L-MT-04-050, Fourth Interval Inservice Inspection Examination Plan, Revision 2

Fourth Interval Inservice Inspection Examination Plan, Revision 2

ML042660307
Person / Time
Site:	Monticello
Issue date:	09/16/2004
From:	Thomas J. Palmisano Nuclear Management Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
L-MT-04-050
Download: ML042660307 (84)

Text

' I II SR Monticello Nuclear Generating Plant

" N Committed to Nudear Operated by Nuclear Management Company, LLC L-MT-04-050 SEP 1 6 2004 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555 Monticello Nuclear Generating Plant Docket 50-263 License No. DPR-22 Monticello Fourth Interval Inservice Inspection Examination Plan. Revision 2 Enclosed with this letter is Revision 2 to the Monticello Nuclear Generating Plant (MNGP) Fourth Interval Inservice Inspection (ISI) Examination plan. Please post changes in your copy of the MNGP Fourth Interval ISI plan. The superseded pages should be destroyed.

This letter makes no new commitments or changes to any existing commitments.

L4omas J. Palmisano Site Vice President, Monticello Nuclear Generating Plant Nuclear Management Company, LLC Enclosure cc: Administrator, Region Ill, USNRC Project Manager, Monticello, USNRC Resident Inspector, Monticello, USNRC 2807 West County Road 75 . Monticello, Minnesota 55362-9637 Telephone: 763.295.5151

• Fax: 763.295.1454 Y4

-Z S, ENCLOSURE MONTICELLO NUCLEAR GENERATING PLANT INSERVICE INSPECTION EXAMINATION PLAN REVISION 2 FOURTH INTERVAL Remove Insert Title/Review and Approval, Rev. 1 Title/Review and Approval, Rev. 2 Record of Revision, pgs i and ii, Rev. 1 Record of Revision, pgs i, ii, iii, iv, Rev. 2 Introduction, pgs 1.2-1 through 1.2-4, Introduction, pgs 1.2-1 through 1.2-4, Rev. 1 Rev. 2 Source Documents, pgs 1.3-1 and 1.3-2, Source Documents, pgs 1.3-1 and 1.3-2, Rev. 1 Rev. 2 Requests for Relief, pgs 1.5-1 through Requests for Relief, pgs 1.5-1 through 1.5-55, Rev. 0 1.5-71, Rev. 1 42 pages follow

- . i NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN NM1)

NUCLEAR MANAGEMENT COMPANY 700 1st Street HUDSON, WISCONSIN 54106 MONTICELLO NUCLEAR GENERATING PLANT 2807 WEST HIGHWAY 75 MONTICELLO, MINNESOTA 55362 INSERVICE INSPECTION EXAMINATION PLAN REVISION 2 I FOURTH INTERVAL MAY 1, 2003 THROUGH MAY 31, 2012 Prepared Ely:

Richard Deopere Section Xl ISI Coordinator, NMC - Monticello I Reviewed E3y:

Gary Park ISI Fleet Lead, NMC Approved By:t) ,

yKevin Shields Supv, Inspection and Materials, NMC - Monticello ANII Review:

Wurt Suleski ANII, Hartford Steam Boiler - CT Revision 2 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN RECORD OF REVISIONS Page Rev.

Review and Approval ........................... 2 i - iv .......................... 2 1.1-1 .................................. 0 1.2-1 through 1.2-4 ............................ 2 1.3-1 and 1.3-2 .......................... 2 1.4-1 through 1.4-2 .......................... 1 1.5-1 through 1.5-71 .......................... 1 1.6-1 through 1.6-4 ......................... 0 1.7-1 through 1.7-3 .......................... 0 Inspection Schedule (Page 1 to 326) .......................... 0 i Revision 2 5/21/2004

U NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN RECORD OF REVISIONS (cont'd)

Summary of Changes, Plan Revision I 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 a Updated revised 3rd Interval extension dates

• Added discussion regarding overlap of 3rd and 4th Intervals

• Changed 4 th Interval start dates Component Selection 0 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 Vill Section to reflect latest modification of Appendix VilI implementation per 10CFR50.55a and remove references to specific Supplements and implementation dates Page 1.2-3 Examination Personnel I Procedures clarified description of examination personnel and procedure requirements.

Clarified to reflect additional use of Mandatory Appendix VilI requirements for UT personnel and procedures as modified by 10CFR50.55a dated September 26, 2002, except where relief has been granted.

• Removed reference to Appendix VilI - 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.

ii Revision 2 I 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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:

a Format i 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

• Format / spacing changes Page 1.2-2 Code Edition Summary a Added provisions for implementation of approved ISI Relief Request #7 to use 2001 Edition of Section Xl 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 0 Capitalized Item Number Page 1.2-4 ISI Plan Overall Description (cont'd)

• Added "Rev. B-A" to TR-1 12657 Page 1.3-1 Source Documents

• Added 1995 Edition, 1995 Addenda of Section XI

• Added 2001 Edition, No Addenda of Section Xl Hi. Revision 2 5/21/2004 I

I NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Page 1.3-2 Source Documents (cont'd)

• Added NRC SER for Relief Request #7 Page 1.5-1 Relief Requests

• 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 iv Revision 2 5/21/2004 I

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN INTRODUCTION

Background:

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (BPV) Code (hereafter referred to as ASME Section Xl,Section XI, or the Code), Section Xl lnservice Inspection (ISI) Program is prepared and maintained by the Nuclear Management Company (NMC). 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 m'aintained separately from this program, and,' although they are not submitted, they are available at the plant site for audit and review.

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 Initerval (Letters to the NRC in May 2002 and January 2003 providing notificatio'n of 3rd Interval extension initiallythrough March 8, 2003 (M2002057) and subsequently May 31, 2003). The 4th Interval will overlap the 3rd interval as permitted by IWA-2430(d)(1),(2),(3), and (4) The 4th Interval begins May 1, 2003 and ends 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 thel996 Addenda (Inspection Program B) and 10CFR50.55a(g)(6)(ii)(A), except where relief has been requested. Per IOCFR50.55a(b)(2)(xi), the requirements of IWB-1220 in the 1989 Edition of ASME Section XI, "Components Exempt from Examination," shall be used for Class 1 piping instead of the 1995 Edition of ASME Section Xl with 1996 Addenda. 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 with 1995 Addenda.

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 Inforrmed Inservice Inspection Evaluation Procedure." The Risk Informed Class i and ,CIass'2 application was also conducted in a manner consistent with ASME Code Case N-578 "Risk Informred 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 Revision 2 5/21/2004

0 NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 (Relief #1) 1989 Edition IWB-1220 (10CFR50.55a)

Class 1 CRD Bolting (B7.80) Augmented program GE SIL. No. 483R2, 10CFR50.55a(b)(2)(xxi)(B) dated September 26, 2002 specifies 1995 Edition with 1995 Addenda Class 2 (Quality Group B) 1995 Edition with 1996 Addenda Risk-informed Program (Relief #1)

Class 3 (Quality Group C) 1995 Edition with 1996 Addenda MC (Metal Containment) 1992 Edition with 1992 Addenda, Subsection IWE Appendix Vill - Mandatory 1995 Edition with 1996 Addenda as modified by 10CFR50.55a dated September 26, 2002 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 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.

1.2-2 Revision 2 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN INTRODUCTION (cont'd)

Examination Personnel I Procedures: Inservice Inspection examination procedures and personnel certifications meet the requirements specified in the 1995 Edition of ASME Section XI 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, except where relief has been granted.

Reporting of Associated Section Xl Programs: The Section XI 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 Xl 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 Xl 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-1 12657, thus identifying the examination method. The examination schedule lists the 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.

1.2-3 Revision 2 5/21/2004

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN INTRODUCTION (cont'd)

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/Tank and NC-ISI-37AN-1, W-2, W-3, W4, 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 Xl 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. I Revision 2 I2.2-4 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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, 1992 Edition with 1992 Addenda, Subsection IWE ASME BPV Code Section Xl, 1995 Edition with 1996 Addenda ASME BPV Code Section XI, 1995 Edition with 1995 Addenda ASME BPV Code Section Xl, 2001 Edition with No Addenda 10CFR56.55a (66FR16391) 10CFR-50.55a(g)(6)(ii)(A)(64FR51370) ASME Section XI, 1995 Edition with 1996 Addenda, Appendix Vil Supplements 10CFR-50.55a(g)(6)(ii)(A)(66FR16391) ASME Section Xl, 1995 Edition with 1996 Addenda, Appendix Vill Supplement 4 Length Sizing Correction Regulatory Guide 1.150, Rev. 1 & Generic Letter 83-15 Regulatory Guide 1.147, Rev. 12, May 1999 Monticello Inservice Inspection Licensee Control Program, 4 AWI-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 NRC Letter, "Monticello Nuclear Generating Plant-Approval of Relief Request Number 8 of the Third 10 Year Inservice Inspection Program,"

(TAC No. M96255), November 19, 1997 NRC Letter, uMNGP-Evaluation of Relief Request No. 12 (for the 3rd 10-Year ISI Program Plan," (TAC No. MB0261), July 27, 2001 NRC Letter, "MNGP-Evaluation of Relief Request No. 13 (for the 3rd 10-Year ISI Program Plan," (TAC No. MB1833), August 22, 2001 1.3-1 Revision 2 5/21/2004

IL NUCLEAR MANAGEMENT COMPANY INSERVICE, INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Source Documents: (cont'd)

Monticello Notification Letter to NRC, "Notification of Extension of 3rd Ten-Year Inservice Testing and Inservice Inspection Intervals," May 30, 2002 NRC Letter, "MNGP-Third 10-Year Interval ISI Program Request for Relief from ASME Code,Section XI Requirements (TAC No. M133904). (Relief Request #14 for 3rd ISI Interval), April 22, 2002 NRC Letter, "Monticello Nuclear Generating Plant - Risk-informed Inservice Inspection Program (TAC MB38 19)" (Relief Request #1 for 4th ISI Interval)

NRC Letter, "Monticello Nuclear Generating Plant - Fourth 10-Year Interval Inservice Inspection Program Plan Relief Request No. 7 (TAC NO. MB6897)"

EPRI Report TR-1 12657, Rev B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," December 1999 1.3-2 Revision 2 5/21/2004. I

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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)

Reactor Vessel Circumferential Welds I (Approved July 27, 2001 for remainder of current 40-Year Operating License) 3 Appendix VIII Supplement 4 0 4 Reactor Vessel Stabilizer Brackets 0 5 Leakage at Bolted Control Rod Drive 0 (CRD) Housing Connections 6 Appendix VII Annual Training 7 Use of 2001 Addenda for Repair/Replacement 0 Program

• Relief No. 1 is approved for use during the 4th ISI Interval and is not being submitted for further NRC Review or approval.

• • Relief No. 2 is approved for use during the remaining time in the current operating license, including the 4th ISI Interval, and is not being submitted for further NRC Approval. It has been revised slightly to correct a weldname nomenclature error and update commitment statements made in Rev. 0.

1.5-1 Revision 1 5/21/2004

-I-NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. I (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 Requirements:

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)'

1.5-2 Revision 1 5/21/2004 I

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 IS 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/Docurnentation 1.5-3 Revision 1 5/21/2004

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 XI 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 Xl 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 Decisionmaking 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:

1.5-4 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 s'upply 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; 1.5-5 Revision 1 5/21/2004

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 [SI PROGRAM REQUIREMENTS 2.1 ASME Section Xl 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-1 12657 provides the requirements for defining the relationship between the RI-ISI program and the remaining unaffected portions of ASME Section XI.

2.2 Augmented Programs The following augmented inspection programs were considered during the RI-ISI application:

1.5-6 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN 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-112657, 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.

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 > 501F, AND Richardson Number > 4 (this value predicts the potential buoyancy of a stratified flow) 1.5-8 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 poteritial 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.

P 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 1.5-9 Revision 1 5/21/2004

I-NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 has 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-112657.

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.

1.5-10 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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-112657 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-112657, 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.

1.5-11 Revision 1 5/21/2004

-U NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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.

l l Class I Piping Welds1l) l Class 2 Piping Welds (2) All Piping Welds (3) l Total Selected Total Selected Total Selected 1 1 817 76 1 901 r 12 1 1718 1 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 1.5-12 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 thetexamination 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-112657, 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.

1.5-13 Revision 1 5/21/2004

IJ-NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 20x 0. 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-112657.

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.

1.5-14 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Risk Consequence Failure Potential System Category RanklI Rank e s DMs Rank In this example if FAC is not considered, the failure potential rank is mediumr 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 Chigh' risk).

FW 5 (3) Medium (High) ' Medium  : TASCS, TT, (FAC) Medium (High)

, In this example if FAC 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 ("mediums risk).

Note

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

1.5-15 Revision 1 5/21/2004

I -

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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-1 12657.

Risk Impact Results Sytr~)ARiskCDF ARiskLERF wl POD lR wlo POD wlPOD l wlo POD RPV 9.00E-1 1 9.OOE-1 I 9.OOE-1 1 9.OOE-1 1 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-1 1 -4.50E-1 1 -4.50E-1 1 -4.50E-1 1 RCR 6.98E-09 6.98E5-09 6.98E-09 6.98E-09 RCIC -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.

1.5-16 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN 3.6.2 Defense-in-Depth The intent of the inspections mandated by ASME Section Xi 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-112657 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-112657 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.

1.5-17 Revision 1 5/21/2004

I-NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl 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 Xl.) 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.

1.5-18 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Decisionmaking Inservice Inspection of Piping" Supporting Onsite Documentation Structural Integrity Calculation/File No. NMC-01-301, "Degradation Mechanism Evaluation for Class I 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-lnformed 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 11Piping Welds, ASME Categories B-F, B-J, C-F-1 and C-F-2," Revision 0 1.5-19 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.1 System Selection and Segment I 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.

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.3 Failure Potential Assessment Summary Stm Thermnal Fatigue Stress Corrosion Cracking . Localized Corrosion Flow Sensitive System _ .

TASCS TT IGSCC TGSCC ECSCC PWSCC MIC - PIT CC E-C FAC RPV RWCU Ms RCR =X_= X RCIC X __ _ _ _ _ _ _ _ _ _ _ _X RHR X _ _ _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _ _ _ _ _ _X Cs__ _ _ __ _ _ x x HPCI X _ _ _ _

FW X X _ _ _ _ __ l X X CCW CRD FPEC PCAC Torus _ _ _ _ _

Note

1. Systems are described in Table 3.1.

1.5-21 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Systemi) Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 With l WithouWi thoutu 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 I 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 (s) 2 44 59 17 20 CS 2 2 a (6) 0 4 4 4(7) 0 6 7 19 23 HPCI 2 2 4 4 3 3 11 11 FW 14(e) 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" n2 R' :ion 1

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C ( C NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Notes for Table 3.4 (cont'd)

5. Of these rive 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 'medium4 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.

1.5-23 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 System, Category I Category 2 Category 3 Category 4 Category 5 Category 6 Category 7 Total Selected Total l 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 CRD _ - - - - __ _ _ _ 10 0 31 0 FPEC _ _ __54 0 _ _ _ _ _ _ _

PCAC 47 0 Torus 49 5 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 Rl-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.

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results SstemC11 l Consequence l Failure Potential, Inspections CDF Impact" 4 , LERF Impact(<)

Category Rank DMs Rank Section Xi2) R-ISI 3) Delta w/ POD wlo POD wI POD wlo POD RPV 4 High None Low 5 3 -2 9.OOE-1 1 9.00E-1 1 9.OOE-1 1 9.OOE-1 1.

RPV 6 Medium None Low 4 0 -4 negigible negligible negiigible negligible RPV 7 Low None Low 2 0 -2 negligible negligible negligible negligible RPV Total 9.00E-11 9.00E-11 9.00E-1I 9.OOE-11 RWCU 4 High None Low 10 9 -1 4.50E-1I 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-1I MS 4 (1) High None (FAC) Low (High) 2 0- -2 9.OOE-1 1 9.OOE-1 1 9.00E-1 1- 9.OOE-1 1 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-1 1 -4.50E-1 1 -4.50E-1 I -4.50E-1 1 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-1I 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 1.5-25 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results System11 j Category Consequence Rank j DMs Failure Potential Rank S

Section X__2)

Inspections RI-lSl1 33 Delta CDF Impact4) w/ POD J w/o POD l LERF Impact(4) w/ POD_Jwlo POD RCIC 4 High None Low 0 2 2 -9.00E-11 -9.002-11 -9.OOE-11 -9.OOE-11 RCIC 5 (3) Medium UT, (FAC) Medium (High) 1 1 0 -1.20E-11 no change -1.20E-12 no change RCIC 5 Medium U 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 U 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 9.002-10 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'r "'3 R-on 1

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results Systemil) Category Consequence Failure Potential Inspections CDF lmpact'4 l LERF impact'4)

Rank DMs Rank Section Xl'2' R.lIS1 '3' Delta w/ POD wlo POD wi 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 -3.00E-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 TU Medium I 0 -1 negligible negligible negligible negligible HPCI Total -6.15E-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 -9.00E-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 TT,(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.OOE-10 FW 2 High TASCS Medium 6 4 -2 -3.24E-00 1.80E-09 -3.24E-09 1.80E-09 FW 2 High TT - Medium- 0 0 0 nochange 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 -1.00E-11 -1.80E-12 -1.OOE-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.OOE-11 -1.80E-12 -1.OOE-12 FW Total . -6.20E-09 3.90E-09 -6.17E-09 3.91 E-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 __negligible negligible neglIgible negligible 1.5-27 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 3.6-1 Risk Impact Analysis Results System 11 Consequence Failure Potential Inspections CDF Impact(4) LERF Impacte4)

Rank DMs Rank Section Xl t 'l RI-IS"_ Delta w/ POD wlo POD wI POD wlo 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 I 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 Xl 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-112657.

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 XI, 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 R `ion 1

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 5-1 Inspection Location Selection Comparison Between 1986 ASME Section Xi Code and EPRI TR-112657 by Risk Region I High Risk Region Medium Risk Region. Low Risk Region System(,) Categoryde2) Weld 1986 SectionXIX2 ) EPRITR-112657 Weld 1986 SectionXI) 2 ) EPRITR-112657 Weld 1986SectionXI121) EPRITR-112657 lCount lVol/SurSurOnly RIlSiI Other 3 Count VolSur SurOnly RI-iSi Other~3 ) Count Vol/Sur SurOnly 3 R-l-SI Other~ )

RPV B-F __5 3 2 1 3 1 2 0 B-J - ... 16 2 3 2 88 5 24 0 RWCU B-F . 1 1 0 1 .

.B-J . . 83 9 15 8 _1 0 0 0 MS B-J ' . . 105 32 1 114 99 21 21 0 B-F . . . . .. . ...1 0 1 .0 B-J SLC - 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 FW B-J 29 9 0 10 41 8 0 6(5)

C-F-2 7 5 0 0 1 1 0 0 1.5-29 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 5-1 (cont'd)

Inspection Location Selection Comparison Between 1986 ASME Section Xl Code and EPRI TR-112657 by Risk Region High Risk Region Medium Risk Region Low Risk Region System(1ode Weld 1986 Section XlI2I EPRI TR-112657 Weld 1986 Section Xlt 12 EPRI TR-112657 Weld 1986 Section Xla2) EPRI TR-112657 Count Vol/Sur SurOnly RI-ISI Other"' Count VollSur Sur Only RI-ISI OtherS' Count Vol/Sur Sur Only RI-ISI Othert32 CCW C-F-2 18 1 0 0 CRID C-17-1 31 28 0 0 C-F-2 I 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 Total B-J 65 16 0 20 459 106 27 50 264 33 59 0 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.

R/L ' 0on 1 51 -.. 2004 N

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NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 5-2 Inspection Location Selection Comparison Between 1986 ASME Section Xi Code and EPRI TR-112657 by Risk Category -

1 system c Risk Consequence Failure Potential- Code l Weld 1986 Section Xli 2)j EPRI TR-112657 Category Rank Rank Ws Rank Category Count Vol/Sur SurOnly RISI Other1 3 )

RPV 4 Medium High None Low B-F .5 3 2 1

_ _ _ _ _ _ __ _1__

_ _ _ _ _ _3__ _ _ _ _ -J 16 2 3 .2 B-F 3 1 2 2 RPV 6 Low Medium None Low

__ __ _ _ _ _ _ _3-J 80 3 22 0 RPV 7 Low Low None Low B-J 8 2 2 0 RWCU 4 Medium High None Low B-F 1 1 0 1

.B-J 83 9 15 8 -

.. RWCU 7.. Low Low None Lw --- J- I1 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-

.il 25 5 12 RCR 7 Low Low None Low B-J 12 0 3 0 1.5-31 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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(t)

Category Risk Rank

{ Consequence Rank DMs Failure Potential Code Weld 11986 Section XI(2)

VolSur Only EPRI TR-112657 RI-ISI Othert3)

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 U Medium C-F-2 20 0 0 2 . . -

RCIC 6 Low Medium None Low 1-J 5 0 2 0 C-F-2 7 1 0 0 RCIC 7 Low Low B-J 9 0 3 0 None Low ___

. C-F-2 4 0 0 0 RHR 2 High High U Medium B-F 1 1 0 0 _

_ _ _ _ _ _ _ _ _ _ _ _ _ B-J 30 4 0 8 RHR 4 Medium High None Low B-F 2 2 0 0

__ _ _ _ _ __ _ _ _ _ _ __ _ _ _ _ _ _ B-i 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 RHR 7 Low Low None Low 1-J 7 4 0 0

. 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 Cs 6 Low Medium None Low B-J a 2 0 0 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 "a9 Re'-ion 1

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C( C C, NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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~'} Risk Consequence Failure Potential Code Weld 1986 Section XI12) EPRI TR-112657 Category Rank Rank DMs Rank Category Count VollSur Sur Only Ri-ISI OtherO31 HPCI 2 High High TT MediumB-F 2 2 0 B3-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 1 FW 2 (1) High (High) High TT, (FAC) Medium (High) C-F-2 4 1 0 1 B-J -2 0 0 1 FW 2 High High TASCS,TT Medium C-F-2 1 0 0 0 ___

B - 12 5 0 4 FW 2 High High TASCS Medium 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 2 ___

FW 4 (1) Medium (High) High None (FAC) Low (High)

C-F-2' 1 I 0 0 FW 4 Medium High None ^ Low' B-J 19 3 0 2 FW .5 (3) Medium (High) Medium TASCS, T, (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 5 Medium Medium TASCS Medium B-J 2 0 0O 1 1.5-33 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Table 5-2 (cont'd)

Inspection Location Selection Comparison Between 1986 ASME Section Xl Code and EPRI TR-112657 by Risk Category Systemt11 Risk Consequence Failure Potential Code Weld 1986 Section XlI 2 EPRI TR-112657

_ _ Category Rank Rank DMs Rank Category(2 ) Count Vol/Sur Sur Only RI-ISI Othert 3 )

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 CRD 7 Low Low None Low C-F- 3 21 0 0

____________________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 Xl. The Code Categories listed in the table are therefore inaccordance 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. 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( '4 ~ '-,ion 1

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_./2004 a

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 2 (Rev. I)

Reactor Vessel Circumferential Shell Welds (note - italicized text clarifies / corrects typographical errors and omissions or describes actions taken to address implementation)

System: Reactor Vessel Class: I Category: B-A Item: B1.I1 Reactor Vessel Circumferential Welds: VCBB-4, VCBB-3 and VCBA-2 (errantly named VCBB-2 on Rev.0)

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 (B1.12) 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 Xl (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 arid 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 1.5-35 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN circumferential reactor 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 1 (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 10Q7 /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 1Ol5/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 10Q6/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.

1.5-36 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN (2) Traininq 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 Operations Procedure 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 Procedure B.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 ro6m 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.

1.5-37 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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-1 and 0255-20-IIC-1 to provide precautions that ensure proper response to a loss of station power (i.e., RWCU and Recirculation pumps are restored along with restoration of 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 Operations Procedure C.4-A for RWCU restoration in order to further inform the operations personnel of the potential of an LTOP event occurring during 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 1.5-38 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 training to operations personnel on the specific scenarios and events evaluated in Reference 1, (paragraph 2.6.1.1-5), including the features of system design and procedural controls 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 103Iyr).

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 S6ction Xl (1995 Edition with 1996 Addenda), Table IWB-2500-1, Examination Category B-A, Item B13.11, Circumferential Welds, are satisfied.

Relief is hereby requested in accordance with 10 CFR 50.55a(a)(3)(1). The proposed alternative examinations provide an adequate level of quality and safety.

1.5-39 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Alternate Examination:

As an alternative to the inspection requirements of ASME Code Section XI (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-B13-01796-1, "Reactor Vessel Fracture Toughness Engineering Evaluation - Task 5.4," March 13, 1996 1.5-40 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 (...'remainder of current 40-year operating license for the unit), (See Reference #7 above).

1.5-41 Revision 1 5/21/2004

A-NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN ATTACHMENT I 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&W B&W Cu, wI% 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 ARTNDT, OF 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 <Ixiob 2x10 ' 8.17x10-'

Probability P(FIE) Attachment 2 1.5-42 Revision 1 5/21/2004

(- ( (

ZC:)

1,I J4 0>

3.=

ATTACHMENT - 2 Circ. Weld Failure Probability vs Mean RTUT Trend Using Limiting CE, CB&I, B&W and Brunswick Data uZ i

1.OOE+00 O-D

<-I i.

0>9 i

1.0OE-.01 z I Legend:

i = CS&8 Limiting Analysis (Ref. 5)

a. 2 = Brunswick Limiting Analysis (Ref. 6)

. iZ. 1.OOE-02 *st-r A"in% I twwtyA --d I... - E) a:. 4 = CE (CEOG) Limiting Analysis (Ref. 5) 5 = B&W Limiting Analysis (Ref. 5) 1 zK 6 j A 1.001E-03 2

4'.1 w i U.Q.

5 i

II . . -4 II '.2 . ., ' ' S =3 II i * *Uc I0 1.OOE-05 mz II U NOTE: Data based on results of NRC i - Trend Une Staff analysis.

i Kn>0 1.00E-06 <

I ;m I - Monilcello Failure Probability vs Mean RTwT Projected Intersection O-C311 ;vj CD i 1.00E-07 .

I I . Z ,

N3<

-A5.i 40 46.9 50 60 70 -. 80 90 100 110

-- I N) E)- Mean RT,,T (F)

C) =)

C) I

.N" i z I

I

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I i

Z NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Monticello Unit 1 - Relief Request No. 3 (Rev. 0)

Appendix Vil 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 Vil, 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 Xl 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 Vil of Section Xl to the 1995 Edition with the 1996 Addenda.

Supplement 4 to Appendix Vil, 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).

1.5-44 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Vil 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 Vil. PDI 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 Vill, Supplement 4, 1995 Edition with the 1996 Addenda. The required implementation date for Supplement 4 was November 22, 2000. Appendix IV to 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 Vil, "Performance Demonstration for Ultrasonic Examiniation 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) 1.5-45 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Vil 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. MB1833) 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 Vil. As discussed above and demonstrated by the PDI, this will provide an acceptable level of quality and safety.

1.5-46 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl, 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 Xl 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, uSummary 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 1.5-47 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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) 1.5-48 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Monticello Unit I - ISI Relief Request No. 4 (Rev. 0)

Reactor Vessel Stabilizer Brackets System: Reactor Vessel Class: 1 Category: B-K Item: B10.10 Code Examination Requirements (ASME Section Xl, 1995 Edition with 1996 Addenda):

Perform surface examination on 100% of the vessel stabilizer bracket to vessel integral attachment welds.

Basis for Relief:

The vessel stabilizer brackets are surrounded by mirror insulation secured with cable hangers and buckles, ventilation ductwork and electrical installations.

The stabilizer brackets do not provide support during normal operation. The brackets stabilize the vessel against local and seismic loads.

Alternative Examination.

Pursuant to 10 CFR 50.55a(a)(3)(i), Monticello proposes to perform a surface examination on the stabilizer brackets if local (et reaction 'forces) or seismic loads are experienced. This proposed alternative to the requirements of Table IWB-2500-1, Category B-K, Item B10.10 will provide an acceptable level of quality and safety.

Status:

This Alternative to the Code was previously approved for 2nd and 3rd Intervals:

• 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 Not yet approved for 4th interval.

Requested for continued use'during 4th interval.

1.5-49 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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):

"Proposed alternatives to the requirements of paragraphs (c), (d), (e), (I, (g), and (h) of this section or portions thereof may 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 O-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.

1.5-50 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 0-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.

1.5-51 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl.

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).

1.5-52 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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) 1.5-53 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 XI 1995 Edition 1996 Addenda, Appendix VII, VII-4240 Examination Category: All Item Number: All

==

Description:==

All NDE Examiners performing ultrasonic volumetric examination in accordance with ASME Section Xl, 1995 Edition 1996 Addenda and Appendix VII, Annual Training.

Component Numbers: All Code and 10 CFR 50.55a Requirement:

ASME Section Xl, 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 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 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 Vil shall receive 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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 Vil - Supplements for accelerated implementation in accordance with ASME Section Xl 1995 Edition, 1996 Addenda.

1.5-54 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 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 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> 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 Xl, 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 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> 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.

1.5-55 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN Justification for Grantinq 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 Xl, 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) 1.5-56 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN RELIEF REQUEST NUMBER: ISI No. 7 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 XI 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 Xl 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.

1.5-57 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl 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 Xl 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 Xl. Each change related to Repair/Replacement Activities is addressed in the attachment to show it will be implemented at Monticello.

1.5-58 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 XI, to govern Repair/Replacement Procedures (IWX-4000) for Class 1,2,3, and MC pressure retaining components and their supports. fUsing the requirements contained in the 2001 Edition of ASME-Section XI for Repairs/Replacements at the Monticello Nuclear Generating Plant will maintain the safety of the plant. The following table indicates th6 implementation of the 2001 Edition for Repair/Replacement Activities.

Article Topic Bases IWA-1 000 Scope and Responsibility 1996 Addenda IWA-2000 Examination and Inspection 1996 Addenda IWA-3000 Acceptance Standards 1996 Addenda IWA-4000 RepairlReplacements 2001 Edition IWA-5000 Pressure Tests (Periodic) 1996 Addenda IWA-5000 Pressure Tests (RepairlReplacements) 2001 Edition IWA-6000 Records 2001 Edition IWA-9000 GlossarV 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-1 000 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-1 000 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 (RepairlReplacements) 2001 Edition 1.5-59 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 wI Appendix J IWF-1 000 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.

1.5-60 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl "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 XI 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 Xl. 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 IOCFR50.55a requires ASME Section Xl Programs to follow the 1995 Edition as amended by the 1996 Addenda of ASME Section Xl 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 XI 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.

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.

1.5-61 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN 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" TS to clarify that thermal metal removal activities fall (Note 1) within 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 Xl 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.

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.

1.5-62 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN ISI RELIEF REQUEST NUMBER: No. 7 Certificate of Reconciliation 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-4411 (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. . -

1.5-63 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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.

1.5-64 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Processina" 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.

1.5-65 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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' I 1.5-66 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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-4611.1 (c) requirements for the Repair/Replacement Program and Plan are redundant with IWA-4150. Deletion of this paragraph'eliminates this redundancy.

IWA-461 1.1(a), (b), and (c) additions are as follows:

i) IWA-4611.1 (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-4611.1 (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-4611.1(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.

1.5-67 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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 Xl, IWA-3000 acceptance criteria is used for NDE of existing metal.

B) IWA-4624.2 invokes Construction Code or Section IlIl 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) 1.5-68 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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-4711. 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 1.5-69 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th 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.

Inthe 1999 Addenda, the requirement to pressure test mechanical joints made in installation of pressure retaining items was deleted from IWA-4540, because Owners 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 Xl, 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 Xl. 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.

1.5-70 Revision 1 5/21/2004

NUCLEAR MANAGEMENT COMPANY INSERVICE INSPECTION MONTICELLO 4th INTERVAL EXAMINATION PLAN NRC Limitation I 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 Xl, 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 Xl, 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) 1.5-71 Revision 1 5/21/2004

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