 1.147, Rev. 13, Jan 2004 Monticello Inservice Inspection Licensee Control Program, 4 AWI-09.04.00 Monticello ASME Section XI Inservice Inspection Program, EWI-09.04.00 GE Nuclear Services Information Letter, SIL. No. 483R2 "CRD Cap Screw Crack Indications," September 5, 1992 Generic Letter 88-01 & NUREG 0313, Rev 2 (IGSCC (M88080A, M88082A)

       **Note: All Monticello welds meet NUREG-0313, Rev. 2. Category A Monticello Notification Letter to NRC, "Notification of Extension of 3rd Ten-Year Inservice Testing and Inservice Inspection Intervals," May 30, 2002 Monticello Notification Letter to NRC, "Change to Inservice Testing Program Plan and Inservice Inspection Examination Plan 10-Year Intervals," January 23, 2003 Monticello Letter to NRC, "Request for Review and Approval of Relief Requests Associated with Fourth 10-Year Interval Inservice Inspection Examination Plan Submittal," December 6, 2002 1.3-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4T' INTERVAL EXAMINATION PLAN Source Documents: (cont'd) Monticello Letter to NRC, "Monticello Fourth Interval Inservice Inspection Examination Plan, Revision 2," September 16, 2004 NRC Letter, "Monticello Nuclear Generating Plant - Risk-Informed Inservice Inspection Program (TAC MB3819)," July 24, 2002 (Relief Request #1 for 4th ISl Interval) NRC Letter, "MNGP-Evaluation of Relief Request No. 12 (for the 3rd 10-Year IS1 Program Plan," (TAC No. MB0261), July 27, 2001 (4 th Interval ISI Relief Request No.2) NRC Letter, "Relief Request Nos. 3 and 6 for the Fourth 10-Year Interval of the Inservice Inspection Examination Plan" (TAC No. MB6896), March 28, 2003 NRC Letter to Nuclear Management Company, "Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No. 5" (TAC No. MB6956), June 9, 2003 NRC Letter, "Monticello Nuclear Generating Plant - Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No. 7 (TAC NO. MB6897)," October 3, 2003 NRC Letter, "Monticello Nuclear Generating Plant - One-Time Inservice Inspection Program Plan Relief Request No. 8 For Leak Testing The "B" And "G" Main Steam Safety Relief Valves (TAC No. MB9538)," June 13, 2003 NRC Letter, "Duane Arnold Energy Center, Monticello Nuclear Generating Plant, Prairie Island Nuclear Generating Plant, Units 1 and 2, Kewaunee Nuclear Power Plant, Point Beach Nuclear Plant, Units 1 and 2, Palisades Nuclear Plant Re: Request for Alternatives to American Society of Mechanical Engineers (ASME) Section Xl, Appendix VIII, Supplement 10 (TAC NOS. MC0814, MC0816, MC0820, MC0821, MC0815, MC0818, MC0819 AND MC0817)," February 26, 2004 (MNGP ISl Relief Request No.9) NRC Letter, "Duane Arnold Energy Center and Monticello Nuclear Generating Plant Re: Request for Authorization to Utilize Code Case N-613-1 (TAC Nos. MC2374 and MC2375)," October 6, 2004 (MNGP ISI Relief Request No.10) NMC Letter to NRC, "10 CFR 50.55a Request GR-04-01; Request For Authorization To Utilize Code Case N-661 (L-HU-04-027)," July 28, 2004, (Proposed MNGP ISI Relief Request No.11) EPRI Report TR-112657, Rev B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," December 1999 1.3-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Source Documents: (cont'd) NRC Commitment M97025A" Response to LER 97-004, Failure to Submit Relief Requests for Limited Inservice Inspection Examinations, dated March 24, 1997." Limited Examination Relief Requests submitted within 12 months. EPRI Report "Risk-Informed Inservice Inspection (RI-ISI) Update - 2007, Monticello Nuclear Generating Plant," dated January 30, 2008 NUREG-1865 "Safety Evaluation Report, Related to the License Renewal of the Monticello Nuclear Generating Plant, Docket No. 50-263" NMC Letter to NRC "10CFR50.55a Request No. 12: Proposed Alternative for Visual Examination Illumination Levels in Accordance with 10CFR 50.55a(a)(3)(i)," dated August 11, 2005 (Proposed MNGP ISI Relief Request No. 12) NRC letter "Monticello Nuclear Generating Plant - Denial of Alternative for Visual Examination Illumination Levels for the Fourth 10-Year Inservice Inspection Interval (TAC NO. MC8102)," February 8, 2006 (MNGP ISI Relief Request No. 12) NMC Letter to NRC "10CFR 50.55a Request No. 13: Relief from Impractical Examination Coverage Requirements Pursuant to 10CFR 50.55a(g)(5)(iii) for the Fourth Ten-Year Inservice Inspection Interval," dated September 27, 2005. (Proposed MNGP ISI Relief Request No. 13) NRC letter "Monticello Nuclear Generating Plant (MNGP) - Fourth 10-Year Interval Inservice Inspection (ISI) Program Plan, Relief Request No. 13 (TAC NO. MC8882)," dated July 18, 2006. NMC Letter to NRC "Request For Authorization To Utilize Code Case N-513-2," dated December 12, 2005 (Proposed MNGP ISI Relief Request No. 14). NOTE that after this relief request was approved, Code Case N-513-2 was approved in Regulatory Guide 1.147 Rev 15 and therefore this Relief Request is no longer needed. NRC letter "Duane Arnold Energy Center, Monticello Nuclear Generating Plant, Palisades Nuclear Plant, Point Beach Nuclear Plant, Units 1 and 2, Prairie Island Nuclear Generating Plant, Units 1 and 2 - Use of ASME Code Case N-513-2 (TAC NOS. MC9478 Through MC9484)," dated July 3, 2006. (MNGP ISI Relief Request No. 14). 1.3-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN Source Documents: (cont'd) NMC Letter to NRC "10CFR 50.55a Request No. 15: Relief from Impractical Examination Coverage Requirements Pursuant to 10 CFR 50.55a(g)(5)(lll) for the Fourth Ten-Year Inservice Inspection Interval," dated September 26, 2007. (Proposed MNGP ISI Relief Request No. 15). NRC letter "Monticello Nuclear Generating Plant (MNGP) - Granting of Relief Regarding Limited Ultrasonic Examination Coverage of Five Welds (TAC NO. MD6854)," dated May 19, 2008. (MNGP ISI Relief Request No. 15) Xcel Energy Letter to NRC "10 CFR 50.55a Request No. 16: Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations," dated March 12, 2010. (Proposed MNGP ISI Relief Request No. 16) Xcel Energy Letter to NRC "10 CFR 50.55a Request 17: Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for Renewed Operating License Term," dated March 12, 2010. (Proposed MNGP ISI Relief Request No. 17). Xcel Energy Letter to NRC "10 CFR 50.55a Request 18: Alternative to Apply ASME Code Case N-705 to the Standby Liquid Control System Tank", dated April 2, 2010. (Proposed MNGP ISI Relief Request No. 18) NRC Email to Xcel Energy, "Monticello-Record of Conference Call Conveying Verbal Approval of Relief Request No. 18 (TAC-ME 3593)", dated April 23, 2010. Xcel Energy Letter to NRC "10 CFR50.55a Request No. 19: Relief from Impractical Examination Coverage Requirements Pursuant to 10 CFR 50.55a(g)(5)(iii) for the Fourth Ten-Year Inservice Inspection Interval," dated May 6, 2010. (Proposed MNGP ISI Relief Request No. 19). NRC Commitment M05008A (Passport AR 00829849) - MNGP site-specific administrative work instructions will be applicable to both safety and non-safety related systems, structures and components that are subject to an aging management review consistent with the current licensing basis during the period of extended operation. NRC Commitment M05009A (Passport AR 00829851) - Site documents that implement aging management activities for license renewal will be enhanced to ensure that an AR is prepared in accordance with plant procedures whenever non-conforming conditions are found (i.e., the acceptance criteria is not met) 1.3-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN Source Documents: (cont'd) NRC Commitment M05010A (Passport AR 00829853) - Revisions will be made to procedures and instructions that implement or administer aging management programs and/or activities for the purpose of managing the associated aging effects for the duration of extended operation. NRC Commitment M05011A (Passport AR 00829856-01) - The MNGP ASME Section XI, Subsection IWF Program will be enhanced to provide inspections of Class MC components consistent with NUREG-1801, Chapter III, Section B1.3. NRC Commitment M05020A (Passport AR 00829890-01) - The BWR Feedwater Nozzle Program will be enhanced so the parameters monitored and inspected are consistent with the recommendations of GE NE-523-A71-0594-A Revision 1. NRC Commitment M05021A (Passport AR 00829893-01) - The BWR Feedwater Nozzle Program will be enhanced so the regions being inspected; examinations techniques, personnel qualifications, and inspection schedule are consistent with the recommendations of GE NE-523-A71-0594-A, Revision 1. NRC Commitment M05022A (Passport AR 00829895-01) - The BWR Feedwater Nozzle Program will be enhanced so that inspections will be scheduled per recommendations of GE NE-523-a71-0594-A, Revision 1. USAR Appendix K, Renewed Operating License - USAR Supplement, Items (K2.1.33, K2.1.3, K2.1.26, K2.1.2, K2.1.28, K2.1.11, K2.1.8, K2.1.7, K2.1.10, K2.1.9, and K5) NUREG-1865, Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant; dated October 2006 (SER sections 3.03.2.2, 3.0.3.2.3, 3.03.2.6, 3.0.3.2.7, 3.0.3.2.8, 3.0.3.2.9, 3.0.3.2.10, 3.0.3.1.6, 3.0.3.1.7, 3.0.3.1.8) 1.3-5

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: The following listed Code Cases are permissible for use at Monticello during the 4th Interval per Reg. Guide 1.147, Rev. 15. The examination schedule will reflect Code Case implementation on an item or category basis, as applicable. Those that are currently applied are underlined in the listing below. Code Case N-307-3 Revised Ultrasonic Examination Volume for Class I Bolting, Table IWB-2500-1, Examination Category B-G-1, When the Examinations are Conducted from the End of the Bolt or Stud, or from the Center-Drilled Hole Code Case N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds, Section Xl, Division 1 Code Case N-491-2 Rules for Examination of Class 1, 2, 3, and MC Component Supports of Light-Water Cooled Power Plants, Section XI, Division 1 (not used for scheduling - used for evaluation provisions on an as-needed basis) Code Case N-498-4 Alternative Rules for 10 Year System Hydrostatic Testing for Class 1, 2, and 3 Systems. (Applicable to Class 3, Category D-B only) Condition of Use for Prior to conducting the VT-2 examination of Code Case N-498-4 Class 2 and Class 3 components not required to operate during normal plant operation, a 10-minute holding time is required after attaining test pressure. Prior to conducting the VT-2 examination of Class 2 and Class 3 components required, provided the system has been in operation for at least 4 hours for insulated components or 10 minutes for non-insulated components. Code Case N-504-3 Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping. 1.4-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Condition of Use for The provisions of Section XI, Nonmandatory Code Case N-504-3 Appendix Q, "Weld Overlay Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping Weldments, must also be met. Code Case-508-3 Rotation of Serviced Snubbers and Pressure Relief Valves for the Purpose of Testing Code Case N-513-2 Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping (This Code Case was previously approved in Relief Request #14. Relief Request #14 is no longer required) Code Case N-517-1 Quality Assurance Program Requirements for Owners Code Case N-522 Pressure Testing of Containment Penetration Piping (applies to subset of Class 2 piping only) Code Case N-526 Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels Code Case N-528-1 Purchase, Exchange, or Transfer of Material Between Nuclear plant Sites Condition of Use for The requirements of 10 CFR Part 21 are to be Code Case N-528-1 applied to the nuclear plant site supplying the material as well as to the nuclear plant site receiving the material that has been purchased, exchanged, or transferred between sites. Code Case N-532-4 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA 6000 1.4-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Code Case N-534 Alternative Requirements for Pneumatic Pressure Testing Code Case N-537 Location of Ultrasonic Depth Sizing Flaws Code Case N-545 Alternative Requirements for Conduct of Performance Demonstration Detection Test of Reactor Vessel Code Case N-546 Alternative Requirements for Qualification of VT-2 Examination Personnel Condition of Use for This Code Case is applicable only to the Code Case N-546 performance of VT-2 examinations and may not be applied to other VT-2 functions such as verifying the adequacy of procedures and training VT-2 personnel Code Case N-552 Alternative Methods-Qualification for Nozzle Inside Radius Section from the Outside Surface Condition of Use for To achieve consistency with the 10CFR50.55a Code Case N-552 rule change published September 22, 1999, incorporating Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," to Section XI; add the following to the Specimen requirements:

                              "At least 50 percent of the flaws in the demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks. Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches.

Add to detection criteria, "The number of false calls must not exceed three." Code Case N-555 Use of Section II, V, and IXCode Cases 1.4-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Code Case N-566-2 Corrective action for leakage Identified at Bolted Connections. (SBLC System only) Code Case N-578* Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B, Section XI, Division 1 (* Not approved by Reg. Guide 1.147, Rev.15, but referenced by Relief Request #1 on page 1.5-2 for Class 1 and 2 piping welds, Category B-F, B-J, C-F-i, and C-F-2) Code Case N-583 Annual Training Alternative Condition for Use for (1) Supplement practice shall be performed on Code Case N-583 material or welds that contain cracks, or by analyzing prerecorded data from material or welds that contain cracks (2) The training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility Code Case N-586-1 Alternative Additional Examination Requirements for Class 1, 2, and 3 Piping, Components, and Supports Code Case N-592 ASNT Central Certification Program Code Case N-597-2 Requirements for Analytical Evaluation of Pipe Wall Thinning Condition of Use for (lengthy list of conditions for use, see Reg. Code Case N-597-1 Guide 1.147 for conditions) Code Case N-598 Alternative Requirements to Required Percentages of examinations (Applied to exam categories in Tables IWB-2412-1, IWC-2412-1, IWD-2412-1, IWE-2412-1, and IWF-2410-2 with exceptions noted in subparagraph (a)) 1.4-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Code Case N-601 Extent of Frequency of VT-3 Visual Examination for Inservice Inspection of Metal Containments (Applied to IWE Program, approved for use per Relief Request MC-7) Code Case N-606-1 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stub tube repairs Condition of Use for Prior to welding, an examination or verification Code Case N-606-1 must be performed to ensure proper preparation of the base metal, and that the surface is properly contoured so that an acceptable weld can be produced. The surfaces to be welded, and surfaces adjacent to the weld, are to be free from contaminants, such as, rust, moisture, grease, and other foreign material or any other condition that would prevent proper welding and adversely affect the quality or strength of the weld. This verification is to be required in the welding procedures. Code Case N-613-1* Ultrasonic Examination of Full Penetration Nozzles in Vessels, Exam Cat. B-D, Item No. B3.90, Reactor Nozzle-to-Vessel Welds, Figs. IWB-2500-7(a), (b), and (c) (This code case was previously approved in Relief Request

                              #10. Relief Request #10 is no longer required).

Code Case N-624 Successive Inspections Code Case N-638 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique 1.4-5

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Code Case N-639 Alternative Calibration Block Material Condition of Use for Chemical ranges of the calibration block may Code Case N-639 vary from the materials specification if (1) it is within the chemical range of the component specification to be inspected, and (2) the phase and grain shape are maintained in the same ranges produced by the thermal process required by the material specification. Code Case N-640 Alternative reference Fracture Toughness for Development of P-T Limit Curves Code Case N-652-1 Alternative Requirements to Categorize B-G-I, B-G-2, and C-D Bolting Examination Methods and Selection Criteria Code Case N-661 Alternative Requirements for Wall Thickness Restoration of Classes 2 and 3 Carbon Steel Piping for Raw Water Service (This Code Case was previously approved under Relief Request

                               #11. Relief Request #11 is no longer required).

Condition of Use for (a) If the root cause of the degradation has not Code Case N-661 been determined, the repair is only acceptable for one cycle. (b) Weld overlay repair of an area can only be performed once in the same location. (c) When through-wall repairs are made by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage Code Case N-685 Lighting Requirements for Surface Examination Code Case N-700 Alternative Rules for Selection of Classes 1, 2, and 3 Vessel Welded Attachments for Examination 1.4-6

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T' INTERVAL EXAMINATION PLAN Section XI Code Cases: (cont'd) Code Case N-702* Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds (Not approved in Reg. Guide 1.147, Rev.15, but referenced by Relief Request #16, submitted, but not yet approved) Code Case N-705* Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks (Not approved in Reg. Guide 1.147, Rev. 15, but referenced by Relief Request #18, submitted, but not yet approved) 1.4-7

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Requests for Relief Relief Request No. Description Rev. 1* Risk Informed Inservice Inspection Plan 0 (Approved July 24, 2002 for 4th Interval) 2** Reactor Vessel Circumferential Welds 1 (Approved July 27, 2001 for remainder of current 40-Year Operating License through Sept. 8, 2010) 3 Appendix VIII Supplement 4 0 (Approved March 28, 2003) 4 Leakage at Bolted Control Rod Drive (CRD) Housing 0 Connections (Approved June 9, 2003) 5 Leakage at Bolted Control Rod Drive (CRD) Housing 0 Connections (Approved June 9, 2003) 6 Appendix VII Annual Training 0 (Approved March 28, 2003) 7 Use of 2001 Edition for Repair/Replacement Program 0 (Approved October 3, 2003) 8 One-time Relief, Class 1 Pressure Test at Less Than System 0 Operating Pressure, Mechanical Joint (Approved June 13, 2003) 9 Use of Alternative Requirements for Appendix VIII, Supplement 0 10 as implemented by PDI (NMC Fleet Relief Request) (Approved February 26, 2004) 10 Use of Code Case N-613-1 (NMC Fleet Relief Request) 0 (Approved October 6, 2004) (Relief Request is no longer needed. Code Case approved for use by NRC in Reg. Guide 1.147 Rev. 15) 11 Use of N-661 (NMC Fleet Relief Request) 0 (Approved March 8, 2005) (Relief Request is no longer needed. Code Case approved for use by NRC in Reg. Guide 1.147 Rev. 15) 1.5-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Requests for Relief Relief Request No. Description Rev. 12 Alternative for Visual Examination Illumination Levels 0 (Denial February 8, 2006) 13 Relief from Impractical Examination Coverage Requirements 0 (Approved July 18, 2006) 14 Authorization to Utilize Code Case N-513-2 (Approved July 3, 0 2006) (Relief Requests is no longer needed. Code Case approved for use by NRC in Reg. Guide 1.147 Rev. 15) 15 Relief from Impractical Examination Coverage Requirements 0 (Approved May 19, 2008) 16 Alternative to Nozzle-to-Vessel Weld and Inner Radius 0 Examinations, Use of Code Case N-702 (Proposed, submitted March 12, 2010) 17 Extension of Permanent Relief from Volumetric Exanimation of 0 Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term (Proposed, Submitted March 12, 2010) 18 Alternative to Apply ASME Code Case N-705 to the Standby 0 Liquid Control System Tank (Proposed, submitted April 2, 2010) (Verbal Approval of Relief Request No. 18 (TAC-ME 3593)", dated April 23, 2010 ) 19 Relief from Impractical Examination Coverage Requirements 0 (Proposed, submitted May 6, 2010) Relief No. 1 was approved for the 4th ISI Interval for implementation on the start date of the 4 th ISI Interval.

• • Relief No. 2 was approved during the 3 rd ISI Interval and is approved for the remaining time in the current operating license, including the 4th ISI Interval.

It has been revised slightly to correct a weldname nomenclature error and update commitment statements made in Rev. 0. 1.5-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 1 (Rev. 0) Risk Informed Inservice Inspection Plan System: Various Class: 1 and 2 Category: B-F Item: ALL B-J ALL C-F-1 ALL C-F-2 ALL Alternative Examination Reauirements: Monticello has implemented Risk Informed Inservice Inspection program for Class 1 and Class 2 piping welds in accordance with EPRI Topical Report TR-112657 Rev. B-A, Final Report, December 1999. Basis for Relief: See attached Risk Informed Program Plan Submittal Rev. 0. Status: Approved July 24, 2002. NRC Letter, "Monticello Nuclear Generating Plant - Risk-Informed Inservice Inspection Program (TAC MB3819)" 1.5-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN RISK-INFORMED INSERVICE INSPECTION PROGRAM PLAN MONTICELLO NUCLEAR GENERATING PLANT - REVISION 0 Table of Contents

1. Introduction 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 1.2 PSA Quality

2. Proposed Alternative to Current Inservice Inspection Programs 2.1 ASME Section XI 2.2 Augmented Programs

3. Risk-Informed ISI Process 3.1 Scope of Program 3.2 Consequence Evaluation 3.3 Failure Potential Assessment 3.4 Risk Characterization 3.5 Element and NDE Selection 3.5.1 Additional Examinations 3.5.2 Program Relief Requests 3.6 Risk Impact Assessment 3.6.1 Quantitative Analysis 3.6.2 Defense-in-Depth

4. Implementation and Monitoring Program

5. Proposed ISI Program Plan Change

6. References/Documentation 1.5-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

1. INTRODUCTION The Monticello Nuclear Generating Plant (MNGP) is nearing the end of its 3rd Inservice Inspection (ISI) Interval as defined by the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Section Xl Code for Inspection Program B. MNGP plans to implement a Risk-Informed Inservice Inspection (RI-ISI) Program concurrent with the start of the 4th ISI interval, which will begin on June 1, 2002. Pursuant to 10 CFR 50.55a(g)(4)(ii), the applicable ASME Section XI Code for the 4th ISI interval will be the 1995 Edition through 1996 Addenda.

The objective of this submittal is to request the use of a risk-informed process for the inservice inspection of Class 1 and 2 piping. The risk-informed inservice inspection (RI-ISI) process used in this submittal is described in Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A "Revised Risk-Informed Inservice Inspection Evaluation Procedure." The RI-ISI application was also conducted in a manner consistent with ASME Code Case N-578 "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B." 1.1 Relation to NRC Regulatory Guides 1.174 and 1.178 As a risk-informed application, this submittal meets the intent and principles of Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" and Regulatory Guide 1.178, "An Approach for Plant-Specific Risk-Informed Decision-making Inservice Inspection of Piping." Further information is provided in Section 3.6.2 relative to defense-in-depth. 1.2 PSA Quality The Monticello Level 1 and Level 2 Probabilistic Safety Assessment (PSA) results that are based on the January 1999 update were used to evaluate the consequences of pipe ruptures for the RI-ISI assessment during power operation. The base PSA Core Damage Frequency (CDF) is 1.5E-5 events per year and the base PSA Large Early Release Frequency (LERF) is 5.5E-7 events per year for the 1999 update. The original IPE result was a CDF of 2.6E-5, which was reported to the NRC in 1992. The PSA model update history is discussed below. The NRC review of the Monticello Individual Plant Examination (IPE) was issued in May 1994. The Staff Evaluation Report (SER) concluded the following regarding the Monticello IPE: 1.5-5

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

• The IPE is complete with respect to the information requested in Generic Letter 88-20 and associated Supplement 1;
• The IPE analytical approach is technically sound and capable of identifying plant-specific vulnerabilities;

    " Monticello employed a viable means to verify that the IPE models reflect the current plant design and operation at the time of submittal to the NRC;
    " The IPE had been peer-reviewed;

• Monticello participated in the IPE process;
• The IPE specifically evaluated the Monticello decay heat removal functions for vulnerabilities;

    " Monticello had responded appropriately to the Containment Performance Improvement program recommendations.

There were no areas of improvement to the PSA model that were identified by the NRC in their review of the plant's IPE submittal. The internal events PSA used for the RI-ISI evaluation is based on a more current version of the PSA than the version used for the IPE. The PSA model was updated in 1994, 1995 and 1999. The major differences in the PSA model between the original IPE and the PSA updates through the 1995 update are that the updated model includes the following:

• Addition of a non-safety 480kv diesel generator that can backfeed through emergency bus 15 to supply battery charges;
• Installation of a hard piped vent that provides an additional means for containment heat removal;
• Improvements to safety relief valve pneumatics (including power supplies);
• Addition of a crosstie for alignment of the diesel fire pump as an additional source of low pressure makeup water; 1.5-6

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

     "    Replacement of an instrument air compressor with one that is not dependent on service water;

• More realistic success criteria for service water by changing from 2 of 3 pumps required for success to 1 of 3 pumps required for success;
• Internal floods initiating event frequency and effects were updated.

The 1999 PSA update was performed to incorporate the effects of power uprate conditions. In 1997, a BWROG PSA Peer Certification Review was performed on the 1995 update PSA model. The overall conclusion was positive and said that the Monticello PSA can be effectively used to support applications involving relative risk significance. The "Facts and Observations" for Monticello have been evaluated, and are being addressed by the Monticello PSA Program. No substantial changes to the RI-ISI consequence conclusions are anticipated due to planned PSA model revisions to address these "Facts and Observations."

2. PROPOSED ALTERNATIVE TO CURRENT ISI PROGRAM REQUIREMENTS 2.1 ASME Section XI ASME Section Xl Examination Categories B-F, B-J, C-F-1 and C-F-2 currently contain the requirements for the nondestructive examination (NDE) of Class 1 and 2 piping components. The alternative RI-ISI program for piping is described in EPRI TR-1 12657. The RI-ISI program will be substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1 and C-F-2) in accordance with 10 CFR 50.55a(a)(3)(i) by alternatively providing an acceptable level of quality and safety. Other non-related portions of the ASME Section Xl Code will be unaffected. EPRI TR-112657 provides the requirements for defining the relationship between the RI-ISI program and the remaining unaffected portions of ASME Section Xl.

1.5-7

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN 2.2 Augmented Programs The following augmented inspection programs were considered during the RI-ISI application:

     " The augmented inspection program for flow accelerated corrosion (FAC) per Generic Letter 89-08 is relied upon to manage this damage mechanism but is not otherwise affected or changed by the RI-ISI program.

• The augmented inspection program for intergranular stress corrosion cracking (IGSCC) as addressed in NRC Generic Letter 88-01 and NUREG-0313, Rev. 2, have been resolved by Monticello's pipe replacement program wherein all susceptible material was replaced with resistant material. All welds are therefore classified as IGSCC Category "A". In accordance with EPRI TR-1 12657, piping welds identified as Category "A" are considered resistant to IGSCC, and as such are assigned a low failure potential provided no other damage mechanisms are present. Examination criteria for these welds will be in accordance with the RI-ISI process.

    " The augmented inspection program for High Energy Line Break (HELB) piping includes 36 Class 1 welds that are classified as ASME Section XI, Examination Category B-J. Although MNGP is not committed to using the NUREG-0800 Standard Review Plan (SRP),

Sections 3.6.1 and 3.6.2 of the SRP are used as guidance in determining appropriate design and examination requirements for specified high energy piping. The 36 Class 1 welds that require examination in accordance with the HELB augmented inspection program are between the containment penetration and the outboard isolation valve in the main steam, high pressure coolant injection, reactor core isolation cooling, reactor water clean-up, residual heat removal and core spray systems. Independent of the HELB program, the RI-ISI application selected 8 of these 36 HELB welds for examination. The remaining 28 HELB welds will continue to be examined in accordance with the HELB augmented inspection program. 1.5-8

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

3. RISK-INFORMED ISI PROCESS The process used to develop the RI-ISI program conformed to the methodology described in EPRI TR-1 12657 and consisted of the following steps:

• Scope Definition
• Consequence Evaluation
• Failure Potential Assessment
• Risk Characterization
• Element and NDE Selection
• Risk Impact Assessment
• Implementation Program
• Feedback Loop A deviation to the EPRI RI-ISI methodology has been implemented in the failure potential assessment for MNGP. Table 3-16 of EPRI TR-1 12657 contains criteria for assessing the potential for thermal stratification, cycling and striping (TASCS).

Key attributes for horizontal or slightly sloped piping greater than 1" nominal pipe size (NPS) include:

1. Potential exists for low flow in a pipe section connected to a component allowing mixing of hot and cold fluids, or

2. Potential exists for leakage flow past a valve, including in-leakage, out-leakage and cross-leakage allowing mixing of hot and cold fluids, or

3. Potential exists for convective heating in dead-ended pipe sections connected to a source of hot fluid, or

4. Potential exists for two phase (steam/water) flow, or

5. Potential exists for turbulent penetration into a relatively colder branch pipe connected to header piping containing hot fluid with turbulent flow, AND AT > 50*F, AND Richardson Number> 4 (this value predicts the potentialbuoyancy of a stratifiedflow) 1.5-9

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN These criteria, based on meeting a high cycle fatigue endurance limit with the actual AT assumed equal to the greatest potential AT for the transient, will identify all locations where stratification is likely to occur, but allows for no assessment of severity. As such, many locations will be identified as subject to TASCS where no significant potential for thermal fatigue exists. The critical attribute missing from the existing methodology that would allow consideration of fatigue severity is a criterion that addresses the potential for fluid cycling. The impact of this additional consideration on the existing TASCS susceptibility criteria is presented below. > Turbulent penetration TASCS Turbulent penetration typically occurs in lines connected to piping containing hot flowing fluid. In the case of downward sloping lines that then turn horizontal, significant top-to-bottom cyclic ATs can develop in the horizontal sections if the horizontal section is less than about 25 pipe diameters from the reactor coolant piping. Therefore, TASCS is considered for this configuration. For upward sloping branch lines connected to the hot fluid source that turn horizontal or in horizontal branch lines, natural convective effects combined with effects of turbulence penetration will keep the line filled with hot water. If there is no potential for in-leakage towards the hot fluid source from the outboard end of the line, this will result. in a well-mixed fluid condition where significant top-to-bottom ATs will not occur. Therefore TASCS is not considered for these configurations. Even in fairly long lines, where some heat loss from the outside of the piping will tend to occur and some fluid stratification may be present, there is no significant potential for cycling as has been observed for the in-leakage case. The effect of TASCS will not be significant under these conditions and can be neglected.

>       Low flow TASCS In some situations, the transient startup of a system (e.g., RHR suction piping) creates the potential for fluid stratification as flow is established. In cases where no cold fluid source exists, the hot flowing fluid will fairly rapidly displace the cold fluid in stagnant lines, while fluid mixing will occur in the piping further removed from the hot source and stratified conditions will exist only briefly as the line fills with hot fluid. As such, since the situation is transient in nature, it can be assumed that the criteria for thermal transients (TT) will govern.

> Valve leakage TASCS 1.5-10

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Sometimes a very small leakage flow of hot water can occur outward past a valve into a line that is relatively colder, creating a significant temperature difference. However, since this is a generally a "steady-state" phenomenon with no potential for cyclic temperature changes, the effect of TASCS is not significant and can be neglected. > Convection heating TASCS Similarly, there sometimes exists the potential for heat transfer across a valve to an isolated section beyond the valve, resulting in fluid stratification due to natural convection. However, since there is no potential for cyclic temperature changes in this case, the effect of TASCS is not significant and can be neglected. In summary, these additional considerations for determining the potential for thermal fatigue as a result of the effects of TASCS provide an allowance for the consideration of cycle severity in assessing the potential for TASCS effects. The above criteria have previously been submitted by EPRI for generic approval (Letter dated February 28, 2001, P.J. O'Regan (EPRI) to Dr. B. Sheron (USNRC), "Extension of Risk-Informed Inservice Inspection Methodology"). 3.1 Scope of Program The systems included in the RI-ISI program are provided in Table 3.1. The piping and instrumentation diagrams and additional plant information including the existing plant ISI program, were used to define the Class 1 and 2 piping system boundaries. 3.2 Consequence Evaluation The consequence(s) of pressure boundary failures were evaluated and ranked based on their impact on core damage and containment performance (i.e., isolation, bypass and large early release). The impact on these measures due to both direct and indirect effects was considered using the guidance provided in EPRI TR-1 12657. 3.3 Failure Potential Assessment Failure potential estimates were generated utilizing industry failure history, plant specific failure history, and other relevant information. These failure estimates were determined using the guidance provided in EPRI TR-112657, with the exception of the previously stated deviation. 1.5-11

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.3 summarizes the failure potential assessment by system for each degradation mechanism that was identified as potentially operative. 3.4 Risk Characterization In the preceding steps, each run of piping within the scope of the program was evaluated to determine its impact on core damage and containment performance (i.e., isolation, bypass and large, early release) as well as its potential for failure. Given the results of these steps, piping segments are then defined as continuous runs of piping potentially susceptible to the same type(s) of degradation and whose failure will result in similar consequence(s). Segments are then ranked based upon their risk significance as defined in EPRI TR-112657. The results of these calculations are presented in Table 3.4. 3.5 Element and NDE Selection In general, EPRI TR-1 12657 requires that 25% of the locations in the high risk region and 10% of the locations in the medium risk region be selected for inspection using appropriate NDE methods tailored to the applicable degradation mechanism. In addition, per Section 3.6.4.2 of EPRI TR-112657, if the percentage of Class 1 piping locations selected for examination falls substantially below 10%, then the basis for selection needs to be investigated. For MNGP, the percentage of Class 1 welds selected per the RI-ISI process is 9.3% (76 of 817 welds), which is not a significant departure from 10%. One additional factor that was considered during the evaluation was that the overall percentage of Class 1 selections included both socket and non-socket welds. Therefore, the percentage of Class 1 selections was 9.3% when both socket and non-socket piping welds were considered. This percentage increases to 13.2% (75 of 567 welds) when considering only those piping welds that are non-socket welded. It should be noted that non-socket welds are subject to volumetric examination, so this percentage does not rely upon welds that are solely subject to a VT-2 visual examination. As stated in TR-1 12657, the existing FAC augmented inspection program provides the means to effectively manage this mechanism. No additional credit was taken for any FAC augmented inspection program locations beyond those selected by the RI-ISI process to meet the sampling percentage requirements. 1.5-12

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN A brief summary is provided below, and the results of the selection are presented in Table 3.5. Section 4 of EPRI TR-1 12657 was used as guidance in determining the examination requirements for these locations. Unit Class I Piping Welds(1 ) Class 2 Piping Welds(2) II All Piping Welds(3) Total Selected Total Selected Total Selected 1 817 76 901 12 1718 88 Notes

1. Includes all Category B-F and B-J locations.

2. Includes all Category C-F-1 and C-F-2 locations.

3. All in-scope piping components, regardless of risk classification, will continue to receive Code required pressure testing, as part of the current ASME Section Xl program. VT-2 visual examinations are scheduled in accordance with the station's pressure test program that remains unaffected by the RI-ISI program.

3.5.1 Additional Examinations The RI-ISI program in all cases will determine through an engineering evaluation the root cause of any unacceptable flaw or relevant condition found during examination. The evaluation will include the applicable service conditions and degradation mechanisms to establish that the element(s) will still perform their intended safety function during subsequent operation. Elements not meeting this requirement will be repaired or replaced. The evaluation will include whether other elements in the segment or additional segments are subject to the same root cause conditions. Additional examinations will be performed on those elements with the same root cause conditions or degradation mechanisms. The additional examinations will include high risk significant elements and medium risk significant elements, if needed, up to a number equivalent to the number of elements required to be inspected on the segment or segments during the current outage. If unacceptable flaws or relevant conditions are again found similar to the initial problem, the remaining elements identified as susceptible will be examined. No additional examinations will be performed if there are no additional elements identified as being susceptible to the same root cause conditions. 3.5.2 Program Relief Requests 1.5-13

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4T" INTERVAL EXAMINATION PLAN An attempt has been made to select RI-ISI locations for examination such that a minimum of >90% coverage (i.e., Code Case N-460 criteria) is attainable. However, some limitations will not be known until the examination is performed, since some locations may be examined for the first time by the specified techniques. In instances where locations are found at the time of the examination that do not meet the >90% coverage requirement, the process outlined in EPRI TR-1 12657 will be followed. None of the existing MNGP relief requests are being withdrawn due to the RI-ISI application. 3.6 Risk Impact Assessment The RI-ISI program has been conducted in accordance with Regulatory Guide 1.174 and the requirements of EPRI TR-1 12657, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements. This evaluation identified the allocation of segments into High, Medium, and Low risk regions of the EPRI TR-112657 and ASME Code Case N-578 risk ranking matrix, and then determined for each of these risk classes what inspection changes are proposed for each of the locations in each segment. The changes include changing the number and location of inspections within the segment and in many cases improving the effectiveness of the inspection to account for the findings of the RI-ISI degradation mechanism assessment. For example, for locations subject to thermal fatigue, examinations will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process. 3.6.1 Quantitative Analysis Limits are imposed by the EPRI methodology to ensure that the change in risk of implementing the RI-ISI program meets the requirements of Regulatory Guides 1.174 and 1.178. The EPRI criterion requires that the cumulative change in core damage frequency (CDF) and large early release frequency (LERF) be less than 1E-07 and 1E-08 per year per system, respectively. 1.5-14

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello conducted a risk impact analysis per the requirements of Section 3.7 of EPRI TR-1 12657. The analysis estimates the net change in risk due to the positive and negative influence of adding and removing locations from the inspection program. A risk quantification was performed using the "Simplified Risk Quantification Method" described in Section 3.7 of EPRI TR-112657. The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) used for high consequence category segments was based on the highest evaluated CCDP (9E-03) and CLERP (9E-03), whereas, for medium consequence category segments, bounding estimates of CCDP (1E-04) and CLERP (1E-05) were used. The likelihood of pressure boundary failure (PBF) is determined by the presence of different degradation mechanisms and the rank is based on the relative failure probability. The basic likelihood of PBF for a piping location with no degradation mechanism present is given as x0 and is expected to have a value less than 1E-08. Piping locations identified as medium failure potential have a likelihood of 20x0 . In addition, the analysis was performed both with and without taking credit for enhanced inspection effectiveness due to an increased POD from application of the RI-ISI approach. The PBF likelihoods and POD values used in the analysis are consistent with those used in the approved RI-ISI pilot applications at Arkansas Nuclear One, Unit 2, and Vermont Yankee, as documented in References 9 and 14 of EPRI TR-1 12657. Table 3.6-1 presents a summary of the RI-ISI program versus ASME Section Xl Code requirements and identifies on a per system basis each applicable risk category. The presence of FAC was adjusted for in the performance of the quantitative analysis by excluding its impact on the risk ranking. However, in an effort to be as informative as possible, for those systems where FAC is present, Table 3.6-1 presents the information in such a manner as to depict what the resultant risk categorization is both with and without consideration of FAC. This is accomplished by enclosing the FAC damage mechanism, as well as all other resultant corresponding changes (failure potential rank, risk category and risk rank), in parenthesis. Again, this has only been done for information purposes, and has no impact on the assessment itself. The use of this approach to depict the impact of degradation mechanisms managed by augmented inspection programs on the risk categorization is consistent with that used in the delta risk assessment for the Arkansas Nuclear One, Unit 2 pilot application. An example is provided below. 1.5-15

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Risk Consequence Failure Potential Category Rank(') Rank DMs Rank In this example if FAC is not considered, the failure potential rank is "medium" instead of "high" based on the TASCS and TT damage mechanisms. When a "medium" failure potential rank is combined with a "medium" consequence rank, it results in risk category 5 ("medium" risk) being assigned instead of risk category 3 ("high" risk). FWV 5 (3) Medium (High) Medium TASCS, TT, (FAC): Medium (High) In this example if FAG were considered, the failure potential rank would be "high" instead of "medium". If a "high" failure potential rank were combined with a "medium" consequence rank, it would result in risk category 3 ("high" risk) being assigned instead of risk category 5 ("medium" risk). Note

1. The risk rank is not included in Table 3.6-1 but it is included in Table 5-2.

1.5-16

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T' INTERVAL EXAMINATION PLAN As indicated in the table below, this evaluation has demonstrated that unacceptable risk impacts will not occur from implementation of the RI-ISI program, and satisfies the acceptance criteria of Regulatory Guide 1.174 and EPRI TR-112657. Risk Impact Results ARiskcDF ARiSkLERF System(w) w/ POD w/o POD w/ POD w/o POD RPV 9.OOE-11 9.OOE-11 9.OOE- 11 9.OOE-11 RWCU 4.50E-11 4.50E-11 4.50E-11 4.50E-11 MS 9.90E-10 9.90E-10 9.90E-10 9.90E-10 SLC -4.50E-11 -4.50E-11 -4.50E-11 -4.50E-11 RCR 6.98E-09 6.98E-09 6.98E-09 6.98E-09 RClC -1.38E-10 -1.10E-10 -9.48E-1 1 -9.20E-1 1 RHR -9.71 E-09 -2.13E-09 -9.72E-09 -2.16E-09 cs 1.22E-09 1.22E-09 1.22E-09 1.22E-09 HPCI -6.15E-10 2.69E-09 -5.88E-10 2.66E-09 FW -6.20E-09 3.90E-09 -6.17E-09 3.91 E-09 ccW negligible negligible negligible negligible CRD negligible negligible negligible negligible FPEC no change no change no change no change PCAC negligible negligible negligible negligible Torus negligible negligible negligible negligible Total -7.40E-09 1.36E-08 -7.30E-09 1.36E-08 Note

1. Systems are described in Table 3.1.

1.5-17

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN 3.6.2 Defense-in-Depth The intent of the inspections mandated by ASME Section Xl for piping welds is to identify conditions such as flaws or indications that may be precursors to leaks or ruptures in a system's pressure boundary. Currently, the process for picking inspection locations is based upon structural discontinuity and stress analysis results. As depicted in ASME White Paper 92-01-01 Rev. 1, "Evaluation of Inservice Inspection Requirements for Class 1, Category B-J Pressure Retaining Welds," this method has been ineffective in identifying leaks or failures. EPRI TR-1 12657 and Code Case N-578 provide a more robust selection process founded on actual service experience with nuclear plant piping failure data. This process has two key independent ingredients, that is, a determination of each location's susceptibility to degradation and secondly, an independent assessment of the consequence of the piping failure. These two ingredients assure defense in depth is maintained. First, by evaluating a location's susceptibility to degradation, the likelihood of finding flaws or indications that may be precursors to leak or ruptures is increased. Secondly, the consequence assessment effort has a single failure criterion. As such, no matter how unlikely a failure scenario is, it is ranked High in the consequence assessment, and at worst Medium in the risk assessment (i.e., Risk Category 4), if as a result of the failure there is no mitigative equipment available to respond to the event. In addition, the consequence assessment takes into account equipment reliability, and less credit is given to less reliable equipment. All locations within the Class 1 and 2 pressure boundaries will continue to receive a system pressure test and visual VT-2 examination as currently required by the Code regardless of its risk classification.

4. IMPLEMENTATION AND MONITORING PROGRAM Upon approval of the RI-ISI program, procedures that comply with the guidelines described in EPRI TR-1 12657 will be prepared to implement and monitor the program. The new program will be integrated into the 4th Inservice Inspection Interval. No changes to the Technical Specifications or Updated Final Safety Analysis Report are necessary for program implementation.

1.5-18

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN The applicable aspects of the ASME Code not affected by this change will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements. Existing ASME Section XI program implementing procedures will be retained and modified to address the RI-ISI process, as appropriate. The monitoring and corrective action program will contain the following elements: A. Identify B. Characterize C. (1) Evaluate, determine the cause and extent of the condition identified (2) Evaluate, develop a corrective action plan or plans D. Decide E. Implement F. Monitor G. Trend The RI-ISI program is a living program requiring feedback of new relevant information to ensure the appropriate identification of high safety significant piping locations. As a minimum, risk ranking of piping segments will be reviewed and adjusted on an ASME period basis. In addition, significant changes may require more frequent adjustment as directed by NRC Bulletin or Generic Letter requirements, or by industry and plant specific feedback.

5. PROPOSED ISI PROGRAM PLAN CHANGE A comparison between the RI-ISI program and ASME Section XI Code 1986 Edition program requirements for in-scope piping is provided in Tables 5-1 and 5-

2. (Since no examination selections had been made for the 4th interval ISI Program prior to the development on the RI-ISI Program, the 3rd Interval selections were used for comparison purposes. The Code of record for the 3rd Interval was the 1986 Edition of ASME Section XI.) Table 5-1 provides a summary comparison by risk region. Table 5-2 provides the same comparison information, but in a more detailed manner by risk category, similar to the format used in Table 3.6-1.

MNGP is implementing the RI-ISI program at the start of the 1st period of its 4th Inspection Interval. As such, 100% of the required RI-ISI program inspections will be completed in the 4th interval. Examinations shall be performed during the interval such that the period examination percentage requirements of ASME Section XI, paragraphs IWB-2412 and IWC-2412 are met. 1.5-19

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

6. REFERENCES/DOCUMENTATION EPRI TR-112657, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," Rev. B-A ASME Code Case N-578, "Risk-Informed Requirements for Class 1, 2, and 3 Piping, Method B, Section XI, Division 1" Regulatory Guide 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis" Regulatory Guide 1.178, "An Approach for Plant-Specific Risk Informed Decision-making Inservice Inspection of Piping" Supporting Onsite Documentation Structural Integrity Calculation/File No. NMC-01-301, "Degradation Mechanism Evaluation for Class 1 and 2 Piping Welds at Monticello Nuclear Generating Plant," Revision 1 Structural Integrity Calculation/File No. NMC-01-302, "Risk-Informed Inservice Inspection Consequence Evaluation of Class 1 and 2 Piping for Monticello Nuclear Power Plant," Revision 1 Structural Integrity Calculation/File No. NMC-01-303, "Risk Ranking Summary, Matrix and Report for the Monticello Nuclear Generating Plant," Revision 0 Structural Integrity Calculation/File No. NMC-01 -304, "Risk Impact Analysis for the Monticello Nuclear Generating Plant," Revision 1 Structural Integrity File No. NMC-01-103-4, Record of Conversation No. ROC-002, "Minutes of the Element Selection Meeting for the Risk-Informed ISI Project at the Monticello Nuclear Generating Plant," Revision 1, June 21, 2001 MNGP Calculation/File No. CA-01-216, "Monticello Nuclear Generating Plant, Risk-Informed Service History Report for Class I and II Piping Welds, ASME Categories B-F, B-J, C-F-1 and C-F-2," Revision 0 1.5-20

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.1 System Selection and Segment / Element Definition System Description Number of Segments Number of Elements RPV - Reactor Pressure Vessel 19 112 RWCU - Reactor Water Clean-Up 10 85 MS - Main Steam 22 204 SLC -Standby Liquid Control 3 35 RCR - Reactor Coolant Recirculation 22 135 RCIC - Reactor Core Isolation Cooling 13 65 RHR - Residual Heat Removal 97 476 CS - Core Spray 36 191 HPCI - High Pressure Coolant Injection 20 158 FW - Feedwater 37 78 CCW - Component Cooling Water 2 18 CRD - Control Rod Drive 7 41 FPEC - Fuel Pool Emergency Cooling 10 54 PCAC - Primary Containment and Atmospheric Control 8 47 Torus - Torus Hard Vent 1 19 Totals 307 1718 NOTE: TABLE 3.2 was not part of the Risk-Informed ISI Program submittal and is intentionally excluded from this document. 1.5-21

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.3 Failure Potential Assessment Summary Thermal Fatigue Stress Corrosion Cracking Localized Corrosion low Sensitive TASCS TT IGSCC TGSCC ECSCC PWSCC MIC PIT CC E-C FAC RPV RWCU MS X SLC RCR X RCIC X X RHR X X CS x x HPCI X FW X X X X CCW CRD FPEC PCAC Torus Note

1. Systems are described in Table 3.1.

1.5-22 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.4 Number of Segments by Risk Category With and Without Impact of FAC High Risk Region Medium Risk Region Low Risk Region System(l) Category 1 Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With Without With Without With Without With Without With Without With Without With Without RPV 6 6 10 10 3 3 RWCU 9 9 1 1 MS 2(2) 0 5 7 14 14 1 1 SLC 1 1 2 2 RCR 10 10 10 10 2 2 RCIC 3(3) 0 2 2 3 6 3 3 2 2 RHR 3 3 15(4) 0 13 13 5(5) 2 44 59 17 20 CS 2 2 1(6) 0 4 4 4(7) 0 6 7 19 23 HPCI 2 2 4 4 3 3 11 11 FW 14(8) 0 14 21 2(9) 0 6 13 1 3 CCW 2 2 CRD 2 2 5 5 FPEC 10 10 PCAC 8 8 Torus 1 1 Total 16 0 31 38 21 0 60 69 16 14 111 127 52 59 Notes

1. Systems are described in Table 3.1.

2. These two segments become Category 4 after FAC is removed from consideration due to no other damage mechanisms being present.

3. These three segments become Category 5 after FAC is removed from consideration due to the presence of other "medium" failure potential damage mechanisms.

4. These fifteen segments become Category 6 after FAC is removed from consideration due to no other damage mechanisms being present.

1.5-23

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN Notes for Table 3.4 (cont'd)

5. Of these five segments, three segments become Category 7 after FAC is removed due to no other damage mechanisms being present.

6. This one segment becomes Category 6 after FAC is removed due to no other damage mechanisms being present.

7. These four segments become Category 7 after FAC is removed due to no other damage mechanisms being present.

8. Of these fourteen segments, seven segments become Category 2 after FAC is removed due to the presence of other "medium" failure potential damage mechanisms, and seven segments become Category 4 after FAC is removed due to no other damage mechanisms being present.

9. These two segments become Category 5 after FAC is removed due to no other damage mechanisms being present.

1.5-24 0 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.5 Number of Elements Selected for Inspection by Risk Category Excluding Impact of FAC High Risk Region Medium Risk Region Low Risk Region t Category 4 Category 5 Category 6 Category 7 System( ) Category 1 Category 2 Category 3 Total Selected Total Selected Total Selected Total [Selected Total Selected Total Selected Total Selected RPV 21 3 83 0 8 0 RWCU 84 9 1 0 MS 105 11(2) 95 0 4 0 SLC 8 1 27 0 RCR 10 3 113 12 12 0 RCIC 12 2 28 3 12 0 13 0 RHR 31 8 67 7 10 1 269 0 99 0 CS 2 1 20 2 35 0 134 0 HPCI 8 2 27 3 33 4 90 0 FW 36 10 38 4(3) 4 2 CCW 18 0 CR0D 10 0 31 0 FPEC 54 0 PCAC 47 0 Torus 19 0 Total 87 24 495 54 75 10 741 0 320 0 Notes

1. Systems are described in Table 3.1.

2. One of these eleven welds was selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for this weld, the FAC examination will be credited toward both programs.

3. Two of these four welds were selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for these welds, the FAC examinations will be credited toward both programs.

1.5-25

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results System(1 ) Category Consequence Failure Potential Inspections CDF Impact(4) LERF Impact(4 ) Rank DMs Rank Section XI(2) RI-ISI(3) Delta w/ POD w/o POD wi POD w/o POD RPV 4 High None Low 5 3 -2 9.OOE-11 9.OOE-11 9.OOE-11 9.OOE-11 RPV 6 Medium None Low 4 0 -4 negligible negligible negligible negligible RPV 7 Low None Low 2 0 -2 negligible negligible negligible negligible RPV Total 9.OOE-11 9.OOE-11 9.OOE-11 9.OOE-11 RWCU 4 High None Low 10 9 -1 4.50E-11 4.50E-11 4.50E-11 4.50E-11 RWCU 7 Low None Low 0 0 0 no change no change no change no change RWCU Total 4.50E-11 4.50E-11 4.50E-11 4.50E-11 MS 4(1) High None(FAC) Low (High) 2 0 -2 9.00E-11 9.O0E-11 9.00E-11 9.OOE-11 MS 4 High None Low 30 10 -20 9.OOE-10 9.OOE-10 9.OOE-10 9.OOE-10 MS 6 Medium None Low 21 0 -21 negligible negligible negligible negligible MS 7 Low None Low 0 0 0 no change no change no change no change MS Total 9.90E-10 9.90E-10 9.90E-10 9.90E-10 SLC 4 High None Low 0 1 1 -4.50E-11 -4.50E-11 -4.50E-11 -4.50E-11 SLC 6 Medium None Low 0 0 0 no change no change no change no change SLC Total -4.50E-11 -4.50E-11 -4.50E-11 -4.50E-11 RCR 2 High CC Medium 10 3 -7 6.30E-09 6.30E-09 6.30E-09 6.30E-09 RCR 4 High None Low 27 12 -15 6.75E-10 6.75E-10 6.75E-10 6.75E-10 RCR 7 Low None Low 0 0 0 no change no change no change no change RCR Total 6.98E-09 6.98E-09 6.98E-09 6.98E-09 1.5-26

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results Inspections CDF Impact(4) LERF Impact(4) System(i) Category Consequence Failure Potential Ran DMs Rank Section Xl(2) Rl-ISl( 3 ) Delta w/POD wlo POD w/POD w/o POD RCIC 4 High None Low 0 2 2 -9.OOE-11 -9.OOE-11 -9.OOE-11 -9.OOE-11 RCIC 5(3) Medium TT, (FAC) Medium (High) 1 1 0 -1.20E-11 no change -1.20E-12 no change RCIC 5 Medium TT Medium 0 2 2 -3.60E-11 -2.00E-11 -3.60E-12 -2.00E-12 RCIC 6 Medium None Low 1 0 -1 negligible negligible negligible negligible RCIC 7 Low None Low 0 0 0 no change no change no change no change RCIC Total -1.38E-10 -1.10E-10 -9.48E-11 -9.20E-11 RHR 2 High TT Medium 5 8 3 -1.03E-08 -2.70E-09 -1.03E-08 -2.70E-09 RHR 4 High None Low 19 7 -12 5.40E-10 5.40E-10 5.40E-10 5.40E-10 RHR 5 Medium TT Medium 4 1 -3 6.00E-12 3.00E-11 6.00E-13 3.00E-12 RHR 6 (3) Medium None (FAC) Low (High) 5 0 -5 negligible negligible negligible negligible RHR 6 Medium None Low 20 0 -20 negligible negligible negligible negligible RHR 7 (5) Low None (FAC) Low (High) 1 0 -1 negligible negligible negligible negligible RHR 7 Low None Low 8 0 -8 negligible negligible negligible negligible RHR Total -9.71 E-09 -2.13E-09 -9.72E-09 -2.16E-09 CS 2 High CC Medium 2 1 -1 9.00E-10 9.00E-10 9.00E-10 9.00E-10 CS 4 High None Low 9 2 -7 3.15E-10 3.15E-10 3.15E-10 3.15E-10 CS 6 (3) Medium None (FAC) Low (High) 0 0 0 no change no change no change no change CS 6 Medium None Low 6 0 -6 negligible negligible negligible negligible CS 7 (5) Low None (FAC) Low (High) 0 0 0 no change no change no change no change CS 7 Low None Low 18 0 -18 negligible negligible negligible negligible CS Total 1.22E-09 1.22E-09 1.22E-09 1.22E-09 1.5-27

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results Failure Potential Inspections CDF Impactj4) LERF Impact(4) System(I) Category Consequence Rank RaCtr ank Section XI(2) RI-ISI(3 ) Delta w/ POD - w/o POD w/ POD w/o POD HPCI 2 High TT Medium 5 2 -3 -5.40E-10 2.70E-09 -5.40E-10 2.70E-09 HPCI 4 High None Low 2 3 1 -4.50E-11 -4.50E-11 -4.50E-11 -4.50E-11 HPCI 5 Medium TT Medium 7 4 -3 -3.OOE-11 3.00E-11 -3.00E-12 3.OOE-12 HPCI 6 Medium None Low 7 0 -7 negligible negligible negligible negligible HPCI 6 Low TT Medium 1 0 -1 negligible negligible negligible negligible HPCI Total -6.16E-10 2.69E-09 -5.88E-10 2.66E-09 FW 2(1) High TASCS, TT, (FAC) Medium (High) 0 1 1 -1.62E-09 -9.OOE-10 -1.62E-09 -9.OOE-10 FW 2 (1) High TASCS, (FAC) Medium (High) 4 1 -3 5.40E-10 2.70E-09 5.40E-10 2.70E-09 FW 2 (1) High "T, (FAC) Medium (High) 2 1 -1 -5.40E-10 9.00E-10 -5.40E-10 9.00E-10 FW 2 High TASCS, TT Medium 0 1 1 -1.62E-09 -9.OOE-10 -1.62E-09 -9.00E-10 FW 2 High TASCS Medium 6 4 -2 -3.24E-09 1.80E-09 -3.24E-09 1.80E-09 FW 2 High TV Medium 0 0 0 no change no change no change no change FW 2 High CC Medium 2 2 0 no change no change no change no change FW 4(1) High None (FAC) Low (High) 6 0 -6 2.70E-10 2.70E-10 2.70E-10 2.70E-10 FW 4 High None Low 3 2 -1 4.50E-11 4.50E-11 4.50E-11 4.50E-11 FW 5(3) Medium TASCS, TT, (FAC) Medium (High) 0 1 1 -1.80E-11 -1.OOE-11 -1.80E-12 -1.00E-12 FW 5 (3) Medium TASCS, (FAC) Medium (High) 0 0 0 no change no change no change no change FW 5 Medium TASCS Medium 0 1 1 -1.80E-11 -1.00E-11 -1.80E-12 -1.OOE-12 FW Total -6.20E-09 3.90E-09 -6.17E-09 3.91E-09 CCW 7 Low None Low 1 0 -1 negligible negligible negligible negligible CCW Total negligible negligible negligible negligible CRD 6 Medium None Low 10 0 -10 negligible negligible negligible negligible CRD 7 Low None Low 21 0 -21 negligible negligible negligible negligible CRD Total I negligible negligible negligible negligible 1.5-28 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results Inspections CDF Impact(4) LERF Impact(4) System(I) Category Consequence Failure Potential Rank DMs Rank 2 Section XI( ) 3 RI-ISI( ) Delta w/ POD w/o POD w/ POD ] w/o POD FPEC 6 Medium None Low 0 0 0 no change no change no change no change FPEC Total no change no change no change no change PCAC 6 Medium None Low 4 0 -4 negligible negligible negligible negligible PCAC Total negligible negligible negligible negligible Torus 6 Medium None Low 1 0 -1 negligible negligible negligible negligible Torus Total negligible negligible negligible negligible Grand Total -7.40E-09 1.36E-08 -7.30E-09 1.36E-08 Notes

1. Systems are described in Table 3.1.

2. Only those ASME Section XI Code inspection locations that received a volumetric examination in addition to a surface examination were included in the count. Inspection locations previously subjected to a surface examination only were not considered in accordance with Section 3.7.1 of EPRI TR-1 12657.

3. Risk Category 4 (1) inspection locations selected for examination by both the FAC and RI-ISI Programs are not included in the count since they do not represent additional examinations.

4. Per Section 3.7.1 of EPRI TR-1 12657, the contribution of low risk categories 6 and 7 need not be considered in assessing the change in risk. Hence, the word "negligible" is given in these cases in lieu of values for CDF and LERF Impact. In those cases where no inspections were being performed previously via Section Xl, and none are planned for RI-ISI purposes, "no change" is listed instead of "negligible."

NOTE: TABLE 4 was not part of the Risk-Informed ISI Program submittal and is intentionally excluded from this document. 1.5-29

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-1 Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Region High Risk Region Medium Risk Region Low Risk Region 1 Code System~ ) 1986 Section Xl( 2 ) EPRI TR-112657 _We(el_ ( Category(2 ) W 2 1986 Section Xl(2 ) EPRI TR-112667 Weld 1986 Section XIl 2) EPRI TR-112657 Weld Weld Count Vol/Sur Sur Only RI-ISI Other1 Count Vol/Sur SurOnly RI-ISI JOthe 3') Count Vol/Sur SurOnly RI-ISI 1OtherP3 ) 5 3 2 1 3 1 2 0 RPV B-F B-J 16 2 3 2 88 5 24 0 RWC U B-F 1 1 0 1 B-J 83 9 15 8 1 0 0 0 MS B-J 105 32 1 110) 99 21 21 0 SLC__ B-F 1 0 _ 1 0 B-J 8 0 3 1 26 0 6 0 RCR B-F 10 10 0 3 2 2 0 0 B-J 111 25 5 12 12 0 3 0 RCIC B-J 14 0 5 0 C-F-2 40 1 0 5 11 1 0 0 B-F 1 1 0 0 2 2 0 0 RHR B-J 30 4 0 8 75 21 0 8 7 4 0 0 C-F-2 361 30 2 0 B-F 2 2 0 1 CS B-J 20 9 0 2 8 2 0 0 C-F-2 161 22 0 0 B-F 2 2 0 0 HPCI B-J 6 3 0 2 9 1 0 0 C-F-2 60 9 0 7 81 7 0 0 B-J 29 9 0 10 41 8 0 6(5) FW C-F-2 7 5 0 0 1 1 0 0 1.5-30 t 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-1 (cont'd) Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Region System() 1 Systm~~ CodeII________I Category 2 Weld High Risk Region 1986 Section XI(2) EPRI TR-112657 Weld Medium Risk Region 1986 Section XI(2) EPRI TR-112657 Weld [1 Low Risk Region 1986 Section XI(2) EPRI TR-112657 Count Vol/Sur SurOnly RI-ISI 'Other3) Count Vol/Sur SurOnly RI-ISI Otherr(3 ) Count RI-IS Vol/Sur Sur Only RI-ISI Othrhe Other(3) CCW C-F-2 18 1 0 0 C-F-1 31 28 0 0 CRD C-F-2 10 3 0 0 FPEC C-F-2 54 0 0 0 PCAC C-F-2 47 4 0 0 Torus C-F-2 19 1 0 0 B-F 15 15 0 4 10 8 2 2 4 1 3 0 B-J 65 16 0 20 459 106 27 50 264 33 59 0 Total C-F-1 31 28 0 0 C-F-2 7 5 0 0 101 11 0 12 762 69 2 0 Notes

1. Systems are described in Table 3.1.

2. Since no examination selections had been made for the 4th interval ISI Program prior to the development of the RI-ISI Program, the 3rd Interval selections were used for comparison purposes. The Code of record for the 3rd Interval was the 1986 Edition of ASME Section XI. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section Xl.

3. The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce substantially less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, MNGP achieved a 9.2% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

4. One of these eleven welds was selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for this weld, the FAC examination will be credited toward both programs.

5. Two of these six welds were selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for these welds, the FAC examinations will be credited toward both programs.

1.5-31

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-2 Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Category System(1) Risk Consequence Failure Potential j Code (2) Weld 2 1986 Section Xl() EPRI TR-112657 Category Rank Rank DMs Rank Category Count Vol/Sur ISur Only RI-ISI IOthe B-F 5 3 2 1 4 Medium High None Low RPV B-J 16 2 3 2 B-F 3 1 2 0 RPV 6 Low Medium None Low B-J 80 3 22 0 RPV 7 Low Low None Low B-J 8 2 2 0 B-F 1 1 0 1 Medium High None Low RWCU 4 B-J 83 9 15 8 RWCU 7 Low Low None Low B-J 1 0 0 0 MS 4(1) Medium (High) High None (FAC) Low (High) B-J 6 2 0 1(4 MS 4 Medium High None Low B-J 99 30 1 10 MS 6 Low Medium None Low B-J 95 21 18 0 MS 7 Low Low None Low B-J 4 0 3 0 SLC 4 Medium High None Low B-J 8 0 3 1 B-F 1 0 1 0 SLC 6 Low Medium None Low B-J 26 0 6 0 RCR 2 High High CC Medium B-F 10 10 0 3 B-F 2 2 0 0 RCR 4 Medium High None Low B.-J 111 25 5 12 RCR 7Low Low None Low B-J 12 0 30 1.5-32 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-2 (cont'd) Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Category Consequence Failure Potential Code --- 1986 Section XI(2) EPRI TR-112657 nWeld ystem(l) Risk j Category R Rank DMs Rank Vol/Sur Sur Only RI-ISI Other(3 ) RCIC 4 Medium High None Low C-F-2 12 0 0 2 RCIC 5 (3) Medium (High) Medium TT, (FAC) Medium (High) C-F-2 8 1 0 1 RCIC 5 Medium Medium TT Medium C-F-2 20 0 0 2 B-J 5 0 2 0 RCIC 6 Low Medium None Low C-F-2 7 1 0 0 Low B-J 9 0 3 0 ___ 7 Low Low None RCIC C-F-2 4 0 0 0 B-F 1 1 0 0 ___ High High TT Medium RHR 2 B-J 30 4 0 8 B-F 2 2 0 0 RHR 4 Medium High None Low B.-J 65 17 0 7 RHR 5 Medium Medium TT Medium B-J 10 4 0 1 RHR 6 (3) Low (High) Medium None (FAC) Low (High) C-F-2 42 5 0 0 RHR 6 Low Medium None Low C-F-2 227 20 0 0 RHR 7(5) Low (Medium) Low None(FAC) Low (High) C-F-2 10 1 0 0 B-J 7 4 0 0 RHR 7 Low Low None Low C-F-2 82 4 2 0 CS 2 High High CC Medium B-F 2 2 0 1 CS 4 Medium High None Low B-J 20 9 0 2 CS 6 (3) Low (High) Medium None (FAC) Low (High) C-F-2 4 0 0 0 B-J 8 2 0 0 CS 6 Low Medium None Low C-F-2 23 4 0 0 CS 7(5) Low (Medium) Low None (FAC) Low (High) C-F-2 13 0 0 0 CS 7 Low Low None Low C-F-2 121 18 0 0 1.5-33

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-2 (cont'd) Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Category System(1 ) Risk Consequence Failure Potential Code [2) Weld 1986 Section XI(2) EPRI TR-112657 Category Rank Rank DMs Rank Category Count Vol/Sur Sur Only RI-ISI ]OtherP ) 3 Medium B-F 2 2 0 0 2 High High TT HPCI B-J 6 3 0 2 HPCI 4 Medium High None Low C-F-2 27 2 0 3 HPCI 5 Medium Medium TT Medium C-F-2 33 7 0 4 HPCI 6 Low Medium None Low C-F-2 81 7 0 0 HPCI 6 Low Low TT Medium B-J 9 1 0 0 FW 2 (1) High (High) High TASCS, TT, (FAC) Medium (High) B-J 1 0 0 1 B-J 1 1 0 1 FW 2 (1) High (High) High TASCS, (FAC) Medium (High) C-F-2 4 3 0 0 B-J 4 1 0 10 High TT, (FAC) Medium (High) C-F 1 1 0 FW 2 (1) High (High) C-F-2 1 1 0 0 B-J 2 0 0 1 High High TASCS, TT Medium FW 2 C-F-2 1 0 0 0 B-J 12 5 0 4 High High TASCS Medium FW 2 C-F-2 1 1 0 0 FW 2 High High TT Medium B-J 1 0 0 0 FW 2 High High CC Medium B-J 8 2 0 2 B-J 18 5 0 25 4(1) Medium (High) High None (FAC) Low (High) FW C-F-2 1 1 0 0 FW 4 Medium High None Low B-J 19 3 0 2 FW 5 (3) Medium (High) Medium TASCS, TT, (FAC) Medium (High) B-J 1 0 0 1 FW 5 (3) Medium (High) Medium TASCS, (FAC) Medium (High) B-J 1 0 0 0 FW 5 Medium Medium TASCS Medium B-J 2 0 0 1 1.5-34

                                       *

• 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 5-2 (cont'd) Inspection Location Selection Comparison Between 1986 ASME Section XI Code and EPRI TR-112657 by Risk Category 2 System(l) Risk Consequence Failure Potential Code Weld 1986 Section xl() EPRI TR-112657 Category Rank Rank DMs D Rank Category() Count VollSur sur Only RI-ISI he CCW 7 Low Low None Low C-F-2 18 1 0 0 CRD 6 Low Medium None Low C-F-1 10 10 0 0 Low C-F-i 21 18 0 0 7 Low Low None CRD C-F-2 10 3 0 0 FPEC 6 Low Medium None Low C-F-2 54 0 0 0 PCAC 6 Low Medium None Low C-F-2 47 4 0 0 Torus 6 Low Medium None Low C-F-2 19 1 0 0 Notes

1. Systems are described in Table 3.1.

2. Since no examination selections had been made for the 4th interval ISI Program prior to the development of the RI-ISI Program, the 3rd Interval selections were used for comparison purposes. The Code of record for the 3rd Interval was the 1986 Edition of ASME Section XI. The Code Categories listed in the table are therefore in accordance with the 1986 Edition of ASME Section XI.

3. The column labeled "Other" is generally used to identify augmented inspection program locations credited per Section 3.6.5 of EPRI TR-1 12657. The EPRI methodology allows augmented inspection program locations to be credited if the inspection locations selected strictly for RI-ISI purposes produce substantially less than a 10% sampling of the overall Class 1 weld population. As stated in Section 3.5 of this template, MNGP achieved a 9.2% sampling without relying on augmented inspection program locations beyond those selected by the RI-ISI process. The "Other" column has been retained in this table solely for uniformity purposes with the other RI-ISI application template submittals.

4. This one weld was selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for this weld, the FAC examination will be credited toward both programs.

5. These two welds were selected for examination by both the FAC and RI-ISI Programs. Since FAC was the only damage mechanism identified for these welds, the FAC examinations will be credited toward both programs.

1.5-35

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 2 (Rev. 1) Reactor Vessel Circumferential Shell Welds (note - italicized text clarifies/ corrects typographicalerrors and omissions or describesactions taken to address implementation) System: Reactor Vessel Class: 1 Category: B-A Item: B1.11 Reactor Vessel Circumferential Welds: VCBB-4, VCBB-3 and VCBA-2 (errantlynamed VCBB-2 on Rev.O) Examination Requirements: A September 8, 1992 revision to 10 CFR 50.55a(g)(6)(ii)(A) contains an augmented examination requirement to perform a one time volumetric examination of essentially 100% (>90%) of all circumferential and axial reactor pressure vessel (RPV) shell assembly welds. This rule revokes previously granted relief requests regarding the extent of volumetric examination on circumferential (B1.11) and longitudinal (81.12) reactor pressure shell vessel welds. 10 CFR 50.55a(g)(6)(ii)(A) requires the augmented examinations to be performed as specified in the ASME Code Section XI (1989 Edition). Monticello requests relief from the inspection of Reactor Vessel Circumferential (B-A) Welds Item B1.11 for the remaining term of the current license for Monticello (during the 4th ISI Interval). Basis For Relief: Monticello reactor vessel circumferential welds were not inspected to the essentially 100% volumetric requirements during the 1st and 2nd ISI inspection intervals. A relief request (RR-01) was granted on the basis of inadequate accessibility and unnecessary radiation exposure during the first two 10 year inspection intervals. Upon submittal of the 3rd Interval ISI Inspection Plan, Rev. 1 (July 29, 1993), continuance for the 1st and 2nd interval relief request (RR-01) was requested. That relief request (RR-01) was denied on the basis of 10 CFR 50.55a(g)(6)(ii)(A), effective September 8, 1992, requiring augmented examination for reactor vessel shell assembly welds. On November 10, 1998, the NRC issued Generic Letter 98-05 "BOILING WATER REACTOR LICENSEES USE OF BWRVIP-05 REPORT TO REQUEST RELIEF FROM AUGMENTED EXAMINATION REQUIREMENTS ON REACTOR PRESSURE VESSEL CIRCUMFERENTIAL WELDS." This generic letter permits licensees to request permanent relief from the inservice inspection requirements of 10 CFR 50.55a(g)(6) for the volumetric examination of circumferential reactor 1.5-36

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN pressure vessel welds if it can be demonstrated that: (1) at the expiration of the license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998, safety evaluation, and (2) operator training and procedures limit the frequency of cold over-pressure events to the amount specified in the staffs July 28, 1998, safety evaluation (Reference 1). The following is our evaluation of these two criteria. (1) Limiting Conditional Failure Probability The values established in Attachment 1 were calculated in accordance with the guidelines of Regulatory Guide 1.99, Revision 2. The chemistry factor for the limiting circumferential weld recorded in Attachment 1 is Monticello (manufactured by Chicago Bridge & Iron (CB&I)) plant specific (Reference 3). This value is slightly higher than the USNRC's value which utilizes Table 1 of Regulatory Guide 1.99, Revision 2. As a result, the Monticello mean RTNDT value of 46.90 F is slightly higher than the USNRC's limiting plant specific analysis mean RTNDT value of 44.50 F listed in Reference 5 for the CB&I reference case. A recent safety evaluation (Reference 6) identified a Brunswick Unit I (manufactured by CB&I) mean RTNDT value of 46.50 F which also exceeded the corresponding CB&I mean RTNDT value specified in Reference 5. To validate the acceptability of the failure probability in this case, the staff performed calculations using the Brunswick Unit 1 value of 46.50 F. The calculations showed only a small increase in failure probability (6 x 10 7/yr for Brunswick vs. 2 x 10 7/yr for the reference case). Since the Monticello mean RTNDT is only slightly higher than the Brunswick Unit 1 mean RTNDT (46.90 F vs. 46.50 F), it is expected that only a small increase in failure probability will result for Monticello. The overall limiting conditional failure probability for circumferential welds across the BWR fleet listed in Reference 5 is 8.17 x 10-5 /yr (calculated by the staff for the Babcock & Wilcox (B&W) reference case). This limiting conditional failure probability is based on reactor vessel data that produced a calculated mean RTNDT of 99.80 F (Reference 5). Since the Monticello mean RTNDT (46.90 F) is less than 99.80 F, it follows that the Monticello conditional failure probability will also be less than the limiting failure probability listed in Reference 5. Attachment 2 provides a plot of mean RTNDT against failure probability using results documented in References 5 and 6. Based on this trend, the conditional failure probability for Monticello is estimated to be less than 1 x 10-6/yr. In conclusion, the above discussion demonstrates that the circumferential welds of the Monticello RPV will continue to satisfy the limiting conditional failure probability listed in Reference 5. 1.5-37

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN (2) Training and Procedures The cold pressurization events considered in Reference 1 (i.e., inadvertent injections, condensate injection, CRD injection, loss of RWCU, actual event) were reviewed to identify the critical operator actions that were assumed to occur to mitigate these events. Procedures and training were reviewed to ensure that those critical operator actions would occur with a high degree of certainty so that the low temperature over pressurization (LTOP) event frequency is maintained less than the amount specified in Reference 1 (i.e., 1 x 10-3/yr). System design was also considered in this review to assure that the associated systems function as described in Reference 1. Results of our review indicate that in general, procedures, training and system design ensure that the evaluations contained in Reference 1 are valid for Monticello. Following are the detailed results of our review:

1. Inadvertent Injections.

The evaluation provided in Reference 1 (paragraph 2.6.1.1) is applicable to Monticello with one exception. The evaluation considered the availability of automatic trips of high pressure injection systems on high water level. Review of Monticello procedures identified that during performance of reactor feedwater pump (RFP) testing during cold shutdown, the high reactor water level trip is bypassed. Measures are taken procedurally to close valves that prevent water from getting to the vessel. Monticello enhanced OperationsProcedure B.06.05-05 to further assure the isolation of flow to the vessel.

2. Condensate Injection.

The evaluation provided in Reference 1, (paragraph 2.6.1.2) is applicable to Monticello. Operating procedures provide precautions which indicate that reactor water level is to be closely monitored when starting a condensate pump. This aids in assuring that an overfill event which could lead to an LTOP event does not occur. In order to assure that operations personnel understand that an overfill event has the potential to lead to an LTOP event, Monticello enhanced Operations ProcedureB.06.05-05 to identify an LTOP event as a potential consequence of an overfill event. Monticello also has high reactor water level and high reactor pressure alarms in the control room that warn operators when high level/pressure limitations are being exceeded which provides further assurance that an LTOP event will not occur due to condensate injection. 1.5-38

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

3. CRD Injection.

The evaluation provided in Reference 1, (paragraph 2.6.1.3), is applicable to Monticello. The evaluation notes that the risk of cold over pressurization due to CRD injection may be higher if a loss of station power were to occur during reactor vessel pressure testing. Monticello revised vessel pressure testing procedures 0255-20-IIA-I and 0255-20-11C-I to provide precautions that ensure properresponse to a loss of station power (i.e., RWCU and Recirculationpumps are restored along with restorationof CRD).

4. Loss of Reactor Water Cleanup (RWCU)

The evaluation provided in Reference 1, (paragraph 2.6.1.4), is applicable to Monticello. Monticello has procedures in place to provide guidance for recovery measures following a scram. In the event that a scram occurs that results in a RWCU isolation, procedural guidance is provided which consists of restoring the RWCU system as soon as the cause of the isolation is identified and resetting the reactor scram as soon as possible in order to limit cold water injection into the vessel. Also, procedural guidance is provided for dealing with recirculation loop or vessel stratification so that an excessive amount of cold water is not distributed throughout the reactor vessel during the restart of a tripped recirculation pump(s). Monticello added a precaution in the OperationsProcedure C.4-A for RWCU restorationin orderto further inform the operations personnel of the potential of an LTOP event occurringduring SCRAM recovery.

5. Actual Event.

General Electric issued RICSIL No. 049, Inadvertent Vessel Pressurization, in response to the actual event discussed in Reference 1, (paragraph 2.6.1.5). Our assessment of the RICSIL indicated that the likelihood of a similar event occurring at Monticello is very low. Procedures require that the reactor vessel remain vented at all times during cold shutdown except as permitted by approved procedures. The reactor vessel pressure test procedure allows the vent valves to be closed during cold shutdown. During the pressure test, strict procedural guidance is provided for administratively monitoring vessel pressure and temperature while controlling CRD injection and RWCU reject in order to assure a smooth, controlled method of increasing or decreasing pressure while vessel temperature is being maintained above the required P-T limits. If reactor pressure exceeds the specified limits, during the test, the CRD pump is immediately tripped. In addition to the above mentioned 1.5-39

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN procedural guidance, a requirement is included to perform an "Infrequent Test or Evolution Briefing" with all essential personnel. This 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 the process in which the test would be aborted if plant systems responded in an adverse manner. The above evaluations show that system design and procedures, including the proposed enhancements, minimize the probability of LTOP events at Monticello. Our review of training indicated that licensed operator training addresses LTOP events. Initial licensed operator simulator training, for example, includes performance of surveillance tests which ensure pressure-temperature curve compliance during plant heatup and cooldown. Additionally Monticello created Request for Training (RFT) 20012810 to provide trainingto operationspersonnel on the specific scenarios and events evaluated in Reference 1, (paragraph 2.6.1.1-5), including the features of system design and proceduralcontrols that prevent such events at Monticello.

== Conclusion:==

The Monticello mean RTNDT value of 46.90 F is less than the mean RTNDT value of 99.80 F corresponding to the B&W limiting reference case. Since the Monticello RTNDT is much less than the limiting RTNDT, the Monticello conditional failure probability will be well below the limiting conditional failure probability of 8.17 x 10-5/yr calculated by the Staff for the corresponding B&W reference case. A thorough review of existing procedures, operator training and system design identified improvement opportunities that Monticello has committed to implement. With the recommended enhancements to existing procedures and operator training and with the current design capabilities of the associated systems, the LTOP event frequency is limited to the amount specified in Reference 1, (1 x 10 3

/yr).

Based on these evaluations the conditions for requesting relief from the inservice inspection requirements of 10 CFR 50.55a(g)(6)(ii)(A), for the volumetric examination of circumferential reactor pressure vessel welds in accordance with ASME Code Section XI (1995 Edition with 1996 Addenda), Table IWB-2500-1, Examination Category B-A, Item B1.11, Circumferential Welds, are satisfied. Relief is hereby requested in accordance with 10 CFR 50.55a(a)(3)(I). The proposed alternative examinations provide an adequate level of quality and safety. 1.5-40

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Alternate Examination: As an alternative to the inspection requirements of ASME Code Section Xl (1995 Edition with the 1996 Addenda) Category B-A, Item B1.11, 100% volume requirement, we propose that the following examination methodology be used. The alternative examination requested maintains essentially 100% (>90%) examination of reactor vessel longitudinal (axial) shell welds, Code Category B-A, Item B.1.12. Two to three percent of the circumferential RPV shell welds Code Category B-A, Item B1.11, Code Category B-A, Item B1.11 will be inspected at the intersections of the axial and circumferential welds. This is consistent with the alternate inspection requirements as specified in GL 98-05. This alternative is capable of detecting weld degradation sufficient to insure the integrity of the reactor pressure vessel boundary, and is the same as that described in the NRC SER (Reference 1). Time Period Relief is Requested For: Relief is presently approved by the NRC for the remaining term of the current Monticello license during the 4th 10 year interval. (Reference 7)

References:

1. NRC Safety Evaluation Report of Topical Report by the Boiling Water Reactor Vessel and Internals Project: "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations, BWRVIP-5," (TAC No. M93925), July 28, 1998.

2. General Electric Report SASR 87-61, DRF137-0010, "Revision of Pressure-Temperature Curves to Reflect Improved Beltline Weld Toughness Estimate for the Monticello Nuclear Generating Plant - Rev. 1," December 1987.

3. NSP Letter to NRC, Submittal of Report on Reactor Pressure Vessel Specimen Test, December 21, 1998.

4. General Electric Report GENE-B1 3-01796-1, "Reactor Vessel Fracture Toughness Engineering Evaluation - Task 5.4," March 13, 1996 1.5-41

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

5. NRC Safety Evaluation Report of Topical Report by the Boiling Water Reactor Vessel and Internals Project: "Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-5 Report (TAC No. MA3395),"

March 7, 2000.

6. Brunswick Steam Electric Plant, Unit No's 1 and 2 - Safety Evaluation for Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i) for Reactor Vessel Circumferential Shell Weld Examinations (TAC No's MA9299 and MA9300).

7. NRC Letter, "Monticello Nuclear Generating Plant - Approval of Relief Request Number 12 of the Third 10 Year Inservice Inspection Program,"

(TAC No. MB0261), July 27, 2001. Status: Approved July 27, 2001 for continued use in 4th Interval (...'remainderof current 40-year operatinglicense for the unit'), (See Reference #7 above). 1.5-42

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN ATTACHMENT 1 Comparison of Monticello RPV Parameters to NRC Limited Plant Specific Parameters Parameter Monticello Parameters USNRC Limiting Plant Description for the Bounding Specific Analyses Circumferential Weld Parameters SER Table 2.6-4 (Reference 5) CB&I B&W Cu, wt% 0.10 (Reference 2) 0.10 0.31 Ni, wt% 0.99 (Reference 2) 0.99 0.59 CF (Chemistry factor) 138.5 (Reference 3) 134.9 196.7 EOL ID 0.51 (Reference 4) 0.51 0.095 Fluence, x 1019 n/cm 2 ARTNDT, °F 112.5 109.5 79.8 RTNDT (u) OF -65.6 (Reference 2) -65 20 Mean RTNDT, OF 46.9 44.5 99.8 Conditional Failure <1x10.6 2x10-7 8.17x10-' Probability P(FIE) Attachment 2 1.5-43

                                                                                                                                       * *z 00 zu
                                                                                                                                         -- I ATTACHMENT -:2 Clrc. Weld Failure Probability vs Mean RTNoT Trend Using Limiting CE, CB&I, B&W and Brunswick Data                r --

1.OOE+00 CA M

         'I.0OOE-01                                                                                                                         :o0 Legend:

1- CB&I Umiting Analysis (Ref. 5) 2= Brunswick Limiting Analysis (Ref. 6):

1.00E-02: 3 C (EVIP-)LlmitingAnaiysLs4ReL5) 4 = CE (CEOG) Limiting Analysis (Ref. 5) z 5 = B&W Limiting Analysis (Ref. 5) zm ILj a1. rn 01 I.OOE-03 0

                                                                                                                                               -1 I.'

0 1 .00E-04: IL, U 0 1.OE&05 0 ie tn bs results rota Ma R" Pe NOTE: 1.OOE-06 Moýnticello Failure Probability vs Mean RTNDT Projected, Intersection =,_M 1, : > 0 1.OOE-07 40 46.9 50 60 70 80 90 100 110* r'n

                                                                                                                                         -u-0 Mean RTNT (F)                                             >

z-4 0 J z

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit 1 - Relief Request No. 3 (Rev. 0) Appendix VIII Supplement 4 System/Component(s) For Which Relief Request Will Be Used Code Class: Class 1

Reference:

ASME, Section XI, Tables IWB-2500-1 (1995 Edition, 1996 Addenda) Examination Category: B-A Item Number: B1.10, B1.20

== Description:== Alternative Requirement to Appendix VIII, Supplement 4 "Qualification Requirements for the Clad/Base Metal Interface of Reactor Vessel" Component Numbers: All Code and 10 CFR 50.55a Requirements: 10 CFR 50.55a(b)(2) was amended on September 22, 1999 to reference Section XI of the ASME Code through the 1995 Edition with the 1996 Addenda (64 FR 51370). This amendment provides an implementation schedule for the supplements to Appendix VIII of Section XI to the 1995 Edition with the 1996 Addenda. Supplement 4 to Appendix VIII, Subparagraph 3.2(c) imposes three statistical parameters for depth sizing. The first parameter, 3.2(c)(1), pertains to the slope of a linear regression line. The linear regression line is the difference between measured versus true value plotted along a through-wall thickness. The second parameter, 3.2(c)(2), pertains to the mean deviation of flaw depth. The third parameter, 3.2(c)(3), pertains to a correlation coefficient. The Final Rule was amended by Federal Register Notice (66FR16391) dated March 26, 2001. This amendment specified the use of a flaw length sizing tolerance criterion of 0.75 inch Root Mean Square (RMS) for reactor vessel qualification to be used in conjunction with the 0.15 inch RMS for depth sizing specified in the Rule in lieu of paragraphs 3.2(a) and 3.2(b). In the Notice, there was no reference to the elimination of the statistical parameters of Paragraph 3.2(c), which were intended for use with paragraphs 3.2(a) and 3.2(b) of Appendix VIII, Supplement 4. There was no amendment statement included to reflect the use of the RMS error calculations for depth and length sizing in lieu of Paragraph 3.2(c). 1.5-45

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Basis for Alternative Examination: This relief request was developed using the Electric Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) ASME Section XI, Appendix VIII Implementation Guideline. It is modeled after the sample request for relief associated with the Supplement 4 published discrepancies: Appendix D, "Sample Request for Relief - Alternative Length Sizing Criteria (Revised)." (Reference 5) The U.S. nuclear utilities created PDI to implement demonstration requirements contained in Appendix VIII. PD1 developed a performance demonstration program for qualifying UT techniques. PDI does not use paragraph 3.2(c) for sizing qualifications. The solution for resolving the differences between the PDI program and the Code was for PDI to participate in the development of a Code case that reflected PDI's program. The Code case was presented to ASME for discussion and consensus building. NRC representatives participated in this process. ASME approved the Code Case and published it as Code Case N-622, "Ultrasonic Examination of RPV and Piping, Bolts and Studs, Section Xl, Division 1." (Reference 6) The NRC first approved the use of Code Case N-622 for Florida Power and Light Company's St. Lucie Plant Unit 2 (TAC No. MA5041). (Reference 7) Operating in parallel with the actions of PDI, the Staff incorporated most of Code Case N-622 criteria in the Rule published in the Federal Register, 64 FR 51370 dated September 22, 1999. This amendment requires the implementation of the ASME Code Section Xl, Appendix VIII, Supplement 4, 1995 Edition with the 1996 Addenda. The required implementation date for Supplement 4 was November 22, 2000. Appendix IVto Code Case N-622 contains the proposed alternative sizing criteria which has been authorized by the Staff. However, the sizing parameters printed in the published Rule differed from the sizing parameters implemented by the PDI Program and Code Case N-622. On January 12, 2000, NRC Staff, representatives from the EPRI Nondestructive Examination Center, and representatives from PDI participated in a conference call. The discussion during the conference call included the differences between Supplement 4, "Qualification Requirements for the Clad/Base Metal Interface of Reactor Vessel," to Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," Paragraph 10 CFR 50.55a(b)(2)(xv)(C)(1) in the rule (Federal Register, 64 FR 51370), and the implementation of Supplement 4 by the PDI Program. (Reference 8) 1.5-46

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN In a public meeting on October 11, 2000 at NRC offices in White Flint, MD, the PDI identified the discrepancy between the PDI Program and statistical parameters required by Subparagraph 3.2(c). The Staff agreed that the inclusion of the statistical parameters of Paragraph 3.2(c) of Supplement 4 to Appendix VIII was an oversight. The NRC agreed that Paragraph 10 CFR 50.55a(b)(2)(xv)(C)(1) should have excluded Subparagraph 3.2(c) as a requirement. (Reference 9) In Subparagraph 3.2(c), the linear regression line is the difference between measured versus true value plotted along a through-wall thickness. For Supplement 4 performance demonstrations, a linear regression line of the data is not applicable because the performance demonstrations are performed on test specimens with flaws located in the inner 15% through-wall. The difference between measured versus true value produce a tight grouping of results that resemble a shotgun pattern. The slope of a regression line from such data is extremely sensitive to small variations, thus making the parameter of Subparagraph 3.2(c)(1) a poor and inappropriate acceptance criterion. The value used in the 3.2(c)(2) is too lax with respect to evaluating flaw depths within the inner 15% of wall thickness. Therefore, Monticello proposes to use the more appropriate criterion of 0.15 inch RMS of 10 CFR 50.55a(b)(2)(xv)(C)(1), that modifies Subparagraph 3.2(a) as the acceptance criteria. Subparagraph 3.2(c)(3) pertains to a correlation coefficient. This value of correlation coefficient is inappropriate for this application since it is based on the linear regression from Subparagraph 3.2(c)(1). The NRC Staff previously approved MNGP use of this Alternative to the Code and 10 CFR 50.55a on August 22, 2001 (TAC No. MB1 833) for use during the 3rd ISI Interval. (Reference 10) Alternative Examination: Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested to use the RMS Error calculations in lieu of the statistical parameters of Subparagraph 3.2(c) in Supplement 4 of the 1995 Edition 1996 Addenda of ASME Section XI Appendix VIII. As discussed above and demonstrated by the PDI, this will provide an acceptable level of quality and safety. 1.5-47

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Implementation Schedule: This alternative is requested for continued use for the 4th Ten-Year Interval of the Inservice Inspection Program for Monticello.

References:

1. ASME Boiler and Pressure Vessel Code, Section XI, 1995 Edition with 1996 Addenda

2. Federal Register, Rules and Regulations, September 22, 1999 (64 FR 51370)

3. Federal Register Notice, Industry Codes and Standards, Amended Requirements, March 26, 2001 (66 FR 16391)

4. Federal Register, Rules and Regulations, September 26, 2002 (67 FR 60520)

5. Performance Demonstration Initiative (PDI), "Guideline for Implementation of Appendix VIII and 10CFR50.55a," Volume One, Programmatic Implementation, Rev. 2, Appendix D, October 14, 2000

6. ASME Section XI Nuclear Code Case N-622, "Ultrasonic Examination of RPV and Piping, Bolts, and Studs"

7. NRC Staff letter to Mr. T. F. Plunkett, Florida Power and Light Company, September 23,1999.

8. Meeting Summary, Teleconference between NRC and representatives from PDI, D.G. Naujock, Metallurgist, NDE & Metallurgy Section, to Edmund J. Sullivan, Chief NDE & Metallurgy Section, Chemical Engineering Branch, Division of Engineering, U.S. NRC, March 6, 2000.

9. NRC Memo, "Summary of Public Meeting Held on October 11, 2000, with PDI Representatives," November 13, 2000

10. NRC Letter to Nuclear Management Company, "MNGP - Evaluation of Relief Request No. 13 for the Third 10-Year Interval Inservice Inspection Program," (TAC No. MB1833), August 22, 2001 1.5-48

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

11. NRC Letter to Nuclear Management Company, "Relief Request Nos. 3 and 6 for the Fourth 10-Year Interval of the Inservice Inspection Examination Plan" (TAC No. MB6896), March 28, 2003 Status:

Approved on March 28, 2003 for use during the 4th Interval. (See Reference 11 above) 1.5-49

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 4 (Rev.0) (restructured and resubmitted) Reactor Vessel Stabilizer Bracket Welds ASME CODE COMPONENT AFFECTED Code Class: Class 1 Code Subsection: IWB Code Examination Category: B-K, Welded Attachments for Vessels, Piping, Pumps, and Valves Code Item No.: B10.10 Parts Examined: Pressure Vessels, Welded Attachments Examination Method: Surface Examination Frequency: 1st Interval and each Successive Interval System: Reactor Pressure Vessel (RPV) Component

Description:

RPV Stabilizer Bracket welds to the RPV ISI Summary Number 102650 Component ID: Vsl Stblzr Lugs, (Quantity of 4) Description of Relief: Proposed alternative to the Code examination frequency requirements APPLICABLE CODE EDITION AND ADDENDA American Society of Mechanical Engineers (ASME) Section XI, 1995 Edition with 1996 Addenda is the Code of Record for the 4th ISI Interval. APPLICABLE CODE REQUIREMENT TABLE IWB-2500-1, CATEGORY B-K, ITEM B10.10, INCLUDING NOTE 2 In each Inspection Interval, each welded attachment and each identified occurrence is required to be examined with a surface examination method (described in IWA-2220.) NOTE (2) The extent of the examination includes essentially 100% of the length of the attachment weld at each attachment subject to examination. 1.5-50

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN REASON FOR REQUEST Monticello Nuclear Generating Plant (MNGP) is a General Electric Type 3 Boiling Water Reactor (BWR-3) with a Mark I Containment. The reactor vessel was designed and built to the 1965 Edition of ASME Section III with Summerl 966 Addenda. Piping systems were designed in accordance with the 1967 Edition of USA Standard (USAS) Code for Pressure Piping B31.1.0. "Power Piping." Construction Permit CPPR-31 was issued on June 19, 1967 and full commercial operation began on June 30, 1971. Plants of this type were designed and erected prior to the examination access requirements of Section XI. The Atomic Energy Commission (AEC) mandated the rules of ASME Section XI in 1971 for all nuclear plants with construction permits issued after January 1, 1971, and in 1976, they mandated use of ASME Section XI for all nuclear plants. 10CFR50.55a(g)(1) states "Fora boiling orpressurized water-cooled nuclear power facility whose constructionpermit was issued before January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (g)(5) of this section to the extent practical." 10CFR50.55a(g)(4) states "... components (including supports) which are classified as ASME Class 1, 2, and Class 3 must meet the requirements, excelt design and access provisions and preservice examination requirements, set forth in Section X1 of editions of the ASME Boiler and Pressure Vessel Code and Addenda ... " [Emphasis Added] Four RPV stabilizer brackets are attached to the Class 1 RPV with full penetration fillet welds at 0°, 90', 180°, and 2700 RPV azimuth at an elevation of 994'-2". The RPV stabilizers are connected with flexible couplings to the brackets on the RPV and also to the biological shield wall. The RPV stabilizers, brackets, and their attachment welds are designed to withstand and resist local loads (jet reaction forces) and seismic loads while allowing axial and radial movement due to normal thermal growth. The RPV stabilizers brackets do not provide structural support during normal operation. The MNGP RPV has never experienced jet reaction forces or seismic events, therefore the stabilizers, brackets, and attachment welds have not experienced the loads for which they are designed. The MNGP Mark I primary containment structure, or drywell, is shaped somewhat like an upside-down light bulb. The RPV stabilizer brackets are located in the higher, necked-down elevations of the drywell. This region of the drywell is a very limited access area; it was not designed with the intention of 1.5-51

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4T' INTERVAL EXAMINATION PLAN providing access and accommodations normally considered necessary for a general work area. The area around the stabilizers is extremely congested. The vessel stabilizer brackets are surrounded by mirror insulation that is secured by cable hangers and buckles, ventilation ductwork with support bracing, and electrical installations such as thermocouples. All of this equipment must be relocated and restored to provide access to the stabilizers for examination of the welds. Additionally, due to the location of the stabilizer brackets and the lack of a working platform at the stabilizer location a complex scaffold installation is required to provide access to the examination location. The photos and reference drawings, attached to this request, show the physical obstacles imposed by the design and construction of the primary containment, RPV, ventilation ducting, RPV stabilizers, containment supports, and other systems. Combined, these obstacles as described below, create an unusually difficult hardship to overcome to provide access for the examination of the stabilizer bracket attachment welds that are specified by the Code. In the course of scaffold installation and removal, interference removal and restoration, insulation removal and restoration, weld preparation, performance of the examination, and health physics monitoring, personnel would be subjected to significant radiation doses found in the drywell for lengthy durations. Dose survey maps taken from the recent refueling outages at this region of containment indicate dose rates in the general area to be 5 - 140 millirem per hour (mrem/hr). It is reasonable to expect that the contact dose rates at the bracket welds would be similar to those experienced at the nearby feedwater (986'-7" elevation) and main steam nozzles (999'-0" elevation). These dose rates range from 20 - 80 mrem/hr in the general nozzle area and 20 - 800 mrem/hr in contact with the components. NMC estimates indicate that radiation exposure to personnel involved in the activities associated with examination of the four RPV Stabilizer Bracket Welds would result in 21.675 person-rem. In summary, NMC has determined that:

1. MNGP is not subject to the access requirements of ASME code as described in 10CFR50.55a(g)(4) due to its age and design.

2. Access to the RPV stabilizer brackets is difficult due to their location, interferences, and surrounding equipment.

3. Radiological dose rates in the area of the RPV stabilizer brackets is high due to the proximity of the brackets to feedwater and main steam nozzles.

1.5-52

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Therefore, pursuant to 10CFR50.55a(a)(3)(ii), NMC has determined that compliance with the Code requirement would result in hardship or unusual difficulty without a compensating increase in quality or safety PROPOSED ALTERNATIVE AND BASIS FOR USE As an alternative to the requirements of the ASME Section Xl Code, Table IWB-2500-1, Category B-K, Item B10.10, NMC proposes to perform a surface examination on the stabilizer brackets if local (jet reaction forces) or seismic design loads are experienced. The stabilizer brackets are located in a very limited access area of the drywell which precludes inadvertent damage to be imparted on the brackets such as rigging, climbing, arc strikes, etc. The RPV stabilizer bracket attachment welds have never experienced loads due to jet reaction forces or seismic events. The stabilizer brackets do not provide support during normal operation. This proposed alternative to the frequency requirements of Table IWB-2500-1, Category B-K, Item B1 0.10 will provide an acceptable level of quality and safety. DURATION OF PROPOSED ALTERNATIVE NMC is requesting relief for the 4th 10-year Interval of the ISI Program for the Monticello Nuclear Generating Plant. PRECEDENTS This relief from the requirements of 1 OCFR50.55a and alternative to the Code was previously approved for 2nd and 3rd 10-year Intervals of the ISI program at Monticello: " NRC Letter, "Monticello - Second Ten-Year Inservice Inspection (ISI) Program," (TAC No. 46510), November 29, 1990, Relief Request No. 51 " NRC Letter, "Evaluation of the Third 10-Year Interval Inservice Inspection Program Plan and Associated Requests for Relief for Monticello," (TAC No. M82545), October 18, 1994, Relief Request No. 2 1.5-53

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Status: Approved on January 6, 2005 for use during the 4th Interval, NRC Letter, "Fourth 10-Year Interval Inservice Inspection Program Plan Request for Relief No. 4" (TAC No. MC2222) 1.5-54

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 5 (Rev. 0) Leakage at Bolted Control Rod Drive (CRD) Housing Connections SYSTEM: Bolted CRD Housing Joint Class: 1 Category: B-P Item: B15.10 Code Examination Requirements: IWA-5250(a)(2): If leakage occurs at a bolted connection on other than a gaseous system, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100. Basis for Relief: 10 CFR Part 50, Section 50.55a(a)(3), which states, (in part):

      "Proposedalternativesto the requirements of paragraphs (c), (d), (e), (0, (g), and (h) of this section or portions thereofmay be used when...

(ii) Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety." The CRD (Control Rod Drive) housings are flanged connections beneath the reactor vessel that are used to secure the 121 CRD mechanisms in position below the vessel. Each of the 121 CRD to CRD housing bolted joints utilizes eight bolts, washers, and nuts to hold the CRD mechanism in position. The joint also utilizes three hollow metal 0-rings to provide a watertight seal capable of withstanding full reactor pressure at normal operating temperatures. The CRD housing joints are VT-2 examined as part of the periodic Reactor Pressure Vessel Leakage and Hydrostatic pressure tests. These tests are conducted with the vessel temperature much less than the design operating temperature. For a typical test, the vessel temperature would be <2120 F, as compared to a normal operating temperature of about 5400 F. It is not unusual for these bolted joints to leak slightly during periodic reactor vessel pressure tests conducted at test temperatures below normal operating temperature. 1.5-55

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN This is a condition identified in the original design of the connection by the Architect/Engineer, General Electric (GE). GE developed guidance to permit evaluation of a leaking CRD housing bolted connection over a period of time, while at test pressure, to determine whether the leak will stop once the vessel heats up to normal operating pressure. This leakage evaluation criteria is incorporated into the VT-2 tests for these joints. Compliance with Code Requirement IWA-5250(a)(2) represents a hardship (burden) in the case of the CRD housing bolted joints because:

1) Examining the bolting would involve the accumulation of considerable personnel radiation exposure, since the work must be performed in a relatively high dose rate area inside the drywell, immediately below the reactor vessel. Typical shutdown dose rates in the vicinity of the bolting flanges would be on the order of 50 to 100 mr/hr.

2) Since the reactor pressure vessel test is critical path item, the additional time needed to depressurize the vessel, remove the bolting, perform the exam, and then re-pressurize the vessel to retest the joint would delay plant startup from an outage by an equivalent amount of time. The cost of such delays is significant, since it is estimated that the cost of extending the duration of an outage is $379,000 per day (including replacement power costs)(this is estimated cost submitted in 1993 (see TAC No.

M82545 referenced in "Status" section) Compliance with Code requirement IWA-5250(a)(2) would not result in a compensating increase in quality or safety because:

1) CRD Housing joint leakage during (relatively) low temperature testing is not unexpected due to the design of the bolted joint. This joint is unusual in that it has hollow metal o-rings that require the CRD housing bolts to be tightened within a specific torque range in order to function properly at normal operating temperature. Thus, the bolts cannot simply be tightened to stop leakage as might be done for a conventional gasketed joint. As noted previously, GE developed guidance to evaluate any CRD housing leakage to determine if the leakage will persist at normal operating temperature/pressure and should therefore be corrected.

1.5-56

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

2) Leakage that is found to be acceptable per the guidance is not considered adverse to quality or safety and need not be corrected prior to startup. This type of analysis is consistent with Section XI.

3) Code paragraph IWB-3142 allows analysis of the leakage for acceptability. Performance of the VT-3 bolting examination does not represent a corrective action for the joint leakage and will not reduce the likelihood of joint leakage upon retest. Therefore, the VT-3 bolting examination does not contribute to increased quality or safety.

4) The bolts in the CRD housing connection are periodically examined when the joint is disassembled, per Table IWB-2500-1, Item B7.80 (1995 Edition with no Addenda per 10CFR50.55A Paragraph (b)(2)(xxi)(B)) and Procedure 9309, "Changeout Selected CRD's -

Maintenance" and Commitment No. M92076A. Four of the eight bolts on each housing joint were replaced with new bolts in 1991 under Work Control Record (WCR) 91-01909. It was also reported in General Electric SIL 483 that only three uniformly distributed housing bolts are required to support the CRD mechanism. These factors provide a high degree of confidence in the long term safety and integrity of the CRD housing joints. Earlier Section Xl code editions invoked by Monticello's 1st and 2nd Ten-Year Inspection Interval Programs did not include the subject examination requirement. During the 3rd Inspection Interval, Relief Request 7 was granted by the NRC in an SER dated October 18, 1994. Alternate Examination: Pursuant to 10 CFR 50.55a(a)(3)(ii), the following alternative is proposed. Any leakage found at a CRD housing bolted joint during a periodic pressure test performed at a temperature much less than operating temperature will be evaluated to determine whether it will stop leaking at operating temperature. If this evaluation shows the leak will stop as temperature increases to normal operating temperature, no further action will be taken. The acceptance criteria is based on guidance provided by General Electric and is included in the VT-2 tests for the joint (Note: This criteria was submitted for NRC review during the Request for Relief process previously approved on October 18, 1994, therefore it is not included in this submittal). If the leakage is determined to be unacceptable according to the General Electric guidelines and the joint is disassembled to correct the leak, any CRD bolting that is reused will be examined by the VT-1 examination method (10 CFR 50.55a(b)(2)(xxi)(B) dated September 26, 2002). 1.5-57

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Upon approval of this Relief Request, MNGP commits to revise the applicable pressure test procedure to perform a VT-1 exam in lieu of a VT-3 exam specified by IWA-5250(a)(2) on all CRD bolting that will be reused when the GE acceptance criteria has been exceeded and disassembly is required to correct the leak. Status: Approved on June 9, 2003 for use during the 4th Interval, NRC Letter to Nuclear Management Company, "Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No. 5" (TAC No. MB6956) 1.5-58

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 6 (Rev. 0) Appendix VII Annual Training System/Component(s) For Which Relief Will Be Used: Code Class: All

Reference:

ASME, Section Xl 1995 Edition 1996 AdcJienda. Appendix VII, VII-4240 Examination Category: All Item Number: All

== Description:== All NDE Examiners performing ultrasonic volumetric examination in accordance with ASME Section XI, 1995 Edition 1996 Addenda and Appendix VII, Annual Training. Component Numbers: All Code and 10 CFR 50.55a Requirement: ASME Section XI, 1995 Edition, 1996 Addenda, Mandatory Appendix VII, Paragraph VII-4240: Supplemental training is required on an annual basis to impart knowledge of new developments, material failure modes, and any pertinent technical topics as determined by the Employer. The extent of this training shall be a minimum of 10 hours per year. A record of attendance and the topics covered during the training shall be maintained; however no examination is required. 10 CFR 50.55a, paragraph (b)(2)(xiv): All personnel qualified for performing ultrasonic examinations in accordance with Appendix VIII shall receive 8 hours of annual hands-on training on specimens that contain cracks. This training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility. Basis For Relief Request: 10 CFR 50.55a was amended in the Federal Register (Volume 64, No. 183 dated September 22, 1999) to require Appendix VIII - Supplements for accelerated implementation in accordance with ASME Section X1 1995 Edition, 1996 Addenda. 1.5-59

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Basis For Relief Request (continued): Paragraph 2.4.1.1.1 in the Federal Register (Volume 64, No. 183 dated September 22, 1999) during rule making contained the following statement: "The NRC had determined that this requirement (10 hours of training on an annual basis) was inadequate for two reasons. The first reason was that the training does not require laboratory work and examination of flawed specimens. Signals can be difficult to interpret and as detailed in the regulatory analysis for this rulemaking, experience and studies indicate that the examiner must practice on a frequent basis to maintain the capability for proper interpretation. The second reason is related to the length of training and its frequency. Studies have shown that an examiner's capability begins to diminish within approximately 6 months if skills are not maintained." Thus, the NRC has determined that 10 hours of annual training is not sufficient practice to maintain skills and that annual Ultrasonic training shall be conducted in accordance with 10 CFR 50.55a(b)(2)(xiv) as amended in the Federal Register (Volume 64, No. 183 dated September 22, 1999) in lieu of ASME Section XI, 1995 Edition, 1996 Addenda, Appendix VII, Subparagraph VII-4240." The latest amendment to 10 CFR 50.55a (Volume 67, No. 187 dated September 26, 2002), paragraph (b)(2)(xiv) further recognizes, and permits use of, analyzing prerecorded data from material or welds that contain cracks for meeting annual training requirements. However, these provisions apply to those sites implementing use of the 1999 Addenda through the latest Edition and Addenda referenced in paragraph (b)(2) of the Rule; Monticello is using the 1995 Edition with the 1996 Addenda as the Code of Record for the 4th ISI Interval. Alternative Requirement: Pursuant to 10 CFR 50.55a(a)(3)(i), Monticello proposes to use the more rigorous and detailed annual training requirements of 10 CFR 50.55a(b)(2)(xiv) in lieu of annual training requirements Appendix VII, paragraph VII-4240. Therefore, all personnel qualified for performing Ultrasonic examinations in accordance with Appendix VIII - Supplements ASME Section XI, 1995 Edition, 1996 Addenda shall receive 8 hours of annual hands-on training on specimens that contain cracks or by analyzing prerecorded data from material or welds that contain cracks. This training will be completed no earlier than 6 months prior to performing ultrasonic examinations at the Monticello Nuclear Generating Unit. 1.5-60

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Justification for Granting Relief: This relief improves the performance of Appendix VIII - Supplement examinations by requiring NDE examiner performing Appendix VIII examinations to demonstrate proficiency by analyzing specimens that contain cracks or prerecorded ultrasonic data from material or welds that contain cracks prior to performing actual examinations. The proposed alternative will simplify record keeping, satisfy the needs of maintaining Ultrasonic examiner skills, and also provides an acceptable level of quality and safety. Implementation Schedule: The proposed alternative is requested for the 4th Ten-Year Interval of the Inservice Inspection Program for Monticello Nuclear Generating Unit.

References:

1. ASME Boiler and Pressure Vessel Code, Section XI, 1995 Edition with 1996 Addenda

2. Federal Register, Rules and Regulations, September 22, 1999 (64 FR 51370)

3. Federal Register, Rules and Regulations, September 26, 2002 (67 FR 60520)

4. NRC Letter to Nuclear Management Company, "Relief Request Nos. 3 and 6 for the Fourth 10-Year Interval of the Inservice Inspection Examination Plan" (TAC No. MB6896), March 28, 2003 Status:

Approved on March 28, 2003 for use during the 4th Interval. (See Reference 4 above) 1.5-61

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 7 (Rev.0) Repair/Replacement, 2001 Edition COMPONENT IDENTIFICATION Code Classes: 1, 2, and 3

References:

IWA, IWB, IWC, IWD, and IWF-4000 (IWX-4000) Examination Category: Not Applicable Item Number: Not Applicable

== Description:== Use of the 2001 Edition of Section Xl to Govern Repair/Replacement Activities and Procedures (IWX-4000). Component Numbers: All Class 1, 2, 3 and MC pressure retaining components and their supports. CODE REQUIREMENT IWX-4000 (ASME Section Xl 1995 Edition with the 1996 Addenda, used for Class 1, 2, and 3 components) provides the rules and requirements for repair/replacement activities associated with pressure retaining components and their supports, including appurtenances, subassemblies, parts of a component, core support structures, metal containments and their integral attachments, and metallic portions of Class CC containments and their integral attachments. IWX-4000 (ASME Section XI 1992 Edition with the 1992 Addenda, used for IWE components) provides the rules and requirements for the repair of pressure retaining components and their supports, including appurtenances, subassemblies, parts of a component, core support structures, metal containments and their integral attachments, and metallic portions of Class CC containments and their integral attachments, by welding, brazing, or metal removal. This article also provides the rules and requirements for the specification and construction of items to be used for replacements and installation of replacement items. 10 CFR 50.55a dated September 6, 1996 required the implementation of Subsections IWE and IWL of the 1992 Edition with the 1992 Addenda. 1.5-62

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN BASIS FOR RELIEF The 1992 Edition with the 1992 Addenda to Section Xl made several changes to Articles IWX-4000. Very few of these changes were technical in nature. Instead, the changes restructured some of the requirements, (ie. Combined IWX-4000 and IWX-7000 into one section) clarified others that were difficult to interpret, and eliminated redundant requirements. Of the actual technical changes made, these changes either added enhancements to the program or added requirements not applicable to Monticello. Meeting both the 1995 with the 1996 Addenda and the 1992 with the 1992 Addenda of ASME Section XI would require the maintenance of two separate repair and replacement programs (one for the IWB, IWC, and IWD components per the 1996 Addenda of ASME Section XI and one for the 1992 Addenda for the containment vessel). Duplicate records to demonstrate compliance with the 1996 Addenda and the 1992 Addenda would also be required. This duplication of programs and records increases the man-hours necessary to maintain the Monticello Repair/Replacement Program without providing any increase in quality or safety. The final rule (Federal RegisterNol. 67, No. 187, dated September 26, 2002) incorporates reference to the 1998 Edition through 2000 Addenda. Attached is a reconciliation of the changes made and a comparison of the 2001 Edition to the 2000 Addenda of Section XI. Each change related to Repair/Replacement Activities is addressed in the attachment to show it will be implemented at Monticello. 1.5-63

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T INTERVAL EXAMINATION PLAN ALTERNATE EXAMINATION This alternative is requested in accordance with 10CFR 50.55a(a)(3)(ii). Monticello Nuclear Generating Plant will use the 2001 Edition of ASME Section Xl, to govern Repair/Replacement Procedures (IWX-4000) for Class 1,2,3, and MC pressure retaining components and their supports. Using the requirements contained in the 2001 Edition of ASME Section Xl for Repairs/Replacements at the Monticello Nuclear Generating Plant will maintain the safety of the plant. The following table indicates the implementation of the 2001 Edition for Repair/Replacement Activities. Article To_9. Bases IWA-1000 Scope and Responsibility 1996 Addenda IWA-2000 Examination and Inspection 1996 Addenda IWA-3000 Acceptance Standards 1996 Addenda IWA-4000 Repair/Replacements 2001 Edition IWA-5000 Pressure Tests (Periodic) 1996 Addenda IWA-5000 Pressure Tests (Repair/Replacements) 2001 Edition IWA-6000 Records 2001 Edition IWA-9000 Glossary 2001 Edition IWB-1000 Scope and Responsibility 1996 Addenda IWB-2000 Examination and Inspection 1996 Addenda IWB-3000 Acceptance Standards 1996 Addenda IWB-5000 Pressure Tests (Periodic) 1996 Addenda IWB-5000 Pressure Tests (Repair/Replacements) 2001 Edition IWC-1000 Scope and Responsibility 1996 Addenda IWC-2000 Examination and Inspection 1996 Addenda IWC-3000 Acceptance Standards 1996 Addenda IWC-5000 Pressure Tests (Periodic) 1996 Addenda IWC-5000 Pressure Tests (Repair/Replacements) 2001 Edition IWD-1000 Scope and Responsibility 1996 Addenda IWD-2000 Examination and Inspection 1996 Addenda IWD-3000 Acceptance Standards 1996 Addenda IWD-5000 Pressure Tests (Periodic) 1996 Addenda IWD-5000 Pressure Tests (Repair/Replacements) 2001 Edition 1.5-64

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Article Topic Bases IWE-1000 Scope and Responsibility 1992 Addenda IWE-2000 Examination and Inspection 1992 Addenda IWE-3000 Acceptance Standards 1992 Addenda IWE-5000 Pressure Tests (Periodic) Appendix J IWE-5000 Pressure Tests (Repair/Replacements) 2001 Edition w/ Appendix J IWF-1000 Scope and Responsibility 1996 Addenda IWF-2000 Examination and Inspection 1996 Addenda IWF-3000 Acceptance Standards 1996 Addenda IWF-5000 Snubber Examinations and Tests 1996 Addenda APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year interval of the Inservice Inspection Program for Monticello Nuclear Generating Plant. 1.5-65

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN Certificate of Reconciliation The Certificate of Reconciliation provides the basis for revisions to the Monticello Nuclear Generating Plant's (MNGP) ASME Section XI "Repair/Replacement Program" (4AWI-09.04.03) in order to meet the 2001 Edition of ASME Section Xl. On September 9, 1996, the Nuclear Regulatory Commission (NRC) issued a revision to 10 CFR 50.55a, implementing subsections IWE and IWL (IWL "Requirements for Class CC Concrete Components of Light-Water Cooled Plants" is not applicable to the Monticello Nuclear Generating Plan) is not of the 1992 edition, including the 1992 addenda of Section Xl of the ASME Code. This required utilities to develop and implement a program for the examination of containments by September 9, 2001. Additionally, it required implementation of an IWE/IWL repair/replacement program effective September 9, 1996. The NMC is updating the MNGP Inservice Inspection (ISI) Program for the fourth ten-year interval to meet the 1995 Edition with the 1996 Addenda. Because of the hardship to maintain two separate Repair/Replacement Programs, this alternative is proposed to allow the use of the 2001 Edition of ASME Section XI. This reconciliation is completed to provide justification for allowing the use of the 2001 Edition for Class 1, 2, 3 and MC pressure retaining components and their supports. The current revision of 10CFR50.55a requires ASME Section Xl Programs to follow the 1995 Edition as amended by the 1996 Addenda of ASME Section XI for Class 1, 2, and 3 components and the 1992 Edition as amended by the 1992 Addenda for Class MC components. There are some general issues to discuss prior to delineating the specific changes that have been made to the ASME Section Xl Code (2000 Addenda to 2001 Edition). By performing the reconciliation from the 1992 Addenda, the reconciliation from the 1996 Addenda is covered as well.

1) The NRC has reviewed and approved with some exceptions the 1998 Edition through 2000 Addenda of the code. This has been included in the Final Rule (dated September 26, 2002). Those specific exceptions made to the rules for repair/replacement activities are included in the implementation of the 2001 Edition.

2) The NMC ISI requirements for MNGP will be based on the 1995 Edition as amended by the 1996 Addenda.

3) The Periodic Pressure Testing requirements will be based on the 1995 Edition as amended by the 1996 Addenda. While the pressure testing requirements for repair/replacement activities will be based on the 2001 Edition.

1.5-66

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN

4) The reconciliation attached addresses the changes contained within the IWA-4000 paragraphs. In addition, any significant changes identified within any related requirements are addressed.

Each change is categorized as: Editorial (E) - Those changes that are of an editorial nature like typographical errors or misspelled words. Technical Significant (TS) - Those changes that effect the technical requirements and either reduce or increase those requirements. These changes are described in more detail as to their applicability to MNGP. Technical (T) - Technical changes that are only used for clarification of an existing requirement. Non-significant (TN) - Those changes that are not technical in nature, but could not be classified as editorial or just a relocation of existing requirements. ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation 2001 Edition IWA-41 10(b) Revised to insert the words "Thermal metal removal" to TS clarify that thermal metal removal activities fall within (Note 1) the scope of IWA-4000 IWA-4230 This was added to relocate the requirements of IWA- TN 4451 "Helical Coil Threaded Inserts". This relocation places these requirements in IWA-4200 "Material" which is appropriate since they deal primarily with helicoil material requirements. IWA-4400 Retitled to "Welding, Brazing, Metal Removal, and TN Installation". This was retitled specify that metal removal rules apply to all Section XI repair activities. IWA-4410 This was rewritten to make its contents consistent with T the revised title. It is also revised to clarify that mechanical metal removal not associated with defect removal is not within the scope of IWA-4400. IWA-441 1 This is a new paragraph titled "Welding and Brazing". T This new paragraph serves to consolidate the requirements applicable only to welding and brazing, and to clarify the distinction between when Construction Code requirements apply and when IWA-4400 requirements apply. 1.5-67

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TM INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4412 This is a new paragraph titled "Defect Removal". This T new paragraph serves to clarify that the requirements of IWA-4420 are mandatory for all defect removal activities, and to direct the user to these requirements. IWA-4413 This is a new paragraph titled "Thermal Metal T Removal". This new paragraph serves to clarify that the requirements of IWA-4461 are mandatory for all thermal metal removal activities, and to direct the user to these requirements. IWA-4420 Revised title to "Defect Removal Requirements". This TN revision makes the title consistent with the changes described below. IWA-4421 Revised to "General Requirements" with the following TN specific changes: i) The second sentence of para. (a) is moved to IWA-4421. ii) The last sentence of para. (a) is dropped, since IWA-4412 now invokes requirements for defect removal and associated NDE. iii) The remainder of the text from IWA-4421 (a), (b), and (c) is reorganized and moved to IWA-441 1(a) and (b), except that the final sentence, "A Report of Reconciliation shall be prepared." has been deleted to make this paragraph consistent with the changes made. 1.5-68

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4422 Revised to "Defect Evaluation and Examination". This TN change makes the title consistent with the content changes described for IWA-4422.1. IWA-4421.1 was changed as follows: i) Title changed to "Defect Evaluation" ii) The first sentence of IWA-4422.1(a) is deleted. The requirement that the defect removal process comply with 4421 is unneeded, as it is redundant with the new IWA-4421 (a) through (d) iii) The third sentence of IWA-4422.1(a) is deleted. This deleted sentence stated, "The component is acceptable for continued service if the resulting section thickness created by the cavity is at least the minimum required thickness." This sentence is deleted for two reasons:

1) It is redundant with the proceeding sentence in IWA-4422.1(a) and

2) It implies that all defect removal operations involve metal removal and creation of a cavity. Several repair types do not involve metal removal or cavity creation.

IWA-4430 This paragraph was deleted. Its contents were TN reworded and relocated to IWA-441 1(f). IWA-4450 This was deleted from the Code in its entirety. Use of TN the ASME Code to mandate compliance with manufacturer's recommendations is considered inappropriate and constitutes the basis for deleting this requirement. IWA-4451 This was renumbered as IWA-4134 and is relocated TN accordingly. This relocation is consistent with the contents of IWA-4451, which address installation of helical-coil threaded inserts. The installation of helical-coil threaded inserts does not fall within the scope of IWA-4400. Table IWA- This table was revised to delete reference to P-1 E 4461.1-1 materials. This revision is editorial in nature, and is incorporated to make Table-4461.1 consistent with IWA-4461.1 and 4461.2. The revision for preheat of P-1 materials prior to thermal metal removal was deleted by a prior revision to IWA-4460, but Table IWA-4461.1 was not revised to reflect this revision. 1.5-69

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4461.4 Title was revised to "Alternatives to Mechanical TS Processing". This change is necessary to (Note 1) accommodate a newly added alternative to mechanical processing after thermal metal removal, which is addressed in IWA-4461.4.2. The two alternatives are addressed in new paragraphs IWA-4461.4.1 and IWA-4461.4.2. IWA-4461.4.1 describes the qualification process whereby thermal metal removal is permitted without subsequent mechanical processing. No changes were made to these requirements other than paragraph renumbering. IWA-4461.4.2 describes the evaluation process where by thermal metal removal is permitted without subsequent mechanical processing. This alternative enables an Owner to perform a documented evaluation to determine whether elimination of mechanical processing is acceptable. A footnote was added to define the term "Mechanical Processinq" IWA-4462 This was revised to "Mechanical Defect Removal TN Processes". IWA-4462(a) is replaced with wording that clarifies the applicability of this paragraph to defect removal activities only. IWA-4500 Title changed to "Examination and Testing" TN IWA-4520(a) This was revised to add two specific exceptions. TS These exceptions are as follows: (Note 1) i) IWA-4521 (a)(1) was revised to exempt Class 3 base material repairs from volumetric examination when full-penetration butt welds in the same location do not require volumetric examination. ii) IWA-4521(a)(2) was revised to invoke the examination requirements of IWA-4600 and 4700 in lieu of Construction Code examinations for all repairs using IWA-4600 or 4700. This exception invokes IWA-4600 NDE requirements for all IWA-4600 welding, and invokes IWA-4700 NDE requirements for IWA-4700 welding. This change clarifies that use of IWA-4600 and IWA-4700 welding alternatives and also mandates use of the associated NDE requirements. 1.5-70

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4600(a) This was revised to delete the words "and TN nondestructive examination requirements". These words are deleted for clarification. The underwater welding alternative requirements of IWA-4660 apply in lieu of Construction Code requirements; however, IWA-4660 invokes Construction Code NDE requirements. Since IWA-4660 invokes Construction Code NDE requirements, it is incorrect to state that 4660's requirements are "in lieu of' Construction Code NDE requirements. IWA-4610 This was revised to "General Requirements for TN Temperbead Welding of all Materials" 1.5-71

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4611 IWA-4611.1 (a), (b) and (c) were deleted and TS alternative requirements were added. (Note 1) i) The defect removal requirements of 4611.1(a) have been moved to IWA-4421.1. The existing 4611 (a), therefore is redundant and is no longer needed. ii) The IWA-4611.1 (b) requirement that "the original defect shall be removed" has been revised to match what the original intent was by the words "the original defect shall be reduced in size to a level that meets the applicable Construction Code NDE acceptance criteria. The requirement for compliance with Construction Code acceptance criteria was added to IWA-4624.2, 4634.2, 4644.2 and 4654.2. iii) The IWA-461 1.1 (c) requirements for the Repair/Replacement Program and Plan are redundant with IWA-4150. Deletion of this paragraph eliminates this redundancy. IWA-4611.1(a), (b), and (c) additions are as follows: i) IWA-461 1.1 (a) now consists of a reference to IWA-4422.1. Use of this reference enables all defect removal activities to rely on a single set of defect removal requirements, eliminating redundancy and reducing complexity. ii) IWA-461 1.1 (b) now includes a reference to the NDE requirements applicable to each of the various repair methods authorized by IWA-4600. This reference is needed because each repair method includes its own unique NDE requirements, and these requirements are different from those used for welding and brazing activities that are not within the scope of IWA-4600. i) IWA-461 1.1(c) now includes a reference to the thermal metal removal requirements of IWA-4413. This reference is needed because the requirements for thermal metal removal apply to all IWA-4600 processes, and because thermal metal removal requirements have been consolidated into IWA-4461, which is referenced by IWA-4413. 1.5-72

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4611 IWA-4611.2(a) was changed as follows: TS (cont'd) i) In the first line, the word "grinding" is replace (Note 1) with "processing". This change is necessary to acknowledge that final grinding is not always required for defect removal. ii) In the sixth line, "IWA-3000" is replaced with "IWB-3500, IWC-3500, or IWD-3000". This change adds a direct reference to the NDE acceptance criteria tables of IWB and IWC (Note: Since IWD tables are 'in course of preparation', the IWD-3000 reference invokes permission to use IWB requirements). By referencing these tables, IWA-4611.2(a) clarifies that the indication may be considered

                           'reduced to an acceptable level' only when the respective table's acceptance criteria has been met.

A new sentence states, "For supports and containment vessels, the provisions of IWA-4422.1 (b) may be used." This sentence is added because ASME Section III Subsections NE and NF do not contain surface examination acceptance criteria for base materials, therefore, no criteria exist for these exams. IWA-4422.1(b) provides an evaluation alternative for these applications. IWA-4620 Title was revised to "Temperbead Welding of Similar TN Materials" IWA-4624 A) IWA-4624.1(a) was added to invoke IWA- TS 4611.2(a), which mandates surface examination (Note 1) prior to welding for all temperbead repairs. This paragraph is added to assure that Section XI, IWA-3000 acceptance criteria is used for NDE of existing metal. B) IWA-4624.2 invokes Construction Code or Section III NDE acceptance criteria on in-processing welding and on the final weld. This assures that all newly installed weld metal complies with Construction Code requirements during installation and at the time of weld completion. IWA-4630 Title was revised to "Temperbead Welding of TN Dissimilar Materials" IWA-4634 This was revised similar to that discussed in IWA-4624 TS above. (Note 1) IWA-4644 This was revised similar to that discussed in IWA-4624 TS above. (Note 1) 1.5-73

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4654 This was revised similar to that discussed in IWA-4624 TS above. (Note 1) IWA-4666 This was revised to impose Construction Code NDE TS requirements on completed underwater welds. This (Note 1) paragraph also provides an alternative to these NDE requirements when the underwater environment renders normal NDE practical. IWA-4711.4 This was revised to clarify the final visual examination TS was to be a VT-1 examination. (Note 1) IWA-4712 This was revised to make its wording consistent with TN IWA-471 1. This change states that use of these requirements is mandatory for Class 1 applications, but use of these requirements in Class 2 and Class 3 applications is also acceptable. IWA-4721.1 This was revised to make its wording consistent with TN IWA-471 1. This change states that use of these requirements is mandatory for Class 1 applications, but use of these requirements in Class 2 and Class 3 applications is also acceptable. IWA-4131.1(a) The change deleted the word "welded" located in TS before the reference to plugs. (Note 2) IWA-4713 This revision adds new requirements for qualification TS of Class 1 mechanical tube plugs. These (Note 2) requirements represent a compilation of the standards and methods that have been used for twenty years to design, qualify, and install steam generator tube plugs. They have proven to provide safe installation and service for mechanical steam generator tube plugs. These requirements include development and qualification of the plug design and of a Plugging Procedure Specification (PPS), and performance qualification for individuals who install the tube plugs 1.5-74

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation IWA-4132 This revision deletes the requirement for pressure TS testing and VT-2 visual examination of relief valves (Note 3) rotated from stock and installed by mechanical means. In the 1999 Addenda, the requirement to pressure test mechanical joints made in installation of pressure retaining items was deleted from IWA-4540, because Owner's operation and maintenance personnel post-installation inspections are adequate without an additional Code-required examination. With the deletion of pressure tests for mechanical connections, a similar exemption is warranted for installation of relief valves by mechanical means. The revision also clarifies that no other IWA-4000 requirements apply to rotation of snubber and relief valves, except those of IWA-4132, and clarifies that use of an ANII is not required. This revision incorporates the provisions of Case N-508-2, "Rotation of Serviced Snubbers and Pressure Relief Valves for the Purpose of Testing, Section XI, Division 1." NOTE 1. It is important to apply the correct acceptance criteria to each repair/replacement activity completed. As reflected in the Final Rule, the NRC recognizes the difference between the NDE of the Construction Codes and ASME Section XI. The other changes were made to clarify the rules as they apply to the mechanical removal process and of a non-technical nature with reordering of paragraphs or moving of requirements to different paragraphs. The MNGP Repair/Replacement Program incorporates these requirements. NOTE 2. NMC has determined that it is important to have all special processes qualified and/or demonstrated to verify the application. Because of the elimination of the word "welded," the alternative requirements provided in IWA-4131.1 are no longer applicable to any tube plugging (mechanical or welded). The MNGP Repair/Replacement Program incorporates these provisions. NOTE 3. Since the code no longer requires a VT-2 Examination on installation of mechanical joints, the NMC has determined that the installation of relief valves rotated from MNGP stock and installed by mechanical means would not require a VT-2 examination. 1.5-75

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN NRC Limitation / NMC Commitments: The NRC staff requires implementation of paragraph IWA-4540(c) of the 1998 edition in lieu of that of the 2001 edition when implementing the 2001 edition of ASME Code, Section XI, Article IWX-4000 for repair and replacement activities. The NRC is planning revisions to the Final rule which may have an effect on this Relief Request. NMC has committed to implement the limitations and modifications to the 1998 edition through 2000 addenda of the ASME Code, Section XI, as stated in 10 CFR 50.55a(b)(2) when implementing the 2001 edition. NMC has further committed to implement any limitations and modifications to the 2001 edition of the ASME Code for its repair and replacement program when the NRC incorporates, by reference, this edition into the regulations.

References:

1. NRC Letter to Nuclear Management Company, "Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No.7" (TAC No. MB6897),

October 3, 2003

2. NRC Letter to Nuclear Management Company, "Issuance of Corrected Page Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No.7" (TAC No. MB6897), December 31, 2003 Status:

Approved on October 3, 2003 for use during the 4th Interval, (See References 1 and 2 above) 1.5-76

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit 1 - ISI Relief Request No. 8 (Rev. 0) One-time Relief, Class I Pressure Test at less than nominal operating pressure. COMPONENT IDENTIFICATION Code Class: 1

References:

IWA-4540(c) IWA-5211 (a) Examination Category: Not Applicable Item Number: Not Applicable

== Description:== System Leakage Pressure Test & accompanying VT-2 Examination at nominal operating pressure following Repair-Replacement activities involving Mechanical Joints. Components: Main Steam Safety Relief Valve Assemblies CODE REQUIREMENTS The 1995 Edition of American Society of Mechanical Engineers (ASME) Section XI with the 1996 Addenda, paragraph IWA-5120(a) states: "Items subjected to repair/replacement activities shall be pressure tested when required by IWA-4500." Paragraph IWA-4540(c) states: "Mechanical joints made in installation of pressure retaining items shall be pressure tested in accordance with IWA-521 1(a)." Paragraph IWA-521 1(a) states: "A system leakage test conducted during operation at nominal operating pressure, or when pressurized to nominal operating pressure and temperature." Paragraph IWB-5210(b) states: "The system pressure tests and visual examinations shall be conducted in accordance with IWA-5000 and this Article. The contained fluid in the system shall serve as the pressurizing medium." 1.5-77

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN BASIS FOR RELIEF Nuclear Management Company, LLC (NMC) Monticello Nuclear Generating Plant (MNGP) recently completed a refueling outage on May 26, 2003. During the refueling outage, MNGP completed the system leakage test required by American Society of Mechanical Engineers (ASME) Section XI, Table IWB-2500-1, Category B-P, Item 15.10 and 10 CFR Part 50 Appendix G, Section IV.A.2.d. Following restart of the unit, the "B" and "G" main steam safety relief valve assemblies (SRVs) have indicated leakage, as determined by higher than normal temperatures in their respective discharge tailpipes. MNGP has decided to conduct a planned unit shutdown and enter a maintenance outage to replace the affected SRV assemblies. The SRV assemblies are connected to the main steam piping with a bolted, mechanical joint. Replacing them for maintenance is considered a Repair-Replacement activity under the rules of ASME Section XI, 1995 Edition with the 1996 Addenda which is the current code of record for the 4th 10-Year ISI Interval. Following repair-replacement, a system leakage test is required by IWA-4540(c). The system leakage test at the nominal pressure associated with the reactor at 100% power would be approximately 1000 psig. MNGP has identified three methods for performing the system leakage test on the mechanical joints associated with the repair-replacement activity that meet the requirements identified above. Several conditions associated with such testing represent an imposition on personnel safety, personnel radiation exposure, and challenges to the normal mode and manner of equipment operation. Method No. 1 would perform the pressure test and VT-2 exam during normal startup procedures. During normal startup with normal power ascension, nominal operating pressure of 1000 psig is reached at a reactor power level of approximately 75%. If access to containment were permitted at this power level, personnel would be exposed to excessive radiation levels, including significant exposure to neutron radiation fields, which is contrary to current station ALARA practices. Establishing the 1000 psig test condition at a more moderate power level (e.g. during plant startup at approximately 7% reactor power) and in the manner needed to address radiation concerns would require altering the normal operational mode of the steam pressure control system. During the performance of plant startup procedures, the electric and mechanical pressure regulator (EPR and MPR) set points are established within their normal operational ranges (approximately 918 psig). Their primary function is to regulate 1.5-78

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4"' INTERVAL EXAMINATION PLAN the main steam system pressures as sensed near the inlet of the high-pressure turbine. Reactor pressure control at the nominal 1000 psig is achieved at higher reactor power levels as a function of the pressure control system and the induced differential pressure across the main steam isolation valves and main steam piping. While it is technically feasible to manipulate these controls to establish the nominal system pressure of 1000 psig at lower power levels, doing so will affect core reactivity and could challenge plant safety systems, such as the reactor protection system (RPS). MNGP has not previously operated the EPR and MPR in this manner. Changing the setpoints outside of the normal range of operation for the purpose of performing this test at nominal operating pressure poses several operational challenges. The lack of experience and predictability of setting pressure regulators outside the normal range of operation could adversely impact personnel and reactor safety. Method No. 2 implements the use of the reactor pressure boundary leakage test which meets the requirements of Table IWB-2500-1, Category B-P, Item 15.10: the reactor pressure vessel (RPV) is filled with coolant and the steam lines are flooded to provide a water-solid condition. Use of this method would result in multiple operational challenges. During a maintenance outage, pressurization for the test would be provided by decay heat and the reactor recirculation pumps. To support the pressurization evolution, the normal decay heat removal system, residual heat removal (RHR) shutdown cooling, would be required to be removed from service and isolated from the vessel to be pressurized. This system is not designed to withstand pressures greater than 185 psig. Thus, the remaining system available for decay heat removal is the reactor water cleanup system (RWCU). Application of ANSI /ANS-1994 decay heat code results in a significant level of decay heat load. The ratio of decay heat input versus the heat removal capacity provided by RWCU is approximately 4:1. Therefore, the decay heat generated by the reactor core will surpass the capacity of RWCU. The heat up rate of the vessel water will cause the temperatures to surpass 2120 F prior to the initiation of the inspections. Method No. 2 would present several operational challenges. The pressure increase would be obtained by balancing the flow into the vessel, which is provided by the control rod drive (CRD) system, with the flow out of the vessel provided by the RWCU system via the dump flow control valve and flow controller. This is the method used during refueling outages to complete the RPV system leakage test. A failure of a non-safety related component, such as the dump valve or flow controller, would cause the interruption of dump flow and 1.5-79

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN would cause the RPV pressure to increase. The RPV pressure would increase until operator action would require the operating CRD pump to be tripped. Due to the amount of decay heat being generated and the RWCU systems heat removal capacity, it is questionable whether the RPV would depressurize and may in fact continue to pressurize until further operator action would be required to depressurize the RPV. Operator actions may include one or more of the following: reestablishing RWCU dump flow; if the failure mechanism was no longer present, opening the main steam line drain valves, SRVs, or head vent line. Any of the last 3 of these actions would probably cause a rapid depressurization transient on the RPV. Extensive valve manipulations, system lineups, and procedural controls are required in order to heat up and pressurize the primary system to establish the necessary test pressure, during plant outage conditions, without the withdrawal of control rods. This test is expected to take approximately 1 day of outage time, and the additional valve lineups and system reconfigurations necessary to support this test impose an additional challenge to the affected systems. A normal plant startup then occurs, after completion and subsequent recover from the test procedure. Method No. 3 would maintain the RPV at its normal level and use decay heat to produce sufficient steam pressure to conduct the test at nominal operating temperature. At the projected time of shutdown for the maintenance outage, MNGP will have a runtime of approximately three weeks since startup from the Cycle 21 refueling outage. The maintenance of the SRV assemblies is projected to be completed within approximately 50 hours after plant shutdown. While the decay heat load is too high for the water-solid method discussed above, there is not sufficient decay heat available to perform the test within a reasonable time period to support completion of the maintenance outage. It would require a minimum of 25 hours to reach the pressure of 1000 psig needed to perform the test required by the Code based upon decay heat projections. Each of the methods discussed above presents a hardship or unusual difficulty to NMC. 1.5-80

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN PROPOSED ALTERNATIVE PROVISIONS Pursuant to 10CFR50.55a(a)(3)(ii), compliance with the required system leakage test under IWA-4540(c) would result in a hardship or unusual difficulty without a compensating increase in the level of quality and safety. The NMC proposes to perform a VT-2 examination on the mechanical joints of the SRV assemblies during the normal operational start-up sequence at a minimum of 900 psig following a 10 minute hold time (for uninsulated components) in lieu of the nominal operating pressure associated with 100% reactor power of approximately 1000 psig. In addition, if there is an unplanned shutdown with a drywell entry before the next refueling outage, another inspection of these bolted connections will be performed to look for any evidence of leakage. Application of this alternative test maintains reasonable levels of personnel safety and reduces the opportunity for the introduction of undesirable operational challenges. While NMC does not expect that leakage will occur, any leakage at the bolted connection to be related to the differential pressure across the connection. A 10% reduction in test pressure is not expected to result in the arrest of a leak that would occur at nominal operating pressure. In the event that leakage would occur at the mechanical joints at higher pressures associated with 100% reactor power, leakage from these mechanical connections would be detected by the drywell monitoring systems, which include drywell pressure monitoring, the containment atmosphere monitoring system (CAM), and the drywell floor drain sumps. Leakage monitoring is required by Monticello Technical Specifications. This alternative method for a system leakage test is particularly applicable for the MNGP maintenance outage, which is of limited scope, and where the only components on the primary system that are being replaced are the main steam "B" and "G" safety relief valve assemblies attached via mechanical connections. The NRC has authorized use of a similar alternative system leakage test method for the Cooper Nuclear Station in 1998 which permitted them to perform a system leakage test at a minimum of 900 psig following replacement of their SRV topworks, a mechanical joint, during a mid-cycle maintenance outage. The approval letter for the Cooper relief request was dated February 26, 1998. 1.5-81

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN CONCLUSION: In summary, the proposed NMC alternative is to perform the system leakage test and VT-2 examination at 900 psig minimum after a 10 minute hold time in lieu of the pressure testing requirements of the 1995 Edition of ASME Section XI with the 1996 Addenda for mechanical joints following repair-replacement activities. In addition, if there is an unplanned shutdown with a drywell entry before the next refueling outage, another inspection of these bolted connections will be performed to look for any evidence of leakage. Considering the hardship and unusual difficulty in performing the available methods for satisfying the code requirements and the ability to detect leakage in primary containment should it occur, this alternative will provide an acceptable verification of the leak integrity of the mechanical joint without putting the plant in a non-conservative operational condition and without unnecessary radiation exposure and safety challenges to personnel. PERIOD FOR WHICH RELIEF IS REQUESTED NMC requests NRC authorization to perform the proposed alternative test on a one-time basis for the system leakage tests following repair/replacement activities on the mechanical joints of SRVs "B"and "G"during the planned maintenance outage. STATUS Approved on June 13, 2003 for use during the 4th Interval, NRC Letter, "Monticello Nuclear Generating Plant - One-Time Inservice Inspection Program Plan Relief Request No. 8 For Leak Testing The "B"And "G" Main Steam Safety Relief Valves" (TAC No. MB9538) 1.5-82

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 9 (Rev. 0) (Fleet Relief Request) Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(i) Appendix VIII - Supplement 10 SYSTEM/COMPONENT(S) FOR WHICH RELIEF IS REQUESTED: Pressure Retaining Piping Welds subject to examination using procedures, personnel, and equipment qualified to ASME Section Xl, Appendix VIII, Supplement 10 criteria. CODE REQUIREMENTS: The following statements or paragraphs are from ASME Section XI, Appendix VIII, Supplement 10 and identify the specific requirements that are included in this request for relief. Item 1 - Paragraph 1.1(b) states in part - Pipe diameters within a range of 0.9 to 1.5 times a nominal diameter shall be considered equivalent. Item 2 - Paragraph 1.1(d) states - All flaws in the specimen set shall be cracks. Item 3 - Paragraph 1.1 (d)(1) states - At least 50% of the cracks shall be in austenitic material. At least 50% of the cracks in austenitic material shall be contained wholly in weld or buttering material. At least 10% of the cracks shall be in ferritic material. The remainder of the cracks may be in either austenitic or ferritic material. Item 4 - Paragraph 1.2(b) states in part - The number of unflawed grading units shall be at least twice the number of flawed grading units. Item 5 - Paragraph 1.2(c)(1) and 1.3(c) state in part - At least 1/3 of the flaws, rounded to the next higher whole number, shall have depths between 10% and 30% of the nominal pipe wall thickness. Paragraph 1.4(b) distribution table requires 20% of the flaws to have depths between 10% and 30%. 1.5-83

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Item 6 - Paragraph 2.0 first sentence states - The specimen inside surface and identification shall be concealed from the candidate. Item 7 - Paragraph 2.2(b) states in part - The regions containing a flaw to be sized shall be identified to the candidate. Item 8 - Paragraph 2.2(c) states in part - For a separate length sizing test, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. Item 9 - Paragraph 2.3(a) states - For the depth sizing test, 80% of the flaws shall be sized at a specific location on the surface of the specimen identified to the candidate. Item 10 - Paragraph 2.3(b) states - For the remaining flaws, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region. Item 11 - Table VIII-S2-1 provides the false call criteria when the number of unflawed grading units is at least twice the number of flawed grading units. RELIEF REQUESTED Relief is requested to use the following alternative requirements for implementation of Appendix VIII, Supplement 10 requirements. They will be implemented through the Performance Demonstration Initiative (PDI) Program. A copy of the proposed revision to Supplement 10 is attached. It identifies the proposed alternatives and allows them to be viewed in context. It also identifies additional clarifications and enhancements for information. It has been submitted to the ASME Code for consideration and as of September 2002 had been approved by the NDE Subcommittee. BASIS FOR RELIEF Item 1 -The proposed alternative to Paragraph 1.1(b) states: "The specimen set shall include the minimum and maximum pipe diameters and thicknesses for which the examination procedure is applicable. Pipe diameters within 1/2 in. (13 mm) of the nominal diameter shall be considered equivalent. 1.5-84

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Pipe diameters larger than 24 in. (610 mm) shall be considered to be flat. When a range of thicknesses is to be examined, a thickness tolerance of +25% is acceptable." Technical Basis - The change in the minimum pipe diameter tolerance from 0.9 times the diameter to within 1/2 inch of the nominal diameter provides tolerances more in line with industry practice. Though the alternative is less stringent for small pipe diameters they typically have a thinner wall thickness than larger diameter piping. A thinner wall thickness results in shorter sound path distances that reduce the detrimental effects of the curvature. This change maintains consistency between Supplement 10 and the recent revision to Supplement 2. Item 2 - The proposed alternative to Paragraph 1.1(d) states: "At least 60% of the flaws shall be cracks, the remainder shall be alternative flaws. Specimens with IGSCC shall be used when available. Alternative flaws, shall meet the following requirements: (1) Alternative flaws, if used, shall provide crack-like reflective characteristics and shall only be used when implantation of cracks would produce spurious reflectors that are uncharacteristic of service-induced flaws. (2) Alternative flaw mechanisms shall have a tip width no more than 0.002 in. (.05 mm). Note, to avoid confusion the proposed alternative modifies instances of the term "cracks" or "cracking" to the term "flaws" because of the use of alternative flaw mechanisms." Technical Basis - As illustrated below, implanting a crack requires excavation of the base material on at least one side of the flaw. While this may be satisfactory for ferritic materials, it does not produce a useable axial flaw in austenitic materials because the sound beam, which normally passes only through base material, must now travel through weld material on at least one side, producing an unrealistic flaw response. In addition, it is important to preserve the dendritic structure present in field welds that would otherwise be destroyed by the implantation process. To resolve these issues, the proposed alternative allows the use of up to 40% fabricated flaws as an alternative flaw mechanism under controlled conditions. The fabricated flaws are isostatically compressed which produces ultrasonic reflective characteristics similar to tight cracks. 1.5-85

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Mechanical fatigue crac re" --- in Base material Item 3 - The proposed alternative to Paragraph 1.1 (d)(1) states: "At least 80% of the flaws shall be contained wholly in weld or buttering material. At least one and no more than 10% of the flaws shall be in ferritic base material. At least one and no more than 10% of the flaws shall be in austenitic base material." Technical Basis - Under the current Code, as few as 25% of the flaws are contained in austenitic weld or buttering material. Recent experience has indicated that flaws contained within the weld are the likely scenarios. The metallurgical structure of austenitic weld material is ultrasonically more challenging than either ferritic or austenitic base material. The proposed alternative is therefore more challenging than the current Code. Item 4 - The proposed alternative to Paragraph 1.2(b) states: "Personnel performance demonstration detection test sets shall be selected from Table VIII-S10-1. The number of unflawed grading units shall be at least 1-1/2 times the number of flawed grading units." Technical Basis - Table VIII-S10-1 provides a statistically based ratio between the number of unflawed grading units and the number of flawed grading units. The proposed alternative reduces the ratio to 1.5 times. This reduces the number of test samples to a more reasonable number from the human factors perspective. However, the statistical basis used for screening personnel and procedures is still maintained at the same level with competent personnel being successful and less skilled personnel being unsuccessful. The acceptance criteria for the statistical basis are in Table VIII-S$ 0-1. Item 5 - The proposed alternative to the flaw distribution requirements of Paragraph 1.2(c)(1) (detection) and 1.3(c) (length) is to use the Paragraph 1.4(b) (depth) distribution table (see below) for all qualifications. Flaw Depth Minimum (% Wall Thickness) Number of Flaws 10-30% 20% 31-60% 20% 61-100% 20% 1.5-86

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Technical Basis - The proposed alternative uses the depth sizing distribution for both detection and depth sizing because it provides for a better distribution of flaw sizes within the test set. This distribution allows candidates to perform detection, length, and depth sizing demonstrations simultaneously utilizing the same test set. The requirement that at least 75% of the flaws shall be in the range of 10 to 60% of wall thickness provides an overall distribution tolerance yet the distribution uncertainty decreases the possibilities for testmanship that would be inherent to a uniform distribution. It must be noted that it is possible to achieve the same distribution utilizing the present requirements, but it is preferable to make the criteria consistent. Item 6 - The proposed alternative to Paragraph 2.0 first sentence states: "For qualifications from the outside surface, the specimen inside surface and identification shall be concealed from the candidate. When qualifications are performed from the inside surface, the flaw location and specimen identification shall be obscured to maintain a "blind test"." Technical Basis - The current Code requires that the inside surface be concealed from the candidate. This makes qualifications conducted from the inside of the pipe (e.g., PWR nozzle to safe end welds) impractical. The proposed alternative differentiates between ID and OD scanning surfaces, requires that they be conducted separately, and requires that flaws be concealed from the candidate. This is consistent with the recent revision to Supplement 2. Items 7 and 8 - The proposed alternatives to Paragraph 2.2(b) and 2.2(c) state: "... containing a flaw to be sized may be identified to the candidate." Technical Basis - The current Code requires that the regions of each specimen containing a flaw to be length sized shall be identified to the candidate. The candidate shall determine the length of the flaw in each region (Note, that length and depth sizing use the term "regions" while detection uses the term "grading units" - the two terms define different concepts and are not intended to be equal or interchangeable). To ensure security of the samples, the proposed alternative modifies the first "shall" to a "may" to allow the test administrator the option of not identifying specifically where a flaw is located. This is consistent with the recent revision to Supplement 2. 1.5-87

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Items 9 and 10 - The proposed alternative to Paragraph 2.3(a) and 2.3(b) state: ". regions of each specimen containing a flaw to be sized may be identified to the candidate." Technical Basis - The current Code requires that a large number of flaws be sized at a specific location. The proposed alternative changes the "shall" to a "may" which modifies this from a specific area to a more generalized region to ensure security of samples. This is consistent with the recent revision to Supplement 2. It also incorporates terminology from length sizing for additional clarity. Item 11 - The proposed alternative modifies the acceptance criteria of Table VIII-S2-1 as follows: TABLE VIII-S7-.1 PERFORMANCE DEMONSTRATION DETECTION TEST ACCEPTANCE CRITERIA Detection Test False .Cal Test Acceptance Critera Acceptance Criteria No. of No. of Maximum Flawed Minimum Unflawed Number Grading Detection Grading of False Units Criteria Units Calls

               .5                ,5             10 212 7                  6             141
              .8                  7             16 ... ....            2 9                 :7            :1                      2 10                   8             2o-9 15                3-- 2 11                   9             2t--17                 3-3 12                    9             24- 18                 3-- 3 13                  10             2-20                   4 3 14                  10             2--23 15                  11             3-&23                  5-    3 16                  12.            3 24           "             4 17                  12             34- 26         ,       6-4 18                  1327                                  -     4 193                                      29                  -- 4 20                  14                       08--        :--

0 1.5-88

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Technical Basis - The proposed alternative is identified as new Table VIl1-SlO-1 above. It was modified to reflect the reduced number of unflawed grading units and allowable false calls. As a part of ongoing Code activities, Pacific Northwest National Laboratory (PNNL) has reviewed the statistical significance of these revisions and offered the revised Table S10-1. ALTERNATIVE EXAMINATION In lieu of the requirements of ASME Section XI, Appendix VIII, Supplement 10, the proposed alternative shall be used. The proposed alternative is described in the enclosure. JUSTIFICATION FOR GRANTING RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), approval is requested to use the proposed alternatives described above in lieu of the ASME Section XI, Appendix VIII, Supplement 10 requirements. Compliance with the proposed alternatives will provide an acceptable level of quality and safety for examination of the affected welds. IMPLEMENTATION SCHEDULE This technical alternative will be used at Duane Arnold Energy Center; Monticello Nuclear Generating Plant; Point Beach Nuclear Plant, Units 1 And 2; Prairie Island Nuclear Generating Plant, Units 1 And 2; Kewaunee Nuclear Power Plant; and Palisades Nuclear Plant during each plant's present Ten-Year Interval of the Inservice Inspection Program. (See Attachment 1 for Interval dates.) STATUS Approved on February 26, 2004 for during the 4th Interval, NRC Letter, "Duane Arnold Energy Center, Monticello Nuclear Generating Plant, Prairie Island Nuclear Generating Plant, Units 1 and 2, Kewaunee Nuclear Power Plant, Point Beach Nuclear Plant, Units 1 and 2, Palisades Nuclear Plant Re: Request for Alternatives to American Society of Mechanical Engineers (ASME) Section XI, Appendix VIII, Supplement 10 (TAC NOS. MC0814, MC0816, MC0820, MC0821, MC0815, MC0818, MC0819 AND MC0817)" 1.5-89

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning 1.0 SCOPE Supplement 10 is applicable to dissimilar A scope statement provides added metal piping welds examined from either clarity regarding the applicable the inside or outside surface. range of each individual Supplement 10 is not applicable to piping Supplement. The exclusion of welds containing supplemental corrosion CRC provides consistency resistant clad (CRC) applied to mitigate between Supplement 10 and the Intergranular Stress Corrosion Cracking recent revision to Supplement 2 (IGSCC). (Reference BC 00-755). Note, an additional change identifying CRC as "in course of preparation" is being processed separately. 1.0 SPECIMEN REQUIREMENTS 2.0 SPECIMEN REQUIREMENTS Renumbered Qualification test specimens shall Qualification test specimens shall meet the No Change meet the requirements listed herein, requirements listed herein, unless a set of unless a set of specimens is specimens is designed to accommodate designed to accommodate specific specific limitations stated in the scope of the limitations stated in the scope of the examination procedure (e.g., pipe size, weld examination procedure (e.g., pipe joint configuration, access limitations). The size, weld joint configuration, access same specimens may be used to limitations). The same specimens demonstrate both detection and sizing may be used to demonstrate both qualification. detection and sizing qualification. 1.1 General. The specimen set shall 2.1 General. Renumbered conform to the following The specimen set shall conform to the requirements. following requirements. 1.5-90 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (a) The minimum number of flaws in a New, changed minimum number of specimen set shall be ten. flaws to 10 so sample set size for detection is consistent with length and depth sizing. (a) Specimens shall have sufficient (b) Specimens shall have sufficient volume Renumbered volume to minimize spurious to minimize spurious reflections that may reflections that may interfere with the interfere with the interpretation process. interpretation process. (b) The specimen set shall include (c) The specimen set shall include the Renumbered, metricated, the the minimum and maximum pipe minimum and maximum pipe diameters and change in pipe diameter tolerance diameters and thicknesses for which thicknesses for which the examination provides consistency between the examination procedure is procedure is applicable. Pipe diameters Supplement 10 and the recent applicable. Pipe diameters within a within 1/2 in. (13 mm) of the nominal revision to Supplement 2 range of 0.9 to 1.5 times a nominal diameter shall be considered equivalent. (Reference BC 00-755) diameter shall be considered Pipe diameters larger than 24 in. (610 mm) equivalent. Pipe diameters larger shall be considered to be flat. When a range than 24 in. shall be considered to be of thicknesses is to be examined, a flat. When a range of thicknesses is thickness tolerance of +25% is acceptable. to be examined, a thickness tolerance of +25% is acceptable. (c) The specimen set shall include (d) The specimen set shall include examples Renumbered, changed "condition" examples of the following fabrication of the following fabrication conditions: to "conditions" condition: 1.5-91

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (1) geometric conditions that normally (1) geometric and material conditions that Clarification, some of the items require discrimination from flaws normally require discrimination from flaws listed relate to material conditions (e.g., counterbore or weld root (e.g., counterbore or weld root conditions, rather than geometric conditions. conditions, cladding, weld buttering, cladding, weld buttering, remnants of Weld repair areas were added as remnants of previous welds, adjacent previous welds, adjacent welds in close a result of recent field experiences. welds in close proximity); proximity, weld repair areas); (2) typical limited scanning surface (2) typical limited scanning surface Differentiates between ID and OD conditions (e.g., diametrical shrink, conditions shall be included as follows: scanning surface limitations. single-side access due to nozzle and (a) for outside surface examination, weld Requires that ID and OD safe end external tapers). crowns, diametrical shrink, single-side qualifications be conducted access due to nozzle and safe end external independently (Note, new tapers paragraph 2.0 (identical to old (b) for inside surface examination, paragraph 1.0) provides for internal tapers, exposed weld roots, and alternatives when "a set of cladding conditions for inside surface specimens is designed to examinations, accommodate specific limitations (e) Qualification requirements shall be stated in the scope of the satisfied separately for outside surface examination procedure."). and inside surface examinations. (d) All flaws in the specimen set shall Deleted this requirement, because be cracks. new paragraph 2.3 below provides for the use of "alternative flaws" in lieu of cracks. 0 1.5-92

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (1) At least 50% of the cracks shall 2.2 Flaw Location. Renumbered and re-titled. Flaw be in austenitic material. At least 50% At least 80% of the flaws shall be contained location percentages redistributed of the cracks in austenitic material wholly in weld or buttering material. At least because field experience indicates shall be contained wholly in weld or one and no more than 10% of the flaws that flaws contained in weld or buttering material. At least 10% of the shall be in ferritic base material. At least buttering material are probable cracks shall be in ferritic material, one and no more than 10% of the flaws and represent the more stringent The remainder of the cracks may be shall be in austenitic base material, ultrasonic detection scenario. in either austenitic or ferritic material. (2) At least 50% of the cracks in 2.3 Flaw Type. Renumbered and re-titled. austenitic base material shall be (a) At least 60% of the flaws shall be Alternative flaws are required for either IGSCC or thermal fatigue cracks, and the remainder shall be placing axial flaws in the HAZ of cracks. At least 50% of the cracks in alternative flaws. Specimens with IGSCC the weld and other areas where ferritic material shall be mechanically shall be used when available. Alternative implantation of a crack produces or thermally induced fatigue cracks. flaws shall meet the following metallurgical conditions that result requirements: in an unrealistic ultrasonic (1) Alternative flaws, if used, shall response. This is consistent with provide crack-like reflective the recent revision to Supplement characteristics and shall only be used 2 (Reference BC 00-755). when implantation of cracks would produce spurious reflectors that are The 40% limit on alternative flaws uncharacteristic of service-induced is needed to support the flaws. requirement for up to 70% axial (2) Alternative flaws shall have a tip flaws. Metricated width no more than 0.002 in. (.05 mm). 1.5-93

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (3) At least 50% of the cracks shall (b) At least 50% of the flaws shall be Renumbered. Due to inclusion of be coincident with areas described in coincident with areas described in 2.1(d) "alternative flaws", use of "cracks" (c) above, above. is no longer appropriate. 2.4 Flaw Depth. All flaw depths shall be greater than 10% of Moved from old paragraph 1.3(c) the nominal pipe wall thickness. Flaw depths and 1.4 and re-titled. Consistency shall exceed the nominal clad thickness between detection and sizing when placed in cladding. Flaws in the specimen set requirements (e.g., sample set shall be distributed as 20% vs. 1/3 flaw depth increments, follows: e.g., original paragraph 1.3(c)) Flaw Depth Minimum (%Wall Thickness) Number of Flaws 10-30% 20% 31-60% 20% 61-100% 20% At least 75% of the flaws shall be in the range of 10 to 60% of wall thickness. 1.2 Detection Specimens. The Renumbered and re-titled and specimen set shall include detection moved to paragraph 3.1(a). No specimens that meet the following other changes requirements. 0 1.5-94

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (a) Specimens shall be divided into Renumbered to paragraph grading units. Each grading unit shall 3.1 (a)(1). No other changes. include at least 3 in. of weld length. If a grading unit is designed to be unflawed, at least 1 in. of unflawed material shall exist on either side of the grading unit. The segment of weld length used in one grading unit shall not be used in another grading unit. Grading units need not be uniformly spaced around the pipe specimen. (b) Detection sets shall be selected Moved to new paragraph 3.1(a)(2). from Table VIII-S2-1. The number of unflawed grading units shall be at least twice the number of flawed grading units. (c) Flawed grading units shall meet Flaw depth requirements moved to the following criteria for flaw depth, new paragraph 2.4, flaw orientation, and type. orientation requirements moved to new paragraph 2.5, flaw type requirements moved to new _paragraph 2.3, "Flaw Type". 1.5-95

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS_____ ________ Current Requirement Proposed Change Reasoning (1)All flaw depths shall be greater Deleted, for consistency in sample than 10% of the nominal pipe wall sets the depth distribution is the thickness. At least 1/3 of the flaws, same for detection and sizing. rounded to the next higher whole number, shall have depths between 10% and 30% of the nominal pipe wall thickness. However, flaw depths shall exceed the nominal clad thickness when placed in cladding. At least 1/3 of the flaws, rounded to the next whole number, shall have depths greater than 30% of the nominal pipe wall thickness. ________________ (2)At least 30% and no more than 2.5 Flaw Orientation. Note, this distribution is applicable 70% of the flaws, rounded to the next (a) For other than sizing specimens at least for detection and depth sizing. higher whole number, shall be 30% and no more than 70% of the flaws, Paragraph 2.5(b)(1) requires that oriented axially. The remainder of the rounded to the next higher whole number, all length- sizing flaws be oriented flaws shall be oriented shall be oriented axially. The remainder of circumferentially. circumferentially. the flaws shall be oriented circumferentially.________________ 1.3 Length Sizing Specimens. The Renumbered and re-titled and specimen set shall include length moved to new paragraph 3.2 sizing specimens that meet the following requirements. (a)All length sizing flaws shall be Moved, included in new paragraph oriented circumferentially. ____________________3.2(a) 1.5-96

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (b) The minimum number of flaws Moved, included in new paragraph shall be ten. 2.1 above (c) All flaw depths shall be greater Moved, included in new paragraph than 10% of the nominal pipe wall 2.4 above after revision for thickness. At least 1/3 of the flaws, consistency with detection rounded to the next higher whole distribution number, shall have depths between 10% and 30% of the nominal pipe wall thickness. However, flaw depth shall exceed the nominal clad thickness when placed in cladding. At least 1/3 of the flaws, rounded to the next whole number, shall have depths greater than 30% of the nominal pipe wall thickness. 1.4 Depth Sizing Specimens. The Moved, included in new specimen set shall include depth paragraphs 2.1, 2.3, 2.4 sizing specimens that meet the following requirements. (a) The minimum number of flaws Moved, included in new paragraph shall be ten. 2.1 1.5-97

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (b) Flaws in the sample set shall not Moved, potential conflict with old be wholly contained within cladding paragraph 1.2(c)(1); "However, and shall be distributed as follows: flaw depths shall exceed the nominal clad thickness when placed in cladding.". Revised for clarity and included in new paragraph 2.4 Flaw Depth Minimum Moved, included in paragraph 2.4 (% Wall Thickness) Number of Flaws for consistent applicability to 10-30% 20% detection and sizing samples. 31-60% 20% 61-100% 20% The remaining flaws shall be in any of the above categories. (b) Sizing Specimen sets shall meet the Added for clarity following requirements. (1) Length-sizing flaws shall be oriented Moved from old paragraph 1.3(a) circumferentially. (2) Depth sizing flaws shall be oriented Included for clarity. Previously as in 2.5(a). addressed by omission (i.e., length, but not depth had a specific exclusionary statement) 1.5-98 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning 2.0 CONDUCT OF 3.0 CONDUCT OF PERFORMANCE Renumbered PERFORMANCE DEMONSTRATION DEMONSTRATION The specimen inside surface and Personnel and procedure performance Differentiate between qualifications identification shall be concealed from demonstration tests shall be conducted conducted from the outside and the candidate. All examinations shall according to the following requirements. inside surface. be completed prior to grading the (a) For qualifications from the outside results and presenting the results to surface, the specimen inside surface and the candidate. Divulgence of identification shall be concealed from the particular specimen results or candidate. When qualifications are candidate viewing of unmasked performed from the inside surface, the specimens after the performance flaw location and specimen identification demonstration is prohibited. shall be obscured to maintain a "blind test". All examinations shall be completed prior to grading the results and presenting the results to the candidate. Divulgence of particular specimen results or candidate viewing of unmasked specimens after the performance demonstration is prohibited. 2.1 Detection Test. Flawed and 3.1 Detection Qualification. Renumbered, moved text to unflawed grading units shall be paragraph 3.1 (a)(3) randomly mixed (a) The specimen set shall include detection Renumbered, moved from old specimens that meet the following paragraph 1.2. requirements. 1.5-99

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (1) Specimens shall be divided into grading Renumbered, moved from old units. paragraph 1.2(a). Metricated. No (a) Each grading unit shall include at least 3 other changes. in. (76 mm) of weld length. (b)The end of each flaw shall be separated from an unflawed grading unit by at least 1 in. (25 mm) of unflawed material. A flaw may be less than 3 in. (76 mm) in length. (c) The segment of weld length used in one grading unit shall not be used in another grading unit. (d) Grading units need not be uniformly spaced around the pipe specimen. (2) Personnel performance demonstration Moved from old paragraph 1.2(b). detection test sets shall be selected from Table revised to reflect a change Table VIII-SIO-1. The number of unflawed in the minimum sample set to 10 grading units shall be at least 1-1/2 times and the application of equivalent the number of flawed grading units. statistical false call parameters to the reduction in unflawed grading units. Human factors due to large sample size. (3) Flawed and unflawed grading units shall Moved from old paragraph 2.1 be randomly mixed. 1.5-100 0

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (b) Examination equipment and personnel Moved from old paragraph 3.1. are qualified for detection when personnel Modified to reflect the 100% demonstrations satisfy the acceptance detection acceptance criteria of criteria of Table VIII S10-1 for both detection procedures versus personnel and and false calls, equipment contained in new paragraph 4.0 and the use of 1.5X rather than 2X unflawed grading units contained in new paragraph 3.1(a)(2). Note, the modified table maintains the screening criteria of the original Table VIII-S2-1. 2.2 Length Sizing Test 3.2 Length Sizing Test Renumbered (a) The length sizing test may be (a) Each reported circumferential flaw in Provides consistency between conducted separately or in the detection test shall be length-sized. Supplement 10 and the recent conjunction with the detection test. revision to Supplement 2 (Reference BC 00-755). 1.5-101

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (b) When the length sizing test is (b) When the length-sizing test is conducted Change made to ensure security conducted in conjunction with the in conjunction with the detection test, and of samples, consistent with the detection test, and less than ten less than ten circumferential flaws are recent revision to Supplement 2 circumferential flaws are detected, detected, additional specimens shall be (Reference BC 00-755). additional specimens shall be provided to the candidate such that at least provided to the candidate such ten flaws are sized. The regions containing a Note, length and depth sizing use that at least ten flaws are sized. flaw to be sized may be identified to the the term "regions" while detection The regions containing a flaw to candidate. The candidate shall determine uses the term "grading units". The be sized shall be identified to the the length of the flaw in each region. two terms define different concepts candidate. The candidate shall and are not intended to be equal determine the lenqth of the flaw in or interchangeable. each region. (c) For a separate length sizing (c) For a separate length-sizing test, the Change made to ensure security test, the regions of each specimen regions of each specimen containing a flaw of samples, consistent with the containinq a flaw to be sized shall to be sized may be identified to the recent revision to Supplement 2 be identified to the candidate. The candidate. The candidate shall determine (Reference BC 00-755). candidate shall determine the the length of the flaw in each region. length of the flaw in each region. (d) Examination procedures, equipment, and Moved from old paragraph 3.2(a) personnel are qualified for length-sizing includes inclusion of "when" as an when the RMS error of the flaw length editorial change. measurements, as compared to the true Metricated. flaw lengths, do not exceed 0.75 in. (19 mm). 2.3 Depth Sizing Test 3.3 Depth Sizing Test Renumbered 1.5-102

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (a) For the depth sizing test, 80% of (a) The depth-sizing test may be Change made to ensure security the flaws shall be sized at a specific conducted separately or in conjunction of samples, consistent with the location on the surface of the with the detection test. For a separate recent revision to Supplement 2 specimen identified to the candidate. depth-sizing test, the regions of each (Reference BC 00-755). specimen containing a flaw to be sized may be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region. (b) For the remaining flaws, the (b) When the depth-sizing test is Change made to be consistent regions of each specimen containing conducted in conjunction with the with the recent revision to a flaw to be sized shall be identified detection test, and less than ten flaws Supplement 2 (Reference BC 00-to the candidate. The candidate shall are detected, additional specimens shall 755). determine the maximum depth of the be provided to the candidate such that at flaw in each region. least ten flaws are sized. The regions of Changes made to ensure security each specimen containing a flaw to be sized of samples, consistent with the may be identified to the candidate. The recent revision to Supplement 2 candidate shall determine the maximum (Reference BC 00-755). depth of the flaw in each region. (c) Examination procedures, equipment, and Moved from old paragraph 3.2(b). personnel are qualified for depth sizing Metricated. when the RMS error of the flaw depth measurements, as compared to the true flaw depths, do not exceed 0.125 in. (3 mm). 1.5-103

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning 3.0 ACCEPTANCE CRITERIA Delete as a separate category. Moved to new paragraph detection (3.1) and __________________________sizing 3.2 and 3.3 3.1 Detection Acceptance Criteria. Moved to new paragraph 3.1 (b), Examination procedures, equipment, reference changed to Table S1l0 and personnel are qualified for from S2 because of the change in detection when the results of the the minimum number of flaws and performance demonstration satisfy the reduction in unflawed grading the acceptance criteria of Table VIII- units from 2X to 1.5X. S2-1 for both detection and false calls. ________________ 3.2 Sizing Acceptance Criteria Deleted as a separate category. Moved to new paragraph on length 3.2 and depth 3.3 (a) Examination procedures, Moved to new paragraph 3.2(d), equipment, and personnel are included word "when" as an qualified for length sizing the RMS editorial change. error of the flaw length measurements, as compared to the true flaw lengths, is less than or equal to 0.75 inch. _______________ 1.5-1 04

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning (b) Examination procedures, Moved to new paragraph 3.3(c) equipment, and personnel are qualified for depth sizing when the RMS error of the flaw depth measurements, as compared to the true flaw depths, is less than or equal to 0.125 in. 4.0 PROCEDURE QUALIFICATION New 1.5-105

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN SUPPLEMENT 10 - QUALIFICATION REQUIREMENTS FOR DISSIMILAR METAL PIPING WELDS Current Requirement Proposed Change Reasoning Procedure qualifications shall include the New. Based on experience gained following additional requirements. in conducting qualifications, the (a) The specimen set shall include the equivalent of 3 personnel sets (i.e., equivalent of at least three personnel a minimum of 30 flaws) is required performance demonstration test sets. to provide enough flaws to Successful personnel performance adequately test the capabilities of demonstrations may be combined to the procedure. Combining satisfy these requirements. successful demonstrations allows (b) Detectability of all flaws in the a variety of examiners to be used procedure qualification test set that are to qualify the procedure. within the scope of the procedure shall Detectability of each flaw within be demonstrated. Length and depth the scope of the procedure is sizing shall meet the requirements of required to ensure an acceptable paragraph 3.1, 3.2, and 3.3. personnel pass rate. The last (c) At least one successful personnel sentence is equivalent to the demonstration shall be performed. previous requirements and is (d) To qualify new values of essential satisfactory for expanding the variables, at least one personnel essential variables of a previously qualification set is required. The qualified procedure acceptance criteria of 4.0(b) shall be met. 1.5-106

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 10 (Rev. 0) (Fleet Relief Request) Duane Arnold Energy Center Monticello Nuclear Generating Plant Alternative to Use Code Case N-613-1

1. ASME Code Component(s) Affected Code Class: Class 1

Reference:

ASME, Section XI Examination Category: B-D Item Number: B3.90

== Description:== Reactor Vessel Full Penetration Nozzle-to-Vessel Welds Component Numbers: See Tables 1 and 2

2. Applicable Code Edition and Addenda

ASME Section XI 1989 Edition, no Addenda is applicable to the Duane Arnold Energy Center (DAEC) Inservice Inspection (ISI) Program for the Third Ten-Year Interval. ASME Section XI 1995 Edition, 1996 Addenda is applicable to the Monticello Nuclear Generating Plant (MNGP) ISI Program for the Fourth Ten-Year Interval.

3. Applicable Code Requirement

Nuclear Management Company (NMC), LLC is currently required to perform inservice examinations of selected reactor vessel nozzle-to-vessel welds in accordance with the requirements of 10 CFR 50.55a, and the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components. Table IWB-2500-1, Examination Category B-D, Item No. B3.90 specifies the examination requirements. Figure IWB-2500-7(b) requires that a minimum volume of material, a distance of ts/2 (one half the reactor vessel shell thickness) adjacent to the weld, be examined.

4. Reason for Request

The required examination volume for the reactor vessel pressure retaining nozzle-to-vessel welds extends far beyond the weld into the base metal, and is unnecessarily large. This proposed alternative would re-define the examination volume boundary to 1/2 inch of base metal on each side of the widest portion of the weld, removing from examination the base metal that was extensively examined during prior inspections, and is not in the high residual stress region associated with the weld. 1.5-107

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

5. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(a)(3)(i), authorization is requested to use the proposed alternative described in ASME Boiler and Pressure Vessel Code Section XI Code Case N-613-1 in lieu of the ASME Section XI Table IWB-2500-1 Examination Category B3.90 requirements. Compliance with the proposed alternative will provide an acceptable level of quality and safety for examination of the affected welds.

In lieu of the ts/2 volume requirement of ASME Section XI, Figure IWB-2500-7(b), NMC proposes to reduce the examination volume next to the widest part of the weld to one-half (1/2) inch from the weld. This refined examination volume is defined in detail within Code Case N-613-1. NMC will use Code Case N-613-1 for the Reactor Pressure Vessel (RPV) nozzles as shown in Figure 2 of the Code Case. The required examination volume for the RPV nozzle-to-vessel welds extends far beyond the weld into the base metal, and is unnecessarily large. This proposed alternative would re-define the examination volume boundary to 1/2 inch of base metal on each side of the widest portion of the weld. This reduction in base metal inspection will not affect the flaw detection capabilities in the weld and heat affected zone. The proposed reduction in exam volume is base metal only. The creation of flaws during plant service in the volume excluded from the proposed reduced examination is unlikely because of the low stress in the base metal away from the weld. The stresses caused by welding are concentrated at or near the weld. Cracks, should they initiate, occur in the high-stressed areas of the weld. These high stress areas are contained in the volume that is defined by Code Case N-613-1 and are thus subject to examination. During previous examinations, no indications exceeding the allowable limits of the preservice or inservice criteria were found in the reactor vessel nozzle to shell examination volumes including the base metal areas proposed for exclusion from examination in this request. The prior thorough examination of the base metal and the examination of the high-stressed areas of the weld provide an acceptable level of quality and safety.

6. Duration of Proposed Alternative This technical alternative will be used at DAEC during the current Third Ten-Year Interval of the Inservice Inspection Program scheduled to end on November 1, 2005, and at MNGP during the current Fourth Ten-Year Interval of the Inservice Inspection Program scheduled to end on May 31, 2012.

The use of Code Case N-613-1 is requested until the NRC publishes the Code Case in a future revision of Regulatory Guide 1.147. 1.5-108

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T"INTERVAL EXAMINATION PLAN STATUS Approved on October 6, 2004 for during the 4th Interval, NRC Letter, "Duane Arnold Energy Center and Monticello Nuclear Generating Plant Re: Request for Authorization to Utilize Code Case N-613-1 (TAC Nos. MC2374 and MC2375)" Table 1 DAEC Nozzle-to-Vessel Welds Within Scope of Request (NOT INCLUDED IN MNGP ISI PLAN) Table 2 MNGP Nozzle-to-Vessel Welds Within Scope of Request Summary Weld Nozzle Full Volume Exam Nondestructive Number Identification Configuration Previously Completed to Examination and Description Extent Achievable (NDE) Method 102652 N-1A NV, Code Case N- Examined in 1994 from UT-0 RPV N-1A 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Suction recordable indications. 102654 N-1B NV, Code Case N- Examined in 2001 from UT-0 RPV N-1 B 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Suction recordable indications. 102656 N-2A NV, Code Case N- Examined in 2001 from UT-0 RPV N-2A 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102658 N-2B NV, Code Case N- Examined in 2001 from UT-0 RPV N-2B 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102660 N-2C NV, Code Case N- Examined in 2000 from UT-0 RPV N-2C 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 1.5-109

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 2 MNGP Nozzle-to-Vessel Welds Within Scope of Request Summary Weld Nozzle Full Volume Exam Nondestructive Number Identification Configuration Previously Completed to Examination and Description Extent Achievable (NDE) Method 102662 N-2D NV, Code Case N- Examined in 1994 from UT-0 RPV N-2D 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102664 N-2E NV, Code Case N- Examined in 1994 from UT-0 RPV N-2E 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102666 N-2F NV, Code Case N- Examined in 2000 from UT-0 RPV N-2F 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102668 N-2G NV, Code Case N- Examined in 1998 from UT-0 RPV N-2G 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102670 N-2H NV, Code Case N- Examined in 1998 from UT-0 RPV N-2H 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102672 N-2J NV, Code Case N- Examined in 1994 from UT-0 RPV N-2J 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102674 N-2K NV, Code Case N- Examined in 2001 from UT-0 RPV N-2K 613-1, RPV shell side only due to UT-45 Nozzle, Recirc Figure 2 nozzle configuration; no UT-60 Riser Inlet recordable indications. 102676 N-3A NV, Code Case N- Examined in 1994 from UT-0 RPV N-3A, Main 613-1, RPV shell side only due to UT-45 Steam Outlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102678 N-3B NV, Code Case N- Examined in 2000 from UT-0 RPV N-3B, Main 613-1, RPV shell side only due to UT-45 Steam Outlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102680 N-3C NV, Code Case N- Examined in 1998 from UT-0 RPV N-3C, Main 613-1, RPV shell side only due to UT-45 Steam Outlet Figure 2 nozzle configuration; One UT-60 indication was recorded with the 600 scan which required evaluation per 1.5-110

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Table 2 MNGP Nozzle-to-Vessel Welds Within Scope of Request Summary Weld Nozzle Full Volume Exam Nondestructive Number Identification Configuration Previously Completed to Examination and Description Extent Achievable (NDE) Method ASME Code Section Xl, 1986 Edition with no Addenda and was found to be acceptable. Indication is contained within the reduced volume of Code Case N-613-1, Figure 2. 102682 N-3D NV, Code Case N- Examined in 2000 from UT-0 RPV N-3D, Main 613-1, RPV shell side only due to UT-45 Steam Outlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102684 N-4A NV, Code Case N- Examined in 1996 from UT-0 RPV N-4A, 613-1, RPV shell side only due to UT-45 Feedwater Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102686 N-4B NV, Code Case N- Examined in 1998 from UT-0 RPV N-4B, 613-1, RPV shell side only due to UT-45 Feedwater Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102688 N-4C NV, Code Case N- Examined in 1994 from UT-0 RPV N-4C, 613-1, RPV shell side only due to UT-45 Feedwater Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102690 N-4D NV, Code Case N- Examined in 2000 from UT-0 RPV N-4D, 613-1, RPV shell side only due to UT-45 Feedwater Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102692 N-5A NV, Code Case N- Examined in 2000 from UT-0 RPV N-5A, Core 613-1, RPV shell side only due to UT-45 Spray Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102694 N-5B NV, Code Case N- Examined in 1994 from UT-0 RPV N-5B, Core 613-1, RPV shell side only due to UT-45 Spray Inlet Figure 2 nozzle configuration; no UT-60 recordable indications. 102375 N-6A NV, Code Case N- Examined in 1996 from UT-0 RPV N-6A, 613-1, RPV shell side only due to UT-45 Spare (formerly Figure 2 nozzle configuration; no UT-60 Reactor Vessel recordable indications. Head Spray) I I I 1.5-111

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4' INTERVAL EXAMINATION PLAN Table 2 MNGP Nozzle-to-Vessel Welds Within Scope of Request Summary Weld Nozzle Full Volume Exam Nondestructive Number Identification Configuration Previously Completed to Examination and Description Extent Achievable (NDE) Method 102377 N-6B NV, Code Case N- Examined in 2000 from UT-0 RPV N-6B, 613-1, RPV shell side only due to UT-45 Spare Figure 2 nozzle configuration; no UT-60 recordable indications. 102379 N-7 NV, Code Case N- Examined in 1998 from UT-0 RPV N-7, 613-1, RPV shell side only due to UT-45 Reactor Vessel Figure 2 nozzle configuration; no UT-60 Head Vent recordable indications. 102696 N-8A NV, Code Case N- Examined in 1994 from UT-0 RPV N-8A, Jet 613-1, RPV shell side only due to UT-45 Pump Figure 2 nozzle configuration; no UT-60 Instrumentation recordable indications. 102698 N-8B NV, Code Case N- Examined in 2001 from UT-0 RPV N-8B, Jet 613-1, RPV shell side only due to UT-45 Pump Figure 2 nozzle configuration; no UT-60 Instrumentation recordable indications. 102700 N-9 NV, Code Case N- Examined in 1996 from UT-0 RPV N-9, Spare 613-1, RPV shell side only due to UT-45 (formerly CRD Figure 2 nozzle configuration; no UT-60 Return) recordable indications. 102623 N-10 NV, Code Case N- Examined in 2000 from UT-0 RPV N-1 0, 613-1, RPV shell side only due to UT-45 Standby Liquid Figure 2 nozzle configuration; no UT-60 Control Injection recordable indications. 1.5-112

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 11 (Rev. 0) (Fleet Relief Request) Request For Authorization To Utilize Code Case N-661 10 CFR 50.55a Request GR-04-01 Proposed Alternative in Accordance with 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected ASME Section Xl, Class 2 and 3 Carbon Steel Piping for Raw Water Service

2. Applicable ASME Section Xl Code Edition and Addenda The applicable Codes for Repair/Replacement activities are as follows:

Monticello (2001 Edition) Prairie Island (1998 Edition with the 2000 Addenda) Point Beach (1998 Edition with the 2000 Addenda) Kewaunee (1998 Edition with the 2000 Addenda) Palisades (1989 Edition) Duane Arnold (1992 Edition with the 1992 Addenda)

3. Applicable Code Requirement

ASME Section Xl 1989 Edition IWA-4120(a) requires that repairs be performed in accordance with the Owner's Design Specification and the original Construction Code of the component or system. IWA-4310 requires that defects be removed or reduced in size in accordance with IWA-4000. ASME Section XI 1992 Edition with the 1992 Addenda IWA-4170(b) requires that repairs and installation of replacement items shall be performed in accordance with the Owner's Design Specification and the original Construction Code of the component or system. IWA-4310 requires that defects be removed or reduced in size in accordance with IWA-4000. 1.5-113

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ASME Section X1 1998 Edition with the 2000 Addenda IWA-4221(a) requires that items used for repair/replacement activities shall meet the applicable Owner's Requirements. IWA-4221 (b) requires that an item to be used for repair/replacement activities shall meet the Construction Code specified in accordance with (1), (2) or (3) below. (1) When replacing an existing item, the new item shall meet the Construction Code to which the original item was constructed. (2) When adding a new item to an existing system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for construction of any originally installed item in that system. (3) When adding a new system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for other systems that perform a similar function. IWA-4422.1 (a) states that a defect is considered removed when it has been reduced to an acceptable size. ASME Section XI 2001 Edition IWA-4221 (a) requires that items used for repair/replacement activities shall meet the applicable Owner's Requirements. IWA-4221 (b) requires that an item to be used for repair/replacement activities shall meet the Construction Code specified in accordance with (1), (2) or (3) below. (1) When replacing an existing item, the new item shall meet the Construction Code to which the original item was constructed. (2) When adding a new item to an existing system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for construction of any originally installed item in that system. (3) When adding a new system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for other systems that perform a similar function. IWA-4422.1(a) states that a defect is considered removed when it has been reduced to an acceptable size. 1.5-114

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN

4. Reason for Request

Relief is requested from replacement or weld repair of wall thinning conditions in Class 2 and 3 carbon steel raw water piping systems to the design specification and the original construction code. Such thinning may be the result of various degradation mechanisms such as erosion, corrosion, cavitation and pitting. The use of Code Case N-661 will provide adequate time so that pipe replacement can be planned to reduce impact on system availability including Maintenance Rule applicability and availability of replacement materials.

5. Proposed Alternative and Basis for Use NMC proposes to implement the requirements of ASME Code Case N-661 as an alternative under 10 CFR 50.55a(a)(3)(i) for Class 2 and 3 raw water piping systems resulting from degradation mechanisms such as erosion, corrosion, cavitation, or pitting as an alternative to the requirements of the ASME Section Xl code as referenced above. These types of defects are typically identified by small leaks in the piping system or by pre-emptive non-code required examinations performed to monitor the degradation mechanisms. The alternative repair technique described in Code Case N-661 involves the application of additional weld metal on the exterior of the piping system that restores the wall thickness requirement. This repair technique is utilized whenever engineering evaluation determines that such a repair is suitable for the particular defect or degradation being resolved. Provisions for use of this Code Case will be addressed in the Repair/Replacement Program for each site.

Provisions for implementation of this Code Case will be addressed on a plant specific basis in each site's Repair/Replacement Program. The provisions will require that adjacent areas be examined to verify that the repair will encompass the entire flawed area and that no other unacceptable degraded locations exist within a representative area. This will be dependent on the degradation mechanism present. An evaluation of the degradation will be performed to determine the re-examination schedule to be conducted over the life of the repair. The repair will be considered to have a maximum service life of two fuel cycles unless the re-examinations conducted during each of the two fuel cycles establish the expected life of the repair. Additionally, the following restrictions will be placed on the use of Code Case N-661 to ensure that the use of the Code Case will provide an acceptable alternative pursuant to 10 CFR 50.55a(a)(3)(i): 1.5-115

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN (a) if the root cause of the degradation has not been determined, the repair is only acceptable for one cycle, (b) weld overlay repair of an area can only be performed once in the same location, and (c) when through-wall repairs are made by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage. The basis for use of the repair technique described in Code Case N-661 is that the ASME Code subcommittee for Section XI determined that this repair technique provides an acceptable alternative to the requirements of IWA-4000 and provides an acceptable level of quality and safety. Therefore, the proposed alternative is justified per 10 CFR 50.55a(a)(3)(i) Code Case N-661 was approved by the ASME Section Xl Code Committee on July 23, 2002; however, it has not been incorporated into NRC Regulatory Guide 1.147 "Inservice Inspection Code Case Acceptability, ASME Section XI Division 1." Therefore NMC requests use of the alternative repair technique described via this relief request. A copy of the ASME Section Xl Code Case N-661 is provided as Enclosure 2 for reference.

6. Duration of Proposed Alternative NMC requests authorization of Code Case N-661 to be used for each plant's 10-year ISI interval (see table below) or until the NRC publishes Code Case N-661 in a future revision of Regulatory Guide 1.147. Upon incorporation into the Regulatory Guide, NMC will review and follow the conditions specified. All other ASME Code, Section XI requirements for which relief was not specifically requested and authorized by the NRC staff will remain applicable including third party review by the Authorized Nuclear Inservice Inspector.

1.5-116

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI Plant Interval Interval Dates Monticello Nuclear Generating Plant 50-263 Fourth 05/01/03 - 05/31/12 Prairie Island Nuclear Generating Plant, Units 1 Fourth 12/21/04-and 2, 50-282 & 50-306 12/20/14 Point Beach Nuclear Plant, Units 1 and 2, 50- Fourth 07/01/02 - 266 and 50-301 06/30/12 Kewaunee Nuclear Power Plant 50-305 Fourth 06/16/04 - 06/16/14 Palisades Nuclear Plant 50-255 Third 05/12/95 - 12/12/06 Duane Arnold Energy Center 50-331 Third 11/01/96 - 11/01/05

7. Precedents NMC has determined that the following previous Authorizations to use Code Case N-661 are directly applicable to this Relief Request:

(1) Letter from NRC to Southern Nuclear Company, "Edwin I. Hatch Nuclear Plant, Units 1 and 2, Joseph M. Farley Nuclear Power Plant, Units 1 and 2, and Vogtle Electric Generating Plant, Units 1 and 2 (TAC Nos. MB8959, MB8960, MB8961, MB8962, MB8963, and MB8964)," dated November 21, 2003, ADAMS Accession No. ML033280037. (2) Letter from NRC to TXU Energy, "Comanche Peak Steam Electric Station, Units 1 and 2 - RE: Relief from the Requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Concerning Relief Requests C-2 and C-7 (TAC Nos. MB7947 and MB7948)," dated February 18, 2004, ADAMS Accession No. ML040490624. STATUS Submitted on July 28, 2004, not yet approved for use, NMC Letter to NRC, "10 CFR 50.55a Request GR-04-01; Request For Authorization To Utilize Code Case N-661(L-HU-04-027)" 1.5-117

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN PAGE LEFT INTENTIONALLY BLANK 1.5-118

10 CFR 50.55a REQUEST NO. 12 PROPOSED ALTERNATIVE FOR VISUAL EXAMINATION ILLUMINATION LEVELS IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i)

1. ASME Code Component(s) Affected ASME Component and System Code Class Control Rod Drive (CRD) System Hydraulic Control Unit 2 (HCU) 26-27 Residual Heat Removal-Service Water (RHRSW) System 3 Pump 12 RHRSW System Pump 14 3 Emergency Diesel Generator-Emergency Service Water 3 (EDG-ESW) System Pump 11 EDG-ESW System Pump 12 3

2. Applicable ASME Section XI Code Edition and Addenda The applicable Code for Repair/Replacement and related activities is the 2001 Edition, No Addenda as authorized in the Monticello Nuclear Generating Plant (MNGP) Fourth Ten-Year Interval Inservice Inspection (ISI) Program 10CFR50.55a Request No. 7, October 3, 2003 (TAC No. MB6897).

3. ADDlicable Code Reauirement The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, IWA-2210, "Visual Examinations," provides the following requirements (meeting either of which would be sufficient to demonstrate adequate illumination) for conducting visual examinations.

IWA-2210(e) specifies the following requirements for illumination levels:

            "It is not necessary to measure illumination levels on each examination surface when the same portable light source or similar installed lighting equipment is demonstrated to provide the illumination specified in Table IWA-2500-1 at the maximum examination distance."

1.5-119

10 CFR 50.55a REQUEST NO. 12 PROPOSED ALTERNATIVE FOR VISUAL EXAMINATION ILLUMINATION LEVELS IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) IWA-221 0(f) states:

             "The adequacy of the illumination levels from battery powered portable lights shall be checked before and after each examination or series of examinations, not to exceed 4 hr between checks. In lieu of using a light meter, these checks may be made by verifying that the illumination is adequate (i.e., no discernable degradation in the visual examination resolution of the procedure demonstration test chart characters)."

4. Reason for Request:

NMC identified five VT-2 examinations, performed in conjunction with post-repair/replacement pressure tests (during the first period of the current 10-year ISI interval), which were not performed in accordance with the illumination requirements of IWA-2210 and could not be re-performed.* The illumination levels were not verified as required. Authorization is requested to utilize an alternative method that was used to verify the illumination levels, as described in Section 5. The five VT-2 examinations were performed using portable battery-powered lights (a standard practice to ensure adequate lighting). However, due to a procedural inadequacy, the illumination levels were not verified before and after each VT-2 examination as specified in IWA-2210(f), nor were the illumination levels demonstrated pursuant to IWA-221 0(e). (This procedural inadequacy is tracked and is being corrected under the MNGP Corrective Action Program.) Consequently, without verifying the illumination levels of the portable battery-powered lights or ambient light conditions, only the knowledge and skill of the certified VT-2 examiner was available to attest that the lighting conditions were adequate for the five VT-2 examinations. Because it could not initially be verified that the lighting levels were acceptable, it was conservatively assumed that the requirements of IWA-2210 were not met for these five VT-2 examinations. Listed below are the five pressure tests, which could not be re-performed,* where the VT-2 examination illumination levels could not initially be determined as acceptable: The replaced components were restored to operable status prior to the identification of this illumination concern with the VT-2 examinations. The pre-service pressure testing required by IWA-4540 (and associated VT-2 examinations) cannot be re-performed because the associated systems were restored to service. 1.5-120

10 CFR 50.55a REQUEST NO. 12 PROPOSED ALTERNATIVE FOR VISUAL EXAMINATION ILLUMINATION LEVELS IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) R/R Plan Examination Number Work Description Date 2003-22-0011 Replace accumulator on 10/26/03 HCU 26-27 2003-22-0012 Replace EDG-ESW Pump 12 with 12/4/03 rebuilt spare from stock. 2003-22-0014 Replace RHRSW Pump 14 with 11/7/03 rebuilt spare from stock. 2003-22-0021 Replace RHRSW Pump 12 with 11/21/03 rebuilt spare from stock. 2004-22-0062 Replace EDG-ESW Pump 11 with 11/15/04 rebuilt spare from stock.

5. Proposed Alternative and Basis for Use The proposed illumination alternative for the five VT-2 examinations is based upon the following three factors. First, the possession of the knowledge and skill by the certified VT-2 examiners, and their attestation that there was sufficient lighting for them to properly perform the examinations. Second, the use of battery-powered portable lights by the VT-2 examiners, which provided additional illumination (beyond ambient) during the examinations. Third, a walk-down of the areas where the VT-2 examinations were conducted was performed by the site Nondestructive Examination (NDE) Level III, to verify the ambient illumination level provided by the permanent plant lighting was adequate to meet the requirements of IWA-2210, as described below.

Upon discovery of the five post-repair/replacement pressure tests, for which the VT-2 examination illumination levels could not be initially be determined as acceptable, the site NDE Level III walked-down each location and performed illumination checks where the VT-2 examinations had been conducted. Illumination checks were performed using a light meter and/or a character card. The illumination checks were done to establish whether the existing permanently installed plant (ambient) lighting provided reasonable assurance of sufficient illumination at each location to comply with IWA-221 0(e) illumination requirements during the performance of the VT-2 examinations. 1.5-121

10 CFR 50.55a REQUEST NO. 12 PROPOSED ALTERNATIVE FOR VISUAL EXAMINATION ILLUMINATION LEVELS IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) The results of these illumination checks provided reasonable assurance that the ambient illumination levels at the components, using the existing permanently installed plant lighting, were sufficient to meet IWA-2210 requirements. Also, during performance of these illumination checks, the NDE Level III examiner did not identify any indications of leakage from the pressure-tested components. Based on the following factors, NMC considers the proposed alternative to be acceptable:

    "       Walk-downs following the VT-2 examinations indicate that the ambient illumination provided by the permanent plant lighting was sufficient,

• The skill and knowledge of the certified VT-2 examiners attesting to a light level sufficient for them to perform their examinations, and

    "       The VT-2 examiners used battery-powered portable lights (which provided additional illumination above ambient) during the VT-2 examinations.

Therefore, NMC considers the proposed method, consisting of the three factors listed above, to be an acceptable alternative to the lighting verification required by IWA-2210 during the performance of the five VT-2 examinations. NMC believes that the proposed alternative provided an acceptable level of quality and safety. NMC requests a one-time approval for the use of the alternative method previously described, that was applied for verifying the illumination levels for the five VT-2 examinations associated with the post-repair/replacement pressure tests identified in this 10 CFR 50.55a request.

6. Duration of Proposed Alternative NMC requests a one-time approval of the alternative method that was applied for determining the illumination levels for the five VT-2 examinations (for the subject components specified in Section 1) for the first period of the current fourth ten-year ISI interval for the MNGP.

1.5-122

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY ASME Code Component(s) Affected Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section Xl, Class 1, Reactor Vessel Nozzle-to-Vessel Shell welds specified below and in-detail in Table A: Recirculation Suction Nozzle N-IA Weld - N-1A NV Recirculation Inlet Nozzle N-2D Weld - N-2D NV Recirculation Inlet Nozzle N-2E Weld - N-2E NV Recirculation Inlet Nozzle N-2J Weld - N-2J NV Main Steam Discharge Nozzle N-3A Weld - N-3A NV Feedwater Inlet Nozzle N-4C Weld - N-4C NV Core Spray Inlet Nozzle N-5B Weld - N-5B NV Jet Pump Instrumentation Nozzle N-8A Weld - N-8A NV

2. Applicable ASME Section Xl Code Edition and Addenda The applicable ASME Section Xl Code for the Monticello Nuclear Generating Plant (MNGP), Fourth Ten-Year Inservice Inspection (ISI) Interval is the 1995 Edition with the 1996 Addenda.

3. AoDlicable Code Reauirement ASME Class I Nozzle-to-Vessel Shell welds are subject to the examination requirements of Subsection IWB Table IWB-2500-1, as shown below, and are required to be examined once within the Fourth Ten-Year Interval:

Code Class: 1

References:

VB-2500, Table IWB-2500-1 IV Examination Category: B-D Item Number: B3.90

Description:

Nozzle-to-Vessel Shell Welds Component Numbers: S ee Section 1 and Table A System: Reactor Vessel Examination Method: Volumetric - Ultrasonic Testing (UT) In lieu of the examination volume depicted in Figure IWB-2500-7(b), the United States Nuclear Regulatory Commission (NRC) has authorized the NMC to use the alternative examination volume requirements of Code Case N-613-1 (Reference 1) for the Nozzle-to-Vessel Shell welds listed in this request. 1.5-123

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY The MNGP Fourth Ten-Year Interval Inservice Inspection Plan also implements Code Case N-460 (Reference 2), which is endorsed by the NRC in Regulatory Guide 1.147 (Reference 3). Code Case N-460 states in part, "when the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent." NRC Information Notice (IN) 98-42 (Reference 4) termed a reduction in coverage of less than 10 percent to be "essentially 100 percent." IN 98-42 states in part,

  'The NRC has adopted and further refined the definition of "essentially 100 percent" to mean "greater than 90 percent"...has been applied to all examinations of welds or other areas required by ASME Section XI.'

4. Impracticality of Compliance Construction Permit CPPR-31 was obtained for the MNGP in 1967. The MNGP systems and components were designed and fabricated before the examination requirements of ASME Section Xl were formalized and published. Because this plant was not specifically designed to meet the requirements of ASME Section XI, full compliance is not feasible or practical within the limits of the current plant design.

10 CFR 50.55a recognizes the limitations to in-service inspection of components in accordance with Section Xl of the ASME Code, that are imposed due to early plants' design and construction, as follows: 10 CFR 50.55a(g)(1): For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g) (4) and (5) of this section to the extent practical. 10 CFR 50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and pre-service examination requirements, set forth in Section Xl of editions of the ASME Boiler and Pressure Vessel Code ... to the extent practical within the limitations of design, geometry and materials of construction of the components. 1.5-124

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY 10 CFR 50.55a(g)(5)(iii): If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in § 50.4, information to support the determinations. The inspection limitations on the subject components are primarily due to inherent nozzle design geometric contours with some additional, minor interference from nearby welded attachments (see Table A). A description of the examination methodology used to provide the maximum obtainable coverage is provided in Section 6 of this request. This methodology is based on ASME Section Xl, Appendix VIII qualification and was applied to the extent practical within the design constraints of the components. Enclosure 3 provides cross-sectional diagrams of the subject welds showing the geometric contour of the component design in relation to the welds and the coverage obtained within the examination volume requirements of Code Case N-613-1, Figure 2.

5. Burden Caused by Compliance Compliance with the examination coverage requirements of ASME Section Xl would require modification, redesign, or replacement of components where geometry is inherent to the component design.

6. Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry and materials of construction.

NMC performed qualified examinations that achieved the maximum, practical amount of coverage obtainable within the limitations imposed by the design of the components. Additionally, as Class 1 examination Category B-P components, a VT-2 examination is performed on the subject components of the Reactor Coolant Pressure Boundary during system pressure tests each refueling outage. This was completed during the 2005 refueling outage and no evidence of leakage was identified for these components. 1.5-125

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the requirements of ASME Section XI Table IWB-2500-1, Category B-D, Item B3.90 and associated Code Cases, and proposes to utilize these completed exams as an acceptable alternative that provides reasonable assurance of continued structural integrity. Basis For Use The MNGP Nondestructive Examination (NDE) procedures incorporate improved inspection techniques qualified under Appendix VIII of the ASME Section XI Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-shell welds. Coverage was obtained by following the scan parameters defined by the MNGP specific Electric Power Research Institute (EPRI) computer modeling report (Reference 5) for each nozzle configuration and angle, and as designated within MNGP NDE procedures. The examinations were performed using a manual contact method from the nozzle outside blend and vessel shell surfaces as discussed in the EPRI modeling report and as stated in MNGP procedures. The shear wave mode of propagation was used for each of the transducer and wedge combinations required for the inner 15 percent of the req~iired parallel scan volume. The refracted longitudinal wave mode of propagation was used for the remaining outer 85 percent of the volume for parallel scans, and all of the perpendicular scans. The subject components received the required examination(s) to the extent practical within the limited access of the component design. For the examinations conducted, satisfactory results were achieved, and no evidence of unacceptable flaws were detected with the improved inspection techniques. Due to the design of these welds it was not feasible to effectively perform a volumetric examination of 100 percent of the volume as described in IWB-2500-7(b). The nozzle-to-vessel welds are accessible from the vessel shell side of the weld, but examinations cannot be performed from the nozzle side of the weld because of the forging curvature. In addition, due to component configuration, certain nozzle-to-vessel weld examinations are further limited by the reactor pressure vessel design obstructions (such as appurtenances). Additional coverage for the limited areas was not achievable or practical, based on the latest qualified ultrasonic technology, nor by other considered examinations methods, such as radiography. MNGP has concluded that if 1.5-126

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY significant degradation existed in the subject welds, it should have been identified by the examinations performed. Additionally, as Class 1 examination category B-P components, VT-2 examinations were performed on the subject components in association with the Reactor Coolant Pressure Boundary system pressure test performed during the 2005 refueling outage, and no evidence of leakage was identified. The materials for the subject components are A533 Cl I nozzle forgings welded to A508 Cl II vessel shell plate. A review of operating experience within the nuclear industry did not reveal any instances of cracking in this location and type of weldment. The MNGP reactor vessel water chemistry is controlled in accordance with the 2004 revision to the BWR Water Chemistry Guidelines (Reference 6). Also a hydrogen water chemistry system is used to reduce the oxidizing environment in the reactor coolant. These additional measures provide added assurance against the initiation of cracking or corrosion from the inside surface of the reactor vessel for the subject components listed in this request. An inerted primary containment environment during operation provides assurance of corrosion protection on the outside surface of the reactor vessel. Based on the above, with due consideration of the earlier plant design, the underlying objectives of the Code required volumetric examinations have been met. The examinations were completed to the extent practical and evidenced no unacceptable flaws present. VT-2 examinations performed on the subject components during system pressure testing each refueling outage (in accordance with examination Category B-P) provide continued assurance that the structural integrity of the subject components is maintained. Additionally, the MNGP Water Chemistry Program and inerted primary containment environment provide added measures of protection for the component materials.

7. Duration of Proposed Alternative NMC requests the granting of this relief for the Fourth Ten-Year Inservice Inspection Interval of the Inservice Inspection Program for the MNGP that is scheduled to end on May 31, 2012.

8. Precedents The NRC has granted relief for the MNGP for previous ten-year inservice inspection intervals, most recently the Third Ten-Year Inservice Inspection Interval (Reference 7). Also, the NRC has granted relief for the Quad Cities Nuclear Power Station, Units 1 and 2 (Reference 8), the Dresden Nuclear Power 1.5-127

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Station, Units 2 and 3 (Reference 9), and the Prairie Island Nuclear Generating Plant, Unit 2 (Reference 10). REFERENCES

1. ASME Section XI Code Case N-613-1, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.10 and B3.90, Reactor Nozzle-To-Vessel Welds, Figures IWB-2500-7(a), (b),

and (c)." 2 ASME Section X1 Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds."

3. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," Revision 13, January 2004.

4. NRC Information Notice 98-42, "Implementation of 10 CFR 50.55a(g) In-service Inspection Requirements."

5. EPRI Internal Report IR-2004-63, "Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations."

6. BWRVIP-1 30, "BWR Water Chemistry Guidelines - 2004 Revision" (EPRI Topical Report TR-1 008192).

7. NRC letter to NMC, "Monticello Nuclear Generating Plant Third 10-Year Interval Inservice Inspection Relief Request No. 16, Parts A, B and C (TAC No.

MB5487)," dated May 19, 2003.

8. Letter from NRC to Commonwealth Edison Company, "Quad Cities, Units 1 and 2 - Relief Request CR-32 for Third 10-Year Inservice Inspection Interval," dated September 6, 2000.

9. Letter from NRC to Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request CR-24 For Third 10-Year Inservice Inspection Interval," dated January 8, 2003.

10. NRC letter to NMC, "Prairie Island Nuclear Generating Plant, Unit 2 - Evaluation of Relief Request No. 16 for the Unit 2 3 rd 10-year Interval Inservice Inspection Program (TAC No. MC1775)," dated October 18, 2004.

1.5-128

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY TABLE A - Category B-D, "Full Penetration Welds of Nozzles in Vessels," Item No. B3.90 Percent Coverage and Limitations for Nozzles N-1A, N-2D, N-2E, N-2J, N-3A, N-4C, N-5B, and N-8A Code System Code Component Category and Component and Percent* Exam and Component ID Examination Volume Coverage Report Item No. Description Required Obtained Limitations Number Limited due to nozzle Reactor Vessel, Nozzle-to-Vessel Weld, configuration. Also, small B-D Recirculation N-AzNV Code Case Nl63-1 83% reduction due to interference 2005UT041 B3.90 Suction Figure 2 from welded thermocouple Nozzle N-1A Figure 2 attachments. Limited due to nozzle B3.90 Reactor Vessel, Nozzle-to-Vessel Weld, configuration. Also, small Recirculation inlet N-2D NV Code Case N-613-1 82% reduction due to interference 2005UT028 Nozzle N-2D Figure 2 from welded thermocouple attachment. Limited due to nozzle .. Reactor Vessel, Nozzle-to-Vessel Weld, coniguraton. zzle 8-0 B3.90 Recirculation Inlet N-2E NV Code Case N-613-1 78% configuration. 2005UT016 Nozzle N-2E Figure 2 Nozzle-to-Vessel Weld, Limited due to nozzle B-D Reactor Vessel, B39 Recirculation Inlet N-2J NV Code Case N-613-1 78% configuration. 2005UT005 83.90 Nozzle N-2J Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Main Steam N-3A NV Code Case N-613-1 83% configuration. 2005UT023 83.90 Discharge Figure 2 Nozzle N-3A Figure 2 Nozzle-to-Vessel Weld, Limited due to nozzle B-D Reactor Vessel, B39 Feedwater Inlet N-4C NV Code Case N-613-1 79% configuration. 2005UT025 83.90 Nozzle N-4C Figure 2 1.5-129

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Code System Code Component Category and Component and Percent* Exam and Component ID Examination Volume Coverage Report Item No. Description Required Obtained Limitations Number Limited due to nozzle B-D Reactor Vessel, Nozzle-to-Vessel Weld, configuration. Also, small B3.90 Creduction from welded to interference duethermocouple 2005UT018 81% Core Spray Nozzle Inlet N-5B N-5B NV Figure N-613-1 Code Case 2 attachments. Limited due to nozzle B-D Reactor Vessel, Nozzle-to-Vessel Weld, configuration. B3.90 Jet Pump N-BA NV Code Case N-613-1 83% 2005UT037 Instrumentation Figure 2 Nozzle N-8A Due to the nozzle design it was not feasible to effectively examine essentially 100 percent of the required examination volume as defined in Figure 2 of Code Case N-613-1. Percentages are conservatively rounded down to the nearest whole number. It should be noted that 100 percent of the inner 15 percent was examined in the parallel scans for all components listed above. 1.5-130

10 CFR 50.55a REQUEST NO. 13 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY EXAM LIMITATIONS IMPOSED BY COMPONENT DESIGN AND CONSTRUCTION This enclosure contains a series of excerpts from the ISI Ultrasonic Testing (UT) reports applicable to the subject components. These excerpts contain sketches depicting the component configuration with physical limitations imposed by the design, e.g., geometrical contour, weld position, interferences, and a cross sectional view depicting the UT coverage and limitations in relation to the required examination volume. Also included is a sketch of a typical nozzle contour and the resulting affect that causes the UT transducer to lose coupling contact when it reaches the nozzle blend radius. COMPONENT REPORT PAGE(S) N-1A NV 2005UT041 Pages 1-3 N-2D NV 2005UT028 Pages 4-5 N-2E NV 2005UT016 Pages 6-7 N-2J NV 2005UT005 Pages 8-9 N-3A NV 2005UT023 Page 10 N-4C NV 2005UT025 Pages 11-12 N-5B NV 2005UT018 Pages 13-14 N-8A NV 2005UT037 Page 15 Typical Nozzle Contour Affecting Transducer Contact Page 16 1.5-131

Coverage drawings excerpted from applicable reports Component- N-IA NV Report # 2005UT041 Supplemental Report Report NO.: 2005UT04.1 Summary No.: 102652 Comments: Parallel scan limitation due to radius area, no contact Axial scan lrLtation duo to transducer sizo and radius area, no contact. R5.2S in

                             ,MonticelloN.l Coverage Plot Parallel scan direction 1.5-132

Component- N-IA NV Report # 2005UT041 Limitation Record Summary No.: 102652 pleport No.: 2005UT041 Description of LimitatVon: 4" Limitation due to thermo-couple. Monticello N I Coverage Plot R525 in Axial scan direction 1.5-133

Component- N-IA NV Report # 2005UT041 Supplemental Report Repor No.: 2005UT041 Summary No.: 102652 Comments: 4" Limitation due to thermo-couple at la 1/2" counter clockwise. Z?:It T& VESSEL W-ELP 1.5-134

Component - N-2D NV Report # 2005UT028 Supplemental Report INW-> Report No.: 200"SUT028 Summary No.: 102662 Comments: Monticello N2D Coverage plots. Axial scan limitation due to transducer size and radius area, no contact. Parallel scan limitation due to radius area, no contact Monticello N2D Coverage Plot Mornicello N2D Coverage Plot Axial scan direction ParaLlel scan diirectdon in 1.5-135

Component - N-2D NV Report # 2005UT028 Supplemental Report Report No.: 2005UT028 Summar, No.: 102662 Comments: 3" Uimitation due to thermo-couple at 79" clockwise. N2D No7ie, to Vessel'Wel-d Limitation SketchA x( MI; 1.5-136

Component- N-2E NV Report # 2005UT016 Supplemental Report Repot No.: 2005UT016 Summary Not.: 10i26o4 Comments: Limitation due to radius area, no contact

• R3.50 in Monticello N2 Coverage Plot Parallel scan direction in 1.5-137

Component- N-2E NV Report# 2005UT016 Supplemental Report Regojit t),j 200SUT016 Summary No.: 102664 Comments: Limitation due to transducer size and radius area, no contacL Monticello N2 Coverage Plot Axial scan direction 1.5-138

Component - N-2J NV Report # 2005UT005 Supplemental Report Report No.: 2005UT005 Summary No.: 102672 Comments: Limitation due to radius area, no contact R3.50 in Monticello N2 Coverage Plot Parallel scan direction 1.5-139

Component - N-2J NV Report # 2005UT005 Supplemental Report Report No.: 2005UT005 Summary No.: 102672 Comments: Limitation due to transducer size and radius area, no contact.

                                      \      R3.miJu
                                             --      Monticello N2 Coverage Plot Axial scan direction
                                                   "* A                6 & "
                                                                                            --- F 6d 60 deg.

_____.__, 1 1.5-140

Component - N-3A NV Report # 2005UT023 Supplemental Report IW Report No.: 2005UT023 Summary No,: 102676 Comments: Monticello N3 Coverage Plot Axial scan lImitation duo to transducer size and radius area, no contact. Parallel scan limitation due to radius area, no contact R3,50 in Monticello N3 Coveraae Plot Monticello N3 Coverage Plof Ay al scan direction Panil el scan direction

                      -   .  - i                        .

1.5-141

Component - N-4C NV Report # 2005UT025 Supplemental Report

~ii~)                                                                               Report,No., 200SUT025 Summary NoI: 102688 Comments;     Limitation due to radius area, no contact.
                '*\          R3.00 in.
               <7,/              Monticello N4 Coverage Plot Parallel scan direction D

5.25 in of coverage Inn=e 15-G F E' 1.5-142

Component - N-4C NV Report # 2005UT025 Supplemental Report tii~ Report No. 2005UT025 Summary Nc,: 10z688 Comments: ULmiftion due to transducer size and radius area, no contact Monticello N4 Coverage Plot Axial scan direction No 60 deg. F F 1.5-143

Component- N-5B NV Report # 2005UT018 Supplemental Report Report No.i 2005UT018 SummaryN*.: 102694 Cornmenra: Monticello N55 Coverage Plot AXclalscan l1initation due to transducer size and radilus area, no contact.

             'Parallel scan Ilmitafion due to radius area, no contact.

R3.50 in Moititicelo N5B Coverage Plot Monticello NSB Coverage Plot Axial scan direction Pairalel scam direction 5.25 in

                                                                                                                       -of=-m   0.78!in 1.5-144

Component- N-5B NV Report# 2005UT018 Supplemental Report Report No.: 2005UT018 Summary No, 102694 Comments: 7" Uimitation due to thermo-cauple top dead center of nozzle. I NS3 Nozzle to Vessel Weld 1.5-145

Component - N-8A NV Report # 2005UT037 Supplemental Report 4, Report No.: 2005UT037 Summary No.: 102696 Comments: Monticello N8 coverage Plots. Axial scan limitation due to transducer size and radius acea, no contact.

              ~Parallel scan limitation due to radius area, no contact.
                         \      in R3.00
                        ",'---Monticllo N8 Coverage Plot                                       Monticello NSCoverage Plot Parallel scan direction Axial scan direction 1.5-146

Typical for Nozzle Limitations Cdoveragi .ffected

                     -        býy4 I.iftoff dueto* radius:

R3aO if Axiallscan shown Amq fz~zu M N2 Nozzle sisown. asleamplea 1.5-147

REQUEST FOR AUTHORIZATION TO UTILIZE CODE CASE N-513-2

1. ASME Code Component(s) Affected ASME Section X1, Moderate Energy Class 2 and Class 3 Piping

2. Applicable ASME Section XI Code Edition and Addenda The applicable code editions are as follows:

-NMC Site Inservice Inspection Repair/Replacement Monticello 1995 Edition with the 1996 2001 Edition Addenda Prairie Island 1998 Edition with the 2000 1998 Edition with the 2000 Addenda Addenda Point Beach 1998 Edition with the 2000 1998 Edition with the 2000 Addenda: Addenda Palisades 1989 Edition 1989 Edition Duane Arnold 1989 Edition 1992 Edition with the 1992

                                         .... __ JAddenda Flaws that exceed the acceptance criteria of the above code editions/addenda are required to be accepted by either a repair/replacement activity or an analytical evaluation.

3. Applicable Code Requirements The applicable code requirements are as follows:

ASME Section Xl 1989 Edition CLASS 3 IWD-3000 states, "This article is in course of preparation. The rules of IWB-3000 may be used." IWB-3132 provides four ways in which an inservice volumetric or surface examination may be accepted..

1. IWB-3132.1, 'Acceptance by Volumetric or Surface Examination"

2. IWB-3132.2, "Acceptance by Repair" 1.5-148

3. IWB-3132.3, "Acceptance by Replacement"

4. IWB-3132.4, "Acceptance by Analytical Evaluation" IWB-3132.2 states, "Components whose volumetric or surface examination reveals flaws that exceed the acceptance standards listed in Table IWB-341 0-1 shall be unacceptable for continued service until the additional examination requirements of IWB-2430 are satisfied, and the flaw shall be either removed by mechanical methods or the component repaired to the extent necessary to meet the acceptance standards of IWB.3000."

IWB-3132.3 states, "As an alternative to the repair requirement of IWB-3132.2, the component or the portion of the component containing the flaw shall be replaced." IWB-3142 provides five ways in which an inservice visual examination may be accepted.

1. IWB-3142.1, "Acceptance by Visual Examination"

2. IWB-3142.2, "Acceptance by Supplemental Examination"

3. IWB-3142.3, "Acceptance by Corrective Measures or Repairs"

4. IWB-3142.4, "Acceptance by Analytical Evaluation"

5. IWB-3142.5, "Acceptance by Replacement" IWB-31.42.3 states, "Components containing relevant conditions shall be acceptable for continued service if the relevant conditions are corrected or the components are repaired to the extent necessary to meet the acceptance standards specified in Table IWB-3410-1."

IWB-31i42.5 states, "As an alternative to either the supplemental examinations of IWB-3142.2, the corrective measures or repairs of IWB-3142.3, or the evaluation of IWB-3142.4, the component or that part of the component containing the relevant condition shall be replaced." CLASS 2 IWC-3122 provides four ways in which an inservice volumetric and surface examination may be accepted.

1. IWC-3122.1, "Acceptance by Examination"

2. IWC-3122.2, "Acceptance by Repair"

3. IWC-3122.3, "Acceptance by Replacement"

4. IWC-3122.4, "Acceptance by Evaluation" IWC-3122.2 states, "Components whose examination reveals flaws that exceed the acceptance standards listed in Table IWC-3410-1 shall be unacceptable for continued service until the additional examination 1.5-149

requirements of IWC-2430 are satisfied, and the flaw shall be either removed by mechanical methods or the component repaired to the extent necessary to meet the acceptance standards of IWC-3000." IWC-3122.3 states, "As an alternative to the repair requirements of 0WC-3122.2, a component or the portion of the component containing the flaw shall be replaced." IWC-3132 provides four ways in which an inservice Visual examination may be accepted.

1. IWC-3132.1, "Acceptance by Supplemental Examination'"

2. IWC-31 32.2, "Acceptance by Corrective Measures or Repairs"

3. IWC-3132.3, "Acceptance by Evaluation"

4. IWC-3132.4, "Acceptance by Replacement" IWC-3132.2 states, "Components containing relevant conditions shall be acceptable for continued service if the relevant conditions are corrected or the components are repaired to the extent necessary to meet the acceptance standards specified in Table IWC-341 0-1 ."

IWC-3132.4 states, "As an alternative to the supplemental examinations of IWC-3132.1, the corrective measures or repairs of IWC-3132.2, or the evaluation of IWC-3132.3, a component or part of a component containing the relevant condition shall be replaced." ASME Section Xl1 995 Edition with the 1996 Addenda CLASS 3 IWD-3000 states, "This Article is in course of preparation. The rules of IWB-3000 may be used." IWB-3132 provides three ways in which an inservice volumetric or surface examination may be accepted.

1. IWB-3132.1, "Acceptance by Volumetric or.Surface Examination",

2. IWB-3132.2, "AcceptanCe by Repair/Replacement Activity", or

3. IWB-31832.3, "Acceptance by Analytical Evaluation".

IWB-3132.2 states, "A component whose volumetric or surface examination detects flaws that exceed the acceptance standards of Table IWB-341 0-1 is unacceptable for continued service until the additional examination requirements of IWB-2430 are satisfied and the component is corrected by a repaidreplacement activity to the extent necessary to meet the acceptance standards of IWB-3000." 1.5-150

IWB-3142 provides four ways in which an inservice visual examination may be accepted.

1. IWB-3142.1, "Acceptance by Visual Examination"

2. IWB-3142.2, "Acceptance by Supplemental Examination"

3. IWB-3142.3, "Acceptance by Corrective Measures or Repair/Replacement Activity"

4. IWB-3142.4, "Acceptance bY Analytical Evaluation" IWB-31 42.3 states, "A com ponent containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair/replacement activity or by corrective measure to the extent necessary to meet the acceptance standards of Table IWB-341 0-1."

CLASS 2 IWC-3122 provides three ways in which an Inservice Volumetric and Surface Examinations may be accepted.

1. IWC-3122.1, "Acceptance by Examination"

2. IWC-3122.2, "Acceptance by Repair/Replacement Activity"

3. IWC-3122.3, "Acceptance by Analytical Evaluation" IWC-3122.2 states, "A component whose examination detects flaws that exceed the acceptance standards of Table IWC-3410-1 is unacceptable for continued service until the additional examination requirements of IWC-2430 are satisfied and the component is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWC-3000."

IWC-3132 provides four ways in which an inservice visual examinations may be accepted.

1. IWC-3132, "Acceptance"

2. IWC-3132.1, "Acceptance by Supplemental Examination"

3. IWC-3132.2, "Acceptance by Corrective Measures or Repair/Replacement Activity",,

4. IWC-3132.3, "Acceptance by Analytical Evaluation" IWC-3132.2 states, "A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair/replacement activity or bycorrective measures to the extent necessary to meet the acceptance standards of Table IWC-341 0-1."

1.5-151

ASME Section XI 1998 Edition with the 2000 Addenda CLASS 3 IWD-3000 states, "This Article is in course of preparation. The rules of IWB-3000 may be used." IWB-3132 provides three ways in which an Inservice Volumetric or Surface Examination may be accepted.

1. iWB-3132.1, "Acceptance by Volumetric or Surface Examination",

2. IWB-3132.2, "'Acceptance by Repair/Replacement Activity', or

3. IWB-3132.3, 'Acceptance by Analytical Evaluation".

IWB-3132.2 states, "A component whose volumetric or surface examination detects flaws that exceed the acceptance standards of Table IWB-341 0-1 is unacceptable for continued service until the additional examination. requirements of IWB-2430 are satisfied and the component is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWB-3000." IWB-3142 provides four ways in which an Inservice Visual examination may be accepted.

1. IWB-3142.1 "Acceptance by Visual Examination"

2. IWB-3142.2 "Acceptance by Supplemental Examination"

3. IWB-3142.3 "Acceptance by Corrective Measures or Repair/Replacement Activity"

4. IWB-3142.4 "Acceptance by Analytical Evaluation" IWB-3!42.3 states, "A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repair/rePlacement activity or by corrective measure to the extent necessary to meet the acceptance standards of Table IWB-3410-1 ."

CLASS 2 IWC-3122 provides three ways in which an Inservice Volumetric and Surface Examinations may be accepted..

1. iWC-3122.1, "Acceptance by Examination"

2. IWC-3122.2, "Acceptance by Repair/Replacement Activity"

.3. IWC-3122.3, "Acceptance by Analytical Evaluation" IWC-3122.2 states, "A component whose examination detects flaws that exceed the acceptance standards of Table IWC-3410-1 is unacceptable for continued service until the additional examination requirements of IWC-2430 1.5-152

are satisfied and the component is corrected by a repair/replacement activity to the extent necessary to meet the acceptance standards of IWC-3000.m IWC-3132 provides four ways in which an inservice visual examination may be accepted.

1. IWC-3132, "Acceptance"

2. IWC-3132.1, "Acceptance by Supplemental Examination"

3. IWC-3132.2, "Acceptance by Corrective Measures or Repair/Replacement Activity"

4. IWC-3132.3, !'Acceptance by Analytical Evaluation IWC-3132.2 states, "A component containing relevant conditions is acceptable for continued service if the relevant conditions are corrected by a repairreplacement activity or by corrective measures to the extent necessary to meet the acceptance standards of Table IWC-3410-1.

4. Reason for Request

Relief is requested from replacement or internal weld repair of wall thinning conditions resulting from various wall thinning degradation mechanisms such as erosion, corrosion, cavitation, and pitting in moderate energy Class 2 and 3 piping systems in accordance with the design specification and the original construction code. The use of Code Case N-513-2 will provide an acceptable method to evaluate flaws on a temporary basis until the next scheduled outage.

5. Proposed Alternative and Basis for Use The Nuclear Regulatory Commission in Regulatory Guide 1.147, 'lnservice Inspection Code Case Acceptability," Revision 14, has accepted Code Case N-513-1 with the following limitations:

1- Specific safety factors in paragraph 4.0 must be satisfied. 2- Code Case N-513 may not be applied to:

i. Components other than pipe and tube.

ii. Leakage through a gasket iii. Threaded connections employing nonstructural seal welds for leakage prevention (through seal weld leakage is not a structural flaw; thread integrity must be maintained). iv. Degraded socket welds 1.5-153

Code Case N-51 3-1 permits flaws in'Class 2 and 3 moderate energy piping on a temporary basis until the next outage if it can be demonstrated that adequate pipe integrity and leakage containment are maintained. The ýCode Case is currently applicable to part-through and through wall planar flaws and part-through nonplanar flaws. Service experience has shown that some piping can suffer degradation from nonplanar flaws, such as pitting and microbiological attack, where local inconsequential leakage can occur. The Code Case can be used for nonplanar through-wall flaws but in a restrictive situation where nonpianar geometry is dominant in one plane. Some plants have Used the intent of N-513 for nonplanar leaking flaws; however, relief requests from code requirements are still required because of the stated limited scope of N-513 in section 3.0 of the Code Case. The Code Case was revised (N-513-2) to extend the application to cover all types of nonplanar flaws. The analysis procedures were expanded to address the general case of through-wall degradation. Code Case N-513-2 has broader applications and therefore has a real direct benefit for operating plants. Code Case N-513-2 includes the incorporation of the improved flaw evaluation procedures for piping that are provided in the new Appendix C of Section XI in the 2002 Addenda. Code Case N-513-2 addresses the limitations posed in Regulatory Guide 1.147 as follows:

1. Paragraph 4.0 was revised to incorporate references to Appendix C for acceptance and eliminated the provision that lower safety factors may be used.

2. 1.0(a) was revised to limit the application of the, code case as specified in the limitation applied in Regulatory Guide 1.147.

NMC considers the proposed alternative of using Code Case N-513-2 to provide an acceptable level of quality and safety in accordance with 10 CFR 50.55a(3)(i).

6. Duration of Proposed Alternative NMC requests approval of Code Case N-513-2 to be used for each plant's 10-year ISI interval (see table 1 below) or until the NRC publishes Code Case N-513-2 in a future revision of Regulatory Guide 1.147. Upon incorporation into the Regulatory Guide, NMC will review and follow the conditions specified. All other ASME Code, Section Xl requirements for which relief was not specifically requested and authorized by the NRC staff will remain applicable including third party review by the Authorized Nuclear Inservice Inspector.

1.5-1 54

Plant Applicable. ISI Interval Dates ASME Section Interval Monticello Nuclear 1995 Edition Fourth 05/01/03 - 05/31/12 .Generating Plant 50-263 with the, 1996 Addenda Prairie Island Nuclear. 1998 Edition Fourth 12121/04 - 12120/14 Generating Plant 50-282- with the 2000 (Unit 1) & 50-306 (Unit 2) Addenda Point Beach Nuclear, Plant 1998 Edition Fourth 07/01/02- P6/30/112: Units I & 2 (50-266 & with the 2000 50-30.1) Addenda ____"_. Palisades Nuclear Plant 1989 Edition Third 05/12/95 - 12/12106 50-255 . Duane Arnold Energy 1989 Edition Third 11/01/96 - 10/31/06.: Center 50-331

7. Precedent Tennessee Valley Authority (TVA) submitted a relief request pursuant to 10 CFR 50.55a(a)(3)(i), for Browns Ferry Nuclear Plant, Units 1, 2 and 3; Sequoyah Nuclear Plant, Units 1 and 2; and Watts Bar Nuclear Plant, Unit 1, dated November 23, 2003 (ADAMS Accession #ML033320222). TVA requested relief from using the specific formula in Code Case N-513, for the maximum allowable flaw width when planar flaw evaluation rules may be applied. As an alternative, TVA proposed the use of the formula for maximum allowable flaw width from Code Case N-513-1, with applicable errata while retaining the use of all the other provisions and requirements in Code Case N-513. The NRC approved this relief request by letter October 6, 2004 (ADAMS Accession #ML042150438). The TVA relief request is similar to the NMC relief request in that the request involves Code Case N-513. However, NMC is requesting relief to use Code Case N-513-2, which incorporates the limitations specified in Regulatory Guide 1.147 on Code Case N-513-.1. In addition, Code Case N-513-2 added a procedure for evaluating non-planar through-wall flaws in moderate energy piping. This revision also includes the improved flaw evaluation procedures for piping added to Section XI, Appendix C, in the 2002 Addenda.

1.5-155

ASME CODE CASE N-513-2, "EVALUATION CRITERIA FOR TEMPORARY ACCEPTANCE OF FLAWS IN MODERATE ENERGY CLASS 2 OR 3 PIPING" 1.5-156

CASE CASES OF ASME BOILER AND PRBSJRE VESSEL CODE N-513-2 Approval Date: February 20, 2004 See Nurnerh: Inde,; for expire on and any reafin'nationdates. Case N-513-2 2.0 PROCEDURE Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping (a) The flaw geomet'y shall be characterized by volu-Section XI, Division 1 metric inspection methods or by physical measurement. The full pipe circumrnference at the flaw location shall be Itquiry.: What requirerments may be used for tempo- inspected to characterize the length and depth of all flaws rary acceptance of flaws, including through-wall flaws. in in the pipe sectiod. mioderate energy Class 2 or 3 piping, without performing a (b) Flaw shall be classified as planar or nonplanar. repair/replacement activity? (c) When multiple flaws, including irregular (com-pound) shape flaws, are detected, the interaction and com-Reply. It is the opinion of the Commitee that the bined area loss of flaws in a given pipe section shall be following requirements may be used to accept flaws, accounted for in the flaw evaluation. (d) A flaw evaluation shall be performed to dtermine including through-wall flaws, in moderate energy Class 2 or 3 piping, without prforming a repairireplacement the conditions for flaw acceptance. Section 3.0 provides activity for a limited time, not exceeding the time to the accepted methods for conducting the required analysis. next schedulcd outage. (e) Frequent periodic inspections of no more than 30 day intervals shall be used to determine if flaws are grow-ing and to establish the time at which the detected flaw 1.0 SCOPE will reach the allowable size. Alterinatively, a flaw growth evaluation may be performed to predict the time at which (a) these lequirements apply to the ASME Section the detected flaw will grow to the allowable size. The II, ANSI B31.1, and ANSI B31.7 piping, classified by flaw growth analysis shall consider the relevant growth the Owner as Class 2 or 3. The provisions of this Case mechanisms such as general corrosion or wastage, fa-do not apply to the following: tigue, or stress corrosion cracking. When a flaw growth (1) pumps, valves, expansion joints and heat ex- analysis is used to establish the allowable time for tempo-changers; rary operation, periodic examinations of no more than (2) socket welds; 90 day intervals shall be conducted to verify the flaw (3) leakage through a flange joint: growth analysis predictions. (4) threaded connections employing nhnstructural (f) For through-wall leaking flaws, leakage shall be seal welds foir leakage protecii6n. observed by daily walkdowos to confirm the analysis (b) The provisions Of the Case apply to Class 2 or 3 conditions used in tie evaluation remain valid. piping whose maximum operating temperature does not (g) If examinations reveal flaw gowth rate to beunac-exceed 200'F (93*C) aid whose maximum operating ceptable, a repair or replacement shall be performed. pres*sre does not exceed 275 psig (1.9 MPa). (it) Repair or replacement shall be performed no later (c) The following flaw evaluation criteria are permit- than when the predicted flaw size from either periodic ted for pipe and tube. The flaw evaluation criteria are inspection orby flaw growth analysis exceeds the accept-permitted for adjoining fittings and flanges to a distance ance criteria of 4,0, or the next scheduled outage, which-of (Rt)V" from the weld centerline. ever occurs first. Repair or replacement shall be in accor-(d) Tie provisions of this Case demonstiate the integ- dance with IWA-4000 or IWA-7000, respectively, in rity of the item a not the consequences of leakge. It Editions and Addenda prior to the 1991 Addenda; and, is the responsibility Of the Owner to demonstrate system in the 1991 Addenda and later, in accordance with operability considering effects of leakage. IWA-4000. Tb,.carmuatadoo' ruet

                            .Wons,~ 1esIE I ws PutbtlesI ofsaraty. '61a I ngonlyto . ,Iaezu,a ioleejtr. ipontba it, consuvtfon dI bori.,,         ino
                                                                                                                                          -1.6or tobnapcrt mIs and rae hrcamnsonent,and Insiovtricnspolrlon farpmznre tnoilgittyofno siromonenent, 3rd trampot La-neand to IInterpretthe1o ruinsWhatoquoottcs adsoo ras~rcoattorfnmtor-.Tt~ Codedoesntotaddrs ahresalfty issues   rnalelao thecoostrueto oftioilamo   ptesstuzoavesseltronsport tantoandnuclear cmpoonixrt.

andThe lnanrica Inspe~talo~f nudeanjce,,Pancras and or.1oopo OootUa. Th. Fit., Oad. h.u~td!.J, ro n - 4r- ,,l. v -d~d-. I-r,..., airier iclEvatilesunocs. In 1 (N-513-2) SUPP. 1 - NC I Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission ofThe American Society of Mechanical Engineers. All rights reserved.. I 1.5-157

CASE (continued) N,51 3-2 CASES OFiSH OIE AND PRESSUREVESSELCODE (bi)'For planar flaws in austenitic piping, the evaluation procedure in Appendix C shall be used. Flaw depths up to .[0%:of wall thickness may be evaluated. When through-wall circumnferenitial. flaws ameevaluated, the for-mulas for evaluation given in C-5320 of Appendix Cmay be used, with the flaw penetration (a/) equal to unity. When through-wall axial flaws are evaluated, the allow-able flaw length is: 2 (1) (2) of= (S' + S,)Y2 (3) la) Circumferential Flaw where

                                                                        'p = pressure for the loading condition D,=pipe outside diameter Of= flow stress S,= Code specified yield strength S,=Code specified uiltimate tensile strength and SF,,, = structural factor on primary memnbrane stress as specified in C-2622 Material properties at the temperature of interest shall bie used.

(c) For planar flaws in ferritic piping, the evaluation procedure of Appendix C shall be used. Flaw depths up to 100% of wall thickness may be evaluated. Whben through-wall circumrferential flaws are evaluated in accor-, Ib) Axl Flaw dance with C-5300 or C-6300, the flaw penetration(a) shall be set to unity. When througb-wall axial flaws are FIG. 1 THROUGH-WALL FLAW GEOMETRY evaluated in accordance with C-5400, the allowable. length is defined by Eqs. (1) through (3), with the appro-. palate structural factors from Appendix C, C-2622. When through-wall flaws are evaluated in accordance with C-7300 or C-7400. the formulas for evaluation given in C-4300 may be used, but with values for F., Fe,, and F (i). Evaluationsi and examination shall be documented applicable to through-wall flaws. Relations for F,., F", in accordance with IWA-6300. The Owner shall docu-ment the use of this Case on the applicable data report and F that take into -accountflaw shape and pipe geometry form., (Rht ratio) shall be used, The appendix to this Case pro-vides equations for Fe,_Fh, and F for a selected range of RI:. Geometry of a through-wall 'crack is shown in Fig. 1. (d) For nonpianar flaws, the pipe is acceptable when 3.0 FLA; W EVALUATIoN the remaining pipe thickness (r~,) is gretater than or equal (a) For planar flaws, the flaw shall be bounded by a to the minimumn wall thickness 1, rectanguilar or circumferential planar area in accordance PJD, with the methods described in Appendix C. rWA-3300 2 (4)

                                                                                                =   (S + .4p) shall be used to determine when multiple proximate flaws areý to be evaluated as a single flbw. The geometry of a        where through-wall planar flaw is shown in Fig. 1.                         P = maximumn operattingpressuire at flaw location SUPP. 1 -- NC                                           2 (A-513-2)

Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of Mechanical Engineers. A rights reserved. I. I.. 1.5-158 0

CASE (continued) CASES OF ASME BOILER AND PRESSURE VESSEL CODE N-51 3-2 t transverse L!() (circumnfarential) A FIG. 2 ILLUSTRATION OF NONPLANAR FLAW DUE TO WALL THINNING S allowable stress at operaing temperature and and t,, is greater tIan 1.13r,,,, tr,* is detemnined by the longitudinal stress limits for tMe Construction satisfying both of the following equations: Code ae satisfied for a uniform wall thickness equal to ., Alternatively, an evaluation may be performed as given Q. L below. The evaluation procedure is afunction oftbledeptb and the extent of the affected area as illustrated in Fig. 2 Ltts 0.3534w (6) 4,t, (1) When the width of wall thinning W. that ex-ceeds :,, fs less than or equal to 0.5 (R,et) t , where R. When the above requirements arc not satisfied, (4) shall is the outside radius and W. is defined in Fig. 2, the be met. flaw, can be classified as a planar flaw and evalu~ated in (4j When the requirements of(1), (2), and (3) above accordance with 3.0(a) through 3.10(), above. When the are not sa&tised, rý!ý is determlined fromnCurve 2 of Fig. above.requirement is not satisfied. (2)saall he mrt. 3.Tn addition, iboe shall satisfy the following equation: (2) When the transverse extent of wall thinning that ex .ceeds 1~,_ 4 .ý,j is not greater than rm',,is determined from Curve I of Fig. 3, where L4,) is defined 4,io " ,.-/ \ I in Fig. 2. When the above requirement is not satisfied, (7) (3) shall be met. (3) When the maxinnsi extent of wall thinning that where o-,, is the nominal pipe longitudinal bending stress exceeds t,, L,~, is less than or equal to 2.65(Rr, resulting from all primary pipe loadings. 3 (N-515-2) SUrP. I-- NC II Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by.i:permission of The ArmriCe, . . SOciety of Mechanical Enginreers. All rights reserved. 1.5-159

CASE (continued) N-513-2 CAMEOF ASME SO!LE AMDPRMMTYt 1'MSEL CODE 1.0

                 .0.8 6

2...4 0.6 0.2 0 0 2' 3 4 5 6 7 8 FIG. 3 ALLOWABLE WALL THICKNESS AND LENGTH OF LOCALLY THINNED AREA (e) When there is through-wall penetration along a of the flaw, or the toml of the flow areas of multiple flaws portion of the thinned wall, as illustrated in Fig. 4, the that are combined into a single flaw for the purpose of flaw may be evaluated by the branch reinforcement evaluation, shall not exceed the lesser of the flow area method. T1e thinned area includiOg the through-wall pen- of the pine or 20 in.2 (130 cre). etration shall be represented by a circular onng at the (t9 Alternatively, when there is through-wall penetra-flaw location. Only the portion of the flaw lying within tion along a portion of the thinned wall as illustrated in Q1 need be considered as illustrated in Fig. 5. When Fig. 4 the flaw may be evaluated as two independent evaluating multiple flaws in accordance with IWA-3330, planar through-wall flaw-one oriented in the axial direc-only the porions of the flaws contained within ta need Lion and the other oriented in the circumferential direc-be considered. tion. The minimum wall thickness (.&, shall be deter-The rinimum wall thickness, ts, shall be deternined mined by Eq. (4). The through-wall lenghts for each by Eq. (4). For evaluation purposes, the adjusted wall flaw are the lenghts L,.-,, and L.*,, where the local wall thickness, tap is the postulated thickness as shown in thickness is equal to t,,t as projected along the axial and Fig. 5. The pipe wall thickness is defined as the thickness circumferontial planes as shown in Fig. 4. The two planar of the pipe in the non-degraded region as shown in Fig. flaws so constructed shall be evaluated to 3.0(a) and 5(a). The diameter of the opening is equal to dadJ as 3.0(b) or 3,0(c), as appropriate. If a flaw growth analysis defined by t=7 as shown in Fig 5(a). T'he postulated value is performed, the growth in flaw dimensions shall con-for !adj shall be greated than iti and shall not exceed the sider both corrosion and crack-growth mechanisms as pipe wall thickness. The t,,* value may be varied between relevant to the application. The flow area of the flaw, or t,,,I,aid the pipe wall thickness to determine whether the total of the flow areas of multiple flaws that are there is a combination of tmj and d j that satisfies the combined into a single flaw for the purpose of evaluation. branch reinforcement requirements. shall not exceed the lesser of the flow area of the pipe The required are= reinforcement for the postulated cir- or 20 in.2 (130 cm2). cular opening, daj and r,,U,as illustrated in Fig. 5(b), (g) In performing a flaw growth analysis, the proce. shall be calculated in accordance with NC-3643,3 or dures in C-3000 may be used as guidance. Relevant ND-3643.3, as appropriate. If a flaw growth analysis is growth rate mechanisms shall be considered. When stress performed, the growth in flaw dimensions shall consider corrosion cracking (SCC) is active, the following growth the degradation mechanism(s) as relevant to the applica- rate equation shall be used: tion. The flaw is acceptable when them is sufficient thick-ness in the degraded area to provide the required area daldt =SrC M8 reinforcement. Comptianea with the primary stress limits where dcz/dc is flaw growth rate in incheslhour, Kt,* is of the Construction Code shall he verified. The flow area the marximnum stress intensity factor under loug-term SUPP. I NC 4 (N-513-2) Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permhission of The American Society of Mechanical Engineers. All rights reserved. 1.5-160

CASE (continued) N,5I3-2 CASES OF ABME BOILER AND PRESSURE VESL CODE Transverse (circumferentiall direction FIG. 4 ILLUSTRATION OF THROUGH-WALL NONPLANAR FLAW DUE TO WALL THINNING 5 (N-513-2) SUP?. I -Nc I Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of Mechanical Engineers. All rights reserved. 1.5-161

CASE (continued) N-513-2 CASE01 ASM BOILER AND PRESSMR ON VESSL CODE Through-wall PenetrationT Adj Pipe (al Adjuatad Wall Thcknless traai b) Equivalent H109 Rleprewntatlon FIG. 5 ILLUSTRATION OF ADJUSTED WALL THICKNESS AND EQUIVALENT HOLE DIAMETER SUPP. 1 - NC 6 (N-513-2) Reprinted from ASME 2004 Edition Code Cases, Nuclear components, by permission of The American Society of Mechanical Engineers. All rights reserved. I 1.5-162

CASE (continued) CASES OF ASME BOILER AHD PRESSURE VESSEL CODE N,513-2 steady state conditions in ksi in. 0.5 ST is a temperature or, if fewer than five, all susceptible and accessible loca-correction factor, and C and n are material constants. tions shall be examined within 30 days of detecting the ilaw. For intergranular SCC in austenitic steels, where T7 (b) When a flaw is detected, an additional sample of 200-F (930C). the:same size as defined in 5(a) shall be examined. C 1.79X I0- (c) This process shall be repeated within 15 days for each successive sample, until no significant flaw is de-n = 2.161 tected or until, 100% of susceptible and accessible loca-For transgranular SCC in austenitic steels, where T < tions have been examined. 200 0F (930C). C 1.79 x 10 S= 3.71 X 10 -a [ 10 (00t'427- 12.Zj 6.0 NOMENCLATURE

      = 2.161 C = cocfficinnt in the crack growth relatonship The temperature Tis the metal temperature in degrees                    D= outside pipe diameter Fahrenheit. The flaw growth rate curves for the above                        F = nondimensional stress intensity factor for SCc growth mechanisms are shown in Figs. 6 and 7.

through-wall axial flaw under hoop stress Other growth rate parameters in Eq. 8 may be used, b= noedimensional stress intensity factor for provided they are supported by appropriate data. through-wall circumferential flaw under pipe (h) For nonferrous materials, noinplanar and planar bending stress flaws may be evaluated following the general approach of 3.0(a) through 3.0(g) above- For pLanar flaws in ductile = nondimensional stress intensity factor for materials, the approach given in 3.0(b),and 3.0(g) may through-wall circumferential flaw under be used; otherwise, the approach given in 3.0(c) and membrane stress 3.0(g) should he applied. Structural factors provided in L = maximum extentt of a local thinned area with t 4.0 shall be used. It is the responsibitiiy of th evaluator to t < o¢,n establish conservative estimates of strength and fracture = length of through-wall crack for the hole pen-toughness for the piping material. etration in the axial direction of the pipe length of through-wall crack for the hole di-ameter penetration in the circumferential di-4.0 ACCEPTANCE CRITERIA rection of the pipe L. = maximum extent of a local thinned area with Piping containing a circumferential planar flaw is ac- t< ,, ceptable for temporary service when flaw evaluation pro-L,ýýJ = axial extent of wall thinning below t,,, vides a margin using the structural factors in Appendix C, C-2621. For axial planar flaws, thc stuctural factors L.(, = circumferential extent of wall thinning t below ,,r. for temporary acceptance are as specified in Appendix C, C-2622. Piping coutaining a nonplanr part through- R = pipe radius wall flaw is acceptable for temporary service if tp Ž t.4ý, outside pipe radius where ti., is determined from 3.0(d). Piping containing S = allowable stress at operating temperature a nonplanar through-wall flaw is acceptable for temporary SF. = structural factor on primary membrane stress service when the flaw conditions of 3.0(e) or 3.0(f) are Sr = coefficient for temperature dependence in the satisfied. crack growth relationship

                                                                              = Code-specified ultimate tensile strength S), Code-specified yield strength 5.0    AtUGMENTED EXAMINATION                                              W. = maximum extent of a local thinned area per-An augmented volumetric examination or physical                               pendicular to L,, with t < rT;,

measurement to assess degradation of the affected system c half crack length shall be performed as follows: dm/dr = flaw growth rate for stress corrosion cracking (a) From the engineering evaluation, tie most suscep- 44 diameter equivalent circular hole at t,4 tible locations shall be identified. A sample size of at d,;= diameter of equivalent circular hole at tfa least five of the most susceptible and accessible locations. = total crack length = 2c 7 (W-513-2) SUPP. 1 -NC Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of Mechanical Engineers. All rights reserved. 1.5-163

CASE (continued) N-513-2 CAE OF 0ASNE BOILER AMl PSS VESSML CODE Ca=allowable axial throu'gh-wall flaw length q=nominal longitudinal bending stress for pri-

         = exponent in the crack growth relationship                        mary loading without stress intensification p     maximum operating pressure at flaw location                     factor r wall thickness                                                = half crack angle for through-wall circunfer-t adjusted wall thickness which is varied for                      ential flaw evaluation puipose in the evaluation of a through-wall nonplanar flaw                         7.0    APPLICABILITY allowable local thickness for anonpianar flaw          This Case is applicable from the 19.83 Editon with the
         =~ minimum wall thickness required for pres-           Winter 1985 Addenda through the 2001 Edition with the sure loading                                        2003 Addenda. References in this Case to Appendix C
         = nominal wall thickness                               shall mean Appenidx C ofthe 2002 Addenda. For editions tp = minimum remaining wall thickacss                     and addenda prior to 2002 Addenda, Class I pipe flaw A = nondimensional half crack length for                  evaluation procedures may be used for other piping through-wall axial flaw                             classes. As a matter of definition, the term "structural
         = material flow stress                                 factor" is equivalent to the term "safety factor" that is o=,j= pipe hoop s*ess due to pressure                      used in earlier editions and addenda.

SUPP. 1 - Nc 8 (N-s13-2) Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of Mechanical Engineers. All rights reserved. 1.5-164

CASE (continued) CASES OF ASME BOILER AIM. PRESSUEM VESSEL CODE N-513-2 1.OE-02 - ___ Austenitlo Pripingi;___ 1.01E-04 ___ - - - ___ C b CD U UAE-06 ____ - - - - - --- tGE-07 ___ _- 1.OE 1 I-10 Stress Intensity Factor, K(ksi M.0-51

                                                                                              -  I  100 GENERAL NOTE: (St conyersion: 1.0 iMnr    7.06 x 10" mMfec; LO Ksl In.0s= 1.099 MFa mO5; C =    F- 3211.8).

FIG. 6 FLAW GROWTH RATE FOR IGSCC IN AUSTENITIC PIPING 9 (N-S13-2) SUPP. 1---NC Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of Mechaniical Engineers. All rights reserved. 1.5-165

CASE (continued) N-513-2 CAME OF ASME BOILER AND PRESSURE VESSEL CODE

                                                                                              ~i f IJ 2OF_

4; 1-~ 1.OE-08 B 1.LEiiz i 10 Stress Intensity Factor, K(ks!in.9-) 100 GENERAL NOTE: (SI conversion: 1.0 *nhr - 7.06x lO'nimnsec, Lo Isf InW' = 1099 MPa imns;=O = VF - 321L,8). FIG. 7 FLAW GROWTH RATE FOR TGSCC IN AUSTENITIC PIPING SUFP. 1 - NC 10(M-513-2) IReprinted from ASMEE 2004 Edition Code Cases, Nuclear Components, by permission of The Society of Mechanical Engineers. All rights reserved. American 1.5-166

                                                                                                  .CASE (continued)

CASE oF aSm Bomtt~ Am4?RESSuRE vESSL CODE N-513-2 APPENDIX I RELATIONS FOR Fm, Fb, AND F FOR THROUGH-WALL FLAWS 1-1.0 DEFINITIONS Ab = -3.26543 + 1.52784 (R -it) - 0.072698 (R/i) 2

                                                                           + .016011 (R) 3 '

Pot through-walI flaws, the crick depth (a) 'will be 2 Bb = 11.36322 - 3.91412 (RIr) + 0.18619 (R/r) replaced with half crack length (c). i thsstress intensity

                                                                           - 0.0o9499 (RI'.

factor equations in C-7300 and C-7400 of Section XI, C4 = -3.18609 + 3.84763 (Rh) -0.18304 (RAI)2 Appendix C. Also, Q Will be set equal to unity in C-7400.

                                                                           + 0.06403 (R*I?)

Equations for F, and F&are accurate for Rht between 1-2.0 CIRCUMFERENTIAL FLAWS 5 and 20 and become increasingly conservative for RIh For a range of R/t between 5 and 20, the following greater than 20. Altemative solutions for F,. and F, may equations for F. and Fb may be used: be used when RIf is greater than 20. F = I+ I A. (ehr)' 3 + B. (oir)15 4C. (01,r)" 1-3.0 AXIAL. FLAWS

          =b I +A&(       rI+ Bb (08.IS+Cb(em"r For internal pressure loading, the following equation where                                                           for F may be used:

49= Half cr .ack angle =c~IR R = Mean pipe radius F = I + 0.072449A + 0,64856A2 - 0.2327A'

        = Pipe wall thickness
        =
                                                                           + 0.038154A4 - 0.0023487AS and where A, = -2.02917 + 1.67763 (RI) - 0.07987 (Rlt)2 c     half crack length
          + 0.00176 (RIt)3 B, = 7.09987 - 4.42394 (RI:) + 0.21036 (R/r)                      A =/(R)V2
          - 0.00463 (Rlt)                                          The equation for F is accurate for A between 0 and 5.

C,. 7.79661 + 5.16676 (R/) - 0.24577 (RI:) 2 Alternative solutions for F may be used when Ais greater 4+0.00841 (RJI)P than 5. 11 fN-513-2) SUPP. 1 - NC Reprinted from ASME 2004 Edition Code Cases, Nuclear Components, by permission of The American Society of.Mechanicar Engineers. All rights reserved. 1.5-167

10 CFR 50.55a REQUEST NO. 15 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY ASME Code Component(s) Affected Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section Xl, Class 1, Reactor Vessel Nozzle-to-Vessel welds specified below and in-detail in Table A: Recirculation Inlet Nozzle N-2B Weld - N-2B NV Recirculation Inlet Nozzle N-2G Weld - N-2G NV Feedwater Inlet Nozzle N-4A Weld - N-4A NV Reactor Head Spare Nozzle N-6A Weld - N-6A NV Capped Control Rod Drive (CRD) Return Nozzle N-9 Weld - N-9 NV

2. Applicable ASME Section XI Code Edition and Addenda The applicable ASME Section Xl Code for the Monticello Nuclear Generating Plant (MNGP), Fourth Ten-Year Inservice Inspection (ISI) Interval is the 1995 Edition with the 1996 Addenda.

3. Applicable Code Requirement

ASME Class 1 Nozzle-to-Vessel welds are subject to the examination requirements of Subsection IWB Table IWB-2500-1, as shown below, and 10 CFR 50.55a(b)(2)(xv)(G). The welds are required to be examined once within the Fourth Ten-Year Interval: Code Class: 1

References:

IVVB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90

Description:

Nozzle-to-Vessel Welds Component Numbers: S ee Section 1 and Table A System: Reactor Vessel Examination Method: Volumetric - Ultrasonic Testing (UT) Examination Volume: Figure IWB-2500-7(b) In August 2005, the Nuclear Regulatory Commission (NRC) issued Regulatory Guide (RG) 1.147, Revision 14, Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1 (Reference 1). In RG 1.147, the NRC identifies the ASME Code Cases that they have determined to be acceptable alternatives to applicable parts of Section Xl, and that these Code Cases may be used by licensees without requesting authorization from the NRC provided that they are used with any identified limitations or modifications. RG 1.147, Table 1 lists the 1.5-1 68

10 CFR 50.55a REQUEST NO. 15 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY following two Code Cases as acceptable to the NRC for use by a licensee with no identified limitations or modifications: 1) Code Case N-460 (Reference 2), and 2) Code Case N-613-1 (Reference 3). Code Case N-460 states in part, "when the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent." NRC Information Notice (IN) 98-42 (Reference 4) termed a reduction in coverage of less than 10 percent to be "essentially 100 percent." IN 98-42 states in part, "The NRC has adopted and further refined the definition of 'essentially 100 percent' to mean 'greater than 90 percent'...has been applied to all examinations of welds or other areas required by ASME Section XI." Code Case N-613-1 provides an alternative examination volume that includes the width of the weld plus one-half inch of adjacent base metal on each side of the widest part of the weld. In comparison, the examination volume required by the Figure IWB-2500-7(b) includes the width of the weld plus the adjacent base metal on each side of the widest part of the weld equal to one-half of the vessel shell wall thickness.

4. Impracticality of Compliance Construction Permit CPPR-31 was obtained for the MNGP in 1967. The MNGP systems and components were designed and fabricated before the examination requirements of ASME Section Xl were formalized and published. Therefore, MNGP was not specifically designed to meet the requirements of ASME Section Xl and full compliance is not feasible or practical within the limits of the current plant design.

10 CFR 50.55a recognizes the limitations to inservice inspection of components in accordance with Section XI of the ASME Code that are imposed due to early plants' design and construction, as follows: 10 CFR 50.55a(g)(1): For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (5) of this section to the extent practical. 1.5-169

10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY 10 CFR 50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and pre-service examination requirements, set forth in Section X1 of editions of the ASME Boiler and Pressure Vessel Code ... to the extent practical within the limitations of design, geometry and materials of construction of the components. 10 CFR 50.55a(g)(5)(iii): If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in § 50.4, information to support the determinations. The inspection limitations on the subject components are due to inherent nozzle design geometric contours (see Table A). A description of the examination methodology used to provide the maximum obtainable coverage is provided in Section 6 of this request. This methodology is based on ASME Section Xl, Appendix VIII qualification and was applied to the extent practical within the design constraints of the components. Enclosure 3 provides cross-sectional diagrams of the subject welds showing the geometric contour of the component design in relation to the welds and the coverage obtained within the examination volume requirements of Code Case N-613-1, Figure 2.

5. Burden Caused by Compliance Compliance with the examination coverage requirements of ASME Section Xl would require modification, redesign, or replacement of components where geometry is inherent to the component design.

6. Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry and materials of construction.

1.5-1 70

10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Nuclear Management Company (NMC) performed qualified examinations that achieved the maximum, practical amount of coverage obtainable within the limitations imposed by the design of the components. Additionally, as Class 1 examination Category B-P components, a VT-2 examination is performed on the subject components of the Reactor Coolant Pressure Boundary (RCPB) during system pressure tests each refueling outage. This was completed during the 2007 refueling outage and no evidence of leakage was identified for these components. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the requirements of ASME Section XI Table IWB-2500-1, Category B-D, Item B3.90, and proposes to utilize these completed exams as acceptable alternatives that provide reasonable assurance of continued structural integrity. Basis for Use The NMC Nondestructive Examination (NDE) procedures incorporate inspection techniques qualified under Appendix VIII of the ASME Section Xl Code by the Performance Demonstration Initiative (PDI) for examination of the subject nozzle-to-vessel welds, and allow the examination volume to meet the provisions of alternative requirements (i.e., Code Case N-613-1). The examinations were performed using a manual contact method from the nozzle outside blend radius and vessel surfaces. Coverage was obtained by following the scan parameters designated within NMC NDE procedures and as defined by MNGP specific Electric Power Research Institute (EPRI) computer modeling reports (References 5 and 6) for each nozzle configuration and angle. It should be noted that that the scans defined by the EPRI report are only applicable to the inner 15 percent of the weld volume when scanning in the parallel direction. The refracted longitudinal wave mode of propagation was applied for all the radial scans of the exam volume, and to the outer 85 percent of the exam volume for parallel scans. The shear wave mode of propagation was applied for each of the transducer and wedge combinations required for the remaining inner 15 percent of the parallel scan exam volume. The subject components received the required examination(s) to the extent practical within the limited access of the component design. One hundred (100) percent coverage was obtained for the inner 15 percent of the examination volume. The examination limitations for the subject components were encountered within the outer 85 percent of the examination volume. For the examinations conducted, satisfactory results were achieved, and no evidence of unacceptable flaws was detected with the inspection techniques. 1.5-171

10 CFR 50.55a REQUEST NO. 15 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Due to the design of these welds it was not feasible to effectively perform a volumetric examination of "essentially 100 percent" of the required volume. The nozzle-to-vessel welds are accessible from the vessel plate side of the weld and are examined to the extent practical, but there are no qualified examinations to obtain coverage of the excluded areas within the outer 85 percent of the examination volume due to the nozzle forging curvature. Additional coverage for the limited areas was not achievable or practical, based on the latest qualified ultrasonic technology, nor by other considered examinations methods, such as radiography. NMC has concluded that if significant degradation existed in the subject welds, it would have been identified by the examinations performed. Additionally, as Class 1 examination category B-P components, VT-2 examinations were performed on the subject components in association with the Reactor Coolant Pressure Boundary system pressure test performed during the 2007 refueling outage. No evidence of leakage was identified during this system test. The materials for the subject components are A508 CI II nozzle forgings welded to A533 Cl I vessel shell plate. A review of operating experience within the nuclear industry did not reveal any instances of cracking in this location and type of weldment. The MNGP reactor vessel water chemistry is controlled in accordance with the 2004 revision to the BWR Water Chemistry Guidelines (Reference 7). Also a hydrogen water chemistry system is used to reduce the oxidizing environment in the reactor coolant. These additional measures provide added assurance against the initiation of cracking or corrosion from the inside surface of the reactor vessel. An inerted primary containment environment during operation provides assurance of corrosion protection on the outside surface of the reactor vessel. The provisions described above as an alternative to the code requirement will continue to provide reasonable assurance of the structural integrity of the subject welds. The examinations were completed to the extent practical and evidenced no unacceptable flaws present. VT-2 examinations performed on the subject components during system pressure testing each refueling outage (in accordance with examination Category B-P) provide continued assurance that the structural integrity of the subject components is maintained. Additionally, the MNGP Water Chemistry Program and inerted primary containment environment provide added measures of protection for the component materials. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NMC requests relief from the ASME Section Xl examination requirements for the subject nozzle-to-vessel welds. 1.5-172

10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY

7. Duration of Proposed Alternative NMC requests the granting of this relief for the Fourth Ten-Year Inservice Inspection Interval of the Inservice Inspection Program for the MNGP that is scheduled to end on May 31, 2012.

8. Precedents The NRC has granted relief for other nozzle-to-vessel shell welds at the MNGP, most recently for the current Fourth Ten-Year Inservice Inspection Interval (Reference 8). Also, the NRC has granted relief for the Quad Cities Nuclear Power Station, Units 1 and 2 (Reference 9), the Dresden Nuclear Power Station, Units 2 and 3 (Reference 10), and the Prairie Island Nuclear Generating Plant, Unit 2 (Reference 11).

1.5-1 73

10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY REFERENCES

1. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1," Revision 14, August 2005.

2. ASME Section XI Code Case N-460, "Alternative Examination Coverage for Class 1 and Class 2 Welds."

3. ASME Section XI Code Case N-613-1, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.10 and B3.90, Reactor Nozzle-To-Vessel Welds, Figures IWB-2500-7(a), (b), and (c)."

4. NRC Information Notice 98-42, "Implementation of 10 CFR 50.55a(g) In-service Inspection Requirements."

5. EPRI Internal Report IR-2004-63, "Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations," dated December 2004.

6. EPRI Internal Report IR-2006-100, "Monticello Nozzle Inner Corner Regions and Nozzle-to-Shell Weld Examinations," dated January 2006.

7. BWRVIP-1 30, "BWR Water Chemistry Guidelines - 2004 Revision" (EPRI Topical Report TR-1 008192).

8. NRC letter to NMC, "Monticello Nuclear Generating Plant (MNGP) - Fourth 10-Year Interval Inservice Inspection (ISI) Program Plan Relief Request No. 13 (TAC No. MC8882)," dated July 18, 2006.

9. Letter from NRC to Exelon Generation Company, LLC, "Quad Cities, Units 1 and 2 - Relief Request CR-39 for Third 10-Year Inservice Inspection Interval (TAC Nos. MC2427 and MC2428)," dated May 10, 2005.

10. Letter from NRC to Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request CR-26 For Third 10-Year Inservice Inspection Interval (TAC Nos. MC3269 and MC3270)," dated October 1, 2004.

11. NRC letter to NMC, "Prairie Island Nuclear Generating Plant, Unit 2 - Evaluation of Relief Request No. 16 for the Unit 2 3 rd 10-year Interval Inservice Inspection Program (TAC No. MC1775)," dated October 18, 2004.

1.5-174

10 CFR 50.55a REQUEST NO. 15 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY TABLE A - Category B-D, "Full Penetration Welds of Nozzles in Vessels," Item No. B3.90 Percent Coverage and Limitations for Nozzles N-2B, N-2G, N-4A, N-6A, and N-9 Code System Code Component Category and Component and Percent* Exam and Component ID Examination Volume Coverage Report Item No. Description Required Obtained Limitations Number Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Recirculation Inlet N-2B NV Code Case N-613-1 78% configuration. 2007UT058 B3.90 Nozzle N-2B Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Recirculation Inlet N-2G NV Code Case N-613-1 78% configuration.2007UT061 B3.90 Nozzle N-2G Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Feedwater Inlet N-4A NV Code Case N-613-1 79% configuration. 2007UT103 B3.90 Nozzle N-4A Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D Top Head Spare N-6A NV Code Case N-613-1 86% configuration. 2007UT104 B3.90 Nozzle N-6A Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to nozzle B-D CRD Return Nozzle N-9 NV Code Case N-613-1 85% configuration.2007UT102 B3.90 (capped) N-9 Figure 2_configuration. Due to the nozzle design it was not feasible to effectively examine essentially 100 percent of the required examination volume as defined in Figure 2 of Code Case N-613-1. Percentages are conservatively rounded down to the nearest whole number. It should be noted that 100 percent of the inner 15 percent was examined for all components listed above. 1.5-175

10 CFR 50.55a REQUEST NO. 15 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY EXAM LIMITATIONS IMPOSED BY COMPONENT DESIGN AND CONSTRUCTION This enclosure contains a series of excerpts from the ISI Ultrasonic Testing (UT) reports applicable to the subject components. These excerpts contain sketches depicting the component configuration with physical limitations imposed by the design, e.g., geometrical contour, weld position, interferences, and a cross sectional view depicting the UT coverage and limitations in relation to the required examination volume. Also included is a sketch of a typical reactor vessel nozzle contour and the resulting effect that causes the UT transducer to lift and lose effective coupling when it reaches the nozzle blend radius. COMPONENT REPORT PAGE(S) N-2B NV 2007UT058 Pages 1-2 N-2G NV 2007UT061 Pages 3-4 N-4A NV 2007UT103 Page 5 N-6A NV 2007UT104 Page 6 N-9 NV 2007UT102 Page 7 Typical Reactor Vessel Nozzle Contour Affecting Page 8 Transducer Contact at blend radius 1.5-176

Coverage drawings excerpted from applicable reports Component N-2B NV Report # 2007UT058 Supplemental Report Nmý) Report No.: 2007UT058 Summary No.: 102658 R3.50 in Monticello N2 Coverage Plot Axial scan direction 1.5-177

Component N-2B NVmR Report # 2007UT058 Supplemental Report NM~) Report No.: 2007UT058 Summary No.: 102658 Monticello N2 Coverage Plot Parallel scan direction 1.5-178

Component N-2G NV Report # 2007UT061 Supplemental Report MV~) Report No.: 2007UT061 Summary No.: 102668 R3.50mi 5 in Monticello N2 Coverage Plot Axial scan direction 5.25 in 1.5-179

Component N-2G NVm Report # 2007UT061 Supplemental Report NMP Report No.: 2007UT061 Summary No.: 102668 Monticello N2 Coverage Plot Parallel scan direction 1.5-180

Component N-4A NV Report # 2007UT103 PJM7&) Suppfementai Report Report No.: 2007UT103 Summary No.: 102684 Comments: Coverage Plots in 1.5-181

Component N-6A NV Report # 2007UT104 NMC) Supplemental Report Report No.: 2007UT104 Summary No.: 102375 Comments: N-SA NV Coverage Plots MonticeUo N6 Coverage Plot Monticello N6 Coverage Plot Axial scan direction Circ scan direction

                     -ý7         AN                                    \,      B      C    D 3.00 in Inn    '-"1""
                                                                        .   ..          7,                          i 1.5-182

S Component N-9 NV Report # 2007UT102 W.M - )I Supplemental Report Report No.: 2007UT102 Summary No.: 102700 Comments: Coverage Plots Moni icello N9 Coverage Plot

                                                                     ýLrc scan direction on low 15Y 1.5-183

Typical Representation of Nozzle Limitations Coverage affected by liftoff due to radius R3.SOin Axial scan shown Area of covcradge 60 &90 5.25 in 60Odel N2 Nozzle shown as example 1.5-184

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 16 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY ALTERNATIVE TO NOZZLE-TO-VESSEL WELD AND INNER RADIUS EXAMINATIONS 1.5-185

10 CFR 50.55a Request No. 16 Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(i) Which Provides an Acceptable Level of Quality or Safety Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations 1.0 ASME Code Component(s) Affected Code Class: I Component Numbers: N2, N3, N5, N6 and N8 Nozzles (See Enclosure 2 for complete list of nozzle identifications.) Examination Category: B-D (Inspection Program B) Item Number: B3.90 and B3.100

Description:

Alternative to ASME Section Xl, Table IWB-2500-1 (for the components described above) 2.0 Applicable ASME Code Edition and Addenda The Monticello Nuclear Generating Plant (MNGP) is currently in the fourth 10-year Inservice Inspection (ISI) Program interval and is committed to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, "Rules for Inservice Inspection of Nuclear Power Plant Components," (ASME Section Xl), 1995 Edition through 1996 Addenda. Additionally, for ultrasonic examinations, ASME Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition through 1996 Addenda is implemented, as required and modified by 10 CFR 50.55a.

3.0 Applicable Code Requirement

Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzles in Vessels - Inspection Program B" Class I nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number B3.90, "Nozzle-to-Vessel Welds," and B3.100, "Nozzle Inside Radius Section." The required method of examination is 1.5-186

volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles require examination each interval. All of the nozzle assemblies identified in Enclosure 2 are full penetration welds to the vessel shell or head.

4.0 Reason for Request

Enclosure 2 provides a complete listing of the applicable Reactor Pressure Vessel (RPV) nozzles. The proposed alternative provides an acceptable level of quality and safety, and the reduction in inspection scope could result in a dose savings of as much as 10 Person-Rem for the unit over the remainder of the interval. 5.0 Proposed Alternative and Basis for Use In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested from performing the required examinations on 100 percent of the nozzle assemblies identified in Table 5-1 below (see Enclosure 2 for complete list of RPV Nozzles). As an alternative for all welds and inner radii identified in Table 5-1, the NSPM proposes to examine a minimum of 25 percent of the MNGP nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, in accordance with ASME Code Case N-702. For the nozzle assemblies identified in Enclosure 2, this would mean that examinations would be required for three of the Recirculation Inlet (N2) nozzles and one from each of the other nozzle groups, as identified below. Table 5-1 MNGP Summary Nozzles Minimum Number Year(s) Nozzle Group per Group Number to be Examined Examined Examined to Date Recirculation Inlet 10 3 6 Three in 2005 (N2) Two in 2007 One in 2009 Main Steam 4 1 2 One in 2005 (N3) One in 2009 Core Spray 2 1 1 One in 2005 (N5) Closure Head 2 1 2 One in 2007 Spare (N6) One in 2009 Jet Pump 2 1 2 One in 2005 Instrumentation (N8) , I I One in 2009 1.5-187

Therefore, upon authorization, no further nozzles in the applicable groups would remain to be inspected for the remainder of the interval. ASME Code Case N-702 stipulates that a VT-1 examination may be used in lieu of the volumetric examination for the inner radii (i.e., Item No. B3.100, "Nozzle Inside Radius Section"). NSPM is only requesting to perform volumetric examinations of the applicable nozzle inner radius sections. NSPM is not requesting use of the VT-1 examination provisions included in the code case in lieu of performing volumetric examinations. The NSPM is not currently using ASME Code Case N-648-1 at the MNGP for the identified components for enhanced magnification visual examination and has no plans of using ASME Code Case N-648-1 on those components in the future. Electric Power Research Institute (EPRI) Technical Report 1003557, "BWRVIP-108: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," (Reference 1) provides the basis for ASME Code Case N-702. The evaluation found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure event are very low (i.e., <1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified. This EPRI report was approved by the NRC in a safety evaluation (SE) dated December 19, 2007 (Reference 2). Section 5.0, "Plant Specific Applicability," of the SE indicates that each licensee who plans to request relief from ASME Code, Section Xl requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVI P-1 08 report as the technical basis for the use of ASME Code Case N-702 as an alternative. The NRC SE further states that each licensee should demonstrate the plant specific applicability criteria from the BWRVIP-108 report to its units in the relief request by showing that all the general and nozzle-specific criteria addressed below are satisfied (i.e., as described in Enclosure 3). (1) The maximum RPV heatup/cooldown rate is limited to less than 115 0F per hour. MNGP Technical Specification (TS) 3.4.9,"RCS Pressure and Temperature (PIT) Limits," provides a limiting condition for operation (LCO). The heatup/cooldown rate is referenced in the MNGP operating procedures where applicable such as scrams and start-ups. The maximum heatup / cooldown rate of 100°F per hour is specified within the MNGP Updated Safety Analysis Report in Table 4.1-1, "Reactor Coolant System Data." 1.5-1 88

For the Recirculation Inlet Nozzles: (2) (pr/t) I CRPV < 1.15, where: p = RPV normal operating pressure, r RPV inner radius, t = RPV wall thickness, and CRPV = 19332 (i.e., 1000 psi x 110 inch / 5.69 inch, based on the BWRVIP-108 recirculation inlet nozzle / RPV finite element method (FEM) model); (3) [p(ro 2 + r, 2 ) / (ro 2 _ ri 2)] / CNOZZLE < 1.15, where: p = RPV normal operating pressure, r, = nozzle outer radius, r = nozzle inner radius, and CNOZZLE = 1637 [i.e., 1000 psi x (13.988 2 + 6.875 2) /(13.988 2 6.875 2)] based on the BWRVIP-108 recirculation inlet nozzle / RPV FEM model]; For the Recirculation Outlet Nozzles: (4) (pr/t) / CRpv < 1.15, where: p = RPV normal operating pressure, r = RPV inner radius, t RPV wall thickness, and CRPV = 16171 (i.e., 1000 psi x 113.2 inch / 7.0 inch, based on the BWRVIP-1 08 recirculation outlet nozzle / RPV FEM model); (5) [p(ro 2 + r, 2) / (ro 2 _ r, 2 )] / CNOZZLE < 1 .15, where: p = RPV normal operating pressure, ro = nozzle outer radius, r= nozzle inner radius, and CNOZZLE = 1977 [i.e., 1000 psi x (22.31 2 + 12.78 2) / (22.31 2 12.78 2)], based on the BWRVIP-108 recirculation outlet nozzle / RPV FEM model]. Note that as stated in the NRC SE, only the recirculation inlet and outlet nozzles need to be checked because the conditional probabilities of failure, P(FIE)s, for the other nozzles are an order of magnitude lower. 1.5-1 89

                                                                                     . ! I Based upon the above information, all requested MNGP RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles, meet the general and nozzle-specific criteria in BWRVIP-108.

Therefore, ASME Code Case N-702 is applicable. Use of ASME Code Case N-702 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for all requested RPV nozzle-to-vessel shell full penetration welds and nozzle inner radii sections, with the exception of the recirculation outlet nozzles. 6.0 Duration of Proposed Alternative The proposed alternative will be applied for the remainder of the fourth 10-year interval of the MNGP ISI Program. 7.0 Precedent The NRC has approved similar requests to adopt an alternative to the ASME Section Xl, Table IWB-2500-1 criteria to allow reduced percentage requirements for nozzle-to-vessel weld and inner radius examinations for the Duane Arnold Energy Center (Reference 3), Perry Nuclear Power Plant, Unit I (Reference 4); the Dresden Nuclear Power Station, Units 2 and 3 (Reference 5), and the Clinton Power Station, Unit No.1 (Reference 6). 1.5-190

8.0 References

1. BWRVIP to NRC letter submitting EPRI report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108)," dated November 25, 2002, as supplemented by letters dated July 25, 2006, September 13, 2007, and November 21, 2007.

2. NRC letter to BWRVIP Chairman (R. Libra), "BWRVIP Safety Evaluation of Proprietary EPRI Report, "BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-1 08)," dated December 19, 2007. (ADAMS Accession No. ML073600374)

3. NRC letter, "Duane Arnold Energy Center - Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations (TAC No. MD8193)," dated August 29, 2008. (ADAMS Accession No. ML082040046)

4. NRC letter, "Perry Nuclear Power Plant, Unit No.1 -Request for Relief Related to Inservice Inspection Relief Request IR-054 (TAC No. MD8458)," dated December 29, 2008. (ADAMS Accession No. ML082960729)

5. NRC letter, "Dresden Nuclear Power Station, Units 2 and 3 - Alternative to Nozzle-to-Vessel Weld and Inner Radius Examinations (TAC Nos. ME0882 and ME0883)," dated November 3, 2009. (ADAMS Package Accession No. ML092940436)

6. NRC letter, "Clinton Power Station, Unit No.1 - Proposed Alternative to 10 CFR 50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections (TAC No. ME0218)," dated August 24, 2009. (ADAMS Accession No. ML092300394) 1.5-191

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 16 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY ALTERNATIVE TO NOZZLE-TO-VESSEL WELD AND INNER RADIUS EXAMINATIONS TABLE OF ASME COMPONENTS AFFECTED 1.5-192

Table of ASME Components Affected Nozzle 0D Nominal Category Item ie Exam Year Nozzle-to-Vessel (NV) Number Number System Size ExmYa Inner Radius (IR) (Inches) (PDI) N-2A (NV) B-D B3.90 Recirculation 12 2009 (inlet) N-2A (IR) B-D B3.100 Recirculation 12 2009 (Inlet) N-2B (NV) B-D B3.90 Recirculation 12 2007 (inlet) N-2B (IR) B-D B3.100 Recirculation 12 2007 (Inlet) N-2C (NV) B-D B3.90 Recirculation 12 (inlet)

      -2C (IR)         B-D     63100        Recirculation    12 N-2C-                      B3.100          (inlet)       12 N-2D (NV)          B-D       B13.10     Recirculation    12        2005 (Inlet)

N-2D (IR) B-D 83.100 Recirculation 12 2005 (inlet) N-21E (NV) B-D B3.90 Recirculation 12 2005 (Inlet) N-2E (IR) 6-0 B3.1 00 Recirculation 12 2005 (Inlet) N-21F (NV) 6-0 B3.90 Recirculation 12 (Inlet) N-2F (IR) B-D 133.100 Recirculation 12 (Inlet) N-2G (NV) B-0 B3.90 Recirculation 12 2007 (Inlet) N-2G (IR) B-D 83.100 Recirculation 12 2007 (inlet) 1.5-193

Table of ASME Components Affected Nominal Nozzle ID Categqory Item Pipe Exam Year Nozzle-to-Vessel (NV) Number Number System Size Inner Radius (IR) (Inches) P) Recirculation 12 N-21-1 (NV) B-D B3.90 (Inlet) N-2H (IR) B-D B3.100 Recirculation 12 (Inlet) N-2J1 (NV) B-D B3.90 Recirculation 12 2005 (Inlet) N-2,1 (IR) B-D B3.1 00 Recirculation 12 2005 (Inlet) N-2K (NV) B-D B3.90 Recirculation 12 (Inlet) N-2K (IR) B-0 B3. 100 Recirculation 12 (Inlet) N-3A (NV) B-D B3.90 Main Steam 18 2005 N-3A (IR) B-D B3.100 Main Steam 18 2005 N-3B (NV) B-D B3.90 Main Steam 18 --- N-3B (IR) B-D B3.100 Main Steam 18 N-3C (NV) B-D B3.90 Main Steam 18 2009 N-3C (IR) B-D B3.100 Main Steam 18 2009 N-3D (NV) B-D B3.90 Main Steam 18 --- N-3D (IR) B-D B3.100 Main Steam 18 N-5A (NV) B-D B3.90 Core Spray 8 --- N-5A (IR) B-D B3.100 Core Spray 8 --- N-5B (NV) B-D B3.90 Core Spray 8 2005 N-5B (IR) B-D B3.100 Core Spray 6 2005 1.5-194

Table of ASME Components Affected Nominal Nozzle ID Cateqory Item Pipe Exam Year Nozzle-to-Vessel (NV) Number Number System Size Inner Radius (IR) (inches) (PDI N-6A (NV) B-D B3.90 Closure Head 6 2007 Spare N-6A (IR) B-D B3.100 Closure Head 6 2007 Spare N-6B (NV) B-D B3.90 Closure Head 6 2009 Spare N-6B (IR) B-D B3.100 Closure Head 6 2009 Spare Jet Pump N-8A (NV) B-D B3.90 Instrumentation 4 2005 Jet Pump 4 2005 N-8A (IR) B-D B3. 100 Instrumentation N-81 (NV) B-D B3.90 Jet Pump 4 2009 Instrumentation Jet Pump N-813 (IR) B-D B3.1 00 Instrumentation 4 2009 1.5-195

0 MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 16 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY ALTERNATIVE TO NOZZLE-TO-VESSEL WELD AND INNER RADIUS EXAMINATIONS VERIFICATION OF PLANT SPECIFIC NOZZLE APPLICABILITY TO APPLY CODE CASE N-702 IN ACCORDANCE WITH BWRVIP-1 08 1.5-196

Verification of Plant Specific Nozzle Applicability to Apply Code Case N-702 to the MNGP in Accordance With BWRVIP-108 (Criteria 1) Maximum RPV heatup/cooldown rate limited to less than 115 0F per hour. The Monticello Updated Safety Analysis Report indicates a maximum heatup/cooldown rate of 100'F per hour. This heatup/cooldown rate is referenced in the MNGP operating procedures, where applicable, such as scrams and start-ups. MNGP Technical Specification 3.4.9,"RCS Pressure and Temperature (P/T) Limits," provides a limiting condition for operation (LCO). Values for Monticello Recirculation Inlet and Outlet Nozzles: Monticello specific values: p = RPV normal operating pressure = 1025 psig r = RPV inner radius = 102.5 inches t = RPV wall thickness = 5.0625 inches r,= Recirculation inlet nozzle outer radius = 14.1875 inches ri= Recirculation inlet nozzle inner radius = 7.0625 inches

    *r,, Recirculation outlet nozzle outer radius = 24.375 inches r= Recirculation outlet nozzle inner radius = 13.0625 inches BWRVIP-108 Constants:

CRpv (Recirculation Inlet Nozzles) = 19332 CNOZZLE (Recirculation Inlet Nozzles) = 1637 CRPV (Recirculation Outlet Nozzles) = 16171 CNOZZLE (Recirculation Outlet Nozzles) = 1977 For the Recirculation Inlet Nozzles: (Criteria 2) (pr/t) / CRPV < 1.15 [(1025 x 102.5)/5.06251/19332 = 1.07351 < 1.15 (Criteria 3) [p(ro 2 + ri 2) / (ro 2 - ri 2)] / CNOZZLE < 1.15 [1025 (14.1875 2 +7.0625 2) /(14.1875 2 -7.0625 2)] 1637= 1.03870 < 1.15 For the Recirculation Outlet Nozzles: (Criteria 4) (pr/t) / CRPV < 1.15 [(1025 x 102.5) / 5.0625] / 16171 = 1.28335 > 1.15 (does not pass) (Criteria 5) [p(ro 2+ ri 2 ) / (r, 2 - r, 2)] / CNOZZLE < 1.15 [1025 (24.375 2 + 13.0625 2) /(24.375 2 - 13.0625 2)] / 1977 = 0.93623 < 1.15 1.5-197

i . .- MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 17 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY EXTENSION OF PERMANENT RELIEF FROM VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS FOR THE RENEWED OPERATING LICENSE TERM 1.5-198

10 CFR 50.55a Request No. 17 Proposed Alternative In Accordance With 10 CFR 50.55a(a)(3)(i) Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term 1.0 ASME Code Component(s) Affected All of the affected reactor pressure vessel (RPV) circumferential shell welds are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, Class 1. Weld Examination Item Number Description Category No. VCBB-1 Circumferential Shell to Bottom Head Weld B-A B1.11 VCBA-2 Circumferential Shell to Shell Weld B-A B13.11 VCBB-3 Circumferential Shell to Shell Weld B-A B1.11 VCBB-4 Circumferential Shell to Shell Weld B-A B13.11 2.0 Applicable ASME Code Edition and Addenda The Monticello Nuclear Generating Plant (MNGP) is currently in the fourth 10-year Inservice Inspection (ISI) Program interval and is committed to the ASME Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," (ASME Section XI), 1995 Edition through 1996 Addenda. Additionally, for ultrasonic examinations, ASME Section Xl, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition through 1996 Addenda is implemented, as required and modified by 10 CFR 50.55a. 3.0 Applicable Code Requirement Table IWB-2500-1, "Examination Category B-A, Pressure Retaining Welds in Reactor Vessel" Examination Category B-A, Item Number B1.11, Circumferential Shell Welds, requires volumetric examination of all circumferential shell welds each interval. 1.5-199

                                                                                   ' " I 4.0 Reason for the Request On July 27, 2001 (Reference 1) the MNGP received U.S. Nuclear Regulatory Commission (NRC) authorization for a technical alternative that eliminated performance of RPV circumferential shell weld examinations for the duration of the full-term operating license that ends on September 8, 2010. The primary basis was an analysis in accordance with Boiling Water Reactor Vessel and Internals Project (BWRVIP) report, BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," (Reference 2) and NRC guidance, which indicated that the limiting conditional failure probability for the circumferential shell welds would be satisfied through the expiration of the current full-term operating license.

Anticipated changes in metallurgical conditions expected over the renewed license period required analysis and further evaluation to demonstrate the continued acceptability for not performing volumetric examinations of these RPV circumferential shell welds over the additional renewed operating license term of 20-years.(1) The analysis was based on BWRVIP-05 and BWRVIP-74, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal" (Reference 3). Information on the projected acceptability for continuing permanent deferral of volumetric examinations on RPV circumferential shell welds was provided in Section 4.2.6 of the License Renewal Application (LRA) and within responses'to NRC staff requests for additional information. On November 8, 2006, the NRC issued Renewed Facility Operating License DPR-22 for the MNGP, with an expiration date of midnight on September 8, 2030 (Reference 4). Accompanying the renewed license was NUREG-1865, "Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant" (Reference 5), which provides a summary of the safety basis for the acceptability of various aspects of the license renewal. NUREG-1865 Section 4.2.6, "RPV Circumferential Weld Examination Relief," discusses specifics of the NRC evaluation for this area and indicated the continued acceptability of continuing application of this alternative for the renewed license period of operation. Subsection 4.2.6.4, "Conclusion," of the NUREG states: The staff concluded that the applicant provided an acceptable demonstration, pursuant to 10 CFR 54.21 (c)(1)(ii), that the analyses of the RPV circumferential weld examination relief have been projected to the end of the period of extended operation. The staff also concluded that the USAR [Updated Safety Analysis Report] supplement contains an Reference 7 indicates that 54 Effective Full Power Years is the realistically expected value at the end of the original full-term (40 year) plus the renewed (20 year) operating license terms, 1.5-200

appropriate summary description of this TLAA [Time-Limited Aging Analysis] evaluation, sufficient to satisfy the requirements of 10 CFR 54.21(d). Section 4.2.6.2, "Staff Evaluation," and Appendix A to NUREG-1 865, indicate that while relief from performance of RPV circumferential shell weld examinations has been determined acceptable from a license renewal technical standpoint, a separate 10 CFR 50.55a (relief) request is necessary to authorize this alternative for the term of the renewed operating license. Accordingly, this 10 CFR 50.55a request is provided to meet the MNGP license renewal commitment to resubmit a request for authorization of permanent relief from the volumetric examination of RPV circumferential shell welds through the 20-year renewed operating license term. The Northern States Power Company - Minnesota (NSPM) is requesting this alternative in accordance with 10 CFR 50.55a(a)(3)(i) on the basis that this proposed alternative provides an acceptable level of quality and safety. 5.0 Proposed Alternative and Basis for Use Proposed Alternative The projected failure frequency of the subject welds at the MNGP has been determined to be sufficiently low for the duration of the renewed operating license term to justify eliminating the examinations required by 10 CFR 50.55a(g) in accordance with ASME Code Section Xl, Table IWB-2500-1, Examination Category B-A, Item No. B1.11, Circumferential Shell Welds. Pursuant to 10 CFR 50.55a(a)(3)(i), and consistent with guidance provided in NRC 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" (Reference 6), and the final license renewal safety evaluation report for proprietary report, BWRVIP-74, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal," (Reference 7), NSPM proposes the following alternate provisions for the subject weld examinations for the 20-year renewed operating license term. The examination requirements of ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No B1.12, for the RPV longitudinal shell welds, also known as vertical or axial welds, will be performed as required to the extent possible. 1.5-201

i !

• As a proposed alternative to the requirements of ASME Code Item No. B1.11 for the RPV circumferential shell welds, the longitudinal weld examinations for ASME Code Item No. B1.12 will include examination on the segment of RPV circumferential welds VCBA-2, VCBB-3, and VCBB-4 that intersects with the longitudinal welds, or approximately 2 to 3 percent of the RPV shell circumferential welds.
• As a proposed alternative to the requirements of ASME Code Item No. B13.11 for RPV circumferential weld VCBB-1, NSPM will perform volumetric examination on approximately 2 to 3 percent of the weld at an accessible location, rather than at the associated longitudinal weld intersections as proposed for the previously mentioned circumferential welds.
• The proposed examination alternative for the RPV circumferential shell welds may be performed from either the internal inside diameter (ID) surface, or from the external outside diameter (OD) surface of the RPV as determined by the MNGP.

  " Examination of the remaining portions of the RPV circumferential shell welds will be permanently deferred through the renewed operating license term.
   " Examination will be completed in accordance with the ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," for the interval's applicable Code of Record edition and addenda as required and modified by 10 CFR 50.55a, "Codes and standards."

Basis for Use The BWRVIP-74 report provides generic guidelines for the appropriate inspection and flaw evaluation recommendations to assure safety function integrity of reactor pressure vessel components continuing from the initial operating license term through the renewed operating license term. The NRC final license renewal safety evaluation for BWRVIP-74 concluded that Appendix E of the July 28, 1998, NRC safety evaluation for BWRVIP-05 conservatively evaluated BWR reactor pressure vessels to 64 Effective Full Power Years (EFPY) which is 10 EFPY greater than what is realistically expected at the end of an additional 20-year license renewal period. 1.5-202

i I The NRC staff analysis provides a technical basis for an alternative from the ASME Code Section Xl requirements for the volumetric examination of RPV circumferential shell welds for the license renewal period. The associated safety evaluation stated that to obtain relief (similar to the conditions promulgated in Generic Letter 98-05 (Reference 6)) each licensee would have to demonstrate that: (1) At the end of the license renewal period, the circumferential welds will satisfy the limiting conditional failure frequency for circumferential welds in Appendix E of the NRC staffs Final Safety Evaluation Report (FSER) for BWRVIP-05, and (2) That they have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC staff's FSER for BWRVIP-05. The following discussion describes how each of these criteria will be met during the renewed operating license period. Demonstrate that Circumferential Welds Will Satisfy the Limiting Conditional Failure Frequency at the End of the License Renewal Period The following discussion is taken from the staff evaluation(2) under Section 4.2.6, "RPV Circumferential Weld Examination Relief," within NUREG-1865, the MNGP license renewal safety evaluation report (SER), and summarizes the basis for use and the acceptability of the proposed alternative. The technical basis for relief is discussed in the staffs final SER concerning the BWRVIP-05 report, "BWR Vessel and Internals Project (BWRVIP), BWR Reactor Pressure Vessel Weld Inspection Requirements," enclosed in the letter dated July 28, 1998, from Mr. G.C. Laines, NRC, to Mr. C. Terry, the BWRVIP Chairman. In this letter, the staff concluded that, because the failure frequency for circumferential welds in BWR plants is significantly below the criterion specified in RG [Regulatory Guide] 1.154, "Format and Content of Plant-Specific Pressurized Thermal Shock Safety Analysis Reports for Pressurized Water Reactors," and below the core damage frequency of any BWR plant, continued inspection of the RPV circumferential welds will result in a negligible decrease in an already acceptably low rate of RPV failure; therefore, elimination of the inservice inspection (ISI) for RPV circumferential welds is justified. The staffs letter indicated that BWR

2. See Subsection 4.2.6.2, "Staff Evaluation" of NUREG-1865.

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applicants may request relief from 10 CFR 50.55a(g) ISI requirements for volumetric examination of circumferential RPV welds by demonstrating that (1) through the expiration of the license period, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in the NRC staffs July 28, 1998 evaluation, and (2) implementation of operator training and established procedures that limit the frequency of cold overpressure events to the frequency specified in the staffs SER. The letter indicated that the requirements for inspection of circumferential RPV welds during an additional 20-year license renewal period will be reassessed, on a plant-specific basis, as part of any BWR LRA [license renewal application]; therefore, the applicant must request relief from inspection of circumferential welds during the license renewal period, pursuant to 10 CFR 50.55a. Section A.4.5 of the BWRVIP-74 report indicates that the staff's SER of the BWRVIP-05 report conservatively evaluated the BWR RPVs to 64 EFPY [effective full power years], which is 10 EFPY greater than realistically expected for the end of the license renewal period. In the July 28, 1998, SER, the staff used the mean [reference temperature of nil-ductility transition] RTNDT value for materials to evaluate failure probability of BWR circumferential welds at 32 and 64 EFPY. The neutron fluence at the clad-weld (inner) interface was used for this evaluation. Since the staff analysis discussed in the BWRVIP-74 report is generic, the applicant submitted plant-specific information to demonstrate that the MNGP RPV beltline materials meet the criteria specified in the report. To demonstrate that the MNGP RPV has not become embrittled beyond the basis for the relief, the applicant, in LRA Table 4.2.6.1, compared 54 EFPY material data for the limiting MNGP circumferential weld with that of the 64 EFPY reference case in Appendix E to the staffs SER on the BWRVIP-05 report. Table 4.2.6-1 on the following page, taken from Subsection 4.2.6.2, "Staff Evaluation," of NUREG-1865, has been modified by addition of a fourth column which shows the effects of a 120% increase in thermal power(3)(4) from the original licensed thermal power (OLTP) on the RPV circumferential weld properties at the end of the 20-year renewed operating license period.

3. The current licensed reactor thermal power (CLTP) is 1775 MWt. The maximum projected power level of 2004 MWt is 120% of the OLTP of 1670 MWt.

4. A power increase request (Reference 13) is under review, but on indefinite hold, pending staff resolution to several industry issues. The values in the fourth column correspond to those presented in Table 2.1-2 in Enclosure 5 of Reference 13.

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Table 4.2.6-1 Effects of Irradiation on RPV Circumferential Weld Properties for MNGP Chicago Bridge MNGP 54 MNGP Value and Iron (CB&I) EFPY 54 EFPY 64 EFPY (120% OLTP) Cu(%) 0.10 0.10 0.10 Ni %) 0.99 0.99 0.99 CF [Chemistry Factor] 134.9 138.5 138.5 Fluence x 10"9 (n/cm2 ) [at 1.02 0.52 0.64 clad/weld interface] ARTNDT (OF) 135.6 113 121 RTNDT (OF) -65 - 65.6 - 65.6 Mean RTNDT (*F) 70.6 47.4 55.8 Probability of a failure 1.78 x 10-5 (1) (1) event (NRC) I I Notes: 1: Ifthe plant-specific mean ARTNDT is less than the mean ARTNDT associated with the limiting case study, the staff concludes that the probability of failure for the plant-specific circumferential weld under review will be less than the conditional probability of failure value for the limiting circumferential weld in the limiting case study. Analysis indicates that assuming 120% of OLTP through the renewed operating license term, that the Fluence, ARTNDT and Mean RTNDT all increase but that the results (see fourth column) are still well within the CB&l 64 EFPY NRC staff acceptance criteria (second column above). The Monticello Updated Safety Analysis Report [USAR], Appendix K, Renewed Operating License - USAR Supplement, provides background and summarizes the bases and results of the license renewal analyses and evaluations. The USAR section on RPV circumferential weld examination relief discusses and compares the MNGP limiting weld parameters to those used in the NRC analysis. Under the Disposition section it states: For MNGP, the chemistry values are the same as those used in the NRC analysis, however, the chemistry factor is higher due to an adjustment to reflect the results from two surveillance capsules. The value of fluence is lower than that used in the NRC analysis. As a result, the shift in reference temperature is lower than the 64 EFPY shift from the NRC analysis. In addition, the unirradiated reference temperature is essentially the same. The combination of unirradiated reference temperature (RTNDT(U)) and shift (ARTNDTW/O margin) yields an Adjusted Reference Temperature (ART) that is lower than the NRC mean analysis value. 1.5-205

Therefore, the RPV shell weld embrittlement due to fluence has a negligible effect on the probabilities of RPV shell weld failure. The Mean RTNDT value at 54 EFPY is bounded by the 64 EFPY Mean RTNDT provided by the NRC. Although a conditional failure probability has not been calculated, the fact that the MNGP values at the end of license are less than the 64 EFPY value provided by the NRC leads to the conclusion that the MNGP RPV conditional failure probability is bounded by the NRC analysis. Based on analysis assuming 120% of the OLTP through the renewed license term, the Mean RTNDT increases to 55.8°F, but remains bounded by the 64 EFPY Mean RTNDT staff acceptance criteria of 70.6 0 F for CB&I vessels. The fact that this value at the end of the renewed operating license term for 120% of OLTP conditions is less than the 64 EFPY staff acceptance criteria demonstrates that the MNGP conditional failure probability remains bounded by the NRC analysis. This conclusion is supported by the following discussion from Section 4.2.6.2, "Staff Evaluation," of NUREG-1 865 which discusses the effects of irradiation on RPV circumferential shell weld properties for the MNGP for the renewed operating license term. The NUREG states: The MNGP material data included amounts of copper and nickel, chemistry factor, the neutron fluence, ARTNDT, initial RTNDT, and mean RTNDT of the limiting circumferential weld at the end of the renewal period. The staff has verified the data for the copper and nickel contents and the initial RTNDT values for the MNGP circumferential beltline weld material by comparing them with the corresponding data in RVID [Reactor Vessel Integrity Database]. The 54 EFPY mean RTNDT value for the MNGP circumferential beltline weld is 47.4°F. The staff checked the applicant's calculations for the 54 EFPY mean RTNDT values for the limiting MNGP circumferential welds using the data presented in LRA Table 4.2.6.1 and found them to be accurate. This 54 EFPY mean RTNDT value for MNGP is bounded by the 64 EFPY mean RTNDT value of 70.6 0 F used by the NRC to determine conditional failure probability of a circumferential weld in a Chicago Bridge and Iron (CB&I) fabricated RPV. The 64 EFPY mean RTNDT value from the staff SER dated July 28, 1998, is for a CB&I weld because CB&I welded the circumferential welds in the [MNGP] RPV. Because the 54 EFPY mean RTNDT value is less than the 64 EFPY value from the staff SER dated July 28, 1998, the staff concluded that the NRC analysis bounds the MNGP RPV conditional failure probability. Since the 54 EFPY analysis results, assuming 120% of OLTP, increase for the Fluence, ARTNDT and Mean RTNDT at the end of the renewed operating license term but are still below the 64 EFPY acceptance criteria specified in the NRC 1.5-206

staff SER dated July 28, 1998, it is concluded that the NRC analysis bounds the MNGP RPV conditional failure probability for these parameters at the end of the renewed operating license term. Implement Operator Traininq and Establish Procedures that Limit the Frequency of Cold Over-Pressure Events to the Amount Specified in the NRC Staff Safe Evaluation for BWRVIP-05 Section 4.2.6.2, "Staff Evaluation, of NUREG-1865 also indicates that to be acceptable the proposed alternative has to include "implementation of operator training and established procedures that limit the frequency of cold overpressure events to the frequency specified in the staffs SER." The NSPM committed to, and revised and upgraded operator training and plant procedures (References 8 and 9) to minimize the frequency for potential cold overpressure events (consistent with the NRC specified frequency) in conjunction with receiving the current relief (Reference 1) from performing RPV circumferential shell weld examinations for the duration of the full-term operating license. Going forward, NSPM proposes to continue these commitments to limit the potential for cold overpressure events for the renewed operating license term. Section 4.2.6.2, "Staff Evaluation," of NUREG-1865 states: The applicant stated that the procedures and training used to limit cold overpressure events will be the same as those approved by the NRC when MNGP requested relief for the current license period. A request for relief during the period of extended operation will be submitted to the NRC before the period of extended operation. Submittal of this 10 CFR 50.55a request satisfies the following commitment, referred to as Item No. 5 in Appendix A to NUREG-1 865, and serves to enforce NSPM's ongoing commitment to implement and maintain operator training and procedural content to preclude cold overpressure events, as prescribed in our present authorized relief for the full-term operating license. The procedures and training used to limit RPV cold overpressure events will be the same as those approved by the NRC when MNGP requested approval of the BWRVIP-05 technical alternative for the term of the current operating license. A request for extension for the 60-year extended operating period will be submitted to the NRC before the period of extended operation. 1.5-207

NSPM has reviewed the above conclusions and has confirmed they are valid for the renewed operating license term of operation. Therefore, the proposed alternative as discussed herein, and as previously evaluated in the NRC safety evaluation for the MNGP for the full-term operating license (Reference 1), provides an acceptable level of quality and safety for the term of the renewed operating license. 6.0 Duration of Proposed Alternative The proposed alternative will be applied for the 20-year term of the renewed operating license. 7.0 Precedent The NRC has authorized similar requests to adopt an alternative to the ASME Section Xl, Table IWB-2500-1, Examination Category B-A, Item. No. B 1.11 criteria for permanent relief from the volumetric examination of RPV circumferential shell welds for the Dresden Nuclear Power Station, Units 2 and 3 and the Quad Cities Nuclear Power Station, Units 1 and 2 (Reference 10), the Nine Mile Point Nuclear Station, Unit No. 2 (Reference 11), and the Oyster Creek Nuclear Generating Station (Reference 12). 1.5-208

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8.0 REFERENCES

1. NRC letter, "Monticello Nuclear Generating Plant - Evaluation of Relief Request Number 12 for the Third 10-Year Interval Inservice Inspection Program (TAC No. MB0261)," dated July 27, 2001.

2. Electric Power Research Institute (EPRI) Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld inspection Recommendations (BWRVIP-05)," dated September 1995.

3. BWRVIP-74-A Report, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal," dated June 2003.

4. NRC letter, "Issuance of Renewed Facility Operating License No. DPR-22 for Monticello Nuclear Generating Plant," dated November 8, 2006.

5. NUREG-1 865, "Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant."

6. NRC 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," dated November 10, 1998.

7. NRC letter, 'Acceptance for Referencing of EPRI Proprietary Report TR-1 13596, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)," and Appendix A, "Demonstration of Compliance with the Technical Information Requirements of the License Renewal Rule (10 CFR 54.21)."' dated October 18, 2001.

8. NMC letter, "Request for Relief No. 12 for the Third 10-Year Interval Inservice Inspection Program," dated October 10, 2000.

9. NMC letter, "Response to NRC Request for Additional Information for Request for Relief No. 12 for the Third 10-Year Interval Inservice Inspection Program,"

dated May 3, 2001.

10. NRC letter, "Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2 - Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential Shell Weld Examinations (TAC Nos.

1.5-209

MC2190, MC2191, MC2192 and MC2193)," dated March 23, 2005. (ADAMS Package Accession No. ML050620359)

11. NRC letter, "Nine Mile Point Nuclear Station, Unit No. 2 - Authorization Under 10 CFR 50.55a(a)(3)(i) for Proposed Alternative Reactor Pressure Vessel Circumferential Shell Weld Volumetric Examinations (TAC No. MD3696)," dated November 5, 2007. (ADAMS Accession No. ML072830047)

12. NRC letter, "Oyster Creek Nuclear Generating Station - Relief Request for Alternative Examination for Reactor Pressure Vessel Circumferential Shell Welds (TAC No. ME0890)," dated September 15, 2009. (ADAMS Accession No.

MIL092520039)

13. NSPM letter, "License Amendment Request: Extended Power Uprate (TAC MD9990)," letter number L-MT-08-052, dated November 5, 2008.

1.5-210

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 18 PROPOSED ALTERNATIVE INACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY ALTERNATIVE TO APPLY ASME CODE CASE N-705 TO THE STANDBY LIQUID CONTROL SYSTEM TANK 1.5-211

10 CFR 50.55a Request No. 18 Proposed Alternative in Accordance With 10 CFR 50.55a(a)(3)(i) Which Provides an Acceptable Level of Quality or Safety Alternative to Apply ASME Code Case N-705 to the Standby Liquid Control System Tank 1.0 ASME Code Component(s) Affected Code Class: 2 Component Numbers: T-200 Examination Category: C-A

Description:

Alternative to ASME Section Xl, Section IWC-3120 for the Standby Liquid Control (SLC) System Tank 2.0 Applicable ASME Code Edition and Addenda The Monticello Nuclear Generating Plant (MNGP) is currently in the fourth 10-year Inservice Inspection (ISI) Program interval and is committed to the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," (ASME Section XI), 1995 Edition through 1996 Addenda. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition through 1996 Addenda is implemented, as required and modified by 10 CFR 50.55a.

3.0 Applicable Code Requirement

Section IWC-3120, "Inservice Volumetric and Surface Exams"

4.0 Reason for Request

While performing a periodic system walkdown, the system engineer identified sodium pentaborate crystallization at the bottom of the SLC Tank on the exposed lip of the tank base. Subsequent investigation identified degradation, i.e., indications of radial cracking in the SLC Tank bottom plate. There was also visible evidence of minute leakage at the tank base. The alternative proposed is to apply the evaluation criteria included in ASME Code Case N-705, "Evaluation 1.5-212

Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks, Section Xl, Division 1." The proposed alternative provides an acceptable level of quality and safety. 5.0 Proposed Alternative and Basis for Use In accordance with 10 CFR 50.55a(a)(3)(i), relief is requested by the Northern States Power Company - Minnesota (NSPM) from ASME Section Xl, IWC-3120, "Inservice Volumetric and Surface Exams," to implement an alternative that allows acceptance of degradation in the SLC Tank, an ASME Class 2 atmospheric pressure component, in accordance with the guidance of ASME Code Case N-705. ASME Code Case N-705 indicates that alternatives to the requirements of IWC-3120, as specified therein, may be used to accept degradation, including through-wall degradation, in a moderate energy ASME Code Class 2 tank for a limited time not to exceed the evaluation period, determined as defined in the code case. NSPM requests relief from IWC-3120, to allow the use of an alternate methodology to evaluate and accept through wall flaws in moderate energy Class 2 tanks. This alternative allows characterization of flaws in accordance with the applicable sections of ASME Code Case N-705, e.g., Section 2.2, "Degradation Characterization," and Section 2.4, "Bounding Flaw Evaluation," to estimate the degradation in inaccessible or uninspectable region(s) of the SLC Tank. Section 6, "Subsequent Examinations and Surveillance," of the code case requires: 1) daily monitoring for tank leakage, and 2) examination of the degradation to verify the predicted growth at one-half of the allowed operating time (unless the time at which the degradation reaches the allowable flaw size is determined to be twice the time to reach the end of the evaluation period). Monitoring of SLC Tank leakage each day in accordance with the operator rounds is proposed by NSPM to meet the first code case requirement (see Section 7.0 herein). NSPM has determined, based upon a flaw evaluation performed by Structural Integrity Associates, Inc. (see Enclosure 2), that the point in time where the limiting degradation reaches the allowable flaw size is far greater than twice the duration of the evaluation period,(1 )26 months, and hence examination before the March 2011 refueling outage is unnecessary. Therefore, this alternate methodology, as permitted in ASME Code Case N-705, would allow for a planned repair of the SLC Tank during the upcoming refueling outage in 2011, vice an unnecessary emergent shutdown and repair.

1. The evaluation period is the time to the next refueling outage - thirteen months.

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NRC draft Regulatory Guide DG-1192, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1," dated June 2009, identifies Code Case N-705 as one of the cases determined as an acceptable alternative to Section XI, for unconditional usage as indicated by its inclusion in Table 1, "Acceptable Section Xl Code Cases," to this draft regulatory guide. A proposed rule revising 10 CFR 50.55a(2) to include the next revision of Regulatory Guide 1.147, i.e., Revision 16, which includes this code case, has been published in the Federal Register, and is awaiting publication as a Final Rule. Use of ASME Code Case N-705 provides an acceptable level of quality and safety in accordance with 10 CFR 50.55a(a)(3)(i) for the evaluation and temporary acceptance of the degradation of the SLC Tank. 6.0 Duration of Proposed Alternative NSPM is requesting authorization for use of this alternative through the next refueling outage, currently scheduled to begin in March 2011. 7.0 Interim Actions Compliance with ASME Code Case N-705, Section 6, "Subsequent Examinations and Surveillance," requires the following action to be implemented.

• NSPM will restore the Standby Liquid Control Tank in accordance with Section XI of the ASME Code by startup from the next refueling outage, currently scheduled to begin in March 2011.

        " Monitoring of Standby Liquid Control Tank leakage each day will be performed in accordance with Procedure 0000-J, "Operations Daily Log -

Part J, Outplant" rounds until the tank is removed from service.

2. Federal Register: June 2, 2009 (Volume 74, Number 104, Proposed Rules),

    "Incorporation by Reference of Regulatory Guide 1.84, Revision 35, and Regulatory Guide 1.147, Revision 16, Into 10 CFR 50.55a," pages 26303-26310.

1.5-214

MONTICELLO NUCLEAR GENERATING PLANT 10 CFR 50.55a REQUEST NO. 18 PROPOSED ALTERNATIVE IN ACCORDANCE WITH 10 CFR 50.55a(a)(3)(i) WHICH PROVIDES AN ACCEPTABLE LEVEL OF QUALITY OR SAFETY ALTERNATIVE TO APPLY ASME CODE CASE N-705 TO THE STANDBY LIQUID CONTROL SYSTEM TANK STRUCTURAL INTEGRITY ASSOCIATES, INC. FLAW EVALUATION FOR THE MONTICELLO STANDBY LIQUID CONTROL TANK

SUMMARY

REPORT MARCH 24, 2010 1.5-215

StructuralIntegrity Associates, Inc. 5215 Hellyer Ave. Suite 210 San Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.structnt.com amiessi@structintcom March 24, 2010 Report No. 1000413.401.RO Quality Program: 0 Nuclear [] Commercial Mr. Jim Bridgeman Xcel Energy Monticello Nuclear Power Plant 2807 W COUNTY ROAD 75 Monticello, MN 55362-9601

Subject:

Flaw Evaluation for Monticello Standby Liquid Control Tank

References:

1. ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Code Case N-705, "Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks."

2. Xcel Energy, Monticello Nuclear Generating Plant, SBLC Tank T-200 Indication Worksheet.

Dear Jim:

This summary report serves as a technical basis to support continued operation of the Monticello Standby Liquid Control (SBLC) tank until the next refueling outage under the provisions of ASME Code Case N-705 [1]. The results summarized herein are based on verified scoping analyses which utilize many conservative assumptions. The conclusions of acceptability for continued operation are not expected to change with the final analyses. INTRODUCTION Leakage was detected at the SBLC tank during a walkdown. Subsequent inspections revealed that the leakage was emanating from a through-wall flaw in the base plate of the tank. The flaw extends radially across the base plate and appears to continue beneath the fillet weld on the outside of the cylindrical shell [2]. In addition, 16 other indications were found along the circumference of the base plate. However, only one flaw was completely through the thickness Annapolis, MD Austin, TX Centennial, CO Charlotte, NC Chattanooga, TN Oakville, Ontario, Canada South Jordan, UT Stonington, CT Uniontown, OH 410-571-0861 512-533-9191 303-792-0077 704-597-5554 423-553-1180 905-829-9817 801-676-0216 860-536-3982 330-899-9753 1.5-216

Mr. Jim Bridgeman / Xcel Energy March 24, 2010 Report No. 1000413.401.RO Page 2 of 3 Flaw Evaluation for Monticello Standby Liquid Control Tank of the base plate. It should also be noted that only five of the 17 indications span the width across the external face of the base plate. No indication was observed in the cylindrical shell. Since the through-wall flaw at the location of the leakage extends past the attachment fillet weld, it was assumed that the flaw could be both in the base plate and also in the cylindrical shell attachment. Thus, in addition to evaluating the flaw in the base plate, a postulated circumferential flaw in the cylindrical shell is evaluated also. TECHNICAL APPROACH The evaluation was performed in accordance with the structural requirements of ASME Section XI Code Case N-705 which has been approved by ASME but currently not approved in regulatory Guide 1.147. Allowable through-wall flaw sizes are calculated for the observed indications in the base plate and a postulated circumferential flaw in the cylindrical shell of the tank. In addition, a stress corrosion crack growth evaluation is performed to ensure that the flaws would not reach the allowable flaw size within the next operating period. RESULTS Allowable Flaw Evaluation The bounding indication in the base plate is modeled as a through-wall flaw extending radially from the OD of the circular plate towards the center of the tank. The hoop stresses in the tank shell are conservatively applied to the base plate. Using the provisions of the structural requirements in Code Case N-705, the allowable flaw was calculated to be 85 inches. For the tank shell, the evaluation was performed assuming a circumferential planar through-flaw in the cylindrical shell. Using the provisions of the structural requirements in Code case N-705, the allowable circumferential flaw in the tank shell was calculated to be 5.6 inches. Stress Corrosion Crack Growth Analysis A flaw growth evaluation was also performed assuming transgranular stress corrosion cracking per the guidelines of Code Case N-705. An initial 2.7 inches long flaw (approximately twice the length from the OD of the base plate to the toe of the fillet weld inside the tank) is assumed for the crack growth analysis. This initial crack size assumption is based on postulating a leak path from the ID of the vessel through the two fillet welds and the shell wall. It was determined that it will take at least 40 years for this initial flaw to reach 3.1 inches. Based on the small crack growth, it is expected that all the indications are well below the allowable flaw sizes. Similarly, the stress corrosion crack growth performed for the postulated circumferential through-wall flaw in the tank cylinder shows that it will take at least 18 years for a 0.35 inches long through-wall flaw to reach the allowable flaw size. No indications were observed in the vertical cylindrical shell during the current inspection. This does not imply that an indication does not exist in the vertical cylindrical shell. 1.5-217 V StructuralIntegrityAssociates, Inc.

Mr. Jim Bridgeman / Xcel Energy March24, 2010 Report No. 1000413.401.RO Page 3 of 3 Flaw Evaluation for Monticello Standby Liquid Control Tank CONCLUSIONS Based on this evaluation, it is concluded that the observed indications in the base plate of the Monticello Standby Liquid Control tank are acceptable for operation until the next scheduled outage which is approximately one year away. A postulated circumferential through-wall flaw in the cylindrical was also found to meet the acceptance criteria in Code Case N-705 for continued operation until the next scheduled outage. Note that, per the requirements of Code Case N-705, daily monitoring is required to ensure that the leakage is within acceptable limits. The removal of a small boat sample at the edge of the base plate, away from the fillet weld of the cylindrical shell, for metallurgical examination will not impact the conclusions of this evaluation. Please contact us if you have any questions. Thank you. Preparedby: Reviewed by: Approved by: G. A. Miessi S. S. Tang, P.E. G. A. Miessi Associate Associate Associate 1.5-218 V StructuralIntegrity Associates, Inc.

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY ASME Code Component(s) Affected Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Class 1, Reactor Vessel Nozzle-to-Vessel welds specified below and in-detail in Table A of Enclosure 2: Recirculation Inlet Nozzle N-2A Weld - N-2A NV Main Steam Outlet Nozzle N-3C Weld - N-3C NV Feedwater Inlet Nozzle N-4B Weld - N-4B NV Reactor Head Spare Nozzle N-6B Weld - N-6B NV Reactor Head Vent Nozzle N-7 Weld - N-7 NV Jet Pump Instrumentation Nozzle N-8B Weld - N-8B NV Standby Liquid Control Inlet / Nozzle N-10 Weld - N-10 NV Core Differential Pressure

2. Applicable ASME Section Xl Code Edition and Addenda The applicable ASME Section Xl Code for the Monticello Nuclear Generating Plant (MNGP), Fourth Ten-Year Inservice Inspection (ISI) Interval is the 1995 Edition with the 1996 Addenda. ASME Section XI, Appendix VIII requirements are implemented as required by, and as modified by, 10 CFR 50.55a. Procedures and personnel are qualified to the Performance Demonstration Initiative (PDI). The PDI Program document meets the requirements of 10 CFR 50.55a up through the 2001 Edition of Section XI.

3. Applicable Code Requirement

ASME Class I Nozzle-to-Vessel welds are subject to the examination requirements of Subsection IWB Table IWB-2500-1, as shown below, and 10 CFR 50.55a(b)(2)(xv)(G). The welds are required to be examined once within the Fourth Ten-Year Interval: Code Class: 1

References:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.90

Description:

Nozzle-to-Vessel Welds Component Numbers: See Section 1 and Enclosure 2 Table A System: Reactor Vessel Examination Method: Volumetric - Ultrasonic Testing (UT) Examination Volume: Figure IWB-2500-7(b) 1.5-219

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY In October 2007, the Nuclear Regulatory Commission (NRC) issued Regulatory Guide (RG) 1.147, Revision 15, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1" (Reference 1). In RG 1.147, the NRC identifies the ASME Code Cases they have determined to be acceptable alternatives to applicable parts of Section Xl, and indicate that licensees may use these Code Cases without requesting authorization from the NRC, provided that they are used with any identified limitations or modifications. RG 1.147, Table 1 lists the following two Code Cases as acceptable to the NRC for use by a licensee with no identified limitations or modifications:

1) Code Case N-460 (Reference 2),

2) Code Case N-613-1 (Reference 3).

Code Case N-460 states in part, "when the entire examination volume or area cannot be examined due to interference by another component or part geometry, a reduction in examination coverage on any Class 1 or Class 2 weld may be accepted provided the reduction in coverage for that weld is less than 10 percent." NRC Information Notice (IN) 98-42 (Reference 4) termed a reduction in coverage of less than 10 percent to be "essentially 100 percent." IN 98-42 states in part, "The NRC has adopted and further refined the definition of 'essentially 100 percent' to mean 'greater than 90 percent'...has been applied to all examinations of welds or other areas required by ASME Section XI." As an alternative to Figure IWB-2500-7(b), Code Case N-613-1 requires an examination volume that includes the width of the weld plus one-half inch of adjacent base metal on each side of the widest part of the weld. In comparison, the examination volume required by the Figure IWB-2500-7(b) includes the width of the weld plus the adjacent base metal on each side of the widest part of the weld equal to one-half of the vessel shell wall thickness.

4. Impracticality of Compliance Construction Permit CPPR-31 was obtained for the MNGP in 1967. The MNGP systems and components were designed and fabricated before the examination requirements of ASME Section Xl were formalized and published. Therefore, MNGP was not specifically designed to meet the requirements of ASME Section Xl and full compliance is not feasible or practical within the limits of the current plant design.

10 CFR 50.55a recognizes the limitations to inservice inspection of components in accordance with Section Xl of the ASME Code imposed due to early plants' design and construction, as follows: 1.5-220

                            °"] I ...... :.

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY 10 CFR 50.55a(g)(1): For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (5) of this section to the extent practical. 10 CFR 50.55a(g)(4): Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) which are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and pre-service examination requirements, set forth in Section Xl of editions of the ASME Boiler and Pressure Vessel Code ... to the extent practical within the limitations of design, geometry and materials of construction of the components. 10 CFR 50.55a(g)(5)(iii): If the licensee has determined that conformance with certain code requirements is impractical for its facility, the licensee shall notify the Commission and submit, as specified in § 50.4, information to support the determinations. The inspection limitations on the subject components are due to inherent nozzle design geometric contours and interference (see Enclosure 2 Table A). A description of the examination methodology used to provide the maximum obtainable coverage is provided in Section 6 of this request. This methodology is based on ASME Section XI, Appendix VIII qualification and was applied to the extent practical within the design constraints of the components. Enclosure 3 provides cross-sectional diagrams of the subject welds showing the geometric contour of the component design in relation to the welds and the coverage obtained within the alternative examination volume requirements of Code Case N-613-1, Figure 2.

5. Burden Caused by Compliance Compliance with the examination coverage requirements of ASME Section Xl would require modification, redesign, or replacement of components where geometry is inherent to the component design.

1.5-221

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY

6. Proposed Alternative and Basis for Use Proposed Alternative In accordance with 10 CFR 50.55a(g)(5)(iii), relief is requested for the components listed in Table A of Enclosure 2 on the basis that the required examination coverage of "essentially 100 percent" is impractical due to physical obstructions and the limitations imposed by design, geometry and materials of construction.

Northern States Power Company - Minnesota (NSPM) performed qualified volumetric examinations that achieved the maximum, practical amount of coverage obtainable within the limitations imposed by the design of the components. In addition to volumetric examinations, as Class 1 Examination Category B-P components, a VT-2 examination is performed on the subject components of the Reactor Coolant Pressure Boundary (RCPB) during system pressure tests each refueling outage. This was completed during the 2009 refueling outage and no evidence of leakage was identified for these components. Pursuant to 10 CFR 50.55a(g)(5)(iii), NSPM requests authorization of an alternative to the requirements of ASME Section XI Table IWB-2500-1, Category B-D, Item B3.90, and proposes to utilize these completed exams as acceptable alternatives that provide reasonable assurance of continued structural integrity. Basis for Use The NSPM Nondestructive Examination (NDE) procedures incorporate inspection techniques qualified under Appendix VIII of the ASME Section XI Code by the PDI for examination of the subject nozzle-to-vessel welds, and allow the examination volume to meet the provisions of alternative requirements (i.e., Code Case N-613-1). The examinations were performed from the Reactor Vessel exterior surface using a manual contact method from the nozzle blend radius, the nozzle-to-vessel shell weld, and vessel shell surface. Coverage was obtained by following the scan parameters designated within NSPM NDE procedures for each nozzle configuration and angle, including those parameters defined by MNGP specific Electric Power Research Institute (EPRI) computer modeling reports (References 5 and 6). It should be noted that that the scans defined by the EPRI report are only applicable to the inner 15 percent of the weld volume when scanning in the parallel (circumferential) direction. The refracted longitudinal wave mode of propagation was applied for all radial (axial) scans of the exam volume. The refracted longitudinal wave mode of propagation was also applied to the outer 85 percent of the exam volume for parallel scans. As required by the NSPM NDE procedures and the EPRI computer modeling report, the 1.5-222

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY shear wave mode of propagation was applied for each of the transducer and wedge combinations required for the remaining inner 15 percent of the parallel scan exam volume. The subject components received the required examination(s) to the extent practical within the limited access of the component design. One hundred percent coverage was obtained for the inner 15 percent of the examination volume for the radial and parallel scans. The examination limitations for the subject components were encountered within the outer 85 percent of the examination volume for the parallel and radial scans. For the examinations conducted, satisfactory results were achieved, and no evidence of unacceptable flaws was detected with the inspection techniques. Due to the design of these welds it was not feasible to effectively perform a volumetric examination of "essentially 100 percent" of the required volume. The nozzle-to-vessel welds are accessible from the vessel plate side of the weld and are examined to the extent practical with qualified techniques, but the curvature of the nozzle forging and proximity to the weld precludes obtaining further coverage of the excluded areas within the outer 85 percent of the examination volume. As required by site procedure, when limitations are encountered that prevent obtaining full coverage of a required volume while performing ISI examinations, the limitations are required to be quantified and recorded. The method used to determine coverage is based on field measurements applied to a two dimensional plot. This allows an informed approximation to be made of the coverage achieved. The methodology is appropriate to the application in that the limitations are physical and the methods applied to the examination are established by the qualified techniques. Variations in the percent coverage obtained from the previous examinations are the result of changes in examination technique andlor required coverage. The current coverage determinations are different from past examinations due to the use of PDI qualified techniques and a reduced exam volume required by use of Code Case N-613-1. Per 10 CFR 50.55a(g)(1) and (4), each of the subject welds 1 were examined to the extent practical during the First, Second, and Third Ten-Year ISI Intervals. Prior to 1997, NSPM did not perform examination coverage determinations or submit relief requests pursuant to 10 CFR 50.55a for limited examinations due to a misinterpretation of 10 CFR 50.55a(g)(4). It was construed that the interferences inherent in the design constituted impracticality and were exempted. As an exception, the N-10 weld was not examined in the First Interval based on approved ISI Relief Request #15 (Reference 14) 1.5-223

                                                              *".1 '" "

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY This misinterpretation was identified during the Third Ten-Year ISI Interval and was reported to the NRC in Licensee Event Report (LER) 97-004 (Reference 13). Going forward from 1998, relief requests for limited exams have been submitted to the NRC. There is no means for retroactively complying with 10 CFR 50.55a for prior Intervals. These issues were documented in the MNGP corrective action program and various corrective actions were taken to prevent recurrence, including submittal of relief requests after completion of every refueling outage when examination coverage is limited and Code examination requirements cannot be met. Details regarding the aforementioned limited exam relief request issues were included with supplemental information submitted to the NRC in March 2008. The supplemental information was provided to assist the NRC with review of MNGP 4th Interval Relief Request #15 (Reference 19) which was subsequently approved in May 2008 (Reference 9). The coverage drawings in Enclosure 3 give a representation of the examination volume and the weld interface line in the same manner as the figure included in Code Case N-613-1. The areas of examination volume coverage are identified by the lightly shaded or cross-hatched areas on the drawings. The remaining areas of the examination volume, with black shading or with no shading or no cross-hatching, represent areas with no coverage. On page 16 of Enclosure 3, a sketch of a typical nozzle is provided with a cross sectional view of the weldment depicting the curvature of the nozzle exterior surface and its effect on transducer liftoff. Although there is some variation, most of the limited coverage is in the nozzle base material with a lesser amount in the weld and base material on the vessel shell side. Additional coverage for the limited areas was not achievable or practical, based on the latest qualified ultrasonic technology, nor by other considered examination methods, such as radiography. NSPM has concluded that if significant degradation existed in the subject welds, itwould have been identified by the examinations performed. A table of examination history and results is provided for prior ISI Intervals in Enclosure 4, including available coverage information and Relief Requests. Additionally, as Class 1 Examination Category B-P components, VT-2 examinations were performed on the subject components in association with the RCPB system pressure test performed during the 2009 refueling outage. No evidence of leakage was identified during this system test. 1.5-224

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY The materials for the subject components are A508 CI I1nozzle forgings welded to A533 Cl I vessel shell plate. The weld filler material for the subject joints was E8018NM. Inner diameter cladding materials are E309-15 for the base layer, and ER308L or E308L-1 5 for subsequent layers. A review of operating experience within the nuclear industry did not reveal any instances of cracking in this location and type of weldment, specifically nozzle-to-vessel shell welds. The MNGP reactor vessel water chemistry is controlled in accordance with the 2008 revision to the BWR Water Chemistry Guidelines (Reference 7). Also a hydrogen water chemistry system is used to reduce the oxidizing environment in the reactor coolant. These additional measures provide added assurance against the initiation of cracking or corrosion from the inside surface of the reactor vessel. An inerted primary containment environment during operation provides assurance of corrosion protection on the outside surface of the reactor vessel. The provisions described above, as an alternative to the code requirement, will continue to provide reasonable assurance of the structural integrity of the subject welds. The examinations were completed to the extent practical and no unacceptable flaws were identified. VT-2 examinations performed on the subject components during system pressure testing each refueling outage (in accordance with Examination Category B-P) provide continued assurance that the structural integrity of the subject components is maintained. Additionally, the MNGP Water Chemistry Program and inerted primary containment environment provide added measures of protection for the component materials. Therefore, pursuant to 10 CFR 50.55a(g)(5)(iii), NSPM requests that the NRC grant relief from the ASME Section XI examination requirements for the subject nozzle-to-vessel welds.

7. Duration of Proposed Alternative NSPM requests the granting of this relief for the Fourth Ten-Year Inservice Inspection Interval of the Inservice Inspection Program for the MNGP that is scheduled to end on May 31, 2012.

1.5-225

i . . 10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY

8. Precedent The NRC has granted relief for other nozzle-to-vessel shell welds at the MNGP, most recently for the current Fourth Ten-Year Inservice Inspection Interval (References 8 and 9). Also, the NRC has granted relief for the Quad Cities Nuclear Power Station, Units 1 and 2 (Reference 10), Dresden Nuclear Power Station, Units 2 and 3 (Reference 11), and Prairie Island Nuclear Generating Plant, Unit 2 (Reference 12).

1.5-226

                                                               .i1 10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii)

INSERVICE INSPECTION IMPRACTICALITY REFERENCES

1. Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1," Revision 15, October 2007.

2. ASME Section Xl Code Case N-460, "Alternative Examination Coverage for Class I and Class 2 Welds."

3. ASME Section XI Code Case N-613-1, "Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.10 and B3.90, Reactor Nozzle-To-Vessel Welds, Figures lWB-2500-7(a), (b), and (c)."

4. NRC Information Notice 98-42, "Implementation of 10 CFR 50.55a(g) In-service Inspection Requirements."

5. EPRI Internal Report IR-2004-63, "Monticello Nozzle Inner Radius and Nozzle-to-Shell Weld Examinations," dated December 2004.

6. EPRI Internal Report IR-2006-100, "Monticello Nozzle Inner Corner Regions and Nozzle-to-Shell Weld Examinations," dated January 2006.

7. "BWRVIP-1 90: BWR Vessel and Internals Project, BWR Water Chemistry Guidelines -

2008 Revision," EPRI Technical Report, TR-1016579, October 2008.

8. NRC letter to NMC, "Monticello Nuclear Generating Plant (MNGP) - Fourth 10-Year Interval Inservice Inspection (ISI) Program Plan Relief Request No. 13 (TAC No.

MC8882)," dated July 18, 2006.

9. NRC letter to NMC, "Monticello Nuclear Generating Plant (MNGP) - Granting of Relief Regarding Limited Ultrasonic Examination Coverage of Five Welds (TAC No.

MD6854)," dated May 19, 2008.

10. Letter from NRC to Exelon Generation Company, LLC, "Quad Cities, Units 1 and 2 -

Relief Request CR-39 for Third 10-Year Inservice Inspection Interval (TAC Nos. MC2427 and MC2428)," dated May 10, 2005.

11. Letter from NRC to Exelon Generation Company, LLC, "Dresden Nuclear Power Station, Units 2 and 3 - Relief Request CR-26 For Third 10-Year Inservice Inspection Interval (TAC Nos. MC3269 and MC3270)," dated October 1, 2004.

1.5-227

10 CFR 50.55a REQUEST NO. 19 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY

12. NRC letter to NMC, "Prairie Island Nuclear Generating Plant, Unit 2 - Evaluation of Relief Request No. 16 for the Unit 2 3rd 10-year Interval Inservice Inspection Program (TAC No. MC1775)," dated October 18, 2004.

13. LER 97-004, "Failure to Submit Relief Requests for Limited Inservice Inspection Examinations," dated March 24, 1997.

14. NRC letter to Northern States Power Company, "Safety Evaluation Report, Monticello Nuclear Generating Plant Inservice Inspection Program," dated April 10, 1981.

15. MNGP Corrective Action Program Action Request (CAP A/R) 01126631, "Available documentation doesnt [sic] support closed M97024A action", origination date February 8, 2008.

16. MNGP Letter to NRC, "Request for Relief No. 11 for the 3rd 10-Year Interval Inservice Inspection Program," dated May 25, 2000.

17. MNGP Letter to NRC, "Supplemental Information Request for Relief No. 11 for the 3rd 10-Year Interval Inservice Inspection Program," dated July 11, 2000.

18. MNGP Corrective Action Program Action Request (CAP A/R) 01013875, '6 limited ISI exams not included in Cycle 19 Relief Request", origination date February 7, 2006.

19. MNGP Letter to NRC, "Response to Request for Additional Information Regarding 10 CFR 50.55a Request No. 15 (RR-15): Relief from Impractical Examination Coverage Requirements Pursuant to 10 CFR 50.55a(g)(5)(iii) for the Fourth Ten-Year Inservice Inspection Interval (TAC No. MD6854)," dated March 21, 2008.

1.5-228

10 CFR 50.55a REQUEST NO. 19 INACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY TABLE A - Category B-D, "Full Penetration Welds of Nozzles in Vessels," Item No. B3.90 2009 Refueling Outage, Percent Coverage and Limitations for Nozzles N-2A, N-3C, N-4B, N-6B, N-7, N-8B, and N-10 Code Code Component Category System and Percent Exam and and Component ID' Examination Volume Coverage Report Item No. Component Description Required Obtained Limitations Number Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to B-D Recirculation Inlet N-2A NV Code Case N-613-1 83% e uto 2009UT033 83.90 Nozzle N-2A Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to B-D Main Steam Outlet N-3C NV3 Code Case N-613-1 83% e uto 2009UT024 B3.90 Nozzle N-3C Figurenozzle configuration.2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to (o N0 B-D Feedwater Inlet 83.9

                                       -4Bnozzle          N-4B NV Nozle             Code Case N-613-1             83%      Limite configuration.

dueato 2009UT026 B3.90 Nozzle N-413 Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to B-D Top Head Spare N-68 NV Code Case N-613-1 87% nozzle configuration. 2009UT021 83.90 Nozzle N-6B Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to B-D Top Head Vent N-7 NV Code Case N-613-1 87% nozzle configuration. 2009UT023 83.90 Nozzle N-7 Figure 2 Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to B-D Jet Pump Instrumentation N-8B NV Code Case N-613-1 83% nozzle configuration. 2009UT028 83.90 Nozzle N-8B Figure 2 B-D Reactor Vessel, Nozzle-to-Vessel Weld, Limited due to Standby Liquid Inlet / Core N-10 NV Code Case N-613-1 85% nozzle configuration 2009UT030 B3.90 Diff. Pressure Nozzle N-10 i s Nl N Figure 2 and closeskirt to vessel proximity weld. 1 With exception of component N-3C NV, no indications were reported for the component's examination. 2 Due to the nozzle design it was not feasible to effectively examine essentially 100 percent of the required examination volume as defined in Figure 2 of Code Case N-613-1. Percentages are conservatively rounded down to the nearest whole number. Previously observed subsurface indication was re-confirmed and re-evaluated as acceptable per Code paragraph IWB-3512-1. No observed change since previous exam in 1998.

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY EXAM LIMITATIONS IMPOSED BY COMPONENT DESIGN AND CONSTRUCTION This enclosure contains a series of excerpts from the ISI Ultrasonic Testing (UT) reports applicable to the subject components. These excerpts contain sketches depicting the component configuration with physical limitations imposed by the design, e.g., geometrical contour, weld position, interferences, and a cross sectional view depicting the UT coverage and limitations in relation to the required examination volume. Also included is a sketch of a typical reactor vessel nozzle contour and the resulting effect that causes the UT transducer to lift and lose effective coupling when it reaches the nozzle blend radius. Detail is also provided to describe the various assembly components including reference to the internal and external reactor vessel surfaces. COMPONENT REPORT PAGE(S) N-2A NV 2009UT033 Pages 2-3 N-3C NV 2009UT024 Pages 4-5 N-4B NV 2009UT026 Pages 6-7 N-6B NV 2009UT021 Pages 8-9 N-7 NV 2009UT023 Pages 10-11 N-8B NV 2009UT028 Pages 12-13 N-10 NV 2009UT030 Pages 14-15 Typical Reactor Vessel Nozzle contour affecting Page 16 transducer contact at blend radius 1.5-230

Coverage drawings excerpted from applicable reports Component N-2A NV Axial (Radial) Scan Plot Report # 2009UT033

  .@7,X*Ene                           :Supplemental Report rn.: 102656                                                     Repma N m   -- IIIi                          -   II i   .. i1.               i ,*---

Monticello N2A Coverage Plot Axial scan direction A 60! deg-.

Component N-2A NV Parallel (Circ) Scan Plot Report # 2009UT033

   ~xcr                           Supplemental Report             Roqmrf ., 2__________

Simary N4~ 1O~S5~ Monticello N2A Coverage Plot Parallel scan direction N) No Inner

Component N-3C NV Axial (Radial) Scan Plot Report # 2009UT024

  *. XcelEnergy           Supp*i*aae*ta Roport lPAnIt NoM: MUTO-2Ut summty.~~

NIQ.!1=200 Monticello N3C Coverage Plot Axial scan direeftion 3' 60 deg.

Component N-3C NV Parallel (Circ) Scan Plot Report # 2009UT024 XcdEnergy Supplemental 1Report Report No,: 2I3OUTM SIummarl No,: 102580 Monticello N3C Coverage Plot Parallel scan direction E

Component N-4B NV Axial (Radial) Scan Plot Report # 2009UT026

  .*. ,x ef                         Supplemental Report R&Po31 NMI. 200AUT020 SuniMary go.:

Monticello N4B Coverage Plot

                      --Axial scan direction 3b No.

Component N-48 NV Parallel (Circ) Scan Plot Report # 2009UT026 Supplemental Report Report No.,: 200SUT6O 8 Monticello N4B Coverage Plot Parallel scan direction ro ci, I of covurage

Component N-6B NV Axial (Radial) Scan Plot Report # 2009UT021

   ,,X-OEnerg'y                      Supplemental Report RaIN M2 2o0UTrD21 Summary ,NO 1o2377 Monticello N6B Coverage Plot Axial scan direction

-4

                            "-.A      B     C      D MOV~gr.

Component N-6B NV Circ (Parallel) Scan Plot Report # 2009UT021 Supplemental. Report ReiOt W-. 'notIM21 SUMImmwyND: 192377 Monticello N6B Coverage Plot Circ scan direction 02 WrocOVM

Component N-7 NV Axial (Radial) Scan Plot Report # 2009UT023

   & xaowwog                   Supplemental Report, Report RD.- 2009UT023 Summary No.: 102379 co~

Monticello N7 Coverage Plot CD Axial scan direction

Component N-7 NV Circ (Parallel) Scan Plot Report # 2009UT023

• XceitE*w SupplerhehtWl Report Report No,~ 2009U102 Suammrt Mý: 102V9~

.0-1 r*o Monticello N7 Coverage Plot 0 Circ scan direction

Component N-8B NV Axial (Radial) Scan Plot Report # 2009UT028 Sopplemental Report Report No,: "OO9UTh28 Monticello NSB-Coverage Plot Axial scan direction .Li-i 3J Arp' of ovaem

60. eg.

Component N-8B NV Parallel (Circ) Scan Plot Report # 2009UT028

    ?9. ReIEerqyo                       Supplemental Report ReW ort I' 2U9th2e W

SuMMrnmv No,;: 102Mn R13.00 in, ontic ello NSB Coverage Plot Parallel scan direction

• 0L rba 4ýh B

No F

Component N-10 NV Axial (Radial) Scan Plot Report # 2009UT030

   &- xceel vrw ti-              Supplemental Report veuw- 200OBUTO-30 Summmi No~,:U1252-3 Monticello N-10 Nozzle Axial Scan Coverage

Component N-10 NV Circ (Parallel) Scan Plot Report # 2009UT030 j, XceIEaer Supplementar Report Report No., 20D0%UT030 Sunitary Naz 102S23 Monticello N-10 Nozzle Circ Scan Coverage °O 1 1 ITUM two

                          /

Typical Representation of Nozzle Limitations Vessel Nozzle Blend Radius Exam Volume: A-B-C-D-E-F-G-H (exte io0 r) Weld Represented By, B-C-F-G -l Scanning past this point onto the nomzle blend racius

 '1 causes Lift-off / [osn of contact Area of No Cover age-Vessel Nozzle                  H 4           F EF          Reactor Vessel Shell (interior surf'ace)                                                     surface)

I(Interior 4in. 1Ain.

                                                                                                        . 1i I  .

10 CFR 50.55a REQUEST NO. 19 IN ACCORDANCE WITH 10 CFR 50.55a(g)(5)(iii) INSERVICE INSPECTION IMPRACTICALITY Table 1 - Historical Examination Information Weld Interval I Exam Coverage I Results for Relief NRC Approval Year Limited Exam Request N-2A 1st I 1974 (Note 1), no flaw indications (Note 2) (Note 2) 2nd I 1982 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 2001 62% coverage, no flaw 3rd, ISI RR-16 TAC No. MB5487, indications May 19, 2003 N-3C 1st (1975 (Note 1), no flaw indications (Note 2) (Note 2) 2nd I 1989 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 1998 51% coverage, acceptable 3rd, ISI RR-10 TAC No. MB3397, mid-wall indication August 4, 1999 N-4B 1st I 1977 (Note 1), no flaw indications (Note 2) (Note 2) 2nd I 1987 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 1998 51% coverage, no flaw 3rd, 11 RR-10 TAC No. MB3397, indications August 4, 1999 N-6B 1st I 1981 (Note 1), no flaw indications (Note 2) (Note 2) 2nd 1991 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 2000 70% coverage, no flaw (Note 3) (Note 3) indications N-7 1st / 1973 (Note 1), no flaw indications (Note 2) (Note 2) 2nd 1984 (Note 1), no flaw indications (Note 2) (Note 2) 3rd 1 1998 89% coverage, no flaw 3rd, ISI RR-10 TAC No. MB3397, indications August 4, 1999 N-8B 1st /1981 (Note 1), no flaw indications (Note 2) (Note 2) 2nd /1991 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 2001 62% coverage, no flaw 3rd, IS] RR-16 TAC No. MB5487, indications May 19, 2003 N-10 1st / N/A N/A 1st, IS[ RR-15 Safety Evaluation Report, Apr 10, 1981 2nd I 1989 (Note 1), no flaw indications (Note 2) (Note 2) 3rd I 2000 55% coverage, no flaw (Note 3) (Note 3) indications Note 1: MNGP did not document Code coverage values for limited exams prior to 1997 (References 9 and 13) Note 2: With the exception of relief identified for the 1st Interval for Nozzle N-1 0 due to inaccessibility at the time (Reference 14), relief was not requested for limited exams on the subject welds prior to 1997 (Reference 9, 13, and 15) Note 3: Limited exam errantly omitted from 3rd Interval ISI Relief Request 11 submitted for the 2000 refueling outage (References 16 and 17), as documented in the MNGP Corrective Action Program (Reference 18) 1.5-246

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN QUALITY GROUP CLASSIFICATION DRAWINGS (ISI BOUNDARY DWGS) ISO # REV SYSTEM / DESCRIPTION BOUNDARY DRAWINGS 1.5-1 0 ISI Index Key 1.5-2 3 Main Steam System 1.5-3 2 Feedwater System 1.5-4 1 Reactor Circulation System 1.5-5 3 Core Spray System 1.5-6 4 Residual Heat Removal System Loop A 1.5-7 5 Residual Heat Removal System Loop B 1.5-8 3 High Pressure Coolant Injection System (Steam Side) 1.5-9 2 High Pressure Coolant Injection System (Water Side) 1.5-10 2 Reactor Core Isolation Cooling (Steam Side) 1.5-11 2 Reactor Core Isolation Cooling (Water Side) 1.5-12 2 Standby Liquid Control System 1.5-13 3 Primary Containment Atmospheric Control System 1.5-14 3 Emergency Diesel Generator Emergency Service Water 1.5-15 3 Emergency Diesel Generator Emergency Service Water 1.5-16 2 RHR Service Water 1.5-17 3 Hydraulic Control Unit 1.5-18 2 Control Rod Drive System (Scram Discharge Piping) 1.5-19 Compressed Air System 1.5-20 2 Demineralized Water System 1.5-21 2 Reactor Water Clean-up & Liquid Radwaste 1.5-22 Traversing In-core Probe System 1.5-23 Excess-Flow Check Valves 1.5-26 1 Primary Containment Sampling Systems 1.5-27 Reactor Vessel Instrumentation 1.6-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI ISOMETRIC DRAWINGS ISO # REV SYSTEM / DESCRIPTION CLASS 1 & 2 DRAWINGS ISI Fig. 0 RX Vessel Interior ISI Fig. 1 RX Vessel Top Head ISI Fig. 2 CRD Location RX Vessel ISI Fig. 3 RX Vessel Bottom Head ISI Fig. 4 Circ. & Long Reactor Vessel Welds ISI Fig. 5 RX Vessel Nozzles ISI Fig. 6 Reactor Vessel Bolting ISI 13142-17-A RHR A Suction ISI 13142-17-B HPCI Water IS1-13142-17-C RHR B ISI-13142-18-A RHR B ISI-13142-18-B RHR B Discharge IS1-13142-18-C RHR B Discharge IS1-13142-19-A HPCI Steam Side Discharge IS1-13142-19-B RCIC Steam Discharge ISI-13142-20-A ISI-13142-20-B Core Spray A Suction Core Spray B Suction 0 ISI-1 3142-26-A Core Spray B Discharge IS1-13142-26-B Core Spray B Discharge ISI-13142-26-C Core Spray B Discharge IS1-13142-26-D Core Spray B Discharge IS1-13142-29-A RX Bldg Cooling Water IS1-13142-31-A Core Spray A Discharge ISI-13142-31-B Core Spray A Discharge IS1-13142-31-C Core Spray A Discharge ISI-13142-31-D Core Spray A Discharge IS1-13142-33-A Main Steam A ISI-13142-34-A Main Steam B ISI-13142-35-A Main Steam C ISI-13142-36-A Main Steam D ISI-13142-37-A RHR A Discharge ISI-13142-37-B Containment Spray ISI-13142-37-C RHR A Discharge IS1-13142-37-D Containment Spray (RHR A) IS1-13142-37-E Containment Spray (RHR A) ISI-13142-40-A HPCI Water Side Discharge ISI-13142-40-B HPCI Water Side Discharge 1.6-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI ISOMETRIC DRAWINGS ISO # REV SYSTEM / DESCRIPTION CLASS 1 & 2 DRAWINGS (continued) ISI-13142-41-A 5 RCIC Water Suction ISI-13142-42-A 6 HPCI Steam Side ISI-13142-43-A 5 RCIC Steam Side ISI-13142-48-A 4 RHR Service Water ISI-13142-48-B 5 RHR Service Water ISI-13142-49-A 4 RHRA ISI-13142-51-A 5 RHR A ISI-13142-51-B 5 RHR B IS1-13142-51-C 2 RHR B ISI-13142-51-D 1 RHR B ISI-13142-52-A 5 Feedwater C & D ISI-13142-53-A 6 Feedwater A & B ISI-13142-62 5 Fuel Pool Emergency Cooling IS1-13142-67 5 Fuel Pool Emergency Cooling ISI-16 4 Jet Pump Instrument Nozzle ISI-19 5 RX Instrument Nozzles ISI-47 4 RCIC Pump ISI-48 5 RHR Pumps ISI-49 5 Core Spray Pump Supports ISI-73880-A 4 RWCU ISI-74209-1-A 4 Recirc. A Drain ISI-74210-1-A 5 Recirc. B Drain ISI-74215-A 6 Standby Liquid Control ISI-782-A 3 RX Head Vent ISI-782-A-A 4 RX Head Vent ISI-786-A 7 Main Steam Condensate Leakoff ISI-7905-32-A 4 RHR HX A ISI-7905-32-B 4 RHR HX B ISI-821-A 3 RX Bottom Head Drain ISI-8292-42-A 4 HPCI Pumps ISI-8292-48-A 0 HPCI Turbine ISI-93268-1 -A 5 CRD Scram Header A ISI-93268-1-B 3 CRD Scram Discharge Header ISI-93268-3-A 6 CRD Scram Header B ISI-94699-A 3 Primary Containment & Atmospheric Control ISI-94879-A 3 Spare Penetration X-47 ISI-94966-A 3 Primary Containment & Atmospheric Control ISI-94966-B 3 Containment Air Purge 1.6-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN ISI ISOMETRIC DRAWINGS ISO # REV SYSTEM / DESCRIPTION CLASS I & 2 DRAWINGS (continued) ISI-97003-A 5 RHR Return Loop A ISI-97003-B 6 RHR A ISI-97004-A 5 RHR Return Loop B ISI-97005-A 6 Recirc. Loop A ISI-97005-B 5 Recirc. Manifold A ISI-97005-C 6 Recirc. Pump A Supports ISI-97006-A 7 Recirc Loop B ISI-97006-B 5 Recirc. Manifold B ISI-97006-C 7 Recirc. Pump B Supports ISI-97007-A 5 RX Instrument Nozzle N-11 B ISI-97008-A 5 RX Instrument Nozzle N-1 1A ISI-97027-A 5 RHR Equalizer ISI-105531-A 3 Standby Gas Treatment & RX Plenum ISI-158074-A 4 Torus Hard Pipe Vent CLASS 3 DRAWINGS ND-ISI-100 2 RHR Service Water ND-ISI-101 2 RHR Service Water ND-ISI-102 1 RHR Service Water ND-ISI-103 2 RHR Service Water ND-ISI-104 1 RHR Service Water ND-ISI-105 1 RHR Service Water ND-ISI-106 2 RHR Service Water ND-ISI-107 2 RHR Service Water ND-ISI-108 2 RHR Service Water ND-ISI-109 2 RHR Service Water ND-ISI-110 1 RHR Service Water ND-ISI-111 3 RHR Service Water ND-ISI-123 2 RHR Service Water ND-ISI-141 Deleted, system eliminated ND-ISI-142 Deleted, system eliminated ND-ISI-144 Deleted, system eliminated ND-ISI-145 Deleted, system eliminated 1.6-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 T' INTERVAL EXAMINATION PLAN ISI ISOMETRIC DRAWINGS ISO # REV SYSTEM I DESCRIPTION CLASS MC DRAWINGS ISI-8291-76 0 Class MC Supports 1.5-81 1 Downcomer Restraints 1.5-82 1 Vent Line & Header Restraints NH-95932-A 1 Ring Header Seismic Restraints NH-95932-B 1 Ring Header Seismic Restraints NH-95932-C 1 Ring Header Seismic Restraints NH-95932-D 1 Ring Header Seismic Restraints NX-8291-34-A I Vent Line & Header NX-8291-34-C 0 Downcomer, Vent Line & Header Supports NON-CODE AUGMENTED NC-ISI-37 4 RCIC Feedwater NC-ISI-51 2 CRD to RWCU 1.6-5

ISI INDEX KEY - - ------- INDICATES NDE REQUIRED N1OR L ASME CODE CLASS 1 INDICATES NO NDE. REQUIRED I*.OR ASME. CODE CLASS:..2 N\ORL ASME. CODE CLASS 3 ASME SAFETY RELATED: COMPONENT BUT CANNOT BE CLASSIFIED UNDER REG. GUIDE 1-26 CRITERIA. ASME NON SAFETY RELATED REF" IFILE NO: (M&SP)- ISI INDEX KEY IS[ DWN: TJH CHKD: .,,T'.I.T APPD: SYSTEM: LINE: DWG: 1.5 5-1 [REV: :0

XDV-2 REACTOR HEAD VENT Zx*ov-i -N-7 CV- 2371 N- 14-, RE ACTOR RV 71 RV-2-71G 'A

SSEL RV-2-71H RV 71C A0-2-SOB AAO-2-86B TO TURBINE . N-3C STOP VALVE 1 F VSTOP AO- 2-O VALVE TO RCfC 80C PEN x-7B 5FE-2-114B FE-2-11.4CA0 PEN X-7C TO HC STEAM SUPPLY STEAM SUPPLY TO TURBINE LA*1 y2-86A A

AO-2-80A RV-2-71E

                                                                     .......-1..'_.....

RV-2-71A N-30 RV-2-71D RV-2-71F AO0-2--8OD A -6 A z**j 2TO TURBINE STOP VALVE L 1 ,. . N-3A VNJ7STOP VALVE PEN X- 7A FE-2-114A' FE-2-114D PEN '<-70 MO-2374 MO-2373 MS-160 MS-161 TO o . . ... _ -... . .

                                                                                                                                    ... L

• _ _-

CONDENSER PEN X-8 L RE F:. NH- 36241 IFILE NO: 02 NS7 MONTICELLO iSI DWN: TJH CHKDC I APPD: PC SYSTEM: MAIN STEAM SYSTEM LINE: DWG: 1.5-2 REV: 03

                                                                          /       \

IREACTORI I VESSEL I I I I I FW 91-2 FW 94-2 `W. 97-2 FW95-2 H I1) __ J 1-8- F%97- FML2-AIF FEEDWATER "- - .i,*]- l 4A T- FEEDWATER A*" P0 PEN X-91A IL - - X-98 REACTOR WATER CLEAN-UP RETURN H-- - - L- - --------

                                                                                                                                   .----                  FRO
                           \   THERMAL TEE                                                                                       - .    -*  ~REACTOR    WATER CLEAN,-UP  RETURN REF- NH-36241                   IFILE NO:

(M&SP)- MONTICELLO 1St DWN: TJH CHI-KD:..iV APPD:7A*/;- SYSTEM: FEEDWATER SYSTEM LINE: DWG: 1.5-3 FREV: 02 I maL

0I

  *RHR SYSTIEM CV-2 i91PE CV-2790                 XR-17 PEX4                                                             -__---_

x-41 I I SI-I L F-I RHR-6-2I RItR-r-i FROM RHR L- -_ - SYSTEM SHIUlDOVN FROM RHR - - - t- -- SYSTEM SHUTIDOWN RETURNI TO CAV RETURN TOSTEM FE-2-109B O-9i r-SYSTEM o0 F'\E-2-109A TO REACTOR P *- "-" SRIiR-9 "TO RIfR WATER LEANUXOV-B

                                                                      --- Lr"----

F-i>3--< - L -"- - -* - SYSTEM SHUTDOWN4 XOV-7 SUPPLY I RWCUL v-MO 2-53i9 (1" UNE) RC-1 MO-2-43A MO 2-53A m6 2-430 I I r -- - - - RECIRCULATION RECIRCULATION ,I-Al DECONTAMI NATION I L F PUMP DECONTAMINATION CON1NECTION I 6-1 L PUIP DECONTAINATION CONHEC TION k / \ / CONNECTION P-200B P-200A A , XR-6-2

                                                                    /1                                    CRW          HI-i NIW.
                                                                            -----    DECONTAMINATION CONNECTIOu             I\LI  . I"ll I r", ') it .

du.(_-r,,,J IFILE1NO: 04 04 A4 XR-7-2 16? (M&SP)- MONTICELLO Isl DWN: TJH *CHKD: ,)-b ' APPD:I..?r.., SYSTEM: REACTOR CIRCULATION SYSTEM LINE: I DWG. 1.5-4 REV: 01

RHR I 1 \ /1 N RINR TEST LINE , PN N _* TORUS TES1 LIE

                                *    ,/(mc)                     \P              l--

X-221A X-210B P IP TORUS RIIG HEADER P-208AI 1.40-1742 CS-1-2 ICS-1-1 O-1741 IS-3-2 A 2-2 CS-2-1 CS-3-1 FROM ORW ORW FROM. CONDENSATE CONDENSATE STORAGE STORAGE REF: NH-36248 IFILE NO: h-4P (M&SP)- MONTICELLO 151 DWN: TJH CHKD: T- APPD:4 v.... SYSTEM: CORE SPRAY SYSTEM LINE: DWG: 1.5-5 I REV: 0.3

RHR-187 RH867

                                                                                                                                                                    ]CST:-189     CST-120 CST-gT-CST-118*

PEN X-39B -99 RHR-90-1 TO RHR 74-1 MO-2022 MO-2020 CONTAINMENT SPRAY __ *_ _ _ _ -T MO2029 RHR- RTCCST-92 F MO ~--:>I- ?CR F 1SERVICE E S-95 WTR 1 - R TO RHR-6-1 I - MO-2014 MO-2012 IRVI2004 I-I - - --- . . IRV-2004 TO PEN rC-LMO-2030 CROSS TIE X-211B FROM CORE 17-a'- SPRAY AO-10-46A "MO7 -200

             /I AIFL RHR-69-1                              MO-2010              MO-     AL06 TORUS-                                                                                                  MO-208        MO2032FE-llAI (MC)                   PE                                                                                             C\;O2F-018 I FROM FPU                             %VX-210B                                                                WATER SURGE TANK                "*                                                  :I' II                PENý,N                 EL X-213-         TORUS RING HDR?,         RHR-8-1                                                                                        RHR-7                              I RV-2031W         PC-                    (cRHR                                                                             CROSS TIE?--.....-T-t><]---                                       I MO-1986 FROM RHR                                                             EMERGENCY FUEL-                      j                                  j L      H                 RO-1998                                     POOL COOLING'-

CST-89 o-1gs8i RO-2000 I TO POST PAS 59-6 PAS 57-5 r--- -.------ --_-_-_-- I-II ACCIDENT IV RETURN FROM PSTALiN POT CIDN - L-- RV-1992 P-202C (13) RHR-2-3 SAMPUNG RHR-3-3 MO-2002 II RHR-5-1 1 IL - RHRSW-17 RHR SERVICE WATER PAS 58-1 PAS 57-6 - -- "2 F- I I FROM CNOS d, "/V19 SERV. WATER RV-1990 -I h---'RHR--I I RHR-I *_ ____l1 RI / -- t. P-202A r --- A#11

                                                                                                               )     IRHR-2-1   - .. RHR-3-1  . .. I                            ---   I H I~       ~     ~      __                 P-3-Y-I--EFt                              -          ---

IH~4 ____RV-_281

                               .HR-18--l1N                RHR-1O-1, FROM                                                                                         REF: NH-36247                                     IFILE NO:

CONDENSATE STORAGE p MONTICELLO SI OWN: JJP CHKD: ) APPD: y SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM LOOP A LINE: DWG: 1.5-6 I REV: 04 IREV: 04

FROM CONDENSATE CST-98 CST-119 SERVICE WATER C IIU~~X-39A PEN _-CST-93 MO-2021 MO-2023 A TO T ....... . ---- > --- + ---- CONTAINMENT SRHR-90-27 RHR-74-2 SPRAY

                                                                                   *- _-t><]- -_,*

PEN r - -X-13A rl RV-2005 I MO-2013 MO-2015 X-1 " TO L -. - - ~.. .- - .- - - -- -- T -- "-<" RECIRCULATING ICST-ý AO-10-46B I RHR-6-2 PUMP 200B IrCST_88 L-- -- -?TOCROSS TIE LI FROM CONDENSATE MO-4085B I I *SERVICE WATER TO . i CROSS-TIE

 -r                                    MO-2007                                                                                                            FROM CORE, IMO-20331         _ ,,,                                                                                                                      rR69SPRAY                       rI(*211A L   L*/\MO-2011                                                                  RHR-69-2                                                                                T1A FE-10-418'-                                                                                                                                                                       TORUS I O-29                                                                                                               )PEN
             -                                                                   I                                                                                                          r-213A FE-10-119B                I RING-NRI
                                                                                                                                                           *'-zz                            7'0.I TRUS RINGHOR I                                                                                                                                       RR-8-2 MO-1987 L -- T---                                                                                                         CV-1 995            '

I ------------- PAS 57-4 PAS 59-5 TO POST CV-1997 C19 RHR-18-2 FROM ACCIDENT / b COND I SAMPUNG RV-I993 I "*"mI STORAGE RETURN FROM

                                                                                                                                .....           R     P-202ID                                                POST ACCIDENT RHR-S-2       I   MO-2003                 RHR-3-4             RHR-2-4        lPF-02_(.4)*                                                                                  SAMPUNG I~                                                                      l                                                                 jLP:AS 58I-22 I\

4 RHRI RHR-4-2 I--- RV-1991 PAS 57--7 I

                               ",IX   HX--*.
                                          *]--- -                  d           RHR-3-2             RHR-2-2                P-202B                     *,
                                                                                    *      -    --   *                    -        --. (f12) z\"         "i*

R RHR-1-2 / I RV-4282 RV. 428 -- -- L , I MO-1989 . L ._ FROM RHR A REF: NH-36246 FILE NO: MONTICELLO ISI DWN: JJP CHKD: APPD:D: SYSTEM: RESIDUAL HEAT REMOVAL SYSTEM LOOP B LINE: DWG: 1.5-7 REV: 05 a a w w

      ...             MO-2034 Ml- 2035 FE 3493                         PEN X-11 HO-B     HO-7  MO-2036 MAIN STEAM BOOSTER PUMP      MAIN PUMP  .                TURBINE VALVES CONTROL STOP
                                                         -      HPCI PUMP
                                                         --   DRIVE TURBINE' P'-209        . "  .CV.2043
                                               .HPCI- 7 fIpCI-60 PS0-2038                    HP1E-HPCI-82    HPCI-10 IRup'uREvý)

HPCI-9

                                    ~F    .         HP-HPCI                                                      CV-2065 Q (20"ý16-)

PCV-3 A PEN COOLING WATER, RETURN REFER TO HPCI (WATER SIDE) REF: NH-36249 IFILE NO: IP MONTICELLO ISI DWN: TJH CHK -;t" J APPD: SYSTEM: HP COOLANT INJECT!ON SYS. (STEAM SIDE) LINE: Dw G. 1.5-8 *1REV: 03

IIPCI-32 ..IO-20)63 r- --- ><]-[ -FROIU CONDEHSATE I

• STORAGE TANIC F -I IPC_-33 TO CONDENSATE RV-2064 STORAGE TANK MAIN I- -

REACTOR (.

                                                                                                                                  /~

• HPCI PUIAP ,

VESSEL MO-207~ DRI,. IURBINE Ao-23-1BI '-1,*j L-~ I I r _MO-2068 _i4- - Ij MO-2067"§ _ _ _i_ _ I r/ T I RWCU FE 2380 L -.-.- FW 98-2_ IW 94-2 I FW 91-2

                                                  -. ---*     - - --   FEEDWATER FW 97-2   -

PEN (SI LEVED) RO 290B HIPCI-42 i-U TOCORE SPRAYIh I TEST UNE \ RO 2066 I X-- E . . MO-2061 HPCI-3

                                                     --.-  ---. I_ MO-2082
                                                                     .--0--

REF: NH-36250 IFILE NO: FOP (M&SP)- MONTICELLO ISI DWN: TJH CHKD: V'QI APPD:-.,---- SYSTEM: HP COOLANT INJECTION SYS. (WATER SIDE) LINE: DWG: 1.5-9 REV: 02

0 4

  /       \                        PEN X-10
 =------    -  -)M      2075   (SLEEVED)
  ~REACTOR~

I VESSEL I+ FROM4CONDENSATE MAINSTEAM STORAGE TANK H-P-207 RCC PUMP DRIVE TURBINEj R0 REACTOR TO BAROMETRI CONDENSER REF: NH-36251 WFILE NO: FG? (M&SP)- MONTICELLO SI DWN: TJH CHKD: iQ\'Q) APPD: Liy,* SYSTEM: RX CORE ISOLATION COOLING (STEAM SIDE) LI NE: DWG: 1,5-10 , EV 0 DWG: 1.5-10 REV: 02

X-227 REF: NH-36252 IFILE NO:

       .WP (M&SP)- MONTICELLO                         ISI DWN: TJH       CHKD:             APPD:,      .

SYSTEM: REACTOR CORE ISOLATION COOLING (WATER SIDE) LINE: I DWG: 1.5-11 I REV: 02

0 TO TE* TANK XPI-13 XP FROM RV-11-39A ACCUMULATOR XP-ll-14A 7-204A PEN X-42 (SLEEVED

          ý4XP-7                                 XI'-

XP L REACTOR' I VESSEL

                             )                                P-2038 ACCUMULATOR T-204B REF: IIH-36253               IFILE NO:

(M&SP)- MONTICELLO IISI DWN: TJH CHKD: I.& APPD:dUi.,jy SYSTEM: STANDBY LIQUID CONTROL SYSTEM LINE: DWG: 1.5-12 JREV: 02

CV-.3268( TO STANDBY GAS TREATMENT AND REACTOR BUILDING PLENUM AO-2387 AO-CV-2385 S AtR PURGE SUPPLY FROM DWV-85 REACTOR BUILDING L IAD-2378 TO STANDBY GAS TREATMENT AO28 AND REACTOR BUILDING PLENUM EZ r- --- NI---II VACUUM DWV-8-2, FROM RELIEF SECONDARY AD-2379 - :::] CONTAINMENT I AO-2396 ~ X-240 DWV D-I

                                  -                                                                                                        VACUUM RELIEF I-     -                                                  FROM SECONDARY
                                                                             -218                                    AO-2380     `]j        CONTAINMENT V8V-488 AO-4539 CV-2384                             ýo                                   NH--36258 N.H-46162 AO-4540 NH- 94896 NH-94897 REF: NH- 1 16629                   VFILE     NO:

Rupture Disk MONTICELLO ISI SPIPE SYSTEM IDWN: BLL CHKDc J APPO:D: ISYSTEM: PRI. CONTAIN. ATMOS. CONT. SYS. [LiNE: I JDWG-: i.5-15 jREV: 03

TO STORM SEWER

                  ,ESEL CENERATORS
                     *12                           TO STORM SEWER WATER PUMPS D

2 REF: NH-36665 IFILE NO: IWP (M&SP)- MONTICELLO ISI DWN: JJP CHKD: APPD:-. P-111A SYSTEM: EMERG. DIESEL GEN. EMER. SERV. WATER LINE: I DWG: 1.5-14 REV: 03 IDWG: 1.5-14 I REV: 03

SW-105-1 ECCS PUMP MOTORS AND VENT UNITS NH-36664 REF: NH-36665 IFILE NO:

                                            '  Sp (M&SP)-  MONTICELLO                ISI EMERGEKCYSERVICE EME        SERVICE     DWN: JJP      CHKD:     "     APPD: c:kJ WATER PUMP       WATER PUMP P-In1c           P-1i1D            SYSTEM: EMERGENCY SERICE WATER LINE:

DWG: 1.5-15 _ ip V v w w

                                                                                            -A

RHR SERVICE WATER PUMPS NH-56664 REF: IIH-36665 IFILE NO: IP (IM&SP)- MONTICELLO ISI DWN: TJH CI-IKD: UJ/'.' APPD:AY, SYSTEM: RHR SERVICE WATER LINE: DWG: 1.5-16 REV: 02

TO EXHAUST FRO1, ORPRIE WATER HEADER WATER HEADER REF: NH.-36245 II-LE NO: 5i (M&SP)- MONTICELLO ISi

                        .DWIN: TJH      CHI(D: !-/\      APPD:,..

SYSTEM: HYDRAULIC CONTROL UNIT LIIN E: DWG: 1.5-17 I RV: o03 0 0

0I FROM HCU, SCRAM RISERS r-- FROM HCU I cv-3-32C cv-3-32A SCRAM RISERS( IF-

                            .L

• FROM HCU SCRAM RISERS I -sc1- -1 SCRAM DISCHARGE VOLUIE-- I I--SCRAM DISCH ARGE VOLUME I I

                                                   ~.DCV-3-33A I

Ao-CV-3-330 REF: NH-36245 IFILE NO: 18 K'*")h (M&SP)- MONTICELLO ISl DWN: TJH CHI(D: I.VlA APPD:,#-.-.- SYSTEM: CONT. ROD DRIVE SYS. (SCRAM DISCH. PIPING) LI NE: QWG: 1.5-18 REV: 02

PEN CV-1478 X-22 PEN

• AI-571 AI-574 CV-7gs6 fINSTRUMENT AIR A89 x-229B At- 8B3 1STRUIAEIIT AIR TO DRYWELL S* TO TORUS

                                                                                    ,AI-631 AI-630 a

PEN AS-78 L AS-79 A-39 I> 4 X-21

                                                                        -- SERVICE AIR TO ORYWELL PEN                                                                                 PEN X-105---G                                                                           X-34A AI-700     AIt-599 AI-584
       ~-N V--~ J).t      --     ~

4-N2TO SUPPLY RV-2--71B AI-708 .AI-598 Al-577 ACC2 SUPPLY S ACCUMULATOR p*" TO RV-2-71F ACUMUAcuuTOR NH-36049-4 NH-36049-10 NH-36049-12 REF: NH-36049-14 IFILE NO: 19 (M&SP)- MONTICELLO ISl DWN: CADWorks CI1D: ) A. SYSTEM: COMPRESSED AIR SYSTEM LINE: DWG: 1.5-19 REV: 01

PEN X-20 DO--51 DM-152 Du 58 OEMIN. WATER 7-- TO ORYWELL DEMIN. WATER SYSTEM TO COOLING FROM REACTOR BUILDING WATER PUMP HEAT EXCHAIGERS

                 ,(0-4230               MO40-4229 MO-1426                RBCC-15
                 * -PEN X-24                    PEN X-23 FROM ORYWELL             TO DRYWELL COOLERS                 COOLERS IIH-36039 REF: NH-36042-2                IFILE NO:     20 PtP        (M&SP)- MONTICELLO                  ISI REACTOR BUILDING COOLING WATER SYSTEM                           DWN: TJH          CI-IKD: ,I, U."    APPD:      K SYSTEM: DEMIN. WATER ýY'STEM & RX BLDG CW LINE:

I DWO: 1.5-20 REV: 02

RC-7--1 R 1 TO' HPCI ..... FROM CLEANUP iciIHEAT EXCHANGERS RC-7-72 R 62 TO RC- - RC-I . RC-104 *O-2397 k MO-2398 FROM REACTOROP-t j - T -- --{1> --..- . RECIRC, LOOP "' '

                                                                                                                                            -          --  TO CLEANUP
                                                                                                                ]*                     PENI                RECIRC. PUMPS X-14 FROM REACTOR ---   -.-      ---

VESSEL DRAIN REACTOR WATER CLEAN-UP SYSTEM PEN AD- 541A AO- 418 X-18 DRW-93 FRO, DRYWIELL -- TO FLOOR DRAIN FLOOR DRAIN SUMP r COLLECTOR TANK( PEN AD-2561A AD- 2561 X-19 cRW- III FROM DRWL EQUIPMENT DRAIN L < ~TO WASTE

                                                                                                  -                                              COLLECTOR TANK NH-36043 N1H-36044 LIQUID RADWASTE                                                                REF: NtH-36254                                 IFILE NO:

(MccSP)- MONTICELLO ISI DWN: TJH CHKD: i,.) APPD:,*',-F-SYSTEM: RX WTR CLEAN-UP & LIQUID RADWASTE LINE: [WO: 1.5-21 REV: 02

0 I

       -     N
  /             N I-  -    -.   -I I                I II I REACTOR I I VESSEL I TO DRIVE MECHANISM NITROGEN PURGE SUPPLY REF:                            IFILE NO:         22 (M&NSP)-    MONTICELLO                   ISI DWN: CADWorks CHKD:.*&r               APPD:    ..

SYSTEM: TRAVERSING IN-CORE PROBE SYSTEM LINE: DWG: 0 WG: 1.5-22 I E:O 1.5-2:2 I REV: Ol

EXCESS-FLOW MANUAL CHECK VALVE VALVE FR

       *1"   LINIE

• I* I' *O* 1 LEHE TO PRIHARY INSTRUMENT SYSTEMDA RACK TYPICAL FOR EXCESS-FLOW CHECK VALVES EXCEPT EXCESS-FLOW CHECK VALVE FOR PENEiRAllON X-28F MANUAL VALVE EXCESS-FLOW CHECK VALVE FOR PENIETRATION X-28F NH-_36241 REF: NH-36242 [FILE NO: 23 (M&SP)- MONTICELLO ISi DWN: CADWorks CHKD: ., 2.- APPD: .

SYSTEM: EXCESS-FLOW CHECK VALVES LINE: DWG: 1.5-23 REV: 01

TO POST ACCiDENT TO OXYGEN TO HYDROGEN-TORUS T0 OXYGEN ANALYZER FROM HYDROGEN-OXYGEN TO HYDROGEN-ANALYZERS OXYGEN ANALYZERS NH- 46162 NH-91197 REF: NH-96042-1 IFILE NO: 26 (M&SP).- MONTICELLO ISI DWN: CADWorks CHKD: , APPD: v SYSTEM: PRIMARY CONTAINMENT SAMPLING SYSTEMS LINE: DWG: 1.5-26 1 REV: 01

14-68 1 REACTOR VESSEL ci- - - -. 4 N-11A N-118 k F5 cJ- ~ t-12A N N-120 4~JN-BA N-Ba ALL BOUNDRIES SHOW, ARE QUALITY GROUP A REF: NH-36242 IFILE NO: 27 kIoS (M&SP)- MONTICELLO ISI DWN: CADWorks CHKID: ,,'s.K- APPD: . SYSTEM: REACTOR VESSEL INSTRUMENTATION LINE: DWG: 1.5-27 REV.: 01

NOTES: INTERIOR ATTACHMENTS W!THIN BELTUNE REGION (B-N- 2) INTERIOR ATTACHMENTS BEYOND BELTLUNE REGION (B-N-2) CORE SUPPORT STRUCTURE (B-N-2')

V. 29'-3 1/2" TOP OF

• ACTIVE FUEL BELTLINE REGION EV. 17'-3 1/2" )

BOTTOM OF ACTIVE PJEL RE I%57 MONTICELL0 isO DWN: BLL C, .KD:-I APPD: AJ D ELEV. 0-0" / SYSTEM: RX VESSEL ITERIOR BOTTOM HEAD DRAIN CRD STUB TUBES-LEVATION ViEW LINE: N/A DWO: St-0 REV: 04

RX VESSEL HEAD VENT Cont. on ISI-782A-A 1013'-0" WELD NO WELD W-1 TOP HEAD CIRC WELD W-2 to TOP HEAD MERIDIONAL WELDS W-7 W-8 TOP HEAD-TO-FLANGE WELD 270" 11 s' = WELD NO. NX-8290-57 (TOP HEAD) O = BOLT NO-NX-8290-71 *(N7 NOZZLE) REF: NX-8290-72 (N6 NOZZLE) IS' MONTICELLO "SI DWN:JJP CHKD: P. APPD: 1~ SYSTEM: RX VESSEL TOP HEAD LINE: N/A DM = DISSIMILAR METAL WELD 180* DWG: ISI FIG. 1 REV: 05 0i I

0 NORTH 180' 51 01 47 l )10101010101 43 0.00 "0010I0t0\ QIo 39 00..0010101010oJ000 6 3/8"

*r 0100o000oo10 31 00000000000oloo                                  UPPER CRD0 HOUSING WELD
                                                                                                 ?        VESSEL 27 -go 0ooO0oooo01O00(i                            -270' 23 00000001010101 19 w0010000010000                                                               BOLTS STUDS 15
                    ý00101000100000                                                             NUTS 11.
                     `,00100,0010,00 it

.07 .03 LOWER CRD. HOUSING WELD 02 06 10 14 18 22 26 30 34 38 42 46 50 VIEW LOOKING DOWN REF: NX-7831-471 (0 = PERIPHERAL CRD HOUSING I%" MONTICELLO ISl CONTROL ROD DRIVE LOCATION & HOUSING WELD DWN: MCWI CHKD: RJ'D APPD: /"*,,-- SYSTEM: CRD LOCATION RX VESSEL (TYPICAL) LINE: N/A DWG: ISI FIG. 2 REV: 04

STANDBY LIQUID CONTROL RX VESSEL BOTTOM HEAD DRAIN Cont. on ISI-74215-A Cont. on ISI-821-A 0" WELD WELD NO. W-1 BOTTOM HEAD DOLLAR PLATE W-2 SEAM WELD BOTTOM HEAD DOLLAR PLATE W-,3 TO SIDE PLATES W-4 TO BOTTOM HEAD SIDE PLATES W-11 MERIDIONAL WELDS 270" 90" W-12 BOTTOM HEAD SKIRT WELD IS' NX-82g0 (BOTTOM HEAD) = WELD NO. CONSI NX-829D-65 (NIO NOZZLE) REF: NX-8290-71 (N15 NOZZLE) KS? MONTICELLO ISI DWN: MCWI CHKD: 12-,&DY- APPD: 9< SYSTEM: RX VESSEL BOTTOM HEAD 180" LINE: N/A DWG: ISI FIG. 3 REV: 03

FLANGE TO VESSEL RX EL. 52'-5 1/4" TOP HEAD VCBC-5 1001'-10 1/4" VL !=0 V 01V LFO- 0 L o=- COURSE 4 VLDB-2 4" 5'- 4" VLDB - 154 4' -- 5" 4,-----, VCBB-4 RX EL. 41'-5 11/16" r II L I 990'-10 11/16" 0 o0 0 000 COURSE 3 L-N2'I, N-SB N-4C N-4B N-4A V5CB N-1 A N RX EL. 30'-6 1/8" VLCB- 2L VLCB-1 V\CRP-- 979'-11 1/8'

                                                                                                                                        *CORE COURSE .2                                                 VLBA-2                                                 VLBA-1 REGION RX EL. 19'-6 9/16"                                                                                                             J 959'-11 9/16"                                   -1 0                                                                               )

VCBA-2 COURSE 1 VLAA-2 r N-2G 0 C 0 N-1B VLAA-1 1A N- -10" RX EL. 8'-7" ' N-88 6'-8" 5-I VCBB-1 958'-0" BOTTOM HEAD BOTTOM HEAD TO VESSEL I I 2 2 I I 1 1 1 6 3 I 360' 330" 300' 270" 240' 210' 180" 150" 120" 90" 60" 30" 0 REF: NX- 9,310 -11 NOTES: !iSP MONTICELLO lSl

1. CIRC. & LONG. WELDS ACCESSIBLE THROUGH NOZZLE WINDOWS DWN' BLL CHKD.ýý APPD:

(APPX. LENGTH ACCESSIBLE AS SHOWN) SYSTEM: CIRC & LONG REACTOR VESSEL WELDS

2. VL - VESSEL STABILIZER LUG INTEGRAL ATTACHMENT LINE: N/A DWG: ISI FIG-4 I REV: 05

N-13, -14 (OFF FLANGE)--

RX EL. 52'-5 1/4" TOP HEAD 1.001'-I0 1%/4" 0O 0 0 0 0 COURSE 4 N-3D N-3C 0 N-38 N-3A 0 RX EL. 41'-5 11/16" N-11B N-11A 990'-1I0 11/16" 0 0 COURSE 3 N-40 0 0 0 N-58 N-4C N-4B 0 0 N-4A N-5A 0N N-9 O RX EL. 30'-6 1/8" N-I12B 979'-11 1/8' N-12A EL. 29'-3 1/2" COURSE 2 CORE RX EL. 19'-6 9/16" TREGION 959'-11 9/1.6" 0 00 0 0 0 0 0 0 COURSE I 0 N-2K N-2J N-2H N-2G N-2F 0 N-2E N-2D N-2C 0 EL. 17'-3 1/2" N- 1A N-28 N-2A N-18 0 RX EL. 8'-7" N-OB N-BA 958'-0" 3 360" 1' I 2 2 1 I I I i I 1 1 330" 300' 270' 240' 210" 180' 150' 120" 90" 60' 30' 0' NOZZLE L SYSt EM N-1A RCAD-l NORTH Recirc Outlet ISI-97005-A SOU TH N-i1 RCB3-1 N-2A RRAO-1 Recirc Recirc Outlet Inlet ISI-97006-A ISI-97006-0 BOTTOM HEAD N'2B RRBD-1 Recirc N-2C RRCDOI Inlet ISI-97006-B Recirc Inlet 151-97006-B N-2D0 RROD-1 Recirc Inlet IS5-97006-B N-2E RRED-I Recirc Inlet ISI-97006-B N-2F 'RRFO-I SN-2G Recirc Inlet 151-97005-8 RRGO-1 Recirc Inlet ISI-97005-B N-21-1 RRHO-'I Recirc Intlt ISI-9 7005-8 N-2J RRJD-1 Recirc Inlet N-2K ISI-97005-8 RRKD-l Recirc Inlet ISI-97005-B N-.3A MSAD-l Moin Steam tSI- 13142-33-A N-3B: MSBD-i Main Steam ISl-13i42-34-A N-3C MSCO-. Main. Steam tSt- 13142-35-A N-30 MSOD-1 Main. Steam ISI-13142-36-A N-9 EL. 986'-O" N-4A FWAD-1 Feedwater N-4B

                                      .ISI-13142-53-A FWS3D-i    Feedwater          ISI- 13142-53-A N-4C     FWCD-1     Feedwoter          ISI-13142-52-A N-40     FWOO- I    Feedwater N-5A ISI- 13142-52-A CSAD- I    Core Spray         tSI-13142-26-A N-SB     CSBD-LI    Core Spray        ISI-13142-31-A REF: NX-9310-11 N-SA     JPAD-I     Jet Pump tnstr. ISI-16 N-SB     JP9D-l     Jet Pump Instr,    ISI-16                                                          hSP       MONTICELLO N-9 N-IIA CRAD-1 VIAE-1 CR0 Return         IS FIG. 5                                                                                                                 ISI Instrumentation    ISI-97008-A N- li I  VIBE-I     Instrumentation    ISI-97007-A                                                    DWN: MCWI           CHKD: 'RfO               APPD:      ,

N-12A VICE-1 'tnstrumenrtation ISI-19 N-12B VIDE-1 Instrumentation ISI- 19 SYSTEM: RX VESSEL NOZZLES N-1i3 VFAE-1 Flange- to-Nozzle ISI FIG. 5 N-14 VFBE-I Flange- to-Nozzle ISI FIG. 5 LINE: N/A DWG: IS FIG. 5 REV: 04

REACTOR VESSEL STUD. WASHER & LIGAMENT LOCATION NUT NORTH 180' WASHER Elw 2 1/2" SET vr-:i t 90" 1 53/4"-- I 0' STUD SOUTH 7/16"0 I BUSHING U REF: NX-8290-63 6"o X 7.4t20" LG 15/16" INS? MONTICELLO ISl DWN: MCWl CHKD:ý i/ APPD: .z-SYSTEM: REACTOR VESSEL BOLTING LINE: N/A DWG:, ISI FIG. 6 REV: 05

TWl 6-1 4"-HE IST) = HANGER NO. E = WELD NO. REF: NX-13142-17 IW MONTICELLO ISI DWN: JJP CHKD: APPD: c:'-. SYSTEM: RHR "A" SUCTI6N LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3,20) DWG: ISI-13142-17-A REV: 06

                                                                                    -0 w                                         w

CONT. ON ISI-8292-42A 1IPCI PUMP SUCTlON NOZZLE

                                  \EL. 9 01--N-                                                       ý10<RTI H-(ANGER NO.

WELD NO. REF:NX-13142-28 IFILE NO: KW (M&SP)- MONTICELLO Isl DWN: TJH CI-KD: W1" - APPD..,,z-H.IGL PRES_._URE COOLANT INJECTION (WATER SIDE SUCr!ON) SYSTEM: HPCI WATER LINE: TWm-14"-HE

• FROU CONMENSATE STORAGE TANK (QUALITY GROUP D - YELLOW)

DWG: ISI-13142-17B I REV: 04.

FROM CONDENSATE STORAGE NO CONT. (NON-SAFETY) TWl 7-14"-HE

                                                                                                        ýgr NORTH TW15-14"-HE TW27- 20"-HE EL. 900'-11" 897'-7" CONT. ON IISI-13142-49A REF: NX-1 3142-17
               = HANGER NO.

PEP MONTICELLO DWN: JJP CHKD:

               = WELD NO.               RHR SUCTION B             SYSTEM: RHR B SUCTION I WkEC. -.irvýnr, IA                           ITEM C3.20)                                               -13142-17-C

CONT. ON ISI-97004A PENE. X-13A TW20-16"-DBGORT

                                                                -TW2O-1                            6-GE S-ýMO-201 S W-17 24
                                                      *-MO-2013 H-tO IA I.WEST SHUT DOWN    COOLING          EL. 935!-0" W-14
                                                   -2033 W-5     'ONT.       ON ISI-1,3142-37A i4l     TW30-14"GE W-2        EL. 927'-10"
                                                  )N-6 12-62 8"-GE                      TW22-14"-GE

_______2__6__N RHR DISCHARGE B REF: NX-131 42-18 I%5? MONTICELLO ISI DWN: MCWI CHKD: iPP: APPO: NOTE: H-3 IS NON-EXISTENT. IT WAS REMOVED PER MOD 920520 SYSTEM: RHR B LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20)

                                                  !DWG:             ISI-123142-18--A                         REV:   05

CONT. ON - ISI-13142-18C Nr

                                                   = HANGER NO.
                                                     -=

WELD NO. ISI APPD: Clg7 5 .1J SYSTEM: RHR "B" DISCHARGE LINE: TW19-10"-GE, TW19-14"-GE I REV: 06 w W Aw Ro a

                                      ýOZRTr RHR "B" HEAT EXCHANGER (E-200B)

CONT. ON IS1-13142-188

                                   = HANGER NO.

ON CS. = WELD NO. 42-18

                                                                                                            ý RTr HPCI TURBINE STEAM DISCHARGE
                                                        *CONT. ON ISI-8292-48-A
                                                                                                         = HANGER .NO.

COýNST

                                                                                                         = WELD NO.

Mre MONTICELLO ISI DWN: MCWI CHKD: POP APPD:,.,,,r SYSTEM: HPCI STEAM SIDE DISCHARGE NOTE: 2" DRAIN LINE EXEMPT PER IWC-1222(a). TORUS CONN. X-221 LINE: NOTED (BAY 6) IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWG: IS1- 13142-19-A I REV: 05

NORTH EU. tORUS PENET7 ' X-212 (BAY 12')- = HANGER NO. S = WELD NO. RE--:NX-131L~2- 19 MONTICELLO !St DWN" BLL CHK APPD: SYSTE.Mt: RCIC STEAM DISCHARGE LINE: RS3-8"-HE IDWG: IS1-13142-19-B REV: 05

                                                                ýORTI TORUS PENET
                                                             = HANGER NO.

CORE SPRAY "A" PUMP P-208A (14-1A) CONT. ON 13142-31C

                                                             = WELD NO.

EL. 897'-10" ISI DWN: MCWI CHKD: 1-* APPD:zo:-.,-- SYSTEM: CORE SPRAY "A" SUCTION LINE: TWlO-122"-HE DWG: ISI-1 3142 A I REV: 05

                       .'4.

TORUS NR Tr NO CONT. I QUALITY GROUP D (YELLOW). CORE SPRAY "B" PUMP P-208B (14-10) CONT. ON ISI-13142-26C HANGER NO. WELD NO. REF: NX-13142-20 rS7 MONTICELLO ISI DWN: MCWI CHKD: D - J,4tl APPD: , SYSTEM: CORE SPRAY "B" SUCTION LINE: TW6-12"-HE DWG: .ISI-13142-20-B REV: O5

CONT. ON ISI-13142-26B 151-3142268EL. 987'-", 100' AZ 'ýSR Tr NOZZLE N-5A SEE ISI FIG. 5 GE: HANGER NO. (G WELD NO. O = BOLT NO. W = VALVE NCO. REF: NX-1 3142-26 MONTICELLO SI DWN:,MCWI CHKD: - APP D":, SYSTEM: CORE SPRAY "B" DISCHARGE DM = DISSIMILAR METAL WELD . = INACCESSIBLE ,LINE: TW7-B"-ED DWG: ISI-13142-26-.AREV: 05

NORTH BLDG RC U ROOM RED..) CONT. ON Sf 131.2-26A. EL. 978'-6" HANGER NO. GEj = WELD NO,

 '!       MONTICELLO                             :SI DWN: BLL          CHIKD:'17hj      APPD:      P46 SYSTEM: CORE SPRAY "B" DISCHARGE NE: TW7-8"-GE ; TW7-10"-GE iDWG:            1SI-13142-26-B              DRE.V:   05

EL. 927'-!0" CONT. ON

                    ,SI-13142-26B NOIRH CONT. ON
                  "   ISI-13142--26D
                                                  =  HANGER    NO.
                                                   = WELD NO REF: NX-i 3`42-26 MONTiCELLO                          IS1

DWN: BLL CYKD: APPD: P  !

SYSTEM: CORE SPRAY "B" DISCHARGE LINE: TW7-1i0"-GE DWG: 151- !31A2.-26-C IIREV: 06 I

NORTH TW8- 8"- GE CONT. ON ISI-131 42-51C

                                                       =  HANGER NO.

CON STA = W D

02 WEL RE. NX- 3142-.26 I%P ONIFICELI0 D IS-1512-26

uW': iSI DWN: BLL CHKD: A PP D: Rkr SYSTEM: CORE SPRAY "S" DISCHARGIE fulW2: ISI-13142-26-D REV: 02

NORTH FROM RX BLDG HEAT EXCHANGERS CW3-8"-HF TO CW PUMPS *NO CONT. (NON-SAFETY) NO CONT-(NON-SAFETY) Y EL: 935'-0"

                                               ?  EL. 935'-0" RBCC-79
                                                                                   ,RBCC-Iv         7 MO- 4229 RBCC- 15 P*EN.3 NO C    CONT.

SNON - SAFETY) TO DRY WELL COOLERS

                                                                                                     **         = HANGER NO.

WELD NO. CONST. / W--8"- HF FROM DRYWELL COOLERS NH- 360z 2-2 NX-! 31!42-29 RE,: NX-1 3142-38 MONTICELLO ISI IDWN: BLL CH KDE=,.j APPD: SYS TE-M: RX BLDG COOLING WATER NOTE: (LINE: CW4-8"-HF &CW3-8"-HF LOCATED ABOVE TORUS BAYS 5 & 6 (INBOARD) IDWG: IS1-13142-29-A I REV: 0o3

1* I

                                                                                         ý SR Tr NOZZLE N-58 SEE ISI FIG. 5 GE     = HANGER   NO.

g74' Rx Cubicle (elevated room b~y SBLC - Ladder access by instrument rack) GE- WELD NO. CONT. ON ISI-13142-31B

                                           'E  L. 97 8'- C 0        = BOLT NO.
                                                                                     = VALVE NO.

REF: NX-13142-31 fG? MONTICELLO ISI

                                      ;DWN: MCWI            CHKD: f2.6i          APPD:  0_

SYSTEM: CORE SPRAY "A" DISCHARGE OM = DISSIMILAR METAL WELD LINE: . = INACCESSIBLE DWG: TWI]-R"-ED ISI-1.3142-31 -A RE

974' Rx Cubicle (elevated room by SBLC - Ladder access by

       -I instrument rock)

H-11B

                                                                        =GHANGER NO.

EL. 931'-11" CONT. ON ISI-13142-31D GE WELD NO. 935'-0" REF: NX-13142-31 1%? MONTICELLO ISI DWN: MCWI CHKD: , APPD: M_1' SYSTEM: CORE SPRAY ':A" DISCHARGE LINE: TW11-1O"-IGE DWC: iI1-1_3142-.3,1-B REV: 0

K~. NOTE: CONT. ON EXAMS REQUIRED ON HANGERS ONLY. tSI-13142-318 NO EXAMS REQUIRED ON WELDS PER TABLE IWC-2500-1, ITEM C5.50. <3/8" NOM WALL. -I

                                                                                                                        =  HANGER   NO.
                                                                                                                          = WELD NO.

__/CORE SPRAY "A- PUMP P-208A (14-1A) CS-9-1 CONT. ON ISI-13142-20A IS[ CH KD: 1ZA1ZD A PPD: t~ SYSTEM: CORE SPRAY "A" DISCHARGE LINE: TWl1-l10"-GE ISI- 13142-.31 -C I-REV: 0.5

OR r H-2 TWE.1 9W1-16 TVM-1 18 CONTNONO 151-13142--37D /CONT. ON W-3 ISI-13142-31B REACTOR T UMO-1749, TOUS r -I ROOM I I X-210B (TORUS BAY 15)

                                                                                             =HANGER   NO.

WELD NO. REF: NX-1 3142-31 lI% MONTICELLO ISI DWN: JJP CHKD: APPD:

• SYSTEM: CORE SPRAY "A" DISCHARGE LINE: TW12-8"-HE DWG: IS1-13142-31-D REV: 02

_ p ia-V W k- --

REACTOR NOZZLE N-3A SEE DWG, ISI FIG, 5 IA Ds = HANGER NO. GET = WELD NO e* = BOLT NO. 7" NX-13142-33 NF-36271 = VALVE NO. NF-36271 -1

                                                  'INACCESSIBLE REF: NH-95008-1,2 I        MONTICELLO

MCWI CHKD:. I' STEAM A IA = INTEGRAL ATTACHMENT (ASME

NO CONT. NOZZLE ISI FIG. 5 IA CONT. ON ISI-13142-42A -8 ~=HANCER NO HPCI (PS18-8"-ED) CONT. ON ISI-782 HEAD VENT (V15-2"-ED) DETAIL A (SE -WELDNO.

                                                                                                                                         = BOLT NO.

k.4SBJI- 2) O CI VALVE NO. W-32 NON-SAFETY w-28 NX-13142-34 u-sNF-36271

                                                                                 *INACCESSIBLE        NF-.36271-1 REF: NH-95008-1,2 MSBJ-3                  KSP MONTICELLO                                  ISI
                                                                       )CONT. ON ISI-786A DWN: MCWI             CHKD: R:b       APPD:         -

SYSTEM: MAIN STEAM B LINE: PS2-18"-ED IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) DWG: ISI-13142-34-A REV: 05 0

0  !.-0 - 0

                                                                                                             ý RTr CONT. ON ISI-13142-43A (PSI7-3"-GD)

NO CONT. RV2-71C -~I

                                                        -26      I
                                                        -3CX IA
                                                                                                            = HANGER

( .J = WELD NO. O = BOLT NO NX-13142-35 MLS I = VALVE NI0. NF-36271 NF-36271-1 REF: NH-95008-1,2 UISCK-14A NS? MONTICELLO NOTE: W-6 WAS PURPOSLEY REMOVED DURING PREVIOUS ISO REVISIONS. I- LAf-lAli CHKD: " W-6 WAS NON-EXISTENT STEAM C IA = INTEGRAL ATTACHMENT (ASME ITEM BI0.20) . = INACCESSIBLE

N-3D SEE ISI-FIG. 5 El. 990'-0"--- 288' AZ -- DWN: JJP CHKD: ,' APPD:- .J SYSTEM: MAIN STEAM D IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) LINE: PS4-18"-ED

*    = INACCESSIBLE                            DWG:          IS1-13142-36-A        I REV: 06 wlm                                       w
                                            ,w w

qw

0 I 4,- EL. 942'-3" 333* Az CONT. ON ISI-97003-A

                                                      .. TW36-16"-DB TW30-16-DB TW30-16-GE EL. 935'-0"
         ..EAST SHUT DOW TORUS ROOM CONT. ON
                                                                                                  ýCONT. ON ISI-1'3142-37B
                                                                                                 \,TW33-12"-GE
                                                                                                                           = HANGER NO.

TW30-1 4"-GE CONT. ON

                                                                         )             ~~ISI-1.3142-37C             s
                                                                                                                           = WELD NO.

FE-10-108-ACOs REF: NX-13142-37 N "S?MONTICELLO ISI DWN: MCWI CHK:D: TQ) APPD: , SYSTEM: RHR "A" DISCHARGE LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWG: ISI-13142-37-A REV: 04

CONT. ON ISI-13142-37E CONT.. ON ISI-13142-37D; EL. 931*-11"  : MO-2006 24

                                                ]A IA IA                                            IA IA CONT. ON ISI-13142-48A                                                                           = HANGER NO.

sWg-8"-GE

                                                                                                          = WELD NO.

REF: NX- 13142--37 MONTICELLO ISI DWN: MCWI CHKD: ýýL) APPO: SYSTEM: CONTAINMENT SPRAY (RHR "A") LINE: TW33-12"-GE IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWG: ISI-1314.2-37-8 REV: 05 S

I .- CONT. ON ISI-13142-378 TW33-12"-CE

                                                                   = HANGER NO.
                                                                   =  WELo NO.

IS' CONT. ON IS1-13142-51A DWN: MCWI CHK,D: C:Ni:-*- APPD: ' TW29-14"-GE SYSTEM: RHR "A" DISCHARGE LINE: TW30-14"-GE DWG: ISI- 131 42-37-C I REV: 05

                           *I-1 1-104 CONT. ON ISI-13142-37B N.            I N.I I         RX. BLDG. EL. 935'-0"
                                           / J-.-       TORUS ROOM CONT. ON ISI-13142-37B EL. 926'-51/2" " X-210B                                                 = HANGER NO.
                                                                         = WELD NO.

REF: NX-1 3142-37 I%? MONTICELLO ISI DWN: JJP CHKD: A PD: ,:::t*. AP SYSTEM: CONTAINMENT SPRAY (RHR "A") LINE: TW34-12"-GE & TW34-10"-HE DWG: IS1-13142-37-D REV: 04 w w

NORTH CONT. ON kS'-13142- 37B EL. 953'- 2

                                                                                                                = HANGER NO.
                                                                                                                = WELD NO.

RFF: NX- 3142-57 MON TI.CELLO ISI DWN- B CH K.:: APPD:

                                                                        ,SYSTEM: CONTAINMENT       SPRAY (RHR  "A")

NOTE: LOCATED IN REACTOR BLDG EL935 ABOVE EAST CRD SCRAM &I INSTRUME'NT RACKS. ILINE: TW33-10"-GE DDWG: ISI-,3142-37-E REV: 01

RTr GE = HANGER I*J0.

                                   =~ WELD NO.

REF: NX-1 31142-40 h%? MONTICELLO ISI DWN: MCWI CHK.D: APPD: AtAD A' SYSTEM: HPCI WATER SIDE DISCHARGE LINE: TW3-12"-ED DWG: ISI-13142-40-A REV: 04

CONT. ON EL. 940'-6" IS-13142-53A W2CIAK-59A SR-598 H-2O "*V-)] CIAK-59 H-12 TWH--50* W-18 1IAK-61A i -5

                                                     .EL.                                              933'-0 34B6"21W 1

7010 STEAM CHASE L TORUS EL. 93v-S" 267 EL.OOM9 DWN: MCWI CHKD: 'i-,k AP.PD:.. O.--/

                                                          .SYSTEM: HPCI WATER SIDE DISCHARGE LINE: TW3-12"-E1 DWG:                 IS2-13!42-40-B                      I REV: 05 IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20)

I

             -ROM CONDENSATE STORAGE TANK NO CONT. (NON-SAFETY)

NORTH TORUS PENE. RFHANGER N1. (BAY 12) CWELD NO. CH-WELD NOP. BL R[F: NX- 3142-41 IDWN" IPMONTCELLO S lDWVN: BLL CiHKD: APPD:I*2

                                          'SYSTEM.: IRCIC WATER SUCT'ON ILINE.- TW5-6'-HE, 017- 6"-HE SD WG:          51114-1AREV:                       05S

FL. 955'-2" /---SA -6 pli PSAJ-8 PSAJý Wl 7 4FWA,Ts- I H2 W22 pýTý- 5 PSAK-7 1142 16 TDAK-5 TIDAJR-4A PSH. p Jý c 9 W21B PSH-1 17 NORTH W6 1.8A 1143 W23 PSAJ-4 DAJR-6 IQ EL 949'-0" W5 A W25 PSAJ-3 Wil S W24 T AJ-8 11 PSAJ-ýO V-2 TDAj-7 1149 k, W21 8 W4 0-2034 13A V, 0- W21A W26 170* D V, PSAF-2C TDAJR-9 2035 TDA 4 KiL?- AZ '1150 B-1 PSAJ 11 13 CY) PEN W20 DM PSAF-28 V-1 wl 3 x - 11 TDAF-3 H3 PSAJ-12 '146 W2 14 W 5, wig A 'DAK-10 PSAj-2A 1140 W14 PSAJ-15 TDAJ-2 'PSH-116 1145 -DAj-12 W16 wl DRYWELL PSAJ-13 9 PSAJ-16 H4 PSAj-I - - - - - - 14A, 10 - - - - - - - - *INACCESSIBLE TDAK-13 CONT. ON 1118 E L. 9 3 SR-705 1S I- 1314 2 - 3 4 A DýT-42 (PS2-',8"-E , ý-P-5 M44 911'-1'.' 1/2' TjAK-46A LT-39 TDA W47 S ROOM 5A

                                                                                                           ý)                                  W31                                       TORLI EL. 905'-6"                     TE)Aj- 41        PSH-709          xg HCICI STEAM TURBINE S-201                                    284                                        W4               DA -40             TDAJ-19          H6 TDAJ-36                                    9 CONT. ON ISI-8292-4-8-A                                                                                 281                                                DAY-17
  'STEAMI INLET)                                                                                            i
                                                                                                            ý36           w33                                                                           HANGER NO.

VIO-2036 W40 DAJR- 21 PSH-104 W46 ToAJ-35 275 TDAJ-40 280 w4 5 6 H7 TDAJ-22 W34 W30 TDAJ-18 TDAJ-14 WELD NO. 282 15 43 TDAJý38 T-DAK-2 PSH-103 2/276 W35 9A H5 W29 TDAJ-15 0= ABOLT NO. H13 283 TDAJ-24 1A

                                                                                                                                                                                                       =VALVE NO.

TDAK-34 AJ- \LS 111 J5 TIDAj-31 PSH-100 4 W36 TDAK-26 REF:NX-13142-42 H12 HPCIROOM TORUS ROOM W31 TDAJ-27 277 PSH-102 1,,rS? MONTi.("ELLO isl TDAK-33

                                                                                                                                          'jDWN: BILL                    C H KID6ýý'Mk-          APPD:             I 7DA     30           TDAJ-28            Hio                    H9

SR-708 TDAK-26A iSYSTEM: HPC! STEAM SIDE SS-707 TDAK-268 DM DISSIIMILAR METAL WELD PSH-101 SR-706 ILINE: PS18-8"--rED

!A fN-IEC-F-,Ai- ATTACHMENT (ASVE ITC-M C3.20) I D WIG: ( s J-13j4.2 -42 -A REV: 06

     !S?

W-14 NO CONT. 12 EXEMPT PER IWC-1221(a)(1)_< NPS 4"

              )-2076         -ý .EL 948W-6"
                       ýW-13 RSAJ-17         2 D HEAD X-1O SIl SG     /=     HANGER NO.
                                                           =  WELD NO.
                                                           =  WELD NO.

REF: NX-13142-43 ItSP MONTICELLO ISI DWN: MCWI CHKD: "pA APPD: ,;,- SYSTEM: RCIC STEAM SIDE LINE: PS17-3"-ED = INACCESSIBLE DWG: ISI-1I3142-43-A REv: 05

CONT. ON

                                            ",ISI- 13142- 48B        ý SR Tr 944'-3" RCIC ROOM
                                                                  = HANGER NO.

CE, CONSf1.

                                                                   = WELD NO.

REF: NX- 13142- 48 KS? MONTICELLO ISI ISI-13142-37B DWN: MCWI CHKD: "1-, APPD: #,O,.,." SYSTEM: RHR SERVICE WATER LINE: SW9-8"-GE DWG: ISI-13142-48-A REV: 04

CONDENSER ROOM swH-igo W1 H4RT (TURBINE BLOC.) SWAJ-2 RCIC ROOM12S-9 (REACTOR BLDG.) ISI-13T42- 48A T, -SWAJ RR1-1 EL~E 9448--DH-X lRHR SERVCE WATER

                                                                             -OT    ONONT.

(c* -"N)= WELD ND.

REF: NX-13142-48 I% ?MONTICELLO ISI DWN" MCWI CHKD: "i,.'L-) APPD: 9,tv-*" SYSTEM: RHR SERVICE WATER LINE: sw-8_-GE DWG: IS1-13142-I48- REV: 05

• 0 0

NORTH

                                  = HANGER NO-
                                  = WELD NO.

REK:N X-'.3142 - 49 ISP MON TI.CELLO iS! DWN: OL'- CHK.D:6ýhtý APPD: SYSTEM: RHR/SDC LUNE: REW1O- 18"-HE DW: S5-1314 2-49-A REV: 04

N CONT. ON ISI-113142-37C ý SR Tr CONT. ON ISI-13142-37C EXCHANGER TW30-14-GE

                                          = HANGER NO.
                                          = WELD NO.

NX-1 3142-51 NOTE:

0 CONT. ON ISI -13142-51D

                                                   -  -'ý  I                                  ýOR Tr 26            MO-2021 TW- 14 W-1.3 CS-7                    TW23-12"-GE EL. 962'-0" TW24-12"-E EL 935-0"
                                                                                        = HANGER NO.

S= WELD NO. MONTICELLO ISI DWN: MCWI CHKD: "1,.> APPD: - SYSTEM: RHR B LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWG: ISI-13142-51-B REV: 05

I El_. 962'-0" NORTH STORUS: ROOM EL 94.3'-6" CONT. ON ISI-i3142-26D

                                                                    = HANGER NO.

1 = WELD NO. EL 929'-6" REF: NX-, i 3142--51 fi %S' MONTICELLO ISI ODWN: BLL CHK[:5::TJ_ APDD:

                            !SYSTEM: RHR B 1LINE: TW24-12"-GE SDWG:

IS1-13!42-511-C REV: 02

NORTH

                                =  HANGER NO.

WELD NO. REF: NX- 13i42- 51 I:KS? MONTICELLO IS! iDWN: BLL CH KD: AP LSYSTEM' RHR B NE: TW23-1I"-GE R __W_: ISI1-i3'.42-51-D IEV: 01

ISP MONTICELLO ISI DWN: JJP CHKD: 0 APPD: '=i.. SYSTEM: FEEDWATER C & D LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) DWG: ISI-13142-52-A REV: 05 w w w v

                                   ýw-ci   FWPB-2   (FWPB-3                                                             N35RTH FWGe)=                                HANGER NO.

EL. 988-3" ._ . = WELD NO. 45' AZ N 4A W-35 SEE I FIG. 5 N-4B F W A 4FW PB-1 FW-36B BOLT NO. FWP[-] 46 VALVE NO. 468 FW 12 H-2 2 - " E

                                -FW2B-1O"-ED----                                 FWAK-21   FW2B-14"-ED IAW-                                                                 -358       -                      NO CONT.

W-1 6 tFA~

                                                                                              "                    , ON-SAFETY
                                               ~91     -   -WS l'  /        ~~

FWAo-22 ~< IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20)

NO CONT W-2 NORTH H-2 EL 1003'-1' ENON-SAFETY / 4 PC-81 H N-O. EL 99 '--' W-4 CNT ONONST... r= WELD NO. H-3 SR-390 3142-62 EL 926'-1

NORTH NO CONT. 2NON SAFETY 1= HANGER NO.

                                                     = WELONO; REW!:0-18"-HE   REF: NX-    42-67 CONT. ON ISI-13142-49A   PIIlP    MONTICELLO                          ISI DDWN: BLL        CHK.Dq-    j    APPD: R
             .SYSTEM: FUgL POOL EMERCENCY COOLINC ILINE: REW11-8"-HG iDWG:       i     ISI-13142-67
                                         ~iI           i REV: 05

I NBA SEE ISI-FIG. 5 EL. 960'-2" 60'AZ DM N8B SEE ISI-FIG. 5 EL 960'-2' 240"AZ Is' C ONST.

                                                                            = WELD NO.

REF: NSP MONTICELLO ISI DWN: MCWI CHKD: 'R-ý APPD: - NOTE: LOCATED IN DRYWELL SYSTEM: JET PUMP INSTRUMENT NOZZLE LINE: DM = DISSIMILAR METAL WELD DWG: ISI-16 I R E V: 04 0 0I

NOZZLE 12A SEE ISI-FIG. 5 NOZZLE 12B SEE ISI-FIG. 5 1" HALF C NO CONT. EL 984'-7" 220"AZ WELD NO. REF: NX-8290-62 N12A REF:NF-97010 N12B REF:NF-97009 Ih5? MONTICELLO ISi DWN: JJP CHKD: APPD: SYSTEM: RX INSTRUMENT NOZZLES LINE: RLM3-1"-DCA & RLM4-1"-DCA DM = DISSIMILAR METAL WELD DWG: ISI-19 REV: 05

f -- TURBINE: MANUFACTURER: TERRY STEAM TURBINE CO. SERIAL NO. 35690 STEAM EXHAUST MATERIAL: ASTM-A217-60T / GR. WC-6 I0 NORTH (NO CONT. EXEMPT HORZ. FLANGE PER. IWC-1221(o)(1) (36 BOLTS)

                                    < NPS 4")

STEA. EXHAUST (CONT. ON ISI-13142-19-B)

                                                                                                      -"      STEAM INLET SUPPORT 'PEDESTALS
                                                                                                    & ANCHOR BOLTING SUPPORT "B" (SUPPORT "A")

VIEW FROM PUMP END (LOOKING NORTH) PUMP SUCTION CONT. ON ISI-13142- 41-A SUCTION -* DISCHARGE 2 BEARING PEDASTALS SUPPORT "D" & ANCHOR BOLTING (2) (SUPPORT "C") VIEW FROM PUMP END (LOOKING NORTH) PUMP: MANUFACTURER: BINGHAM SERIAL NO. 27060B MATERIAL: ASTM-126 / GR. WCB LOOKING DOWN REF:

                                                                      -IS         MONTICELLO                              ISI RE-ACTOR CORE INJECTION COOLANT (RCIC)-PUMP                           DWN: MCWI            CHKD,: T,"7C        APPD:9-i.z       "
                                                       & TURBINE     SYSTEM: RCIC PUMP & TURBINE RCIC ROOM EL. 896'-D" LINE:

DWG: ISI-47 REV: 04

0 0 MOTOR HOUSING

                                                                    -1   PIECE CASING ASTM A216 GR.WCB DISCHARGE (NO BACKSIDE WELD ON SUPPORT INTEGRAL ATTACHMENT) 2 MOUNTING FEET ASTM A216 GR.WCB BOLTS SIDE VIEW (TYPICAL)

RHR PUMPS EL. 896'-D" REF: NX-7905-18

                                                                  *lS*P      MONTICELLO                     iSI SUPPORT NAME   PUMP NAME LOCATION                                  DWN: MCWJ           CHKD: iJo)

SUPPORT "A" P-202A APPD: -j RHR "A" ROOM SUPPORT "B" P-202B RHR "B" ROOM SYSTEM: RHR PUMPS SUPPORT "C" P-202C RHR "A" ROOM LINE: SUPPORT "0" P-202D RHR "B" ROOM DWG: ISI-48 REV: 05

DISCHARGE jiv IL SIDE VIEW (TYPICAL) CORE SPRAY PUMPS EL. 896'-0" REF: NX-7833-33 N? MONTICELLO DWN: MCWI CHKD: :,1*) APPD: SUPPORT NAME PUMP NAME SYSTEM: CORE SPRAY PUMP SUPPORTS LOCATIONOM CONNECTS TO SUPPORT "A" P-208A RHR "A" ROOM NOZZLE N-50 LINE: SUPPORT "B" P-208B RHR "B" ROOM NOZZLE N-5A ISI-49

                                                                                              ýORT ON ISI-821-A
                                                                                           = HANGER NO.
                                                                                           = WELD NO.

CONT. ON ISI-97006-A EL. 953'-11" 9 NF-73880 S;REF: NF-97006 AI 1I S MONTICELLO DWN: MCWI CHK D: i SYSTEM: RWCU DM = DISSIMILAR METAL WELD IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) LiNL.: REW3-4 -ED/EDB/VCA . = INACCESSIBLE )WO: ISI- 73880-A

SR Tr A CONT. ON "ISI-97005-A

                                                                     "  REW13A-28'DCA N'.

LI N. N. N. NO CONT. NON-SAFETY - XR-7-1 IWB- 1"

                                                                                                           *13IŽ~    = WELD NO.

Kco~> REF: NQ-74209-1 0 = BOLT NO. l%57 MONTICELLO SI 1 DWN: MCWI CHKD: j14L-b APPD: t5f^7' SYSTEM: RECIRC, "A" DRAIN NOTE: LOCATED IN. DRYWELL BELOW RECIRC PUMP P-200A SUCTION LINE LINE: REW28-2"-ECB DWG: ISI-74209-1-A. REV: 04

4 -0 e CONT. REW29-.2ED NO CONT. EXEMPT PER IWB-1220(b)(1).e< NPS 1" CONT. ON / REW29-2ECB ISI-97006-A 2%2"x2:

                                 ?r'<  1   -      TCCT
                                                                                                  = WELD NO.

O " =BOLT NO. REF: NQ-74210-1 NS? MONTICELLO ISI DWN:. MCWl CHKD: "D*P&-> APPD:D SYSTEM:: RECIRC. "B" DRAIN NOTE: LINE: REW29-2"--ECB LOCATED IN DRYWELL BELOW RECIRC PUMP P-200B SUCTION LINE DWG: ISI-74210-1-A ,_ , REV: 05

DM ýOR Tr EL 950'-0"

                                                                                               = HANGER NO.
                                                                                               = WELD NO.

FSK-785 O = BOLT NO. NO-74215 REF: NX-7831- 523 tSP MONTICELLO i DWN: MCWi CHKD: Pi1) AP PD: NOTE: 'NO CONT. SYSTEM: STANDBY LIQUID CONTROL EXEMPT PER

• EXEMPT PER IWB-1220(b)(1). NPS < 1". IWC-1222(o) LINE: CH2-1.5" INACCESSIBLE DM = DISSIMILAR METAL WELD DWG: ISI-74215-A

4 CONT. ON ISI-'782A-A EL. 1004'-7" -r'

                                                           = HANGER NO.
                                                           = WELD NO.

Mr MONTICELLO ISI DWN: MCWI CHKD: 'R-".. APPD: 0,,- NOTE: IRLV: 03 I RV 0 NOTE:

NO CONT. (I' LINE)

                                                                                    = HANGER NO.

CONT. ON ISI-782-A = WELD NO. eI - BOLT NO. REF: FSK-782A IS' MONTICELLO ISI DWN; MCWI CHKD: . APPD: _.. SYSTEM: RX HEAD VENT NOTE: LINE: V15-2"-ED LOCATED ON REFUEL FLOOR DURING REFUELING OUTAGE DWC: ISI-782--A-A REV: 04 0

H-3 MSD- NORTH SR577 ,,..0 CL..4 ON SAFETY 4- 1 SM W-7 JDP CHK : APP3 W MTA DM~W2 WL = ISIMLA

     ~
    **    ~ ~      ~   ~~        MAJ1          PS-2 EXMTPRIW-20b(1,<NS1 D7              S-86AAE:0

G PAD 1 1/2" T 24" x 16" RlED. (.50"T) SECTION A-A 0o 4`-0 5/6-IA IA IA S= WHANGER NO. 180' SECTION B-B WELD NO. REF: NX- 7905-,32 KS? MONTICELLO ISI 4 ' - 4 3 / 4 ' O.D. - j -.

                                 - --       /u , ....         DWN: MCWI        CHKD: V         P P D:

RHR HEAT EXCHANGER E200_ A 'SYSTEM: RHR HX "A" LINE: IA = INTEGRAL ATTACHMENT (ASME ITEM C3.10) I JDWG: JSI-7905-32-A REV: 04

                                                                                         , 0
                                                                 -REINFORCING PAD 35" O.D. x 1 1/2" T
                                                               -24" x 16" RED.

(.500"T) INLET SECTION A-A 0* 4'-O 5/6' IA IA IA

                                                                                      = HANGER NO.

SECTION B-B = WELD NO. REF: NX-7905-32 I MONTICELLO 4*-4 3/4" 0,D. DWN: MCWI CHKD: SYSTEM: RHR HX "B" RHR HEAT EXCHANGER E2008 LINE: IA. = :INTEGRAL ATTACHMENT (ASME ITEM C3.10)

                                                                       'I-7905-32-1

f -. CONT. ON Iý SR Tr ISi FIG. 3

• INACCESSIBLE I HANGER NO

                                                              - WELD NO.

O* - BOLT NO. NL-74238 REF: FSK-821 [SP MONTICELLO DWN: MCWl CHKD: "*v* A PPD: SYSTEM: Rx BOTTOM HEAD DRAIN LINE: REW31-2"-ED DWG: ISI-821 -A

                                                                                                .- S NORTH DVMX BYRON JACKSON 671-5-1185 ASTM A216 GR. WCB

_ SHAFT DISCHARGE NOZZLE SUPPORT A (SUPPORT D) GE = WELD NO. VIEW LOOKING EAST REF: NX-B292-42 HPCI PUMP ROOM ISP MONTICELLO IsI DWN: MCWI CHKD: --- : APPD: ,#O, EL. 896'-0" SYSTEM: HPCI PUMPS HIGH PRESSURE COOLANT INJECTION (HPCI) PUMPS LINE: DWG: ISI-8292-42-A REV: 04

MANUFACTURER: TERRY STEAM TURBINE CO. SERIAL NO. 35932 STEAM CHEST COVER BLANK FLANGE BOLTING STEAM INLET STEAM DISCHARGE (CONT. ON ISI-13142-42-,A) (CONT. ON I6I-13142-19-A) STEA*M INLET FLANGE BOLTING-

                            -BOLTS)

S6 (24-BOLTS) TOP VALVE GOVERNOR END -//* AMp', PUMP END- p VIEW FROM PUMP END (LOOKING SOUTH) NX-8292-54 REF: NX-8292-48 HIGH PRESSURE COOLANT INJECTION (HPCI) TURBINE HPCI ROOM EL. 896'-0" _NSP MONTICELLO IsI DWN: MCWI CHKD: pR,.tD APPD: " SYSTEM: HPCI TURBINE LINE: DWG: ISI-8.292-48-A 0

0 NORTH CRD15A-4"-DB CRD14A-6"-DB CRD13A-4"-DB IA a* = BOLT NO. NF-93268-1 SE_ = HANGER NO. NF-93268-4 REF:NX-13142-77 WELD NO. IiP MONTICELLO ISI DWN: MCWl CHKD:cýýj APPD: R NOTE: LOCATED ON WEST SIDE OF Rx BLDG AT 935' ELEVATION SYSTEM: CRD SCRAM HEADER "A" DM = DISSIMILAR METAL WELD LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWC: ISI-93268-1-A REV: 05

                                                                  ý SR Tr TANK 'A'
                                                             = HANGER     NO.
                                                              - WELD NO.

CONST. NF-93268-1 NF-93268-3 CRD- 16A- 2"CCD RX BLDG. (WEST) REF: NF-93268-4 EL. 935'-0" IN5? MONTICELLO ISI DWN: MCWl CHKD: "D-,?- APPD: O SYSTEM: CRD SCRAM HEADER DISCHARGE LINE: CRD-16A-2"CCD & CRD-16B-2"CCD DWG: ISI-93268-1-B I REV: 03

NORTH

                                                                                           )

IA CDBK-47 rL CRD 14B- 6"'- DB

                                                                                            = HANGER NO.
                                                                                            = WELD NO.

NF- 93268- I = BOLT NO. NF-93268-3 REF: NX-13142-77 I%57 MONTICELLO ISI DWN:/ MCWI CHKD: APPD: A (.A9 NOTE: LOCATED ON EAST SIDE OF Rx BLDG @ 935' ELEVATION SYSTEM: CRD SCRAM HEADER "B" DM = DISSIMILAR METAL WELD LINE: NOTED IA = INTEGRAL ATTACHMENT (ASME ITEM C3.20) DWG: ISI-93268-3-A RE,/: 06

NORTH Cont. Line CP5-,8"-HE 2AO--2896 EL. 927'-9"'J DOUBLE STRUT Torus Pen X-205 = HANGER NO. Boy 4 6111L = WELD NO. NH- 36258 NH-94878 REF: NH-94699 I%5~l'P MONTICELLO ISI DWN: BLL CHKD: -j 1 APPPD:

                                  'SYSTEM: PRI. CONT.   & ATMOS. CONTROL

[L;NE: CP5-18"-HE iDWG: !Si-94699-A REV: 03

NORTH Drywell Pen X-47 HANGER NO.

                                                   '     = WELD NO.

NH- 91.230 NH-36258 REF: NH-94879 ISP M MONTICELLO IS! DWN: BLL CHKD.OI.J APPD: :RAV

                 'SYSTEM:    SPARE PENETRATION X-47 LDNE..:. C48-6.-CBISI-94879-.A *
                -IDWO:                                      REV:    03 -J i

I-I INACCESSIBLE

NORTH

                                                                                                      = HANGER NO.
                                                                                                      = WELD NO.

lSP MONTICELLO ISi DWN: 3LL CHKD.-::*. APPD:, SYSTEM: PRI. CONT. & ATMOS. CONTROL NOTE: LIN CP4-20-Fc- & CP3-18"-HE LOCATED ABOVE TORUS BAYS 12 & 13 (ABOVE INBOARD HANDRAIL OF CATWALK) *DWC: ISI--94966-A JRFV: 03

SOUTH Cont. Line Stb, CPi-18"-HE Drywel! Pen X-26

                                                     = HANGER NO.
                                                     = WELD NO.

I F- 36258 REF: N-H-9496 I5PMONTICELLO

           -DWN.:: OLL          C-HKD:Crxi-.. APPD:

SYSTM: CONTAINIME-NT I UG LN: L. Pl1-8"-HE GO~: IS4-94ý966-B REV:.0

N 0 ,(TW4O-4"--DBA) 1 1x4RHAJR-13 H-4 I DETAIL A Ds HANGER NO. GE WELD NO.

                                                                                        = BOLT NO.

NF-97003

                                                                                  'l    = VALVE NO.

NF-74551 NF-96201 REF: NX-1 3142-49 M7 MONTICELLO iSI V-3 - - DWN: JJP CHKD: -fp APPD:'J

                        ": CONT. ON       SYSTEM: RHR SUPPLY LOOP "A" ISI-13142-49-A LINE: RW-1O-18"-ED DM = DISSIMILAR METAL WELD  (REW10-18-HE)  DWG:              ISI-97003-B                    REV: 06

W-19 EL. 964'-0" 13W3-2"DA RBJR-4 W1 SO-10-46 W-22*'TB" RB 1 RHwJ-2° ,I-BA

• CO T.ON

                                                                                   -2*]              I E-917006-8A_ DCAN S RHF-JR-12 7=           WELD NO.

a¢[ 1* A'* IQ i"ISI-97027II-A Y

                                                               -W1-6--\                                                      =   VALVE NO.

REF: NF-97004 DRYWELL EL_942'-.3' ( MONTICELLO 1S1 CONT. ON " M0-2015 WEST SHUT 18" Az ON DWN: JJP CHKD: APPD:*--,, DM = DISSIMILAR ME

• = INACCESSIBLE

0. Nr Q 151GE= HANGER I,O. CONST (GE = WELD NO. CONST O BOLT NO. m = VALVE NO). lI"u MONTICELLO ISi NOTE: LOCATED IN DRYWELL DWN: JJP CHKD: $' APPD:'*-,..J

• B2= BOLTS NUMBERED CLOCKWISE, BOLT (# 1) IN-LINE WITH INLET PIPING. SYSTEM: RECIRC LOOP "A" DM = DISSIMILAR METAL WELD LINE: REW13A-28"-DCA IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) DWG: ISI-97005-A IREV: 06

NORTH

                                                                                               )A
                                                                                                 =  HANGER NO.

CONTI. ON RAJR3~3

                                                                                                  = WELD NO.

iSI-97005-A J'%- NF-96201 RFW13A-28"-DCA N 'X-7831-37-1 H8 RE--: NF-97005 I[A RA-39 A RMVAIR- 13> MONTICELLO is! SS7AR 4 DWN: BLL CH KD: C1t,ý-, APPD:. SYSTEM: RECIRC. MANIFOLD "A" DMV= :D'SSIMILAR METAL WELD IL!NE: REW32-22" A =INTEGRAL ATTACHMENT (ASME ITEM B10.20) DWG: ISI-97005-BPEV: 05

0 IA

                                                                                          = HANGER NO.

NF-97005 NF- 96201 NX- 7831-34 REF: NX-7831-37-3 MSP VONTICELLO lS PLAN 'A-A' EPYWIL DLAN 'B-3' DDWN: BLL C H KD't=` i APPD: 9 3 5'- 0 SYSTEM: PECIRC PUMP A SUPPORTS ELL.952'-0" LINE: IA = INTEGRAL ATTACHMENT (ASME !TEM 810,30) DWO: ISI-97005- C REV: 06

deg Az. NORTH CONT. ON ISI-97006-B CONT. ON N-lB \ ISI-97004-A TW20-18"-DCA SEE ISI-FIG. 5 CONT. ON ISI-73880-A-REW3-4"-VCA EL. 954'-0" IA IA GE = HANGER NO. GE = WELD NO. CONST IA a = BOLT NO. IA NF-97006 W = VALVE NO. NF-96202 REF: NX-7831-37-1 I%57 MONTICELLO NOTE: LOCATED IN DRYWELL DWN: JJP CHKD:

• B2= BOLTS NUMBERED CLOCKWISE. BOLT (# 1) IN-LINE WITH INLET PIPING.

SYSTEM: RECIRC LOOP "B" DM = DISSIMILAR METAL WELD LINE: REW13B-28"-DCA IA = INTEGRAL ATTACHMENT (ASME ITEM B10.20) DWG: ISI-97006-A p

NORTH

                                                                                     = -HANGER NO.

IA

                                                                                     = WELD NO.
                                                                                            !SI jSYSTEM: RECIRC MANIFOLD "B" DM ,= DISSIMILAR METAL WELD DRY WE LL ILINE: REW32-22" IA  = INTEGRAL AITTACHMENT (ASME ITEM B10.20)           I[DWG:           !SI-9:7006-B    REV:    05

NORTH. IA NO. GE = HANGER~ NFF- 97006: NF-96202 N;X-78.31 -34 R*-F: NX-7831-37-3 PAS? MONTICELLO IS? DVRYWELL P AN 'B-B' DWN: BLL CKO:a-A4t-3 APED: -A PLAN 'A-A' SYSTEM: RECIRC PUMP B SUPPORTS EL. 935'-0" EL. 952'-C" L NE: 1 IA = INTEGRAL ATTACHMENT (ASME !TEM .210.30)

                                                        ~DWG:               6Si-97006-C        ]REV:    071-

NORTH

                          \CONDENSING CHAMBER CONDENSING CHAMBER SEE ISI-FIG. 5 (GE       WELD NO.

CONS, REF: NX-8290-62 REF: NF-97007 IS? MONTICELLO ISI DWN: JJP CHKD: -PAPPD:CA.. NOTE: SYSTEM: RX INST. NOZZLE N-11B LOCATED IN DRYWELL ILINE: RLM2- 1 1/2"-DCA DM = DISSIMILAR METAL WELD DWG: ISI-97007-A - REV: 05

AS C SEE DETAL EL 994'-7" 40"AZ CHAMBER CONDENSING SEE ISI-FIG. 5 DETAIL

                                                                        = WELD NO.

REF: NX-8290-62 REF: NF-97008 P"'1 MONT~CELLO ISI DWN: JJP CHKD: APPD: c/-ji1 NOTE: SYSTEM: RX INST. NOZZLE N-11A LOCATED IN DRYWELL UNE: RLM--1/2"-DCA DM = DISSIMILAR METAL WELD DWG: ISI-97008-A REV: 05 A& A& w L,- w Na -- low

00 W-19 ~CONT. ON ISI-97003-BSRT HE-(RHR LOOP A SUCTION) W-20 REW1O-18" RHEjR-31 W-10 W-1E JR-OI RI-fEJR--2L EL 953'- 1Q*-*,-- Wý17.RRHERR-14 ._E_-6f-/ CONT. ON ISI-97004-A MO4DM0 -406a (RHR LOOP RETURN) TW20-16" W 1 R H---W_ RT.95R-15 R(RHR LOOP A RETURN) RHEKR-12 W-2-16 C RHEKR-1O

                                                                                    \WHS-202/

WHJ-3-25 9 W-30 W1 W-9 1

                                                                                                                             %L,4-O                ELN.

CE = HANGER NO, R RHHERR-2 RHEER-2 EL 93'-RH-Jw-11 REF: NF-97027 MONTICELLO

                                                                                                                  'CNT                               ISI DWN: MCWI        CHKD: .LOORE-TUR APPA:

SYSTEM: RHR EQUALIZER NOTE: LINE: TW4O-4"W-DBA LOCATED IN DRYWELL DWO: IS1-97027-A JREV: 05

NORTH Drywell Den X-25 W-2

                                                                              *,,ýEL. 988'- 6" EL. 985*-0" EL. 977-6 1/2" A
                                                                                                                 = HANGER NO.
                                /  "".'AO- 2387 NO CONT..*-                    N NON-SAFETY                                 5:Cv- 2385                                       = WELD NO.

NH-91230 NH-36258 REF: NH-i 05531 MONJTICELLO ISi

                                                                           ,DV\ BLL            CH.KD--   %J, APPD: 9.i NOTE:

COMPONENTS BELOW 985' ARE lN RWCU ROOM iSYSTEM: STANDBY GAS TRTMNT & RX PLENUM LMN: CP2-18"-HE

 = INACCESSIBLE                                                            'DWG               !SI-105531-A          REV: 03

NORTH Torus Pen X-240 EL. 935'-0"

                                                                      = HANGER NO.

Aillsk = WELD=NO t/Ru pt

              .I
             .P    ,J isk l

EL. 925'-5" rap MON T ICELLO ISI DWN: BLL COHKD:Cd-Wy, APPD: R SYSTEM: TORUS HARD PIPE VENT ILiNE: HJPV- 8"- HE DWC: IS1-158074--A REV: 04

       /..-
                                                                        ý RTr K'

(RHR-A-5HEAT EXCH,' 10-2A (E200A) EL. 920'-5 1/" HANGER NO. CONST REF: NX-13142-16 KS? MONTICELLO ISI DWN: MCWI CHKD: , APPD: . SYSTEM: RHR SERVICE WATER NOTE: LOCATED IN RHR "A" ROOM LINE: SW11-16"-GF DWG: ND-ISI-100 REV 02

i 174

                                                                                            -0
                       --    EL. 923'-10 RHR-SW-3-2 SW10-18"-CF CONT ON DWG ND-ISI-102
                                             \    , _ýEL.897-3" Zý           ýHANGER NO.

CON"S I REF: NX- 131-42-50 KS? MONTICELLO ISI DWN: MCWI CHKD:, fKý APPD: .,,, SYSTEM: RHR SERVICE WATER NOTE: LOCATED IN REACTOR BLDG LINE: AS NOTED DWG: ND-ISI-101 .[REV:: 02

I

                                                                                               ýOR T1
                                                                           - CONT ON ND-ISI-110 RCIC ROOM NO
                             ýTORUS ROOM CONT ON DWG ND-ISI- 101
                                                                                            - HANGER  NO.

EL. 897'-3"/ REF: NX- 131 42- 50 KS? MONTICELLO -ISI DWN: MCWI CHKD: 'PJ",O APPD: - SYSTEM: RHR SERVICE WATER NOTE: LOCATED IN REACTOR BLDG LINE: SWIO-18"-GF DWG: ND-ISI-102 REV: o*

64v - 0

                                                                  ")-, CONT ON: DWG I   ND-ISI-104
                                                                                                ýSR Tr SW9-18"-OGF EL. 925'-2 7/8" ROOM RHR "A" ROOM is' HANGER NO.

EL.' 925-2 1/8" CONST REF: NX-13142-93

           ,1RHR "A"                                   ISP MONTICELLO                               ISI HEAT EXCH/                                DWN: MCWI         CHKD: 1R*[z   APPD:   ,z--

S(E200A) SYSTEM: RHR SERVICE WATER NOTE: LINE: AS NOTED LOCATED IN REACTOR BLDG DWG: ND-ISI-103 REV: 02

ORTr EL. 935*-0" CONT ON DWG EL. 931'-0" F:X BLDG FL ND-ISI-105 TORUS ROOM RCIC ROOM IsP- HANGER NO. CONT ON DWG ND-ISI- 103 REF: NX- 13142-9.3 NSP MONTICELLO ISI DWN: MCWl CHKD: *.,,-, APPD: ,#',- SYSTEM: RHR SERVICE WATER NOTE: LOCATED IN REACTOR BLDG LINE: SW9-18"--GF ODW;: ND-ISI-104 REV: 01

AD

                                                                                ý SR Tr EL. 944'-3" CONT ON DWG   CONT ON DWG CONT ON DWG                                                                  ND-ISI-106 ND-ISI-104
                                                                             = HANGER NO. I REF: NX-13142--12 ISP    MONTICELLO                           ISI DWN: MCWI       CHKD: R'J       APPD:      .

SYSTEM: RHR SERVICE WATER BLDG LINE: Sw9-18"-GF NOTE: LOCATED IN TURBINE DWG: ND-ISI-105 REV: 01

S* **

                                                                                              ý OgR Tr EL. 925'-3"J IsH CONT ON DWG                                              =HANGER     NO.

ND-ISI-105 REF: NX- 13142--12 MONTICELLO ISI DWN" MCWI CHKD: -P4,ý63 APPD: o SYSTEM: RHR SERVICE WATER NOTE: LINE: SW9-18"-GF LOCATED IN TURBINE BLDG IDWG: ND-ISI-106 REV: 02

1(-4 - 0 I

                                                                               ýSR Tr SW9-12"-GF P-109A RHR SERVICE WATER PUMP
                                                                            = HANGER NO.

ISP MONTICELLO ISI EL 25-3" CONT ON DWG DWN: MCWI CHKD: j-tAT* APP: D: ND-ISI- 106 SYSTEM: RHR SERVICE WATER NOTE: LINE: AS NOTED LOCATED IN TURBINE BLDG

DWG: ND-ISI-107 REV: 02

35SR SR SR T 1.SSW SR 3SW-3-2 -P-109D RHR SERVICE WATER PUMP 384 SW10-1 P-IOgB 2"-GF RHR W O5 RHR-SW-1-4 SERVICE WATER PUMP SWlO0- 18" -QF is *'m. RHR-SW-2-4 UNG HORIZ. RHR-SW-4-2-2 RUN ___RHR

                                                                            -SW-2-2
                                                                                           = HANGER NO.

0. 14

                                                               = HANGER NO.

REF: NX-1 3142--92 I%5 MONTICELLO ISf DWN: MCWI CHKD: *,,AC) APPD:D:s,- SYSTEM: RHR SERVICE WATER NOTE: LINE: SW1O-18"-GF I LOCATED IN TURBINE BLDG DWG: ND-ISI-109 -FREV: 02

I ."'

                                              "<I I
                                                                                                      ýSR Tr EL. 931'-D" DOOR CONT ON DWG CONT ON DWG
                                                                                - /" ND-ISI-109 TURBINE BLDG CONDENSER ROOM
                                                                                                  = HANGER NO.

REF: NX-13142-92 T%5' MONTICELLO ISI DWN: MCWI CHKD: . APPD: , .- SYSTEM: RHR SERVICE WATER NOTE: LOCATED IN TURBINE BLDG LINE: SW1O-18"-GF DWG: ND-ISI-110 REV: 01

                                                                                                             -0
                                                                                                                 ýORTr NO CONT.

EL. 939'-9" RHR-SW-12 TURBINE BLDG CONT. ON ND-ISI-106 CONT. ON NO-ISI- 105

                                                     'I-CLASS 2 CONT. ON ISI-13142-48-B isi        HANGER NO.

NH-91 187-1 REF: NX-13142--48 NS? MONTICELLO ISI DWN: MCWI CHKD: I.-

R ) A PPD: ,, ..- SYSTEM: RHR SERVICE WATER NOTE: LINE: SW9-8"-GF LOCATED IN TURBINE BLDG (FLOOR EL. 931'-o") DWG: ND-ISI-I1 REV: 0,3

                                                                                                ý OsTr ESW9-6"-HBD ESW-35 ESW9-.3"-HBD EL. 921'-O 207 3S1 208                                                                HANGER NO.

CONST. REF: NF-93484 ISP MONTICELLO ISI DWN: MCWi CHKD.: _jt7p AP.PD: p,- NOTE: LOCATED IN TURBINE BLDG (FLOOR EL. 911'-0" SYSTEM: RHR. SERVICE WATER .= SUPPORT IS EXEMPT PER IWF-.1230 (IWD-1220(o)) LINE: AS NOTED DWG: ND-ISI-123 REV: 02

I 7

   -      .- c        -
                                                                                                            !-i I1 ou Inboard
                                                                                                                   -===    :==-

Dsac~otlon 4~~~~~~~~~~2-

                                                                                                                                       =

andOutboor okn, s ASMRCLASS l N.

                                                                                                                                                    -- p.-

MV7R* MCSUPPORTS 1-8220.-78. N*-LQIQI*Q.4.* Nt*LQ400N=4 Q M* *A* PlantIS0 NH-73025-1.-2. 1---0--*- 2 N NH-73028. NH-73027. NH-73028 79 Irvwoil D &lIfuFonua. St*aUz. Sldrt &AnchorBolL INX-8291-38, 1.5.78 NX-82911-0415-0 10-1911 8.r Hano.. NL-05931-3_NX-8221-52LNH-95932 9CCS Header Sslme Resbani / Struts NL-95919,20., 21. 22, 23, 24. 2, 27. 28., 29, 30, 31, 32 0 Vent H.adwColumns NH-767800NX-8291-25. -34 40 8RARS .1 D ownooer BradnolRasmabrI. NX-8201-25, NK-86911. NH-Si1155-2. NH-94692-1 71& 11BLIE STBINTSBACET HýFHBSJ

     ~

NORTH ~ SUPPORT)910 H-3SB -5 H-7FS H-2SR H-4S7MS EL STABILUZER BARS VENT HEADER DEFLECTOR- VNUNE'SUPESO CHNR H-2MS - DOWNCOMER* H-2FS-BIO TO CONTAINMENT TRUS 1 -"DRYWELL BO-OM MALESTABIUZER MALESTABUZER

            -.SUPPRESSION
                                                                                                                                            )        .       (TYPICAL   8 LACES)                         C I*E      ED 8

HANERS NBSPPRTS SUPPORT(* - d LATEAL RESTRAINT 1 ... SADLERING QTORUS-SEISMIC RESTRAINTS GIRDER DRAIN SSM RTAIS H-4S11-1i4 DIAMBER SUPPORT AND SKIRTSS ANCHOR BOLTSQ REF:NF-36423 IFILE NO: REF:NX-8291-371 REF:NX-8291-26 E -V (M&SP)- MONTI IS' is REF:NH-91990 REF:NH-73028 DWN: JJP CHKD: ,-r APPD:4:w-ý,J REF:NH-73027 -SYSTEM: . REF=:NH-726 IREF. NX-8291-76 REF:-NH-73025 IDWG: 151-8.291 -- 76 291-76 RI-V: IIIVnf n9

TORUS SHELL X-5 (TYP.) VENT LINE & HEADER DRYWELL X-202 (TYP.) PROTECTIVE COVER (NON-CLASS) X-208 (TYP.) VACUUM BREAKER___- - PRIMARY STEAM SRV DISCHARGE UNE (NON-CLASS) DRY*rELL VENT UNE DRAIN X-207(TYP.) REF:NH-91155 BRACING REF:NH-86911

                                       \-LATERAL                             REF:NX-7823-19 RESTRAINT (TYP. 2 PER DOWNCOMER) REF:NX-8291-34 & 25       IFILE NO:

A,B,C.D IP i1I(M&SP)- MONTI ME DRYWEU. & TORUS IS] 10m. BMfDRYF F DEI Wt"MMAWS IS AT MIV FACE FLANHM BO~ING A= V.VES & BLMM FLANGES ARE CLAS Z DWN: JJP CHKD: Z APPD: ALL SRACES ARE WA MCEXCT UI NO'/h. SYSTEM: PRIMARY CONTAINMENT WORK IN CONJUNCTIOC WNTH DWGS: 1.5.55 ; 1.5.70; 1.5.82 LINE: NX-8291-34-A; NX-8291-34-B DWG: 1.5.81 FREV: 01 pw V w IR w a ---

q NON-VENT BAY I 916'-1 1/2" a DOWNCOMER BRACING VENT LINE & HEADER RESTRAINTS EF (INBOARD & OUTBOARD, TYP. EACH NON-VENT BAY) REF: NH-94692-1 & 2 REF: NX-8291-34 IFILE NO: RiS (M&SP)- MON'TI ME DRYWELL & TORUS ISI DWN: JJP CHKD: APPD: -- '.- SYSTEM: PRIMARY CONTAINMENT LINE: NOTE. WORK IN CONJUNCTION WITH DWGS: 1.5MB ; NX-8291-34-A ; NX-8291-34-B FDWG: 1.5.82 REV: o1

90, - -___ - -___ - -___ - -___ - -___ - _ __

                                                              .-              /                                             NORTH I         I      H-4C-BAY 4     SR-3 64 H-4A TORUS PEN X-204A CONT. ON DWG 25-7     NX-8291-79-A(B).

S-4 C-224A BAY R-358 o3BAY5 H-2A = SUPPORT

                                                                                                                     ~=   PLATE REF: NL-95931                  C       = PEN 2REF:            N X- 8291-51 BAY 2                          215-B    REF:NH-95932              IFILE NO:

hbP (M&SP)- MONTI ISl ODWN: JJP CHKD: - APPD: D-:

• REF: NL-95926 SYSTEM: PRIMARY CONTAINMENT REF: NL-95927 LINE: 20" RING HEADER BAYS 1-4 REF: NL-95928 DWG: NH-95932-A REV: 01 ip p 14 W

BAY 8 NORTH C-319 REF: TORUS PEN X-204B CONT. ON DWG BAY 6 NX-R291-79-E(-).\ REF NE-8 91-5 _EF367)95

                                                                            = SUPPORT

= PLATE CIS

                                                                            =  PEN REF: NH-95931 FFILE NO:

BAY 5

                                        %5    (M&SP)- MONTI                          isI DWN: JJP        CHKD:          APPD:     -/t    -

SYSTEM: PRIMARY CONTAINMENT REF: NL-95929 LINE: 20" RING HEADER BAYS 5-8 REF: NL-95930 DWG: NH-95932-B REV: 01

BAY 9 BAY 10 4,- NORTH 11 TORUS PEN X-204C CONT. ON DWG NX-8291-79-C(B).

                                                              '(12 I=   SUPPORT II                *-
                                                      -[-270'                 = PLATE ISI    =  EN REF: NL-95931                ILCBI  NO REF: NX-8291-51 REF: NH-95932              FFILE NO:

hSP (M&SP)- MONTI ISI DWN: JJP CHKD: 4, APPD: .f-zl SYSTEM: PRIMARY CONTAINMENT REF: NL-95919 LINE: 20" RING HEADER BAYS 9-12 REF: NL-95920 PDWG: NH-95932-C REV: 01 k6w - Rw -9 L-1 is w w

BAY 13 4 NORTH BAY 14 TORUS PEN X-204.D CONT. ON DWG NX-8291-79-D(B). X-224E

                                                                                    = SUPPORT CX-224E
                                                                                    = PLATE 15 H-16C                                                                     CB   = PEN S-5                                        REF: NL-95931 H-16D REF: NH-95932             IFILE NO:

Ip (M&SP)- MONTI ISI BAY 16 DWN: JJP CHKD: APPD:<qf,:,,J REF: NL-95921 SYSTEM: PRIMARY CONTAINMENT REF: NL-95922 LINE: 20" RING HEADER BAYS 13-16 REF: NL-95923 REF: NL-95924 DWG: NH-95932-D REV: 01

C = SUPPc RT 0U cIs = PEN

                                                                                                                 =BPLATE REF: NX- 8291-34             FILE NO:

VENT LINE & HEADER IEP (M&SP)- MONTI IWE DRYWELL & TORUS ISI DWN: JJP CHKD: 4 APPD: "1', 0 NOTE: 1. ISI NUMBERS ARE INDICATED ON DWG: NX-8291-34-B SYSTEM: PRIMARY CONTAINMENT

2. WORK IN CONJUNCTION WITH DWG'S: NX-8291-34-B ; 1.5.81 ; 1.5.82 LINE:

3. MC SUPPORT NUMBERS ARE INDICATED ON DWG: NX-8291-34-C. DWG: NX-8291-34-A REV: 01 w w

                                                                   *--                                                   r-l w

BAY # DOWNCOMER DOWNCOMER VENT HEADER ASS'Y VENT HEADER ASS'Y COLUMNS COLUMNS I ISi CBI iSI CBI ISI CBI 1 TO 2 .... H-1 NOTE 3 1 H-1 A,B.C,D NOTE 1 H-1 EF NOTE 2 2 TO 3 H-2 NOTE 3 2 H-2 A,B NOTE 1 3 TO 4 H-3 NOTE 3 3 H-3 A,B,C,D NOTE 1 H-3 E,F NOTE 2 4 TO 5 H-4 NOTE 3 4 H-4 A,B NOTE 1 5 TO 6 H-5 NOTE 3 5 H-5 A,BC,D NOTE 1 H-5 E,F NOTE 2 6 TO 7 H-6 NOTE 3 6 H-6 A,B NOTE 1 7 TO 8 H-7 NOTE 3 7 H-7 A,B,C,D NOTE 1 H-7 E,F NOTE 2 8 TO 9 H-8 NOTE 3 8 H-8 AB NOTE 1 9 TO 10 H-9 NOTE 3 9 H-9 AB,CD NOTE 1 H-9 E,F NOTE 2 10 TO 11 H-10 NOTE 3 10 H-10 A,B NOTE 1 11 TO 12 H-11 NOTE 3 11 H-11 AB,C,D NOTE 1 H-11 E,F NOTE 2 ... 12 TO 13 H-12 NOTE 3 12 H-12 A,B NOTE 1 13 TO 14 H-13 NOTE 3 13 H-13 AB,C,D NOTE 1 H-13 E,F NOTE 2 14 TO 15 H-14 NOTE 3 14 H-14 A,B NOTE 1 15 TO 16 H-15 NOTE 3 15 H-15 A,B,C,D NOTE 1 H-15 EF NOTE 2 16 TO 1 H-16 NOTE 3 16 H-16 A.B NOTE 1 DOWNCOMER, VENT LINE & HEADER RESTRAINTS VENT HEADER SUPPORT COLUMNS REF:NX-8291-34 ILE NO: IWFP MONTI IWE DRYWELL & TORUS ISI DWN: JJP CHKD: #f' APPD: -v-ktJ NOTE 1: WORK IN CONJUCTION WITH DWGS: NX-8291-25, NH-8691. SYSTEM: PRIMARY CONTAINMENT 2: WORK IN CONJUCTION WITH DWGS: NH-94692-1. LINE: 3: WORK IN CONJUNCTION WITH DWGS: NX-8291-34-A; 1.5.81; 1.5.82; NH-76790. DWG: NX-8291-34-C REV: 00

NO CONT. NON-SAFETY

                                                 *FW-91 -1 W-1 FROM FW2A-14"ED P T W-3      FROM FW5-4"ED
               *FW-94-1 CONT. ON Y ISI-13142-52A EL. 940'-6" W

EL. 939'-o* STEAM CHASE

                                                                                  = WELD NO.

REF: NX-13142-30 MONTICELLO ISI DWN: MCWI CHKD:D:)f APPD: e,5o.- SYSTEM: RCIC FEEDWATER LINE: D)WG: NC-ISI-37 II Rl* V 04

                                                         , 7-111ý 9 0

[ABOVE CR0 REBUILD AREA I OR Tr NO CONT. A NO CONT. 3" x 0.438"T DM 4 3.N

                                                                      " x 0;300"T NO CONT.

CONTROL ROD DRIVE TO REACTOR. WATER. CLEANUP RWELD C NO. REF: NX-1 3142-51 NSP MONTICELLO ISl DWN: M.CWI. CHKD: 'F&-J'¢t APPD: 6-*) SYSTEM: CR0 TO RWCU LI NE: DM = DISSIMILAR METAL WELD DWG: I NC-ISI-51 REV: 02

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN INSPECTION PLAN AND SCHEDULE TABLE Periods Period 1 (May 1, 2003 to May 31, 2005) 0 Shortened due to 1 year extension of 3 rd Interval Period 2 (June 1, 2005 to May 31, 2009) Period 3 (June 1, 2009 to May 31, 2012) Scheduled Outages Interval - Period - Outage # - Year 4th 1 1 2003 4th 1 2 2005 4th 2 1 2007 4th 2 2 2009 4th 3 1 2011 Above noted outage dates are subject to change during the interval. Inspection percentages are based on periods in accordance with Program B and Code Case N-598, regardless of currently scheduled outage dates. Notes:

1. Components may be rescheduled within the same period.

2. Longitudinal welds are examined with the associated circumferential welds.

3. IWB, IWC, and IWD integral attachment weld examinations for Category B-K, C-C, and D-A items will be scheduled at the same time as the IWF component examination, to the extent practical.

4. The Code Category B-G-2 Item B7.80 used in the previous ISI Interval Plan for CRD Bolting inspections was removed from the Section XI 1995 Edition, 1996 Addenda. 10CFR50.55a requires examination of reused CRD Bolting to the 1995 Edition. Therefore, Code Category B-G-2 Item B7.80 has been reinstated in the schedule.

5. The Code Category B-G-2 Item B7.70 valve bolting is only scheduled when a Class 1 valve is required to be disassembled for maintenance or repair. (See Note (2) for Table IWB-2500-1, Categories B-G-2 and B-M-2)

6. The Code Category B-G-2 Item B6.180 pump bolting is only scheduled when a Class 1 pump is required to be disassembled for maintenance or 1.7-1

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 INTERVAL EXAMINATION PLAN repair. (See Note (3) for Table IWB-2500-1, Cat. B-G-2 and Note (2) for Table IWB-2500-1, Cat.B-L-2)

7. The letter designations used in the schedule columns are as follows:

s = scheduled to be examined c = examination was completed for Interval (credit is taken) b = multiple examination or re-occurring examination scheduled during interval B = multiple or re-occurring examination was completed p = item is a partially completed examination (further examination expected) e = expanded scope examination E = expanded scope examination was completed a = additional expanded scope examination A = additional expanded scope examination was completed r = item was rescheduled d = item was deferred until later in the Interval h = examination scheduled for successive periods (follow-up exam) H = examination scheduled for successive periods was completed I = (lower case L) examination was limited (less than code required percentage achievable) Pressure Testing Notes: A. The system leakage tests described by Section XI Category B-P, Item B15.10, B15.50, B.15.60 and B.15.70 are performed each refueling outage in accordance with Monticello procedures 0255-20-11C-1 and 0255-20-11C-2, except for the end of Interval test. The boundary is configured with all valves in the position required for normal reactor operation startup, however the VT-2 boundary extends to the 2nd closed valve. For the ten-year exam at the end of the Interval, 0255-20-11C-3 is used to configure the boundary instead of 0255-20-IIC-1. For that test, the systems are configured to pressurize all Class 1 pressure retaining components. All Class 1 pressure retaining components, including the Reactor Vessel, piping, pumps, and valves, within the Reactor Coolant Pressure Boundary (RCPB) are included in the examination boundary. Items B15.10, B15.50, B15.60, and B15.70 are all combined into a single summary number in the Plan under Item B15.10 for the RCPB and are not listed nor scheduled separately. The Class 1 systems included in these tests are Reactor Vessel (including vent and drain), Reactor Vessel Instrumentation, Reactor Recirculation, Main Steam, Feedwater, Core Spray, Residual Heat Removal, High Pressure Coolant Injection, Reactor Core Isolation Cooling, Standby Liquid Control, Reactor Water Cleanup, 1.7-2

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Control Rod Drive, and the Excess Flow Check Valves. Boundary Drawings for these systems are 1.5-2, -3, -4, -5, -6, -7, -8, -9, -10, -11, - 12, -17, -21, -23, -27 Record information of examination results is maintained by the Plant. B. The system leakage tests described by Section XI, Category C-H, Items C7.10, C7.30, C7.50 and C7.70 are performed in accordance with the Monticello procedure identified under the procedure columns of the Examination and Schedule table. All Class 2 pressure retaining components for Code Item numbers C7.10, C7.30, C7.50, and C7.70, including piping, pumps, valves, and vessels, are included within the examination boundary of the applicable procedure. Therefore, on a system basis, the components are all combined into a single summary number in the Plan under Code Item C7.10 for that system and are not listed nor scheduled separately. Record information of examination results is maintained by the Plant. C. (reserved - this note not currently in use) D. See Code Case N-498-4 (System Leakage Test in lieu of System Hydrostatic Test, Applicable to Class 3 only). E. See Code Case N-522 (Testing in accordance with Appendix J Program) F. The system leakage tests described by Section XI, Category D-B, Items D2.10, D2.20, D2.30, D2.40, D2.50, D2.60, D2.70, and D2.80 are performed in accordance with the Monticello procedure identified under the procedure columns of the Examination and Schedule table. All Class 3 pressure retaining components for Code Item numbers D2.10, D2.30, D2.50, and D2.70, including piping, pumps, valves, and vessels, are included within the examination boundary of the applicable procedure. Therefore, on a system basis, the components are all combined into a single summary number in the Plan for the System Leakage Test under Code Item D2.10 for that system and are not listed nor scheduled separately. Likewise, for Code Item numbers D2.20, D2.40, D2.60, and D2.80, they are combined under D2.20 for the System Hydrostatic Test. Record information of examination results is maintained by the Plant. 1.7-3

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN NDE Notes: AA. Reactor Pressure Closure Studs, Item B6.30, will be examined when removed near the end of the Interval. BB. (reserved - this note not currently in use) CC. Dissimilar metal weld. DD. Encapsulated weld. EE. The internal surface visual examination is done during maintenance: Atwood Morrill Globe Valve - Main Steam: VT-3 examination of one valve. FF. The internal surface visual examination is done during maintenance: Target Rock Relief Valves - Main Steam: VT-3 examination of one valve. GG. The internal surface visual examination is done during maintenance: Anchor Check Valves - Feedwater: VT-3 examination of one valve. HH. The internal surface visual examination is done during maintenance: Anchor Gate Valves - Various Systems: VT-3 examination of one valve in each size. II. The internal surface visual examination is done during maintenance: Chapman Crane Gate Valves - Recirculation System: VT-3 examination of one valve. JJ. Welds in core spray and containment spray were added at the end of the 2nd Interval per EGG-MS-8969. KK. (reserved - this note not currently in use) LL. (reserved - this note not currently in use) MM. Augmented examination for MEB 3-1 as amended by Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements (HELB Superpipe - requires 100% volumetric examination of welds in accordance with ASME Section IWA-2400 requirements once per 10 years). NN. (reserved - this note not currently in use) 1.7-4

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Damage Mechanism / Risk Categories for Risk-Informed ISI Item Description Item Description Number Number Elements Subject to High Cycle Elements Subject to Thermal Fatigue Category 1 - High Risk R1.12-1 Mechanical Fatigue Category 1 - High Risk R1.12-2 Elements Subject to High Cycle Ri. 11-2 Elements Subject to Thermal Fatigue Mechanical Fatigue Category 2 - High Risk Category 2 - High Risk Elements Subject to High Cycle R1.1 1-3 Elements Subject to Thermal Fatigue Mechanical Fatigue Category 3 - High Risk Category 3 - High Risk Subject to Thermal Fatigue R1. 12-4 Elements Subject to High Cycle R. 11-4 Elements Mechanical Fatigue Category 4 - Medium Risk Category 4 - Medium Risk 11-5 Elements Subject to Thermal Fatigue R1.12-5 Elements Subject to High Cycle Category 5a or b - Medium Risk Mechanical Fatigue Category 5a or b - Medium Risk R1.12-6 Elements Subject to High Cycle Ri. 11-6 Elements Subject to Thermal Fatigue Mechanical Fatigue Category 6a or b - Low Risk Category 6a or b - Low Risk Elements Subject to High Cycle Elements Subject to Thermal Fatigue R1.12-7 R1.11-7 Mechanical Fatigue Category 7a thru d - Low Risk Cateqory 7a thru d - Low Risk Elements Subject to Erosion Elements Subject to Crevice R1.14-1 R1.13-1 Cavitation Corrosion Cracking Cateqorv 1 - Hicqh Risk Cateqorv 1 - Hicqh Risk Elements Subject to Erosion R1.14-2 Elements Subject to Crevice R1.13-2 Cavitation Corrosion Cracking Category 2 - High Risk Category 2 - High Risk Elements Subject to Erosion Elements Subject to Crevice R1.13-3 Cavitation Corrosion Cracking Category 3 - High Risk Category 3 - High Risk Elements Subject to Erosion Elements Subject to Crevice R1.14-4 R1.13-4 Cavitation Corrosion Cracking Category 4 - Medium Risk Category 4 - Medium Risk Elements Subject to Erosion Elements Subject to Crevice Ri1.14-5 R1.13-5 Cavitation Corrosion Cracking Category 5a or b - Medium Risk Category 5a or b - Medium Risk Elements Subject to Erosion Elements Subject to Crevice R1.13-6 Cavitation Corrosion Cracking Category 6a or b - Low Risk Category 6a or b - Low Risk Elements Subject to Erosion Elements Subject to Crevice R1.14-7 R1.13-7 Cavitation Corrosion Cracking Cateqory 7a thru d - Low Risk Catelorv 7a thru d - Low Risk 1.7-5

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4 TH INTERVAL EXAMINATION PLAN Damage Mechanism / Risk Categories for Risk-Informed ISI Item Item Description Description Number Number Elements Subject to IGSCC or Ri. 15-1 Elements Subject to PWSCC R1.16-1 TGSCC Category 1 - High Risk Category 1 - High Risk Elements Subject to IGSCC or R1. 15-2 Elements Subject to PWSCC R1.16-2 TGSCC Category 2 - High Risk Category 2 - High Risk Elements Subject to IGSCC or Ri1.15-3 Elements Subject to PWSCC R1.16-3 TGSCC Category 3 - High Risk Category 3 - High Risk Elements Subject to IGSCC or Ri. 15-4 Elements Subject to PWSCC R1.16-4 TGSCC Category 4 - Medium Risk Category 4 - Medium Risk Elements Subject to IGSCC or RI1.15-5 Elements Subject to PWSCC R1.16-5 TGSCC Category 5a or b - Medium Risk Category 5a or b - Medium Risk Elements Subject to IGSCC or Ri1.15-6 Elements Subject to PWSCC R1.16-6 TGSCC Category 6a or b - Low Risk Category 6a or b - Low Risk Elements Subject to IGSCC or Elements Subject to PWSCC R1.15-7 R1.16-7 TGSCC Category 7a thru d - Low Risk Category 7a thru d - Low Risk Elements Subject to MIC or Pitting Elements Subject to FAC R1.17-1 R1.18-1 Category 1 - High Risk Category 1 - High Risk Elements Subject to MIC or R1.17-2 Pitting Elements Subject to FAC Category 2 - High Risk Category 2 - High Risk Elements Subject to MIC or R1.17-3 Pitting R1.18-3 Elements Subject to FAC Category 3 - High Risk Category 3 - High Risk Elements Subject to MIC or R1.18-4 Elements Subject to FAC R1.17-4 Pitting Category 4 - Medium Risk Category 4 - Medium Risk Elements Subject to MIC or R1.17-5 Pitting R1.18-5 Elements Subject to FAC Category 5a or b - Medium Risk Category 5a or b - Medium Risk Elements Subject to MIC or R1.17-6 Pitting R1.18-6 Elements Subject to FAC Category 6a or b - Low Risk Category 6a or b - Low Risk Elements Subject to MIC or R1.17-7 Pitting Elements Subject to FAC R1.18-7 Category 7a thru d - Low Risk Category 7a thru d - Low Risk 1.7-6

NORTHERN STATES POWER CO - MINNESOTA INSERVICE INSPECTION MONTICELLO 4' INTERVAL EXAMINATION PLAN Damage Mechanism / Risk Categories for Risk-Informed ISI Item Item Description Description Number Number Elements Not Subject to a Damage Ri. 19-1 Elements Subject to ECSCC R1.20-1 Mechanism Category 1 - High Risk Category 1 - High Risk Elements Not Subject to a Damage R1.19-2 Elements Subject to ECSCC R1.20-2 Mechanism Category 2 - High Risk Category 2 - High Risk Elements Not Subject to a Damage Ri. 19-3 Elements Subject to ECSCC R1.20-3 Mechanism Category 3 - High Risk Category 3 - High Risk Elements Not Subject to a Damage R1.19-4 Elements Subject to ECSCC R1.20-4 Mechanism Category 4 - Medium Risk Category 4 - Medium Risk Elements Not Subject to a Damage R1.19-5 Elements Subject to ECSCC Category 5a or b - Medium Risk R1.20-5 Mechanism Category 5a or b - Medium Risk Elements Not Subject to a Damage Ri1.19-6 Elements Subject to ECSCC R1.20-6 Mechanism Category 6a or b - Low Risk Category 6a or b - Low Risk Elements Not Subject to a Damage R1.19-7 Elements Subject to ECSCC Mechanism R1.20-7 Category 7a thru d - Low Risk Category 7a thru d - Low Risk 1.7-7

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 1 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 oq 0 r- o" 0 Category, DwgIlSO No. Item No., Comp. Desc.

                                                                                                           ~-r.4   NJ N~               ox    w uL uL                                         LI.

Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-A 102637 ISI Fig 4 ISI B1.11 VCBA-2 Circ Weld AUG 1 Reactor Vesse OWN PRE B-A 102636 Ml1_4-PlRF22 / IS[ / UT / / PEI- ISI Fig 4 ISI B----------- s - - - B1.11 VCBB-1 02.03.15 B. HeadNesse AUG 1 Reactor Vesse Ml_14-P3_RF25 / ISI / UT/ / FP-PE- OWN NDE-406 PRE B-A 102639 ISI Fig 4 IS - B1.11 VCBB-3 Circ Weld AUG 1 Reactor Vesse OWN PRE B-A 102640 ISI Fig 4 ISI B1. 11 VC BB-4 C irc We ld @ N4 C A UG . . . . . . . . . . . . . . 1 Reactor Vesse OWN . . . ... . . ... . . . PRE B-A 105005 ISI Fig 4 IS - B1.11 VCBB-4 Circ Weld @ N4B AUG 1 Reactor Vesse OWN PRE B-A 102642 ISI Fig 4 ISI----------------- - - - - 61.12 VLAA-1 M1_I4-P3_RF25 / ISI / UTI / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102643 M1_I4-P2_RF23 / ISI / UT / / PEI- ISI Fig 4 ISI - p - - - - - - 61.12 VLAA-2 02.03.15 Long Seam AUG 1 Reactor Vesse Ml_14-P3_RF25 / ISI / IUT/ / FP-PE- OWN NDE-406 PRE B-A 102644 ISI Fig 4 IS[ .- ..---------- - - - - B1.12 VLBA-1 Ml_14-P3_RF25 / ISI / UT/ / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE . . . . . . . . . . . . . . B-A 102645 ISI Fig 4 ISI S B1.12 VLBA-2 M1_14-P3_RF25 / ISI / UT/ / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 2 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M. W) o, o= oN N C. o o 0 N N N N Category, DwgIlSO No.

                                                                                                                .- N                     U)

Item No., Comp. Desc. LL N w w w, LLw, LL of Class Summary NoJCompID/System Scope I Method I Procedure Code Case B-A 102646 ISI Fig 4 ISI----------------- s - - - B1.12 VLCB-1 M1_14-P3_RF25 / ISI / UT / / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102647 ISI Fig 4 ISI----------------- s - - - B1.12 VLCB-2 M1_I4-P3_RF25 / ISI / UT/ / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102648 ISI Fig 4 ISI----------------- s - - - B1.12 VLDB-1 M1_I4-P3_RF25 / ISI / UT/ / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102649 ISI Fig 4 ISI----------------- s - - - B1.12 VLDB-2 M1_I4-P3_RF25 / IS[ / UT/ / FP-PE- Long Seam AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102380 ISI Fig I SI------------------ s - - - B1.21 W-1 M1_14-P3_RF25 / ISI / IUT / FP-PE- T.H. Circ. Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102627 ISI Fig 3 ISI----------------- s - - - B1.21 W-3 M1_I4-P3_RF25 / ISI / UT / / FP-PE- B.H. Circ Weld AUG 1 Reactor Vesse NDE-406 OW N - - . . . . PRE B-A 102381 ISI Fig I IS1 ---------------- s - - - B1.22 W-2 Ml_i4-P3_RF25 / ISI / UT/ / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102382 ISI Fig 1 SI------------------ s - - - B13.22 W-3 M1_14-P3_RF25 / ISI / UT / / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102383 ISI Fig 1 ISI S B1.22 W-4 M1_14-P3_RF25 / ISI / UT / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 3 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 r- 0) a C- N Category, Dwg/ISO No. N N Itn Item No., Comp. Desc. U. U-N. w w w-Class Summary No.IComplD/System Scope I Method I Procedure Code Case B-A 102384 ISI Fig 1 ISI--------- - - S - - - B1.22 W-5 M1_I4-P3_RF25 / ISI / UT/ / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE.406 OWN iP RE B-A 102385 ISi Fig 1 ISI------------- -- s - - - B1.22 W-6 M1_I4-P3_RF25 / ISI / UT / / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102386 ISI Fig 1 ISi----------------- s - - - B13.22 W-7 M1_I4-P3_RF25 / ISI / UT/ / FP-PE- T.H. Meridional Weld AUG 1 Reactor Vesse NDE406 OWN PRE B-A 102625 ISI Fig 3 ISI----------------- s - - - B1.22 W-1 Ml_14-P3_RF25 / ISI / UT/ / FP-PE- B.H. Dollar PI L.S, AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102626 ISI Fig 3 ISI----------------- s - - - 81.22 W-2 MI 14-P3_RF25 / ISI / UT / / FP-PE- B.H. Dollar PI L.S. AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102628 ISI Fig 3 ISI----------------- s - - - 81.22 W-4 M1_14-P3_RF25 / ISI / UT / / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102629 ISI Fig 3 ISI----------------- s - - - 81.22 W-5 M1_14-P3_RF25 / ISI / UT/ / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE.-406 OWN PRE B-A 102630 ISI Fig 3 ISI----------------- s - - - 81.22 W-6 Ml_14-P3_RF25 / ISI / UT/ / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102631 II Fig 3 ISI S 81.22 W-7 M1_I4-P3_RF25 / ISI / UT / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 4 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 N Category, DwgIlSO No. LL o4 o o w cm N N U- O4 Item No., Comp. Desc. U. IL w w U-nL Class Summary No.lCompID/System Scope I Method I Procedure Code Case B-A 102632 ISI Fig 3 SI- ---------------- s - - - B1.22 W-8 M1_4-P3_RF25 / ISI / UT/ / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102633 ISI Fig 3 SI----------------- s - - - B13.22 W-9 M1_I4-P3_RF25 / ISI / UT / / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102634 ISI Fig 3 ISI----------------- s - - - B1.22 W-1 0 M1_I4-P3_RF25 / ISI / UT / / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102635 ISI Fig 3 SI----------------- s - - - B13.22 W-1 1 M1_I4-P3_RF25 / ISI / UT / / FP-PE- B.H. Meridional Weld AUG 1 Reactor Vesse NDE-406 OWN PRE B-A 102641 ISI Fig 4 ISI r d - - - d s - - - B1.30 VCBC-5 M1_i4-P3_RF25 / ISI / UT / / UT- Vessel/Flange Weld AUG 1 Reactor Vesse Vendor N-623,N-623,N-623,N-623 OWN PRE B-A 102387 Ml_14-P3_RF25 / ISI/MT/ / PEI- ISI Fig 1 ISI r d - r d - - s - - - 11.40 W-8 02.02.01 T.H. Flange Weld AUG 1 Reactor Vesse M1_I4-P3 RF25 / ISI / UT / / FP-PE- N-623,N-623,N-623,N-623,N-623 OWN NDE-406 PRE B-D 102374 ISI Fig 1 ISI c B3.100 N-6AIR M1_i4-P2_RF231/SI/UTI/ FP-PE- N-6AInnerRadius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102376 ISI Fig 1 ISI----------- c - - r - - - B3.100 N-6B IR M1_I4-P2_RF2411SIIUT / FP-PE- N-6B InnerRadius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102378 ISI Fig 1 ISI c B3.100 N- 7 IR M1_ 4-P2_RF24 / ISI UT/ LI FP-PE- N- 7 Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 5 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Ua N 0 C., Category, DwglISO No. 0L 0L N N w 0 e'L c, Item No., Comp. Desc. U,, e': w. Class Summary No.lComplDISystem Scope I Method I Procedure Code Case B-D 102622 ISI Fig 3 ISI- ----- c r - - - B3.100 N-10 IR Ml1_I4-P2_RF24 / ISI / UT/ / FP-PE- N-10 Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102651 ISI Fig 5 ISIC B3.100 N- 1A IR M1_I4-PlRF22 / ISI / UT / / PEI- N- 1A Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102653 ISI Fig 5 ISI-- ------- s - - - B3.100 N- 1B IR Ml_14-P3_RF25 ISI I UT/I /FP-PE- N-1B Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE .. . . ... . . ... . . . B-D 102655 ISI Fig 5 ISI ---------- c - r - - - B3.100 N-2AIR M1_I4-P2_RF241ISI/UTI/ FP-PE- N-2AInnerRadius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102657 ISI Fig 5 ISI - - - c r - - - B3.100 N-2B IR M1_I4-P2_RF23/ISI/UTI /FP-PE- N-2B InnerRadius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102659 ISI Fig 5 ISI ---------------- s - - - B3.100 N- 2C IR Ml_14-P3_RF25 / ISI / UT/ / FP-PE- N- 2C Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102661 ISI Fig 5 ISI r c B3.100 N- 2D IR Ml_14-PIRF22 / ISI / UT / / PEI- N- 2D Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102663 ISI Fig 5 ISI r c B3.100 N- 2E IR M1_14-PlRF22 / ISI / UT / / PEI- N- 2E Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102665 I1I Fig 5 iSI S B3.100 N- 2F IR M1_I4-P3_RF25 / ISI / UT I FP-PE- N- 2F Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 6 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3

                                                                                                                       'o   Lo o     0             o      3 o    0             0    0 N    N Category,                                                                          Dwg/ISO No.                         T-   N NN Item No.,                                                                          Comp. Desc.                         LL   LL N

U. N IL LL w Class Summary No.lComplD/System Scope I Method I Procedure Code Case B-D 102667 ISI Fig 5 IS- - - c r B3.100 N- 2G IR M1_14-P2_RF23 / ISI / UT / / FP-PE- N- 2G Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102669 ISI Fig 5 ISI ---------- r s - - - B3.100 N- 2H IR M1_14-P3_RF25 /1SI/UT/ / FP-PE- N- 2H Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102671 ISI Fig 5 ISI r c B3 .10 0 N- 2 J IR M1 _I4-P l _R F2 2 /IS I / UT / / P E I- N- 2 J Inn e r R a d ius A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.03.17 OWN PRE B-D 102673 ISI Fig 5 Is[ ---------- r s - - - B3.100 N- 2K IR Ml_14-P3_RF25 / ISI / UT / / FP-PE- N- 2K Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102675 ISI Fig 5 ISI r c B3.100 N- 3A IR M1_4-PlRF22 / ISI / UT / / PEI- N- 3A Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102677 ISI Fig 5 ISI----------------- s - - - B3.100 N- 3B IR M1_I4-P3_RF25 /IiSI / UT / / FP-PE- N- 3B Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102679 ISI Fig 5 ISI----------- c B3.100 N- 3C IR M1_14-P2_RF24 / ISI / UT / / FP-PE- N- 3C Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102681 ISI Fig 5 ISI----------------- s - - - B3.100 N- 3D IR M1_I4-P3_RF25 I ISI / UT // FP-PE- N- 3D Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102683 ISI Fig 5 ISI c B3.100 N-4A IR M1_14-P2_RF23 / ISI / UT I I FP-PE- N- 4A Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE Printed 6/510

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 7 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M~ ILa o 0 1 0 0 o 0 Category, DwgIISO No. LO*

                                                                                                                         - '    Z                                C14 Item No.,                                                                                Comp. Desc.                      N Li.

CI LL N Li. N U. Class Summary No.IComplDISystem Scope I Method I Procedure Code Case w w B.D 102685 ISI Fig 5 ISI- -c ------ B3 .10 0 N- 4 B IR M l_14 -P 2 _R F 2 4 / IS I / UT / / F P-P E - N- 4 B Inn e r R a d ius AUG . . . . . . . . . . . . . . 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102687 ISI Fig 5 ISI r c B3.100 N- 4C IR M1_4-PlRF22 /1SI/UT / / PEI- N- 4C Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102689 ISI Fig 5 ISI ---------------- s - - - B3.100 N- 4D IR M1_I4-P3_RF25 / ISI / UT / / FP-PE- N- 4D Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102691 ISI Fig 5 ISI ---------------- s - - - B3.100 N- 5A IR M1_14-P3_RF25 / ISI / UT / / FP-PE- N- 5A Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102693 ISI Fig 5 ISI r c B3.100 N- 5B IR M1_14-P1_RF22 / ISI / UT / / PEI- N- 5B Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102695 ISI Fig 5 Is] r c B3.100 N- 8A IR Ml_14-PlRF221/SI/UT/ /PEI- N- 8A Inner Radius AUG 1 Reactor Vesse 02.03.17 OWN PRE B-D 102697 ISI Fig 5 ISI ---------- c - r - - - B3.100 N- 8B IR Ml_14-P2_RF24 / IS/ UT/ / FP-PE- N- 8B Inner Radius AUG 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102699 ISI Fig 5 ISI - - - c B3 .10 0 N- 9 IR Ml _14-P 2_ R F 2 3 1/S I / UT I / F P-P E - N- 9 Inn e r R a d ius A UG . . . . . . . . . . . . . . 1 Reactor Vesse NDE-UT-03 OWN PRE B-D 102375 1S I-ig 1 1*51 C B3.90 N- 6A NV M1_I4-P2_RF23/ ISI / UT I I PEI- N- 6A Noz/Head Weld AUG 1 Reactor Vesse 02.03.15 N-613-1 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 8 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 o CD o 0 N e4 04 Category, DwgIlSO No. C14 C14 o CM (1 U.4 In Item No., Comp. Desc. L1 N

w. It*

LL4 w-Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-D 102377 ISI Fig 1 ISI----------- c - - r - - - B3.90 N- 6B NV M1_ 4-P2_RF24 / ISI / UT / / FP-PE- N- 6B NozJHead Weld AUG 1 Reactor Vesse NDE-406 N-613-1,N-613-1 OWN PRE B-D 102379 ISI Fig 1 ISI----------- c B3.90 N- 7 NV M1_I4-P2_RF24 / ISI / UT / / FP-PE- N- 7 Noz/Head Weld AUG 1 Reactor Vesse NDE-406 N-613-1 OWN PRE B-D 102623 ISI Fig 3 ISI----------- c r - - - B3.90 N-10 NV M1_I4-P2_RF24 / ISI UT/ /FP-PE- N-10 NozNsl Weld AUG 1 Reactor Vesse NDE-406 N-613-1,N-613-1 OWN PRE B-D 102652 ISI Fig 5 ISI r c B3.90 N- 1A NV MI 14-P1_RF22/ISIIUT/ /PEI- N- 1A Vsl/Noz Weld AUG 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN PRE B-D 102654 ISI Fig 5 SI----------------- s - - - B3.90 N- 1B NV M1_I4-P3_RF25 / ISI /UT / / FP-PE- N- 1B Vsl/Noz Weld AUG 1 Reactor Vesse NDE-406 N-613-1 OWN PRE B-D 102656 ISI Fig 5 ISI----------- c - r - - - B3.90 N- 2A NV M1_14-P2_RF24 / ISI / UT / / FP-PE- N- 2A Noz/Vsl Weld AUG 1 Reactor Vesse NDE-406 N-613-1,N-613-1 OWN PRE B-D 102658 ISI Fig5 SI - - - - c - - - r - - - B3.90 N- 2B NV Ml_14-P2_RF23 11/SIUT/I PEI- N- 2B Noz / Vsl Weld AUG .............. 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102660 ISI Fig 5 SI------------------ S - - - B3.90 N- 2C NV Ml_14-P3_RF25 / ISI / UT / / FP-PE- N- 2C NozNsl Weld AUG 1 Reactor Vesse NDE-406 N-613-1 OWN PRE B-D 102662 ISI Fig 5 Is' r c B3.90 N- 2D NV M1_14-PlRF22 / ISI / UT/ / PEI- N- 2D Noz/Vsl Weld AUG 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN PRE Printed 6

0 Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 9 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o* o o= o o o o o N N Category, DwgllSO No. N4 C4 N C4 N4 Item No., Comp. Desc. U. N Iu. N U. N U. N U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-D 102664 ISI Fig 5 ISI r c B3.90 N- 2E NV M1_I4-PIRF22 / SI/UT / / PEI- N- 2E Noz/Vsl Weld AUG 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN PRE B-D 102666 ISI Fig 5 SI----------------- s - - - B3.90 N- 2F NV M1_4-P3_RF25 / ISI / UT / / FP-PE- N- 2F Noz/Vsl Weld AUG 1 Reactor Vesse NDE-406 N-613-1 OWN PRE B-D 102668 ISI Fig5 I - - - c r B3.90 N- 2G NV M1 14-P2_RF23 / ISI / UT / / PEI- N- 2G NozNsI Weld AUG 1 ReactorVesse 02.03.15 N-613-1,N-613-1 OWN PRE B-D 102670 ISI Fig 5 ISI ---------- r s - - - B3.90 N- 2H NV Ml 14-P 3_R F25 / S I I UT I / FP -P E- N- 2 H Noz/V sl We ld A UG . . . . . . . . . . . . . . 1 Reactor Vesse NDE-406 N-613-1,N-613-1 OWN PRE B-D 102672 ISI Fig 5 ISI r c B3.90 N-2J NV M1_4-PlRF22/ISl/UT / PEI- N-2J Noz/Vsl Weld AUG 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN PRE B-D 102674 ISI Fig 5 SI- ---------- r s - - - B3.90 N- 2K NV M1_I4-P3_RF25 /ISI/ UT / / FP-PE- N- 2K NozNsl Weld AUG 1 Reactor Vesse NDE-406 N-613-1,N-613-1 OWN PRE B-D 102676 ISI Fig 5 ISl r c B3.90 N- 3A NV Ml _14-PiRF22 / ISI / UT / / PEI- N- 3A Vsl/Noz Weld AUG 1 R e a c to r V e ss e 0 2 .0 3. 15 N-6 13 -1,N-6 13 -1 O WN . . . . . . . . . . . . . . PRE B-D 102678 ISI Fig 5 ISI----------------- s - - - B3.90 N- 3B NV M1_I4-P3_RF25 / ISI / UT / / FP-PE- N- 3B Vsl/Noz Weld AUG 1 Reactor Vesse NDE-406 N-613-1 OWN PRE B-D 102680 M1_I4-P2_RF24 / ISI / UT/ / FP-PE- ISI Fig 5 ISI c B3.90 N- 3C NV NDE-406 N- 3C Noz/Vsl Weld AUG 1 Reactor Vesse MlI4-P2_RF24 / ISI / UT / / PEI- N-613-1 OWN 02.03.16 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 10 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00 0 0 0 Category, N' I N4I *I N C4 1 DwgIISO No. LN Item No., Comp. Desc. - 4 C4 C-LI. L w* o: . U. LL w* w w0 Class Summary No.lComplD/System Scope I Method I Procedure Code Case B-D 102682 ISI Fig 5 ISI -- -------------- s - - - B3.90 N- 3D NV M1_l4-P3_RF25 / ISI / UT / / FP-PE- N- 3D VsVNoz Weld AUG . . . ... . . ... . . . 1 Reactor Vesse NDE-406 N-613-1 OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-D 102684 ISI Fig 5 ISI - - - - c B3.90 N- 4A NV M1_I4-P2_RF23 / ISI / UT / / PEI- N- 4A NozNsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse 02.03.15 N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102686 ISI Fig 5 ISI---------- c--------- B3.90 N- 4B NV Ml_14-P2_RF24 /1SI/UT/I /FP-PE- N- 4B Noz/Vsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse NDE-406 N-613-1 OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-D 102688 ISI Fig 5 ISI r c . ... . . ... . . . B3.90 N- 4C NV M1_I4-P 1RF22 /ISI/UT/I /PEI- N- 4C Noz/Vsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102690 ISI Fig 5 ISI -- ---- ---- ------ S - - - B3.90 N- 4D NV MI_14-P3_RF25 / ISI / UT / / FP-PE- N- 4D Noz/Vsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse NDE-406 N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102692 ISI Fig 5 ISI ---- ------------ s - - - B3.90 N- 5A NV M1_I4-P3_RF25 / ISI / UT / / FP-PE- N- 5A NozNsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse NDE-406 N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102694 ISI Fig 5 ISI r c . ... . . ... . . . B3.90 N-5B NV M1_14-PIRF221/S[IUT/ /PEI- N-58 NozA/sl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN . . . ... . . ... . . . P RE . . . . . . . . . . . . . . B-D 102696 ISI Fig 5 ISI r c . ... . . ... . . . B3.90 N- 8A NV M1_14-PlRF221/SI/UT/ /PEI- N- 8A NozNsl Weld AUG . . . ... . . ... . . . 1 Reactor Vesse 02.03.15 N-613-1,N-613-1 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-D 102698 ISI Fig 5 ISI c r B3.90 N- 8B NV M1_I4-P2_RF24 / ISI / UT/ /FP-PE- N- 8B Vsl/Noz Weld AUG 1 Reactor Vesse N-613-1,N-613-1 NDE-406 OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 11 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o a 0 N N N Category, DwgIlSO No. NI W um N N' Item No., Comp. Desc. N_ L NN uL uL U-w~ w Class Summary No.IComplDISystem Scope I Method / Procedure Code Case B-D 102700 ISI Fig 5 ISI - - - - c B3.90 N- 9 NV M1_I4-P2_RF23 / ISI / UT / / PEI- N- 9 Ves/Noz Weld AUG 1 Reactor Vesse 02.03.15 N-613-1 OWN PRE B-G-1 102710 ISIFig 6 ISI r d - - r d - - s - - - B6.10 Nuts M1_I4-P3_RF25 / ISI /VT / / FP-PE- 64 RPV Nuts AUG 1 Reactor Vesse NDE-510 OWN PRE B-G-1 101971 ISI-97005-A ISI r c B6.180 B-2 RC Pump A Studs, Nuts AUG 1 Recirculation A N-307-2,N-307-2 OWN PRE - c B-G-1 102076 M1_ 4-P2_RF23 / ISI / UT/ / PEI- ISI-97006-A ISI r r - - c B6.180 B-2 02.03.06 RC Pump B Studs, Nuts AUG 1 Recirculation B M1_14-P2_RF23 / ISI / VT / / PEI- N-307-2,N-307-2,N-307-2 OWN 02.05.01 M1_14-P2_RF23 / PSI /VT IPEI-02.05.01 PRE - - - - c B-G-1 101992 ISI-97005-A ISI - c B6.190 P200-A M1_I4-P1_RF22 / ISI V-/ / PEI- Flange Surface AUG 1 Recirculation A 02.05.01 OWN PRE B-G-1 107184 ISI-97006-A ISI - - - - c B6.190 P200-B Ml_14-P2_RF23 / ISI / VT / PEI- Flange Surface AUG 1 Recirculation B 02.05.01 OWN PRE B-G-1 102711 ISI Fig 6 ISI r d - - r d - - s - - - B6.20 Studs 64 RPV Studs, In-Place AUG M1_I4-P3_RF25 / ISI / UT/ /FP-PE-1 Reactor Vessel NDE-408 N-307-2,N-307-2,N-307-2,N-307-2,N-307-2 OWN PRE B-G-1 102712 I1I t-ig 6 AUG r r r r - 5s 64 RPV Studs, Removec AUG B6.30 Studs 1 Reactor Vessel M1_14-P3_RF25 / ISI / / / N-307-2,N-307-2,N-307-2,N-307-2,N-307-2 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 12 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 m 0 oý 0 o 0 N o N Category, DwgIlSO No. N C-4N IN Item No., Comp. Desc. w N .LL w U- U, Nx U-Class Summary No./ComplD/System Scope I Method I Procedure Code Case B-G-1 102709 ISI Fig6 ISI r d - r d s - - - B6.40 Threads in Flange M1_I4-P3_RF25 ISI / UT / / PEI- Threads in Flange AUG 1 Reactor Vesse 02.03.07 OWN PRE B-G-1 102708 ISI Fig 6 ISI r d - - r d - - s - - - B6.50 Bushings M1_ 4-P3_RF25 ISI / VT /FP-PE- 64 Bushings AUG 1 Reactor Vesse NDE-510 OWN . . . ... . . ... . . . PRE B-G-1 102713 Ml 14-P2_RF24/PSI/VT/VT-1 /FPISI Fig6 ISI r d - r d - s B6.50 Washers PE-NDE-510 64 Washers AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI / VT / VT-1 / FP- OWN PE-NDE-510 PRE-------- c--------c B-G-2 100395 ISI Fig I ISI ---------------- s - - - B7.10 B-2 M1-14-P1_RF21 / AUG / PT /PEI- N-6A Flange Bolts AUG 1 Rx Ves Head Coolinc 02.01.01 OWN Ml-14-PlRF21 /PSI/ VTI/ PEI-02.05.01 Ml-14-PlRF21 /PSI/ VT I PEI-02.05.01 Ml_14-P3_RF25 / ISIIVT / / FP-PE-NDE-510 PRE p B-G-2 102373 ISI Fig 1 ISI r c B7.10 B- I MI_14-Pl-RF22 / ISI /VT / / PEI- N-6B Flange Bolts AUG 1 Top Head 02.05.01 OWN PRE B-G-2 100566 ISI-13142-33-A ISI c B7.50 B- 1 M1_I4-PlRF21 I IS[ /VT/ / PEI- Flange Bolts @ W-12 AUG 1 Main Steam A 02.05.01 OWN PRE B-G-2 100567 ISI-13142-3I3-A [SI C B7.50 B-2 Ml_14-P2_RF23 / ISI / VT// PEI- Flange Bolts @ W-14 AUG 1 Main Steam A 02.05.01 OWN PRE Printed /

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 13 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o o o e o o o a 0 N N N N4 N Category, DwgIISO No. -, NI U) Item No., Comp. Desc. IL N LL N L. Nm Class w1 ILwg Ix It U-Summary No.IComplDISystem Scope I Method I Procedure Code Case B-G-2 100568 M1_I4-PlRF21 / ISI/VT/I PEI- ISI-13142-33-A SI C B7.50 B- 3 02.05.01 Flange Bolts @ W-16 AUG 1 Main Steam A M1_I4-P2_RF23 /PSI/ VT /PEI- OWN 02.05.01 PRE - - - C B-G-2 100569 IS1-13142-33-A IS1----------------- s B7.50 B-4 M1_14-P3 RF25 / ISI/VT/I FP-PE- Flange Bolts @ W-18 AUG 1 Main Steam A NDE-510 OWN PRE B-G-2 100570 IS1-13142-33-A ISI ---------------- s - - B7.50 B- 5 M1_14-P3_RF25/ ISI VTI FP-PE- Flange Bolts @ W-20 AUG 1 Main Steam A NDE-510 OWN PRE B-G-2 100571 IS1-13142-33-A ISI----------------- s - - - 87.50 8-6 M14-P3_RF25 / ISI/VT / / FP-PE- Flange Bolts @ W-22 AUG 1 Main Steam A NDE-510 OWN PRE B-G-2 100621 Ml-14-PlRF21 / ISI / VT/ / PEI- ISI-13142-34-A ISI c B7.50 B- 1 02.05.01 Flange Bolts @ W-13 AUG 1 Main Steam B Ml_14-PlRF22 / PSI /VT// PEI- OWN 02.05.01 PRE - c . . . . . . . . . . . B-G-2 100622 ISI-13142-34-A ISI - - - - c B7.50 B-2 M1_I4-P2_RF23 /1SI/VT/I PEI- Flange Bolts @ W-15 AUG 1 Main Steam B 02.05.01 OWN PRE B-G-2 100623 Ml_14-P1_RF21 / ISI/VT / PEI- ISI-13142-34-A ISI c B7.50 8- 3 02.05.01 Flange Bolts @ W-16 AUG 1 Main Steam E M1_I4-PlRF22 / PSI / VT/ PEI- OWN 02.05.01 PRE c B-G-2 100674 ISI-13142-35-A ISI C B7.50 B- 3 Ml_14-P2_RF24 / ISI/VT / / FP-PE- Flange Bolts @ W-16 AUG 1 Main Steam C OWN NDE-510 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 14 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 0 0 0 0 0 0CIAl 0C1 0 C1410 N* Category, Dwg/ISO No.

                                                                                                                 -     N          C,)   V              UI Item No.,                                                                       Comp. Desc.                      N U.

N1 L. N U. N LL

                                                                                                                                                        .11 LL Class       Summary No./ComplDISystem     Scope I Method I Procedure            Code Case B-G-2       100675                                                              ISI-13142-35-A      ISI            c B7.50       B- 4                          M1_I4-PlRF21 / ISI /VT/ / PEI-        Flange Bolts @ W-18 AUG 1           Main Steam C                  02.05.01                                                  OWN PRE B-G-2       100676                                                              IS1-13142-35-A      ISI            c B7.50       B- 5                          Mi_14-P1_RF21 / IS[ / VT / / PEI-     Flange Bolts @ W-14 AUG 1           Main Steam C                  02.05.01                                                  OWN PRE B-G-2       100725                                                              IS1-13142-36-A      ISI            c B7.50       B- 1                          M1_I4-PlRF21 / ISI /VT/    / PEI-     Flange Bolts @ W-13 AUG 1           Main Steam C                  02.05.01                                                  OWN PRE B-G-2       100726                                                              ISI-13142-36-A      ISI-------------------               c B7.50       B- 2                          M1_14-P2_RF24 / ISIVT / / FP-PE- Flange Bolts @ W-15      AUG 1           Main Steam C                  NDE-510                                                   OWN PRE B-G-2       100727                        M1_I4-PlRF21 / ISI/VT / I PEI-        ISI-13142-36-A      ISI            c       -   - -        B B7.50       B-S3                          02.05.01                              Flange Bolts @ W-17 AUG 1           Main Steam C                  M1_I4-P2 RF24 / ISI / VT / VT-i / FP-                     OWN PE-NDE-510 Mi_14-P2_RF24 / IS[ / VT / VT-1 / FP-PE-NDE-510 PRE B-G-2       100728                                                              ISI-13142-36-A      ISI--------                           c-----------

B7.50 B- 4 M1_14-P2_RF24 / ISI/VT / / FP-PE- Flange Bolts @ W-19 AUG 1 Main Steam C NDE-510 OWN PRE B-G-2 100729 ISI-13142-36-A ISI ---------------- s B7.50 B- 5 M1_I4-P3_RF25 / ISI / VT / / FP-PE- Flange Bolts @ W-21 AUG 1 Main Steam C NDE-510 OWN PRE B-G-2 100730 ISI-13142-36-A ISI s B7.50 B- 6 M1_14-P3_RF25 / ISI / VT / / FP-PE- Flange Bolts @ W-23 AUG 1 Main Steam C NDE-510 OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 15 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00C 00C 0 0N* 0C14 0 N 0C Category, DwgIlSO No.

                                                                                                            .-e I  C.1 U.IX        eco -tX Item No.,                                                                     Comp. Desc.                     NC e'4 L.U- w.
                                                                                                                    %4 LL N           U.

CIA LL Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-G-2 101392 ISI-74209-1A ISI----------------- s - - - B7.50 B- 1 M1_I4-P3_RF25 / ISI /VTI / FP-PE- Flange Bolts AUG 1 Recirc A Drain Line NDE-510 OWN PRE B-G-2 101408 ISI-74210-1A ISI c r B7.50 B- 1 MlI14-P2_RF23 / ISI /VT/ / PEI- Flange Bolts AUG 1 Recirc B Drain Line 02.05.01 OWN PRE B-G-2 101491 M1_-4-PlRF21 /PSI// VT / / PEI- ISI-782A-A -------- c------ B7.50 B- 1 02.05.01 Flange Bolts AUG 1 Head Vent Ml14-PlRF22 / PSI /VT/ / PEI- OWN 02.05.01 M1 14-P2_RF24 / ISI IVTI IFP-PE-NDE-510 PRE c B . .. . . . .. . . . . B-G-2 101492 ISI-782A-A ISI r c B7.50 B- 2 Ml_14-P2_RF23 / PSI / VT/ IPEI- Flange Bolts AUG 1 Head Vent 02.05.01 OWN PRE c B B B-G-2 101616 ISI-821A ISI--- ----- s - B7.50 B- 1 M1_I4-P3_RF25 / ISI /VT// FP-PE- Flange Bolts AUG 1 Bottom Head Drair NDE-510 OWN

                                                        -                  10PRE B-G-2       101850                                                            ISI-97003-A         ISI------------------- c B7.50       B- 3                          M1_I4-P2_RF24 / ISI / VT / / FP-PE- Flange Bolts @ W 11 AUG 1           RHR Return A                  NDE-510                                                 OWN PRE B-G-2       101851                                                            ISI-97003-A         ISI-----------------                           s    -   -  -

B7.50 B- 4 M1_14-P3_RF25 / ISI / VT / / FP-PE- Flange Bolts @ W 22 AUG 1 RHR Return A NDE-510 OWN PRE B-G-2 101891 ISI-97003-B ISI C B7.50 B- 2 M1_4-P2_RF24 / ISI / VT I I FP-PE- Flange Bolts @ W-9 AUG 1 RHR Suction A NDE-510 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 16 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o a o a 0 Category, N DwgIlSO No. N N Item No., Comp. Desc. U. U. It N* w w Class Summary No./ComplDlSystem Scope I Method I Procedure Code Case B-G-2 101892 ISI-97003-B IS1- ---------------- s - - - B7.50 B- 3 M1_I4-P3_RF25 / ISI/VT// FP-PE- Flange Bolts @ W-14 AUG 1 RHR Suction A NDE-510 OWN PRE B-G-2 101930 ISI-97004-A ISI----------- c B7.50 B- 3 M1_I4-P2_RF24 / ISI / VT / / FP-PE- Flange Bolts AUG 1 RHR Return B NDE-510 OWN PRE B-G-2 101931 ISI-97004-A SI------------------ s - - - B7.50 B- 4 Ml_14-P3_RF25 /ISI / VT / / FP-PE- Flange Bolts AUG 1 RHR Return B NDE-510 OWN PRE B-G-2 101972 ISI-97005-A ISI----------- c - r - - - B7.50 B- 3 M1_14-P2_RF24 / ISI / VT / VT-1 / FP- Flange Bolts @ W-17 AUG 1 Recirculation A PE-NDE-510 OWN PRE B-G-2 102077 ISI-97006-A ISI c B7.50 B- 3 M1_I4-PlRF21 ISI / VT/ / PEI- Flange Bolts @ W-16 AUG 1 Recirculation B 02.05.01 OWN PRE B-G-2 100424 ISI-13142-26-A ISI B7.70 B- 1 Valve Bolts AUG 1 Core Spray B OWN PRE . . . ... . . ... . . . B-G-2 100425 ISI-13142-26-A ISI B7.70 B- 2 Valve Bolts AUG . . . ... . . ... . . . 1 Core Spray B OWN PRE . . . ... . . ... . . . B-G-2 100426 ISI-13142-26-A ISI Valve Bolts AUG B7.70 B- 3 1 Core Spray B OWN . . . ... . . ... . . . PRE B-G-2 100492 ISI-13142-31-A ISI B7.70 B- 1 Valve Bolts AUG 1 Core Spray , OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 17 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o, O CN 0 N N Category, Dwg/lSO No. N Item No., Comp. Desc. N Ii. u. U. N uL N LL LL w-Class Summary No.ICompID/System Scope I Method I Procedure Code Case B-G-2 100493 ISI-13142-31-A ISI B7.70 B-2 Valve Bolts AUG 1 Core Spray A OWN PRE B-G-2 100494 IS1-13142-31-A ISI B7.70 B-3 Valve Bolts AUG 1 Core Spray A OWN PRE B-G-2 100572 ISI-13142-33-A ISI C B7.70 8-7 M1 14-P1_RF21 IPSIIVTIVT-1 Valve Bolts AUG 1 Main Steam A PEI-02.05.01 OWN PRE p ISI-13142-33-A ISI B-G-2 100573 B7.70 B- 8 Valve Bolts AUG 1 Main Steam A OWN PRE B-G-2 100574 ISI-13142-33-A ISI B7.70 8- 9 Ml_14-P2_RF23 / PSI / VT / PEI- Valve Bolts AUG 1 Main Steam P 02.05.01 OWN PRE ---- IS1-13142-33-A ISI B-G-2 100575 M1_14-PlRF21 /PSI/VTI /PEI-B7.70 B-10 02.05.01 Valve Bolts AUG 1 Main Steam P Ml-14-P1_RF22 / PSI /VT/ I PEI- OWN 02.05.01 PRE p c B-G-2 100624 ISI-13142-34-A ISI B7.70 B-4 Valve Bolts @ W-26 AUG 1 Main Steam 8 OWN PRE B-G-2 100625 IVe-1 B14t ISI 8- 5 Valve Bolts AUG B7.70 1 Main Steam 2 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 18 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 oN a0 00 00 Category, N w-DwgIISO No. Item No., Comp. Desc. -1 LL Ii. U L2L Class Summary No.IComplDISystem Scope I Method I Procedure Code Case - B-G-2 100626 M1_ 4-PlRF21 I PSI/VT / /PEI- IS1-13142-34-A ISI B7.70 B- 6 02.05.01 Valve Bolts AUG 1 Main Steam E M1_ 4-PlRF22 / PSI / VT / / PEI- OWN 02.05.01 Ml_I4-P2_RF23 / PSI IVTI / PEI-02.05.01 PRE c B B - - B-G-2 100629 ISI-13142-34-A SI------------------ s - - - B7.70 B-7 Ml_14-PlRF21 /PSI/VT / PEI- Valve Bolts AUG 1 Main Steam B 02.05.01 OWN . . . .. . . . .. . . . . M1_14-PlRF22 / PSI /VT/ / PEI-02.05.01 Ml_14-P2_RF23 / PSI /VT/VT-1 I PEI-02.05.01 Ml_14-P3_RF25 / ISI IVTIVT-1 IFP-PE-NDE-510 PRE c B - - B B-G-2 100672 ISI-13142-35-A ISI - c B7.70 B- 1 Valve Bolts AUG 1 Main Steam C OWN PRE - c B-G-2 100673 ISI-13142-35-A ISI c B7.70 B-2 Valve Bolts AUG 1 Main Steam C OWN PRE c B-G-2 100677 ISI-13142-35-A iSI B7.70 B-6 Valve Bolts AUG 1 Main Steam C OWN PRE B-G-2 100679 ISI-13142-35-A ISI B7.70 B-7 Valve Bolts AUG 1 Main Steam C OWN PRE B-G-2 100731 ISI-13142-36-A ISI B7.70 B- 7 Valve Bolts AUG 1 Main Steam C OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 19 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 C.) o U) o CD- 0) 0 (N (N (N Category, DwgIlSO No. (N1 w (N w Ln

                                                                                                      - 1   C-Item No.,                                                                      Comp. Desc.                              U. U.

e.J U. Class Summary NoJComplD/System Scope / Method I Procedure Code Case B-G-2 100732 M1_14-P2_RF24 / ISI / VT / VT-1 /FP- IS1-13142-36-A ISI--------- c---------- B7.70 B- 8 PE-NDE-510 Valve Bolts AUG 1 Main Steam C M1_4-P2_RF24 / PSI / VT / VT-1 / FP OWN PE-NDE-510 PRE-------- c--------c B-G-2 100733 ISI-13142-36-A ISI B7.70 B- 9 Valve Bolts AUG 1 Main Steam C OWN PRE B-G-2 100734 ISI-13142-36-A ISI B7.70 B-10 Valve Bolts AUG 1 Main Steam C OWN PRE B-G-2 100973 ISI-13142-42-A ISI B7.70 B- 1 Valve Bolts AUG 1 HPCI Steam OWN PRE B-G-2 100974 M1_14-PlRF21 IPSIIVTI IPEI- IS1-13142-42-A ISI t c B7.70 B- 2 02.05.01 Valve Bolts AUG 1 HPCI Steam M1_14-P2_RF23 / ISI /VT/ v-r-1/ PEI, OWN 02.05.01 M1_14-P2_RF23 /PSI/ VTIVT-1 I PEI-02.05.01 PRE c B B B-G-2 101191 ISI-13142-52-A ISI B7.70 B- 1 Valve Bolts AUG 1 Feedwatei OWN PRE B-G-2 101192 ISI-13142-52-A ISI B7.70 B-2 Valve Bolts AUG 1 Feedwatei OWN PRE .. . . ... . . ... . . . B-G-2 101193 ISI-13142-52-A ISI t B7.70 B-3 Valve Bolts AUG 1 Feedwate= OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 20 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 M' Ii, D CD o 0 o a N 1N1 N N Category, DwgIISO No. 04

                                                                                               -    N Item No.,                                                                    Comp. Desc.        N LL LL N       N   N LL LL Class      Summary No./ComplDISystem      Scope I Method I Procedure         Code Case                   -

B-G-2 101246 ISI-13142-53-A ISI B7.70 B- 1 Valve Bolts AUG 1 Feedwatei OWN PRE B-G-2 101247 ISI-13142-53-A ISI B7.70 B- 2 Valve Bolts AUG 1 Feedwatei OWN PRE B-G-2 101248 ISI-13142-53-A ISI B7.70 B-3 Valve Bolts AUG 1 Feedwatei OWN PRE B-G-2 101848 ISI-97003-A ISI B7.70 B- 1 Valve Bolts AUG 1 RHR Return A OWN PRE B-G-2 101849 ISI-97003-A SI c B7.70 B-2 M1_14-P1_RF21 /ISI /VT /PEI- Valve Bolts AUG 1 RHR Return A 02.05.01 OWN PRE B-G-2 101852 ISI-97003-A IS - B7.70 B- 5 Valve Bolts AUG 1 RHR Return A OWN PRE B-G-2 101890 ISI-97003-B ISI B7.70 B- 1 Valve Bolts AUG 1 RHR Suction A OWN PRE B-G-2 101893 ISI-97003-B ISI B7.70 B-4 Valve Bolts AUG 1 RHR Suction A OWN PRE B-G-2 101894 ISI-970V3-B ISI B7.70 B- 5 Valve Bolts AUG 1 RHR Suction A OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 21 of 370 ASME Section Xl (1995 EditiOn, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 r- 0D oo aW o3 0 o3 Category, DwgIlSO No. t"4l *1 C4 ~C2 N Comp. Desc. u- C' M~ ~ C4, C.4 L.L Item No., UL IL wn Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-G-2 101928 ISI-97004-A SI - t . .. . . . .. . . . . B7.70 B- 1 Valve Bolts AUG . . . ... . . ... . . . 1 RHR Return B OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-G-2 101929 ISI-97004-A ISI - - - - c . B7 .7 0 B- 2 M1 14 -P 2 _R F2 3 IS I / T / / P E I- Va lv e Bo lts A UG . . . . . . . . . . . . . . 1 RHR Return B 02.05.01 OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . ..- . B-G-2 101932 ISI-97004-A ISI . . . ... . . ... . . . B7.70 B- 5 Valve Bolts AUG . . . ... . . ... . . . 1 RHR Return B OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B -G -2 10 19 7 0 IS I-9 7 0 0 5 -A IS I . . . . . . . . . B7.70 B- 1 Valve Bolts AUG . . . ... . . ... . . . 1 Recirculation A OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 101973 ISI-97005-A ISIs[ . . . ... . . ... . B7.70 B- 4 Valve Bolts AUG . . . ... . . ... . . . 1 Recirculation A OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102075 ISI-97006-A ISI . . . ... . . ... . . . B7.70 B- 1 Bolts AUG . . . ... . . ... . . . 1 Recirculation B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102078 ISI-97006-A SI . . . ... . . ... . . . B7.70 B- 4 Bolts AUG . . . ... . . ... . . . 1 Recirculation B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . - B-G-2 102392 IS[ Fig 2 1SI . . . ... . . ... . . . B7.80 02-23B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-G-2 102395 IS[ -ig 2 Ii B7.80 02-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 22 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 00 00 00 00 N4 N 01 Category, DwgIlSO No. N M It 1

                                                                                                     ;;        I1 Item No.,                                                                     Comp. Desc.            UL SN LL N4 LL N

U. C-4 Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-G-2 102398 ISI Fig 2 SI - - B7.80 02-31B CRD Housing Bolts AUG 1 CRD Housings OW N . . . .. . . . .. . . . . PRE B-G-2 102401 ISI Fig 2 ISI B7.80 06-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102404 ISI Fig 2 ISi . . . .. . . . .. . . . . B7.80 06-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE ---------- c B-G-2 102407 ISI Fig 2 IS - B7.80 06-23B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102408 ISI Fig 2 ISI B7.80 06-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE ---------- c B-G-2 102412 ISI Fig 2 ISI B7.80 06-31B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102415 ISI Fig 2 ISi B7.80 06-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE c------------ B-G-2 102418 ISI Fig 2 ISI B7.80 06-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102421 ISI Fig 2 ISI B7.80 10-11B CRD Housing Bolts AUG 1 CRD Housings OWN PRE C Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 23 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 00 0 0 0 00 Category, Dwg/ISO No. -1

                                                                                                          *.I N1I1        I, *"

C14~ C14 M104 Item No., Comp. Desc. C4 CN LI. L . NN U. L I. LN U-Class w* w" o: w Summary No.IComplDISystem Scope I Method I Procedure Code Case B-G-2 102424 ISI Fig 2 iS1 B7.80 10-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102425 ISI Fig 2 iS - B7.80 10-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102426 M1_14-P1_RF21 /PSI/ VT / / PEI- ISI Fig 2 ISI B7.80 10-23B 02.05.01 CRD Housing Bolts AUG 1 CRD Housings Ml-14-PlRF22 / PSI / VT / / PEI- OWN 02.05.01 PRE c B B-G-2 102427 ISI Fig 2 ISI B7.80 10-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102428 ISI Fig 2 ISI B7.80 10-31B M1 I4-PlRF21 /PSI/VT/ /PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c B-G-2 102429 ISI Fig 2 ISI B7.80 10-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102432 ISI Fig 2 ISI B7.80 10-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Fig 2 PR -SI B-G-2 102435 B7.80 10-43B CRD Housing Bolts AUG 1 CRD Housings OWN PRE-------- c--------c B-G-2 102438 ISI Fig 2 ISI B7.80 14-07B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval IS Plan (Rev. 4) Page 24 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 e0 0 001 0C, 0) 0 o Category, DwglISO No. 41 04 0l N 01 Item No., Comp. Desc. -N U.LL C.4 L L

                                                                                                                                  )L C14 LL.

Cn NL Class Summary No.ICompID/System Scope I Method I Procedure Code Case B-G-2 102441 ISI Fig 2 ISI B7.80 14-11B CRD Housing Bolts AUG 1 CRD Housings OWN PRE c------------ B-G-2 102442 ISI Fig 2 ISI 87.80 14-15B CRD Housing Bolts AUG 1 CRD Housings OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . B-G-2 102443 ISI Fig 2 ISI . . . . . . . . .. . . . . B7.80 14-19B CRD Housing Bolts AUG 1 CRD Housings OW N -- . . .. . . . .. . . . . PRE ---------- c B-G-2 102444 ISI Fig 2 ISI B7.80 14-23B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN PRE B-G-2 102445 ISI Fig 2 ISI B7.80 14-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102446 ISI Fig 2 ISI B7.80 14-31B Ml_14-PlRF21 /PSI/ VT /PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c B-G-2 102447 ISI Fig 2 IsI B7.80 14-35B CRD Housing Bolts AUG 1 CRD Housings OWN - - .- .. . . . ... . . . PRE ------ - -c B-G-2 102448 ISI Fig 2 ISl B7.80 14-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102451 ISI Fig 2 ISI B7.80 14-43B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/@

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 25 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o o oM N Category, DwglISO No. oI0

                                                                                                     - I , I1              I     et Item No.,                                                                   Comp. Desc.                   L u.U                                  01 LL Class      Summary No.IComplDISystem      Scope I Method I Procedure        Code Case B-G-2      102454                                                           ISI Fig 2         ISI B7.80      14-47B                                                           CRD Housing Bolts AUG 1          CRD Housings                                                                       OWN PRE B-G-2      102457                                                           ISI Fig 2         ISI B7.80      18-07B                                                           CRD Housing Bolts AUG       .     .   .   ..      .    .   .    ...        .   .    .

1 CRD Housings OWN PRE B-G-2 102458 ISI Fig 2 ISI B7.80 18-11B CRD Housing Bolts AUG . . . .. . . . .. . . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE B-G-2 102459 ISI Fig 2 ISI B7.80 18-15B M1 14-P1 RF21 /PSI/ VTI IPEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c ISI Fig 2 ISI B-G-2 102460 B7.80 18-19B Ml14-P1_RF21 / PSI IVT- I PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c B-G-2 102461 ISI Fig 2 ISI B7.80 18-23B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102462 ISI Fig 2 IS] B7.80 18-27B CRD Housing Bolts AUG 1 CRD Housings OWN . . . ... . . ... . . . PRE B-G-2 102463 ISI Fig 2 ISI B7.80 18-31B CRD Housing Bolts AUG 1 CRD Housings OWN PRE -------------- c B-G-2 102464 ISI1-Fig Z Is1 B7.80 18-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 26 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 [1 0) o 0 o CD o N N N N Category, DwgIlSO No.

                                                                                                          - 1N1                       M~I  -e                wI Item No.,                                                                     Comp. Desc.                  N LL LL N                  N LL LL N                 N L-Class      Summary No.IComplDISystem       Scope I Method I Procedure         Code Case B-G-2       102465                                                            iSI Fig 2         ISI B7.80       18-39B                         Ml_14-Pl_RF21 /PSI   VT/ IPEI-     CR0 Housing Bolts AUG          ...............

1 CRD Housings 02.05.01 OWN PRE c . . . . . . . . . . . . . B-G-2 102466 ISI Fig 2 1SI5 - - . . . . B7.80 18-43B CRD Housing Bolts AUG .............. 1 CRD Housings OWN PRE-------- c--------c B-G-2 102469 ISI Fig 2 ISI - B7.80 18-47B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN PRE B-G-2 102472 ISI Fig 2 ISI B7.80 22-03B CRD Housing Bolts AUG 1 CRD Housings OWN PRE-------- c--------c B-G-2 102475 ISI Fig 2 ISI B7.80 22-07B CRD Housing Bolts AUG 1 CRD Housings OWN PRE --- ------- c B-G-2 102476 ISI Fig 2 ISI B7.80 22-11B CRD Housing Bolts AUG 1 CRD Housings OWN PRE-------- c--------c B-G-2 102477 ISI Fig 2 ISI .. . . .. . . . .. . . . . B7.80 22-15B M1_I4-PlRF21 /PSI /VT/ /PEI- CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings 02.05.01 OWN PRE c ISI Fig 2 ISI . . . ... . . ... . . . B-G-2 102478 87.80 22-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102479 ISI Fig 2 ISI B7.80 22-23B CRD Housing Bolts AUG I CRD Housings OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 27 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 00 C 0 0 00 00 Category, DwgIlSO No.

                                                                                                          - C4 'J        e1  ,V1       U, Item No.,                                                                    Comp. Desc.                                                U-LL
                                                                                                         .L u.           U. L.

Class Summary No./CompID/System Scope I Method I Procedure Code Case B-G-2 102480 ISI Fig 2 ISI B7.80 22-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102481 ISI Fig 2 ISI 87.80 22-31B M1_I4-P1_RF21 /PSI VT /PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c B-G-2 102482 ISI Fig 2 ISI B7.80 22-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102483 1SIFig 2 ISI B7.80 22-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102484 ISI Fig 2 ISI B7.80 22-43B CRD Housing Bolts AUG 1 CRD Housings OWN PRE------ ----- c B-G-2 102487 ISI Fig 2 ISI B7.80 22-47B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102490 ISI Fig 2 IsI B7.80 22-51B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102493 ISI Fig 2 ISI B7.80 26-03B CRD Housing Bolts AUG 1 CRD Housings OWN PRE---------- c B-G-2 102494 ISI Fig 2 ISI B7.80 26-07B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 28 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 00M N1 N4 Category, DwgIlSO No. 1 N N w- of Iof of U) Item No., Comp. Desc. u.L U* U.. I.II IL Class Summary No./CompID/System Scope I Method I Procedure Code Case B-G-2 102495 SI Fig 2 Is[ B7.80 26-11 B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102496 ISI Fig 2 ISI B7.80 26-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE . . . . . . . . . . . . . . B-G-2 102497 ISI Fig 2 ISI B7.80 26-19B CRD Housing Bolts AUG .............. 1 CRD Housings OWN .............. PRE B-G-2 102498 ISI Fig 2 ISI B7.80 26-23B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102499 ISI Fig 2 ISI B7.80 26-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102500 ISI Fig 2 ISI B7.80 26-31B M1_ 4-P1_RF21 / PSI /VT/ / PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE c B-G-2 102501 ISI Fig 2 ISI B7.80 26-35B CRD Housing Bolts AUG .............. 1 CRD Housings OWN P RE . . . . . . . . . . . . . . B-G-2 102502 ISI Fig 2 ISI B7.80 26-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102503 SI Fig 2 ISI B7.80 26-43B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/540

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 29 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 0 CD' Category, DwgIISO No. O41 C41 N N N 1

                                                                                                   .-     r'I C4 N            w  w Item No.,                                                                    Comp. Desc.            U. L*.

LL Class Summary No./CompID/System Scope I Method I Procedure Code Case B-G-2 102504 ISI Fig 2 ISI B7.80 26-47B CRD Housing Bolts AUG 1 CRD Housings OWN PRE- c------------ B-G-2 102507 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 26-51B CRD Housing Bolts AUG 1 CRD Housings OWN . . . ... . . ... . . . PRE B-G-2 102510 ISI Fig 2 ISI B7.80 30-03B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102513 ISI Fig 2 ISI B7.80 30-07B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102514 ISI Fig 2 IS1 B7.80 30-11B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102515 ISI Fig 2 ISI B7.80 30-15B CRD Housing Bolts AUG I CRD Housings OWN PRE B-G-2 102516 ISI Fig 2 ISI B7.80 30-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102517 ISI Fig 2 ISI B7.80 30-23B CRD Housing Bolts AUG 1" CRD Housings OWN PRE B-G-2 102518 ISI Fig 2 ISI B7.80 30-31 B M1_14-PlRF21 / PSI/VT I PEI- CRD Housing Bolts AUG 1 CRD Housings 02.05.01 OWN PRE C Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 30 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 00 0 04 1 0C4 00 C41 C-4 N Category, DwgllSO No. tIn

                                                                                                   -      N               C'.    -*

Item No., Comp. Desc. C4 LL UL U LI. L w* Class Summary No./CompiDISystem Scope I Method I Procedure Code Case B-G-2 102519 ISI Fig 2 ISI B7.80 30-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102520 ISI Fig 2 ISI . . . .. . . . .. . . . . B7.80 30-39B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . .. . . . .. . . . . PRE . . . ... . . ... . . . B-G-2 102521 ISI Fig 2 ISI B7.80 30-43B CRD Housing Bolts AUG 1 CRD Housings OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . B-G-2 102524 ISI Fig 2 ISI B7.80 30-47B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102527 ISI Fig 2 ISI B7.80 30-51B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102530 ISI Fig 2 ISI B7.80 34-07B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102531 ISI Fig 2 ISI B7.80 34-11B CRD Housing Bolts AUG 1 CRD Housings OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . B-G-2 102532 ISI Fig 2 ISI 87.80 34-15B CRD Housing Bolts AUG . . . .. . . . .. . . . . 1 CRD Housings OWN PRE B-G-2 102533 ISI Fig 2 ISI B7.80 34-19B CRD Housing Bolts AUG I CRD Housings OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 31 of 370 ASME Section X1 (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M. fl) C1 N R Category, DwglISO No.

                                                                                                       .-       N                                              N Item No.,                                                                    Comp. Desc.                N uL uL
                                                                                                                 "                   N LL uL IJ N

Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-G-2 102534 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 34-23B CRD Housing Bolts AUG .. . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-G-2 102535 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 34-27B CRD Housing Bolts AUG . . . . 1 CRD Housings OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-G-2 102536 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 34-31B CRD Housing Bolts AUG . . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102537 ISI Fig 2 ISl . . . ... . . ... . . . B7.80 34-35B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102538 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 34-39B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE ----------------- B-G-2 102541 ISI Fig 2 ISI B7.80 34-43B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102544 ISI Fig 2 ISI B7.80 34-47B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE ------ c B-G-2 102547 ISI Fig 2 ISI . . . .. . . . .. . . . . B7.80 38-07B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-G-2 102550 i01 rig z Isl B7.80 38-11B CRD Housing Bolts AUG 1 CRD Housings OWN PRE C Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 32 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 V*) La o 0 N Category, DwgIlSO No. N w N LL e.J 0: Item No., Comp. Desc. LL LL w w In Class Summary No./CompID/System Scope I Method I Procedure Code Case B-G-2 102551 ISI Fig 2 IsI B7.80 38-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102552 ISI Fig 2 IS[ B7.80 38-19B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102553 ISI Fig 2 IsI B7.80 38-23B CRD Housing Bolts AUG 1 CRD Housings OWN -- PRE B-G-2 102554 ISI Fig 2 iSi B7.80 38-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102555 M1_I4-PlRF21 / PSI / VT / I PEI- ISI Fig 2 ISI B7.80 38-31B 02.05.01 CRD Housing Bolts AUG 1 CRD Housings M1_I4-PlRF22 / PSI VTI PEI- OWN -- 02.05.01 PRE c B B-G-2 102556 ISI Fig 2 SI . . . .. . . . .. . . . . B7.80 38-35B CRD Housing Bolts AUG 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-G-2 102557 ISI Fig 2 ISI B7.80 38-39B CRD Housing Bolts AUG 1 CRD Housings OWN . . . ... . . ... . . . PRE B-G-2 102563 ISI Fig 2 ISI B7.80 38-47B CRD Housing Bolts AUG 1 CRD Housings OWN PRE . . . . . . . . . . . . . . B-G-2 102566 15I Fig 2 ISI B7.80 42-11B CRD Housing Bolts AUG I CRD Housings OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 33 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cn U) oo D a 0 N N4 N Category, DwgIiSO No. 0 U4 LL mI

                                                                                                  '-    04 Item No.,                                                                    Comp. Desc.           N u.. N1 LL* u_

LL o o U. Class Summary No.ICompID/System Scope I Method I Procedure Code Case B-G-2 102569 ISI Fig 2 ISI B7.80 42-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102570 ISI Fig 2 ISI B7.80 42-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102571 ISI Fig 2 ISI B7.80 42-23B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102572 ISI Fig 2 Is[ B7.80 42-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102573 ISI Fig 2 ISI B7.80 42-31B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102574 ISI Fig 2 ISi B7.80 42-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102577 ISI Fig 2 ISI B7.80 42-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102580 ISI Fig 2 ISI B7.80 42-43B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102583 ISI Fig 2 ISt B7.80 46-15B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 34 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 I,- o* o a

c. U.

o w N N Category, DwgIISO No. N N in Item No., Comp. Desc. N uL uL N U-Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-G-2 102586 ISI Fig 2 ISI B7.80 46-19B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102589 ISI Fig 2 ISI B7.80 46-23B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102590 ISI Fig 2 ISI B7.80 46-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102593 ISI Fig 2 ISI B7.80 46-31B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102596 ISI Fig 2 IS1 B7.80 46-35B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102599 ISI Fig 2 ISI B7.80 46-39B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102602 ISI Fig 2 ISI - - - - - . . . . B7 .8 0 5 0 -2 3 B C R D Ho u s ing Bo lt s A UG . . . . . . . . . . . . . . 1 CRD Housings OWN PRE B-G-2 102605 ISI Fig 2 ISI B7.80 50-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102608 ISI Fig 2 ISI B7.80 50-31B CRD Housing Bolts AUG 1 CRD Housings OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 35 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 coo oD 00 00 000 04 Category, Dwg/ISO No. N N 04 N N Item No., Comp. Desc. I x wI w w4 IL Class Summary No.IComplD/System Scope I Method I Procedure Code Case B-G-2 102714 iSI Fig 2 IS-B7.80 30-27B CRD Housing Bolts AUG 1 CRD Housings OWN PRE B-G-2 102715 ISI Fig 2 ISI . . . ... . . ... . . . B7.80 38-43B CRD Housing Bolts AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . B-K 102636 ISI Fig 3 SI----------------- s - - - B10.10 W-12 M1_I4-P3_RF25 /ISIIMT/ /PEI- B.H. to Skirt Weld AUG . . . ... . . ... . . . 1 Reactor Vesse 02.02.01 OWN . . . ... . . ... . . . PRE B-K 102650 ISI Fig 4 ISI r c B10.10 Vsl Stblzr Lug @ 0 deg. M1 _I4-PlRF22 / ISI /V r / / PEI- Stblz Lug @ 0 deg / R R4 - O ne tim e V isual A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.02 OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . B-K 106105 ISI Fig 4 ISI - c B10.10 Vsl Stblzr Lug @ 90 deg. M1 4-P 1_RF22 / ISI / VT // PEI- Stblz Lug @ 90 deg / RR4 - One time Visual AUG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.02 OWN . . . ... . . ... . . . PRE B-K 106106 ISI Fig 4 SI - c Stblz Lug @ 180 deg / RR4 - One time B 10 .10 Vsl Stblzr Lug @ 18 0 de g . Ml_14 -P I_ R F 22 / IS I /v -/ / P EI- Visual A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.02 OWN PRE B-K 106107 ISI Fig 4 SI - c Stblz Lug @ 270 deg / RR4 - One time B10.10 Vsl Stblzr Lug @ 270 deg. Ml-14-PlRF22 / ISI/VT/IPEI- Visual AUG 1 Reactor Vesse 02.05.02 OWN . . . ... . . ... . . . PRE B-K 100576 ISI-13142-;33-A lb51 c B10.20 H- 1 M1_14-PlRF21 / ISI / MT/ / PEI- Dbl Spring / 4 Lugs AUG 1 Main Steam A 02.02.01 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 36 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 c't In IO O~ 0 0 0 0 1(4 C1 04I1 U-Category, DwgIlSO No.

                                                                                                               -   N                 C4     C4                 N Ink Item No.,                                                                     Comp. Desc.                 U..      L.                 UL     IL                 U-Class       Summary No./CompID/System      Scope I Method I Procedure         Code Case B-K         100584                                                            ISI-13142-33-A      ISI 810.20      H- 8                                                              Dbi Spring / 4 Lugs AUG 1           Main Steam A                                                                          OWN PRE B-K         100630                                                            ISI-13142-34-A      ISI B10.20      H- 1                                                              Dbl Spring /4 Lugs  AUG 1           Main Steam B                                                                          OWN PRE B-K         100634                                                            ISI-13142-34-A      ISI B10.20      H- 5                                                              Dbi Spring / 4 Lugs AUG 1           Main Steam B                                                                          OWN PRE B-K         100681                                                            ISI-13142-35-A      ISI       .        .    .    ..       .     .    .     ..      .     .     .      .

B10.20 H- 1 Dbl Spring / 4 Lugs AUG .. . . ... . . ... . . . 1 Main Steam C OWN - - - - . . . . PRE . . . ... . . ... . . . B-K 100685 ISI-13142-35-A ISI B10.20 H- 5 Dbl Spring /4 Lugs AUG . . . .. . . . .. . . . . 1 Main Steam C OW N . . . .. . . . .. . . . . PRE B-K 100735 ISI-13142-36-A ISI B10.20 H- 1 Dbl Spring / 4 Lugs AUG 1 Main Steam E OWN PRE B-K 100742 ISI-13142-36-A ISi B10.20 H- 7 Dbl Spring / 4 Lugs AUG - . . .. . . . .. . . . . 1 Main Steam C OW N . . . .. . . . ... . . . PRE . . . ... . . ... . . . B-K 101197 ISI-13142-52-A ISI- -------------- - s - - - B10.20 H- 3 M1_I4-P3_RF25 ISIMT /PEI- Dbl Spring / 4 Lugs AUG . . . ... . . ... . . . 1 Feedwatei 02.02.01 OW N . . . .. . . . .. . . . . PRE B-K 101203 ISI-13142-52-A ISI B10.20 H- 8 Dbl Spring / 4 Lugs AUG 1 Feedwatei OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 37 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 O IO !O O~ Category, DwgIlSO No. N C- N 1N cm

                                                                                                                     -wr   N4 w       M o:*   V w.

Item No., Comp. Desc. C1 N1 N 04 UL U LL U. LL: Class Summary No.ICompiD/System Scope / Method I Procedure Code Case B-K 101251 IS1-13142-53-A ISI 810.20 H- 3 Dbi Spring / 4 Lugs AUG 1 Feedwatel OWN PRE B-K 101256 IS1-13142-53-A ISI---------- c---------- B10.20 H- 8 M1_14-P2_RF24 /ISI/ MT// PEI- DbI Spring / 4 Lugs AUG 1 Feedwatel 02.02.01 OWN PRE B-K 101363 ISI-73880-A ISI B10.20 H- 3 Dbl Spring /4 Lugs AUG 1 Reactor WMr Cleanup OWN PRE B-K 101976 ISI-97005-A ISI B10.20 H- 2 Dbl Spring /4 Lugs AUG 1 Recirculation A OWN PRE B-K 101981 ISI-97005-A ISI B10.20 H- 5 Clevis / Lugs / Constant-Suppor AUG 1 Recirculation A OWN PRE B-K 101983 ISI-97005-A ISI B10.20 H- 6 Clevis / Lugs / Constant-Suppor AUG 1 Recirculation A OWN PRE ISI-97005-A ISI B-K 101984 B10.20 H- 7 Double Spring AUG 1 Recirculation A OWN PRE B-K 101986 ISI-97005-A ISI B10.20 H- 8 Snubber / Lugs AUG 1 Recirculation A OWN PRE B-K 101989 ISI-97005-A ISI c B10.20 H-1 0 M1_14-P1_RF21 / ISI / PT / / PEI- Dbl Spring / 4 Lugs AUG I Recirculation A 02.01.01 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 38 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 00D00 Category, DwgIlSO No. U.J Item No., Comp. Desc. U UJ M le 0 w0 LL UL UL U Class Summary No.IComplD/System Scope I Method I Procedure Code Case B-K 102019 ISI-97005-B ISI B10.20 H- 2 Snubber/Lugs AUG 1 Recirc Manifold A OWN PRE B-K 102023 ISI-97005-B ISI B10.20 H-6 Spring / Lugs AUG 1 Recirc Manifold A OWN PRE B-K 102028 ISI-97005-B ISI B10.20 H-11 Spring / Lugs AUG 1 Recirc Manifold A OWN PRE B-K 102029 ISI-97005-B ISI r------------s - - - B10.20 H-12 M1_I4-P3_RF25 ISI PT! /PEI- Snubber/Lugs AUG 1 Recirc Manifold A 02.01.01 OWN PRE B-K 102081 ISI-97006-A ISI c B 10 .2 0 H- 3 M1 P I_ R F2 1 I ISI IP T I P EI- Db l S p rin g / 4 Lu g s A UG . . . . . . . . . . . . . . . 1 Recirculation B 02.01.01 OWN PRE B-K 102082 ISI-97006-A ISI B10.20 H-5 Clevis / Lugs / Constant-Suppor AUG 1 Recirculation B OWN PRE B-K 102087 ISI-97006-A IS1 B10.20 H-6 Clevis / Lugs / Constant-Suppor AUG 1 Recirculation B OWN PRE B-K 102089 ISI-97006-A ISI B10.20 H-7 Dbl Spring / Lugs AUG 1 Recirculation B OWN PRE B-K 102090 ISub-9706u-A ISI Snubber /Lugs AUG B10.20 H- 8 1 Recirculation B OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 39 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 I.- CD o CD o 0 C~4 Category, DwgllSO No. 1 N Item No., Comp. Desc. N, L. N' II. LL LL of Class Summary No.lComplDISystem Scope I Method ) Procedure Code Case B-K 102093 ISI-97006-A ISI B10.20 H-10 Dbi Spring AUG 1 Recirculation B OWN PRE B-K 102125 ISI-97006-B ISI B10.20 H- 2 Snubber/Lugs AUG 1 Recirc Manifold B OWN PRE B-K 102127 ISI-97006-B ISI B10.20 H-5 Spring / Lugs AUG 1 Recirc Manifold B OWN PRE B-K 102136 ISI-97006-B IS1 B10.20 H-10 Spring / Lugs AUG 1 Recirc Manifold B OWN PRE B-K 102138 ISI-97006-B IS] B10.20 H-12 Snubber AUG 1 Recirc Manifold B OWN PRE B-K 107644 ISI-97004-A ISI B10.20 H-1 Variable / SlidE AUG 1 RHR Return B OWN PRE B-K 107645 ISI-97003-A ISi . . . ... . . ... . . . B10.20 H-1 Variable Spring AUG . . . .. . . . .. . . . . 1 RHR Return A OWN PRE B-K 107646 ISI-97003-A ISI B10.20 H-2 Variable Spring AUG 1 RHR Return A OWN PRE B-K 107647 ISI-97003-B ISI B10.20 H-5 Variable SlidE AUG 1 RHR Suction A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 40 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M4. a C".= n N0 N Category, DwgIlSO No. ai Item No., Comp. Desc. N IL. N LL N U. N UL LA. Class Summary No.lCompID/System Scope I Method I Procedure Code Case B-K 107660 ISI-13142-42-A ISI . . . ... . . ... . . . B10.20 H- I Variable Spring AUG 1 HPCI Steam OWN . . . ... . . ... . . . PRE B-K 107661 IS1-13142-43-A ISI B10.20 H- 1 Variable Spring AUG . . . ... . . ... . . . 1 RCIC Steam OWN . . . ... . . ... . . . PRE B-K 107662 ISI-13142-33-A IS1 . . . ... . . . B10.20 H-4 Varible / Spring AUG - - - - . . . . 1 Main Steam A OWN . . . ... . . ... . . - PRE . . . ... . . . B-K 107663 ISI-13142-36-A ISl B10.20 H- 4 Variable / Spring AUG 1 Main Steam C OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-K 107665 ISI-13142-52-A SI . . . ... . . ... . . . B10.20 H-2 Variable / Spring AUG 1 Feedwatei OWN PRE B-K 107666 ISI-13142-53-A ISI . . . ... . . ... . . . 810.20 H- 2 Variable / Spring AUG . . . ... . . ... . . . 1 Feedwatei OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-K 107667 IS1-13142-26-A SI . . . ... . . ... . . . B10.20 H- 2 Restraint Slide AUG . . . ... . . ... . . . 1 Core Spray B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-K 102063 ISI-97005-C SI . . . ... . . ... . . . 810.30 H- 1 Rod/Clevis Grip/Lugs/Constant-Suppor AUG . . . ... . . ... . . . 1 Recirc Pump A OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-K 102064 ISI-97005-C ISI 810.30 H- 2 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump A OWN PRE - I Printed /

Monticello Nuclear Generating Plant 4th Interval IS[ Plan (Rev. 4) Page 41 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o 0 04 o o0 Category, DwgIlSO No. o N Ln Item No., Comp. Desc. uL U. LL La. UJ. w, Class Summary No./ComplDISystem Scope / Method I Procedure Code Case B-K 102065 ISI-97005-C ISI - c B10.30 H- 3 M1_I4-P1_RF22 ISI /PT// PEI- Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump A 02.01.01 OWN PRE B-K 102172 ISI-97006-C ISI B10.30 H- 1 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump B OWN PRE B-K 102173 ISI-97006-C ISI B10 .30 H- 2 R o d /C le v is G rip/L u g s/C o n sta nt-S u p po r AU G . . . . . . . . . . . . . . . 1 Recirc Pump B OWN . . . ... . . ... . .- PRE B-K 102174 ISI-97006-C ISI B10.30 H- 3 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump B OWN PRE B-L-2 102099 M1_I4-PlRF21 /PSII/ VT / / PEI- ISI-97006-A ISI - - - - c B12.20 P200-B 02.05.02 Pump Casing Internal Surfaces AUG 1 Recirculation B Ml-14-P1_RF22 / PSI / VT / / PEI- OWN 02.05.04 M1_14-P2_RF23 / ISI/VT/ /PEI-02.05.04 PRE c B B-L-2 107183 ISI-97005-A ISI - c B12.20 P200-A M1_14-PIRF21 / PSI / VT/ PEI- Pump Casing Internal Surfaces AUG 1 Recirculation A 02.05.02 OWN PRE c B B-M-2 100431 ISI-13142-26-A ISI B12.50 V- 1 Valve Int Surfaces AUG 1 Core Spray B OWN PRE B-M-2 100432 ISI-13142-26-A ISI B12.50 V- 2 Valve Int Surfaces AUG 1 Core Spray B OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 42 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 o eo o e 040 N N Category, DwglISO No.

                                                                                                    *- I    %                        0ý Item No.,                                                                     Comp. Desc.                              w   N 0:   w Class      Summary No.IComplDISystem       Scope I Method I Procedure         Code Case B-M-2       100433                                                            ISI-13142-26-A     ISI B12.50     V- 3                                                               Valve Int Surfaces AUG 1          Core Spray B                                                                          OWN PRE B-M-2       100497                                                            ISI-13142-31-A     ISI B12.50     V- 1                                                               Valve Int Surfaces AUG 1          Core Spray P                                                                          OWN PRE B-M-2      100498                                                             ISI-13142-31-A     SI   c B12.50     V-2                             M1_I4-P1_RF21 ISIIVTI /PEI-        Valve Int Surfaces AUG 1          Core Spray A                    02.05.04                                              OWN PRE B-M-2      100499                                                             ISI-13142-31-A     ISI B12.50     V- 3                                                               Valve Int Surfaces AUG 1          Core Spray A                                                                          OWN PRE B-M-2      100585                                                             ISI-13142-33-A     ISI B12.50     V- 1                                                               Valve Int Surfaces AUG 1          Main Steam A                                                                          OWN PRE B-M-2       100586                                                            ISI-13142-33-A     ISI B12.50     V- 2                                                               Valve Int Surfaces AUG 1          Main Steam A                                                                          OWN PRE    .     .  . ...       .  . ...       .   . .

B-M-2 100587 M1-4-P 1_RF21 / ISI IVT/I! PEI- ISI-13142-33-A ISI c B12.50 V- 3 02.05.04 Valve Int Surfaces AUG 1 Main Steam A Ml_14-PI_RF22 / PSI / VTI PEI- OWN 02.05.04 PRE - c B-M-2 100588 ISI-13142-33-A ISI B12.50 V- 4 Valve Int Surfaces AUG 1 Main Steam A OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 43 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 CO) o 103 O o a o 0 o a 0

                                                                                                    -          NN           N Category,                                                                    DwgIlSO No.                 N Item No.,                                                                    Comp. Desc.            L. IL N

u. N

u. N U-w.

Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-M-2 100635 ISI-13142-34-A IS1 c B12.50 V- 1 M1_I4-P1_RF21 / PSI/ / Valve Int SurfaceE AUG 1 Main Steam B Ml_14-PlRF22 / PSI VT/ PEI- OWN 02.05.04 Ml-14-PIRF22 /PSI /VT/ /PEI-02.05.04 PRE p c B-M-2 100636 M1_14-PlRF21 / ISI IVT IPEI- ISI-13142-34-A ISI c B12.50 V- 2 02.05.04 Valve Int SurfaceE AUG 1 Main Steam B M1_14-PlRF22 / PSI /VT /PEI- OWN 02.05.04 PRE - c B-M-2 100637 ISI-13142-34-A ISI B12.50 V- 3 Valve Int SurfaceE AUG 1 Main Steam B OWN PRE B-M-2 100638 IS1-13142-34-A ISI B12.50 V- 4 Valve Int SurfaceE AUG 1 Main Steam E OWN PRE B-M-2 100687 ISI-13142-35-A ISI B12.50 V- I Valve Int SurfaceE AUG 1 Main Steam C OWN PRE B-M-2 100688 ISI-13142-35-A ISI B12.50 V- 2 Valve Int Surfaces AUG 1 Main Steam C OWN PRE B-M-2 100689 Ml-14-PlRF21 / PSI IVTI ISI-VT- ISI-13142-35-A ISI - c B12.50 V- 3 3.0 Valve Int SurfaceE AUG 1 Main Steam C Ml-14-P1_RF22 / ISI /VT/ / PEI- OWN 02.05.04 PRE p B-M-2 100690 IS1-13142-35-A ISI C B12.50 V- 4 Valve Int SurfaceE AUG 1 Main Steam C OWN MI-14-Pl-RF21 / PSI / /I/ PRE p Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 44 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o CD N 04 04 Category, DwgIISO No. Item No., Comp. Desc.

                                                                                                        ~-141  N N.

w w In LI. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-M-2 100743 IS1-13142-36-A ISI B12.50 V- 1 Valve Int Surface- AUG 1 Main Steam C OWN PRE B-M-2 100744 ISI-13142-36-A ISI B12.50 V- 2 Valve Int Surfaces AUG 1 Main Steam C OWN PRE B-M-2 100745 ISI-13142-36-A ISI B12.50 V- 3 Valve tnt Surfaces AUG 1 Main Steam C OWN PRE B-M-2 100746 IS1-13142-36-A ISI---------------------c B12.50 V- 4 M1_14-P2_RF24 /ISI VT/ VT-3 /FP- Valve Int Surfaces AUG 1 Main Steam C PE-NDE-530 OWN PRE B-M-2 100991 ISI-13142-42-A ISI B12.50 V- 1 Valve Int Surfaces AUG 1 HPCI Steam OWN PRE B-M-2 100992 ISI-13142-42-A ISI - I - C B12.50 V- 2 M1_14-P2_RF23/ ISI / VT/ IPEI- Valve Int Surfaces AUG 1 HPCI Steam 02.05.04 OWN PRE B-M-2 101204 Ml1_4-P1_RF22 / ISI/V -/ / PEI- ISI-13142-52-A IS! - C B12.50 V- 2 02.05.04 Valve Int Surfaces AUG 1 Feedwatei Ml-14-P1_RF22 / ISI / V-r / / PEI- OWN 02.05.04 PRE B-M-2 101205 ISI-13142-52-A ISI B12.50 V- 1 Valve Int Surfaces AUG 1 Feedwatei OWN .. . . ... . . ... . . . PR E . . . . . . . . . . . . . . . B-M-2 101206 vIt-13142-5S-A IS1 c - - V- 3 Valve Int Surfaces, AUG B12.50 M1_14-PlRF21 / ISI /VT/ PEI-I Feedwatei 02.05.04 OWN PRE Printed 6/5*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 45 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 00 00 Category, DwgIlSO No. w oI wI Ln

                                                                                                          -N             .~   u Item No.,                                                                    Comp. Desc.                                                      C-U.

wt Class Summary No.IComplDISystem Scope / Method I Procedure Code Case B-M-2 101257 ISI-13142-53-A ISI B12.50 V- 1 Valve Int Surfaces AUG 1 Feedwatei OWN PRE B-M-2 101258 ISI-13142-53-A SI - c B12.50 V- 2 Ml_14-PlRF22 /ISI/ T / /PEI- Valve Int Surfaces AUG 1 Feedwatei 02.05.04 OWN PRE B-M-2 101259 IS1-13142-53-A IS - B12.50 V- 3 Valve Int Surfaces AUG 1 Feedwatei OWN PR E . . . . . . . . . . . . . . . B-M-2 101857 ISI-97003-A ISI - . . ... . . ... .. . . B12.50 V- 1 Valve Int Surfaces AUG 1 RHR Return A OWN PRE B-M-2 101858 ISI-97003-A ISt c B12.50 V- 2 M1 14-PlRF21 /ISI /VT /PEI- Valve Int Surfaces AUG 1 RHR Return A 02.05.04 OWN PRE B-M-2 101859 ISI-97003-A ISI B12.50 V- 3 Valve Int Surfaces AUG 1 RHR Return A OWN PRE B-M-2 101900 ISI-97003-B ISI B12.50 V- I Valve Int Surfaces AUG 1 RHR Suction A OWN PRE B-M-2 101901 ISI-97003-B ISI B12.50 V-2 Valve Int Surfaces AUG 1 RHR Suction A OWN PRE B-M-2 101902 ISI-97003-B ISI B12.50 V- 3 Valve nt Surfaces AUG 1 RHR Suction A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 46 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 00 0 00 0 CI CI Category, DwgIISO No. r w1 N 1N4 w Ix 04 NC.4 V) - NL Item No., Comp. Desc. N LL N LL CI LL IL Nl of Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-M-2 101937 ISI-97004-A SI - t B12.50 V- I Valve Int SurfaceE AUG 1 RHR Return B OWN PRE B-M-2 101938 ISI-97004-A ISI - - - - c B12.50 V- 2 MI 14-P2_RF23 ISI VT V--3 PEIValve Int Surface. AUG 1 RHR Return B 02.05.04 OWN PRE B-M-2 101939 ISI-97004-A IS - B12.50 V- 3 Valve Int SurfaceE AUG 1 RHR Return B OWN PRE B-M-2 101993 ISI-97005-A IS - B12.50 V- 1 Valve Int SurfaceE AUG 1 Recirculation A OWN PRE B-M-2 101994 ISI-97005-A IS - B12.50 V- 2 Valve Internal AUG 1 Recirculation A OWN PRE B-M-2 102100 ISI-97006-A ISI B12.50 V- 1 Valve Int Surfaces AUG 1 Recirculation B OWN PRE B-M-2 102101 ISI-97006-A ISI B12.50 V- 2 Valve Int Surfaces AUG 1 Recirculation B OWN PRE B-N-1 102716 ISI Fig 0 ISI - B - - - B - s M1_I4-PlRF22 I ISI / V- / / PEI- Top Guide (areas made access., fuel cell B13.10 C- 1A vacated) AUG 02.05.05 1 Reactor Vesse Ml_14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1_14-P3_RF25 / ISI/VT / / PEI-02.05.05 PRE Printed 6/5*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 47 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 co .n r,oo N! Category, DwgIISO No. NI, NIJ NIJ. U4 04 U) Item No., Comp. Desc. uL LL UL U- NL w-Class Summary No.ICompiD/System Scope I Method I Procedure Code Case B-N-1 102718 M1 _14-P1_RF22 / ISI / VT / / PEI- ISI Fig 0 SI - B - B - - - s - - - B 13 .10 C- 3A 0 2 .0 5.0 5 S h ro ud S h e lf 0 -18 0 d e l A UG . . . . . . . . . . . . . . 1 Reactor Vesse M1_I4-P2_RF23 / ISI /VT/VT-3 / PEl, OWN 02.05.05 M1_I4-P3_RF25 / ISI /VT/ I PEI-02.05.05 PRE B-N-1 102719 ISI Fig 0 ISI - B - - - B - - s - - - MI_14-PlRF22 / ISI / VT / / PEI- Surveillance Sample Holder & Bracket @ 30 B13.10 C- 4A 02.05.05 deg AUG 1 Reactor Vesse MlI4-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1_I4-P3_RF25 / ISI /VT/ I PEI-02.05.05 PRE B-N-1 102723 ISI Fig 0 ISI - B - - - B s - - - M1_14-P1_RF22/ISI/VT- / / PEI- Sim Dryer Holddown Bracket @ 35 deg (in B13.10 C- 8A 02.05.05 Closure Head) AUG 1 Reactor Vesse Ml_14-P2_RF24 / ISI / V- / / PEI- OWN 02.05.05 M1_14-P3_RF25 /ISI/VT / / PEI-02.05.05 PRE B-N-1 102724 I1I -Ig U ISI - B - - - Bi - - S - - - M1_I4-PIRF22 / ISI/VT / / PEI-B13.10 C-9A 02.05.05 Steam Dryer Support Bracket @ 35 deg. AUG 1 Reactor Vesse Ml_14-P2_RF24 / ISI /VT/ I PEI- OWN 02.05.05 M1_I4-P3_RF25 / ISI /VTI I PEI-02.05.05 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 48 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00O 0 0 00O 0 0 0

                                                                                                                                  'I 1         11

• NI Category, DwgIlSO No.

                                                                                                                               .-   041      M      -I Item No.,                                                                       Comp. Desc.                                    N1 ILL N1 ILL N1 U-N ILL U.

N Class Summary No./ComplD/System Scope I Method I Procedure Code Case B-N-1 107678 ISI Fig0 I - B - - - B s - - - M1_I4-P1_RF22/ ISI / VT / I PEI- Surveillance Sample Holder & Bracket @ 121 B13.10 C-4B 02.05.05 deg AUG 1 Reactor Vesse M1_14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 Ml_14-P3_RF25 / ISI/ IVT I/ PEI-02.05.05 PRE B-N-1 107679 ISI Fig 0 ISI - B - - - B - - s - - - Ml_14-P1_RF22 / ISI / VT / / PEI- Surveillance Sample Holder & Bracket @ 301 B13.10 C-4C 02.05.05 deg AUG 1 Reactor Vesse Mll4-P2_RF24 / ISI / V- / / PEI- OWN 02.05.05 Ml_14-P3_RF25 / ISI /VT // PEI-02.05.05 PRE B-N-1 107690 M1_14-P1_RF22/ ISI/VT/ / PEI- ISI Fig 0 SI - B B - - - s - - - B13.10 C- 3B 02.05.05 Shroud Shelf 180-360 del AUG 1 Reactor Vesse M1_I4-P2_RF23 / ISI / VT / VT-3 / PEI OWN 02.05.05 Ml14-P3_RF25 / ISI / VT / / PEI-02.05.05 PRE B-N-1 107696 M1_I4-PlRF22 / ISI / VT / / PEI- ISI Fig 0 ISI - B - - - B s - - - B13.10 C- ic 02.05.05 Guide Rod Bracket @ 175 decl AUG 1 ReactorVesse Ml_14-P2_RF24 / ISI /V-r/ / PEI- OWN 02.05.05 M1_I4-P3_RF25 / ISI /IVT/ / PEI-02.05.05 PRE B-N-1 107697 M1 _4-PI1RF22 / ISI / VT // PEI- ISI Fig U IS' B B S B13.10 C- 1D 02.05.05 Guide Rod Bracket @ 355 dec AUG 1 Reactor Vesse OWN Ml_14-P2_RF24 / ISI / VT/ / PEI-02.05.05 M1_I4-P3_RF25 / ISI / VT/ / PEI-02.05.05 PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 49 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M U) o a Co o C1 C1 0 N N N N Category, DwgIlSO No. Item No., Comp. Desc. N SN N N U. n U. LL LL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-1 107723 ISI Fig0 SI - B - - - B s - - - M1_14-P1_RF22 /ISI/VFT/ PEI- Stm Dryer Holddown Bracket @ 145 deg (in B13.10 C- 8B 02.05.05 Closure Head) AUG 1 Reactor Vesse Ml_14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1I14-P3_RF25 / ISI / VT / / PEI-02.05.05 PRE B-N-1 107724 ISI Fig0 1 - B - - - B - s - - - M1_14-P1_RF22 / ISI / Vr / / PEI- Stm Dryer Holddown Bracket @ 215 deg (in 813.10 C- BC 02.05.05 Closure Head) AUG 1 Reactor Vesse M1 14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1 14-P3_RF25 / ISI IVTI I PEI-02.05.05 PRE B-N-1 107725 ISI Fig 0 ISI - B - - - B s - - - M1 14-P1_RF22 / ISI / VT / / PEI- Stm Dryer Holddown Bracket @ 325 deg (in B13.10 C- 80 02.05.05 Closure Head) AUG 1 Reactor Vesse M1 14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1 14-P3_RF25 / ISI/V-r / / PEI-02.05.05 PRE B-N-1 107730 ISI Fig 0 ISI - B - - - B - - S - - - M1 14-P1_RF22 / ISI IVTI I PEI-B13.10 C- 9B Steam Dryer Support Bracket @ 145 deg. AUG 02.05.05 1 Reactor Vesse M1I14-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 M1 14-P3_RF25 / ISI /VTI I PEI-02.05.05 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 50 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M' La 11- a% 0' HI HI4 N I1 N N Category, Dwg/ISO No.

                                                                                                                           -     N          0~    -e        LO Item No.,                                                                     Camp. Desc.                                   N LI.

C-UL N U. C4 L N IL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-1 107731 ISI Fig 05Is - B - - - B - - s - - - M1 14-P1 RF22/151S/VTI /PEI-B13.10 C-9C 02.05.05 Steam Dryer Support Bracket @ 215 deg. AUG 1 Reactor Vesse M1_I4-P2_RF24 / ISI / VT / / PEI- OWN 02.05.05 Ml_14-P3_RF25 / ISI / VT / / PEI-02.05.05 PRE B-N-1 107732 ISI Fig 0 ISI - B - - - B - - s - - - M1_14-P1_RF2211SIIVT/I/PEI-B13.10 C- 9D 02.05.05 Steam Dryer Support Bracket @ 325 deg. AUG 1 Reactor Vesse M1_4-P2_RF24 / ISI / VT // /PEI- OWN 02.05.05 M1_4-P3_RF25 / ISI /VTr /I PEI-02.05.05 PRE B-N-1 107738 ISI Fig 0 ISI B B B B s Core Plate (areas made access., fuel cell B13.10 C-12A Ml_14-P1_RF22 / ISI/VT/ / PEI- vacated) AUG 1 Reactor Vesse 02.05.05 OWN M1_14-P2_RF23 / ISI IVT/ / PEI-02.05.05 Ml_14-P2_RF24 / ISI / VT / / PEI-02.05.05 M1_14-P3_RF25 / ISI / VT / I PEI-02.05.05 PRE B-N-2 102717 ISI Fig 0 ISI ----- - s- -s Jet Pump Pair 1/2 Riser Support Pads! B13.20 C- 2A Ml 14-P3_RF25 / ISI / VT/ /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107681 ISI Fig 0 ISI S - - - Jet Pump Pair 3/4 Riser Support Pads! B13.20 C-2B M1_I4-P3_RF25 / ISI / V /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed 6/50

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 51 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3

                                                                                                                                   .,) Ifl M LO         o    a oýl NQ        o    w        a N1   N*

Category, DwgIISO No. - N1 Item No., Comp. Desc. N N LL N Ug U. N Code Case Ix w1 Class Summary No./CompID/System Scope I Method I Procedure B-N-2 107682 ISI Fig 0 ISI----------------- S - - - Jet Pump Pair 5/6 Riser Support Pads / B13.20 C- 2C M1_I4-P3_RF25 / ISI /v/ I PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107683 ISI Fig 0 SI----------------- s - - - Jet Pump Pair 7/8 Riser Support Pads / B13.20 C- 2D M1_I4-P3_RF25/ ISIV/ / PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107684 IS[ Fig 0 SI----------------- s - - - Jet Pump Pair 9/10 Riser Support Pads / B13.20 C- 2E M1_4-P3_RF25 ISI VT /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107685 ISI Fig 0 ISI----------------- s - - - Jet Pump Pair 11/12 Riser Support Pads/ B13.20 C- 2F Ml_14-P3_RF25 /ISI VT/ /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107686 ISI Fig 0 ISI----------------- s - - - Jet Pump Pair 13/14 Riser Support Pads / B13.20 C- 2G M1_4-P3_RF25 ISI /VT /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107687 ISI Fig 0 ISI----------------- s - - - Jet Pump Pair 15/16 Riser Support Pads / B13.20 C- 2H M1_ 4-P3_RF25 ISI VTI /PEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107688 ISI -ig U Ibi 5 - - - Jet Pump Pair 17/18 Riser Support Pads I B13.20 C-2J Ml_14-P3_RF25 / SI / VT I / PEt- Welds AUG 1 ReactorVesse 02.05.05 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 52 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cq UO I- M~ 0 0 01 0 N N N N Category, DwgIlSO No. -t 1 N MI Item No., Comp. Desc. LL W N4 U. N U. N U. N LA. Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-2 107689 ISI Fig 0 ISI-- ------- s - - - Jet Pump Pair 19/20 Riser Support Pads / B13.20 C-2K Ml-14-P3_RF251ISIIV-r IPEI- Welds AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107693 ISI Fig 0 SI----------------- s - - - Surveillance Sample Holder Lower Weld @ B13.20 C-4D M1_ 4-P3_RF251ISIIVTI /PEI- 30deg AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107694 ISI Fig 0 ISI----------------- s - - - Surveillance Sample Holder Lower Weld @ B13.20 C-4E M1_I4-P3_RF251iSIIVTI /PEI- 120 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE . . . ... . . ... . . . B-N-2 107695 ISI Fig 0 ISI -- ------------ - s - - - Surveillance Sample Holder Lower Weld @ B13.20 C-4F M1 14-P3_RF25/ISI/VTI IPEI- 300 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107698 ISI Fig 0 ISI ---------------- s - - - B13.20 C- 1E M1 14-P3_RF25 / ISI / VTI /PEI- Guide Rod Bracket Weld @ 175 del; AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107699 ISI Fig 0 SIS[-- ------- s B13.20 C- 1F M1_ 4-P3_RF25 / ISI / VT / / PEI- Guide Rod Bracket Weld @ 355 deý AUG 1 Reactor Vesse 02.05.05 OWN PRE . . . ... . . ... . . . B-N-2 102720 ISI Fig 0 SI - c B13.30 C- 5A M1_I4-PlRF22 / ISI / VT / / PEI- Core Spray Piping Bracket Weld @ 30 deg. AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 102721 ISIFig 0 ISI S B13.30 C- 6A M1_I4-P3_RF25 IISI/VTI IPEI- FW Sparger Bracket Welds @ 0 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed O

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 53 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 oo am N Category, DwgIlSO No. u41 N u4 N Item No., Comp. Desc. Nt N N N N IL Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-N-2 107691 ISI Fig 0 ISI--------- -------- S - - - B13.30 C- 3C M1 14-P3_RF25 ISI/VT /PEI- Shroud Shelf H-9 Weld 0-180 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107692 ISI Fig 0 ISI----------------- s - - - B13.30 C- 3D M1_I4-P3_RF25 ISI/VTr /PEI- Shroud Shelf H-9 Weld 180-360 dec AUG 1 Reactor Vesse 02.05.05 OWN PR E . . . . . . . . . . . . . . B-N-2 107700 ISI Fig 0 ISI - c Core Spray Sparger Bracket Weld @ 150 813.30 C-5B M1_4-P1_RF221/SIIVTI/ PEI- deg. AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107701 ISI Fig 0 IS! - c B13.30 C- 5C M1_14-P1_RF22/ISI/VT/ /PEI- Core Spray Piping Bracket Weld@ 210 deg.AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107702 ISI Fig 0 ISI - c B13.30 C- 5D M1 14-PlRF22 / ISI /VT/ / PEI- Core Spray Piping Bracket Weld @ 330 deg.AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107703 ISI Fig 0 ISI----------------- s - - - B13.30 C- 6B M1 14-P3_RF25 / ISI / Vr / / PEI- FW Sparger Bracket Welds @ 90 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107704 ISI Fig 0 ISI----------------- s - - - B13.30 C- 6C M1_I4-P3_RF25 / ISI / VT / / PEI- FW Sparger Bracket Welds @ 180 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107705 ISI Fig 0 ISI S B13.30 C-6D M1 1_-P3_RF25/ISI/VT/ IPEI- FW Sparger Bracket Welds @ 270 deg AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 54 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 00 0 0 0 00 0 0 N Category, DwgIlSO No. 04 C14 cl 1 C4 OfN C4 C4 Item No., Comp. Desc. uL U-wg ULw U. I1 w IL of Class Summary No./ComplD/System Scope I Method I Procedure Code Case B-N-2 107706 M1_14-P2_RF24 / ISI / VT / / PEI- ISI Fig 0 ISI----------- A h - - - B13.30 C- 7B 02.05.05 Shroud Support Leg Weld @ 10 del; AUG 1 Reactor Vesse Ml_14-P3_RF25 / ISI / VT / VT-3 / PEI OWN 02.05.05 PRE B-N-2 107707 M1_I4-P2_RF24 / ISI / VT / / PEI- ISI Fig 0 ISI- ----- A h - - - B13.30 C- 7C 02.05.05 Shroud Support Leg Weld @ 30 deC AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI / VT / VT-3 / PEI. OWN 02.05.05 PRE B-N-2 107708 M1_I4-P2_RF24 / ISI /VTI / / PEI- ISI Fig 0 ISI- ----- A h - - - B13.30 C- 7D 02.05.05 Shroud Support Leg Weld @ 60 del AUG 1 Reactor Vesse M1_l4-P3_RF25 / ISI / VT / V--3 / PEI. OWN 02.05.05 PRE B-N-2 107709 M1_I4-P2_RF24 / ISI / VT / / PEI- ISI Fig 0 SI-- --------- A - - h - - - B13.30 C- 7E 02.05.05 Shroud Support Leg Weld @ 90 de[ AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI /VT / V--3 / P EI. OWN 02.05.05 PRE B-N-2 107710 Ml_14-P2_RF24 / ISI / VT/ / PEI- ISI Fig 0 ISI ---------- A - - h - - - B13.30 C- 7F 02.05.05 Shroud Support Leg Weld @ 120 decý AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI / VT / VT-3 / PEI. OWN - - 02.05.05 PRE B-N-2 107711 M1_I4-P2_RF24 / ISI / V- / / PEI- ISI Fig 0 ISI ---------- A - - h - - - B13.30 C- 7G 02.05.05 Shroud Support Leg Weld @ 150 dec( AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI / VT / VT-3 / PEI. OWN 02.05.05 PRE B-N-2 107712 Ml_14-P2_RF24 / ISI / VT / / PEI- ISI Fig 0 ISI A h B13.30 C- 7H 02.05.05 Shroud Support Leg Weld @ 170 deC AUG I Reactor Vesse Ml_14-P3_RF25 / ISI / VT / VT-3 / PEI. OWN 02.05.05 PRE Printed O

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 55 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 e, un o o o o o o N Category, DwgIlSO No. M -*

                                                                                                                               -I   C-      N' C'J Item No.,                                                                        Comp. Desc.                                   C1. C4     cLL LL         ,L w,

wL w Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-2 107713 M1_I4-P2_RF24 / ISI IVTI I PEI- ISI Fig 0 ISI----------- A - - h - - - B13.30 C- 7J 02.05.05 Shroud Support Leg Weld @ 190 de( AUG 1 Reactor Vesse Ml14-P3_RF25 / ISI / VT / VT-3 / PEI. OWN 02.05.05 PRE B-N-2 107714 Ml_14-P2_RF24 / IS /IVT/VT-3 / PEIlISI Fig 0 ISI H - - H - - h - - - B13.30 C- 7K 02.05.05 Shroud Support Leg Weld @ 210 deC AUG 1 Reactor Vesse Ml_14-P3_RF25 / ISI / VT / VT-3 / PEI, OWN 02.05.05 PRE B-N-2 107715 M1_14-P2_RF24 / ISIIVT/ / PEI- ISI Fig 0 ISI----------- E - - h - - - B13.30 C- 7L 02.05.05 Shroud Support Leg Weld @ 240 deg AUG 1 Reactor Vesse MI1_4-P3_RF25 / ISI/VT / VT-3 / PEI OWN 02.05.05 PRE B-N-2 107716 M1_l4-P2_RF24/ISIIV-FI /PEI- ISI Fig 0 ISI----------- A - - h - - - B13.30 C- 7M 02.05.05 Shroud Support Leg Weld @ 270 de(c AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI / VT I VT-3 / PEI, OWN 02.05.05 PRE B-N-2 107717 M1_I4-P2_RF24/SII/VT / PEI- ISI Fig 0 ISI----------- A - - h - - - B13.30 C- 7N 02.05.05 Shroud Support Leg Weld @ 300 deý AUG 1 Reactor Vesse M1_I4-P3_RF25 I ISI / VT / VT-3 / PEI. OWN - 02.05.05 PRE B-N-2 107718 M1_I4-P2_RF24/SI /VTI /PEI- ISI Fig 0 ISI ---------- A - - s - - - B13.30 C- 7P 02.05.05 Shroud Support Leg Weld @ 330 de( AUG 1 Reactor Vesse M1_I4-P3_RF25 / ISI /VT/ / PEI- OWN 02.05.05 PRE B-N-2 107719 M1_I4-P2_RF24 ISI / VT// PEI- ISI Fig 0 ISI A S B13.30 C- 70 02.05.05 Shroud Support Leg Weld @ 350 de; AUG 1 Reactor Vesse Ml_14-P3_RF25 / ISI / VT/ PEI- OWN 02.05.05 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 56 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 r- 4M o 0 o o o 0 o o Category, DwgIlSO No. N 1 4 N N4 N

                                                                                                                              -      C4                                                 U, Item No.,                                                                    Comp. Desc.                                       N      N                      N     C4                    N wo UL     L.                     LL uL                       Li.

Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-2 107720 ISI Fig 0 Is . . . ... . . s m Surveillance Sample Holder Upper Weld @ B13.30 C-4G Ml_14-P3_RF25 /11SI /VT I PEI- 30 deg AUG . . . ... . . ... . . . 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-N-2 107721 ISI Fig 0 ISI- - .- -- - - - .----s - - - Surveillance Sample Holder Upper Weld @ B13.30 C- 4H M1_I4-P3_RF25 ISII/VT / PEI- 120 deg AUG - - - - - . .- . ... . . . 1 Reactor Vesse 02.05.05 OWN -- . . ... . . ... . . . PRE . . . ... . . ... . . . B-N-2 107722 ISI Fig 0 SIS - . .-. . .-.-.---- s - - - Surveillance Sample Holder Upper Weld @ B13.30 C-4J M1_I4-P3_RF2511SIIV-I /PEI- 300 deg AUG . . . ... . . ... . . . 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-N-2 107726 ISI Fig 0 ISI- -.------------ - S - - - Stm Dryer Holddown Bkt Weld @ 35 deg (in B13.30 C- 8E M11_4-P3_RF25 /ISI VT PEI- Closure Hd) AUG 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . B-N-2 107727 ISI Fig 0 ISI -- ------------ - s - - - Stm Dryer Holddown Bkt Weld @ 145 deg B13.30 C- 8F M 14-P3_RF25 / ISI /VT / / PEI- (in Closure Hd) AUG . . . ... . . ... . . . 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-N-2 107728 ISI Fig 0 ISI- -.------------ - s - - - Stm Dryer Holddown Bkt Weld @ 215 deg 813.30 C- 8G M 14-P3_RF25 / ISI / V-I- PEI- (in Closure Hd) AUG . . . ... . . ... . . . 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . B-N-2 1U07729 ISI Fig U ISI S - - - Stm Dryer Holddown Bkt Weld @ 325 deg B13.30 C-8H M1_-4-P3_RF25 /IS[/V-/ PEI- (in Closure Hd) AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed 6/5*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 57 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 co U) o 0 0 0 co 1 a 0, Category, DwgIlSO No. N N- N

                                                                                                                                        -      N_

Item No., Comp. Desc. N LL N1 LL N N LL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case B-N-2 107733 ISI Fig 0 ISI - - s - - - Steam Dryer Support Bracket Weld @ 35 B13.30 C-9E Ml_14-P3_RF2511SIIVTI /PEI- deg. AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107734 ISI Fig 0 SI----------------- s - - - Steam Dryer Support Bracket Weld @ 145 B13.30 C- 9F M1_I4-P3_RF25 /ISi/VT/ PEI- deg. AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107735 ISI Fig 0 ISt----------------- s - - - Steam Dryer Support Bracket Weld @ 215 B13.30 C-9G M1_I4-P3_RF251/IS[VTI /PEI- deg. AUG 1 Reactor Vesse 02.05.05 OWN PR E . . . . . . . . . . . . . . B-N-2 107736 ISI Fig 0 IS- ---------------- S - - - Steam Dryer Support Bracket Weld @ 325 B 13 .3 0 C- 9 H M1_ I4-P 3 _R F 2 5 1/SiI VT I / P EI - deg . A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107740 ISI Fig 0 ISI c B13 .3 0 C -13 M1 _14-P l_ R F 2 2 / IS I/ VT / /P E I- Bo tto m He a d Dra in W e ld (N-15i A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107741 ISI Fig 0 ISI c B13 .3 0 C -14 M1 _4 -P 1_R F 2 2 / IS I /VT / P E I- Bo tto m He a d C R D S tu b T u b e ,s A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.05 OWN PR E . . . . . . . . . . . . . . B-N-2 107742 ISI Fig 0 ISI c B13 .30 C -15 M1 _4-P 1_R F 22 / IS I/VT / / P E I- Botto m Head C R D Ho using/Stu b T ube ,' A UG . . . . . . . . . . . . . . 1 Reactor Vesse 02.05.05 OWN . . . ... . . ... . . . PRE B-N-2 107743 ISI Fig 0 ISI C B13.30 C-16 M1_ 4-P1_RF221SII/VTI IPEI- Bottom Head .Incore Housint AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 58 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 cl U) Q C0 o V 0 0 0 N Category, DwgIISO No.

                                                                                                                                    -    I1       UI   LI1 Item No.,                                                                    Comp. Desc.                                             N     N4      w    N          N U. LI.                      IL Class       Summary No.IComplD/System     Scope I Method I Procedure         Code Case                                                                                     -

B-N-2 102722 ISI Fig 0 ISI---------------- s - - - B13.40 C- 7A Ml_14-P3_RF25 /ISI / VT /PEI- Shroud Support LegE AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107680 ISI Fig 0 ISI- -------------- - S - - - B13.40 C- 1B M1_I4-P3_RF25 ISI /VT// PEI- Top Guide AUG 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107739 ISIFig 0 ISI B B - B B s - - - Core Plate (areas made access., fuel cell B13.40 C-12B Ml-14-PlRF22/ ISI/VT / PEI- vacated) AUG 1 Reactor Vesse 02.05.05 OWN Ml_14-P2_RF23 / ISI/VT / / PEI-02.05.05 M1l_4-P2_RF24 / ISI/VT / I PEI-02.05.05 M1_I4-P3_RF25 / ISI/VTI/ / PEI-02.05.05 PRE B-N-2 107744 ISI Fig 0 ISI B B B - - - s M1_14-P1_RF22 / IS[ /VT / / PEI- Fuel Support Cast. (areas made access., fue B13.40 C-17A 02.05.05 cell vacated) AUG 1 Reactor Vesse Ml_I4-P2_RF23 / ISI/VT / / PEI- OWN 02.05.05 M1_I4-P3_RF25 / ISt / VT / PEI-02.05.05 PRE B-N-2 107745 ISi Fig 0 ISI B - - - 5 - - - Peripheral Fuel Spt (areas made access, fuel B13.40 C-17B M1_I4-P3_RF25 / ISI/ VT /PEI- cell vacated) AUG 1 Reactor Vesse 02.05.05 OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 59 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3

c. InI O O~

U,_ NI eqI N N Category, DwgIlSO No. -N V) - Comp. Desc. 'N N N N N Item No., UJ U. L. Lj. U. Class Summary No.IComplDISystem Scope / Method ! Procedure Code Case B-N-2 107746 ISI Fig 0 ISI B B B B s - - - Ctrl Rod Guide Tube - Int (when access, fuel B13.40 C-18 MlI4-P1_RF22 / ISI / VT / / PEI- cell vacated) AUG 1 Reactor Vesse 02.05.05 OWN - - - ,- - Ml_14-P2_RF23 / ISI/VT / / PEI-02.05.05 M1_I4-P2_RF24 / ISI/VT / / PEI-02.05.05 Ml_14-P3_RF25 / ISI / VT / / PEI-02.05.05 PRE B-N-2 107747 ISI Fig 0 SI----------------- s - - - B13.40 C- 3E Ml_14-P3_RF25 / ISI / VT / / PEI- Shroud Shelf 0-180 deg AUG .............. 1 Reactor Vesse 02.05.05 OWN PRE B-N-2 107748 ISI Fig 0 SI- - ----- - -s -s B13.40 C- 3F M1_I4-P3_RF25 /ISI /VT/ /PEI- Shroud Shelf 180-360 deg AUG .............. 1 Reactor Vesse 02.05.05 OWN PRE B-0 102390 ISI Fig 2 ISI B14.10 02-23 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE . . . . . . . . . . . . . . B-0 102391 ISI Fig 2 ISI 814.10 02-23 (Upper) CRD Housing Weld AUG .............. 1 CRD Housings OWN - - - - .... PRE . . . ... . . ... . . . B-O 102393 ISI Fig 2 ISIts - - - B14.10 02-27 (Lower) M1_I4-P3_RF25 /ISI/ PT / PEI- CRD Housing Weld AUG 1 CRD Housings 02.01.01 OWN PRE B-0 102394 IS] Fig 2 ISt S B14.10 02-27 (Upper) M1_14-P3_RF25 / ISI / PT/ PEI- CRD Housing Weld AUG 1 CRD Housings 02.01.01 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval IS[ Plan (Rev. 4) Page 60 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Ct' LO- oD 0 0 0 0 N N C4 4 Category, DwgIlSO No. "~1 0

                                                                                                        - 1N                                 In Item No.,                                                                    Comp. Desc.                  C-
                                                                                                           .L LL C1          N IL C4 U.

N U. Of' W* Ix W Class Summary No.ICompID/System Scope I Method I Procedure Code Case B-0 102396 ISI Fig 2 SI----------------- s - - - B14.10 02-31 (Lower) M1_I4-P3_RF25 / ISI / PT / / PEI- CRD Housing Weld AUG 1 CRD Housings 02.01.01 OWN PRE B-0 102397 ISI Fig 2 ISI----------------- s - - - B14.10 02-31 (Upper) M1 14-P3_RF25 / ISI / PT / / PEI- CRD Housing Weld AUG 1 CRD Housings 02.01.01 OWN PRE . . . ... . . ... . . . B-0 102399 ISI Fig 2 ISI - - - , B14.10 06-15 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-O 102400 ISI Fig 2 ISI B14.10 06-15 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE e-0 102416 ISI Fig 2 ISI B14.10 06-39 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102417 ISI Fig 2 ISI B14.10 06-39 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102419 ISl Fig 2 ISI B14.10 10-11 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102420 ISI Fig 2 ISI B14.10 10-11 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102433 ISI Fig 2 ISI B14.10 10-43 (Lower) CRD Housing Weld AUG I CRD Housings OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 61 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M~ UO o 0 0 o 0 N C4 N N 1NC Category, DwglISO No. N1 N Item No., Comp. Desc. LL '- LL LI.  : Nx U-Class Summary No./CompiD/System Scope I Method I Procedure Code Case Wm0 B-0 102434 iSI Fig 2 ISi B14.10 10-43 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102436 ISI Fig 2 ISI B14.10 14-07 (Lower) CRD Housing Weld AUG . . . ... . . ... . . . 1 CRD Housings OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . B-0 102437 ISI Fig 2 ISI B14.10 14-07 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN . . . ... . . ... . . . PRE B-0 102452 ISI Fig 2 IS1 . . . .. . . . .. . . . . B14.10 14-47 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102453 ISI Fig 2 ISi B14 .10 14 -4 7 (Up p e r) C R D Ho us ing We ld A UG . . . . . . . . . . . . . . 1 CRD Housings OWN PRE . . . ... . . ... . . . B-C 102470 ISI Fig 2 -IS B14.10 22-03 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102471 ISI Fig 2 ISI B14.10 22-03 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102488 ISI Fig 2 ISI B14.10 22-51 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE . . . ... . . ... . . . B-O 102489 ISI Fig 2 ISI B14.10 22-51 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 62 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 LO 00 0 0 0 00 0 0 Category, N' C'4 DwglIS0 No. - .1I I N. C4I

                                                                                                                                       -qI                 Ino Item No.,                                                                    Comp. Desc.             0'4 N' LL W    LL emJ N LL   IL a:                            W. W                     IL W.

Class Summary No.ICompID/System Scope I Method I Procedure Code Case B-0 102491 ISI Fig 2 ISI B14.10 26-03 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-O 102492 ISI Fig 2 ISI B14.10 26-03 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE . . . ... . . ... . . . B -O 10 2 5 0 5 IS I F ig 2 ISI . . . . . . . . . . . . . . B14.10 26-51 (Lower) CRD Housing Weld AUG -- . . .. . . . .. . . . . 1 C R D Ho u sing s O WN . . . . . . . . . . . . . . PRE . . . . . . . . . . . . . . B-0 102506 ISI Fig 2 IS - B14.10 26-51 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE -- B-0 102508 ISI Fig 2 ISI B14.10 30-03 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE . . . ... . . ... . . . B-0 102509 ISI Fig 2 ISI B14.10 30-03 (Upper) CRD Housing Weld AUG 1 C RD H o usings O WN . . . . . . . . . . . . . . PRE B -0 10 2 52 5 IS[ Fig 2 ISI . . . . . . . . . . . . . . B14.10 30-51 (Lower) CR D Housing W eld AUG . . . . . . . . . . . . . . 1 C R D Ho u sin g s O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . B -0 10 2 5 2 6 ISI F ig 2 SI . . . . . . . . . . . . . . B14.10 30-51 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102545 ISI Fig 2 ISI B14.10 38-07 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 63 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 V), If oý 0 o 0 N 04 Category, Dwg/ISO No. N N-0a N Item No., Comp. Desc. C4 LL N LA. LL W of N Class Summary No./ComplDISystem Scope I Method I Procedure Code Case B-0 102546 ISI Fig 2 ISI B14.10 38-07 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102561 ISI Fig 2 IS - B14.10 38-47 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102562 ISI Fig 2 ISI B14.10 38-47 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-O 102564 ISI Fig 2 IS - B14.10 42-11 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102565 ISI Fig 2 ISI B14.10 42-11 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102578 ISI Fig 2 IS - B14.10 42-43 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-O 102579 ISI Fig 2 ISi B14.10 42-43 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102581 ISI Fig 2 IS - B14.10 46-15 (Lower) CRD Housing Weld AUG 1 CRD Housings OWN PRE B-0 102582 ISI Fig 2 ISI B14.10 46-15 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 64 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 In) tOI'- Oa~ Category, DwgllSO No. N W 1 U.* 041 N 0ý N W) U.1 N W I' N C'4 -4 Item No., Comp. Desc. U, N Class Summary No.IComplDISystem Scope / Method / Procedure Code Case B-0 102597 ISI Fig 2 ISI B14.10 46-39 (Lower) CRD Housing Weld AUG . . . .. . . . .. . . . . 1 C R D Ho u sings OW N . . . . . . . . . . . . . . PR E - - - - . . . . B -O 10 2 5 9 8 IS I Fig 2 ISI . . . . . . . . . . . . . . B14.10 46-39 (Upper) CRD Housing W eld AUG . . . .. . . . .. . . . . 1 C R D Ho us ing s O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . B-O 102600 ISI Fig 2 II- -- -------------- s - - - B14.10 50-23 (Lower) M1_14-P3_RF25 /ISI /PT/ /PEI- CRD Housing Weld AUG . . . ... . . ... . . . 1 C R D Housings 02.01.01 O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . B-O 102601 ISI Fig 2 ISI - - ------- s - - - B14.10 50-23 (Upper) M1_I4-P3_RF25 I SI / PT / / PEI- CRD Housing Weld AUG . . . .. . . . ... . . . 1 C R D Housings 02.01.01 O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . B -O 10 2 6 0 3 ISI Fig 2 ISI . . . . . . . . . . . . . . B14.10 50-27 (Lower) CRD Housing W eld AUG . . . . . . . . . . . . . . 1 C R D Ho u sin g s O WN . . . . . . . . . . . . . . PRE B-O 102604 ISI Fig 2 ISI B14.10 50-27 (Upper) CRD Housing Weld AUG 1 CRD Housings OWN PRE B- 102606 ISI Fig 2 ISI B14.10 50-31 (Lower) CRD Housing Weld AUG . . . . . . . . . . . . . . 1 C R D Ho u sing s O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . B-0 10267U SI- Fig 2 ISI B14.10 50-31 (Upper) CRD Housing Weld AUG I CRD Housings OWN PRE 0 Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 65 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 U) 1- O~ 00 0 0 00 0 0lN Category, Dwg/ISO No. N Comp. Desc. LLU. I. Item No., W WE W W LL W) Class Summary No.IComplDISystem Scope I Method I Procedure Code Case h-I-I-B-P B15.10 107055 Reactor Coolant Pressure Boundar 1.5-2 Boundary System Leakage Pressure Tes ISI AUG BB - B B s- - 1 Reactor Coolant Pressure Boundary M1_I4-PlRF21 / ISI /V T / 0255 OWN IIC-2 Ml-14-PlRF22 / ISI / V- / / 0255 IIC-1 Ml_14-PlRF22 / ISI IVTI /0255 IIC-2 Ml _14-P2_RF23 / ISI / Vr / / 0255 IIC-2 Ml_14-P2_RF24 / ISI / V- / / 0255 IIC-1 Ml_I4-P2_RF24 / ISI I/VTI / 0255 IIC-2 M1 _l4-P3_RF25 / ISI IVTI / 0255 IIC-1 MI_14-P3_RF25 / ISI IVTI /0255 IIC-2 PRE C-A 101600 ISI-7905-32A ISI r c C1.10 W-1 M1_I4-PlRF22 / ISI / UT / / PEI- Top Flange to Shel AUG 2 RHR Heat Exchanger P 02.03.01 OWN PRE C-A 101601 ISI-7905-32A ISI ---------------- s - - - C1.10 W-2 Ml 14-P3_RF25 ISI IUT/ FP-PE- Lower Flange/Shel AUG 2 RHR Heat Exchanger P NDE-401 OWN PRE C-A 101611 ISI-7905-32B ISI Top Flange to Shel AUG Cl.10 W-1 2 RHR Heat Exchanger E OWN PRE C-A 101612 Lw Fa-7905e-32B ISI W-2 Lower Flange/Shel AUG C1.10 2 RHR Heat Exchanger E OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 66 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 N N N C4 N Category, DwgIlSO No. N N N N Item No., Comp. Desc. W W N L-N N Class Summary No.ICompID/System Scope I Method I Procedure Code Case C-A 101602 ISI-7905-32A ISI--------------------- C1.20 W-3 M114-P2_RF24 / 1SI/ UT/ / FP-PE- Lower Head/Flange AUG 2 RHR Heat Exchanger A NDE-401 OWN . . . ... . . ... . . . PRE C-A 101603 ISI-7905-32A ISI ---------------- s - - - Ci .20 W-4 M1_I4-P3_RF25 ISI /IUT/ I FP-PE- Lower Head Circ Welo AUG . . . ... . . ... . . . 2 RHR Heat Exchanger A NDE-401 OWN . . . ... . . ... . . . PRE C-A 101613 ISI-7905-32B ISI C1.20 W-3 Lower Head/FlangE AUG 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . C-A 101614 ISI-7905-32B IS! . . . ... . . ... . . . C1 .20 W-4 Lower Head Circ Welo AUG 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE C-B 101594 ISI-7905-32A ISI r c C2.31 N- 1 Mi1_4-PlRF22 / ISI / MT / / PEI- Shell-Pad-Noz AUG . . . ... . . ... . . . 2 RHR Heat Exchanger A 02.02.01 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . C-B 101595 ISI-7905-32A IS1 r r - - - c C2.31 N- 2 Ml_14-P2_RF24 / ISI /IMT / PEI- SheII-Pad-Noz AUG . . . ... . . ... . . . 2 RHR Heat Exchanger A 02.02.01 OWN PRE C-B 101604 ISI-7905-32B ISI . . . ... . . ... . . . C2.31 N -1 Shell-Pad-No2 AUG . . . ... . . ... . . . 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . C-B 101605 ISI-7905-32B ISI . . . ... . . ... . . . C2.31 N- 2 SheII-Pad-No2 AUG . . . ... . . ... . . . 2 RHR Heat Exchanger E OWN PRE C-C 101596 ISI-7905-32A ISI C3.10 Support A SupportA,E200A, 0' AUG 2 RHR Heat Exchanger P OWN PRE Printed O

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 67 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 03in M W) 0 N 0 N N4 Category, DwgIISO No. N L- IL IN Item No., Comp. Desc. N N N u.~ LL* LL N IL LI-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case C-C 101597 ISI-7905-32A ISI C 3 .10 S u p p o rt B S u p p o rt B,E2 0 0A 18 0 ' A UG . . . . . . . . . . . . . . 2 RHR Heat Exchanger P OWN . . . ... . . ... . . . PRE C-C 101598 ISI-7905-32A ISI - c C3.10 Support C M1 14-PlRF22 /ISIVIMT /PEI- Support C,E200A 315' AUG 2 R HR He a t E x c h a n g e r P 0 2 .0 2 .0 1 O WN . . . . . . . . . . . . . . PRE C-C 101606 ISI-7905-32B ISI . . . ... . . ... . . . C 3 . 10 S u p p o rt A S u pp o rt A ,E2 0 0 B 0 ' A UG . . . . . . . . . . . . . . 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE C-C 101607 ISI-7905-32B ISI . . . ... . . ... . . . C 3 .10 S up p o r t B S u p p o rt B,E2 0 0B 18 0 ' A UG . . . . . . . . . . . . . . 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE C-C 101609 ISI-7905-32B ISI . . . ... . . ... . . . C 3 .10 S u p p o rt C S u p p o r t C ,E 2 0 0 B 3 15' AU G . . . . . . . . . . . . . . 2 RHR Heat Exchanger E OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . C-C 100060 ISI-13142-17-A ISI . . . ... . . ... . . . C3.20 H- 5 DbI Spr / Half Clamp AUG 2 RHR Suction A OWN PRE . . . . . . . . . . . . . . C-C 100062 ISI-13142-17-A ISI C3.20 H- 6 Dbl Spr / Half Clamp AUG . . . ... . . ... . . . 2 RHR Suction A OWN . . . ... . . ... . . . PRE C-C 100133 ISI-13142-17-C ISI - -------------- s - - - C 3 .2 0 H- 6 M1 _I4 -P 3 R F 2 5 1 S I IMT I P E I- Dbl S pr / Ha lf C la m p A UG . . . . . . . . . . . . . . 2 RHR Suction B 02.02.01 OW N . . . . PRE C-C 100135 ISI-1314a-17 Cl- Is[1 Dbl Spr / Half Clamp AUG C3.20 H-8 2 RHR Suction B OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 68 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o 0 N Nm N CM Category, DwgIlSO No. C4 1

                                                                                                         -I Item No.,                                                                    Comp. Desc.                N U.

N LL N LI. N LI. N U-Class Summary No.ICompiD/System Scope I Method I Procedure Code Case C-C 100136 ISI-13142-17-C ISI - c C3.20 H- 9 M1_I4-PlRF22 /ISI /MT// PEI- DbI Strut / 8 Lugs AUG 2 RHR Suction B 02.02.01 OWN PRE C-C 100179 ISI-13142-18-A ISI c C3.20 H- 9 M1_I4-P1_RF21 / ISI /MT /I PEI- Dbi Spring / 4 Lugs AUG 2 RHR Discharge B 02.02.01 OWN PRE C-C 100265 ISI-13142-19-A ISI C3.20 H- 3 DbI Spring / 4 Lugs AUG 2 HPCI Steam Disch OWN PRE C-C 100270 ISI-13142-19-A ISI ---------------- s - - - C3.20 H- 7 M1_I4-P3_RF25 1SIIMTI /PEI- DbI Spring /4 Lugs AUG ............ - - 2 HPCI Steam Disch 02.02.01 OWN .............. PRE C-C 100788 ISI-13142-37-A ISI C3.20 H-7 Dbi Spring / 4 Lugs AUG 2 RHR Discharge A OWN PRE C-C 100814 ISI-13142-37-B ISI C3.20 H- 2 Strut / 8 Lugs AUG 2 Containment Spral OWN PRE C-C 100815 ISI-13142-37-B ISI - - - - C C3.20 H- 3 M1_ 4-P2_RF23 /ISI /MT/ /PEI- Strut / 8 Lugs AUG 2 Containment Spra) 02.02.01 OWN PRE C-C 100818 ISI-13142-37-B ISI C3.20 H- 5 Snubber /Lugs AUG 2 Containment Spral OWN PRE C-C 100820 IS1-13142-37-B ISI C3.20 H-6 Dbi Spr/Clamp&SaddlE AUG 2 Containment Spra) OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 69 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 oo o e o V LO 04 Category, DwgIISO No. ,- N'* N N- N Item No., Camp. Desc. U. N w N 1j. wo Class Summary No.IComplDISystem Scope I Method I Procedure Code Case C-C 100821 ISI-13142-37-B ISI C3.20 H- 7 Dbi Spring / 4 Lugs AUG 2 Containment Spral OWN PRE C-C 100917 ISI-13142-40-B ISI C3.20 H- 6 DbI Strut / 4 Lugs AUG 2 HPCI Water Side Dsch OWN PRE C-C 100978 ISI-13142-42-A ISI - - - - c C3.20 H- 3 Ml_14-P2_RF23 /IS /MT I / PEI- Dbi Spring / 4 Lugs AUG . . . ... . . ... . . . 2 HPCI Steam 02.02.01 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . C-C 101167 ISI-13142-51-B SI . . . ... . . ... . . . C3.20 H- 2 Dbi Spring / 4 Lugs AUG . . . ... . . ... . . . 2 Containment Spral OWN PRE C-C 101721 ISI-93268-1A ISI C3.20 H-20 Tank Support AUG 2 CRD Scram Header A OWN PRE C-C 101804 ISI-93268-3A ISI C3.20 H- 7 Restraint / 4 Lugs AUG 2 C R DS c ra m He a d e r B OW N - - . . . . . . . . PR E . . . . . . . . . . . . . . C-C 101815 ISI-93268-3A SI . . . . . . . . . . . . . . C3.20 H-14 Tank Support AUG . . . ... . . ... . . . 2 CRD Scram Header B OWN PRE . . . ... . . ... . . . C-C 106979 ISI-13142-67 ISI C 3.20 H- 6 Restraint Hangei A UG . . . . . . . . . . . . . . 2 Fuel Pool Emergency Coolin' OWN PRE C-C 107608 ISI-13142-17-A ISI C3.20 H- 3 Slide Pipe AUG 2 RHR Suction A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 70 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o o o o C 0 Co 0 N N Category, DwgIlSO No. MI It 1

                                                                                                     -I    NIJ                   C4 Item No.,                                                                                                                         it Comp. Desc.              N LL.

N-IL N U. N LL LL Class Summary NoJComplDISystem Scope I Method I Procedure Code Case C-C 107609 ISI-13142-17-A ISI C3.20 H-8 Slide AUG 2 RHR Suction A OWN PRE C-C 107610 ISI-13142-17-A ISI C3.20 H-9 Slide AUG 2 RHR Suction A OWN PRE C-C 107611 ISI-13142-17-B ISI C3.20 H- 3 Restraint AUG 2 HPCI Water Side Sctn OWN PRE C-C 107612 ISI-13142-40-B ISI C3.20 H-10 Variable Spring AUG 2 HPCI Water Side Dsch OWN PRE C-C 107613 IS1-13142-20-B ISI C3.20 H- 1 Variable AUG 2 Core Spray B OWN PRE C-C 107614 ISI-13142-20-B ISI C3.20 H- 2 Slide Hangei AUG 2 Core Spray B OWN PRE C-C 107615 ISI-13142-20-B IS - C3.20 H- 4 Variable / Suppor AUG 2 Core Spray H OWN PRE C-C 107616 ISI-13142-20-B ISI - - - - c C3.20 H- 5 M1_14-P2_RF23 /ISI / MT / I PEI- Variable / Suppor AUG 2 Core Spray B 02.02.01 OWN PRE C-C 107617 ISI-13142-20-A ISI H- 1 Variable Spr / SlidE AUG C3.20 2 Core Spray A OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 71 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o 0 N M -e Category, DwgIlSO No. mI cN- w w Item No., Comp. Desc. LL UL (.4 U-Class . Summary NoJComplDISystem Scope I Method I Procedure Code Case C-C 107618 ISI-13142-20-A ISI ---------------- s - - - C3.20 H- 2 M1_14-P3_RF25 I ISI/ MT / PEI- Slide AUG 2 Core Spray A 02.02.01 OWN PRE C-C 107619 ISI-13142-20-A ISI C3.20 H- 4 Variable Spr/SlidE AUG 2 Core Spray A OWN PRE C-C 107620 ISI-13142-51-A ISI C3.20 H- 2 Restraint Hangei AUG 2 RHR A OWN PRE C-C 107621 ISI-13142-51-A ISI C3.20 H- 4 Restraint Hangei AUG 2 RHR A OWN PRE C-C 107622 ISI-13142-51-A ISI C3.20 H- 6 Restraint Hangei AUG 2 RHR A OWN PRE C-C 107623 ISI-13142-51-A ISI C3.20 H- 7 Restraint Hangel AUG 2 RHR A OWN PRE C-C 107624 ISI-13142-37-C ISI C3.20 H- 3 Variable Spring AUG 2 RHR Discharge A OWN PRE C-C 107625 ISI-13142-37-A ISI C3.20 H- 4 Dbl Spring / Clamp AUG 2 RHR Discharge A OWN PRE C-C 107626 IVi-1ae14-S-pi Is1 Variable Spring AUG C3.20 H- 1 2 Containment Spral OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 72 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o 0 o 0 Category, DwgIISO No. N 04N N N N L0) Item No., Comp. Desc. w w w w u-Class Summary No.ICompiD/System Scope I Method I Procedure Code Case C-C 107627 ISI-13142-37-B IS - C3.20 H-4 Variable Spring AUG 2 Containment Spral OWN PRE C-C 107628 ISI-13142-26-C IS1 C3.20 H- 1 Restraint Hangel AUG 2 CORE SPRAY B OWN PRE C-C 107629 ISI-13142-26-C isi C3.20 H- 6 Restraint Hangei AUG 2 Core Spray B Discharge OWN PRE C-C 107630 ISI-13142-26-C ISI C3.20 H- 9 Restraint Hangei AUG 2 Core Spray B Discharge OWN PRE C-C 107631 ISI-13142-26-B IS1 C3.20 H- 2 Restraint Hangei AUG 2 CORE SPRAY B OWN PRE C-C 107632 IS1-13142-18-B ISI C3.20 H- 1 Variable SlidE AUG 2 RHR Discharge B OWN PRE C-C 107633 IS1-13142-18-A 0SI C3.20 H- 5 Dbi Spr/U-BItSaddle AUG 2 RHR Discharge B OWN PRE C-C 107634 IS1-13142-18-A ISI C3.20 H- 6 Dbl Spr/U-BIt/Saddle AUG 2 RHR Discharge B OWN PRE C-C 107635 IS1-13142-18-A Is' C3.20 H- 8 Dbi Spr/U-BltlSaddle AUG 2 RHR Discharge B OWN PRE Printed 6/5/2

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 73 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 0 0 0 0 0 0 Category, DwgIISO No. CI N N1 CI1 we

                                                                                                     -1 14     Cl)I

• 04 w w w w Item No., Comp. Desc. N1 N1 N NM LA. U. LL LL LL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case C-C 107636 ISI-13142-37-D ISI C3.20 H- 1 Restraint Hangei AUG 2 RHR A OWN PRE C-C 107637 ISI-13142-37-D ISI C3.20 H- 2 Restraint Hangei AUG 2 RHR A OWN PRE C-C 107638 IS1-13142-31-B ISI C3.20 H-1 B Restraint Hangei AUG 2 Core Spray A DischargE OWN PRE C-C 107639 ISI-13142-26-B ISI C3.20 H- 4 Restraint Hangei AUG 2 CORE SPRAY B OWN PRE C-C 107640 ISI-13142-51-D ISI C3.20 H- 1 Restraint Hangei AUG 2 RHR B OWN PRE C-C 107641 ISI-13142-51-D ISI C3.20 H- 2 Restraint Hangei AUG 2 RHR B OWN PRE C-C 107642 ISI-13142-51-C ISI C3.20 H- 1 Restraint Hanger AUG 2 RHR B OWN PRE C-C 107643 IS1-13142-18-A ISI C3.20 H-1I Seismic Restraint AUG 2 RHR Discharge B OWN PRE C-C 107648 ISI-13142-31-C ISI C3.20 H- 1 Restraint Hangei AUG 2 Core Spray A DischargE OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 74 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 cO In O~ O~ 00 0 0C C 0 0 00C 0 N 1 1 N IN Category, DwgIlSO No. N u. Item No., Comp. Desc. 04 U L U-04 C4 U N IL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case C-C 107649 ISI-13142-49-A ISI C3.20 H- 2 DbI Spring / Tee AUG 2 RHR Suction A OWN PRE C-C 107650 ISI-13142-17-C ISI C3.20 H-4 M 1-14-P2_RF23 /PSI /PT/ /PEI- Slide AUG 2 RHR Suction B 02.01.01 OWN PRE - - - - c C-C 107651 ISI-13142-17-C ISI C3.20 H- 2 Slide AUG 2 RHR Suction B OWN PRE C-C 107652 ISI-13142-17-C ISI C3.20 H- 3 Slide AUG 2 RHR Suction B OWN PRE C-C 107653 ISI-13142-62 ISI C3.20 H- 5 Restraint Hangel AUG 2 Fuel Pool Emergency Coolini OWN PRE C-C 107654 ISI-13142-62 ISI C3.20 H- 4 Restraint Hanger AUG 2 Fuel Pool Emergency Coolinj OWN PRE C-C 107656 ISI-13142-37-D ISI C3.20 H- 3 Double Snubber AUG 2 RHR A OWN PRE C-C 107657 ISI-13142-51-C ISI C3.20 H- 5 Seismic Snubber AUG 2 RHR B OWN PRE C-C 107658 1Sn-13142-51-A ISI C3.20 H- 1 Snubber / Dbl Strui AUG 2 RHRA OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 75 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o CD Co 0ý N C4 N 1N" Lo N Category, DwgIISO No.

                                                                                                    - 1N Item No.,                                                                    Comp. Desc.             C-UL N

LL N LL N4 IL cm LL w, Class Summary No.lComplD/System Scope I Method ) Procedure Code Case C-C 107659 ISI-13142-37-C ISI C3.20 H-4 Dbi Strut / Snubber AUG 2 RHR Discharge A OWN PRE C-C 107664 ISI-13142-19-B ISI C3.20 H- 2 Restraint Hanger AUG 2 RCIC Steam Discharge OWN PRE C-C 107668 ISI-13142-40-A ISI C3.20 H-14 Seismic Restraint AUG 2 HPCI Water Side Dsch OWN PRE C-C 107669 ISI-13142-17-B ISI C3.20 H-4 Restraint AUG 2 HPCI Water Side Sctn OWN PRE C-C 107670 ISI-13142-40-B ISI C3.20 H- 3 Seismic Restraint AUG 2 HPCI Water Side Dsch OWN PRE C-C 107671 ISI-13142-19-B ISI C3.20 H- 5 Restraint Hanger AUG 2 RCIC Steam DischargE OWN PRE C-C 107672 ISI-13142-31-C ISI C3.20 H- 4 Seismic Restraint AUG 2 Core Spray A DischargE OWN PRE C-C 107673 ISI-13142-48-A ISI C3.20 H- 4 Restraint Hanger AUG 2 RHR SERVICE WATER OWN PRE C-C 107674 ISI-13142-67 ISI C3.20 H- 1 Restraint Hangei AUG 2 Fuel Pool Emergency Cooling OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 76 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M La o a o 0 N C4 N N IN Category, DwgllSO No. 4, Item No., Comp. Desc. LL IL N uL N uL U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case C-C 107675 IS1-13142-67 ISI C3.20 H- 2 Restraint Hangei AUG 2 Fuel Pool Emergency Coolini OWN PRE C-C 107676 ISI-13142-67 ISI C3.20 H- 5 Restraint Hanger AUG 2 Fuel Pool Emergency Cooling OWN PRE C-C 107677 IS1-13142-67 ISI C3.20 H- 9 Restraint Hangei AUG 2 Fuel Pool Emergency Coolint OWN PRE C-C 107753 ISI-13142-37-D ISI C3.20 W-7A Reinforcing Plate-to-PipE AUG 2 RHR A OWN PRE C-C 107754 ISI-13142-37-D ISI C3.20 W-7B Reinforcing Plate-to-Pipt AUG 2 RHR A OWN PRE C-C 102745 ISI-48 ISI C3.30 RHR Support C Pump Support AUG 2 RHR Pumps OWN PRE C-C 102747 ISI-48 ISI C3.30 RHR Support D Pump Support AUG 2 RHR Pumps OWN PRE C-C 102751 ISI-48 ISI---------------- - s - - - C3.30 RHR Support A M1 14-P3_RF25 /ISI /MT / / PEI- Pump Support AUG 2 RHR Pumps 02.02.01 OWN PRE C-C 102752 ISI-48 ISI C3.30 RHR Support B Pump Support AUG 2 RHR Pumps OWN PRE Printed /

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 77 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 C4 04 0 04 i e rI I I 1 N Category, DwgIlSO No. tIn Item No., Camp. Desc. -L w LL w LL w w N L-Class Summary No.ICompID/System Scope I Method I Procedure Code Case C-C 102753 ISI-49 IS1 r r c - - C3.30 Support, Pump A M1_I4-P2_RF23 ISI/MTI I PEI- Pump Support AUG 2 Core Spray Pumps 02.02.01 OWN PRE C-C 102754 ISI-49 ISI C3.30 Support, Pump B Pump Support AUG 2 Core Spray Pumps OWN PRE C-H 100003 1.5-12 Boundarý ISI - B - - - B - - s - - - C7.10 SBLC Leakage Pressure Tes AUG MI-14-PlRF22 / ISI / VT/ / 0255-02-System 2 SBLC ilC-1 OWN M1_I4-PlRF22 / ISI / VT / / 0255 IIC-2 M1 1_4-P2_RF24 / ISI / VT / / 0255 IIC-1 M 1_I4-P2_RF24/ ISI / V- / / 0255 IIC-2 M1_I4-P3_RF25 / ISII VT/ / 0255 IIC-1 MlI4-P3_RF25 / ISI / VT / / 0255 IIC-2 PRE C-H 100008 Ml_14-P2_RF24 / ISI / VT/ / 0255-03.1.5-5 Boundarý IS! - B - - - B - - s - - - C7.10 Core Spray System Loop J tIC-1 System Leakage Pressure Tes AUG 2 Core Spray Loop Ml 14-P3_RF25 / ISI / V--/ /0255 OWN IPC-1 PRE - - - - - - - - - - - - C-H 100010 1.5-7 Boundarý ISI B B S C7.10 RHR System Loop E M1 14-P 1_RF22 / ISI / / System Leakage Pressure Tes AUG 2 RHR Loop B Ml_4-P2_RF24 / ISI /VT/ / 0255 OWN IIC-2 M1_I4-P3_RF25 / ISI IVTI / 0255 IIC-2 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 78 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 0 N Category, Dwg/ISO No. C- M It Item No., Comp. Desc. U. S0N L. N U. N U. N U. Class Summary No.IComplD/System w, Scope I Method I Procedure Code Case C-H 100013 1.5-6 Boundary ISI - B - - - B - - s - - - C7.10 RHR System Loop A 025 5 04 System Leakage Pressure Tes AUG M1 14Pl-F22/ ISI / VT/ / 0255-04-Sse Ml_14-P1_RF22 / ISI/VT / / 0255 2 RHR~oopA IC-1 IW IIC-3 Ml_14-P2_RF24 / ISI / VT / / 0255 IIC-1 M1_14-P3_RF25 / ISI/VT / / 0255 llC-3 Ml1_4-P3_RF25 /1ISI/VT I /0255 IIC-3 PRE C-H 100023 1.5-8 Boundary ISI - B - - - B - - s - - - C7.10 HPCI MI1_4-PlRF22 / ISI / System Leakage Pressure Tes AUG 2 HPCI M1_I4-P2_RF24 / ISI /VT/ / 0255 OWN iiC-1 Ml_14-P3_RF25 / ISI /VT/ 0255 1IC-1 PRE C-H 100025 1.5-10 Boundary ISI - B - - - B - - s - - - C7.10 RCIC M1_I4-PlRF22 / ISI / / System Leakage Pressure Tes AUG 2 RCIC M1_I4-P2_RF24 / ISI/VT / / 0255 OWN Ic-1 M1_14-P3_RF25 / ISI/VT / / 0255 IIC-1 PRE C-H 100037 1.5-3 Boundary ISI B B S C7.10 Feedwater Systery Ml_14-PlRF22 / ISI / / / System Leakage Pressure Tes AUG 2 Feedwater P M1_14-P2_RF24/ISI/VT/ / 0255 OWN IIC-1 Ml_14-P3_RF25 / ISI /VTI /0255 'IC-1 PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 79 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M L o o o* o* 04 N M le Category, DwgIlSO No. 01 N' 04 N N-

                                                                                                                   - 1N1 Item No.,                                                                           Comp. Desc.                     N uL N4 L.        LLt N    LL N

IL w Class Summary No.ICompID/System Scope / Method I Procedure Code Case C-H 100038 1.5-3 Boundary ISI - B - - - B - - s - - - C7.10 Feedwater Systery M1_I4-PlRF22 / ISI // System Leakage Pressure Tes AUG 2 Feedwater E Ml_14-P2_RF24 / ISI /VT/ / 0255 OWN IIC-2 M1_I4-P3_RF25 / ISI IVTI/ 0255 IIC-2 PRE C-H 100044 M1_14-PlRF22 / ISI/VT / / 0255-16-1.5-21 Boundar) ISI - B - - - B - - s - - - C7.10 Drywell Floor Drair 1lC-1 System Leakage Pressure Tes AUG 2 Drywell Floor Drair M1_I4-P2_RF24 / ISI /VT / / 0255 OWN IIC-1 M1_I4-P3_RF25 / IS] / VT / / 0255 'Ic-1 PRE C-H 100046 1.5-21 Boundary ISI - B - - - B - - s - - - C7.10 Drywell EQT Drain Ml_14-P1_RF22 / IS[ / / System Leakage Pressure Tes AUG 2 Drywell EQT Drain M1_I4-P2_RF24 / ISI / VT / / 0255 OWN IIC-2 Ml_I14-P3_RF25 / ISI /VT/ /0255 IIC-2 PRE C-H 106300 1.5-13 Boundar ISI - t - - t t - - t - - - C7.10 Pri. Containment Vac. Relie System Leakage Pressure Tes AUG 2 Pri. Containment Vac. Relie N-522,N-522,N-522,N-522 OWN PRE C-H 106400 1.5-13 Boundary St - t - - t t - - t - - - C7.10 Pri. Containment Air Purge Suppli System Leakage Pressure Tes AUG 2 Pri. Containment Air Purge Suppil N-522,N-522,N-522,N-522 OWN PRE C-H 106500 1.5-13 Boundary SI - t - - t t - - t - - - C7.10 Torus Air Purge Supply System Leakage Pressure Tes AUG 2 Torus Air Purge SuppIl N-522,N-522,N-522,N-522 OWN PRE C-H 106601 1 .5-13 Bounaarý ISI t t t t C7.10 PCAC System Leakage Pressure Tes AUG 2 PCAC N-522,N-522,N-522,N-522 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 80 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o a [8 8 o a N N N Category, DwgIlSO No. C4 MI leI 0o N

                                                                                                                   - 1N Item No.,                                                                          Comp. Desc.                      N    N        N4
                                                                                                                                  *l N

U. U-Code Case w wL wg w1 Class Summary No.IComplDSystem Scope I Method I Procedure C-H 106700 1.5-13 Boundarý ISI - It t - - - - C7.1 0 Standby Gas Treatmen System Leakage Pressure Tes AUG 2 Standby Gas Treatmen N-522,N-522,N-522,N-522 OWN PRE C-H 106800 1.5-13 Boundar) ISI - t t - t - - - C7.10 Torus HPV System Leakage Pressure Tes AUG 2 Torus HPV N-522,N-522,N-522,N-522 OWN PRE C-H 107136 M1_t4-P1_RF22 / ISI / VT / / 0255 1.5-17 Boundary ISI - B - - - B - - s - - - C7.10 CRDH IIC-2 System Leakage Pressure Tes AUG 2 CRDH M1_14-P2_RF24 / ISI / VT / / 0255 OWN IIC-2 Ml_14-P3_RF25 / ISI / VT / / 0255 IIC-2 PRE C-H 107140 1.5-19 Boundar) ISI - t - - t t - - t - - - C7.1 0 Compressed Air Systerr System Leakage Pressure Tes AUG 2 Compressed Air N-522,N-522,N-522,N-522 OWN PRE C-H 107144 1.5-20 Boundarý ISI - t - - t t - - t - - - C7.10 Demin Water Systerr System Leakage Pressure Tes AUG 2 Demin Water Syterr N-522,N-522,N-522,N-522 OWN PRE C-H 107148 1.5-20 Boundarý ISI - t - - I t - - t - - - C7.10 RBCCW System Leakage Pressure Tes AUG 2 RBCCW N-522,N-522,N-522,N-522 OWN PRE C-H 107156 M1 14-PlRF22 / ISI /VT1 /10255 1.5-21 Boundar ISI - B - - - B s - - - C7.10 RWCU iIC-2 System Leakage Pressure Tes AUG 2 RWCU M1_t4-P2_RF24 / ISIIVT / / 0255 OWN IIC-2 M1_I4-P3_RF25 / ISI / VT / / 0255 IIC-2 PRE C-H 107164 1.5-22 Boundarý ISI t t t t - C7.10 Tranversing In-Core Probe Systen System Leakage Pressure Tes AUG 2 Tranversing In-Core Probe Systen N-522,N-522,N-522,N-522 OWN PRE Printed /

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 81 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 o W o M N N o 0 0

                                                                                                                 -41   (4 Category,                                                                         DwgIlSO No.

M -e Item No., Comp. Desc. W I 04 UN LL W. Class Summary No./CompID/System Scope I Method I Procedure Code Case C-H 107172 1.5-23 Boundarý ISI - I - - t t - - t - - - C7.10 Excess-Flow Check Valve! System Leakage Pressure Tes AUG 2 Excess-Flow Check Valve! N-522,N-522,N-522,N-522 OWN PRE C-H 107176 1.5-26 Boundarý SI - t - - t t - - t - - - C7.10 PC Sample Systerr System Leakage Pressure Tes AUG 2 Primary Containment Sample Syster N-522,N-522,N-522,N-522 OWN PRE C-H 107607 M1 _14-P2_RF24 / ISI IVT/ / 0255-03.1.5-5 Boundary ISI - B - - - B - - s - - - C7.10 Core Spray System Loop E IIC-2 System Leakage Pressure Tes AUG 2 Core Spray Loop E Ml_14-P3_RF25 / ISI IVT/ /0255 OWN IIC-2 PRE D-A 102756 nd-isi-101 ISI D1.20 SS-562 Dbi Strut / Snubber AUG 3 RHR Service Water OWN PRE D-A 102757 nd-isi-101 ISI D1.20 SWH-43 Stanchion AUG 3 RHR Service Water OWN PRE D-A 102758 nd-isi-102 ISI D1.20 SR-79 Seismic Restraini AUG 3 RHR Service Water OWN PRE D-A 102759 nd-isi-102 ISI D1.20 SWH-304 Double Spring AUG 3 RHR Service Water OWN PRE D-A 102760 nd-isi-103 ISI D1.20 SR-459 Seismic Restraini AUG 3 RHR Service Water OWN PRE D-A 102761 nd-isi-1 0b ISI D1.20 SR-457 M1_I4-P2_RF24 / PSI / VT /VT-1 / FP Anchor AUG 3 RHR Service Water PE-NDE-510 OWN PRE c Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 82 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cl) w r- M 001 0.0 1 0I0 00 04 Nl N N Category, DwgIlSO No. LI Item No., Comp. Desc. LL NN LL N LL N4 LL U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case D-A 102762 nd-isi-107 ISI D1.20 SR-105 M1_14-P2_RF24 /PSI /VT/VT-1 /FP Stanchion AUG 3 RHR Service Water PE-NDE-510 OWN PRE- ------ c D-A 102763 And-isi-107 ISI 01.20 SWH-72A Stanchion AUG OWN 3 RHR Service Water PRE D-A 102764 nd-isi-107 ISI 01.20 SWH-728 Stanchion AUG 3 RHR Service Water OWN PRE . . . ... . . ... . . . D0-A 102765 nd-isi-108 ISI ---------------- s - - - D1.20 IS-SWH-65 Ml_14-P3_RF25 ISI VT/ /FP-PE- Stanchion AUG 3 RHR Service Water NDE-510 OWN PRE nd-isi-1PR -SI D-A 102766 Stanchion AUG D1.20 IS-SWH-66 OWN 3 RHR Service Water PRE D-A 102767 nd-isi-108 ISI c r 01.20 SR-106 Ml_14-PIRF21 /ISI/VT /PEI- Stanchion AUG 3 RHR Service Water 02.05.01 OWN PRE . . . ... . . ... . . . D-A 102768 nd-isi-109 ISI D1.20 SR-458 Anchor AUG . . . ... . . ... . . . 3 RHR Service Water OWN ............... PRE .. . . ... . . ... . . . D-A 102769 nd-isi-1 11 ISI D1.20 SR-88 Stanchion AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . D-B 100021 M1_I4-PlRF22 / ISIS v / 0255 1.5-16 Boundarý ISI B c D2.10 RHR SW Div. I tIC-1 System Leakage Pressure Tes AUG 3 RHR SWA MlI4-P2_RF24 / 1SIVT/ !' 0 2 5 5 -0 5 -<No Code Case,<No Code Case OWN tIC-1 PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 83 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 Cn Lu I-- OD 0 0 0 0 0 C-- C" 0 N C41

                                                                                                                          *1I          0 M

0cl V NI Category, DwgIlSO No. SN C-4 0 Item No., Comp. Desc. u. L U. LL,,

                                                                                                                            ,L,        UI    UJ.

L N Li. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case D-B 100022 Ml1_4-PlRF22/ISIIVTI / 0255-05.1.5-16 Boundary ISI - B - - - c D2.10 RHR SW Div. II IIC-2 System Leakage Pressure Tes AUG 5 3 RHR SW B M1_I4-P2_RF24 / ISI /VT/ / 02 5 5 <No Code Case,<No Code Cast OWN IIC-2 PRE D-B 100031 1.5-14 Boundary ISI - B - - - c D2.10 EDG-ESWA Div 1 Ml-14-P1_RF22 / ISI / VT / / 0255-1 1-System Leakage Pressure Tes AUG 3 EDG-ESWA IIC-3-1 <No Code Case OWN M1_14-Pl RF22 / ISI / VT / / 0255 IIC-3-2 M1_ 4-P2_RF24 / ISI / VT / / 0255 IIC-3-1 Ml 14-P2_RF24 / ISI / VT / / 0255 IIC-3-2 PRE D-B 100032 1.5-15 Boundar ISI - B - - - c 02 55 11 D2.10 EDG-ESW B Div. II Ml-14-P1_RF22 / ISI /VT/ / - -System Leakage Pressure Tes AUG 3 EDG-ESW B IIC-4-1 <No Code Case,<No Code Cast OWN M1_I4-PIRF22/]SIIVTI /0255 IIC-4-2 M1_I4-P2_RF24 / ISI IVTI /0255 IIC-4-1 M1_I4-P2_RF24 / ISI /VTI / 0255 IIC-4-2 PRE D-B 100033 M1 _14-P1_RF22 / ISI IVT/ / 0255-11.1.5-14 Boundary ISI - B - - - c D2.10 ESW A Div. 1 IIC-1 System Leakage Pressure Tes AUG 02 55 3 ESW A M1_I4-P2_RF24 / ISI /VT/ / <No Code Case,<No Code Cast OWN 1IC-1 PRE D-B 100034 Ml1_4-PlRF22 / iSI /VT/I 0255-11.1.5-15 Boundary ISI - B - - - c D2.10 ESW B Div. II IIC-2 System Leakage Pressure Tes AUG 02 55 3 ESW B M1_I4-P2_RF24 / ISI / VT / / <No Code Case,<No Code Cast OWN .............. IIC-2 PRE D-B 107102 Ml-14-PlRF22 / ISI / VT / 0255-06.1.5-8 Boundary ISI B C D2.10 HPCI IIC-1 System Leakage Pressure Tes AUG 3 HPCI M1_I4-P2_RF24 / ISI / VT/ I 02 5 5- 0 6 -<No Code Case OWN IiC-1 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 84 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o o o o e o o N NI N Category, DwgIlSO No. N N N

                                                                                                                           -1    C4        N    N4 Item No.,                                                                            Comp. Desc.                            04 UL N

Li. Li. U. Class Summary No./ComplD/System Scope I Method I Procedure Code Case D-B 107111 M1 _14-PIRF22 / ISI / VTI / /0255-08-1.5-10 Boundary 1 - B - - -Sc D2.10 RCIC IiC-1 System Leakage Pressure Tes AUG 3 RCIC Ml_14-P2_RF24 / ISI /VT-/ / 02 5 5-08-<No Code Case OWN IIC-1 PRE D-B 100017 1.5-16 Boundary SI----------------- s - - - D2.20 RHR SW Div. I M1_I4-P3_RF25 / ISI /VT/ / 0 2 5 5 HYdrostatic Pressure Test AUG 3 RHR SW A IlC-1 N-498-4 OWN PRE D-B 100020 1.5-16 Boundary SI----------------- s - - - D2.20 RHR SW Div. II M1_4-P3_RF25 / ISI /VVT/ / 0 2 5 5 - 0 5 -Hydrostatic Pressure Test AUG 3 RHR SW B liC-2 N-498-4 OWN PRE D-B 100027 M1_I4-P3_RF25 / ISI IVTI / 0255-11-1.5-14 Boundary ---------------- s - - - D2.20 EDG-ESW A Div. I IIC-3-1 Hydrostatic Pressure Test AUG 0255 11 3 EDG-ESW A MI_14-P3_RF25 / ISI /VT/ / - -N-498-4 OWN IIC-3-2 PRE D-B 100028 M1_I4-P3_RF25 / ISI / VT / / 0255-11.1.5-15 Boundary ISI----------------- s - - - D2.20 EDG-ESW B Div. II IIC-4-1 Hydrostatic Pressure Test AUG 3 EDG-ESW B M1_I4-P3_RF25 / ISI / VT / 10 2 5 5 - 1 1 -N-498-4 OWN IIC-4-2 PRE D-B 100029 1.5-14 Boundary ISI----------------- s - - - D2.20 ESW A Div. I M1 _14-P3_RF25SI IS/ VT / / 0 2 5 5 - 1 1 -Hydrostatic Pressure Test AUG 3 ESW A IIC-1 N-498-4 OWN PRE D-B 100030 1.5-15 Boundary ISI----------------- s - - - D2.20 ESW B Div. II Ml_14-P3_RF25 / ISI / VT / / 0 2 5 5 Hydrostatic Pressure Test AUG 3 ESW B Ilc-2 N-498-4 OWN PRE D-B 107597 1.5-8 Boundary SI----------------- s - - - D2.20 HPCI M1_I4-P3_RF25 / ISI /VT/ / 0 2 5 5-06-Hydrostatic Pressure Test AUG 3 HPCI iC-1 N-498-4 OWN PRE D-B 107598 1.5-10 Boundarý ISI s D2.20 RCIC M1_I4-P3_RF25 / ISI /VTI / 0 2 5 5- 0 8 -Hydrostatic Pressure Test AUG 3 RCIC IIC-1 N-498-4 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 85 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o 0 0 NI Category, DwgllS0 No. co le C.- N N N Item No., Comp. Desc. LL w U. 0: w w LL n Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 100427 ISI-13142-26-A SI------------ c Fl.10a H-1 M1_I4-P2_RF241ISIIVTI /FP-PE- Restraint/Clamp AUG 1 Core Spray B NDE-530 OWN PRE F-A 100428 IS1-13142-26-A ISI F1.10a H- 2 Restraint Slide AUG 1 Core Spray B OWN PRE F-A 101361 ISI-73880-A ISI F1.10a H-2 Box Restrainl AUG 1 Reactor Wtr Cleanup OWN PRE F-A 101420 ISI-74215A ISI Fl.10a H- 1 Box Restrainl AUG 1 Standby Liquid Cntr OWN PRE F-A 101421 ISI-74215A ISI F1.10a H- 2 Box Restraint AUG 1 Standby Liquid Cntr OWN PRE F-A 101422 ISI-74215A ISI----------------- s - - - Fl.10a H- 3 M1_14-P3_RF25 /ISI/VT/ /FP-PE- Box Restraini AUG 1 Standby Liquid Cntr NDE-530 OWN PRE F-A 101423 ISI-74215A ISI Fl.10a H- 4 Box Restraint AUG 1 Standby Liquid Cntr OWN PRE F-A 101460 ISI-782A ISI Fl.10a H- 1 Hanger AUG 1 Head Veni OWN PRE F-A 101461 It~-f *LA IS1 C Fl.10a H- 2 M1_I4-PlRF22 / iSi/ VT/ / PEI- Hanger AUG 1 Head Veni 02.05.02 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 86 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 Cl LO O O 00 00 00 00 0 CN N N N Category, DwglISO No. N

                                                                                                   -         N                  M~ le Item No.,                                                                    Comp. Desc.             N        NI                 N     N1                 LL N

U. U. U. U. LL Class Summary No.IComplDlSystem Scope / Method I Procedure Code Case F-A 101520 ISI-786A ISI B - - -.. . Fl.10a H-1 M1_I4-PlRF21 ISIIVTI /PEI- Box Support AUG . . . ... . . ... . . . 1 MS Condensate Lkof 02.05.02 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 101521 ISI-786A ISI . . . . . . . . . . . . . . Fl.10a H-2 Box Support AUG . . . . . . . . .. . . . . 1 MS Condensate Lkof OWN PRE . . . ... . . ... . . . F-A 101522 iSI-786A SI - c Fl.10a H- 3 Ml_14-P1_RF22 /ISI VT IPEI- Linear Supporl AUG 1 MS Condensate Lkof 02.05.02 OWN . . . ... . . ... . . . PRE F-A 101523 ISI-786A ISI . . . . . . . . . . . . . . F1.10a H-4 Box Support AUG 1 MS Condensate Lkof OW N . . . .. . . . .. . . . . PRE F-A 101618 ISI-821A ISI- s------- Fl.10a H-2 M1_I4-P3_RF2511SIIVTI /FP-PE- Box Restrainl AUG . . . ... . . ... . . . 1 Bottom Head Drair NDE-530 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 101620 ISI-821A ISI . . . . . . . . . . . . . . Fl.10a H- 4 Box Restraint AUG 1 Bottom Head Drair OWN . . . .. . . . .. . . . . PRE F-A 101621 ISI-821A SI . . . . . . . . . . . . . . F1.10a H- 5 Box Restraint AUG . . . . . . . . . . . . . . 1 Bottom Head Drair OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . F-A 101622 ISI-821A SI . . . . . . . . . . . . . . F1.10a H- 6 Box Restraint AUG 1 Bottom Head Drair OWN PRE F-A 101974 ISI-97005-A ISI Fl.10a H- 1 Seismic Restraint AUG 1 Recirculation A OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 87 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 oo o 04

                                                                                                        ,-    ¢N       co   I*

Category, Dwg/ISO No. U4 04 N w w LL w Item No., Comp. Desc. w Of N. U-Class Summary No./ComplD/System Scope I Method I Procedure Code Case F-A 101980 ISI-97005-A ISI-- ------- s - - - Fl.10a H- 4 M1_14-P3_RF25 / ISI / VT / / FP-PE- Seismic Restraint AUG 1 Recirculation A NDE-530 OWN PRE F-A 101988 ISI-97005-A SI E F1.10a H- 9 Ml_I14-PlRF21 / ISI / VT / / PEI- Seismic Restraint AUG 1 Recirculation A 02.05.02 OWN PRE F-A 101991 ISI-97005-A ISI E Fl.10a H-12 M1_14-P1_RF21 / ISI / VT / / PEI- Seismic Restraint AUG 1 Recirculation A 02.05.02 OWN PRE F-A 102018 ISI-97005-B SI-- -c-------- C Fl.10a H- 1 M1_I4-P2_RF24 / ISI/VT / / FP-PE- Restraint AUG 1 Recirc Manifold A NDE-530 OWN PRE F-A 102020 ISI-97005-B ISI Fl.10a H- 3 Restraint AUG 1 Recirc Manifold A OWN PRE F-A 102021 ISI-97005-B SI] ---------------- s - - - Fl.10a H-4 M1 14-P3_RF251iSI/VT/ IFP-PE- Restraint AUG 1 Recirc Manifold A NDE-530 OWN PRE F-A 102022 ISI-97005-B IS - Fl.10a H- 5 Restraint AUG 1 Recirc Manifold A OWN PRE F-A 102024 ISI-97005-B ISI Fl.10a H- 7 Restraint AUG 1 Recirc Manifold A OWN PRE F-A 102025 ISI-train-t 1I I F1.10a H- 8 Restraint AUG 1 Recirc Manifold A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 88 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 00 0 Category, DwgIlSO No. I ,I e'I , 04 N. cm CJ. La Item No., Comp. Desc. N N N N NL L wj LA. U LA. X Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102026 ISI-97005-B ISI Fl.10a H- 9 Restraint AUG 1 Recirc Manifold A OWN PRE F-A 102027 ISI-97005-B ISI - - - - c F1.10a H-10 M1_I4-P2_RF23 /ISI/v-VF PEI- Restraint AUG 1 Recirc Manifold A 02.05.02 OWN PRE F-A 102030 ISI-97005-B ISt Fl.10a H-13 Restraint AUG 1 Recirc Manifold A OWN PRE F-A 102079 ISI-97006-A ISI Fl.10a H- 1 Restraint AUG 1 Recirculation B OWN PRE F-A 102085 ISI-97006-A ISt c r Fl.10a H-4 Ml 4-PIRF21 IISIIVT/ /PEI- Restraint AUG 1 Recirculation B 02.05.02 OWN PRE F-A 102092 ISI-97006-A ISI Fl.10a H- 9 Restraint AUG 1 Recirculation B OWN PRE F-A 102098 ISI-97006-A ISt . . . ... . . ... . . . Fl.10a H-12 Restraint AUG . . . ... . . ... . . . 1 Recirculation B OWN PRE F-A 102124 ISI-97006-B IS - F1.10a H- 1 Restraint AUG 1 Recirc Manifold B OWN PRE F-A 102130 ISI-97006-13 IS1 Fl.l0a H- 4 Restraint AUG 1 Recirc Manifold B OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant - 4th Interval ISI Plan (Rev. 4) Page 89 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0, LO o 0 N N N Category, DwgIlSO No. N NI Item No., Comp. Desc. N U. N LL N N N U-Class Summary No./ComplD/System Scope I Method I Procedure Code Case F-A 102131 ISI-97006-B ISI Fl.10a H- 3 Restraint AUG .............. 1 Recirc Manifold B OWN PRE F-A 102132 ISI-97006-B ISI c F1.10a H- 6 M1_I4-PlRF22 ISI /VT /PEI- Restraint AUG 1 Recirc Manifold B 02.05.02 OWN PRE F-A 102133 ISI-97006-B ISI Fl,10a H- 7 Restraint AUG 1 Recirc Manifold B OWN PRE F-A 102134 ISI-97006-B IS - F1.10a H- 8 Restraint AUG 1 Recirc Manifold B OWN PRE . . ... PSI-97006-B ISI . . . ... F-A 102135 AUG Restrainl F1.10a H- 9 OWN 1 Recirc Manifold B PRE F-A 102137 ISI-97006-B ISI F1.10a H-11 Restraint AUG 1 Recirc Manifold B OWN PRE - - - -- - -

                                                                               -97006-B        I  -       -.---                                  -

F-A 102139 Fl.10a H-13 Ml_14-P3_RF251iSIIVTI /FP-PE- Restraint AUG 1 Recirc Manifold B NDE-530 OWN ............... PRE F-A 105019 ISI-782A ISI F1.10a H- 5 Hanger AUG 1 Head Vent OWN PRE .............. F-A 105020 ISiI-782A~ Is1 C F1.10a H-5A M1_14-P2_RF23 / SII /VT I PE- Hanger AUG I Head Vent 02.05.02 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 90 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 cl '0 O (D 0 0 0 0 0 N 1 NI

                                                                                                         -     N Category,                                                                      DwgIlSO No.         -      M     0 N
                                                                                                                     '4 10       14 Item No.,                                                                      Comp. Desc.         N LL CI U.

N4 LL N% LI. LI. Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 105021 ISI-782A ISI F1.10a H-7 Hanger AUG 1 Head Vent OWN PRE F-A 105023 ISI-782A-A ISI Fl.10a H- I M1_14-P1_RF21 / PSI IVTI / PEI- Hanger AUG 1 Head Vent 02.05.02 OWN M1_ 4-PlRF22 / PSI /VTI/ / PEI-02.05.02 M1_14-P2_RF23 I PSI IVTI / PEI-02.05.02 M1_14-P2_RF24 / PSI /VTI / FP-PE-NDE-530 Ml_14-P3_RF25 / PSI IVTI/ /FP-PE-NDE-530 PRE ISI B B B B b - - - 105025 ISI-782A-A F-A F1.10a H-3 M1_I4-P1_RF21 IPSIIVTI/ PEI- Hanger AUG 1 Head Vent 02.05.02 OWN Ml_14-PlRF22 / PSI / VT / / PEI-02.05.02 M1_14-P2_RF23 / PSI/V-I / PEI-02.05.02 Ml_14-P2_RF24 / PSI IVTI I FP-PE-NDE-530 M1_14-P3_RF25 / PSI /VT/ / FP-PE-NDE-530 PRE B B B B b F-A 105052 ISI-821A ISI F1.10a H- 9 Support AUG 1 Bottom Head Drair OWN PRE F-A 100578 Ml_I4-P2_RF24 / PSI /VT/ /FP-PE- ISI-13142-33-A ISI Fl.10b H-2 NDE-530 Snubber / Clamp AUG I Main Steam A M1_I4-P3_RF25 / PSI / VT /FP-PE- OWN NDE-530 PRE C b Printed 6

0 Monticello Nuclear Generating Plant 4th Interval IS[ Plan (Rev. 4) Page 91 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00 0 0 00 0 0 N N N4 N Category, DwgIlSO No. - N M) - Item No., Comp. Desc. N IL N L. N ML Nl IL U-Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 100579 IS1-13142-33-A iSI Fl.10b H-3 M1 14-Pl RF21 /PSIIVTI /PEI- Snubber/Clamp AUG 1 Main Steam A 02.05.02 OWN PRE c F-A 100582 M1 14-Pl RF22 / PSI / VT / /PEI- IS1-13142-33-A ISI c F1.10b H-S6 02.05.02 Snubber/ Clamp AUG 1 Main Steam, M1_I4-P2_RF23 / ISI VTI/ PEI- OWN 02.05.02 M1_I4-P2_RF23 / PSI /VT/ I PEI-02.05.02 PRE B B F-A 100583 M1_I4-P2-RF24 / PSI / VT/ / FP-PE- ISI-13142-33-A ISI F1.10b H- 7 NDE-530 Snubber / Clamp AUG 1 Main Steam A M1_I4-P3_RF25 / PSI IVT / FP-PE- OWN NDE-530 PRE-c - - - - - F-A 100631 M1_I4-P2_RF24 / PSI VTI I FP-PE- ISI-13142-34-A ISI F1.10b H- 2 NDE-530 Snubber/ Clamp AUG 1 Main Steam B M1lI4-P3_RF25 / PSI /VT / / FP-PE- OWN NDE-530 PRE- --- - - b - - - F-A 100632 IS1-13142-34-A ISI Fl.10b H-3 M1_4-P1_RF21 IPSI/VT/ /PEI- Snubber / Clamp AUG 1 Main Steam B 02.05.02 OWN PRE H F-A 100682 Ml 14-P1_RF22 /PSI /VT/ /PEI- IS1-13142-35-A ISI Fl.10b H-2 02.05.02 Snubber / Clamp AUG . . . ... . . ... . . . 1 Main Steam C Ml_14-P2_RF23 / PSI / VT / / PEI- OWN 02.05.02 PRE c - B F-A 100683 IS1-13142-35-A ISI c F1.10b H- 3 M1_l4-P2_RF23 / PSI / VT / / PEI- Snubber / Clamp AUG 1 Main Steam C 02.05.02 OWN PRE -B - B F-A 100737 1IS1-13142-36-A, ISI F1.10b H-2 Ml_14-PlRF21 /PSI/IVT /PEI- Snubber/Clamp AUG 1 Main Steam C 02.05.02 OWN PRE C Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 92 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 C.) La lr. im 030 0 0 0 0 0 0 C4J C14 C'4 0' Category, DwgIISO No. mle

                                                                                                        -      ('C41 Item No.,                                                                     Comp. Desc.                LL      LL       LL    ILL     U.

Class Summary No.lComplD/System Scope I Method I Procedure Code Case F-A 100740 ISI-13142-36-A ISI F1.10b H-5 M1_I4-PFRF221PSI/VTI /PEI- Snubber / Clamp AUG 1 Main Steam C 02.05.02 OWN PRE - B F-A 100741 Ml_14-PI_RF21 / PSI IVTI I PEI. ISI-13142-36-A ISI----------------- s - - - Fl.10b H- 6 02.05.02 Snubber / Clamp AUG 1 Main Steam C M1_I4-P3_RF25 / ISI /VT IFP-PE- OWN NDE-530 PRE H F-A 101042 ISI-13142-43-A ISI Fl.10b H-2 Strut / Clamp AUG 1 RCIC Steam OWN PRE F-A 101198 Ml_14-P2_RF23/PSIIVT/ /PEI. IS1-13142-52-A ISI F1.10b H-4 02.05.02 Snubber / Clamp AUG 1 Feedwatei Ml_14-P2_RF24 /PSI /VT/ /FP-PE- OWN NDE-530 PRE - - - - c B F-A 101201 M1_14-P2_RF23 / PSI / VT / / PEI- ISI-13142-52-A ISI Fl.10b H-7 02.05.02 Snubber / Clamp AUG 1 Feedwatei M1_I4-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 PRE - - - - c B F-A 101253 Ml_14-P2_RF23/ PSI / VT / / PEI- ISI-13142-53-A SI - c Fl.10b H- 5 02.05.02 Snubber/Clamp AUG 1 Feedwatel M1_I4-P2_RF24 / PSI /VT/ / FP-PE- OWN NDE-530 PRE - t c B F-A 101255 Ml_14-P2_RF23 / PSI / VT/ / PEI- ISI-13142-53-A ISI F1.lOb H-7 02.05.02 Snubber / Clamp AUG 1 Feedwatei Ml_14-P2_RF24 /PSI/VT / FP-PE- OWN NDE-530 PRE -Ic B - Printed 1

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 93 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3

                                                                                                       '-        t-   o.

N N NI 04 cm C.1 Category, DwgIISO No. N Item No., Comp. Desc. NN LL U. CM U. N U. IL Class Summary NoJComplDlSystem Scope / Method I Procedure Code Case F-A 101855 ISI-97003-A ISI Fl.10b H-3 MI_14-P1_RF21 IPSI/VT/ /PEI- Snubber/Clamp AUG 1 RHR Return A 02.05.02 OWN M1_I4-PIRF22/PSI/VTI /PEI-02.05.02 - M1 14-PIRF22/PSI/VT/ /PEI-02.05.02 M1_I4-P2_RF23 / PSI /VT I PEI-02.05.02 PRE c B B - - F-A 101856 ISI-97003-A ISI F1.10b H- 4 Snubber / Clamp AUG 1 RHR Return A OWN PRE F-A 101896 ISI-97003-B ISI Fl.10b H- 2 M1_14-PlRF22 / PSI / VT / PEI- Snubber/Clamp AUG 1 RHR Suction A 02.05.02 OWN .............. M1_I4-P2_RF23 / PSI IVT/ /PEI-02.05.02 M1_I4-P2_RF23 / PSI /VTI / PEI-02.05.02 Ml_14-P2_RF24 / PSI /VTI / FP-PE-NDE-530 PRE - c B B F-A 101897 ISI-97003-B ISI Fl.10b H-3 Snubber / Clamp AUG 1 RHR Suction A OWN PRE F-A 101935 M1_I4-P2_RF23 /PSI/ VT/ / PEI- ISI-97004-A ISI F1.10b H- 3 02.05.02 Snubber / Clamp AUG 1 RHR Return B Ml114-P2_RF24 / PSI / VT// FP-PE- OWN .............. NDE-530 PRE B

                                                                                                                - c F-A         101936                        M1_14-P2_RF23 / PSI / VT I / PEI-  ISI-97004-A     ISI                        C Fl.lOb      H- 4                          02.05.02                           Snubber / Clamp AUG 1           RHR Return B                                                                     OWN M1_I4-P2_RF24 / ISI/VT/ /   FP-PE-NDE-530 Ml_4-P2_RF24 / PSI /VTI /FP-PE-NDE-530 PRE                   c    B   -   -

Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 94 of 370 ASME Section X1 (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 N Category, DwgIISO No. ox w of0 10 Item No., Comp. Desc. U. LL LL. L Nl L-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102195 IS1-97027-A I -- Fl.10b H- 2 Snubber / Clamp AUG 1 RHR Equalizer OWN PRE F-A 102196 ISI-97027-A ISI ---------------- s - - - F1.10b H- 3 M1 14-P3_RF25/ ISI/VT/ / FP-PE- Snubber/Clamp AUG 1 RHR Equalizer NDE-530 OWN PRE F-A 105051 ISI-821A ISI Fl.10b H- 8 Strut / Clamp AUG 1 Bottom Head Drair OWN PRE F-A 107528 ISI-97005-A ISI - r - c Fl.10b H- 8 M1_I4-P2_RF23 / ISI / VT /PEI- Snubber/Lugs AUG 1 Recirculation A 02.05.02 OWN Ml_14-P2_RF23 / PSI/ VT/ / PEI-02.05.02 Ml_14-P2_RF23 / PSI /VT/ / PEI-02.05.02 M1_4-P2_RF24 / PSI /VT/ / FP-PE-NDE-530 PRE - c B B F-A 107530 M1_I4-P2_RF24 / PSI / VT / / FP-PE- ISI-97005-B ISI Fl.10b H-2 NDE-530 Snubber/Lugs AUG 1 Recirc Manifold A M1_I4-P3_RF25 / PSI / VT / / FP-PE- OWN NDE-530 PRE- ----- c b - - - F-A 107533 ISI-97005-B ISI B F1.10b H-12 M1 _14-PlRF21 / ISI /VT / I PEI- Snubber/Lugs AUG . . . ... . . ... . . . 1 Recirc Manifold A 02.05.02 OWN PRE F-A 107547 M1_I4-P2_RF23 / PSI / VT / / PEI- I51-97006-A ISI S F1.10b H- 8 02.05.02 Snubber/ Lugs AUG 1 Recirculation B OWN Ml_I14-P2_RF24 / PSI / VT / / FP-PE-NDE-530 M1_14-P3_RF25 I ISI / VT/ /FP-PE-NDE-530 PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 95 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M an Cý4 a CDI o 0

o. o 0 N N N N N Category, DwgIlSO No. N ua Item No., Comp. Desc. NN N1 N LL Class LL of, L.

w, wg U. LL Ix* N: Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 107549 ISI-97006-B ISI Fl.10b H-2 M1_I4-Pl RF21 /PSIIVTI /PEI- Snubber/Lugs AUG 1 Recirc Manifold B 02.05.02 OWN PRE c F-A 107552 Ml-14-Pl-RF21 / PSI / VT / / PEI- ISI-97006-B ISI Fl.10b H-12 02.05.02 Snubber AUG . . . . .. . . . .. . . . . 1 Recirc Manifold B Ml1_4-PlRF21 / PSI / VT/ / PEI- OWN 02.05.02 PRE c F-A 100495 ISI-13142-31-A ISI Fl.10c H-1 Spring / Clamp AUG 1 Core Spray A OWN PRE F-A 100496 ISI-13142-31-A ISI ------------------- c FI.10c H-2 M1_I4-P2_RF24 / ISI /VT / / FP-PE- Spring / Clamp AUG Core Spray A NDE-530 OWN PRE F-A 100580 ISI-13142-33-A ISI -r ------------ s - - - Fl.10c H-4 M1_14-P3_RF25/ISI/VT/ /FP-PE- Variable/SprinC AUG 1 Main Steam A NDE-530 OWN PRE F-A 100581 ISI-13142-33-A ISI Fl.10c H-5 Spring / Clamp AUG 1 Main Steam A OWN PRE . . . ... . . ... . . . F-A 100633 ISI-13142-34-A IS - Fl.10c H- 4 Spring / Clamp AUG 1 Main Steam B OWN PRE F-A 100684 IS1-13142-35-A SI-s Fl.c H-4 M1_I4-P3_RF2511SIIVT- /FP-PE- Spring/Clamp AUG 1 Main Steam C NDE-530 OWN PRE F-A 100738 ISI-13142-36-A ISI F1.10c H- 3 Spring / Clamp AUG 1 Main Steam C OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 96 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 04 N N Category, DwgIlSO No. Nm N Item No., Comp. Desc. LI. U. IL. w w Class Summary NoJComplD/System Scope I Method I Procedure Code Case F-A 100739 ISI-13142-36-A ISI F1.10c H- 4 Variable / Spring AUG 1 Main Steam C OWN PRE . .. . ... . . ... . . . F-A 100975 ISI-13142-42-A ISI Fl.10c H- 1 Variable Spring AUG 1 HPCI Steam OWN PRE F-A 101041 IS1-13142-43-A ISI F1.10c H- 1 M1_14-PlRF21 / PSI /VT-/ PEI- Variable Spring AUG 1 RCIC Steam 02.05.02 OWN PRE H F-A 101194 ISI-13142-52-A ISI F1.10c H-i1 Spring / Clamp AUG .............. 1 Feedwatei OWN PRE F-A 101196 ISI-13142-52-A ISI F1.10c H- 2 Variable / Spring AUG 1 Feedwatei OWN PRE - - - F-A 101199 ISI-13142-52-A ISI F1.10c H-5 Spring / Clamp AUG 1 Feedwatei OWN PRE F-A 101200 ISI-13142-52-A SI-- -- s- -- F1.10c H- 6 M1_I4-P3_RF25 ISI VT/ /FP-PE- Spring / Clamp AUG 1 Feedwatei NDE-530 OWN PRE . . . . . . . . . . . . . . F-A 101249 ISI-13142-53-A ISI Fl.10c H- 1 Spring AUG ............. 1 Feedwatei OWN ............... PRE F-A 101250 VIa-13142-53-A ISI FI.IOc H- 2 Variable / Spring: AUG 1 Feedwatei OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 97 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M La C, 0 0 0 N N C4 Category, DwgIlSO No. LL. w) Item No., Comp. Desc. NN uL U. NN uL IjL N4 U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 101252 ISI-13142-53-A IS1 F1.10c H-4 Spring / Clamp AUG 1 Feedwatei OWN PRE F-A 101254 ISI-13142-53-A ISI F1.10c H- 6 Spring / Clamp AUG 1 Feedwatei OWN PRE F-A 101360 ISI-73880-A ISI Fl.10c H- 1 Spring / Clamp AUG 1 Reactor Wtr Cleanup OWN PRE F-A 101617 ISI-821A ISI Fl.10c H- 1 Clamp AUG 1 Bottom Head Drair OWN PRE F-A 101619 ISI-821A ISI Fl.10c H- 3 Clamp AUG 1 Bottom Head Drair OWN PRE F-A 101623 ISI-821A SI------------------ S - - - Fl.10c H-7 Ml_14-P3_RF25 ISI VT/ IFP-PE- Clamp AUG 1 Bottom Head Drair NDE-530 OWN PRE F-A 101853 ISI-97003-A ISI F1.10c H- I Variable Spring AUG 1 RHR Return A OWN PRE F-A 101854 ISI-97003-A ISI F1.10c H-2 Variable Spring AUG 1 RHR Return A OWN PRE F-A 101895 ISI-97003-B ISI FI.IOc H- 1 Variable Spr /IClamr AUG 1 RHR Suction A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 98 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 Cq LO t, CD 0 0 00 0 0 00 NI NI NI N OIn Category, DwgIlSO No. - N MI

• Item No., Comp. Desc. N1 LL U.

NI N N LL L. U4 LL Class Summary No.IComplDISystem W, Scope I Method I Procedure Code Case F-A 101898 ISI-97003-B ISI F1.10c H-4 Variable Slide/Clamf AUG 1 RHR Suction A OWN PRE F-A 101899 ISI-97003-B IS1 F1.!0c H- 5 Variable Slide AUG 1 RHR Suction A OWN PRE F-A 101933 ISI-97004-A ISI F1.10c H- I Variable / Slide AUG 1 RHR Return B OWN PRE F-A 101934 ISI-97004-A ISI - - - - c F1.10c H-2 M1_I4-P2_RF23/ ISI /VT/ /PEI- Variable/Slide/Clamf AUG 1 RHR Return B 02.05.02 OWN PRE F-A 101975 ISI-97005-A SI c F1.10c H-11 M1_I4-PlRF21 /ISI /VT / PEI- Sway Brace AUG 1 Recirculation A 02.05.02 OWN PRE F-A 101978 ISI-97005-A ISI Fl.10c H- 3 Sway Brace AUG 1 Recirculation A OWN PRE F-A 102080 ISI-97006-A ISI F1.10c H- 2 Sway Brace AUG 1 Recirculation B OWN PRE F-A 102096 ISI-97006-A ISI F1.10c H-11 Sway Brace AUG 1 Recirculation B OWN PRE F-A 102194 ISI-97027-A ISI Fl.1Oc H- 1 Spring / Clamp AUG 1 RHR Equalizer OWN PRE Printed 6/5/4

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 99 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o a o a a 0 S01 0 o 0 N N4 Category, DwgIlSO No. -N MI -e N Item No., Comp. Desc. w w N U. N LA. w.IL Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 102197 ISI-97027-A IS- ---------------- s - - - Fl.10c H-4 M1_I4-P3_RF2511/SII/TI /FP-PE- Spring/Clamp AUG 1 RHR Equalizer NDE-530 OWN PRE F-A 105022 M 1_4-P2_RF23 / ISI / VT /PEI- ISI-782A ISI - - - - C Fl.1c H-17 02.05.02 Spring / Clamp AUG 1 Head Ven Ml1_I4-P2_RF23 / PSI / VT / I PEI- OWN 02.05.02 Ml_14-P2_RF24 / PSI / VT / / FP-PE-NDE-530 PRE - - - - c B F-A 105024 ISI-782A-A IS - Fl.10c H-2 M1_I4-PlRF21 /PSI/VT / PEI- Spring / Clamp AUG 1 Head Vent 02.05.02 OWN Ml_14-PlRF22 / PSI / VT / / PEI-02.05.02 M1_I4-P2_RF23 / PSI /VT/ / PEI-02.05.02 Ml_14-P2_RF24 / PSI /VT/ / FP-PE-NDE-530 Ml 14-P3_RF25 / PSI / VT / /FP-PE-NDE-530 PRE B B - B B b - - - F-A 107511 ISI-13142-33-A SI c Fl.10c H- I Ml_14-PlRF21 / ISI /VT/ /PEI- Dbl Spring / 4 Lugs AUG 1 Main Steam A 02.05.02 OWN PRE F-A 107512 ISI-13142-33-A ISI F1.10c H- 8 Dbi Spring / 4 Lugs AUG 1 Main Steam A OWN PRE F-A 107513 I,..I-13142-3I4-A I1I C FI.10c H- 1 Ml-14-PlRF21 / ISI / VT / / PEI- Dbi Spring / 4 Lugs AUG 1 Main Steam 1 02.05.02 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 100 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M O o a o 0 N o41 N*1 N N Category, DwgIlSO No. M,I -e 1 U-Item No., Comp. Desc. C1 oL LL U. N U/. LL U. U- w Class Summary No.lComplD/System Scope / Method I Procedure Code Case F-A 107514 IS1-13142-34-A ISI F1.10c H- 5 Dbl Spring /4 Lugs AUG 1 Main Steam E OWN PRE F-A 107515 ISI-13142-35-A ISI Fl.10c H- 1 DbI Spring /4 Lugs AUG 1 Main Steam C OWN PRE F-A 107516 ISI-13142-35-A ISI F1.10c H- 5 M1_14-P1_RF21 PSI / VT/I PEI- Dbl Spring / 4 Lugs AUG 1 Main Steam C 02.05.02 OWN PRE c F-A 107517 ISI-13142-36-A ISI F1.10c H- 1 DbI Spring /4 Lugs AUG 1 Main Steam C OWN PRE F-A 107518 ISI-13142-36-A ISr F1.10c H- 7 Dbi Spring / 4 Lugs AUG 1 Main Steam C OWN PRE F-A 107519 ISI-13142-52-A SI - r ----------- s - - - Fl.10c H-3 M1_14-P3_RF2511SI1VTI IFP-PE- DblSpring/4 Lugs AUG 1 Feedwatej NDE-530 OWN PRE F-A 107520 ISI-13142-52-A ISI .............. F1.10c H- 8 Dbi Spring / 4 Lugs AUG 1 Feedwatei OWN PRE . . . ... . . ... . . . F-A 107521 ISI-13142-53-A IS - F1.10c H- 3 Dbl Spring / 4 Lugs AUG 1 Feedwatei OWN . . . ... . . ... . . . PRE F-A 107522 151-13142-53-A ISI C F1.IOc H- 8 M1 14-P2_RF24 lISt / / / I FP-PE- Dbi Spring/ 4 Lugs AUG 1 Feedwatei NDE-530 OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 101 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 La o 0 oo a0 0 o 0 N N1 N CN N Category, DwgIlSO No. u4 iz4

                                                                                                                - 1N.4                       in Item No.,                                                                    Comp. Desc.                                         N   N        N w    w        U. u-        U-Class      Summary No.IComplDISystem      Scope I Method I Procedure         Code Case F-A         107523                                                           ISI-73880-A                     ISI   -    -  -  -  c F1.10c      H- 3                          M1 14-P2_RF23    ISI/I /V- PEI-    Dbl Spring / 4 Lugs             AUG 1           Reactor Wtr Cleanup           02.05.02                                                           OWN PRE F-A         107524                                                           ISI-97005-A                     ISI  E F1.10c      H- 2                          M1_I4-PlRF21 /ISI/VT/ / PEI-       Dbl Spring / 4 Lugs             AUG 1           Recirculation A               02.05.02                                                           OWN PRE F-A         107525                                                           ISI-97005-A                     ISI F1.10c      H-5                                                              Clevis / Lugs / Constant-Suppor AUG 1          Recirculation A                                                                                   OWN PRE F-A         107526                                                           ISI-97005-A                     ISI  c F1.10c     H- 6                           M1 14-PlRF21 / ISI / VT / I PEI-   Clevis / Lugs I Constant-Suppor AUG 1          Recirculation A                02.05.02                                                           OWN PRE F-A         107527                                                           ISI-97005-A                     ISI  E F1.10c     H- 7                           M1_I4-Pl-RF21 / ISI/VTI / PEI-     Double Spring                   AUG 1          Recirculation A                02.05.02                                                           OWN PRE F-A         107529                                                           ISI-97005-A                     ISI  c     r F1.M1c     H-1                            M1_4-PlRF21 /ISI/VT /PEI-          Dbl Spring / 4 Lugs             AUG 1          Recirculation A                02.05.02                                                           OWN PRE F-A         107531                                                           ISI-97005-B                     IS1 Fl.10c     H-M6                                                              Spring / Lugs                   AUG 1           Recirc Manifold A                                                                                OWN PRE F-A         107532                                                           ISI-97005-B                     ISI Fl.10c     H-11                                                              Spring / Lugs                   AUG 1          Recirc Manifold A                                                                                 OWN PRE F-A         107543                                                           ISI-97006-A                     ISI  C F1.10c     H- 3                           M1_I4-PlRF21 / ISI IVTI IPEI-      Dbl Spring / 4 Lugs             AUG 1          Recirculation B                02.05.02                                                           OWN PRE Printed  6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 102 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 0 0 0 0 0 0 0 N N*1 1 ,10 Category, DwgIlSO No. LI Item No., Comp. Desc. C-4 C14 N4 N w. UL LI w, w L. nw w LI Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 107544 ISI-97006-A ISI F1.10c H- 5 Clevis / Lugs / Constant-Suppor AUG 1 Recirculation B OWN PRE . . . ... . . ... . . . F-A 107545 ISI-97006-A ISI FI 10c H-6 Clevis I Lugs / Constant-Suppor AUG 1 Recirculation B OWN PRE F-A 107546 ISI-97006-A ISI c Fl.10c H-7 M1_I4-P1_RF21 IISIIVTI /PEI- Dbl Spring / Lugs AUG 1 Recirculation B 02.05.02 OWN . . . ... . . ... . . . PRE F-A 107548 ISI-97006-A ISI E Fl.10c H-10 M1_14-P1_RF21 /ISI /VT/ /PEI- DbI Spring AUG 1 Recirculation B 02.05.02 OWN PRE F-A 107550 ISI-97006-B ISI Fl.10c H- 5 Spring / Lugs AUG 1 Recirc Manifold B OWN PRE ISI-97006-B PSI F-A 107551 Spring I Lugs AUG F1.10c H-10 OWN 1 Recirc Manifold B PRE F-A 100098 ISI-13142-17-B IS - - - - c F1.20a H- 1 Ml_14-P2_RF23 /ISI/ VT I I PEI- Rod / Clevis AUG .............. 2 HPCI Water Side Sctn 02.05.02 OWN PRE F-A 100100 ISI-13142-17-B ISi . . . ... . . ... . . . F1.20a H- 3 Restraint AUG OWN 2 HPCI Water Side Sctn PRE F-A 100101 ISI-13142-17-B ISI F1.20a H- 4 Restraint AUG 2 HPCI Water Side Sctn OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 103 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0D0 I 0 I 0 NI Category, DwgIlSO No. 004C1 0404 C-4 N Item No., Comp. Desc. U-wN nU LL U U. I.N U. N I~, LL Code Case wg Class Summary No.IComplD/System Scope / Method I Procedure F-A 100169 ISI-13142-18-A ISI F1.20a H-2 Box Restrain1 AUG 2 RHR Discharge B OWN PRE F-A 100171 ISI-13142-18-A SI------------------ s - - - F1.20a H-4 M1 _I4-P3_RF251ISII V-FI/FP-PE- Box Restrainl AUG 2 RHR Discharge B NDE-530 OWN PRE F-A 100176 ISI-13142-18-A ISi F1.20a H-1I Seismic Restraint AUG 2 RHR Discharge B OWN PRE F-A 100364 IS1-13142-20-B ISI F1.20a H- 4 Variable / Suppor AUG 2 Core Spray B OWN PRE F-A 100365 IS1-13142-20-B ISI - - - - c F1.20a H- 5 M1_14-P2_RF23/ ISI /VT/ /PEI- Variable / Suppor AUG 2 Core Spray B 02.05.02 OWN PRE F-A 100876 ISI-13142-40-A ISI F1.20a H- 2 Box Restrainl AUG 2 HPCI Water Side Dsch OWN PRE F-A 100877 ISI-13142-40-A ISI F1.20a H- 3 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100880 IS1-13142-40-A ISI F1.20a H-6 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100881 IS1-13142-40-A ISI F1.20a H-7 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 104 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 1 N cm Category, Dwg/ISO No. N La N Item No., Comp. Desc. C1 uL LL N N N U LL U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 100883 ISI-13142--40-A ISI F1.20a H- 9 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100884 ISI-13142-40-A ISI F1.20a H-10 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100885 ISI-13142-40-A ISI c F1.20a H-1I Ml1_4-P1_RF21 I/SIIVT/ /PEI- Rod/Clamp AUG 2 HPCI Water Side Dsch 02.05.02 OWN PRE F-A 100886 ISI-13142-40-A ISI F1.20a H-12 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100888 ISI-13142-40-A ISI F1.20a H-14 Seismic Restrainl AUG . . . ... . . . 2 HPCI Water Side Dsch OWN PRE F-A 100912 ISI-13142-40-B ISI F1.20a H- 1 Rod / Clamp AUG 2 HPCI Water Side Dsch OWN PRE ISI F-A 100914 ISI-13142-40-B F 1.2 0 a H- 3 S e is m ic R e s tra int AUG . . . . . . . . . . . . . . 2 HPCI Water Side Dsch OWN PRE F-A 100916 Ml_14-PlRF21 / ISI/VTI / PEI- ISI-13142-40-B ISI E S F1.20a H- 5 02.05.02 Box Restraini AUG 2 HPCI Water Side Dsch Ml1_4-PlRF21 /PSI/VT/ /PEI- OWN 02.05.02 M1 14-P3_RF25 / ISI / VT /FP-PE-NDE-530 PRE Printed 6/510

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 105 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00 C 0 0 CMi 0 0 C14 0 C%1 0 C14 N Category, DwgIlSO No. Item No., Comp. Desc.

                                                                                                - I, N

NI. Nq Cq) N I.I ~-N. C4 hL LL U. LL LL n, Class Summary No./ComplDISystem Scope / Method I Procedure Code Case F-A 100986 ISI-13142-42-A ISI F1.20a H-10 Rod / Clamp AUG 2 HPCI Steam OWN PRE F-A 100988 IS-13142-42-A ISI F1.20a H-12 Box Restraint AUG 2 HPCI Steam OWN PRE F-A 101708 ISI-93268-1A ISI F1.20a H- 7 Box Restrain1 AUG 2 CRD Scram Header A OWN PRE F-A 101711 ISI-93268-1A ISI F1.20a H-10 Restraint AUG 2 CRD Scram Header A OWN PRE F-A 101712 ISI-93268-1A SI c F1.20a H-11 M1_14-P1_RF21 IISI/VTI /PEI- Rod/Clamp AUG 2 CRD Scram Header A 02.05.02 OWN PRE F-A 101802 ISI-93268-3A ISI F1.20a H- 5 Box Restrainl AUG 2 CRD Scram Header B OWN PRE F-A 101803 ISI-93268-3A ISI F1.20a H-6 Rod / Clamp AUG 2 CRD Scram Header B OWN PRE F-A 101806 ISI-93268-3A ISI F1.20a H-8 Box Restraini AUG 2 CRD Scram Header B OWN PRE F-A 101812 ISI-93268-3A ISI F1.20a H-13 Box Restraini AUG 2 CRD Scram Header B OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 106 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 00 00 0 0 0 0 NI NI eIq Category, DwgIlSO No.

                                                                                                                 -     N      CI VI Item No.,                                                                     Comp. Desc.                         C14 U.

CMJ U. CJ CM LL LL IL U. w, Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 106811 IS1-13142-19-B ISI F1.20a H- 2 Restraint Hanget AUG 2 RCiC Steam Discharge OWN PRE F-A 106812 ISI-13142-19-B ISI F1.20a H- 3 Restraint Hangei AUG 2 RCIC Steam Discharge OWN PRE F-A 106813 IS1-13142-19-B IS - F1.20a H-4 Restraint Hangei AUG 2 RCIC Steam Discharge OWN PRE F-A 106814 ISI-13142-19-B ISI----------------- - - - - F1.20a H-5 Ml_14-P3_RF251ISIIVTI /FP-PE- RestraintHangei AUG 2 RCIC Steam Discharge NDE-530 OWN PRE F-A 106815 ISI-13142-19-B ISI F1.20a H- 6 Restraint Hanget AUG 2 RCIC Steam Discharge OWN PRE F-A 106821 ISI-13142-26-B ISI F1.20a H- 2 Restraint Hange= AUG 2 CORE SPRAY B OWN PRE F-A 106822 ISI-13'142-26-B IS - F1.20a H- 3 SEISMIC RESTRAINT HANGER AUG 2 CORE SPRAY B OWN PRE F-A 106823 IS1-13142-26-B ISI F1.20a H- 4 Restraint Hangei AUG 2 CORE SPRAY B OWN PRE F-A 106824 ISI-13142-26-B ISI F1.20a H- 5 SEISMIC RESTRAINT HANGER AUG 2 CORE SPRAY B OWN PRE Printed 615/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 107 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 C') LO o 0 Co ooo o o o N N Category, DwgIISO No. N

                                                                                                          - 1       .

Item No., Comp. Desc. N U. N1 IL w N U-LL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 106825 ISI-13142-26-B ISI - - - - c F1.20a H-6 M1_I4-P2 RF23 / ISIVT / PEI- RESTRAINT HANGER TWH-121 AUG . . . ... . . ... . . . 2 CORE SPRAY B 02.05.02 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 106826 IS1-13142-26-B ISI . . . .. . . . .. . . . . F1.20a H- 7 SEISMIC RESTRAINT SR-646 AUG . . . ... . . ... . . . 2 CORE SPRAY B OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . ISI . . . ... . . ... . . . 106827 ISI-13142-26-B F-A F1.20a H- 8 RESTRAINT HANGER TWH-102 AUG 2 CORE SPRAY B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 106830 ISI-13142-26-C ISI F1.20a H- 1 Restraint Hangei AUG . . . .. . . . .. . . . . 2 CORE SPRAY B OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 106831 ISI-13142-26-C ISI F1.20a H-2 Seismic Restraint AUG 2 Core Spray B Discharge OWN PRE F-A 106832 ISI-13142-26-C ISI c r . .. . . . .. . . . . F1.20a H- 3 M1_14-PlRF21 /ISI /VT/ PEI- Restraint Hange= AUG 2 Core Spray B Discharge 02.05.02 OWN PRE . . . ... . . ... . . . F-A 106833 ISI-13142-26-C ISI . . . . . . . . . . . . . . F1.20a H-4 Seismic Restraint AUG 2 Core Spray B Discharge OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . F-A 106834 ISI-13142-26-C ISI F1.20a H- 5 SEISMIC RESTRAINT SR-647 AUG 2 Core Spray B Discharge OWN PRE F-A 106835 ISI-13142-26-C ISI F1.20a H-6 Restraint Hangel AUG 2 Core Spray B Discharge OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 108 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 001 0 0,0 00C Category, DwgIlSO No. C1 ICI 4 I

                                                                                                          -4       N                      C)   V                      (M Item No.,                                                                       Comp. Desc.                 N4 LL N

LL N4 LL N4 U Lig L0 x Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 106836 ISI-13142-26-C ISI c r F1.20a H-7 M1_14-Pl _RF21 / ISI / VT /PEI- Seismic Restraint AUG 2 Core Spray B DischargE 02.05.02 OWN PRE F-A 106837 IS1-13142-26-C ISI F1.20a H- 8 Seismic Restraint AUG . . . ... . . ... . . . 2 Core Spray B DischargE OWN .............. PRE F-A 106838 ISI-13142-26-C SI . . . ... . . ... . . . F1.20a H- 9 M1_I4-P2_RF23 / PSI / VT I I PEI- Restraint Hangei AUG 2 Core Spray B DischargE 02.05.02 OWN .............. PRE - - - - c F-A 106847 ISI-13142-26-D ISI F1,20a H-i1 Restraint Hangei AUG . . . . . . . OWN 2 Core Spray B DischargE PRE F-A 106850 ISI-13142-31-B ISI - --- s s------------ F1.20a H-1A M1_I4-P3_RF25 / ISI/VT/ FP-PE- Seismic Restraint AUG .............. 2 Core Spray A Discharge NDE-530 OWN PRE . . . ... . . ... . . . F-A 106851 ISI-13142-31-B ISI . . . ... . . ... . . . F1.20a H-1B MlI4-P2_RF24 / PSI/ VT FP-PE- Restraint Hangei AUG .............. 2 Core Spray A DischargE NDE-530 OWN .............. PRE-------- c--------c ISI-13142-31-B ISI . . . ... . . ... . . . F-A 106852 F1.20a H-IC Restraint Hangel AUG . . . . .. . OWN . . . ... . . ... . . . 2 Core Spray A DischargE P RE . . . . . . . . . . . . . . F-A 106853 ISI-13142-31-B ISI . . . ... . . ... . . . F1.20a H-iD Seismic Restraint AUG . . . . .. . OWN . . . ... . . ... . . . 2 Core Spray A DischargE P RE . . . . . . . . . . . . . . F-A 106854 IS1-13142-31-B ISI F1.20a H-1E Restraint Hangei AUG 2 Core Spray A DischargE OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 109 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cl) U) oý 0 0~ 0) 0 o0 e 0 o -4 N C, Category, DwgIlSO No. - 1N N N U, Item No., Comp. Desc. Ii. U- uL LiL N U-Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 106860 ISI-13142-31-C ISI . . . ... . . ... . . . F1.20a H- 1 Restraint Hangei AUG 2 Core Spray A Discharge OWN PR E . . . . . . . . . . . . . . F-A 106861 IS1-13142-31-C ISI F1.20a H- 2 Seismic Restrainl AUG 2 Core Spray A Discharge OWN PRE F-A 106862 IS1-13142-31-C ISI F1.20a H- 3 Restraint Hangei AUG 2 Core Spray A Discharge OWN PRE F-A 106863 ISI-13142-31-C ISI F1.20a H- 4 Seismic Restraint AUG 2 Core Spray A Discharge OWN PRE F-A 106864 ISI-13142-31-C IS - F1.20a H- 5 Restraint Hangei AUG 2 Core Spray A Discharge OWN PRE F-A 106865 ISI-13142-31-C IS - F1.20a H- 6 Seismic Restraint AUG 2 Core Spray A Discharge OWN PRE F-A 106866 ISI-13142-31-C ISI F1.20a H- 7 Restraint Hangei AUG 2 Core Spray A Discharge OWN PRE F-A 106867 ISI-13142-31-C IS! F1.20a H- 8 Restraint Hangei AUG 2 Core Spray A Discharge OWN PRE F-A 10l6880] RSI-1r142-t1-Ie I:I Restraint Hangei AUG F1.20a H- 1 2 Core Spray A Discharge OWN PRE Printed *6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 110 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 co U) o o o N o 0 N N Category, DwgIlSO No. 1 ,41 N N4 Item No., Comp. Desc. N LL N LL N

u. N ul. N U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 106881 ISI-13142-31-D ISI F1.20a H- 2 M1_I4-P2_RF23 PSI VT /PEI- Restraint Hangei AUG 2 Core Spray A Discharge 02.05.02 OWN PRE - - - - c F-A 106890 ISI-13142-41-A ISI F1.20a H- 1 Restraint Hanget AUG 2 RCIC WATER SUCTION OWN P RE . . . . . . . . . . . . . . .

F-A 106891 IS1-13142-41-A ISI F1.20a H-2 Restraint Hangei AUG 2 RCIC WATER SUCTION OWN PRE F-A 106895 ISI-13142-48-A ISI F1.20a H- I Restraint Hangei AUG 2 RHR SERVICE WATER OWN PRE F-A 106896 MI-14-PlRF21 /11SIVTI /PEI- ISI-13142-48-A ISI c F1.20a H- 2 02.05.02 Restraint Hangei AUG 2 RHR SERVICE WATER M1 14-PlRF21 / PSI IVTI IPEI- OWN 02.05.02 PRE F-A 106897 ISI-13142-48-A ISI F1.20a H- 3 Restraint Hangei AUG 2 RHR SERVICE WATER OWN PRE F-A 106898 ISI-13142-48-A ISI F1.20a H- 4 Restraint Hangej AUG 2 RHR SERVICE WATER OWN PRE F-A 106899 ISI-13142-48-A ISI F1.20a H- 5 Restraint Hangei AUG 2 RHR SERVICE WATER OWN PRE F-A 106900 ISI-13142-48-A ISI F1.20a H- 6 Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 111 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 Ir- a) w0 oD e o o N N Category, DwgIlSO No. U) w- N. Item No., Comp. Desc. N U. N LL IL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 106901 ISI-13142-48-A ISI Fl.20a H- 7 Restraint Hangei AUG 2 RHR SERVICE WATER OWN PRE F-A 106902 ISI-13142-48-A ISI F1.20a H- 8 Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE F-A 106910 ISI-13142-48-B ISI F1.20a H- I Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE F-A 106911 ISI-13142-48-B ISI F1.20a H-2 Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE F-A 106912 ISI-13142-48-B ISI F1.20a H-3 Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE F-A 106913 ISI-13142-48-B ISI c F1.20a H- 4 M1_14-Pl RF21 /SI/ VT/ /PEI- Seismic Restraint AUG 2 RHR SERVICE WATER 02.05.02 OWN PRE F-A 106914 ISI-13142-48-B ISI F1.20a H- 5 Seismic Restraint AUG 2 RHR SERVICE WATER OWN PRE F-A 106915 ISI-13142-48-B ISI F1.20a H-6 Restraint Hangei AUG 2 RHR SERVICE WATER OWN PRE F-A 106916 IS I-13142-48-B ISI c F1.20a H-7 M1_I4-PlRF21 / ISI/VT/ /PEI- Restraint Hangei AUG 2 RHR SERVICE WATER 02.05.02 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 112 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 oD 0 0 N N DwgIlSO No. ol V Category, N C' C141

                                                                                                         - INI                       in Item No.,                                                                    Comp. Desc.                   Nl LL N

LL LL U. N4 U. Class Summary No.IComplDISystem Scope / Method I Procedure Code Case F-A 106920 ISI-13142-51-A ISI F1.20a H- 2 Restraint Hangei AUG 2 RHR A OWN PRE F-A 106921 ISI-13142-51-A ISI F1.20a H- 3 Seismic Restraint AUG 2 RHR A OWN PRE F-A 106922 ISI-13142-51-A IS1 F1.20a H- 4 Restraint Hanget AUG 2 RHR A OWN PRE F-A 106923 IS1-13142-51-A IS1 F1.20a H- 5 Seismic Restraint AUG 2 RHR A OWN PRE F-A 106924 ISI-13142-51-A ISI F1.20a H-6 Restraint Hangei AUG 2 RHR A OWN PRE F-A 106946 ISI-13142-51-C IS-I -------------- - s - - - F1.20a H- 1 M1 _14-P3_RF25 /ISI Vr / /FP-PE- Restraint Hangei AUG 2 RHR B NDE-530 OWN PRE F-A 106947 IS1-13142-51-C iSI----------------- s - - - Fl.20a H- 2 Ml_14-P3_RF25 / 1SI/VT I / FP-PE- Seismic Restraint AUG 2 RHR B NDE-530 OWN PRE F-A 106948 ISI-13142-51-C ISI F1.20a H- 3 Seismic Restraint AUG OWN 2 RHR B PRE F-A 106949 ISI-13142-51-C ISI F1.20a H-4 M1_i4-P2_RF24 /PSI /VT/VT-3 / FP Restraint Hangei AUG 2 RHR B PE-NDE-530 OWN PRE c Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 113 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M La 0 om N C. Category, Dwg/ISO No.

                                                                                                          .-   c1N            w    w               V)

Item No., Comp. Desc. LL LIL N-Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 106971 ISI-13142-51-D ISI .. . . .. . . . .. . . . . F1.20a H- I M1_I4-P3_RF25 / PSI /VTI/VT-3 / FP Restraint Hanger AUG .. . . ... . . ... . . . 2 RHR B PE-NDE-530 OWN - - - - .... PRE----------r -- s--- ISI-13142-51-D PSI .. . . ... . . ... . . . F-A 106972 F1.20a H--2 M14-P2_RF24 / PSI /VT/VT-3 /FP Restraint Hangei AUG 2 RHR B PE-NDE-530 OWN PRE ---------- c F-A 106973 ISI-13142-51-D ISI F1.20a H- 3 Seismic Restrainl AUG 2 RHR B OWN PRE F-A 106974 IS1-13142-51-D ISI . . . .. . . . .. . . . . F1.20a H-4 Restraint Hangei AUG .. . . .. . . . .. . . . . 2 RHR B OWN PRE F-A 106980 IS1-13142-62 ISI F1.20a H- 1 Restraint Hangei AUG .. . . .. . . . .. . . . . 2 Fuel Pool Emergency Coolin! OWN PRE . . . . . . . . . . . . . . F-A 106961 IS1-13142-62 ISI ---------------- s - - - Fl.20a H-12 M14-P3_RF25 ISI VTr /FP-PE- Restraint Hangei AUG 2 Fuel Pool Emergency Coolini NDE-530 OWN PRE F-A 106982 ISI-13142-62 ISI F1.20a H- 3 Seismic Restrainl AUG 2 Fuel Pool Emergency Coolin! OWN PRE F-A 106983 ISI-13142-62 ISI F1.20a H- 4 Restraint Hanger AUG 2 Fuel Pool Emergency Cooling OWN PRE F-A 106984 ISI-n314H-62 Restraint Hanger AUG F1.20a H- 5 2 Fuel Pool Emergency Cooling OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 114 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M U) f- C" oN CD o M N o V C3, N Category, DwgIlSO No. Item No., Comp. Desc. N11 N uL LL N LL w wL cc Class Summary No.lCompID/System Scope I Method I Procedure Code Case F-A 106985 IS1-13142-62 ISI F1.20a H- 6 Seismic Restrainl AUG 2 Fuel Pool Emergency Coolin! OWN PRE F-A 106986 ISI-13142-62 ISI F1.20a H- 7 Seismic Restraint AUG 2 Fuel Pool Emergency Coolin! OWN PRE F-A 106990 IS1-13142-67 ISI c . . . . . . . . . . . . . F1.20a H- 1 M1_4-PlRF21 /ISI /VT/ PEI- Restraint Hangei AUG 2 Fuel Pool Emergency Coolin! 02.05.02 OWN . . . .. . . . ... . . . PRE F-A 106991 ISI-13142-67 IS1 . . . . . . . . . . . . . . F1.20a H-2 Restraint Hanget AUG 2 Fuel Pool Emergency Cooling OWN PRE F-A 106992 ISI-13142-67 ISI F1.20a H- 3 Restraint Hangei AUG 2 Fuel Pool Emergency Cooling OWN PRE . . . ... . . ... . . . F-A 106993 IS1-13142-67 IS1 . . . . . . . . . . . . . . F1.20a H- 4 Restraint Hangel AUG 2 Fuel Pool Emergency Cooling OWN PRE . . . ... . . ... . . . F-A 106994 ISI-13142-67 SI . . . . . . . . . . . . . . F1.20a H- 5 Restraint Hangel AUG . . . .. . . . .. . . . . 2 Fuel Pool Emergency Cooling OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . F-A 106995 ISI-13142-67 ISI . . . . . . . . . . . . . . . F1.20a H- 6 Restraint Hangej AUG 2 Fuel Pool Emergency Cooling OWN PRE F-A 106996 ISI-13142-67 ISI c F1.20a H- 7 M1_I4-P2_RF23 / ISI / VTr/ PEI- Seismic Restraint AUG 2 Fuel Pool Emergency Cooling 02.05.02 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 115 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M on 0) o3 0 0 o N N N Category, Dwg/ISO No. N CM cm co -e Item No., Comp. Desc. N U1 N- N Class Summary No.lComplDISystem Scope I Method I Procedure Code Case w w w* LL w L-F-A 106997 ISI-13142-67 ISI F1.20a H- 8 Restraint Hangei AUG 2 Fuel Pool Emergency Cooling OWN PRE F-A 106998 ISI-13142-67 ISI F1.20a H- 9 Restraint Hangei AUG 2 Fuel Pool Emergency Cooling OWN PRE F-A 106999 ISI-13142-67 IS - F1.20a H-10 Seismic Restraint AUG 2 Fuel Pool Emergency Coolin! OWN PRE F-A 107028 ISI-13142-37-D IS - F1.20a H- I Restraint Hangei AUG 2 RHR A OWN PRE F-A 107029 ISI-13142-37-D ISI F1.20a H- 2 Restraint Hangei AUG . . . ... . . ... . . . 2 RHR A OWN PRE F-A 107048 ISI-13142-37-E ISI . . . .. . . . .. . . . . F1.20a H- 1 Restraint Hangei AUG 2 RHR A OWN PRE F-A 107050 ISI-13142-37-E ISI F1.20a H- 3 Restraint Hangei AUG 2 RHR A OWN PRE F-A 107587 ISI-93268-1A IS - F1.20a H-20 Tank Support AUG 2 CRD Scram Header A OWN PRE F-A 10758U RsI-93n4u-3A iS[ Restraint / 4 Lugs AUG F1.20a H- 7 2 CRD Scram Header B OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 116 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 0 0 00Q 0 0 0 0 0 N1 Category, DwgIlSO No. N N M N* in Item No., Comp. Desc. NI w Ix. N4 Ix N N L-Class Summary No.IComplD/System Scope I Method I Procedure Code Case F-A 107589 ISI-93268-3A ISI - - - - c F1.20a H-14 M1_I4-P2_RF23 /ISI /VT/ /PEI- Tank Support AUG 2 CRD Scram Header B 02.05.02 OWN PRE F-A 107600 ISI-13142-18-B ISI F1.20a H-2A Seismic Restraint AUG 2 RHR Discharge B OWN PRE F-A 107603 ISI-13142-51-A ISI F1.20a H- 7 Restraint Hangel AUG 2 RHR A OWN PRE ISI F-A 107750 ISI-13142-26-B F1.20a H-1A M1_I4-P2_RF23 /PSI /VT /VT-3/ SEISMIC RESTRAINT HANGER AUG 2 CORE SPRAY B PEI-02.05.02 OWN PRE - - - - c F-A 100056 ISI-13142-17-A ISI F1.20b H- 1 Strut / Clamp AUG 2 RHR Suction A OWN PRE F-A 100057 ISI-13142-17-A ISI c r F1.20b H-2 M1_14-PlRF21 I/SI/VT/ /PEI- Slide / Clamp AUG 2 RHR Suction A 02.05.02 OWN PRE F-A 100058 ISI-13142-17-A ISI F1.20b H- 3 Slide Pipe AUG 2 RHR Suction A OWN PRE . . . ... . . ... . . . F -A 10 00 5 9 IS I-13 14 2 A ISI . . . . . . . . . . . . . . F1.20b H-4 Strut / Clamp AUG 2 RHR Suction A OWN PRE F-A 100065 IS1-13142-17-A ISI F1.20b H- 8 Slide AUG 2 RHR Suction A OWN PRE Printed 6/5/1

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 117 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3

                                                                                                  -~ I C4 Category,                                                                     DwgIlSO No.                                          La U-C1   N         N    N          N Item No.,                                                                     Comp. Desc.          w    w         N    w U.

N Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 100066 IS1-13142-17-A ISI F1.20b H- 9 Slide AUG 2 RHR Suction A OWN PRE F-A 100129 ISI-13142-17-C ISI F1.20b H- 2 Slide AUG 2 RHR Suction B OWN PRE F-A 100130 ISI-13142-17-C ISI F1.20b H-3 Slide AUG 2 RHR Suction B OWN PRE F-A 100131 Ml_14-P2_RF23 /PSI /VT/ /PEI- ISI-13142-17-C ISI F1.20b H- 4 02.05.02 Slide AUG 2 RHR Suction B MI1_4-P2_RF24 I PSI I VT / I FP-PE- OWN NDE-530 PRE - - - - c B F-A 100132 ISI-13142-17-C ISI F1.20b H- 5 M1_I4-P2_RF23 / PSI/ VT// PEI- Strut / Clamp AUG 2 RHR Suction B 02.05.02 OWN PRE - - - - c F-A 100174 ISI-13142-18-A ISI . . . . .. . . . .. . . . . F1.20b H- 7 Strut / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100209 ISI-13142-18-B ISI F1.20b H- 3 Strut / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100246 ISI-13142-18-C ISI F1.20b H- 1 Strut / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100263 Ml_14-P2_RF23 / PSI / VT /PEI- ISI-13142-19-A ISI F1.20b H- 1 02.05.02 Snubber / Clamp AUG 2 HPCI Steam Disch Ml_14-P2_RF24 / PSI / VTI /FP-PE- OWN NDE-530 PRE c B Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 118 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M LO 1'0 0 0 0 0 0,0 00a N 1 C'41 CIS Category, Dwg/ISO No. 1 N4 U-A Item No., Comp. Desc. LL Cs(4 u-('4 jjL N' UL Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 100264 M1_I4-PlRF21 /ISI/VTrI PEI- ISI-13142-19-A ISI c r - - - F1.20b H- 2 02.05.02 Dbl Snubber / Clamp AUG 2 HPCI Steam Disch Ml_14-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 M1_14-P3_RF25 / PSI /VT/ / FP-PE-NDE-530 PRE- ----- c b - - - F-A 100267 ISI-13142-19-A iS[ F1.20b H- 4 Strut / Clamp AUG 2 HPCI Steam Disch OWN PRE F-A 100269 ISI-13142-19-A ISI F1.20b H- 6 Strut / Clamp AUG 2 HPCI Steam Disch OWN PRE F-A 100272 ISI-13142-19-A ISI F1.20b H- 8 Strut / Clamp AUG 2 HPCI Steam Disch OWN PRE F-A 100273 ISI-13142-19-A ISI F1.20b H- 9 Strut / Clamp AUG 2 HPCI Steam Disch OWN PRE F-A 100332 ISI-13142-20-A ISf- ---------------- s - - - F1.20b H-2 M1_I4-P3_RF2511SIIVTI /FP-PE- Slide AUG 2 Core Spray P NDE-530 OWN PRE F-A 100333 ISI-13142-20-A ISI F1.20b H- 3 Snubber AUG 2 Core Spray A OWN PRE F-A 100335 ISI-13142-20-A ISI F1.20b H- 5 Strut / Clamp AUG 2 Core Spray A OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 119 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o oaa a

                                                                                                    -~ 10C4 N   N Category,                                                                    Dwg/ISO No.             N   N N

Item No., Comp. Desc. w w N N U-w, Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 100362 IS1-13142-20-B ISI F1.20b H- 2 M1_I4-P1_RF22 / PSI /VT/ / PEI- Slide Hanger AUG 2 Core Spray B 02.05.02 OWN PRE - c F-A 100363 ISI-13142-20-B ISI F1.20b H- 3 Snubber/Clamp AUG 2 Core Spray B OWN PRE ISI-13142-31-B ISI F-A 100515 F1.20b H- 1 Valve Strap AUG 2 Core Spray A OWN PRE F-A 100784 ISI-13142-37-A ISI F1.20b H-3 M1_I4-Pl RF21 IPSIIVTI /PEI- Snubber AUG 2 RHR Discharge A 02.05.02 OWN PRE c F-A 100786 ISI-13142-37-A ISI F1.20b H-5 Dbl Strut / Clamp AUG 2 RHR Discharge A OWN PRE F-A 100787 ISI-13142-37-A ISI F1.20b H-6 Strut / Clamp AUG 2 RHR Discharge A OWN PRE F-A 100823 ISI-13142-37-B iSI F1.20b H-8 Strut / Clamp AUG 2 Containment Spral OWN PRE F-A 100825 ISI-13142-37-B ISI F1.20b H-10 Dbl Strut/Dbl Clamp AUG 2 Containment Spral OWN PRE F-A 100850 ISt-13142-37-b ISI F1.20b H- 4 Dbl Strut/ Snubber AUG 2 RHR Discharge A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 120 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012)

                                                                                                                        ,Period I                  Period 2                 Period 3 o       0                     o   0 a      o                     o N

0 4 N I " N Category, DwgIISO No.

                                                                                                                  .-      N                                              La Item No.,                                                                             Comp. Desc.                  N u.

N UL N U. N U. N UL Class Summary No.IComplD/System Scope I Method I Procedure Code Case F-A 100851 ISI-13142-37-C iSI . . . ... . . ... . . . F1.20b H- 5 Dbl Strut / Clamp AUG . . . ... . . ... . . . 2 RHR Discharge A OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . F-A 100878 ISI-13142-40-A ISI . . . ... . . ... . . . F1.20b H-4 Strut/ Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 100879 ISI-13142-40-A ISI . . . ... . . ... . . . F 1.2 0 b H- 5 M1 _14 - P2 _RF2 3 / P S I / VT / / P EI - S n ub b e r / C la m p A UG . . . . . . . . . . . . . . 2 HPCI Water Side Dsch 02.05.02 OWN . . . ... . . ... . . . PRE - t - - c F-A 100882 ISI-13142-40-A ISI . . . ... . . ... . . . F1.20b H- 8 Strut / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 100887 ISI-13142-40-A ISI F1.20b H-13 Strut / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . F -A 10 0 913 IS I-13 14 2 B ISI - - . . . . . . . F1.20b H-2 Strut / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 100919 Ml-14-PIRF21 / ISI/VT/ / PEI- ISI-13142-40-B ISI E F1.20b H- 7 02.05.02 Strut / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch Ml14-PIRF21 / PSI/T / PEI- OWN 02.05.02 PR E . . . . . . . . . . . . . . F -A 10 0 92 1 Ml_ 14 -P 1_ R F 2 1 /I S I / VT / / P E I- ISI-13 142 B ISI c . . . . . . . . . . . . . F1.20b H- 8 02.05.02 Strut / Clamp / Bo) AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch M1_14-P1_RF21 / PSI/VT / PEI- OW N . . . .. . . . .. . . . . 02.05.02 PRE F-A 100924 S0r-ut I -lap-D F1.20b H-11 Strut I Clamp AUG 2 HPCI Water Side Dsch OWN PRE 0 Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 121 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 e3 Uto r- a* 00 0 0 0 0 0 04 S-040C1 M1 C*1. Category, DwgIlSO No. - C1 M~ V NI Item No., Comp. Desc. 11 U. U. N LL N U. U. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 100925 ISI-13142-40-B IS1 F1.20b H-12 Strut / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch OWN PRE F-A 100980 ISI-13142-42-A ISI F1.20b H- 4 DbI Strut/Dbl Clamp AUG 2 HPCI Steam OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 100985 ISI-13142-42-A ISI F1.20b H- 9 Strut / Clamp AUG 2 HPCI Steam OWN PRE F-A 100987 ISrI-1314 A ISI F1.20b H-11 Snubber / Clamp AUG 2 HPCI Steam OWN Ml_14-PlRF21 / PSI / VT / / PEI-02.05.02 Ml-14-PlRF21 / PSI/ VT / / PEI-02.05.02 Ml-14-PlRF22 /P I VT / I PEI-02.05.02 Ml-14-PlRF22 / PSI IVT/ / PEI-02.05.02 Ml_14-P2_RF23 / PSI / VT / / PEI-02.05.02 Ml_14-P2_RF23 / PSI /VT/ / PEI-02.05.02 M1_14-P2_RF24 / PSI /VT/ / FP-PE-NDE-530 Ml_14-P2_RF24 / PSI /VT/ / FP-PE-NDE-530 Ml_14-P3_RF25 / PSI / VT / / FP-PE-NDE-530 M1_14-P3_RF25 / PSI / VT / VT-3 / FP PE-NDE-530 PRE IB B - -lB B - -lb - - - Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 122 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 0 0 W-0 C11 N N N1 Category, DwgIlSO No. NV I. cmNIJ.M Cw IJC4 ftem No., Comp. Desc. L NL NL N U. w 0: 1 ww N Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 101106 IS1-13142-49-A ISI----------------- s - - - F1.20b H-3 M1_I4-P3_RF251iSIIVTI /FP-PE- DblStrut/DbIClamp AUG 2 RHR Suction A NDE-530 OWN PRE F-A 101107 ISI-13142-49-A ISI F1.20b H- 4 Snubber / Clamp AUG 2 RHR Suction A OWN PRE F-A 101110 ISI-13142-49-A ISI c r F1.20b H- 7 M1_14-PlRF21 / ISF/ VT / / PEI- Dbi Strut/Dbi Strut AUG 2 RHR Suction B 02.05.02 OWN PRE F-A 101113 M1 _ 4-PlRF21 / ISI / VT // PEI- ISI-13142-49-A ISI c r F1.20b H-10 02.05.02 Snubber / Clamp AUG 2 RHR Suction B M1 4-PlRF21 / PSI/ VT/ / PEI- OWN 02.05.02 PRE F-A 101114 ISI-13142-49-A ISI Fl.20b H-11 M1 4-P1_RF22/PSI/VT/ /PEI- Slide AUG 2 RHR Suction B 02.05.02 OWN PRE - c F-A 101139 M1_I4-P2_RF24 / PSI / VT / / FP-PE. ISI-13142-51-A Isi F1.20b H- 1 NDE-530 Snubber/ Dbi Strut AUG 2 RHR A M1_I4-P3_RF25 /PSI/ VT/ / FP-PE- OWN -- NDE-530 PRE---------- c b - - - F-A 101165 ISI-13142-51-B ISI . . . .. . . . .. . . . . F1.20b H- I Strut / Clamp AUG . . . ... . . ... . . . 2 Containment Spral OWN PRE F-A 101169 ISI-13142-51-B ISI F1.20b H- 4 Strut / Clamp AUG 2 Containment Spral OWN PRE F-A 101170 ISI-13142-51-B ISI F1.20b H-5 Dbi Strut/Dbl Clamp AUG 2 Containment Spra, OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 123 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0 0 0 01 0 N N N Category, DwgIlSO No. -I I4 I " I C-4 C1 cli CI In Item No., Comp. Desc. L. U. LU. N LL n, w* w* w-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 101173 ISI-13142-51-B ISI F1.20b H- 8 Strut / Clamp AUG 2 Containment Spral OWN PRE F-A 101709 ISI-93268-1A ISI F1.20b H- 8 Strut / Clamp AUG 2 CRD Scram Header A OWN PRE F-A 101710 ISI-93268-1A ISI - - - - c F1.20b H-9 M1_I4-P2_RF23/ ISI/v- / PEI- Strut / Clamp AUG 2 CRD Scram Header A 02.05.02 OWN PRE F-A 101713 MI 14-P2_RF23 / PSI /VT/ I PEI- ISI-93268-1A SI F1.20b H-12 02.05.02 Snubber/Clamp AUG 2 CRD Scram Header A Ml_14-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 PRE -t--c B ISI-93268-1A ISI F-A 101714 F1.20b H-13 Strut / Clamp AUG 2 CRD Scram Header A OWN PRE F-A 101715 Ml_14-P2_RF23 / PSI /VT/ / PEI- tSI-93268-1A ISI F1.20b H-14 02.05.02 Snubber/Clamp AUG 2 CRD Scram Header A M1_I4-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 PRE - t c- -B F-A 106041 ISI-94966-A ISI F1.20b H- 1 Double Rigid Strul AUG 2 Vacuum Relief& CGCS Outlet Div.1 OWN PRE F-A 106042 ISI-94966-A ISI F1.20b H- 2 Ml_14-P2_RF23 /PSI /VT/ /PEI- Snubber / Clamp AUG 2 Vacuum Relief & CGCS Outlet Div.1 02.05.02 OWN PRE - - - - c F-A 106043 ISI-94966-A ISI S F1.20b H- 3 M1_14-P3_RF25 / ISI/ VT / / FP-PE- Dbi Spring / U-boll AUG 2 Vacuum Relief & CGCS Outlet Div.1 NDE-530 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 124 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 400C 0 0 4 " N 1 Category, DwgIlSO No. Item No., Comp. Desc. LL LL LL LL N L-Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 106044 ISI-94966-A ISI F1.20b H-4 Ml_14-P2 RF23 /PSI//T/ / PEI- Snubber/ Clamp AUG .. . . ... . . ... . . . 2 Vacuum Relief &CGCS Outlet Div.1 02.05.02 OWN .. . . ... . . ... . . . PRE - - - - c F-A 106046 ISI-94966-A IS1 F1.20b H- 6 Dbl Strut / U-boll AUG 2 Vacuum Relief &CGCS Outlet Div.1 OWN PRE F-A 106052 ISI-94966-B IS - F1.20b H- 1 Dbl Strut / U-boll AUG 2 Containment Air Purge OWN PRE F-A 106053 ISI-94966-B ISI FI.20b H-2 Strut / Clamp AUG 2 Containment Air Purge OW N .. . . .. . . . .. . . . . PR E - - . . . . . . . . . . . F-A 106065 ISI-94699-A ISI F1.20b H-1/H-2 Double Snubbei AUG 2 CGCS OUTLET DIV 2 OWN PRE F-A 106077 ISI-105531-A ISI F1.20b H- 1 Base Plate AUG 2 Standby Gas Trtmnt & Rx Plenun OWN PRE F-A 106097 ISI-158074-A ISI F1.20b H- 1 Hanger AUG 2 Torus HPV OWN PRE F-A 106810 IS1-13142-19-B ISI F1.20b H- 1 M1_14-P2_RF23 / PSI / VT / I PEI- Double Snubbei AUG 2 RCIC Steam Discharge 02.05.02 OWN Ml_14-P2_RF23 / PSI /VT/ / PEI-02.05.02 M1_I4-P2_RF24 / PSI /VT/ / FP-PE-NDE-530 Ml_14-P2_RF24 / PSI /VT I/FP-PE-NDE-530 PRE I- t - c B - - I - - 0 Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 125 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o o 0 w w N N N Category, DwgIlSO No. LL IL Item No., Comp. Desc. N: N N U. w* Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 106820 IS1-13142-26-B ISI F1.20b H- 1 M1_14-P2_RF23 / PSI / VT/ PEI- Snubber AUG 2 CORE SPRAY B 02.05.02 OWN PRE - I - c F-A 106950 ISI-13142-51-C ISI FM.20b H- M14-PlRF21 1 / PSI/ VT / PEI- Seismic Snubber AUG 2 RHR B 02.05.02 OWN .............. PRE c F-A 107030 M1_4-P1_RF22 / PSI / VT / I PEI- IS1-13142-37-D ISI F1.20b H- 3 02.05.02 Double Snubbei AUG 2 RHR A M1-14-P _RF22 / PSI / VT / / PEI- OWN 02.05.02 PRE - c F-A 107049 M1_I4-P2_RF23 / PSI / VI / / PEI- ISI-13142-37-E ISI F1.20b H-2 02.05.02 Snubber AUG 2 RHR A M1_4-P2_RF24 / PSI I VT / / FP-PE- OWN NDE-530 M1_I4-P2_RF24 / PSI IVTI / FP-PE-NDE-530 PRE - - - - c B F-A 107568 ISI-13142-17-C ISI Fl.20b H-9 Ml_14-Pl_RF221PSII/VTI/ PEI- DbI Strut / 8 Lugs AUG 2 RHR Suction B 02.05.02 OWN PRE - c F-A 107573 ISI-13142-37-B ISI - - - - c F1.20b H-2 M1 14-P2_RF23/ISI/VTI IPEI- Strut / 8 Lugs AUG 2 Containment Spra* 02.05.02 OWN PRE F-A 107574 ISI-13142-37-B ISI - - - - c Fl.2Db H-3 M1_4-P2_RF23 / S1i VT I PEI- Strut/8 Lugs AUG 2 Containment Spral 02.05.02 OWN PRE F-A 107575 ISI-1314/-J f-Is IjI Fl.20b H-5 Ml_14-PlRF21 / PSI/VT/ /PEI- Snubber/Lugs AUG 2 Containment Spra) 02.05.02 OWN PRE c Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 126 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M. La oý 0 Ir. o o* aV o 0 0 N C4 N N Category, Dwg/ISO No. 4 N

                                                                                                       -    1                             Ln Item No.,                                                                    Comp. Desc.                N       N LL -          w      N        U-Class      Summary No.IComplDISystem      Scope I Method I Procedure         Code Case F-A        107578                         Ml_14-PIRF21 / ISI/VT/ /PEI-       ISI-13142-40-B       ISI     c      r F1.20b     H- 6                           02.05.02                           Dbl Strut / 4 Lugs   AUG      .      .  .  ..  .      .  . ..   .    .   . .

2 HPCI Water Side Dsch M1_I4-P1_RF21 / PSI /VTI /PEI- OWN 02.05.02 PRE F-A 100064 IS1-13142-17-A ISI ---------------- s - - F1.20c H-7 M1_14-P3_RF251/SIIVTI IFP-PE- Spring/Clamp AUG 2 RHR Suction A NDE-530 OWN PRE F-A 100099 ISI-13142-17-B ISI F1.20c H-2 Dbi Spring / Clamp AUG 2 HPCI Water Side Sctn OWN PRE F-A 100128 ISI-13142-17-C ISI F1.20c H- 1 Spring / Clamp AUG 2 RHR Suction B OWN PRE F-A 100134 ISI-13142-17-C ISI F1.20c H-7 Dbl Spring / Clamp AUG 2 RHR Suction B OWN PRE F-A 100168 ISI-13142-18-A ISI - - - c F1.20c H- 1 M1 14-P2 RF23 ISI VT/I PEI- Spring / Clamp AUG 2 RHR Discharge B 02.05.02 OWN PRE F-A 100172 ISI-13142-18-A ISI F1.20c H- 5 DbI Spr/U-BIt/Saddle AUG 2 RHR Discharge B OWN PRE F-A 100173 ISI-13142-18-A ISI F1.20c H-6 DbI Spr/U-BIt/Saddle AUG 2 RHR Discharge B OWN PRE F-A 100175 ISI-13142-18-A ISI F1.20c H-8 Dbl Spr/U-BIt/Saddle AUG 2 RHR Discharge B OWN PRE Printed 6/5/*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 127 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 N N N 4N N Category, DwgIlSO No. - N- () -e W'a Item No., Comp. Desc. U. L. LL I. N1 L-w, w, Q: a: Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 100178 ISI-13142-18-A ISI F1.20c H-10 Spring/Clamp/SlidE AUG 2 RHR Discharge B OWN PRE F-A 100207 IS1-13142-18-B ISI F1.20c H- 1 Variable Slide AUG 2 RHR Discharge B OWN PRE F-A 100208 ISI-13142-18-B ISI F1.20c H- 2 Variable Claml AUG 2 RHR Discharge B OWN PRE F-A 100210 ISI-13142-18-B ISI F1.20c H- 4 Spring / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100247 ISI-13142-18-C ISI F1.20c H- 2 Spring / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100248 ISI-13142-18-C ISI F1.20c H- 3 Spring / Clamp AUG 2 RHR Discharge B OWN PRE F-A 100268 ISI-13142-19-A ISI F1.20c H- 5 Spring / Clamp AUG 2 HPCI Steam Disch OWN PRE F-A 100331 Ml-14-PlRF21 / ISI IVTI / PEI- ISI-13142-20-A ISI c F1.20c H- 1 02.05.02 Variable Spr / SlidE AUG 2 Core Spray A Ml-14-P1_RF21 / PSI/ VT / PEI- OWN 02.05.02 PRE F-A 100334 IS I-13142-20-A ISI F1.20c H- 4 Variable Spr/SlidE AUG 2 Core Spray A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 128 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M W) a N cm N Category, DwgIlSO No. NN

                                                                                                        -N                 V~)I Item No.,                                                                     Comp. Desc.            U-     U.              N     w              1L.

Class Summary No.ICompiD/System Scope I Method I Procedure Code Case F-A 100361 ISI-13142-20-B ISI F1.20c H- 1 Variable AUG 2 Core Spray E OWN . . . ... . . ... . . . PRE F-A 100429 IS1-13142-26-A ISI F1.20c H- 3 Spring / Clamp AUG 2 Core Spray E OWN PRE F-A 100430 IS1-13142-26-A ISI F1.20c H- 4 Spring / Clamp AUG 2 Core Spray 1 OWN PRE F-A 100516 ISI-13142-31-B ISI F1.20c H- 2 Dbl Spring / Clamp AUG 2 Core Spray A OWN PRE F-A 100782 ISI-13142-37-A ISI F1.20c H- 1 Spring / Clamp AUG 2 RHR Discharge A OWN PRE . . . ... . . ... . . . F-A 100783 ISI-13142-37-A ISI F1.20c H- 2 Spring / Clamp AUG . . . .. . . . .. . . . . 2 RHR Discharge A OWN . . . ... . . ... . . . PRE F-A 100785 ISI-13142-37-A ISI . . . ... . . ... . . . F1.20c H- 4 Dbl Spring / Clamp AUG 2 RHR Discharge A OWN . . . .. . . . ... . . . PRE F-A 100790 ISI-13142-37-A ISI F1.20c H- 8 Variable / Clamr AUG 2 RHR Discharge A OWN PRE F-A 100813 IS1-13142-37-B IS5 S F1.20c H- 1 M1_14-P3_RF25 / ISi/ / FP-PE- Variable Spring I" AUG 2 Containment Spral NDE-530 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 129 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M* Lo o e o o o e o o N N N 1N N U,I Category, DwgIISO No. Item No., Comp. Desc. N N4 ,'a LIL U. L. LL LL Class Summary No.IComplDISystem Scope I Method I Procedure Code Case - F-A 100817 IS1-13142-37-B ISI F1.20c H- 4 Variable Spring AUG 2 Containment Spral OWN PRE F-A 100824 IS1-13142-37-B ISI - - - - c F1.20c H- 9 M1 14-P2_RF23 /ISI /VT /PEI- Spring / Clamp AUG 2 Containment Spral 02.05.02 OWN PRE F-A 100826 ISI-13142-37-B ISI F1.20c H-11 Spring / Clamp AUG 2 Containment Spral OWN PRE F-A 100827 ISI-13142-37-B ISI F1.20c H-12 Spring / Clamp AUG 2 Containment Spral OWN PRE F-A 100847 ISI-13142-37-C ISI F1.20c H- 1 Spring / Clamp AUG 2 RHR Discharge A OWN PRE F-A 100848 ISI-13142-37-C ISI F1.20c H- 2 Spring / Clamp AUG 2 RHR Discharge A OWN PRE F-A 100849 ISI-13142-37-C ISI F1.20c H-3 Variable Spring AUG 2 RHR Discharge A OWN PRE F-A 100875 ISI-13142-40-A ISI F1.20c H- 1 Spring / Clamp AUG 2 HPCI Water Side Dsch OWN PRE F-A 100915 1*1-1314Z-4U-B ISI F1.20c H-4 Spring / Clamp AUG 2 HPCI Water Side Dsch OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 130 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 I.- o* o M aý La o3 0 0 0 NN N Category, DwgIISO No.

                                                                                                     - 1N4 Item No.,                                                                     Comp. Desc.             LL   IL N

ILL NLL. U. Class wt w w, Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 100922 ISI-13142-40-B ISI c F1.20c H- 9 M1_I4-PlRF21 ISI /VT/ IPEI- Dbl Spring / Clamp AUG . . . ... . . ... . . . 2 HPCI Water Side Dsch 02.05.02 OWN PRE F-A 100923 ISI-13142-40-B ISI F1.20c H-10 Variable Spring AUG 2 HPCI Water Side Dsch OWN PRE F-A 100977 ISI-13142-42-A ISI F1.20c H- 2 Spring / Clamp AUG 2 HPCI Steam OWN PRE F-A 100981 ISI-13142-42-A ISI F1.20c H- 5 Spring / Clamp AUG 2 HPCI Steam OWN PRE F-A 100982 ISI-13142-42-A iSI F1.20c H- 6 M1_I4-P3_RF25/ PSI /VT /VT-3 /FP Spring / Clamp AUG 2 HPCI Steam PE-NDE-530 OWN PRE---------------- s - - - F-A 100983 ISI-13142-42-A ISI F1.20c H- 7 Spring / Clamp AUG 2 HPCI Steam OWN PRE F-A 100984 ISI-13142-42-A ISI F1.20c H- 8 Spring / Clamp AUG 2 HPCI Steam OWN PRE F-A 100989 ISI-13142-42-A IS I F1.20c H-13 Spring / Clamp AUG 2 HPCI Steam OWN PRE F-A 100990 ISI-13142-42-A ISI F1.20c H-14 Spring / Clamp AUG 2 HPCI Steam OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 131 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 oco 0 NI 0 N1 C11 04 Category, DwgIlSO No. 4

                                                                                                    - I                               w     N Comp. Desc.             N         N Item No.,                                                                                            u.        u.                                                  L.

Class Summary No.IComplD/System Scope I Method I Procedure Code Case F-A 101104 ISI-13142-49-A ISI . . . . . . . . . . . . . . F1.20c H- I Spring / Clamp AUG . . . ... . . ... . . . 2 RHR Suction A OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . F-A 101105 ISI-13142-49-A SI . . . ... . . ... . . . F1.20c H- 2 DbI Spring / Tee AUG . . . ... . . ... . . . 2 RHR Suction A OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 101108 ISI-13142-49-A ISI c . . ... . . ... . . . F1.20c H-5 Ml_14-P1_RF21 /ISII/VT /PEI- Spring / Clamp AUG . . . .. . . . .. . . . . 2 RHR Suction A 02.05.02 OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 101109 ISI-13142-49-A ISI . . . ... . . ... . . . F1.20c H- 6 Spring / Clamp AUG . . . ... . . ... . . . 2 RHR Suction B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 101111 ISI-13142-49-A SI ... . ... . . ... . . . F1.20c H- 8 Spring / C lam p A UG . . . . . . . . . . . . . . 2 RHR Suction B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 101112 ISI-13142-49-A ISI . . . ... . . ... . . . F1.20c H- 9 Clamp AUG 2 RHR Suction B OWN PRE F-A 101168 ISI-13142-51-B ISI F1.20c H- 3 Dbi Spring / Clamp AUG 2 Containment Spral OWN PRE F-A 101171 ISI-13142-51-B ISI F 1.20c H- 6 S pring / C lam p A UG . . . . . . . . . . . . .- 2 Containment Spral OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . .

                                                                                               .II                                 r F-A        101172                                                            Sr-in 1/   Il  -am Spring / Clamp     AUG F1.20c     H- 7 2          Containment Spraý                                                                    OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 132 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 00 0 0 0C14 0IN 0C-4 0 C4 N Category, DwgIlSO No. - N C14 M .1 1o Item No., Comp. Desc. N14 N LI. U-N LL N LL N L-Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 106020 IS1-13142-29-A ISI F1.20c H- 1 Dbi Strut / U-Boll AUG 2 RBCCW OWN PRE F-A 107564 ISI-13142-17-A ISI F1.20c H- 5 Dbl Spr / Half Clamp AUG 2 RHR Suction A OWN PRE F-A 107565 ISI-13142-17-A ISI F1.20c H- 6 Dbl Spr / Half Clamp AUG 2 RHR Suction A OWN PRE F-A 107566 ISI-13142-17-C SI------------------ s - - - F1.20c H-6 Ml_14-P3_RF2511SIIVTI IFP-PE- DblSpr/HalfClamp AUG 2 RHR Suction B NDE-530 OWN PRE F-A 107567 ISI-13142-17-C ISI F1.20c H- 8 DbI Spr / Half Clamp AUG 2 RHR Suction B OWN PRE F-A 107569 ISI-13142-18-A ISI c F1.20c H- 9 M1_14-PlRF21 ISI VTI /PEI- DbI Spring / 4 Lugs AUG 2 RHR Discharge B 02.05.02 OWN PRE F-A 107570 ISI-13142-19-A ISI F1.20c H- 3 Dbl Spring / 4 Lugs AUG 2 HPCI Steam Disch OWN PRE F-A 107571 ISI-13142-19-A SI------------------ s F1.20c H- 7 Ml_14-P3_RF25 / ISI / VT FP-PE- Dbl Spring / 4 Lugs AUG 2 HPCI Steam Disch NDE-530 OWN PRE F-A 107572 IS1-13142-37-A ISI F1.20c H- 7 Dbl Spring / 4 Lugs AUG 2 RHR Discharge A OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 133 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 U,} rOP- Om 0 0 0 0 "14 Category, DwglISO No. -I oI N I I Item No., Comp. Desc. N uL N uL N uL C14 LL a-Class Summary No.IComplDISystem Scope I Method ) Procedure Code Case F-A 107576 ISI-13142-37-B ISI F1.20c H- 6 Dbi SprlClamp&SaddlE AUG 2 Containment Spral OWN PRE F-A 107577 ISI-13142-37-B ISI F1.20c H- 7 Dbl Spring / 4 Lugs AUG 2 Containment Spral OWN PRE F-A 107579 ISI-13142-42-A ISI - - - - c F1.20c H- 3 M1_14-P2_RF23 i tSI / VT// PEI- Dbl Spring / 4 Lugs AUG 2 HPCI Steam 02.05.02 OWN PRE F-A 107580 ISI-13142-51-B ISI F1.20c H- 2 Dbl Spring / 4 Lugs AUG 2 Containment Spral OWN PRE F-A 107601 ISI-13142-18-B ISI F1.20c H-2B Variable Claml AUG 2 RHR Discharge B OWN PRE F-A 102229 nd-isi-101 ISI F1.30a SR-563 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102230 nd-isi-101 ISi F1.30a SR-81 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102232 nd-isi-101 ISI F1.30a SWH-308 Base Plate AUG 3 RHR Service Water OWN PRE F-A 102233 nd-lsi-lul ISI F1.30a SWH-309 Base Plate AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 134 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 0~ 6D 1` 0) 00 0 0 010 00 04 Category, DwgIlSO No. C- N CA N1 Lo

                                                                                                         *-      C4                   M~ V Item No.,                                                                      Comp. Desc.                N       N                   N     N                   N.

LL LL uL IL U) N Class Summary No.IComplDISystem Scope I Method I Procedure Code Case I-F-A 102234 nd-isi-101 IS[ F1.30a SWH-310 Base Plate AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102235 nd-isi-101 ISI . . . .. . . . .. . . . . F1.30a SWH-311 Base Plate AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102236 nd-isi-101 ISi . . . .. . . . .. . . . . F1.30a SWH-312 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 102237 nd-isi-101 IS] . . . .. . . . .. . . . . F1.30a SWH-313 Seismic Restraint AUG 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102240 nd-isi-102 ISI------------------- c F1.30a SR-213 M1_I4-P2_RF24 / ISi / VT / / FP-PE- Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water NDE-530 OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 102241 nd-isi-102 ISI- ---------------- s - - - Fl.30a SR-79 Ml 14-P3_RF25 / iSi /VT / / FP-PE- Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water NDE-530 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102242 nd-isi-102 ISI . . . .. . . . .. . . . . F1.30a SR-80 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102244 nd-isi-102 ISI - - - - c F1.30a SWH-305 M1 14-P2_RF23 /ISI/VT I / PEI- Base Plate AUG . . . ... . . ... . . . 3 RHR Service Water 02.05.02 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102245 nd-isi-1 02 ISI F1.30a SWH-306 Base Plate AUG 3 RHR Service Water OWN PRE Printed 6/510

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 135 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 0040 0 00 00 0 I*l *I *1 cI N Category, DwgIlSO No. - N M -e in Item No., Comp. Desc. U.

                                                                                                     -N   04 U.

Of LL LL

                                                                                                                             .      N4 U-Class       Summary No.IComplD/System     Scope I Method I Procedure         Code Case F-A         102246                                                           nd-isi-102        ISI F1.30a      SWH-307                                                          Base Plate        AUG 3           RHR Service Water                                                                  OWN PRE F-A         102247                                                           nd-isi-102        ISI F1.30a     SWH-31                                                            Seismic Restrainl AUG 3           RHR Service Water                                                                  OWN PRE F-A         102248                                                           nd-isi-102        ISI F1.30a     SWH-32                                                            Restraint         AUG 3           RHR Service Water                                                                  OWN PRE F-A         102249                                                           nd-isi-103        ISI  -       - -  -    r      -  - -  s    -   -  -

F1.30a SR-394 M1_14-P3_RF25 /ISI/ VT / / FP-PE- Seismic Restraint AUG 3 RHR Service Water NDE-530 OWN PRE F-A 102250 nd-isi-103 ISI F1.30a SR-395 Seismic Restrainl AUG 3 RHR Service Water OWN PRE F-A 102251 nd-isi-103 ISI F1.30a SR-396 Ml_14-P2_RF23 / PSI/ V/ VT-3 Seismic Restraint AUG 3 RHR Service Water PEI-02.05.02 OWN PRE - - - - c F-A 102252 nd-isi-103 ISI F1.30a SR-459 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102254 nd-isi-103 ISI F1.30a SR-84 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102255 nd-isi-103 ISI F1.30a SWH-161 Rod & Strut AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 136 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o o o o Category, N N IN N Dwg/ISO No. MI V1 0n N Item No., Comp. Desc. of w N N LL

w. LL a: U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102256 nd-isi-104 ISI - -

F1.30a SR-460 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102257 nd-isi-104 SI . . . ... . . ... . . . F1.30a SR-461 Seismic Restrainl AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 102258 nd-isi-104 ISI . . . ... . . ... . . . F1.30a SR-462 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102259 nd-isi-104 SI . . . ... . . ... . . . F1.30a SR-463 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102260 nd-isi-104 ISI . . . ... . . ... . . . F1.30a SR-473 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102261 nd-isi-104 ISI . . . .. . . . ... . . . F1.30a SWH-45 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102262 nd-isi-105 ISI . . . ... . . ... . . . F1.30a SR-481 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102263 nd-isi-105 IS[ . . . ... . . ... . . . F1.30a SR-482 Seismic Restraint AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . F-A 102264 nd-isi-105 IsI F1.30a SR-90A Rod AUG 3 RHR Service Water OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 137 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o e o N N N N N Category, DwgIISO No. La N_ Item No., Comp. Desc. N N N N Lz. UL L-Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 102265 nd-isi-105 ISI - - - ... . . ... . . . F1.30a SW-21 Rod AUG 3 RHR Service Water OWN PR E . . . . . . . . . . . . . . F-A 102266 nd-isi-105 ISI F1.30a SW-22 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN PR E . . . . . . . . . . . . . . F-A 102267 nd-isi-105 ISI . . . ... . . ... . . . F1.30a SWH-483 Rod AUG 3 RHR Service Water OWN PRE . . . ... . . ... . . . F-A 102268 nd-isi-106 ISI . . . ... . . ... . . . F1.30a SR-18 Rod AUG 3 RHR Service Water OWN PRE F-A 102269 nd-isi-106 IS I F1.30a SR-398 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102272 nd-isi-106 ISI F1.30a SR-95 Seismic Restraint AUG . . . .. . . . .. . . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE F-A 102273 nd-isi-106 ISI . . . ... . . ... . . . F1.30a SR-97 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102274 nd-isi-106 ISI F1.30a SW-15 Rod AUG 3 RHR Service Water OWN PRE F-A 102275 na-IsI-1 U0 ijI Fl.30a SW-16 Rod AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 138 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 c'* "1C I*. O* 0 0 00O 0 0 00O Category, C1 C- 04 N N DwgIlSO No. I .1c I V 04 Item No., Comp. Desc. C1 LL. N UL N LL N N LL Class w* Summary No.lCompID/System Scope I Method I Procedure Code Case F-A 102276 nd-isi-106 ISI F1.30a SW-4A Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102277 Ml_14-P2_RF24 /PSI /VTT/VT-3 / FP nd-isi-107 ISI----------------- s - - - F1.30a IS-SWH-75 PE-NDE-530 Rod AUG 3 RHR Service Water Ml_14-P3_RF25 / ISI / vr / VT-3 / FP- OWN PE-NDE-530 PRE- ------ c F-A 102278 nd-isi-107 ISI F1.30a IS-SWH-76 M1_14-P2_RF24 / PSI / VT / VT-3 / FP Rod AUG 3 RHR Service Water PE-NDE-530 OWN PRE---------- c F-A 102281 nd-isi-107 ISI F1.30a SR-399 M1_I4-P2_RF24 / PSI / VT / VT-3 / FP Seismic Restraint AUG 3 RHR Service Water PE-NDE-530 OWN PRE---------- c F-A 102282 nd-isi-107 ISI F1.30a SR-451 M1 14-P2_RF24/ PSI /VT/VT-3 / FP Seismic Restrain AUG 3 RHR Service Water PE-NDE-530 OWN PRE---------c nd-isi-107 PSI F-A 102283 FM.30a SR-452 Ml_14-P2_RF24 / PSI / VT / VT-3 / FP Seismic Restraint AUG 3 RHR Service Water PE-NDE-530 OWN PRE---------- c F-A 102284 nd-isi-107 ISI F1.30a SR-453 M14-P2_RF24 /PSI /VTVT-3 FP Seismic Restraint AUG 3 RHR Service Water PE-NDE-530 OWN PRE---------c nd-isi-107 PSI F-A 102285 Fl.30a SR-454 M1_I4-P2_RF24 / PSI /VT/ V--3 / FP Seismic Restraint AUG 3 RHR Service Water PE-NDE-530 OWN PRE-------------- F-A 102286 Ml_14-P2_RF24 / PSI / VI / VT-3 / FP nd-isi-107 ISI F1.30a SR-455 PE-NDE-530 Seismic Restraint AUG 3 RHR Service Water Ml_14-P2_RF24 / PSI / VT / VT-3 / FP OWN PE-NDE-530 PRE - c Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 139 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 oD 0 0 Category, DwgIlSO No. w N n, Item No., Comp. Desc. N 0: N U-Class Summary No.IComplD/System Scope I Method I Procedure Code Case F-A 102288 nd-isi-107 iSI F1.30a SR-469 M1_14-P2_RF24 / PSI /VT / VT-3 / FP Seismic Restrainl AUG 3 RHR Service Water PE-NDE-530 OWN PRE- ------ c F-A 102291 Ml1_4-PlRF21 / ISI /VT/ / PEI- nd-isi-107 ISI c r F1.30a SWH-8 02.05.02 Dead Weight Suppori AUG 3 RHR Service Water M1 _I4-P2_RF24 / PSI VT/VT-3 FP OWN PE-NDE-530 PRE- ------ c F-A 102294 nd-isi-108 ISI F1.30a ISI-SWH-69 Rod AUG 3 RHR Service Water OWN PRE F-A 102295 nd-isi-108 ISI F1.30a ISI-SWH-70 Dead Weight Support AUG 3 RHR Service Water OWN PRE F-A 102296 nd-isi-108 ISI c r F1.30a SR-100 M1_14-P1_RF21 /ISI VT/ /PEI- Double Rod AUG 3 RHR Service Water 02.05.02 OWN PRE F-A 102297 nd-isi-108 ISI F1.30a SR-103 Double Rod AUG 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . nd-isi-108 ISI . . . ... . . ... . F-A 102298 F1.30a SR-1 04 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102307 nd-isi-108 IS - F1.30a SWH-11 Rod AUG 3 RHR Service Water OWN PRE F-A 102308 no-isi-0uo II1 F1.30a SWH-9 Rod AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 140 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o- o~ o 0 o o N N N N Category, DwgIISO No. N*

                                                                                                  -     N41         MI -*             Ia Item No.,                                                                    Comp. Desc.           N     N          N ui   N

u. N LI.

w* w LL U-Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 102309 nd-isi-108 ISi Fl.30a SWH-9A Rod AUG 3 RHR Service Water OWN PRE F-A 102315 nd-isi-109 ISI F1.30a SR-94 Seismic Restrainl AUG 3 RHR Service Water OWN PRE F-A 102316 nd-isi-109 ISI F1.30a SR-96 Seismic Restraint AUG 3 RHR Service Water OWN PRE F-A 102317 nd-isi-109 ISI F1.30a SW-24 Rod AUG 3 RHR Service Water OWN PRE F-A 102318 nd-isi-109 ISi F1.30a SW-26 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE F-A 102319 nd-isi-109 ISI F1.30a SW-4B Seismic Restraini AUG 3 RHR Service Water OWN PRE F-A 102320 nd-isi-109 ISI . . . .. . . . .. . . . . F1.30a SWH-490 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE F-A 102321 nd-isi-109 ISI . . . .. . . . .. . . . . F1.30a SWH-498 Rod AUG . . . ... . . ... . . . 3 RHR Service Water OWN PRE F-A 102323 nd-isi-110 IS' F1.30a SR-91 Seismic Restraint AUG 3 RHR Service Water OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 141 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M 0La C o 0 C 0 0 0 a N 1.N Category, DwgllSO No. C-4 CM

                                                                                                                  '4
                                                                                                           - 1          V~) It        N Item No.,                                                                    Comp. Desc.                    u=

w u= It Nq uL uL N I. Nf Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102324 nd-isi-1 10 ISI F1.30a SW-29 Rod AUG 3 RHR Service Water OWN PRE F-A 102325 nd-isi-110 IS1 F1.30a SWH-493 Rod AUG 3 RHR Service Water OWN PRE F-A 102327 nd-isi-123 ISI F1.30a H-201 Rod AUG 3 RHR Service Water OWN PRE F-A 102328 nd-isi-123 ISI----------- c F1.30a H-202 M1 14-P2_RF24 /ISI /VT/ /FP-PE- Rod AUG 3 RHR Service Water NDE-530 OWN PRE F-A 102329 nd-isi-123 iSI F1.30a H-203 Rod AUG 3 RHR Service Water OWN PRE F-A 102330 nd-isi-123 IS I F1.30a H-204 Rod AUG 3 RHR Service Water OWN PRE F-A 102331 nd-isi-123 ISI F1.30a H-205 Rod AUG 3 RHR Service Water OWN PRE F-A 102228 nd-isi-100 ISI F1.30b SWH-180 Clevis AUG 3 RHR Service Water OWN PRE F-A 102231 nd-isi-101 ISI F1.30b SS-562 Dbl Strut / Snubber AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th interval ISI Plan (Rev. 4) Page 142 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 c¢) 10 o 04 0 C4 M le 0 Category, DwgIlSO No. CM C4

                                                                                                    -1      C41 Item No.,                                                                      Comp. Desc.            Of     C4                                       In Cl UL      L.                U. U.

Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 102238 nd-isi-101 ISI F1.30b SWH-375 Strut AUG 3 RHR Service Water OWN PRE . . . ... . . ... . . . F-A 102239 nd-isi-101 ISI . . . ... . . ... . . . F1.30b SWH-43 Stanchion AUG . . . ... . . ... . . . 3 RHR Service Water OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102253 nd-isi-103 ISI ---------------- s - - - F1.30b SR-464 M1_I4-P3_RF251/SIIVTI /FP-PE- Strut AUG . . . ... . . ... . . . 3 RHR Service Water NDE-530 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102270 nd-isi-106 SI . . . ... . . ... . . . F1.30b SR-450 M1 14-P2_RF24 / PSi / VT / VT-3 / FP Strut AUG . . . ... . . ... . . . 3 RHR Service Water PE-NDE-530 OWN . . . ... . . ... . . . PRE-------- c--------c F-A 102271 nd-isi-106 IS] . . . ... . . ... . . . F1.30b SR-457 Ml_14-P2_RF24 / PSI / VT / VT-3 / FP Anchor AUG . . . ... . . ... . . . 3 RHR Service Water PE-NDE-530 OWN PRE-------- c--------c F-A 102279 nd-isi-107 ISI F1.30b SR-105 M1_14-P2_RF24 /PSI /VT/V/-3 / FP Stanchion AUG 3 RHR Service Water PE-NDE-530 OWN . . . ... . . ... . . . PRE- ----- c F-A 102280 nd-isi-107 ISI . . . ... . . ... . . . F1.30b SR-397 M1_I4-P2_RF24 / PSI / VT / VT-3 / FP Double Strut AUG . . . ... . . ... . . . 3 RHR Service Water PE-NDE-530 OWN . . . ... . . ... . . . PRE-------- c-------c F-A 102287 nd-isi-107 ISI . . . ... . . ... . . . F1.30b SR-456 Ml_14-P2_RF24 / PSI / VT / VT-3 / FP Strut AUG 3 RHR Service Water PE-NDE-530 OWN PRE c------------ F-A 102289 nd-isi-1 07 ISI F1.30b SWH-72A Stanchion AUG 3 RHR Service Water OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 143 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M3 Ln o o o o 04 Lo Category, Dwg/ISO No. U-w LL w

                                                                                                -      N Item No.,                                                                    Comp. Desc.          N4 LL N

U u1 LL w Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 102290 nd-isi-107 SI F1.30b SWH-72B Stanchion AUG 3 RHR Service Water OWN PRE F-A 102292 nd-isi-108 ISI F1.30b IS-SWH-65 Stanchion AUG 3 RHR Service Water OWN . . . ... . . ... . . . PRE F-A 102293 nd-isi-108 ISI F1.30b IS-SWVH-66 Stanchion AUG 3 RHR Service Water OWN PRE F-A 102299 nd-isi-108 ISI c r F1.30b SR-106 Ml_14-P1 RF21 /ISl/V-/ /PEI- Stanchion AUG 3 RHR Service Water 02.05.02 OW N . . . .. . . . .. . . . . PRE F-A 102300 nd-isi-108 ISI F1.30b SR-382 Strut AUG 3 RHR Service Water OWN PRE F-A 102301 nd-isi-108 IS I F1.30b SR-383 Strut AUG 3 RHR Service Water OWN PRE F-A 102302 nd-isi-108 ISI Fl.30b SR-384 Strut AUG 3 RHR Service Water OWN PRE F-A 102303 nd-isi-108 ISI F1.30b SR-385 Ml 14-P2_RF23 /PSI /v /VT-3/ Strut AUG 3 RHR Service Water PEI-02.05.02 OWN PRE - - - - c ~F-A 102304 nd-isi-108 IS[ Fl.30b SR-386 Strut AUG 3 RHR Service Water OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 144 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 000 0 000 0 CNN C1 Nl 04 Category, DwgIlSO No. - ,I I , .1 Item No., Comp. Desc. -. LL C U-C4 U-C1 IL N LL Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 102305 nd-isi-108 ISI F1.30b SR-400 Strut AUG 3 RHR Service Water OWN PRE F-A 102306 nd-isi-108 ISI F1.30b SR-401 Strut AUG 3 RHR Service Water OWN PRE F-A 102310 nd-isi-109 ISI F1.30b SR-337 Strut AUG 3 RHR Service Water OWN PRE F-A 102311 nd-isi-109 ISI F1.30b SR-402 Strut AUG 3 RHR Service Water OWN PRE F-A 102312 nd-isi-109 ISI F1.30b SR-458 Anchor AUG 3 RHR Service Water OWN PRE F-A 102313 nd-isi-109 ISI F1.30b SR-491 Double Strul AUG 3 RHR Service Water OWN PRE F-A 102314 nd-isi-109 ISI F1.30b SR-492 Strut AUG 3 RHR Service Water OWN PRE F-A 102322 nd-isi-110 ISI F1.30b SR-494 Strut AUG 3 RHR Service Water OWN PRE F-A 102326 nd-isi-111 ISI c F1.30b SR-88 M1_I4-P2_RF23 / ISI /VTI IPEI- Stanchion AUG 3 RHR Service Water 02.05.02 OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 145 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M La N 10I Category, DwgIISO No. C,4 C)J e'm C4l U-Item No., Comp. Desc. U. U. w~ w U. Class Summary No.IComplDISystem Scope I Method / Procedure Code Case F-A 102243 nd-isi-102 ISI - - - - C F1.30c SWH-304 M1_4-P2_RF23 / ISi/VT/ PEI- Double Spring AUG 3 RHR Service Water 02.05.02 OWN PRE F-A 101682 ISI-8292-42A ISI - - - - C F1.40a DVMX Supp A M1_I4-P2_RF23 / ISI / VT / / PEI- DVMX Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101683 ISI-8292-42A ISI - - - - C F1.40a DVMX Supp B M1_14-P2_RF23 / IS[ / VT / / PEI- DVMX Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101684 ISI-8292-42A IS[ - - - - c F1.40a DVMX Supp C M1_ 4-P2_RF23 / ISI / VT / / PEI- DVMX Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101685 ISI-8292-42A ISI---------------------C F1.40a DVMX Supp D M1_I4-P2_RF24 / ISI / VT/ / FP-PE- DVMX Pump Supporl AUG 2 HPCI Pumps NDE-530 OWN PRE F-A 101686 ISI-8292-42A ISI c r F1.40a DVS Supp A M1 4-PlRF21 / ISI / VT / / PEI- DVS Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101687 ISI-8292-42A ISI c r F1.40a DVS Supp B M1 4-PlRF21 / ISI / VT / / PEI- DVS Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101688 ISI-8292-42A ISI - - - - C F1.40a DVS Supp C Ml 14-P2_RF23 / ISI /V- / / PEI- DVS Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE F-A 101689 ISI-8292-42A ISI C F1.40a DVS Supp D M1 14-P2_RF23 / ISI / VT / / PEI- DVS Pump Support AUG 2 HPCI Pumps 02.05.02 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 146 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00300 0- 01 CD 0. O Category, DwgIlSO No. N (

                                                                                                                                   "l      c             N I. I.I      I,. I Item No.,                                                                    Comp. Desc.                               C-L.

N L. C1 C4 U. U. LL 0: i-Class Summary NoiComplDISystem Scope I Method I Procedure Code Case F-A 102074 ISI-97005-C ISI-- ------- s - - - F1.40a H-10 MI1_4-P3_RF251ISIIVTI /FP-PE- Restraint AUG 1 Recirculation A NDE-530 OWN PRE F-A 102181 ISI-97006-C ISI F1.40a H-10 Restraint AUG 1 Recirc Pump B OWN PRE F-A 102741 ISI-47 ISI r C F1.40a Pump Supp A Ml-14-PlRF22 / ISI / Pump Supporl AUG 2 RCIC Turbine & Pump Ml_14-P2_RF23 / ISl / VT// PEI- OWN 02.05.02 PRE F-A 102742 ISI-47 IS1 r c F1.40a Pump Supp B M1 14-P1_RF22 /I I/// Pump Supporl AUG 2 RClC Turbine & Pump Ml_14-P2_RF23 /ISI/v/ V PEI- OWN 02.05.02 PRE F-A 102743 ISI-47 ISI r c . . ....- - - - F1.40a Pump Supp C M1_14-P1_RF22 I ISI / / Pump Support AUG 2 RCIC Turbine & Pump Ml 14-P2_RF23 / ISI/VT/ /PE- OWN 02.05.02 PRE F-A 102744 ISI-47 ISI r c F1.40a Pump Supp D M1 _4-P1_RF22 / ISI / I Pump Support AUG 2 RCIC Turbine & Pump M1 14-P2_RF23 / ISI / VT// PE- OWN 02.05.02 PRE F-A 102838 ISI-8291-76 ISI F1.40a H-1 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE F-A 102839 ISI-8291-76 ISI F1.40a H-5 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE Printed 6/5/@

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 147 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o a o a 04 N N N Category, DwgIlSO No. LO N Item No., Comp. Desc. U. 4 N LL cm LL N LL LL w Class Summary No./CompID/System Scope I Method I Procedure Code Case F-A 102840 ISI-8291-76 ISI F1.40a H-8 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE F-A 102841 ISI-8291-76 IS - F1.40a H-9 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE . . . ... . . ... . . . F-A 102842 ISI-8291-76 IS[ F1.40a H-10 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE F-A 102843 ISI-8291-76 ISI F1.40a H-13 ECCS Suction Header Hangers (Item 8 AUG MC Primary Containmen OWN PRE F-A 107562 ISI Fig 3 IS- ---------------- s - - - F1.40a W-12 M1_14-P3_RF25 / ISI / v-r / VT-3 FP- B.H. to Skirt Weld AUG 1 Reactor Vesse PE-NDE-530 OWN PRE F-A 107581 ISI-7905-32A ISI----------------- s - - - F1.40a SupportA M1_4-P3_RF25/iSI/V- // FP-PE- SupportA,E200A, 0' AUG 2 RHR Heat Exchanger P NDE-530 OWN .............. PRE F-A 107582 ISI-7905-32A IS] -------------------- c F1.40a Support B M1 14-P2_RF24 / ISI /VT / / FP-PE- Support B,E200A 180' AUG 2 RHR Heat Exchanger P NDE-530 OWN PRE F-A 107583 ISI-7905-32A IS[ - c F1.40a Support C M1_-4-PlRF22 / ISI / VT / / PEI- Support C,E200A 315' AUG 2 RHR Heat Exchanger P 02.05.02 OWN PRE F-A 107584 ISI-79U0-32B I11 Support A Support A,E200B 0' AUG F1.40a 2 RHR Heat Exchanger E OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 148 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 3 o e o 0 o e 0 N Category, DwglISO No. N 04 In N Item No., Comp. Desc. w Iw u. aL u. LL U-Class Summary No./CompID/System Scope I Method I Procedure Code Case F-A 107585 ISI-7905-32B IS1 F1.40a Support B Support B,E200B 180c AUG 2 RHR Heat Exchanger E OWN PRE F-A 107586 ISI-7905-32B ISI F1.40a Support C Support C,E200B 315' AUG 2 RHR Heat Exchanger E OWN PRE F-A 107590 ISI-48 ISI F1.40a RHR Support C Pump Supporl AUG 2 RHR Pumps OWN PRE F-A 107591 ISI-48 ISI F1.40a RHR Support D Pump Support AUG 2 RHR Pumps OWN PRE F-A 107592 ISI-48 ISI----------------- s - - - F1.40a RHR Support A M1_I4-P3_RF25 ISi VTr /FP-PE- Pump Supporl AUG 2 RHR Pumps NDE-530 OWN PRE F-A 107593 ISI-48 ISI F1.40a RHR Support B Pump Supporl AUG 2 RHR Pumps OWN PRE c ISI-49 ISI r r - - F-A 107594 F1.40a Support, Pump A M1_4-P2_RF23 /ISI VT PEI- Pump Support AUG 2 Core Spray Pumps 02.05.02 OWN PRE F-A 107595 ISI-49 ISI F1.40a Support, Pump B Pump Support AUG 2 Core Spray PumpE OWN PRE F-A 107604 IS1-8292-48-A ISI C F1.40a Turbine Suppt A Ml_14-PIRF21 / ISI IVTI IPEI- HPCI Turbine Support A AUG 2 HPCI 02.05.02 OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval IS[ Plan (Rev. 4) Page 149 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o o a oo 0a 0 N N4 N Category, DwgIlSO No. N w.- N1 0:,j Nwq I0 Item No., Comp. Desc. N: U. U. u N LL L. N IL Class Summary No./ComplD/System Scope I Method I Procedure Code Case F-A 107605 ISI-8292-48-A ISI ---------------- s - - - Fl.40a Turbine Suppt B M1 14-P3_RF25 ISI /VT / / FP-PE- HPCI Turbine Support B AUG 2 HPCI NDE-530 OWN PRE F-A 102776 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-1C 1) AUG MC Primary Containmen OWN PRE F-A 102777 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-2C 1) AUG MC Primary Containmen OWN PRE F-A 102778 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-3C 1) AUG MC Primary Containmen OWN PRE F-A 102779 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-4C 1) AUG MC Primary Containmen OWN PRE F-A 102780 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-5C 1) AUG MC Primary Containmen OWN PRE F-A 102781 151-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-6C 1) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 150 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 o a 0 Category, C-4 N oq V N DwgIlSO No. N- N

                                                                                                                              -1N                                    U, Item No.,                                                                    Comp. Desc.                                   N LL LL N

w w N LI-0. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 102782 ISI-8291-76 IS - Torus Inboard and Outboard Columns (Item F1.40b H-7C 1) AUG MC Primary Containmen OWN PRE F-A 102783 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-8C 1) AUG MC Primary Containmen OWN PRE F-A 102784 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-9C 1) AUG MC Primary Containmen OWN .. . . ... . . ... . . . PRE F-A 102785 ISI-8291-76 IS[ . . . . .. . . . ... . . . . Torus Inboard and Outboard Columns (Item F1.40b H-10C 1) AUG MC Primary Containmen OWN PRE F-A 102786 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-11C 1) AUG MC Primary Containmen OWN PRE F-A 102787 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item Fl.40b H-12C 1) AUG MC Primary Containmen OWN PRE F-A 102788 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-13C 1) AUG MC Primary Containmen OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 151 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 o O o 0 o e o 0 0 0 N N U0 Category, DwgIISO No. 0- " Item No., Comp. Desc. N. C14 N-C14 C4 U. N U. U-A Class Summary No./ComplD/System Scope I Method I Procedure Code Case F-A 102789 ISI-8291-76 ISI . . . ... . . ... . . . Torus Inboard and Outboard Columns (Item F1.40b H-14C 1) AUG MC Primary Containmen OWN PRE . . . . . . . . . . . . . . F-A 102790 ISI-8291-76 ISI . . . ... . . ... . . . Torus Inboard and Outboard Columns (Item F1.40b H-15C 1) AUG . . . ... . . ... . . . MC Primary Containmen OWN PR E . . . . . . . . . . . . . . F-A 102791 ISI-8291-76 ISI Torus Inboard and Outboard Columns (Item F1.40b H-16C 1) AUG . . . ... . . ... . . . MC Primary Containmen OWN PR E . . . . . . . . . . . . . . F-A 102792 ISI-8291-76 is[ . . . ... . . ... . . . F1.40b H-IS Torus Saddles (Item 2; AUG . . . ... . . ... . . . MC Primary Containmen OWN PR E . . . . . . . . . . . . . . F-A 102793 ISI-8291-76 ISI F1.40b H-2S Torus Saddles (Item 2, AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PRE F-A 102794 ISI-8291-76 ISI . . . ... . . ... . . . F1.40b H-3S Torus Saddles (Item 2' AUG MC Primary Containmen OWN PR E . . . . . . . . . . . . . . F-A 102795 ISI-8291-76 ISI F1.40b H-4S Torus Saddles (Item 2; AUG . . . ... . . ... . . . MC Primary Containmen OWN PRE F-A 102796 ISI-8291-76 ISI F1.40b H-5S Torus Saddles (Item 2; AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 152 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o a r~- OM o a a Category, C- N DwgIlSO No. *- ('1I Item No., Comp. Desc. N 04 0: w uL LL LL 0. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 102797 ISI-8291-76 ISI F1.40b H-6S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102798 ISI-8291-76 ISI F1.40b H-7S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102799 ISI-8291-76 IS[ F1.40b H-8S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102800 ISI-8291-76 ISI F1.40b H-9S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102801 ISI-8291-76 ISI F1.40b H-1OS Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102802 ISI-8291-76 ISI F1.40b H-11S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102803 ISI-8291-76 ISI Fl.40b H-12S Torus Saddles (Item 2, AUG MC Primary Containmen OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102804 ISI-8291-76 ISI F1.40b H-13S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102805 ISI-8291-76 ISI F1.40b H-14S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 153 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o Category, DwgIlSO No. N N to V HIi N4 N Item No., Comp. Desc. N- N LL LL w N Class Summary No./ComplDISystem Scope I Method I Procedure Code Case F-A 102806 ISI-8291-76 ISI F1.40b H-15S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102807 ISI-8291-76 ISI F1.40b H-16S Torus Saddles (Item 2, AUG MC Primary Containmen OWN PRE F-A 102808 ISI-8291-76 ISI Fl.40b H-4R Torus Seismic Restraints (Item 3: AUG MC Primary Containmen OWN PRE F-A 102809 ISI-8291-76 ISI F1.40b H-8R Torus Seismic Restraints (Item 3: AUG MC Primary Containmen OWN PRE F-A 102810 ISI-8291-76 ISI F1.40b H-12R Torus Seismic Restraints (Item 3: AUG MC Primary Containmen OWN PRE F-A 102811 ISI-8291-76 ISt F1.40b H-16R Torus Seismic Restraints (Item 3: AUG MC Primary Containmen OWN PRE F-A 102812 ISI-8291-76 ISI F1.40b H-1T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE F-A 102813 1ISI-8291-76 ISI F1.40b H-2T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 154 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 oo o a 0 C1 N 0 0 Category, DwgllSO No.

                                                                                                                         -      0              N    N-V I Item No.,                                                                    Comp. Desc.                                  NC LL. LL
                                                                                                                                 "             MI w.IL Class       Summary No.ICompID/System     Scope I Method I Procedure         Code Case F-A         102814                                                           ISI-8291-76                              ISI F1.40b      H-3T                                                             Bio Shield to Containment Truss (Item 4) AUG MC          Primary Containmen                                                                                        OWN PRE F-A         102815                                                           ISI-8291-76                              ISI F1.40b      H-4T                                                             Bio Shield to Containment Truss (Item 4) AUG MC          Primary Containmen                                                                                        OWN  ..............

PRE F-A 102816 ISI-8291-76 ISI . . . ... . . ... . . . F1.40b H-5T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE F-A 102817 ISI-8291-76 ISI F1.40b H-6T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE F-A 102818 ISI-8291-76 ISI F1.40b H-7T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE F-A 102819 ISI-8291-76 ISI F1.40b H-8T Bio Shield to Containment Truss (Item 4) AUG MC Primary Containmen OWN PRE F-A 102820 ISI-8291-76 ISI F1.40b H-1MS Drywell Male Stabilizers (Item 5 AUG MC Primary Containmen OWN PRE F-A 102821 ISI-8291-76 IS1 F1.40b H-2MS Drywell Male Stabilizers (Item 5 AUG MC Primary Containmen OWN PRE Printed 6/5/@

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 155 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o e o e oo a4 a N, NI N Category, DwgIlSO No. uIn

                                                                                                                         -     04           N   N           IN Item No.,                                                                    Comp. Desc.                                  w     w N   N,          U-Class      Summary No./ComplDISystem      Scope I Method I Procedure         Code Case F-A         102822                                                           ISI-8291-76                              ISI F1.40b      H-3MS                                                            Drywell Male Stabilizers (Item 5         AUG MC          Primary Containmen                                                                                        OWN PRE F-A         102823                                                           ISI-8291-76                              ISI F1.40b     H-4MS                                                             Drywell Male Stabilizers (Item 5         AUG MC         Primary Containmen                                                                                         OWN PRE F-A         102824                                                           ISI-8291-76                              ISI    .     . .  ...        .   . ...        .   . .

F1.4 0b H-5 MS Dryw e ll Ma le Sta b iliz e rs (Ite m 5 AUG . . . . . . . . . . . . . . MC Primary Containmen OWN PRE F-A 102825 ISI-8291-76 IS - F1.40b H-6MS Drywell Male Stabilizers (Item 5 AUG MC Primary Containmen OWN PRE F-A 102826 ISI-8291-76 ISI F1.40b H-7MS Drywell Male Stabilizers (Item 5 AUG MC Primary Containmen OWN PRE F-A 102827 ISI-8291-76 ISI Fl.40b H-8MS Drywell Male Stabilizers (Item 5 AUG MC Primary Containmen OWN PRE F-A 102828 ISI-8291-76 ISI Fl.40b H-1FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102829 ISI-8291-76 ISI F1.40b H-2FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102830 ISI-8291-76 IS1 F1.40b H-3FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 156 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 I- o* o e o e 1 N N N N Category, DwgIlSO No. U, Item No., Comp. Desc. u. uL N U. N IL N II-of Ir* Class Summary No./ComplD/System Scope / Method I Procedure Code Case F-A 102831 ISI-8291-76 ISI F1.40b H-4FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102832 ISI-8291-76 ISI F1.40b H-5FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102833 ISI-8291-76 ISI F1.40b H-6FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102834 ISI-8291-76 ISI F1.40b H-7FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102835 ISI-8291-76 ISI F1.40b H-8FS Drywell Female Stabilizers (Item 6 AUG MC Primary Containmen OWN PRE F-A 102836 ISI-8291-76 ISI F1.40b H-1 Drywell Support Skirt & Anchor Bolts (Item 7)AUG MC Primary Containmen OWN PRE F-A 102837 ISI-8291-76 ISI F1.40b H-2 Drywell Support Skirt &Anchor Bolts (Item 7)AUG MC Primary Containmen OWN PRE F-A 102844 IS;I-829 1-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-2A (Item 9) AUG MC Primary Containmen OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 157 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Mq LD o 0 oo aN o 0 N N4 N N N Category, DwgIlSO No. to Item No., Comp. Desc. w N LL LL. N w N ILL Class Summary No./ComplD/System Scope I Method I Procedure Code Case F-A 102845 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-2B (Item 9) AUG MC Primary Containmen OWN PRE F-A 102846 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-2C (Item 9) AUG MC Primary Containmen OWN PRE F-A 102847 ISI-8291-76 ISi ECCS Header Snubber Supports/Struts F1.40b H-2D (Item 9) AUG MC Primary Containmen OWN PRE F-A 102848 ISI-8291-76 iSI ECCS Header Snubber Supports/Struts F1.40b H-4A (Item 9) AUG MC Primary Containmen OWN PRE F-A 102849 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-4B (Item 9) AUG MC Primary Containmen OWN PRE F-A 102850 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-4C (Item 9) AUG MC Primary Containmen OWN PRE F-A 102851 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-4D (Item 9) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 158 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Category, Dwg/ISO No. N N1 N N1 Ul Item No., Comp. Desc. LL U. U. U. N1 L-Class Summary No.IComplDISystem Scope I Method / Procedure Code Case F-A 102852 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-6A (Item 9) AUG MC Primary Containmen OWN PRE F-A 102853 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-6B (Item 9) AUG MC Primary Containmen OWN PRE F-A 102854 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-6C (Item 9) AUG MC Primary Containmen OWN PRE F-A 102855 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-6D (Item 9) AUG MC Primary Containmen OWN PRE F-A 102856 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-12A (Item 9) AUG MC Primary Containmen OWN PRE F-A 102857 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-12B (Item 9) AUG MC Primary Containmen OWN PRE F-A 102858 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-12C (Item 9) AUG MC Primary Containmen OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 159 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 oe 0- aý o o aý N N4 Category, DwgIlSO No. oIn Item No., Comp. Desc. -1 LL N U. N LL N U-N U-wJ Class Summary No.ICompID/System Scope I Method / Procedure Code Case F-A 102859 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-12D (Item 9) AUG MC Primary Containmen OWN PRE F-A 102860 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-14A (Item 9) AUG MC Primary Containmen OWN PRE F-A 102861 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-14B (Item 9) AUG MC Primary Containmen OWN PRE F-A 102862 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-14C (Item 9) AUG MC Primary Containmen OWN PRE F-A 102863 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts Fl.40b H-14D (Item 9) AUG MC Primary Containmen OWN PRE F-A 102864 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-16A (Item 9) AUG MC Primary Containmen OWN PRE F-A 102865 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-16B (Item 9) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 160 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 r- M~ o' 0 a N C4 M le to Category, DwgIlSO No. N- N 'a Item No., Comp. Desc. U- LL NI w w LL IL U-Class Summary NoJComplDlSystem Scope I Method / Procedure Code Case F-A 102866 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-16C (Item 9) AUG MC Primary Containmen OWN PRE F-A 102867 ISI-8291-76 ISI ECCS Header Snubber Supports/Struts F1.40b H-16D (Item 9) AUG MC Primary Containmen OWN PRE F-A 102868 ISI-8291-76 ISI F1.40b H-1 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE F-A 102869 ISI-8291-76 ISI F1.40b H-2 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE .. . . ... . . ... . . . F-A 102870 ISI-8291-76 SI .. . . ... . . ... . . . F1.40b H-3 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE . . . ... . . ... . . . F-A 102871 ISI-8291-76 ISI . . . ... . . ... . . . F1.40b H-4 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE F-A 102872 ISI-8291-76 ISI F1.40b H-5 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE F-A 102873 ISI-8291-76 ISI F1.40b H-6 Vent Header Columns (Item 10 AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . F-A 102874 V Hr-8u91-7t IS1 Vent Header Columns (Item 10 AUG F1.40b H-7 MC Primary Containmen OWN PRE Printed 6/5/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 161 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Mo*in e o o oq N o N N Category, DwglISO No. T- N N Nq wI N Item No., Comp. Desc. LL U. NN U. U. L-Class Summary No./ComplDISystem Scope I Method / Procedure Code Case F-A 102875 ISI-8291-76 ISI . . . .. . . . ... . . . F1.40b H-8 Vent Header Columns (Item 10 AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102876 ISI-8291-76 ISI . . . .. . . . .. . . . . F1.40b H-9 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102877 ISI-8291-76 ISI . . ..... . . ... . . . F1.40b H-10 Vent Header Columns (Item 10 AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . F-A 102878 ISI-8291-76 ISI . . . .. . . . .. . . . . F1.40b H-11 Vent Header Columns (Item 10 AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . F-A 102879 ISI-8291-76 SI . . . .. . . . .. . . . . F1.40b H-12 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE - , F-A 102880 ISI-8291-76 ISI - . . .. - - - - - - - - F1.40b H-13 Vent Header Columns (Item 10 AUG . . . ... . . ... . . . MC Primary Containmen OWN . . . ... . . ... . . . PRE F-A 102881 ISI-8291-76 ISI F1.40b H-14 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE F-A 102882 ISI-8291-76 ISI F1.40b H-15 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE F-A 102663 ISI-8291-76 II Fl.40b H-16 Vent Header Columns (Item 10 AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 162 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 M on N N Category, DwgIlSO No.

                                                                                                                                 '-    I  C Lo Item No.,                                                                    Comp. Desc.                                          N uL uL CM             w    w              N It Class      Summary No./ComplDISystem      Scope I Method I Procedure         Code Case F-A        102884                                                            ISI-8291-76                                     ISI      .      .   .   ..    .    .   .   ..      .    .   .   .

F1.40b H-1A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102885 ISI-8291-76 ISI F1.4 0 b H-1B Dow nco m e r Bracing / R estra ints (Ite m 11) A UG . . . . . . . . . . . . . . MC Primary Containmen OWN PRE F-A 102886 ISI-8291-76 is] Fl.40b H-1C Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102887 ISI-8291-76 ISI . . . .. . . . .. . . . . F1.40b H-1D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102888 ISI-8291-76 ISI F1.40b H-1E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102889 ISI-8291-76 ISI F1.40b H-1F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102890 ISI-8291-76 ISI F1.40b H-2A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 163 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 1' M CD 0 o 0 0 NI N N Category, Dwg/ISO No. w,, N, . N", Comp. Desc. w w Item No., ii ui N wn LA. Class Summary No.ICompID/System Scope I Method I Procedure Code Case F-A 102891 ISI-8291-76 ISI F1.40b H-2B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102892 ISI-8291-76 ISI F1.40b H-3A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102893 IS1-8291-76 ISI F1.40b H-3B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102894 ISI-8291-76 SI F1.40b H-3C Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102895 ISI-8291-76 ISI F1.40b H-3D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102896 ISI-8291-76 ISI F1.40b H-3E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102897 ISI-8291-76 ISI F1.40b H-3F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 164 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 00 0 0 0 0 0 0 0 N C- UL4 N Category, DwglISO No. LIn Item No., Comp. Desc. LL C1 LL C%1 N U- II U-Class Summary No.lCompID/System Scope I Method I Procedure Code Case F-A 102898 ISI-8291-76 ISI F1.40b H-4A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102899 ISI-8291-76 ISI F1.40b H-4B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102900 ISI-8291-76 ISI F1.40b H-5A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102901 ISI-8291-76 ISI F1.40b H-5B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102902 ISI-8291-76 ISI F1.40b H-5C Downcomer Bracing I Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102903 ISI-8291-76 ISI F1.40b H-5D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102904 ISI-8291-76 ISI Fl.40b H-5E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6/5/1

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 165 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M La N aM 01 Category, DwglISO No. 0- 0c M -e N~ N' LO L 0U Item No., Comp. Desc. u- LL (,L w.,

                                                                                                                                          .L w'       LL.

w, Class w* w Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102905 ISI-8291-76 ISI F1.40b H-5F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102906 ISI-8291-76 ISI F1.40b H-6A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102907 ISI-8291-76 ISI F1.40b H-6B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102908 ISI-8291-76 ISI F1.40b H-7A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102909 ISI-8291-76 ISI F1.40b H-7B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102910 ISI-8291-76 ISI F1.40b H-7C Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102911 ISI-8291 ISI F1.40b H-7D Downcomer Bracing I Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 166 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 1-- 0) 0 NI Category, DwgIISO No. C- N N N4 Item No., Comp. Desc. U- NL N N N U. w ix Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102912 ISI-8291-76 ISI . . . .. . . . .. . . . . F1.40b H-7E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102913 ISI-8291-76 ISI F1.40b H-7F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102914 ISI-8291-76 ISI F1.40b H-8A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102915 ISI-8291-76 ISI F1.40b H-8B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE . . . . . . . . . . . . . . F-A 102916 ISI-8291-76 ISI Fl.40b H-9A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102917 ISI-8291-76 ISI Fl.40b H-9B Downcomer Bracing ! Restraints (Item 11) AUG MC Primary Containmen OWN

                                                                                                                -    PRE F-A         102918                                                           II1-8291-76                              ISI I-F1.40b      H-9C                                                             Downcomer Bracing / Restraints (Item 11) AUG MC          Primary Containmen                                                                                        OWN PRE Printed    6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 167 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00 00 0 Category, DwgIlSO No. CD I 04II1 IV) 0 I-q CD Item No., Comp. Desc. U- LL w' w' LL LL

w. w.

Ln LL Class Summary No.lComplDISystem Scope I Method I Procedure Code Case F-A 102919 ISI-8291-76 ISI F1.40b H-9D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102920 ISI-8291-76 ISI F1.40b H-9E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102921 ISI-8291-76 ISI F1.40b H-9F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102922 ISI-8291-76 IS - F1.40b H-10A Downcomer Bracing I Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102923 ISI-8291-76 ISI F1.40b H-lOB Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102924 ISI-8291-76 IS - F1.40b H-11A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102925 ISI-8291-76 ISI F1.40b H-11B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 168 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 C3 o* 0, 0 N 1N N NI .I N Category, DwgIlSO No.

                                                                                                                         ;;I.INN I.I C14     I.. V' M~     ..

Item No., Comp. Desc. uL UL Nl LL N U. LL U-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 102926 ISI-8291-76 ISI Fl.40b H-11C Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102927 ISI-8291-76 ISI Fl.40b H-11D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102928 ISI-8291-76 ISI F1.40b H-11E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102929 ISI-8291-76 Isl Fl.40b H-11F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102930 IS1-8291-76 ISI F1.40b H-12A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102931 ISI-8291-76 ISI Fl.40b H-12B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102932 ISI-8291-76 ISI F1.40b H-13A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 169 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cl) W) I- CD 0 0 0 0 N N1 N C% N Category, DwgIlSO No. mI

                                                                                                                          -   O1     C17   -4      N Item No.,                                                                    Comp. Desc.                                   N i.

N4 U. N4 LL N4 I. U-Class Summary No.IComplD/System Scope / Method I Procedure Code Case w* w* w w F-A 102933 ISI-8291-76 ISI F1.40b H-13B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102934 IS1-8291-76 ISI F1.40b H-13C Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102935 ISI-8291-76 ISI F1.40b H-13D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102936 ISI-8291-76 ISI F1.40b H-13E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102937 ISI-8291-76 ISI F1.40b H-13F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102938 ISI-8291-76 ISI F1.40b H-14A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102939 I;51-8291-76 ISI F1.40b H-14B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 170 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o o o 0 o a o e 0 N N N N N Category, DwgIlSO No. In. Item No., Comp. Desc. N U. N U. Nm N LL U. N U-Class Summary No./ComplDISystem Scope / Method I Procedure Code Case F-A 102940 ISI-8291-76 ISI F1.40b H-15A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102941 ISI-8291-76 ISI F1.40b H-15B Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102942 ISI-8291-76 ISI F1.40b H-15C Downcomer Bracing I Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102943 ISI-8291-76 ISI F1.40b H-15D Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102944 ISI-8291-76 ISI F1.40b H-15E Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102945 ISI-8291-76 ISI F1.40b H-15F Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 102946 ISI-8291-76 ISI F1.40b H-16A Downcomer Bracing / Restraints (Item 11) AUG MC Primary Containmen OWN PRE 0 Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 171 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o M La 0 a IN Nl Category, DwgIISO No. *- N NN C-4 U, Comp. Desc. N N4 N-Item No., IJ_ LL N Nx LL Summary No.ICompID/System Scope I Method I Procedure Code Case wt Class F-A 102947 ISI-8291-76 ISI F1.40b H-16B Downcomer Bracing/ Restraints (Item 11) AUG MC Primary Containmen OWN PRE F-A 107540 M1_ 4-P2_RF23 / PSI I VT/ / PEI- ISI-97005-C ISI-- ------- s - - - F1.40b H- 7 02.05.02 Snubber/Lugs AUG 1 Recirc Pump A M1_I4-P2_RF24 / PSI / VT / / FP-PE- OWN .............. NDE-530 M1_I4-P3_RF25 / ISI /VTI IFP-PE-NDE-530 PRE - - - - c B F-A 107541 Ml_14-P2_RF23 / PSI / VT / I PEI- ISI-97005-C ISI - c F1.40b H- 8 02.05.02 Snubber/Lugs AUG 1 Recirc Pump A M1_I4-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 PRE - t - c - B F-A 107542 M1_I4-P2_RF23 / PSI / VT / I PEI- ISI-97005-C ISI ---------------- s - - - F1.40b H- 9 02.05.02 Snubber / Lugs AUG 1 Recirc Pump A M1_I4-P2_RF24 / PSI / VT/ / FP-PE- OWN NDE-530 M1_14-P3_RF25 / ISI /VTI FP-PE-NDE-530 PRE - - - - CB F-A 107559 M1_I4-P2_RF23 / PSI / VT / I PEI- ISI-97006-C ISI F1.40b H- 7 02.05.02 Snubber/Lugs AUG . . . . .. . . . .. . . . . 1 Recirc Pump B M1_I4-P2_RF24 / PSI / VT / / FP-PE- OWN NDE-530 PRE - I B- - - F-A 107560 ISI-97006-C ISI . . . . . . . . . . . . . . Fl.40b H-8 Ml_14-PlRF21 /PSI/VT/I /PEI- Snubber / Lugs AUG 1 Recirc Pump B 02.05.02 OWN PRE c F-A 107561 ISI-97006-C ISI F1.40b H-9 M1_14-PlRF21 IPSI/VTI /PEI- Snubber/Lugs AUG 1 Recirc Pump B 02.05.02 OWN PRE H Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 172 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 C *) U, I O~ 0 Category, DwgIlSO No. . C4

                                                                                                                                    *1 N

1 N

                                                                                                                                                               -e
                                                                                                                                                               ,ol N

Item No., Comp. Desc. IL I* N I. N UJ. N UJ. N U-Class Summary No.IComplD/System Scope I Method I Procedure Code Case F-A 106108 ISI Fig 4 ISI----------------- s - - - F1.40c H-2SB M1_I4-P3_RF251ISI/VTI /FP-PE- VslStbi/Lug@ 90deg. AUG 1 Reactor Vesse NDE-530 OWN PRE .. . . ... . . ... . . . F-A 106109 ISI Fig 4 ISI- -------------- - s - - - F1.40c H-3SB M1_I4-P3_RF25 ISI / VT/ I FP-PE- VsI Stblzr / Lug @ 180 deg AUG 1 Reactor Vesse NDE-530 OWN PRE ISI Fig-4 I-S-- - ---- --- - - - F-A 106110 FM.40c H-4SB M14-P3_RF25 / ISI / VT / / FP-PE- VsI Stblzr / Lug @ 270 deg AUG 1 Reactor Vesse NDE-530 OWN PRE F-A 107534 ISI-97005-C ISI - - - - c F1.40c H- 1 Ml_14-P2_RF23 / ISI / VT / / PEI- Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump A 02.05.02 OWN PRE F-A 107535 ISI-97005-C ISI - - - - c F1.40c H- 2 Ml_14-P2_RF23 / ISI /VT / / PEI- Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump A 02.05.02 OWN PRE F-A 107536 ISI-97005-C ISI c F1.40c H- 3 Ml_14-PlRF22 /ISI/ VT/ /PEI- Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump A 02.05.02 OWN PRE F-A 107553 ISI-97006-C ISI .............. F1.40c H- 1 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump B OWN ............... PRE F-A 107554 ISI-97006-C ISI F1.40c H- 2 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump B OWN PRE F-A 107555 Ml_14-P1_RF21 / ISI /VTI PEI- ISI-97006-C ISI C F1.40c H- 3 02.05.02 Rod/Clevis Grip/Lugs/Constant-Suppor AUG 1 Recirc Pump B Ml_14-P2_RF23 / PSI I VT/ PEI- OWN 02.05.02 PRE C 0 Printed 6/5/20

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 173 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 0Cq 0 Lo 0 0(D. 01 Category, DwgIlSO No. - I N C.)I e La C4 Item No., Comp. Desc. U.. U UL LL LL w-Class Summary No.IComplDISystem Scope I Method I Procedure Code Case F-A 107563 ISI Fig 4 ISI r r----------- S - - - F1.40c H-1SB M1_I4-P3_RF25 / ISI / VT// FP-PE- VsI Stblzr / Lug @ 0 deg. AUG . . . ... . . ... . . . 1 Reactor Vesse NDE-530 OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . NC 105009 NC-ISI-37 ISI B B - - B B - - S - - - NCR95-068 W-1 Pipe-to-Pipe AUG . . . ... . . ... . . . NC RClC FeedwateW Ml-14-PlRF21 / ISI / UT / / PEI- PWpe.to.Pipe 02.03.01 OWN M1_I4-PlRF22 / ISI / UT/ / PEI-02.03.01 M1_I4-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 Ml_14-P2_RF23 / ISI / UTI / FP-PE-NDE-401 M1 14-P2_RF24 / ISI / IUT / / FP-PE-NDE-401 M1 14-P3_RF25 I ISI / UT / / FP-PE-NDE-401 PRE NC 105010 NC-ISI-37 iSI Bi B - - Ii BB - -I S - - - NCR95-068 W-2 Pipe-to-Tee AUG Ml_14-PlRF21 / ISI / UT/ / PEI- OWN NC RCIC Feedwatei 02.03.01 M1_I4-PlRF22 / IS[ / UT / / PEI-02.03.01 Ml_14-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 M1_I4-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 Ml_14-P2_RF24 / ISI / UTI / FP-PE-NDE-401 Ml_14-P3_RF25 / ISI / IUT / / FP-PE-NDE-401 PRE I. -..... . -. -.--- Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 174 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M~ La 1'. M 000 000 0, C1 el C1 N N Category, DwglISO No. -I CI cI - I U, Item No., Comp. Desc. CN NM LL uL N U. N-U. N U-Class Summary No./CompID/System Scope I Method I Procedure Code Case MEW aI NC 105011 NC-ISI-37 ISI B B B B s - NCR95-068 W-3 Tee-to-Pipe AUG M1-14-PlRF21 / ISI / UT / / PEI-NC RCIC Feedwatet OWN 02.03.01 Ml-14-PIRF22 / ISI / UT / / PEI-02.03.01 Ml_14-P2 RF23 / ISI / UT/ / FP-PE-NDE-401 Ml_14-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 Ml_14-P2_RF24 / ISI / UT/ / FP-PE-NDE-401 M1_I4-P3_RF25 / ISI / IUT/ / FP-PE-NDE-401 PRE 4- + 4 NC 105012 NC-ISI-37 ISI B B- B B- S - - - NCR95-068 W-4 Tee-to-Elbow AUG Ml_14-PlRF21 / ISI / UT/ / PEI-NC RCIC Feedwatej OWN 02.03.01 Ml-14-PlRF22 / ISI / UT/ / PEI-02.03.01 Ml_14-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 M1_I4-P2_RF23 / ISI / UT / / FP-PE-NDE-401 Ml_14-P2_RF24 / ISI / UT/ / FP-PE-NDE-401 Ml14-P3_RF25 / ISI / UT/ / FP-PE-NDE-401 PRE Printed 6/*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 175 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cq La oo 0 e oo o e N N N4 N Category, DwgIlSO No. - N. L0 04 Comp. Desc. N N1 Item No., L1L LL N N NL w-Class Summary NoJComplDISystem Scope I Method I Procedure Code Case NC 105013 NC-ISI-37 ISl B B - - B B - - S - - - NCR95-068 W-12 Pipe-to-Pipe AUG Ml-14-PlRF21 /ISI/UT/ /PEI-NC RCIC Feedwatei OWN 02.03.01 M114-P1_RF22 / ISI / UT/ / PEI-02.03.01 M1_I4-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 M1_I4-P2_RF23 / ISI / UT / FP-PE-NDE-401 M1I14-P2_RF24 / ISI / UT / FP-PE-NDE-401 M1I14-P3_RF25/ISI/UT/ IFP-PE-NDE-401 PRE I I. 1 4 NC 105014 NC-ISI-37. ISI B B - B B - - Is - - - NCR95-068 W-12A Pipe-to-Tee AUG M114-PlRF21 / ISI / UT / PEI-NC RCIC Feedwatei OWN 02.03.01 Ml-14-PlRF22 / ISI / UT // PEI-02.03.01 M1 14-P2_RF23/ISI/UT/ /FP-PE-NDE-401 M1_14-P2_RF23 / ISI / UT/ / FP-PE-NDE-401 M1I14-P2_RF24 / ISI / UT / / FP-PE-NDE-401 Ml14-P3_RF25 / ISI / UT / FP-PE-NDE-401 PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 176 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Category, DwglISO No. 0 N 0 C1 N1

                                                                                               - 1  C41     V*1  V        t Item No.,                                                                     Comp. Desc.        N    N      N    CN Class       Summary No.IComplDISystem     Scope I Method I Procedure          Code Case NC          105016                                                            NC-ISI-51     ISI  B    B      B    B   -      s   -   -  -

NCR95-068 W-11 Pipe-to-Tee AUG NC CRD TO RWCU OWN Ml1_4-PlRF21 / ISI / PT/ / PEI-02.01.01 Ml_14-PlRF21 / ISI / UT/ / ISI-UT-16 Ml-14-PlRF22 / ISI / PT/ / PEI-02.01.01 M1_I4-PIRF22 / ISI / UT / / PEI-02.03.11 Ml_14-P2_RF23 / ISI / UT / IPEI-02.03.11 Ml_14-P2_RF23 / ISI / UT/ / PEI-02.03.11 M1_14-P2 RF23 / ISI / UT / PEI-02.03.11 M1_14-P2_RF23 / ISI / UT / PEI-02.03.11 Ml_14-P2_RF23 / ISI / UT / PEI-02.03.11 Ml_14-P2_RF23 / ISI / UT / PEI-02.03.11 M1_I4-P2_RF24 / ISI / UT / / FP-PE-NDE-410 Ml_14-P3_RF25 / ISI / UT/ / FP-PE-NDE-410 PRE

  *

• Printed 6/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 177 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period 1 Period 2 Period 3 N1 N N N 1 N Category, DwgIISO No. - N M U) Item No., Comp. Desc. U. U.I U.. U.. N w-Class Summary No./ComplDlSystem Scope I Method I Procedure Code Case 9-  ! - I -S .- NC 105017 NU-ISI-51 ISI B B B B - - - NCR95-068 W-12 Pipe-to-Tee AUG NC CRD TO RWCU Ml-14-PlRF21 / ISI / PT/ / PEI- OWN 02.01.01 M1_I4-PlRF21 / ISI I/UT / / ISI-UT-16 Ml-14-Pl-RF22 / ISI / PT// PEI-02.01.01 Ml-14-Pl-RF22 / ISI I UT// PEI-02.03.02 Ml_14-P2_RF23 / ISI / UT // FP-PE-NDE-402 Ml_14-P2_RF23 / ISI / UT / / FP-PE-NDE-402 Ml_14-P2_RF24 / ISt / UT / / FP-PE-NDE-402 Ml_14-P3_RF25 / ISI / UT/ / FP-PE-NDE-402 PRE I------- -- Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 178 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 Cl) ILa o o C 0 o a o, 0 Category, DwglISO No. - IN- NL Item No., Comp. Desc. w N U. LL 04 w, Class Summary No.IComplDISystem Scope I Method I Procedure Code Case

                                                                                                        -I NC          105018                                                            NC-ISI-51      ISI  B    B      B   B        S   -   -  -

NCR95-068 W-13 Tee-to-Pipe AUG NC CRD TO RWCU Ml-14-PIRF21 / ISI / PT / / PEI- OWN 02.01.01 Ml-14-PIRF21 / ISI / UT// ISI-UT-16 Ml-14-PlRF22 / ISI / PT / I PEI-02.01.01 Ml-14-PlRF22 / ISI / UT/ / PEI-02.03.02 M1_14-P2_RF23 / ISI / UTI / FP-PE-NDE-402 M1_I4-P2_RF23 / ISI / UT/ / FP-PE-NDE-402 Ml_14-P2_RF24 / ISI / UT/ / FP-PE-NDE-402 M1_I4-P3_RF25 / ISI / UT/ / FP-PE-NDE-402 PRE NC 105015 NX-7879-6 ISI NC-SAC SBLC M1_I4-P2_RF24 / OWN / VT I I FP- Tank Internals AUG NC SBLC PE-NDE-530 OWN C PRE Printed 6/50

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 179 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M' La o 0 o 0 04 Category, DwgIISO No.

                                                                                                     - IC11       C.) -4 Item No.,                                                                     Comp. Desc.              C~4 N~      C-4 04i       C14 LL L.       U. IL         U.

Class Summary No./ComplDISystem Scope I Method I Procedure Code Case R-A 100993 ISI-13142-42-A ISI R1.11-2 W-1 Sweepolet-to-Elbov AUG 1 HPCI Steam OWN PRE R-A 100994 ISI-13142-42-A ISI R1.11-2 W-2 Elbow-to-PiPE AUG 1 HPCI Steam OWN PRE R-A 100995 IS1-13142-42-A ISI R1.11-2 W-3 Pipe-to-Ventur AUG 1 HPCI Steam OWN PRE R-A 100996 IS1-13142-42-A ISI R1.11-2 W-4 Venturi-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 100997 ISI-13142-42-A ISI R1.11-2 W-5 Pipe-to-Valvw AUG 1 HPCI Steam OWN PRE R-A 101007 ISI-13142-42-A ISI R1.11-2 W-15 Pipe-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 101008 ISI-13142.42-A SI------------------ s - - - R1.11-2 W-16 M1_14-P3_RF251ISIIUTI IFP-PE- Pipe-to-Pipe AUG 1 HPCI Steam NDE-401 OWN PRE R-A 101009 IS1-13142-42-A ISI c R1.11-2 W-1 7 M1_14-P2 RF24 / ISI / UT / / FP-PE- Pipe-to-ValvE AUG 1 HPCI Steam NDE-401 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 180 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o3 01 N Category, DwgIlSO No. N N N

                                                                                                                                                 %0 Item No.,                                                                     Comp. Desc.          w       w L-w  N:

Class Summary NoIComplDISystem Scope I Method I Procedure Code Case R-A 101208 IS1-13142-52-A ISI R1.11-2 W-2 Tee-to-Pipe AUG . . . ... . . ... . . . 2 Feedwatei OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101209 ISI-13142-52-A SI . . . ... . . ... . . . R1.11-2 W-3 Pipe-to-Pipe AUG . . . ... . . ... . . . 2 Feedwatei OWN PRE . . . ... . . ... . . . R-A 101210 ISI-13142-52-A ISI R1.11-2 W-4 Valve-to-PipE AUG 2 Feedwatei OWN PRE R-A 101211 ISI-13142-52-A ISI . . . ... . . ... . . . R1.11-2 W-5 Valve-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei OWN . . . ... . . ... . . . PRE . . . ISI-13142-52-A ISI . . . ... . . ... R-A 101212 R1.11-2 W-6 Pipe-to-Pipe AUG 1 Feedwatei OWN . . . ... . . ... . . . PRE R-A 101227 IS1-13142-52-A ISI - c R1.11-2 W-21 M1_I4-PlRF22/ ISI /UT/ /PEI- Pipe-to-ElboA AUG . . . ... . . ... . . . 1 Feedwatel 02.03.01 OWN - ... . . . PRE . . . ... . . ... . . . R-A 101228 ISI-13142-52-A IS1 - c . ... . . ... . . . R1.11-2 W-22 M1_14-P1_RF22 / ISI / UT/ /PEI- Elbow-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei 02.03.01 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101229 IS1-13142-52-A ISI - c R1.11-2 W-23 M1_ 4-PlRF22 / ISI / UT /PEI- Pipe-to-Safe Eno AUG 1 Feedwatei 02.03.01 OWN PRE R-A 101241 ISI-13142-52-A Is' R1.11-2 W-35 Pipe-to-Elbovu AUG 1 Feedwatei OWN PRE Printed 6/*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 181 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M U, o0* o0 C. 0 C14 C14 04 C4 C-4 Category, DwgIISO No. '- 1C'4 Comp. Desc. C14 C14 e'4 ('4 ('4 Item No., LIL L. L. LL LL Class Summary No./ComplD/System Scope I Method I Procedure Code Case R-A 101242 ISI-13142-52-A ISI . . . .. . . . .. . . . . R1.11-2 W -36 Elbow-to-Pipe AUG . . . .. . . . .. . . . . 1 Feedwatei OWN . . . ... . . ... . . . PRE . . . . . . . . . . . . . . R-A 101243 IS1-13142-52-A ISI R1.11-2 W-37 Pipe-to-New Safe Enc AUG . . . .. . . . .. . . . . 1 Feedwatel OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 101260 IS1-13142-53-A ISI . . . .. . . . .. . . . . R1.11-2 W-1 Valve-to-TeE AUG . . . . . . . . .. . . . . 2 Feedwatei OW N . . . . . . . . . . . . . . PRE . . . . . . . . . . . . . . R-A 101261 IS1-13142-53-A ISI . . . .. . . . .. . . . . R1.11-2 W-2 Tee-to-Pipe AUG . . . . . . . . .. . . . . 2 Feedwatei OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101262 IS1-13142-53-A ISI . . . .. . . . .. . . . . R1.11-2 W-3 Pipe-to-Pipe AUG . . . . . . . . . . . . . . 2 Feedwatei OW N . . . .. . . . .. . . . . PR E . . . . . . . . . . . . . . R-A 101263 IS1-13142-53-A ISI . . . .. . . . .. . . . . R1.11-2 W-4 Pipe-to-ValvE AUG . . . .. . . . .. . . . . 2 Feedwatei OWN PRE . . . ... . . ... . . . R-A 101264 ISI-13142-53-A SI[ . .-. ...-.-- .--- s - - R1.11-2 W-5 M1_I4-P3_RF25 ISI UT / FP-PE- Valve-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei NDE-401 OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 101265 ISI-13142-53-A SI-- ---------------- s - - - R1.11-2 W-6 M1_t4-P3_RF25 / ISI / UT/ / FP-PE- Pipe-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei NDE-401 OW N . . . .. . . . .. . . . . PRE . . . . . . . . . . . . . . R-A 101268 ISI-13142-53-A ISI R1.11-2 W-9 Valve-to-Elbo, AUG 1 Feedwatei OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 182 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o'q LD o a o a Category, DwgIlSO No. N N of N

                                                                                                    -    N Item No.,                                                                    Comp. Desc.             oN    N       N  N uL LL       IL Class      Summary No./ComplD/System      Scope I Method I Procedure         Code Case R-A        101269                                                            ISI-13142-53-A   ISI R1.11-2    W-10                                                              Elbow-to-ValvE   AUG 1          Feedwatei                                                                          OWN PRE R-A        101270                                                            ISI-13142-53-A   ISI R1.11-2    W-11                                                              Valve-to-PipE    AUG 1          Feedwatei                                                                          OWN PRE R-A        101280                                                            ISI-13142-53-A   ISI R1.11-2    W-21                                                              Pipe-to-Elbo%    AUG 1          Feedwatei                                                                          OWN PRE R-A        101281                                                            ISI-13142-53-A    SI------------------             s   -   -

R1.11-2 W-22 M1_I4-P3_RF25/ ISI / UT/ /FP-PE- Elbow-to-Pipe AUG 1 Feedwatei NDE-401 OWN PRE R-A 101282 ISI-13142-53-A ISI----------------- s R1.11-2 W-23 M1_I4-P3_RF25 ISI UT/ /FP-PE- Pipe-to-Safe Eno AUG 1 Feedwatei NDE401 OWN PRE R-A 101285 ISI-13142-53-A ISI R1.11-2 W-26 Tee-to-Reducei AUG 1 Feedwatei OWN PRE R-A 101286 IS1-13142-53-A ISI---------------------- R1.11-2 W-27 M1_I4-P2_RF24 /ISI / UT / / FP-PE- Reducer-to-Pipe AUG 1 Feedwate= NDE-401 OWN PRE R-A 101294 ISI-13142-53-A IS1 R1.11-2 W-35 Pipe-to-Elbow AUG 1 Feedwatei OWN PRE R-A 101295 ISI-13142-53-A ISI R1.11-2 W-36 Elbow-to-Pipe AUG 1 Feedwatei OWN PRE Printed 6/5@

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 183 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o a 0 0 o D o 0 a N N Category, DwglISO No. 04 04 N Item No., Comp. Desc. N LL N1 C-4 Code Case w w w1 U. wg Class Summary No.lComplDISystem Scope I Method I Procedure R-A 101296 IS1-13142-53-A ISI . . . ... . . ... . . . R1.11-2 W-37 Pipe-to-New Safe Enc AUG . . . ... . . ... . . . 1 Feedwatei OWN P RE . . . . . . . . . . . . . . R-A 101869 ISI-97003-A ISI . . . ... . . ... . . . R1.11-2 W-10 Pipe-to-Weldolel AUG . . . ... . . ... . . . 1 RHR Return A OWN PRE R-A 101870 ISI-97003-A ISI . . . ... . . ... . . . R1.11-2 W-11 Weldolet-to-Flange AUG . . . ... . . ... . . . 1 RHR Return A OWN . . . ... . . ... . . . PR E . . . . . . . . . . . . . . R-A 101871 ISI-97003-A ISI c r - - - - - R 1.1 1-2 W -12 Ml_14 -P l_ R F 2 1 / IS I / UT / / P E I- P ipe -to -E lbo w A UG . . . . . . . . . . . . . . 1 RHR Return A 02.03.01 OWN PRE R-A 101872 ISI-97003-A ISI c R1.11-2 W-13 M1_I4-P1_RF21 / ISI / UTI / PEI- Elbow-to-Pipe AUG . . . .. . . . .. . . . . 1 RHR Return A 02.03.01 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101873 ISI-97003-A ISI R1.11-2 W-14 Pipe-to-Branch AUG 1 RHR Return A OWN PRE ISI-97003-A ISI - - - - c R-A 101874 R1.11-2 W-15 Ml_14-P2_RF23 I ISI / UT / / FP-PE- Pipe-to-Elbow AUG 1 RHR Return A NDE-401 OWN . . . ... . . ... . . . PRE R-A 101875 ISI-97003-A ISI ------------------- c R1.11-2 W-16 Ml_14-P2_RF24 / ISI / IUT / / FP-PE- Elbow-to-Pipe AUG . . . .. . . . .. . . . . 1 RHR Return A NDE-401 OWN . . . ... . . ... . . . PRE R-A 101876 ISI-97003-A ISI R1.11-2 W-17 Pipe-to-Elbow AUG 1 RHR Return A OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 184 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 I- M~ o 01 N C4 N 0 a N N Category, DwgllSO No. NN U) Item No., Comp. Desc. w w C1 N N L,, Ii-IJL UL Class Summary NoJComplDISystem Scope I Method I Procedure Code Case R-A 101877 ISI-97003-A ISI R1.11-2 W-18 Elbow-to-Pipe AUG 1 RHR Return A OWN PRE R-A 101878 ISI-97003-A IS1 R1.11-2 W-19 Pipe-to-Elbow AUG . . . ... . . ... . . . 1 RHR Return A OWN PRE . . . ISI-97003-A ISI . . . ... . . ... R-A 101879 R1.11-2 W-20 Elbow-to-Pipe AUG 1 RHR Return A OWN PRE .. . . ... . . ... . . . R-A 101880 ISI-97003-A ISI . . . ... . . ... . . . R1.11-2 W-21 Pipe-to-Branch AUG 1 RHR Return A OWN PRE R-A 101881 ISI-97003-A IS1 R1.11-2 W-22 Branch-to-Flange AUG 1 RHR Return A OWN PRE R-A 101882 ISI-97003-A ISI R1.11-2 W-23 Pipe-to-Valve AUG 1 RHR Return A OWN PRE R-A 101883 ISI-97003-A ISI R1.11-2 W-24 Valve-to-PipE AUG 1 RHR Return A OWN . . . ... . . ... . . . PRE R-A 101884 ISI-97003-A IS[ . . . .. . . . .. . . . . R1.11-2 W-25 LS D Pipe-to-Pipe AUG . . . ... . . ... . . . 1 RHR Return A OWN PRE R-A 101949 ISI-97004-A ISl S R1.11-2 W-1 0 M1_14-P3_RF25 / ISI / UT / I FP-PE- Pipe-to-ElboN AUG 1 RHR Return B NDE-401 OWN PRE Printed 6/50

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 185 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M La a N Co 00 N Category, Dwg/ISO No. C' C-4N N N LLI Item No., Comp. Desc. w w w w N1 Ix U-Class Summary No./ComplDISystem Scope I Method I Procedure Code Case R-A 101950 ISI-97004-A SI-- - --------- c R1.11-2 W-11 M1_I4-P2 RF24 ISI UT /FP-PE- Elbow-to-PipE AUG 1 RHR Return B NDE-401 OWN PRE R-A 101951 ISI-97004-A ISI R1.11-2 W-12 Pipe-to-Weldolel AUG 1 RHR Return B OWN PRE R-A 101952 ISI-97004-A ISI R1.11-2 W-13 Pipe-to-Weldolel AUG 1 RHR Return B OWN PRE R-A 101953 ISI-97004-A ISI R 1 .11-2 W-14 W e ld o le t-to -Fla ng E A UG . . . . . . . . . . . . . . 1 RHR Return B OWN . . . ... . . ... . . . PRE R-A 101954 ISI-97004-A SI c R1.11-2 W-15 M1_I4-P1_RF21 /IS / UT / / PEI- Pipe-to-Elbow AUG 1 RHR Return B 02.03.01 OWN PRE R-A 101955 ISI-97004-A IS1 c R1.11-2 W-16 M1 14-P1_RF21 / ISI / UT / / PEI- Elbow-to-Pipe AUG 1 RHR Return B 02.03.01 OWN PRE R-A 101956 ISI-97004-A iSI R1.11-2 W-17 Pipe-to-Elbow AUG 1 RHR Return B OWN PRE R-A 101957 ISI-97004-A ISI R1.11-2 W-18 Elbow-to-Pipe AUG 1 RHR Return B OWN PRE R-A 101958 ISI-97004-A ISI R1.11-2 W-19 Pipe-to-Elbow AUG 1 RHR Return B OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 186 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 o a o a o a o a CW) N1 N La. N N N Category, DwgIlSO No. Item No., Comp. Desc.

                                                                                                             ~-Nl  I N

i N N4 N U. uL LL UL U-Class Summary No./ComplDISystem Scope / Method I Procedure Code Case R-A 101959 ISI-97004-A ISI R1.11-2 W-20 Elbow-to-Pipe AUG 1 RHR Return B OWN PRE R-A 101960 ISI-97004-A IS1 R1.11-2 W-21 Pipe-to-Weldolel AUG 1 RHR Return B OWN PRE . . . . . . . . . . . . . . R-A 101961 ISI-97004-A ISI . . . .. . . . ... . . . R 1.11-2 W-2 2 We ld o le t-to -F la n g E A UG . . . . . . . . . . . . . . 1 RHR Return B OWN PRE R-A 101962 ISI-97004-A ISI R1.11-2 W-23 Pipe-to-ValvE AUG 1 RHR Return B OWN PRE R-A 101963 ISI-97004-A ISi . . . ... . . ... . . . R1.11-2 W-24 Valve-to-Pipe AUG 1 RHR Return B OWN . . . ... . . ... . . . PRE R-A 100055 IS1-13142-42-A ISI R1.11-5 W-26 Elbow-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 100303 ISI-13142-19-B ISI . . . .. . . . .. . . . . R1.11-5 W-1 Flange-to-Tee AUG 2 RCIC Steam Discharge OWN . . . ... . . ... . . . PRE R-A 100304 IS1-13142-19-B ISI R1.11-5 W-2 Tee-to-Endcap AUG 2 RCIC Steam Discharge OWN PRE R-A 100305 IS1-13142-19-B ISI R1.11-5 W-3 Tee-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE Printed 6/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 187 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o e) 0 NN N Category, Dwg/ISO No. N N 01 Item No., Comp. Desc. w It w N LL u. Class Summary No./ComplDISystem Scope I Method I Procedure Code Case R-A 100306 ISI-13142-19-B IS - R1.11-5 W-4 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE R-A 100307 IS1-13142-19-B IS - c R1.11-5 W-5 Elbow-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE - c R-A 100308 ISI-13142-19-B IS1 - c R1.11-5 W-6 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE - c R-A 100309 IS1-13142-19-B ISI R1.11-5 W-7 Elbow-to-TeE AUG 2 RCIC Steam Discharge OWN PRE R-A 100310 ISI-13142-19-B ISI R1.11-5 W-8 Tee-to-Flange AUG 2 RCIC Steam Discharge OWN PRE R-A 100311 ISI-13142-19-B ISI R1.11-5 W-9 Tee-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE R-A 100312 ISI-13142-19-B IS1 R1.11-5 W-10 Pipe-to-ValvE AUG 2 RCIC Steam Discharge OWN PRE R-A 100313 ISI-13142-19-B IS1 R1.11-5 W-11 Valve-to-PipE AUG 2 RCIC Steam Discharge OWN PRE R-A 100314 IS1-13142-19-B IS' R1.11-5 W-1 2 Pipe-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 188 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 00 0C10 0 NI Category, DwgIlSO No. w w w o

                                                                                                '-      C"                    cl   -                    cm Item No.,                                                                     Comp. Desc.        N       N                      4  04                    U-L4 U.      U.                    U. IL Class      Summary No.IComplDISystem       Scope I Method I Procedure         Code Case R-A         100315                                                            IS1-13142-19-B ISI R1.11-5    W-13                                                               Pipe-to-ValvE  AUG    .      .     .      ...           .     .     ...          .     .     .

2 RCIC Steam Discharge OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100316 ISI-13142-19-B ISI . . . ... . . ... . . . R1.11-5 W-14 Valve-to-ElboA AUG . . . ... . . ... . . . 2 RCIC Steam Discharge OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100317 ISI-13142-19-B ISI . . . ... . . ... . . . R1.11-5 W-15 Elbow-to-ValvE AUG . . . ... . . ... . . . 2 RCIC Steam Discharge OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100318 ISI-13142-19-B ISI . . . ... . . ... . . . R1.11-5 W-16 Valve-to-PipE AUG . . . ... . . ... . . . 2 RCIC Steam Discharge OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100319 IS1-13142-19-B ISI . . . ... . . ... . . . R1.11-5 W-17 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE . . . ... . . ... . . . R-A 100320 ISI-13142-19-B ISI r c . ... . . ... . . . R1.11-5 W-18 M1_14-P1_RF22 /ISI /UT/ PEI- Elbow-to-Pipe AUG . . . ... . . ... . . . 2 RCIC Steam Discharge 02.03.01 OWN . . . ... . . ... . . . PRE . . . . . . . ISI-13142-19-B PSt . . . ... . . ... . . . R-A 100321 Pipe-to-Elbo- AUG . . . ... . . . R1.11-5 W-19 OWN . . . ... . . ... . . . 2 RCIC Steam Discharge PRE . . . ... . . ... . . . R-A 100322 ISI-13142-19-B ISI . . . ... . . ... . . . R1.11-5 W-20 Elbow-to-Pipe AUG . . . ... . . ... . . . 2 RCIC Steam Discharge OWN PRE . . . ... . . ... . . . R-A 1uu3z3 PIe-1t14o-1P-p ISI R1.11-5 W-21 Pipe-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE Printed 6/*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 189 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 q to I 04 Category, DwgIISO No. 04 N C4 w-C4 La Item No., Comp. Desc. N LL N U-N U-C,4 IL N LL. Class Summary No.ICompID/System Scope I Method I Procedure Code Case R-A 100324 IS1-13142-19-B ISI R1.11-5 W-22 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE R-A 100325 ISI-13142-19-B ISI R1.11-5 W-23 Elbow-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE R-A 100326 ISI-13142-19-B ISI R1.11-5 W-24 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE R-A 100327 IS1-13142-19-B IS I R1.11-5 W-25 Elbow-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE R-A 100328 ISI-13142-19-B ISi R1.11-5 W-26 Pipe-to-Elbow AUG 2 RCIC Steam Discharge OWN PRE R-A 100329 IS1-13142-19-B ISI R1.11-5 W-27 Elbow-to-Pipe AUG 2 RCIC Steam Discharge OWN PRE R-A 100330 ISI-13142-19-B ISI R1.11-5 W-28 Pipe-to-Torus Pent. AUG 2 RCIC Steam Discharge OWN PRE - - . . . ... . . . R-A 101010 ISI-13142-42-A isI R1.11-5 W-18 Valve-to-Tee AUG 2 HPCI Steam OWN PRE R-A 101011 ISI-13142-42-A ISI R1.11-5 W-19 Tee-to-Pipe AUG 2 HPCI Steam OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 190 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o e o e N o4 a Category, DwglISO No. oq V N N N N In Item No., Comp. Desc. N. cLL 'L N N N N N L-Class Summary No./ComplD/System Scope I Method I Procedure Code Case R-A 101012 IS1-13142-42-A ISI R1.11-5 W-20 Pipe-to-Cap AUG 2 HPCI Steam OWN PRE R-A 101013 ISI-13142-42-A ISI c . . . . . . . . . . . . . R1.11-5 W-21 M1 _14-PlRF21 /ISI /UT/ /PEI- Tee-to-Pipe AUG . . . .. . . . .. . . . . 2 HPCI Steam 02.03.01 OWN PRE R-A 101014 ISI-13142-42-A ISI R1.11-5 W-22 Pipe-to-Pipe AUG 2 HPCI Steam OWN P RE . . . . . . . . . . . . . . R-A 101015 ISI-13142-42-A SI c R1.11-5 W-23 M1_14-PM_RF21 /ISI UT /PEI- Pipe-to-Elbow AUG 2 HPC I Steam 02.03.01 O WN . . . . . . . . . . . . . . PRE R-A 101016 ISI-13142-42-A ISI R 1.11-5 W-24 Elbow-to-Pipe A UG . . . . . . . . . . . . . . 2 HP C I S te a m O WN . . . . . . . . . . . . . . PRE . . . ... . . ... . . . R-A 101017 ISI-13142-42-A ISI R1.11-5 W-25 Pipe-to-Elbow AUG . . . ... . . ... . . . 2 HPCI Steam OWN PRE R -A 10 10 19 ISI-13 14 2-4 2-A ISI . . . . . . . . . . . . . . R1.11-5 W-27 Pipe-to-Pipe AUG 2 HP C I Ste a m O WN . . . . . . . . . . . . . . P RE . . . . . . . . . . . . . . R-A 10 102 0 ISI-13 142-4 2-A SI . . . . . . . . . . . . . . R1.11-5 W-28 Pipe-to-Elbow AUG 2 HP C I Ste a m O WN . . . . . . . . . . . . . . PRE .. . . ... . . ... . . . R-A 101021 IS1-13142-42-A ISI R1.11-5 W-29 Elbow-to-Pipe AUG 2 HPCI Steam OWN PRE Printed 6/0

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 191 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 (., U, o o o o1 M o 0 01N 0 Category, DwgIlSO No. .- IN CM4 LIn Item No., Comp. Desc. N U. N U. N U. N U. U. Class Summary No./ComplD/System Scope I Method I Procedure Code Case R-A 101022 ISI-13142-42-A ISI . . . ... . . ... . . . R1.11-5 W-30 Pipe-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101023 ISI-13142-42-A ISI R1.11-5 W-31 Pipe-to-Elbow AUG 2 HPCI Steam OWN PRE R-A 101024 ISI-13142-42-A ISI R1.11-5 W-32 Elbow-to-PipE AUG 2 HPCI Steam OWN PRE R-A 101025 ISI-13142-42-A ISI R1.11-5 W-33 Pipe-to-Elbow AUG 2 HPCI Steam OWN PRE R-A 101026 ISI-13142-42-A ISI R1.11-5 W-34 Elbow-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101027 ISI-13142-42-A ISI R1.11-5 W-35 Pipe-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101028 ISI-13142-42-A ISI R1.11-5 W-36 Pipe-to-Elbow AUG 2 HPCI Steam OWN PRE R-A 101029 ISI-13142-42-A ISI R1.11-5 W-37 Elbow-to-PipE AUG 2 HPCI Steam OWN PRE 101030 ISI-13142-42-A ISI R1.11-5 W-38 Pipe-to-Elbow AUG 2 HPCI Steam OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 192 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 In* I-o CD 00 04 Category, DwgllSO No. c.J

                                                                                                   - 1      -41              M       11             V*)

Item No., Comp. Desc. C4 L. LL N LI IL NL Class Summary No.IComplDISystem Scope I Method I Procedure w-Code Case R-A 101031 ISI-13142-42-A ISI R1.11-5 W-39 Elbow-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101032 IS1-13142-42-A IS1 R1.11-5 W-40 Pipe-to-Elbow AUG 2 HPCI Steam OWN PRE R-A 101033 ISI-13142-42-A ISI R1.11-5 W-41 Elbow-to-Tee AUG 2 HPCI Steam OWN PRE R-A 101034 ISI-13142-42-A ISI R1.11-5 W-42 Tee-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101035 IS1-13142-42-A IS1 R1.11-5 W-43 Pipe-to-Cap AUG 2 HPCI Steam OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101036 IS1-13142-42-A IS1 R1.11-5 W-44 Tee-to-Pipe AUG . . . ... . . ... . . . 2 HPCI Steam OWN .............. PRE R-A 101037 ISI-13142-42-A IS1 R1.11-5 W-45 Pipe-to-Valve AUG 2 HPCI Steam OWN PRE R-A 101038 ISI-13142-42-A IS1 R1.11-5 W-46 Valve-to-Pipe AUG 2 HPCI Steam OWN PRE R-A 101039 ISI-13142-42-A ISI R1.11-5 W-47 Pipe-to-Red Elbow AUG 2 HPCI Steam OWN PRE Printed 6/*

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 193 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 M La o o 0 o 1 N N N N Category, DwgIlSO No. ~- C. Item No., Comp. Desc. U. LL U. U. u-Class Summary No./ComplDlSystem Scope / Method / Procedure Code Case R-A 101040 ISI-13142-42-A SI . . . ... . . ... . . . R1.11-5 W-48 Red Elbow-to-FlangE AUG . . . ... . . ... . . . 2 HPCI Steam OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101213 ISI-13142-52-A SI . . . ... . . ... . . . R1.11-5 W-7 Pipe-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei OWN . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 101214 IS1-13142-52-A SI-- ---------------- s - - - R 1.11-5 W-8 -P 3 _R F2 5 / ISI/ UT / / F P-P E - P ipe -t o -V alve M1 _I4 A UG . . . . . . . . . . . . . . 1 Feedwatei NDE-401 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101266 IS1-13142-53-A ISI R1.11-5 W-7 Pipe-to-Pipe AUG . . . ... . . ... . . . 1 Feedwatei OWN . . . ... . . ... . . . PRE R-A 101267 IS1-13142-53-A ISI- ---------------- s - - - R 1.11-5 W-8 -P 3 _R F 2 5 / ISI / UT / /F P-P E - P ipe -to-V a lve M1 _I4 A UG . . . . . . . . . . . . . . 1 Feedwatel NDE-401 OWN PRE R-A 101863 ISI-97003-A ISI . . . ... . . ... . . . R1.11-5 W-4 Pipe-to-ElboA AUG 1 RHR Return A OWN . . . ... . . ... . . . PRE R-A 101864 ISI-97003-A ISI R 1.11-5 W-5 E lbow-to-P ipe AUG - - . . . . . . 1 RHR Return A OWN PRE R-A 101865 ISI-97003-A SI . . . ... . . ... . . . R1.11-5 W-6 Pipe-to-Elbow AUG . . . ... . . ... . . . 1 RHR Return A OWN . . . ... . . ... . . . PRE R-A 101866 ISI1-97003-A Isi C R1.11-5 W-7 M1_14-P2_RF23 / ISI / UT / FP-PE- Elbow-to-Pipe AUG 1 RHR Return A NDE-401 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 194 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 o 0 oNJ C4 C, om N N Category, DwgIISO No. "l C4 La

                                                                                               - 1    N Item No.,                                                                    Comp. Desc.         Nl    C4                 cL   *U.

uL uL wl w U-Class Summary No.lCompID/System Scope I Method I Procedure Code Case R-A 101867 ISI-97003-A Is[ R1.11-5 W-8 Pipe-to-ValvE AUG 1 RHR Return A OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 101943 ISI-97004-A ISI . . . ... . . ... . . . R1.11-5 W-4 Pipe-to-ElbovN AUG . . . ... . . ... . . . 1 RHR Return B OWN . . . . . . . ... . . . PRE . . . ... . . ... . . . R-A 101944 ISI-97004-A ISI . . . ... . . ... . . . R1.11-5 W-5 Elbow-to-Pipe AUG 1 RHR Return B OWN PRE . . . ... . . ... . . . R-A 101945 ISI-97004-A ISI . . . ... . . ... . . . R1.11-5 W-6 Pipe-to-Elbow AUG . . . ... . . ... . . . 1 RHR Return B OWN . . . ... . . ... . . . PRE R-A 101945 ISI-97004-A ISI . . . ... . . ... . . . R1.11-5 W-7 Elbow-to-Pipe AUG . . . ... . . ... . . . 1 RHR Return B OWN PRE . . . ... . . ... . . . R-A 101947 ISI-97004-A ISI . . . ... . . ... . . . R1.11-5 W-8 Pipe-to-ValvE AUG . . . ... . . ... . . . 1 RHR Return B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 105006 ISI-13142-42-A ISI c . . ... . . ... . . . R1.11-5 W-21A Ml_I14-PIRF21 /ISI/UT/ /PEI- Pipe-to-Pipe AUG - - - - . . . . 2 HPCI Steam 02.03.01 OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R'A 105007 IS1-13142-42-A ISI c . . ... . . ... . . . R1,11-5 W-21B M1_I4-P1_RF21 / ISI / UT/ / PEI- Pipe-to-Pipe AUG 2 HPCI Steam 02.03.01 OWN PRE R-A 100998 S6I-13142-42-A ISI R1.1 1-6 W-6 Valve-to-PipE AUG HPCI Steam OWN PRE Printed 6

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 195 of 370 ASME Section XI (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period I Period 2 Period 3 M' U) CD 0 o o1 04 0 C-4 0 0 N N N N" 0404, Category, DwgIlSO No. N

                                                                                               .- 1N            V)   -e Item No.,                                                                    Comp. Desc.        N La. UL N          N U.

NC LIL U. n Class Summary No.IComplD/System Scope I Method I Procedure Code Case R-A 100999 ISI-13142-42-A ISI . . . ... . . ... . . . R1.11-6 W-7 Pipe-to-Elbow AUG 1 HPCI Steam OWN PRE R-A 101000 ISI-13142-42-A SI R1.11-6 W-8 Elbow-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 101001 ISI-13142-42-A ISI R1.11-6 W-9 Pipe-to-Elbow AUG 1 HPCI Steam OWN PRE R-A 101002 ISI-13142-42-A ISI R1.11-6 W-10 Elbow-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 101003 ISI-13142-42-A ISI R1.11-6 W-11 Pipe-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 101004 ISI-13142-42-A SI R1.11-6 W-12 Pipe-to-Elbow AUG 1 HPCI Steam OWN PRE R-A 101005 ISI-13142-42-A ISI R1.11-6 W-13 Elbow-to-PipE AUG 1 HPCI Steam OWN PRE R-A 101006 ISI-13142-42-A ISI R1.11-6 W-14 Pipe-to-Pipe AUG 1 HPCI Steam OWN PRE R-A 100114 IS1-13142-17-B IS1 Valve-to-PipE AUG R1.20-4 W-1 3 2 HPCI Water Side Sctn OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 196 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1, 2003 to May 31, 2012) Period 1 Period 2 Period 3 I* It I O C1 Category, C1 N N 04 DwgIlSO No. N- M le1. C-4 Item No., Comp. Desc. L %1 N L 11 LL. LL UL LI. LI w* Class Summary No.lComplDlSystem Scope I Method ) Procedure Code Case - R-A 100115 IS1-13142-17-B ISI- ---------------- s - - - R1.20-4 W-14 M1_I4-P3_RF25 /ISI /UT/ /FP-PE- Pipe-to-Tee AUG 2 HPCI Water Side Sctn NDE-401 OWN PRE R-A 100116 IS1-13142-17-B ISI . . . .. . . . ... . . . R1.20-4 W-15 Valve-to-PipE AUG . . . ... . . ... . . . 2 HPCI Water Side Sctn OWN PRE R-A 100117 IS1-13142-17-B ISI R1.20-4 W-16 Pipe-to-ValvE AUG . . . ... . . ... . . . 2 HPCI Water Side Sctn OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100118 ISI-13142-17-B ISI R1.20-4 W-17 Valve-to-ElboA AUG . . . ... . . ... . . . 2 HPCI Water Side Sctn OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100119 IS1-13142-17-B ISI . . . .. . . . ... . . . R1.20-4 W-18 Elbow-to-TeE AUG 2 HPCI Water Side Sctn OWN PRE . . . ... . . ... . . . R-A 100120 IS1-13142-17-B ISI . . . ... . . ... . . . R1.20-4 W-19 Tee-to-Elbow AUG . . . ... . . ... . . . 2 HPCI Water Side Sctn OWN PRE R-A 100121 lSI-13142-17-B ISI R1.20-4 W-20 Elbow-to-ValvE AUG 2 HPCI Water Side Sctn OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100122 IS1-13142-17-B ISI R1.20-4 W-21 Valve-to-PipE AUG . . . ... . . ... . . . 2 HPCI Water Side Sctn OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100123 IS1-13142-17-B ISI R1.20-4 W-22 Pipe-to-FlangE AUG 2 HPCI Water Side Sctn OWN PRE Printed 6/9

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 197 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 o e o o a N N N N Category, DwgIlSO No. N N N Item No., Comp. Desc. U. NN U. N N U, w Class Summary No.ICompID/System Scope I Method I Procedure Code Case R-A 100124 ISI-13142-17-B ISI R1.20-4 W-23 Flange-to-PipE AUG 2 HPCI Water Side Sctn OWN PRE R-A 100125 ISI-13142-17-B IS - R1.20-4 W-24 Pipe-to-Flange AUG 2 HPCI Water Side Sctn OWN PRE R-A 100126 ISI-13142-17-B ISI R1.20-4 W-25 Flange-to-Reducei AUG 2 HPCI Water Side Sctn OWN PRE R-A 100127 IS1-13142-17-B IS1 R1.20-4 W-26 Reducer-to-Pump AUG 2 HPCI Water Side Sctn OWN PRE R-A 100423 ISI Fig 1 iSI R1.20-4 W-22 Flange-to-NozzlE AUG 1 Rx yes Head Cooliný OWN PRE R-A 100439 M1_I4-P2_RF23/AUG / UT/ IFP- ISI-13142-26-A ISI R1.20-4 W-6 PE-NDE-401 Valve-to-PipE AUG - - - - c 1 Core Spray B MI 14-P2_RF23 /AUG / UT/ / FP- OWN PE-NDE-401 M1_14-P2_RF23 / AUG / UT / / FP-PE-NDE-401 PRE R-A 100440 M1 14-P2_RF23 /AUG / UT I / FP- ISI-13142-26-A ISI R1.20-4 W-7 PE-NDE-401 Pipe-to-Pipe AUG - - - - c 1 Core Spray B M1_I4-P2_RF23/AUG / UT / / FP- OWN PE-NDE-401 M1I14-P2_RF23/AUG / UT/ / FP-PE-NDE-401 PRE R-A 100441 ISI-13142-26-A ISI R1.20-4 W-8 M1 14-P3_RF25/AUG / UT/ /FP- Pipe-to-Pipe AUG S Core Spray B PE-NDE-401 OWN PRE Printed 6/5/2010

Monticello Nuclear Generating Plant 4th Interval ISI Plan (Rev. 4) Page 198 of 370 ASME Section Xl (1995 Edition, 1996 Addenda) (May 1,2003 to May 31, 2012) Period I Period 2 Period 3 Cl) iLD 00 00 00 0 04 Category, DwgIlSO No. I .1"II Itl C' - A N C4) La Item No., Comp. Desc. LL n U.. w LJL n, LI n V4 L-Class Summary No.IComplD/System Scope I Method I Procedure Code Case R-A 100442 IS1-13142-26-A ISI R1.20-4 W-9 Pipe-to-Pipe AUG . . . ... . . ... . . . 1 Core Spray B OWN . . . ... . . ... . . . PRE . . . ... . . ... . . . R-A 100447 ISI-13142-26-A ISi . . . .. . . . .. . . . . R1.20-4 W-14 Valve-to-PipE AUG 1 Core Spray B OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 100448 IS1-13142-26-A ISI . . . .. . . . ... . . . R1.20-4 W-15 Pipe-to-Elbo% AUG 1 Core Spray B OWN PRE . . . . . . . . . . . . . . R-A 100449 ISI-13142-26-A ISI . . . .. . . . .. . . . . R1.2 0 -4 W -16 E lb o w -to -V aIv E A UG . . . . . . . . . . . . . . 1 Core Spray B OW N . . . .. . . . .. . . . . PRE R-A 100450 ISI-13142-26-A IS1 . . . .. . . . .. . . . . R1.20-4 W-17 Valve-to-Bent Pipe AUG . . . ... . . ... . . . 1 Core Spray B OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 100451 ISI-13142-26-A ISi . . . .. . . . .. . . . . R1.20-4 W-18 Bent Pipe-to-Bent Pipe AUG . . . ... . . ... . . . 1 Core Spray B OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . ISI-13142-26-A ISi . . . ... . . ... . . . R-A 100452 R1.20-4 W-19 Bent Pipe-to-Safe Enc AUG . . . ... . . ... . . . 1 Core Spray B OW N . . . .. . . . .. . . . . PRE . . . ... . . ... . . . R-A 100453 ISI-13142-26-A SI . . . .. . . . .. . . . . R1.20-4 W-20 Safe End-to-NozzlE AUG 1 Core Spray B OWN PRE R-A 100500 M1_14-P2_RF23 /AUG /UT FP-I' ISV-13142-31-A ISI R1.20-4 W-1 PE-NDE-401 Valve-to-PipE AUG C I Core Spray A Ml_14-P2_RF23/AUGI/UT I FP- OWN}}