ML091310666

Third 10-year Interval Inservice Inspection Program and Associated Proposed Alternatives and Relief Requests

ML091310666
Person / Time
Site:	Millstone
Issue date:	04/28/2009
From:	Price J A Dominion Nuclear Connecticut
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
09-187, NSSL/MLC, FOIA/PA-2011-0115
Download: ML091310666 (314)

Text

Dominion Nuclear Connecticut, Inc.5000 Dominion Boulevard, Glen Allen, Virginia 23060 Web Address: www.dom.com April 28, 2009 U.S. Nuclear Regulatory Commission Serial No.09-187 Attention:

Document Control Desk NSSL/MLC RO Washington, DC 20555 Docket No. 50-423 License No. NPF-49 DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 THIRD 10-YEAR INTERVAL INSERVICE INSPECTION PROGRAM AND ASSOCIATED PROPOSED ALTERNATIVES AND RELIEF REQUESTS Pursuant to 10 CFR 50.55a(g)(4), Dominion Nuclear Connecticut, Inc. (DNC) submits the Millstone Power Station Unit 3 (MPS3) inservice inspection (ISI) program for the third 10-year interval applicable to Class 1, 2, and 3 components and component supports.

The ISI program, included as Attachment 1 to this letter, describes the programmatic aspects of ISI examination of components and component supports.

The attached ISI program does not address the piping examination requirements that will result from implementation of risk-informed technology.

DNC is presently developing a risk-informed program for the examination of piping that will be applicable to Class 1 piping, as a minimum. Upon development of the risk-informed program and required supporting documents, a supplemental submittal to the Third 10-Year Interval ISI Program will be provided.

This supplemental submittal will address the risk-informed scope and modifications to the examination requirements of affected piping components.

The delayed submittal of the risk-informed portion of the ASME Section Xl program for MPS3 was discussed via teleconference with the NRC on May 8, 2008 and agreed upon in a May 13, 2008 teleconference with Mr. Siva Lingam.The ISI program has been developed in accordance with the requirements of the 2004 Edition, with no addenda, of Section Xl of the ASME Boiler and Pressure Vessel Code.MPS3 will also comply with the limitations and modifications to these requirements stated in 10 CFR 50.55a(b) related to the implementation of the 2004 ASME Code. The third 10-year ISI interval began on April 23, 2009 and MPS3 implemented the program on that date.Pursuant to 10 CFR 50.55a (a)(3)(i) and/or (ii) and 10 CFR 50.55a(g)(5)(iv), DNC is also requesting the use of alternative examination or testing requirements in place of and/or requesting relief from certain examination or testing requirements of the 2004 ASME Code. The proposed alternatives or relief requests from specific 2004 code requirements are provided in Attachment

2. Prior NRC approval is required before the relief requests can be implemented.

The MPS3 Third 10-year Interval ISI Plan and associated requests for the use of alternative or relief from specific 2004 ASME Code requirements have been reviewed and approved by the station's Facility Safety Review Committee.

DNC requests review and approval of relief request IR-3-01 by April 1, 2010 in order to utilize the relief request in the first outage (3R13) of the third 10-year interval if needed.4 Serial No.09-187 Docket No. 50-423 MPS3 Third 10-Year Interval ISI Plan Page 2 of 3 Forecasting anticipated needs based on operating experience, DNC requests review and approval of relief request IR-3-04 by August 31, 2009.DNC requests review and approval of the remaining relief requests by June 1, 2010. The remaining portions of the ISI program are within the provisions of the ASME Boiler And Pressure Vessel Code and require no NRC approval for implementation.

If you have any questions or require additional information, please contact Wanda Craft at (804) 273-4687.Sincerely, J. Al n Price Vic esident -Nuclear Engineering Attachments:

1. MPS3 Third 10-Year Interval Inservice Inspection (ISI) Plan 2. MPS3 Third 10-Year Interval Inservice Inspection (ISI) Relief Requests Commitments made in this letter: 1. None Serial No.09-187 Docket No. 50-423 MPS3 Third 10-Year Interval ISI Plan Page 3 of 3 cc: U.S. Nuclear Regulatory Commission (w/o attachments)

Region I 475 Allendale Road King of Prussia, PA 19406-1415 Ms. C. J. Sanders (w/o attachments)

NRC Project Manager, Mail Stop 8B3 U.S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852-2738 Mr. S. W. Shaffer (w/o attachments)

NRC Senior Resident Inspector Millstone Power Station Serial No.09-187 Docket No. 50-423 MPS3 Third 10-Year Interval ISI Plan ATTACHMENT 1 MILLSTONE POWER STATION UNIT 3 THIRD 10-YEAR INTERVAL INSERVICE INSPECTION (ISl) PLAN DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3 MILLSTONE UNIT 3 INSERVICE INSPECTION PROGRAM MANUAL REVISION 2 THIRD TEN-YEAR INTERVAL DOMINION NUCLEAR CONNECTICUT MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL ABSTRACT This document describes the third Ten-Year Inservice Inspection Program for Millstone Power Station Unit 3 (MPS3). This summary addresses the requirements of the ASME Boiler and Pressure Vessel (B&PV) Code,Section XI, Division 1, 2004 Edition.Additional requirements for augmented inspections are also addressed.

Included are relief requests and tables identifying the components subject to examination by Code Examination Category and Code Item Number.Pursuant to 10 CFR 50.55a(g)4(ii)

Dominion Nuclear Connecticut (DNC) is required, as a minimum, to develop this program to the 2001 Edition, through the 2003 Addenda of ASME Section XI. However, as allowed by 10 CFR 50.55a(g)(4)(iv), the third inspection interval was prepared to the requirements of the 2004 Edition of ASME Section XI with no Addenda.An alternative to the ASME Section XI requirements for the Inservice Inspection of Class 1 piping, Category B-J and B-F was implemented during the second interval based on the Risk Informed technology developed in accordance with the Westinghouse Owners Group Topical Report "WCAP 14572, Revision 1-NP-A". The request to use this alternative was submitted to the Nuclear Regulatory Commission on July 25, 2000 with approval received on March 12, 2002.The ISI Program does not currently address the piping examination requirements that will result from the implementation of risk-informed technology for the third inspection interval.

DNC is presently updating the risk-informed program for the examination of piping that will be applicable to Class 1 piping, as a minimum. Upon the completion of the update of the risk-informed program, and required supporting documents, a supplemental submittal to the third interval 1SI Program will be provided.

This supplemental submittal will address the risk-informed scope and modifications to the examination requirements of affected piping components.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE OF CONTENTS 1. INTRODUCTION

1.1 General

1.2 Applicable Editions and Addenda to Section XI 1.3 Historical ISI Program Information

1.4 System

Classification

1.5 Inspection

Program 1.5.1 Schedule 1.5.2 Additional Examinations

1.6 Responsibilities

1.6.1 Nuclear

Engineering, ISI/Material Group 1.7 Personnel Certification 1.8 Code of Federal Regulations Modifications.

Limitations and Augmented Examination Requirements

2. INSPECTION PLAN 2.1 Class 1 Components

2.1.1 Exemption

Basis 2.1.2 Component/Piping Examinations

2.2 Class

2 Components

2.2.1 Exemption

Basis 2.2.2 Component/Piping Examinations

2.3 Class

3 Components

2.3.1 Exemption

Basis 2.3.2 Component/Piping Examinations

2.4 Component

Supports 2.4.1 Exemption Basis 2.4.2 Examinations i

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2.5 Augmented Examinations

2.5.1 Reactor

Coolant Pump Flywheels 2.5.2 High Energy System Break Exclusion Area (BEA)2.5.3 Class 2 Excluded Systems 2.5.4 Reactor Vessel Bottom Mounted Instrument Nozzles 2.5.5 Reactor Vessel Head Penetrations

2.5.6 Additional

Examinations for Class 1 Alloy 600/82/182 Pressure Retaining Welds 2.6 Description of ISI Schedule Summary 2.6.1 ASME Section XI Inservice Inspection Examination Summary Tables 2.6.2 Risk-Informed Inservice Inspection Examination Summary Table 2.6.3 Augmented Examination Summary Table 2.6.4 Break Exclusion Area (BEA) Weld Listing 2.7 Procedures

2.8 Examination

Zone Listing Table 2.1 MPS3 Class 1, 2, 3 Component and Component Support ISI Examination Summary Table 2.2 MPS3 Class 1 Risked-Informed ISI Examination Summary Table 2.3 MPS3 Third Interval Augmented Examination Summary Table 2.4 MPS3 Third Interval Break Exclusion Region (BEA) Weld Listing Table 2.5 List of Examination Procedures Table 2.6 Examination Zone Listing 3. CODE CASES Table 3.1 Code Case Summary 4. EVXLUATION CRITERIA AND CALIBRATION STANDARDS 4.1 Class 1 Acceptance Standards 4.2 Class 2 Acceptance Standards 4.3 Class 3 Acceptance Standards 4. EVALUATION CRITERIA AND CALIBRATION STANDARDS (continued) ii MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 4.4 Acceptance Standards for Component Supports (Classes 1, 2, and 3)4.5 Calibration Standards 4.6 Analytical Evaluation of Flaws 4.7 Unanticipated Operating Events Table 4.1 Class 1 Acceptance Standards Table 4.2 Class 2 Acceptance Standards Table 4.3 Class 3 Acceptance Standards Table 4.4 Calibration Standards 5. SYSTEM PRESSURE TESTS 5.1 General Requirements Table 5.1-1 System Pressure Test Schedule 5.2 Visual Examination Requirements

5.3 Class

1 System Pressure Test Requirements

5.4 Class

2 and Class 3 System Pressure Test Requirements

5.5 Repair

and Replacement Pressure Test Requirements

5.6 Implementing

Instructions

5.7 Corrective

Measures 6. REPAIR/REPLACEMENT ACTIVITIES

6.1 Establishment

of a Baseline 7. RECORDS AND REPORTS 7.1 Preparation

7.2 Submittal

7.3 Retention iii MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 8. COMMITMENTS 8.1 Tracking/Logging of Correspondence 8.2 Code Cases 8.3 Relief Requests 8.4 Commitment

/ Correspondence Tracking Log Table 8.1 MPS3 ISI Commitment

/ Correspondence Tracking Log 9. RELIEF REQUESTS 9.1 Alternative Requirements 9.1.1 Code Cases Not Approved for Use 9.1.2 Hardship or Unusual Difficulty

9.2 Impractical

Requirements

9.3 Format

9.4 Relief Request Summary Tables Table 9.4-1 First Interval Relief Request Summary Table Table 9.4-2 Second Interval Relief Request Summary Table Table 9.4-3 Third Interval Relief Request Summary Table 10. (Reserved for Later Use)11. DRAWINGS 12. IMPLEMENTATION OF ASME SECTION XI, APPENDIX VIII iv MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 1. INTRODUCTION

1.1 General

1.1.1 This manual describes the MPS3 third Ten-Year Inservice Inspection Plan.The inspection plan, which consists of ASME Class 1, Class 2, and Class 3 systems and components (and their supports) has been developed utilizing applicable portions of the following documents:

• 10 CFR 50.2, Definitions

• 10 CFR 50.55a, Codes and Standards* ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2004 Edition 0 ASME Section III, Rules for Construction of Nuclear Power Plants, 1971 Edition with the Summer 1973 Addenda up to and including the 2004 Edition 0 USNRC Standard Review Plan (SRP 6.6, Section 11-7)9 USNRC Regulatory Guides: 1. RG 1.26, Rev. 3, February 1976 2. RG 1.65, Rev. 0, October 1973 3. RG 1.83, Rev. 1, July 1975 4. RG 1.84 Rev. 34, October 2007 5. RG 1.147, Rev. 15, October 2007 6. RG 1.150, Rev. 1, February 1983 7. RG 1.193, Rev 2, October 2007* MPS3 FSAR MPS3 Technical Specifications WCAP 14572, Revision 1-NP-A, "Westinghouse Owners Group Application of Risk-Informed Methods to Piping Inservice Inspection Topical Report".1-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL.ASME Code Cases: 1. N-432-1 2. N-460 3. N-504-3 4. N-513-2 5. N-526 6. N-532-4 7. N-537 8. N-545 9. N-552 10. N-566-2 11. N-586-1 12. N-600 13. N-613-1 14. N-624 15. N-638-1 16. N-639 17. N-648-1 18. N-651.19. N-658 20. N-661 21. N-663 22. N-683 23. N-686 24. N-695 25. N-696 26. N-722 27. N-729-1 28. N-731 29. N-770 Note: Code Case details may be found in Section 3.1-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 1.1.2 The programs listed below are addressed in separate program documents:

• Inservice Testing of Pumps and Valves" Class MC.and Class CC Components (Containment Inservice Inspection Program)* Examination of Steam Generator Tubing" Snubber Examination and Testing 1.1.3 Administrative implementing procedures utilized in implementation of the ISI program are listed below: 0 MP-24-ISI-FAP01, Inservice Inspection Unresolved Indication Reporting 0 MP-24-ISI-FAP02.3, ASME Section XI Pressure Tests Program for Unit 3 0 EN 3-1090, Elevated Pressure Test* SP 3601F.7, Reactor Coolant System Leak Test 0 SP 31129, Inservice Inspection Implementation 0 CEN 101C, Management of ASME Section XI Inservice Inspection Program o ER-AA-RRM-100, ASME Section XI Repair/Replacement.

Program Fleet Implementation Requirements 0 ER-AA-ISI-10, ASME Section XI Inservice Inspection Program* ER-AA-ISI-100, DNC Inservice Inspection Program* ER-AA-ISI-101, DNC Inservice Inspection Program Preparation and Change Control Process o ER-AA-SPT-10, ASME Section XI System Pressure Test Program* ER-AA-SPT-100, ASME Section XI System Pressure Test Program Fleet Implementation Requirements 0 ER-AA-ISI-RI-10, ASME Section XI Risk Informed Inservice Inspection Program* ER-AA-ISI-RI-100, DNC Risk Informed Program 1.2 Applicable Editions and Addenda to Section XI 1.2.1 Pursuant to Title 10 of the Code of Federal Regulations, Part 50, Paragraph 50.55a (10 CFR 50.55a), Final Rules dated September 10, 2008 and October 2, 2008, the inspection requirements applicable to nondestructive examination are based on the rules set 1-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL forth -in the 2004 Edition of ASME Section XI, hereinafter referred to as ASME Section XI.1.2.2 Pursuant to Title 10 of the Code of Federal Regulations, Part 50, Paragraph 50.55a (10 CFR 50.55a), Final Rules dated September 10, 2008 and October 2, 2008, the implementation of ASME Section XI, Appendix VIII "Performance Demonstration for Ultrasonic Examination Systems" is based on the 2001 Edition of ASME Section XI.1.2.3 As permitted by paragraph 50.55a(g)(4)(iv), DNC may elect to meet the requirements set forth in editions and addenda of ASME Section XI which become effective subsequent to the 2004 Edition of ASME Section XI. NRC approval is required prior to implementing these later editions or addenda (Reference NRC Regulatory Issue Summary RIS-2004-12).

Editions and Addenda of ASME Section XI or ASME Code Cases that are adopted will be identified in the appropriate sections of thisinspection program.* It is the intent of DNC to apply appropriate revisions of ASME Section XI, with NRC approval, which improve the overall quality of MPS3's inspection program. Those changes that are applied will be identified in this Section of the ISI Program.1.3 Historical ISI Program Information 1.3.1 The base code of record for the preservice inspection was the 1980 Edition through Winter 1980 Addenda of ASME Section XI.1.3.2 The base code of record for the first Ten-Year ISI inspection interval was the 1983 Edition through Summer 1983 Addenda of ASME Section XI, except for Class 2 piping welds which were updated to the 1983 Edition through Winter 1985 Addenda.1.3.3 The first 10-Year Inspection Interval began on April 23, 1986 and ended on October 23, 1999. The interval was extended by 18 months based on the unit being out of service continuously greater than 6 months as allowed by IWA-2430(e) and extended an additional 1 year as allowed by IWA-2430(d).

Ref. NRC correspondence letter B 16015 dated November 20, 1996 and letter B17355 dated July 14, 1998.1.3.4 The second 10-Year Inspection Interval began on April 23, 1999 (overlapping with the end of the first interval) and is scheduled to end on April 22, 2009. The interval was adjusted by six months 1-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL per NRC Letter S/N 07-0340, dated May 03, 2007. The Code of record was the 1989 Edition of ASME Section XI.1.4 System Classification 1.4.1 The construction permit for MPS3 was issued in August1974.

The operating license was issued in January 1986. Northeast Nuclear Energy Company was the owner of record, and Stone &Webster was the Architect Engineer and installer of record.1.4.2 The system classifications for the Inservice Inspection Plan are based on the requirements of 10 CFR 50 and Regulatory Guide 1.26.1.4.3 Class 1 system boundaries were developed based on the 10 CFR 50.2, Reactor Coolant Pressure Boundary definition.

1.4.4 Class

2 and Class 3 system boundaries were developed based on Regulatory Guide 1.26.1.4.5 System boundary diagrams are listed below and depict the specific boundaries for the Class 1, Class 2, and Class 3 systems. These are controlled documents in accordance with site procedures.

Applicable Class Drawing Number Class 1 Boundaries 25212-20997 Class 2 Boundaries 25212-20998 Sht. 1 & 2 Class 3 Boundaries 25212-20999 Sht. 1, 2, & 3 System pressure test boundaries are described in Section 5.1.5 Inspection Program 1.5.1 Schedule Examinations for the third ten-year interval are scheduled in accordance with Inspection Program B, as described in IWA-2400 of ASME Section XI, for the Class 1, Class 2, and Class 3 systems, components, and supports.

Where the original schedule conflicts with other activities during the refueling outage, examinations may be rescheduled as long as the requirements of IWX-2412 are met.The sequence of examinations established for this inspection interval have been scheduled as close as practical to that of the 1'-5 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL previous inspection intervals.

The duration between examinations of the second interval and the third interval may exceed 10 years due to the extended down time of the unit and interval extensions allowed by IWA-2430.The Third Inspection Interval will begin on April 23, 2009 and Will end April 22, 2019.The refueling outage schedule and corresponding inspection periods are listed below: Refueling Outage Year Scheduled Inspection Period 3R13 Spring 2010 1 3R14 Fall 2011 1 3R15 Spring 2013 2 3R16 Fall 2014 2 3R17 Spring 2016 3 3R18 Fall 2017 3 3R19 Spring 2019 3 Examination Period Dates First Period -April 23, 2009 to April 22, 2012 Second Period -April23, 2012 to August 22, 2015 Third Period -August 23, 2015 to April 22, 2019 DNC may increase or decrease the inspection interval by as much as one year as allowed by ASME Section XI, Article IWA-2400.IWA-2430(d)(2) allows performance of examinations in overlapping intervals, as long as code credit for the given exam is taken for only one of the intervals..

DNC may increase or decrease the inspection period by as much as 1 year to enable an inspection to coincide within a plant outage as allowed by ASME Section XI, IWA-2430(d)(1).

1.5.2 Additional

Examinations

[Commitment Table 8.1, File B 16368]1-6 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL NOTE: If in the expansion to perform additional examinations it is found that ASME Section XI requirements cannot be met in the performance of these exams, then relief shall be sought as specified in 3.1 of this Program Manual.A. Class 1 Additional examinations for Class 1 equivalent components (IWB) shall be in accordance with the requirements of IWB-2430 or the alternative requirements of Code Case N-586-1 in lieu of those in 1WB-2430(a).

The additional examination samples are defined as those items (welds, areas, or parts) in a particular examination category and item number. The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.(1) Examinations performed in accordance with Table IWB-2500-1, except for Examination Category B-P, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional number of welds, areas, or parts included in the inspection item equal to the number of welds, areas, or parts included in the inspection item that were scheduled to be performed during the present inspection period. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

(2) If the additional examinations required by 1WB-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1, the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining number of welds, areas, or parts of similar material and service subject to the same type of flaws or relevant conditions.

(3) For the inspection period following the period in which the examinations of IWB-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with 1WB-2400.1-7 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL (4) For steam generator tubing, additional examinations shall be governed by plant Technical Specifications.

(5) If welded attachments are examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of Table IWB-3410-1, additional examinations shall be performed, if determined necessary, based ont an evaluation by DNC.An alternative to the requirements of IWB-2430 was previously implemented during the second inspection interval for Class 1 Risk Informed piping (Category R-A)examinations as submitted in Relief Request 1-RI-ISI-01 dated July 25, 2000. The Class 1 Risk Informed program is presently being updated for the third inspection interval and will be submitted following completion.

This submittal includes a request to continue to use the following alternative to IWB-2430;(1) An engineering evaluation shall be performed to determine 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 determine whether other elements on the segment or segments are subject to the same root cause and degradation mechanism.

Additional examinations will be performed on these elements up to a number equivalent to the number of elements required to be inspected on the segment or segments.(2) If unacceptable flaws or relevant conditions are again found in the additional examinations similar to the initial condition, the remaining elements identified as susceptible will be examined.(3) No additional examinations will be performed if there are no additional elements identified as being susceptible to the same service related root cause conditions or degradation mechanisms.

1-8 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL B. Class 2 Additional examinations for Class 2 equivalent components (1WC) shall be selected in accordance with IWC-2430 or the alternative requirements of Code Case N-586-1 in lieu of those in 1WC-2430(a).

The additional examination samples are defined as those items (welds, areas, or parts)in a particular examination category.

The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.(1) Examinations performed in accordance with Table IWC-2500-1, except for Examination Category C-H, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-3410-1 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional number of welds, areas, or parts included in the inspection item equal to 20% of the number of welds, areas, or parts included in the inspection item that are scheduled to be performed during the interval.

The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

(2) If the additional examinations required by IWC-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-3410-1, the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining number of welds, areas, or parts of similar material and service subject to the same type of flaws or relevant conditions.

(3) For the inspection period following the period in which the examinations of IWC-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with IWC-2400.(4) If welded attachments are examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of Table 1WC-3410-1, additional examinations shall be performed, if determined necessary, based on an evaluation by DNC.1-9 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL C. Class 3 Additional examinations for Class 3 equivalent components (1WD) shall be selected in accordance with IWD-2430 or the alternative requirements of Code Case N-586-1 in lieu of those in IWD-2430(a).

The additional examination samples are defined as those items (welds, areas, or parts)in a particular examination category.

The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.(1) Examinations performed in accordance with Table IWD-2500-1, except for Examination Category D-B, that reveal flaws or relevant conditions exceeding the acceptance standards of IWD-3000 shall be extended to include additional examinations during the current outage.The additional examinations shall include an additional number of welds, areas, or parts included in the inspection item equal to 20% of the number of welds, areas, or parts included in the inspection item that are scheduled to be performed during the interval.

The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

(2) If the additional examinations required by IWD-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of IWD-3000, the examinations shall be further extended to include additional examinations during the current outage. The extent of the additional examinations shall be determined by DNC based upon an engineering evaluation of the root cause of the flaws or relevant conditions.

DNC's corrective actions shall be documented in accordance with IWA-6000.(3) For the inspection period following the period in which the examinations of IWD-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with IWD-2400.(4) If welded attachmentsare examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of 1-10 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL IWD-3000, additional examinations shall be performed, if determined necessary, based on an evaluation by DNC.D. Component Supports Additional examinations for Class 1, 2, 3 and MC component supports shall be selected in accordance with IWF-2430 or the alternative requirements of Code Case N-586-1 in lieu of those in lWF-2430(a).

The additional examination samples are defined as those component supports in a particular examination category.

The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.NOTE: When an inservice examination of a component support reveals conditions described in ASME.Section XI, and the component support has been analyzed and/or tested to substantiate its integrity for its intended service, and has been found to be acceptable and corrective measures have been performed to restore the support to its original condition, then additional support examinations are not required. (Reference ASME Inquiry XI- 1-86-30R2.)

(1) Component support examinations performed in accordance with Table IWF-2500-1 that reveal flaws or relevant conditions exceeding the acceptance standards of 1IWF-3400 shall be extended, during the current outage, to include the component supports immediately adjacent to those component supports for which corrective action is required.

The additional examinations shall be extended, during the current outage, to include additional supports within the system, equal in number and of the same type and function as those scheduled for examination during the inspection period.(2) When the additional examinations required by lWF-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining component supports within the system of the same type and function.(3) When the additional examinations required by IWF-2430(b) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, the examinations shall 1-11 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL be extended, during the current outage, to include all nonexempt supports potentially subject to the same failure modes that required corrective actions in accordance with IWF-2430(a) and (b). Also, these additional examinations shall include nonexempt component supports in other systems when the support failures requiring corrective actions indicate non-system-related support failure modes.(4) When the additional examinations required by IWF-2430(c) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, DNC shall examine, during the current outage, those exempt component

.supports that could be affected by the same observed failure modes and could affect nonexempt components.

1.6 Responsibilities

1.6.1 Nuclear

Engineering, ISI/Materials Group The ISllMaterials Group is responsible for the establishment and implementation of the inservice inspection (ISI) program. The following procedures detail the responsibilities related to implementation of the ISI Program:* ER-AA-ISI-10, ASME Section XI Inservice Inspection Program* ER-AA-ISI-100, DNC Inservice Inspection Program" ER-AA-ISI-101, Dominion Inservice Inspection Program Preparation and Change Control Process" ER-AA-ISI-102, Dominion Inservice Inspection IDDEAL Software Suite" ER-AA-SPT-10, ASME Section XI System Pressure Test Program* ER-AA-SPT-100, ASME Section XI System Pressure Test Program Fleet Implementation Requirements

• ER-AA-ISI-RI-10, ASME Section XI Risk Informed Inservice Inspection Program" ER-AA-ISI-RI-100, DNC Risk Informed Program* ER-AA-ISI-RI-101, The Dominion Risk Informed Period Update Process The ISI Program Owner (DNC) retains primary responsibility for implementation of the ISI program.1-12 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 1.7 Personnel Certification Personnel shall be qualified and certified to perform nondestructive examination (NDE) in accordance with IWA-2300 and applicable site procedures.

Personnel performing VT-2 examinations shall be certified in accordance with 1WA-2300.1.8 Code of Federal Regulations Modifications, Limitations and Augmented Examination Requirements The following mandatory and optional Code of Federal Regulations Limitations, Modifications and Augmented Examination Requirements are included in 10 CFR 50.55a as published on September 10, 2008 and amended on October 2, 2008. Only those 10 CFR 50.55a Limitations, Modifications and Augmented Examination Requirements applicable to the 2004 Edition of Section XI nondestructive examination requirements for Class 1, 2, and 3 components and component supports are listed.These Limitations, Modifications and Augmented Examination Requirements were reviewed for inclusion in the ISI Program Manual and dispositioned as follows: 1.8.1 MPS3 will not implement the option in 10 CFR 50.55a(b)(2)(i), to utilize ASME Section XI, 1974 Edition with Addenda through Summer 1975 and ASME Section XI 1977 Edition with Addenda through Summer 1978.1.8.2 MPS3 will not utilize the option in 10 CFR 50.55a(b)(2)(ii), to examine Class 1 piping per ASME Section XI, 1974 Edition with the Summer 1975 Addenda.1.8.3 As allowed by 10 CFR 50.55a(b)(2)(iii), steam generator tubing at MPS3 will be examined in accordance with plant Technical Specification 3/4.4.5 in lieu of Article IWB-2000.1.8.4 MPS3 will not utilize the option in 10 CFR 50.55aib)(2)(iv), to examine Class 2 piping per ASME Section XI, 1974 Edition with the Summer 1975 Addenda and the 1983 Edition through the Summer 1983 Addenda.1.8.5 The MPS3 design includes a concrete containment subject to ASME Section XI, Subsection IWL requirements.

Therefore the mandatory modification in 10 CFR 50.55a(b)(2)(viii) applies to MPS3. Subsection IWL requirements and 10 CFR 50.55a 1-13 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL requirements are addressed under Containment Inspection Program MP-24-CII-PROG.

1.8.6 The MPS3 design includes a metal containment subject to ASME Section XI, Subsection IWE requirements.

Therefore the mandatory modification in 10 CFR* 50.55a(b)(2)(ix) applies to MPS3. Subsection IWE requirements and 1OCFR50.55.a requirements are addressed under Containment Inspection Program MP-24-CII-PROG.

1.8.7 As required by 10 CFR 50.55a(b)(2)(x), MPS3 will apply the station Appendix B Quality Assurance Program of NQA- 1 to Section XI activities.

1.8.8 The requirements for performing underwater welding as stated in 10 CFR 50.55a(b)(2)(xii) are not addressed in the MPS3 ISI Program. Repair Replacement activities are addressed in Dominion Fleet Repair Replacement Program ER-AA-RRM-100.

1.8.9 As allowed by 10 CFR 50.55a(b)(2)(xiv), for Appendix VIII Qualified Personnel, MPS3 will use the annual practice requirements in VII-4240 of Section XI Appendix VII in place of the 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 (when deemed appropriate) as discussed in 10 CFR 50.55a(b)(2)(xiv).

When utilizing this option, the annual practice requirements will be performed on material or welds that contain cracks, or by analyzing prerecorded data from material or welds that contain cracks. All training will'be completed no earlier than 6 months prior to performing ultrasonic examinations.

The implementation of ASME Section XI, Appendix VII requirements is addressed in Section 12 of this ISI Program Manual and Dominion Fleet Procedure No. ER-AA-NDE- 121.1.8.10 MPS3 will not implement the optional Appendix VIII specimen set and qualification provisions in paragraphs (b)(2)(xv)(A) to (b)(2)(xv)(M) in accordance with 10 CFR 50.55a(b)(2)(xv).

The implementation of ASME Section XI, Appendix VIII requirements is addressed in Section 12 of this ISI Program Manual and.Dominion Fleet Procedure No. ER-AA-NDE-122.

Note that the* alternative requirements of Code Case N-695 will be utilized-in lieu of those in Appendix VIII, Supplement

10. In addition, the alternative requirements of Code Case N-696 will be utilized in lieu of those in Appendix VIII, Supplements 2, 3 and 10.1.8.11 As required by 10 CFR 50.55a(b)(2)(xvi)(A) and 10 CFR 50.55a(b)(2)(xvi)(B), MPS3 examinations performed from one side of a ferritic vessel weld and examinations performed from 1-14 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL one side of a ferritic or stainless steel pipe will be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations.

The implementation of ASME Section XI, Appendix VIII requirements is addressed in Section 12 of this ISI Program Manual and Dominion Fleet Procedure No. ER-AA-NDE-122.1.8.12 As required by 10 CFR 50.55a(b)(2)(xviii)(A), Level I and II nondestructive examination personnel at MPS3 will be recertified on a 3-year interval in lieu of the 5-year interval specified in IWA-2314(a) and IWA-2314(b) of the 2004 Edition. The certification of visual examination personnel is addressed in Dominion Fleet Procedure No. ER-AA-NDE-123.

1.8.13 As required by 10 CFR 50.55a(b)(2)(xviii)(B), paragraph IWA-2316 of the 2004 Edition will only be used to qualify personnel that observe for leakage during system leakage and hydrostatic tests conducted in accordance with IWA-521 1(a) and (b). The certification of visual examination personnel is addressed in Dominion Fleet Procedure No. ER-AA-NDE-123.

1.8.14 As required by 10 CFR 50.55a(b)(2)(xviii)(C), when qualifying visual examination personnel for VT-3 visual examinations under paragraph IWA-2317 of the 2004 Edition, the proficiency of the training must be demonstrated by administering an initial qualification examination and administering subsequent examinations on a 3-year interval.

The certification of visual examination personnel is addressed in Dominion Fleet Procedure No. ER-AA-NDE-123.

1.8.15 As required by 10 CFR 50.55a(b)(2)(xix), MPS3 will apply the rules in IWA-2240 of Section XI, 1997 Addenda in lieu of the IWA-2240 requirements in Section XI, 2004 Edition for the substitution of alternative examination methods.1.8.16 As required by 10 CFR 50.55a(b)(2)(xx)(B), the NDE provision in IWA-4540(a)(2) of the 2002 Addenda of Section XI will be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2004 Edition of ASME Section XI. Repair Replacement activities are addressed in Dominion Fleet Repair Replacement Program Procedure No. ER-AA-RRM-100.

1.8.17 As required by 10 CFR 50.55a(b)(2)(xxi)(A), the provisions of Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items Nos. B3.120 and B3.140 of 1-15 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Inspection Program B in the 1998 Edition will be applied by MPS3. As allowed by 10 CFR 50.55a(b)(2)(xxi)(A), a visual examination with enhanced magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, 2004 Edition, with a limiting assumption on the flaw aspect ratio (i.e., a/l = 0.5), may be performed in place of an ultrasonic examination.

1.8.18 The requirements of 10 CFR 50.55a(b)(2)(xxi)(B) for Table IWB-2500-1, Examination Category B-G-2, Item B7.80, Pressure Retaining Control Rod Drive (CRD) Housing Bolting are not.applicable to MPS3. The MPS3 design has threaded connections with canopy seals rather than Item No. B7.80 CRD bolting.1.8.19 MPS3 will not implement the provision in IWA-2220, "Surface Examination" that allows the use of an ultrasonic examination method. The use of this provision is prohibited by 10 CFR 50.55a(b)(2)(xxii).

1.8.20 Prohibiting the use of IWA-4461.4.2 for eliminating mechanical processing of thermally cut surfaces as stated in 10 CFR 50.55a(b)(2)(xxiii) is not addressed in the MPS3 ISI Program. Repair Replacement activities are addressed in Dominion Fleet Repair Replacement Program Procedure ER-AA-RRM-100.

1.8.21 MPS3 will comply with 10 CFR 50.55a(b)(2)(xxiv) which prohibits the use of Appendix VIII and the supplements to Appendix VIII and Article 1-3000 in the 2002 Addenda through the 2004 Edition.1.8.22 Prohibiting the use of IWA-4340 for the mitigation of defects by modification as stated in 10 CFR 50.55a(b)(2)(xxv) is not addressed in the MPS3 ISI Program. Repair Replacement activities are addressed in Dominion Fleet Repair Replacement Program ER-AA-RRM-100.

1.8.23 Placing restrictions on the pressure testing of replaced components and appurtenances per IWA-4540(c) as stated in 10 CFR 50.55a(b)(2)(xxvi) is not addressed in the MPS3 ISI Program. Repair Replacement activities are addressed in Dominion Fleet Repair Replacement Program ER-AA-RRM-100.

1.8.24 10 CFR 50.55a(b)(2)(xxvii) modifies the requirements of IWA-5242 for insulation removal from 17-4 PH or 410 stainless steel studs or bolts aged at a temperature below 1 100'F or those having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per 1-16 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL square inch or higher. These requirements will be implemented at MPS3.1.8.25 For MPS3, the examination and testing of snubbers, including attachment hardware, is performed per Technical Specification 3/4.7.10 in lieu of the requirements for snubbers in Section XI and IWF-5300(a) and-(b). As such, there is no need to implement the optional criteria of 10 CFR 50.55a(b)(3)(v).

1.8.26 MPS3 will not implement the option in 10 CFR 50.55a(g)(4)(iii) to perform surface examinations on High Pressure Safety Injection System welds specified in Table IWB-2500-1, Examination Category B-J, Item Numbers B9.20, B9.21, and B9.22.1.8.27 The implementation schedule for ASME Section XI, Appendix VII as stated in 10 CFR 50.55a(g)(6)(ii)(C) was met during the Second Interval at MPS3.1.8.28 MPS3 will meet the criteria of 10 CFR 50.55a(g)(6)(ii)(D) for the performance of reactor vessel head inspections, except for (5) Core Exit Thermocouple (CET) Penetration Nozzles for which relief is being requested similar to second interval Relief Request IR-2-46.1.8.29 MPS3 will meet the criteria of 10 CFR 50.55a(g)(6)(ii)(E) for the performance of reactor coolant pressure boundary visual inspections.

1-17 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2. INSPECTION PLAN 2.1 Class 1 Components

2.1.1 Exemption

Basis The classification basis for Class 1 components is described in Section 1.4. The exemptions of IWB-1220 are then applied to determine the componentsI subject to the examination requirements of Table IWB-2500-1.

ASME Section XI, IWB-1220 exemptions for Class 1 components are listed below:* components that are connected to the reactor coolant system and are part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under normal plant operating conditions is within the capacity of makeup systems that are operable from on-site emergency power. The emergency core cooling systems are excluded from the calculation of makeup capacity." components and piping segments Nominal Pipe Size (NPS) 1 and smaller, except for steam generator tubing;" components and piping segments which have one inlet and one outlet, both of which are NPS 1 and smaller;" components2 and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the Outside Diameter (OD) of NPS 1 pipe" reactor vessel head connections and associated piping, NPS 2 and smaller, made inaccessible by control rod drive penetrations

• welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by guard pipe..An alternative to the ASME Section XI requirements for the Inservice Inspection of Class 1 piping, Category B-J and B-F was implemented during the second interval based on the Risk Informed technology developed in accordance with the Westinghouse Owners Group Topical Report "WCAP 14572, Revision 1-NP-A". The request to use this alternative was The exemptions from examination in IWC-1220 may be applied to those components permitted to be Class 2 in lieu of Class 1 by the regulatory authority having jurisdiction at the plant site.2 For heat exchangers, the shell side and tube side may be considered separate components.

2-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL submitted to the Nuclear Regulatory Commission on July 25, 2000 with approval received on March 12, 2002.The ISI Program does not currently address the piping examination requirements that will result from the implementation of risk-informed technology.

DNC is presently updating the risk-informed program for the examination of piping that will be applicable to Class 1 piping, as a minimum. Upon development of the risk-informed program, and required supporting documents, a supplemental submittal to the third interval 1SI Program will be provided.

This supplemental submittal will address the risk-informed scope and modifications to the examination requirements of affected piping components.

Applicable portions of the systems listed below are included in the Class 1 boundary and are depicted by Zone drawings 1 through 52.System pressure testing of Class I components is described in Section 5.0. Zone 999 is used for system pressure tests, but not shown on zone boundary drawings.2.1.2 Component/Piping Examinations The examination schedule for Class 1 components during the third inspection interval is included in the ISI Database.

Major Class 1 components and related information is provided in the following sections.A. Reactor Pressure Vessel (RPV)RPV shell examinations are conducted from the internal surfaces utilizing remote examination equipment and techniques.

Those areas of the bottom head which are obstructed from a complete Inside Diameter (ID) examination may be supplemented from the OD as practical.

The flange to upper shell weld is examined from the upper shell ID and manually from the flange seal surface. The annular area surrounding the bolt holes is examined manually from the flange mating surface. Closure studs are manually examined utilizing surface and alternative volumetric techniques, as applicable.

Alternative volumetric examination of the closure.studs is in accordance with Category B-G-1, Note 7.2-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Core support lugs and the applicable portions of the vessel ID are examined visually utilizing a remote examination inspection system.The RPV examinations are performed to meet the requirements of ASME Section XI and the intent of Regulatory Guide 1.150.B. Reactor Vessel Closure Head The RPV closure head is ultrasonically examined utilizing manual techniques from the OD surface. The RPV closure head examinations are performed to meet the requirements ASME Section XI and the intent of Regulatory Guide 1.150.The peripheral CRDM welds and extension tube welds will be ultrasonically examined utilizing manual techniques from the OD or will be examined utilizing surface examination methods from the ID.C. Steam Generators (Primary Side)Applicable steam generator welds (e.g., the tube sheet to channel head weld and nozzle to vessel welds) are ultrasonically examined utilizing manual techniques from the OD surface. Primary manway bolting is visually examined.The nozzle inner radius sections are ultrasonically examined utilizing manual techniques from the OD surface.Alternatively, as allowed by 10 CFR 50.55a(b)(2)(xxi)(A) a visual examination with enhanced magnification may be performed.

Note: Examination Category B-D, Item number B3.140 of ASME Section XI, 1998 Edition applies, as required by 10 CFR 50.55a(b)(2)(xxi)(A).

D. Pressurizer Applicable pressurizer welds are ultrasonically examined utilizing manual techniques from the OD surface. Manway bolting is visually examined.The support skirt-to-shell weld is examined utilizing surface examination methods. Relief was requested for the inside surface of the weld during the first ten-year inspection interval 2-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL due to component geometricconfiguration and Code criteria.Supplemental best effort UT was performed from the OD surface during the second interval examination as a commitment from second interval Relief Request IR-2-26 Rev. 1. Based on the new criteria in the 2004 Edition of ASME Section XI, surface examination of the inside surface of the weld is no longer required, so no third interval Relief Request is needed.The nozzle-to-shell welds are ultrasonically examined utilizing manual techniques from the OD surface.The nozzle inner radius sections are ultrasonically examined utilizing manual techniques from the OD surface.Alternatively, as allowed by 10 CFR 50.55a(b)(2)(xxi)(A) a visual examination with enhanced magnification may be performed.

Note: Examination Category B-D, Item number B3.120 of ASME Section XI, 1998 Edition applies as required by 10 CFR 50.55a(b)(2)(xxi)(A).

The six pressurizer nozzle to safe-end welds consist of Inconel 82/182 material and have received a full structural weld overlay that also encompasses their adjacent safe-end to pipe welds.The weld overlays will be examined using the examination requirements of Code Case N-770. See Relief Request IR-3-05.E. Pressure Boundary Piping An alternative to the ASME Section XI requirements for the Inservice Inspection of Class 1 piping, Category B-J and B-F was implemented during the second interval based on the Risk Informed technology developed in accordance with the Westinghouse Owners Group Topical Report "WCAP 14572, Revision 1-NP-A". The request to use this alternative was submitted to the Nuclear Regulatory Commission on July 25, 2000 with approval received on March 12, 2002.The'lSI Program does not currently address the piping examination requirements that will result from the implementation of risk-informed technology.

DNC is presently updating the risk-informed program for the examination of piping that will be applicable to Class 1 piping, as a minimum.2-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Upon development of the risk-informed program, and required supporting documents, a supplemental submittal to the third interval ISI Program will be provided.

This supplemental submittal will address the risk-informed scope and modifications to the examination requirements of affected piping components..

MPS3 will meet the criteria of 10 CFR 50.55a(g)(6)(ii)(E) for the performance of reactor coolant pressure boundary visual inspections.

F. Valve Bolting The main coolant isolation valve bolting and flange surfaces (when disassembled) are examined utilizirig volumetric and visual (VT-1) methods in accordance with Examination Category B-G- 1.Other Class 1 valve bolting is less than 2-inch diameter and is examined utilizing visual (VT-1) methods in accordance with Examination Category B-G-2.Visual examination (VT-3) of the valve internal pressure boundary surfaces will be performed in accordance with Examination Category B-M-2 upon disassembly.

G. Welded Attachments Applicable welded attachments are examined utilizing volumetric or surface examination methods, as appropriate.

H. Reactor Coolant Pumps (RCPs)The RCP studs and nuts are examined utilizing ultrasonic and visual (VT-1) examination methods, respectively.

The pump interior and flange surface require visual examination (VT-3 and VT-1, respectively) in the event the pump is disassembled.

I. Steam Generator Eddy Current Testing (ECT)The steam generator tubing surveillance requirements are contained in the MPS3 Plant Technical Specifications.

The examination requirements are based on EPRI PWR Steam Generator Examination Guidelines, Rev. 7, dated October 2007.2-5 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2.2 Class 2 Components

2.2.1 Exemption

Basis The classification basis for Class 2 components is described in Section 1.4. The exemptions of IWC-1220 are then applied to determine the components subject to the examination requirements of Table IWC-2500-1.

ASME Section XI, IWC-1221, IWC-1222 and IWC-1223 exemptions for Class 2 components are listed below: IWC-1221:

Components Within RHR, ECC, and CHR Systems or portions of systems: a) For systems, except high pressure safety injection systems in pressurized water reactor plants: 0 components and piping segments NPS 4 and smaller* components and piping segments which have one inlet and one outlet, both of which are NPS 4 and smaller 0 components and- piping segments which have multiple inlets or multiple outlets, whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 4 pipe b) For high pressure safety injection systems in pressurized water reactor plants: a components and piping segments NPS 11/2 and smaller 0 components and piping segments which have one inlet and one outlet, both of which are NPS 11/2 and smaller* components and piping segments which have multiple inlets or multiple outlets whose. cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 11/2/2 pipe c) Vessels, piping, pumps, valves, other components, and component connections of any size in statically pressurized, passive (i.e., no pumps) safety injection systems of pressurized water reactor plants.d) Piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions.

2-6 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL IWC-1222:

Components Within Systems or Portions of Systems Other Than RHR, ECC, and CHR Systems: a) For systems, except auxiliary feedwater systems in pressurized water reactor plants: 9 components and piping systems NPS 4 and smaller* components and piping segments which have one inlet and one outlet, both of which are NPS 4 and smaller e components and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 4 pipe b) For auxiliary feedwater systems in pressurized water reactor plants: %e components and piping segments NPS 11/2/ and smaller* components and piping segments which have one inlet and one outlet, both of which are NPS 11/22 and smaller 9 components and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional does not exceed the cross-sectional area defined by the OD of NPS 11/2/ pipe c) Vessels, piping, pumps, valves, other components, and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig and at a temperature equal to or less than 200'F.d) Piping and other components of any size beyond the last shutoff valve in open ended portions of systems that do not contain water during normal plant operating conditions (1) RHR, ECC, and CHR systems are the Residual Heat Removal, Emergency Core Cooling, and Containment Heat Removal Systems, respectively.

(2) For heat exchangers, the shell side and tube side may be considered separate components.

(3) Statically pressurized, passive safety injection systems of pressurized water reactor plants are typically called: (a) accumulator tank and associated system (b) safety injection tank and associated system (c) core flooding tank and associated system 2-7 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL IWC-1223:

Inaccessible Welds a) Welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by a guard pipe.Applicable portions of the systems listed below are included in the Class 2 boundary and are depicted by Zone drawings 53 through 133.System pressure testing of Class 2 components is described in Section 5.0. Zone 999 is used for system pressure tests, but not shown on zone boundary drawings.2.2.2 Component/Piping Examinations The examination schedule for Class 2 components during the third inspection interval is included in the ISI Database.

Major Class 2 components and related information are provided in the following sections.A. Steam Generators (Secondary Side)The steam generator welds and nozzle inner radius sections (FWS) are examined utilizing manual or automated ultrasonic techniques from the vessel OD of one steam generator (SG"A") as permitted by Examination Category C-A and C-B, Notes 3 and 4, respectively.

Relief request IR-3-02 addresses the component geometry limitations that preclude examination of the main steam nozzle inner radius.B. Residual Heat Removal Heat Exchangers The RHR heat exchanger welds are examined utilizing surface and ultrasonic examination methods as appropriate.

The examinations may be limited to one vessel or distributed among the vessels as permitted by Examination Category C-A and C-B, Notes 3 and 4, respectively.

The inlet and outlet nozzles were fabricated with reinforcing plate for the nozzle area and are examined utilizing surface and visual (VT-2) methods in accordance with the requirements of Code Category C-B, Item numbers C2.31 and C2.33.2-8 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL The support skirt to heat exchanger weld is examined utilizing surface examination techniques.

C. Pressure Retaining Bolting The Main Steam Isolation Valve (MSIV) bonnet bolting and Safety Injection Pump casing studs are subject to the volumetric examination requirements of Examination Category C-D.D. Pressure Retaining Welds in Piping Circumferential piping system welds are examined utilizing surface and volumetric examination methods as required by Examination Categories C-F-I and C-F-2. Code Case N-663 is being utilized by MPS3 for Class 2 piping weld examinations.

This Code Case is approved for use in Regulatory Guide 1.147, Revision 15.Ultrasonic examination of thin-walled material less than 0.375" inch thickness, is excluded from NDE requirements under table IWC-2500-1, but is part of the weld selection count as noted in Table 2.2. A 7.5 percent sample of these welds in RHR, ECCS, CHR systems has been scheduled for examination in accordance with Table 2.3, see Program Section 2.5.3.2.3 Class 3 Components

2.3.1 Exemption

Basis In accordance with IWD-1210, examination requirements of Subsection IWD only apply to pressure retaining components and their welded attachments on Class 3 systems in support of the following functions:

• Reactor shutdown 0 Emergency core cooling 0 Containment heat removal 0 Atmosphere cleanup* Reactor residual heat removal 0 Residual heat removal for spent fuel storage pool The classification basis for Class 3 components is described in Section 1.4. The exemptions of IWD-1220 are then applied to 2-9 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL determine the components subject to the examination requirements of Table IWD-2500-1.

ASME Section XI, IWD-1220 exemptions for Class 3 components are listed below: IWD-1222:

Components Exempt From Examination

• components 1 and piping segments NPS 4 and smaller" components and piping segments which have one inlet and one outlet, both of which are NPS 4 and smaller," components' and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 4 pipe* components that operate at a pressure of 275 psig or less and at a temperature of 200'F or less in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal, and emergency core cooling* welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by guard pipe.Applicable portions of the systems listed below are included in the Class 3 boundary and are depicted by Zone drawings 134 through 183.System pressure testing of Class 3 components is described in Section 5.0. Zone 999 is used for system pressure tests, but not shown on zone boundary drawings.2.3.2 Component/Piping Examinations The examination schedule for Class 3, components during the third inspection interval is included in the ISI Database.Class 3 welded attachments require visual examination (VT-1) in accordance with Table-IWD-2500-1, Examination Category D-A.For heat exchangers, the shell side and tube side may be considered separate components.

2-10 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2.4 Component Supports 2.4.1 Exemption Basis Component supports to be examined shall be the supports of those Class 1, Class 2, Class 3 and Class MC components not exempted under IWB-1220, 1WC-1220, IWD-1220 and IWE-1220.

Refer to Subsections

2.1 through

2.3 for information related to Class 1, 2, and 3 selection basis. Note that the MPS3 design does not include any Class MC component supports requiring examination.

The exemptions of IWF-1230 are applied to determine the components subject to the examination requirements of Table IWF-2500-1.

ASME Section XI, IWF-1230 exemptions for Component Supports are listed below: IWF-1230:

Supports Exempt From Examination Supports exempt from the examination requirements of IWF-2000 are those connected to piping and other items exempted from volumetric, surface, or VT-1 or VT-3 visual examination by IWB-1220, IWC-1220, IWD-1220, and IWE-1220.

In addition, portions of supports that are inaccessible by being encased in concrete, buried underground, or encapsulated by guard pipe are also exempt from the examination requirements of IWF-2000.2.4.2 Examinations The examination schedule for component supports during the third inspection interval is included in the ISI Database.Component and piping supports require visual examination (VT-3)in accordance with Table-1WF-2500-1, Examination Category F-A.2.5 Augmented Examinations The augmented examinations described in this section are summarized in Table 2.5-1.2.5.1 Reactor Coolant Pump Flywheels The reactor coolant pump flywheels are examined in accordance with Technical Specification 4.4.10 based on the Safety Evaluation Report (SER) for WCAP 14535, dated September 12, 1996. Each RCP flywheel receives an in-place ultrasonic examination, or a 2-11 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL surface examination if the flywheel is disassembled, at least once every 10 years.2.5.2 High Energy System Break Exclusion Area (BEA)Class 2 high energy piping systems within the designated BEA, as described in Final Safety Analysis Report (FSAR) Section 3.6 and FSAR Figures 3.6-8 through 3.6-17, shall receive augmented examinations in accordance with FSAR Section 3.6.2.1.2.2.F.

These augmented examinations will be performed in accordance with the risk-informed methodology established in WCAP- 14572, revision 1-N-A, Addenda 1. The welds identified for examination are documented in Engineering Record of Correspondence (ERC)25212-ER-08-0044, Rev 0.Piping greater than four inch NPS4 shall receive a surface and volumetric examination.

Piping less than or equal to four inch NPS shall receive a surface examination.

The examination volumes are defined in accordance with Examination Category C-F- 1 and C-F-2 for Class 2 piping welds and are specifically denoted in the inspection plan as requiring augmented examination.

The systems affected are: 0 Main Steam System (MSS)* Main Feedwater System (FWS)* Turbine Plant Drains (DTM)A list of weld numbers for piping subject to the augmented requirements for BEA is included in Table 2.4.2.5.3 Class 2 Excluded Systems The welds in the Chemical and Volume Control System (CHS) and High Pressure Safety Injection (SIH) for Examination Category C-F-1 were excluded from examination because of nominal wall thickness.

A 7.5% sample of each system prorated among line sizes is scheduled as an augmented examination during the third inspection interval.2-12 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2.5.4 Reactor Vessel Bottom Mounted Instrument Nozzles (58)Effective January 1, 2009, the mandatory examination requirements of Code Case. N-722 apply replacing the requirements of NRC Bulletin 2003-02.* 100% bare metal visual examination required every other refuel outage 2.5.5 Reactor Vessel Head Penetrations Effective December 29, 2008, the mandatory examination requirements of Code Case N-729-1 apply replacing the requirements of NRC Order EA-03-009.

• 100% bare metal visual examination of all penetrations required every third refuel outage or 5 calendar years, whichever is less, provided a VT-2 visual examination is performed under the insulation through multiple access points in outages where the 100% bare metal visual is not completed.(Note: Requirement is for units with EDY as defined in Code Case N-729-1, of less than 8 and with no flaws detected that were unacceptable for continued service, which apply to MPS3.).* Volumetric

/ Surface examination for all nozzles required every 8 calendar years or before Reinspection Years (RIY) as defined in Code Case N-729-1, equals 2.25, whichever is less.2.5.6 Additional examinations for Class 1 Alloy 600/82/182 Pressure Retaining Welds.Effective January 1, 2009, the mandatory examination requirements of Code Case N-722 apply. Examination requirements and frequency are identified in Table 1 of this Code Case.2.6 Description of IS1 Schedule Summaries The third ten-year ISI Schedule tables contain component examination information for MPS3. The inspection schedule tables are prepared from the computerized database utilized to schedule and track ASME Section XI and augmented components for examination.

2-13 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 2.6.1 ASME Section XI Inservice Inspection Examination Summary Tables The Class 1, 2 and 3 components and component supports requiring examination per ASME Section XI are shown in Table 2.1. The ASME Section XI Inservice Inspection Examination Summary Table 2.1 provides the following information:

• Examination Category This column lists the examination category as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1.

• Item Number This column lists the Code Item No. as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1.

Description of Components Examined This column lists a description of the components examined as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1.

Number of Components This column lists the total population of components potentially subject to examination.

The number of components actually examined during the inspection interval will be based upon the Code requirements for the subject item number." Examination Method(s)The column lists the examination method(s) required byASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, and IWF-2500-1." Request Number(s)This column provides a listing of applicable Requests for Alternatives or Relief Requests.

If a request number is 2-14 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL identified, see the corresponding Request for Alternative or Relief Request in Section 9.2.6.2 Risk-Informed Inservice Inspection Examination Summary Table The Class 1 piping welds which will require examination upon completion of the risk informed program update, are shown in Table 2.2. The Risk-Informed Inservice Inspection Examination SummaryTable

2.2 provides

the following information:

• Examination Category This column lists the examination category as identified in ASME Section XI Code Case N-577-1.* Item Number This column lists the Code Item No. as identified in ASME Section XI Code Case N-577-1.* Parts Examined This column provides a description of the elements to be examined, which are classified by their potential degradation mechanism." Number of Components This column lists the total population of components that are subject to examination and will be populated upon completion of the risk informed program update. Upon completion of the risk informed program update, the number of components actually examined during the inspection interval will be based on Westinghouse WCAP 14572, Revision 1-NP-A.* Examination Method(s)The column lists the examination method(s) required in accordance with the RI-ISI application which will be based on Westinghouse WCAP 14572, Revision 1-NP-A.*2-15 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Request Number(s)This column provides a listing of applicable Requests for Alternatives or Relief Requests.

If a request number is identified, see the corresponding Request for Alternative or Relief Request in Section 9.2.6.3 Augmented Examination Summary Table MPS3 components requiring augmented examinations are shown in Table 2.3. The Augmented Examination Summary Table 2.3 provides the following information:

• Program Section This column lists the section of this ISI Plan that addresses the subject augmented examinations" Description of Parts or Components Examined This column provides a description of the elements to be examined." Database Item No.This column provides a reference to the Item No. used for the augmented examination in the ISI Database." Examination Method(s)The column lists the examination method(s) required in accordance with the implementing augmented examination criteria.* System This column provides a reference to the system that the components are in.* Zone No. or Size This column provides a reference to the Zone No. that the components are in or the line size.2-16 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Total Item Nos.This column lists the total population of components potentially subject to examination.

The number of components actually examined during the inspection interval will be in accordance with implementing augmented examination criteria.* Comments This column provides any comments associated with the item.2.6.4 Break Exclusion Area (BEA) Weld Listing Table 2.4 lists the welds subject to examination in the break exclusion region as documented in ERC 25212-ER-08-0044 broken down by Zone No. and Drawing No.2.7 Procedures Nondestructive examination procedures are listed in Table 2.5. Other NDE procedures may be utilized provided they are reviewed and accepted in accordance with applicable station procedure.

2.8 Examination

Zone Listing The inspection plan has been divided into areas of interest identified as"zones". Each zone is defined by a drawing which identifies and locates the components requiring examination for the subject zone (except Zone 999 is not shown on drawings, but is discussed in Section 5.0).A listing of examination zones and their associated drawings is provided in Table 2.6. The examination zone boundary drawings are controlled plant drawings and are listed below: Class 1 Boundaries 25212-20997 Class 2 Boundaries 25212-20998 Sht. 1 and 2 Class 3 Boundaries 25212-20999 Sht. 1, 2, and 3 2-17 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Shell Welds B 1.11 'Circumferential 2 Volumetric B-A B 1.12 Longitudinal 9 Volumetric Prsr Head Welds Pressure Retaining B 1.21 Circumferential 3 Volumetric Welds in B 1.22 Meridional 8 Volumetric Reactor Vessel B 1.30 Shell-to-Flange Weld I Volumetric IR-3-08 B 1.40 Head-to-Flange Weld 1 Volumetric and Surface B 1.51 Repair Welds in the Beltline Region N/A Volumetric 2-18 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Pressurizer Shell to Head Welds B2.1 I- Circumferential 2 Volumetric B2.12 Longitudinal 2 Volumetric Pressurizer Head Welds B2.21 Circumferential 0 Volumetric B2.22 Meridional 0 Volumetric Steam Generator (Primary Side) Head Welds B-B B2.31 Circumferential 0 Volumetric Pressure Retaining B2.32 Meridional 0 Volumetric Welds in Vessels Other Than Reactor B2.40 Tubesheet to Head Weld 4 Volumetric Vessels Heat Exchangers (Primary Side) -Head Welds B2.51 Circumferential 0 Volumetric B2.52 Meridional 0 Volumetric Heat Exchangers (Primary Side) -Shell B2.60 Tubesheet-to-Head Welds 0 Volumetric B2.70 Longitudinal Welds 0 Volumetric B2.80 Tubesheet-to-Shell Welds 0 Volumetric 2-19 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Reactor Vessel B3.10 Nozzle-to-Vessel Welds N/A Volumetric B3.20 Nozzle Inside Radius Section (Examined with B3.90) N/A Volumetric Pressurizer B3.30 Nozzle-to-Vessel Welds N/A Volumetric B-D Full Penetration B3.40 Nozzle Inside Radius Section N/A Volumetric/

Welded VT-I Visual Nozzles in Vessels, Steam Generator (Primary Side)Inspection B3.50 Nozzle-to-Vessel Welds N/A Volumetric Program A B3.60 Nozzle Inside Radius Section N/A Volumetric/

VT-I Visual Heat Exchangers (Primary Side)B3.70 Nozzle-to-Vessel Welds N/A Volumetric B3.80 Nozzle Inside Radius Section N/A Volumetric 2-20 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Reactor Vessel B3.90 Nozzle-to-Vessel Welds 8 Volumetric B3.100 Nozzle Inside Radius Section (Examined with B3.90) 8 Volumetric.Pressurizer B3. 110 Nozzle-to-Vessel Welds 6 Volumetric B-D Full Penetration B3.120 Nozzle Inside Radius Section 6 Volumetric/

Welded VT-1 Visual Nozzles in (See Note 1)Vessels, Steam Generator (Primary Side)Inspection B3.130 Nozzle-to-Vessel Welds 8 Volumetric Program B B3.140 Nozzle Inside Radius Section 8 Volumetric/

VT-i Visual (See Note 1)Heat Exchangers (Primary Side)B3.150 Nozzle-to-Vessel Welds 0 Volumetric B3.160 Nozzle Inside Radius Section 0 Volumetric 2-21 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Reactor Vessel B5.10 NPS 4 or Larger, Nozzle-to-Safe End Butt Welds B5.20 Less than NPS 4 Nozzle-to-Safe End Butt Welds B5.30 Nozzle-to-Safe End Socket Welds Pressurizer B-F B5.40 NPS 4 or Larger, Nozzle-to-Safe End Butt Welds See Note 2 See Note 2 Pressure B5.50 Less than NPS 4 Nozzle-to-Safe End Butt Welds and Table 2.2 and Table 2.2 Retaining B5.60 Nozzle-to-Safe End Socket Welds (For All Item (For All Item Dissimilar Numbers) Numbers)Metal Welds Steam Generator In Vessel B5.70 NPS 4 or Larger, Nozzle-to-Safe End Butt Welds Nozzles B5.80 Less than NPS 4 Nozzle-to-Safe End Butt Welds B5.90 Nozzle-to-Safe End Socket Welds Heat Exchangers B5. 100 NPS 4 or Larger, Nozzle-to-Safe End Butt Welds B5.1 10 Less than NPS 4, Nozzle-to-Safe End Butt Welds B5.120 Nozzle-to-Safe End Socket Welds 2-22 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request, Category Number Components Method(s)

Number(s)Reactor Vessel B6.10 Closure Head Nuts 54 VT-I Visual B6.20 Closure Studs 54 Volumetric B6.40 Threads In Flange 54 Volumetric B6.50 Closure Washers, Bushings 54 VT-i Visual Pressurizer B-G-l B6.60 Bolts and Studs 0 Volumetric Pressure Retaining B6.70 Flange Surface, when connection disassembled 0 VT-I Visual Bolting B6.80 Nuts, Bushings, and Washers 0 VT-I Visual Greater Than 2 in. In Steam Generators Diameter B6.90 Bolts and Studs 0 Volumetric B6.100 Flange Surface, when connection disassembled 0 VT-i Visual B6.1 10 Nuts, Bushings, and Washers 0 VT-1 Visual Heat Exchangers

-B6.120 Bolts and Studs 0 Volumetric B6.130 Flange Surface, when connection disassembled 0 VT-i Visual B6.140 Nuts, Bushings, and Washers 0 VT-I Visual 2-23 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Pivint!B6.150 Bolts and Studs 0 Volumetric B6.160 Flange Surface, when connection is disassembled 0 VT-,1 Visual B-G-I B6.170 Nuts, Bushings, and Washers 0 VT-i Visual Pressure Retaining Bolting Pumps Greater Than B6.180 Bolts and Studs 4 Volumetric 2 in. In B6.190 Flange Surface, when connection is disassembled 4 VT-I Visual Diameter (cont'd) B6.200 Nuts, Bushings, and Washers 0 VT-i Visual Valves B6.210 Bolts and Studs 8 Volumetric B6.220 Flange Surface, when connection disassembled 8 VT-i Visual B6.230 Nuts, Bushings, and Washers 8 VT-I Visual 2-24 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Reactor Vessel B7.10 Bolts, Studs, and Nuts 0 VT-I Visual Pressurizer B7.20 Bolts, Studs, and Nuts 1 (16 Studs & VT-I Visual 16 Nuts)B-G-2 Steam Generator Pressure B7.30 Bolts, Studs, and Nuts 16 (128 Studs VT-i Visual Retaining and 128 Nuts Bolting, Heat Exchangers 2 in. and Less In B7.40 Bolts, Suds, and Nuts 0 VT-I Visual Diameter pipin8 B7.50 Bolts, Studs, and Nuts 8 VT-1 Visual Pumps B7.60 Bolts, Studs, and Nuts 4 VT-i Visual Valves B7.70 Bolts, Studs, and Nuts 32 VT-I Visual 2-25 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number _ Components Method(s)

Number(s)NPS 4 or Lar2er B9.11 Circumferential Welds Less Than NPS 4 See Note 2 See Note 2 B9.21 Circumferential Welds Other Than PWR High Pressure and Table 2.2 and Table 2.2 B-J Safety Injection Systems (For All Item (For All Item Pressure Retaining B9.22 Circumferential Welds of PWR High Pressure Safety Numbers) Numbers)Welds in Piping Injection Systems Branch Pipe Connection Welds B9.31 NPS 4 or Larger B9.32 Less Than NPS 4 B9.40 Socket Welds Pressure Vessels B-K B 10.10 Welded Attachments 9 Surface Welded Attachments Pipifl2 for Class 1 B 10.20 Welded Attachments 9 Surface Vessels, Piping, Pumps Pumps, and Valves B 10.30 Welded Attachments 0 Surface Valves B 10.40 Welded Attachments 0 Surface 2-26 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)B-L-1 Pumps Pump Casing Welds B12.10 Pump Casing Welds 0 VT-I Visual B-L-2 Pumps Pump 4PV-3Viua Casings B 12.20 Pump Casing 4 VT-3 Visual B-M-l Valves Valve Body Welds B 12.30 Valves, Less than NPS 4, Valve Body Welds 0. Surface B-M-2 Valves Valve Body B 12.50 Valves Exceeding NPS 4, Valve Internal Surfaces 39 VT-3 Visual B-N-i Reactor Vessel Interior of B 13.10 Vessel Interior I VT-3 Visual Reactor Vessel 2-27 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Reactor Vessel (BWR)B-N-2 B 13.20 Interior Attachments Within Beltline Region N/A VT-i Visual Welded Core B 13.30 Interior Attachments Beyond Beltline Region N/A VT-3 Visual Support Structures B 13.40 Core Support Structure N/A VT-3 Visual and Interior Attachments to RetactrmVessls tReactor Vessel (PWR)B 13.50 Interior Attachments Within Beltline Region 0 VT-I Visual B 13.60 Interior Attachments Beyond Beltline Region 7 VT-1 Visual B-N-3 Removable Reactor Vessel Core Support B 13.70 Core Support Structure 2 VT-3 Visual Structures Reactor Vessel (BWR)B-0 B 14.10 Welds in CRD Housing N/A Volumetric or Surface Pressure Retaining Reactor Vessel (PWR)Welds in Control Rod B 14.20 Welds in CRD Housing 40 Volumetric Housings Periphereral or Surface Housings B 14.21 Welds in In-Core Instrumentation Nozzle Housings >NPS 2 0 Volumetric or Surface 2-28 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)B-P All Pressure B 15.10 Pressure Retaining Components Class 1 VT-2 Visual IR-3-09 Retaining Pressure IR-3-10 Components Boundary IR-3-11 B-QB 16.10 Steam Generator Tubing in Straight Tube Design N/A Volumetric Steam Generator Tubing B 16.20 Steam Generator Tubing in U-Tube Design 4 Steam Volumetric Generators C-A C1.10 Shell Circumferential Welds 12 Volumetric Pressure Retaining Welds CI.20 Head Circumferential Welds 6 Volumetric in Pressure Vessels C 1.30 Tubesheet-to-Shell Weld 6 Volumetric 2-29 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category [ Number Components Method(s)

Number(s)Nozzles in Vessels < 1/2 in. Nominal Thickness C2.11 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Weld 0 Surface Nozzles Without Reinforcing Plate in Vessels > 1/2 in.Nominal Thickness C2.21 Nozzle-to Shell (Nozzle to Head or Nozzle to Nozzle) Weld 8 Volumetric C-B and Surface Pressure Retaining C2.22 Nozzle Inside Radius Section 8 Volumetric IR-3-02 Nozzle Welds in Vessels Nozzles With Reinforcing Plate in Vessels > 1/2 in. Nominal Thickness C2.31 Reinforcing Plate Welds to Nozzle and Vessel 8 Surface C2.32 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Welds 0 Volumetric When Inside of Vessel is Accessible C2.33 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Welds 4 VT-2 Visual When Inside of Vessel is Inaccessible 2-30 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item 1Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)C-C Pressure Vessels Welded C3.10 Welded Attachments 32 Surface Attachments Piping for Class 2 C3.20 Welded Attachments 44 Surface Vessels, PumDs Piping, Pumps, C3.30 Welded Attachments 6 Surface and Valves Valves C3.40 Welded Attachments 0 Surface Pressure Vessels C4.10 Bolts and Studs 0 Volumetric C-D Piping Pressure Retaining C4.20 Bolts and Studs 0 Volumetric Bolting Greater than 2 Pumps in. In Diameter C4.30 Bolts and Studs 2 Volumetric Valves C4.40 Bolts and Studs 4 Volumetric 2-31 MILLSTONE POWER STATION UNIT. 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Piping Welds A 3/8 in. Nominal Wall Thickness for Piping >NPS 4 C5.11 Circumferential Weld 1334 Volumetric and Surface (See Note 3)C-F-1 Piping Welds > 1/5 in. Nominal Wall Thickness Pressure Retaining for Piping NPS 2 and NPS 4 Welds in Austenitic C5.21 Circumferential Weld 216 Volumetric Stainless Steel and Surface or High Alloy (See Note 3)Piping C5.30 Socket Welds 209 Surface (See Note 3)Piping Branch Connections of Branch Piping > NPS 2 C5.41 Circumferential Weld 5 Surface (See Note 3)2-32 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)Piping Welds > 3/8 in. Nominal Wall Thickness for Piping >NPS 4 C5.51 Circumferential Weld 392 Volumetric and Surface (See Note 3)C-F-2 Pipin2 Welds> 1/5 in. Nominal Wall Thickness for Piping >Pressure NPS 2 Retaining C5.61 Circumferential Weld 0 Volumetric and Surface in Carbon or Low Alloy Steel (See Note 3)Piping C5.70 Socket Welds 0 Surface (See Note 3)Pipe Branch Connections of Branch Piping > NPS 2 C5.81 Circumferential Weld 40 Surface (See Note 3)2-33 MILLSTONE POWER STATION UNIT 3 Ih4SERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s), C-G Pumps Pressure C6.10 Pump Casing Welds 2 Surface Retaining Welds in Valves Pumps and Valves C6.20 Valve Body Welds 20 Surface C-H All Pressure C7.10 Pressure Retaining Components Class 2 VT-2 Visual IR-3-06 Retaining Pressure Components Boundary Pressure Vessels D 1.10 Welded Attachments 14 VT-1 Visual D-A Welded Attachments D1.20 Welded Attachments 124 VT-1 Visual for Class 3 Vessels, Piping, Pumps Pumps and Valves D1.30 Welded Attachments 0 VT-1 Visual Valves D 1.40 Welded Attachments 0 VT-1 Visual 2-34 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.1 MPS3 CLASS 1, 2,3 COMPONENT AND COMPONENT SUPPORT ISI EXAMINATION

SUMMARY

Examination Item Description of Components Examined Number of Examination Request Category Number Components Method(s)

Number(s)D-B All Pressure D2. 10 Pressure Retaining Components Class 3 VT-2 Visual IR-3-07 Retaining Pressure Components Boundary F1.10 Class I Piping Supports 375 VT-3 Visual IR-3-01 F-A F1.20 Class 2 Piping Supports 393 VT-3 Visual IR-3-01 Supports (See Note 4) F1.30 Class 3 Piping Supports 711 VT-3 Visual IR-3-01 F1.40 Supports Other Than Piping Supports 72 VT-3 Visual IR-3-01 (Class 1, 2, 3, and MC)Notes: 1. In accordance with 10 CFR 50.55a(b)(2)(xxi)(A), the 1998 Edition of ASME Section XI without Addenda must be applied for Examination Category B-D, Item Nos. B3.120 and B3.140. Although the 1999 Addenda eliminated Code Item Nos. B3.120 and B3.140, these Code Item Numbers were still active in the 1998 Edition, and therefore are included in the MPS3 ISI Program Plan. However, per 10 CFR 50.55a(b)(2)(xxi)(A) a visual examination with enhanced magnification that has a resolution sensitivityto detect a 1r-mil width wire or crack,.utilizing the allowable flaw criteria in Table IWB-3512-1 may be performed in place of an ultrasonic examination.

2. In 2002, MPS3 implemented a risk-informed inservice inspection program for Class 1 piping welds (i.e., Examination Categories B-F and B-J). As part of this application, the Class 1 circumferential piping welds were assigned alternate Examination Category and Code Item Numbers that are consistent with ASME Section XI Code Case N-577-1. The alternate risk-informed Examination Category and Code Item Numbers are shown in Table 2.2. The Class 1 circumferential piping welds that were previously listed in Table 2.1 have been moved to 2-35 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Table 2.2. Note that the total number of Class 1 circumferential piping welds in the ISI Program remains unchanged.

Only the designations have been changed to reflect the Examination Category and Code Item Numbers established in Code Case N-577-1.3. For the Third Interval, MPS3 is implementing Code Case N-663 for Class 2 piping welds. Code Case N-663 states that surface-examinations on these piping welds may be limited to areas identified by the Owner as susceptible to outside surface attack.4. The examination and testing of snubbers, including attachment hardware, shall be conducted in accordance with Technical Specification 3/4.7.10.

Details are provided in Relief Request No. IR-3-01 2-36 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.2 MPS3 CLASS 1 RISK-INFORMED ISI EXAMINATION

SUMMARY

Examination Item Parts Examined Number of Examination Request Category Number Components Method(s)

Number To be populated Vlmti R 1.11 Elements Subject to Thermal Fatigue following Volumetric 2 Elements Subject to High Cycle Mechanical completion of Risk R 1.12 Fatigue Informed Program Visual, VT-2 Fatigue_ _Update R1.13 Elements Subject to Erosion Cavitation Volumetric R1.14 Elements Subject to Crevice Corrosion Volumetric Cracking R 1.15 Elements Subject to Primary Water Stress Volumetric Corrosion Cracking (PWSCC)R-A Elements Subject to Intergranular or R 1.16 Transgranular Stress Corrosion Cracking Volumetric (IGSCC, TGSCC)Elements Subject to Localized Visual, R 1.17 Microbiological Corrosion VT-3 on Internal[Microbiologically-Induced Corrosion Surfaces,_ _ (MIC) or Pitting] or Volumetric R1.18 Elements Subject to Flow Accelerated Per FAC Program Corrosion (FAC)R1.19 Elements Subject to External Chloride Surface , Stress Corrosion Cracking (ECSCC)R1.20 Elements Not Subject to a Damage Volumetric Mechanism 2-37 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.3 MPS3 THIRD INTERVAL AUGMENTED EXAMINATION

SUMMARY

Program Description of Parts or Database Exam System Zone No. Total No. Comments Section Components Examined Item No. Method or Size Items J Section 2.5.1 Reactor Coolant Pump Flywheels RG 1.14 Vol. or RCS 008, 009, 5 T.S. 4.4.10 Surface 010,&011 Section 2.5.2 High Energy Break Exclusion B.E.A. Vol. & Various Various 216 Area (BEA) PipingWelds-Welds Surface> 4 NPS Section 2.5.2 High Energy Break Exclusion B.E.A. Surface Various Various 221 Area (BEA) Branch Conn. and Misc. Welds > 4 NPS and Piping Welds < 4 NPS Section 2.5.3 Circumferential Welds Excluded C5.11 Vol. & CHS 6" 66< 3/8" Nom. Thick. Surface Section 2.5.3 Circumferential Welds Excluded C5.11 Vol. & CHS 8" 62< 3/8" Nom. Thick. Surface Section 2.5.3 Circumferential Welds Excluded C5.11 Vol. & SIH 6" 49< 3/8" Nom. Thick. Surface Section 2.5.3 Circumferential Welds Excluded C5.11 Vol. & SIH 8" 47< 3/8" Nom. Thick. Surface Section 2.5.4 Reactor Vessel Bottom Mounted B 15.80 Bare Metal RPV 001 58 Instrument Nozzles Section 2.5.5 Reactor Vessel Head Penetrations B4.10 Bare Metal RPV 001 79 Code Case B4.20 Vol. & N-729-1 Surface 2-38 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.3 MPS3 THIRD INTERVAL AUGMENTED EXAMINATION

SUMMARY

(Continued)

Program Description of Parts or Database Exam System Zone No. Total No. Comments Section Components Examined Item No. Method or Size Items Section 2.5.6 Class I Alloy 600/82/182 B 15.90 Visual (VE) Various Various 19 Code Case Pressure Retaining Welds B 15.95 N-722 B15.100 B 15.120 B15.210 2-39 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL MPS3 THIRD TABLE 2.4 INTERVAL BREAK EXCLUSION AREA (BEA)WELD LISTING Zone No. 057 System: MS Drawing 25212-20959 MSS-502-FW-6 Zone No. 058 System: MS Drawing 25212-20960 MSS-503-FW-8 Zone No. 059 System: MS Drawing 25212-20961 MSS-503-FW-4 Zone No. 060 System: MS Drawing 25212-20962-MSS-501-FW-6 Zone No. 061 System MS Drawing 25212-20975 DTM-34-FW-1 MSS-28-FW-1 MSS-28-FW-4 MSS-29-FW-1 MSS-29-FW-3 MSS-33-FW-1-GM MSS-33-FW-1-MM MSS-33-FW-2 Zone No. 062 System: MS Drawina 25212-20976

..... .. f LI DTM-3 1 -FW- 1 MSS-27-FW-1 MSS-27-FW-4 MSS-30-FW-1 MSS-30-FW-3 MSS-34-FW-1-MM MSS-34-FW-2 MSS-34-FW-1-PM Zone No. 063 System MS Drawing 25212-20977 I LI DTM-28-FW-1 MSS-26-FW-2 MSS-26-FW-3 MSS-31-FW-2' MSS-31-FW-3 MSS-35-FW-1 MSS-35-FW-1-GM MSS-35-FW-1-RM Zone No. 064 System MS Drawing 25212-20978 I DTM-25-FW-1 MSS-25-FW-2 MSS-25-FW-3 MSS-32-FW-2 MSS-32-FW-3 MSS-36-FW-1-HM MSS-36-FW-1-XM MSS-36-FW-1 Zone No. 065 System FWS FWS-23-FW-1 Zone No. 066 System FWS FWS-21-FW-1 Drawing 25212-20963 Drawing 25212-20964 2-40 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.4 MPS3 THIRD INTERVAL BREAK EXCLUSION AREA (BEA)WELD LISTING (continued)

Zone No. 067 System FWS Drawing 25212-20965 FWS-22-FW-1 Zone No. 068 System FWS Drawing 25212-20966 FWS-24-FW-1 Zone No. 069 System FWS Drawing 25212-20979 FWS-11-FW-70 FWS-11-FW-54 FWS-11-FW-7-BM FWS-11-FW-74 FWS-12-FW-27 FWS-11-FW-5-CM FWS-11-FW-7-CM FWS-11-FW-5-BM Zone No. 070 System FWS Drawing 25212-20980 FWS713-FW-6-BM FWS-13-FW-97 FWS-13-FW-65 FWS- 13-FW-67 FWS-13-FW-8-BM FWS- 13-FW-8-CM FWS-13-FW-77 FWS- 14-FW-6 Zone No. 071 System FWS Drawina 25212-20981 FWS-15-FW-6-BM FWS- 15-FW-76 FWS- 15-FW-80 FWS-15-FW-64 FWS- 15-FW-7-LM FWS- 15-FW-8-BM FWS-15-FW-8-CM FWS-16-FW-30 Zone No. 072 System FWS Drawing 25212-20982 FWS-17-FW-65 FWS-17-FW-70 FWS-17-FW-7-CM FWS- 17-FW-5-BM FWS- 17-FW-92 FWS- 18-FW-35 FWS- 17-FW-5-CM FWS-17-FW-7-BM Zone No. 122 System MS Drawing 252 12-20996 DTM-394045-FW-6 DTM-394045-FW-83 DTM-394048-FW-16 DTM-394048-FW-17-1 DTM-394048-FW-21-1 DTM-394048-FW-22-1 DTM-394048-FW-25 DTM-394048-FW-26 DTM-394048-FW-4 DTM-394052-FW-1 MSS-507-FW-3 MSS-508-1-FW-84 (MSS-509-1-SW-6 DTM-394052-FW-2 DTM-394052-FW-3 DTM-394052-FW-4 MSS-509-FW-2-1 MSS-508-FW-2-1 2-41 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.5 LIST OF EXAMINATION PROCEDURES Procedure No. Application ER-AA-NDE-PT-300 Liquid Penetrant Examination ER-AA-NDE-MT-200 Magnetic Particle Examination ER-AA-NDE-VT-600 Enhanced VT-I Visual Examination of Nozzle Inside Radius Sections of Steam Generator Nozzles ER-AA-NDE-VT-601 VT- I Visual Examination MP-VE-2 VT-2 Visual Examination ER-AA-NDE-VT-602 VT-2 Visual Examination ER-AA-NDE-VT-603 VT-3 Visual Examination ER-AA-NDE-VT-607 VE Examination of Class 1 Alloy 600/82/182 Welds MP-VE-4 Remote Visual Examination of the Reactor Vessel Internals ER-AA-NDE-UT-800 Appendix VIII Qualified Equipment Tables for PDI ER-AA-NDE-UT-801 Ultrasonic Examination of Ferritic Piping Welds ER-AA-NDE-UT-802 Ultrasonic Examination of Austenitic Piping Welds ER-AA-NDE-UT-803 Ultrasonic Through Wall Sizing of Pipe welds ER-AA-NDE-UT-805 Ultrasonic Straight Beam Examination of Studs and Bolts.ER-AA-NDE-UT-808 Ultrasonic Examination of Weld Overlaid Welds ER-AA-NDE-UT-810 Ultrasonic Examination of Dissimilar Metal Welds MP-PDI-UT-4 Ultrasonic Examination of Studs from the Heater Hole MP-XT-4 Weld Marking Datum Points and Identifications MP-UT-6 Ultrasonic Examination of RCP Flywheels MP-UT-7 Ultrasonic Examination of Vessel Weld MP-UT-46 Ultrasonic Examination of Steam Generator Primary Nozzle and Feedwater Nozzle Inner Radius Sections.VPROC-NDE02-012 (Vendor) Ultrasonic Examination of Pressurizer Spray, Relief and Safety Nozzle Inner Radius Sections VPROC- ISI01-001 (Vendor) Radiographic Examination of Welds 2-42 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING Zone Class 1 Systems Drawing No. NDE Hanger No.1 Reactor Pressure Vessel 20900 20907 Sh 1 2 RPV Closure Head 20900 N/A 3 Steam Generator 1A Primary Side 20901 N/A 4 Steam Generator 1B Primary Side 20902 N/A 5 Steam Generator IC Primary Side 20903 N/A 6 Steam Generator ID Primary Side 20904 N/A 7 Pressurizer 20905 20993 Sh 7 8 Reactor Coolant Pump 1A 20906, 20914 20993 Sh 8&8A 9 Reactor Coolant Pump 1 20907, 20915 20993 Sh 9&9A 10 Reactor Coolant Pump IC 20908, 20916 20993 Sh 10&10A 11 Reactor Coolant Pump ID 20909, 20917 20993 Sh 1 1&1l1A 12 Reactor Coolant Piping Loop 1 20910 N/A 13 Reactor Coolant Piping Loop 2 20911 N/A 14 Reactor Coolant Piping Loop 3 20912 N/A 15 Reactor Coolant Piping Loop 4 20913 N/A 16 Pressurizer Surge Line 20918 20993 Sh 16 17 Pressurizer Spray Line from 20919 20993 Sh 17 Loop IA 18 Pressurizer Spray Line from 20920 20993 Sh 18 Loop lB 19 Pressurizer Spray Line Combined 20921 20993 Sh 19 20 Pressurizer Safety Line 20922 20993 Sh 20 21 Pressurizer Relief Line 20923 20993 Sh 21 22 SafetyInjection Loop A (10") 20924 20993 Sh 22 23 Safety Injection Loop B (10") 20925 20993 Sh 23 24 Safety Injection Loop B (6") 20926 20993 Sh 24 25 Safety Injection Loop C (10") 20927 20993 Sh 25 26 Safety Injection Loop D (10") 20928 20993 Sh 26 27 Residual Heat Removal & 6" HPSI 20929 20993 Sh 27 Combined (Loop A)28 Residual Heat Removal & 6" HPSI 20930 20993 Sh 28 Combined (Loop D)29 Loop Bypass Loop A (8"/2") 20931 20993 Sh 29 30 Loop Bypass Loop B (8"/2") 20932 20993 Sh 30 31 Loop Bypass Loop C (8"/2") 20933 20993 Sh 31 32 Loop Bypass Loop D (8"/2") 20934 20993 Sh 32 33 Loop Bypass Loop A (3"/2") 20935 20993 Sh 33 2-43 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING Zone Class 1 Systems Drawing No. NDE Hanger No._34 Loop Bypass Loop B (3"/2"9) 20936 20993 Sh 34 35 Loop Bypass Loop C (3"/2") 20937 20993 Sh 35 36 Loop Bypass Loop D (3"/2") 20938 20993 Sh 36 37 HPSI Loop A (1-1/2") 20939 20993 Sh 37 38 HPSI Loop B (1-1/2") 20940 20993 Sh 38 39 Int. Head Safety Injection (SIH)(6")

20941 20993 Sh 39 40 HPSI Loop C (1-1/2") 20942 20993 Sh 40 41 HPSI Loop D (1-1/2") 20943 20993 Sh 41 42 Reactor Drain Loop A 20944 20993 Sh 42 43 Reactor Drain Loop B 20945 20993 Sh 43 44 Reactor Drain Loop C 20946 20993 Sh 44 45 Reactor Drain Loop D 20947 20993 Sh 45 46 Reactor Drain Combined 20948 20993 Sh 46 47 Chemical & Volume Control 20949 20993 Sh 47___ Loop A (2" EL)48 Chemical & Volume Control 20950 20993 Sh 48 Loop A (3" CL)49 Chemical & Volume Control 20951 20993 Sh 49 Loop A (2" EL)50 Chemical & Volume Control 20952 20993 Sh 50 Loop C (2" EL)51 Chemical & Volume Control 20953 20993 Sh 51 Loop C (3" EL)52 Chemical & Volume Control 20954 20993 Sh 52 Loop D (2" EL)999 Pressure Test (See Section 5.0 for_____ description) 2-44 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems Drawing No. NDE Hanger No. [53 Steam Generator A Secondary Side 20955 20993 Sh 53 54 Steam Generator B Secondary Side 20956 20993 Sh 54 55 Steam Generator C Secondary Side 20957 20993 Sh 55 56 Steam Generator D Secondary Side 20958 20993 Sh 56 57 Main Steam Loop A 20959 20993 Sh 57 Inside Containment 58 Main Steam Loop B 20960 20993 Sh 58 Inside Containment 59 Main Steam Loop C 20961 20993 Sh 59 Inside Containment 60 Main Steam Loop D 20962 20993 Sh 60 Inside Containment 61 Main Steam Loop A 20963 20993 Sh 61 Inside Containment 62 Main Steam Loop B 20964 20993 Sh 62 Inside Containment 63 Main Steam Loop C 20965 20993 Sh 63 Inside Containment 64 Main Steam Loop D 20966 20993 Sh 64 Inside Containment 65 Main & Aux Feedwater Loop A 20963 20993 Sh 65 Inside Containment 66 Main & Aux Feedwater Loop B 20964 20993 Sh 66 Inside Containment 67 Main & Aux Feedwater Loop C 20965 20993 Sh 67 Inside Containment_

68 Main & Aux Feedwater Loop D 20966 20993 Sh 68 Inside Containment 69 Main & Aux Feedwater Loop A 20979 20993 Sh 69 Outside Containment 70 Main & Aux Feedwater Loop B 20980 20993 Sh 70 Outside Containment 71 Main & Aux Feedwater Loop C 20981 20993 Sh 71 Outside Containment 72 Main & Aux Feedwater Loop D 20982 20993 Sh 72 Outside Containment 73 Residual Heat Removal Heat 20983 20993 Sh 73 Exchanger "A" 2-45 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems Drawing No. NDE Hanger No.74 Residual Heat Removal Heat 20984 20993 Sh 74 Exchanger "B" 75 Safety Injection Tank A to Loop A 20967 20993 Sh 75 76 Safety Injection Tank B to Loop B 20968 20993 Sh 76 77 Safety Injection Tank C to Loop C 20969 20993 Sh 77 78 Safety Injection Tank D to Loop D 20970 20993 Sh 78 79 Safety Inject. Penetration 93 to 20971 20993 Sh 79 Loop A, B 80 Safety Inject. Penetration 94 to 20972 20993 Sh 80 Loop C, D 81 Safety Inject. Penetration 95 to 20973 20993 Sh'81 Loop B, D 82 Combined RHR to Loops A, D 20974 20993 Sh 82 83 RHR Pump 1A to Heat Exchanger 20860 20993 Sh 83 84 RHR Pump LB to Heat Exchanger 20861 20993 Sh 84 85 RHR Exchanger A Discharge 20862 20993 Sh 85 Header 86 RHR Train B 20863 20993 Sh 86 87 Low Pressure Safety Injection 20864 20993 Sh 87 88 Low Pressure Safety Injection 20865 20993 Sh 88 89 Low Pressure Safety Injection 20866 20993 Sh 89 90 HPSI 8" & 6" Pumps PlA to PlB 20867 20993 Sh 90 91 Low Pressure Safety Injection 20868 20993 Sh 91 92 CVCS From Charging 20869 20993 Sh 92 Pumps B & C 93 CVCS From Charging Pump A 20870 20993 Sh 93 94 HPSI to Charging Pumps 20871 20993 Sh 94 95 Quench Spray From Chemical 20872 20993 Sh 95 Refuel Tank 96 Quench Spray From Storage Tank 20873 20993 Sh 96 to Pump _97 Quench Spray From Pumps to 20874 20993 Sh-97 Penetration 98 Quench Spray Headers 20875 20993 Sh 98 Inside Containment 99 Recirculation Spray System D 20876 20993 Sh 99 100 Recirculation Spray System B 20877 20993 Sh 100 101 Recirculation Spray System C 20878 20993 Sh 101 2-46 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems 2 Drawing No. NDE Hanger No. I 1 I 102 Recirculation Spray System A 20879 120993 Sh 102 103 Containment Sump to Valve V-10 20880 N/A 104 Containment Sump to Valve V-7 20881 N/A 105 Containment Sump to Valve V-4 20882 N/A 106 Containment Sump to Valve V-1 20883 N/A 107 CMS 20997 N/A 108 HCS 20998 N/A 109 CVS 20999 N/A 110 Deleted _111 Low Pressure Safety Inject. To 20888 N/A Zone 092 112 Low Pressure Safety Inject. from 20889 20993 Sh 112 Zone 080 113 Low Pressure Safety Inject. from 20890 20993 Sh 113 Zone 082 114 RHR from Heat Exchanger "A" 20891 N/A 115 RSS to LPSI 20892 N/A 116 RSS to RHR 20893 N/A 117 RHR Penetration 092 to PIB 20894 20993 Sh 117 118 RHR Penetration 091 to PlA 20895 20993 Sh 118 119 Low Pressure Safety Inject. from 20896 N/A 3QSS-TK-1 120 Steam Generator Blowdown 20994 N/A 121 Deleted 122 Main Steam Supply to Steam-Driven 20996 N/A Auxiliary Feed Pump Turbine 123 Chemical & Volume Control 20985 N/A 124 Chemical & Volume Control 20986 N/A 125 Chemical & Volume Control 20987 N/A 126 Chemical & Volume Control 20988 N/A 127 Intermediate Safety Inject. Outside 20047 N/A Containment Penetrations 128 HPSI to Penetration No. 51 20990 N/A 129 CHS Discharge from Pumps 20991 N/A P&B & P&C 129A CHS Alt. Min-Flow Lines from P3B 20992 Sh 2 & 3 20993 Sh 185 and P3C 130 CHS Discharge from Pump P3A 20992 N/A 2-47 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems Drawing No. NDE Hanger No. a 1 _130A CHS Alt. Min-Flow Lines from P3A 20992 Sh 1 20993 Sh 184 131 CHS Discharge Cross Connect 20353 N/A 132 CHS Discharge Outside CTMT 20352 N/A 133 CHS Discharge to Penetration No. 20354 N/A 26 999 See Section 5.0 for description 134 FWA-TK1 to FQA P1A (Suction) 20993 Sh 134 135 FWA from Penetration No. 81 20993 Sh 135 136 FWA Pump 1B to Valve No. 22 20993 Sh 136 137 FWA-TK1 to Valve 995 20993 Sh 137 138 FWA-TK1 to FQA Pump 2 20993 Sh 138 (Suction)139 Valve 915 (FWA) to Zone 116 20993 Sh 139 140 FWA Pump IA to Wall Penetration 20993 Sh 140 Zone 142 141 FWA from Valve 13 to Valve 10 20993 Sh 141 142 FWA from Valve 41 to CTMT 20993 Sh 142 Penetration No. 80 143 FWA from Valve 33 to Valve 36 20993 Sh 143 144 FWA from Valve 24 to CTMT 20993 Sh 144 Penetration No. 82 145 FWA TK- Ito FWA Pump 1B 20993 Sh 145 (Suction)146 FWA Pump 2 (Discharge) to CTMT 20993 Sh 146 Penetration No. 79 147 FWA from Valve 984 to FWA-TK-1 20993 Sh 147 148 FWA from Valve 986 to FWA-TK-1 20993 Sh 148 149 FWA CTMT Penetration 79 and 80 20993 Sh 149 150 FWA CTMT Penetration No. 81 to 20993 Sh 150 CL.3 Break 151 FWA CTMT Penetration No. 82 to 20993 Sh 151 CL.2 Break 152 SFC E1A andE1B to Spent Fuel 20993 Sh 152 Pool 153 SFC Discharge Pumps PIA & P1B 20993 Sh 153 to SFC E1A &E1B 154 Spent Fuel Pool to Pumps IA & lB 20993 Sh 154 155 CCP from Zone 158 to SFC-E1C 20993 Sh 155 2-48 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems Drawing No. NDE Hanger No.156 RHS-E1B to SFC-E1B (CCP) 20993 Sh 156 157 CCP Penetration 42 to CCP PIB 20993 Sh 157 (Suction)158" CCP Heat Exchanger 1B to 20993 Sh 158 Penetration No. 40 159 Pump E1B to SFC-E1B Discharge 20993 Sh 159 (CCP)160 CCP IA Suction 20993 Sh 160 161 SFC-E1A Discharge to Valve 107 20993 Sh 161 (CCP)162 CCP Penetration to System CL. 20993 Sh 162 Break, 163 3CCP-E1A to 3CCP-E1C Discharge 20993 Sh 163 164 SFC-E1A Suction to RHS-EIA 20993 Sh 164 (Split System)165 CCP-P1A Discharge to Pump E1A 20993 Sh 165 (Suction)166 Heat Exchanger CHS-E2 (Suction) 20993 Sh 166 from V-17 167 CCP Penetration 41 to CHS-E2 20993 Sh 167 Pump 168 CCP Penetration 40 to Valve 958 20993 Sh 168 169 SWP from EGS No. E2A and E2B 20993 Sh 169 Pumps to Yard 170 SWP Section to E1A and E1B 20993 Sh 170 171 SWP from EGS No. E2A and E2B 20993 Sh 171 Pumps to Yard 172 SWP from Zone 173 to Yard 20993 Sh 172 173 3HVK Chiller IA to SWP-P2A 20993 Sh 173 (Suction)174 SWP Valve 9 Control Building Yard 20993 Sh 174 175 SWP Discharge from P2A and P2B 20993 Sh 175 to Chill IA and 1B_176 Pumps E1A and E1C Discharge to 20993 Sh 176 Yard 177 RSS-EIB and ElD to Yard 20993 Sh 177 178 Yard to RSS-E1B and ElD 20993 Sh 178 179 RSS-E1A and E1C to Valve 24 and 20993 Sh 179 Valve 304 2-49 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 2.6 EXAMINATION ZONE LISTING (continued)

Zone Class 2 Systems Drawing No. NDE Hanger No.180 Comp. Cooling HX-1A, IB, & 1C 20993 Sh 180 (Suction)181 SWP Strainers lB & ID to Yard 20993 Sh 181 182 SWP Strainers IA & IC to Yard 20993 Sh 182 183 CCP-HX IA, lB & IC to Circ. 20993 Sh 183 Water Discharge Tunnel 999 Pressure Test (See Section 5.0 for description) 2-50 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 3. CODE CASES Code Cases approved for use in revisions to Regulatory Guide 1.147 that are issued during the inspection interval may be used for the duration of the inspection interval without specific written NRC approval subject to the limitations provided in the Regulatory Guide.Code Cases that are not approved for use in revisions to Regulatory Guide 1.147 that are issued during the inspection interval shall require specific written NRC approval prior to use. Requests for approval to use Code Cases not listed in Regulatory Guide 1.147 shall be submitted to the NRC in accordance with 10 CFR 50.55a(a)(3)(i) or (ii).Code Cases shown in Table 3.1 form an integral part of this program and are identified with either the revision of Regulatory Guide 1.147 in which they were approved for use, or the NRC SER letter that provided approval to use the Code Case.As Code Cases are applied, they are referenced in the appropriate Sections of this manual.TABLE 3.1 CODE CASE

SUMMARY

TABLE Code Title Documentation

/ Notes Case N-432-1 Repair Welding Using Automatic or Code Case N-432-1 is listed as acceptable Machine Gas Tungsten-Arc Welding to the NRC in Rev. 15 of Regulatory Guide (GTAW) Temper Bead Technique 1.147.N-460 Alternative Examination Coverage for , Code Case N-460 is listed as acceptable to Class 1 and Class 2 Welds the NRC in Rev. 15 of Regulatory Guide 1.147.N-504-3 Alternative Rules for Repair of Class 1, Code Case N-504-3 is listed as 2, and 3 Austenitic Stainless Steel conditionally acceptable in Rev. 15 of Piping Regulatory Guide 1.147.N-513-2 Evaluation Criteria For Temporary Code Case N-513-2 is listed as acceptable Acceptance of Flaws in Moderate to the NRC in Rev. 15 of Regulatory Guide Energy Class 2 or 3 Piping 1.147.N-526 Alternative Requirements for Successive Code Case N-526 is. listed as acceptable to Inspections of Class 1 and 2 Vessels the NRC in Rev. 15 of Regulatory Guide 1.147.3-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 3.1 CODE CASE

SUMMARY

TABLE Code Title Documentation

/Notes Case N-532-4 Alternative Requirements to Repair and Code Case N-532-4 is listed as acceptable Replacement Documentation to the NRC in Rev. 15 of Regulatory Guide Requirements and Inservice Summary 1.147.Report Preparation and Submission as Requested by IWA-4000 and IWA-6000 N-537 Location of Ultrasonic Depth-Sizing Code Case N-537 is listed as acceptable to Flaws the NRC in Rev. 15 of Regulatory Guide 1.147.N-545 Alternative Requirements for Conduct Code Case N-545 is listed as acceptable to of Performance Demonstration the NRC in Rev. 15 of Regulatory Guide Detection Test of Reactor Vessel 1.147.N-552 Alternative Methods-Qualification For Code Case N-552 is listed as conditionally Nozzle Inside radius Section From acceptable to the NRC in Rev. 15 of Outside Surface Regulatory Guide 1.147.N-566-2 Corrective Action for Leakage Identified Code Case N-566-2 is listed as acceptable at Bolted Connections to the NRC in Rev. 15 of Regulatory Guide 1.147.N-586-1 Alternative Additional Examination Code Case N-586-1 is listed as acceptable Requirements for Class 1, 2, and 3 to the NRC in Rev. 15 of Regulatory Guide Piping, Components, and Supports 1.147.N-600 Transfer of Welder, Welding Operator, Code Case N-600 is listed as acceptable to Brazer and Brazing Operator the NRC in Rev. 15 of Regulatory Guide Qualifications Between Owners 1.147.N-613-1 Ultrasonic Examination of Penetration Code Case N-613-1 is listed as acceptable Nozzles in Vessels, Examination to the NRC in Rev. 15 of Regulatory Guide Category B-D, Item Nos. B3.10 and 1.147.B3.90, Reactor Nozzle to Vessel Welds, Figs. IWB-2500-7(a), (b), and (c).N-624 Successive Inspections Code Case N-624 is listed as acceptable to the NRC in Rev. 15 of Regulatory Guide 1.147.N-638-1 Similar and Dissimilar Metal Welding Code Case N-638-1 is listed as Using Ambient Temperature Machine conditionally acceptable to the NRC in Gas Tungsten-Arc Welding (GTAW) Rev. 15 of Regulatory Guide 1.147 Temper Bead Technique 3-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 3.1 CODE CASE

SUMMARY

TABLE Code Title Documentation

/ Notes Case N-639 Alternative Calibration Block Material Code Case N-639 is listed as conditionally acceptable to the NRC in Rev. 15 of Regulatory Guide 1.147 N-648-1 Alternative Requirements for Inner Code Case N-648-1 is listed as Radius Examination of Class 1 Reactor conditionally acceptable to the NRC in Vessel Nozzles Rev. 15 of Regulatory Guide 1.147.N-651 Ferritic and Dissimilar Metal Welding Code Case N-651 is listed as acceptable to Using SMAW Temper Bead Technique the NRC in Rev. 15 of Regulatory Guide Without Removing the Weld Bead 1.147.Crown of the First Layer N-658 Qualification Requirements for Code Case N-658 is listed as acceptable to Ultrasonic Examination of Wrought the NRC in Rev. 15 of Regulatory Guide Austenitic Piping Welds 1.147.N-661 Alternative Requirements for Wall Code Case N-661 is listed as conditionally Thickness Restoration For Class 2 and 3 acceptable to the NRC in Rev. 15 of Carbon Steel Raw Water Service Regulatory Guide 1.147.N-663 Alternative Requirements for Class 1 Code Case N-663 is listed as acceptable to and 2 Surface Examinations the NRC in Rev. 15 of Regulatory Guide 1.147. (See Note 1)N-683 Method for Determining Maximum Code Case N-683 is listed as acceptable to Allowable False Calls When Performing the NRC in Rev. 15 of Regulatory Guide Single Sided Access Performance 1.147.Demonstration in Accordance With Appendix VIII, Supplements 4 and 6 N-686 Alternative Requirements for Visual Code Case N-686 is listed as acceptable to Examinations, VT-1, VT-2, and VT-3 the NRC in Rev. 15 of Regulatory Guide 1.147.N-695 Qualification Requirements for Code Case N-695 is listed as acceptable to Dissimilar Metal Piping Welds the NRC in Rev. 15 of Regulatory Guide 1.147.N-696 Qualification Requirements for Code Case N-696 is listed as acceptable to Appendix VIII Piping Examinations the NRC in Rev. 15 of Regulatory Guide Conducted From the Inside Surface 1.147.3-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 3.1 CODE CASE

SUMMARY

TABLE Code Title Documentation

/ Notes Case N-722 Additional Examinations for PWR The use of Code Case N-722 is mandated Pressure Retaining Welds in Class 1 by 10 CFR 50.55a(g)(6)(ii)(E).

Additional Components Fabricated With Alloy conditions apply.600/82/182 Materials N-729-1 Alternative Examination Requirements The use of Code Case N-729-1 is mandated for PWR Reactor Vessel Upper Heads by 10 CFR 50.55a(g)(6)(ii)(D).

Additional With Nozzles Having Pressure-conditions apply.Retaining Partial Penetration Welds N-731 Alternative Class 1 System Leakage See Relief Request IR-3-09 Test Pressure Requirements N-770 Additional Examinations for PWR See Relief Request IR-3-05 (See Note 2)Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 Materials NOTES 1) The criteria of Code Case N-663 will be superseded by the Risk-Informed Inservice Inspection application performed on Class 1 piping. However, Code Case N-663 could be applied to Class 2 piping since it has not undergone a RI-ISI application.

Application of Code Case N-663 involved an evaluation of Class 2 piping for susceptibility to external (OD) cracking.

During the second interval, MPS3 implemented Code Case N-663 for Class 2 piping welds in Relief Request IR-2-36. For the third interval, DNC will implement Code Case N-663 for Class 2 components only.2) Although Code Case N-770 had not been published as of Supplement 6 to the 2007 Code Cases, it had been approved by the Board of Nuclear Codes and Standards and is scheduled for publication.

Per IR-3-05 DNC will implement the examination criteria of Code Case N-770 at MPS3 for the inservice inspection of the existing Pressurizer weld overlays that were originally applied during the second interval through Relief Requests IR-2-39 and IR-2-47.3-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 4. EVALUATION CRITERIA AND CALIBRATION STANDARDS 4.1 Class 1 Acceptance Standards Class 1 acceptance standards are listed in Table 4.1 4.2 Class 2 Acceptance Standards Class 2 acceptance standards are listed in Table 4.2 4.3 Class 3 Acceptance Standards Class 3 acceptance standards are listed in Table 4.3.4.4 Acceptance Standards for Component Supports (Classes 1, 2, and 3)Acceptance standards to verify component support structural integrity will be in accordance with IWF-3400.4.5 Calibration Standards Calibration blocks and standards are listed in Table 4.4. Calibration blocks for piping are located in the ISI calibration cage in the waste disposal building.

Vessel calibration blocks are stored in the warehouse and need to be requisitioned prior to the Outage.4.6 Analytical Evaluation of Flaws Flaws exceeding the size of allowable flaws defined in IWX-3500 may be evaluated in accordance with IWB-3600 to determine acceptability for continued serviPe. Flaws which meet the acceptance criteria of IWB-3600 shall be. re-examined during the next three inspection periods per IWB-2420(b). If the flaw indications remain essentially unchanged during the next three inspection periods, the component examination schedule may revert to the original schedule of successive inspections.

NOTE: Per IWF-3122.3(b), when an inservice examination of a component support reveals conditions described in IWF-3410(a), and the component support has been analyzed and/or tested to substantiate its integrity for its intended service, and has been found to be acceptable, and corrective measures to restore the support to its original condition have been performed, then successive support examinations are not required.4-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 4.7 Unanticipated Operating Events IWB-3720 requires an engineering evaluation to be performed following an operating event outside the normal operating pressure and temperature limits defined in MPS3 Plant Technical Specifications.

ASME Section XI, Appendix E, may be utilized to evaluate the event. The evaluation procedures are subject to acceptance by the regulatory authority.

TABLE 4.1 CLASS 1 ACCEPTANCE STANDARDS Examination Category Components and Parts Examined Acceptance Standard B-A Welds in Reactor vessel IWB-3510 B-B Welds in other vessels IWB-3510 B-D Nozzle welds in vessels IWB-3512 B-F Rx vessel dissimilar metal welds See Exam. Cat. R-A B-G-1 Class 1 Bolting >2'" IWB-3515, IWB-3517 B-G-2 Class 1 Bolting <2" ) IWB-3517 B-J Welds in piping See Exam. Cat. R-A B-K Welded attachments lWB-3516 B-L-1 Welds in pump casings IWB-3518 B-L-2 Pump casings IWB-3519 B-M- 1 Welds in valve bodies IWB-3518 B-M-2 Valve bodies IWB-3519 B-N-I Reactor vessel interior IWB-3520.2 B-N-2 Reactor vessel integrally welded core IWB-3520.

1, IWB-3520.2 support structures B-N-3 Reactor vessel removable core support IWB-3520.2 structures B-O Welds in CRD housings, IWB-3523 B-P Pressure Retaining Components IWB-3522 R-A Risk Informed Piping Examinations IWB-3142, IWB-3514 (will be addressed in the RI-ISI Application) 4-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 4.2 CLASS 2 ACCEPTANCE STANDARDS Examination Category Components and Parts Examined Acceptance Standard C-A Welds in pressure vessels lWC-3510 C-B Nozzle welds in vessels IWC-3511 Welded attachments for Class 2 Pressure C-C vessels, Piping, Pumps, and Valves lWC-3512 C-D Class 2 Bolting >2" in Diameter IWC-3513 Welds in austenitic stainless steel or high C-F-I alloy piping IWC-3514 (lWB-3514)

C-F-2 Welds in carbon or low alloy steel piping IWC-3514 (IWB-3514)

C-G Welds in Pumps and Valves IWC-3515 (IWB-3518)

C-H Pressure Retaining Components TWC-3516 (IWB-3522)

TABLE 4.3 CLASS 3 ACCEPTANCE STANDARDS Examination Category Components and Parts Examined Acceptance Standard Welded Attachments for Vessels, Piping, D-A Pumps, and Valves IWD-3000 (IWC-3500)

D-B All Pressure Retaining Components IWD-3000 (lWC-3500) 4-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 4.4 CALIBRATION STANDARDS Std. No. Description Heat Numbers Use UT-1 11" thick clad HT No. C4068-2 RPV -(SA-533 GR B Ci. 1)UT-2 9" thick clad B9804-3 RPV UT-3 7" thick clad B9804-3 RPV Closure Head (SA-533 GR B C1.1)UT-4 5" thick clad C4372-2 S/G (SA-533 GR B C1.1) Primary Weld UT-5 3" thick HT No. C4068-2 RCP Safe Ends UT-6 33" x 8" x 11" HT No. RPV Flange Ligament Area (SA-508 C1.2)125J596VA1 UT-7 2 V2" thick HTNo. 5160C-1 Main Coolant piping (CCSS) (SA-351 GR CF8A)UT-7A 2 /2" thick HT No. 147895-1 Main Coolant piping (CCSS) (SA-351 GR CF8A)UT-8 3 1/22" thick HT No. T 11340 Steam Generator Secondary Side Welds (SA-533 GR A, C1.2)UT-9 5" thick clad HT No. B-9804-3 RPV Shell (SA-533 GR A, C1.4)UT-10 3" thick clad HT No. BZ66 Pressurizer Shell (SA-533 GR A, C1.2)UT-12 3/4" T HT No. 894124W Vertical Residual Heat Exchangers (SA-240 TP 304)UT-15 4 2" dia. V" T SB167 CRDM Tube Weld UT-16 3" dia. 20" Long HT No. L-3269- Main Coolant Valve Studs (SA-453 Condition K3, K4 B, GR 660)(SA-453, Condition B, GR 660)(NON-PDI)UT-16A 3" dia. 20" Long Main Coolant Valve Studs (SA-453 Condition B, GR 660)(SA-453, Condition B, GR 660)UT-18 14" Sch. 160 HTNo. 20873 Press Surge Line Safe End (SA-182 GR F316).UT-19 6" Sch. 160 HT No. 78809 Press Relief and Safety Line Safe Ends (SA-182 GR F316)UT-20 4"' Sch. 160 HT No. 42018 Press Spray Line Safe End (SA-182 GR F316)'UT-21 4" Sch. 160 HT No. M9593 Class 1 Piping (SA-376 TP 316)4-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 4.4 CALIBRATION STANDARDS Std. No. Description Heat Numbers Use UT-22 6" Sch. 160 HT No. 0631-29-2 Class 1 Piping (SA-376 TP 316)UT-23 8" Sch. 160 HT No. M0035 Class 1 Piping (SA-376 TP 316)UT-24 10" Sch. 160 HT No. 1091-5 Class 1 Piping (SA-376 TP 316)2 UT-25 12" Sch. 160 HT No. 1081-44-1 Class 1 Piping (SA-376 TP 316)UT-26 12" Sch. 160 HT No. 534982 Class 1 Piping (SA-376 TP 316)UT-27 14" Sch. 160 HT No. 3-383 Class 1 Piping (SA-376 TP 316)UT-28 4.5" dia. HT No. 3P4028 RCP Studs (A-540-B24)

_ _ _(NON-PDI)

UT-28A 4.5" dia RCP Studs (A-540-B24)

UT-29 HT No. A-0159-5 RCP Flywheel (A516 GR 70)UT-31 32" 1.292" T HT No. 65817 Main Steam Piping (SA-155, C1.1, GR KC-70)UT-32 30" 1.250" T HT No. 89626 Main Steam Piping (SA-155, C1.1, GR KC-70)UT-33 6" Sch. 160 HT No. 12060 Auxiliary Feedwater Piping (SA-106, GRC)UT-34 31.5" 2.0" T HT No. 803P7323 Main Steam Piping (SA- 155, C 1.1, GR KC-70)UT-35 20" Sch. 100 HT No. N37344 Main Feedwater Piping (SA-106, GR-B)MP1-UT-35 10" Sch. 80 HT No. 39891 LPSI Outside Containment (Zone 88) (SA-240 TP 316)UT-36 18" Sch. 100 HT No. 52510 Main Feedwater Piping (SA-106, GR-B)UT-37 16" Sch. 100 HT No. N16677 Main Feedwater Piping (SA-106, GR-B)UT-38 8" Sch. 100 HT No. 26737 Main Steam Piping (SA-106, GR-B.Auxiliary Feedwater Piping)UT-40 57.700" LG HT No. 83289 RPV Stud 57.570" Long UT-41 35" LG x 2 1/2/" HT No. 220690 MSIV Stud dia. (NON-PDI)UT-41A 35" LG x 2 1/22" HT No. 220690 MSIV Stud dia.UT-LW-3 10" x 1 1/2" T -SA-240 Type 304 Thin Walled Class 2 Piping (Retired)UT-44 9" dia. X 6" HT No. -n/s Pressurizer Inner Radius 4-5 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 4.4 CALIBRATION STANDARDS Std. No. Description Heat Numbers Use UT-45 2.125" dia. x HT No.-90723 SA-193 Gr. B7 SIH pump Stud (NON-9.25". PDI) Replaced with UT-47 UT-46 8"x 12"x 14" HT No. R4980 SG Feedwater Nozzle Inner Radius UT-47 2.125" dia. x HT No.-90723 SA-193 Gr. B7 Stud. SIH Pump Casing Stud.9.25" QA-6326 ASME Carbon Steel piping welds Alternative Block QA-6327 ASME Carbon Steel piping welds Alternative Block QA-6328 ASME 304 Stainless piping welds Alternative Block QA-6329 ASME 304 Stainless piping welds Alternative Block QA-6330 ASME 316 Stainless piping welds Alternative Block QA-6331 ASME 316 Stainless'piping welds Alternative Block MP3-UT-49 8" Circ Scan HT No. Pressurizer

-Safety, Spray and Relief Nozzle Block 727018/238363 Weld Overlay MP3-UT-50 8" Circ Scan HT No. Pressurizer

-Safety, Spray and Relief Nozzle Block 727018/238363 Weld Overlay MP3-UT-51 16" Circ Scan HT No. Pressurizer

-Surge Nozzle Weld Overlay Block 727018/238363 MP3-UT-52 16" Circ Scan HT No. Pressurizer

-Surge Nozzle Weld Overlay Block 727018/238363 4-6 MILLSTONE POWER STATION UNIT 3 IN4SERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 5. SYSTEM PRESSURE TESTS 5.1 General Requirements Pressure retaining components within Class 1, Class 2 and Class 3 system boundaries are subject to system pressure tests in accordance with the following portions of ASME Section XI: Classification Applicable Code Requirements General IWA-5000 Class 1 IWB-2500-1 (Examination Category B-P) and IWB-5000.Class 2 IWC-2500-1 (Examination Categories:

C-B[Item No. C2.33] and C-H) and IWC-5000 Class 3 IWD-2500-1 Examination Category D-B and IWD-5000.Table 5.1-1, System Pressure Test Schedule, lists each system within the scope of this program along with the required test type and schedule.The Class 1, Class 2 and Class 3 boundaries shown on the applicable piping and instrumentation diagrams (P&IDs) shall be used in determining test boundaries.

When pressure testing is performed at normal operating conditions, the Operations department shall be responsible for verification that system conditions are normal. Generally, the normal operating conditions are satisfied when the system is operating in a normal alignment with system pump(s) pressurizing the system.System pressure tests have been designated as Zone 999 in the ISI database (see Section 2.7) and are not shown on Zone boundary drawings.ASME Section XI, Table IWA-5210-1, lists applicable references to determine the appropriate type of system pressure test and related test parameters for each Code Class system. Typical system pressure test parameters are: pressure (as determined by type of test), temperature, holding time, and test boundary.

System hydrostatic pressure testing, if required, will be performed on a case-by-case basis in accordance with the applicable code of record and site procedure, EN 31090, Elevated Pressure Test.5-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Buried components require a pressure loss, change in flow, or unimpaired flow test in accordance with the requirements of IWA-5244.

In some instances, it is impractical to perform pressure loss or change in flow tests on buried Class 2 and 3 piping segments bounded.by butterfly valves. Relief Request Nos. IR-3-06 and IR-3-07 propose that a verification of unimpaired flow test be performed for these piping segments.Records of the visual examination conducted during a system leakage test shall include the procedure documenting the system test condition and system pressure boundary.

Any source of leakage or evidence of structural distress shall be itemized, and the location and corrective action documented.

Ferritic steel components shall meet the test temperature requirements specified by fracture prevention criteria or determined by DNC, as applicable.

Austenitic steel components do no require test temperature restrictions.

5-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 5.1-1 SYSTEM PRESSURE TEST SCHEDULE System Applicable Test Type and Schedule P&IDs Period 1 Period 2 Period 3 BDG Steam Generator Blowdown EM-123A, EM-144A Leakage Leakage Leakage CCE Charging Pump Cooling EM-105A Leakage Leakage Leakage CCI Safety Injection Pump Cooling EM-i 14A Leakage Leakage Leakage CCP Reactor Plant Component Cooling EM-121A, EM-121B Leakage Leakage Leakage Water CHS Boric Acid EM-I104A, EM-104C Leakage Leakage Leakage CHS Volume Control EM-102A, EM-102B, EM-102D, EM-102E, Leakage Leakage Leakage EM-102E, EM-102F, EM-103A, EM-104A, EM-i 12A, EM-i 13B, EM-144B, EM- 144C EGF Emergency Diesel Fuel 2 EM- 117A Leakage Leakage Leakage FWA Auxiliary Feedwater EM-130B, EM-130C, EM-130D,,EM-131A Leakage Leakage Leakage FWA Auxiliary Feedwater (alternate EM- 120B Leakage Leakage Leakage supply from.CST)FWS Feedwater EM- 130C, EM- 130D Leakage Leakage Leakage HVK Control Building Chilled Water EM- 151 D, EM- 151E Leakage Leakage Leakage MSS Main Steam (supply to FWA EM-123A Leakage Leakage Leakage turbine)MSS Main Steam (from S/Gs to EM-I23A, EM-123B, EM-123D, EM-130C, Leakage Leakage Leakage MSIVs) EM-130D, EM-145A QSS Quench Spray 3 EM- 115A Leakage Leakage Leakage QSS Refueling Water Storage Tank EM-104A, EM- 112A, EM- 113A, EM- 13B, Leakage Leakage Leakage EM-i 15A MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR.INTERVAL TABLE 5.1-1 SYSTEM PRESSURE TEST SCHEDULE System Applicable Test Type and Schedule P&IDs Period 1 Period 2 1 Period 3 RCS Reactor Coolant EM-102A, EM-102B, EM-102C, EM-102D, Leakage 4 Leakage 4 Leakage 4 EM-102E, EM-102F, EM-103A, EM-i 12A, EM-I 12B, EM-I 13A, EM-I 13B, EM- 144B, EM-155A RHS Residual Heat Removal EM-1 12A, EM-1 12C, EM-144B Leakage Leakage Leakage RSS Containment Recirculation Spray 5 EM-i 12C Leakage Leakage Leakage SFC Spent Fuel Pool Cooling and EM-1Il A Leakage Leakage Leakage Purification SIH High Pressure Safety Injection EM-104A, EM-i 12A, EM-i 12B, EM-i 13A, Leakage Leakage Leakage EM-i 13B SIL SIL Accumulators 6 EM-1 12B, EM- 144B Leakage Leakage Leakage SWP Service Water (SWP side of RSS EM-133B Leakage Leakage Leakage heat exchangers)

SWP Service Water (accessible piping EM-133A, EM-133B, EM-133D Leakage Leakage Leakage and components)

SWP Service Water (buried piping) 8 EM-132A, EM-133A, EM-133B, EM-133D Unimpaired Unimpaired Unimpaired Flow Flow Flow Penetration 28 (DAS) EM-106C Leakage 7 Leakage 7 Leakage 7 Penetration 27 (DGS) EM-107A Leakage 7 Leakage 7 Leakage 7 Penetration 59 (SFC) EM-111A Leakage 7 Leakage 7 Leakage 7 Penetration 60 (SFC) EM-I lA Leakage 7 Leakage 7 Leakage 7 Penetration 99 (Sil) EM-113A Leakage 7 Leakage 7 Leakage 7 Penetration 15 (PGS) EM- 119A Leakage 7 Leakage 7 Leakage 7 Penetration 70 (CCP) EM- 1219A Leakage 7 Leakage 7 Leaka e 7 Penetration 38 (CDS) EM-122A Leakage 7 Leakage 7 Leakage 7 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 5.1-1 SYSTEM PRESSURE TEST SCHEDULE System Applicable Test Type and Schedule P&IDs Period 1 Period 2 Period 3 Penetration 72 (CDS) EM-122A Leakage 7 Leakage 7 Leakage 7 Penetration 45 (CDS) EM-122A Leakage 7 Leakage 7 Leakage 7 Penetration 116 (CDS) EM-122A 'Leakage 7 Leakage 7 Leakage 7 Penetration 56 (FPW) EM-146B Leakage7 Leakage7 Leakage7 Penetration 12A (SSR) EM-144B Leakage Leakage Leakage Penetration 12B (SSR) EM-144B Leakage Leakage Leakage Penetration 13A (SSR) EM-144B Leakage Leakage Leakage Penetration 13D (SSR) EM-146B Leakage Leakage Leakage Penetration 115 (SSP) EM-155A Leakage Leakage Leakage Penetration 120 (SSP) EM-155A Leakage Leakage Leakage Note 1 The purification and boron thermal regeneration sub-systems of CHS are exempt from pressure testing because they are not in the scope of Table IWD-2500-1, Examination Categories D-A, D-B, and D-C.Note 2 The emergency diesel fuel system is included in the pressure test program per Technical Specification Amendment No. 110.Note 3 The open-ended portion of QSS spray header is exempt from pressure testing per IWC-5222(b).

Note 4 Leakage test of the RCS shall be performed each refueling outage.Note 5 The open-ended portion of RSS spray header is exempt from pressure testing per lWC-5222(b).

Note 6 The nitrogen supply to SIL accumulator isolation valves is exempt from pressure testing because it is an air system which is not required to be Class 2 per Regulatory Guide 1.26.Note 7 Exempt from pressure testing per IWA-5 10(c).Note 8 Verification of unimpaired flow will be performed for buried Class 3 piping segments bounded by butterfly valves as defined in Relief Request No. IR-3-07.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 5.2 Visual Examination Requirements Personnel performing VT-2 examinations shall be certified in accordance with IWA-2300.

The VT-2 visual examination is conducted to locate the evidence of leakage from pressure retaining components, or abnormal leakage from components with or without leakage collection systems. The requirements of IWA-5240 are applicable for performance of VT-2 visual examination.

5.3 Class

1 System Pressure Test Reqiuirements Class 1 systems are subject to system leakage tests per IWB-5220.Specific leakage testing requirements are as follows:* Test Pressure shall be in accordance with IWB-5221: (a) The system leakage test shall be conducted at a pressure not less than the pressure corresponding to 100% rated reactor power.(b) The system test pressure and temperature shall be attained at a rate in accordance with the heat-up limitations specified for the system.* Test Boundaries shall be in accordance with IWB-5222: (a) The pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity.(b) The pressure retaining boundary during the system leakage test conducted at or near the end of each inspection interval shall extend to all Class 1 pressure retaining components within the system boundary.For some piping segments it is impractical to extend the Class 1 pressure testing boundaries during this system leakage test. The applicable piping segments and proposed alternatives for conducting the system pressure tests are addressed in Relief Request No. IR-3-09.Test Temperature shall be in accordance with 1WB-5240: (a) The minimum test temperature for either the system leakage or system hydrostatic test shall not be lower than the minimum temperature for the associated pressure specified in the plant Technical Specifications.(b) The system test temperature shall be modified as required by the results obtained from each set of material surveillance specimens withdrawn from the reactor vessel during the service lifetime.(c) For tests of systems or portions of systems constructed entirely of austenitic steel, test temperature limitations are not required to meet fracture prevention criteria.

In cases where the components of the 5-6 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL system are constructed of ferritic and austenitic steels that are nonisolable from each other during a system leakage or system hydrostatic test, the test temperature shall be in accordance with IWB-5230(a).Refer to Table 5.1-1 for information related to system pressure testing of Class 1 components.

5.4 Class

2 and Class 3 System Pressure Test Requirements Class 2 and Class 3 systems are subject to system leakage tests per IWC-5220 and IWD-5220, respectively.

Specific leakage testing requirements are as follows: Test Pressure shall be in accordance with IWC-5221 and 1WD-5221, respectively.(a) The system leakage test shall be conducted at the system pressure obtained while the system, or portion of the system, is in service performing its normal operating function or at the system pressure developed during a test conducted to verify system operability (e.g., to demonstrate system safety function or satisfy technical specification surveillance requirements).

Test Boundaries shall be in accordance with IWC-5222 and IWD-5222, respectively.(a) The pressure retaining boundary includes only those portions of the system required to operate or support the safety function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required.(b) Items outside the boundaries of (a), and open ended discharge piping, are excluded from the examination requirements.

• Test Temperature shall be in accordance with IWC-5240 and IWD-5221, respectively.(a) In systems containing ferritic steel components for which fracture toughness requirements were neither specified nor required in the construction of the components, the system test temperature shall be determined by DNC.(c) No limit on system test temperature is required for systems comprised of components constructed entirely of austenitic steel materials.

Refer to Table 5.1-1 for information related to system pressure testing of Class 2 and Class 3 components.

I 5-7 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 5.5 Repair and Replacement Pressure Test Requirements The MPS3 repair and replacement program is performed in accordance with Dominion Fleet Repair Replacement Program Procedure ER-AA-RRM-100. System pressure tests following repair and replacement activities are in accordance with IWA-5214 and IWA-4540.5.6 Implementing Instructions System pressure tests are performed in accordance with plant procedures U3-24-ISI-FAP02.3, ASME Section XI Pressure Test Program and EN 31090, Elevated Pressure Test, or other plant approved procedures or approved work orders.The procedure or work order shall include the following: " a drawing or description of the test boundary* test method and techniques

• .test pressure and temperature

• holding time* method for documentation The individual responsible for test performance shall ensure the following, as required: e certification of test personnel* preparation of pressurizing equipment* calibration of test equipment* preparation of components/systems

5.7 Corrective

Measures The sources of leakage detected during the performance of a system pressure test shall be located and evaluated for corrective measures as follows:* buried components with leakage losses in excess of limits acceptable for continued service shall be repaired or replaced 0 if leakage occurs at a bolted connection, the bolting shall be evaluated for joint integrity in accordance with the requirements of Code Case N-566-2 5-8 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM *THIRD TEN-YEAR INTERVAL repairs or replacements of components shall be performed in accordance With Dominion Fleet Repair Replacement Program Procedure ER-AA-RRM-100 5-9 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 6. REPAIR/REPLACEMENT ACTIVITIES Repair/replacement activities associated with pressure retaining components and their supports shall be performed in accordance with Dominion Fleet Repair Replacement Program Procedure ER-AA-RRM-100.

6.1 Establishment

of a Baseline A preservice examination following repair/replacement activities shall be performed prior to the return of the system to operation to establish a baseline and to provide data on initial conditions supplementing comparison with subsequent examinations.

The preservice examination of the item(s) used for the replacement or repair, including any applicable joints connecting the item to the system shall be made in accordance with IWX-2200.

This examination shall therefore make use of the same methods, techniques, and types of equipment as those which are planned to be used later on.Shop and field examinations performed during construction may form part of the baseline examinations where examination after final installation and testing is n6t practical, provided that: 1. Such examinations are conducted under similar conditions and with equipment and techniques equivalent to those that are planned to be employed during subsequent inservice examinations.

2. Examinations conducted before a hydrostatic (or pneumatic) pressure test is followed by a confirmatory examination after the test on a sample of inspection areas to demonstrate that no significant change has occurred.3. In the case of components classified as pressure vessels only, examinations are performed after the hydrostatic (or pneumatic) test.4. The shop and field examination records are documented and identified in a form consistent with this manual and the planned records of subsequent inservice inspections.

When improvements are made in the methods, techniques, or new equipment is used in the inservice inspection program, a new baseline shall be established.

The new baseline shall be developed for a particular component or portion of a component when the next scheduled examination of the component occurs. To the extent feasible, a correlation between the new and previous baseline shall be established using calibration data and component examination data.6-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM'THIRD TEN-YEAR INTERVAL 7. RECORDS AND REPORTS DNC will maintain adequate inspection, examination, test, flaw evaluation, and repair/replacement activity records such as radiographs, diagrams, drawings, calculations, examination and test data, description of procedures used, and evidence of personnel qualifications in accordance with IWA- 1400(k) and IWA-6000.DNC will retain all inspection, examination, test, and repair and replacement records for the service lifetime of the component or system in accordance with IWA- 1400(l).7.1 Preparation

7.1.1 Plans

and Schedules DNC will prepare plans and schedules for inservice examinations and tests to meet the requirements of IWA-6000.7.1.2 Examinations 7.1.2.1 Analytical Evaluations Analytical evaluations for acceptance of flaws found by volumetric, surface 0rvisual examinations shall be prepared in accordance with IWA-6340(e), IWB-3132.3, IWB-3142.4, IWC-3122.3, or IWC-3132.3.

7.1.2.2 Summary Reports Inservice inspection summary reports will be prepared at the completion of each inspection conducted during a refueling outage. The summary report shall contain the information required by Code Case N-532-4.7.1.3 Tests DNC will prepare records of visual examinations conducted during a system leakage test as required by IWA-5300 7-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 7.1.4 Repair/Replacement Activities Reports and Records required by Code Case N-532-4 shall be prepared by DNC under Dominion Fleet Repair Replacement Program Procedure ER-AA-RRM-100.7.1.5 Unresolved Indications Unresolved Indications will be reported in accordance with MP-24-ISI-FAPO1

7.2 Submittal

7.2.1 Plans

and Reports In accordance with IWA-1400(c), DNC will file with the enforcement and regulatory authorities plans and reports for inservice examinations and tests prepared as described in 7.1.1.7.2.2 Inservice Inspection Summary Reports*In accordance with Code Case N-532-4, at the completion of the inservice inspection conducted during each refueling outage, DNC shall file with the enforcement and regulatory authorities an inservice inspection summary report prepared as described in 7.1.2.2. This report shall be submitted within 90 days following the completion of each refuel outage 7.2.3 Analytical Evaluations In accordance with IWB-3134(b), IWB-3144(b), IWC-3125(b), and IWC-3134(b), DNC shall submit to the regulatory authority analytical evaluations of examination reports prepared as described in 7.1.2.1.7-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 7.3 Retention 7.3.1 Inspection Records and Reports DNC shall retain inspection records and reports in accordance with Code Case N-532-4, and the Dominion Quality Assurance Topical Report.7.3.2 Items Subject to Review by Authorities DNC shall retain records which are subject to review by enforcement and regulatory authorities in accordance with the following Articles: IWA-4150 IWB-3134 IWB-3144 IWC-3125 IWC-3134 Repair/Replacement plan required by IWA-4150(c).

Repair program and reexamination results for flaws found by volumetric and surface 1 examinations and accepted by repair in Class 1 components.

Repair program and reexamination results for flaws found by visual examinations and accepted by repair in Class 1 components.

Repair program and reexamination results for flaws found by volumetric and surface examinations and accepted by repair in Class 2 components.

Repair program and reexamination results for flaws found by visual examinations and accepted by repair in Class 2 components.

7-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 8. COMMITMENTS This Program addresses the requirements of the ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition. In addition to meeting those requirements, the Program also addresses examinations that are performed to specific approved Code Cases, and Augmented Examinations identified in regulatory commitments to the NRC. DNC will document those examinations performed in lieu of the code (Code Case), and those exams performed in addition to the code (NRC Commitments).

8.1 Tracking,/Logging of Correspondence A review was performed July, 1996 on the first interval ISI Program in accordance with 10 CFR 50.54. A search was performed of all correspondences that might be related to meeting regulatory commitments made regarding the ISI Program. A MPS3 ISI commitment/correspondence Tracking File of correspondence and/or regulatory commitments prior to July, 1996 was created. Correspondence after July, 1996 relevant to the ISI Program shall be filed in this Tracking File. In addition, a summary of this Correspondence shall be maintained in the MPS3 ISI commitment/correspondence Tracking Log, Table 8.1.8.2 Code Cases Section 3 of this document provides guidance in the application of Code Cases. Upon the decision to apply a specific Code Case, if NRC approval is required a copy of all correspondence leading to approval shall be placed in the MPS3 ISI commitment/correspondence Tracking File and a summary added to Table 8.1.8.3 Relief Requests Section 9 of this document provides guidance in the utilization and submittal of Relief Requests.Upon submittal of a Relief Request to the NRC, a copy of the correspondence package shall be placed in the MPS3 ISI commitment/correspondence Tracking File and a summary added to Table 8.1.Once a response is received from the NRC, a copy of the correspondence shall be placed into the MPS3 ISI commitment/correspondence Tracking File and a summary added to Table 8.1.8-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 8.4 Commitment

/ Correspondence Tracking Log Augmented Examinations that are performed outside of the scope of Section XI requirements shall be referred to as Augmented Examinations.

These Augmented Examinations might come at the direction of the regulatory authority, or might be a result of developing industry issues.All correspondence leading to the addition of Augmented Examinations shall be placed in the MPS3 ISI commitment/correspondence Tracking File and a summary added to Table 8.11.All correspondence with the NRC leading to the commitment of performing Augmented Examinations shall be placed in the MPS3 ISI Commitment

/Correspondence Tracking File and a summary added to Table 8.1.8-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (Sheet I of 16)File No. Date Orig. Commitment Compliance Verification DJ720001 04/26/82 NU Augmented ISI on aux feed pump These welds are included in the current steam line normally pressurized interval augmented ISI Program.between cont. and first down stream normally closed valve DB54000 05/03/83 NRC Confirm 100% UT of all welds on The ISI Program includes all BEA high energy lines between outboard systems in FSAR S3.6, technical isolation valve and last (outboard) review team will verify boundaries.

restraint.

DD41000 08/01/83 NU NU confirmed conformance to Reg. RPV ISI verified to meet 1.150 for the Guide 1.150 rev. 1 for RPV ISI first inspection interval.inspections.

DD41000 08/01/83 NU NU states that only UT responses Superseded by subsequent greater than 50 percent DAC will be commitment.

recorded for data consistency.

DD41000 08/01/83 NU NU confirms that high energy fluid Covered in other commitments.

system will receive Augmented ISI per requirements of SRP 6.6 par 11.7 DD41000 08/01/83 NU NU states for MSS and FWA, the ISI Program for MS & FWA does not Augmented ISI extends from go up to first whip restraint, only penetration inboard to first rigid inboard penetration weld is in restraint.

augmented,;URI to resolve.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 2 of 16)File No. Date Orig. Commitment Compliance Verification CQ93000 03/01/84 NU All pipe welds in BEA as per FSAR The ISI Program included a MT for S3.6 will receive Augmented Exam, welds < 4" and UT/MT > 4".UT/MT on pipe >4", not on pipe <4".CS67000 05/01/84 NU Added compensatory measures to UT Superseded by subsequent proc. to include higher sensitivity of commitment.

3/4 t SDH, record > 20% DAC CU37000 06/18/84 NU Revise UT proc for CS & SS to say NU-UT-2, and NU-LW-1 require"any indication of suspected flaw recording of any suspected flaw regardless of amplitude will be regardless of amplitude.

recorded." CF90001 03/07/85 NU The ISI Program for cast SS welds ISI Coordinator tries to select should select welds with best STATIC/CSS joints and uses 0 degree acoustical properties.

UT to avoid noisy CCSS CG90000 05/07/85 NU The ISI Program for cast SS welds ISI Coordinator selected sample based will select welds with best access for upon accessibility; and all CCSS welds UT of weld and volume, are ground flush CL6 10001 07/25/85 NRC- NU commitment to perform CCSS NU sponsored a round robin piping UT using the best available demonstration, technical paper and technology, continues evaluation.

BG71000 03/21/88 NU Four (4) rigid feedwater sys piping All piping supports on the FWS in supports will be visually inspected cont. were inspected by 1991 as per ISI during next outage of sufficient Program.duration.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 3 of 16)File No. Date Orig. Commitment Compliance Verification BG41000 08/05/88 NU RR No. IR-6 only inspect one side of Exam done 1989, MT performed on the pressurizer support skirt to shell outside of skirt, see work order No.weld (surface exam) M3-89-02275 BA62000 05/04/89 NU Visual exam of RC pump casing 1987 done, for pump A reference data pressure boundary surfaces will be sheet No. 3-87-vt-216 performed if pump is disassembled, see IR-4 BA62000 05/04/89 NU Visual exam of bolting and flange Commitment was to VT any pump surface if a RCP is disassembled (IR- flange when disassembled, UIR 808 4) tracked. Program updated with change No. 15 to show completion.

A07335 07/25/89 NU Perform additional weld inspection if LLRT Procedure SP3612B.4-41 was (RCR- 19260) leakage of sufficient rate to induce revised to include statement to contact thermal stresses is detected in valves ISI Coordinator to perform additional SIH*MV8801A/B weld exams if leakage exceeds limits.Subject welds were added to the ISI Program as Augmented

-EEN88-08 exams.A1350001 02/08/91 NRC RR14 residual heat exchanger SH/FL Zone 074 of the ISI Program shows a UT accessible portion and PT UT and PT performed in 1989 inaccessible portion MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 4 of 16)File No. Date Orig. Commitment Compliance Verification A09339 02/08/91 NRC NRC approved relief requests IR-2,IR- Verified relief requests incorporated 3, IR-4, IR-7,IR-8,IR-9, IR-1O,IR-11, into ISI Program IR-13, IR-14 AK21000 04/30/91 NRC ABB will include questionable and ABB/AMDATA confirmed that marginal indications in training and marginal surface exam test samples are test samples. included in training.AK21000 04/30/91 NRC CES modify monitoring procedure to CES monitoring procedure has been increase field observations modified many times to improve training and monitoring.

B 13932 10/30/91 NU Upgrade ISI Program for Class 2 Confirmed both program and database components to ASME Section XI 83 are to the 83 Winter 85 Addenda Code Edition through Winter 85 Addenda for the first inspection interval.(NRC, granted A10880, 3/3/93)AP41000 10/31/91 NU Perform UT of RPV BF nozzle/SE SWRI used a SLIC 40 transducer for weld with 45R1 on lower 2/3 and the exam, UIR No. 806 tracked 65R1 on upper 1/3 completion.

Revised IR-9 (Rev. 2) to reflect. .AD42000 06/01/92 NU IR-1, rev 2, NU investigate tooling Welds 102-151,101-154A,B,C&D advances and determine, if have code coverage less than 90%.supplemental manual UT can be Relief request IR-26 accepted by NRC.performed or a request for relief of inaccessible areas will be sent.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 5 of 16)File No. Date Orig. Commitment Compliance Verification AD42000 06/01/92 NU IR-9 NU will demo that OD cracks G. Miemiec/SLS witnessed demo with can be detected on safe end weld in NRC in attendance 3/95 lieu of pt.AD42000 06/01/92 NU IR-1 rev3 UT of accessible areas, in Leak test done after every outage. 10-service hydro test and inservice year in accordance with Code Case N-leakage test 498-1.GB87000 02/11/93 NRC Upgraded computer sys (ISI) will Currently on-line installed week of 01/17/93 GB 87000 02/11/93 NRC New Computer program will have UIR No. 857 tracked completion.

option to review data sheets First implementation scheduled for RFO-7.A 10880 03/03/93 NRC NRC approved relief requests IR- 1, Verified relief requests incorporated IR-6,IR-9, IR-12, IR-18 into ISI Program A10880 03/03/93 NRC NRC approved use of N-323, N-436- ISI coordinator verified these Code 1, N-437,N-460, N-461 Cases were incorporated into ISI program.JC380001 05/26/93 NU NU will inspect RPV studs and MSIV Superseded by a subsequent studs to ASME code case N-307-1 commitment/correspondence JC380001 05/26/93 NU Will perform 7.5% sample UT of 7.5% in program. Exams completed in chemical and volume control and high RFO6 pressure safety injection systems A11306 11/05/93 NRC Approved use of N-491 for the 1st Included in ISI Program Interval.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 6 of 16)File No. Date Orig. Commitment Compliance Verification JG020001 11/05/93 NU Approves N-307-1 for RPV bolts, RPV bolting exam was done in 1995 at disapproves it for MSIV bolts West in accordance with N-307-1 JG730001 02/04/94 NU 16 thin wall SS weld did not get full See JV910001, duplicate UT coverage relief request to follow JG730001 02/04/94 NU 4 integral attachment welds id not See JV910001, duplicate receive UT coverage relief request to follow JJ990001 06/17/94 NU NU commits to performing UT of MSIV studs done to Appendix VI, SIH class 2 bolts using appendix VI 1983 pump studs Completed in 1999 (RFO6)in lieu of N-307-1 JJ990001 06/17/94 NU NU will update ISI program plan to Included under section 5.7 of MP3 ISI incorporate code case N498 "alternate Program, First Interval.

Included in 10 hydro pressure test rules for Class 1 Section 5.0 of the MP3 ISI Program, and 2 systems. Second Interval.A11760 08/03/94 NRC Approved use of N-498 for the 1st This Code Case is now superseded in Interval.

program by N-498-1 JO330001 12/16/94 NU NU requests relief IR-19 from S/G Completed as part of the system leak steam nozzle IR & proposes to do tests.visual exam during system leak test as per N-498 JO510001 12/23/94 NU Tech Spec change (Amendment 110) Test performed during 1995 outage will include augmented inspection of under AWO M3-85-03733 for the first DG fuel oil system pressure test as per inspection interval.

Scheduled for each 1983 table IWD 2500-1 period of the second interval.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 7 of 16)File No. Date Orig. Commitment Compliance Verification A12085 01/18/95 NRC NRC approved use of N-416-1 and N- Included in ISI Program 498-1 for the 1st Interval.A12289 05/04/95 NRC NRC approved relief request IR-19 Verified relief request incorporated into ISI Program JV91001 06/30/95 NRC NU commits to perform a top to Performed by ISI Coordinator in 1995 bottom review of first 10 year interval and ESAR report MP3-PES-97-0015.

JV910001 06/30/95 NRC NRC determined that the 16 SS pipe B-K-I resolved by IR-20 (01/12/96)

UTs and 4 integral attachments were DRM changed ISI Program for SS counted lbut relief was never requested welds JV9 10001 06/30/95 NRC ISI Engineer will include all Added to program with change 5 to correspondence to/from NRC in include Correspondence/

Commitment Program Manual log.JV9 10001 06/30/95 NRC Within 6 months after outage, ISI CEN 101B was upgraded to include ENG will perform and document this requirement in NOV. 95 ASME code verifications JV910001 06/30/95 NRC ISI Engineer will issue changes to the CEN 101B was upgraded to include ISI program within six months after this requirement in NOV. 95.outage Additionally added to SP31129.JV9 10001 06/30/95 NRC NU will notify NRC if essential 100% Issued Relief Request IR-21 and IR-22 for RPV UT coverage was not and IR-27.examined JV910001 06/30/95 NRC NRC note that large scale audits of the UIR No. 805 tracked completion.

program maybe could have picked up these discrepancies

-I MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 8 of 16)File No. Date Orig. Commitment Compliance Verification A 12790 04/05/96 NRC Approved use of Code Case N-535 for Incorporated in ISI Program Manual the ISt Interval A12893 6/28/96 NRC NRC approves Reliefs IR-20, IR-21, Incorporated in ISI Program Manual and IR-22.A13004 09/13/96 NRC Approved use of Code Case N-546 for Incorporated in ISI Program Manual the I" Interval (Change 18)B 16368 04/24/97 NNECO 1) ISI Program will be revised to Incorporated in ISI Program Manual include guidance on ASME (Change 12)Section XI requirements for additional and successive exams.2) ISI documents will be revised to include guidance on requirements for requesting relief IAW 10CFR50.55a A1 3313 5/30/97 NRC Repair/Replacements performed after Incorporated in ISI Program Manual 9/9/96, on Containment to be (Change 13)performed in accordance with requirements of IWE and 1WL of 1992 Edition and Addenda of Section XI.A13670 2/17/98 NRC NRC approves Relief Request IR-26 Incorporated in ISI Program Manual B 17409 8/19/98 NNECO NNECO withdraws Relief Requests ISI Program updated to reflect.IR-23, IR-24, AND IR-25.A13884 8/20/98 NRC NRC approves Code Case N-389-1 Incorporated in ISI Program Manual.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 9 of 16)File No. Date Orig. Commitment Compliance Verification B17355 7/14/98 NNECO NNECO informed NRC of ISI Interval First Interval Closed Oct. 23, 1999.extension.

B17871 9/23/98 NNECO Request to use Code Case N-532 (JR- ISI Manual updated.2-10)A 14053 2/22/99 NRC NRC approves Relief Request IR-27 Incorporated in ISI Program Manual 3/03/99 NRC NRC approves T.S. amendment 167 Incorporated into Tech. Specs.for T.S. 4.7.10.B 17752 4/7/99 NNECO Submitted 2 nd Interval ISI Program In NRC review.A14068 4/16/99 NRC NRC approves T.S Amendment 169 Incorporated into Tech. Specs. And ISI for T.S. 4.4.10. Program manual updated requirements.

B 17598 4/22/99 NNECO Request Relief to use the 1998 Edition In NRC Review.of ASME Section XI for IWE/IWL.(RR-E1)(RR-L1)

A14092 5/11/99 NRC NRC approves Relief Request Incorporated in ISI Program Manual IR-9 (Rev.2)B 17867 .9/17/99 NNECO Submittal of Relief Request IR-28 Approved B13883 9/27/99 NNECO Submitted RFO6 Outage Summary N/A__Report.A15133 11/3/99 NRC NRC Approves Relief Request IR-28 Incorporated in ISI Program Manual B 17927 12/13/99 NNECO Amended Relief Request for Retracted 1WE/1WL to include the 1999 Add.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 10 of 16)File No. Date Orig. Commitment Compliance Verification B17885 2/14/00 NNECO Request to use Code Case N-623 (IR- Approved 2-12)B 17985 2/11/00 NNECO Retraction of the Request to use the N/A 1999 Addenda for IWE/IWL.B 18005 2/25/00 NNECO Clarify that the term "certified" is Containment Inservice Inspection being replaced with "Trained and Manual updated.Qualified" for the IWE/1WL Program.A15255 4/21/00 NRC NRC approval granted to use 1998 Containment Inservice Inspection Edition of Section XI for IWE/IWL..

Manual has been updated to reflect the__requirements of the 1998 Ed.B 18098 5/31/00 NNECO Submittal of Relief Request IR-2-13 Approved and IR-2-14, B 18135 5/31/00 NNECO Submittal of Relief Request for Approved temporary Non-Code repair of a Class 3 Service Water leak.B 18146 6/21/00 NNECO Revised Request to use Code Case N- Approved 566 to N-566-1 (IR-2-06 Rev.1)B 18105 6/28/00 NNECO Inform NRC of the Intent to suspend Approved examination of Class 1 (B-J, B-F)piping examinations and begin RI-ISI exams in the Second Period of the Second Interval.A15322 7/24/00 NRC NRC SEVAL approval for the 2nd ten Incorporated into the ISI Program year inspection interval.

Manual MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 11 of 16)File No. Date Orig. Commitment Compliance Verification B18104 7/25/00 NNECO Submittal of the Class 1, Risk Approved.Informed ISI Program.A15335 8/24/00 NRC NRC approval granted to use Code Incorporated.

Case N-532 (IR-2-10)B 18202 8/25/00 NNECO Submittal of Relief Requests IR-2-15, Revised and resubmitted under B 18253 16, 17, and 18 for Appendix VIII dated 11/08/00 implementation.

A 15337 8/28/00 NRC NRC Approval of Relief Request Incorporated.

IR-2-12 (Code Case N-623)A15356 10/04/00 NRC NRC Approval of Relief Request Incorporated.

IR-2-13, IR-2-14 (Limited exam coverage).

B 18253 11/08/00 NNECO Submittal of Revision to Relief Approved Request IR-2-15, 2-16, 2-18 (Appendix VIII Implementation).

A15383 11/14/00 NRC NRC approval of Relief Request Incorporated.

RR-E2 for Containment ISI Program.B 18269 11/16/00 NNECO Response to NRC RAI concerning N/A intent to suspend B-J & B-F exams under original submittal B 18105.A15384 11/27/00 NRC Approval for Relief Request N/A associated with temporary SWP Non-Code repair MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 12 of 16)File No. Date Orig. Commitment Compliance Verification A 15423 1/26/01 NRC SER for Appendix VII Relief Incorporated Requests IR-2-15, 2-17, and 2-18.A 15429 2/02/01 NRC Approval to suspend examinations of AR 00010956-02 to track commitment Class 1 (B-J, B-F) welds for the 1st to complete a min. of 50% of the Class period. 1 Risk Informed welds by the end of the 2 nd period.B 18470 9/26/01 Dom Response to request for additional N/A information for the Risk-Informed ISI program.B 18490 9/28/01 Dom Submittal of Relief Requests IR-2-2 1, Approved 22, 23, 24, 25, 26 A15653 3/12/02 NRC NRC SER approval for Class 1 Risk Incorporated into the ISI Program Informed ISI Program B 18621 4/02/02 Dom Millstone Power Station Response to N/A NRC Bulletin 2002-01 (Rx vessel head degradation).

B 18657 5/30/02 Dom Submittal of revised relief Request Approved IR-2-25 rev 1, and IR-2-26 rev. 1 as a response for request for additional information via NRC telecon.A15742 8/20/02 NRC NRC SER approval for Relief Incorporated into the ISI program Requests IR-2-21, 22, 23, 24, 25 rev. 1, and 26 rev. 1.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 13 of 16)File No. Date Orig. Commitment Compliance Verification B 18790 10/18/02 Dom Submittal of the first period inservice N/A inspection summary report (OAR-1).B 18796 11/26/02 Dom Submittal of Relief Request IR-2-27, Approved for the one time use of a non-ASME certificate holder for fabrication of FW Sys. Components B 19000 11/17/03 Dom Ninety day response to NRC Bulletin Completed (RFO9) inspection under 2003-02 AWO M3-03-06218 B 19021 03/30/04 Dom Request to use Code Case N-663 Approved S/N 04-315 (5/04/04)S/N 04-315 05/13/04 NRC NRC approval to use Code Case Incorporated into ISI Program N-663 S/N 04-027 06/12/04 NRC NRC approval of Relief Request Incorporated into ISI Program IR-2-27 nd S/N 04-447 08/04/04 Dom Submittal of 2 period (RF08 & N/A RF09) Owners Activity Report (OAR-i).S/N 04-535- 09/23/04 Dom Request to use temporary Non-Code Approved S/N 05-066 (9/22/05)repair in Service Water Sys. Brazed joints to 3CCI*E1A S/N 04-717 11/17/04 Dom -NRC review / closeout of Millstone 90 N/A day response to NRC Bulletin 2003-02 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 14 of 16)File No. Date Orig. Commitment Compliance Verification S/N 05-100 -5/9/05 Dom Request to use Non-Code temporary Withdrawn repair of SWP brazed joint. Relief, Request IR-2-38 S/N 05-343 6/20/05 Dom Request to use later Edition and Addenda of ASME Section XI for repair and Replacement Program.S/N 05-708A 10/19/05 Dom Request to use weldoverlay on Approved S/N 06-064 (1/23/06)Pressurizer weld. Relief Request IR-2-39.S/N 05-789 12/15/05 Dom NRC Bulletin 2004-01. Inspection of N/A Alloy 600 PZR penetrations and steam space piping connections, 60 day response.S/N 06-031 1/24/06 Dom Submittal of 3R10 OAR-1 Summary N/A Report S/N 06226 5/11/06 Dom Request to use alternative UT sizing Approved S/N 07-366 (5/1/07 techniques.

Relief Request IR-2-42 S/N 06226 5/11/06 Dom Request to use Code Case N-696. Approved S/N 07-369 (5/2/07 Relief Request IR-2-43 S/N 06-226 5/11/06 Dom Request to use PDI qualified Approved S/N 07-368 (5/1/07)procedures, equipment, and personnel for Non-PDI RPV Flange to shell weld. Relief Request IR-2-44 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 15 of 16)File No. Date Orig. Commitment Compliance Verification S/N 07-340 5/3/07 Dom Inservice Inspection and Testing 10- N/A Year Interval changes.S/N 07-0207 7/2/07 Dom Proposed Tech. Spec. change for Approved S/N 08-0498 (7/31/08)ISI/IST deleting reference to ISI and change of reference code from ASME Section XI to OM for IST.S/N 07-397 7/13/07 Dom Results of Reactor Vessel Head N/A inspection required by NRC order EA-03-009.S/N 07-0494 7/13/07 Dom Inspection and mitigation of Alloy N/A 82/182 PZR butt welds and additional information on completion of weld overlays.S/N 07-0530 8/03/07 Dom OAR- 1 3R 11 Outage Summary Report N/A S/N 0338 ' 11/09/07 Dom Alternative request for pressure testing Approved S/N 08-0437 (7/10/08)of Class 2 Buried Piping S/N 0338 11/09/07 Dom Alternative request for pressure testing Approved S/N 08-0437 (7/10/08)of Class 3 Buried Piping S/N 07-0416 2/25/08 Dom Alternative request to use Section 1l11 Approved S/N 08-0111 (2/25/08)Code Cases N-756 and N-757.

MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 8.1 MPS3 ISI COMMITMENT

/ CORRESPONDENCE TRACKING LOG (continued)(Sheet 16 of 16)File No. Date Orig. Commitment Compliance Verification S/N 08-0498 7/31/08 NRC NRC Approval for Tech Spec change N/A to remove Inservice Inspection from Tech Specs RA-08-026 10/27/08 Dom MPR-139 MRP Deviation Accepted in MRP meeting minutes Notification dated December 2, 2008 S/N 08-0523 10/27/08 Dom NRC Notification of MPR-139 N/A deviation.

S/N 09-050 2/13/09 Dom OAR-I 3R12 Outage Summary Report N/A MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 9. RELIEF REQUESTS 9.1 Alternative Requirements 9.1.1 Code Cases Not Approved for Use ASME publishes Code Cases which explain the intent of Code rules or provide for alternative requirements under special circumstances.

Section 3 of this document describes the use of approved Code Cases. If it is determined that it would be beneficial to utilize Code Cases which provide alternative requirements to the inspection requirements of ASME Section XI as documented in this program manual which have not been approved for use in Regulatory Guide 1.147, relief requests will be filed in accordance with 10 CFR 50.55a(a)(3).

9.1.2 Hardship

or Unusual Difficulty

[Commitment Table 8.1, File B 16368]During Inservice Inspection, there are cases where compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. An example of this is encountering significant radiation exposure during the performance of an examination.

If such circumstances are identified, relief requests will be filed in accordance with 10 CFR 50.55a(a)(3)

9.2 Impractical

Requirements During Inservice Inspection, there are cases where component configuration and/or interferences prohibit coverage of the code required volume or surfaces.

In each case where .such limitations have been encountered the details are documented in a relief request. If additional conditions are encountered where the inspection requirements of ASME Section XI as documented in this program manual cannot be met, relief requests will be filed in accordance with 10 CFR 50.55a(g)(5)(iii).

9-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 9.3 Format Relief requests are numbered sequentially, with relief requests issued during the third interval indicated by the number "3", i.e., IR-3-XX. Each relief request will be formatted in accordance with the Nuclear Energy Institute (NEI) Standard Format for Requests Pursuant to 10 CFR 50.55, Revision 1, dated June 7, 2004.9.4 Relief Request Summary Tables 9.4.1 Relief Requests which were submitted during the First Interval are listed in Table 9.4-1 and are included as historical information.

The Table includes an evaluation of the applicability of a similar Relief Request to the Second Interval Program.9.4.2 Relief Requests which were submitted during the Second Interval ISI Program are listed in Table 9.4-2 and are included as historical information.

9.4.3 Relief

Requests applicable to the Third Interval ISI Program are listed in Table 9.4-3. Copies of Relief Requests are in Attachment

2. Additional relief requests, as deemed necessary, will be submitted pursuant to the requirements of 10 CFR 50.55a(a)(3) or 10 CFR 50.55a(g)(5)(iii).

9-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (Sheet 1 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-1 (Rev. 3) Pressure B-A B13.12 Reactor Vessel Pressure Retaining Volumetric Volumetric Exam of Granted 03/03/93 Retaining Welds Shell Longitudinal Welds in the Exam of 100% accessible portions in Reactor Beltline Region: of length of all of welds Vessel Welds 6, 7, & 8 welds B-A B1.21 Reactor Vessel Pressure Retaining Volumetric Bottom Head Circumferential Weld Exam of (lower shell-to-bottom head accessible 101-141) length of all welds Reference Document B13932 (10/30/91)(s)

NRC Letter A10880 (3/3/93) includes Safety Evaluation of 1st 10 Year ISI Program Summary Report, Rev. 3 (G)Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.IR-2 (Rev. 2) Pressure B-A B1.22 Reactor Vessel Pressure Retaining Volumetric Volumetric Exam of Relief not 02/08/91 Retaining Welds Closure Head Meridional Weld Exam of accessible portions required in Reactor No. 101-104D accessible of weld Vessel Closure length of all Head welds B-A B1.40 Reactor Vessel Head-to-flange Weld Volumetric Volumetric and Granted Head Flange No. 101-101 and surface surface exam of exam accessible portions of weld Reference Document NU Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Evaluation of 1st 10 Yr Prog. & TER (SAIC Report)(G)

Note: IR-2 Rev. 2 resubmitted in B13932 (10/30/91)

NRC Response A10880 (3/3/93) did not grant further relief Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test The Meridional Head weld would not require relief based on Code Case N-460. Relief request should be submitted following performance of examinations during the second Ten-Year interval r 9-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 2 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed *Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-3 Pressure B-B B2.11 Pressurizer Pressure Retaining Volumetric Volumetric Exam of Granted 02/08/91 Retaining Welds Circumferential shell-to-head accessible portions in Vessels Other welds of weld Than Reactor 03-007-SW-J Vessels 03-007-SW-F Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Evaluation of 1st 10 yr. Program & TER.(SAIC Report)(G)

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test The Shell to Upper Head weld would not require relief based on Code Case N-460. Relief request should be submitted following performance of examinations during the second Ten-Year interval IR-4 Internal B-L-2 B12.20 Pump Casings Internal surfaces of pumps Visual, VT-3 Class 1 pumps will Granted 02/08/91 Surfaces of receive a visual Pump Casings exam (VT-3) when and Valve they are Bodies disassembled B-M-2 B12.50 Valve Bodies Internal surfaces of valve Visual, VT-3 Class 1 valves will bodies receive a visual examination (VT-3)when they are disassembled Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Evaluation of 1st 10 yr. Program & TER (SAIC Report)(G)

Second Interval Applicability A similar relief request will not be required for the second Ten-Year ISI interval.

The 1989 Code has addressed this concern in the footnotes of Table IWB-2500-1, Examination Category B-L-2 and B-M-2 by adding the following: "Examination is required only when a pump or valve is disassembled for maintenance, repair, or volumetric examination.

Examination of the internal pressure boundary shall be performed to the extent practicable.

Examination is required only once during the inspection interval." 9-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 3 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-6 Integrally B-H B8.20 Pressurizer Integrally Welded Surface Exam Inspect per Code Granted 3/3/93 Welded

Attachment:

Case N-323 Attachments to 03-007-SW-X Vessels Reference Document B13932(10/30/91)(S)

NRC Letter A10880 (3/3/93) includes Safety Eval of 1st 10 Year ISI Program Summary Report, Rev. 3 (G)Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval.

Code Case N-323 which is referenced in this relief request is not applicable to the 1989 Edition of the Code. Evaluation of alternate examination methods and submittal of the relief request should be made prior to performance of the examinations when the limitation and examination coverage percentage can be accurately determined.

IR-7 C-B C2.21 Steam Generators Full penetration nozzle-to-Volumetric Volumetric and Granted 2/08/91 vessel welds: and surface surface on 03-003-SW-R S/G A accessible areas 03-003-SW-T S/G A Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Eval~of 1st 10 yr. Program & TER (SAIC Report)(G)

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-5 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 4 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-8 B-D B3.1 10 Pressurizer Nozzle-to-vessel welds: Volumetric Volumetric exam of Granted 2/08/91 03-007-SW-A accessible portions 03-007-SW-B of weld 03-007-SW-C 03-007-SW-D 03-007-SW-E 03-007-SW-S B-D B3.130 Steam Generator Nozzle-to-vessel welds: Volumetric Volumetric exam of 03-003-SW-V Inlet accessible portions 03-003-SW-U Outlet of weld 03-004-SW-V Inlet 03-004-SW-U Outlet 03-005-SW-V Inlet 03-005-SW-U Outlet 03-006-SW-V Inlet 03-006-SW-U Outlet Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G)

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval for the Pressurizer Vessel to Nozzle welds; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-6 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 5 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-9 (Rev. 2) B-J B9.11 Piping pressure Safe end-to- pipe welds Surface and Full volumetric exam Granted 3/03/93 boundary nominal volumetric from the ID surface; pursuant with 2/08/91 pipe size (NPS) 4" OD surfaces will be demo or larger visually examined witness by NRC in 3/95 B-F B5.10 Reactor Vessel Nozzle-to-safe end butt welds Surface and Full volumetric exam See above.volumetric from the ID surface;OD surface will be visually examined.Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G).

Note: IR-9, Rev. 1, resubmitted in B13932 (10/30/91)

NRC Response A10880 (3/3/93)(G).

Revision 2 submitted in B17593 (12-21-98).

NRC Response A14092 (5/11/99) (G).Second Interval Applicability A similar relief request (IR-2-03) has been incorporated in the Second Interval Program.IR-10 Centrifugally B-J B9.11 Piping welds NPS Centrifugally cast stainless Volumetric Volumetric and Granted 2/08/91 cast stainless 4" or larger steel component-to fitting and surface surface on steel component welds: accessible portions to fitting welds LP4-EC-1 -SW-B RCS-20-FW-37 RCS-20-FW-38 RCS-20-FW-39 Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G).

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval for two of the welds. Welds LP4-EC-2-SW-B and RCS-20-FW-39 will not require relief based on Code Case N-460. Since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-7 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 6 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-11 Pressure B-J B9.1 1 Piping welds NPS Pressure retaining weld in Volumetric Volumetric and Granted 2/08/91 retaining welds 4" or larger Class 1 and surface surface on in piping SIL-6.6-SW-B accessible portions Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91-)

includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G).

Second Interval Applicability A similar relief request will not be required for the second Ten-Year ISI interval based on Code Case N-460.IR-12 Pressure C-F-1 C5.1 1 Piping welds NPS Welds: Volumetric Volumetric and Granted (not 3/03/93 retaining welds C-F-2 C5.51 4" or larger MSS-33-FW-1-GM and surface surface on required for in-piping C5.81 MSS-33-FW-1-HM accessible portions FWS weld)MSS-30-FW-2-7M FWS-17-6-SW-E SIL-25-FW-2 SIL-25-FW-1-7M Reference Document B13932 (10/30/91)(S)

NRC Letter A10880 (3/3/93) includes safety Eval of 1st 10 yr ISI program summary report, Rev. 3 (G)Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.IR-13 (Rev. 1) C-C C3.20 Piping Integrally welded attachments Surface Surface exam on the Granted 2/08/91 for piping weld accessible portions 3-SIL-4-PSR-040 Reference Document Letter (5/22/86)(S)

NRC A09339 Letter (2/8/91) includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G).

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-8 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 7 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-14 C-A C1.10 Residual heat shell to flange circumferential Volumetric Volumetric on the Granted 2/08/91 exchanger welds accessible portion;inaccessible portion will receive a liquid penetrant exam Reference Document Letter (5/22/86)(S)

NRC Letter A09339 (2/8/91) includes Safety Eval of 1st 10 yr. Program & TER (SAIC Report)(G).

Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.IR-18 B-H B8.20 Pressurizer Integrally welded attachments Volumetric or Perform in same Granted 3/03/93 surface period vs. over int.Reference Document B13932 (10/30/91)(S)

NRC Letter A10880 (3/03/93) includes Safety Eval of 1st 10 yr ISI program summary report, Rev 3 (G)Second Interval Applicability A similar relief request (IR-2-04) has been incorporated in the Second Interval Program.IR-19 Inner radius of C-B C2.22 SG main steam Inner radius of main steam Volumetric Visual exam during Granted 5/04/95 main steam nozzle nozzle off of SGs leak test nozzles of SGs Reference Document B15064 (12/16/94)(S), B15175 (3/29/95)(S), A12289 (5/04/95)(G)

Second Interval Applicability A similar relief request (IR-2-05)has been incorporated in the Second Interval Program.9-9 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 8 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-20 Integrally B-K-1 B10.10 Piping Integrally welded attachments Surface Surface exam on Granted 6/28/96 welded aft. for piping welds: accessible portions Welds RCS-504B-PSSH507 RCS-504B-PSSH508 Reference Document B15475 (1/12/96)(S), A12893 (6/28/96) (G) Note: NRC to reevaluate if Code Case N-509 is requested for use in the future.Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.IR-21 Pressure B-D B3.90 RV nozzle welds 107-121A, 105-121A, Volumetric Vol. exam on Granted 6/28/96 retaining welds B3.100 105-121B, 107-121B, accessible portions in reactor vessel 107-121C, 105-121C, including sys. Leak 105-121 D, 107-121D test.Reference Document B15475 (1/12/96)(S), Al 2893 (6/28/96) (G)Second Interval Applicability A similar-relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-10 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 9 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-22 Pressure B-F B5.70 SG nozzle to safe RCS-LP1-FW-4 Volumetric Vol. and surface Granted 6/28/96 retaining end welds RCS-LP1-FW-5 and surface exam on accessible dissimilar metal RCS-LP2-FW-4 portions including welds of SGs RCS-LP2-FW-5 sys. Leak test.RCS-LP3-FW-4 RCS-LP3-FW-5 RCS-LP4-FW-4 RCS-LP4-FW-5 Reference Document B15475 (1/12/96)(S), A12893 (6/28/96) (G)Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval; however, since Code Case N-498-1 has been utilized the proposed alternate examinations should not include a hydrostatic test. Relief request should be submitted following performance of examinations during the second Ten-Year interval.IR-23 Exemption from See relief See relief Class 1, 2, 3 See relief request See relief All repairs will be Withdrawn N/A repair request request components request done in accordance requirements of with our QA program IWA-4000 for piping valves, fit., nominal pipe size 1" and smaller and associated supports Reference Document B15120 (12/15/95)(S), B15659 (6/07/96), B17409 (W)Second Interval Applicability Letter B15120 also requested permission to use this relief request in the Second Interval Program..9-11 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 10 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-24 Exemption from Class 1, 2, 3 QA program uses Withdrawn N/A Repair/ components only qual. welders &replacement procedures requirements of IWA-4000 and IWA-7000 for seal welds Reference Document B15120 (12/15/95)(S), B15659 (6/07/96), B17409 (W)Second Interval Applicability Letter B15120 also requested permission to use this relief request in the Second Interval Program.IR-26 Reactor B-A B1.21 Lower Head 102-151 Volumetric Volumetric exam on Granted 2/17/98 Pressure Vessel Circumferential accessible portions;Lower Head Welds Inservice system Welds leakage test.B1.22 Lower Head 101-154A Meridional Welds 101-154B 101-154C 101-154D Reference Document B 16647 (7/31/97) (S), A 13670 (2/17/98) (G)Second Interval Applicability A similar relief request will be required for the second Ten-Year ISI interval.

Relief request should be submitted following performance of examinations during the second Ten-Year interval.9-12 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-1 FIRST INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 11 of-11)Exam.Cat.System or Item No. Component Relief Request IR-27 Title Reactor.Pressure Vessel Shell to Flange Weld Volume or Area To Be Examined 101 -121 B-A B1.30 Shell to Flange Weld Required Licensee Proposed Method Alternative Volumetric Vol. exam on accessible portions.Inservice leakage test.Status Date Approved Granted 2-22-99 Reference Document B17468 (9-23/98) (S)Second Interval Applicability IR-28 Relief from Visual Examination F-A F1.40 CL. 1 RPV Supports RVS-1 RVS-2 RVS-3 ,RVS-4 VT-3 VT-3 accessible portions including surrounding insulation.

Granted 11/3/99 Reference Document B17867 (9/17/99) (S), A15133 (11/3/99) (G)Second Interval Applicability A similar will be required for the second Ten Year ISI Interval.NOTE: For the reference documents, the following are applied: (S) = Submitted, (W) = Withdrawn, (G) = Granted, (D) = Denied 9-13 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (Sheet 1 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-01 Utilization of B-D, B-F Various Various Various As set forth in Alternatives as listed Granted 7/24/2000 Code Cases B-J, B-P the 1989 in the referenced N-389-1 C-F-1 Edition of the Code Cases 1 C-F-2 ASME Boiler N-416-1 C-2 C-H and Pressure N-491-1, D-A, D-B Vessel Code N-498-1, D-C, F-A N-521, N-522, and N-524 Reference Document: B17752 (4/7/99)(S), Al 5322 (7/24/00)(G)

IR-2-02 Snubber Snubbers Per IWF-5300, As set forth in the Granted 7/24/2000 Examination and as set forth in current Technical Testing the first Specifications in addenda to accordance with ASME/ANSI Generic Letter 90-09 OM-1987 '.(OMa-1988), Part 4 Reference Document: B17752 (4/7/99)(S), A15322 (7/24/00)(G)

IR-2-03 Proposed B-J B9.11 Piping pressure Safe end-to- pipe welds Surface and Full volumetric exam Granted 7/24/2000 Alternate boundary nominal volumetric from the ID surface Examination to pipe size (NPS) 4" Reduce or larger Exposure B-F B5.10 Reactor Vessel Nozzle-to-safe end butt welds Surface and Full volumetric exam volumetric from the ID surface Reference Document: B17752 (4/7/99)(S), Al 5322 (7/24/00)(G) 9-14 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 2 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component

.Examined Method Alternative j Approved IR-2-04 Proposed B-H B8.20 Pressurizer Integrally welded attachments Volumetric or Perform in same Granted 7/24/2000 Alternate surface period vs. over int.Examination to Reduce Exposure Reference Document: B17752 (4/7/99)(S), A15322 (7/24/00)(G)

IR-2-05 Proposed C-B SG main steam Inner radius of main steam Volumetric Visual exam during Granted 7/24/2000 Alternate to nozzle nozzle off of SGs leak test Impractical Examination Reference Document: B1 7752 (4/7/99)(S), Al 5322 (7/24/00)(G)

IR-2-06 (Rev 1) Utilization of all Leakage Identified As set forth in Alternative as listed Granted 7/24/2000 Code Case at Bolted Connections IWA- in the referenced N-566-1 5250(a)(2)

Code Case Reference Document: B17752 (4/7/99)(S), B18146 (6/21/00(S), A15322 (7/24/00)(G)

IR-2-07 Utilization of all all all Alternative as listed Granted 7/24/2000 Alternative in the referenced Requirements Code Case for ISI Intervals per Code Case N-535 Reference Document: B17752 (4/7/99)(S), A15322 (7/24/00)(G) 9-15 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 3 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-08 Utilization of Various Various Class 1 Systems Insulated Pressure Retaining VT-2 Alternative as listed Granted 7/24/2000 Code Case Bolted Connections in the referenced N-533 Code Case with additional requirement of a 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> hold time.Reference Document: B17752 (4/7/99)(S), A15322 (7/24/00)(G)

IR-2-09 Utilization of Various Various Class 1, 2 & 3 Qualification of VT-2 VT-2 Alternative as listed Granted 7/24/2000 Code Case Systems Examination Personnel in the referenced N-546 Code Case with additional requirements to test examiners Reference Document: B17752 (4/7/99)(S), Al 5322 (7/24/00)(G)

IR-2-10 Utilization of All All Class 1, 2, & 3 Repair and Replacement and _ Alternative as listed Granted 8/24/00 Code Case Systems Inservice Summary in the referenced N-532 Documentation Preparation.

Code Case Reference Document: B17871 (9/23/98)(S), A15335 (8/24/00)(G)

IR-2-11 Utilization of B-D B3.120 CL 1 Nozzle UT Not Code Case B3.140 inner Radius Submitted N-619 9-16 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 4 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-13 Nozzle to Shell C-B C2.21 Steam Generator Nozzle-to-Vessel Weld Volumetric Volumetric and Granted 10/04/00 Weld 03-53-SW-R and surface Surface exam of accessible portions, including VT-2 Reference Document: B18098 (5/31/00)(S), A15356 (10/4/00)(G)

IR-2-14 Shell-to-Flange C-A C1.10 B-RHR Heat Shell -to-Flange Weld Volumetric Volumetric Granted 10/04/00 weld Exchanger 03-074-004 accessible portions and surface exam of the areas inaccessible for volumetric., including VT-2.Reference Document B18098 (5/31/00)(S), A15356 (10/4/00)(G)

IR-2-15 Alternative B-A B1.10 Reactor Pressure RPV Shell Welds, Length Sizing Length Sizing Qual. granted 1/26/01 Length Sizing B1.20 Vessel Head Welds Subject to Qual. Criteria -Criteria to be 0.75 Criteria Append. VIII 1/4 inch + 1 inch RMS.inch Reference Document B18202 (8/25/00)(S), B18253 (11/8/00)(S)

A15423 (1/26/01)(G)

IR-2-16 Single Sided B-J, C-F-1 Austenitic welds Single side weld exam UT Document restricted Withdrawn access coverage on OAR-1 report in lieu of Relief Request Reference Document B18202 (8/25/00)(S

), B18253 (11/8/00 )(W)9-17 MILLSTONE POWER STATION-UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 5 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-17 Annual Various Various Components UT Annual UT training Granted 1/26/01 ultrasonic subject to ASME shall be conducted retraining Appendix VII lAW 50.55a(b)(2)(xiv) in lieu of Sec. XI Reference Document B18202 (8/25/00)(S), B18253 (11/8/00)(S)

A15423 (1/26/01)(G)

IR-2-18 Delay Various Various Components Written practice will Granted 1/26/01 implementation subject to ASME be updated to meet (Expired of CP-189 Appendix VII CP-189 by August 1, 8/1/01)2001 Reference Document B18202 (8/25/00)(S), B18253 (11/8/00)(S)

A15423 (1/26/01)(G)

IR-2-19 Request to use Various Various Class 1, 2, and 3 Pressure retaining bolted -VT-2 As listed in the Not Code Case Systems connections of borated referenced Code Submitted N-616 systems -Case Reference Document IR-2-20 Request to use Various Various Class 1, 2, and 3 Alternative Additional As listed in the Not Code Case piping, components examination requirements referenced Code Submitted N-586 and supports Case Reference Document 1-RI-ISI-01 Class 1 Risk B-J, B-F Various Class 1 piping Piping welds -Surface and As stated in the Granted 3/12/02 Informed welds Volumetric Class 1 ISI Risk Program Examination Informed Program Implementation as applicable Reference Document B1 8124 (7/25/00)(S), B1 8470 (9/26/01)(S)

Al 5653 (3/12/02)(G) 9-18 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 6 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-21 Pressure B-A B1.40 Rx Vessel Head Head-to-Flange Weld No. Volumetric Volumetric and Granted 8/20/02 retaining welds Flange 101-101 and surface surface examination in Reactor Examination of accessible Vessel Closure portions of the weld Head Reference Document B18490 (9/27/01)(S), A15742 (8/20/02)(G)

I IR-2-22 Full penetration B-B B1.40 Steam Generator

-Nozzle-to-Shell Weld No. -Volumetric Volumetric Granted 8/20/02 Welds of Nozzle-to-Shell 03-003-SW-U, examination of the Nozzles in 03-003-SW-V accessible portions Vessels 03-004-SW-U of the weld 03-004-SW-V Reference Document B18490 (9/27/01)(S), A15742 (8/20/02)(G)

IR-2-23 Pressure B-F B5.70 Steam Generator Nozzle-to-Pipe Weld No. -Volumetric Volumetric and Granted 8/20/02 Retaining Nozzle-to-Pipe RCS-LP3-FW-4 and surface surface examination Dissimilar Metal welds RCS-LP3-FW-5 examination of the accessible Welds RCS-LP4-FW-4 portions of the weld RCS-LP4-FW-5 Reference Document B18490 (9/27/01)(S), A15742 (8/20/02)(G)

IR-2-24 Pressure C-B C2.21 Steam Generator -Nozzle-to-Head Weld No. -Volumetric Volumetric and Granted 8/20/02 retaining Nozzle Nozzle-to-Head 03-053-SW-T and Surface surface examination Welds in Weld examination of the accessible Vessels portions of the weld Reference Document B1 8490 (9/27/01)(S), Al 5742 (8/20/02)(G)

IR-2-25 (Rev.1) Pressure C-F-1 C5.10 Circumferential Pipe -Pipe welds as listed in --Volumetric Volumetric and Granted 8/20/02 Retaining Welds Welds Relief Request and Surface surface examination in Austenitic examination of the accessible Stainless Steel portions of the weld Reference Document B18490 (9/27/01)(S), B18657 (5/30/02)(S), A15742 (8/20/02)(G) 9-19 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 7 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-26 (Rev. 1) Integral B-H B8.20 Pressurizer Support -Skirt-to Shell Weld No. -Surface Surface examination Granted 8/20/02 Attachments in Skirt-to-Shell Weld 03-007-SW-X examination of the accessible Vessels portion and a best effort volumetric (UT) examination.

Reference Document B1 8490 (9/27/01)(S), Bi 8657 (5/30/02)(S), Al 5742 (8/20/02)(G)

IR-2-27 Use of proposed C-F-2 C5.81 Feedwater piping Piping and welds as listed in Provisions of Granted 6/12/04 alternative to C-C C3.20 and attachment Relief Request Appendix B to be allow one time welds used along with the use of Non- participation of the ASME ANI in lieu of Certificate holder to perform R/R fabrication activities.

Reference Document B18796 (11/26/02)(S), S/N 04-027 (1/12/04)(G)

IR-2-28 Request to use N/A N/A N/A Use of 1998 Edition Not N/A 1998 Edition of of Section Xl Submitted Section XI for Repair/Replacement IR-2-29 Request to use Various Various All Various Various Use of 1998 Edition Not N/A 1998 Edition of of Section XI Submitted Section XI, IWX-2430 for choosing additional exams.IR-2-30 THRU Numbers Not N/A IR-2-35 Used 9-20 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 8 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-36 Request to use C-F-1 C5.10 Class 2 piping In Accordance with ASME MT/PT Examinations limited Granted 5/04/04 Code Case C-F-2 Thru weldsSection XI. to those areas N-663 C5.40 and identified as C550 susceptible to Thru outside surface C5.82 attack.Reference Document B19021 (3/30/04)(S), S/N 04-315 (5/04/04)(G)

RR-89-52 Request to use N/A N/A Service Water to Monitor leakage and Granted 9/22/05 Temporary Non- 3CCI*E1A Periodic follow-up Code repair in NDE until permanent Service Water repair can be Sys. Brazed performed during the joints next shutdown.Reference Document S/N 04-535 (9/23/04)(S), S/N 05-666 (9/22/2005(G)

IR-2-37 Temporary Non- N/A N/A Withdrawn N/A Code Repair of a Brazed joint in Service Water System Drain Line Reference Document S/N 05-100 (5/9/05) (S), S/N 05-100A (12/15/05 (W)IR-2-38 Alternative N/A N/A SWP Class 3 Perform Structural Granted 02/28/07 Brazed Joint brazed joints integrity assessment Assessment with supplemental Methodology LUT in lieu of immediate repair.Reference Document S/N 05-201 (6/9/05) (S), S/N 07-0153 (02/28/07) (G)9-21 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUESt

SUMMARY

TABLE (continued)(Sheet 9 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-39 Rev. 1 Use of Weld R-A R1.11 Pressurizer nozzle Full structural weld Granted 01/23/06 Overlay and R1.15 to safe end weld overlay Associated 03-X-5641-E-T Alternative Repair Technique Reference Document S/N 05-708A (10/19/05) (S), S/N 06-064 (01/23/06) (G)IR-2-40 Proposed C-H C7.30 oSS Buried piping IWA-5244(a)

IWA-5244(c)

Granted 07/10/08 Alternative From SIL iWA-5244(b)

Full flow testing Pressure Test Requirements of FWA ASME Section Xl, IWA5244 Buried Piping (Class 2 Piping)Reference Document S/N 07-0338 (11/09/07) (S), S/N 08-0437 (07/10/08) (G)IR-2-41 Proposed D-B D2.10 SWP Buried piping IWA-5244(a)

IWA-5244(c)

Granted 07/10/08 Alternative From IWA-5244(b)

Full flow testing Pressure Test Requirements of ASME Section Xl, IWA5244 Buried Piping (Class 3 Piping)Reference Document S/N 07-0338 (11/09/07) (S), S/N 08-0437 (07/10/08) (G)IR-2-42 Request to Use R-A R1.15 RCS Reactor vessel nozzle to safe UT Use of .224 RMS Granted -05/01/07 Alternative end welds error in lieu of .125 Sizing Criteria for depth sizing Reference Document S/N 06-226 (05/11/06) (S), S/N 07-366 (05/01/07) (G)9-22 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 10 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-43 Request Use of R-A R1.15 RCS Reactor vessel nozzle to safe UT Use of ASME Code Granted 05/02/07 ASME Code end welds Case N-696 for Case N-696 qualification requirements for Appendix VIII piping examinations conducted from the inside surface Reference Document S/N 06-226 (05/11/06) (S), S/N 07-369 (05/02/07) (G)IR-2-44 Request the B-A B1.30 RCS Reactor vessel shell to flange UT Perform UT Granted 05/01/07 Use of PDI weld examinations using Qualified procedures, Procedures, personnel, and Personnel and equipment qualified Equipment for in accordance with Non-Appendix Appendix VIII (PDI).VIII Reactor Vessel Shell-to-Flange Weld Reference Document S/N 06-226 (05/11/06) (S), S/N 07-368 (05/01/07) (G)IR-2-45 Request B-P B15.51 Various Class 1 system pressure VT-2 See Relief request Granted 09/27/07 Alternative B 15.71 boundary Class 1 Pressure Test Requirements Reference Document S/N 06-305 (08/17/06 (S), S/N 07-0676 (09/27/07) (G)9-23 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-2 SECOND INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 11 of 11)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-2-46 Relaxation of N/A N/A RCS Reactor Vessel Head UT Granted 05/02/07 the requirements of Order EA-03-009 Regarding Reactor Vessel Head Inspections Reference Document S/N 06-251 (05/16/06( (S), S/N 07-0370 (05/02/07) (G)IR-2-47 Request Use of N/A RCS Pressurizer Nozzle Full structural weld Granted 05/03/07 Weld Overlays to Safe end welds overlay as an Alternative Repair Technique Reference Document S/N 06-731 (10/17/06)(S), S/N 07-371 (05/03/07) (G)IR-2-48 Request to Use N/A N/A Not N/A Code Case N- Submitted 532-4 Reference Document N/A IR-2-49 Alternative N/A N/A Class 1, 2, and 3 Alternative rules for Granted 02/25/08 Request to Use valves acceptability for ASME Section Class 1 (N-756) and III Code Cases Class 2, 3 (N-757)N-756 and N- NPS 1 and smaller 757 Reference Document S/N 07-0416 (07/25/07) (S), S/N 08-0111 (02/25/08) (G)NOTE: For the reference documents, the following are applied: (S) = Submitted, (W) = Withdrawn, (G) = Granted, D) = Denied 9-24 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-3 THIRD INTERVAL RELIEF REQUEST

SUMMARY

TABLE (Sheet 1 of 3)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-3-01 Snubber N/A N/A Snubbers Snubber and attachment Use Technical Examination and hardware examination and Specification 4.7.10 Testing testing Reference Document IR-3-02 Relief from C-B C2.22 SG main steam Inner radius of main steam Volumetric Visual exam during volumetric exam of nozzle nozzle off of SGs leak test nozzle inside radius sections of Main Steam nozzles Reference Document IR-3-03 Not Submitted Reference Document IR-3-04 Alternative brazed N/A N/A SWP Class 3 Perform Structural joint assessment brazed joints integrity assessment methodology with supplemental UT in lieu of immediate repair.Reference Document IR-3-05 Examination of R-A R1.11 Six (6) Pressurizer Volume of welds defined in Volumetric Use of examination Weld Overlays R1.15 nozzle to safe end Code Case N-770 .criteria of Code welds and one (1) Case N-770 Pressurizer safe end to pipe weld Reference Document 9-25 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-3 THIRD INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 2 of 3)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-3-06 Pressure testing C-H C7.30 QSS Buried piping IWA-5244(a)

IWA-5244(b)(2) of Class 2 SIL IWA-5244(b)

Full flow testing buried piping FWA Reference Document IR-3-07 Pressure testing D-B D2.10 SWP Buried piping IWA-5244(a)

IWA-5244(b)(2) of Class 3 IWA-5244(b)

Full flow testing buried piping Reference Document IR-3-08 Implementation B-A B1.30 RCS Reactor vessel shell to flange UT Perform UT of Appendix VIII weld examinations using Supplements 4 procedures, and 6 -use of personnel, and PDI qualified equipment qualified procedures, in accordance with personnel, and Appendix VIII (PDI).equipment for non-Appendix VIII RPV flange-to-vessel weld Reference Document 9-26 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL TABLE 9.4-3 THIRD INTERVAL RELIEF REQUEST

SUMMARY

TABLE (continued)(Sheet 3 of 3)Exam. System or Volume or Area To Be Required Licensee Proposed Status Date Relief Request Title Cat. Item No. Component Examined Method Alternative Approved IR-3-09 Use of B-P B15.10 Various Class 1 system pressure VT-2 Visual Perform VT-2 alternative boundary examination of pressure testing certain components criteria for the at a reduced system leakage pressure and other test conducted components in their at or near the normally isolated end of the condition inspection interval on Class 1 piping Reference Document IR-3-10 Alternative B-P B15.10 Reactor Coolant Hot leg and cold leg. nozzle to VE Visual Perform Volumetric examination system pipe welds Exams in lieu of criteria for the Visual Examination visual examination of Reactor Coolant System hot leg and cold leg nozzle-to-safe end welds Reference Document IR-3-11 Alternative B-P B15.10 RPV flange leak-off RPV flange leak-off piping VT-2 Visual Perform VT-2 visual pressure testing piping examination each requirements for outage on the the RPV flange unpressurized leak-off piping subject piping Reference Document 9-27 K MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 10. Reserved for Future Use 10-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 11. DRAWINGS The applicable zone drawings for component welds and supports, and their associated Program Plan listings are maintained on site as part of the Inservice Inspection Program Manual.11-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL 12. IMPLEMENTATION OF ASME SECTION XI, APPENDIX VIII Program Basis 10 CFR 50.55a, as amended by the Federal Register Published September 22, 1999, Volume 64, Number 183 (Final Rule) required expedited implementation of Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems".

The effective date for the expedited implementation was November 22, 1999.Effective Code DNC will implement Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems" in accordance with ASME Section XI, 2001 Edition for the 3rd Inspection Interval of MPS3.Implementation Schedule DNC has implemented Appendix VIII in accordance with the following schedule as defined in 10 CFR 50.55a: SUPPLEMENT 1 2 3 4 5 6 7 8 9 10 11 12 13 14 QUALIFICATION REQUIREMENTS Evaluating Electronic Characteristics of Ultrasonic Systems Wrought Austenitic Piping Welds Ferritic Piping Welds Clad/Base Metal Interface of Reactor Vessel Nozzle Inside Radius Section Reactor Vessel Welds Other Than Clad/Base Metal Interface Nozzle-to-Vessel Welds Bolts and Studs Cast Austenitic Piping (In Course of Preparation)

Dissimilar Metal Welds Full Structural Overlaid Wrought Austenitic Piping Welds Coordinated Implementation of Selected Aspects of Supplements 2, 3, 10 and 11 Coordinated Implementation of Selected Aspects of Supplements 4, 5, 6 and 7 Supplement 14: Qualification Requirements for Coordinated Implementation of Supplements 10, 2, and 3 for Piping Examinations Performed from the Inside Surface IMPLEMENTATION DATE May 22, 2000 May 22, 2000 May 22, 2000 November 22, 2000 November 22, 2002 November 22, 2000 November 22, 2002 May 22, 2000 N/A November 22, 2002 November 22, 2001 November 22, 2002 November 22, 2002 November 22, 2002 12-1 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Performance Demonstration Program DNC will utilize personnel qualified through the Performance Demonstration Initiative (PDI) Program. PDI is an organization comprised of all the US nuclear utilities formed to provide an efficient, cost effective and technically sound implementation of Appendix VIII performance demonstration requirements.

Performance Demonstration Program Administrator DNC will utilize the EPRI NDE Center as the Performance Demonstration Administrator (PDA) in accordance with the PDI Program Description Document (PDD), Revision 4 1.Supplement 1: Evaluating Electronic Characteristics of Ultrasonic Systems This Supplement defines the steps necessary to interchange the pulsars and/or receivers in an ultrasonic examination system without the need for requalification.

The PDI Program does not support this activity, but PDI will supply DNC with technical guidance as required on a case-by-case basis.Supplement 2: Qualification Requirements for Wrought Austenitic Piping Welds The PDI Program is in full compliance with Supplement 2 as modified by 10 CFR 50.55a for examinations conducted from outside the piping. The PDI Program does not address examinations conducted from inside the pipe except for RPV nozzle-to-pipe welds. These examinations are included in Supplement 14, which is described later. Welds containing corrosion-resistant cladding (CRC) that is part of the pressure-retaining boundary, such as that typically applied' to mitigate cracking are excluded.

A recent Code inquiry indicates that these welds should be examined in accordance with Appendix IT until qualification requirements are developed.

Supplement 3: Qualification Requirements for Ferritic Piping Welds The PDI Program is in full compliance with Supplement 3 as modified by 10 CFR 50.55a for examinations conducted from outside the pipe. The PDI Program does not address examinations conducted from inside the pipe except for RPV nozzle-to-pipe welds.Supplement 4: Qualification Requirements for the Clad/Base Metal Interface of Reactor Vessel The PDI Program is in full compliance with Supplement 4 as modified by 10 CFR 50.55a.Supplement 5: Qualification Requirements for Nozzle Inside Radius Section The PDI Program is in full compliance with Supplement 5 as modified by 10CFR50.55a for examinations conducted from inside the reactor vessel.The PDI Program is in full compliance with Code Case N-552 as modified by Regulatory Guide 1.147. For examinations conducted from outside the reactor 12-2 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL vessel. Code Case N-552 is applicable to both the nozzle inside radius region and the inner 15% of nozzle-to-shell welds when scanning for flaws oriented transverse to the weld. Code Case N-552 requires that all nozzles be modeled, including the inner 15% of the nozzle-to-vessel weld, and introduces maximum metal path and maximum misorientation angles as new essential variables that must be qualified.

Code Case N-552 is identified as an acceptable alternative in Regulatory Guide 1.147 and should be identified in the ISI Program. Modeling will be the responsibility of DNC.Supplement 6: Qualification Requirements for Reactor Vessel Welds Other Than Clad/Base Metal Interface The PDI Program is in full compliance with Supplement 6 as modified by 10 CFR 50.55a. In addition, the PDI Program considers that demonstrations conducted on clad vessel specimens in accordance with Supplement 4 exceed the requirements of Supplement 6 and may be used for examination of the inner 10%of unclad components without a relief request. Demonstrations performed on unclad vessel specimens, however, may not be used for the examination of clad vessels.Supplement 7: Qualification Requirements for Nozzle-to-Vessel Welds The PDI Program is in full compliance with Supplement 7 as modified by 10 CFR 50.55a. The title of this Supplement is somewhat misleading:

originally intended as a stand-alone Supplement, the examination coverage requirements of 10 CFR 50.55a have resulted in this Supplement being applicable only to examinations conducted from the nozzle bore (for example, PWR inlet and outlet nozzles).

Examinations of the inner 15% conducted from the outside for flaws oriented transverse to the weld are qualified according to Supplement 5 using Code Case N-552 as above. Otherwise, Supplement 4 and 6 are applicable.

Supplement 8: Qualification Requirements for Bolts and Studs The PDI Program provides personnel only demonstrations for bolting. This demonstration covers a range of bolting types and sizes, using the typical equipment and other essential variables.

Other variables to be used in each examination shall be demonstrated during calibration, prior to the examination.

Supplement 9: Qualification Requirements for Cast Austenitic Piping Welds In 'accordance with Appendix VIII, specific qualification requirements are in the course of preparation, and the requirements of Appendix III are applicable.

Supplement 10: Qualification Requirements for Dissimilar Metal Piping Welds The PDI Program is not in compliance with Supplement

10. PDI qualifications are conducted in accordance with Code Case N-695. This Code Case is identified in RG 1.147, as acceptable and should be included in the ISI Program. A relief request will be required if the examination procedure is incapable of depth sizing flaws to an accuracy of 0.125 root mean square (RMS).12-3 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL Supplement 11: Qualification Requirements for Full Structural Overlaid Wrought Austenitic Piping Welds The PDI Program is in compliance with Supplement 11.Supplement 12: Requirements for Coordinated Implementation of Selected Aspects of Supplements 2, 3, 10, and 11 The PDI Program is in full compliance with Supplement 12 as modified by 10 CFR 50.55a for the coordinated implementation of Supplements 2 and 3 for examinations conducted from the outside. It does not support the coordinated implementation of Supplements 10 and 11; these are performed on an individual basis. Supplement 14 is appropriate on a limited basis for examinations conducted from the inside surface.Supplement 13: Requirements for Coordinated Implementation of Selected Aspects of Supplements 4, 5, 6, and 7 The PDI Program does not support this Supplement.

Supplements 4, 5, 6, and 7 qualifications are performed on an individual basis.Supplement 14: Qualification Requirements for Coordinated Implementation of Supplements 10, 2, and 3 for Piping Examinations Performed from the Inside Surface Supplement 14 is functionally applicable only to PWR vessels. It is a new Supplement established by the PDI Program for a coordinated implementation of the qualifications required for the typical examinations performed from the ID of PWR nozzles. Supplement 14 uses the more technically stringent Supplement 10 qualification as a base and then incorporates a limited number of Supplement 2 and Supplement 3 samples. Qualification requirements for examination conducted from the inside surface are conducted in accordance with Code Case N-696.Single Sided Access for Austenitic Welds The Final Rule requires that if access is available, the weld shall be scanned in each of the four directions (parallel and perpendicular to the weld) where required.Coverage credit may be taken for single sided exams on ferritic piping. However, for austenitic piping, a procedure must be qualified with flaws on the inaccessible side of the weld. There are currently no qualified single side examination procedures that demonstrate equivalency to two-sided examination procedures on austenitic piping welds. Current technology is not capable of reliably detecting or sizing flaws on the far side of an austenitic weld for configurations common to US nuclear applications.

The PDI Program conforms with the Final Rule regarding single side access for piping. PDI Performance Demonstration Qualification Summary (PDQS)certificates for austenitic piping list the limitation that single side examination is performed on a best effort basis. The best effort qualification is provided in place of a complete single side qualification to demonstrate that the examiners qualification and the subsequent weld examination is based on application of the 12-4 MILLSTONE POWER STATION UNIT 3 INSERVICE INSPECTION PROGRAM THIRD TEN-YEAR INTERVAL best available technology.

When the examination area is limited to orfe side of an austenitic weld, examination coverage does not comply with 10 CFR 50.55a(b)(2)(xv)(A) and proficiency demonstrations do not comply with 10 CFR 50.55a(b)(2)(xvi)(B) and full coverage credit may not be claimed.DNC will document the affected austenitic welds for which best effort one sided exams are encountered and the percentage of the weld examined.DNC will submit a request for relief of one sided examination of austenitic welds.Credit will be taken for single sided access for ferritic welds and the weld shall be scanned in each of the four directions (parallel and perpendicular to the weld)where accessible.

Responsibilities The Unit ISI Program Owner (DNC) is responsible for implementing the Appendix VIII program.The Site Level III is responsible for certification and proficiency maintenance of NDE personnel.

12-5 Serial No.09-187 Docket No. 50-423 MPS3 Third 10-Year Interval ISI Relief Requests ATTACHMENT 2 MILLSTONE POWER STATION UNIT 3 THIRD 10-YEAR INTERVAL INSERVICE INSPECTION (ISI)RELIEF REQUESTS DOMINION NUCLEAR CONNECTICUT, INC.MILLSTONE POWER STATION UNIT 3' 10 CFR 50.55a Request Number IR-3-01 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--1. ASME Code Components Affected ASME Code Classes: Code Classes 1, 2, 3 and MC

References:

ASME Section XI, IWF-5300 ASME/ANSI-OM (Part 4)Examination Category:

N/A Item Number: N/A

Description:

Snubber Examination and Testing Components:

Code Class 1, 2, 3 and MC Snubbers 2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement IWF-5300 (a) states: "Inservice examinations shall be performed in accordance with ASME/ANSI OM, Part 4, using the VT-3 visual examination method descrilbed in IWA-2213." IWF-5300 (b) states: "Inservice tests shall be performed in accordance with ASME/ANSI OM, Part 4." IWF-5300 (c) states: "Integral and nonintegral attachments for snubbers, including lugs, bolting, pins, and clamps, shall be examined in accordance with the requirements of this Subsection." Table 1WA-1600-1 references ASME/ANSI-OM (Part 4) Revision 1987 with OMa-1988.4. Reason for Request OMa-1988 imposes surveillance requirements for visual inspection and functional testing of snubbers.

A visual inspection is the observation of the condition of the installed snubbers to identify those that are damaged, degraded, or inoperable as caused by physical means, leakage, corrosion, or environmental exposure.

To verify that a snubber can operate within specific performance limits, functional testing is required that typically involves removing the snubber and testing it on a specially designed test stand. Functional testing provides a 95% confidence that 90-100% of the snubbers operate within the specified acceptance limits. The performance of Page 1 of 2 10 CFR 50.55a Request Number IR-3-01 (Continued) visual inspections is a separate process that complements the functional testing program and provides additional confidence in snubber operability.

The Code specifies a schedule for snubber visual inspections that is based on the number of inoperable snubbers found during the previous visual inspection.

Because the current schedule for snubber visual inspections is based only on the number of inoperable snubbers found during the previous inspection, irrespective of the size of the snubber population, the visual inspection schedule is excessively restrictive.

A significant amount of resources must be spent, and plant personnel subjected to unnecessary radiological exposure to comply with the visual examination requirements.

5. Proposed Alternative and Basis for Use As an alternative to performing inservice examination and testing in accordance with ASME/ANSI OM, Part 4, as required by lWF-5300, Millstone Power Station Unit 3 (MPS3) will apply the visual and functional testing requirements that are prescribed by MPS3 Technical Specification 4.7.10 (including sampling and frequency requirements) to the snubbers identified above.An alternate schedule for visual inspections has been developed that maintains the same confidence level as the existing schedule.

and generally will allow performance of inspections and corrective actions during plant outages. This schedule is given in Table 4.7-2, invoked from MPS3 Technical Specification 4.7.10.b.

Because it will reduce future occupational radiation exposure and is highly cost effective, this is consistent with NRC policy. The alternative inspection schedule is based on the number of unacceptable snubbers found during the previous inspection in proportion to the sizes of the various snubber populations or categories.

While the schedule of examinations is to be determined from MPS3 Technical Specification 4.7.10, the examinations are still to be performed using VT-3 visual examination certified personnel.

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.7. Precedents A similar alternative was approved for use at MPS3 during the second inservice inspection 10-year interval (see Relief Request IR-2-02), in NRC letter A15322, dated July 24, 2000, ADAMS Accession No. ML003730922.

Page 2 of 2 10 CFR 50.55a Request Number IR-3-02 Relief Request in Accordance with 10 CFR 50.55a(g)(5)(iii)

--Inservice Inspection Impracticality--

1. ASME Code Component(s)

Affected ASME Code Class: Code Class 2

Reference:

ASME Section XI, Table IWC-2500-1 Examination Category:

C-B Item Number: C2.22

Description:

Relief from Volumetric Examination of Nozzle Inside Radius Sections of Main Steam Nozzles Components:

Component Identification Numbers: 03-053-SW-T-IR 03-054-SW-T-IR 03-055-SW-T-IR 03-056-SW-T-IR

2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement ASME Section XI 2004 Edition Table IWC-2500, Item C2.22 requires a volumetric examination of the nozzle inside radius of nozzles without reinforcing plate in vessels greater than 1/22 inch nominal thickness, as defined by Figure IWC-2500-4 (a), (b), or (d). In the case of multiple vessels of similar design, size, and service the required examinations may be limited to one vessel or distributed among the vessels.4. Impracticality of Compliance In evaluating the referenced ASME Code, the concern is that the inside radius of the Main Steam Nozzle is considered susceptible to flaw initiation and growth due to high thermal and mechanical stresses associated with the vessel and connected piping systems. In the case of the MPS3 steam generator nozzles, the nozzle is a one-piece forging containing a set of seven holes bored parallel to the nozzle centerline (see Figure 1 in Attachment 1). This nozzle design does not match the typical figures in Figure IWC-2500-4.

Since the ligaments between the holes distribute the loads throughout the nozzle forging, the primary stress Page 1 of 2 10 CFR 50.55a Request Number IR-3-02 (Continued) resulting from the pressure is significantly lower than those for a typical nozzle. The plate ligaments provide both reinforcement and additional insulation by directing the flow away from the nozzle inner radius. Furthermore, the local stresses resulting from the thermal gradients in the inner radius are expected to be low, since the nozzle is only exposed to saturated steam resulting in a low transfer between the nozzle and the coolant.The design of the nozzles consists of seven, 8-1/2 inch bore holes, which precludes a meaningful ultrasonic examination of the area of interest.

As a result, a volumetric examination is impractical.

The design of the nozzle precludes visual examination of the nozzle inside radius even if access to this area were possible.

Access to the steam generator main steam nozzle from inside the steam generator is restricted by the upper deck plate and moisture separators; therefore, visual examinations from inside the steam generator are impractical.

The unique design of the steam generator main steam outlet nozzle results in low stresses when compared to nozzles with typical inner radius configurations.

Stresses in the nozzle inner radius region are less than 68 percent of the ASME Code allowable for each design condition.

This was determined by a review of a proprietary Westinghouse Steam Generator Stress Report. In addition, the stresses are considerably lower than those of other nozzles, such as the steam generator feedwater nozzle.5. Burden Caused by Compliance In order to provide access to perform an inner radius examination, significant modifications would have to be made to the current steam generator design. These modifications would be cost prohibitive and would yield little, if any, safety benefit.6. Proposed Alternative and Basis for Use Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested from the ASME Section XI requirements described above. Instead, Dominion Nuclear Connecticut (DNC) will perform visual examination during the system leakage tests as required by Section XI.7. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.8. Precedents Similar relief requests were approved for the First Interval (Relief Request IR- 19) in NRC letter A12289, dated May 4, 1995 and the Second Interval (Relief Request IR-2-05) in NRC letter A15322, dated July 24, 2000, ADAMS Accession No. ML003730922.

Page 2 of 2 10 CFR 50.55a Request Number IR-3-02 Attachment 1 Figure 1: Steam Generator Steam Outlet Nozzle Forging Attachment 1, Page 1 of 1 10 CFR 50.55a Request Number IR-3-04 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--1. ASME Code Components Affected ASME Code Class: Code Class 3

References:

ASME Section XI, IWA-4000, IWA-5250 and IWD-3000 Examination Category:

N/A Item Number: N/A

Description:

Alternative Brazed Joint Assessment Methodology Components:

Service Water System Brazed Piping Joints, 3 inches Nominal Size and Smaller Figure 1 in Attachment A shows a typical brazed joint. Attachment B provides additional details concerning applicable brazed joint materials, configuration and brazing.2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement If leakage of a Class 3 brazed connection is discovered during the course of normal operation, IWA-4000, Repair/Replacement Activities, applies and the joint must be repaired or replaced in accordance with that article. However, if the leakage is discovered during a scheduled leak test, the joint must be evaluated and repaired in accordance with IWD-3000 as clarified by the following: " IWD-3000 does not have acceptance criteria for Class 3 components.

IWD-3500,"ACCEPTANCE STANDARDS" refers to IWC-3500, "ACCEPTANCE STANDARDS".

IWC-3516, "Examination Category C-H, All Pressure Retaining Components" states, "These standards are in the course of preparation.

The standards of IWB-3522 may be applied."" IWB-3522.1 establishes the acceptance standard for Visual Examination, VT-2, in which leakage of non-insulated and insulated piping is listed as a relevant condition.

lWB-3522.1 states that such relevant conditions that may be detected during the conduct of system pressure tests shall require correction to meet the requirements of IWB-3142 and IWA-5250 prior to continued service." IWA-5250, "Corrective Action," in the context of a system leak test, requires identification of the source of leakage for evaluation of its corrective action which may include repair/replacement activities.

Page 1 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued)" IWB-3142, "Acceptance," permits acceptance of visually identified conditions under the requirements of IWB-3142.2, "Acceptance by Supplemental Examination."" IWB-3200, "SUPPLEMENTAL EXAMINATIONS," permits supplemental surface or volumetric examinations to determine the extent of the unacceptable conditions and the need for corrective measures, repairs, analytical evaluation, or repair/replacement activities.

4. Reason for Request In the course of plant operation, brazed joints are sometimes observed to be leaking at very low rate ("weepage")

through a defect in the braze bond between the pipe and fitting.Applicable Code requirements depend on whether the leak is discovered in the course of normal plant operation or during a scheduled leak test.'Section XI and Section III of the ASME Code do not have rules applicable to evaluation of weepage through brazed joints caused by defects in braze bonding between piping and fittings.

Section XI, IWD-3000, has no acceptance standards and refers to the rules of IWB-3000. However, 1WB-3000 has no rules pertaining to brazed joints. Therefore,Section XI does not have rules specific to examination and acceptance of relevant conditions observed in brazed joints. Lacking such rules, the leaking joint must be repaired in accordance with IWA-5250(a)(3) if found during a Code required system leakage test or IWA-4000 during any other mode of system operation.

A safe alternative to the requirement to immediately repair a brazed joint with leakage can include a deferred, but planned, repair/replacement activity that permits continued plant operation based on an evaluation of continued acceptable integrity and functionality of the brazed joint.. With this approach, sections of piping containing brazed joints can be replaced with welds or flanges in a systematic and planned manner and without unnecessary unavailability of safety related systems or components as well as unnecessary plant shutdowns.

5. Proposed Alternative and Basis for Use It is proposed that in lieu of the immediate repair requirement of IWA-5250 or IWA-4000, DNC perform a supplemental ultrasonic test (UT) examination and comparison with alternative acceptance criteria.

The UT examination will establish the extent of braze bond within the joint. The UT results will be compared with pre-established brazed joint bond levels required for structural integrity of the specific piping under consideration and that account for the design basis loadings applicable to the condition.

This will establish the basis for determining joint integrity to the extent required for system operability.

The lack of full braze bonding originates from construction, or fabrication, and is not progressive over time. However, the proposed methodology provides for continued monitoring until a resolution of the nonconforming condition (e.g., weepage) occurs through repair/replacement activities.

Periodic monitoring of the joint and its leakage verifies that ASME Code Interpretation XI- 1-92-19 Page 2 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) assumptions used for the assessment remain valid. The overall methodology has been validated by performance of physical testing on an array of simulated bond configurations, as well as several brazed joints salvaged from MPS3 piping. Consequently, the request provides an acceptable level of quality and safety commensurate with the original licensing and design basis of MPS3 as well as the provisions of 10 CFR 50.55a(a)(3)(i).

5.1 SCOPE

The alternative is limited to brazed service water piping (typically constructed of copper-nickel or Monel piping and cast bronze fittings) or on-skid equipment piping that has a design pressure of 150 psig or less and a design temperature of 150 degrees Fahrenheit or less. The piping nominal size is limited to three inches maximum.Basis: The limitation of pipe sizes to three inches or less ensures that the alternative is applied to piping for which it was intended, and is comparable to the range of pipe sizes (2 and 3 inches) included in the physical testing described in Attachment D. The limitation to service water systems ensures that the operating pressure and temperature are well within the moderate energy range. The fluid Contents of the piping are comparable to the ones examined for potential corrosion effects.5.2 EXAMINATION As permitted by LWB-3200, "Supplemental Examinations," the brazed joint will be examined by UT using a straight beam technique that monitors the relative strengths of signals returned from the internal diameter (ID) of the pipe and the fitting. This technique was derived from and is consistent with the technique standardized by the U.S. Navy for use on brazed shipboard piping.2 The UT procedure in Attachment E is provided for reference only and is subject to change. The UT procedure will require that technicians be certified in accordance with ANSI / ASNT CP- 189, 1995 Edition. Only Level II or III certified technicians may perform or review the braze readings and they must be familiar with brazed joint geometry and signal response characteristics..

As a prerequisite, the examination surface must be suitably prepared to obtain satisfactory sound transmission.

The joint circumference is marked at a number of locations such that they are spaced no greater than 1 inch apart. For the actual examination a straight beam longitudinal wave signal is required.

At each marked location the percent bond is recorded based on the relative strengths of signals received from the pipe ID and fitting ID. The procedure provides instructions to distinguish between fittings of the "face fed" and "insert" type, the latter of which have an internal groove in which a ring of braze filler material is inserted before brazing.The MPS3 UT procedure will provide for documentation of the braze bond readings on suitable data sheets which also include the calibration data. The data sheets are reviewed 2 NAVSEA 0900-LP-001-7000, "Fabrication and Inspection of Brazed Piping Systems", dated January 1, 1973.Page 3 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) by a certified Level II or III reviewer.

The data sheets are then forwarded to Engineering for assessment.

Basis for Nondestructive Examination Technique:

The alternative UT examination is based on requirements tor UT examination contained in the U.S. Navy standard for brazed piping. It uses basic straight beam UT technology, and was utilized to confirm the quality of critical piping systems in the submarine fleet of the U.S. Navy. A brazed joint is considered acceptable without further evaluation by the standard if the average bond is 60 percent or more.Consistent with the reference standard, the MPS3 procedure will require this work to be performed by certified UT technicians, using calibrated equipment and approved couplants.

It will require examination at multiple locations around the circumference of the fitting. It will require review of the data by a Level II or III technician.

The UT procedure will be reviewed and approved by a Level III technician in accordance with Dominion quality requirements.

Previous trial demonstrations show that individual bond readings at a location on the fitting may vary but the average reading is consistent among qualified examiners.

5.3 ASSESSMENT

An assessment of the joint using this methodology includes the following considerations:

• system performance and indirect effects assessments,* adjustment of bond readings to account for uncertainties,* a review of design basis stress analysis of the piping to determine required joint strength,* comparison of the adjusted bond readings with the prequalified bond levels that have been shown empirically by physical testing to assure structural integrity.

5.3.1 SYSTEM

EFFECTS As a prerequisite to structural assessment, knowledgeable engineering personnel assess the effect of the leak on the system and other nearby equipment.

Typically a brazed joint with a defect in the braze material bonding will leak only drops per minute. The actual leak rate will be estimated and compared to service water system margins for loss or diversion of flow. In addition, a walkdown will be performed to identify any nearby equipment that may be affected by dripping or impingement spray from the leak. If required, a drip collection device or spray shield will be installed and maintained for the duration that the leak continues.

Basis: ASME Code,Section XI code cases such as N-513-2 permit continued operation of low energy systems with minor leakage when justified by evaluation of system performance.

Similarly, the proposed alternative permits continued operation provided that the leakage rate will not adversely affect required flows and the Page 4 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) leakage or spray will not adversely affect safety related equipment.

Typical flow from a weeping brazed joint is in terms of drops per minute. Even in a theoretical worst case of a joint having a total lack of braze material, the close tolerance between the pipe and fitting prevents significant flow. The total diametric clearance of a braze joint is about 0.005 inches. For a 3 inch pipe, the maximum possible flow area would be nominally 0.28 square inches (e.g., 3.14 x 3.5 x 0.0025) through which the upper bound flow rate at 100 psig would be about 6 gpm, a very small rate in comparison to service water pump capacity.

More realistic estimates and actual leak rates would be much lower. Therefore, the maximum potential for braze joint leakage is very small. In addition, the proposed alternative requires a specific evaluation to assure that leakage does not unacceptably reduce system margins. Therefore, the system will meet all functional requirements and maintain an equivalent level of quality and safety.5.3.2 ACCEPTANCE THRESHOLD AND ADJUSTMENT OF BOND READINGS If the average measured bond reading is 60 percent or above, then no further assessment is required since the bond strength exceeds piping strength.

If the average is less than 60 percent, then the bond readings as documented in the UT procedure are adjusted downwards on a sliding scale, such that all readings at 10 percent and below are assumed to be zero, and readings above 10 percent are adjusted using the following formula: badj = 100 x (reading -10)/( 100 -10) units of percent For example, a 50 percent UT reading would be adjusted to 44 percent bond level for assessment purposes.

For simplicity, the adjustment may be applied to the average of the UT readings, or alternatively to each of the UT readings prior to averaging.

The average of the adjusted readings is then used for assessment purposes.

For bond readings that are significantly non-uniform around the circumference of the braze, an effective (lower) bond is computed based on the equivalent moment of the adjusted bond areas.If the average adjusted bond reading is above 55 percent then the joint strength is considered equal to or better than the piping and steps 5.3.3 and 5.3.4 below are skipped.Basis for acceptance threshold and adjustments of readings: Acceptance of average UT bond readings of 60 percent or more is the same as the acceptance criteria in the U.S. Navy standard that has been used for critical shipboard piping systems. The U. S. Navy criteria are applicable to systems rated 300 psig and greater. The 60 percent threshold criterion is therefore conservative for systems with design conditions 150 psig or less. For further confirmation of the 60 percent threshold, testing has shown that if true bond in the joint exceeds Page 5 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) 30 percent then the piping collapse load occurs before any bond failure. The testing performed for MPS3 is described in Attachment D. There is no braze bond failure mode because the piping deforms plastically to relieve the imposed load, and this occurs at loads greater than the maximum load permitted by the licensing basis analysis of the piping. The downward adjustment of bond readings, beyond what is required by the U.S. Navy standard, is an introduced conservatism used to help correlate the data from actual piping samples and accounts for uncertainties in bond readings.5.3.3 CONSTRUCTION CODE QUALIFICATION STRESS ANALVSIS REVIEW The Construction Code qualification stress analysis of record is reviewed to determine design basis loadings at the subject braze joint. Pressure, deadweight, and safe shutdown earthquake (SSE) loadings are included.

The loads are either used directly or expressed in terms of equivalent pipe stress so that stress analysis outputs may be used directly.

The stress intensification factor (SIF) that may have been applied in Construction Code stress analysis is not required to be included in the summation of nominal stresses used for assessment.

Basis for Stress Analysis Review: The review of stress analysis required by this proposal is a data gathering activity required to determine the primary loads imposed on the brazed joint. The primary loads consist of maximum operating pressure, deadweight, SSE seismic, and any transient dynamic loads that have been defined for the piping. Since the stress analysis is the calculation of record for qualifying the piping in accordance with licensing basis requirements, it is an acceptable source of input for assessing the structural integrity of brazed joints.The use of Construction Code stress values implicitly treats piping torsion loads as equivalent to bending moments. This is conservative because in the bonded joint the torsional shear is actually half that calculated on an equivalent pipe stress basis.5.3.4 COMPARISON OF ADJUSTED BOND TO REQUIRED BOND Equation 3 in Figure 2 of Attachment A was developed to give the allowable loading for an equivalent bondlevel.

The equation is used for a comparison that is needed only when the average bond is less than 60 percent. When an equivalent adjusted bond of a brazed joint is determined, as described in section 5.3.2, an allowable loading (Smax(badj) ) can be obtained from the equation.

This is the safe loading level that the joint can withstand.

If the joint load demand that has been determined in section 5.3.3 is less than the allowable ( Seq < Smax(badj)), then the brazed joint is concluded to have adequate structural integrity for continued service. The comparison is quantified as shown in Figure 2.An example of a structural assessment performed for a hypothetical leaking brazed joint is included in Attachment C. The example is for a joint with 55 Page 6 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) percent average measured bond, which is adjusted to an effective minimum bond of 43 percent for bending loads. This effective bond level results in a joint load capability of 11.0 ksi nominal pipe stress. The 11.0 ksi load capability is adequate for the design basis loads of this example since the joint load demand is only 4.4 ksi. Therefore, the example structural assessment concludes the joint can be left in service provided it is monitored until its permanent repair/replacement activity is completed.

If a joint does not have adequate bond by this assessment, this comparison for determining the adequacy of structural integrity of the joint is not applicable.

Prompt repair/replacement of the joint, or temporary non-Code repairs subject to NRC review and approval may still be an option, consistent with considerations in Regulatory Issue Summary 2005-20 for the resolution of degraded and nonconforming conditions.

Basis for Comparison of an Adjusted to Required Bonding: Brazed joints with reduced bond levels can retain a significant strength that is adequate for the structural integrity of the joint. Dominion has sponsored tests at an independent testing facility to demonstrate the correlation between reduced bond levels and joint strength.

The tests and their results are described in Attachment D.The correlation developed by the testing conservatively determines a required bond level for a given intensity of joint loading. The results of these tests support the use of the comparison shown in Figure 2 of Attachment A for the structural integrity analysis.The estimated joint strength obtained using Equation 3 in Figure 2 is confirmed conservative by test results. Each of the tested joints achieved a collapse load well above that which would be predicted for a 5 ksi braze shear strength.

This also confirms the conservatism of the 5 ksi maximum braze shear stress assumption that is used as an input to the Equation 3, shown in Figure 2.With the adjustment of bond readings imposed by this methodology, and a joint load capacity that is based on a 5 ksi shear stress, the tests demonstrate that a margin of greater than 1.5 exists between test results and estimated allowable joint load capacity from the actual piping removed from plant service. This margin provides an equivalent factor of safety (FS) to that provided by the ASME Code, Sections III and XI.The ASME Code,Section III, Appendix F has been accepted by the NRC for evaluation of degraded conditions.

3 Appendix F, paragraph F1331.1 (a) permits primary stress at levels up to 0.7Su and in paragraph (c) it permits primary membrane plus bending stress at levels up to (1.5)(0.7Su)

= 1.05S,. These result in a maximum FS of 1.4 relative to ultimate strength.

In shear across a section, paragraph F133 1.1 (d) limits shear to 0.42Su for a FS of 1.37 relative to (1 / 43)Su.3 Generic Letter 91-18, Rev. 1, "Information to Licensees Regarding NRC Inspection Manual Section on Resolution of Degraded and Nonconforming Conditions," October 8, 1997.Page 7 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued)

The 5 ksi shear limit used at the braze bond is well below this Appendix F limit of 0.42Su.The ASME Code,Section XI permits acceptance of planar flaws for which Appendix C in paragraph C-3320(b) requires a safety factor of 1.39 for circumferential flaws, and paragraph C-3420(a) requires a safety factor of 1.50 for axial flaws, both for emergency and faulted loads. These same safety factors are also permitted in Code Case N-513-1, which has been accepted by the NRC for evaluation of flaws.Considering the ASME Code references described above, a FS of 1.5 for design basis loadings in ductile materials provides an equivalent and acceptable level of safety as compared to the plant design basis and permitted methodologies for evaluation of flaws.5.4 MONITORING:

The proposed alternative assessment methodology requires periodic monitoring to assure that the assumptions of the assessment remain valid. This monitoring will be in addition to the normal daily plant operator rounds during which personnel are observant for signs of leakage. The monitoring will be by visual observation of the appearance of the joint and its leak rate. The frequency of the monitoring will be approximately once every three months, not to exceed 120 days between observations.

The monitoring will continue as described until the joint is repaired or replaced.

If there are changes in the nonconforming condition of an evaluated brazed joint with weepage that may impact its assessment for adequate structural integrity or its functionality, a Condition Report will be generated in accordance with the Millstone Power Station Corrective Action Program and the UT readings on the joint will be repeated and reassessed.

Monitoring Basis: The degree and frequency of periodic monitoring is conservative because the braze defect that permits this form of leakage stems from original construction, or fabrication, and is not the result of a progressive degradation mechanism.

Conditions that are applicable to the use of this methodology stem from defects in braze material inside a socket joint and will have a very low leak rate. Leakage is commonly considered weepage, at drops per minute or simply the appearance of moisture and salt deposits.In MPS3 operating experience, there have been no conditions where the piping disengaged from brazed fitting sockets. Consequently, no conditions have been observed that would have impacted the ability to maintain adequate system flow. This positive operating experience is due to the inherent structural integrity of brazed joints in service water systems.To further address the potential for degradation, a search and review of external operating experience was performed.

Braze failures in closed loop and electrical cooling systems such as generator stator cooling have been attributed to corrosion.

However, there was no operating experience indicating progressive failure for open loop seawater systems.To confirm the conclusion that no progressive failure mechanism applies, DNC had two specimens that had already been removed from Millstone Power Station seawater service, Page 8 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued) and that were reported to have low bonding, disassembled and examined.

The surface examination of the separated fitting and pipe surfaces did not reveal evidence of braze metal corrosion product. Since these examined joints are typical of plant construction and have seen nearly 20 years of service with no degradation of the bond, it is concluded that periodic visual monitoring of leak rate for this condition is acceptable, and monitoring may be scheduled on a quarterly basis. The specified response to altered conditions such as increased weepage will ensure that degradation to system functional margins does not occur.5.5 REPAIR / REPLACEMENT:

If the assessment can conclude that a brazed joint with leakage retains adequate structural integrity and functionality, an operability determination can be used to document an operable but not fully qualified status. A timely repair/replacement activity can be planned, commensurate with safety, and in accordance with 10 CFR Part 50, Appendix B. Consistent with the Millstone Power Station Corrective Action Program, the permanent Code repair/replacement for this nonconforming condition will be considered timely when completed during the next cold shutdown of sufficient duration, or the next refueling outage, whichever comes first. However, a time frame for a repair/replacement activity that could exceed the next refueling outage interval will be explicitly justified in the operability determination depending on factors that can include the time required for design, review, approval, or procurement of materials, availability of equipment, or the need to be in a hot or cold shutdown mode to implement the action.If a joint does not have adequate bond by this assessment, the methodology for determining the adequacy of structural integrity of the joint is not applicable.

Prompt repair/replacement of the joint, or temporary non-Code repairs subject to NRC review and approval may still be an option, consistent with considerations in Regulatory Issue Summary 2005-20 for the resolution of degraded and nonconforming conditions.

Basis: The bases for continued operation prior to repair of the joint are: system functionality is maintained as justified in section 5.3.1 above, structural integrity of the joint is maintained as justified in section 5.3.4, and there is no progressive braze bond failure mechanism that would alter these conclusions over time. Compensatory actions for the condition are administratively controlled under the Millstone Power Station Corrective Action Program. These include but are not necessarily limited to the periodic monitoring of leakage for the condition or housekeeping measures to contain weepage from affected piping. The application of this methodology will be consistent with considerations of Regulatory Issue Summary 2005-20 for the resolution of degraded and nornconforming conditions.

The permanent repair/replacement of the brazed joint assessed using this methodology will be in accordance with ASME Code,Section XI, 1WA-4000.Page 9 of 10 10 CFR 50.55a Request Number IR-3-04 (Continued)

5.6 AUGMENTED

EXAMINATION:

Up to five similar brazed joints will be selected for augmented leakage examination.

The additional joints will be selected based on consideration of adjacency, opposite train, fitting type, or other factors that may be evident from the specific condition.

Selection of fewer than five joints for an augmented examination is acceptable if the population of similar joints not previously examined is fewer than five. If leakage is observed in similar joints, the resolution of each nonconforming condition will be evaluated in accordance with the Millstone Power Station Corrective Action Program, and the extent of condition will be documented and addressed.

Basis: The examination of the additional joints is consistent with current practice for the resolution of degraded and nonconforming conditions, (e.g., application of ASME Code Case N-513-2).

Augmented examinations provide information regarding the extent of condition being evaluated and are consistent with current Millstone Power Station procedures for responding to leakage in service water piping.6. Duration of Proposed Alternative This proposal requests approval for the use of an alternative brazed joint assessment methodology for the third 10-year Inservice Inspection (1SI) interval, which starts on April 23, 2009, and is scheduled to be completed on April 22, 2019.7. Precedents A similar request for relief was granted in the Second Interval (Relief Request IR-2-38) per letter 07-0153 dated February 28, 2007, ADAMS Accession No. ML070580514 Page 10 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A FIGURES DOMINION MILLSTONE POWER STATION UNIT 3 Attachment A Page 1 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued)

TABLE OF CONTENTS Figure 1: Typical Brazed Joint Configuration

........................................................

3 Figure 2: Equations for Brazed Joint Assessment

....................

....4 Comparison of Brazed Joint Load vs. Capacity Figure 3: Two Inch Couplings:

Fabricated Samples at (a) 30% (above) and ..... 5 (b) 60% bond Figure 4: Two Inch Joints: Two Fabricated Samples with 12% Bond ..........

6 Figure 5: Arc Segment Disbondment, (a) 90 (above) and ......................................

7 (b) 126 Degrees Arc Figure 6: Two Inch Braze Field Sample Test Curve ............................................

8 Figure 7: Three Inch Braze Field Sample Test Curve .............................................

9 1 Figure 8: Test Results for Specially Fabricated Joints ..........................................

10 Figure 9 Test Results for Joints Removed From Service ....................................

10 Attachment A Page 2 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued)

Weepage location Pipe 4-t t f t t t t t f t f Internal Fluid Pressure A jA Figure 1: Typical Brazed Joint Configuration Attachment A Page 3 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued)

Se S maxkbadj)(I)(2)Seq S Sp + Sdl + Ssse.+ Sdyn Sip = longitudinal pressure stre Sdl = deadload stress Ssse. = SSE seismic stress Sdyn = dynamic stress (if defin, Unintensified pipe stresses from Code qualification analysis it 2=ins "max Smax(badj) b dj (3)D = pipe outside diameter-ins = insert.depth of fitting socket excluding any insert grove Zpipe= piping section modulus Tmax = 5000 psi (maximum braze shear stress)baj = adjusted effective bond Figure 2: Equations for Brazed Joint Assessment, Comparison of Brazed Joint Load vs. Capacity Attachment A Page 4 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued)-A5 0X QA 2000 Coece 1~ 500 -.i~1 2, 4ý00i 0) 0. 3. 0.75 1 Z S 13 25 I., 75.: " Figure 3: Two Inch Couplings:

Fabricated Samples at (a) 30% (above) and (b) 60% bond Attachment A Page 5 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued) 40U00'215(k)I0MM) --Raw ICorrected 0 I) 1K2 0.5 (:,75 1 1,25 1-5 75LT~-s HI50-5Z 0;5 75 1 25 I~ 75 I Figure 4: Two Inch Joints: Two'Fabricated Samples with 12% Bond Attachment A Page 6 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued) 5000 Raw 3500 400~ ~~0 -C rec 05(XI, a 0 1-sx -/..." 43;50 I I:,w" 15I ýI) 1 L)Le.pla~ccelu 1 1 21 Figure 5: Arc Segment Disbondment, (a) 90 (above) and (b) 126 Degrees Arc Attachment A Page 7 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued) 40}.N t'L.30 20 10 2. Coupling Joilnit371,127%uboaid 0 0.255 M.5 0.75 1 1.25: 1.-5 .L75 2 2" Field Samples Dis [lA kment (in)Figure 6: Two Inch Braze Field Sample Test Curve Attachment A Page 8 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued) 40 30 I..CE 4A-~ 31A 10 -....J2, 4.5%-13,47%-J4A, 15%-19, 38%-J9j, 48%131A, 21%boiid readings 0 0.1 3" FieldSamples 0.2 0.3 0.4 0.5 Displacenwnt (in)Figure 7: Three Inch Braze Field Sample Test Curve Attachment A Page 9 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment A (Continued) 0 U)U)U)I.'.45-40-35 30 25-20 15-_________~.

No bond failure, pipe icollapse governs H.! No bond failure, pipe ..................................( collapse.governs 1 l 2" test C 3" test 2.4 Sh Limit..2" Shear Limit.-.- 3" Shear Limit S10 5 1~0% 20% 40% 60% 80%,100%Percent True Bond Figure 8: Test Results for Specially Fabricated Joints 40 j 35 bond failure, pipe A 2" Field Samples Gcollapse governs* 30 -! 0 3"Field Samples 20. -3" Shear Limit C, 10____-- -2.4 Sh Limit:.. ..............

.....'..0% 20% 40% 60% 80% 100%Adjusted Percent Bond Figure 9 -Test Results for Joints Removed From Service Attachment A Page 10 of 10 10 CFR 50.55a Request Number IR-3-04 Attachment B BRAZED JOINT CONFIGURATION AND MATERIALS DOMINION MILLSTONE POWER STATION UNIT 3 Attachment B Page 1 of 3 10 CFR 50.55a Request Number IR-3-04 Attachment B (Continued)

BRAZED JOINT CONFIGURATION AND MATERIALS 1.0 MATERIALS:

Typical materials of construction of brazed piping are copper-nickel (SB-466) or nickel alloy (SB-165) annealed piping, and cast bronze fittings and valves (SB-61 or SB-62) dimensioned to MIL-F-1183.

The brazing alloy is SFA 5.8 BAg-l, BAS-la, or BAg-7. Construction Code minimum properties of the piping and fitting materials are: Material Item Sh, ksi Yield, Ultimate, ksi ksi SB466 Pipe 8.7 13 38 CDA706 SB-165 Pipe 17.5 28 70 SB-61 fitting 8.5 16 34 SB-"62 fitting 7.5 14 30 2.0 CONFIGURATION:

As shown in Figure 1 of Attachment A, a typical brazed joint fitting has a deep socket for inserting the pipe. Although it appears similar to a socket welded joint, the fabrication and structural behavior are quite different.

Whereas the socket weld achieves its joint strength by a fillet weld, resulting in fusion of similar material between the pipe and the outer face of the fitting, the braze achieves its strength by surface bonding of the outside of the pipe to the inside of the fitting socket using a dissimilar metal braze filler of silver alloy. The resulting braze filler metal is very thin (approximately I to 5 mils). The load transfer between pipe and fitting is thus primarily by shear through the braze filler. It is noted that there is no inherent stress concentration factor like that normally applicable to socket welds because there is no significant pipe wall bending induced by the shear load transfer over a length that is several wall thicknesses long.The following has been excerpted from a standard piping handbook.4 The length of lap in a joint, the shear strength of the brazing alloy, and the average percentage of the brazing surface area that normally bonds are the principal factors determining the strength of brazed joints. The shear strength may be calculated by multiplying the width by the length of lap by the percentages of bond area and by taking into consideration the shear strength of the alloy used.For the standard braze joint fittings used at MPS3, the joint overlap is about four to one. The smallest overlap occurs in a 3 inch joint, with an overlap length of 3.6 times pipe wall thickness.

4 Crocker and King, Piping Handbook, 5 th Edition, McGraw-Hill Book Company, page 7-212 Attachment B Page 2 of 3 10 CFR 50.55a Request Number IR-3-04 Attachment B (Continued)

3.0 BRAZED

JOINT FUNCTIONAL CHARACTERISTICS:

Since the piping loads causing longitudinal stress in the pipe are all transferred by shear stress through the brazed bond, the shear stress in the brazed bond is directly related to longitudinal pipe stress divided by a factor equal to the overlap ratio. Thus for a fully bonded brazed joint, the shear stress is about one fourth of the piping longitudinal stress. If the bond is only 50 percent of maximum then the bond shear stress will be about half the piping longitudinal stress.Given that piping and brazing filler metals have similar strength, a brazed joint has more than enough residual strength to tolerate moderate bond imperfections.

Consequently, the joint is not the weak link in the piping assembly.Consistent with this inherent over-design of brazed joints, the Construction Codes, such asSection III of the ASME Code and ANSI B3 1.1, require only visual inspection of the resulting bond. ND-5360, Visual Acceptance Standards for Brazedp oints, states "Brazing metal shall give evidence of having flowed uniformly through a joint by the appearance of an uninterrupted, narrow, visible line of brazing alloy at the joint." Surface exams such as by liquid penetrant are not required.

Volumetric exams are not specified or even defined for brazed joints.If the lack of bond is severe then the brazed joint becomes the weak link in the piping assembly.It fails by shear failure of the brazed bond. Brazing with a lower level of bond may however still be acceptable if the piping design basis loads are low enough. A brazing material defect with weepage is not the result of a flaw in the pipe or fitting pressure boundary.

The pressure-retaining boundary retains its structural integrity.

Although the shear load transfer between the pipe and fitting is clearly a pressure boundary function, the brazing material functions more as a sealant between the connected components and less like a pressure boundary.With regard to structural integrity, imperfections in the sealant function of the braze material are permissible, provided its load transfer function retains adequate margin. Thus, because there is no direct degradation of the pressure boundary, the available flaw evaluation methodologies such as in ASME Code Case N-513-2 or Generic Letter 90-05, are not directly applicable.

In addition, the characterization of braze imperfections is very different from the planar flaws or loss of wall thickness that are addressed in ASME Code,Section III, IWA-3000.Attachment B Page 3 of 3 10 CFR 50.55a Request Number IR-3-04 Attachment C EXAMPLE STRUCTURAL ASSESSMENT DOMINION MILLSTONE POWER STATION UNIT 3 AttaChment C Page 1 of 5 10 CFR 50.55a Request Number IR-3-04 Attachment C (Continued)

Braze Bond Structural Assessment Joint 1A (examnple only)Part61 Basic Data (dashed boxes are inputs Inputs: Line No: Sys Function:

IA supply to XXX-IA Piping tso:'CP.0123456 Joint.t 1AI Side of Joint: "Upstream Jt. Orientation: ,Mark I is up hlputs: Pipe Dia' -2.375 in Norm. Well Thk 0.156 in Pipe Mat'.'SB 466 CDA 706 FiltingMat'lSBi lor.62 Ref: Bond Strength:

5,000 psi Bond Adjustment 10%Measured Ave. Bond 55% (calculated.

For bond measurements, see sheet 'UT Readings')

55 % >= 60% ? No, Detailed assessment required Part 2 Bond Data Summary Offsets based on adjusted bond: Dxx 0.098 in Dyy -0.205 in Doffset 0.227 in 19% of pipe radius Alpha 12.0 degrees -rotation angle of principal axes Calculated effective bond data are in principal axes system, and are based on adusted bond.Actual Adjusted BXX 58% 54%Byy 4% 43%Bbend 49% 43%Bpress 55% 50%Note: Plot Is figurative only, actual braze bond is cylindrical, not through-wall.

I ,5 6 17 Attachment C Page 2 of 5 10 CFR 50.55a Request Number IR-3-04 Attachment C (Continued)

Braze Bond Structural Assessment Joint 1A Part 3 Calculated Bond LoadCapability D tnom Pipe Z Linsert Smax(100%)

2.375 in 0.156 in 0.566 inW3 0.656 in 25,662 psi Lookup Tb: Linsert per MVISpec DAnom D.od Linsert 3/4 1i05 11/32 1 1.315 7/16 1.5 1.9 5/8.2 2.375 21132 2,5 2.875 25/32 3 3.5 53/64 Load Capability (Allowable Nominal Pipe Stress)(Based on bond levels from Part 2)Actual Adjusted Sxx 14,997 13,746 psi Syy 12,538 10,975 psi Sallow 12,538 10,975, psi stress based on shear allow, and percent bond Smaý4.adj)

=ba , g4' ( J' max Part 4 Pipe Stress Data Stress Catc NP-XI901 Rev] CCN Rev. 5 CCN 4 Line No: 3SWP-002-999-3 Sys Function:

A supply to XXX-1A Piping Iso: CP-0123456 Joint: 1A Pipe Dia 2.375 in Nom. Wall Thk 0.156 in Pipe Matl SB 466 CDA 706 Fitting Mat't SB 61 or 62 A.pressure 1.865 inA2 Z.pipe 0.566 inW2 In_ pu_.L --- --- --Stress 10 1 P -A Alt. Stfess Node In/a S1 Z )f" MAX p SIF Used: 2.1 "-= p Eff.Put. SIF 1.57 IP+S-S 1 9 'a+ i-~ -~ --b --- inptsPO Design Pressure; 100 pslg Max Op. Pressure I 100 psig ! Calculated Nominal Stresses SIp I 761 psi ISp_offset 75 psi Eq. 8 (P*DL) 2500 psi ' Sust'd 8 1830 psi Eq, 19(N/J), 45 0 0 psi , NU 9 3100 psi Eq. 9F (Design BasisO 1 6500 psi I Faulted 9F' 4370 psi Max Nominal 4370 psi Part 5 Structural Integrity Determination Joint 1A Joint Load Capability Design Basis Load 10,975 psi 4,370 psi (from Part 3)(krom Part 4)Check: 4,370 < 10,976 ==> Braze Is adequate for design basis loads Monitor until repair/replacement Attachment C Page 3 of 5 10 CFR 50.55a Request Number IR-3-04 Attachment C (Continued)

• Braze Bond Measurements Joint 1A Reading 1 2'3 4 5 6 7 9 10 11 12 13 14 15 16 17 18 19 20 Bond Adjustment 10%Angle Meas. Bond Adj Bond PlotValue Adj Plot 01 30%1 22% o.825 0.806 18. 40%, 33% 0.850 0.833.36! 40%! 33% 0.850 o.8 541 35%! 28% 0.838 0,819 721 700% 67% 0.925 o.917 90% so%, 44% 0.875 0.861 1081 80%o 78% 0.950 0.944 126j 90%j 89%, 0.975 0.972 144, 90%. 89% 0.975 0.972 162! .80 ! 78% 0.950 0.944 1801 20%1 11% 0.800 0.778 198' SO'N 44% 0.875 0.861 216! 80%9 78% 0.950 0.944 2341 70%1 67% 0.925 0.917'252i 50%j 44% 0.875 0.861 270i 50%' 44% 0.875 0.861 288, 40%! 33% o.850 0.833 3061 45%( 39% 0.863 0.847 324' 50%' 44% 0.875 0.861 3421 40%; 33% 0.850 0.833 R Rmin 1 0.75 Max Min 1 0.75 1 0.75 1 0.75 1 0.75 1 0.75 1 0.75 1 .0.75 1 0.75.1 0.75 1 0.75 1 0,75 1 0.75 1 0.75 1 0.75 1 0.75 1 0.75 1 0.75 1 0,75 1 0.75 Nreadings dTheta degrees 20 18 Ave Min Max 55% 50%20% 11%90% 89%Attachment C Page 4 of 5 10 CFR 50.55a Request Number IR-3-04 Attachment C (Continued)

Braze Bond Calculations Joint 1A SOsat .Nacn 10% 20 0 "eges 0.000 tad Equivalent bond based on measured bond readings, vWtout adjustment A"gte Meas. eond 0 30%18 40%3B 40%54. "35%72 70%80 50%t08 80%.126 90%144 90%162 80%180 20%It's .50%216 80%234 70%252 50%270 50%288 40%306 45%324 50%342 40%Bpre5%55'%an(tlheta) db0os 00caOs2 dbein00zs aln(thata) bWan. dbsin"20- o .3 0 300 0.000 0.000 0.0o 0.000 0.951 0.380 0382 0.118 0.309 0,124 0,038: 0.809 0,324 0.262 0.190 0.588 0 235 0.138 0.588 0.208 0.121 0,166 0.809 0.283 0.229 0.309 0.216 0.087 0.206 0.951 0.86 0.833 0.000 0.000 0.000 0,000 1.000 0.800 0,800.309: -0.247 0.076 -0.236 0.981 0.701 0,724-4,538 -0.429 0311 .0,428 0.809 0.728 0,89-0.809 -0.720 0.580 .0.428 0.588 0.529 0,311-0.951 -V.71 0,724 -0,235 0.309 0.247 0,078-1.000 .020 0200 0.00D 0.0 0.000 0.000-0.951 -0.476 0.452 0,147 4.309 .0.165 0.048-.809 .0.647 0.524 0.380 -0.588 -0.470 0.276-0,58 -0.411 01242 0,333 ,-0.89 .0.866 0.438.0,309 -..155 0,O48 O.147 -0.051 .0.478 0:452 0,000 0.000 0.000 0,000 1-.000 .0.500 0.8O0: 0.309 0.124 0,038 -0,118 .0.951 .0.380 0.382 0.588 0,265 0.155 -0,214 -0.80 .0.364 '0298: 0.609 0.405 0.37 -0,238 .0.588 -.294 0.173 0.91 0.380 0.362 -0,118 .0.309 .0,124 0.030 0,000 0.0076 6.180 0 0,326 0.00I 0.037 5.840 checkw0 ry BpW Bpxy ewek-0 rx Bpxx ,.141 0.258 -0.016 0.550 0.068 0202 RofSet YOffsal BY' BWMyy S Xnofset BXX 0.18W -0,168 0.247 -0.011 01538 0.080 0.290 BByy 49% Bave 54% 5Bxx 5w%s5..p 0.244 sx8Lp 0.292 49% 58%

BxyO tan o0 2emrta sin 291pha lsn check alpha.0.043 -0.011 0.519 0,888 0.481 0.519 0.239 rad FALSE FALSE 13.7 deg M b-0 Equivalent bond based on adjusted bond readings Airo A.4. Bond cJtheta)e dcos dbecos^2 dbelncos ar(40eta) dbsin db'sln'2 0 22% 1.00 0.222 0.222 0.000 0.000 0.000 0.000 i8 .33% 0.051 0.3V7 0.302 0.098 0.309 0.103 0.032 38 33% 0.809 0.270. 0.218 0.159 0.588 0.196 0.115 54 28% 0.588 0.183 0.006 0,132 0.09 0.225 0.182 72 87% 0.309 0.206 0.084 0,198 0.951 0.634 0.403 90 44% 0.00 0.000 0.00 0.00 1.000 0,444 0.444 108 78% .0.309 .0.2.A0 0.074 -0.22 0,81, 0.740 0.704 128 40% -0,586 -0.822 0.307 -0,423 0.80 0.719 0.582 144 89% .0.80 -0.719 0.582 -.0423 0.588 0.D22 0.307 182 78% 40.951 -0.740 0.704 -0229 0.309 0.240 0.074 18o 11% .1.000 4.11 o.111 o.000 0.000 0,000 0,000 198 44% 40.951 -0.423 0.402 0"13A -0.309 -OA37 0.042 216 78% -0.80 29 0.509 0.370 .0.588 -0.457 0.269 234 .A6% `.58 4392 0,230 0,317 -4,0o .0.539 0.438 252 44% -0.309 .0.137 0.042 0.131 .-0,951 -0.423 0.402 270 44% 0.000 0,000 0.000 0.000 -1.000 4.444 0.44, 288 33% 0.309 0.103 0.032 0.049 -0.951 .-0317 0.302 306 39% 0.588 0.220 0.134 -0.185 4.009 4.315 0.285 324 44% O.J90 .0.380 0.291. 0.211 -0.588 -0.28t 0.154 342 33% 0.951 0.317 0.302 -0098 -0.309 -0.103 0.032 0.80 -0.086 4.622 -0,32 0.000 8378 BmIM chack=0 ly em Bpxy chckx0 rx OBPX 60% 0.173 0.231 -.018 04500 0.083 Reffxel Yofxet Byy Bey Bwyl'Bxx Xoffet Bu 0.227 -0.205 0.216 -0,011 0,482 0.098 0,265 BByy 43% Bare 46% BBxx 53%8yy.. 0.214 SUp 0.268 43% S4%8&fS.s0 Oxy-0 Wta 2s o~s 2abhsin 2ilphtan Ched .Ik-0.049 -0.011 0,445 0.914 0.406 0.445 0.209 rad FALSE FALSE 12.0.deg Mehwu~d Oonda Adluste Bonds1 ROid Values Ceaiated at AOOI siet SM914d vauits cakulaleM eA -Offet aftol ycofsel Ryy X(Offe Rexj Yoffuel 8" xaffel BXXj 40.188 490% 0.08 58% -.0205 43% 0.098 63%1 Attachment C Page 5 of 5 10 CFR 50.55a Request Number IR-3-04 Attachment D MECHANICAL TESTS DOMINION MILLSTONE POWER STATION UNIT 3 Attachment D Page 1 of 4 10 CFR 50.55a Request Number IR-3-04 Attachment D (Continued)

MECHANICAL TESTS

1.0 BACKGROUND

The correlation developed by the testing conservatively determines a required bond level for a given intensity of joint loading. The results of these tests support the use of the comparison shown in Figure 2, Attachment A, for the structural integrity analysis.2.0 TEST SAMPLE DESIGNS The effort to empirically confirm required bond levels for varying intensities of joint loadings consisted of three separate series of mechanical tests: a) specially fabricated joints with a controlled average bond level, b) specially fabricated joints that had disbandment on a contiguous arc-segment of the joint, and c) field sample piping joints, salvaged from piping removed from the plant.All joints were tested in three-point bending with the brazed, fitting in the middle of the configuration.

2.1 Specially

Fabricated Joints With a Controlled Average Bond Level: By a combination of machining and use of insert-groove type fittings, a series of test joints were fabricated with equivalent bond levels of 12. 30. 40 and 60 percent. The machining removed only about 30 mils of pipe thickness so that piping strength was not significantly affected.

The samples were fabricated for 2-inch and for 3-inch joints.Three examples of each size and bond level were fabricated, for a total of 24 samples.(Of the 24 samples in this category, one of the 40 percent bond samples was subsequently found to have less than the fully intended bond and is excluded from the results.)2.2 Specially Fabricated Joints That had Disbondment on a Contiguous Arc-Segment of the Joint: These test items were intended to explore the effect of having a significantly non-uniform distribution of bond area around the circumference of the joint. Six samples were fabricated with disbandment segment angles of 36, 48, 72.90, 108 and 126 degrees. The average bond levels for these, assuming perfect bond except in the disbonded area, ranged from 90 percent down to 65 percent, respectively.

2.3 Field

Sample Piping Joints: These joints were salvaged from piping that were removed from the plant after about 20 years of service, and screened by Ultrasonic Testing (UT). Piping joints with the lowest of measured bond were selected for testing.Attachment D Page 2 of 4 10 CFR 50.55a Request Number IR-3-04 Attachment D (Continued)

The nine items selected for testing included the following:

Description Quantity 2 inch couplings 3 3 inch couplings 2 3 inch tee (run sides) 1 3 inch flanges 3 The couplings and the tee included two brazed joints subjected to test loads. The test flanges were mated to full strength flanges not under test.3.0 MECHANICAL TEST RESULTS The results from testing on each of the series of tests are described in the balance of this section.The referenced figures are included in Attachment A. A test report has been incorporated into the Millstone Station plant records.3.1 Specially Fabricated Joints With a Controlled Average Bond Level: For the intentionally disbanded joints, all joints with 30 percent or better true bond achieved full piping collapse strength with no failure of the bond. Refer to Figure 3. As testing of each joint continued above the piping collapse load, one of the 40 percent true bond joints had indications of bond failure. The 12 percent true bond joints all experienced bond failure before reaching piping collapse load, but still withstood a minimum of 37 percent of the piping collapse load. Refer to Figure 4. All test items achieved their test collapse load at a load well above that which would be predicted for a 5 ksi braze shear strength.3.2 Specially Fabricated Joints that had Disbondment on a Contiguous Arc-Segment of the Joint: From 36 through 72 degrees of segment disbondment, the test items all achieved full piping collapse load. The test items from 90 through 126 degrees disbondment exhibited progressively lower collapse load, as shown in Figure 4. At 126 degrees disbondment, the test item achieved about 60 percent of the piping collapse load. The load deflection curves for these joints did not exhibit any indications of bond failure, however at the extremes of deflection (well above the level that would be acceptable for application of this methodology) the higher angle joints were significantly distorted.

For such large levels of deflection it was apparent that the close mechanical fit-up of the pipe in socket configuration contributed to joint bending strength.

All test items achieved their test collapse load at a load well above that which would be predicted for a 5 ksi braze shear strength.Attachment D Page 3 of 4 10 CFR 50.55a Request Number IR-3-04 Attachment D (Continued)

3.3 Field

Sample Piping Joints: The field sample test items exhibited considerable variation in collapse load for roughly similar UT bond readings.

The variations were expected for the field samples. Figures 6 and 7 show the displacement load curve for the tested field samples. Bond failure limited the collapse load in the two-inch Joints 37 and 39, and the three-inch Joints 3 and 9. The load curve for Joint 9 has a slight discontinuity at 11.9 ksi that is conservatively considered to indicate initial bond failure, even though the load continues above this point. The collapse load for other samples was limited by the piping collapse load, which is equivalent to about 21 ksi. Even with the low UT bond readings the field samples developed at least 50 percent of the piping collapse load. The higher than expected collapse load for some of the three-inch joints is believed to be partly due to the thickness of filler metal present as a fillet at the face of some of the joints. All test items achieved their test collapse load at a load well above that which would be predicted for a 5 ksi braze shear strength and the adjusted percent bond used in this methodology.

The adequacy of the 5 ksi shear stress assumed in the methodology in Equation 3 of Figure 2, Attachment A, for estimating joint strength is confirmed by the testing margins shown in the following table.Table 1: Test Load vs. Bond Shear Capacity Average Adjusted Test Shear Test /Test Joint AdjUT Collapse Capacity Shear Load, ksi Load, ksi Margin 36 65 61 22.8 15.8 1.44*37 27 19 11.6 4.9 2.41 39 55 50 19.6 13.0 1.52 2 45 39 27.3 9.0 3.02*3 47 41 22.6 9.6 2.38*4A 15 5 27.3 1.3 23.59*9 38 31 11.9 7.2 1.69 91 48 42 28.6 9.8 2.95*31A 21 12 32.0 2.8 11.61** Piping collapse load reached before bond failure or deflection run out.The data in Table 1 are plotted in Figure 9, Attachment A. Of the joints that were limited by bond failure prior to reaching piping collapse load, the minimum margin factor was 1.52. This minimum margin appears in Joint 39, with a 50 percent adjusted average bond. Review of detailed bond readings around the circumference of Joint 39 gives an equivalent adjusted bond of 43 percent for the bending axis used during the test, corresponding to a margin factor of 1.74 for this test case.Attachment D Page 4 of 4 10 CFR 50.55a Request Number IR-3-04 Attachment E ULTRASONIC TEST PROCEDURE Provided For Reference Only (subject to change)DOMINION MILLSTONE POWER STATION UNIT 3 (Attachment E Page 1 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

Non Destructive Examination Procedure AMll Millstone Station ULTRASONIC EXAMINATION PROCEDURE FOR EXAMINATION OF BRAZED JOINTS -MILLSTONE UNIT 3 SERVICE WATER PIPING MP-UT-45 Rev. 000-01 Approval Date: Effective Date: 07/24/07 07/31/07 Attachment E Page 2 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

Ultrasonic Examination Procedure for Examination Of Brazed Joints -Millstone Unit 3 TABLE OF CONTENTS 1 P U R P O S E ...................................................................................................

..3 2 P R E R EQ U IS IT E S .........................................................................................

3 3 D E F IN IT IO N S ..............................................................................................

..4 4 IN ST R U C T IO N S ...........................................................................................

4 4.1 Exam ination Preparartion

........................................................................

4 4.2 Exam ination M ethod ...............................................................................

4 4.3 Required Docum entation .......................................................................

6 5 R EV IEW A N D SIG N-O FF ..............................................................................

7 6 R E F E R E N C E S ............................................................................................

..7 7

SUMMARY

OF CHANGES .....................................

7 Figure 1 ....................................................

8 Attachment 1 -Ultrasonic Calibration Data Sheet ...........................................

9 Attachment 2 -Brazed Joint Sketch Sheet Sheet ....................

.......................

10 Attachment E Page 3 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

1. PURPOSE 1.1 Objective This procedure describes equipment and procedures that shall be used in the ultrasonic inspection of brazed pipe joints.1.2 Applicability 1.2.1 This procedure contains all the specific application requirements for the examination of Millstone Unit 3 service water system brazed joints to determine percentage of bonded areas.1.3 Discussion 1.3.1 In ultrasonic examination of brazed pipe joints, ultrasonic waves are transmitted from a search unit into the brazed joint to determine the amount of braze bond present beneath the search unit.1.3.2 Brazed joints shall be examined by the straight-beam (compressional wave)method as illustrated in Figure 1. Signals, if present along the base line, occur successively (reading from left to right) from the following sources; the insert groove (if present), the fitting inside diameter, the pipe inside diameter and possible multiple reflections.

1.3.3 To examine a brazed joint, the transducer is placed over the bonded area of the joint and moved around the circumference in increments and in a number of passes determined by the number of lands, land or engagement area width and the crystal size. The percent of bond and pattern are determined for each increment, land or pass and the total joint.2. PREREQUISITES

2.1 General

2.1.1 The outer surface of the fitting socket shall be prepared sufficiently to obtain satisfactory sound transmission and shall not be rounded in the longitudinal direction and should be relatively parallel to the pipe surface.2.1.2 For joint configurations that cannot be satisfactorily ultrasonically examined, this procedure is not applicable.

2.2 Personnel

Requirements 2.2.1 Only Level II, or Level III personnel may independently perform, interpret, evaluate and report examination results.2.2.2 Levels II and III shall be certified in accordance with Reference 6.1.2.2.3 The UT examiners shall have sufficient knowledge and training to determine ultrasonically the bond in brazed joints.2.2.4 UT examiners shall demonstrate ability to recognize such technical deficiencies as insufficient beam penetration (transmission), poor transducer contact and interfering contact surface roughness from patterns displayed on the ultrasonic screen.Attachment E Page 4 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E ,(Continued) 2.2.5 UT examiners shall maintain proficiency for examination of brazed joints by performing an examination of a brazed joint at least every six months.2.2.6 Examiners who do not meet the requirement of 2.2.5 above shall demonstrate their ability to examine brazed joints prior to performing examinations in the field. See Table 1 for initial examiner qualification and proficiency requirements.

2.3 Measuring

and Test Equipment 2.3.1 All measuring and test equipment shall meet the requirements of WC-8.2.4 Examination Limitations

2.4.1 Examiners

shall identify potential examination coverage limitations prior to performing the examination.

3. DEFINITIONS 3.1 Face of Fitting -The annulus surrounding the socket end.3.2 Insert Groove -The groove in the fitting socket prepared to contain the brazing alloy ring.3.3 Land, Fitting -That portion of the fitting on the side of the insert groove nearest the middle of the fitting.3.4 Land, Center -That portion of the fitting between the grooves in a multiple insert fitting.3.5 Land, Pipe -That portion of fitting on the side of the insert groove toward the end of the fitting.3.6 Examiner -A person that has sufficient knowledge in determining bond.3.7 Level III Examiner -The person in charge of ensuring examiners are qualified and have sufficient knowledge in determining bond.4. INSTRUCTIONS

4.1 Examination

Preparation

4.1.1 After

preparing the surface of the fitting, lay out the circumference as follows: a) Marking shall be accomplished using a permanent marker on the fitting surface, in increments not exceeding one inch. If the joint is to be re-examined, vibro-etching may be advisable but is not mandatory b) Markings shall be numbered clockwise as viewed facing the fitting from the pipe.4.2 Examination Method 4.2.1 The straight beam longitudinal wave method shall be used.4.2.2 The position of reflections along the base line of the viewing screen shall be indexed for signals from an insert groove, the inside diameter of the fitting, and the inside diameter of the pipe.Attachment E Page 5 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued) 4.2.3 For fittings containing insert grooves, place the transducer so that the active area is over one land only. Mark the first back reflection of the insert groove, inside diameter of fitting (no bond) and the inside diameter of the pipe (bond) at the left edge of the signal, on the face of the viewing screen. If necessary, check the back reflections with the reference calibration standard to ensure positive signal identification.

4.2.4 The amplitude of any one signal shall not reach a saturation point on viewing screen presentation.

4.2.5 For fittings which contain no insert grooves, place the transducer so that the active area: covers 1/2 of the OD of the fitting in the engagement area.4.2.6 Reflection markings and scope presentations will be as above except there will be no ring groove signal.4.2.7 The continuous or static scan technique shall be used.4.2.8 In the continuous scan, the transducer is moved in a continuous movement from one increment mark to the next increment mark. The bond and no-bond signals are mentally averaged while scanning the increment.

The bond for the increment is estimated to the nearest five percent in accordance with 4.2.9 through 4.2.11.4.2.9 In the static scan, the transducer is placed on the increment mark. The bond and no-bond signals are recorded for the increment.

The bond for the increment is estimated to the nearest five percent in accordance with 4.2.9 through 4.2.11.4.2.10 Readings for joints with inside pipe diameters less than 1-1/2 inches shall be taken at four equally spaced intervals in the increment, and for joints with inside pipe diameters greater than 1-1/2 inches, the readings shall be taken at three equally spaced intervals in the increment.

4.2.11 These increments shall be measured on the outside diameter of the fitting.4.2.12 Bond indications shall be recognized as to the percentage of bond without actually referring to the formula:% bond = 100 (bond amplitude (bond amplitude plus no-bond amplitude)

Attachment E Page 6 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued) 4.2.13 Increments for which no ultrasonic reading can be obtained shall be marked as follows: a) "X" -Increments which are inaccessible due to fitting configuration.

b) "NA" -Increments which are inaccessible due to piping, configuration or location.c) "NP" -Increments in which there is a lack of ultrasonic penetration.

d) Increments of the above type shall be assigned percent bond values as follows: "NA" = 0% bond"NP" and "X" = Increments up to a total length not exceeding 20 percent of the circumference of the land shall be assigned a percentage bond value equal to that of the lowest readable increment adjacent to the "X" or "NP" increments or 60 percent whichever is the least."X" or "NP" increments in excess of 20 percent of the circumference shall be assigned a bond value of 0 percent.The examiner may, at his discretion, shift the incremental scale so that the minimum number of increments contain "X", "NP" or "NA" values.NOTE: Within the 20 percent limitation, two or more adjoining "X" and/or "NP;' increments are considered a group of increments if the average of the remaining increments is 60 percent or more.The outermost two of any group within the 20 percent maximum limitation shall be rated on the basis of the adjacent readable increment.

The inner increments of the group shall be assigned a zero value for calculation purposes.4.2.14 The bond for the land (or pass of a no insert fitting) is the average of the readings for all increments in the land.S4.2.15 The percentage bond for the joint is that percentage of the total design faying surface which is bonded.4.3 Required Documentation 4.3.1 The UT calibration data shall be documented on Attachment 1.4.3.2 A sketch for each component detailing the increment locations shall be documented on Attachment 2.Attachment E Page 7 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

5. REVIEW AND SIGN-OFF The intent of this section is to clarify who is responsible to sign off on the examination data sheet.5.1 The Examiner shall print name, sign, and date the data sheet. The examiner shall then submit the completed data sheet to the appropriate reviewer.5;2 Reviewer's sign-off box can be signed only by Dominion Level II or III personnel (or their designee's) certified in the ultrasonic method.5.2.1 Review of the data sheet is intended to provide reasonable assurance of accuracy, thoroughness and procedure compliance.

5.2.2 The reviewer should compare the examiners data sheet against the AWO and other known parameters of the component(s) being examined.5.2.3 Review of the examination data sheet shall take place as soon as possible, and prior to the close-out of the AWO. The examination data sheet shall then be forwarded to the appropriate AWO package and/or job supervisor.

6. REFERENCES 6.1 ANSI/ASNT CP-1 89, 1991 Edition 6.2 WC-8, "Control and Calibration of Measuring and Test Equipment" 6.3 Granted Relief Request IR-2-38, "Structural Integrity Assessment Methodology for Brazed Joints (TAC NO. MC8893) -Millstone Power Station, Unit No. 3 7.

SUMMARY

OF CHANGES 7.1 Revision 000-01 7.1.1 Deleted paragraph

1.2.1 which

stated that procedure was for Engineering use only until NRC acceptance of relief request.7.1.2 Added paragraphs 2.2.5 and 2.2.6 to address proficiency of examiners.

7.1.3 Added

Table 1 which establishes frequency, number of samples required, and acceptance criteria for initial qualification, maintenance of proficiency, and requalification.

Attachment E Page 8 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

Table 1 Brazed Joint Examiner Qualification Qualification Type No. of Samples Period of Qualification Initial Qualification 6 3 Years Proficiency 3 6 months Requalification 6 3 Years Acceptance Criteria: Initial Qualification:

The percent bond of the six test specimens as reported by the examinee shall be compared to the true bond and accepted on the following basis; the arithmetic average of the six test specimens shall not deviate by more than 8% from the true bond and no single specimen shall deviate by more than 15% from the true bond.Proficiency Maintenance:

The percent bond of the three test specimens as reported by the examinee shall be compared to the true bond and accepted on the following basis; the arithmetic average of the three test specimens shall not deviate by more than 15% from the true bond.Requalification:

Same as initial qualification.

Attachment E Page 9 of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

Figure 1 FITTING INSERT GROOVE/F GRAZING ALLOY ULTRASONIC WAVE&UNBOUNO REGION IN OUTER LAND Attachment E Page 10 of 12 10 CFR 50.55a Request Number IR-3-04.Attachment E (Continued)

ATTACHMENT 1 ULTRASONIC CALIBRATION DATA SHEET Page _ of Plant: Unit: AWO Number: Purpose: Cal Block Temp Cal Block Number Thermometer S/N & Due Date DWG No.Search Unit Instrument

& Settings 100 Manufacturer Mfg. / Model Style or Type Serial Number 80 Frequency Range Size & Shape Material Velocity 60 Mode T or C Delay Search Unit Angle Pulser 40 Measured Angle Reject Serial Number Frequency 90 Cable Type, Length Damping No. of Connectors Zero Value Pulse Rep Rate 2 4 6 8 10 Gain Setting Attachments (Check) Calibrations Time CRT Setup Inches Sketch Sheet Initial Calibration Metal Path Supplements Final Calibration Depth Final Calibration Couplant Data Brand Batch Number SAP Batch Mgmt. No.Component ID Component Type Comments Examiner (Print & Sign) Level Date Examiner (Print & Sign) Level Date Reviewer (Signature)

Level Date Attachment E Page 1I of 12 10 CFR 50.55a Request Number IR-3-04 Attachment E (Continued)

Attachment 2 Millstone Power Station I BRAZED JOINT SKETCH SHEET PAGE OF Examined by (print/sign)

_ Level Date Millstone Power Station Reviewer (sign) -Level Date Attachment E Page 12 of 12 10 CFR 50.55a Request Number IR-3-05 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level of Quality and Safety--1. ASME Code Components Affected ASME Code Class: Code Class 1

References:

WCAP 14572, Revision 1-NP-A, Second Interval Relief Requests IR-2-39 and IR-2-47 Examination Category:

R-A Item Numbers: R1. 11 (Safe End-to-Pipe Welds)Ri. 15 (Nozzle-to-Safe End Welds)Description:

Examination of Weld Overlays Components:

Dissimilar Metal Piping Welds with Alloy 82/182 Weld Metal and Adjacent Welds which have had a Full Structural Weld Overlay Applied. See Below for List of Welds.1. Weld No. 03-X-5551-X-T:

Weld overlay encapsulating Pressurizer surge nozzle-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-SL-FW-4).2. Weld No. 03-X-5641-E-T:

Weld overlay encapsulating Pressurizer spray nozzle-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-517-FW-12).3. Weld No. 03-X-5644-A-T:

Weld overlay encapsulating Pressurizer safety nozzle at 810 azimuth-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-516-FW-1)

4. Weld No. 03-X-5648-B-T:

Weld overlay encapsulating Pressurizer safety nozzle at 147'azimuth-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-516-FW-3)

5. Weld No. 03-X-5649-C-T:

Weld overlay encapsulating Pressurizer safety nozzle at 2120 azimuth-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-516-FW-5)

6. Weld No. 03-X-5650-D-T:

Weld overlay encapsulating Pressurizer relief nozzle at 2780 azimuth-to-safe end dissimilar metal weld and the adjacent safe end-to-pipe weld (Weld No. RCS-513-FW-1) 2.. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)Page 1 of 5 10 CFR 50.55a Request Number IR-3-05 (continued)

3. Applicable Code Requirement The inservice inspection of the subject welds was initially performed in accordance with ASME Section'XI, IWB-2500, Examination Categories B-F and B-J.An alternative to the ASME Section XI requirements for the inservice inspection of Class 1 piping, Category B-J and B-F welds was implemented during the second interval based on the Risk-Informed technology developed in accordance with the Westinghouse Owners Group Topical Report "WCAP 14572, Revision 1-NP-A". The request to use this alternative was submitted to the Nuclear Regulatory Commission on July 25, 2000 with approval received on March 12, 2002.During the second interval, full structural weld overlays were applied to the subject welds.Inservice inspection for the weld overlays was performed in accordance with approved relief requests IR-2-39 (for Weld No. 03-X-5641-E-T) and IR-2-47 (for the remainder of the listed weld overlays.4. Reason for Request Currently, there is no comprehensive criteria for a licensee to perform inservice examination of weld overlays applied as a repair or for preemptive measures due to susceptibility of the underlying weld to PWSCC.The applications of the weld overlays at MPS3 were one time Relief Requests in the second interval based on the guidance of Code Case N-504-2 for Relief Request IR-2-39 and Code Case N-740 for Relief Request IR-2-47. For the third interval, the subsequent examination of the weld overlays needs to be considered.

DNC proposes to combine the examination criteria for the weld overlays identified in Relief Requests IR-2-39 and IR-2-47 into a one examination criteria as described below.5. Proposed Alternative and Basis for Use 5.1 Each weld overlay has been examined once during the first or second refueling outage following application of the weld overlay. The weld overlay examinations showed no indication of crack growth or new cracking and will be placed into a unique population within the ISI Program to be examined on a sample basis. Twenty-five percent of this population shall be added to the ISI Program as new welds in accordance with IWB-2412(b).5.2 The 25% sample shall consist of the same welds in the same sequence during successive intervals to the extent practical (note that' all welds experience pressurizer temperatures).

5.2.1 ,These examinations may be deferred to coincide with the vessel nozzle examinations required by Category B-D.5.2.2 Examinations during future intervals may be deferred to the end of the interval, provided no additional repair/replacement activities have been performed on the examination item, and no flaws or relevant conditions requiring successive examination in accordance with Attachment 1 are contained in the mitigated weld.Page 2 of 5 10 CFR 50.55a Request Number IR-3-05 (continued) 5.3 The examinations shall be volumetric (ultrasonic) and shall meet the applicable requirements of Appendix VIII. The requirements for the examination volume and required thicknesses shall be as described in Attachment 1, Figures I (a) "Examination Volume in Full Structural Weld Overlays" and 1 (b) "Definition of Thickness t 1 and t 2 for Application of IWB-3514 Acceptance Criteria." 5.4 Acceptance Criteria 5.4.1 General 5.4.1.1 The volumetric examinations shall be evaluated by comparing the examination results with the acceptance standards in 5.4.2.5.4.1.2 Volumetric examination results shall be compared with recorded results of the preservice examination and prior inservice examinations.

Acceptance of welds for continued service shall be in accordance with 5.4.2.5.4.2 Acceptance 5.4.2.1 Acceptance by Volumetric Examination 5.4.2.1.1 A weld whose volumetric examination confirms the absence of flaws shall be acceptable for continued service.5.4.2.1.2 Flaws shall meet the acceptance standards of IWB-3514 or be accepted for continued service in accordance with 5.4.2.2 or 5.4.2.3.5.4.2.1.3 A weld with new planar surface flaws or unexpected or'unacceptable growth of existing flaws shall be accepted for continued service in accordance with the provisions of 5.4.2.2 or 5.4.2.3.5.4.2.2 Acceptance by Repair/Replacement Activity 5.4.2.2.1 A weld whose volumetric examination reveals a flaw not acceptable for continued service in accordance with the provisions of 5.4.2.3 is unacceptable for continued service until the additional examinations of 5.4.3 are satisfied and the weld is corrected by repair/replacement activity in accordance with IWA-4000.5.4.2.2.2 For weld overlay examination volumes (Figure 1(a)) with unacceptable indications in accordance with 5.4.2.3.2, the weld overlay shall be removed, including the original defective weld, and the weld shall be corrected by repair/replacement activity in accordance with IWA-4000.5.4.2.3 Acceptance by Evaluation 5.4.2.3.1.

Previously evaluated flaws that were mitigated by the full structural weld overlay need not be reevaluated nor have additional successive or additional examinations performed unless the previously evaluated flaws have grown or new planar flaws have been identified.

The flaw is not considered to have grown if Page 3 of 5 10 CFR 50.55a Request Number IR-3-05 (continued) the size difference is within the measurement accuracy of the NDE technique employed.5.4.2.3.2 A weld overlay whose volumetric examination detects planar flaw growth or new planar flaws that exceed the acceptance standards of IWB-3514 is acceptable for continued service without repair/replacement activity if the weld overlay meets the acceptance criteria of IWB-3600 and the additional examinations of 5.4.3 are performed.

If a planar flaw is detected in the outer 25% of the original weld/base metal thickness for the examination volume it is acceptable for continued service if the crack growth calculations and structural design and sizing calculations required for original weld overlay acceptance show or are revised to show acceptability of the detected flaw. Any indication in the weld overlay material characterized as stress corrosion cracking is unacceptable.

5.4.3 Additional

Examinations 5.4.3.1 Examinations of additional weld overlays during the current outage are required if unacceptable planar flaws are detected in the weld overlay thickness, or if this examination reveals crack growth into the examination volume larger than predicted by the previous 5.4.2.3 analysis.

The number of additional weld examinations shall be equal to the number of overlaid welds originally scheduled to be performed during the present inspection period.5.4.3.2 If the additional examinations required by 5.4.3.1 reveal unacceptable flaws (5.4.2.3.2), the remaining weld overlays shall be volumetrically examined during the current interval.6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which*begins on April 23, 2009, and is scheduled to end on April 22, 2019.7. Precedents This is a first time request and DNC knows of no known examples of licensees applying to use the criteria in N-770 for the inservice inspection of weld overlays at this time. This request is being submitted because of the need to apply consistent examination requirements for weld overlays within the Third Inservice Inspection Interval.

The alternative requirements proposed in this request are derived from those in Code Case N-770,"Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1." Code Case N-770 has been approved by ASME (ASME C&S Connect Record No. 08-9).Only those requirements'pertinent to the inservice inspection of full structural weld overlays were used (Code Case N-770, Table 1, Item F).Page 4 of 5 10 CFR 50.55a Request Number IR-3-05 (continued)

8. References 8.1 2004 Edition, No Addenda, ASME Code,Section XI.8.2 ASME Code Case N-770, "Alternative Examination Requirements and Acceptance Standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without Application of Listed Mitigation Activities,Section XI, Division 1" (Approved by ASME January 26, 2009).8.3 Dominion Request for Relief IR-2-39, Revision 1 "Use of Weld Overlay and Associated Alternative Repair Techniques", dated October 19, 2005, ADAMS Accession No.ML052930108 8.4 NRC Letter, "Millstone Power Station Unit No. 3 -Issuance of Relief from Code Requirements (TAC NO. MC8609)", dated January 20, 2006, ADAMS Accession No.ML053260012

8.5 Dominion

Request for Relief IR-2-47, Revision 1 "Use of Weld Overlay as an Alternative Repair Technique", dated March 28, 2007, ADAMS Accession No.ML070880565 8.6 NRC Letter, "Request for Approval to Use IR-2-47 for Dissimilar Metal Weld Overlays as an Alternative Repair Technique (TAC NO. MD3379)", dated May 3, 2007, ADAMS Accession No. ML071210024 Page 5 of 5 10 CFR 50.55a Request Number IR-3-05 Attachment 1 Inservice Inspection Requirements For Full Structural Weld Overlay.(Figures 1(a) and 1(b) are shown on the next page.)(a) The weld overlay examination volume in Fig. 1(a) shall be ultrasonically examined to determine the acceptability of the weld overlay and to determine if any new or existing cracks have propagated into the outer 25% of the original weld or base material or into the overlay. The angle beam shall be directed perpendicular and parallel to the piping axis, with scanning performed in four directions.(b) The weld overlay shall meet the inservice examination standards of IWB-3514.

In applying the acceptance standards to planar indications, the thickness tl or t 2 , defined in Fig. 1(b), shall be used as the nominal wall thickness in 1WB-3514, provided the base material beneath the flaw (i.e., safe end, nozzle, or piping material) is not susceptible to PWSCC.For susceptible material, tj shall be used. If the acceptance standards of IWB-3514 cannot be met, the weld overlay shall meet the acceptance standards of IWB-3600.

Any indication characterized as. stress corrosion cracking in the weld overlay material is unacceptable.(c) As an alternative to (a), for inservice inspection, the weld examination volume in ASME Section XI, Figure IWB-2500-8(c) may be ultrasonically examined.

If cracking is detected extending beyond the weld examination volume, the weld examination of (a) and (b) above shall be performed to determine the acceptability of the weld overlay.(d) If inservice examinations of (a), (b), or (c) reveal crack growth, or new cracking in the weld overlay or outer 25% of original weld/base material meeting the acceptance standards, the weld overlay examination volume shall be reexamined during the first or second refueling outage following discovery of the crack growth or new cracking.

The weld overlay examination volume shall be subsequently examined two additional times at the period of one or two refueling outages, i.e., a total of 3 examinations within 6 refueling outages.(e) If the examinations.

required by (d) reveal that the flaws remain essentially unchanged for three successive examinations, the weld examination schedule may revert to the sample and schedule of examinations identified in 5.1.Attachment 1, Page 1 of 2 10 CFR 50.55a Request Number IR-3-05 Attachment 1 (Continued)

Minimum 112 in. (13 mrm)Minimum V12 in. (13 ram) (Ntle 1)t As-found Flaw Examination Volume A-B-C-D GENERAL NOTE: The weld includes the nozzle or safe end butter where applied.NOTE 1: For axial and circumferential flaws, the axial extent of the examination volume shall extend at least 1/22 in.(13 mm.) beyond the as-found flaw and at least 1/22 in. (13 mm.) beyond the toes of the original weld, including weld end butter; where applied, plus any PWSCC-susceptible base material in the nozzle and safe-end.Fig. 1(a): Examination Volume in Full Structural Weld Overlays ti Mininmm V2 irn. (13 rnni) See Figure 2(a) Note I Examination Volume A-B-C-D GENERAL NOTES: (a) The nominal wall thickness is tj for flaws in the examination volume A-B-C-D and t 2 for flaws outside examination volume A-B-C-D.(b) For flaws that are in examination volume A-B-C-D and extend outside this examination volume, the thickness t 1 shall be used.(c) The weld includes the nozzle or safe end butter, where applied, plus any PWSCC-susceptible base material in the nozzle and safe-end.Fig. 1(b): Definition of Thickness t 1 and t 2 for Application of IWB-3514 Acceptance Standards Attachment 1, Page 2 of 2 10 CFR 50.55a Request Number IR-3-06 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

--Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality or Safety--1. ASME Code Components Affected ASME Code Class: Code Class 2

Reference:

ASME Section XI, IWA-5244(b)

Examination Category:

C-H Item Number: C7.10

Description:

Class 2 buried segments of piping from the MPS3 Refueling Water Storage Tank (RWST) to the Quench Spray (QSS) and Safety Injection (SI) Systems, and from the Demineralized Water Storage Tank (DWST) to the AFW System Components:

See Applicable Line Numbers Below The applicable piping line numbers are identified as follows: " Line 3SIL-024-152-2

-One 24" (common train) suction line of the low pressure safety injection (SIL) / residual heat removal (RHS) pumps from the RWST that also provides suction to the high pressure safety injection (SIH) and charging (CHS) pumps.(Reference P&ID 25212-26912, Sheet 1)" Lines 3QSS-014-022-2/3QSS-014-026-2

-Two 14" suction lines for the "A" and "B" train QSS pumps from the RWST. (Reference P&ID 25212-26915)" Lines 3FWA-008-001-3

/ 3FWA-008-007-3

/ 3FWA-010-013-3

-Three separate 8" and 10" suction lines for the "A", "B" and swing train AFW pumps from the DWST.(Reference P&ID 25212-26930, Sheet 2)Excerpts of P&ID drawings are provided in Attachment 1 for information only.All of the subject buried segments are constructed of corrosion resistant Type 304 stainless steel. During original installation additional preventative measures were taken to mitigate corrosion and protect the external piping surfaces with the application of two coats of a silicone protective coating.2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)Page 1 of 4 10 CFR 50.55a Request Number IR-3-06 (Continued)

3. Applicable Code Requirement The 2004 Edition of ASME Section XI requires that for "Buried Components" the pressure test requirements of 1WA-5244(b) will be applied as follows: For buried components where a VT-2 visual examination cannot be performed, the examination requirement is satisfied by the following:

(1) The system pressure test for buried components that are isolable by means of valves shall consist of a test that determines the rate of pressure loss. Alternatively, the test may determine the change inflow between the ends of the buried components.

The acceptable rate of pressure loss or flow shall be established by the Owner.(2) The system pressure test for nonisolable buried components shall consist of a test to confirm that flow during operation is not impaired.(3) Test personnel need not be qualified for VT-2 visual examination.

4. Reason for Request An alternative is requested from performing the pressure testing using the pressure loss or change in flow methods described in IWA-5244(b)(1) for the buried piping segments of the Class 2 Quench Spray (QSS), Safety Injection (SI) and Auxiliary Feedwater (AFW) Systems.The alternative is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.The buried piping segments of the Class 2 QSS, SI, and AFW Systems are provided with a single, normally locked-open valve at their respective storage tanks. These are butterfly type valves for the QSS and SI piping from the Refueling Water Storage Tank (RWST) and gate type valves for the AFW piping segments from the Demineralized Water Storage Tank (DWST). These tanks are of such a large capacity that a small increase in level due to leakage through the tank boundary valve could not be detected.

A pressure decay test on these buried piping segments that do not have double isolation valves with a drain test connection to quantify internal valve seat leakage could not differentiate between boundary valve internal seat leakage and external pressure boundary leakage from the buried piping segment. To perform an accurate pressure drop test, extensive maintenance or system modification would be required.

For example, the storage tanks would need to be drained and additional valves installed, or the valves would need to be removed from the system and blind flanges installed.

The configuration of the piping segments do not provide for a sufficient straight length of pipe to properly install a flowmeter for accurate flow measurement at the storage tank ends of the buried pipe segments.

Therefore, it is not possible to compare a change in flow between the ends of the buried components.

There is no annulus provided in which the areas at the ends of the buried components could be visually examined and there is no access to the buried sections without excavation.

Page 2 of 4 10 CFR 50.55a Request Number IR-3-06 (Continued)

5. Proposed Alternative and Basis for Use 5.1 Proposed Alternative DNC proposes to use, as an alternative to the requirements of IWA-5244(b)(1), a verification of unimpaired flow in accordance with IWA-5244(b)(2) to provide an acceptable level of quality and safety. For each segment of the subject buried pipe, periodic flow testing will be performed in accordance with Inservice Test (IST) Program surveillance procedures.

These surveillance procedures require flow to be measured, recorded and compared to established acceptance criteria to provide the assurance that flow is not impaired during operation.

During each flow test, the pump draws suction from the storage tank through the associated buried sections of piping.5.2 Basis for Use Flow testing of the two QSS pumps is performed quarterly and will use the established minimum flow rate specified in the IST procedures as the acceptance criteria for the pressure testing of the associated 14" QSS buried pipe segments.

The flow rate is currently specified as 3,950 gallons per minute (gpm).Flow testing of the two RHS pumps is performed during each Refuel Outage and will use the established minimum flow rate specified in the IST procedures as the acceptance criteria for the associated single 24" SIL buried pipe segment. The flow rate is currently specified as 4,000 gpm.Flow testing of the three AFW pumps is performed each refueling outage and will use the established minimum flow rates specified in the IST procedures as the acceptance criteria for the associated 8" and 10" AFW buried pipe segments.

The flow rate of the two motor driven pumps will be used as the acceptance criteria for the 8" segments and the flow rate of the turbine driven pump will be used as the acceptance criteria for the 10" segment.These flow rates are currently specified as 490 gpm for each of the two motor driven pumps and 750 gpm for the turbine driven pump.If during the IST surveillances the minimum flows cannot be achieved, the pump(s)would be declared inoperable and a condition report initiated in accordance with the Millstone Power Station Corrective Action Program with further corrective actions as required to restore the pump(s) and/or system to an operable status.Additionally, the level in the RWST and DWST are monitored periodically to satisfy Technical Specification requirements.

In the case of the DWST, monitoring is performed once every 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in the case of the RWST, monitoring is performed once every seven days. The RWST and DWST also have a low level alarm in the control room.The existing tank isolation valves are administratively locked open during all modes of operation, thus the buried sections of piping are continuously exposed to the static head pressure of their respective storage tanks. Tank level losses due to leakage from buried piping would be promptly identified.

Page 3 of 4 10 CFR 50.55a Request Number IR-3-06 (Continued)

5.3 Discussion

on Acceptable Level of Quality or Safety or Hardship Without A Compensating Increase in the Level of Quality or Safety Based on the hardship described above and the use of inservice flow testing to verify that flow through the buried piping is unimpaired, and in conjunction with the periodic storage tank level monitoring, this alternative provides an acceptable level of quality and safety.The hardship is considered to be without a compensating increase in the level of quality or safety for the following reasons: " These surveillance procedures require flow to be measured, recorded and compared to established acceptance criteria to provide the assurance that flow is not impaired during operation." During each flow test, the pump draws suction from the storage tank through the associated buried sections of piping." The verification of unimpaired flow in accordance with IWA-5244(b)(2) provides an acceptable level of quality and safety considering the proposed alternative and its basis for use.6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.7. Precedents A similar request for relief of buried Service Water piping was previously approved for Byron Station, Units 1 and 2 (Relief 13R-07) and Braidwood Station Units 1 and 2 (Relief 12R-46) by letter dated January 16, 2007, ADAMS Accession No. ML063260074.(Reference TAC NOS. MD1757, MD1758, MD1759 and MD1760). In addition, a similar relief request for the second interval at Millstone 3 (IR-2-40) was approved in letter 08-0437, dated July 10, 2008, ADAMS Accession No. ML081720069.

Page 4 of 4 10 CFR 50.55a Request Number IR-3-06 Attachment 1 A to (A Lo- L V) 0 >-'I 2 >- LOCY) V) <F- < 0 o::o n~ m< 0 0' W M ~~3RCS C'u B C TSPB/ MB2: W) POWER M,2 ALOCAL CONTROL REFUELING WATER STORAGE TANK 3QSS*TKI(X-)

HS 3SIL*V1(Z-)

X""~LO 3-SIL-024-153-2(Z-)

ý3-SIL-024-151-2(Z-)

ESF BLDG I 3-SIL-008-154-2(Z-)

V/- 3-SIL-024-1-2(Z-) SEL-016-2-2(Z-)

MB ESF FUEL \0o BLDG BLDG HIGH PRESS SAFETY INJECTION 5"- / EM-113A A-3-3-SIH-008-85-2(Z-)

26913 SH 1 3-RHS-012-43-2(A-)

CC)0 LC D "-3--RHS-500 MLC>::-83-2(B- ) _ nCL 602 *v9(A-)I Enlarged Portion of P&ID 25212-26912, Slheetl 1 I_.I I ~ ~CL 1502 V1O(A- )SET AT 140 PSIG 3-RHS-C CL 60---CL 15('.-3-RHS-004-47-2(ATM STRUCT'LESF BUILDING (B-)37B Attachment 1, Page 1 of 3 10 CFR 50.55a Request Number IR-3-06 Attachment 1 (Continued)

YARDz_4 4 NSIDE (BURIED) PIT_T932 I *V87(X-IALV RWST VALVE PIT 2 MB2 I Ii~1i" TEST NOZZLES*3-QSS-750-802-2(X-) QSS-750-800-2(X-)

--" \ 3QSS-PIA 3QSS-PIB K REACTOR PLANT CHILLED WTR SYS I. OPEN 3CDS-SOV26

-- E-1 -F-).26922 SH I* 'LT933 3-QSS-750-803-2(X-)

REFUELING WATER STORAGE TANK*TKI(X-)Z2 AB2 A/S 3-QSS-750-801-2(X-)

3-QSS-014-21-2(A-)

?ESSURE INJECTION A B-2 2_SHI 3-QSS-014-19-2(8-)

\CjiZ. 3-QSS--006-64-2(Z-)

• 3-QSS-006-46-2(Z-)-IN VALVE PIT I I_I "~IN YARD i ....S (BURIED) I,-QSS-750-47-4--

3-QSS-150-51-4 RAACT LIQ WASTE& AERATED DRAINS MB2____________

26906 SH 3 REACTOR PLANT CHILLED WTR SYS CLOSE 3CDS-SOV26

---EýM-122A E-2 26922 SH I V A'*934(A- ) 32 I(A-) *v52(A--I I .v o I Enlarged Portion of P&ID 25212-26915 t...3-1 F Attachment 1, Page 2 of 3 10 CFR 50.55a Request Number IR-3-06 Attachment 1 (Continued)

STEAM GENERATOF EM-131A G-7 26931 SN. I I Attachment 1, Page 3 of 3 10 CFR 50.55a Request Number IR-3-07 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

--Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality or Safety--1. ASME Code Components Affected ASME Code Class: Code Class 3

Reference:

ASME Section XI, TWA-5244(b)

Examination Category:

D-B Item Number: D2.10

Description:

MPS3 buried pipin1g segments consist of two trains of buried 30" Service Water System (SWS) supply piping from the intake structure to the auxiliary building including branch lines to loads in the emergency diesel generator, control, and emergency safeguards buildings.

This piping is carbon steel material clad with copper nickel and encased in concrete.Components:

See Applicable Line Numbers Below The applicable piping line numbers are identified as follows (Reference P&ID 25212-26933, Sheets 2 and 4. Excerpts of P&ID drawings are provided in Attachment 1 for information only.): 3SWP-030-191-3

/3SWP-030-190-3 (sheets 2 and 4)3SWP-026-65-3

/3SWP-026-57-3 (sheets 2 and 4)3SWP-012-24-3

/3SWP-012-13-3 (sheet 4)3SWP-010-25-3

/3SWP-010-40-3 (sheet 4)3SWP-006-31-3

/3SWP-0067226-3 (sheet 4)2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)Page 1 of 4 10 CFR 50.55a Request Number IR-3-07 (Continued)

3. Applicable Code Requirement The 2004 Edition of ASME Section XI requires that for "Buried Components" the pressure test requirements of IWA-5244(b) will be applied as follows: For buried components where a VT-2 visual examination cannot be performed, the examination requirement is satisfied by the following:

(1) The system pressure test for buried components that are isolable by means of valves shall consist of a test that determines the rate of pressure loss. Alternatively, the test may determine the change inflow between the ends of the buried components.

The acceptable rate of pressure loss or flow shall be established by the Owner.(2) The system pressure test for nonisolable buried components shall consist of a test to confirm that flow during operation is not impaired.(3) Test personnel need not be qualified for VT-2 visual examination.

4. Reason for Request An alternative is requested from performing the pressure testing using the pressure loss or change in flow methods described in IWA-5244(b)(1) for the buried piping segments of the MPS3 SWS on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.The buried piping segments of the MPS3 Class 3 SWS piping are bounded by butterfly valves that are not designed or expected to provide an adequate leak tight boundary that is necessary for an accurate pressure decay test, and extensive maintenance or system modification would be required to conduct this test. For example, the valves would need to be replaced with valves that have better leakage control characteristics, or the valves would need to be removed from the system and blind flanges installed.

The configuration of the piping segments do not provide for a sufficient straight length of pipe to properly install a flowmeter for accurate flow measurement at the ends of the buried pipe segments.

Therefore, it is not possible to compare a change in flow between the ends of the buried components.

There is no annulus provided in which the areas at the ends of the buried components could be visually examined and there is no access to the buried sections without excavation.

5. Proposed Alternative and Basis for Use 5.1 Proposed Alternative DNC proposes to use, as an alternative to the requirements of IWA-5244(b)(1), a verification of unimpaired flow in accordance with 1WA-5244(b)(2) to provide an acceptable level of quality and safety. For each segment of the subject buried pipe, periodic flow testing will be performed in accordance with Inservice Test (IST) Program surveillance procedures.

These surveillance procedures require flow to be measured, Page 2 of 4 10 CFR 50.55a Request Number IR-3-07 (Continued) recorded and compared to established acceptance criteria to provide the assurance that flow is not impaired during operation.

5.2 Basis

for Use Flow testing of the four MPS3 SWS pumps is performed quarterly and will use the established minimum flow rate specified in the IST procedures as the acceptance criteria for the pressure testing of the associated SWS buried pipe segments.

The flow rate is currently specified as 8820 gallons per minute (gpm).If during the IST surveillances the minimum flows cannot be achieved, the pump(s)would be declared inoperable and a condition report initiated in accordance with the Millstone Power Station Corrective Action Program with further corrective actions, as required, to restore the pump(s) and/or system to an operable status.Additionally, internal visual inspection is performed periodically during plant refueling outages, of the accessible buried pipe segments that are 18 inches in diameter and greater, to ensure the piping, coating, or lining is not experiencing unacceptable degradation.

Discussion on Acceptable Level of Quality or Safety or Hardship Without A Compensating Increase in the Level of Quality or Safety Based on the hardship described above and the use of inservice flow testing to verify that flow through the buried piping is unimpaired, and in conjunction with the periodic internal Visual inspections this alternative provides an acceptable level of quality and safety.The hardship is considered to be without a compensating increase in the level of quality or safety for the following reasons: " These surveillance procedures require flow to be measured, recorded and compared to established acceptance criteria to provide the assurance that flow is not impaired during operation." Internal visual inspections ensure the piping, coating, or lining is not experiencing unacceptable degradation.

• The verification of unimpaired flow in accordance with 1WA-5244(b)(2) provides an acceptable level of quality and safety considering the proposed alternatives and their basis for use.6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.Page 3 of 4 10 CFR 50.55a Request Number IR-3-07 (Continued)

7. Precedents A similar request for relief of buried Service Water piping was previously approved for Byron Station, Units 1 and 2 (Relief 13R-07) and Braidwood Station Units 1 and 2 (Relief 12R-46) by letter dated January 16, 2007, ADAMS Accession No. ML063260074.(Reference TAC NOS. MD1757, M01758, MD1759 and MD1760). In addition, a similar relief request for the second interval at Millstone 3 (TR-2-41) was approved in letter 08-0437, dated July 10, 2008, ADAMS Accession No. ML081720069.

Page 4 of 4 10 CFR 50.55a Request Number IR-3-07 Attachment 1 L i .Ri INSIDE TO BL,-ABOVE FLOOR UNDER TIJR9 BLED z-3-SWP-026-12O-/

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DROEG CHWIIN%'RE14 CGNCRETE ENCA3EU I I I v v I v b I Aaý!ý ýCCCI C AAAAICRA--- -N N-- ---I-I-N-PIýHOS (-'I NC.-. 9i? CIRCULATING FIDllP (~}(~D ".y lG!Pl CI-TN ID I E SAG2i 1731 UAACR I 5.CNN CI ERC,4SE Attachment 1, Page 4 of 4 10 CFR 50.55a Request Number IR-3-08 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

--Alternative Provides Acceptable Level ,of Quality and Safety--1. ASME Code Components Affected ASME Code Class:

References:

Code Class 1 ASME Section XI, Table IWB-2500-1 ASME Section XI, Appendix VIII, Supplements 4 and 6, 10 CFR 50.55a(b)(xv).

Examination Category:

B-A Item Number:

Description:

Components:

B11.30 Implementation of Appendix VIII, Supplements 4 and 6 -Use of PDI Qualified Procedures, Personnel and Equipment for Non-Appendix VIII RPV Shell-to-Flange Weld Reactor Pressure Vessel (RPV) Shell-to-Flange Weld No. 101-121 2. Applicable Code Edition and Addenda ASME Section XI, 2001 Edition (No Addenda) per 1OCFR50.55a(b)(2)(xv) for Appendix VIII.3. Applicable Code Requirement The 2001 Edition with No Addenda of the American Society of Mechanical Engineers (ASME Code)Section XI, "Rules for Inservice Inspection of Nuclear Power Plant Components," Subsection IWA-2232, requires ultrasonic (UT) examinations be performed in accordance with Mandatory Appendix I. Paragraph 1--2110(b) of Mandatory Appendix I requires that examination of the RPV shell-to-flange weld to be in accordance With ASME Code,Section V, Article 4.Page lof 2 10 CFR 50.55a Request Number IR-3-08 (Continued)

4. Reason for Request The use of this alternative will allow the use of Performance Demonstration Initiative (PDI)qualified procedures for the performance of the ultrasonic testing examination of the reactor pressure vessel (RPV) shell-to-flange weld from the vessel side of the weld in accordance with ASME Code,Section XI, Division 1, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6. This alternative would be used in lieu of Article 4 of Section V requirements.

5. Proposed Alternative and Basis for Use DNC proposes to perform ultrasonic examinations of the RPV shell-to-fl ange weld using procedures, personnel, and equipment that have been demonstrated and qualified in accordance with ASME Section XI, 2001 Edition, No Addenda, Appendix VIII, Supplements 4 and 6 as amended by 10 CFR 50.55a and the PDI Program. Since the examination will be performed from a single side due to the weld configuration, all procedures, personnel, and equipment will be qualified for single sided access for examination of this weld.Appendix VIII requirements were developed and adopted to ensure the effectiveness of ultrasonic examinations within the nuclear industry by means of a rigorous, item specific performance demonstration containing flaws of various sizes, locations, and orientations.

The performance demonstration process has established with a high degree of confidence, the capability of personnel, procedures, and equipment to detect and characterize flaws that could be detrimental to the structural integrity of the RPV. The PDI approach has demonstrated that for detection and characterization of flaws in the RPV the ultrasonic examination techniques are equal to or surpass the requirements of the ASME Section V, Article 4 ultrasonic examination requirements.

Though Appendix VIII is not required for the RPV shell-to-flange weld examination, the use of Appendix VIII, Supplements 4 and 6 criteria for detection and sizing of flaws in this weld will be equal to or exceed the requirements of ASME Section V, Article 4. Therefore, the use of the proposed alternative will continue to provide an acceptable level of quality and safety, and approval is requested pursuant to 10 CFR 50.55a(a)(3)(i).

6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.7. Precedents A similar relief request (RR ISI-30) was previously approved for Union Electric Company for its Callaway Plant, Unit 1 on April 7, 2004 (ADAMS Accession Nos. ML032340608 and ML041 000516). In addition, Relief Request IR-2-44 was approved for Millstone 3 for use during the Second Interval in a letter from the NRC (Serial No. 07-0368) dated May 1, 2007 (ADAMS Accession No. ML070740465).

Page 2 of 2 10 CFR 50.55a Request Number IR-3-09 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(a)(3)(ii)

--Alternative Provides Acceptable Level of Quality and Safety and Compliance with the Specified Requirements Result in a Hardship without a Compensating Increase in the Level of Quality and Safety--1. ASME Code Components Affected ASME Code Class:

References:

Examination Category: Item Number:

Description:

Systems: Code Class 1 ASME Section XI, Table IWB-2500-1 and IWB-5222(b)

B-P B15.10 Use of Alternative Pressure Testing Criteria for the System Leakage Test Conducted at or Near the End of the Inspection Interval on Class 1 Piping Pressurizer Auxiliary Spray Reactor Head Vent Reactor Coolant Low Pressure Safety Injection High Pressure Safety Injection Residual Heat Removal 41 Reactor Coolant Pressure Boundary (RCPB) piping segments primarily consisting of small bore < 2" Nominal Pipe Size (NPS)piping vents, drains, and branch (VTDB) lines and connections.

Additional segments are portions of larger diameter piping 6", 8", 10" and 12" NPS located between check valves and isolated, required to be isolated at operation or otherwise continually under pressure and monitored for loss of pressure.

Details related to these piping segments are provided by Tables 1 through 5 in Section 4.0 and Attachment 1.Component(s):

2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement 1WB-2500, Table IWB-2500-1, Code Category B-P, Item Number B 15.10 requires that all Class 1 pressure retaining components be Visual, VT-2 examined each refueling outage. The required system pressure test can be either a hydrostatic test or a system leakage test. The system leakage test is performed at a pressure not less than the pressure corresponding to Page 1 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued) 100% rated reactor power. Per 1WB-5222(a), the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity.

Per IWB-5222(b), the pressure retaining boundary during the system leakage test conducted at or near the end of the interval shall extend to all Class 1 pressure retaining components within the system boundary.4. Reason for Request The 1WB-5222(b) system pressure test which is performed once per ten-year interval is the subject of this request. The 1WB-5222(b) test requires that portions of the Class 1 system which are not normally pressurized to the RCS pressure associated with 100% rated reactor power be pressurized to that pressure.

Reasons for this request are grouped in Component Groups 1 through 5 and described in Sections 4.1 through 4.5. The following is a summary list of names for these groups of piping segments.Group Piping Segment Summary Name 1 Small bore < 2" NPS piping vents, drains, and branch (VTDB) lines and connections in the Reactor Coolant and Reactor Head Vent Systems 2 Low Pressure Safety Injection (LPSI) header pipe segments 3 Safety Injection to RCS Cold and Hot Legs 4 Residual Heat Removal (RHS) Suction 5 Auxiliary Pressurizer Spray Page 2 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

4.1 Component

Group 1: "Small bore < 2" NPS piping vents, drains, and branch (VTDB)lines and connections in the Reactor Coolant and Reactor Head Vent Systems" -The component Group 1 piping segments 1-29 are shown in Table 1. Additional details are in Attachment 1, Table 1.Table 1: Affected Piping Segments of Group 1 Seg t(Description Segment Boundary1 Dia. Length______________eegetnondr (in) (ft)M3-1 RCS Loop 1 Fill Line 3RCS*V24 to 3RCS*V23 2 <1 (AV8036A)M3-2 RCS Loop 2 Fill Line 3RCS*V99 to 3RCS*V100 2 <1 (AV8036C)M3-3 RCS Loop 3 Fill Line 3RCS*V68 to 3RCS*V67 2 <1 (AV8036B)M3-4 RCS Loop 4 Fill Line 3RCS*V140 to 3RCS*V141 2 <1 (AV8036D)M3-5 Loop 1 Drains to Primary 3RCS*V202

/ 3RCS*V203 2 81 Drain Header to Line 3-RCS-002-148-1 M3-6 Loop 2 Drains to Primary 3RCS*V205

/ 3RCS*V206 2 65 Drain Header to Line 3-RCS-002-148-1 M3-7 Loop 3 Drains to Primary 3RCS*V208

/ 3RCS*V209 2 83 Drain Header to Line 3-RCS-002-148-1 M3-8 Loop 4 Drains to Primary 3RCS*V211

/ 3RCS*V212 2 137 Drain Header to Line 3-RCS-002-148-1 M3-9 Primary Loop Drain Header 3RCS*V213 to 3RCS*V198 2, 1 236 and 3RCS*V898 M3-10 Primary Loop Drain Header 3RCS*V895 to 3RCS*V898 1, 3/4 2 Drain and 3RCS*V899 M3-11 Loop 1 T-Cold Stop Valve 3RCS*V989 to flange 3/4 <1 Disk Pressure Connection M3-12 Loop 1 T-Cold Stop Valve 3RCS*V990 to flange 3/4 <1 Disk Pressure Connection M3-13 Loop 1 T-Hot Stop Valve 3RCS*V991 to flange 3/4 <1 Disk Pressure Connection M3-14 Loop 1 T-Hot Stop Valve 3RCS*V992 to flange 33/44 <1 Disk Pressure Connection M3-15. Loop 3 T-Cold Stop Valve 3RCS*V979 to flange 3/4 <1 Disk Pressure Connection M3-16 Loop 3 T-Cold Stop Valve 3RCS*V980 to flange 3/4 <1 Disk Pressure Connection M3-17 Loop 3 T-Hot Stop Valve 3RCS*V981 to flange 33/44 <1 Disk Pressure Connection M3-18 Loop 3 T-Hot Stop Valve 3RCS*V982 to flange 3/4 <1 Disk Pressure Connection Page 3 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

Table 1: Affected Piping Segments of Group 1 (Continued)

Segment Description Segment Boundary ) Dia. Length (in) (ft)M3-19 Loop 2 T-Cold Stop Valve 3RCS*V984 to flange 3/4 <1 Disk Pressure Connection M3-20 Loop 2 T-Cold Stop Valve 3RCS*V985 to flange 3/4 <1 Disk Pressure Connection M3-21 Loop 2 T-Hot Stop Valve 3RCS*V986 to flange 3/4 <1 Disk Pressure Connection M3-22 Loop 2 T-Hot Stop Valve 3RCS*V987 to flange 3/4 <1 Disk Pressure Connection M3-23 Loop 4 T-Cold Stop Valve 3RCS*V974 to flange 33/44 <1 Disk Pressure Connection M3-24 Loop 4 T-Cold Stop Valve 3RCS*V975 to flange 3/4 <1 Disk Pressure Connection M3-25 Loop 4 T-Hot Stop Valve 3RCS*V976 to flange 3/4 <1 Disk Pressure Connection M3-26 Loop 4 T-Hot Stop Valve 3RCS*V977 to flange 3/4 <1 Disk Pressure Connection M3-27 Reactor Vessel Head Vent 3RCS*V958 to flange 1 1 Line M3-28 Reactor Vessel Head Vent 3RCS*V956 to flange 1 1 Line Drain M3-29 Loop 1 TC Instrument Line 3RCS*V33 to 3RCS*V34 2, 3/4 3 and V35 NOTE: 1. The segment boundaries are described in terms of valve-to-valve unless otherwise annotated for a flange or line designation.

Each of these VTDB lines and connections are equipped with manual valves, which provide double isolation of the RCPB. These valves are generally maintained closed during normal operation.

The piping outboard of the first isolation valve is not normally pressurized.

Under normal operating conditions, these VTDB lines and connections, except for the low pressure safety injection (LPSI) VTDB lines and connections, are subject to reactor coolant system (RCS) pressures and temperatures only if leakage through the inboard valves occurs. For the LPSI VTDB lines and connections, leakage at inboard valves will only result in pressures associated with the pressure of the safety injection tanks.Because these VTDB lines and connections typically do not have test connections that would allow them to be individually pressure tested without design modifications, it will be necessary to open the inboard valves to pressurize these VTDB lines and connections to perform the IWB-5222(b) system pressure test. Pressurization by this method defeats the double isolation feature and presents significant personnel safety concerns for the Page 4 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued) personnel performing the test on the valves that are at normal RCS pressure and temperature.

Performing this test with the inboard. isolation valves open requires several man-hours to position or cycle these valves for the test and restore the valves after the test is complete.Most of these valves are located in close proximity to. the RCS loop piping and thus require personnel entry into high radiation areas within the containment.

Based on previous outage data, estimated radiation exposure associated with valve alignment and'realignment would result in an additional 1.9 man-Rem. Since this test would be performed near the end of an outage when all RCPB work has been completed, the time required opening and closing the valves on these VTDB lines and connections could impact the outage schedule.4.2 Component Group 2: "Low Pressure Safety Injection (LPSI) header pipe segments" -The component Group 2 piping segments (30 through 33) are shown in Table 2.Additional details are in Attachment 1, Table 2.Table 2: Affected Piping Segments of Group 2 Segment Description Segment Boundary Dia. Length (in) (ft)M3-30 Loop 1 LPSI Header 3RCS*V30 to 3S1L*V15 & 6 15 3SEL*V987 10 M3-31 Loop 2 LPSI Header 3RCS*V107 to 3SIL*V19 & 6 20 3SIL*V985 10 M3-32 Loop 3 LPSI Header 3RCS*V71 to 3SIL*V17 & 6 20 3SIL*V986 10 M3-33 Loop 4 LPSI Header 3RCS*V146 to 3SIL*V21 & 6 20 3SIL*V984 10 NOTE: 1. Segment boundary is described in terms of valve-to-valve.

The pipe segments of Group 2 are part of the LPSI system at MPS3 and are continuously pressurized because they are in the injection flow path from the safety injection tanks.In order to perform the IWB-5222(b) system pressure test on these pipe segments it would be necessary to connect jumpers circumventing the inboard check valve boundaries from the RCS. This is a significant personnel safety hazard that results in an estimated additional 0.2 man-Rem of unnecessary personnel radiation exposure.Page 5 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

4.3 Component

Group 3: "Safety Injection to RCS Cold and Hot Legs" -The component Group 3 piping segments (34 through 38) are shown in Table 3. Additional details are in Attachment 1, Table 3.Table 3: Affected Piping Segments of Group 3 Segment Description Segment Boundary ( Dia. Length (in) (ft)M3-34 SI to Loop 1 Hot Leg 3SIH*V1 10 to 3RCS*V26 2, 6 200 M3-35 SI to Loop 2 Hot Leg 3SIH*V 112 to 3RCS*V102 2, 6 208 M3-36 SI to Loop 3 Hot Leg 3SIL*V27 and 3SIL*V26 to 2,6,8 275 3RCS*V69 M3-37 SI to Loop 4 Hot Leg 3SIL*V29 and 3SIL*V28 to 2, 6, 8 101 3RCS*V142 M3-38 SI to (4) Cold Legs 3SIH*V5 to 3RCS*V29, 1.5, 3 758 3RCS*V106, 3RCS*V70, and 3RCS*V145 NOTE: 1. Segment boundary is described in terms of valve-to-valve.

The pipe segments of Group 3 are in high pressure and low pressure Safety Injection (HPSI, LPSI) systems, in portions of piping between check valves that are not normally pressurized during plant operation.

In order to pressurize these segments to perform the 1WB-5222(b) system pressure test it would be necessary to connect jumpers circumventing the inboard check valve.boundaries from the RCS. This is a significant personnel safety hazard and results in an estimated additional 0.375 man-Rem of unnecessary personnel radiation exposure.4.4 Component Group 4: "Residual Heat Removal (RHS) Suction" -The component Group 4 piping segments (39, 40) are shown in Table 4. Additional details are in Attachment 1, Table 4.Table 4: Affected Piping Segments of Group 4 Segment Description Segment Boundary ( Dia. Length_________

____________________________________________________ (in) J (ft)M3-39 A RHS Suction Line 3RCS*V999 to 3RHS*V997 12 59 (8701A)M3-40 B RHS Suction Line 3RCS*V998 to 3RHS*V996 12 59 (8702B)NOTE: 1. Segment boundary is described in terms of valve-to-valve.

The pipe segments 39 and 40 are part of the RHS system, which is not pressurized during normal plant operation.

In order to pressurize this segment to perform the IWB-5222(b)

Page 6 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued) system pressure test it would be necessary to open the isolation valves 3RHS*8701C

("A" train) and 3RHS*8702C

("B" train). These isolation valves are required to be closed when the plant is in Modes 1, 2, and 3, as described in the MPS3 Final Safety Evaluation Report (FSAR) Section 5.4.7.1. Alternatively, temporary high pressure hoses with a hydrostatic pump would need to be installed to pressurize these segments during a refuel outage, which would introduce a significant personnel safety hazard if the connection or hose fails in the presence of inspection personnel.

4.5 Component

Group 5: "Auxiliary Pressurizer Spray" -The component Group 5 piping segment (41) is described in Table 5 and additional details are in Attachment 1, Table 5.Table 5: Affected Piping Segments of Group 5 Segment Description Segment Boundary ' Dia. Length (in) (ft)M3-41 Auxiliary Pressurizer Spray 3RCS*V174 (AV8145) to 2 230 3RCS*V175 NOTE: 1. Segment boundary is described in terms of valve-to-valve.

Segment 41 is part of the MPS3 auxiliary pressurizer spray line, which is not normally pressurized.

In order to pressurize this segment to perform the IWB-5222(b) system pressure test it would be necessary to open the normally closed upstream isolation valve 3RCS*MV8145.

Water in this line is supplied from the charging system with an operating pressure greater than the RCS normal operating pressure.

Opening this valve would allow water in the auxiliary pressurizer spray line, which is at containment ambient temperature, to pass through a check valve into the main spray header and through the spray nozzle into the pressurizer.

With the RCS at normal operating temperature, this test would create a thermal shock transient to the spray nozzle, which has been evaluated to be in excess of 320 degrees F. The pressurizer stress report for MPS3 has evaluated spray nozzle shock as a design basis transient, but based on the temperature severity, this test could result in the most severe case of which only 10 cycles were considered in the stress analysis.5. Proposed Alternative and Basis for Use The alternatives and basis for the MPS3 request are organized by component group discussions in the balance of this section. The provisions of 10 CFR 50.55a(a)(3)(i) permit requests for alternatives in some of these component groups because the request provides for an acceptable level of quality and safety. For other component groups, the provisions of 10 CFR 50.55a(a)(3)(ii) permit alternative requests when the specified requirements result in a hardship without a compensating increase in the level of quality and safety. The following list summarizes the application of these provisions in this request for relief. Refer to Tables 1 through 5 of both Section 4.0 and Attachment 1 for additional details regarding piping segment component groups and requested alternatives.

Page 7 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

Group 10 CFR 50.55a Provision Section 2 (a)(3)(ii) 5.1 2 (a)(3)(i) 5.2 3 (a)(3)(ii) 5.3 4 (a)(3)(ii) 5.4 5 (a)(3)(ii) 5.5 5.1 Alternative and Basis for Component Group 1 MPS3 Group 1 segments 1-29 are VTDB lines and connections that are equipped with manual valves, which provide double isolation of the RCPB. As an alternative to the IWB-5222(b) system pressure test requirements for these RCPB pipe segments, this request proposes to perform an ASME Code Section XI, Table IWB-2500-1 and 1WB-5221 system leakage test with the isolation valves in the normally closed position.

This examination will be performed at the nominal operating pressure associated with 100%reactor power after satisfying the ASME Code required hold time.Basis for approval of this alternative includes the following information:

5.1.1. The non-isolable portion of the RCPB VTDB lines and connections will be pressurized and will be visually examined as required.

Only the isolable portion of these small diameter VTDB lines and connections will not be pressurized, but a VT-2 examination will still be performed in these cases.5.1.2. A typical VTDB line and connection includes two manual valves or one manual valve, separated by a short piece of pipe or a pipe nipple, which is connected to the RCPB via another short pipe nipple. These connections are typically socket welded and the welds receive a surface examination after installation.

The piping and valves are normally heavy walled. The VTDB lines and connections are not subject to high or cyclic loads and design ratings are greater than RCPB operating pressure.5.1.3. MPS3 uses the ASME Code Section XI, 2004 Edition (No Addenda) for itsSection XI Repair/Replacement program activities, but the requirements exclude components or connections NPS 1 and smaller from the pressure test requirement after welded repairs. Therefore, requiring a pressure test and visual examination of the NPS 1 and smaller of Group 1 Class 1 RCPB VTDB lines and connections once each 10-year interval is unwarranted considering that pressure testing a repair weld on the same connections is not required by the ASME Code,Section XI.Considering this information and the implications for personnel safety and radiation exposure that would occur as a result of meeting the ASME Code Section XI, 2004 Edition, pressure test requirements, compliance with the pressure test requirements for Groups 1 RCPB VTDB lines and connections results in an unnecessary hardship without a sufficient compensating increase in the level of quality and safety. Therefore, DNC requests approval of this alternative pursuant to 1OCFR50.55a(a)(3)(ii).

Page 8 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

5.2 Alternative

and Basis for Component Group 2 Group 2 segments (30 through 33) are part of the LPSI system and are continuously pressurized and monitored for loss of pressure because they are in the open injection flow path from the safety injection tanks. As an alternative to the pressure test requirements for these RCPS pipe segments, DNC proposes to use the pressure associated with the statically pressurized passive safety injection system.The basis for approval of this alternative includes the following information.

ASME approved Code Case N-731, "Alternative Class 1 System Leakage Test Pressure RequirementsSection XI, Division 1," on February 22, 2005 (see Attachment 2), because it believed that detection of leakage from a through-weld or through-wall flaw is affected by pressure, temperature, and time, with time being the controlling factor. Since the requirements of Code Case N-731 limit its application to safety injection systems that must be under pressure for an entire operating cycle there appears to be no reason to have the pressure elevated to full RCS pressure to prove leakage integrity for this piping.Because this alternative for Group 2 is specific to the LPSI piping at MPS3 that is continuously under pressure for the entire operating cycle, continually monitored for loss of pressure, and is included in the scope of the ASME approved Code Case N-73 1, use of this alternative provides an acceptable level of quality and safety. Therefore, DNC requests approval to use Code Case N-731 pursuant to 10 CFR 50.55a(a)(3)(i).

5.3 Alternative

and Basis for Component Group 3 The Group 3 piping segments 34-38 of MPS3 are part of the safety injection system that are located between check valves that isolate these segments from RCS pressure.

As an alternative to the 1WB-5222(b) system pressure test for these RCPB pipe segments, DNC proposes to perform this test using a reduced test pressure during the full flow check valve tests of these segments, during the refuel outage with the RCS depressurized.

The basis for approval of this alternative is included in the following information.

In order to pressurize these segments to meet the IWB-5222(b) system pressure test requirements, it would be necessary to connect jumpers (high pressure hose)circumventing the inboard check valve boundaries from the RCS. This is a significant personnel safety hazard and will result in unnecessary personnel radiation exposure.Considering this information, compliance with these requirements for Group 3 RCPB pipe segments results in unnecessary hardship without sufficient compensating increase in the level of quality and safety. Therefore, DNC requests approval of this alternative pursuant to 10 CFR 50.55a(a)(3)(ii).

Page 9 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

5.4 Alternative

and Basis for Component Group 4 The Group 4 piping segments 39 and 40 are in the RHS system and are not pressurized during normal plant operation.

As an alternative to the IWB-5222(b) system pressure test requirements for these RCPB pipe segments, DNC proposes to perform this test using a reduced test pressure prior to the valves being closed, isolating these segments in the normal preparation for mode change during startup.The basis for approval of this alternative is included in the following information.

a) In order to pressurize this segment to meet the IWB-5222(b) system pressure test requirements, it would be necessary to open the isolation valves 3RHS*8701 C ("A" train) and 3RHS*8702C

("B" train). These isolation valves are required to be closed when the plant is in Modes 1, 2, and 3 as described in the MPS3 FSAR Section 5.4.7.1 and plant operational procedures.

b) Alternatively, to install temporary high pressurehoses with a hydrostatic pump to pressurize these segments during the refuel outage would add additional personnel exposure and introduce a significant personnel safety hazard if the connection or hose fails in the presence of inspection personnel.

Considering this information, compliance with the 1WB-5222(b) system pressure test requirements for Group 4 RCPB pipe segments, results in unnecessary hardship without sufficient compensating increase in the level of quality and safety. Therefore, DNC requests approval of this alternative pursuant to 10 CFR 50.55a(a)(3)(ii).

5.5 Alternative

and Basis for Component Group 5 The RCPB pipe segment 41 at MPS3 is part of the auxiliary pressurizer spray line, which is not normally pressurized during plant operation.

As an alternative to the IWB-5222(b) system pressure test requirements for these RCPB pipe segment, DNC proposes to perform this test at a reduced pressure when pressurizer spray is initiated for normal plant cooldown in accordance with plant operating procedures.

The basis for approval of this alternative is included in the following information.

In order to pressurize this segment to meet the IWB-5222(b) system pressure test requirements, it would be necessary to open the normally closed upstream isolation valve 3RCS*MV8145.

Opening this valve would allow water in the auxiliary pressurizer spray line, which is at containment ambient temperature, to pass through a check valve into the main spray header and through the spray nozzle into the pressurizer.

With the RCS at normal operating temperature this would create a thermal shock transient to the spray nozzle.Considering this information, compliance with the IWB-5222(b) system pressure test requirements for the Group 3 piping segment for MPS3 results in an unnecessary hardship and adverse impact to plant equipment without a sufficient compensating increase in the level of quality and safety. Therefore, DNC requests approval of this alternative pursuant to the provisions of 10CFR50.55a(a)(3)(ii).

Page 10 of 11 10 CFR 50.55a Request Number IR-3-09 (Continued)

6.0 Duration

of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019.7.0 Precedents With the exception of the use of ASME Code Case N-73 1, similar alternatives to the hydrostatic test requirements of the ASME Code Section XI, 1989 Edition, IWB-5222 have been approved for Indian Point Units 2 and 3 and for the 1998 Edition with the 2000 Addenda for the 10-year ISI interval leakage test pressure per IWB-5220 for Surry Units 1 and 2. The differences in the Code requirements taken to establish the basis for the approval of these requests result in the same approved objective of not having to pressure test portions of the Class 1 RCPB to full RCS pressure and reduced pressures for other piping segments.Both of the safely evaluations for these precedent requests are cited below from ADAMS.1. Indian Point Nuclear Generating Units 2 and 3, dated December 7, 2005, (ADAMS Accession No. ML053110525).

2. Surry Units 1 and 2, Relief, dated November 1, 2005, (ADAMS Accession No.ML052930032).

In addition, Relief Request IR-2-45 was approved for MPS3 for use during the second inspection interval on September 27, 2007, (ADAMS Accession No. ML072620318).

8.0 References

1. 2004 Edition, ASME Code,Section XI, No Addenda.2. ASME Code Case N-731, "Alternative Class 1 System Leakage Test Pressure RequirementsSection XI, Division 1".Page 11 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 1 ATTACHMENT 1 USE OF ALTERNATIVE CLASS 1 PRESSURE TEST REQUIREMENTS ALTERNATIVE REQUEST IR-3-09 PIPE SEGMENT DETAILS OF COMPONENT GROUPS 1 THROUGH 5 Attachment 1, Page 1 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 1 -COMPONENT GROUP 1 PIPING DETAILS Dose (4)Segment (2) Other ISI (3) Savings Number (1) Drawing Line Number Exmiaton Request Estmat (MR)M3-1 26902 Sh. 1 3-RCS-002-128-1 None Relief is from using valve 3RCS*V024 to 5 pressurize downstream pipe and valve.M3-2 26902 Sh. 2 3-RCS-002-136-1 None Relief is from using valve 3RCS*V099 to 10 pressurize downstream pipe and valve.M3-3 26902 Sh. 4 3-RCS-002-131-l None Relief is from using valve 3RCS*V068 to 10 pressurize downstream pipe and valve.M3-4 26902 Sh. 5 3-RCS-002-14-1 None Relief is from using valve 3RCS*V140 to 10 pressurize downstream pipe and valve.M3-5 26902 Sh. 6 3-RCS-002-172-1 None Relief is from using valve 3RCS*V203 to 15 3-RCS-002-173-1 pressurize downstream pipe and valve.M3-6 26902 Sh. 6 3-RCS-002-172-1 None Relief is from using valve 3RCS*V206 to 15 3-RCS-002-173-1 pressurize downstream pipe and valve.Examinations of welds M3-7 26902 Sh. 6 3-RCS-002-172-1 RCS-176-FW-31, FW- Relief is from using valve 3RCS*V209 to 15 3-RCS-002-173-1 32, FW-33, FW-34, pressurize downstream pipe and valve.and FW-38 Examinations of welds M3-8 26902 Sh. 6 3-RCS-002-172-1 RCS-178-FW-9, FW- Relief is from using valve 3RCS*V212 to 3-RCS-002-173-1 10, FW- 11, FW- 12, pressurize downstream pipe and valve. 15 and 407302-FW-5 3-RCS-002-148-1 Relief is from using any of the loop drain M3-9 26902 Sh. 6 3-RCS-001-1 13-1 None isolation valves to pressurize downstream 15 3-RCS-001-245-1 pipe and valves.3-RCS-001-246-1 Relief is from using valve 3RCS*V898 to M3-10 26902 Sh. 6 3-RCS-001-247-1 None pressurize downstream pipe and valves. 15 3-RCS-750-214-1 pressurize downstream pIpeandvalves.

Attachment 1, Page 2 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 1 -COMPONENT GROUP 1 PIPING DETAILS (Continued)

Dose (4)Segment (2) (3) Savings Number (1) Drawing Line Number Other ISI Examinations Request Estimate (MR)M3-11 26902 Sh. 1 3RCS*MV8002A(5) None Relief is from using valve 3RCS*V989 to 100 pressurize downstream pipe and flange.M3-12 26902 Sh. 1 3RCS*MV8002A(5) None Relief is from using valve 3RCS*V990 to 100 pressurize downstream pipe and flange.M3-13 26902 Sh. 1 3RCS*MV8001A(5) None Relief is from using valve 3RCS*V991 to 100 pressurize downstream pipe and flange.3-14 26902 Sh. 1 3RCS*MVI8001A(5) None Relief is from using valve 3RCS*V992 to 100 pressurize downstream pipe and flange.M3-15 26902 Sh. 2 3RCS*MV8002C(5) None Relief is from using valve 3RCS*V979 to 100 pressurize downstream pipe and flange.M3-16 26902 Sh. 2 3RCS*MV8002C(5) None Relief is from using valve 3RCS*V980 to 100 pressurize downstream pipe and flange.M3-17 26902 Sh. 2 3RCS*MV8001C(5) None Relief is from using valve 3RCS*V981 to 100 pressurize downstream pipe and-flange.

M3-18 26902 Sh. 2 3RCS*MV8001C(5) None Relief is from using valve 3RCS*V982 to 100 pressurize downstream pipe and flange.M3-19 26902 Sh. 4 3RCS*MV8002B(5) None Relief is from using valve 3RCS*V984 to 100 pressurize downstream pipe and flange.M3-20 26902 Sh. 4 3RCS*MV8002B(5) None Relief is from using valve 3RCS*V985 to 100 pressurize downstream pipe and flange.M3-21 26902 Sh. 4 3RCS*MV8001B(5) None Relief is from using valve 3RCS*V986 to 100 M3_21 292S4 3Nnpressurize downstream pipe and flange.M3-22 26902 Sh. 4 3RCS*MV8001B(5) None Relief is from using valve 3RCS*V987 to 100 pressurize downstream pipe and flange. _M3-23 26902 Sh. 5 3RCS*MV8002D (5) NoneRelief is from using valve 3RCS*V974 to 100 None pressurize downstream pipe and flange.Attachment 1, Page 3 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 1 -COMPONENT GROUP 1 PIPING DETAILS (Continued)

Dose (4)Segment (2) (3) Savings Number (1) Drawing Line Number Other ISI Examinations Request Estimate (MR)M3-24 26902 Sh. 5 3RCS*MV8002D(5) None Relief is from using valve 3RCS*V975 to 100 pressurize downstream pipe and flange.M3-25 26902 Sh. 5 3RCS*MV8001D(5) None Relief is from using valve 3RCS*V976 to 100 pressurize downstream pipe and flange.M3-26 26902 Sh. 5 3RCS*MV8001D(5) None Relief is from using valve 3RCS*V977 to 100 pressurize downstream pipe and flange.M3-27 26902 Sh. 6 3RCS-001-226-1 None Relief is from using valve 3RCS*V958 to 100 pressurize downstream pipe and flange.M3-28 26902 Sh. 1 3RCS-001-225-1 None Relief is from using valve 3RCS*V956 to 2 pressurize downstream pipe and flange.M3-29 26902 Sh. 1 3RCS-002-23-1 None Relief is from using valve 3RCS*V33 to 50 3RCS-750-116-02 pressurize downstream pipe and flange.NOTES 1. Schedule 160 piping is used for all segments of Group 1, unless annotated otherwise by Note 5. Material of piping segments is Austenitic stainless steel, SA376, Type 316.2. Design Pressure is 2485 psig; Normal Operating Pressure:

None, and remains normally isolated.3. Proposed Test Pressure:

None.4. The estimated accumulated dose savings for the use of alternative requirements in the component groups from Request IR-3-09 is a total of 2.5 man-Rem at MPS3.5. Affected segment material is SA312, F304. This is a part of valve assembly.Attachment 1, Page 4 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 2 -COMPONENT GROUP 2 PIPING DETAILS Dose (4)Segment Drawing Line Number (1), (2) Other ISI Examinations Request (3) Savings Number IEstimate I I (MR)M3-30 26902 Sh. 1 3-SIL-006-139-1 Request JAW with 50 26912 Sh. 2 3-SIL-010-45-1 (5) None Code Case N-731 26902 Sh. 2 3-SIL-006-145-1 Weld examinations of welds SIL-6-5-SW-M3-31 26912 Sh. 2 3-SIL-010-49-1 (5) E, SIL-6-6-SW-B, SIL-6-FW-8, and SIL Code Case N-731w FW-9 M3-32 26902 Sh. 4 3-SIL-006-140-1 Request JAW with 50 26912 Sh. 2 3-SIL-010-47-1 (5) None Code Case N-731 M3-33 26902 Sh. 5 3-SIL-006-146-1 Weld examinations of welds SIL-7-5-SW-Request IAW with 26912 Sh. 2 3-SIL-010-51-1 (5) B, SW-E, SW-F, SW-G, SIL-7-FW-7 and Code Case N-731 50 FW-8 NOTES: 1. 6"- Schedule 160 piping is used for segments of Group 2, unless annotated otherwise by Note 5 (10"- Schedule 140).Material of piping segments is Austenitic stainless steel, SA376, Type 316.2. Design pressure is 2485 psig; Nominal Operating Pressure is 650 psig.3. Proposed Minimum Test Pressure:

636 psig.4. The estimated accumulated dose savings for the use of alternative requirements in the component groups from Request IR-3-09 is a total of 2.5 man-Rem at MPS3.5. 10" -Schedule 140 piping.'Attachment 1, Page 5 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 3 -COMPONENT GROUP 3 PIPING DETAILS Dose (4)Segment Drawing Line Number (1), (2) Other ISI Examinations Request (3) Savings SLine NEstimate (MR)Between check valves -relief from 26902 Sh. 1 3-SII-002-145-1 Welds 407254-FW-49, FW- installing and removing temporary M3-34 26902 Sh. I 3-SIH-006-5-1 50, 5W-63-1ad FW- jumper hoses from downstream of 75 26913 Sh. 2 3-SIJ-006-59-l 50, FW-63-1, .and FW-85 check valve 3RCS*V26 to pressurize upstream piping Between check valves -relief from 26902 Sh. 2 3-SIH-002-146-l installing and removing temporary M3-35 26913 Sh. 2 3-SIH-006-62-1 None jumper hoses from downstream of 75 check valve 3RCS*V102 to pressurize upstream piping 3-SIL-002-20-1 Between check valves -relief from 26902Sh. 4 3-SIL-006-21-1 installing and removing temporary M3-36 26912 Sh. 1 3-SIL-006-161-1 None jumper hoses from downstream of 75 3-SIL-008-155-1 check valve 3RCS *V69 to pressurize upstream piping 3-SIL-002-24-1 Between check valves -relief from 26902 Sh. 5 3-SIL-006-25-1 installing and removing temporary M3-37 26912 Sh. 1 3-SIL-006-162-1 None jumper hoses from downstream of 75 3-STL-008-156-1 check valve 3RCS*V 142 to pressurize

_______________

upstream piping Attachment 1, Page 6 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 3 -COMPONENT GROUP 3 PIPING DETAILS (Continued)

____Segment Drawing Line Number (1), (2) Other ISI Examinations Request (3) Savings Number Estimate (MR)3-SIH-150-141-1 3-SIH-150-142-1 3-SIH1-150-143-1 26902 Sh. 1 Between check valves -relief from 2692 S. 23-SIH1-150-144-1 26902 Sh. 2 3-SIH-150-29-1 Welds 407023-FW-2, FW-3, installing and removing temporary M3-38 26902 Sh. 4 3-SIH-150-30-1 FW-4, 408046-FW-3 and jumper hoses from downstream of 75 26902 Sh. 5 3-SIH-150-30-1 FW-4 check valve 3RCS*V029, V106, V70 26913 Sh. 1 3-SIH-003--1 or V145 to pressurize upstream piping.3-5111-003-26-1 3-SIH-003-114-1 3-SIH-150-28-1 NOTES: 1. Schedule 160 piping is used for segments of Group 3. Material of piping segments is Austenitic stainless steel, SA376, Type 316.2.Segment: Design Pressure: Maximum Operating:

34/35 2485 psig 2235 psig 36/37 2485 psig 1747 psig 38 2735 psig 2725 psig 3. Proposed Test Pressure will be associated with full flow check valve test with the RCS depressurized.

There is no pressure instrumentation monitoring these segments.

The actual pressure is that achieved during full flow testing using the Safety Injection and Charging pumps.4. The estimated accumulated dose savings for the use of alternative requirements in the component groups from Request IR-3-09 is a total of 2.5 man-Rem at MPS3.Attachment 1, Page 7 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment 1 (Continued)

TABLE 4 -COMPONENT GROUP 4 PIPING DETAILS Dose (4)Segment Drawing Line Number (1), (2) Other ISI Examinations Request (3) Savings Number Estimate (MR)SWelds RHS-501-3-SW-5, Relief is from using valve 3RHS*V999 M3-39 26912 Sh. 1 3-RHS-012-33-1 RHS-501-FW-3 and FW-5 to pressurize downstream pipe and None valve.Reliefis from using valve 3RHS*V998 M3-40 26912 Sh. 1 3-RHS-012-35-1 None to pressurize downstream pipe and None valve.NOTES: 1. 12" -Schedule 160 piping is used for segments of Group 4. Material is Austenitic stainless steel, SA376, Type 316.2. Design Pressure is 2485 psig; Maximum Operating Pressure is 441psig.3. Proposed Minimum Test Pressure:

340 psia Attachment 1, Page 8 of 9 10 CFR 50.55a Request Number IR-3-09 Attachment I (Continued)

TABLE 5 -COMPONENT GROUP 5 PIPING DETAILS Dose (4)Segment Drawing Line Number (1), (2) Other ISI Examinations Request (3) Savings Number Estimate (MR)26902 Sh 1 Relief is from using valve 3RCS*V174 M3-41 26902 Sh3 3-RCS-002-150-1 None (AV8145) to pressurize upstream None_ _ 90 I _ _ _ _piping and valve.NOTES: 1. 2" -Schedule 160 piping is used for segments of Group 5. Material is Austenitic stainless steel, SA376, Type 316.2. Design Pressure is 2485 psig; Minimum Operating Pressure is 325psia. during normal plant cooldown.3. Proposed Minimum Test Pressure:

325psia 4. The estimated accumulated dose saving for the use of alternative requirements in the component groups from Request IR-3-09 is a total of 2.5man-Rem at MPS3.Attachment 1, Page 9 of 9 10OCFR 50.55a Request Number IR-3-09 Attachment 2 ASME CODE CASE N-731*. ALTERNATIVE CLASS 1 SYSTEM LEAKAGE TEST PRESSURE REQUIREMENTS SECTION XI, DIVISION 1*(1 PAGE)(I DOMINION MILLSTONE POWER STATION UNIT 3*Reprinted from ASME 2007 BPVC, Code Cases, Nuclear Components, by permission of The American Society of Mechanical Engineers.

All rights reserved.Attachment 2, Page 1 of 2 10 CFR 50.55a Request Number IR-3-09 Attachment 2 (Continued)

CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: February 22, 2005 The ASME Boiler and Pressure Vessel Standards Committee took action to eliminate Code Case expiration dates effective March 11, 2005. This means that all Code Cases listed in this Supplement and beyond will remain available for use until annulled by the ASME Boiler and Pressure Vessel Standards Committee.

CASE N-731 Case N4731 Alternative Class 1 System Leakage Test Pressure RequirementsSection XI, Division 1 Inquiry: What alternative Class I system leakage test pressure requirements may be used for portions of Class I systems that are continuously pressurized during an operating cycle by a statically-pressurized passive safety injection system of a pressurized water reactor, in lieu of the requirements of IWB-522 I(a)?Reply: It is the opinion of the Committee that. for por-tions of Class I safety injection systems that are continu-ously pressurized during an operating cycle, the pressure associated with a statically-pressurized passive safety injection system of a pressurized water reactor maybe used.The Comrittoe's function is to establish mutes of safety. relating only to pressure Integrity, governing the construction of boiloiS, pressure vessuls, transport tanks and fluclear components, and Inservice inspection for pressure Integrity of nuclear componctnts and transport tanks. and to interpret those rules when arise regarding their Intent. This Code does not address other safety issues relating tothe construction of boilers, pressure vessels, transport tanks end nuclear comnponents, and the insornvico inspection of nuclear coenponents and transport tanks. The user of the Code should refer to other pertinent codes, standards, laws,. regulations or other relevant documents.

1 (N-731)Attachment 2, Page 2 of 2 10 CFR 50.55a Request Number IR-3-09 Attachment 3 USE OF ALTERNATIVE CLASS 1 PRESSURE TEST REQUIREMENTS REOUEST IR-3-09 PIPING AND INSTRUMENTATION DIAGRAMS Attachment 3, Page 1 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued)

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3, Page 3 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued) i LI -1:-bUL -I I.--r or I -F F-----Min --- ----- ---4 P"414 Q>-0,ý -- -J-LA A W~-1-ý ,. ~ -------- g--------~ -- -- I- ------IIL j Li (OA CAA.2. r 4> -' - -_ ----i I A7~VfiTt A -L PIPING N I-R-EITI-N 0-111 IS& 1211 A f.27-M12 js&w nW(,. mN.i2179-FM-102C.

Attachment 3, Page 4 of 11 10 _CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued) t~nAY ... A i -tL qi A21-s ( F4 I, 6-ZA 711 V----------L, _jE c~~A--I..-----

-- -- -- -----F ~-. I.9 j zz ýli~riTT 4.~f...... vzv A IAJCLEAA SAFETY RELATED QA GA7 .* 1 FSAR FIGURE QA OPERATIONS CRITICAL-~ ~ ýK -T.TT-A~ .J4LRrEML~IS&W DWG. NO.12173-FM-102D Attachment 3, Page 5 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued)

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L T , --' ----..................A"r .- , > .I G.m IL I ; /-" --"~~ ~ --F--- --. -_ _, -...., IL3,- -ANA Attachment 3, Page 6 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued) 3-SLt 13-tl 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued)

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IS&W DWG. NO.12179-EM-102F Attachment 3, Page 7 of I 1 10 CFR 50.55a Request Number IR-3-09 Attachment 3 IV F i A- --43/4LA~ ~ --,- =7~r -K ----J. jz_>FL-----------~=,I Lz.q -rXf 1 J S S L _12- d --5 2 L TS -.-..........m I: 'F-Hg"£t~ rn-A j5 '>1-i C F--4;-'A -K 4 I- "A I F I>~ '4'I £ &i~i.I ~ AFt-'F 'F Az amr"rr\IF z'A F' U, Attachment 3, Page 8 of 11 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued)

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AA I!I'k-& DIG rý.12179-EM---

t TT o FIG~ ___ran-AZT C)IT I~INF-- DWG.FV~ NO119-M 1 Attchen 3,Pg 1 f1 10 CFR 50.55a Request Number IR-3-09 Attachment 3 (Continued)

L z-SL.A1 AA Ai~ IS r ý N 4 igg Z~iii___ T-'5 -p..-7t s~t. 3 '_ K Q J _ C yyyyyyy k-b 2 LC Ls-7-I In)-A_4 "'0 < W/L FSRFGR CTIR7CN CRITICA 11 HI .w IS&W DWO. NO.]2IT£-EM-tl3B JS&W DWG. NO).1217B-EM-1138 Attachment 3, Page 11 of 11 10 CFR 50.55a Request Number IR-3-10 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(i)

-Alternative Provides Acceptable Level Of Quality And Safety -1. ASME Code Components Affected ASME Code Class:

References:

Examination Category: Item Number:

Description:

Code Class 1 ASME Section XI, IWA-5241 and Table IWB-2500-1 10 CFR 50.55a(g)(6)(ii)(E), Code Case N-722 B-P B15.10 Alternative Examination Criteria for the Visual Examination of Reactor Coolant System Hot Leg and Cold Leg Nozzle-to-Safe End Welds Reactor Pressure Vessel (RPV) Nozzle-to-Safe End Welds: Components:

Inlet Nozzles: 301-121-A 301-121-B 301-121-C 301-121-D Outlet Nozzles: 302-121-A 302-121-B 302-121-C 302-121-D 2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement Examination Category B-P, "All Pressure Retaining Components," requires that Visual, VT-2 examinations are performed each refueling outage. IWA-5241(b) allows this examination to be performed without insulation removal, stating, "only the examination of the surrounding area (including floor areas or equipment surfaces located underneath the components) for evidence of leakage shall be required." Page 1 of 6 10 CFR 50.55a Request Number IR-3-10 (Continued) 10 CFR 50.55a(g)(6)(ii)(E)(1), states (in part) that "all licensees of pressurized water reactors shall augment their inservice inspection program by implementing ASME Code Case N-722, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 600/82/182 MaterialsSection XI, Division 1," dated: July 5, 2005, Attachment 1.ASME Code Case N-722, requires a Visual, VE "Bare Metal Visual" of all listed locations (See Attachment 1, Code Case N-722), which includes RPV outlet (hot leg) and inlet (cold leg) nozzle welds (Item Nos. B 15.90 and B 15.95, respectively).

4. Reason for the Request The ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition, No Addenda,[Reference 8.1] is the Code Edition to be used for the Third Inservice Inspection Interval at MPS3, which starts on April 23, 2009, and is scheduled to end on April 22 2019. The requirements for Class 1 Visual, VT-2 examinations for leakage of pressure retaining components are included in this Code Edition under Examination Category B-P, "All Pressure Retaining Components." Visual, VT-2 examinations are performed each refueling outage. For insulated components, IWA-5241(b) allows this examination to be performed without insulation removal, stating, "only the examination of the surrounding area (including floor areas or equipment surfaces located underneath the components) for evidence of leakage shall be required." Based on these requirements, visual examinations for leakage may be performed on Class 1 components with insulation in place.However, both the industry as represented in this case by the Materials Reliability Program (MRP) and the NRC staff along with ASME have concluded that a visual examination for leakage performed on insulated components is inadequate for the identification of leakage that can potentially occur as a result of primary water stress corrosion cracking (PWSCC) in items made with Alloy 600/82/182 materials.'

This alternative request is needed to address this inadequacy.

The NRC staff provided their position on these visual examination requirements in the 10 CFR 50.55a rulemaking issued September 10, 2008 (effective date October 10, 2008)under 73 FR 52748. It included the mandatory use of ASME Code Case N-722 Attachment 1 with conditions specified in the 10 CFR 50.55a(g)(6)(ii)(E), "Reactor Coolant Pressure Boundary Visual Inspections".

Meeting these requirements is a concern because the reactor pressure vessel (RPV) nozzles at MPS3 have an insulation package surrounding each nozzle and its corresponding nozzle-to-safe end Alloy 82/182 welds which makes them inaccessible for these required bare metal visual examinations.

Prior to the fall 2008 MPS3 refueling outage, DNC submitted a letter to the NRC' notifying the staff that DNC was implementing a deviation from the requirements of MRP-139[Reference 8.2]. This deviation was to not perform the required bare metal visual 1 DNC Letter to NRC, "Dominion Nuclear Connecticut, Inc. Millstone Power Station Unit 3 Electric Power Research Institute MRP-1 39 Deviation Notification," dated October 27, 2008, ADAMS Accession No.ML083010233 Page 2 of 6 10 CFR 50.55a Request Number IR-3-10 (Continued) examinations specified in MRP-139 based on the restricted access to the MPS3 RPV nozzle welds caused by the insulation package surrounding each nozzle. Attachment 2 contains the Technical Evaluation M3-EV-08-0016 that provided the basis for not performing these bare metal visual examinations.

The restrictions from performing the bare metal visual examinations of MRP-139 are the same as the restrictions that this alternative request has been written to address now that the requirements of Code Case N-722 [Reference 8.3]apply.This alternative would be used in lieu of the requirements of Code case N-722.5. Proposed Alternative and Basis for Use For purposes of this request, the following Table is provided to show the examinations that will be performed under this requested alternative.

These examinations are exactly what were derived from the flaw tolerance evaluation contained in Attachment 2.These modifications to the requirements of Code Case N-722 Attachment 1 are shown in Table 1, below with the modifications bolded and underlined for ease of identification.

Table 1: MPS3 Class 1 PWR Components Alloy 600/82/182 To Be Examined In Accordance with Code Case N-722 Attachment 1 with Bolded and Underlined Modifications Supported By Attachment 2 of this Request Deferral of First Successive Inspection Parts Examination Examination Acceptance Inspection Inspection To end of Item No. Examined Requirements Method Standards Interval Intervals Interval B 15.90 Hot leg All 4 Hot leg Volumetric aIb) IR-3-10 Every Same as for Not nozzle-to-nozzle-to safe other 1 St interval permissible pipe end welds Para. 5.1 refueling connections outage(c)B 15.95 Cold leg All 4 Cold leg Volumetric ab) IR-3-10 Every 3 rd Same as for Not Nozzle-to nozzle-to safe refueling 1 st interval permissible pipe end welds Para. 5.1 outage(d)connections Notes related to examination of hot leg nozzle to pipe connection (B 15.90) and cold leg nozzle to pipe connection (B 15.95): (a) UT will be performed from the inside diameter of these welds in lieu of the Visual, VE.(b) All UT examinations will meet the appropriate supplement of Section XI, Appendix VIII of the ASME Boiler and Pressure Vessel Code. The required weld volume shall be as shown on Fig. IWB-2500-8(c) of ASME Section XI, 2004 Edition (No Addenda), [Reference 8.1].Page 3 of 6 10 CFR 50.55a Request Number IR-3-10 (Continued)(c) UT will be performed every other refueling outage. These welds were last examined in the spring of 2007 (3R11) outage and will be due in the spring 2010 (3R13) outage based on the analysis in Attachment 2, which supports UT approximately every 36 months.(d) UT will be performed every 3 rd refueling outage. They were last examined in the spring of 2007 (3R 11) and will be due in the fall 2011 (3R14) outage based on the analysis in Attachment 2, which supports UT approximately every 54 months.5.1 Acceptance Standards 5.1.1 Evaluation of Examination Results 5.1.1.1 General (a) The volumetric examinations performed in accordance with 1WA-2200 shall be evaluated by comparing the examination results with the acceptance standards in 5.1.2.(b) Volumetric examination results shall be compared with recorded results of the preservice examination and prior inservice examinations.

Acceptance of welds for continued service shall be in accordance with 5.1.2.5.1.2 Acceptance 5.1.2.1 Acceptance by Volumetric Examination (a) A weld whose volumetric examination confirms the absence of flaws shall be acceptable for continued service.(b) A weld with planar surface flaws in the butt weld or base metal inside surface shall be accepted for continued service in accordance with the provisions of 5.1.2.2 or 5.1.2.3. Other flaws shall meet the acceptance standards of IWB-3514 or be accepted for continued service in accordance with 5.1.2.2 or 5.1.2.3.5.1.2.2 Acceptance by Repair/Replacement Activity or Corrective Measures (a) A weld whose volumetric examination reveals a flaw not acceptable for continued service in accordance with the provisions of 5.1.2.3 is unacceptable for continued service until the additional examinations of 5.2 are satisfied and the weld is corrected by Page 4 of 6 10 CFR 50.55a Request Number IR-3-10 (Continued) repair/replacement activity in accordance with TWA-4000 or by corrective measures beyond the scope of this relief request (e.g. stress improvement).

5.1.2.3 Acceptance by Evaluation (a) A weld whose volumetric examination detects planar surface flaws in the butt weld or base metal inside surface, or other flaws (5.1.2.1(b))

in the required examination volume that exceed the acceptance standards of IWB-3514, is acceptable for continued service if an analytical evaluation meets the requirements of IWB-3600 and the additional examinations of 5.2 are performed in the current outage.(b) Any weld containing a planar surface flaw in the butt weld/base metal inside surface will be reexamined at an every refueling outage frequency, unless mitigated by an approved mitigation technique.

5.2 Additional

Examinations

5.2.1 Examinations

which reveal unacceptable flaws as defined in 5.2.1 (a) and (b), below shall be extended to include examinations of additional welds during the current outage. The use of IWB-3514 is for the purpose of determination of scope expansion and not for the purposes of determining acceptability of the flaws. Acceptability of flaws is determined in accordance with 5.1.(a) Planar surface flaws in the butt weld or base metal inside surface exceeding the surface flaw sizes of IWB-3514 are revealed.(b) Examination volumes that reveal axial crack growth beyond the specified examination volume.5.2.2 The number of additional weld examinations shall be equal to the number of welds for that Inspection Item of Table 2 originally scheduled to be performed during the present inspection period. The additional examinations shall be selected from the same Inspection Item and where applicable, from welds of similar materials, construction, and the same or higher operating temperatures.

However, if the original examination was for Inspection Item B 15.95 of Table 1, the additional examinations shall include first, additional welds from Inspection Item B 15.90, if any remain, and second, additional weld(s) from Inspection Item B 15.95 to reach the required number of additional examinations.

Page 5 of 6 10 CFR 50.55a Request Number IR-3-10 (Continued) 5.2.3 If the additional examinations required by 5.2.1 reveal flaws exceeding the requirements of 5.2. 1(a), or (b) the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining number of welds for that Inspection Item in Table 2, at the same or higher operating temperature conditions.

In addition a 25% sample of welds of that Inspection Item at lower operating temperatures shall be sampled. If the examinations of this sample of welds at lower operating temperature reveal flaws exceeding the requirements of 5.2. 1(a), or (b), the examinations shall be further extended to include all welds of that Inspection Item, regardless of operating temperature, within the scope of this relief request.6. Duration of the Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled to end on April 22, 2019, and will be used as a basis to continue scheduling the sequence examinations into the next inspection interval or until the items associated with this request are either replaced, repaired, or mitigated.

7. Precedents Because of the unique insulation package that surrounds the MPS3 RPV nozzles, there are no precedents for this specific type of request.8. References 8.1 ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition, No Addenda 8.2 Material Reliability Program, "Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP 139)", TR-1010087, EPRI, Palo Alto, CA, dated: August 2005 8.3 ASME Code Case N-722, "Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/182 Materials,Section XI, Division 1", dated: July 5, 2005 Page 6 of 6 10 CFR 50.55a Request Number IR-3-10 Attachment 1 ASME CODE CASE N-722, ADDITIONAL EXAMINATIONS FOR PWR PRESSURE RETAINING WELDS IN CLASS 1 COMPONENTS FABRICATED WITH ALLOY 600/82/182 MATERIALS SECTION XI. DIVISION 1 DATED: JULY 5,2005"Reprinted from ASME 2007 BPVC, Code Cases, Nuclear Components, by permission of the American Society of Mechanical Engineers, All rights reserved." Attachment, 1 Page 1 of 4 10 CFR 50.55a Request Number IR-3-10 Attachment I (Continued)

CASES OF ASME BOILER AND PRESSURE VESSEL CODE Approval Date: July 5, 2005 The ASME Boiler and Pressure Vessel Standards Committee took action to eliminate Code Case expiration dates effective March i1, 2005. This means that all Code Cases listed in this Supplement and beyond will remain available for use until annulled by the ASME Boiler and Pressure Vessel Standards Committee.

CASE N-722 Case N-722 Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated With Alloy 6001824182 MaterialsSection XI, Division 1 inquiry: What examinations, in addition to those of Table IWB-2500-1, may be performed to provide addi-tional detection capability for pressure boundary leakage in pressurized water reactor plants having pressure retaining partial or full penetration welds in Class I components fabricated with Alloy 600/82/182 material?Reply: It is the opinion of the Committee that in addition to the examination requirements of Table IWB-2500-1 the additional examinations of Table 1 shall be performed for pressurized water reactor plants having partial or full penetration welds in Class 1 components fabricated with Alloy 600/82/182 material.The Committee's function is to establish rules of safety, relating only to pressure integrity, governing the construction of boilers, pressure vessels, transport "ahs and nuclear components, and Inservice inspection for pressure Integrity of nuclear components and transport tanks, and to Interpret these roles when questions arise regarding their Intent. This Code does not address other safety issues relating to the construction of boilers, pressure vessels, transport tanks and nuclear components, end the Inservice inspection of nuclear components and transport tanks. The user of the Code should refer to other pertinent codes, standarda, laws, regulations or other relevant documents.

1 (N-722)Attachment, 1 Page 2 of 4 Mo TABLE 1 EXAMINATION CATEGORIES 0 S CO GO CO C A 5.-)N SQ N)CLASS 1 PWR COMPONENTS CONTAINING ALLOY 600182f182' Extent and Frequency of Examination Deferral of Item Examination Examination Acceptance Successive Inspection to No. Parts Examined 2 Requirements Method 3,4, 1 Standard First Inspection Interval Inspection Intervals End of Interval* Reactor Vessel 2 B15.80 RPV bottom-mounted instrument penetrations All penetrations Visual, VE IWB-3522 Every other refueling outage Same as for 1st interval Not permissible B15.90 Hot leg nozzle-to-pipe connections All connections Visual, VE IWB-3522 Each refueling outage Same as for lst interval Not permissible B15.95 Cold leg nozzle-to-pipe connections All connections Visual, VE IWB-3522 Once per interval 6 , 7 Same as for 1st interval Not permissible B15.100 Instrument connections All connections Visual, VE IWB-3522 Once per interval t 2' Same as for lst interval Not permissible Steam Generators B15.110 Hot leg nozzle-to-pipe connections All connections Visual, VE IWB-3522 Each refueling outage Same as for 1st interval Not permissible 815.115 Cold leg nozzle-to-pipe connections All connections Visual, VE IWB-3522 Once per intervals, 6 Same as for 1st interval Not permissible B15.120 Bottom channel head drain tube penetration All penetrations Visual, VE IWB-3522 Once per interval', 7 Same as for 1st interval Not permissible B15.130 Primary side hot leg instrument connections All connections Visual, VE IWB-3522 Each refueling outage Same as for lst interval Not permissible B15.135 Primary side cold leg instrument connections All connections Visual, VE IWB-3522 Once per intervals', Same as for 1st interval Not permissible Pressurizer B15.140 Heater penetrations All penetrations Visual, VE IWB-3522 Each refueling outage Same as for 1st interval Not permissible B15.150 Spray nozzle-to-pipe connections All connections Visual, VE IWB-3522 Each refueling outage -Same as for 1st interval Not permissible B15.160 Safety and relief nozzle-to-pipe connections All connections Visual, VE IWB-3522 Each refueling outage Same as for 1st'interval Not permissible B15.170 Surge nozzle-tn-pipe connections All connections Visual, VE IWB-3522 Each refueling outage Same as for lst interval Not permissible B15.180 Instrument connections All connections Visual, VE IWB-3522 Each refueling outage Same as for lst interval Not permissible B15.190 Drain nozzle-to-pipe connections All connections Visual, VE IWB-3522 Each refueling outage Same as for lst interval Not permissible Piping B15.200 Hot leg instrument connections All connections Visual, VE IWB-3522 Each refueling outage Same as for Ist interval Not permissible B15,205 Cold leg instrument connections All connections Visual, VE IWB-3522 Once per interval', 7 Same as for ist interval Not permissible B15.210 Hot leg full penetration welds All welds Visual, VE IWB-3522 -Each refueling outage Same as for ist interval 'Not permissible B15.215 Cold leg full penetration welds All welds Visual, VE IWB-3522 Once per interval 6 , ý Same as for 1st interval Not permissible C)ceo ceo 0 ceo w 0 sm r C)0 CD CD t-n tin:2 (B=

TABLE It-EXAMINATION CATEGORIES (CONT'D) 0 NOTES: (1) Alloy 600/82/182 are equivalent to UNS N06600 (SB-163, S8-166, SB-167, SB-168 and S1-564), UNS N06082 (SFA 5.14 ERNiCr-3) and UNS W86182 (SFA 5.11 ENiCrFe-3).

I (2) The reactor vessel closure head is not addressed in this Case.( (3) The Visual Examination (VE) performed on Alloy 600/82/182 components for evidence of pressure boundary leakage and corrosion on adjacent ferritic steel components shall consist of the 0 e ft.7' following:

• (a) A direct VE of.the bare-metal surface performed with the insulation removed. Alternatively, the VE may be performed with insulation in place using remote visual inspection equipment Z eD 0that provides resolution of the component metal surface equivalent to a hare-metal direct VE.53 (b) The VE may be performed when the system or component is depressurized.

C1 -(t .(c) The direct VE shall be performed at a distance not greater than 4 ft (1.2 m) from the component and with a demonstrated illumination level sufficient to allow resolution of lower case Cs characters having a height of not greater thau 0.105 in 12.7 mm).-4. (4) Personnel performing the VE shall be qualified as VT-2 visual examiners and shall have completed a minimum of four (4) hours of additional training in detection of borated water leakage-t from Alloy 600182/182 components and the resulting boric acid corrosion of adjacent ferritic steel components.

ft.4. (5) An ultrasonic examination, performed from the component inside or outside surface in accordance with the requirements of Table IWB-2500-1 and Appendix VIII (1995 Edition with the .'1996 Addenda or later) shall be acceptable in lieu of the VE requirement of this table. -(6) VE shall be performed in accordance with the.schedule in IWB-2400.

1 (7) The detection of evidence of pressure leakage at a VE location shall require the VE of all components within the Examination Item No. prior to reactor startup. These additional VEs shall r 0 not affect the original VE schedule of the components within the Examination Item No. -00 C, m>.zg 0~

10 CFR 50.55a Request Number IR-3-10 Attachment 2 Dominion Nuclear Connecticut, Inc Millstone Power Station Unit 3 Electric Power Research Institute MRP-139 Notification and Technical Evaluation for Technical Justification for deviation from Mandatory Requirements of MRP-139 Millstone Unit 3 M3-EV-08-0018 Rev.0 -9/25/08 MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR Attachment 2, Page 1 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

ADominiollif Dominion Nud-cr Coiinecticut, Inc.5000 Dominion Boulevard, Glen Ailen, Virginia 23060 Web Address: www.domn.com October 27, 2008 Mr. Dennis P. Weakland Materials Reliability Program -EPRI c/o Jennifer Ma Administrative Assistant ANT, MRP & SGMP 3420 Hillview Avenue Palo Alto, CA 94304 Memo No.NLOS/GAW RA-08-026 RO DOMINION NUCLEAR CONNECTICUT.

INC.MILLSTONE POWER STATION UNIT 3 MRP-139 DEVIATION NOTIFICATION In accordance with the Nuclear Energy Institute (NEI) Guideline for the Management of Materials Issues (NEI 03-08, Rev. 1), Dominion Nuclear Connecticut, Inc. (DNC) is providing a report supporting the deviation from the requirements of Electric Power Research Institute (EPRI) Materials Reliability Program (MRP): Primary System Piping Butt Weld Inspection and Evaluation Guidelines (MRP-1 39) at Millstone Power Station Unit 3 (MPS3).The deviation report was approved by senior management on September 29, 2008 and is included as an enclosure to this letter. Specifically, the deviation relates to the mandatory visual examination requirements contained in the MRP-139 Table 6-2. It is expected that this deviation will continue while the MRP-1 39 requirement remains in effect, or until the locations are mitigated to prevent propagation of potential primary water stress corrosion cracking (PWSCC).If you have any questions regarding this report, please contact Mr. Geoffrey Wertz at (804) 273-3572.Sincerely, an Price President

-Nuclear Engineering

Enclosure:

TECHNICAL EVALUATION for Technical Justification for Deviation from Mandatory Requirements of MRP-1 39, Millstone Unit Three, M3-EV-08-0018 Rev. 0, September 25, 2008.Attachment 2, Page 2 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

Memo No. RA-08-026 MRP-139 Deviation Notification Enclosure Enclosure TECHNICAL EVALUATION for Technical Justification for Deviation from Mandatory Requirements of MRP-139 Millstone Unit Three M3-EV-08-0018 Rev. 0 9125108 MILLSTONE POWER STATION UNIT 3 DOMINION NUCLEAR CONNECTICUT, INC.Attachment 2, Page 3 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

QA Non-QA L-DB or LB document change required?

yes nI no TECHNICAL EVALUATION for Technical Justification for Deviation from Mandatory Requirements of MRP-139 Millstone Unit Three M3-EV-08-0018 Rev. 0 9/25/2008 Total Number of Pages = 36 Glenn Gardner Jý, je,ýýPreparer Robert Schonenberg "r, Independent Reviewer Date Date Date Martin Van Haltem / 7~71zt~.-

17 , 4 Z Engineering Approver Additional Concurrence per NEI 03-08 Addendum E Rev. 3 William McBrineA / ,f.L A '4¢,. f9 Independent Materials Expert -Altran Solutions Alan Price R spoible Dominion Vice President 9,&p/ao0 r' bate Date Attachment 2, Page 4 of 39, 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

MS-EV-08-0018 page 2 of 36 Technical Justification for Deviation from Mandatory Reauirements of MRP-139 Rev. 00 Section paie TABLE of CONTENTS 2 1.0 PURPOSE 3

2.0 BACKGROUND

3 3.0 DISCUSSION 5 4.0 SAFETY-SIGNIFICANCE 8

5.0 CONCLUSION

8 6.0 LIST OF ATTACHMENTS 8 Pages in body 8 Pages in attachments 28 Total pages 36 Attachment 2, Page 5 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 page 3 of 36 Technical Justiffcation for Deviation from Mandatory Requirements of MRP-139 Rev. 00 1.0 PURPOSE This technical evaluation (TE) documents the technical justification for Millstone Unit 3 (MPS3)to deviate from certain bare metal visual examination (yE) requirements of MRP-139 [I], the industry-endorsed guideline for management of Alloy 600 issues on piping and nozzle butt welds. The TE is intended for independent materials expert concurrence and transmittal to the Materials Reliability Program (MRP) for notification in accordance with the industry initiative on materials, NEI 03-08 [2].

2.0 BACKGROUND

2.1 Materials

Aging Issue Primary Water Stress Corrosion Cracking (PWSCC) of nickel-based alloys has been an on-going industry issue for several years. The cracking occurs in susceptible materials when subjected to high stress levels in the PWR reactor coolant environment CI]. The susceptible materials include weld filler materials Al 82/A82, which are utilized at Millstone Unit 3 to weld the stainless steel safe-ends to the reactor vessel nozzles. Both inlet (RCS cold leg) and outlet (RCS hot leg)nozzles are potentially affected by PWSCC at the nozzle to safe end welds. The subject weld joints include Al 82 buttering on the ferritic vessel nozzle and Al 82 weld filler between the buttering and the forged stainless steel safe end. References

[5] and [6) provide greater detail on these locations.

K 2.2 Applicable MNP-139 Requirements The applicable MRP-139 requirements for managing PWSCC at the nozzle welds are presented in the Table I on the next page along with the current inspection status, just prior to the 3R12 refueling outage (RFO) in Fall 2008. MPS3 is on an 18 month refueling cycle.Attachment 2, Page 6 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-001 8 To,.-hnr~l .h ftnr frnmr~,rt M~Innd tnr, IilrrnAntk of MRP-1R~Q page 4 of 36 PRyv An(Table .1 -MRP-139 Inspection Requirements Hot Leg (Outlet) Cold Leg (Inlet)MRP Requirements Current Status & MRP Requirements Current Status &Tbl 6-1 Cat. D and Next req'd Tbl 6-1 Cat. E and Next req'd Thl 6-2 Cat. J inspection Thl 6-2 Cat. K inspection Volume- Every 5 yrs 3R1 1, Spring 07 Every 6 yrs 3R11, Spring 07.tric (UT) 3R14, Fall 11 3RI5, Spring 13 Bare Every RFO except Not inspected Within 3 RFO (4.5 Not inspected Metal ones with 3R12, Fall 08 yrs) of volumetric 3R12, Fall 08 Visual Volumetric exam Examina-tion (VE)The tabular listing of requirements is simplified but presents MRP-139 schedule requirements accurately.

The current inspection status shows that the subject welds were UT inspected from the ID during 3R1 1 in the spring, 2007. No indications were recorded.

At issue is the fact that nozzle inaccessibility prevents bare metal visual examinations as required by MRP-139 Table 6-2 for visual examination categories J and K.2.3 Millstone 3 Unique Design Features and Nozzle Accessibility The vessel nozzle accessibility for Millstone 3 is very difficult because of the insulation package design at the nozzles. The insulation package comprises at least 14 heavy blocks weighing from 200 lbs to 1200 lbs each, bolted in place, in a very restricted location under the pit seal of the reactor vessel flange. A sketch of the package is shown in Attachment 2 on page 10. Scaffolding must be erected and each of the blocks needs to be rigged in and out. Removal of these blocks to permit the bare metal visual examination is estimated to require 105 work hours per nozzle, with a dose impact of 3.69 Rem per nozzle. The total dose impact for examination of the eight vessel nozzles is approximately 29.5 Rem.2.4 Previous Evaluations Technical Evaluation M3-EV-05-0024

[6] performed a similar evaluation for the initial bare metal visual examination of the nozzles required prior to the issuance of MRP-139. The TE documents an extensive review of original fabrication radiography of the nozzle to safe end welds. The review showed that, "For the nozzle to safe end welds, this review did not show any unusual results. All the welds had some porosity and slag inclusions but they were within acceptable limits. There were not any multiple reader sheets or repair weld numbers indicating weld repairs." Attachment 2, Page 7 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 page 5,of 36 Technical Justification for Deviation from Mandatory Reauirements of MRP-139 Rev. 00 Technical Evaluation M3-EV-07-0026

[5], in addition to mapping out a mitigation plan for A600 locations, documents a complete listing of them along with any repair records available from Westinghouse records. Records showed only minor local repairs were performed.

2.5

References:

1 "Materials Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guidelines (MRP-139)", Technical Report 1010087, EPRI, Palo Alto, CA: 2005. (retrievable from Portal -Virginia)2 NEI 03-08, "Guideline for the Management of Materials Issues", Nuclear Energy Institute (NEI), Rev. I dated April 2007, with Addendum E Rev. 3 dated April 2008 3 CR-08-07092, "Millstone Unit 3 Can Not Do a Mandatory Requirement of MRP-139", initiated 6/18/2008 4 LTR-PAFM-08-127 Rev. 2, "Technical Justification for Deviation from MRP-1 39 Visual Inspection Schedules for Millstone Unit 3 Reactor Vessel Inlet and Outlet Nozzles", dated July 2008, 0 2008, Westinghouse Electric Company LLC (Attachment 3)5 M3-EV-07-0026 Rev. 0, "Technical Evaluation For The Control And Remediation Plan For Alloy 600 MPS 3", dated 6/22/2007 6 M3-EV-05-0024 Rev. 0, "Justification for the Deferral of Visual Examination of the Millstone Unit 3 Reactor Vessel Nozzle to Safe End Welds", dated 6/28/2007 7 ASME BPV Code Section XI, 1989 Edition, no Addenda 8 ASME BPV Code Section XI, 1998 Edition with 2000 Addenda 3.0 DISCUSSION In summary of the issue at hand, MRP-139 requires bare metal visual examination of the nozzle to safe end welds of both the inlet and outlet of the RPV during 3R12 in Fall 2008. These are mandatory requirements under NEI 03-08. However in view of the almost 30 Rem dose impact of the inspections, ALARA principles compel an alternative approach unless the examinations provide an essential increment of assurance and safety that cannot be otherwise obtained.

As a result, MPS3 has developed a justification for waiving the visual examinations while concurrently increasing volumetric inspection frequency, thereby achieving the same objective and intent of the original MRP-139 requirement.

The basis and intent of the visual examination requirements in MRP-139 are discussed in Section 6.10 and 6.11 for examination categories J and K respectively.

Section 6.10.3 states "Visual examination capable of detecting any leakage must be performed in lieu of UT inspections." Section 6.11.3 has a similar statement for category K welds. Visual examinations are required only when UT examinations are not performed.

Therefore the intent of the examination is to detect leakage, as a supplement to the primary strategy of relying on volumetric examinations to Attachment 2, Page 8 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 page 6 of 36 RAy. O0 confirm the absence of initiated flaws. In effect, the visual examinations address any uncertainties regarding the possibility of not detecting existing flaws and of crack growth rates for the PWSCC mechanism.

The approach of this deviation is that such uncertainties may equivalently be addressed by a higher inspection frequency for the volumetric examinations.

Therefore, in lieu of the required visual examination schedule MPS3 plans to rely on volumetric examinations that will be performed on a schedule consistent with the results of a flaw tolerance evaluation.

A table of the inspection plan is provided below and is justified in the following text.Table 2 -Comparison of MRP-139 and MPS3 Inspection Plan Hot Leg (Outlet) Cold Leg (Inlet)MRP Requirements Next and MRP Requirements Next and Thl 6-1 Cat. D and subsequent Thl 6-1 Cat. E and subsequent Thl 6-2 Cat. J inspections Tbl 6-2 Cat. K inspections Volume- Every 5 yrs 3R13, Spring 10 Every 6 yrs 3R14, Fall 12 and trio (UT) and every other every third RFO RFO (every 3 yrs) (every 4.5 yrs)Bare Every RFO except Not required Within 3 RFO (4.5 Not required Metal ones with yrs) of volumetric Visual Volumetric exam Examin-ation (VE)As shown in the above table, the planned volumetric (UT) inspection schedule is at a greater frequency than the generic requirement of MRP-139, compensating for the lack of visual examinations in intervening outages. This schedule will be followed until revised due to mitigation of the affected welds or being superseded by regulatory action. The basis for this schedule is documented in the flaw tolerance evaluation performed by Westinghouse

[4] and included as attachment 3 to this TE. The Westinghouse evaluation is discussed below.The flaw tolerance evaluation postulates an initial flaw and projects its subsequent growth in the interval between examinations based on accepted flaw growth correlations and the limits of flaw stability identified in ASME Section XI IWB-3640.

The 1989 Edition [7] is the basis for the current I1I program at MPS3, while the 1998 Edition 18], which is approved by the NRC, is used for the flaw tolerance evaluation.

The acceptability of a flaw tolerance evaluation as a basis for an alternative to the MRP-139 inspection schedule is based on the example of Section XI Appendix L acceptance of flaw tolerance for actual flaws, and its prior use in similar deviation reports submitted to the MRP.The initial flaw assumption for the flaw tolerance evaluation relies on having performed a recent volumetric examination, with no recordable indications, performed in accordance with qualified Attachment 2, Page 9 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 page 7 of 36 Technical Justification for Deviation from Mandatory Requirements of MRP-139 Rev, 00 UT techniques and techniques that meet MRP-139 Section 5.1.5 coverage requirements.

In the 3R I1 examinations there were no exceptions to coverage requirements, and none are expected in the future. Based on the clean examination results with no detected flaws, a postulated initial circumferential flaw with 10% through-wall depth and limited length is assumed. The axial flaw length assumed is governed by the width'of the weld. The assumed stress field with no major repairs was assumed based on the lack of such repairs for MPS3, as discussed in Section 3.4 of this TE. The flaw growth correlation used is referenced to NUREG/CR-6964 and is consistent with the'MRP-139 recommendation in Section 2.6.2 for A182 materials.

The uprated reactor power RCS temperatures are conservatively used in the flaw growth analysis.The results of the flaw tolerance evaluation are summarized in the flaw growth limit curves contained in Attachment

3. Figure 3-1 shows that the axial flaw growth governs for the inlet (cold leg) nozzle but is not limiting for long periods up to 72 months. For conservatism and to limit the deviation from MRP-139, a limit of 54 months (4.5 years) inspection interval is specified.

For the outlet nozzle (hot leg) the circumferential flaw governs and the higher temperature reduces the allowable inspection interval to less than 46 months. For conservatism a 36 month (3.0 years) inspection interval is specified for this nozzle.In summary, the plant specific flaw tolerance analysis shows with reasonable margin that the selected inspection frequencies for the inlet and outlet nozzles will ensure that an initiating flaw will not propagate to the extent that IWB-3640 limits are exceeded.

In addition, it shows even greater margin against propagation to pressure boundary leakage. It is only this through-wall condition that is detectable by bare metal visual examination.

Therefore, an alternative that waives visual examinations for times prior to challenging IVB-3640 limits does not introduce any significant increment of risk, while allowing a nearly 30 Rem reduction in personnel exposure.

It is thus a justified deviation to MRP-139 requirements.

As a final remark, this evaluation and notification does not meet the usual MRP expectation regarding timeliness of notification.

However the original examination plan for 3R12 had been developed under the assumption that the NRC would soon issue a revision to lGCFR 50.55a requiring inspections in accordance with ASME Code Case N-722, plus additional stipulations that would accompany the rulemaking.

Since the mandated Code Case would have overriding effect on MRP-139, a relief request was prepared in anticipation of this new rule, and no deviation would be required per MRP-139 Section 5.1.7. However, the NRC issuance of the rule change was delayed beyond its original scheduled date such that there is no assurance that review of the proposed relief request would be completed prior to the Fall 2008 outage. Therefore the plan to seek a relief request was modified to instead develop the justification for a deviation from the MRP-139 mandatory visual examination requirements and provide a deviation report to the MRP in accordance with NEI 03-08 Addendum E. The late notification to the MRP is therefore unavoidable.

Attachment 2, Page 10 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 page 8 of 36 Technical Justification for Deviation from Mandatory Requirements of MRP-1 39 Rev. 00 4.0 SAFETY-SIGNIFCANCE This technical evaluation performs an evaluation only and does not implement any change.Since there is no required change to the Technical Specifications, there is no change to the facility, there is no change to a procedure, and there is no test involved with this technical evaluation, no 50.59 screen is required.

5.0 CONCLUSION

Based on the discussions and information presented, the following is concluded." Visual examination of the MPS3 vessel nozzle welds would entail nearly 30 Rem dose impact to personnel and could be avoided by ajustified deviation to MRP-139" Elimination of visual examinations are compensated by an increased frequency of volumetric examinations" A plant specific flaw tolerance evaluation demonstrates that maximum flaw growth during the period between volumetric inspections will remain with ASME Code limits" The objective and intent of MRP-139 visual inspection requirements for the nozzles are satisfied by the alternative volumetric inspection plan Therefore this deviation from the mandatory visual examination requirements is justified for the MPS3 RPV nozzle welds.6.0 ATTACHMENTS I. Independent Reviewer's Comment Sheet (I page)2. Sketch of RPV Nozzle Insulation Package Design (I page)3. Flaw tolerance evaluation LTR-PAFM-08-127 Rev. 2 (Westinghouse non-proprietary)

(19 pages)4. Independent materials expert concurrence opinion letter (6 pages)Attachment 2, Page 11 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attachment I page 9 of 36 I UUiiII I .JULIjIIGLIoII Ir UeVI oILIoUI.lq.l]

IVianyUaLjury LequlrltllelL-UI Ivlr3r"- l.3 rK v. VU Independent Reviewer Comment and Resolution Sheet(s)(ER/EV) No. M3-EV-05-0016 Rev. 0 Page 9 of 10 Independent Reviewer:

Robert Schonenberg Date IComment No. ER/EV Section Comment 1 Maiscellaneous clarifications Comments incorpomted D A-7 7/-4 Attachment 2, Page 12 of 39 C S z In In C In Q C-e EDl C-(1) 0 4, CL CD-W Ne E (D E 9c co (U LilaF LU A Fn//tqvIE;en C en-4 0.)C C 0 S U TRANSCO PRODUCTS, INC, PROCEDURE NO. 50004012.

PAGE 19 I FIGURE I.

10 CFR'50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attachment 3 page 1 I of 36 Technical Justification for Deyiaton from-Mandatory Requirements of MRP-139 Rev. 00 Reference 4 -Flaw Tolerancte Evaluation.

(19 pages follow)Attachment 2, Page 14 of 39 10 CFR 5 0.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0'18 Rev 0 Att. 3 Pg 12 WESTINGHOUSE NON-PROPRIETARY CLASS 3 LTR-PAFM-08-127 Revision 2 Technical Justification for Deviation from MRP-139 Visual Inspection Schedules for Millstone Unit 3 Reactor Vessel Inlet and Outlet Nozzles September 2008 Author: S. F. Hankinson*, Piping Analysis and Fracture Mechanics Verifier:

A. Udyawar*, Piping Analysis and Fracture Mechanics Approved:

S. A. Swamy*, Manager, Piping Analysis and Fracture Mechanics*Electronically approved records are authenticated In the electronic document management system.© 2008 Westinghouse Electric Company LLC All Rights Reserved fWestinghouse Attachment 2, Page 15 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 13 Revision Record Revision Date , Description 0 August 2008 Original Issue 1 September 2008 Incorporate third party review comment by revising the technical justification for the assumed circumferential flaw aspect ratio in Section 2.3 2 September 2008 Incorporate Westinghouse comment on the third party review comment by revising the technical justification for the assumed circumferential flaw aspect ratio in Section 2.3 with concurrence from Dominion.Page 2 of 19 Attachment 2, Page 16 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 14 1.0 Introduction Recent field experiences and the potential for Primary Water Stress Corrosion Cracking (PWSCC) at the Alloy 82/182 dissimilar metal (DM) butt welds require reassessment of the examination frequency and the overall examination strategy for these butt welds.MRP-139 (Reference

1) provided the inspection and evaluation guidelines for the primary system piping dissimilar metal butt welds. Millstone Unit 3 had performed a volumetric and 100% surface examination of the reactor vessel inlet and outlet nozzle to safe end dissimilar metal butt welds during the Spring 2007 outage and no Indications were detected.

For the butt welds at the outlet nozzles, since they are being exposed to the hot leg temperatures, are not made of PWSCC resistant material and also have not been mitigated, visual inspection Is required per MRP-139 In every outage when volumetric examinations are not being performed, until these butt welds are replaced or mitigated.

A less frequent visual inspection schedule Is required for the inlet nozzles per MRP-139 due to the lower normal operating temperature at these nozzles.Flaw tolerance analyses have been performed for the Millstone Unit 3 reactor vessel inlet and outlet nozzle DM welds in order to provide technical justification for deviating from the MRP-139 visual inspection requirements, by not performing visual inspection of the reactor vessel inlet and outlet nozzle butt welds for at least two operating cycles (36 months). The following provides a discussion of the methodology, results and conclusion of the flaw tolerance analysis for both nozzles.2.0 Methodology

2.1 Maximum

End-of-Evaluation Period Flaw Size The maximum end-of-evaluation period flaw sizes for axial and circumferential inside surface flaws at the Alloy 82/182 welds of the Inlet and outlet nozzle are determined using the IWB-3640 evaluation procedure and acceptance criteria in the ASME Section XI Code (Reference

2) including the use of Z-factor for flux welds, The nozzle geometry (Reference

3) for the reactor vessel nozzles Is shown in kTable 2-1. The piping reaction loads from various loading conditions that are used In determining the most limiting end-of-evaluation period flaw sizes are summarized in Appendix A and taken from References 3, 4. and 5.Page 3 of 19 Attachment 2, Page 17 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 15 Table 2-1 Millstone Unit 3 Reactor Nozzle Geometry and Operating Parameters (Reference 3)Inlet Nozzle Outlet Nozzle Outside Diameter (in) 32 'Ia2 34 7/32 Inside Diameter (in) 27 15/32 28 31/32 Thickness (in) 2.500 2.625 Normal Operating 556.4 622.6 Temperature

(°F)2.2 PWSCC Crack Growth Analysis The Millstone Unit 3 reactor vessel inlet and outlet nozzle to safe end dissimilar metal weld regions are made of nickel based alloys. This nickel based alloy material (Alloy 821182) is susceptible to PWSCC crack growth mechanism.

The PWSCC crack growth rate used In the crack growth analysis is based on the EPRI recommended crack growth curves for Alloy 182 material (Reference

6) and shown below.exp. .4 2(l/T 11/Toor}(K)O wt hRe where: da dt-Crack growth rate in m/sec= Thermal activation energy for crack growth =130 kJ/mole (31.0 kcal/mole)

R = Universal gas constant = 8.314 x 10-3 kJ/mole-K (1.103 x 10.3 kcal/mole-°R)

T Absolute operating temperature at the location of crack (K or *R)Tref = Absolute reference temperature used to normalize data = 5981.15 K (1076.67-R) a Crack growth amplitude= 1.50 x 10.12 at 3250C (617-F)= Exponent = 1.6 K = Crack tip stress Intensity factor (MPa 4m)Page 4 of 19 Attachment 2, Page 18 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-B9V-08-0018" Rev 0 Att. 3 Pg 16 It should be noted that the PWSCC crack growth mechanism is applicable only to the inside surface flaws since they are exposed to the primary water environment.

The stresses used for PWSCC evaluations included normal operating condition piping reaction loads, pressure, and residual stresses at the DM welds. The normal operating temperatures for the inlet and outlet nozzles are 556.4 0 F and 622.6 0 F respectively, The Impact of fatigue crack growth mechanism is considered in the flaw tolerance analysis.

Fatigue crack growth is negligible, especially for short plant operation duration (2 to 3 refueling cycles) when compared to that due to PWSCC because the locations of interest at the inlet and outlet nozzles are not subjected to any significant thermal transient loadings.The residual stresses considered In the analyses were based on the reactor vessel nozzle residual stress profiles from Reference 7 for the case with no Inside surface weld repair. This Is acceptable since a review of all the available manufacturing records for the reactor vessel did not show any significant Inside surface weld repairs made to either the inlet or outlet nozzle dissimilar metal welds (Reference 8).Using the applicable stresses at the DM welds, the crack tip stress intensity factors can be determined based on the stress intensity factor expressions from References 9 and 10. The through-wall stress distribution profile is represented by a cubic polynomial:

a (x) =A 0 + A 1 x + A 2 x 2 + A 3 X 3 where: Ao, A,, A 2 , and A 3 are the stress profile curve fitting coefficients, x is the distance from the wall surface where the crack initiates, and a is the stress perpendicular to the plane of the crack.The stress intensity factor calculations for semi-elliptical inside surface flaws with various aspect ratios (flaw iengthldepth) for axial and circumferential flaws are performed.

The Influence coefficient at any points on the crack front can be obtained by using an interpolation method. The crack tip stress Intensity factors can be expressed in the general form as follows: , =)O j (ac, aft, t/R, D) Aj a)where: a: Crack Depth Page 5 of 19 Attachment 2, Page 19 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 17 c: Half Crack Length Along Surface t: Thickness of Cylinder R: Inside Radius (D Angular Position of a Point on the Crack Front Gj: G, is Influence coefficient forj a stress distribution on crack surface (i.e., Go, G 1 , G 2 , G 3).Q: The shape factor of an elliptical crack, which is the square of the complete elliptical integral of the second kind or Shape Factor=[ J(cos2D +1--sin2 C))2 d(l]2.Q is approximated by: 0 C Q = 1 + 1.464(alc)-'

6' for a/c < 1 or Q = 1 + 1.464(c/a)"6' for a/c > 1.Once the crack tip stress intensity factors are determined, PWSCC crack growth calculations can be performed using the crack growth rate discussed in Section 2.2 for the applicable normal operating temperature.

2.3 Maximum

Undetected Flaw size The initial flaw size used In the flaw tolerance analysis is assumed to be the maximum undetected flaw size since no Indications were detected during the Spring 2007 volumetric and surface examination.

The maximum undetected flaw depth is assumed to be 10% of the wall thickness.

This assumed flaw depth is similar to the In-service inspection acceptance criteria in Table IWB-3514-2 of the ASME Section Xl Code for returning components into service and therefore is a conservative and reasonable assumption.

An aspect ratio (flaw length/depth) of 2 is assumed for the axial flaw since PWSCC is limited to the width of the A82/182 weld. For the circumferential flaw, an aspect ratio (AR) of 6 is assumed. As for the circumferential flaw, an Initial flaw depth of 0.25 inch (10% through wall) and initial flaw length of 1.5 inches (aspect ratio of 6) is conservatively assumed for the inlet nozzle. Assuming the same aspect ratio for the outlet nozzle, the initial flaw length is assumed to be 1.58 inch. Since no detectable flaws were found in the dissimilar metal welds of these nozzles during the spring 2007 volumetric examination, it is considered highly unlikely, with a qualified volumetric examination, a flaw of this size would go undetected.

3.0 Flaw Tolerance Analysis Results Figures 3-1 to 3-4 display the maximum allowable initial flaw size for the axial and circumferential flaws at the nozzle to safe end Alloy 82/182 welds for the Inlet and Outlet nozzles based on the IWB-3640 acceptance criteria.

The horizontal axis displays the flaw depth to length ratio or the inverse of the flaw aspect ratio. The vertical axis shows, the flaw depth to wall thickness ratio (a/t). The flaw evaluation chart displays allowable flaw size curves for plant operation duration up to 54 months. If the flaw parameters of a given flaw fall below the allowable flaw size curve for a given plant operation duration, then the flaw will not grow to the maximum end-of-evaluation period allowable flaw size within that plant operation duration.

For comparison purposes, the maximum undetected Page 6 of 19 Attachment 2, Page 20 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 18 flaw size as discussed In Section 2.3 is also shown in Figures 3-1 to 3-4 to show the available margins for this assumed initial flaw size.Figures 3-5 to 3-8 display the maximum allowable initial flaw size for the axial and circumferential flaws at the nozzle to safe end Alloy 82/182 welds for the Inlet and Outlet nozzles based on leakage instead of limit load failure. Leakage Is assumed to occur once the Initial inside surface flaw becomes a 100% through-wall flaw. If the flaw parameters of a given flaw fall below the allowable flaw size curve for a given plant operation duration, then the flaw will not grow to a 100% through-wall flaw within that plant operation duration.

For comparison purposes, the maximum undetected flaw size as discussed in Section 2.3 is also shown in Figures 3-5 to 3-8 to show the available margins for this assumed initial flaw. The margins shown are slightly larger than those based on the IWB-3640 acceptance criteria.As shown in Figures 3-3 and 3-4, the flaw tolerance result for the outlet nozzle is more limiting and continued plant operation duration of only 36 months is acceptable for the assumed undetected flaw size. There is adequate margin for the inlet nozzle (Figures 3-1 and 3-2) for continued plant operation duration of 54 months. Additionally, this margin is demonstrated by the 72 month curves identified in the Inlet nozzle flaw tolerance charts. Since no Indications were detected during the Spring 2007 refueling outage, crack growth due to PWSCC for the maximum undetected flaw size would not reach the end-of-evaluation period allowable flaw size per IWB-3640 or result in leakage for continued plant operation of at least 36 months for the reactor vessel Inlet and outlet nozzles.PWSCC crack growth curves for the limiting reactor vessel outlet nozzles are shown in Figures 3-9 to 3-10 for axial (AR=2) and circumferential flaw (AR=6) respectively with the initial flaw size equals to the assumed maximum undetectable flaw size. The horizontal axis displays the service life In effective full power months (EFPM), while the vertical axis shows the flaw depth to wall thickness ratio (alt). These curves demonstrated the service life required to reach the IWB-3640 acceptable flaw size and a 100% through-wall flaw.Based on the IWB-3640 end-of-evaluation period allowable flaw size, it would take at least 48 EFPM for an axial flaw (AR=2), with an initial flaw depth of a/t=0.10, to reach the end-of-evaluation period allowable flaw depth. For a circumferential flaw (AR--6) with the same initial flaw depth, it would take 46.2 EFPM to reach the end-of evaluation period allowable flaw depth. The service life required is therefore more than 2 operating cycles (36 months) at Millstone Unit 3. Also as illustrated in Figures 3-9 and 3-10, the service life required to reach 100% through-wall thickness is slightly longer.Page 7 of 19 Attachment 2, Page 21 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 19 Figure 3-1 Maximum Initial Acceptable Axial Flaw (IWB-3540 Criterfa)Based on PWSCC Growth (Mllstone Unit 3 RV Inlet Nozle)I 0.40 30 months 54 months 72 manths 0.3D 1 ...0.10 Maximum Assumed Undetected Flaw Size 0.00 .. .1 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Depth / Flaw Length Ratio Figure 3-2 Maximum Initial Acceptable Circumferentlal Flaw (MWB-3640 Criteria)Based on PWSCC Growth (Millstone Unit 3 RV Inlet Nozzle)0.80 ....0.70 ____ ____30 " wnths 54 months 72 months 0.40 _ _0.30 ,__i0.0Maxim~um Assumed.2 t"Undetected Flaw Size 0.t0 0.00 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.46 0.50 Flaw Depth I Flaw Length Ratio Page 8 of 19 Attachment 2, Page 22 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 20 Figure 3-3 Maximum Initial Acceptable Axial Flaw (IWB.3640 Criteria)Based on PWSCC Growth (Millstone Unit 3 RV Outlet Nozzle)0.25 -jO.2 0 -0.15 A0.10 0.05 0.00 0.10 0.15 020 0.25 0.30 0.35 0.40 0.45 0.50 Flaw Depth I Flaw Length Ratio Figure 3-4 Maximum Initial Acceptabla Circumferential raw (IWO-3640 Criteria)Based on PWSCC Growth (Millstone Unit 3 RV Outlet Nozzle)0.70 0.60 0.30 0.20 U 0.00 0.10 0.15 0.20 0.25 0.30 0.35 Flaw Depth I Flaw Length Ratio 0.40 0.45 0.50 Page 9 of 19 Attachment 2, Page 23 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 21 Figure 3-5 Maximum Inlital Accoptable Axial Flaw (Leakage Criteria)Based on PWSCC Growth (Millstone Unit 3 RV Inlet Nozzle)-0.50 ____ ___o 36 months 54 months 72 months oI o u.W L =.0.30 _ __....0.10 Maximum Assumed Undetected Maw Size 0.00 '-0.10 0.15 0.20 0.25 0.30 0.35 Flaw Depth I Flaw Length Ratio 0A0 0.45 0.50 Figure 3-6 Maximum Initial Acceptable Circumferential Flaw (Leakage Criteria)Based on FWSCC Growth (Millstone UnIt 3 RV Inlet Nozzle)0.10 0.15 0.20 0.25 0.30 0.35 Flow Depth I Flaw Length Ratio 0.40 0.45 0.50 Page 10 of 19 Attachment 2, Page 24 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0' Att. 3 Pg 22 Figure 3-7 Maximum Initial Acceptable Axial Flaw (Leakage Criteria)Based on PWSCC Growth (Millstone Unit 3 RV outlet Nozzle)0.10 0.15 0.20 0.2S 0.30 0.35 0.40 0.45 0.50 Flaw Depth I Flaw Length Ratio Figure 3-8 Maximum Initial Acceptable Circumferential Flaw (Leakage Criteria)Based on PWSCC Growth.(Millstone Unit 3 RV Outlet Nozzle)oo 0.40= 0.35 0.20 S 0.25 0.30 0.35 Flaw Depth I Flaw Length Ratio Page 11 of 19 Attachment 2, Page 25 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 23 Figure 3-9 PWSCC Axial Crack Growth Curves for Outlet Nozzle Alloy 82/182 Weld Region Millstone Unit 3 RV Outlet Nozzle PWSCC Growth (Axial Flaw, Aspect Ratio of 2, Normal Operating Temperature 622.6 Deg F)-r- -- ... .. "1_0...0.2 1.0 0.9 i 068- --; ; ;__, ..+ -

IW B-3640 Cdteda (a/t;=0.75)

.; __ _" _. __ _-----------

-...p o- ,- --- , -, --I_ V.4., ,, ...._ _ _al -- Rio- -- .1 -- --- --- 010- -- -----01 , ' , ,,I li~~~~i~~~i.,~~

---

.. ....' .--.--L -.. i._n n -., .;_.. ..... .. .........

....+ _0 10 20 30 Effective Full Power Months 40 50 Page 12 of 19 Attachment 2, Page 26 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 P9 24 Figure 3-10 PWSCC Clrcumferential Crack Growth Curves for Outlet Nozzle Alloy 82/182 Weld Region Millstone Unit 3 RV Outlet Nozzle PWSCC Growth (Circumferential Flaw, Aspect Ratio of 6, Normal Operating Temperature

= 622.6 Deg F).... 100% Through-Wall

7. , , , , , a .i .-'- ... .--r -r ---'r ---r -r J ¢ , ....:" --,. ....---.. .. ........9 I , , 0 a a a U 0.C (U I..1.0 0.9 0.7nteaa-t 074) , , 07 -i .. .. -,- -462- -' nths- ... .r --,. ....0 5,. .... -0.54-" 'I .... ... ... " ., !0.3 L ----r -r ----r ...r -r --r + -r ....r ------ff ....r --r -....r .......---- ---0.3 Initial alt Ratio= 0.10 -. ........-". ... ... ....,. ..,.. .'----0.04 0 10 20 30 Effective Full Power Months 40 50 Page 13 of 19 Attachment 2, Page 27 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 25 4.0 Discussion and Conclusion The required visual inspection schedules for the inlet and outlet nozzles are shown in Table 6-2 of MRP-139. For the outlet nozzle, visual inspection is required in every outage when volumetric examinations are not being performed until the nozzle is being mitigated or replaced.

The required volumetric examination for the outlet nozzle is every 5 years per Table 6-1 of EPRI Report MRP-139. Based on the MRP-139 volumetric examination schedule, the Millstone Unit 3 outlet nozzle would perform volumetric inspection every 3 refueling cycles since the refueling cycle interval for Millstone Unit 3 is 18 months. Based on the flaw tolerance results shown in Figures 3-3 and 3-4, it is acceptable to deviate from the MRP-139 visual inspection schedule by performing a visual inspection every other refueling outage when volumetric examinations are not being performed Instead of every outage.For the inlet nozzle, visual inspection Is required once every three refueling cycles until the nozzle Is being mitigated or replaced.

The required volumetric examination for the inlet nozzle Is every 6 years per Table 6-1 of EPRI Report MRP-139. Based on this volumetric examination schedule, the Millstone Unit 3 inlet nozzle would be Inspected every four refueling outages. Per Table 6-2 of MRP-139, deterministic analysis can be used as a basis to allow the inlet nozzle DM welds to be visually examined at a frequency less than once every three refueling outages. Based on the flaw tolerance analysis performed, the results shown in Figures 3-1 and 3-2 demonstrated that there is adequate margin to support deviation from the MRP-139 visual inspection schedule for the inlet nozzle.In summary, since no indications were detected during the Spring 2007 refueling outage, crack growth due to PWSCC for the maximum undetected flaw size would not reach the end-of-evaluation period allowable flaw size per IWB-3640 or result in leakage for continued plant operation duration of at least 36 months for the inlet and outlet nozzles.Based on the results of the flaw tolerance analysis, it Is technically justified to seek a less frequent visual inspection schedule than those required in MRP-139 for both the reactor vessel inlet and outlet nozzle dissimilar metal weld regions.Page 14 of 19 Attachment 2, Page 28 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 26 5.0 References

1. Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139), EPRI, Palo Alto, CA: 2005. 1010087. (EPRI Proprietary Document)2. Rules for Inservice Inspection of Nuclear Power Plant Components, ASME Boiler &Pressure Vessel Code,Section XI, 1998 Edition through 2000 Addenda.3. Dominion Document 25212-ER-08-0024 dated 5/1412008, "Reactor Vessel Nozzle Evaluation Input for NEU-08-29" (Dominion Proprietary Document)4. Stone and Webster Engineering Corporation (SWEC) Document No. 12179-NP(B)-

x7001A -"Pipe Stress Analysis:

Reactor Coolant System Piping and Associated Branch Connections loops 1 and 3 ASME III Code Class 1 and 2" -Revision 0 -Boston, MA.5. Stone and Webster Engineering Corporation (SWEC) Document No. 12179-NP(B)-

x7002A -"Pipe Stress Analysis:

Reactor Coolant System Piping and Associated Branch Connections loops 2 and 4 ASME III Code Class 1 and 2' -Revision 0 -Boston, MA.6. NUREGICR-6964 ANL-07/12, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments.

7. Materials Reliability Program: Alloy 821182 Pipe Butt Weld Safety Assessment for US PWR Plant Designs (MRP-113), EPRI, Palo Alto, CA: 2005. 1009549. (EPRI Proprietary Document)8. Westinghouse Letter LTR-PCAM-07-21, "Millstone Unit 3 -PWROG PA-MSC-0233 -Task 2 Customer Deliverable

-Primary Pressure Boundary Alloy 600/82/182 Fabrication Detail," February 23, 2007. (Westinghouse Proprietary Document)9. Raju, I. S. and Newman, J. C., "Stress Intensity Factor Influence Coefficients for Internal and External Surface Cracks in Cylindrical Vessels," ASME Publication PVP. Volume 58,1982, pp. 37-48.10. Mettu, S. R., Raju, I. S., and Forman, R. G., NASA Lyndon B. Johnson Space Center report no. NASA-TM-111707, "Stress Intensity Factors for Part-through Surface Cracks in Hollow Cylinders," in Structures and Mechanics Division, July 1992.Page 15 of 19 Attachment 2, Page 29 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 27 Appendix A Millstone Unit 3 Reactor Vessel Inlet and Outlet Nozzle Loads Page 16 of 19 Attachment 2, Page 30 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 28 Table A-1 Inlet Nozzle Loads -Table 1 of 3 Fodrces Moments (in-Kips)Loading Reference (kips)Fx Mx My Mz+28 0 +1621 0 Thermal 4905 -921 -8105 Operating Pressure 3 1374 113 -316 -69 Inlet Deadweight 3 4 3 -401 1 -1108 Nozzle OBE Inertia 4 40 1354 2345 2526 OBESAM 4 39 71 112 181 SSE Inertia 4 49 1350 2409 2356 SSE SAM 4 61 108 177 274 Note: SAM = Seismic Anchor Motion Table A-2 Inlet Nozzle Loads -Table 2 of 3 Forces Moments (in-Kips)Loading Reference (kips)Fx Mx My Mz+34 0 +1711 +7277 Thermal , 4909 -1371 0 Operating Pressure 3 1376 -305 533 -113 Inlet Deadweight 3, 5 3 69 90 Nozzle OBE Inertia 5 42 1377 2150 2113 OBE SAM 5 39 71 213 162 SSE Inertia 5 53 1369 2189 1972 SSE SAM 5 61 108 335 245 Note: SAM = Seismic Anchor Motion Page 17 of 19 Attachment 2, Page 31 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att 3 Pg 29 Table A-3 Inlet Nozzle Loads -Table 3 of 3 Forces Loading Reference (kips) Moments (in-Kips)Fx Mx My Mz+34 0 +1711 +7277 Thermal 3, 5 4909 -1371 0 Operating Pressure 3 1376' -305 533 -113 Deadweight 3,5 3 69 90 OBE Inertia 5 42 1377 2150 2113 OBE SAM 5 39 71 213 162 Inlet SSE Inertia 5 53 1369 2189 1972 Nozzle SSE SAM 5 61 108 335 245+245 +913 +3909 +1705-533 -1132 -6865 -1003+387 +7756 +52699 0-500 0 -24678 -14413+74 +927 +3630 +2164-441 -1101 -4790 -1436 Note: SAM = Seismic Anchor Motion Table A-4 Outlet Nozzle Loads -Table I of 3 Forces Forces Moments (in-Kips)Loading Reference (kips)Fx Mx My Mz+35 +43 +3813 0 Thermal 3, 4 348 -2570 -13444 Operating Pressure 3 1511 -27 341 -1095 Outlet Deadweight 3,4 2 29 -114 -728 Nozzle OBE Inertia 4 194 547 3813 3670 OBESAM 4 45 166 225 366 SSE Inertia 4 254 630 3544 3351 SSE SAM 4 71 251 358 554 Note: SAM = Seismic Anchor Motion Page 18 of 19 Attachment 2, Page 32 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Rev 0 Att. 3 Pg 30 Table A-5 Outlet Nozzle Loads -Table 2 of 3 Forces Moments (in-Kips)Loading Reference (kips) Mmt(Fx Mx my Mz-27 +738 +2071 0 Thermal 3, 5 +90 0 -4889 +12545 Operating Pressure 3 1509 70 -649 -237 Outlet Deadweight 3, 5 0 -27 +50 -2550 Nozzle OBE Inertia 5 182 862 3488 3239 OBE SAM 5 50 154 227 319 SSE Inertia 5 236 965 3214 2956 SSE SAM 5 80 234 360 483 Note: SAM = Seismic Anchor Motion Table A-6 Outlet Nozzle Loads -Table 3 of 3 Forces Moet(i-ps Loading Reference (kips) Moments (in-Kips)Fx Mx My Mz Teml35 +90 0 +2071 + 12545-27 -738 -4889 0 Operating Pressure 3 1509 70 -649 -237 Deadweight 3,5 0 -27 +50 -2550 OBE Inertia 5 182 862 3488 3239 OBE SAM 5 50 154 227 319 Outlet SSE Inertia 5 236 965 3214 2956 Nozzle SSE SAM 5 80 234 360 483 Bek9-RR 3+803 +3810 +34494 0-14 -3428 0 -48087 Bek1- 3+492 +1238 +2538 +5391-55 -1661 -4114 -5296 Brak11-Srg 3+567 +4086 1+387331 +42348 Break 11 -Surge 3-24 25 0 0-274 -4250 0 0 Note: SAM = Seismic Anchor Motion Page 19 of 19 Attachment 2, Page 33 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attachment 4 page 31 of 36 Technical Justification for DevIatIon from Mandatory Requirements of MRP-139 Rev. 00 Independent Materials Expert Concurrence W. McBrine, Altran Solutions (5 pages follow)Attachment 2, Page 34 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attach. 4 Pg. 32 aLTRan SOLU-IONS www.altransolut[ons.com 451 D Street Phone: 617-204-1000 Fax: 617-204-1010 Boston, MA 02210 September 25, 2008 08-0419-L-001 Mr. Steven D. Janes Dominion Nuclear Connecticut Millstone Power Station Rope Ferry Road (Route 156)Waterford, CT 06385

SUBJECT:

Transmittal of Altran Design Verification Report 08-0419-VR-001, Rev. 0, "Third Party Review of the Technical Justification for Deviation from MRP-139 Visual Inspection Schedule, Millstone Point Unit 3" REF: Dominion Purchase Order 70187510, dated 911112008.

Dear Mr. Janes:

Please find enclosed the original copy of Altran Design Verification Report 08-0419-VR-001, Rev. 0. This report documents the third party review that Altran performed on Dominion Nuclear Connecticut Technical Evaluation M3-EV-08-0018, Rev. 0, "Technical Evaluation for Technical Justification for Deviation from Mandatory Requirements of MRP-139, Millstone Unit Three." Altran appreciates the opportunity to be of service to Dominion Nuclear. If you have any questions or comments, please do not hesitate to call Bill McBrine at (617) 204-1000.Very truly yours, ALTRAN CORPORATION Chock, Jr. (Technical Lead- Mechanical Engineering William J. McBrine Technical Manager -Materials Engineering Enclosure Attachment 2, Page 35 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attach. 4 Pg. 33 ALTRAN VERIFICATION REPORT VRNo.: 08-0419-VR-001 Project No. 08-0419 Page 1 of 4 Design, Analysis,.Test, or Examination Verified:

<n/a>Document Verified&Dominion Nuclear Connecticut Technical Evaluation M3-EV-08-0018, Rev. 0, "Technical Evaluation for Technical Justification for Deviation from Mandatory Requirements of MRP-139, Millstone Unit Three." Method of Verification:

j_ Independeat Review _Alternate Calculation Testing Qualification Summary of Verification At the request of Dominion Nuclear Connecticut, Altran Solutions Corporation performed a third-party review of a technical justification for deviation from the MRP-139 visual inspection schedules for Millstone Point Unit 3 (MP3) reactor vessel input and output nozzles. The results of this review are discussed in the following sections.Documents Reviewed 1. Dominion Nuclear Connecticut, "Technical Evaluation for Technical Justification for Deviation from Mandatory Requirements of MRP-139, Millstone Unit Three". Tech.Eval. No. M3-EV-08-0018, Rev. 0, September, 2008.2. Westinghouse Letter LTR-PAFM-08-127 (Non-Proprietary), "Technical Justification for Deviation from MRP-139 Visual Inspection Schedules for Millstone Unit 3 Reactor Vessel Inlet and Outlet Nozzles", Rev. 2, September, 2008.Background Recent field experiences and the potential for Primary Water Stress Corrosion Cracking (PWSCC) at the Alloy 82/182 dissimilar metal (DM) butt welds require reassessment of the examination frequency and the overall examination strategy for these butt welds. EPRI MRP-139 provides the inspection and evaluation guidelines for the primary system piping dissimilar metal butt welds. For butt welds at the outlet nozzles, that are exposed to hot leg temperatures, are not made of PWSCC resistant material and also have not been mitigated, MRP-139 requires visual inspection is required at every outage when volumetric examinations are not being performed, until these welds are replaced or mitigated.

A less frequent visual inspection schedule is required Attachment 2, Page 36 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attach. 4 Pg. 34 Altran Solutions Page 2 of4 Verification Report 08-0419-VR-001, Rev. 0 for the inlet nozzles per MRP-139 due to the lower normal operating temperature at these nozzles.MP3 performed a volumetric examination of the reactor vessel inlet and outlet nozzle to safe end dissimilar metal butt welds during the Spring 2007 outage. At that time, no indications were detected.

To provide relief from the MRP-139 visual inspection schedule requirements, Westinghouse performed a flaw tolerance analysis (see Document 1) of the MP3 RV inlet and outlet nozzle DM welds. This analysis demonstrated that the next visual inspection of the reactor vessel inlet and outlet nozzle butt welds would not be necessary for at least two operating cycles (36 months).Technical Approach Altran's review of the two documents assessed the adequacy and presentation of the following: " Criteria (i.e., applicability to the requirements of MRP-139)" Methodology

• Selection of suitable input" Tabulated results" Conclusions of the evaluation.

Reference Documents As part of the review process, the following documents were examined.

These documents are either commercially available or comprise the design basis of Millstone Unit 3.1. Electric Power Research Institute, Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139), EPRI Report 1010087.(EPRI Proprietary Document).

Palo Alto, CA: 2005.2. American Society of Mechanical Engineers, "Rules for Inservice Inspection of Nuclear Power Plant Components", ASME Boiler & Pressure Vessel Code,Section XI, 1998 Edition through 2000 Addenda.3. Argonne National Laboratory, Crack Growth Rates and Metallographic Examinations of Alloy 600 and Alloy 82/182 from Field Components and Laboratory Materials Tested in PWR Environments, NUREG/CR-6964.

Argonne, IL: U.S. Nuclear Regulatory Commission, Office of Nuclear Regulatory Research, 2008.4. Nuclear Energy Institute, "Guidelines for the Management of Materials Issues", NEI 03-08, Rev. I. Washington, DC: April 2007.5. Electric Power Research Institute, Material Reliability Program: Alloy 82/182 Pipe Butt Weld Safety Assesment for US PWP Plant Designs (MRP-1 13), EPRI Report 10 07029.(EPRI Proprietary Document).

Palo Alto, CA: 2004.Attachment 2, Page 37 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attach. 4 Pg. 35 Altran Solutions Verification Report 08-0419-VR-001, Rev. 0 Page 3 of 4 Conclusions As a result of the review, Altran Solutions Corporation has made the fbllowing findings: 1. Altran concurs that the deviation set forth in Technical Evaluation M3-EV-08-0018 satisfies the objective and intent of MRP-139.2. Altran further finds that the technical arguments in support of Technical Evaluation M3-EV-08 are satisfactory, and that they accurately incorporate the basis provided in Westinghouse Document LTR-PAFM-08-127.

Qualifications of Reviewers The third-party review was conducted by William MoBrine, PE, with contributions from Edmund Dunn, Sc.D. and Bahaa Elaidi, Ph.D. A short summary of team member qualifications is provided in Attachment A. Full professional resumes are available upon request.Statement of Concurrence Having performed a third-party review of Domnion Technical Evaluation M3-EV-08-0018, Rev.0 in the role of Independent Materials Expert, Altran Solutions hereby states its concurrence with the technical evaluation and the results herein.Mr. McBrine has affixed his endorsement as Independent Materials Expert to Millstone Technical Evaluation M3-EV-08-0018.

Wi1lm! .Merine, PS. Techical Lead A- MEidi, Ph.D., Contributor Edmund M. Dunn, Sc-D., Contributor Date Date Date Attachment 2, Page 38 of 39 10 CFR 50.55a Request Number IR-3-10 Attachment 2 (Continued)

M3-EV-08-0018 Attach. 4 Pg. 36 Altran Solutions Page 4 of 4 Verification Report 08-0419-VR-001, Rev. 0 ATTACHMENT A

SUMMARY

OF QUALIFICATIONS OF REVIEWERS William J. MeBrine, PE Technical Lead William McBrine is the Technical Manager of the Materials Engineering Group at Altran Solutions.

Mr. McBrine has 30 years of experience in the nuclear power industry with particular expertise in addressing structural integrity issues. He has extensive experience in the assessment of degraded mechanical components, including failure analysis, flaw evaluations and remaining life prediction.

He has led projects investigating Alloy 600 issues including the prediction of SCC crack growth rate and influencing factors. Mr. McBrine also has extensive experience in stress analysis, fracture mechanics and qualifications to ASME B&PV Sections III and XI requirements.

Bahaa A. Elaidi, Ph.D.Dr. Elaidi is the Technical Manager of Structural Engineering and Engineering Mechanics at Altran Solutions.

He has over 25 years of experience in applied mechanics, failure analysis, and root cause evaluation, with a diverse background in analysis, inspection, and repair of civil and mechanical systems and components.

Previous applicable work includes investigation of cracking in steam generator tubes, establishment of critical flaw sizes welded joints of piping and spent fuel canisters, failure analyses and life assessment of nuclear plant components, and analytical modeling of flaws and crack growth.Edmund M. Dunn, Sc.D.Edmund M. Dunn has over 30 years of experience in Materials Science and Engineering with core expertise in solidification metallurgy, brazing and welding. His experience includes Bettis Atomic Power Laboratory, and GTE Laboratories.

While at Bettis, his work included studies on factors affecting stress corrosion cracking and weld hot cracking in reactor plant materials (Ni-Cr-Fe Alloy 600).He has served as chair of a national committee, the TMS/AIME Solidification Committee and is the author or coauthor of numerous papers and five patents. His work has included materials selection, market evaluation, process improvement, and failure analysis.

He is a member of ASM and TMS/AMIE.

He has been Secretary of ASM International Boston Chapter.Dr. Dunn received an Sc.D. in Materials Science and Engineering from MIT, a B.S. in Materials Engineering from RPI, and an MBA from the University of California at Berkeley.Attachment 2, Page 39 of 39 10 CFR 50.55a Request Number IR-3-11 Proposed Alternative In Accordance with 10 CFR 50.55a(a)(3)(ii)

--Hardship or Unusual Difficulty without a Compensating Increase in Level of Quality or Safety--1. ASME Code Components Affected ASME Code Class: Code Class 1

References:

ASME Section XI, Table IWB-2500-1 and IWB-5222 Examination Category:

B-P (All Pressure Retaining Components)

Item Number: B15.10

Description:

Alternative Pressure Testing Requirements for the RPV Flange Leak-Off Piping Components:

NPS 1 RPV Flange Seal Leak-Off Piping 2. Applicable Code Edition and Addenda ASME Section XI, 2004 Edition (No Addenda)3. Applicable Code Requirement IWB-2500, Table 1WB-2500-1, Code Category B-P, Item Number B 15.10 requires that all Class 1 pressure retaining components be Visual, VT-2 examined each refueling outage. The required system pressure test can be either a hydrostatic test oir a system leakage test. The system leakage test is performed at a pressure not less than the pressure corresponding to 100% rated reactor power. Per IWB-5222(a), the pressure retaining boundary during the system leakage test shall correspond to the reactor coolant boundary, with all valves in the position required for normal reactor operation startup. The visual examination shall, however, extend to and include the second closed valve at the boundary extremity.

Per IWB-5222(b), the pressure retaining boundary during the system leakage test conducted at or near the end of the interval shall extend to all Class, 1 pressure retaining components within the system boundary.4. Reason for Request As discussed in 3, "Applicable Code Requirements," ASME Section XI, 2004 Edition (No Addenda) requires that Class 1 pressure boundary piping shall be pressure tested after each refueling outage. The Reactor Pressure Vessel (RPV) head flange seal leak detection piping is shown in Attachment

1. The piping is separated from the reactor coolant pressure boundary by one passive membrane, which is an o-ring located on the inner vessel flange. A second o-ring is located on the outside of the tap in the vessel flange. Failure of the inner o-Page 1 of 3 10 CFR 50.55a Request Number IR-3-11 (Continued) ring is the only condition under which this line is pressurized.

Therefore, the line is not expected to be pressurized during the system pressure test following a refueling outage.The configuration of this piping precludes system pressure testing while the vessel head is removed because the configuration of the vessel tap coupled with the high test pressure prevents the tap in the flange from being temporarily plugged or connected to other piping.The opening in the flange is smooth walled, making the effectiveness of a temporary seal very limited. Failure of a temporary test seal could possibly cause ejection of the device used for plugging or connecting to the vessel flange.The configuration also precludes pressurizing the line externally with the head installed.

The top head of the vessel contains two grooves that hold the o-rings. The o-rings are held in place by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were to be performed with the head on, the inner o-ring would be pressurized in a direction opposite to its design function.

This test pressure would result in a net inward force on the inner o-ring that would tend to push it into the recessed cavity that houses the retainer clips. The thin o-ring material would likely be damaged by the inward force.Purposely failing or not installing the inner o-ring in order to perform a pressure test would require purchasing a new o-ring set and the time and radiation exposure associated with removing and reinstalling the RPV head to replace the o-rings would be an undue hardship.Considering this information, compliance with the IWB-5222(b) system pressure test requirements results in unnecessary hardship without sufficient compensating increase in the level of quality and safety.5. Proposed Alternative and Basis for Use In lieu of the requirements of IWB-5222(b), a VT-2 visual examination will be performed each outage on the unpressurized subject piping as part of the Class 1 leakage test. If the inner o-ring should leak during the operating cycle it will be identified by an increase in temperature of the leak-off line above ambient temperature because this is an indication of o-ring seal leakage. This high temperature would actuate an alarm in the Control Room, which would be closely monitored by procedurally controlled operator actions allowing identification of any further compensatory actions required.

This leakage would be collected in the primary drain transfer tank.Additionally, the flange seal leak-off line is essentially a leakage collection/detection system and the line would only function as a Class 1 pressure boundary if the inner o-ring fails, thereby pressurizing the line. If any significant leakage does occur in the leak-off line piping itself during this time of pressurization then it would clearly exhibit boric acid accumulation and be discernable during the proposed VT-2 visual examination that will be performed unpressurized as proposed in this request.6. Duration of Proposed Alternative This relief is requested for the duration of the Third Inservice Inspection Interval, which begins on April 23, 2009, and is scheduled'to end on April 22, 2019.Page 2 of 3 10 CFR 50.55a Request Number IR-3-11 (Continued)

7. Precedents North Anna Unit 1 Relief Request SPT-013, "Examination Category B-P Pressure Retaining Components in the Reactor Coolant System," approved by NRC letter dated February 9, 2006, ADAMS Accession No. ML060450517 Page 3 of 3 10 CFR 50.55a Request Number IR-3-11 Attachment 1 Reactor Pressure Vessel Seal Leak-Off Details DOMINION Millstone 3 Page 1 of 3 10 CFR 50.55a Request Number IR-3-11 Attachment 1 (Continued)

Figure 1: REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF LINE CONFIGURATION Page 2 of 3 10 CFR 50.55a Request Number IR-3-11 Attachment 1 (Continued)

QC b I'ilt ftv llu Lelk/ Inrirfl wiqen Sti Petfiu W A l~etail A Vessel Flange Sectionol viaew Figure 2 REACTOR PRESSURE VESSEL HEAD FLANGE LEAK-OFF LINE DETAILS Page 3 of 3