ML022610153

Fourth Interval Inservice Inspection Program Plan, Attachments a and B

ML022610153
Person / Time
Site:	Dresden
Issue date:	09/06/2002
From:	Hovey R Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
RHLTR: #02-0054
Download: ML022610153 (138)

Text

ATTACHMENT A Fourth Interval Dresden Nuclear Power Station ISI Program Plan

DRESDEN NUCLEAR POWER STATION UNITS 2 & 3 ISI Program Plan Fourth Ten-Year Inspection Interval Commercial Service Dates:

Unit 2 - 06/09/72 Unit 3 - 11/16/71 Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450 Exelon Generation Company (EGC, LLC) 220 Exelon Way Kennett Square, PA 19348

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval REVISION CONTROL SHEET Revision Date Revision Summary 0 8/30/02 Initial Issue.

Note: This ISI Program Plan (Sections 1 - 10 inclusive) is controlled by the Dresden Nuclear Power Station Engineering Programs Group.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval REVISION

SUMMARY

Section Effective Pages Revision Date i to iii 0 8/30/02 1.0 1-1 to 1-6 0 8/30/02 2.0 2-1 to 2-13 0 8/30/02 3.0 3-1 to 3-2 0 8/30/02 4.0 4-1 to 4-2 0 8/30/02 5.0 5-1 0 8/30/02 6.0 6-1 to 6-7 0' 8/30/02 7.0 7-1 to 7-31 0 8/30/02 8.0 8-1 to 8-3 0 8/30/02 9.0 9-1 to 9-2 0' 8/30/02 10.0 10-1 to 10-3 0 8/30/02 Note 1: Sections 6.0 and 9.0 of this Fourth Interval ISI Program Plan incorporate the requirements of Revision 3 of the CISI Plan. No technical changes were made during this incorporation. These sections are being statused as Revision 0 simply to address the minor formatting changes required to generate the Fourth Interval ISI Program Plan.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE OF CONTENTS Section Description Page

1.0 INTRODUCTION AND BACKGROUND

1-1 to 1-6 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2-1 to 2-13 3.0 COMPONENT ISI PLAN 3-1 to 3-2 4.0 SUPPORT ISI PLAN 4-1 to 4-2 5.0 SYSTEM PRESSURE TESTING ISI PLAN 5-1 6.0 CONTAINMENT ISI PLAN 6-1 to 6-7 7.0 INSERVICE INSPECTION

SUMMARY

TABLES 7-1 to 7-31 8.0 RELIEF REQUESTS FROM ASME SECTION XI 8-1 to 8-3 9.0 CONTAINMENT INSPECTION RELIEF REQUESTS FROM 9-1 to 9-2 ASME SECTION XI

10.0 REFERENCES

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IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval

1.0 INTRODUCTION AND BACKGROUND

1.1 Introduction This Inservice Inspection (ISI) Program Plan details the requirements for the examination and testing of ISI Class 1, 2, and 3 pressure retaining components and supports at Dresden Nuclear Power Station (DNPS),

Units 2, 3, and 2/3. Common (Unit 2/3) components are included in the Unit 3 sections, reports, and tables. This ISI Program Plan also includes Risk-Informed Inservice Inspections (RISI), augmented inservice inspections, and pressure testing requirements imposed on or committed to by DNPS. Procedure ER-AA-330, "Conduct of Inservice Inspection Activities," implements the ASME Section XI ISI Program.

The Fourth Inservice Inspection Interval is effective from January 20, 2003, through January 19, 2013, for DNPS Unit 2 and January 20, 2003, through January 19, 2013, for DNPS Unit 3. These effective interval dates are based on the assumption that DNPS will be approved to extend plant operation under the license renewal application. Paragraph IWA-2430(d)(1) of ASME Section XI allows an inspection interval to be extended or decreased by as much as one year, and Paragraph IWA-2430(e) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The extension may be taken for a period of time not to exceed the duration of the outage. See Table 1.1-1 at the end of this section for extensions that apply to DNPS's Fourth Interval.

The Fourth Inservice Inspection Interval is divided into three successive inspection periods as determined by calendar years of plant service within the inspection interval. Table 1.1-1 identifies the period dates for the Fourth Inservice Inspection Interval as defined by Inspection Program B.

In accordance with IWA-2430(d)(3), the inspection periods specified in Table 1.1-1 may be decreased or extended by as much as 1 year to enable inspections to coincide with DNPS's refueling outages.

1.2 Background The Commonwealth Edison Company, now known commercially as Exelon Generation Company (Exelon), obtained construction permits to build DNPS on January 10, 1966, for Unit 2, CPPR-18, and on October 14, 1966, for Unit 3, CPPR-22. The docket numbers assigned to DNPS are 50-237 for Unit 2 and 50-249 for Unit 3. After satisfactory plant construction and preoperational testing was completed, Exelon was granted a full power operating license for Unit 2, DPR-19, and subsequently commenced commercial operation on June 9, 1972; the full 1-1 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval power operating license for Unit 3, DPR-25, was granted and commercial operation commenced on November 16, 1971.

DNPS's piping systems and associated components were designed and fabricated before the examination requirements of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI were formalized and published. Since this plant was not specifically designed to meet the requirements of ASME Section XI, literal compliance is not feasible or practical within the limits of the current plant design. Limitations are likely to occur due to conditions such as accessibility, geometric configuration, and/or metallurgical characteristics.

For some inspection categories, an alternate component may be selected for examination and the code statistical and distribution requirements can still be maintained. If Code required examination selection criteria cannot be met, a relief request will be submitted in accordance with 10 CFR 50.55a.

1.3 Third Interval ISI Program Pursuant to the Code Of Federal Regulations, Title 10, Part 50, Section 55a, Codes andstandards,(10 CFR50.55a), Paragraph (g), Inservice inspection requirements,licensees were required to update their ISI programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI program was required to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii). Accordingly, the Inservice Inspection requirements applicable to the Third Inservice Inspection Program should have been based on the rules set forth in the 1986 Edition of ASME Section XI.

However, ComEd by letter dated April 26, 1991, requested NRC approval to meet the requirements set forth in the 1989 Edition, 1989 Addenda of ASME Section XI prior to its incorporation by reference into 10 CFR 50.55a(b)(2). NRC approval was received to utilize the 1989 Edition, No Addenda under the letter from R. J. Barrett to T. J. Kovach dated January 28, 1992, "Inservice Inspection Program Update - Dresden, Units 2 and 3." Therefore, the DNPS Third Interval ISI Program Plan was developed in accordance with the requirements of the 1989 Edition, No Addenda of ASME Section XI. The ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI, and utilized Inspection Program B.

Subsection IWE was added to the ISI Program midway through the Third Interval to address Containment Inservice Inspections (CISI). These 1-2 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval requirements were mandated by the Federal Register in 1996 and marked the beginning of the First Interval for CISI inspections. Implementation of the CISI Program is discussed in Section 6.0.

DNPS adopted the EPRI Topical Report TR-112657, Rev. B-A methodology, which was supplemented by Code Case N-578-1, for implementing risk-informed inservice inspections. The RISI program was in effect from the middle of the Third Period through the end of the Third Interval. This approach replaced the categorization, selection, and examination volume requirements of ASME Section XI Categories B-F, B J, C-F-I, and C-F-2 applicable to DNPS with Category R-A as defined in Code Case N-578-1.

1.4 Fourth Interval ISI Program Per 10 CFR 50.55a(g), licensees are required to update their ISI programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI program is required to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii).

The DNPS Fourth Interval ISI Program Plan was developed in accordance with the requirements of 10 CFR 50.55a, and the 1995 Edition with the 1996 Addenda of ASME Section XI. This ISI Program Plan addresses Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI, and utilizes Inspection Program B as defined therein. Implementation of Subsection IWE is discussed in Section 6.0.

DNPS has adopted the EPRI Topical Report TR-1 12657, Rev. B-A methodology, which was supplementedby Code Case N-578-1, for implementing risk-informed inservice inspections. The RISI program will be in effect for the entire Fourth Inspection Interval. This approach replaces the categorization, selection, and examination volume requirements of ASME Section XI Categories B-F, B-J, C-F-I, and C-F-2 applicable to DNPS with Category R-A as defined in Code Case N-578-1.

Implementation of the RISI program will be in accordance with relief request I4R-02.

1.5 Code Cases Per Footnote 6 of 10 CFR 50.55a, ASME Code Cases that have been determined to be suitable for use in ISI Program Plans by the NRC are listed in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1." The approved Code Cases 1-3 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval in Regulatory Guide 1.147, being utilized by DNPS are included in Section 2.1.1 of this document. The latest version of Regulatory Guide 1.147 incorporated into this document is Revision 13. As this guide is revised, newly approved Code Cases will be assessed for plan implementation at DNPS.

Footnote 6 also states that the use of other Code Cases (than those listed in Regulatory Guide 1.147) may be authorized by the Director of the office of Nuclear Reactor Regulation upon request pursuant to 10 CFR 50.55a(a)(3). Code Cases not approved for use in Regulatory Guide 1.147, which are being utilized by DNPS through associated relief requests that are included in Sections 8.0 and 9.0.

This ISI Program Plan will utilize the Draft Regulatory Guide DG-1091 (Proposed Revision 13 of Regulatory Guide 1.147) with the anticipation that the Final Revision 13 of Regulatory Guide 1.147 will be approved prior to the start of the Fourth Inspection Interval. DNPS will review the Final Revision 13 of Regulatory Guide 1.147 for ISI program impact at which time it is published.

1.6 Relief Requests In accordance with 10 CFR 50.55a, when a licensee either proposes alternatives to ASME Section XI requirements which provide an acceptable level of quality and safety, determines compliance with ASME Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, or determines that specific ASME Section XI requirements for inservice inspection are impractical, the licensee shall notify the NRC and submit information to support the determination.

The submittal of this information will be referred to in this document as a "relief request." Relief requests for the Fourth Interval are included in Sections 8.0 and 9.0 of this document. The text of the relief requests contained in Sections 8.0 and 9.0 will demonstrate one of the following:

the proposed alternatives provide an acceptable level of quality and safety per 10 CFR 50.55a(a)(3)(i), compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per 10 CFR 50.55a(a)(3)(ii), or the code requirements are considered impractical per 10 CFR 50.55a(g)(5)(iii).

Per 10 CFR 50.55a Paragraphs (a)(3) and (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate relief requests and "may grant such relief and may impose such alternative requirements as it 1-4 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval determines is authorized by law and will n6t endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility."

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 1.1-1 DNPS Unit 2 and Unit 3 ISI Interval/Period/Outage Matrix Unit 2 Period Interval Period Unit 3 Outage Outage Duration Start Date to End Date Start Date to Start Date to End Date Outage Duration Outage Number End Date Number D2R18 Scheduled llt l't Scheduled D3R118 11/03 1/20/03 to 1/19/06 1/20/03 to 1/19/06 11/04 D2R19 Scheduled 4 th (Unit 2) 2nd Scheduled D3R19 11/05 1/20/03 to 1/20/06 to 1/19/10 11/06 D2R20 Scheduled 2nd 1/19/13 Scheduled D3R20 11/07 DR1Shdld4* 1/20/06 to 1/19/10 (Unitto3Sceud

3) 3rd 11/08 Scheduled D3R21 D2R21 Scheduled 1/20/03 R2 11/09 1/19/13t 1/20/10 to 1/19/13 11/10 D2R22 Scheduled 3'd Scheduled D3R22 11/11 1/20/10 to 1/19/13 1 1 11/12 Note I: The Unit 3 Third Inspection Interval was extended by 80 days as permitted by IWA-2430(d). This extension is being carried forward to the Fourth Interval to accommodate both Units 2 and 3 having the same interval start date. As required by IWA-2430(d)(1), successive intervals shall not be altered by more than one year from the original pattern. This means that for the remainder of the Fourth Interval, only 285 days are available to use under the IWA-2430(d) extension.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2.1 ASME Section XI Examination Requirements As required by 10 CFR 50.55a, this program Was developed in accordance with the requirements detailed in the 1995 Edition, 1996 Addenda, of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices, Inspection Program B of IWA-2432, and approved alternatives through relief requests and safety evaluation reports (SERs).

The ISI program implements Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," ASME Section XI 1995 Edition with the 1996 Addenda as required by 10 CFR 50.55a(g)(6)(ii)(C). Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in piping, bolting, and the reactor pressure vessel. Each organization (e.g.

owner or vendor) will be required to have a written program to insure compliance with the requirements. These requirements are implemented through the Performance Demonstration Initiative (PDI) Program according to the schedule defined in 10 CFR 50.55a(g)(6)(ii)(C).

For the Fourth Inspection Interval, DNPS's inspection program for ASME Section XI Categories B-F, B-J, C-F-l, and C-F-2 will be govemedby risk-informed requirements. The RISI program methodologyis described in the EPRI Topical Report TR-1 12657, Rev. B-A. To supplement the EPRI Topical Report, Code Case N-578-1 (as applicable per Relief Request 14R-02) is also being used for the classification of piping structural elements under the RISI program. The RISI program scope will be implemented as an altemativeto the 1995 Edition with the 1996 Addenda, ASME Section XI examination program for Class I B-F and B-J welds and Class 2 C-F-1 and C-F-2 welds in accordance with 10 CFR 50.55a(a)(3)(i). The basis for the resulting risk categorizationsof the non exempt Class 1 and 2 piping systems at DNPS is defined and maintained in the Final Report, "Risk Informed Inservice Inspection Evaluation," as referenced in Section 10.0 of this document.

The CISI Program Plan per Subsection IWE has been incorporated into Section 6.0 of this ISI Program Plan and is not discussed in this section.

2.1.1 ASME Section XI Code Cases As referenced by 10 CFR 50.55a Footnote 6 and allowed by NRC Regulatory Guide 1.147, Revision 13, the following Code Cases are being incorporated into the DNPS ISI Program.

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IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval N-307-2 Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, When the Examinations Are Conducted From the End of the Bolt or Stud or From the Center-Drilled Hole N-416-2 Alternative Pressure Test Requirements for Welded Repairs, Fabrication Welds for Replacement Parts and Piping Subassemblies, or Installation of Replacement Items by Welding, Class 1, 2, and 3 Code Case N-416-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

(1) Additional surface examinations should be performed on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section III.

(2) A 4-hour hold time must be maintained prior to the VT-2 visual examination.

(See Technical Approach and Position number I4T-03).

N-460 Alternative Examination Coverage for Class I and Class 2 Welds N-498-1 Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1, 2, and 3 Systems Code Case N-498-1 is only being implemented as it pertains to Class 3 systems. (The portions of the Case that address Class 1 and 2 systems have been incorporated into the ASME Section XI code of record, 1995 Edition with 1996 Addenda, applicable to the DNPS Fourth Interval.)

N-504-2 Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping N-516-2 Underwater Welding Code Case N-516-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

When welding is to be performed on high neutron fluence Class I material, then a mock-up, using material with similar fluence levels, should be welded 2-2 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval to verify that adequate ciack prevention measures were used.

N-523-2 Mechanical Clamping Devices for Class 2 and 3 Piping N-526 Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels N-532 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000 Code Case N-532 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

An Owner's Activity Report Form OAR-1 is required to be prepared and certified upon completion of each refueling outage. The Code Case does not designate a time frame for submission to the regulatory authority.

Thus, the OAR-1 must be submitted within 90 days.

Applicable IWA-4000 and IWA-6000 references from the 1995 Edition, with the 1996 Addenda of ASME Section XI, will be utilized in place of the code references specified in Code Case N-532. A matrix of those reference paragraphs was added in Code Case N-532-1 for various Code years.

N-546 Alternative Requirements for Qualification of VT-2 Examination Personnel Code Case N-546 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

(1) Qualify examination personnel by test to demonstrate knowledge of Section XI and plant specific procedures for VT-2 visual examination.

(2) Requalify examination personnel by examination every three years.

(3) This Code Case is applicable only to the performance of VT-2 examinations.

N-588 Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels 2-3 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval N-598 Alternative Requirements to Required Percentages of Examinations N-623 Deferral of Inspections of Shell-to-Flange and Head-to Flange Welds of a Reactor Vessel N-624 Successive Inspections N-640 Alternative Reference Fracture Toughness for Development of P-T Limit Curves Additional Code Cases may be invoked in the future based on new Plan requirements or revisions to Regulatory Guide 1.147. Any Code Cases invoked in the future shall be in accordance with those approved for use in the latest published revision of Regulatory Guide 1.147 at that time.

2.2 Augmented Examination Requirements Augmented examination requirements are those examinations that are performed above and beyond the requirements of ASME Section XI. Below is a summary of those examinations performed by DNPS that are not specifically addressed by ASME Section XI, or the examinations that will be performed in addition to the requirements of the Code on a routine basis during the Fourth Inspection Interval.

2.2.1 Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping," Revision 2 / Supplement 1 to Generic Letter 88 01, and NUREG 0313, "Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping,"

Revision 2, and BWRVIP-75 "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules," as conditionally approved by NRC SER dated May 14, 2002.

These documents discuss the examination requirements for Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping. References to GL 88-01 within the ISI programh refer to the comprehensive commitments to all of these documents.

DNPS has committed to the requirements of these documents as discussed in Updated Final Safety Analysis Report (UFSAR) Section 5.2.3.5. The original DNPS commitment concerning Generic Letter 88-01 was sent to the NRC in a letter from W. E. Morgan (CECo) to the NRC dated July 29, 1988. The NRC reviewed this commitment in letters from T. M. Ross (NRC) to T. J. Kovach (CECo) dated May 22, 1989 and from B. L. Siegel (NRC) to T. J. Kovach (CECo) dated August 23, 1990.

The outboard RWCU piping has been excluded from Generic Letter 88-01. The basis for this exclusion is documented in a letter from P. L.

Piet (CECo) to the T. E. Murley (NRC) dated August 20, 1993 and in a 2-4 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval letter from R. M. Pulsifer (NRC) to D. L. Iarrar (ComEd) dated September 22, 1994.

RISI regulations are being invoked for DNPS in this ISI Program Plan.

Under these new guidelines, Class 1 and 2 piping structural elements are inspected in accordance with EPRI Topical Report TR- 12657, Rev. B-A and Code Case N-578-1. Per this Topical Report and Code Case, welds within the plant that are assigned to IGSCC Categories B through G will continue to meet existing IGSCC schedules, while IGSCC Category A welds will be subsumed into the RISI program.

2.2.2 Alternate BWR Feedwater Nozzle Inspection Requirements, dated October 1995 This document discusses BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking. The alternate approach was developed and submitted to the NRC by the Boiling Water Reactor Owners' Group (BWROG). The NRC conditionally accepted these alternate requirements in an SER dated June 5, 1998.

DNPS initially committed to the requirements of NUREG 0619 as stated in the Third Interval ISI Program Plan. DNPS revised this commitment to utilize the BWROG alternate inspections in a letter from J. M. Heffley (ComEd) to the NRC dated May 21, 1999.

2.3 System Classifications and P&ID Boundary Drawings The ISI Classification Basis Document details those systems that are ISI Class 1, Class 2, or Class 3 that fall within the inservice inspection scope of examinations.

Below is a summary of the classification criteria used within the Basis Document.

Each safety related fluid system containing water, steam, air, oil, etc. included in the DNPS UFSAR was reviewed to determine which safety functions they perform during all modes of system and plant operation. Based on these safety functions, the systems and components were evaluated per classification documents. The systems were then designated as ISI Class 1, Class 2, Class 3, or non-classed accordingly. This evaluation followed the guidelines of UFSAR Section 5.2.4 for ISI Class 1 and UFSAR Section 6.6 for ISI Classes 2 and 3.

Safety related portions of systems are defined by the Piping and Instrumentation Diagrams (P&IDs) with an "S" flag.

When a particular group of components is identified as performing an ISI Class 1, Class 2, or Class 3 safety function, the components are further reviewed to assure the interfaces (boundary valves and boundary barriers) meet the criteria set by 10 CFR 50.2, 10 CFR 50.55a(c)(1), 10 CFR 50.55a(c)(2), and Regulatory Guide 1.26. Although DNPS is not committed to or licensed in accordance with these documents, Standard Review Plan (SRP) 3.2.2, "System Quality Group Classification", and American National Standards Institute/American Nuclear 2-5 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Society (ANSI/ANS) 52.1-1983, "Nuclear Safety Criteria for the Design of Stationary Boiling Water Reactor Plants", were also used for guidance in determining the classification boundaries when 10 CFR and Regulatory Guide 1.26 did not address a given situation. The valve positions shown on the system flow diagrams are assumed to be the normal positions during system operation unless otherwise noted.

At the time the construction permits for DNPS Units 2 and 3 were issued, ASME Section III covered only pressure vessels, primarily nuclear reactor vessels. The majority of piping, pumps, and valves were designed and installed according to the rules of USAS B31.1.0-1967 Edition, "Power Piping." Consequently, the DNPS ISI Program has essentially no ASME Section III Class 1, 2, or 3 piping systems.

ISI classification boundaries are defined by the P&IDs with a classification flag.

The designators 1, 2, and 3, respectively, are used for classifying nonexempt ASME Section XI components.

The systems and components subject to the examinations of Articles IWB-2000, IWC-2000, IWD-2000 and IWF-2000, and pressure tests of Articles IWB-5000, IWC-5000 and IWD-5000 are identified on the P&IDs as detailed in Table 2.3-1.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 2.3-1 DNPS P&ID BOUNDARY DRAWINGS UNIT 2 UNIT 3 & 2/3 TITLE M-12, SH. 1 & 2 M-345, SH. 1 & 2 Diagram of Main Steam Piping (MS)

M-14 M-347 Diagram of Reactor Feedwater Piping (FW)

M-22 M-355 Diagram of Service Water Piping (CCSW) &

(DGSW)

M-25 M-356 Diagram of Pressure SuppressionPiping (PS)

M-26, SH. 1, 2 & 3 M-357, SH. 1, 2 & 3 Diagram of Nuclear Boiler& Reactor RecirculationPiping (RR)

M-27 M-358 Diagram of Core Spray Piping (CS)

M-28 M-359 Diagram of Isolation Condenser Piping (ISCO)

M-29, SH. 1 & 2 M-360, SH. 1 & 2 Diagram of L.P. Coolant InjectionPiping (LPCI)

M-30 M-361 Diagram of Reactor Water Cleanup Piping (RWCU)

M-32 M-363 Diagram of Shutdown Reactor Cooling Piping (SDC)

M-33 M-364 Diagram of Standby Liquid Control Piping (SBLC)

M-34, SH. 2 M-3 65, SH. 2 Diagram of Control Rod Drive HydraulicPiping (CRD)

M-35, SH. I M-366 Diagram of Demineralized Water System Piping (DW)

M-39 M-369 Diagram of Reactor Building Equipment Drains M-51 M-374 Diagram of H.P. Coolant Injection Piping (HPCI)

M-517, SH. I M-517, SH. 2 & 3 Diesel Generator Engine Cooling Water System (DGCW)

M-1234, SH. I M-1239, SH. 1 P&ID Liquid Sampling M-3121 M-3121 Diagram of Control Room HVAC 2-7 Revision 0

151 Program Plan ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 2.4 ISI Isometric Drawings for Nonexempt ISI Class Components and Supports ISI isometric drawings were developed to detail the ISI Code Class 1, 2, and 3 components (welds, bolting, etc.) and support locations at DNPS. SPT isometric drawings were also developed to show those components subject to pressure testing. The ISI isometric and SPT isometric drawings are listed in Table 2.4-1.

DNPS's ISI program, including the database, basis document, and schedule, addresses the non-exempt components which require examination and testing.

A summary of DNPS Units 2 and 3 ASME Section XI nonexempt components and supports is included in Section 7.0.

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1SI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 2.4-1 DNPS ISI ISOMETRIC AND PRESSURE TESTING DRAWINGS Sheet I of 3 UNIT 2 UNIT 3 & 2/3 TITLE ISI-101, SH. 1 & 2 ISI-112, SH. 1 & 2 ISI Class 1 Main Steam Piping ISI-101, SH. 3 ISI- 112, SH. 3 ISI Class 1 Main Steam Drain Piping ISI-102, SH. 1 & 2 ISI-113, SH. I & 2 ISI Class 1 Reactor FeedwaterPiping ISI-103, SH. I to 4 ISI-I 14, SH. 1 to 4 ISI Class 1 Nuclear Boiler and Reactor RecirculationPiping ISI-104 ISI-115 ISI Class 1 Core Spray Piping ISI-105 ISI-116 ISI Class I Isolation CondenserPiping ISI-106, SH. 1 & 2 ISI- 117, SH. 1 & 2 ISI Class 1 Low Pressure Coolant InjectionPiping ISI-107,SH. 1 & 2 ISI- 118,SH. 1 & 2 ISI Class 1 Reactor Water Cleanup Piping ISI-108 ISI-119 ISI Class 1 Reactor Shutdown CoolingPiping ISI-109 ISI-120, SH. 1 & 2 ISI Class 1 Standby Liquid Control ISI-110 ISI-121 ISI Class 1 Control Rod Drive Hydraulic Piping ISI-111 ISI-122 ISI Class 1 High Pressure Coolant Injection Piping ISI-126 ISI-123 ISI Class 1 Reactor Head Vent Piping ISI-127 ISI-124 ISI Class 1 ReactorHead Spray Piping ISI-128, SH. 1 to 3 ISI-125,SH. I to 3 ISI Class 1 ReactorPressure Vessel ISI-200, SH. I & 2 ISI-205, SH. I & 2 ISI Class 2 Core Spray Piping ISI-200, SH. 3 ISI-205, SH. 3 & 4 ISI Class 2 Core Spray Piping Obstructions ISI-201 ISI-206 ISI Class 2 Isolation CondenserPiping ISI-202, SH. I to 4 ISI-207, SH. I to 4 ISI Class 2 Low Pressure Coolant Injection Piping ISI-202, SH. 5 & 6 ISI-207, SH. 5 & 6 ISI Class 2 Low Pressure Coolant Injection Piping Obstructions ISI-202, SH. 7 ISI-207, SH. 7 ISI Class 2 Low Pressure Coolant Injection Piping ISI-202, SH. 8 ISI-207, SH. 8 ISI Class 2 LPC1ICCSW Heat Exchangers ISI-203, SH. 1 & 2 ISI-208, SH. I & 2 ISI Class 2 High Pressure Coolant Injection Piping ISI-203, SH. 3 & 4 ISI-208, SH. 3 & 4 ISI Class 2 High Pressure Coolant Injection Piping Obstructions ISI-204 ISI-209 ISI Class 2 Reactor Water Cleanup and Reactor FeedwaterPiping 2-9 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 2.4-1 DNPS ISI ISOMETRIC AND PRESSURE TESTING DRAWINGS Sheet 2 of 3 UNIT 2 UNIT 3 & 2/3 TITLE ISI-213, SH. 1 ISI-212, SH. 1 ISI Class 2 CRD Scram Discharge Volume, East Bank ISI-213, SH. 2 ISI-212, SH. 2 ISI Class 2 CRD Scram Discharge Volume, West Bank ISI-214 ISI-215 ISI Class 2 ECCS Ring Header ISI-300, SH. I to 6 ISI-302, SH. I to 4 ISI Class 3 ContainmentCooling Service Water Piping ISI-301, SH. I to 7 ISI-303 ISI Class 3 Diesel Generator Service Water Piping ISI-304, SH. I to 9 ISI-307 ISI-305 ISI Class 3 Isolation Condenser and Vent Piping ISI-308, SH. 1 & 2 ISI-306, SH. 1 & 2 ISI Class 3 Safety ReliefValve DischargePiping ISI-501, SH. 1 ISI-551, SH. I System Pressure Test Walkdown Isometric Reactor Head Cavity (589' Elevation)

ISI-501, SH. 2 ISI-551, SH. 2 System Pressure Test Walkdown Isometric Drywell Fourth Floor (576'-7 1/8" Elevation)

ISI-501, SH. 3 ISI-55 1, SH. 3 System Pressure Test Walkdown Isometric Drywell Third Floor (562'-0" Elevation)

ISI-501, SH. 4 ISI-551, SH. 4 System Pressure Test Walkdown Isometric Drywell Second Floor (537'-1 1/4" Elevation)

ISI-501, SH. 5 ISI-551, SH. 5 System Pressure Test Walkdown Isometric Drywell First Floor (515'-5 / 4" Elevation)

ISI-501, SH. 6 ISI-55 1, SH. 6 System Pressure Test Walkdown Isometric Drywell Basement (502'-4" Elevation)

ISI-501, SH. 7 ISI-55 1, SH. 7 System Pressure Test Walkdown Isometric Lower Head CRD Area (502'-4" Elevation)

ISI-501, SH. 8 ISI-551, SH. 8 System Pressure Test Walkdown Isometric Instrumentation ISI-502, SH. I ISI-552, SH. 1 System Pressure Test Walkdown Isometric Isolation Condenser Piping ISI-502, SH. 2 & 3 ISI-552, SH. 2 & 3 System Pressure Test Walkdown Isometric Isolation Condenser, Instrumentation, and Vent Piping ISI-503 ISI-553 System Pressure Test Walkdown Isometric Shutdown Cooling Piping 2-10 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 2.4-1 DNPS ISI ISOMETRIC AND PRESSURE TESTING DRAWINGS Sheet 3 of 3 UNIT 2 UNIT 3 & 2/3 j TITLE ISI-504 ISI-554 System Pressure Test Walkdown IsometricMSIV Room-X Area ISI-505, SH. 1 ISI-555, SH. I System Pressure Test Walkdown Isometric Control Rod Drive Hydraulic Piping, East Bank ISI-505, SH. 2 ISI-555, SH. 2 System Pressure Test Walkdown Isometric Control Rod Drive Hydraulic Piping, West Bank ISI-506, SH. 1 to 3 ISI-556, SH. I to 3 System Pressure Test Walkdown Isometric Standby Liquid Control Piping ISI-507, SH. I & 2 ISI-557, SH. I & 2 System Pressure Test Walkdown Isometric Core Spray Piping ISI-508 ISI-558 System Pressure Test Walkdown Isometric Head Flange Seal Leak Detection ISI-509, SH. 1 & 2 ISI-559, SH. 1 & 2 System PressureTest Walkdown IsometricL.P.

Coolant Injection Piping ISI-509, SH. 3 & 4 ISI-559, SH. 3 & 4 System Pressure Test Walkdown Isometric Containment Cooling Service Water Piping ISI-509, SH. 5 Not Applicable System Pressure Test Walkdown Isometric Control Room HVAC ISI-509, SH. 6 ISI-559, SH. 5 System Pressure Test Walkdown Isometric CCSW (In the course of preparation) Supply to HPCI/LPCI Room Coolers ISI-510, SH. 1 & 2 ISI-560, SH. 1 & 2 System Pressure Test Walkdown Isometric H. P.

Coolant Injection Piping ISI-510, SH. 3 ISI-560, SH. 3 System Pressure Test Walkdown Isometric Torus Level Instrumentation ISI-51I ISI-561 System Pressure Test Walkdown Isometric Safety Relief Valve Discharge Piping ISI-512 ISI-562 SH. 1 (Unit 3) & 2 System Pressure Test Walkdown Isometric Diesel (Unit 2/3) Generator Service Water ISI-513, SH. 1 ISI-563, SH. 1 System Pressure Test Walkdown Isometric Emergency Core Cooling, Torus Ring Header I1SI-513, SH. 2 ISI-563, SH. 2 System Pressure Test Walkdown Isometric ECCS Keep Fill Pump and Piping 2-11 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 2.5 Technical Al'prohch and Positions When the requirements of ASME Section XI are not easily interpreted, DNPS has reviewed general licensing/regulatoryrequirements and industry practice to determine a practical method of implementing the Code requirements. The technical approach and position documents contained in this section have been provided to clarify DNPS's implementationofASME Section XI requirements. An index which summarizes each technical approach/positionis included in Table 2.5-1.

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IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX/SUMMARIES Position Revision Status' (Program) Description Number Date 14T-01 0 Active (SPT) System Leakage Testing of 8/30/02 Non-Isolable Buried Components W4T-02 0 Active (SPT) Valve Seats as Pressurization 8/30/02 Boundaries W4T-03 0 Active (SPT) Alternative Pressure Test 8/30/02 Requirements following Repair and Replacement Activities 14T-04 0 Active (ISI) Reactor Coolant Makeup Systems 8/30/02 Note 1: Technical Approach and Position Status Options: Active - Current ISI Program Technical Approach is being utilized at DNPS; Deleted - Technical Approach is no longer being utilized at DNPS.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-01 (Page 1 of 1)

COMPONENT IDENTIFICATION Code Class: 3

Reference:

IWA-5244(b)(2)

Examination Category: N/A Item Number: N/A

==

Description:==

System Leakage Testing of Non-Isolable Buried Components Component Number: Non-Isolable Pressure Retaining Buried Components CODE REOUIREMENT IWA-5244 requires non-isolable buried components be tested to confirm that flow during operation is not impaired.

POSITION Article IWA-5000 provides no guidance in setting acceptance criteria for what can be considered "adequate flow". In lieu of any formal guidance provided by the Code, DNPS has established the following acceptance criteria:

For opened ended lines on systems that require Inservice Testing (IST) of pumps, adherence to IST acceptance criteria is considered as reasonable proof of adequate flow through the lines.

This acceptance criteria will be utilized as proof of adequate flow in order to meet the requirements of IWA-5244(b)(2).

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-02 (Page 1 of 1)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3

Reference:

IWA-5221 IWA-5222 Examination Category: B-P, C-H, D-B Item Number: B15.10, B15.30, B15.50, B15.70, C7.10, C7.30, C7.50, C7.70, D2.10, D2.30, D2.50, D2.70

==

Description:==

Valve Seats as Pressurization Boundaries Component Number: All Pressure Testing Boundary Valves CODE REQUIREMENT IWA-5221 requires the pressurization boundary for system leakage testing extend to those pressure retaining components under operating pressures during normal system service.

System leakage testing is performed in lieu of hydrostatic pressure testing (Paragraph IWA-5222) at or near the end of each inspection interval in accordance with Code Case N-498-1 for Class 1, 2, and 3 systems. Code Case N-498-1 requires the pressurization boundary extend to all Class 1 components during the system leakage test, and extend to all Class 2 and 3 components included in those portions of systems required to operate or support the system safety function up to and including the first normally closed valve.

POSITION DNPS's position is that the test pressurization boundary extends up to the valve seat of the valve utilized for isolation. For example, in order to pressure test the Class 1 components, the valve that provides the Class break would be utilized as the isolation point. In this case the true pressurization boundary, and Class break, is actually at the valve seat.

Any requirement to test beyond the valve seat is dependent only on whether or not the piping on the other side of the valve seat is ISI Class 1, 2, or 3.

The extension of the pressurization boundary during a pressure test would require an abnormal valve line-up. Extending the boundary would require the over pressurization of low pressure piping at systems that have a high/low pressure interface (such as LPCI and Core Spray).

In order to simplify examination of classed components, DNPS will perform a VT-2 visual examination of the entire boundary valve body and bonnet (during pressurization up to the valve seat).

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-03 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 1,2, and 3

Reference:

IWA-4540 Examination Category: N/A Item Number: N/A

==

Description:==

Alternative Pressure Test Requirements following Repair and Replacement Activities Components Number: All Class 1, 2, and 3 Pressure Retaining Components CODE REQUIREMENT IWA-4540(a) requires a system hydrostatic test be performed after welding on the pressure retaining boundary, or installation of an item by welding or brazing, except as exempted by IWA 4540(b).

POSITION Hydrostatic tests conducted at elevated pressures are difficult to perform and often represent a true hardship without any compensating increase in plant safety. Some of the difficulties associated with a hydrostatic test include complicated or abnormal valve line-ups, gagging or removing relief valves, additional maintenance on valve internals not normally used for isolation, and substantially increased radiation exposure.

For this reason, DNPS will utilize ASME Code Case N-416-2 as an alternative to the ASME Section XI repair/replacementhydrostatic pressure testing requirement. Code Case N-416-2 is conditionally approved for use by the NRC in Revision 13 of Regulatory Guide 1.147. DNPS will implement this Case in accordance with the Regulatory Guide imposed conditions as follows:

(1) Additional surface exams will be performed on the root (pass) layer of Class 3 butt and socket welds when the surface exam method is used in accordance with Section III.

(2) A 4-hour hold time must be maintained prior to the VT-2 visual exam. [Note: This condition is consistent with established regulatory position.]

The Technical Approach established here is to clarify the hold time condition stated in (2) above.

The regulatory position referenced affects several pressure testing activities and requirements included in future Code editions and other ASME Code Cases. The position is stated in both related Safety Evaluation Reports and in Federal Rulemaking. For the purpose of establishing this position, DNPS will take guidance from the latest NRC Proposed Rulemaking affecting 10 CFR 50 as published in the Federal Register, Volume 66, Number 150, dated August 3, 2001.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-03 (Page 2 of 2)

POSITION (Continued)

Section 2.2.7, System Leakage Tests, and Section 3, Paragraph (b)(2)(xx) both state the NRC position regarding hold times. The position as stated in Section 3 reads "a 10-minute hold time for non-insulated systems and components or a 4-hour hold time for insulated systems and components will be required after attaining system operating pressure".

DNPS will utilize this regulatory position for the purpose of clarifying RG 1.147, Code Case N-416-2, Condition (2) as stated above. As such, the 4-hour hold time referenced in Condition (2) will only be implemented for those pressure tests conducted after repair/replacementactivities on insulated components. If the system or component is not normally insulated, or if the insulation is removed for the purpose of conducting the system leakage test, a 10-minute hold time will be used after attaining test pressure.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-04 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3

Reference:

IWA-4131.1 (a)(2),IWB-1220(a), 10 CFR 50.55a(c)(2)(i)

Examination Category: N/A Item Number: N/A

==

Description:==

Reactor Coolant Makeup Systems Components Number: Class 1 Pressure Retaining Components CODE REQUIREMENT IWA-413 1.1 (a)(2) provides size criteria for application of alternative repair replacement requirements for small items no larger than NPS 1. The alternative requirements are applied to components of a size and design such that, in the event of postulated failure during normal plant operating conditions, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by normal reactor coolant makeup systems operable from on-site emergency power.

IWB-1220(a) provides an exemption from volumetric and surface examination for components that are connected to the reactor coolant system and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the reactor coolant system under normal plant operating conditions is within the capacity of makeup systems that are operable from on-site emergency power. The emergency core cooling systems are excluded from the calculation of makeup capacity.

10 CFR 50.55a(c)(2)(i) allows components which are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in 10 CFR 50.2 need not meet the requirements for Class 1 components when, in the event of postulated failure of the component during normal reactor operation, the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system.

POSITION DNPS credits the control rod drive (CRD) system as the reactor coolant makeup system. Dresden UFSAR Section 3.1.2.4.4 states that during normal operation, water level in the reactor vessel is maintained by the feedwater system. For small breaks in the reactor coolant pressure boundary, the makeup capability is provided by the feedwater system and the control rod drive (CRD) system. The emergency core cooling system provides inventory makeup for breaks beyond the capability of the feedwater system.

While the CRD system is not normally fed from the emergency source, it is capable of being re aligned to emergency buses. Since the CRD system is operable from on-site emergency power, the capabilities of this system are utilized for determining size of components within the capability of the reactor coolant makeup system.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITIONNUMBER: 14T-04 (Page 2 of 2)

POSITION (Continued)

The CRD system is capable of providing makeup to the reactor coolant system utilizing CRD cooling water flow through each of the CRD units, seal purge flow through the reactor recirculation pump seals, and reactor vessel head cooling injected through the head spray nozzle in the reactor head. Since reactor vessel head cooling injected through the head spray nozzle does not supply makeup water to the reactor coolant system during normal reactor operation, this source of makeup is not credited. Therefore, the normal makeup to the reactor coolant system is as follows:

CRD Cooling Water Flow 58 gpm Recirculation Pump Seal Purge Flow 5 rpm 63 gpm In determining the size of the liquid and steam lines (DNPS Calculation XCE040.0201), with in the capability of the makeup system, liquid lines are defined as those which penetrate the reactor pressure vessel (RPV) below the normal water level and steam lines as those which penetrate the RPV above the normal water level. The line diameters based on the reactor coolant system makeup capability are as follows:

Liquid Line Break Size 0.533" I.D.

Steam Line Break Size 0.876" I.D.

These line sizes are utilized when applying the following code requirements.

IWA-413 1.1 (a)(2) limits the application of alternative requirements for small items to those line sizes bounded by the above calculation.

IWB-1220(a) is not applied since this exemption is bound by IWB-1220(b)(1) which exempts piping NPS 1 and smaller, and IWB-1220(b)(2) which exempts components and their connections in piping ofNPS 1 and smaller.

No reactor coolant pressure boundary components are excluded from Class I requirements in accordance with 10 CFR 50.55a(c)(2)(i) due to the small line sizes that can be excluded.

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151 Program Plan ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 3.0 COMPONENT ISI PLAN The DNPS Component ISI Plan includes ASME Section XI nonexempt pressure retaining welds, piping structural elements, pressure retaining bolting, attachment welds, pump casings, and valve bodies of 1SI Class 1, 2, and 3 components that meet the criteria of Subarticle IWA-1300. These components are identified on the P&IDs listed in Section 2.3, Table 2.3-1. Procedure ER-AA-330-002, "Inservice Inspection of Welds and Components," implements the ASME Section XI Component ISI Plan. This Component ISI Plan also includes component augmented inservice inspection examinations specified by documents other than ASME Section XI as referenced in Section 2.2 of this document.

3.1 DNPS Nonexempt ISI Class Components The DNPS ISI Class 1 components subject to examination are those which are not exempted under the criteria in Subarticle IWB-1220 of the 1989 Edition, No Addenda of ASME Section XI (see Section 3.1.2 below). The DNPS ISI Class 2 and 3 components subject to examination are those not exempted under the criteria of the Subarticles IWC-1220 and IWD-1220 of the 1995 Edition, 1996 Addenda of ASME Section XU. A summary of DNPS Units 2, 3, and 2/3 ASME Section XI nonexempt components is included in Section 7.0.

3.1.1 Identification of ISI Class 1, 2, and 3 Nonexempt Components ISI Class 1, 2, and 3 components are identified on the ISI isometric drawings listed in Section 2.4, Table 2.4-1. Welded attachments are also identified by controlled DNPS support drawings.

3.1.2 10 CFR 50.55a(b)(2)(xi) specifies that the 1989 Edition, No Addenda of ASME Section XI, Subarticle IWB-1220 shall be used in lieu of the 1995 Edition, 1996 Addenda of ASME Section XI, Subarticle IWB-1220.

IWB-1220, Components Exempt from Examination (1989 Edition, No Addenda) - The following components (or parts of components) are exempted from the volumetric and surface examination requirements of IWB-2500 per the Code paragraph referenced:

(a) Components that are connected to the Reactor Coolant System and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the Reactor Coolant System, under normal plant operating conditions, is within the capacity of makeup systems which are operable from on-site emergency power; 3-1 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Those systems which provide rehctor coolant makeup and application of the exemptions are documented in Technical Approach and Position 14T-04.

(b)(1) piping of NPS 1 and smaller; (b)(2) components and their connections in piping of NPS 1 and smaller; (c) reactor vessel head connections and associated piping, NPS 2 and smaller, made inaccessible by control rod drive penetrations.

3.2 Risk-Informed Examination Requirements Piping structural elements that fall under RISI Category R-A are risk ranked as High (1, 2, and 3), Medium (4 and 5), and Low (6 and 7). Per the EPRI Topical Report TR-l 12657, Rev. B-A and Code Case N-578-1, piping structural elements ranked as High or Medium Risk are subject to examination while piping structural elements ranked as Low Risk do not receive any examinations (except for pressure testing). Thin wall welds that were excluded from volumetric examination under ASME Section XI rules per Table IWC-2500-1 are included in the element scope that is potentially subject to RISI examination at DNPS. '

Piping structural elements may be excluded from examination (other than pressure testing) under the RISI Program if the only degradation mechanism present for a given location is inspected for under certain other DNPS programs such as the Flow Accelerated Corrosion (FAC) or Intergranular Stress Corrosion Cracking (IGSCC) Programs. These piping structural elements will remain part of the FAC or IGSCC programs which already perform "for cause" inspections to detect these degradation mechanisms. Piping structural elements susceptible to FAC or IGSCC along with another degradation mechanism (e.g., thermal fatigue) are retained as part of the RISI scope and are included in the element selection for the purpose of performing exams to detect the additional degradation mechanism.

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151 Program Plan ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 4.0 SUPPORT ISI PLAN The DNPS Support Program includes the supports of ASME Section XI nonexempt ISI Class 1, 2, and 3 components as described in Section 3.0. Procedure ER-AA-330-003, "Visual Examination of Section XI Component Supports," implements the ASME Section XI Support ISI Plan.

4.1 DNPS Nonexempt ISI Class Supports The DNPS ISI Class 1, 2, and 3 nonexempt supports are those which do not meet the criteria of Subarticle IWF-1230. A summary of DNPS Units 2, 3, and 2/3 ASME Section XI nonexempt supports is included in Section 7.0.

4.1.1 Identification of ISI Class 1, 2, and 3 Nonexempt Supports ISI Class 1, 2, and 3 supports are identified on the ISI isometric drawings listed in Section 2.4, Table 2.4-1. Supports are also identified by controlled DNPS support drawings.

4.2 Snubber Examination and Testing Requirements 4.2.1 ASME Section XI Paragraphs IWF-5200(a) & (b) and IWF-5300(a) & (b) require VT-3 Visual Examinations and Inservice Tests of snubbers to be performed in accordance with the Operation and Maintenance of Nuclear Power Plants (OM), Standard ASME/ANSI OM, Part 4. As allowed by 10 CFR 50.55a(b)(3)(v), DNPS will use Subsection ISTD, "Inservice Testing of Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants," ASME OM Code, 1995 Edition with the 1996 Addenda, in lieu of the requirements for snubbers in ASME Section XI, Paragraphs IWF-5200(a) & (b) and IWF-5300(a) & (b). Procedure ER-AA-330-004, "Visual Examination of Technical Specification Snubbers," implements visual examination of snubbers. Procedure ER-AA-330-010, "Snubber Functional Testing," implements functional testing of snubbers.

The ASME Section XI ISI Program uses Subsection IWF to define support inspection requirements. The ISI Program maintains the Code Class snubbers in the populations subject to inspection per Subsection IWF.

This is done to address the related requirements of Paragraphs IWF-5200(c) and IWF-5300(c). (See Section 4.2.2 below.)

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 4.2.2 ASME Section XI Paragraphs IWF-5200(c) and IWF-5300(c) require integral and nonintegral attachments for snubbers to be examined in accordance with Subsection IWF of the Code. This results in VT-3 visual examination of the snubber attachment'hardware including lugs, bolting, pins, and clamps.

The ASME Section XI ISI Program uses Subsection IWF to define the inspection requirements for all Class 1, Class 2, and Class 3 supports, regardless of type. The ISI Program maintains the Code Class snubbers in the support populations subject to inspection per Subsection IWF. This is done to facilitate scheduling, preparation including insulation removal, and inspection requirements of the snubber attachment hardware (e.g.,

lugs, bolting, pins, and clamps) per Paragraphs IWF-5200(c) and IWF-5300(c).

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1SI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 5.0 SYSTEM PRESSURE TESTING ISI PLAN The DNPS System Pressure Testing (SPT) Program includes all pressure retaining ASME Section XI, ISI Class 1, 2, and 3 components, with the exception of those specifically exempted by Paragraphs IWC-5222(b) and IWD-5240(b). All RISI piping structural elements, regardless of risk classification, remain subject to pressure testing as part of the current ASME Section XI program.

The SPT Program performs system pressure tests and visual inspections on the ISI Class 1, 2, and 3 pressure retaining components to verify system and component structural integrity. This program conducts both Periodic and Interval (10-year frequency) pressure tests as defined in ASME Section XI Inspection Program B. Procedure ER-AA-330-001, "Section XI Pressure Testing," implements the ASME Section XI System Pressure Testing ISI Plan.

All components subject to ASME Section XI System Pressure Testing are shown on the P&IDs listed in Section 2.3, Table 2.3-1, and the System Pressure Test Walkdown Isometric Drawings listed in Table 2.4-1.

5.1 Risk-Informed Examinations of Socket Welds Socket welds selected for examination under the RISI program are to be inspected with a VT-2 exam each refueling outage per ASME Code Case N-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socket welds selected for inspection under the RISI program shall be pressurized each refueling outage.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 7.0 INSERVICE INSPECk ION

SUMMARY

TABLES The following Tables 7.0-1 and 7.0-2 provide a summary of the ASME Section XI component, support, system pressure testing, and augmented examinations and tests for the Fourth Interval at DNPS Units 2, 3, and 2/3.

The format of the Inservice Inspection Summary Tables is as depicted below and provides the following information:

Examination Item Number Description Exam Total Number of Relief Notes Category(with (or Augmented Requirements Components by Request!

Category Number or System (or Test TAP Description) Risk Category Block Number) Number Number) I I I (I) (2) 1 (3) (4) (5) (6) (7)

(1) Examination Category (with Examination Category Description):

Provides the examination category and description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1. Only those examination categories applicable to DNPS are identified.

Examination Category "N/A" is used to identify Augmented ISI examinations and other DNPS commitments.

Examination Category "R-A" from Code Case N-578-1 is used in lieu of ASME Section XI Examination Categories B-F, B-J, C-F-I, and C-F-2 to identify Class 1 and 2 piping structural elements for the RISI program.

(2) Item Number (or Augmented Number or Risk Category Number):

Provides the item number as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1. Only those item numbers applicable to DNPS are identified.

Specific abbreviations such as BWROG and GL88-01 have been developed to identify Augmented ISI examinations and other DNPS commitments.

For piping structural elements under the RISI program, the Risk Category Number (e.g., 1-5) is used in place of the Item Number.

(3)

Description:

Provides the description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval For Augmented inspection commitments, a description of the Augmented requirement is provided.

For Risk-Informed piping examinations, a statement of the Risk Category is provided.

(4) Exam Requirements:

Provides the examination method(s) required by ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1.

Provides the examination requirements for augmented components from DNPS commitments or Relief Requests.

Provides the examination requirements for piping structural elements under the RISI Program in accordance with the EPRI Topical Report TR-1 12657, Rev. B-A and Code Case N-578-1.

(5) Total Number Of Components by System (or Test Block Number):

Provides the system designator (abbreviations). See Section 2.3, Table 2.3-1 for a list of these systems.

This column also provides the number of components within a particular system for that Item Number, Augmented Number or Risk Category Number.

Provides the unique alphanumeric identificationnumber of each test block for System Pressure Testing Categories. The number consists of the unit designation (2, 3, or 2/3), a two letter system designation, followed by two numeric characters that uniquely identify the test block within the system (i.e., 2CS01 would be the number 01 test block in the Unit 2 Core Spray system).

(6) Relief Request/TAP Number:

Provides a listing of Relief Request/Technical Approach & Position (TAP) numbers applicable to specific components, the ASME Section XI Item Number, Augmented Number, or Risk Category Number. Relief requests that generically apply to all components, or an entire class are not listed. If a Relief Request! TAP number is identified, see the corresponding relief request in Sections 8.0 and 9.0 or the technical approach and position in Section 2.5.

(7) Notes:

Provides a listing of program notes applicable to the ASME Section XI Item Number, Augmented Number, or Risk Category Number. If a program note number is identified, see the corresponding program note at the end of the Table 7.0-2.

7-2 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-A BI.11 Circumferential Shell Welds Volumetric RPV: 4 14R-04 Pressure Retaining Welds BI.12 Longitudinal Shell Welds Volumetric RPV: 18 14R-10 in Reactor Vessel B 1.21 Circumferential Head Welds Volumetric RPV: 1 B1.22 Meridional Head Welds Volumetric RPV: 6 B1.30 Shell-to-Flange Weld Volumetric RPV: I L B 1.40 Head-to-Flange Weld Volumetric & RPV: I Surface B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 31 Full Penetration Welds B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 31 14R-01 of Nozzles in.Vessels B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-I RPV: 92 Pressure Retaining B6.20 Closure Studs, in place (Reactor Vessel) Volumetric RPV: 92 Bolting, Greater Than B6.30 Closure Studs, when removed (Reactor Vessel) Volumetric & RPV: 92 2 in. In Diameter Surface B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 92 B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-1 RPV: 92 B6.180 Bolts and Studs (Pumps) Volumetric -RR: 2 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-I RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-1 RR: 2 7-3 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-I ISCO: 1 Pressure Retaining MS: 13 Bolting, 2 in. and Less RHS: 4 In Diameter RHSP: 1 RHV: 3 RR: 4 RVBD: I SBLC: 2 SDC: 5 B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-I LPCI: 2 MS: 8

__ _RR: 6 7-4 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-K B10.10 Welded Attachments to Pressure Vessels Surface RPV: 5 Welded Attachments for B 10.20 Welded Attachments to Piping Surface CS: 7 Vessels, Piping, FW: 10 Pumps, and Valves HPCI: 5 ISCO: 5 LPCI: 6 MS: 23 RHS: 2 RHV: 4 RR: 9 RWCU: 3 SBLC: 2 SDC: 10 B10.30 Welded Attachments to Pumps Surface RR: 6 B 10.40 Welded Attachments to Valves Surface RR: 2 B-L-2 B12.20 jPump Casing Visual, VT-3 RR: 2 Pump Casings I _I B-M-1 B12.40 Valve Body Welds (NPS 4 or Larger) Volumetric - MS: 4 Pressure Retaining Welds in Valve Bodies B-M-2 B12.50 Valve Body (Exceeding NPS 4) Visual, VT-3 CS: 6 Valve Bodies FW: 6 HPCI: 2 ISCO: 4 LPCI: 6 MS: 21 RR: 6 RWCU: 3 SDC: 7 7-5 Revision 0

IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-N-f R B13.10 VesselInterior Visual, VT-3 RPV: 1 Interior of Reactor Vessel I B1 n V B-N-2 B13.20 Interior Attachments Within Beltline Region Visual, VT-i RPV: 26 Welded Core Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region Visual, VT-3 RPV: 32 Attachments to Reactor Vessels B 13.40 Core Support Structure Visual, VT-3 RPV: I B-0 B14.10 1Welds in CRD Housing Volumetric or RPV: 32 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings 7-6 Revision 0

IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

TAP Number B-P B115.10 Reactor Vessel - System Leakage Test Visual, VT-2 2RCO1 14R-09 All Pressure B115.50 Piping - System Leakage Test Visual, VT-2 2SC01 14T-02 Retaining Components B 15.60 Pumps - System Leakage Test Visual, VT-2 14T-03 (Periodic) B 15.70 Valves - System Leakage Test Visual, VT-2 7-7 Revision 0

I51 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number C-A C1.30 Tubesheet-to-Shell Welds Volumetric LPCI: 4 Pressure Retaining Welds in Pressure Vessels _

C-B C2.21 Nozzle-to-Shell (or Head) Welds without Reinforcing Plates Volumetric & ISCO: 4 Pressure Retaining in Vessels, Greater than 1/2" Nominal Thickness Surface Nozzle Welds in C2.31 Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface ECCS: 4 Vessels with Reinforcing Plates in Vessels, Greater than 1/2" LPCI: 8 Nominal Thickness C2.33 Nozzle-to-Shell (or Head) Welds with Reinforcing Plates Visual, VT-2 ECCS: 4 when Inside of Vessel is Inaccessible for Vessels, Greater LPCI: 4 than 1/2" Nominal Thickness C-C C3.10 Welded Attachments to Pressure Vessels Surface LPCI: 4 Welded Attachments C3.20 Welded Attachments to Piping Surface CRD: 2 For Vessels, Piping, CS: 14 Pumps, and Valves ECCS: 16 HPCI: 17 ISCO: 2 LPCI: 25 RWCU: I C3.30 Welded Attachments to Pumps Surface CS: 2 HPCI: 1 LPCI: 4 7-8 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

TAP Number C-H C7.10 Pressure Vessels-System Leakage Test Visual, VT-2 2CS01 14R-05 All Pressure C7.30 Piping-System Leakage Test Visual, VT-2 2ECO1 14R-07 Retaining C7.50 Pumps-System Leakage Test Visual, VT-2 2EC02 14R-09 Components C7.70 Valves-System Leakage Test Visual, VT-2 2HPOI 14T-02 (Periodic) 2HP02 14T-03 2HP03 2LPOI 2N1B01 2RCO1 2RC02 2SC01 2SC02 2SC03 7-9 Revision 0

IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number D-A D11.10 Welded Attachments to Pressure Vessels Visual, VT-1 ISCO: 3 Welded Attachments D1.20 Welded Attachments to Piping Visual, VT-1 CCSW: 3 For Vessels, Piping, SRVD: 16 Pumps, and Valves 7-10 Revision 0

1ST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

TAP Number D-B D2.1O Pressure Vessels-System Leakage Test Visual, VT-2 2CCO1 14R-06 All Pressure D2.30 Piping-System Leakage Test Visual, VT-2 2DGOI 14R-09 Retaining D2.50 Pumps-System Leakage Test Visual, VT-2 2DG02 14T-O I Components D2.70 Valves-System Leakage Test Visual, VT-2 21C0 1 14T-02 (Periodic) 21C02 14T-03 D-B D2.20 Pressure Vessels-System Hydrostatic Test Visual, VT-2 2CCOI 14R-06 All Pressure D2.40 Piping-System Hydrostatic Test Visual, VT-2 2DGO1 14R-09 Retaining D2.60 Pumps-System Hydrostatic Test Visual, VT-2 2DG02 14T-O I Components D2.80 Valves-System Hydrostatic Test Visual, VT-2 21C0I 14T-02 (Interval) 21C02 14T-03 7-11 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components Relief Request/

TAP Number E-A El.11 Accessible Surface Areas General Visual 349 MCR-02 Containment Surfaces El.12 Wetted Surfaces of Submerged Areas General Visual 16 MCR-02 El.20 BWR Vent System General Visual 45 MCR-02 Accessible Surface Areas El.30 Moisture Barriers General Visual 4 MCR-02 E-C E4.11 Visible Surfaces Detailed Visual 0 MCR-02 Containment Surfaces Requiring E4.12 Surface Area Grid UT Thickness 0 MCR-02 Augmented Examination Grid Line Intersections and Minimum Wall Thickness Locations 7-12 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number F-A F1.10 Class 1 Piping Supports Visual, VT-3 CS: 15 Supports FW: 14 HPCI: 8 ISCO: 7 LPCI: 8 MS: 44 RHS: 8 RHV: 11 RR: 21 RVBD: 4 RWCU: 8 SBLC: 13 SDC: 16 F1.20 Class 2 Piping Supports Visual, VT-3 CRD: 28 CS: 61 ECCS: 34 HPCI: 86 ISCO: 21 LPCI: 81 RWCU: 1 7-13 Revision 0

IS! Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number F-A F1.30 Class 3 Piping Supports Visual, VT-3 CCSW: 116 Supports DGSW: 71 (Continued) ISCO: 3 SRVD: 47 F1.40 Supports Other Than Piping Supports Visual, VT-3 CCSW: 4 (Class 1, 2, 3, and MC) CS: 2 DGSW: 1 HPCI: 2 LPCI: 8 RPV: 5 RR: 22 7-14 Revision 0

IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number N/A BWROG Alternate BWR Feedwater Nozzle Inspection Requirement Volumetric RPV: 4 Augmented Components (BWROG)

Components GL88-01 Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category C: 94 4 Austenitic Stainless Steel Piping Components Category D: 65 Category E: 38 Category F: 2 Category G: 2 7-15 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-1 Examination Category Risk Description Exam Total Number of Relief Request/ Notes (with Category Description) Category Requirements Components by TAP Number Number System R-A 1 Risk Category 1 See Notes FW: 25 14R-02 3 Risk-Informed Piping 2 Risk Category 2 See Notes CS: 19 5 Examinations ISCO: 11 LPCI: 34 SDC: 47 3 Risk Category 3 See Notes MS: 57 4 Risk Category 4 See Notes CS: 37 ECCS: 56 HPCI: 73 ISCO: 4 LPCI: 56 MS: 117 RPV: 4 RR: 19 RWCU: I1 SBLC: 23 5 Risk Category 5 See Notes HPCI: 32 ISCO: 3 RPV: 6 RVBD: 7 SBLC: 13 7-16 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-A BI.11 Circumferential Shell Welds Volumetric RPV: 4 14R-04 Pressure Retaining Welds Bl.12 Longitudinal Shell Welds Volumetric RPV: 14 14R-10 in Reactor Vessel BI.21 Circumferential Head Welds Volumetric RPV: 1 B 1.22 Meridional Head Welds Volumetric RPV: 6 B1.30 Shell-to-Flange Weld Volumetric RPV: I B 1.40 Head-to-Flange Weld Volumetric & RPV: I Surface B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 31 Full Penetration Welds B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 31 I4R-01 of Nozzles in Vessels B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-1 RPV: 92 Pressure Retaining B6.20 Closure Studs, in place (Reactor Vessel) Volumetric RPV: 92 Bolting, Greater Than B6.30 Closure Studs, when removed (Reactor Vessel) Volumetric & RPV: 92 2 in. In Diameter Surface B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 92 B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-1 RPV: 92 B6.180 Bolts and Studs (Pumps) Volumetric RR: 2 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-l RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-1 RR: 2 7-17 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-I ISCO: I MS: 13 Pressure Retaining RHS: 4 Bolting, 2 in. and Less RHSP: I In Diameter RHV: 3 RR: 6 RWCU: 2 SBLC: 2 SDC: 5 B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-I LPCI: 2 MS: 8 RR: 4 7-18 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-K B110.10 Welded Attachments to Pressure Vessels Surface RPV: 5 Welded Attachments for B 10.20 Welded Attachments to Piping Surface CS: 3 Vessels, Piping, FW: 14 Pumps, and Valves HPCI: 7 ISCO: 4 LPCI: 2 MS: 23 RHS: 2 RR: 4 RWCU: 3 SDC: 4 B10.30 Welded Attachments to Pumps Surface RR: 6 3B10.40 Welded Attachments to Valves Surface RR: 2 B-L-2 B12.20 IPump Casing Visual, VT-3 RR: 2 Pump Casings I I B-M-1 B12.40 Valve Body Welds (NPS 4 or Larger) Volumetric MS: 4 Pressure Retaining Welds in Valve Bodies B-M-2 B12.50 Valve Body (Exceeding NPS 4) Visual, VT-3 CS: 6 Valve Bodies FW: 6 HPCI: 2 ISCO: 4 LPCI: 6 MS: 21 RR: 4 RWCU: 4 SDC: 7 7-19 Revision 0

IS Program Plan Dresden Nuclear Power Station Units 2 & 3. Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number B-N-e R B13.10 Vessel Interior Visual, VT-3 Interior of Reactor Vessel IIIIII RPV: I B-N-2 B13.20 Interior Attachments Within Beltline Region Visual, VT-I RPV: 46 Welded Core Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region Visual, VT-3 RPV: 32 Attachments to Reactor Vessels B 13.40 Core Support Structure Visual, VT-3 RPV: 1 B-O B14.10 Welds in CRD Housing Volumetric or RPV: 32 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings I I 7-20 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

TAP Number B-P B15.10 Reactor Vessel - System Leakage Test Visual, VT-2 3RC01 14R-09 All Pressure B 15.50 Piping - System Leakage Test Visual, VT-2 3SC01 14T-02 Retaining Components B 15.60 Pumps - System Leakage Test Visual, VT-2 14T-03 (Periodic) B 15.70 Valves - System Leakage Test Visual, VT-2 7-21 Revision 0

IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number C-A C1.30 Tubesheet-to-Shell Weld Volumetric LPCI: 4 Pressure Retaining Welds in Pressure Vessels C-B C2.21 Nozzle-to-Shell (or Head) Weld without Reinforcing Plates Volumetric & ISCO: 4 Pressure Retaining in Vessels, Greater than 1/2" Nominal Thickness Surface Nozzle Welds in C2.31 Reinforcing Plate Welds to Nozzle and Vessel for Nozzles Surface ECCS: 4 Vessels with Reinforcing Plates in Vessels, Greater than 1/2" LPCI: 8 Nominal Thickness C2.33 Nozzle-to-Shell (or Head) Welds with Reinforcing Plates Visual, VT-2 ECCS: 4 when Inside of Vessel is Inaccessible for Vessels, Greater LPCI: 4 than 1/2" Nominal Thickness C-C C3.10 Welded Attachments to Pressure Vessels Surface LPCI: 4 Welded Attachments C3.20 Welded Attachments to Piping Surface CRD: 7 For Vessels, Piping, CS: 16 Pumps, and Valves ECCS: 16 FW: 1 HPCI: 17 ISCO: 2 LPCI: 25 C3.30 Welded Attachments to Pumps Surface CS: 2 HPCI: 1 I_ I I LPCI: 4 7-22 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

ITAP Number C-H C7.10 Pressure Vessels-System Leakage Test Visual, VT-2 3CSO1 14R-05 All Pressure C7.30 Piping-System Leakage Test Visual, VT-2 3ECO1 14R-07 Retaining C7.50 Pumps-System Leakage Test Visual, VT-2 3EC02 14R-09 Components C7.70 Valves-System Leakage Test Visual, VT-2 3HP01 14T-02 (Periodic) 3HP02 14T-03 3HP03 3LPOI 3NBO1 3RCO0 3RC02 3SCOI 3SC02 3SC03 7-23 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number D-A DI.10 Welded Attachments to Pressure Vessels Visual, VT-I ISCO: 3 Welded Attachments D13.20 Welded Attachments to Piping Visual, VT-i CCSW: 13 2 For Vessels, Piping, DGSW: 0+1 Pumps, and Valves SRVD: 16 7-24 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Test Block Approved Notes (with Category Description) Number Requirements Number Relief Request/

TAP Number D-B D2.10 Pressure Vessels-System Leakage Test Visual, VT-2 3CCOI I4R-06 All Pressure D2.30 Piping-System Leakage Test Visual, VT-2 3DGO1 14R-09 Retaining D2.50 Pumps-System Leakage Test Visual, VT-2 3DG02 14T-O 1 Components D2.70 Valves-System Leakage Test Visual, VT-2 2/3DGO1 14T-02 (Periodic) 2/3DG02 14T-03 2/3DG03 31C01 31C02 D-B D2.20 Pressure Vessels-System Hydrostatic Test Visual, VT-2 3CCO1 14R-06 All Pressure D2.40 Piping-System Hydrostatic Test Visual, VT-2 3DGO1 I4R-09 Retaining D2.60 Pumps-System Hydrostatic Test Visual, VT-2 3DG02 14T-01 Components D2.80 Valves-System Hydrostatic Test Visual, VT-2 2/3DGOI 14T-02 (Interval) 2/3DG02 14T-03 2/3DG03 3IC01 31C02 7-25 Revision 0

1ST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components Relief Request/

TAP Number E-A E1.1 1 Accessible Surface Areas General Visual 280 MCR-02 Containment Surfaces El.12 Wetted Surfaces of Submerged Areas General Visual 16 MCR-02 E1.20 BWR Vent System General Visual 45 MCR-02 Accessible Surface Areas E1.30 Moisture Barriers General Visual 4 MCR-02 E-C E4.11 Visible Surfaces Detailed Visual 0 MCR-02 Containment Surfaces Requiring E4.12 Surface Area Grid UT Thickness 21 MCR-02 Augmented Examination Grid Line Intersections and Minimum Wall Thickness Locations 7-26 Revision 0

1ST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number F-A Fl.10 Class 1 Piping Supports Visual, VT-3 CS: 12 Supports FW: 18 HPCI: 10 ISCO: 7 LPCI: 8 MS: 43 RHS: 6 RHV: 4 RR: 13 RVBD: 3 RWCU: 11 SBLC: 33 SDC: 13 FI.20 Class 2 Piping Supports Visual, VT-3 CRD: 29 CS: 53 ECCS: 34 FW: 1 HPCI: 75 ISCO: 23 LPCI: 80 RWCU: I 7-27 Revision 0

IS! Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

System TAP Number F-A F1.30 Class 3 Piping Supports Visual, VT-3 CCSW: 108 1 Supports DGSW: 78+80 2 (Continued) ISCO: 3 SRVD: 47 F17.40 Supports Other Than Piping Supports Visual, VT-3 CCSW: 4 1 (Class 1, 2, 3, and MC) CS: 2 2 DGSW: 1+1 HPCI: 2 LPCI: 12 RPV: 5 RR: 22 7-28 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Approved Notes (with Category Description) Number Requirements Components by Relief Request/

I System TAP Number N/A BWROG Alternate BWR Feedwater Nozzle Inspection Requirement Volumetric RPV: 4 Augmented Components (BWROG)

Components GL88-01 Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category C: 50 4 Austenitic Stainless Steel Piping Components Category D: 13 7-29 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Unit 3 Inservice Inspection Summary Table 7.0-2 Examination Category Risk Description Exam Total Number of Relief Request/ Notes (with Category Description) Category Requirements Components by TAP Number Number System R-A 1 Risk Category 1 See Notes FW: 25 I4R-02 3 Risk-Informed Piping 2 Risk Category 2 See Notes CS: 9 5 Examinations ISCO: 4 LPCI: 16 SDC: 44 3 Risk Category 3 See Notes MS: 67 4 Risk Category 4 See Notes CS: 51 ECCS: 56 HPCI: 66 ISCO: 3 LPCI: 40 MS: 110 RHV: 1 RPV: 4 RR: 80 RWCU: 1I SBLC: 38 5 Risk Category 5 See Notes HPCI: 29 ISCO: 7 "RPV: 6 SBLC: 17 7-30 Revision 0

IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Inserviee Inspection Summary Table Program Notes Note P Note Summary I ISI snubber visual examinations are performed in accordance with the ASME OM Code, Subsection ISTD Program. The number of DNPS Unit 2, 3, and 2/3 supports identified includes snubbers for visual examination of the integral and nonintegral attachments per Paragraphs IWF-5200(c) and IWF-5300(c). The snubbers are scheduled and administratively tracked in the ISI Program; however, the ASME OM Code, Subsection ISTD Program will be the mechanism for actually performing the visual examinations scheduled within the ISI Program.

2 The Unit 3 population counts include those components that are common to both units (typically designated as "2/3") and are listed in Table following a "+"

symbol.

3 For the Fourth Inspection Interval, DNPS's Class I and 2 piping inspection program will be governed by risk-informedregulations. The RISI program m\ethodology is described in the EPRI Topical Report TR-I 12657, Rev. B-A and Code Case N-578-1. The RISI program scope will supercede the 1995 Edition with the 1996 Addenda of the ASME Section XI Code examination program for Class 1 B-F and B-J welds and Class 2 C-F-I and C-F-2 welds in accordance with 10 CFR 50.55a(a)(3Xi).

4 IGSCC Category A welds subsumed into the RISI program.

5 Examination requirements within the RISI program are determinedby the various degradationmechanisms present at each individual piping structural element.

See EPRI TR- 112657, Rev. B-A and Code Case N-578-1 for specific exam method requirements.

7-31 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests written per 10 CFR 50.55a(a)(3)(i) for situations where alternatives to ASME Section XI requirements irovide an acceptable level of quality and safety; per 10 CFR 50.55a(a)(3)(ii) for situations where compliance with ASME Section XI requirements results in a hardship or an unusual difficulty without a compensating increase in the level of quality and safety; and per 10 CFR 50.55a(g)(5)(iii) for situations where ASME Section XI requirements are considered impractical.

The following NRC guidance was utilized to determine the correct 10 CFR 50.55a paragraph citing for DNPS relief requests. 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(a)(3)(ii) provide alternatives to the requirements of ASME Section XI, while 10 CFR 50.55a(g)(5)(iii) recognizes situational impracticalities.

10 CFR 50.55a(a)(3)(i): Cited in relief requests when alternatives to the ASME Section XI requirements which provide an acceptable level of quality and safety are proposed. Examples are relief requests which propose alternative non-destructive examination (NDE) methods and/or examination frequency.

10 CFR 50.55a(a)(3)(ii): Cited in relief requests when compliance with the ASME Section XI requirements is deemed to be a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components solely to provide access for examinations, and development of sophisticated tooling that would result in only minimal increases in examination coverage.

10 CFR 50.55a(g)(5)(iii): Cited in relief requests when conformance with ASME Section XI requirements is deemed impractical. Examples of impractical requirements are situations where the component would have to be redesigned, or replaced to enable the required inspection to be performed.

An index for DNPS relief requests is included in Table 8.0-1. The "14R-XX" relief requests are applicable to ISI and SPT.

8-1 Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 8.0-1 RELIEF REQUEST INDEX Sheet 1 of 2 Relief Revision Status 2 (Program) Description / Approval Summary' Request Date3 14R-01 0 Submitted (ISI) Inspection of Standby Liquid Control nozzle 8/30/02 inner radius. Revision 0 Submitted.

I4R-02 0 Submitted (ISI) Alternate Risk-Informed Selection and Examination Criteria for Category B-F, B-J, C-F-I, and C-F-2 Pressure Retaining Piping Welds.

Revision 0 Submitted.

14R-03 0 Submitted (ISI) Alternative Requirementsto ASME SectionXI, 8/30/02 Appendix VII, SubsubarticleVII-240, "Annual Training". Revision 0 Submitted.

14R-04 0 Submitted (ISI) Alternative Requirements to Appendix VIII, 8/30/02 Supplement 4, "Qualification Requirements for the Clad/Base Metal Interface of Reactor Pressure Vessel". Revision 0 Submitted.

I4R-05 0 Submitted (SPT) Exemption from pressure testing Reactor Pressure Vessel Head Flange Seal Leak Detection System. Revision 0 Submitted.

14R-06 0 Submitted (SPT) Testing Frequency for Isolation Condenser 8/30/02 Shell Side and Vent Piping. Revision 0 Submitted.

I4R-07 0 Submitted (SPT) Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators.

Revision 0 Submitted.

14R-08 0 Submitted (ISI) Pressure Retaining Welds in Piping Subject to 8/30/02 Appendix VIII, Supplement 11. Revision 0 Submitted.

I4R-09 0 Submitted (SPT) Alternative Rules for Corrective Measures if 8/30/02 Leakage Occurs at Bolted Connections. Revision 0 Submitted.

14R-10 0 Authorized (ISI) Exemption from Volumetric Examination of All Reactor Pressure Vessel Circumferential Welds.

Authorized per SER dated 2/25/00.

8-2 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval TABLE 8.0-1 RELIEF REQUEST INDEX Sheet 2 of 2 Relief Revision Status 2 (Program) Description / Approval Summary'1 3

Request Date I4R-1 1 0 Submitted (ISI) Evaluation Criteria for Temporary Acceptance 8/30/02_

_ of Flaws. Revision 0 Submitted.

Note 1: The NRC grants relief requests pursuant to 10 CFR 50.55a(g)(6)(i) when Code requirements cannot be met and proposed alternatives do not meet the criteria of 10 CFR 50.55(a)(3). The NRC authorizes relief requests pursuant to 10 CFR 50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level of quality and safety or under (3)(ii) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety.

Note 2: This column represents the status of the latest revision. Relief Request Status Options: Authorized Approved for use in an NRC SER (See Note 1); Granted - Approved for use in an NRC SER (See Note 1);

Authorized Conditionally - Approved for use in an NRC SER which imposes certain conditions; Granted Conditionally - Approved for use in an NRC SER which imposes certain conditions; Denied - Use denied in an NRC SER; Expired - Approval for relief has expired; Withdrawn - Relief has been withdrawn by the station; Not Required - The NRC has deemed the relief unnecessary in an SER or RAT; Cancelled - Relief has been cancelled by the station prior to issue; Submitted - Relief has been submitted to the NRC by the station and is awaiting approval.

Note 3: The revision listed is the latest revision of the subject relief request. The date this revision became effective is the date of the approving SER which is listed in the fourth column of the table. The date noted in the second column is the date of the ISI Program Plan revision when the relief request was incorporated into the document.

8-3 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-01 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

IWB-2500 Table IWB-2500-1 Examination Category: B-D Item Number: B3.100

==

Description:==

Inspection of Standby Liquid Control Nozzle Inner Radius.

ComponentNumber: Unit 2: N 12-1 Unit 3: N12-1 CODE REQUIREMENT IWB-2500 states that components shall be examined and tested as specified in Table IWB-2500-1.

Table IWB-2500-1 requires a volumetric examination to be performed on the inner radius section of all reactor pressure vessel nozzles each inspection interval.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(g)(5)(iii),relief is requested on the basis that conformance with the Code requirements is impractical.

The Standby Liquid Control (SBLC) nozzle, as shown in Figure 14R-01.1, is designed with an integral socket to which the boron injection piping is fillet welded. The SBLC nozzle is located near the bottom of the vessel in an area which is inaccessible for ultrasonic examinations from the inside of the vessel. Therefore, ultrasonic examinations would need to be performed from the outside diameter of the vessel. As shown in Figure 14R-01.1, the ultrasonic scan would need to travel through the full thickness of the vessel into a complex cladding/socket configuration. These geometric and material reflectors inherent in the design prevent a meaningful examination from being performed on the inner radius of the SBLC nozzle.

In addition, the inner radius socket attaches to piping which injects boron at locations far removed from the nozzle. Therefore, the SBLC nozzle inner radius is not subjected to turbulent mixing conditions that are a concern at other nozzles.

PROPOSED ALTERNATE EXAMINATION As an alternate examination, DNPS will perform a VT-2 visual examination of the subject nozzles each refueling outage in conjunction with the Class 1 System Leakage Test.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER:, 14R-01 (Page 2 of 2)

FIGURE 14R-01.1 2 INCH STANDBY LIQUID CONTROL NOZZLE Vessel (Mn-Mo)

I.D Cladding Safe-End Nozzle Pipe (Mn-Mo) 3/4" R Min 6 3/8" Revision 0

IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: I4R-02 (Page 1 of 5)

COMPONENT IDENTIFICATION Code Class: 1 and 2 Examination Category: B-F, B-J, C-F-I, and C-F-2 Item Number: B5.10, B5.20, B9.11, B9.21, B9.31, B9.32, B9.40, C5.1 1, C5.41, C5.51, C5.70, and C5.81

==

Description:==

Alternate Risk-Informed Selection and Examination Criteria for Category B-F, B-J, C-F-i, and C-F-2 Pressure Retaining Piping Welds Component Number: Pressure Retaining Piping

Reference:

1) Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure"

2) W. H. Bateman (USNRC) to G. L. Vine (EPRI) letter dated October 28, 1999 transmitting "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-1 12657, Revision B, July 1999)"

3) Initial Risk-Informed Inservice Inspection Evaluation Dresden Nuclear Power Station Units 2 and 3 (Dated July 2000)

4) American Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B"

5) A. J. Mendiola (USNRC) to 0. D. Kingsley (Exelon) letter dated September 5, 2001 transmitting "Safety Evaluation of Third Interval Risk-Informed Inservice Inspection Program Relief Request" CODE REOUIREMENT Table IWB-2500-1, Examination Category B-F, requires volumetric and/or surface examinations on all welds for Items B5.10 and B5.20.

Table IWB 2500-1, Examination Category B-J, requires volumetric and/or surface examinations on a sample of welds for Items B9.11, B9.21, B9.31, B9.32, and B9.40. The weld population selected for inspection includes the following:

Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 2 of 5)

CODE REQUIREMENT (Continued)

I1. All terminal ends in each pipe or branch run connected to vessels.

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

a. primary plus secondary stress intensity range of 2.4Sr for ferritic steel and austenitic steel.

b. cumulative usage factor U of 0.4.

3. All dissimilar metal welds not covered under Category B-F.

4. Additional piping welds so that the total number of circumferential butt welds, branch connections, or socket welds selected for examination equals 25% of the circumferential butt welds, branch connection, or socket welds in the reactor coolant piping system. This total does not include welds excluded by IWB-1220.

Table IWC-2500-1, Examination Categories C-F-1 and C-F-2 require volumetric and/or surface examinations on a sample of welds for Items C5.11, C5.41, C5.51, C5.70, and C5.81. The weld population selected for inspection includes the following:

1. Welds selected for examination shall include 7.5%, but not less than 28 welds, of all dissimilar metal, austenitic stainless steel and high alloy welds (Category C-F-i) or of all carbon and low alloy steel welds (Category C-F-2) not exempted by IWC-1220. (Some welds not exempted by IWC-1220 are not required to be nondestructively examined per Examination Categories C-F-1 and C-F-2. These welds, however, shall be included in the total weld count to which the 7.5% sampling rate is applied.) The examinations shall be distributed as follows:

a. the examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt dissimilar metal, austenitic stainless steel and high alloy welds (Category C-F-i) or carbon and low alloy welds (Category C-F-2) in each system;

b. within a system, the examinations shall be distributed among terminal ends, dissimilar metal welds, and structural discontinuities prorated, to the degree practicable, on the number of nonexempt terminal ends, dissimilar metal welds, and structural discontinuities in the system; and

c. within each system, examinations shall be distributed between line sizes prorated to the degree practicable.

Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 3 of 5)

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative utilizing Reference 1 along with two enhancements from Reference 4 will provide an acceptable level of quality and safety.

As stated in "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-1 12657, Revision B, July 1999)" (Reference 2):

"The staff concludes that the proposed RI-ISI program as described in EPRI TR-1 12657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10 CFR 50.55a for the proposed alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection."

The initial DNPS RISI Program was submitted during the Third Period of the Third Interval for both Units 2 and 3. This initial RISI program was developed in accordance with EPRI TR-1 12657, Revision B-A, as supplemented by Code Case N-578-1. The program was approved for use by the USNRC via Safety Evaluation as transmitted to Exelon on September 5, 2001 (Reference 5).

The transition from the 1989 Edition to the 1995 Edition with the 1996 Addenda of ASME Section XI for DNPS's Fourth Interval does not impact the currently approved Risk-Informed ISI evaluation process used in the Third Interval, and the requirements of the new Code edition/addenda will be implemented as detailed in the DNPS ISI Program Plan.

The Risk Impact Assessment completed as part of the original baseline RISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI program to the new RISI methodology. For the Fourth Interval ISI update, there is no transition occurring between two different methodologies, but rather, the currently approved RISI methodology and evaluation will be maintained for the new interval. As such, the initial screening of the risk impact assessment is not a part of the living program process and is not required to be continually updated.

The actual evaluation and ranking procedure including the Consequence Evaluation and Degradation Mechanism Assessment processes of the currently approved (Reference 5) RISI Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-1 12657, Revision B-A. These portions of the RISI Program are reevaluated as major revisions of the site PRA occur and modifications to plant configuration are made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, and Element Selection steps define the living program process applicable to the RISI Program.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: I4R-02 (Page 4 of 5)

PROPOSED ALTERNATE PROVISIONS The proposed alternative originally implemented in the "Risk Informed Inservice Inspection Plan, Dresden Units 2 and 3" (Reference 3), along with the two enhancements noted below, provide an acceptable level of quality and safety as required by 10 CFR 50.55a(a)(3)(i). This original program along with these same two enhancements is currently approved for DNPS's Third Inspection Interval as documented in Reference 5.

The Fourth Interval RISI Program will be a continuation of the current application and will continue to be a living program as described in the Basis For Relief above. No changes to the evaluation methodology as currently implemented under EPRI TR-1 12657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISI Selected Examinations" of EPRI TR-1 12657, DNPS will utilize the requirements of Subarticle -2430, "Additional Examinations" contained in Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in Code Case N-578-1 provides a more refined methodology for implementing necessary additional examinations.

To supplement the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of EPRI TR-1 12657, DNPS will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations" contained in Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 1995 Edition with the 1996 Addenda of ASME Section XI (DNPS's code of record for the Fourth Interval) will be utilized which parallel those referenced in the Code Case for the 1989 Edition. Table 1 of Code Case N-578-1 will be used as it provides a more detailed breakdown for examination method and categorization of parts to be examined.

The DNPS RISI Program, as developed in accordance with EPRI TR-1 12657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as "High" risk (i.e., Risk Category 1, 2, and 3) and 10% of the elements that are categorized as "Medium" risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI TR-1 12657 while the guidance for the examination method and categorization of parts to be examined are provided by the EPRI TR-1 12657 as supplemented by Code Case N-578-1.

In addition to this risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to receive Code required pressure testing as part of the current ASME Section XI program. VT-2 visual examinations are scheduled in accordance with the DNPS pressure testing program, which remains unaffected by the RISI program.

Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 5 of 5)

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

Revision 0

151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER:. 14R-03 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: All

Reference:

ASME Section XI, Appendix VII, Subsubarticle VII-4240, "Annual Training" Examination Category: All categories for components subject to Ultrasonic Examination Item Number: All item numbers for components subject to Ultrasonic Examination

==

Description:==

Alternative Requirements to ASME Section XI, Appendix VII, Subsubarticle VII-240, "Annual Training" Component Number: All Components Subject to Ultrasonic Examination CODE REOUIREMENT 10 CFR 50.55a, "Codes and Standards," Paragraph (b)(2) incorporates by reference, the 1995 Edition and Addenda through 1996 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for use in preparing inservice inspection programs. Subsubarticle VII-4240, "Annual Training," of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII, requires a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> annual training.

10 CFR 50.55a(b)(2)(xiv), "Appendix VIII personnel qualification," requires that all personnel qualified to perform ultrasonic examinations in accordance with ASME Section XI, Appendix VIII, shall receive 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of annual hands-on training on specimens that contain cracks. This training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from the training provision of Subsubarticle VII-4240 of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII, that requires a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> annual training. The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

On September 22, 1999, the USNRC published a final rule in the Federal Register (64 FR 51370) to amend 10 CFR 50.55a(b)(2), to incorporate by reference the 1995 Edition and addenda through the 1996 Addenda, of ASME Section XI. The change included the requirement to have a minimum of 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of annual training contained in Subsubarticle VII-4240 of ASME Section XI.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-03 (Page 2 of 3)

BASIS FOR RELIEF (Continued)

Additionally, the September 22, 1999, Federal Register notice amended 10 CFR 50.55a(b)(2)(xiv). The amended 10 CFR 50.55a(b)(2)(xiv) requires that all personnel qualified to perform ultrasonic examinations in accordance with Appendix VIII shall receive 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of annual hands-on training on specimens that contain cracks. This training must be taken no earlier than 6 months prior to performing examinations at a licensee's facility. Paragraph 2.4.1.1.1 in the Federal Register notice contained the following statement which includes a discussion of the Electric Power Research Institute (EPRI) Performance Demonstration Initiative (PDI) program.

"The USNRC had determined that this requirement (i.e., Subsubarticle VII-4240) was inadequate for two reasons. The first reason was that the training does not require laboratory work and examination of flawed specimens. Signals can be difficult to interpret and, as detailed in the regulatory analysis for this rulemaking, experience and studies indicate that the examiner must practice on a frequent basis to maintain the capability for proper interpretation. The second reason is related to the length of training and its frequency. Studies have shown that an examiner's capability begins to diminish within approximately 6 months if skills are not maintained. Thus, the USNRC had determined that 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> of annual training is not sufficient practice to maintain skills, and that an examiner must practice on a more frequent basis to maintain proper skill level... The PDI program has adopted a requirement for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of training, but it is required to be hands-on practice. In addition, the training must be taken no earlier than 6 months prior to performing examinations at a licensee's facility. PDI believes that 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> will be acceptable relative to an examiner's abilities in this highly specialized skill area because personnel can gain knowledge of new developments, material failure modes, and other pertinent technical topics through other means. Thus, the USNRC has decided to adopt in the Final Rule the PDI position on this matter. These changes are reflected in 10 CFR 50.55a(b)(2)(xiv) of the final rule."

Implementation of the training requirements contained in Subsubarticle VII-4240 of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII and 10 CFR 50.55a(b)(2)(xiv) will result in redundant training programs. The approval of this Relief Request, to qualify our personnel to perform ultrasonic examinations in accordance with 10 CFR 50.55a(b)(2)(xiv), will simplify record keeping, satisfy the need to maintain skills, and provide an acceptable level of quality and safety.

PROPOSED ALTERNATIVE PROVISIONS Annual ultrasonic training shall be conducted in accordance with 10 CFR 50.55a(b)(2)(xiv) in lieu of Subsubarticle VII-4240 of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII. The annual ultrasonic training shall require that all personnel qualified for performing ultrasonic examinations in accordance with ASME Section XI, Appendix VIII, shall receive 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of annual hands-on training on specimens that contain cracks. This training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility.

Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-03 (Page 3 of 3)

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-04 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

ASME Section XI, Table IWB-2500-1 ASME Section XI, Appendix VIII, Supplement 4, Subparagraph 3.2(c)

Examination Category: B-A Item Number: B1.10, B1. 11, Bl.12, B1.20, B1.21, B1.22, B1.50, B1.51

==

Description:==

Alternative Requirements to Appendix VIII, Supplement 4, "Qualification Requirements for the Clad/Base Metal Interface of Reactor Pressure Vessel" Component Number: All Components CODE REQUIREMENT 10 CFR 50.55a(b)(2) incorporates by reference, the 1995 Edition with the 1996 of ASME Section XI for use in preparing inservice inspection programs.

Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, Supplement 4, requires that the ultrasonic testing (UT) performance demonstration results be plotted on a two dimensional plot with the measured depth plotted along the ordinate axis and the true depth plotted along the abscissa axis. For qualification, the plot must satisfy the statistical parameters identified in Subparagraph 3.2(c).

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from the statistical parameters identified in Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, Supplement 4. The basis of the relief requests is that the proposed alternatives would provide an acceptable level of quality and safety.

On September 22, 1999, the USNRC published a final rule in the Federal Register (64 FR 51378) to amend 10 CFR 50.55a(b)(2), to incorporate by reference the 1995 Edition and addenda through the 1996 Addenda, of ASME Section XI. The change included the provisions of Subparagraph 3.2(a), 3.2(b) and 3.2(c) of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VIII, Supplement 4.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-04 (Page 2 of 2)

BASIS FOR RELIEF (Continued)

Additionally, the September 22, 1999, Federal Register amended 10 CFR 50.55a(b)(2)(xv)(C)(1). The amended 10 CFR 50.55a(b)(2)(xv)(C)(1), requires a depth sizing acceptance criterion of 0.15 inch Root Mean Square (RMS) to be used in lieu of the requirements of Subparagraph 3.2(a) and 3.2(b) of ASME Section XI, Appendix VIII, Supplement 4.

On March 26, 2001, the USNRC published a correction to the September 22, 1999, final rule in the Federal Register (66 FR 16390). The USNRC identified that an error had occurred in the published wording of 10 CFR 50.55a(b)(2)(xv)(C)(1). The corrected 10 CFR 50.55a(b)(2)(xv)(C)(1), requires a depth sizing acceptance criterion of 0.15 inch Root Mean Square (RMS) to be used in lieu of the requirements of Subparagraph 3.2(a) and a length sizing requirement of 0.75 inch RMS to be used in lieu of the requirements 3.2(b) of ASME Section XI, Appendix VIII, Supplement 4.

The statistical parameters to be used in flaw sizing specified in Subparagraph 3.2(c) of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VIII, Supplement 4, rely upon the depth sizing acceptance criteria used in Subparagraph 3.2(a) and the length sizing acceptance criteria used in Subparagraph 3.2(b). For Supplement 4 UT performance demonstrations, the linear regression line of the data required by Subparagraph 3.2(c) is not applicable because the performance demonstrations are performed on test specimens with flaws located on the inner 15% through-wall. Additionally, the Subparagraph 3.2(c) specified value for evaluating the mean deviation of flaw depth is not restrictive enough for evaluating flaw depths within the inner 15% of wall thickness. We propose to use the 10 CFR 50.55a(b)(2)(xv)(C)(1) RMS calculations of Subparagraph 3.2(a), which utilizes an RMS value of 0.15 inch depth and the RMS calculations of Subparagraph 3.2(b), which utilizes an RMS value of 0.75 inch length, in lieu of the statistical parameters of 3.2(c).

PROPOSED ALTERNATIVE PROVISIONS The RMS calculations of Subparagraph 3.2(a) of ASME Section XI, Appendix VIII, Supplement 4, which utilize an RMS value of 0.15 depth and the RMS calculations of Subparagraph 3.2(b),

which utilizes an RMS value of 0.75 length shall be used in lieu of the statistical parameters of Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, Supplement 4.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 1 of 6)

COMPONENT IDENTIFICATION Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H Item Number: C7.10, C7.30, C7.50, C7.70

==

Description:==

Exemption From Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System Component Number:

Unit No. Drawing Test Block No.

Unit 2 M-26 SH. 1 2NB01 Unit 3 M-357 SH. 1 3NB01 CODE REOUIREMENT Table IWC-2500-1 requires a Visual VT-2 examination to be performed during a system leakage test.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

The Reactor Pressure Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver plated O-ring located on the vessel flange. A second O-ring is located on the opposite side of the tap in the vessel flange (See Figure 14R-05.2).

This line is required during plant operation in order to indicate failure of the inner flange seal 0 ring. Failure of the O-ring would result in the annunciation of a High Level Alarm in the control room. On this annunciation, control room operators would quantify the leakage rate from the O-ring and then isolate the leak detection line from the drywell sump by closing the AO 2(3) -220-51 valve (see Figure 14R-05.1). This action is taken in order to prevent steam cutting of the O-ring and the vessel flange. Failure of the inner O-ring is the only condition under which this line is pressurized.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 2 of 6)

BASIS FOR RELIEF (Continued)

The configuration of this system precludes manual testing while the vessel head is removed because the odd configuration of the vessel tap (See Figure 14R-05.2) coupled with the high test pressure requirement (1000 psig minimum), prevents the tap in the flange from being temporarily plugged. The opening in the flange is only 3/16 of an inch in diameter and is smooth walled making a high pressure temporary seal very difficult. Failure of this seal could possibly cause ejection of the device used for plugging into the vessel.

A pneumatic test performed with the head installed is precluded due to the configuration of the top head. 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 spaced 150 apart. The retainer clips are contained in a recessed cavity in the top head (see Figure 14R-05.3). Ifa pressure test was performed with the head on, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation.

This test pressure would result in a net inward force on the O-ring that would tend to push it into the recessed cavity that houses the retainer clips.' The O-ring material is only .050" thick with a silver plating thickness of .004" to .006" and could very likely be damaged by this deformation into the recessed areas on the top head.

In addition to the problems associated with the O-ring design that preclude this testing it is also questionable whether a pneumatic test is appropriate for this line. Although the line will initially contain steam if the inner 0-ring leaks, the system actually detects leakage rate by measuring the level of condensate in a collection chamber. This would make the system medium water at the level switch. Finally, the use of a pneumatic test performed at a minimum of 1000 psig would represent an unnecessary risk in safety for the inspectors and test engineers in the unlikely event of a test failure, due to the large amount of stored energy contained in air pressurized to 1000 psig.

System leakage testing of this line is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. It is extremely impractical to purposely fail the inner O-ring in order to perform a test.

Based on the above, DNPS requests relief from the ASME Section XI requirements for system leakage testing of the Reactor Pressure Vessel Head Flange Seal Leak Detection System.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval kELIEF REQUEST NUMBER: 14R-05 (Page 3 of 6)

PROPOSED ALTERNATE EXAMINATION A VT-2 visual examinationwill be performed on the line during vessel flood-up in a refueling outage. The static head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the line. This examination will be performed with the frequency specified by Table IWC-2500-1 for a System Leakage Test (once each inspection period).

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: i4R-05 (Page 4 of 6)

FIGURE 14R-05.1 HEAD FLANGE SEAL LEAK DETECTION SCHEMATIC 14R-05 2 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 5 of 6)

FIGURE 14R-05.2 FLANGE SEAL LEAK DETECTION LINE DETAIL QHigh Pressure Detection Leak Monitoring Tap Inner Flange Seal Ring See Detail "A" Detail "A" Vessel Flange Sectional View Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 6 of 6)

FIGURE 14R-05.3 VESSEL TOP HEAD O-RING RETAINER DETAIL A

Flange Leak-Off line SECTION A-A Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: I4R-06 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: 3

Reference:

Table IWD-2500-1 Examination Category: D-B Item Number: D2. 10, D2.20, D2.30, D2.40, D2.50, D2.60, D2.70, D2.80

==

Description:==

Testing Frequency for Isolation Condenser Shell Side and Vent Piping Component Number:

Unit No. Drawing Test Block No.

Unit 2 M-28, M-39 21C0 1, 21C02 Unit 3 M-359, M-369 31C01, 31C02 CODE REQUIREMENT IWD-5222(b) states that in the case of atmospheric storage tanks, the nominal hydrostatic head, developed with the tank filled to its design capacity shall be acceptable as the system test pressure.

Table IWD-2500-1 requires a system leakage test once each period and a system hydrostatic test once each inspection interval (DNPS utilizes Code Case N-498-1 to perform a system leakage test in lieu of the hydrostatic test).

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

The Isolation Condenser does not have a design level. Instead, it has a design pressure. The design pressure of the shell side of the Isolation Condenser is 25 psig. It is impossibleto develop this pressure in the Isolation Condenser shell side because the condenser is vented to the atmosphere through a nonisolable line and the condenser is only 12 ft in height. A hydrostatic head of 25 psig would correspond to approximately 58 ft of water.

Although the shell side of the condenser is designed for 25 psig, the system is normally operated with the Isolation Condenser water at a level greater than or equal to 6 ft. It would be an abnormal activity and would be impractical to fill the Isolation Condenser to the top simply to achieve a few more pounds of static head pressure. This water would have to be drained and processed by radwaste.

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BASIS FOR RELIEF (Continued)

The 3.0 psig difference in test pressure (with the Isolation Condenser filled to the top vs. the minimum operating level) is so slight that a test with the Isolation Condenser filled to the normal operating level is essentially the same as a test with the Isolation Condenser filled to the top.

Once every five years, DNPS performs a normal operational test of the isolation condenser in accordance with the plant Technical Requirements Manual. This test adequately challenges the isolation condenser and associated piping to allow inspectors to conduct a VT-2 examination that meets the System Leakage Test requirements ofIWD-2500-1 and Code Case N-498-1. However, since this test is only performed once every five years, the requirementto conduct a VT-2 examination once per period cannot be met.

To complete the interval inspection requirements of the isolation condenser and associated piping, normal static conditions (filled greater than or equal to 6 ft per the discussion above) exist under which a VT-2 visual examination can be performed.

PROPOSED ALTERNATE EXAMINATION The Isolation Condenser and associated piping in Test Blocks 2(3)ICO1 and 2(3)IC02 will be VT-2 visually examined during the Technical Requirements Manual 5-year operational test. This test will cover two of the three inservice inspectionperiods. Additionally, Test Block 2(3)IC02 will be VT-2 examined under normal static conditions (filled greater than or equal to 6 ft) during the remaining period in which the normal operational test is not performed. These pressure tests will be performed as alternatives to the periodic and interval testing requirements of IWD-2500-1 and Code Case N-498-1.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: i4R-06 (Page 3 of 3)

FIGURE 14R-06.1 ISOLATION CONDENSER SYSTEM SIMPLIFIED SCHEMATIC To Floor Drains F(3)-1301-10 eS Fill Wat~er Supply Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-07 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H Item Number: C7.10, C7.30, C7.50, C7.70

==

Description:==

Continuous Pressure Monitoring of the Control Rod Drive (CRD)

System Accumulators Component Number:

UnitNo. Drawing Test BlockNo.1 Unit 2 M-34 Sh.2 2RC02 Unit 3 M-365 Sh. 2 3RC02 CODE REOUIREMENT Table IWC-2500-1 requires a Visual VT-2 examination to be performed during a system leakage test.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

As required by DNPS Technical Specifications, the CRD System Accumulator pressure must be greater than or equal to 940 psig to be considered operable. The accumulator pressure is continuously monitored by system instrumentation. Since the accumulators are isolated from the source of make up nitrogen, the continuous monitoring of the CRD accumulators functions as a pressure decay type test. Should accumulator pressure fall below 1000 psig, an alarm is received in the control room. The pressure drop for the associated accumulator is then recorded, and the accumulator is recharged in accordance with DNPS procedures. If an accumulator requires charging more than twice in a thirty day period, then a leak check is performed to determine the cause of the pressure loss. When leakage is detected, corrective actions are taken to repair the leaking component as required by DNPS procedures.

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BASIS FOR RELIEF (Continued)

Since monitoring the nitrogen side of the accumulators is continuous, any leakage from the accumulator would be detected by normal system instrumentation. An additional Visual VT-2 examination performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a Visual VT-2 would require applying a leak detection solution to 177 accumulators per unit resulting in additional radiation exposure without any added benefit in safety. This inspection would thus not be consistent with ALARA practices.

Relief is requested from the Visual VT-2 examination requirements specified in Table IWC-2500-1 for the nitrogen side of the CRD System Accumulators on the basis that DNPS Technical Specification Surveillance requirements exceed the code requirement for a Visual VT-2 Examination.

PROPOSED ALTERNATE PROVISIONS As an alternate to the Visual VT-2 examination requirements of Table IWC-2500-1, DNPS will perform continuous pressure decay monitoring in conjunction with Technical Specifications for the nitrogen side of the CRD Accumulators including attached piping.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspectioninterval of the Inservice Inspection Program for DNPS Units 2 and 3.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08 (Page 1 of 16)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

ASME Section XI, Appendix VIII, Supplement 11 ,"Qualification Requirements For Full Structural Overlaid Wrought Austenitic Piping Welds" Examination Category: B-J Item Number: B9.11

==

Description:==

Pressure Retaining Welds in Piping Subject to Appendix VIII, Supplement 11 (Note: Also Identified in USNRC Generic Letter 88-01 as Category E)

Component Number: Weld Overlay Components Subject to Ultrasonic Examination CODE REQUIREMENT The Code requirements for which relief is requested are all contained within Appendix VIII, Supplement 11. For example, paragraph 1.1 (d)(1), requires that all base metal flaws be cracks.

Paragraph 1.1 (e)(1) requires that at least 20% but less than 40% of the flaws shall be oriented within +20 degrees of the pipe axial direction. Paragraph 1.1 (e)(1) also requires that the rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws. Paragraph 1.1(e)(2)(a)(1) requires that a base grading unit shall include at least 3 in. of the length of the overlaid weld and the outer 25 percent of the overlaid weld and base metal on both sides. Paragraph 1.1 (e)(2)(a)(3) requires that for unflawed base grading units, at least 1 inch of unflawed overlaid weld and base metal shall exist on either side of the base grading unit.

Paragraph 1.1 (e)(2)(b)(1) requires that an overlay grading unit shall include the overlay material and the base metal-to-overlay interface of at least 6 sq. in. The overlay grading unit shall be rectangular, with minimum dimensions of 2 in. Paragraph 3.2(b) requires that all extensions of base metal cracking into the overlay material by at least 0.1 in. be reported as being intrusions into the overlay material.

Specific Code requirements for which relief is requested are identified in the right hand column of the attached table.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

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BASIS FOR RELIEF (Continued)

Paragraph 1.1(d)(1), requires that all base metal flaws be cracks. As illustrated below, implanting a crack requires excavation of the base material on at least one side of the flaw.

While this may be satisfactory for ferritic materials, it does not produce a usable axial flaw in austenitic materials because the sound beam, which normally passes only through base material, must now travel through weld material on at least one side, producing an unrealistic flaw response. To resolve this issue, the PDI program revised this paragraph to allow use of alternative flaw mechanisms under controlled conditions. For example, alternative flaws shall be limited to when implantation of cracks precludes obtaining an effective ultrasonic response, flaws shall be semi-elliptical with a tip width of less than or equal to 0.002 inches, and at least 70% of the flaws in the detection and sizing test shall be cracks and the remainder shall be alternative flaws.

(E, tion Mechanical fatigue crack I cin area Base material Relief is requested to allow closer spacing of flaws provided they don't interfere with detection or discrimination. The existing specimens used to date for qualification to the Tri party (USNRC/BWROG/EPRI) agreement have a flaw population density greater than allowed by the current Code requirements. These samples have been used successfully for all previous qualifications under the Tri-party agreement program. To facilitate their use and provide continuity from the Tri-party agreement program to Supplement 11, the PDI Program has merged the Tri-party test specimens into their weld overlay program. For example: the requirement for using IWA-3300 for proximity flaw evaluation in paragraph 1.1(e)(1) was excluded, instead indications will be sized based on their individual merits; paragraph 1.1 (d)(1) includes the statement that intentional overlay fabrication flaws shall not interfere with ultrasonic detection or characterization of the base metal flaws; paragraph 1.1(e)(2)(a)(1) was modified to require that a base metal grading unit include at least 1 in. of the length of the overlaid weld, rather than 3 inches; paragraph 1.1(e)(2)(a)(3) was modified to require sufficient unflawed overlaid weld and base metal to exist on all sides of the grading unit to preclude interfering reflections from adjacent flaws, rather than the 1 in. requirement of Supplement 11; paragraph 1.1 (e)(2)(b)(1) was modified to define an overlay fabrication grading unit as including the overlay material and the base metal-to-overlay interface for a length of at least 1 in., rather than the 6 sq. in. requirement of Supplement 11; and paragraph 1.1 (e)(2)(b)(2) states that overlay fabrication grading units designed to be unflawed shall be separated by unflawed overlay material and unflawed base metal-to-overlay interface for at least I in. at both ends, rather than around its entire perimeter.

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BASIS FOR RELIEF (Continued)

Additionally, the requirement for axially oriented overlay fabrication flaws in paragraph 1.1(e)(1) was excluded from the PDI Program as an improbable scenario. Weld overlays are typically applied using automated gas tungsten arc welding techniques with the filler metal being applied in a circumferential direction. Because resultant fabrication induced discontinuities would also be expected to have major dimensions oriented in the circumferential direction axial overlay fabrication flaws are unrealistic.

I The PDI Program revised paragraph 2.0 to permit the overlay fabrication flaw test and the base metal flaw tests be performed separately.

The requirement in paragraph 3.2(b) for reporting all extensions of cracking into the overlay is omitted from the PDI Program because it is redundant to the RMS calculations performed in paragraph 3.2(c) and its presence adds confusion and ambiguity to depth sizing as required by paragraph 3.2(c). This also makes the weld overlay program consistent with the Supplement 2 depth sizing criteria.

There are, however, some additional changes that were inadvertently omitted from the Code Case. The most important change is paragraph 1.1(e)(2)(a)(1) where the phrase "and base metal on both sides," was inadvertently included in the description of a base metal grading unit. The PDI program intentionally excludes this requirement because some of the qualification samples include flaws on both sides of the weld. To avoid confusion several instances of the term "cracks" or "cracking" were changed to the term "flaws" because of the use of alternative flaw mechanisms. Additionally, to avoid confusion, the overlay thickness tolerance contained in paragraph 1.1 (b) last sentence, was reworded and the phrase "and the remainder shall be alternative flaws" was added to the next to last sentence in paragraph 1.1 (d)(1). These changes are identified by bold print in the third column of the attached table.

PDI has submitted these changes as a Code Case and they have been approved, but the Code Case will not be published until later in 2002. A detailed comparison matrix (Table 14R-08.1) between Supplement 11, the proposed ASME Section XI Code Case N-654 (provided for information only), and the PDI Program provides supporting documentation. The first column identifies the current requirements in the 95 Edition and 96 Addenda of Supplement 11, while the second (middle) column identifies the changes made by the Code Case.

PROPOSED ALTERNATE EXAMINATION In lieu of the requirements of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VIII, Supplement 11, DNPS will use the PDI Program.

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APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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IST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 5 of 16)"

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bc00- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only) 1.0 SPECIMEN REQUIREMENTS No Change No Change Qualification test specimens shall meet the requirements listed herein, unless a set of specimens is designed to accommodate specific limitations stated in the scope of the examination procedure (e.g., pipe size, weld joint configuration, access limitations). The same specimens may be used to demonstrate both detection and sizing qualification.

1.1 General. The specimen set shall No Change No Change conform to the following requirements.

(a) Specimens shall have sufficient volume No Change No Change to minimize spurious reflections that may interfere with the interpretation process.

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IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 6 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(b) The specimen set shall consist of at No Change (b) The specimen set shall consist of at least three specimens having different least three specimens having different nominal pipe diameters and overlay nominal pipe diameters and overlay thicknesses. They shall include the thicknesses. They shall include the minimum and maximum nominal pipe minimum and maximum nominal pipe diameters for which the examination diameters for which the examination procedure is applicable. Pipe diameters procedure is applicable. Pipe diameters within a range of 0.9 to 1.5 times a nominal within a range of 0.9 to 1.5 times a diameter shall be considered equivalent. If nominal diameter shall be considered the procedure is applicable to pipe equivalent. If the procedure is applicable to diameters of 24 in. or larger, the specimen pipe diameters of 24 in. or larger, the set must include at least one specimen 24 specimen set must include at least one in. or larger but need not include the specimen 24 in. or larger but need not maximum diameter. The specimen set must include the maximum diameter. The include at least one specimen with overlay specimen set shall include specimens thickness within -0.1 in. to +0.25 in. of the with overlay thickness within +0.1 in. of maximum nominal overlay thickness for the minimum nominal overlay thickness which the procedure is applicable, and within -0.25 in. of the maximum nominal overlay thickness for which the procedure is applicable.

(c) The surface condition of at least two No Change No Change _

specimens shall approximate the roughest surface condition for which the examination procedure is applicable.

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IST Program Plan Dresden Nuclear Power Station Units 2 & 3,Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 7 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(d) Flaw Conditions (1) Base metal flaws. All flaws must be in (1) Base metal flaws. All flaws must be in (1) Base metal flaws. All flaws must be or near the butt weld heat-affected zone, or near the butt weld heat-affected zone, cracks in or near the butt weld heat-affected open to the inside surface, and extending at open to the inside surface, and extending at zone, opený to the inside surface, and least 75% through the base metal wall. least 75% through the base metal wall.

extending at least 75% through the base Intentional overlay fabrication flaws shall Intentional overlay fabrication flaws shall metal wall. Flaws may extend 100% not interfere with ultrasonic detection or not interfere with ultrasonic detection or through the base metal and into the overlay characterization of the cracking. Specimens characterization of the base metal flaws.

material; in this case, intentional overlay containing IGSCC shall be used when Specimens containing IGSCC shall be used fabrication flaws shall not interfere with available. At least 70 percent of the flaws when available. At least 70 percent of the ultrasonic detection or characterization of in the detection and sizing tests shall be flaws in the detection and sizing tests shall the cracking. Specimens containing IGSCC cracks. Alternative flaw mechanisms, if be cracks and the remainder shall be shall be used when available, used, shall provide crack-like reflective alternative flaws. Alternative flaw characteristics and shall be limited by the mechanisms, if used, shall provide crack following: like reflective characteristics and shall b-e limited by the following:

(a) Flaws shall be limited to when (a) Flaws shall be limited to when implantation of cracks precludes obtaining implantation of cracks precludes obtaining a realistic ultrasonic response. an effective ultrasonic response.

(b) Flaws shall be semi-elliptical with a tip (b) Flaws shall be semi-elliptical with a tip width of less than or equal to 0.002 inches. width of less than or equal to 0.002 inches.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 8 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bc00- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(2) Overlay fabrication flaws. At least 40% No Change No Change of the flaws shall be non-crack fabrication flaws (e.g., sidewall lack of fusion or laminar lack of bond) in the overlay or the pipe-to-overlay interface. At least 20% of the flaws shall be cracks. The balance of the flaws shall be of either type.

(e) Detection Specimens (1) At least 20% but less than 40% of the (1) At least 20% but less than 40% of the (1) At least 20% but less than 40% of the base metal flaws shall be oriented within base metal flaws shall be oriented within flaws shall be oriented within +20 deg. of +20 deg. of the pipe axial direction. The +20 deg. of the pipe axial direction. The the pipe axial direction. The remainder remainder shall be oriented remainder shall be oriented shall be oriented circumferentially. Flaws circumferentially. Flaws shall not be open circumferentially. Flaws shall not be open shall not be open to any surface to which to any surface to which the candidate has to any surface to which the candidate has the candidate has physical or visual access, physical or visual access. physical or visual access.

The rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws.

(2) Specimens shall be divided into base (2) Specimens shall be divided into base (2) Specimens shall be divided into base and over-lay grading units. Each specimen metal and overlay fabrication grading units. metal and overlay fabrication grading units.

shall contain one or both types of grading Each specimen shall contain one or both Each specimen-shall contain one or both units. types of grading units. Flaws shall not types of grading units. Flaws shall not interfere with ultrasonic detection or interfere with ultrasonic detection or characterization of other flaws. characterization of other flaws.

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IS1 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 9 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(a)(1) A base grading unit shall include at (a)(1) A base metal grading unit shall (a)(1) A base metal grading unit shall least 3 in. of the length of the overlaid include at least 1 in. of the length of the include at least I in. of the length of the weld. The base grading unit includes the overlaid weld. The base metal grading unit overlaid weld. The base metal grading unit outer 25% of the overlaid weld and base includes the outer 25% of the overlaid weld includes the outer 25% of the overlaid weld.

metal on both sides. The base grading unit and base metal on both sides. The base The base metal grading unit shall not shall not include the inner 75% of the metal grading unit shall not include the include the inner 75% of the overlaid overlaid weld and base metal overlay inner 75% of the overlaid weld and base weld and base metal overlay material, or material, or base metal-to-overlay interface, metal overlay material, or base metal-to- base metal-to-overlayinterface.

overlay interface.

(a)(2) When base metal cracking penetrates (a)(2) When base metal cracking penetrates (a)(2) When base metal flaws penetrate into the overlay material, the base grading into the overlay material, the base metal into the overlay material, the base metal unit shall include the overlay metal within grading unit shall not be used as part of any grading unit shall not be used as part of any 1 in. of the crack location. This portion of overlay fabrication grading unit. overlay fabrication grading unit.

the overlay material shall not be used as part of any overlay grading unit.

(a)(3) When a base grading unit is designed (a)(3) Sufficient unflawed overlaid weld (a)(3) Sufficient unflawed overlaid weld to be unflawed, at least 1 in. of unflawed and base metal shall exist on all sides of the and base metal shall exist on all sides of the overlaid weld and base metal shall exist on grading unit to preclude interfering grading unit to preclude interfering either side of the base grading unit. The reflections from adjacent flaws. reflections from adjacent flaws.

segment of weld length used in one base grading unit shall not be used in another base grading unit. Base grading units need not be uniformly spaced around the specimen.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 10 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bc00- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(b)(1) An overlay grading unit shall include (b)(l) An overlay fabrication grading unit (b)(1) An overlay fabrication grading unit the overlay material and the base metal-to- shall include the overlay material and the shall include the overlay material and the overlay interface of at least 6 sq. in. The base metal-to-overlay interface for a length base metal-to-overlay interface for a length overlay grading unit shall be rectangular, of at least 1 in. of at least 1 in.

with minimum dimensions of 2 in.

(b)(2) An overlay grading unit designed to (b)(2) Overlay fabrication grading units (b)(2) Overlay fabrication grading units be unflawed shall be surrounded by designed to be unflawed shall be separated designed to be unflawed shall be separated unflawed overlay material and unflawed by unflawed overlay material and unflawed by unflawed overlay material and unflawed base metal-to-overlay interface for at least base metal-to-overlay interface for at least base metal-to-overlay interface for at least 1 in. around its entire perimeter. The 1 in. at both ends. Sufficient unflawed 1 in. at both ends. Sufficient unflawed specific area used in one overlay grading overlaid weld and base metal shall exist on overlaid weld and base metal shall exist on unit shall not be used in another overlay both sides of the overlay fabrication both sides of the overlay fabrication grading unit. Overlay grading units need grading unit to preclude interfering grading unit to preclude interfering not be spaced uniformly about the reflections from adjacent flaws. The reflections from adjacent flaws. The specimen. specific area used in one overlay specific area used in one overlay fabrication grading unit shall not be used in fabrication grading unit shall not be used in another overlay fabrication grading unit. another overlay fabrication grading unit.

Overlay fabrication grading units need not Overlay fabrication grading units need not be spaced uniformly about the specimen. be spaced uniformly about the specimen.

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SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(b)(3) Detection sets shall be selected from (b)(3) Detection sets shall be selected from (b)(3) Detection sets shall be selected from Table VIII-S2-1. The minimum detection Table VIII-S2-1. The minimum detection Table VIII-S2-1. The minimum detection sample set is five flawed base grading sample set is five flawed base metal sample set is five flawed base metal units, ten unflawed base grading units, five grading units, ten unflawed base metal grading units, ten unflawed base metal flawed overlay grading units, and ten grading units, five flawed overlay grading units, five flawed overlay unflawed overlay grading units. For each fabrication grading units, and ten unflawed fabrication grading units, and ten unflawed type of grading unit, the set shall contain at overlay fabrication grading units. For each overlay fabrication grading units. For each least twice as many unflawed as flawed type of grading unit, the set shall contain at type of grading unit, the set shall contain at grading units. least twice as many unflawed as flawed least twice as many unflawed as flawed grading units. For initial procedure grading units. For initial procedure qualification, detection sets shall include qualification, detection sets shall include the equivalent of three personnel the equivalent of three personnel qualification sets. To qualify new values of qualification sets. To qualify new values of essential variables, at least one personnel essential variables, at least one personnel qualification set is required. qualification set is required.

(f) Sizing Specimen (1) The minimum number of flaws shall be (1) The minimum number of flaws shall be (1) The minimum number of flaws shall be ten. At least 30% of the flaws shall be ten. At least 30% of the flaws shall be ten. At least 30% of the flaws shall be overlay fabrication flaws. At least 40% of overlay fabrication flaws. At least 40% of overlay fabrication flaws. At least 40% of the flaws shall be cracks open to the inside the flaws shall be open to the inside the flaws shall be cracks open to the inside surface. For initial procedure qualification, surface. For initial procedure qualification, surface. sizing sets shall include the equivalent of sizing sets shall include the equivalent of three personnel qualification sets. To three personnel qualification sets. To qualify new values of essential variables, at qualify new values of essential variables, at least one personnel qualification set is least one personnel qualification set is required. required.

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1ST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 12 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(2) At least 20% but less than 40% of the No Change No Change flaws shall be oriented axially. The remainder shall be oriented circumferentially. Flaws shall not be open to any surface to which the candidate has physical or visual access.

(3) Base metal cracking used for length No Change (3) Base metal flaws used for length sizing sizing demonstrations shall be oriented demonstrations shall be oriented circumferentially. circumferentially.

(4) Depth sizing specimen sets shall No Change (4) Depth sizing specimen sets shall include at least two distinct locations where include at least two distinct locations where cracking in the base metal extends into the flaws in the base metal extend into the overlay material by at least 0.1 in. in the overlay material by at least 0.1 in. in the through-wall direction. through-wall direction.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 13 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bc00- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only) 2.0 CONDUCT OF The specimen inside surface and The specimen inside surface and PERFORMANCE DEMONSTRATION identification shall be concealed from the identification shall be concealed from the The specimen inside surface and candidate. All examinations shall be candidate. All examinations shall be identification shall be concealed from the completed prior to grading the results and completed prior to grading the results and candidate. All examinations shall be presenting the results to the candidate. presenting the results to the candidate.

completed prior to grading the results and Divulgence of particular specimen results Divulgence of particular specimen results presenting the results to the candidate, or candidate viewing of unmasked or candidate viewing of unmasked Divulgence of particular specimen results specimens after the performance specimens after the performance or candidate viewing of unmasked demonstration is prohibited. The overlay demonstration is prohibited. The overlay specimens after the performance fabrication flaw test and the base metal fabrication flaw test and the base metal demonstration is prohibited. flaw test may be performed separately. flaw test may be performed separately.

2.1 Detection Test. Flawed and unflawed grading units shall be Flawed and unflawed grading units shall be Flawed and unflawed grading units shall be randomly mixed. Although the boundaries randomly mixed. Although the boundaries randomly mixed. Although the boundaries of specific grading units shall not be of specific grading units shall not be of specific grading units shall not be revealed to the candidate, the candidate revealed to the candidate, the candidate revealed to the candidate, the candidate shall be made aware of the type or types of shall be made aware of the type or types of shall be made aware of the type or types of grading units (base metal or overlay grading units (base metal or overlay grading units (base or overlay) that are fabrication) that are present for each fabrication) that are present for each present for each specimen. specimen. specimen.

2.2 Length Sizing Test No Change No Change (a) The length sizing test may be conducted separately or in conjunction with the detection test.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 14 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only)

(b) When the length sizing test is No Change No Change conducted in conjunction with the detection test and the detected flaws do not satisfy the requirements of 1.1(f), additional specimens shall be provided to the candidate. The regions containing a flaw to be sized shall be identified to the candidate.

The candidate shall determine the length of the flaw in each region.

(c) For a separate length sizing test, the No Change No Change regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the length of the flaw in each region.

(d) For flaws in base grading units, the (d) For flaws in base metal grading units, (d) For flaws in base metal grading units, candidate shall estimate the length of that the candidate shall estimate the length of the candidate shall estimate the length of part of the flaw that is in the outer 25% of that part of the flaw that is in the outer 25 % that part of the flaw that is in the outer 25 %

the base wall thickness, of the base metal wall thickness. of the base metal wall thickness.

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151 Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 15 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only) 2.3 Depth Sizing Test. The candidate shall determine the depth of The candidate shall determine the depth of For the depth sizing test, 80% of the flaws the flaw in each region. the flaw in each region.

shall be sized at a specific location on the surface of the specimen identified to the candidate. For the remaining flaws, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region.

3.0 ACCEPTANCE CRITERIA Examination procedures are qualified for Examination procedures are qualified for 3.1 Detection Acceptance Criteria. detection when all flaws within the scope of detection when all flaws within the scope of Examination procedures, equipment, and the procedure are detected and the results of the procedure are detected and the results of personnel are qualified for detection when the performance demonstration satisfy the the performance demonstration satisfy the the results of the performance acceptance criteria of Table VIII-S2-1 for acceptance criteria of Table VIII-S2-1 for demonstration satisfy the acceptance false calls. Examination equipment and false calls. Examination equipment and criteria of Table VIII-S2-1 for both personnel are qualified for detection when personnel are qualified for detection when detection and false calls. The criteria shall the results of the performance the results of the performance be satisfied separately by the demonstration demonstration satisfy the acceptance demonstration satisfy the acceptance results for base grading units and for criteria of Table VIII-S2-1 for both criteria of Table VIII-S2-1 for both overlay grading units. detection and false calls. The criteria shall detection and false calls. The criteria shall be satisfied separately by the demonstration be satisfied separately by the demonstration results for base metal grading units and for results for base metal grading units and for overlay fabrication grading units, overlay fabrication grading units.

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1ST Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-08.1 (Page 16 of 16)

SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL Extracted from: The Proposed Alternative to STRUCTURAL OVERLAID http://www.boilercode.org/PDF/bcOO- Supplement 11 Requirements WROUGHT AUSTENITIC PIPING 756R.pdf WELDS (Provided for Information Only) 3.2 Sizing Acceptance Criteria. No Change No Change Examination procedures, equipment, and personnel are qualified for sizing when the results of the performance demonstration satisfy the following criteria.

(a) The RIMS error of the flaw length No Change (a) The RMS error of the flaw length measurements, as compared to the true flaw measurements, as compared to the true flaw lengths, is less than or equal to 0.75 inch. lengths, is less than or equal to 0.75 inch.

The length of base metal cracking is The length of base metal flaws is measured measured at the 75% through-base-metal at the 75% through-base-metal position.

position.

(b) All extensions of base metal cracking This requirement is omitted. This requirement is omitted.

into the overlay material by at least 0.1 in.

are reported as being intrusions into the overlay material.

(c) The RMS error of the flaw depth (b) The RMS error of the flaw depth (b) The RMS error of the flaw depth measurements, as compared to the true flaw measurements, as compared to the true flaw measurements, as compared to the true flaw depths, is less than or equal to 0.125 in. depths, is less than or equal to 0.125 in. depths, is less than or equal to 0.125 in.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3

Reference:

IWA-5250(a)(2)

Examination Category: N/A Item Number: N/A

==

Description:==

Alternative Rules for Corrective Measures if Leakage Occurs at Bolted Connections Component Number: All Pressure Retaining Bolted Connections CODE REOUIREMENT IWA-5250(a)(2) states that if leakage occurs at a bolted connection, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3 100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Removal of pressure retaining bolting at mechanical connections for VT-3 visual examination and subsequent evaluation in locations where leakage has been identified is not always the most prudent course of action to determine condition of the bolting and/or the root cause of the leak.

The requirement to remove, examine, and evaluate bolting in this situation does not allow consideration of other factors which may indicate the condition of mechanical joint bolting.

Other factors which should be considered in an evaluation of bolting condition when leakage has been identified at a mechanical joint include, but are not limited to:

"* Bolting material

"* Corrosiveness of process fluid

"* Service age ofjoint bolting materials

"* Leakage location

"* Leakage history at connection

"* Visual evidence of corrosion at connection (connection assembled)

"* Plant / industry studies of similar bolting materials in a similar environment

"* Condition and leakage history of adjacent components Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 2 of 3)

BASIS FOR RELIEF (Continued)

An example at DNPS is the complete replacement of bolting materials (e.g., studs, bolts, nuts, washers, etc.) at mechanical joints during plant outages. In some cases, when the associated system process piping is pressurized during plant start-up, leakage is identified at these joints.

The cause of this leakage is often due to thermal expansion of the piping and bolting materials at the joint and subsequent process fluid seepage at the joint gasket. In most of these cases, proper re-torquing of the joint bolting stops the leakage. Removal of any of the joint bolting to evaluate for corrosion would be unwarranted in this situation. ASME Section XI Code Interpretation XI-1-92-01 has recognized that this situation exists, and has clarified that the requirements of IWA-5250(a)(2) do not apply.

PROPOSED ALTERNATE PROVISIONS DNPS proposes the following alternative, consistent with the methodology of Code Case N-566 2, to the requirements of IWA-5250(a)(2), which will provide an equivalent level of quality and safety when evaluating leakage and bolting material condition at Class 1, 2, and 3 bolted connections.

As an alternative to the to the requirements ofIWA-5250(a)(2), one of the following requirements will be met for leakage at bolted connections.

(a) The leakage will be stopped, and the bolting and component material will be reviewed for joint integrity as described in (c) below.

(b) If the leakage is not stopped, the DNPS will evaluate the structural integrity and consequences of continuing operation, and the effect on the system operability of continued leakage. This engineering evaluation will include the considerations listed in (c) below.

(c) The evaluation of (a) and (b) above is to determine the susceptibility of the bolting to corrosion and failure. This evaluation will include the following:

(1) the number and service age of the bolts; (2) bolt and component material; (3) corrosiveness of process fluid; (4) leak location and system function; (5) leakage history at the connection or other system components; (6) visual evidence of corrosion at the assembled connection.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 3 of 3)

PROPOSED ALTERNATE PROVISIONS (Continued)

If any of the above parameters indicates a need for further examination, corrective action will be taken in accordance with IWA-5250(a)(2).

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-11 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 2 and 3

Reference:

IWC-3122.3 IWC-3132.3 IWC-3600 IWD-3000 Examination Category: N/A Item Number: N/A

Description:

Evaluation Criteria for Temporary Acceptance of Flaws, Component Number: Moderate Energy Class 2 and 3 Piping CODE REQUIREMENTS IWC-3122.3 states that a component whose volumetric or surface examination detects flaws may be acceptable for continued service without a repair/replacement activity if an analytical evaluation is performed in accordance with IWC-3600. Similar requirements for visual examinations are contained in IWC-3132.3.

In the 1995 Edition with the 1996 Addenda of ASME Section XI, IWC-3600, Analytical Evaluation of Flaws, and IWD-3000, Acceptance Standards, are in the course of preparation and state that the requirements of IWB may be used.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives would provide an acceptable level of quality and safety.

ASME Section XI Code Case N-513 is conditionally approved for use in Revision 13 of Regulatory Guide 1.147; however, this Case is not applicable to the 1996 Addenda which is DNPS's code of record for the Fourth Inspection Interval. Code Case N-513-1 has since been issued in Supplement 11 of the 1998 Edition and is currently applicable through the 2001 Edition. This revision of the Code Case is not yet approved for use by the USNRC.

Code Case N-513-1 revises the base case to expand the temporary acceptance methodology from Class 3 moderate energy piping to Class 2 and 3 moderate energy piping. Both cases provide requirements which may be followed for temporary acceptance of flaws in ASME Section III, ANSI B3 1.1, and ANSI B31.7 piping designated as Class 2 or 3. This acceptance is limited to moderate energy piping defined as piping whose maximum operating temperature does not exceed 2007F and whose maximum operating pressure does not exceed 275 psig. The provisions of the case demonstrate the integrity of the item containing the flaw for a limited period of time until appropriate repair/replacement or additional examination activities can be performed.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval RELIEF REQUEST NUMBER: 14R-11 (Page 2 of 2)

PROPOSED ALTERNATE PROVISIONS When using analytical evaluation as the method of acceptance for flaws in moderate energy Class 2 or 3 piping, DNPS will follow the provisions of Code Case N-513-1 without performing a repair/replacement activity. This acceptance will be temporary and will remain in affect for a limited time, not exceeding the time to the next scheduled outage.

DNPS may implement this method or one of the other methods contained in ASME Section MI to accept detected flaws; however, in no case will the temporary evaluation process be applied to (a) components other than pipe or tube, (b) leakage through a gasket, (c) threaded connections with nonstructural seal welds for leakage prevention, or (d) degraded socket welds.

When applying the methods of Code Case N-513-1, the specific safety factors contained in Paragraph 4.0 of the Case will be satisfied. These conditions are consistent with those contained in 10 CFR 50.55a(b)(2)(xiii) regarding the use of Code Case N-513.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for DNPS Units 2 and 3.

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ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval

10.0 REFERENCES

The references used to develop this Inservice Inspection Program Plan include:

1) Code of Federal Regulations, Title 10, Part 50, Paragraph 50.55a, "Codes and Standards"

2) Code of Federal Regulations, Title 10, Part 50, Paragraph 2, "Definitions," the definition of "Reactor Coolant Pressure Boundary"

3) SECY-96-080, Issuance of Final Amendment To 10 CFR 50.55a To Incorporate By Reference The ASME Boiler And Pressure Vessel Code (ASME Code),

Section XI, Division 1, Subsection IWE and IWL

4) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," the 1989 Edition with the No Addenda

5) ASME Boiler and Pressure Vessel Code Section XI, Division 1, Subsection IWE, 1992 Edition with the 1992 Addenda

6) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," the 1995 Edition with the 1996 Addenda

7) ASME Boiler and Pressure Vessel Code Section XI, Division 1, Subsections IWE, 1998 Edition, No Addenda

8) ASME OM Code, 1995 Edition with the 1996 Addenda, Subsection ISTD, "Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants"

9) American Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk Informed Requirements for Class 1, 2, or 3 Piping, Method B"

10) USAS B31.1.0-1967, "Power Piping"

11) Regulatory Guide 1.26, Revision 3, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive Waste- Containing Components of Nuclear Power Plants"

12) Regulatory Guide 1.147, Revision 13, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1

13) BWROG "Alternate BWR Feedwater Nozzle Inspection Requirements," dated October 1995 10-1 Revision 0

ISI Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval

14) USNRC NUREG 0313, Revision 2, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping"

15) Generic Letter 88-01, Revision 2, dated January 25, 1988, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping"

16) Generic Letter 88-01, Supplement 1, dated February 4, 1992, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping"

17) Generic Letter 90-09, "Alternate Requirements for Snubber Visual Inspection Intervals and Corrective Actions"

18) DNPS Units 2 and 3 Updated Final Safety Analysis Report (UFSAR)

19) DNPS Units 2 and 3 Technical Specifications (TS)

20) DNPS Units 2 and 3 Technical Requirements Manual (TRM)

21) DNPS Units 2 and 3 ISI Classification Basis Document, Fourth Ten-Year Inspection Interval

22) DNPS Units 2 and 3 ISI Selection Document, Fourth Ten-Year Inspection Interval

23) EPRI Containment Inspection Program Guide

24) EPRI Topical Report TR-1 12657, Rev. B-A, Final Report, "Revised Risk Informed Inservice Inspection Evaluation Procedure," July 1999

25) USNRC SER related to EPRI Topical Report TR-1 12657, Rev. B, Final Report, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 1999,"

dated October 28, 1999

26) CoinEd Risk-Informed Inservice Inspection Project "Definition of RISI Scope for DNPS Units 2 and 3, Revision 1," dated April 21, 2000

27) CornEd Risk-Informed Inservice Inspection Evaluation (Final Report) for DNPS Units 2 and 3

28) ER-AA-330, "Conduct of Inservice Inspection Activities"

29) ER-AA-330-001, "Section XI Pressure Testing"

30) ER-AA-330-002, "Inservice Inspection of Welds and Components"

31) ER-AA-330-003, "Visual Examination of Section XI Component Supports" 10-2 Revision 0

IS Program Plan Dresden Nuclear Power Station Units 2 & 3, Fourth Interval

32) ER-AA-330-004, "Visual Examination of Technical Specification Snubbers"

33) ER-AA-330-007, "Visual Examination of Section XI Class MC Surfaces and Class CC Liners"

34) ER-AA-330-009, "ASME Section XI Repair/Replacement Program"

35) ER-AA-330-010, "Snubber Functional Testing"

36) DNPS Reactor Coolant Pressure Boundary Normal Makeup Calculation, XCE.040.0201 10-3 Revision 0

ATTACHMENT B ISI Schedule Summary Report

DRESDEN NUCLEAR POWER STATION UNITS 2 & 3 ISI Schedule Summary Fourth Ten-Year Inspection Interval Commercial Service Dates:

Unit 2 - 06/09/72 Unit 3 - 11/16/71 Dresden Nuclear Power Station 6500 North Dresden Road Morris, Illinois 60450 Exelon Generation Company (EGC, LLC) 220 Exelon Way Kennett Square, PA 19348

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval

1.0 INTRODUCTION

The tables contained in this document provide a summary of Class 1, 2, and 3 components and their supports scheduled during the Fourth Inservice Inspection Interval for Dresden Nuclear Power Station (DNPS) Units 2 and 3.

For Class 1, 2, and 3 components and their supports subject to examination per ASME Section XI, Subsections IWB, IWC, IWD, and IWF, the required examinations in each examination category shall be completed during the inspection interval in accordance with Tables IWB(C)(D)-2412-1 and IWF-2410-2, as modified by Code Case N-598.

Examinations required by Subsection IWB that may be' deferred until the end of an inspection interval, as specified in Table IWB-2500- 1, are not required to meet the distribution requirements of Table IWB-2412-1. For risk ranked piping structural elements, the examination requirements of Code Case N-578-1 are provided in Paragraph -2410.

All Subsection IWB, IWC, IWD, IWF, and Code Case N-578-1 examinations, except as noted above for deferrable exams, shall meet the following:

Inspection Inspection Minimum Maximum Interval Period Examinations Examinations Scheduled, % Scheduled, %

4th 1 16 50 2 5 0(l) 75 3 100 100 (1) If the first period completion percentages for any examination category exceeds 34%, at least 16% of the required examinations shall be performed in the second period.

Exceptions to the above distribution percentages and examination requirements of items or welds added to the Inspection Program shall meet the requirements of ASME Section XI, Paragraphs IWB(C)(D)-2412 and IWF-2410. Examination categories with less than three items or welds to be examined shall meet the requirements of ASME Section XI, Paragraphs IWB(C)(D)-2412.

For Class 1, 2, and 3 pressure retaining components, subject to examination per ASME Section XI, Subsections IWB, IWC, and IWD, the required examinations in each examination category shall be completed during the inspection period in accordance with Tables IWB(C)(D)-2500-1, as amended by Code Case N-498-1.

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ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval 2.0

SUMMARY

TABLES The following tables provide a summary of components scheduled during the fourth inspection interval for each examination category. A table is provided for each examination category and inspection frequency required within that category (Inspection Program B, deferrable, etc.). A description of the information contained in each column is provided below.

Title:

The tile of each table lists the examination category and specifies whether the applicable components are scheduled in accordance with Inspection Program B, are deferrable, inspected each inspection period or refuel outage.

Item: The applicable code item listed in Table IWB-2500-1. For risk elements, Examination Category R-A, risk categories are utilized.

1. of Comp: Number of components contained in each applicable code item listed in Table IWB-2500-1. For risk elements, Examination Category R-A, risk categories are utilized.

Total Comp Selected: Number of components selected for examination. When the number of components selected for examination is less than the total number of components (i.e. one of multiple,

% selection), a note is provided to explain the scheduling requirements.

N Selected Per Period: Details the distribution of Total Comp Selected during each inspection period. Note that the number of components is not cumulative.

% of Comp Selected: This column is only utilized when components are selected on a percentage basis. For those items which a percentage is selected a note is provided to explain the scheduling requirements. (Total Comp Selected/ # of Comp)

% Selected Per Period: Details the cumulative percent complete of Total Comp Selected for each inspection period based on the # Selected PerPeriodcolumn. (# Selected PerPeriod/ Total Comp Selected) 2

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category B-A Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 131.11 4 Selected 01 0 0 0 131.12 18 18 00 131.21 1 1 00 18 1

B1.22 ef codp 6 6 0 0 63 Total 29 25 0 0 25 NIA NIA 100%

Notes:

(1) In accordance with Relief Request 14R-10, RPV circumferential welds (B1.11) are permanently deferred for up to 32 effective full power years of operation.

Examination Category B-A Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B1.40 1 1 0 1 M Bl~

Total 2 2 0 1 , 1 0% , 50% 100%

Examination Category B-D Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B3.90 31 31 10 6 15 B3.100 31 31 10 6 15 Total 62 62 20 12 30 3 52% 100%

Notes:

(1) Per Examination Category B-D, Note 2, at least 25% but not more than 50% of the nozzles shall be examined by the end of the first inspection period, and the remainder by the end of the inspection interval.

3

151 Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category B-G-1 Frequency: Deferred Examinations

- r - - -

Item # of Total # Selected Per Period %of Comp 0/ý Poloptati Pgar Parie-A Comp Comp 1 2 3 Selected 1 2 3 Selected B6.10 92 92 0 0 92 B6.20 92 92 0 0 92 B6.30 92 5 0 0 5 B6.40 92 92 0 0 92 B6.50 92 92 0 0 92 B6.180 2 01 0 0 0 B6.190 2 F- 0 0 0 B6.200 2 01 0 0 0 Total 466 373 0 0 373 NIA N/A 100%

Notes:

(1) Two components represent all of the bolts and associated nuts, bushings, and washers on each reactor recirc pump. The examination of these components is limited to only one of the pumps per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 1.

Also, per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 2, examination is required only when a pump is disassembled for maintenance, repair, or volumetric examination. Examination is required only once per inspection interval.

(2) The examination of the Recirc Pump Flange Surfaces is limited to only one of the pumps per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 1. Also, per this item number examination is required only when a pump is disassembled. Examination is required only once per inspection interval.

(3) Per Table IWB-2500-1, reactor vessel closure studs are required to be examined when removed. The five scheduled components represent the population of studs which are removed each refuel outage to allow access to the refuel pool. Additional studs would be inspected if removed. Examination is required only once per inspection interval.

Examination Category B-G-2 Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3

_Selected B7.50 34 34 8 10 16 Total 34 34 8 10 16" 23% 53% 100%

4

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule SEmmaary Examination Category B-G-2 Frequency: Deferred Examinations' Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B7.70 16 _0T 0 0 0 Total 16 0 0 0 0 NIA NIA 0%

Notes:

(1) Per Examination Category B-G-2, Note 2 and Examination Category B-M-2, Note 2, examination is required only when a valve is disassembled for maintenance, repair, or volumetric examination. Also, per Examination Category B-G-2, Note 2 and Examination Category B-M-2, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Examination is required only once per inspection interval within each valve group.

Examination Category B-K Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B10.10 5 20 1 1 B10.20 86 92 2 3 4 11%

B10.30 6 I- 0 1 0 17%

B10.40 2

-z 1 0 0 50%

Total 99 13 3 5 23% 61% 100%

Notes:

(1) Per Examination Category B-K, Note 4, for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. One attachment of each "type" has been selected for the single reactor vessel.

(2) Per Examination Category B-K, Note 5, for piping, pumps, and valves, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF-2510 shall be examined.

Examination Category B-L-2 Frequency: Deferred Examinations 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period I Comp Comp 1 2 3 Selected 1 2 3 2 2 Selected B12.20 _2 S 1 0 0 0 Total 2 0 0 0 0_NIA NIA 0%

Notes:

(1) The examination of pump casings is limited to only one of the pumps performing similar functions in the system per Examination Category B-L-2, Note 1. Also, per Examination Category B-L-2, Note 2, examination is required only when a pump is disassembled for maintenance, repair, or volumetric examination. Examination is required only once per inspection interval.

5

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category B-M-1 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 I Selected B12.40 4 1 0 0 1 Total 4 1 0 0 1 NIA N/A 100%

Notes:

(1) Per Examination Category B-M-I, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Only one examination is required since all B12.40's are within the same group.

Examination is required only once per inspection interval within each valve group. ,

Examination Category B-M-2 Frequency: Deferred Examinations 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B12.50 61 01 0 0 0 Total 61 0 0 0 0 NIA NIA 0%

Notes:

(1) Per Examination Category B-M-2, Note 2, examination is required only when a valve is disassembled for maintenance, repair, or volumetric examination. Also, per Examination Category B-M-2, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Examination is required only once per inspection interval within each valve group.

Examination Category B-N-1 Frequency: Each Inspection Period1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected E313.10 1 1 1 1 1 Total I I I I I N/A I N/A NA Notes:

(1) Per Table IWB-2500-1, Examination Category B-N-I, accessible areas of the vessel interior are to be examined each inspection period.

Examination Category B-N-2 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B13.20 26 26 0 0 26 B13.30 32 32 0 0 32 B13.40 1 1 0 0 1 Total 59 59 0 0 i59 N/A NIA 1100%

6

151 Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category B-0 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 i_ _ Selected B14.10 32 41_ 0 0 4 Total 32 4 0 0 4 NIA NIA 100%

Notes:

(1) Per Table IWB-2500-1, Examination Category B-0, 10% of peripheral CRD Housings (32 listed components represent the peripheral CRD housings) are required to be examined.

Examination Category B-P Frequency: Each Refuel Outage' Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 Selected ___ 3 B15.101 B15.50 B15.60 2 2 4 4 2 B15.70 _

Total 2 2 4 4 2 NIA NIA NIA Notes:

(1) Per Table IWB-2500-1, Examination Category B-P a system leakage test is performed during each refuel outage.

(2) During the first and second inspection periods, each test is performed twice since two refuel outages are scheduled.

Examination Category C-A Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C1.30 4 21 0 1 1 Total 4 2 0 1 1 0% 50% 100%

Notes:

(1) Per Examination Category C-A, Note 3, 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. Two vessel welds are present on each of the 'A' and 'B' LPCI heat exchangers, and thus the two welds associated with one heat exchanger are required to be examined.

7

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category C-B Frequency: Inspection Program B I . . .

Total I I# ..

Item #of # Selected Per Per iod  % of Comp  % Selected Per Period Comp Comp I 2 3 Selected 1 2 3 Selected C2.21 4 41 0 0 4 C2.31 12 _8W 2 4 2 Total 16 12 2 4 6 17% 50% 1 ,100%

Notes:

(1) Per Examination Category C-B, Note 4, 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 A total of four components are associated with a single vessel (Isolation '

Condenser). Therefore, all four components are selected.

(2) Per Examination Category C-B, Note 4, 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. A total of twelve components are distributed as follows: four components associated with the torus/ECCS suction header, and four components distributed across each of the 'A' and 'B' LPCI heat exchangers. All four components from the torus/ECCS ring header and four components from one LPCI heat exchanger ('B') are selected for examination.

Examination Category C-B Frequency: Each Inspection Period1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected 02.33 8 6 6 6 6 Total 8 6 6 6 NIA NIA NIA Notes:

(1) Per Examination Category C-B, Note 5, the components shall be examined during performance of the system pressure test, which is performed each inspection period.

(2) Per Examination Category C-B, Note 4, 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. A total of eight components are distributed as follows: four components associated with the torus/ECCS suction header, and two components distributed across each of the 'A' and 'B' LPCI heat exchangers. All four components from the torus/ECCS ring header and two components from one LPCI heat exchanger ('B') are selected for examination.

8

151 Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category C-C Freauencv: Inspection Proaram B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C3.10 4 0 0 1 C3.20 77 9z 2 4 3 12%

C3.30 7 1 0 0 1 14%

Total 88 11 2 4 5 18% 54% 100%

Notes:

(1) Per Examination Category C-C, Note 4, for multiple vessels of similar design, function, and service, only one attachment of only one of the multiple vessels shall be selected for examination. Two welded attachments are present on each of the 'A' and 'B' LPCI heat exchangers, and one of these welds is required to be examined.

(2) Per Examination Category C-C, Note 5, for piping, pumps, and valves, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF-2510 shall be examined.

Examination Category C-H Frequency: Each Inspection Period 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 C7.30C7.1__0 13 Selected 12 2 12 12 12 j C7.50 C7.70 Total 13 12 12 12 12 NIA NIA N/A Notes:

(1) Per Table IWB-2500-1, Examination Category C-H a system leakage test is performed during each inspection period.

(2) Of the 13 components, one test block (2RC02) does not require a pressure test in accordance with Relief Request 14R-07.

Examination Category D-A Frequency: Inspection Program B Item # of Total # Selected Per Period %of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected D1.10 3 3 0 0 3 D1.20 19 F31_ 1 2 0 16%

Total 22 6 1 2 3 17% 50% 100%

Notes:

(1) Per Examination Category D-A, Note 3, for multiple vessels of similar design, function, and service, the welded attachments of only one of the multiple vessels shall be selected for examination. A total of three components are associated with a single vessel (Isolation Condenser). Therefore, all three components are selected.

(2) Per Examination Category D-A, Note 3, for welded attachments of piping, pumps, and valves, a 10% sample shall be selected for examination.

9

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category D-B Frequency: Each Inspection Period'

-I .. .. T .. . - -- -- --

Item Sof Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected D2. 10 D2.20 D2.30 D2.40 5 5 5 5 5 D2.50 D2.60 D2.70 D2.80 Total 5 5 5 5 5 NIA NIA NIA Notes:

(1) Per Table IWB-2500-1, Examination Category D-B and Code Case N-498-1 a system leakage test is performed during each inspection period.

Examination Category F-A Frequency: Inspection Program B 1 .. - T .. I - --

Item of Total # Selected Per Period %of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected I *t----------t F1.10 177 45' 16 11 18 25%

-. I - Tt -- t F1.20 4 312 + 49W 15 *4*

15 I.

19 I.

16%

F1.30 237 25- 6 6 13 1 no/.

Total 77 142 4I 3 + 6 31% 57% 100%

Notes:

(1) Per Table IWF-2500-01, Examination Category F-A, 25% of Class 1 supports are required to be inspected.

(2) Per Table IWF-2500-01, Examination Category F-A, 15% of Class 2 supports are required to be inspected.

(3) Per Table IWF-2500-01, Examination Category F-A, 10% of Class 3 supports are required to be inspected.

(4) Per Examination Category F-A, Note 3, for multiple components other than piping within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined.

10

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 2 Schedule Summary Examination Category R-A Freauencv: Inspection Proaram B Risk # of Total # Selected Per Period  % of Comp  % Selected Per Period Cat'y Comp1 Comp 1 2 3 Selected 1 2 3 Selected' 1 25(0) 7(0)2 2 2 3 28%

2 108(3) 29(0)2 9 10 10 26%

3 2(55) 0(6)z 0 0 0 28%

4 350(50) 42(3)4 7 15 20 11%

5 41(20) 4(4)- 1 1 2 13%

Total 526(92) 82(13) 19 28 1 35 23% 57% 100%

Notes:

(1) The number of components listed in parentheses represents socket welds. Since socket welds are inspected each refuel outage they are not included in the # Selected Per Periodor

% Selected Per Perioddistribution.

(2) Per EPRI TR-1 12657 and Code Case N-578-1, 25% of the elements in each high risk category (1, 2, or 3) are selected for examination.

(3) Per EPRI TR-112657 and Code Case N-578-1, 10% of the elements in each medium risk category (4 or 5) are selected for examination.

(4)  % of Comp Selected includes socket welds in the calculation (Total Comp Selected / # of Comp).

11

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category B-A Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected 131.11 4 01 0 0 0 131.12 14 14 0 0 14 131.21 1 1 0 0 1 31.22 6 6 0 0 6 Total 25 21 0 0 21 NIA NIA 100%

Notes:

(1) In accordance with Relief Request 14R-10, RPV circumferential welds (B1.11) are permanently deferred for up to 32 effective full power years of operation.

Examination Category B-A Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B1.30 1 1 0 0 1 B1.40 1 1 0 1 0 Total 2 2 0 1 1Do 0% 50% 100%

Examination Category B-D Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B3.90 31 31 8 12 11  !

B3.1 00 31 1 31 8 12 11 Total 62 62 16 24 22 26% 64% 100%

Notes:

(1) Per Examination Category B-D, Note 2, at least 25% but not more than 50% of the nozzles shall be examined by the end of the first inspection period, and the remainder by the end of the inspection interval.

12

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category B-G-1 Frequency: Deferred Examinations r - 1 - - -r Item #of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B6.10 92 92 0 0 92 B6.20 92 92 0 0 92 B6.30 92 5_ 0 0 5 B6.40 92 92 0 0 92 B6.50 92 92 0 0 92 B6.180 2 0" 0 0 0 B6.190 2 0' 0 0 0 B6.200 2 0 0 0 0 Total 466 373 0 0 373 NIA N/A 100%

Notes:

(1) Two components represent all of the bolts and associated nuts, bushings, and washers on each reactor recirc pump. The examination of these components is limited to only one of the pumps per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 1.

Also, per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 2, examination is required only when a pump is disassembled for maintenance, repair, or volumetric examination. Examination is required only once per inspection interval.

(2) The examination of the Recirc Pump Flange Surfaces is limited to only one of the pumps per Examination Category B-G-1, Note 3 and Examination Category B-L-2, Note 1. Also, per this item number examination is required only when a pump is disassembled. Examination is required only once per inspection interval.

(3) Per Table IWB-2500-1, reactor vessel closure studs are required to be examined when removed. The five scheduled components represent the population of studs which are removed each refuel outage to allow access to the refuel pool. Additional studs would be inspected if removed. Examination is required only once per inspection interval.

Examination Category B-G-2 Frequency: Inspection Proqram B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B7.50 37 37 9 12 16 Total 37 37 9 12 1624 57% 100%

13

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category B-G-2 Frequency: Deferred Examinations 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B7.70 14 01 0 0 0 Total 14 0 0 0 0r NIA NIA 0%

Notes:

(1) Per Examination Category B-G-2, Note 2 and Examination Category B-M-2, Note 2, examination is required only when a valve is disassembled for maintenance, repair, or volumetric examination. Also, per Examination Category B-G-2, Note 2 and Examination Category B-M-2, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Examination is required only once per inspection interval within each valve group.

Examination Category B-K Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B1o.10 5 20 1 1 B10.20 66 8; 2 3 3 12%

B10.30 6 F- 0 0 1 17%

B10.40 2 1z 1 0 0 50%

Total 79 12 3 4 5 25% 58% 100%

Notes:

(1) Per Examination Category B-K, Note 4, for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. One attachment of each "type" has been selected for the single reactor vessel.

(2) Per Examination Category B-K, Note 5, for piping, pumps, and valves, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF-2510 shall be examined.

Examination Category B-L-2 Frequency: Deferred Examinations 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B12.20 2 01 0 0 0 Total 2 0 0 0 0 NIA NIA 0%

Notes:

(1) The examination of pump casings is limited to only one of the pumps performing similar functions in the system per Examination Category B-L-2, Note 1. Also, per Examination Category B-L-2, Note 2, examination is required only when a pump is disassembled for maintenance, repair, or volumetric examination. Examination is required only once per inspection interval.

14

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category B-M-1 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B12.40 4 11 0 0 I Total 4 1 0 0 1 NIA N/A 100%

Notes:

(1) Per Examination Category B-M-1, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Only one examination is required since all B12.40's are within the same group.

Examination is required only once per inspection interval within each valve group.

Examination Category B-M-2 Frequency: Deferred Examinations' Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B12.50 60 0 0 0 0 Total 60 0 0 0 0 NIA NIA 0%

Notes:

(1) Per Examination Category B-M-2, Note 2, examination is required only when a valve is disassembled for maintenance, repair, or volumetric examination. Also, per Examination Category B-M-2, Note 3, examinations are limited to at least one valve within each group of valves that are of the same size, design, manufacturing method, and function. Examination is required only once per inspection interval within each valve group.

Examination Category B-N-1 1

Frequency: Each Inspection Period Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected 1 E313.10 1 1 1 1 1 Total I 1 1 1 1 N/A NNIN 1NIA 1 NIA_

Notes:

(1) Per Table IWB-2500-1, Examination Category B-N-I, accessible areas of the vessel interior are to be examined each inspection period.

Examination Category B-N-2 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B13.20 46 46 0 0 46 B13.30 32 32 0 0 32 B13T40 1 1 0 0 7 Total 79 79 0 0 79*NA N/A 100%

15

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category B-0 Frequency: Deferred Examinations Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B14.10 32 41 0 0 4 Total 32 4 0 0 4 NIA NIA 100%

Notes:

(1) Per Table IWB-2500-1, Examination Category B-0, 10% of peripheral CRD Housings (32 listed components represent the peripheral CRD housings) are required to be examined.

Examination Category B-P Frequency: Each Refuel Outage 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected B15.10 B15.50 2 2 2 42 42 B15.60 B15.70 Total 24 4 NIA NIA NIA Notes:

(1) Per Table IWB-2500-1, Examination Category B-P a system leakage test is performed during each refuel outage.

(2) During the second and third inspection periods, each test is performed twice since two refuel outages are scheduled.

Examination Category C-A Frequency: Inspection Program B Item # of Total F Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C1.30 4 -- 21_ 0 1 1 Total 4 2 0 1 1 0% 50% 100%

Notes:

(1) Per Examination Category C-A, Note 3, 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. Two vessel welds are present on each of the 'A' and '8' LPCI heat exchangers, and thus the two welds associated with one heat exchanger are required to be examined.

16

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category C-B 1 1 Frequency: Inspection Proqram B I -

Item # of Total # Selected Per Period %of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C2.21 4 4-C2.31 Total 12 16 8

12

{r 022 2,

2 2

4 4 6 17% 50% 100%

Notes:

(1) Per Examination Category C-B, Note 4, 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. A total of four components are associated with a single vessel (Isolation ,

Condenser). Therefore, all four components are selected.

(2) Per Examination Category C-B, Note 4, 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. A total of twelve components are distributed as follows: four components associated with the torus/ECCS suction header, and four components distributed across each of the 'A' and 'B' LPCI heat exchangers. All four components from the torus/ECCS ring header and four components from one LPCI heat exchanger ('A') are selected for examination.

Examination Category C-B Frequency: Each Inspection Period 1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C2.33 8 6;1 6 6 6 Total 8 6 6 6 6 NIA NIA NIA Notes:

(1) Per Examination Category C-B, Note 5, the components shall be examined during performance of the system pressure test, which is performed each inspection period.

(2) Per Examination Category C-B, Note 4, 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. A total of eight components are distributed as follows: four components associated with the torus/ECCS suction header, and two components distributed across each of the 'A' and 'B' LPCI heat exchangers. All four components from the torus/ECCS ring header and two components from one LPCI heat exchanger ('A') are selected for examination.

17

1SI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category C-C Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected C3.10 4 0 o 0_1 C3.20 84

• 3 3 3 11%

C3.30 7 1 o110 0 1 14% f Total 95 11 3 3 5 27% 54% 100%

Notes:

(1) Per Examination Category C-C, Note 4, for multiple vessels of similar design, function, and service, only one attachment of only one of the multiple vessels shall be selected for examination. Two welded attachments are present on each of the 'A' and 'B' LPCI heat exchangers, and one of these welds is required to be examined.

(2) Per Examination Category C-C, Note 5, for piping, pumps, and valves, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF-2510 shall be examined.

Examination Category C-H Frequency: Each Inspection Period1 Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp

________Selected Comp 1 2 3 Selected 1 2 3 C7.1 C7.300 13 12 2 12 12 12 C7.50 C7.70 Total 13 12 12 12 12 N/A NIA NIA Notes:

(1) Per Table IWB-2500-1, Examination Category C-H a system leakage test is performed during each inspection period.

(2) Of the 13 components, one test block (3RC02) does not require a pressure test in accordance with Relief Request 14R-07.

Examination Category D-A Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected D1.10 3 31 0 0 3 D1.20 30 4; 2 2 0 1 13%

Total 7 28% 57% 100%

Notes:

(1) Per Examination Category D-A, Note 3, for multiple vessels of similar design, function, and service, the welded attachments of only one of the multiple vessels shall be selected for examination. A total of three components are associated with a single vessel (Isolation Condenser). Therefore, all three components are selected.

(2) Per Examination Category D-A, Note 3, for welded attachments of piping, pumps, and valves, a 10% sample shall be selected for examination.

18

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category D-B Frequency: Each Inspection Period' Item # ot Total # Selected Per Peiriod  % of Comp F O/ Selected Pr Prind Comp Comp 1 2 3 Selected 1 2 3 Selected D2.10 D2.20 D2.30 D2.40 8 8 8 8 8 D2.50 D2.60 D2.70 D2.80 Total 8 8 8 8 8 NIA NIA NIA Notes:

(1) Per Table IWB-2500-1, Examination Category D-B and Code Case N-498-1 a system leakage test is performed during each inspection period.

Examination Category F-A Frequency: Inspection Program B Item # of Total # Selected Per Period  % of Comp  % Selected Per Period Comp Comp 1 2 3 Selected 1 2 3 Selected F1.10 181 47' 15 14 18 26%

F1.20 296 47z 13 12 22 16%

F1.30 316 647- 12 16 11%

F1.40 49 26_4 7 3 16 Total 842 154 41 41 72 53% 100%

Notes:

(1) Per Table IWF-2500-01, Examination Category F-A, 25% of Class 1 supports are required to be inspected.

(2) Per Table IWF-2500-01, Examination Category F-A, 15% of Class 2 supports are required to be inspected.

(3) Per Table IWF-2500-01, Examination Category F-A, 10% of Class 3 supports are required to be inspected.

(4) Per Examination Category F-A, Note 3, for multiple components other than piping within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined.

19

ISI Program Schedule Dresden Nuclear Power Station Units 2 & 3, Fourth Interval Dresden Unit 3 Schedule Summary Examination Category R-A Frequency: Inspection Program B Risk # of Total # Selected Per Period % of Com p  % Selected Per Period Cat'y Comp, Comp 1 2 3 Selected" 2 3 Selected1 I 1 25(0) 7(0)2 2 2 3 28%

2 70(3) 19(0)W 6 6 7 26%

3 2(65) 0(10), 0 0 0 27%

4 411(49) 46(5)3 12 18 16 11%

5 43(16) 5(2)_ 1 2 ]E2 12%

Total 551(103) 77(17) 21 28 28 27% 64% 100%

Notes:

(1) The number of components listed in parentheses represents socket welds. Since socket welds are inspected each refuel outage they are not included in the # Selected PerPeriodor

% Selected PerPerioddistribution.

(2) Per EPRI TR-112657 and Code Case N-578-1, 25% of the elements in each high risk category (1, 2, or 3) are selected for examination.

(3) Per EPRI TR-112657 and Code Case N-578-1, 10% of the elements in each medium risk category (4 or 5) are selected for examination.

(4)  % of Comp Selected includes socket welds in the calculation (Total Comp Selected / # of Comp).

20