W3F1-2017-0082, Submittal of the Inservice Inspection Plan and Containment Inservice Inspection Plan for the Fourth 10-Year Interval

Submittal of the Inservice Inspection Plan and Containment Inservice Inspection Plan for the Fourth 10-Year Interval

ML17338B265
Person / Time
Site:	Waterford
Issue date:	12/04/2017
From:	Jarrell J Entergy Operations
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
W3F1-2017-0082
Download: ML17338B265 (263)

Text

Entergy Operations, Inc.

17265 River Road Killona, LA 70057-3093 Tel 504-739-6685 jjarrel@entergy.com John P. Jarrell III Manager, Regulatory Assurance Waterford 3 10 CFR 50.55a W3F1-2017-0082 December 4, 2017 U.S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555

Subject:

Submittal of the Inservice Inspection Plan and Containment Inservice Inspection Plan for the Fourth 10-Year Interval Waterford Steam Electric Station, Unit 3 (Waterford 3)

Docket No. 50-382 License No. NPF-38

Dear Sir or Madam:

As required by paragraph IWA-1400(c) of Section XI of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Entergy Operations, Inc. (Entergy) hereby submits the approved Inservice Inspection Plan and Containment Inservice Inspection Plan for the Fourth 10-year interval at Waterford 3.

Entergy is providing these plans for information only. Entergy is not requesting NRC approval of the plans.

The plans contain no new commitments.

If you have any questions or require additional information, please contact the Regulatory Assurance Manager, John P. Jarrell, at (504) 739-6685.

Sincerely, JPJ/MMZ Attachments: 1. Waterford 3 Fourth 10-Year Interval Inservice Inspection Plan

2. Waterford 3 Fourth 10-Year Interval Containment Inservice Inspection Plan

W3F1-2017-0082 Page 2 cc: Mr. Kriss Kennedy, Regional Administrator U.S. NRC, Region IV RidsRgn4MailCenter@nrc.gov U.S. NRC Project Manager for Waterford 3 April.Pulvirenti@nrc.gov U.S. NRC Senior Resident Inspector for Waterford 3 Frances.Ramirez@nrc.gov Chris.Speer@nrc.gov Louisiana Department of Environmental Quality Office of Environmental Compliance Surveillance Division Ji.Wiley@LA.gov

Attachment 1 to W3F1-2017-0082 Waterford 3 Fourth 10-Year Interval Inservice Inspection Plan (196 pages)

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 1 of 91 PROGRAM SECTION FOR ASME SECTION XI, DIVISION 1 WF3 INSERVICE INSPECTION PROGRAM ENTERGY NUCLEAR ENGINEERING PROGRAMS APPLICABLE SITES ALL SITES:

AN1 CNS GGNS IP3 PAL PNP VTY Specific Sites:

AN2 HQN IP2 JAF PLP RBS WF3 Yes Safety Related:

No

Program Section No.: SEPISMO4 Page No.: Page 2 ot 91 REVIEW AND CONCURRENCE SHEET Program Section

Title:

ASME SECTION XI, DIVISION 1, WF3 INSERVICE INSPECTION PROGRAM Prepared By: Date:

Checked By: Date:

ANII: Date: 7 /_

Reviewed By (or NA)

Concurred: Date: tY-ResponsibupeMsor

/

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 3 of 91 REVISION STATUS SHEET REVISION

SUMMARY

REVISION DESCRIPTION 0 Initial Revision Issued For Use. Based on CEP-ISI-104 and an update of the last revision with the following changes made.

1 Updated program to complete replacement of MRP-139 Augmented inspection program with the N-770-1 requirements as allowed by 10 CFR 50.55a.

Updated Figures WTR-1-4100, WTR-1-4101, WTR-1-4200 and WTR-1-4201 to reflect Cold Leg RTD weld Component IDs that require N-722-1 examination. This was previously captured in A600 Program.

2 Updated W3-R&R-006 Augmented Inspection discussion to show a 9 year examination frequency (Ref. CR-WF3-2011-6576. Added MRP-146 Augmented Inspection Program, added W3-ISI-020, W3-ISI-021 and W3-ISI-023 to Table 3.4.1, incorporated SEP-ISI-104-PCN-2, SEP-ISI-104-PCN-3, SEP-ISI-104-PCN-4, inserted reference to 77FR3073 to Section 2.1B and updated Section 2.2 to not refer to Alloy 600 Iddeal database.

3 Updated WTR-1-1300 per SEP-ISI-104 PCN-5.

Updated Figure WTR-1-4205 to reflect addition of welds IDs 22-031X2, 22-033X2, and 22-064.

4 Update Code Case N-532-4 to N-532-5 on page

32. Add Reference to CEP-RR-001 on Page 38.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 4 of 91 REVISION STATUS SHEET REVISION

SUMMARY

REVISION DESCRIPTION 5 Deleted Code Case N-664 from pg. 39 (reference: CR-HQN-2015-01353 CA-008) 6 Revised wording in Code Case N-648-1 to the latest revision of the Code Case on pg. 39.

7 Updated SEP-ISI-104 to incorporate ASME Boiler and Pressure Vessel Code Section XI 2007 Edition, 2008 Addenda for Fourth Ten Year Interval and reformatted for consistency with Entergy Inservice Inspection Programs prepared by Iddeal Concepts, Inc. This is a complete revision of the document therefore revision bars were not used.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 5 of 91 Table of Contents Section Description Page Number NA Cover Page 1 NA Revision and Concurrence Sheet 2 NA Revision Status Sheet 3 NA Table of Contents 5 1.0 ASME Section XI Inservice Inspection Program 8 1.1 Purpose 8 1.2 General 11 1.3 Inspection Intervals 11 1.4 ASME Section XI Code of Record for the Fourth Inservice 12 Inspection Interval 1.5 ASME Section XI Inservice Inspection Program 14 Description 1.6 Records, Reports, and Submittals 16 1.7 Administrative Controls 18 2.0 ASME Code Case Applicability 19 2.1 Adoption of Code Cases 19 2.2 Regulatory Guide 1.147, Revision 17 Approved Code 21 Cases 2.3 Code Cases Adopted Via NRC Approved Requests 23 2.4 Code Cases Required by 10 CFR 50.55a 23 3.0 Relief Requests 24 3.1 Request for Alternatives 24

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 6 of 91 Table of Contents Section Description Page Number 3.2 Relief Request Required due to Impracticality or Limited 24 Examinations 3.3 Requests to use Later Edition and Addenda of ASME 25 Section XI 4.0 Developmental References 27 5.0 Augmented Examination Programs 27 5.1 Augmented Examinations for External Commitments 28 5.2 Owner Elected Examinations for Internal Commitments 33 5.3 License Renewal Examinations for Aging Management 33 Commitments 6.0 ASME Systems & Examination Boundaries 33 6.1 ASME Class 1, 2 and 3 33 6.2 Calibration Standards 35 7.0 Application Criteria and Code Compliance 36 7.1 ASME Section XI 36 7.2 Augmented Programs 45 8.0 Inservice Inspection Data Management Software 71 9.0 Examination Evaluation Criteria 73 9.1 Successive Inspections 73 9.2 Additional Examinations 75 10.0 Risk-Informed ISI Program 80 TABLES Table WF3 ISI Component Scheduling 11 1.3-1 Table Code Cases Adopted from Regulatory Guide 1.147 21 2.2-1 Table Code Cases Adopted Via NRC Approved Requests 23 2.3-1

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 7 of 91 Table of Contents Section Description Page Number Table Code Cases Required by 10 CFR 50.55a 23 2.4-1 Table Waterford Unit No.3 Fourth Interval Relief Requests 26 3.0-1 Table 7.1 Waterford Unit No.3 Code Category Summary 47 FIGURES Figure 1 WF3 Program Plan 10 Figure 2 WF3 4th ISI Interval Schedule 12 APPENDIX Appendix Waterford Unit No.3 ASME Code Class Systems 86 A ISI Boundary Diagrams Appendix Waterford Unit No.3 Class 1, Class 2 and Class 3 ISI 88 B Isometrics and Equipment Drawings

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 8 of 91 1.0 ASME Section XI Inservice Inspection Program 1.1 Purpose Implementation of an Inservice Inspection Program in accordance with the requirements of ASME Section XI, Division 1, is mandated by the United States Code of Federal Regulations, Title 10, Part 50, Section 55a (10 CFR 50.55a).

This Program Section contains the details of the ASME Section XI, Division 1, Inservice Inspection (ISI) Program for the Waterford 3 Steam Electric Station (WF3).

This Inservice Inspection Program is for the fourth 120-month Inservice Inspection Interval. The coordination of refueling outages and periods within the current Inservice Inspection interval for WF3 is shown in Figure 2 WF3 4th ISI Interval Schedule.

The scope of this Program Section includes the inspection of Class 1, 2, and 3 pressure retaining components and their supports. This Program Section also includes augmented inservice examinations that require NDE inspections and are maintained and implemented by the ISI Program. Other augmented examinations are maintained and implemented in independent programs due to their complexity, subject matter, or because the examinations are not related to the ISI Program.

Changes to the contents of this Program Section shall be performed in accordance with Nuclear Management Manual Procedure EN-DC-174.

This inservice inspection plan is controlled by 10 CFR 50.55a. The original and any changes that meet 10 CFR 50.55a do not require a Process Applicability Determination per EN-LI-100.

This plans objective is to provide a traceable link between the governing code requirements and the implementing procedures in order to ensure that Regulatory rules and ASME Section XI Code requirements for the inservice inspection of safety related systems, components and structures are being fulfilled.

These instructions provide the necessary guidance for the Corporate & Station Engineering Programs personnel at Entergys Waterford 3 Steam Electric Station (WF3) to ensure the following:

1) Conformance to Title 10, Section 50.55a of the Code of Federal Regulations (10 CFR 50.55a).

2) Conformance to the 2007 Edition 2008 Addenda of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code.

3) Conformance to Entergy Corporation and Station policies, practices and procedures.

4) The necessary technical content included in WF3 Inservice Inspection Program and implementing procedures.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 9 of 91

5) The proper ASME Section XI Code required examinations, tests and administrative procedures are implemented.

6) The proper ASME Section XI Code request for alternatives and relief requests are submitted to and approved by the regulatory authority.

7) Component repair and replacement activities are performed in accordance with ASME Section XI Code requirements.

8) The proper examination, test and repair and replacement records and reports are maintained and submitted.

Figure 1 WF3 Program Plan illustrates how this program plan effectively functions as a central source to help ensure all regulatory rules and Entergy requirements are incorporated into the Inservice Inspection Program. This program plan provides a useful aid in program self-assessments, procedure preparation and/or revision, management quick reference, and program familiarization.

Program Section No.: SEP-ISI-104 Revision No.: 5 Page No.: Page 10 of 91 Figure 1 WF3 Program Plan Examination ASME Code Method Section XI 2007 Edition/2008 Addenda Visual Code of Inspection Federal Program ANSI Regulations Specifications Inservice Title 10, Part Inspection Surveillance 50.55a Procedures Implement Nuclear Plant Technical Plan Inspections Regulatory Specifications Commission 6.5.7 & 6.5.9 Testing Specifications Regulatory Final Safety Guides Analysis Report 1.14 5.2.4 and 6.6 Administrative 1.26 Procedures 1.84 1.147 Relief Requests 1.193 1.200 Limited Examinations Authorized Nuclear Compliance Records & Reports Completed Inservice Inspector

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 11 of 91 1.2 General The Waterford 3 Steam Electric Station nuclear steam supply system is a Combustion Engineering two loop Pressurized Water Reactor design. Entergy obtained a Construction Permit to build WF3 on November 14, 1974. After satisfactory plant construction and pre-operational testing was completed, a full power operating license, NPF-38, was granted on March 16, 1985. WF3 subsequently commenced commercial operation on September 24, 1985. The NRC Docket Number assigned to WF3 is 050-00382.

1.3 Inspection Intervals The initial 120-month ISI Program commenced with commercial operation on September 24, 1985. However, by authorization from the Director of the Office of Nuclear Regulation (Reference ILN 94-0177 and GNRI 96-00244) the first 120-month interval was extended until June 30, 1997. The second 120-month interval began on July 1, 1997 and continued through June 30, 2007, but was extended 11 months to May 31, 2008, as allowed by ASME Section XI, IWA-2430(c)(1). The third 120-month interval began on May 31, 2008 and continued to November 30, 2017 thus reclaiming 6 months of the 11 month extension that occurred in the 2nd interval.

The Fourth Inspection Interval is divided into inspection periods consistent with ASME Section XI Table IWB-2411-1 (Reference Table 1.3-1). The inspection periods are scheduled as follows:

1st Period: From December 1, 2017 through November 30, 2020 (3 Years) 2nd Period: From December 1, 2020 through November 30, 2024 (4 Years) 3rd Period: From December 1, 2024 through November 30, 2027 (3 Years)

In accordance with ASME Section XI, IWA-2430(c)(3), that portion of an inspection interval described as an inspection period may be reduced or extended by as much as 1 year. This adjustment shall not alter the requirements for scheduling inspection intervals.

Outages during the 4th inservice inspection interval are scheduled in Figure 2 WF3 4th ISI Interval Schedule:

TABLE 1.3-1 WF3 ISI Component Scheduling INSPECTION INSPECTION PERIOD MINIMUM MAXIMUM INTERVAL CALENDAR YEARS OF EXAMINATIONS EXAMINATIONS PLANT SERVICE WITHIN COMPLETED, % CREDITED, %

THE INTERVAL 4th 3 16 50 4th 7 50 1 75 4th 10 100 100

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 12 of 91 1

If the first period completion percentage for any examination category exceeds 34%, at least 16% of the required examinations shall be performed in the second period.

Figure 2 WF3 4th ISI Interval Schedule 1RF22 1RF23 1RF24 1RF25 1RF26 1RF27 S 2019 F 2020 S 2022 F 2023 S 2025 F 2026 1/2017 1/2018 1/2019 1/2020 1/2021 1/2022 1/2023 1/2024 1/2025 1/2026 1/2027 1/2028 st nd rd 1 Period 2 Period 3 Period 36 Months 48 Months 36 Months 12/01/2017 12/01/2020 12/01/2024 TH 11/30/2027 Start 4 ISI TH End 4 ISI Interval Interval 1.4 ASME Section XI Code of Record for the Fourth Inservice Inspection Interval Entergy is required to update the ASME Section XI, ISI Program for WF3 once every ten years. The regulations in 10 CFR 50.55a(g)(4) establish the effective ASME Code edition and addenda to be used by licensees for performing inservice inspections of components (including supports). Paragraph 50.55a(g)(4)(ii) requires the use of the latest edition and addenda that has been incorporated by 10 CFR 50.55a(a), 12 months before the start of the 120-month inspection interval. This is considered the Code of Record.

CFR82FR32934 was published in July 2017 requiring the implementation of Code Cases N-729 and N-770 to N-729-4 and N-770-2 respectively. This Final Rule became effective August 17, 2017.

Based on 10 CFR 50.55a(a)(1)(ii)(C), the WF3 Fourth Ten-Year Interval ISI Program Plan is required to meet the 2007 Edition of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI with the 2008 Addenda with the associated conditions contained in 10 CFR 50.55a(b)(2) as defined below. 10 CFR 50.55a conditions applicable to the System Pressure Testing, Containment

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 13 of 91 Inservice Inspection and Repair/Replacement Activities are address in their applicable Program Sections.

10 CFR 50.55a(b)(2)(x), WF3 will apply the station 10 CFR 50 Appendix B Quality Assurance Program to ASME Section XI activities.

10 CFR 50.55a(b)(2)(xviii)(A), the Level I and II nondestructive examination personnel must be recertified on a 3-year interval in lieu of the 5 year interval in IWA-2314(a) and IWA-2314(b).

10 CFR 50.55a(b)(2)(xix), the provisions for substituting alternative examination methods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 must be applied. The provisions in IWA-4520(b)(2) and IWA-4521 of the 2008 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, allowing the substitution of ultrasonic examination for radiographic examination specified in the Construction Code, are not approved for use.

10 CFR 50.55a(b)(2)(xxi)(A), the provisions of Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.120 and B3.140 of the 1998 Edition must be applied when using the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section. A visual examination with magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph(a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e.,

a/l=0.5), may be performed instead of an ultrasonic examination.

10 CFR 50.55a(b)(2)(xxii), The use of the provision in IWA-2220, Surface Examination, of Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, that allows use of an ultrasonic examination is prohibited.

10 CFR 50.55a(b)(2)(xxviii), Analysis of flaws, Licensees using ASME BPV Code, Section XI, Appendix A must use the following conditions when implementing Equation (2) in A-4300(b)(1): For R < 0, KI depends on the crack depth (a), and the flow stress f). The flow stress is defined by f = 1 2( ys + ult), where ys is the yield strength and ult is the ultimate tensile strength in units ksi (MPa) and a is in units in. (mm). For -2 R 0 and Kmax - Kmin 0.8 x 1.12 f a), S = 1 and KI = Kmax. For R < -2 and Kmax -

Kmin 0.8 x 1.12 f a), S = 1 and KI = (1 - R) Kmax/3. For R < 0 and Kmax - Kmin > 0.8 x 1.12 f a), S = 1 and KI = Kmax - Kmin.

ASME Code Cases N-729-4, N-722-1 and N-770-2 with the applicable conditions as identified in 10 CFR 50.55a(g)(6)(ii)(D),(E) and (F).

1.5 ASME Section XI Inservice Inspection Program Description The ASME Section XI Inservice Inspection Program for the Fourth Inspection Interval is comprised of the following individual programs:

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 14 of 91 ASME Section XI and Augmented ISI Program for pressure retaining components and their supports ASME Section XI System Pressure Test Program ASME Section XI Containment Inservice Inspection Program ASME Section XI Repair and Replacement Program These programs are separate documents or procedures, each bearing the title of the program. The ISI Programs are administered through EN-DC-174, Engineering Program Sections.

For the purpose of clarification, the following definitions for terms used in the ASME Section XI Inservice Inspection program are:

PROGRAM The term program is used to provide the structure and methods for accomplishment of the object and is to be considered synonymous with the term plan as used in ASME Section XI, 2007 Edition, 2008 Addenda, Article IWA-2400.

SCHEDULE The term schedule is used to define the detailed scheme of performing the required inspections over the ten-year interval.

1.5.1 Ten-Year Inservice Inspection (ISI) Program Description The Ten-Year Inservice Inspection Program details WF3 compliance with ASME Code, Section XI 2007 Edition, 2008 Addenda, Articles IWA, IWB, IWC, IWD, and IWF for examination of Class 1, 2, and 3 pressure retaining items and their supports. This document defines the Class 1, 2, and 3, components and the Code required examinations for each ASME Section XI examination category, and the augmented inspection scope.

The purpose of the Ten-Year ISI Program is to periodically perform nondestructive examination of ASME Class 1, 2, or 3 safety related components and supports in order to identify the presence of service related degradation.

The administrative procedures and Inspection Schedule described in the Ten-Year ISI Program, combined with applicable WF3, Entergy, and approved vendor procedures, constitute the ISI portion of the Ten-Year ISI Program required by FSAR Sections 5.2.4 and 6.6.

For convenience, the Ten-Year ISI Program also contains augmented examinations that are outside the scope of 10 CFR 50.55a. These examinations may result from activities including, but not limited to, commitments made to the NRC outside the scope of 10 CFR 50.55a, internal commitments, License Renewal commitments, or industry initiatives.

These examinations may use ASME Section XI techniques and procedures as defined in the commitment, however, the requirements of ASME Section XI, including reporting, do

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 15 of 91 not apply. These augmented examinations are not required by the ASME Code and are not included in the summary totals for the specific examination category.

The Ten-Year ISI Program schedule is contained in the IDDEAL Software Suite computer database. This information is retained at the plant site and is available for review.

1.5.2 Repair and Replacement Program Description ASME Section XI requirements for Class 1, 2 and 3 and MC pressure retaining components and supports repairs and replacements are not contained in this plan. CEP-RR- 001, ASME Section XI Repair/Replacement Program is the implementing procedure which describes the implementation process for the repair and replacement provisions of ASME Code, Section XI Article IWA-4000.

1.5.3 Snubber Program Description The Snubber Program requirements are not contained in this plan. Snubbers (pin-to-pin portion) will be examined and tested in accordance with plant Snubber Program contained in, SEP-WF3-IST-002. Snubber support hardware such as lugs, bolting, pins and clamps between the snubber pins and component (or the snubber pin and the building structure) will be examined in accordance with Section XI, Subsection IWF, per this Program Section.

1.5.4 Pressure Test Program Description ASME Section XI requirements for Class 1, 2 and 3 pressure testing are not contained in this plan. CEP-PT-001, ASME Section XI Pressure Testing Program and SEP-PT-WF3-001, Inservice Inspection Pressure Testing (PT) Program Section, satisfies the requirements established in ASME, Section XI, Articles IWA-5000, IWB-5000, IWC-5000, and IWD-5000.

1.5.5 Containment Inspection Program Description ASME Section XI requirements for MC components and component supports are not contained in this plan. SEP-CISI-104, Containment Inservice Inspection Program implements the requirements of the ASME Code, Section XI Articles IWE (Class MC components).

1.5.6 Pump and Valve Program Description The Inservice Testing (IST) Program requirements are not contained in this plan. The WF3 Inservice Testing Program implements the requirements of the ASME OM Code.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 16 of 91 1.5.7 Steam Generator Tube Surveillance Program Description In accordance with ASME Section XI 2007 Edition, 2008 Addenda IWB-2500-1 Examination Category B-Q, the Steam Generator Tube Inspection Program is in accordance with Technical Specification 6.5.9 in lieu of ASME Section XI.

1.6 Records, Reports, and Submittals The requirements of Code Case N-532-5 shall be met in lieu of the requirements of IWA-6210(c), (d), (e), and (f), IWA-6220, IWA-6230(b), (c), and (d), and IWA-6240(b). All other requirements of IWA-6000 and the other requirements of Section XI applicable to records and reporting shall be met as described below.

1.6.1 Records Records of examinations, tests, and repair replacement activities shall be prepared in accordance with the governing procedure or program document.

Records and reports required by Section XI and this program are safety related and shall be maintained as a permanent plant record. They shall be filed in a manner to ensure accessibility by the Authorized Nuclear Inservice Inspector (ANII). The storage and maintenance of these records are under the auspices of the Quality Assurance Program.

Records and reports shall be either the original or a reproduced legible copy.

Alternatively, they may be maintained in an electronic format provided the requirements of IWA-6320(a) are met. Radiographs may be microfilmed or digitally reproduced. Digital reproduction shall be in accordance with IWA-6320(b).

Inservice Inspection Records, as a minimum shall include:

Record Index Preservice and inservice inspection plans and schedules Preservice and inservice inspection reports Records and reports of repair/replacement activities Records of flaw acceptance by analytical evaluation Records of regions in ferritic Class 1 components with modified acceptance criteria Nondestructive examination procedures Nondestructive examination records Pressure test procedures Pressure test records

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 17 of 91 1.6.2 Reports and Submittals Reports and submittals comprise those documents or records that are required to be submitted to the NRC either on a recurring basis or as a result of activities described by Section XI. Reports shall be as described below. Reports and submittals are provided to the NRC for Information Only to meet a specific Section XI requirement. NRC approval is not required.

In accordance with IWA-1400(c), the initial ISI Program shall be provided to the NRC.

Program Change Notices (PCN) is not required to be submitted.

The Summary Reports required by ASME Section XI, 2007 Edition, 2008 Addenda IWA-6000 are replaced by the Owners Activity Report described by Code Case N-532-5. Form OAR-1 will be prepared, maintained, and submitted in accordance with the applicable Program Section. The Owners Activity Report shall be prepared to meet the requirements of the Code Case and submitted to the NRC within 90 calendar days of completing each refueling outage. The report shall include all applicable activities of the operating cycle not included in the previous report.

Analytical evaluations that accept items for continued service in accordance with IWB-3132.3, IWB-3142.4, IWC-3122.3, IWC-3132.3, and IWD-3600 shall be submitted to the NRC. If the evaluation is performed during a refueling outage, the submittal may be included with the Owners Activity Report provided the cover letter makes specific reference to the contained analytical evaluation. Evaluations performed during the operating cycle between refueling outages shall be submitted independent of the Owners Activity Report. As a guide, these submittals should be timely and within 30 calendar days of completing the analytical evaluation.

Evaluation procedures used in analytical evaluations of flaws in austenitic and ferritic piping shall be provided to the NRC in accordance with IWB-3640, IWC-3640, and IWD-3640. A onetime submittal is acceptable provided subsequent revisions are provided and the scope and application of the procedure does not change. The cover letter should indicate that the procedure is only submitted once (not for each use) and what criteria will be used to determine when additional submittals will be made.

When acceptance standards have been modified as a result of using smaller flaw sizes in developing the Pressure/Temperature Limit Curves in accordance with IWB-3410.2, the modified acceptance standards shall be filed with the NRC.

1.6.3 Incomplete or Missing Examinations Resolution and tracking of incomplete or missed examinations shall be in accordance with the Program Change Notice (PCN) and Condition Report process as described below and EN-DC-174, Engineering Program Sections:

1.6.3.1 PCN Requirements

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 18 of 91 A PCN to the program shall be developed in a timely manner. The PCN should contain the following:

a) The affected component.

b) The examination requirement that has not been, or will not be, met.

c) The reason for incomplete or missed examination.

d) Alternate examinations performed or recommended, (i.e. best effort vs. VT-1, etc.) to include reasons why this is acceptable.

e) For partial examinations, the extent of coverage the component received or will receive using approved methods.

1.6.3.2 Request for Relief If a Request for Relief is required for an incomplete or missing examination, the following steps shall be completed:

a) The site should develop and submit the necessary documentation to the NRC1.

b) If the Request for Relief is approved by the NRC, the approval should be incorporated into the program plan by a PCN.

c) If the Request for Relief is disapproved, the site should initiate a Condition Report.

1.6.3.3 Missed Examinations Which Cannot Be Made Up If it is discovered that a required examination has been missed and cannot be performed before the end of the inspection period in accordance with the Program Plan, a Condition Report should be initiated.

NOTE:

The period includes any extensions made within the ASME Section XI and 10 CFR 50.55a provisions. The end of the period does not occur until any applicable extensions are completed.

1.7 Administrative Controls This program is written and implemented as a Program Section meeting the requirements of EN-DC-174, Program Sections. This Program Section will be maintained through revisions and Program Change Notices in accordance with EN-DC-174.

1 For information on what the NRC expects to see in Coverage Relief Request refer to the presentation made by Stephen Cumblidge to the Industry/NRC NDE technical Information Exchange Public Meeting on January 13-15, 2015 (ML15013A266)

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 19 of 91 This program will be maintained current to reflect applicable changes to plant configuration through the review process defined by EN-DC-115, Engineering Change Process.

Additions of items to the program resulting from changes to the plant configuration will be incorporated and scheduled for inspection or examination in accordance with IWB/IWC/IWD-2411 and IWF-2410.

Organizational and divisions of responsibilities for the management and implementation of the ISI Program are defined in EN-DC-351 Inservice Inspection Program Duties and Responsibilities and CEP-ISI-100 ASME Section XI, Division 1 Fleet Administrative Controls for Inservice Inspection Program.

Several elements required by Section XI to be contained in the ISI Program are maintained in the Iddeal Concepts Software Suite. This software and its contents are part of the formal ISI program and are controlled in accordance with Program Section CEP-COS-0100, Control and Use of Iddeal Concepts Software.

The initial ISI Program Plan, revisions, and PCNs are subject to review by the Authorized Nuclear Inservice Inspector (ANII).

2.0 ASME Code Case Applicability This section contains ASME Section XI Code Cases applicable to the WF3 Fourth Ten Year Inspection Interval. ASME Section XI Code Cases applicable to System Pressure Testing, Containment Inservice Inspection, Repair/Replacement Activities, and NDE are addressed in their associated Program Sections; CEP-PT-001, SEP-CISI-104, CEP-RR-001, and CEP-NDE-100, respectively.

2.1 Adoption of Code Cases ASME Section XI Code Cases adopted for ISI and related NDE activities for the Fourth Interval are listed in Tables 2.2-1, 2.3-1, and 2.4-1. The use of Code Cases is in accordance with ASME Section XI, IWA-2440, 10 CFR 50.55a, and Regulatory Guide 1.147. As permitted by ASME Section XI and Regulatory Guide 1.147 or 10 CFR 50.55a, ASME Section XI Code Cases may be adopted and used as described below:

2.1.1 Adoption of Code Cases Listed for Generic Use in Regulatory Guide 1.147 Code Cases that are listed for generic use in the latest revision of Regulatory Guide 1.147 may be included in the ISI program provided any additional conditions specified in the Regulatory Guide are also incorporated. Table 2.2-1 identifies the Code Cases approved for generic use and adopted for the fourth interval.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 20 of 91 2.1.2 Adoption of Code Cases Not Approved in Regulatory Guide 1.147 Certain Code Cases that have been approved by the ASME Board of Nuclear Codes and Standards may not have been reviewed and approved by the NRC Staff for generic use and listed in Regulatory Guide 1.147. Use of such Code Cases may be requested in the form of a Request for Alternative in accordance with 10 CFR 50.55a(a)(z). Once approved, these Requests for Alternatives will be available for use until such time that the Code Cases are adopted into Regulatory Guide 1.147, at which time compliance with the provisions contained in the Regulatory Guide is required.

Table 2.3-1 identifies those Code Cases that have been requested through Requests for Alternatives. For convenience to the user of this ISI Program, the appropriate internal correspondence number is provided to assist in their retrieval from Document Control. All other Requests for Alternatives and Relief Requests (those not associated with NRC approval of Code Cases) are addressed in Section 3.0.

2.1.3 Adoption of Code Cases Mandated by 10 CFR 50.55a Code Cases required by rule in 10 CFR 50.55a are incorporated into the ISI Program and implemented at the specified schedule. Code Cases currently required by 10 CFR 50.55a are identified in Table 2.4-1.

2.1.4 Use of Annulled Code Cases As permitted by Regulatory Guide 1.147(B), Code Cases that have been adopted for use in the current inspection interval that are subsequently annulled by ASME may be used for the remainder of the interval.

2.1.5 Code Case Revisions Initial adoption of a Code Case requires use of the latest revision of that Code Case listed in Regulatory Guide 1.147. However, if an adopted Code Case is later revised and approved by the NRC, then either the earlier or later revision may be used as permitted by Regulatory Guide 1.147(B). An exception to this provision would be the inclusion of any conditions on the later revision necessary to enhance safety. In this situation, the conditions imposed on the later revision must be incorporated into the program.

2.1.6 Adoption of Code Cases Issued Subsequent to Filing the Inservice Inspection Plan Code Cases issued by ASME subsequent to filing the Inservice Inspection Plan with the NRC may be incorporated within the provisions of paragraphs 2.1.1 or 2.1.2 by revision to the ISI Plan. Any subsequent Code Cases shall be incorporated into the program and identified in either Table 2.2-1 or 2.3-1, as applicable, prior to their use.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 21 of 91 2.1.7 Non Inservice Inspection Code Cases Only Code Cases applicable to ISI and related nondestructive examination requirements for Class 1, 2, and 3 components and component supports are included in Table 2.2-1, 2.3-1 and 2.4-1. Code Cases applicable to System Pressure Testing, Containment Inservice Inspection, Repair/Replacement Activities, and NDE are addressed in their associated Program Sections; CEP-PT-001, SEP-CISI-104, CEP-RR-001, and CEP-NDE-100, respectively.

2.1.8 Code Cases not approved for use by the NRC Code Cases that have been approved by the ASME Board of Nuclear Codes and Standards may not be approved by the NRC Staff for generic use. These Code Cases are listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use. However, the NRC may approve their use in specific cases. Code Cases listed in Regulatory Guide 1.193 will not be used at WF3 without an approved Request for Alternative in accordance with 10 CFR 50.55a(a)(z).

2.2 Regulatory Guide 1.147, Revision 17 Approved Code Cases Table 2.2 Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Conditions Number N-526 Alternative Requirements for Successive None Inspections of Class 1 and 2 Vessels N-532-5 Repair/Replacement Activity None Documentation Requirements and Inservice Inspection Summary Report Preparation and Submission N-586-1 Alternative Additional Examination None Requirements for Classes 1, 2, and 3 Piping, Components, and Supports N-613-1 Ultrasonic Examination of Penetration None Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Nozzle-to-Vessel Welds, Figs. IWB-2500-7(a), (b), and (c).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 22 of 91 Table 2.2 Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Conditions Number N-639 Alternative Calibration Block Material Chemical ranges of the calibration block may vary from the materials specification if (1) it is within the chemical range of the component specification to be inspected, and (2) the phase and grain shape are maintained in the same ranges produced by the thermal process required by the material specification.

N-648-1 Alternative Requirements for Inner Radius In lieu of a UT Examination of Class 1 Reactor Vessel examination, licensees Nozzles may perform a VT-1 examination in accordance with the code of record for the Inservice Inspection Program utilizing the allowable flaw length criteria of Table IWB-3512-1 with limiting assumptions on the flaw aspect ratio.

N-716-1 Alternative Classification and Examination None Requirements N-735 Successive Inspection of Class 1 and 2 None Piping Welds N-765 Alternatives to Inspection Interval None Scheduling Requirements of IWA-2430 N-773 Alternative Qualification Criteria for Eddy None Current Examinations of Piping Inside Surfaces

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 23 of 91 2.3 Code Cases Adopted Via NRC Approved Requests The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 17 and require a request for alternative prior to implementation. Refer to Section 3.0 of this plan for the applicable requests.

Table 2.3 Code Cases Adopted Via NRC Approved Requests Code Case Request for Title Number Alternative No.

NA NA NA 2.4 Code Cases Required by 10 CFR 50.55a The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 17, but are mandated in 10 CFR 50.55a.

Table 2.4 Code Cases Required by 10 CFR 50.55a Code Case Title Notes Number N-722-1 Additional Examinations for PWR Pressure Conditions specified Retaining Welds in Class 1 Components in paragraphs Fabricated With Alloy 600/82/182 Materials (g)(6)(ii)(E)(2) through (4) of 10 CFR 50.55a.

N-729-4 Alternative Examination Requirements for PWR Conditions specified Reactor Vessel Upper Heads With Nozzles in paragraphs Having Pressure-Retaining Partial-Penetration (g)(6)(ii)(D)(2)

Welds through (4) of 10 CFR 50.55a.

N-770-2 Alternative Examination Requirements and Conditions specified Acceptance Standards for Class 1 PWR Piping in paragraphs and Vessel Nozzle Butt Welds Fabricated With (g)(6)(ii)(F)(2)

UNS N06082 or UNS W86182 Weld Filler through (13) of 10 Material With or Without Application of Listed CFR 50.55a.

Mitigation Activities

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 24 of 91 3.0 Relief Requests Throughout this program, the term relief request is used interchangeably referring to submittals to the NRC requesting permission to deviate from either an ASME Section XI requirement, a 10 CFR 50.55a rule, or to use provisions from Editions or Addenda of Section XI not approved by the NRC as referenced in 10 CFR 50.55a(b). However, when communicating with the NRC and in written requests to deviate, the terms as defined below must be used for clarity and to satisfy 10 CFR 50.55a. Submittals to the NRC must clearly identify which of the below rules are being used to request the deviation.

Table 3.0-1 contains an index of Relief Requests written in accordance with 10 CFR 50.55a(a)(g)(5)(iii) and 10 CFR 50.55a(z). The applicable Entergy submittal and NRC Safety Evaluation Report (SER) correspondence numbers are also included for each request.

Only relief requests associated with the ISI program and the supporting nondestructive examination requirements are listed in Table 3.0-1. Relief Requests applicable to System Pressure Testing, Containment Inservice Inspection, Repair/Replacement Activities, and NDE are addressed in their associated Program Sections; CEP-PT-001 or SEP-PT-WF3-001, SEP-CISI-104, CEP-RR-001, and CEP-NDE-100, respectively.

3.1 Request for Alternatives When seeking an alternative to the rules contained in 10 CFR 50.55a(b) through (h) the request is submitted under the provision of 10 CFR 50.55a(z). Once approved by the Director, Office of Nuclear Reactor Regulation, the alternative may be incorporated into the ISI program. These types of requests are typically used to request use of Code Cases, Code Edition, or Addenda not yet approved by the NRC. Request for Alternatives must be approved by the NRC prior to their implementation or use. Within the provisions of 10 CFR 50.55a(z) there are two specific methods of submittal:

10 CFR 50.55a(z)(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety.

10 CFR 50.55a(z)(2) Hardship without a compensating increase in quality and safety.

Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

3.2 Relief Request Required due to Impracticality or Limited Examinations 10 CFR 50.55(a)(g)(5)(iii) and (iv) allows relief to be requested in instances when a Code requirement is deemed impractical with (iv) being specific to examination requirements that are determined to be impractical. The provisions of these two paragraphs are typically used to address impracticalities like limited examination coverage. Under 10 CFR 50.55(a)(g)(5)(iv), relief requests for impracticality determinations must be submitted

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 25 of 91 for NRC review and approval not later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.

In cases where conformance with an ASME Section XI requirement for inservice inspection is considered impractical, Entergy will notify the NRC and submit information to support the determination, as required by 10 CFR 50.55a(g)(5). The submittal of this information will be referred to as a Request for Relief.

In the event that the entire examination volume or surface (as defined in the ASME Code) cannot be examined due to interference by another component or part geometry, then in accordance with ASME Section XI IWA-2200(c) a request for relief will be submitted.

ASME Section XI IWA-2200(c) does not apply to the bare metal visual examination required by Code Case N-729-4. When performing visual examinations to meet N-729-4, the requirements of N-729-4 shall be used for determining allowance for partial examination of the required surface.

3.3 Requests to use Later Edition and Addenda of ASME Section XI On July 28, 2004, the NRC published Regulatory Issue Summary (RIS) 2004-12, Clarification on Use of Later Editions and Addenda to ASME OM Code and Section XI.

This RIS clarifies the NRC position on using Editions and Addenda of Section XI, in whole or in part, later than those specified in the ISI program. If the desired Edition or Addenda are referenced in 10 CFR 50.55a(b)(2), the request is submitted following the guidance of the RIS. These types of request are not required to demonstrate hardship, difficulty, or provide evidence of quality and safety. They do need to ensure that all related requirements are also used. Requests to use edition and/or addenda of ASME Section XI that are referenced in 10 CFR 50.55a(b)(2) that are later than the initial Code of Record established for the ISI program shall be submitted under the provisions of 10 CFR 50.55a(g)(4)(iv).

The guidance of RIS 2004-12 does not apply to the use of later Edition/Addenda of ASME Section XI, Appendix I and VIII. In accordance with the provisions of 10 CFR 50.55a(g)(4)(ii), Appendix I and VIII from the latest Edition/Addenda of ASME Section XI listed in 10 CFR 50.55a(a) may be adopted anytime within the interval without additional NRC approval.

Program Section No.: SEP-ISI-104 Revision No.: 5 Page No.: Page 26 of 91 Table 3.0-1 Waterford Unit No.3 Fourth Interval Relief Requests Entergy NRC SER Relief Request Relief Request Description Correspondence Correspondence W3-ISI-006 Request to extend the second and third 10-year In-Service Inspection W3F1-2013-0062 Revision 1 Interval examinations of the reactor vessel welds (Category B-A) as well as the nozzle-to-vessel welds and inner radius sections (Category B-D) until 2015 and 2035, respectively, plus or minus one refueling cycle.

W3-ISI-023 ASME Code Case N-770-1 Weld Examination Coverage W3F1-2013-0044 TAC NO. MF2815 Requirements.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 27 of 91 4.0 Developmental References 4.1 Title 10, Code of Federal Regulations, Part 50 4.2 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section XI, 2007 Edition, 2008 Addenda 4.3 Regulatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity 4.4 Regulatory Guide 1.147 Revision 17, Code Case Applicability 4.5 Regulatory Guide 1.26, Quality Group Classifications 4.6 Waterford Unit 3 Technical Specifications 4.7 Waterford Unit 3 Final Safety Analysis Report (FSAR) 4.8 CEP-COS-100, Control and Use of Iddeal Concepts Software 4.9 CEP-COS-0110, Control and Use of the of ScheduleWorks Module of IDDEAL Software 4.10 EN-DC-174, Engineering Program Sections 4.11 CEP-PT-001, ASME Section XI Pressure Testing Program 4.12 CEP-RR-001, ASME Section XI Repair/Replacement Program 4.13 SEP-PT-WF3-001, WF3 Period Pressure Testing Program 4.14 EN-DC-351 Inservice Inspection Program Duties and Responsibilities 4.15 CEP-ISI-100 ASME Section XI, Division 1 Fleet Administrative Controls for Inservice Inspection Program 5.0 Augmented Examination Programs The Ten-Year ISI Program contains augmented examination plans and schedules that may be inside or outside the scope of 10 CFR 50.55a. These examinations may result from activities including, but not limited to, commitments made to the NRC per 10 CFR 50.55a, commitments made outside of 10 CFR 50.55a and made based on other document requirements, internal commitments, license renewal commitments, or industry initiatives. These examinations may use ASME Section XI techniques and procedures as defined in the commitment, however, the requirements of ASME Section XI, including reporting, do not apply when these examinations are required outside of 10 CFR 50.55a.

Augmented examinations that are not required by 10 CFR 50.55a and ASME Code requirements, such as Section XI and its Code Cases, are not required to be included in the summary totals for a specific examination category. Changes to these examination requirements and schedules are outside of 10 CFR 50.55a and are required to be evaluated using the 10 CFR 50.59 screening/review process unless otherwise specified.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 28 of 91 Augmented examinations are considered in one of the following three categories:

Augmented Examinations for External Commitments; Owner Elected Examinations for Internal Commitments; and License Renewal Examinations for Aging Management Commitments. These three types of augmented examinations are described in Sections 5.1, 5.2 and 5.3 below. In addition, each section includes a table in Section 7.0 which lists the applicable augmented programs that are described in detail within each section.

5.1 Augmented Examinations for External Commitments This type of augmented examination is conducted to meet a commitment made to a source outside the utility. Typically these are commitments made to the NRC in response to regulatory documents such as Generic Letters, Bulletins and NUREGs. The augmented examinations for external commitments covered in this program are as follows:

5.1.1 RI-BER

Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping (Examination of High Energy Line Break Piping)

Source Document: USNRC Branch Technical Position, MEB 3-1 dated November 1975 and NUREG-0800, Section 3.6.2.

Commitment No.: P-4845.

Associated Document: Engineering Reports ENTP-19Q-302, WF3-ME-07-00001, WF3-SA-07-00001 and WF3-SA-07-00002.

Purpose:

The purpose of this augmented program is to perform examinations on piping subject to High Energy Line Break (HELB) analysis criteria. During previous intervals examinations were required on 100% of the subject welds per NUREG-0800, Section 3.6.2, Branch Technical Position MEB 3-1, Paragraph B.1.b(7). For this interval the examinations will be maintained and performed under the risk-informed break exclusion region (RI-BER) application.

Scope: For Main Steam and Feedwater piping, all of the circumferential and longitudinal welds from both Steam Generators to the first rigid restraint past the outer containment isolation valve.

For Moderate Energy Safety Injection piping, the augmented ISI requirements will be imposed on the two 14-inch shutdown cooling lines and the four 8-inch LPSI lines penetrating containment.

This was defined by Revisions 0-2 of the first interval ISI Plan. The lines were identified by line number and the welds were identified as requiring augmented

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 29 of 91 examination. This was approved in NRC SER dated January 29, 1987. On May 11, 2001, the NRC completed an inspection at Waterford Steam Electric Station, Unit 3.

The design basis for the high energy line break exclusion area on the main steam and main feedwater systems as documented in the FSAR and the NRCs SER of July 1981 was reviewed. The purpose of the review was to determine whether the design bases of the systems were met by the installed and tested configurations. Inspection Report IR-2001-03 (ILN 01-0097, dated 6/01/01) documented that no findings of significance were identified.

Method: Ultrasonic examinations will be performed per the RI-BER application, except as restricted by part geometry or access.

Industry Code or Standards: ASME Code Section XI, 2007 Edition, 2008 Addenda.

Frequency: Ultrasonic examinations will be performed once per 10 year interval per the RI-BER application.

Acceptance Criteria or Standard: Flaws detected during examination shall be evaluated by comparing the examination results to the acceptance standards established in ASME Section XI, IWB-3514.

Regulatory Basis: The examination criteria of NUREG-0800, Section 3.6.2 are superseded by the RI-BER application.

5.1.2 RG 1.14 RCP FLYWHEEL: Reactor Coolant Pump Flywheel Integrity Source Document: NRC Regulatory Guide 1.14.

Commitment No.: P-9944.

Associated Document: Technical Specification 6.5.7.

Purpose:

The reactor coolant pump (RCP) motor flywheels are examined due to a concern about high-energy missiles inside containment that could potentially damage and cause the simultaneous failure of multiple trains of multiple safety-related systems.

Scope: The scope includes the examination of all four RCP flywheels.

Method: Surface and volumetric examinations of all four RCP flywheels shall be conducted in accordance with WF3 Technical Specification 6.5.7. These examinations are to be performed to the extent possible through the access ports in the motor housings without disassembly of the motors.

Industry Code or Standards: ASME Code Section XI, 2007 Edition, 2008 Addenda.

Frequency: All 4 RCP flywheels shall be inspected once every ISI 10-Year Interval.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 30 of 91 Acceptance Criteria or Standard: Any flaws detected during examination shall be forwarded to Entergy Engineering for resolution.

Regulatory Basis: The regulatory basis for this augmented examination program is NRC Regulatory Guide 1.14.

5.1.3 W3-R&R-006: Volumetric Examinations of Weld Overlay Including Upper 25%

of Original Weld Source Document: W3-R&R-006 (TAC NO. MD5388) 08-0069.

Commitment No.: A-27070 & A-27148.

Associated Document: W3F1-2013-0062, CR-WF3-2011-6576, CR-WF3-2009-2572, EC 15163.

Purpose:

Ensure cracking does not propagate into the upper 25% of the original base metal of dissimilar metal welds that contain alloy that have been mitigated by full structural weld overlays. The weld overlay examination shall be ultrasonically examined during the first or second refueling outage following application. Welds mitigated in RF15.

Scope: Pressurizer Nozzle to Surge Line, Pressurizer Nozzle to Relief Valve (2 locations), Pressurizer Spare Nozzle (capped), Pressurizer Nozzle to Spray Line, Hot Leg Loop 1 Nozzle to Surge Line, Hot Leg Loop 1 Nozzle to Drain Line, Hot Leg Loop 1 Nozzle to Shut Down Cooling (SDC) Line and Hot Leg Loop 2 Nozzle to SDC Line.

Method: Perform Ultrasonic Examinations of the proposed preemptive full structural weld overlays. The examination volume A-B-C-D of Figure 2 for each weld overlay will be examined using ultrasonic examination procedures and personnel qualified in accordance with ASME Section XI, Appendix VIII (as implemented through PDI).

However, as of the submittal date of this request, PDI-qualified procedures to examine the upper 25% of the cast austenitic stainless steel are not available.

Therefore, until such procedures are qualified, Entergy will perform best effort inservice inspections of this region. Once qualified, Entergy will implement these PDI procedures at the next scheduled inservice inspection as required by Attachment 2, paragraph 3.0(c) of W3-R&R-006.

Industry Code or Standards: ASME Code Section XI, 2007 Edition, 2008 Addenda may be used for guidance.

Frequency: ISI of the DMW overlays with cast stainless steel safe ends will be performed at a higher frequency than those without cast stainless steel safe ends.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 31 of 91 First, the initial UT inspection of the DMW weld overlays will be performed during the first or second refueling outage after installation. Then, instead of being inspected on a twenty-five percent sample basis, these weld overlays, with the exception of the Pressurizer Safety/Relief Valves weld overlays, will be UT examined once every ten years from the date of installation until such time when a technique is qualified to examine cast stainless steel in accordance with the PDI Program. Then also, instead of being inspected on a twenty-five percent sample basis, the Pressurizer Safety/Relief Valves weld overlays will be UT examined once every nine years from the date of installation until such time when a technique is qualified to examine cast stainless steel in accordance with the PDI Program. Once the subject DMW overlays are UT examined by a qualified PDI technique (for cast stainless steel) and no planar flaws are detected, then the DMW overlays will be placed into a population in which 25% of that population is examined once every 10 years. Until such time, the UT inspections of the subject DMW overlays (with cast stainless steel safe ends) will not be credited to satisfy the 25% sample inspection requirement. Note the updating of the commitment and communication of this change to the NRC is being tracked within CR-WF3-2011-6576.

Acceptance Criteria or Standard: The acceptance standards for the ultrasonic examination of the weld overlay are specified in Table IWB-3514-2. However, if the weld overlay fails to meet the acceptance standards of Table IWB-3514-2, it can be accepted based on an analytical evaluation meeting the requirements and acceptance criteria of IWB-3600. However, flaws identified as primary water stress corrosion cracking (PWSCC) cannot be accepted by an IWB-3600 analytical evaluation. Cracks in the outer 25% of the base metal shall meet the design analysis requirements of 2.0 in Attachment 2 to W3-R&R-006.

Regulatory Basis: W3-R&R-006.

5.1.4 NUREG-0612 HEAVY LOADS: Control of Heavy Loads at Nuclear Power Plants Source Document: NUREG-0612.

Commitment No.: P8577, P8583, P8584, P8589, P8590, P8591, P8592, & P8594.

Associated Documents: NOECP-256, Non Destructive Examination of Reactor Vessel Head, Upper Guide Structure, and Core Support Barrel Lifting Rig Assemblies.

Purpose:

The purpose of this augmented examination is to perform surface examinations of load bearing welds on special lifting devices in accordance with NUREG-0612. Visual examinations of the components shall be performed per NOECP-256 and are not included in the WF3 ISI program.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 32 of 91 Scope: As defined in NOECP-256, the special lifting devices addressed by this augmented examination are the RVH Lift Rig, Upper Guide Structure Lift Rig, and Core Support Barrel Lift Rig.

Method: Load bearing welds on the subject components shall be subject to a surface examination.

Industry Code or Standards: ANSI N14.6-1978 and ANSI B30.9-1971 were used to develop this program.

Frequency: The surface examination shall be performed once every ten years.

Acceptance Criteria or Standard: Acceptance criteria are stated in N14.6-1978, Section 5.5.

Regulatory Basis: The regulatory basis for this augmented examination program is NRC NUREG-0612.

5.1.5 MRP-146

Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines Source Document: MRP-146 Rev 1, EC-48006.

Commitment No.: N/A.

Associated Documents: LO-HQNLO-2007-68, WSES-18Q-301, SEP-ISI-104-PDCN-003.

Purpose:

The purpose of this augmented examination is to perform ultrasonic examinations of portions of piping that were assessed to possibly be susceptible to thermal fatigue in normally stagnant, non-isolable RCS branch lines. This is per guidance developed in MRP-146 Rev 0, MRP-146 Rev 1 and MRP-146S, Rev 0.

Scope: Assessment of WF3 piping was performed and documented in WSES-18Q-301 and EC-48006. There were three lines identified as applicable and requiring inspection per MRP-146 Rev 1, Section 2.4 (1RC2-46RL1A, 1RC2-48RL2A, 1RC2-47RL1B).

Method: Perform volumetric examination on the subject line applicable welds.

Industry Code or Standards: MRP-146 and MRP-170.

Frequency: The volumetric examination shall be performed once every other outage (MRP-146 Rev 1, Section 2.1.5.5).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 33 of 91 Acceptance Criteria or Standard: MRP-146 Rev 1, Section 2.4.

Regulatory Basis: The regulatory basis for this augmented examination program is Needed Requirement of an NEI 03-08 issue program.

5.2 Owner Elected Examinations for Internal Commitments This type of augmented examination is conducted to meet a commitment made internally at the plant. These typically involve examinations resulting from Condition Reports or similar documents that identify conditions in the plant that warrant monitoring, but are outside the scope of ASME Section XI.

5.2.1 There are no Owner Elected Examinations for Internal Commitments for WF3 at this time.

5.3 License Renewal Examinations for Aging Management Commitments This type of augmented examination is conducted to ensure the integrity of components during the license extension period. These typically involve components or criteria that are beyond those of ASME Section XI, but in some cases may supplement existing Code or regulatory requirements.

5.3.1 There are no License Renewal Examinations for Aging Management Commitments for WF3 at this time.

6.0 ASME Systems & Examination Boundaries 6.1 ASME Class 1, 2 and 3 This section defines those systems that are designated as ASME Class 1, 2, or 3 and provides justification for their inclusion or exclusion within the Fourth Ten-Year Inspection Interval NDE Program. Several of WF3 systems or portions of systems are excluded from system and component NDE as allowed by Articles IWB-1200, IWC-1200, and IWD-1200.

However, these portions of systems are not excluded from the pressure testing requirements of ASME Code, Section XI, except as allowed by Articles IWA-5000, IWB-5000, IWC-5000, and IWD-5000. System information for the NDE Program is provided in Appendix A, Waterford 3 ASME Code Class Systems ISI Boundary Diagrams. The 2007 Edition, 2008 Addenda of the ASME Code, Section XI defines the inspection requirements for each of the ASME Code Classes within the fourth inspection interval, which started on December 1, 2017.

Per IWA-1400(a) of the 2007 Edition, 2008 Addenda of Section XI, it is the owner's responsibility to determine the appropriate Code Classes for each component and to identify the system boundaries subject to inspection. IWA-1400(a), footnote 1, states that classification criteria are specified in 10 CFR 50. This reference is to footnote 7 of 10 CFR 50.55a which references Regulatory Guide 1.26 and Section 3.2.2 of NUREG-0800.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 34 of 91 The component classifications of the ASME Code (Class 1, 2, or 3) determine the rules and requirements for inspection and define the Section XI examination boundaries.

Because early vintage nuclear plants were designed and constructed before Section III of the ASME Boiler and Pressure Vessel Code was incorporated into 10 CFR 50.55a, the ASME Section XI Code classifications for ISI may differ from the original design classifications. Therefore, while the ASME Code classifications determine the rules for repairs and replacements and the component inspection requirements, repairs and replacements are generally performed to meet the specifications of the original design code.

In order to avoid confusion between ASME Code Class for design (typically ASME Section III) and ASME Code Class for inservice inspection (ASME Section XI), Entergy is utilizing ISI Class 1, 2, and 3 terminology for the application of ASME Section XI inservice inspection requirements.

Components considered to be optionally classified as ASME Code Class may be excluded from the Section XI ISI Program per IWA-1320(e).

Determination of the ASME Code Class (or equivalent to ASME Code Class for components not designed in accordance with ASME Section III requirements) for WF3 was performed as follows:

The ASME Code Class of components was established in accordance with the safety classification criteria of ANSI N18.2, "Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants", 1973, and ANSI N18.2a, "Revision and Addendum to Nuclear Safety Criteria for the Design of Stationary Pressurized Water Reactor Plants",

1975 as discussed in FSAR Section 3.2.2.

ISI Boundary Diagrams, ISI Isometrics and ISI Equipment Drawings were developed to identify the ASME Code Class 1, 2, and 3 components (welds, bolting, etc.) and component supports which are subject to examination per ASME Section XI. The ISI Boundary Diagrams, ISI Isometrics and ISI Equipment Drawings were developed to support the ISI Program and are not to be used for design information.

The systems and components included in the Inservice Inspection Program are identified in the W3 ISI Boundary Diagrams. A listing of these ISI Boundary Diagrams is included on drawing P-100 Sheet 1. These systems and components are referred to as ISI Class 1, ISI Class 2 or ISI Class 3 and are subject to the requirements of ASME Section XI, Subsections IWB, IWC, IWD and IWF.

Typically, systems and components at WF3, which is designed in accordance with ASME Section III, Class 1, 2, and 3 requirements are included in the Inservice Inspection Program. Some of these systems and components are excluded or exempted from Section XI requirements because they are optionally upgraded to Code Class as discussed in IWA-1320(e) or because they do not perform the functions listed in IWD-1210(a) to (f). Explanation of the implementation of Section XI requirements (including exemptions and exclusions) are referenced on the Piping Line Coding Legends and Exemption and Exclusion Flag Legends P-100 Sheet 1.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 35 of 91 The ISI Class 1, 2, and 3 components which are not exempt from examination per ASME Section XI, IWB-1220, IWC-1220, IWD-1220, and IWF-1230 are identified on the ISI Boundary Diagrams in accordance with the Pipe Line Codes on drawing P-100 Sheet 1. Nonexempt components are not included within flags corresponding to Section XI exemptions listed in the plant Exemption and Exclusion Flag Legend.

The ISI Class 1, 2, and 3 components which are exempt from examination are those which meet the criteria of ASME Section XI, IWB-1220, IWC-1220 and IWD-1220.

Component supports which meet the criteria of IWF-1230 are also exempt from examination. Exempt components are identified on the ISI Boundary Diagrams in accordance with the Exemption and Exclusion Flag Legends on drawing P-100 Sheet

1. Exempt component supports are supports of components which are exempt from examination per IWB-1220, IWC-1220, and IWD-1220.

ASME Section XI, IWC-1222(b) presents separate exemption criteria for Class 2 auxiliary feedwater systems in pressurized water reactor plants. For WF3, there is a conflict with the nomenclature for this piping regarding the application of IWC-1222(b) requirements.

At WF3, there are two systems that support this function, one designated as Auxiliary Feedwater (AFW) and the other designated as Emergency Feedwater (EFW).

Waterford Design Basis Document W3-DBD-20, Main Feedwater, indicates that the AFW system is non-safety related and Waterford Design Basis Document W3-DBD-3 Emergency Feedwater System indicates that the EFW is safety related.

Because only the safety related portions of the auxiliary feedwater systems (i.e.,

Emergency Feedwater) have been classified as ISI Class 2, the components in these systems will be subject to the exemption criteria of ASME Section XI, IWC-1222(b) and subsequent examination criteria of Table IWC-2500-1. The non-safety related components performing the auxiliary feedwater function at WF3 (i.e., AFW system) are not classified for ASME Section XI inservice inspections and are not subject to the requirements of IWC-1222(b).

An index of ISI Boundary Diagrams, ISI Isometrics and ISI Equipment Drawings is included in Appendix A and Appendix B.

6.2 Calibration Standards 6.2.1 Calibration standards are used to calibrate ultrasonic examination equipment. These standards were used during preservice examinations and during the previous ISI intervals, and are maintained at WF3. Calibration standards are listed in the IDDEAL database.

Calibration block thickness is the same nominal thickness as the material being examined. Any calibration block thickness that is within + 25% of the material being examined may be considered to be the same nominal thickness.

Alternative calibration blocks may be used if the referencing procedure allows the use of such blocks. Additions or changes to the calibration blocks listed in the IDDEAL

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 36 of 91 database shall be requested with the use of a Program Database Change Notice (PDCN) in accordance with Program Section CEP-COS-0110.

7.0 Application Criteria and Code Compliance 7.1 ASME Section XI The following provides a summary of the application of ASME Code, Section XI, 2007 Edition, 2008 Addenda to the Waterford Unit No.3 Ten-Year Program for the Fourth Inspection Interval. Class 1, 2 and 3 components and component supports were selected for examination per the requirements of ASME Section XI 2007 Edition, 2008 Addenda Tables IWB-2500-1, IWC-2500-1, IWD-2500-1 and IWF-2500-1, respectively. The application and distribution of examinations for this interval is based upon Articles IWB-2411, IWC-2411, IWD-2411 and IWF-2410 of Section XI. Augmented examinations will align with the ASME Section XI interval unless stated otherwise in the document that governs the specific criteria for the examinations. There may be instances when augmented examinations and ASME Section XI examinations coincide for the same component. Examinations should be performed concurrently to optimize examination resources to the extent practical.

If there are less than three items or welds to be examined in an Examination Category, the items or welds may be examined in any two periods, or in any one period if there is only one item or weld in lieu of the percentage requirements in IWB-2411-1, IWC-2411-1 or IWD-2411.

Administration and control of Section XI examination records and personnel qualification requirements are included in CEP-NDE-0100, Administration and Control of NDE.

The results of this application are summarized by ASME Category and Item number and are contained within Table 7.1. Table 7.1 only contains those ASME Item numbers that are relevant to WF3.

7.1.1 EXAMINATION CATEGORY B-A - PRESSURE RETAINING WELDS IN REACTOR VESSEL The Reactor Vessel examinations are scheduled to be examined in the third interval to meet the 2007 Edition, 2008 Addenda of the ASME Code, Section XI as required by 10 CFR 50.55a. Once Request for Alternative W3-ISI-006 is approved for the Fourth Interval, these examinations can be removed from the schedule.

7.1.2 EXAMINATION CATEGORY B-B - PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS The 2007 Edition, 2008 Addenda of the ASME Code Section XI, Examination Category B-B requires that pressurizer and primary side of steam generators be examined during the inspection interval. The examination may be limited to one vessel among the group of vessels performing a similar function.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 37 of 91 STEAM GENERATORS There is one Class 1 Examination Category B-B, tube sheet to head weld on each steam generator (S/G). Table IWB-2500-1, Category B-B, Note (1) allows the examination be limited to one vessel among a group of vessels performing a similar function or 50%. One of the two tubesheet to head welds is selected for examination, thus meeting the one vessel among a group of vessels performing a similar function or 50% requirement.

PRESSURIZER Section XI requires examination of both shell to head welds and one foot of one intersecting longitudinal weld per head. Both shell to head welds are selected meeting the 100%

requirement. There are two longitudinal welds on each shell to head weld (total 4). One foot of one longitudinal weld on each head requires examination per ASME Section XI. 1 foot each of two of the longitudinal welds is selected for examination or 50%.

This meets the Examination Category B-B examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.3 EXAMINATION CATEGORY B-D - FULL PENETRATION WELDED NOZZLES IN VESSELS This category applies to the reactor vessel, pressurizer, and steam generators. ASME Code, Section XI does not allow the deferral of these examinations, except for nozzle to reactor vessel welds. Table IWB-2500-1, Category B-D, Note 3, allows partial deferral if the nozzle weld is examined by the straight beam ultrasonic method from inside the nozzle bore, the remaining examinations required from the shell inside diameter may be performed at or near the end of the interval.

The RV Nozzle-to-Vessel and Nozzle Inner Radius Sections are scheduled for examination during the Third Period of the Fourth Interval meeting the requirements of ASME Section XI.

Once NRC approval of Request for Alternative W3-ISI-006 is obtained for the Fourth Interval, these examinations may be removed from the schedule.

10 CFR 50.55a(b)(2)(xxi)(A) requires that the provisions of Table IWB-2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Code Items B3.120 and B3.140 of the 1998 Edition must be applied when using the 2007 Edition, 2008 Addenda. A visual examination with magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, with a limiting assumption on the flaw aspect ratio (i.e.,

a/l=0.5), may be performed instead of an ultrasonic examination. All of the pressurizer and steam generator nozzle to vessel welds and nozzle inner radii are selected for examination meeting the 100% examination requirement. Each Steam Generator has three integrally cast nozzles. Therefore, all six of the inner radii are scheduled for examination.

This meets the Examination Category B-D examination requirements in the 2007 Edition, 2008 Addenda of Section XI as modified by 10 CFR 50.55a(b)(2)(xxi)(A).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 38 of 91 7.1.4 EXAMINATION CATEGORY B-F - PRESSURE RETAINING DISSIMILAR METAL WELDS IN VESSEL NOZZLES This category addresses Nozzle-to-Safe End Welds and Piping Welds. WF3 has developed a Code Case N-716-1 RI-ISI program. All Examination Category B-F Welds have been re-categorized as R-A welds in accordance with Code Case N-716-1. Therefore no examinations will be performed per Examination Category B-F.

7.1.5 EXAMINATION CATEGORY B-G PRESSURE RETAINING BOLTING, GREATER THAN 2 in. (50 mm) IN DIAMETER The RV Studs, Nuts, Threads in Flange, Washers, and Bushings are scheduled. For Volumetric examination of Pump Studs, one of four sets of reactor coolant pump studs is selected for volumetric examination. For the visual examination of pump bolting, one of the reactor coolant pumps will be examined, only if disassembled. This is the same requirement as for Examination Category B-L-2. All other B-G-1 components are scheduled for examination.

This meets the Examination Category B-G-1 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.6 EXAMINATION CATEGORY B-G PRESSURE RETAINING BOLTING, 2 in. (50 mm)

AND LESS IN DIAMETER This category includes pressurizer manway, steam generator manway, reactor coolant pump heat exchanger, reactor coolant pump seal bolting, flange bolting and valve bolting.

Examinations will be conducted as required in Table IWB-2500-1 in the 2007 Edition, 2008 Addenda of Section XI. This bolting will only be examined if associated connections are disassembled.

For components other than piping, bolting examinations will be required only when the component is examined under Examination Category B-B, B-L-2, or B-M-2. WF3 has grouped B-G-2 bolting in accordance with components examined under Examination Category B-B, B-L-2, or B-M-2.

This meets the Examination Category B-G-2 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.7 EXAMINATION CATEGORY B-J - PRESSURE RETAINING WELDS IN PIPING This category addresses piping welds. WF3 has developed a Code Case N-716-1 RI-ISI program. All Examination Category B-J Welds have been re-categorized as R-A welds in accordance with Code Case N-716-1. Therefore no examinations will be performed per Examination Category B-J.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 39 of 91 7.1.8 EXAMINATION CATEGORY B-K - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category B-K of the ASME Code Section XI, 2007 Edition, 2008 Addenda requires examination of Welded Attachments. For vessel welded attachments, Note 4 allows 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. For single vessels, only one welded attachment shall be selected for examination. WF3 has one welded attachment on the Reactor Vessel, one welded attachment on the Pressurizer and one welded attachment on each Steam Generator. The Reactor Vessel welded attachment, the Pressurizer welded attachment and one of the two S/G welded attachments are selected for examination or 75% of all vessel welded attachments.

For piping, pumps and valves, inspection of 10% of the total population of welded attachments is required. 10% of all reactor coolant pump welded attachments are selected for examination. WF3 does not have any piping or valve welded attachments.

This meets the Examination Category B-K examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.9 EXAMINATION CATEGORY B-L PUMP CASINGS This category involves reactor coolant pumps and requires visual, VT-3 examination on pump internal surfaces when disassembled. The reactor coolant pumps are the only ASME Code Class 1 pumps.

B-L-2 requires examination of one pump in each group of pumps performing similar functions in the system. However, this examination is only required when a pump is dissembled for maintenance, or repair. None of the four Reactor Coolant Pump internal surfaces have been selected. This requirement will be met during disassembly for maintenance, or repair.

This meets Examination Category B-L-2 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.10 EXAMINATION CATEGORY B-M VALVE BODIES Examinations will be conducted as required in Table IWB-2500-1. There are twenty two valves in WF3 within this examination category. The twenty two valves are placed into six groups and are identified in the Program Type field in the IDDEAL Software Suite, computer program.

B-M-2 examinations are limited to at least one valve within each group of valves that are of the same size, constructural design, and manufacturing method, and that perform similar functions in the system. WF3 Class 1 valves have been separated into 6 valve groups.

However, this examination is only required when a valve is dissembled for maintenance, or repair. No valve body internal surfaces have been selected. This requirement will be met during disassembly for maintenance, or repair.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 40 of 91 This meets Examination Category B-M-2 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.11 EXAMINATION CATEGORY B-N REACTOR VESSEL INTERIOR These examinations will be conducted each inspection period.

This meets the Examination Category B-N-1 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.12 EXAMINATION CATEGORY B-N WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSELS Examination will be conducted during the third period to coincide with the 10 year inspection of the reactor vessel.

This meets the Examination Category B-N-2 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.13 EXAMINATION CATEGORY B-N REMOVABLE CORE SUPPORT STRUCTURES Examination will be conducted during the third period to coincide with the 10 year inspection of the reactor vessel.

This meets the Examination Category B-N-3 examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.14 EXAMINATION CATEGORY B-O - PRESSURE RETAINING WELDS IN CONTROL ROD DRIVE AND INSTRUMENT NOZZLE HOUSINGS The ASME Code, Section XI, 2007 Edition, 2008 Addenda, requires volumetric or surface examination of 10% of peripheral control rod drive housings during the inspection interval.

There are 28 peripheral Control Rod Drive Housings (CRDH) on the reactor vessel head with 4 welds each Control Rod Drive Housing. There are 10 Incore Instrumentation (ICI)

Nozzles on the reactor vessel head with 1 weld each per ICI. To meet the Code requirements 3 of the CRDH Housings have been selected for examination, each with 4 welds for a total of 12 welds requiring examination and 1 ICI Nozzle for a total of 1 weld.

This meets the Examination Category B-O examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

WF3 does not contain any instrument nozzle housings, welds in In-Core Instrumentation Nozzle (ICI) Housings > NPS 2 (DN50).

7.1.15 EXAMINATION CATEGORY B-P - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at WF3 meets the requirements of ASME Code, Section XI, 2007 Edition, 2008 Addenda for Class 1 systems. Details of the component listing are contained in the Pressure Test Program Plan.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 41 of 91 7.1.16 EXAMINATION CATEGORY B-Q - STEAM GENERATOR TUBING The Steam Generator Tube inspection program at WF3 is governed by Technical Specification 6.5.9.

7.1.17 EXAMINATION CATEGORY C-A - PRESSURE RETAINING WELDS IN PRESSURE VESSELS This category applies to the steam generators and shutdown cooling heat exchangers. WF3 has developed a Code Case N-716-1 RI-ISI program. The shutdown cooling heat exchangers were evaluated as Low Safety Significant (LSS) and therefore no examinations are required. The steam generators were evaluated as High Safety Significant (HSS) and therefore required examinations in accordance with Category C-A. The Extent of Examination in Table IWC-2500-1, Examination Category C-A only requires examination of Cylindrical-shell-to-conical-shell-junction welds and shell (or head)-to-flange welds, Head to shell welds and Tubesheet-to-shell welds. Note (3) states: 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 or 50%. There is one circumferential head weld on each Steam Generator, therefore, 1 of 2 Steam Generator Head welds is selected for 50%. There is one tubesheet-to-shell weld on each Steam Generator. One tubesheet-to-shell weld on one Steam Generator is selected for examination for 50%.

There are 6 welds on each Steam Generator that are not considered Cylindrical-shell-to-conical-shell-junction welds or shell (or head)-to-flange welds and thus are exempt from examination.

This meets the Examination Category C-A examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.18 EXAMINATION CATEGORY C-B - PRESSURE RETAINING NOZZLE WELDS IN VESSELS This category applies to steam generators and shutdown cooling heat exchangers. WF3 has developed a Code Case N-716-1 RI-ISI program. The shutdown cooling heat exchangers were evaluated as Low Safety Significant (LSS) and therefore no examinations are required. The steam generators were evaluated as High Safety Significant (HSS) and therefore required examinations in accordance with Category C-B. Note (1) in Table IWC-2500-1, Category C-B, excludes manways and handholes. Note (3) requires that nozzles selected initially for examination shall be reexamined in the same sequence over the service life of the component, to the extent practical. Note (4) allows that 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 Feedwater nozzle to vessel welds (including inner radius) on one Steam Generator are selected for examination for 50%.

This meets the Examination Category C-B examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 42 of 91 7.1.19 EXAMINATION CATEGORY C-C - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category C-C of the ASME Code Section XI, 2007 Edition, 2008 Addenda requires examination of Welded Attachments. For vessel welded attachments, Note 4 allows 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. There are two steam generators with four welded attachment welds each. One of the steam generator welded attachment welds is selected for examination or 12.5% of all vessel welded attachment welds.

For piping, pumps and valves, inspection of 10% of the total population of welded attachments is required. Percentage of all piping and pump welded attachments scheduled for examination is based on IWF-2500-1 support selection.

WF3 has no valve welded attachments.

This meets the Examination Category C-C examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.20 EXAMINATION CATEGORY C-D - PRESSURE RETAINING BOLTING GREATER THAN 2 IN. (50 MM) IN DIAMTER WF3 does not contain any Class 2 pressure retaining bolting greater than 2 in. (50 mm) in diameter.

7.1.21 EXAMINATION CATEGORY C-F PRESSURE RETAINING WELDS IN AUSTENITIC STAINLESS STEEL OR HIGH ALLOY PIPING This category addresses Class 2 piping welds. WF3 has developed a Code Case N-716-1 RI-ISI program. Examination Category C-F-1 welds has been re-categorized as R-A welds in accordance with Code Case N-716-1.Therefore no examinations will be performed per Examination Category C-F-1.

7.1.22 EXAMINATION CATEGORY C-F PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL PIPING This category addresses Class 2 piping welds. WF3 has developed a Code Case N-716-1 RI-ISI program. Examination Category C-F-2 Welds have been re-categorized as R-A welds in accordance with Code Case N-716-1. Therefore no examinations will be performed per Examination Category C-F-2.

7.1.23 EXAMINATION CATEGORY C-H - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at WF3 meets the requirements of ASME Code, Section XI, 2007 Edition, 2008 Addenda for Class 2 systems. Details of the component testing are contained in the Pressure Test Program Plan.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 43 of 91 7.1.24 EXAMINATION CATEGORY D-A - WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category D-A of the ASME Code Section XI, 2007 Edition, 2008 Addenda requires examination of Welded Attachments. WF3 has no pump or valve welded attachments.

There are two Component Cooling Water Heat Exchangers with one welded attachment each, therefore the welded attachment on one Component Cooling Water Heat Exchanger is scheduled for 50%.

For piping, inspection of 10% of the total population of welded attachments is required and have been selected. The selections are proportional to the total number of nonexempt welded attachments in each system.

This meets the Examination Category D-A examination requirements in the 2007 Edition, 2008 Addenda of Section XI.

7.1.25 EXAMINATION CATEGORY D-B - ALL PRESSURE RETAINING COMPONENTS The pressure testing program at WF3 meets the requirements of ASME Code, Section XI, 2007 Edition, 2008 Addenda for Class 3 systems. Details of the component listing are contained in the Pressure Test Program Plan.

7.1.26 EXAMINATION CATEGORY F-A - SUPPORTS Examination Category F-A of the ASME Code Section XI, 2007 Edition, 2008 Addenda requires 25% of Class 1 Piping Supports, 15% of Class 2 Piping Supports, and 10% of Class 3 Piping Supports to be examined during the inspection interval. 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. The supports have been separated by type as defined in Note (1) to Examination Category F-A. This type has been added to the Item Number to clearly identify each support by type. Twenty-five percent (25%) of the Class 1 supports have been selected and are prorated by type and system. Fifteen percent (15%) of the Class 2 supports have been selected and are prorated by type and system. Ten percent (10%) of the Class 3 supports have been selected and are prorated by type and system.

For supports, other than piping supports, the components have been scheduled as follows:

Table IWF-2500-1 Note (3) requires only one component of multiple components of similar design, function, and service, to be scheduled for inspection.

Seventeen of the forty non piping supports are selected for examination or 42.5% of all non piping supports.

The selection of these component supports for examination meets the requirements of Examination Category F-A Examination Requirements in the ASME Code Section XI, 2007 Edition, 2008 Addenda.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 44 of 91 7.1.27 EXAMINATION CATEGORY R-A Note The PRA quality for items IF-B2, IF-C3c, IF-C6, IF-C8, IF-D5a, ID-D7, and IF-E5a has been determined to not meet Appendix II of Code Case N-716-1. Prior to completing examinations for the first period, these items must meet the Quality requirements of Appendix II or:

1. Using the provisions of 10 CFR 50.55a(z), obtain approval from the NRC to use N-716-1 without meeting the Quality requirements, or

2. Perform the examinations required by Examination Categories B-F, B-J, C-F-1, C-F-2, C-A, and C-B.

Note The scheduling of Examination Category R-A contained in the IDDEAL Software Suite and summarized in Table 7.1 has been completed based on preliminary evaluations meeting Code Case N-716-1. When the N-716-1 compliance report is complete, the schedule and Table 7.1 requires verification. Verification and changes, if required, should be completed prior to completing examinations for the first period.

The RIS-B Program utilizing ASME Code Case N-716-1 has been substituted for the current program for Class 1 and 2 piping (Examination Categories B-F, B-J, C-F-1,C-F-2, C-A, and C-B).

7.1.28 CODE CASE N-722 ADDITIONAL EXAMINATIONS FOR PWR PRESSURE RETAINING WELDS IN CLASS 1 COMPONENTS FABRICATED WITH ALLOY 600/82/182 MATERALS Code Case N-722-1 provides additional examination requirements beyond those required by ASME Table IWB-2500-1 for PWR pressure retaining welds in ASME Class 1 components fabricated with Alloy 600/82/182 material. The selection of these welds for examination meets the requirements of Code Case N-722-1, as mandated in 10 CFR 50.55a(g)(6)(ii)(E).

7.1.29 CODE CASE N-729 ALERNATIVE EXAMINATION REQUIREMENTS FOR PWR REACTOR VESSEL UPPER HEADS WIH NOZZLES HAVING PRESSURE -RETAINING PARTIAL-PENETRATION WELDS Code Case N-729-4 provides alternative examination requirements for PWR Reactor Vessel Upper Heads with nozzles having pressure retaining partial penetration welds. The Reactor Vessel Head will be visually examined every third refueling outage, or 5 calendar

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 45 of 91 years, whichever is less and the Nozzles and Partial-Penetration Welds in the head will be volumetrically and or surface examined once per inspection interval.

The selection of these welds for examination meets the requirements of Code Case N-729-1 examination requirements as mandated in 10 CFR 50.55a(g)(6)(ii)(D).

7.1.30 CODE CASE N-770 ALTERNATIVE EXAMINATION REQUIREMENTS AND ACCEPTANCE STANDARDS FOR CLASS 1 PWR PIPING AND VESSEL NOZZLE BUTT WELDS FABRICATED WITH UNS N06082 OR UNS W86182 WELD FILLER MATERIAL WITH OR WIHOUT APPLICATION OF LISTED MITIGATION ACTIVITIES Code Case N-770-2 provides alternative examination requirements and acceptance standards for Class 1 PWR Piping and Vessel Nozzle Butt Welds fabricated with UNS N060802 or UNS W86182 material with or without mitigation. In accordance with Table 1 of the Code Case, WF3 has twenty RCS cold leg unmitigated butt welds (Inspection Item B) that will be visually examined once per interval and volumetrically examined every second inspection period not to exceed 7 years.

The selection of these welds for examination meets the requirements of Code Case N-770-1 as mandated in 10 CFR 50.55a(g)(6)(ii)(F) with implementation of WF3 Relief Request W3-ISI-023.

7.2 Augmented Programs Note The scheduling of HELB RI-BER welds contained in the IDDEAL Software Suite and summarized in Table 7.1 has been completed based on preliminary evaluations meeting EPRI TR-1006937. When the BER compliance report is complete, the schedule and Table 7.1 requires verification. Verification and changes, if required, should be completed prior to completing examinations for the first period.

7.2.1 NUREG-0800, HELB RI-BER: Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants - Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping (Examination of High Energy Line Break Piping)

Ultrasonic examinations will be performed on each of the subject welds once per 10 year interval. Twenty-Four (24) welds are identified and are scheduled for examination to meet this augmented requirement.

7.2.2 RG 1.14 Reactor Coolant Pump Flywheel Integrity Surface and volumetric examinations of all four RCP flywheels shall be conducted in accordance with WF3 Technical Specification 6.5.7. Examinations are to be performed to

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 46 of 91 the extent possible through the access ports in the motor housings without disassembly of the motors. All 4 RCP flywheels shall be examined once every ISI 10-Year Interval and are scheduled to meet this augmented requirement.

7.2.3 W3- R&R-006 Volumetric Examination of Weld Overlay Including Upper 25% of Original Weld The full structural weld overlays (FSWO) on nine welds (9) shall be examined in accordance with Appendix VIII, of the 2007 Edition, 2008 Addenda of ASME Section XI, as modified by the Performance Demonstration Initiative (PDI) Program. The weld overlay examination volume A-B-C-D in Figure 2 of WF3-R&R-006 was added to the third ten-year inspection plan. 100% of the FSWO population shall be examined until PDI (Performance Demonstration Initiative) approved techniques are developed in accordance with the frequency defined in 5.1.3 of this document. Examination of these welds is scheduled to meet augmented requirements during the fourth ten-year interval.

7.2.4 NUREG-0612 Heavy Loads: Control of Heavy Loads at Nuclear Power Plants The special lifting devices addressed by this augmented examination are the Reactor Vessel Head Lift Rig, Upper Guide Structure Lift Rig and Core Support Barrel Lift Rig. The component load bearing welds shall be subject to a surface examination and performed once every ten years. Eleven (11) welds have been scheduled during this interval to meet this augmented requirement.

7.2.5 MRP-146 Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines Volumetric examination conducted in accordance with MRP-146 Rev.1. Examinations are to be performed every other refueling outage. Volumetric techniques and personnel qualifications shall be as specified in MRP-146 Revision 1, Section 2.4. WF3 has three weld locations (31-003,33-003 and 35-003) out of twelve weld locations applicable to examination and scheduled for examination in refueling outages 1R22, 1R24 and 1R28.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 47 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period B-A Pressure Retaining Welds in Reactor Vessel Reactor Vessel B-A B1.11 Circumferential Shell Volumetric 3 0 100% 03 0 0 3 Welds Reactor Vessel B-A B1.12 Volumetric 9 0 100% 9 0 0 9 Longitudinal Shell Welds Reactor Vessel B-A B1.21 Circumferential Head Volumetric 1 0 100% 1 0 0 1 Welds Reactor Vessel B-A B1.22 Volumetric 6 0 100% 6 0 0 6 Meridional Head Welds Reactor Vessel Shell-to-B-A B1.30 Volumetric 1 0 100% 1 0 0 1 Flange Weld Category Total 20 0(1) 20 0 0 20 Cumulative Percentage 20 0 0 0 100%

Note 1: Once NRC approval of Request for Alternative W3-ISI-006 is obtained for the Fourth Interval, these examinations can be Notes for Cat. B-A removed from the schedule.

B-B Pressure Retaining Welds in Vessels Other Than Reactor Vessels Pressurizer Shell-to-B-B B2.11 Head Welds Volumetric 2 2 100% 2 1 0 1 Circumferential Pressurizer Shell-to- 2 (1 ft. of one B-B B2.12 Volumetric 4 2 50%(1) 1 0 1 Head Welds Longitudinal weld per head

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 48 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Steam Generators (2)

B-B B2.40 (Primary Side) Volumetric 2 1 50% 1 0 1 0 Tubesheet-to-Head Weld Category Total 8 5 5 2 1 2 Cumulative Percentage 8 5 40% 60% 100%

Note 1: Two pressurizer heads each having two longitudinal welds. Section XI requires examination of one longitudinal weld per head.

Two of the longitudinal welds are selected for examination or 50%.

Notes for Cat. B-B Note 2: The examination may be limited to one vessel among the group of vessels performing a similar function. (Ref. Table IWB-2500-1, Examination Category B-B, Note 1).

B-D Full Penetration Welded Nozzles in Vessels Reactor Vessel Nozzle-B-D B3.90 Volumetric 6 6 100%(3) 6 0 0 6 to-Vessel Welds Reactor Vessel Nozzle B-D B3.100 Volumetric 6 6 100%(3) 6 0 0 6 Inside Radius Section Pressurizer Nozzle-to-B-D B3.110 Volumetric 5 5 100% 5 4 1 0 Vessel Welds Pressurizer Nozzle Volumetric or B-D B3.120 5 5 100% 5 4 1 0 Inside Radius Section(1)(2) Visual Steam Generators Volumetric or B-D B3.140 (Primary Side) Nozzle 6 6 100% 6 0 0 6 (1)(2) Visual Inside Radius Section Category Total 28 16 28 8 2 18

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 49 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Cumulative Percentage 28 28 50% 62.5% 100%

Note 1: 10 CFR 50.55a(b)(2)(xxi)(A) requires that the provisions of Table IWB-2500-1, Examination Category B-D, Code Items B3.120 and B3.140 in the 1998 Edition must be applied.

Note 2: 10 CFR 50.55a(b)(2)(xxi)(A) allows use of a visual examination with magnification that has a resolution sensitivity to detect a 1-Notes for Cat. B-D mil width wire or crack, utilizing the allowable flaw length criteria in Table IWB-3512-1, with a limiting assumption on the flaw aspect ratio (i.e., a/l=0.5), may be performed instead of an ultrasonic examination.

Note 3: Once NRC approval of Request for Alternative W3-ISI-006 is obtained for the Fourth Interval, these examinations can be removed from the schedule.

B-G-1 Pressure Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter Reactor Vessel Closure B-G-1 B6.10 Visual, VT-1 2 2 100%(5) 2 1 0 1 Head Nuts Reactor Vessel Closure B-G-1 B6.20 Volumetric 2 2 100%(4)(5) 2 1 0 1 Studs Reactor Vessel Threads B-G-1 B6.40 Volumetric 2 2 100%(5) 2 1 0 1 in Flange Reactor Vessel Closure B-G-1 B6.50 Visual, VT-1 2 2 100%(5) 2 1 0 1 Washers, Bushings B-G-1 B6.180 Pump Bolts and Studs Volumetric 64 16 25%(2)(4) 16 0 0 16 Pumps Flange Surface B-G-1 B6.190 when connection Visual, VT-1 64 16 25%(1)(3) 0 0 0 0 disassembled Pumps Nuts, Bushings, (3)

B-G-1 B6.200 Visual, VT-1 64 16 25% 0 0 0 0 and Washers Category Total 200 56 24 4 0 20

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 50 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period (6)

Cumulative Percentage 200 56 24 16.6% 16.6% 100%

Notes for Cat.B-G-1 Note 1: Not required unless disassembled. Examination includes 1 in. (25 mm) annular surface of flange surrounding each stud (Ref.

Table IWB-2500-1, Examination Category B-G-1, Note 6).

Note 2: Volumetric examination of bolting of heat exchangers, pumps, or valves may be conducted on one heat exchanger, one pump, or one valve among a group of heat exchangers, pumps, or valves that are similar in design, type, and function. In addition, when the component to be examined contains a group of bolted connections of similar design and size, such as flanged connections, the examination may be conducted on one bolted connection among the group. (Ref. Table IWB-2500-1, Examination Category B-G-1, Note 3).

Note 3: Visual examination of bolting for heat exchangers, pumps, and valves, is required only when the component is examined under Examination Category B-B, B-L-2, or B-M-2. Examination of a bolted connection is required only once during the interval. (Ref.

Table IWB-2500-1, Examination Category B-G-1, Note 4).

Note 4: When bolts or studs are removed for examination, surface examination meeting the acceptance standards of IWB-3515 may be substituted for volumetric examination. (Ref. Table IWB-2500-1, Examination Category B-G-1, Note 7).

Note 5: Bolting divided into 2 groups - 1 thru 27 and 28 thru 54. The first group scheduled in the first period and the second group scheduled in the third period.

Note 6: Cumulative Percentage is provided for information as some of these items are not required to be examined unless they are disassembled.

B-G-2 Pressure Retaining Bolting, 2 in. (50 mm) and Less in Diameter Pressurizer Bolts, Studs, B-G-2 B7.20 Visual, VT-1 1 1 100%(1) 0(2) 1 0 0 and Nuts Steam Generators Bolts, B-G-2 B7.30 Visual, VT-1 4 2 50%(1) 0(2) 0 0 0 Studs, and Nuts Heat Exchangers Bolts, (1) (2)

B-G-2 B7.40 Visual, VT-1 4 1 25% 0 0 0 0 Studs, and Nuts

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 51 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Piping Bolts, Studs, and (1) (3)

B-G-2 B7.50 Visual, VT-1 2 1 50% 0 0 0 0 Nuts Pumps Bolts, Studs, and (1) (2)

B-G-2 B7.60 Visual, VT-1 4 1 25% 0 0 0 0 Nuts B7.70 Valves Bolts, Studs, and B-G-2 Visual, VT-1 2 1 50%(1) 0(2)(4) 0 0 0 GR. 1 Nuts B7.70 Valves Bolts, Studs, and B-G-2 Visual, VT-1 4 1 25%(1) 0(2)(4) 0 0 0 GR. 2 Nuts B7.70 Valves Bolts, Studs, and B-G-2 Visual, VT-1 1 1 100%(1) 0(2)(4) 0 0 0 GR. 3 Nuts B7.70 Valves Bolts, Studs, and B-G-2 Visual, VT-1 4 1 25%(1) 0(2)(4) 0 0 0 GR. 4 Nuts Category Total 26 10(1) 0% 0 1 0 0 Cumulative Percentage 26 10 0 10% 10% 10%

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 52 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Note 1: Bolting is required to be examined only when a connection is disassembled or bolting is removed. (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 1).

Note 2: For components other than piping, examination of bolting is required only when the component is examined under Examination Notes for Cat.B-G-2 Category B-B, B-L-2, or B-M-2. Examination of bolted connection is required only once during the interval (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 2).

Note 3: The examination of flange bolting in piping systems may be limited to one bolted connection among a group of bolted connections that are similar in design, size, function, and service. Examination is required only when a flange is disassembled.

Examination of a bolted connection is required only once during the interval. (Ref. Table IWB-2500-1, Examination Category B-G-2, Note 3).

Note 4: Only one valve of each group of valves is required as outlined in B-M-2.

B-K Welded Attachments for Vessels, Piping, Pumps, and Valves Pressure Vessels B-K B10.10 Surface 4(4) 3 75%(1)(2) 3 1 0 2 Welded Attachments Pumps Welded B-K B10.30 Surface 4(4) 1 25%(3) 1 0 1 0 Attachments Category Total 8 4 4 4 1 1 2 Cumulative Percentage 8 4 0 25% 50% 100%

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 53 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Notes for Cat. B-K Note 1: 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. For single vessels, only one welded attachment shall be selected for examination.

Note 2: There is one welded attachment on the Reactor Vessel and one Pressurizer welded attachment, each requiring examination.

There are two Steam Generators each with one welded attachment, with 1 of 2 requiring examination. As such, 3 of the 4 welded attachments or 75% require examination.

Note 3: There are 4 Reactor Coolant Pumps, each with a welded attachment. 10% of the Reactor Coolant Pumps welded attachments associated with the component supports selected for examination under IWF-2510 shall be examined.

Note 4: Examination is required whenever component support member deformation, e.g., broken, bent, or pulled out parts, is identified during operation, refueling, maintenance, examination, or testing. (Ref. Table IWB-2500-1, Examination Category B-K, Note 6).

B-L-2 Pump Casings Pump Casing B-L-2 B12.20 Visual, VT-3 4 1 25%(1)(2) 0 0 0 0 (B-L-2)

Category Total 4 1 0 0 0 0 Cumulative Percentage 4 1 0 0% 0% 0%

Note 1: Examinations are limited to at least one pump in each group of pumps performing similar functions in the system.

Notes for Cat. B-L-2 Note 2: Examination is required only when a pump is disassembled for maintenance, or repair.

B-M-2 Valve Bodies B12.50 Valve Body, Exceeding (1)(2)

B-M-2 Visual, VT-3 2 1 50% 0 0 0 0 Gr. 1 NPS 4 (DN 100) (B-M-2)

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 54 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period B12.50 Valve Body, Exceeding (1)(2)

B-M-2 Visual, VT-3 4 1 25% 0 0 0 0 Gr. 2 NPS 4 (DN 100) (B-M-2)

B12.50 Valve Body, Exceeding B-M-2 Visual, VT-3 4 1 25%(1)(2) 0 0 0 0 Gr. 3 NPS 4 (DN 100) (B-M-2)

B12.50 Valve Body, Exceeding B-M-2 Visual, VT-3 4 1 0%(1)(2) 0 0 0 0 Gr. 4 NPS 4 (DN 100) (B-M-2)

B12.50 Valve Body, Exceeding B-M-2 Visual, VT-3 4 1 0%(1)(2) 0 0 0 0 Gr. 5 NPS 4 (DN 100) (B-M-2)

B12.50 Valve Body, Exceeding B-M-2 Visual, VT-3 4 1 0%(1)(2) 0 0 0 0 Gr. 6 NPS 4 (DN 100) (B-M-2)

Category Total 22 6 0 0 0 0 Cumulative Percentage 22 0 0% 0% 0%

Note 1: Examinations are limited to at least one valve within each group of valves that are of the same size, constructural design, and manufacturing method and that perform similar functions in the system. WF3 valves are grouped into 6 groups and one valve Notes for Cat. B-M-2 per group is required, if disassembled.

Note 2: Examination is required only when a valve is disassembled for maintenance, or repair.

B-N-1 Reactor Vessel Interior Reactor Vessel, Vessel B-N-1 B13.10 Visual, VT-3 1 3(1)(2) 100% 3 1 1 1 Interior (B-N-1)

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 55 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period (1)

Category Total 1 3 3 1 1 1 (3)

Cumulative Percentage 1 3 100% 100% 100%

Note 1: Examination of this item number is required each period. Therefore the number required during the interval is three times the total number of components. This is also reflected in the category total.

Notes for Cat B-N-1 Note 2: Areas to be examined shall include the spaces above and below the reactor core that are made accessible for examination by removal of components during normal refueling outages.

Note 3: Cumulative Percentages are provided for information only and are calculated for each period.

B-N-2 Welded Core Support Structures and Interior Attachments to Reactor Vessels Reactor Vessel (PWR)

Interior Attachments B-N-2 B13.50 Visual, VT-1 6 6 100% 6 0 0 6 Within Beltline Region (B-N-2)

Reactor Vessel (PWR)

Interior Attachments B-N-2 B13.60 Visual, VT-3 16 16 100% 16 0 0 16 Beyond Beltline Region (B-N-2)

Category Total 22 22(1) 100% 22 0 0 22(1)

Cumulative Percentage(2) 22 22 0% 0% 100%

Note 1: Examinations will be conducted during the third period to coincide with the 10 year Reactor Vessel inspection.

Notes for Cat. B-N-2 Note 2: Cumulative Percentage is provided for information only as these items can be deferred to the end of the interval.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 56 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period B-N-3 Removable Core Support Structures Reactor Vessel (PWR)

B-N-3 B13.70 Core Support Structure Visual, VT-3 1 1 100% 1 0 0 1 (B-N-3)(2)

Category Total 1 1(1) 100% 1 0 0 1(1)

Cumulative Percentage(3) 1 1 0% 0% 100%

Note 1: Examinations will be conducted during the third period to coincide with the 10 year Reactor Vessel inspection.

Notes for Cat. B-N-3 Note 2: The structure shall be removed from the reactor vessel for examination.

Note 3: Cumulative Percentage is provided for information only as these items can be deferred to the end of the interval.

B-O Pressure Retaining Welds in Control Rod Drive and Instrument Nozzle Housings 10%

Reactor Vessel (PWR)

Volumetric (1) peripheral B-O B14.20 Welds in Control Rod 112 12 12 0 0 12 or surface CRD Drive (CRD) Housings Housings Reactor Vessel (PWR)

Welds in In-Core Volumetric 10% ICI B-O B14.21 10 1 1 0 0 1 Instrumentation Nozzles or surface housings

>NPS 2 Category Total 112(1) 13 13 0 0 13

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 57 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Note 1: There are 28 peripheral CRDHs for a total of 112 welds. 10% of 28 peripheral CRDHs rounded up is 3. The 3 CRDHs each Notes for Cat. B-O have 4 welds which total 12 welds for examinations.

C-A Pressure Retaining Welds in Pressure Vessels Pressure Vessels Shell C-A C1.10 Volumetric 2 0 0%(1)(2) 0 0 0 0 Circumferential Welds Pressure Vessels Head C-A C1.20 Volumetric(5) 2 1 50%(1)(3) 1 0 0 1 Circumferential Welds Pressure Vessels C-A C1.30 Volumetric (5) 4 1 25%(1)(4) 1 1 0 0 Tubesheet-to-Shell Weld Category Total 8 2 2 1 0 1 Cumulative Percentage 8 2 50% 50% 100%

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 58 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Notes for Cat. C-A Note 1: 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 and includes essentially 100% of the weld length. (Ref. Table IWC-2500-1, Examination Category C-A, Notes 2 & 3).

Note 2: There is one circumferential shell to flange weld on each of the two Shutdown Cooling Heat Exchangers, implementation of Code Case N-716-1 evaluated the Shutdown Cooling Heat Exchangers as Low Safety Significant (LSS) therefore no examinations are required or 0%.

Note 3: There is one circumferential head weld on each Steam Generator (two total), therefore 1 of 2 head welds is required to be examined or 50%.

Note 4: There is one tubesheet-to-shell welds on each Steam Generator (two total), and one tubesheet-to-shell weld on each Shutdown Cooling Heat Exchanger (two total). Implementation of Code Case N-716-1 evaluated the Shutdown Cooling Heat Exchangers as LSS, therefore 1 of 4 tubesheet-to-shell welds (one on one Steam Generator) are required to be examined for 25%.

Note 5: For welds in vessels with nominal wall thickness of 0.2 in. (5 mm) or less, a surface examination may be applied in lieu of a volumetric examination. (Ref. Table IWC-2500-1, Examination Category C-A, Note 1).

C-B Pressure Retaining Nozzle Welds in Vessels Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 Surface and C-B C2.21 mm) Nominal Thickness 2 1 50%(1)(2) 1 0 0 1 volumetric Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Weld Nozzles Without Reinforcing Plate in Vessels > 1/2 in. (13 C-B C2.22 Volumetric 2 1 50%(1)(3) 1 0 0 1 mm) Nominal Thickness Nozzle Inside Radius Section

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 59 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Nozzles With Reinforcing Plate in Vessels > 1/2 in.

(13 mm) Nominal (1)(4)

C-B C2.31 Surface 8 0 0% 0 0 0 0 Thickness Reinforcing Plate Welds to Nozzle and Vessel Nozzles With Reinforcing Plate in Vessels > 1/2 in.

(13 mm) Nominal Thickness Nozzle-to-C-B C2.33 Visual, VT-2 4 0 0%(1)(5) 0 0 0 0 Shell (Nozzle to Head or Nozzle to Nozzle) Welds Inside of Vessel is Inaccessible Category Total 16 2 2 0 0 2 Cumulative Percentage 16 2 0% 0% 100%(6)

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 60 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Notes for Cat. C-B Note 1: In case of multiple vessels of similar design, size, and service, the required examinations may be limited to one vessel or distributed among the vessels. (Ref. Table IWC-2500-1, Examination Category C-B, Note 4).

Note 2: There is one Feedwater nozzle on each Steam Generator, therefore 1 of 2 nozzles is required to be examined or 50%.

Note 3: There is one Feedwater nozzle inner radius on each Steam Generator; therefore, 1 of 2 inner radii is required to be examined or 50%.

Note 4: There are four nozzles on each Shutdown Cooling Heat Exchanger; implementation of Code Case N-716-1 evaluated the Shutdown Cooling Heat Exchangers as LSS, therefore 0 of 8 nozzles are required to be examined or 0%.

Note 5: There are two nozzles on both Shutdown Cooling Heat Exchanger, implementation of Code Case N-716-1 evaluated the Shutdown Cooling Heat Exchangers as LSS, therefore 0 of 4 nozzles are required to be examined or 0%.

Note 6: After implementation of Code Case N-716-1 which does not require any examinations of the Shutdown Cooling Heat Exchangers, there are only two feedwater nozzles in each item number. Per IWC-2411(a) If there are less than three items or welds to be examined in any two periods, or in any one period if there is only one item or weld, in lieu of the percentage requirements of Table IWC-2411-1. Since only one nozzle is required to be examined (this includes both the nozzle-to-vessel weld and the inner radius section) it is scheduled in the third period.

C-C Welded Attachments for Vessels, Piping, Pumps, and Valves Pressure Vessels C-C C3.10 Surface 8 1 12.5% 1(1)(2)(4) 0 1 0 Welded Attachments(1)

Piping Welded C-C C3.20 Surface 216 6 2.7% 6(3)(4) 2 2 2 Attachments Pump Welded C-C C3.30 Surface 9 1 10% 1(3)(4) 0 0 1 Attachments Category Total 234 8 8 2 3 3 Cumulative Percentage 234 8 25% 62.5% 100%

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 61 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Notes for Cat. C-C Note 1: 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. For single vessels, only one welded attachment shall be selected for examination. (Ref.

Table IWC-2500-1, Examination Category C-C, Note 4).

Note 2: There are four welded attachments on each Steam Generator; therefore 1 of 8 welded attachments is required to be examined for 12.5%.

Note 3: 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. (Ref. Table IWC-2500-1 Examination Category C-C, Note 5). There are a total of 682 piping supports, 108 of those supports have been selected for examination under IWF-2510. Of those supports selected for examination only 56 have welded attachments, 10% or 6 would require examination.

Note 4: Examination is required whenever component support member deformation, e.g., broken, bent, or pulled out parts, is identified during operation, refueling, maintenance, examination, or testing. (Ref. Table IWC-2500-1, Examination Category C-C, Note 6).

D-A Welded Attachments for Vessels, Piping, Pumps, and Valves Pressure Vessel Welded D-A D1.10 Visual, VT-1 2 1 50% 1(1)(2)(3) 0 1 0 Attachments Piping Welded D-A D1.20 Visual, VT-1 233 24 10% 24(1)(3) 8 8 8 Attachments Category Total 235 25 25 8 9 8 Cumulative Percentage 235 25 32% 68% 100%

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 62 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Notes for Cat. D-A Note 1: Selected samples of welded attachments shall be examined each inspection interval. All welded attachments selected for examination shall be those most subject to corrosion, as determined by the Owner, such as the welded attachments of the Service Water or Emergency Service Water systems. For multiple vessels of similar design, function and service, the welded attachments of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. The attachment selected for examination on one of the multiple vessels or the single vessel, as applicable, shall be an attachment under continuous load during normal system operation, or an attachment subject to a potential intermittent load (seismic, water hammer, etc.) during normal system operation if an attachment under continuous load does not exist. For welded attachments of piping, pumps, and valves, a 10% sample shall be selected for examination. This percentage sample shall be proportional to the total number of nonexempt welded attachments connected to the piping, pumps and valves in each system subject to examination. (Reference Table IWD-2500-1, Category D-A, Note 3).

Note 2: There are two Component Cooling Water Heat Exchangers with welded attachments, considered a group, therefore the welded attachments on one Component Cooling Water Heat Exchanger is scheduled for 50%.

Note 3: Examination is required whenever component support member deformation, e.g., broken, bent, or pulled out parts is identified during operation, refueling, maintenance, examination, or testing. (Ref. Table IWD-2500-1, Examination Category D-A, Note 4).

F-A Supports Class 1 Piping Supports -

F-A F1.10A Visual, VT-3 145 37 (1) 37 12 10 15 One Directional Class 1 Piping Supports -

F-A F1.10B Visual, VT-3 32 8 (1) 10 3 3 4 Multi-directional Class 1 Piping Supports -

F-A F1.10C Visual, VT-3 38 10 (1) 11 4 3 4 Thermal Movement Category F-A, Item F1.10 Total 215 55 25% 58 19 16 23 Class 2 Piping Supports -

F-A F1.20A Visual, VT-3 302 46 (1) 47 13 14 20 One Directional Class 2 Piping Supports -

F-A F1.20B Visual, VT-3 239 36 (1) 39 13 9 17 Multi-directional

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 63 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Class 2 Piping Supports -

F-A F1.20C Visual, VT-3 141 22 (1) 22 7 6 9 Thermal Movement Category F-A, Item F1.20 Total 682 104 15% 108 33 29 46 Class 3 Piping Supports -

F-A F1.30A Visual, VT-3 304 31 (1) 33 10 10 13 One Directional Class 3 Piping Supports -

F-A F1.30B Visual, VT-3 545 55 (1) 56 11 21 24 Multi-directional Class 3 Piping Supports -

F-A F1.30C Visual, VT-3 32 4 (1) 5 0 1 4 Thermal Movement Category F-A, Item F1.30 Total 881 90 10% 94 21 32 41 Supports other than F-A F1.40 Piping Supports (Class Visual, VT-3 40 18 45% 18(2) 4 7 7 1,2,3, and MC)

Category F-A, Item F1.40 Total 40 18 45% 18(2) 4 7 7 Category Total 1818 267 278 77 84 117 Cumulative Percentage 1818 267 28.8% 60.2% 104%

Notes for Cat. F-A Note 1: The total percentage sample shall be comprised of supports from each system, where the individual sample sizes are proportional to the total number of non-exempt supports of each type and function within each system. (Ref. Table IWF-2500-1, Examination Category F-A, Note 2).

Note 2: 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. (Ref. Table IWF-2500-1, Examination Category F-A, Note 3).

R-A Risk Informed Piping Welds - Code Case N-716-1

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 64 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period LSS Welds that have not (1)

R-A R0.00 been Evaluated for N/A 1440 0 0% 0 0 0 0 Degradation Welds Subject to R-A R1.11 Volumetric 340 66 19.7%(1) 66 20 19 27 Thermal Fatigue Welds Subject to R-A R1.11/16 Thermal Fatigue and Volumetric 8 2 25% 2 0 1 1 IGSCC Welds Subject to Primary R-A R1.15 Water Stress Corrosion Volumetric 16 0 0%(1) 0 0 0 0 Cracking (PWSCC)

Welds not Subject to a R-A R1.20 Volumetric 569 27 4.5%(1) 27 8 10 9 Damage Mechanism Category Total 2373 95 (1) 95 28 30 37 Cumulative Percentage 2373 95 29.4% 61.0% 100%

Notes for Cat. R-A Note 1: WF3 has initiated Risk-Informed ISI (RI-ISI) contained in ASME Code Case N-716-1, approved in Regulatory Guide 1.147 Revision 17 and listed in Table 2.2-1.

A-600 Code Case N-722-1 PWR Class 1 Components Containing Alloy 600/82/182(1)

Pressurizer Spray Every B9.21 B-J nozzle-to-pipe Visual, VE 2 2 Refueling 12 4 4 4 (B15.150) connections Outage Piping Cold Leg A-600 B15.205 Visual, VE 12 12 100% 12 6 3 3 Instrument Connections B9.11 Piping Cold leg Full (2)

B-J Visual, VE 8 8 100% 8 0 0 0 (B15.215) Penetration Welds

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 65 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period B9.21 Piping Cold leg Full (2)

B-J Visual, VE 6 6 100% 6 0 0 0 (B15.215) Penetration Welds R1.11 Piping Cold leg Full (2)

R-A Visual, VE 6 6 100% 6 0 0 0 (B15.215) Penetration Welds Category Total 34 14(2) 24(3) 10 7 7 Notes for Cat. A- Note 1: Frequency of examination, examination method, and personnel qualifications are in accordance with Code Case N-722-1, 600/N-722-1 Table 1 as required by 10 CFR 50.55a(g)(6)(ii(E)).

Note 2: VE examinations are not scheduled because these locations are volumetrically examined under Code Case N-770. This meets Note 5 of Code Case N-722-1.

Note 3: This total only represents visual examinations not replaced by volumetric examinations.

A-600 Code Case N-729-4 PWR Reactor Vessel Upper Head (RVUH)(1)

RVUH with nozzles and partial-penetration welds A-600 B4.30 Visual, VE 3 3 100% 3(2) 0 3 3 of PWSCC-resistant materials RVUH Nozzles and partial-penetration welds Volumetric A-600 B4.40 3 3 100% 3(3) 0 3(4) 0 of PWSCC-resistant &/or Surface materials in head Category Total 6 6 6 0 6 3 Notes for Cat. A- Note 1: These examinations are in accordance with Code Case N-729-4, Table 1 as required by 10 CFR 50.55a(g)(6)(ii)(D).

600/N-729-4 Note 2: Every third refueling outage, or 5 calendar years, whichever is less.

Note 3: Not to exceed one Inspection Interval (normally 10 calendar years).

Note 4: Examination required prior to 2023. Includes examination of the Vent Nozzle J-Groove Weld (02-S-01X1).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 66 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Code Case N-770-2 Class 1 PWR Pressure Retaining Dissimilar Metal Piping and Vessel Nozzle Butt A-600 Welds Containing Alloy 82/182(1)

Volumetric every N-770-1 Category B - second Unmitigated butt weld at Volumetric 8 8 inspection 16 8 0 8 N-770-2 B9.11 cold leg operating period not to Item B temperatures > 525°F exceed 7 and <580°F years(1)

Visual once Visual(2) 8 8 8(2) 0 0 0 per Interval Volumetric every N-770-1 Category B - second Unmitigated butt weld at Volumetric 6 6 inspection 12 6 0 6 N-770-2 B9.21 cold leg operating period not to Item B temperatures >525°F exceed 7 and < 580°F years(1)

Visual once Visual(2) 6 6 6(2) 0 0 0 per Interval Volumetric every N-770-1 Category B - second Unmitigated butt weld at Volumetric 6 6 inspection 12 6 0 6 N-770-2 R1.11 cold leg operating period not to Item B temperatures >525°F exceed 7 (1) and < 580°F years (2) Visual once (2)

Visual 6 6 6 0 0 0 per Interval

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 67 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period (2) (3)

Category Total 40 20 40 20 0 20 Notes for Cat. A- Note 1: Frequency of examination, examination method, and personnel qualifications are in accordance with Code Case N-770-2, Table 600/N-770-2 1 as required by 10 CFR 50.55a(g)(6)(ii)(F).

Note 2: VE examinations are not scheduled because these locations are volumetrically examined. This meets Note 3 of Code Case N-770-2.

Note 3: This total only represents visual examinations not replaced by volumetric examinations Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants -

NUREG-0800, Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of HELB RI-BER Piping (Examination of High Energy Line Break Piping) Examination of High Energy Line Break Piping Need to verify once the updated RI-BER compliance report is completed R1.11 Welds subject to thermal R-A (FW) Volumetric 9 5 55% 5 0 4 1 fatigue RC 2 R1.11 Welds subject to thermal R-A (SI) Volumetric 28 9 32% 9 4 1 4 fatigue RC 2 R1.18 R-A (FW) Welds subject to a FAC Volumetric 8 0 0% 0 0 0 0 RC 4 R1.20 Welds not subject to a R-A (FW) Volumetric 29 2 6.8% 2 0 1 1 Degradation Mechanism RC 4 R1.20 Welds not subject to a R-A (MS) Volumetric 47 8 17% 8 3 4 1 Degradation Mechanism RC 4

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 68 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period R1.20 Welds not subject to a R-A (SI) Volumetric 6 0 0% 0 0 0 0 Degradation Mechanism RC 4 R1.11 Welds subject to thermal R-A (SI) Volumetric 8 0 0% 0 0 0 0 fatigue RC 5 R1.11 Welds subject to thermal R-A (SI) Volumetric 20 0 0% 0 0 0 0 fatigue RC 6 Category Total 155 24 24(1) 7 10 7 Notes for NUREG- Note 1: Examinations required once per 10 year interval.

0800 HELB RG 1.14 Reactor Coolant Pump Flywheel Integrity Volumetric/

RG RG 1.14 RCP Flywheels 4 4 100% 4 1 1 2 Surface Category Total 4 4 100% 4(1) 1 1 2 Notes for RG 1.14 Note 1: Inservice inspection of each operating (4) reactor coolant pump flywheel shall be performed once every ISI ten year interval.

W3-R&R-006 Volumetric Examinations of Weld Overlay Program Including Upper 25% of Original Weld Welds Subject to R-A R1.11 Volumetric 1 1 100% 1 1 0 0 Thermal Fatigue

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 69 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Welds Subject to R-A R1.11/15 Thermal Fatigue and Volumetric 2 2 100% 2 2 0 0 PWSCC Welds Subject to Primary R-A R1.15 Water Stress Corrosion Volumetric 6 6 100% 3 3 0 3 Cracking (PWSCC)

Category Total 9 9 9 62) 0 31)

Notes for Note 1: Perform examinations once every 9 years until Performance Demonstration Initiative (PDI) approved techniques are developed.

W3-R&R-006 Note 2: Perform examinations once every 10 years until PDI approved techniques are developed.

NUREG-0162 Control of Heavy Loads at Nuclear Power Plants Heavy Loads Load Bearing Welds X-Aug N/A Surface 11 11 100% 11 6 3 2 Lifting Devices Category Total 11 11 11(1) 6 3 2 Notes for NUREG-Note 1: Surface examination on critical welds shall be performed once every ten years.

0612 Heavy Loads Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch MRP-146 Lines Circumferential Welds in Non-Isolable RCS R-A R1.11 Volumetric 3 9(1)(2) 100% 9 3 3 3 Branch Lines Susceptible to Thermal Fatigue Category Total 3 9(1)(2) 9(1)(2) 3 3 3

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 70 of 91 Table 7.1 Waterford Unit No.3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components in Percentage Scheduled Examined Examined Examined Number During Item No. Required During the in First in Second in Third Interval Interval Period Period Period Note 1: Volumetric examinations shall be performed once every other refueling outage.

Notes for MRP-146 Note 2: MRP-146 Revision 1 requires multiple examinations within the interval for these components. The number of exams required per period has been adjusted to reflect these multiple examinations.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 71 of 91 8.0 Inservice Inspection Data Management Software The Inservice Inspection Data Management software is comprised of a series of program modules assembled in a comprehensive software package entitled the Iddeal Software Suite.

This software suite is NETWORK based within the Entergy system and is accessible by any computer within Entergy.

The Iddeal Software Suite is a quality level B software. It is the Controlled copy of the ISI Schedule. The software has limited access to protect the contents of the databases. Access to the software is limited through the use of a user defined password controlled via Central Engineering Procedure CEP-COS-100.

Access to the Iddeal Software Suite is obtained through the Windows Citrix' icon entitled the Iddeal Software Suite. There are two version of the Iddeal Software Suite being utilized.

The following programs are being used for the version 8 IDDEAL SOFTWARE SUITE:

IDDEAL - IDDEAL is used to track and progress completion of outage examinations.

Generates various reports to status examinations.

CERTWORKS - Maintains and tracks the personnel certifications of inspectors.

EQUIPWORKS - Maintains and tracks the NDE equipment certifications.

SNUBWORKS - Used to track all work associated with plant snubbers.

The following modules are being used for the version 9 IDDEAL SOFTWARE SUITE.

COMPONENTS - is used to identify all of the components that exist in the plant and being considered for the ISI Program.

LIBRARIES - is used to identify all of the requirements that could be applied to the components.

PROGRAMS - is used to create and maintain program manuals, scheduling, and completion of examinations for credit to those programs.

ACTIVITIES - is used to maintain and track examinations within an outage or other activity.

EQUIPMENT - is used to maintain and track the equipment used in performing the examinations (i.e. calibration blocks)

The Programs Module contains the WF3 information necessary to perform ISI and Augmented Programs. Maintenance of the Module is controlled via Central Engineering Procedure CEP-COS-0110. All changes to data in the WF3 ISI Database must be made in accordance with CEP-COS-0110.

The controlled Iddeal Software Suite (both versions) houses the information below which meets the requirements of ASME Section XI, IWA-2420(b).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 72 of 91 Identification of the components selected for examination and test, including successive exams from prior periods; The Code requirements by examination category and item number for each component and the examination or test to be performed and the extent of the examination or test; Identification of drawings showing items that require examination; List of examination procedures; Description of alternative examinations and identification of components to be examined using alternative methods; Identification of calibration blocks used for ultrasonic examination of components.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 73 of 91 9.0 Examination Evaluation Criteria Evaluation of reportable indications detected during the inservice inspection of components and component supports will be performed in accordance with Article IWA-3000 of ASME Section XI, 2007 Edition, 2008 Addenda and CEP-NDE-3000. Specific evaluation requirements for Class 1, 2, and 3 components are included in IWB-3000, IWC-3000, and IWD-3000, respectively. Specific evaluation requirements for Class 1, 2, and 3 component supports are included in IWF-3000. Indications detected may be evaluated by other nondestructive methods, where practical, to assist in the determination of flaw characteristics (e.g. size, shape, location, orientation, etc.) before final disposition is made.

Evaluation of relevant indications detected during the augmented examination of components and component supports will be performed in accordance with this section and the applicable document that governs each augmented examination.

9.1 Successive Inspections 9.1.1 Successive Inspections on Class 1 components (excluding Category R-A, see 9.1.4) will be performed in accordance with IWB-2420 which includes the following criteria in IWB-2420(b),

(c), (d) and (e):

(b) If a component is accepted for continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the Inspection Program of IWB-2400. Alternatively, acoustic emission may be used to monitor growth of existing flaws in accordance with IWA-2234.

(c) If the reexaminations required by IWB-2420(b) reveal that the flaws or relevant conditions remain essentially unchanged for three successive inspection periods, the component examination schedule may revert to the original schedule of successive inspections.

(d) For steam generator tubing, the successive examinations shall be governed by the plant Technical specifications.

(e) If welded attachments are examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of Table IWB-3410-1, successive examinations shall be performed, if determined necessary, based on an evaluation by the Owner.

9.1.2 Successive Inspections on Class 2 components (excluding Category R-A see 9.1.4) will be performed in accordance with IWC-2420 which includes the following criteria in IWC-2420(b),

(c) and (d):

(b) If a component is accepted for continued service in accordance with IWC-3122.3 or IWC-3132.3, the areas containing flaws or relevant conditions shall be reexamined during the next inspection period listed in the schedule of the Inspection Program of

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 74 of 91 IWC-2400. Alternatively, acoustic emission may be used to monitor growth of existing flaws in accordance with IWA-2234.

(c) If the reexaminations required by IWC-2420(b) reveal that the flaws or relevant conditions remain essentially unchanged for the next inspection period, the component examination schedule may revert to the original schedule of successive inspections.

(d) If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of Table IWC-3410-1 successive examinations shall be performed, if determined necessary, based on an evaluation by the Owner.

9.1.3 As an alternative to 9.1.1 and 9.1.2 for Class 1 and 2 vessels, the requirements of Code Case N-526 may be used to eliminate successive examinations. The requirements of Code Case N-526 are outlined below.

(a) The flaw is characterized as subsurface in accordance with Figure 1 of Code Case N-526.

(b) The NDE technique and evaluation that detected and characterized the flaw, with respect to both sizing and location, shall be documented in the flaw evaluation report.

(c) The vessel containing the flaw is acceptable for continued service in accordance with IWB-3600, and the flaw is demonstrated acceptable for the intended service life of the vessel.

9.1.4 For Examination Category R-A, Class 1 and 2 piping welds, the requirements of Code Case N-716-1 shall be used for successive examinations.

(a) If Examination Category R-A Piping Welds are accepted for continued service by analytical evaluation in accordance with IWB-3132.3 or IWB-3142.4, the areas containing the flaws or relevant conditions shall be reexamined during the next three inspection periods..

(b) If Class 2, 3, or non Class vessels, pumps or valves, and pressure-retaining bolting are accepted for continued service by analytical evaluation in accordance with IWC-3122.3 or IWC-3132.3, as applicable, the areas containing flaws or relevant conditions shall be reexamined during the next inspection period.

(c) If the reexaminations required by (a) reveal that the flaws or relevant conditions remain essentially unchanged, the examination schedule shall revert to the original schedule of successive inspections.

(d) If the reexaminations required by (b) reveal that the flaws or relevant conditions remain essentially unchanged during the successive inspection period, the examination scheduled shall revert to the original schedule of successive inspections.9.1.5 Successive Inspections on Class 3 components will be performed in accordance with IWD-2420 which includes the following criteria in IWD-2420(b), (c) and (d):

(b) If components are accepted for continued service by evaluation in accordance with IWD-3000, the areas containing flaws or relevant conditions shall be reexamined during

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 75 of 91 the next inspection period listed in the schedule of the Inspection Program of IWD-2400.

(c) If the reexaminations required by IWD-2420(b) reveal that the flaws or relevant conditions remain essentially unchanged for the next inspection period, the component examination schedule may revert to the original schedule of successive inspections.

(d) If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of IWD-3000, successive examinations shall be performed, if determined necessary based on an evaluation by the Owner.

9.1.6 Successive Inspections on Class 1, 2 and 3 component supports will be performed in accordance with IWF-2420 which includes the following criteria in IWF-2420(b) and (c):

(b) When a component support is accepted for continued service in accordance with IWF-3112.2 or IWF-3122.2, the component support shall be reexamined during the next inspection period listed in the schedule of the Inspection Programs of IWF-2410.

(c) When the examinations required by IWF-2420(b) do not require additional corrective measures during the next inspection period, the inspection schedule may revert to the requirements of IWF-2420(a).

9.1.7 Successive Inspections on Augmented examinations will be performed in accordance with its Augmented Program requirements document.

9.1.8 Plant specific successive Inspections required to be performed in accordance with IWB-2420, IWC-2420, IWD-2420 and IWF-2420 are identified in the IDDEAL database module by an v (scheduled successive inspection) or an V (completed successive inspection).

9.2 Additional Examinations 9.2.1 Additional examinations of Examination Category R-A welds shall be determined in accordance with 9.2.7.

All other Section XI additional examinations shall be determined in accordance with 9.2.2, 9.2.3, 9.2.4, 9.2.5 as applicable, or 9.2.6 may be used as an alternative. Additional examinations for augmented components and component supports will be performed in accordance the applicable document that governs each augmented examination.

9.2.2 Additional examinations on Class 1 components will be performed in accordance with IWB-2430 which includes the following criteria in IWB-2430(a), (b), (c) (d) and (e):

(a) Examinations performed in accordance with Table IWB-2500-1, except for Examination Category B-P, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 76 of 91 number of welds, areas, or parts2 included in the inspection item3 equal to the number of welds, areas, or parts included in the inspection item that were scheduled to be performed during the present inspection period. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

(b) If the additional examinations required by IWB-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1, the examinations shall be further extended to include additional examinations during the current outage.

These additional examinations shall include the remaining number of welds, areas, or parts of similar material and service subject to the same type of flaws or relevant conditions.

(c) For the inspection period following the period in which the examinations of IWB-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with IWB-2400.

(d) For steam generator tubing, additional examinations shall be governed by plant Technical Specifications.

(e) If welded attachments are examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of Table IWB-3410-1 additional examinations shall be performed, if determined necessary, based on an evaluation by the Owner.

9.2.3 Additional examinations on Class 2 components will be performed in accordance with IWC-2430 which includes the following criteria in IWC-2430(a), (b), (c) and (d):

(a) Examinations performed in accordance with Table IWC-2500-1, except for Examination Category C-H, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-3410-1 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional number of welds, areas, or parts4 included in the inspection item5 equal to 20% of the number of welds, areas, or parts included in the inspection item that are scheduled to be performed during the interval. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

2 Welds, areas or parts are those described or intended in a particular inspection item of Table IWB-2500-1.

3 An inspection item, as listed in Table IWB-2500-1, may comprise a number of welds, areas, or parts of a component required to be examined in accordance with the inspection plan and schedule (IWA-2420).

4 Welds, areas or parts are those described or intended in a particular inspection item of Table IWC-2500-1.

5 An inspection item, as listed in Table IWC-2500-1, may comprise a number of welds, areas, or parts of a component required to be examined in accordance with the inspection plan and schedule (IWA-2420).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 77 of 91 (b) If the additional examinations required by IWC-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of Table IWC-3410-1, the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining number of welds, areas, or parts of similar material and service subject to the same type of flaws or relevant conditions.

(c) For the inspection period following the period in which the examinations of IWC-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with IWC-2400.

(d) If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of Table IWC-3410-1 additional examinations shall be performed, if determined necessary, based on an evaluation by the Owner.

9.2.4 Additional examinations on Class 3 components will be performed in accordance with IWD-2430 which includes the following criteria in IWD-2430(a), (b), (c) and (d):

(a) Examinations performed in accordance with Table IWD-2500-1, except for Examination Category D-B, that reveal flaws or relevant conditions exceeding the acceptance standards of IWD-3000 shall be extended to include additional examinations during the current outage. The additional examinations shall include an additional number of welds, areas, or parts6 included in the inspection item7 equal to 20% of the number of welds, areas, or parts included in the inspection item that are scheduled to be performed during the interval. The additional examinations shall be selected from welds, areas, or parts of similar material and service. This additional selection may require inclusion of piping systems other than the one containing the flaws or relevant conditions.

(b) If the additional examinations required by IWD-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of IWD-3000, the examinations shall be further extended to include additional examinations during the current outage. The extent of the additional examinations shall be determined by the Owner based upon an engineering evaluation of the root cause of the flaws or relevant conditions. The Owners corrective actions shall be documented in accordance with IWA-6000.

(c) For the inspection period following the period in which the examinations of IWD-2430(a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with IWD-2400.

(d) If welded attachments are examined as a result of identified component support deformation, and the results of these examinations exceed the acceptance standards of Table IWD-3000, additional examinations shall be performed, if determined necessary, based on an evaluation by the Owner.

6 Welds, areas or parts are those described or intended in a particular inspection item of Table IWD-2500-1.

7 An inspection item, as listed in Table IWD-2500-1, may comprise a number of welds, areas, or parts of a component required to be examined in accordance with the inspection plan and schedule (IWA-2420).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 78 of 91 9.2.5 Additional examinations on Class 1, 2 and 3 component supports will be performed in accordance with IWF-2430 which includes the following criteria in IWF-2430(a), (b), (c) and (d):

(a) Component support examinations performed in accordance with Table IWF-2500-1 that reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, and that require corrective measures, shall be extended, during the current outage, to include the component supports immediately adjacent to the component supports for which corrective measures are required. The additional examinations shall be extended, during the current outage, to include additional supports within the system, equal in number and of the same type and function as those scheduled for examination during the inspection period.

(b) When the additional examinations required by IWF-2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, and that require corrective measures, the examinations shall be further extended to include additional examinations during the current outage. These additional examinations shall include the remaining component supports within the system of the same type and function.

(c) When the additional examinations required by IWF-2430(b) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, and that require corrective measures, the examinations shall be extended, during the current outage, to include all nonexempt supports potentially subject to the same failure modes that required corrective measures in accordance with IWF-2430(a) and (b). Also, these additional examinations shall include nonexempt component supports in other systems when the support failures requiring corrective actions indicate non-system-related support failure modes.

(d) When the additional examinations required by IWF-2430(c) reveal flaws or relevant conditions exceeding the acceptance standards of IWF-3400, and that require corrective measures, the Owner shall examine, during the current outage, those exempt component supports that could be affected by the same observed failure modes and could affect nonexempt components.

9.2.6 The following criteria of Code Case N-586-1 may be applied for the performance of additional examinations on Class 1, 2 or 3 components or component supports.

(a) An engineering evaluation shall be performed. Topics to be addressed in the engineering evaluation shall include:

1) A determination of the root cause of the flaws or relevant conditions.

2) An evaluation of applicable service conditions and degradation mechanisms to establish that the affected welds, areas, or supports will perform their intended safety functions during subsequent operation.

3) A determination of which additional welds, areas, or supports could be subject to the same root cause conditions and degradation mechanisms. This may require the

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 79 of 91 inclusion of piping systems other than the one containing the original flaws or relevant conditions.

(b) Additional examinations shall be performed on those welds, areas, or supports subject to the same root cause conditions and degradation mechanisms. No additional examinations are required if the engineering evaluation concludes that either:

1) There are no additional welds, areas, or supports subject to the same root cause conditions, or

2) No degradation mechanism exists.

(c) Any required additional examinations shall be performed during the current outage.

(d) The engineering evaluation shall be retained in accordance with IWA-6000.

9.2.7 Additional examinations on Category R-A components shall be performed in accordance with Code Case N-716-1 6(b)(1), (2), and (3).

(b)(1) Examination performed, excluding Examination Categories B-P and C-H, that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1, Table IWC-3410-1, or IWD-3000, as applicable, shall be extended to include a first sample of additional examinations during the current outage.

(a) The inspection item to be examined in the first sample of additional examinations shall include HSS inspection items with the same postulated degradation mechanism in systems whose materials and service conditions are similar to the element that exceeded the acceptance standards.

(b) The number of examinations required is the number of HSS inspection items with the same postulated degradation mechanism scheduled for the current inspection period. If there are not enough HSS inspection items to equal this number, the Owner shall include remaining HSS inspection items and LSS inspection items within the Class 1, 2, and 3 systems subject to examination in accordance with Section XI, up to and including this number that are subject to the same degradation mechanism. This might require additional analysis of LSS components (e.g., degradation mechanism evaluation).

(b)(2) If the additional examination required by (b)(1) reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1, Table IWC-3410-1, or IWD-3000 as applicable, the examinations shall be extended to include a second sample of additional examinations during the current outage.

(a) The second sample of additional inspection items to be examined shall include remaining HSS inspection items subject to the same degradation mechanism.

(b) The Owner shall also examine LSS inspection items identified with the same postulated damage mechanism within the Class 1, 2, and 3 systems subject to examination in accordance with Section XI or document the basis for their exclusion.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 80 of 91 This might require additional analysis of LSS components (e.g., degradation mechanism evaluation).

(b)(3) For the inspection period following the period in which the examination of (b)(1) and (b)(2) were completed the examinations shall be performed as originally scheduled in accordance with IWB-2400.

9.2.8 Plant specific additional Examinations required to be performed in accordance with IWB-2430, IWC-2430, IWD-2430 and IWF-2430 are identified in the IDDEAL database module by an a (scheduled additional examination) or an A (completed additional examination).

10.0 Risk-Informed ISI Program WF3 has initiated Risk-Informed ISI (RI-ISI) contained in ASME Code Case N-716-1, approved in Regulatory Guide 1.147 Revision 17 and listed in Table 2.2-1.

10.1 RISK-INFORMED ISI PROCESS The process used to develop the RI-ISI Program conformed to the methodology described in Code Case N-716-1 and consisted of the following steps:

Safety Significance is identified by code case Degradation Mechanism Assessment Piping Weld Selection Risk Impact Assessment Implementation Program Periodic Update 10.1.1 N-716-1 Safety Significance Determination The piping and instrumentation diagrams and additional plant information including the existing plant ISI Program were used to define the N-716-1 program scope.

Components are assigned safety-significance categories, which are used to determine the treatment requirements.

(a) High safety-significant (HSS) components shall include the following:

(1) Class 1 portions of the reactor coolant pressure boundary (RCPB), with the exception of components described in (a) or (b) below:

(a) 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.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 81 of 91 (b) The component is or can be isolated from the reactor coolant system by two valves in series (both closed, both open, or one closed and the other open). Each open valve must be capable of automatic actuation and, assuming the other valve is open, its closure time must be such that, in the event of postulated failure of the component during normal reactor operation, each valve remains operable and the reactor can be shut down and cooled down in an orderly manner, assuming makeup is provided by the reactor coolant makeup system only.

(2) Applicable portions of the normal shutdown cooling pressure boundary function.

That is, Class 1 and 2 components of systems or portions of systems needed to utilize the normal shutdown cooling flow path either:

(a) As part of the RCPB from the reactor pressure vessel (RPV) to the second isolation valve (i.e., farther from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds; or (b) Other systems or portions of systems from the RPV to the second isolation valve (i.e., farther from the RPV) capable of remote closure or to the containment penetration, whichever encompasses the larger number of welds.

(3) Class 2 portions of the steam generators and Class 2 feedwater system components greater than NPS 4 (DN 100) of pressurized water reactors (PWRs) from the steam generator to the outer containment isolation valve.

(4) Components larger than NPS 4 (DN 100) within the break exclusion for high-energy piping systems as defined by the Owner.

(5) Any piping segment or component (including piping segments or components grouped or subsumed within existing plant initiating event groups) whose contribution to core damage frequency (CDF) is greater than 1E-06, or whose contribution to large early release frequency (LERF) is greater than 1E-07, based upon a plant-specific probabilistic risk assessment (PRA) of pressure boundary failures (e.g., pipe whip, jet impingement, spray, and inventory losses). This may include Class 1 and 2 components exempt from volumetric and surface examination by Section XI, or Class 3, or Non-Class components.

(b) Low safety significant (LSS) components shall include all other Class 1, 2, 3, or Non-Class components not classified HSS in accordance with (a) above.

10.1.2 Degradation Mechanism Assessment The susceptibility of each HSS piping weld to the degradation mechanisms listed in Table 2 of Code Case N-716-1 shall be determined. Table 1 of Code Case N-716-1 item numbers are assigned based upon the degradation mechanism evaluation. As described in section 10.1.1(5) above, Code Case N-716-1 augments the generic HSS

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 82 of 91 components with a search for plant-specific HSS components based upon the plant specific PRA pressure boundary failures.

10.1.3 Piping Weld Selection (a) HSS vessels, pumps, valves, and pressure-retaining bolting shall be selected and examined in accordance with Section XI.

(b) Ten percent of the HSS piping welds shall be selected for examination. The examination requirements for these locations are defined in Table 1. The existing plant IGSCC (Generic Letter 88-01, Categories B through G) inspection program may be credited toward the 10% requirement, provided the requirements of this Case are met. Selection of piping welds for examination shall be as follows:

(1) The susceptibility of each HSS piping weld to the degradation mechanisms listed in Table 2 shall be determined. HSS piping welds shall be assigned an item number in Table 1 based upon the results of the degradation mechanism evaluation. HSS piping welds identified as not susceptible shall be assigned Item No. R1.20 of Table 1.

(2) Examinations shall be prorated equally among systems to the extent practical, and each system shall individually meet the following requirements. At least 25% of the piping welds in each item number and item number combination (e.g., R1.11 and R1.16) shall be selected, excluding Item Numbers R1.15 and R1.20.

(3) For the RCPB:

(a) Ten percent of the piping welds shall be selected.

(b) At least two thirds of the examinations shall be located between the first isolation valve (i.e., isolation valve closest to the RPV) and the RPV. If this cannot by accomplished due to a limited number of available welds between the first isolation valve and the RPV, 25%

of the piping welds between the RPV and the first isolation valve shall be selected, not to exceed 10% of the RCPB piping welds.

(4) At least 10% of the piping welds in that portion of the RCPB outside containment (e.g., portions of the main feedwater system in BWRs) shall be selected.

(5) For each system within the break exclusion region, at least 10% of the piping welds shall be selected.

(6) If the examinations selected above exceed 10% of the total number of HSS piping welds, the examinations may be reduced by prorating among the requirements of (2), (3), (4), and (5) above, to the extent practical, such that at least 10% of the HSS piping welds are examined.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 83 of 91 (7) If the examinations selected above are not at least 10% of the HSS piping welds, additional welds shall be selected so that the total number selected for examination is at least 10% of the HSS piping welds. The additional piping welds may be selected using the criteria of (2), (3), (4), and (5) above.

(8) When selecting welds for examination, the following shall be considered:

(a) plant-specific cracking experience (b) weld repairs (c) random selection (d) minimization of worker exposure (9) For raw water systems identified as susceptible to localized corrosion, excluding crevice corrosion, in lieu of the above 10% sample, areas to be examined may be determined in accordance with section 3.6.7.1 or section 3.6.7.2 of EPRI TR-112657, Revision B-A. As discussed in TR-112657, if representative segments are defined, inspections shall include HSS segments.

(10) For combinations of material type and service condition with less than 10 yr. of operating experience, the sample shall be at least 25% of the locations (e.g. welds) containing the material.

(c) LSS components are exempt from the volumetric, surface, and VT-1 and VT-3 visual examination requirements of Section XI.

10.1.4 Risk Impact Assessment The Risk Informed Program has been conducted in accordance with Regulatory Guide 1.174 and the requirements of Code Case N-716-1, and the risk from implementation of this program is expected to remain neutral or decrease when compared to that estimated from current requirements.

This evaluation categorized segments as high safety significant or low safety significant in accordance with Code Case N-716-1, and then determined what inspection changes are proposed for each system. The changes include changing the number and location of inspections and in many cases improving the effectiveness of the inspection to account for the findings of the degradation mechanism assessment.

For example, examinations of locations subject to thermal fatigue will be conducted on an expanded volume and will be focused to enhance the probability of detection (POD) during the inspection process.

10.1.5 Quantitative Analysis

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 84 of 91 Code Case N-716-1 criterion requires that the cumulative change in CDF and LERF be less than 1E-07 and 1E-08 per year per system, and the total increase in CDF and LERF should be less than 1E-06 and 1E-07 respectively.

10.1.6 Implementation and Monitoring Program The applicable aspects of the ASME Code not affected by this code case will be retained, such as inspection methods, acceptance guidelines, pressure testing, corrective measures, documentation requirements, and quality control requirements.

Existing ASME Section XI program implementing procedures will be retained and modified, as appropriate.

The monitoring and corrective action program will contain the following elements:

A) Identify, B) Characterize, C) Evaluate

1) determine the cause and extent of the condition identified and

2) develop a corrective action plan or plans, D) Decide, E) Implement, F) Monitor and G) Trend 10.1.7 Periodic Update Examination selections shall be reevaluated on the basis of inspection periods that coincide with the inspection program requirements of IWA-2431. For the inspection program, the third period reevaluation will serve as the subsequent inspection interval reevaluation.

The performance of each inspection period reevaluation may be accelerated or delayed by as much as one year. Each reevaluation shall consider the cumulative effects of previous reevaluations. The reevaluation shall determine if any changes to the HSS/LSS categorization or examination selections need to be made by evaluation of the following:

(a) plant design changes (e.g., physical: new piping or equipment installation; programmatic: power uprating/18 to 24 month fuel cycle; and procedural; operating procedure changes).

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 85 of 91 (b) changes in postulated conditions or assumptions (e.g., check valve seat leakage greater than previously assumed).

(c) examination results (e.g., discovery of leakage or flaws).

(d) Failures in HSS or LSS components (e.g., plant-specific or industry occurrences of through-wall or through-weld leakage, failure due to a new degradation mechanism, or a non-postulated mechanism).

(e) PRA updates that would increase the scope (e.g., new initiating events, new system functions, more detailed model used, and initiating event and data changes).

(f) the impact of (a) through (e) on the change-in-risk evaluation.

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 86 of 91 Appendix A Waterford Unit No.3 ASME Code Class Systems ISI Boundary Diagrams Systems ISI Boundary Diagram No.

(System Designator)

Legend, Notes and Drawing Index Sheet P-100 Sheet 1 Main & Extraction Steam Penetration 1 & 2 P-151 Sheet 1 Main & Extraction Steam Detail A & Detail B P-151 Sheet 4 Instrument Air - Reactor Auxiliary Building Outboard Penetration 9 P-152 Sheet 4 Instrument Air - Reactor Containment Building Inboard Penetration 9 P-152 Sheet 6 Instrument Air - Essential Air P-152 Sheet 9 Feedwater & Emergency Feedwater Penetration 3 & 4 P-153 Sheet 4 Service Air Penetration 8 P-157 Sheet 1 Component Cooling Water P-160 Sheet 1 Component Cooling Water P-160 Sheet 2 Component Cooling Water P-160 Sheet 3 Component Cooling Water P-160 Sheet 4 Component Cooling Water P-160 Sheet 5 Component Cooling Water P-160 Sheet 6 Fire & Make Up Water Penetration 60 & 61 P-161 Sheet 1 Fire & Make Up Water Penetration 7 & 71 P-161 Sheet 2 Sampling P-162 Sheet 1 Sampling P-162 Sheet 2 Sampling P-162 Sheet 3 Waste Gas Analyzer P-162 Sheet 5 Containment Spray P-163 Sheet 1 Emergency Diesel Generator Fuel Oil System, Lube Oil System & P-164 Sheet 1 Penetration 63 & 65 Emergency Diesel Generator Starting Air System & Penetration 49 P-164 Sheet 2 Blowdown P-164 Sheet 5 TRB MISX - Main Steam Line Drain Penetration 5 & 6 P-165 Sheet 3 N2, H2, CO2 Penetration 14 P-166 Sheet 1 N2, H2, CO2 P-166 Sheet 2

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 87 of 91 Appendix A Waterford Unit No.3 ASME Code Class Systems ISI Boundary Diagrams Systems ISI Boundary Diagram No.

(System Designator)

Safety Injection P-167 Sheet 1 Safety Injection P-167 Sheet 2 Safety Injection P-167 Sheet 3 Safety Injection P-167 Sheet 4 Chemical Volume Control System Penetration 26 & 27 P-168 Sheet 1 Chemical Volume Control System Penetration 44 P-168 Sheet 2 Chemical Volume Control System P-168 Sheet 3 Spent Fuel Pool P-169 Sheet 1 Waste Management Penetration 31 P-170 Sheet 4 Boron Management Penetration 43 P-171 Sheet 1 Boron Management P-171 Sheet 2 Reactor Coolant System P-172 Sheet 1 Sump Pump Penetration 42 P-173 Sheet 3 RCP Seals P-179 Sheet 1 HVAC Air Flow Penetration 10, 11, 45, 46, 47, 48 & 302 P-853 Sheet 1 HVAC Chilled Water Flow P-853 Sheet 3 HVAC Chilled Water Flow P-853 Sheet 4 HVAC Chilled Water Flow P-853 Sheet 5 HVAC Chilled Water Flow P-853 Sheet 6 HVAC Chilled Water Flow P-853 Sheet 11

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 88 of 91 Appendix B Waterford Unit No.3 Class 1, Class 2 and Class 3 ISI Isometrics and Equipment Drawings Drawing Number Drawing Title ASME Code Class WTR-1-1100 Sheet 1 Reactor Vessel 1 WTR-1-1000 Sheet 2 Reactor Vessel 1 WTR-1-1200 Reactor Vessel Internals 1 WTR-1-1300 Reactor Vessel Closure Head 1 WTR-1-1400 R.V. Studs, Nuts, and Washers 1 WTR-1-2100 Pressurizer 1 WTR-1-3101 Steam Generator #1 1 WTR-1-3102 Steam Generator #1 - Bottom View 1 WTR-1-3201 Steam Generator #2 1 WTR-1-3202 Steam Generator #2 - Bottom View 1 WTR-1-4100 Reactor Coolant Piping Loop 1A & Hot Leg 1 WTR-1-4101 Reactor Coolant Piping - Loop 1B 1 WTR-1-4102 Shutdown Cooling From Loop 1, Class 1 1 WTR-1-4103 Safety Injection to Loop 1A, Class 1 1 WTR-1-4104 Safety Injection to Loop 1B, Class 1 1 WTR-1-4105 H.P.S.I. Line to Loop 1 Hot Leg, Class 1 1 WTR-1-4106 Hot Leg Drain - Loop 1 1 WTR-1-4107 Cold Leg Drain - Loop 1A 1 WTR-1-4108 Charging Line to Loop 1A 1 WTR-1-4109 Cold Leg Drain - Loop 1B 1 WTR-1-4200 Reactor Coolant Piping - Loop 2A & Hot Leg 1 WTR-1-4201 Reactor Coolant Piping - Loop 2B & Hot Leg 1 WTR-1-4202 Shutdown Cooling From Loop 2, Class 1 1 WTR-1-4203 Safety Injection to Loop 2A, Class 1 1 WTR-1-4204 Safety Injection to Loop 2B, Class 1 1 WTR-1-4205 H.P.S.I. Line to Loop 2 Hot Leg, Class 1 1 WTR-1-4206 Cold Leg Drain - Loop 2A 1 WTR-1-4207 Letdown Line from Loop 2B 1 WTR-1-4208 Charging Line to Loop 2A 1

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 89 of 91 Appendix B Waterford Unit No.3 Class 1, Class 2 and Class 3 ISI Isometrics and Equipment Drawings Drawing Number Drawing Title ASME Code Class WTR-1-4500 Pressurizer Surge 1 WTR-1-4501 Pressurizer Safety Valve Piping 1 WTR-1-4502 Combined Pressurizer Spray 1 WTR-1-4503 Pressurizer Spray from Loop 1A 1 WTR-1-4504 Pressurizer Spray from Loop 1B 1 WTR-1-4505 Pressurizer Auxiliary Spray 1 WTR-1-4506 Pressurizer Auxiliary Spray 1 WTR-1-5100 Reactor Coolant Pump 1A 1 WTR-1-5200 Reactor Coolant Pump 1B 1 WTR-1-5300 Reactor Coolant Pump 2A 1 WTR-1-5400 Reactor Coolant Pump 2B 1 WTR-2-1100 Shutdown Cooling Heat Exchanger A 2 WTR-2-1200 Shutdown Cooling Heat Exchanger B 2 WTR-2-3101 Steam Generator #1 - Top View 2 WTR-2-3201 Steam Generator #2 - Top View 2 WTR-2-4100 Main Steam Header A - Inside Containment 2 WTR-2-4101 Main Steam Header A - Outside Containment 2 WTR-2-4102 Main and Emergency Feed Header A 2 WTR-2-4103 Containment Spray and Safety Injection Suction Header A 2 WTR-2-4104 Low Pressure Safety Injection Pump A Suction 2 WTR-2-4105 Shutdown Cooling from Loop 2, Class 2 - Outside Containment 2 WTR-2-4106 Shutdown Cooling from Loop 2, Class 2 - Inside Containment 2 WTR-2-4107 High Pressure Safety Injection Pump A Suction 2 WTR-2-4108 Containment Spray Pump A Suction 2 WTR-2-4109 LPSI Pump A & CS Pump A Discharge to SDCHX A 2 WTR-2-4110 SDCHX A Discharge 2 WTR-2-4111 Safety Injection to Loop 2A, Class 2 2 WTR-2-4112 Combined Safety Injection to Loop 2, Class 2 2 WTR-2-4113 Safety Injection to Loop 2B, Class 2 2

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 90 of 91 Appendix B Waterford Unit No.3 Class 1, Class 2 and Class 3 ISI Isometrics and Equipment Drawings Drawing Number Drawing Title ASME Code Class WTR-2-4114 Safety Injection from SI Tank 1A, Class 2 2 WTR-2-4115 Safety Injection from SI Tank 1B, Class 2 2 WTR-2-4116 High Pressure Safety Injection Pump A Discharge 2 WTR-2-4118 Containment Spray A Discharge Piping 2 WTR-2-4119 Containment Spray A Discharge Piping 2 WTR-2-4120 Containment Spray A Discharge Piping 2 WTR-2-4121 HPSI A Hot Leg Injection 2 WTR-2-4122 HPSI A and B to RCS Loop 2A (Cold Leg) 2 WTR-2-4123 HPSI to RCS Loop 2A (Cold Leg) 2 WTR-2-4124 HPSI A & B to RCS Loop 1A (Cold Leg) 2 WTR-2-4125 HPSI A & A/B Recirc. 2 WTR-2-4200 Main Steam Header B - Inside Containment 2 WTR-2-4201 Main Steam Header B - Outside Containment 2 WTR-2-4202 Main and Emergency Feed Header B 2 WTR-2-4203 Containment Spray and Safety Injection Suction Header B 2 WTR-2-4204 Low Pressure Safety Injection Pump B Suction 2 WTR-2-4205 Shutdown Cooling from Loop 1, Class 2 - Outside Containment 2 WTR-2-4206 Shutdown Cooling from Loop 1, Class 2 - Inside Containment 2 WTR-2-4207 High Pressure Safety Injection Pump B Suction 2 WTR-2-4208 Containment Spray Pump B Suction 2 WTR-2-4209 LPSI Pump B and CS Pump B Discharge to SDCHX B 2 WTR-2-4210 SDCHX B Discharge 2 WTR-2-4211 Safety Injection to Loop 1A, Class 2 2 WTR-2-4212 Safety Injection to Loop 1B, Class 2 2 WTR-2-4213 High Pressure Safety Injection Pump A/B Suction 2 WTR-2-4214 Safety Injection from SI Tank 2A, Class 2 2 WTR-2-4215 Safety Injection from SI Tank 2B, Class 2 2 WTR-2-4216 High Pressure Safety Injection Pump B Discharge 2 WTR-2-4217 High Pressure Safety Injection Pump A/B Discharge 2

Program Section No.: SEP-ISI-104 Revision No.: 7 Page No.: Page 91 of 91 Appendix B Waterford Unit No.3 Class 1, Class 2 and Class 3 ISI Isometrics and Equipment Drawings Drawing Number Drawing Title ASME Code Class WTR-2-4218 Containment Spray B Discharge Piping 2 WTR-2-4219 Containment Spray B Discharge Piping 2 WTR-2-4220 Containment Spray B Discharge Piping 2 WTR-2-4221 HPSI B Hot Leg Injection 2 WTR-2-4222 HPSI A & B to RCS Loop 2B (Cold Leg) 2 WTR-2-4223 HPSI A & B to RCS Loop 1B (Cold Leg) 2 WTR-2-4224 HPSI B & A/B Recirc. To RWSP 2 WTR-2-4225 HPSI to RCS Loop 1B (Cold Leg) 2 WTR-2-4226 HPSI Pump A/B Recir. 2 WTR-2-5100 Containment Spray Pump A 2 WTR-2-5200 Containment Spray Pump B 2 WTR-2-5300 High Pressure Safety Injection Pump A 2 WTR-2-5400 High Pressure Safety Injection Pump B 2 WTR-2-5500 High Pressure Safety Injection Pump A/B 2 WTR-2-5600 Low Pressure Safety Injection Pump A 2 WTR-2-5700 Low Pressure Safety Injection Pump B 2 ESSE- NG-200 and NG-201 3 1564- 49, 1347, 1466, 1493, 1506, 4274 and 6580 3 4305- 3256, 3257, 3261, 3264, 5314, 5676, 5677, 5678, 5679, 5680, 3 5685, 5686, 5691, 5692, 5693, 5694, 5697, 5698, 5699, 5700, 5701, 5702, 5706, 5707, 5708, 5709, 5710, 5711, 5712, 5713, 6238, 6239, 6237, 6240, 6241, 6242, 6243, 6244, 6245, 6246, 6247, 6248, 6249, 6250, 6251, 6252, 6253, 6254, 6255, 6257, 6258, 6259, 6260, 6263, 6264, 6268, 6274, 6275, 6276, 6278, 6279, 6354, 6359, 6360, 6363, 6364, 6367, 6368, 6369, 6371, 6372, 6373, 6374, 6375, 6376, 6377, 6412, 6415, 6597, 6598, 6603, 6604, 6605, 6606, 6633, 6634, 6636, 6912, 6913, 6914, 6915, 6916, 8301, 8338, 8339, 8340, 8341, 8888, 8889, 8890, 8891, 8892, 9536, 9537, 9538, 9542, 9540 and 9598 8469- 176, 180, 187, 191, 195, 202, 203, 205, 206, 216, 315, 316, 3 317, 318, 319, 321, 324, 328, 329, 334, 335, 337, 338, 339, 340, 341, 342, 343, 345, 346, 347, 351, 352, 355, 356, 376, 1694, 1695 and 1696

wIR1 1 00 REACIOP VESSEL SHEET 1 OE 2 REACTOR VESSEL NOZZLES F II

WTR1 1 00 REACIOF VESSEL SHEET 2 OF 2 REACTOR VESSEL FLANGE LIGAMENTS BOTTOM HEAD 1 BO 01 OD 01005 01 L1 5 01 L1 1 01003 010DB Dl L1 01 L42 0l0 01002 UPPER SHELL SEGMENTS MIDDLE SHLL SEGMENTS LOWER SHELL SEGMENTS 0101 B 01010 B 7D YO 9D 7D 01017 Dl 009 01019 01 01 3 01011 ri, MAUC. WTD_1_i,flflcufl1.

WIR1 1200 REACTOR VESSEL INTERNALS 01 054 ACCESSIBLE AREA OF INTERIOR OF REACTOR VESSEL WITH FUEL AND CORE SUPPORT BARREL IN PLACE.

01055 ACCESSIBLE AREA OF CORE SUPPORT BARREL AND INTERIOR OF REACTOR VESSEL WITH FUEL AND CORE SUPPORT BARREL REMOVED.

SURVEILLANCE CAPSULE HOLDER 01056 THRU o;061 CORE STABILIZING LUGS (SNUBBERS) 01039 THRU 01044 CORE STOP LUGS 01045 THRU 01053 FLOW BAFFLE Dl 062 1800 1 Lm1 c.

t -(

SURV. CAP.

HOLDER (IYP. 6 9Q0 PLACES)

SNUBBER

)TYP. 6 PLACES)

CORE STOP LUG (TYP). 9 PLACES FILl\ME: WIR 1200.dgn

WTR1 1300 REt\CTCR VESSEL flr XL r,r iE r

0 SEE DETAIL A

/ 7 Y

7 1 8O U -

OLTAIL A Do EEDM YIELD DESIGNATIONS JNSIHIIVENT NELD DEEIGNATIGIIS 02VD1 Xl THROUGH 02V87X1 02115071 THROUGH 02WSJX1 PERIPHERAL 02T92X1 THROUGH Dl Ti 0171 02X60X1 THROUGH 027-8771 PER1PHERA 02U92X1 THROUGH 02010171 L

021501 THROUGH 0218771 PER IPHERAL U2ZSOXi 1 HRDUGFI 0228771 PER I PIIERAL

WTR1 J 1OO R.V. STUDS NLTS AND WASHERS N UT STUD LI 01 NOf T H P LI 01 N51 I

015 01 THRU 01 S5z1 WASHER fl 01 W01 I Ii N U 01 W5 F]LENAME: VIIR11400.dVn

]TRJ 2100 PRESSUR I ZER TOP HEAD 27O 05013 05B03 05018 05004 05-005 05020 05Oil 05012 6

05007 05017 900 05006 HEATER CONNECTION WELD DESIGNATIONS Al 05H0i Di 05Hi 1 01 05H21 A2 05H02 D2 05Hi 2 62 05H22 05005 01 05H03 03 05Hi 3 63 O5H23 B2 05HD4 D4 05Hi 4 64 05H24 03 05H05 El 05HiS Hi 05H25 64 05 4 00 02 05 H 10 HO 05 H OS Cl 05H07 Fl 05H17 H3 05H27 05004 C2 05H08 E2 05Hi B H4 05H28 C3 05H09 F3 05HiS Ii 05H29 C4 05HJo F4 D5H20 12 05H30 BOTTOM HEAD 05003 05C01 1 8O 05A0l 05ADO 05002 05001 05-014 05 021 2700 . 900 05009 --

FILENAME: WTR12100.dgn

WTR-1-3101 STEAM GENERATOR #1 Key Bracket to < Upper Shell to Elliptical Upper Shell Weld Head Weld 03-068 03-071 SG#l Upper Key Support @ 0° Axis03-072 SG#1 Upper Key Support @ 18D°.xis Snubber Lug to ]pper Shell Weld 03-069 SG#1 Snubber Lug t 90 03-070 SG#1 Snubber Lug @ 270 03-073 Feedwater Nozzle to Shell Weld 03-074 Feedwater Nozzle Inside Radius E- 03-075 Tubesheet to Lower Shell Weld Primary Nozzle Insi e Radius (Hot Leg)03-086 °03-076 Channel Head to ubesheet Girth Weld Primary Nozzle to Trnsition Primary Nozzle Inbide Radius (Cold Legs)

Piece Butt Weld (I-lot Leg)03-077 03-085 @ 45° Axis i 03-079 @ 315° Axis I I I

°-- I

_ f N I Support Pedestal to Channel° / °° Primary Nozzle to Transition Piece Butt Weld Head Weld 03-083 -. / -

(Cold Legs)03-078 45° Axis Primary Manway Bolting 03-080 @315° Axis03-081 @ 0° Axis03-082 @ 112.5° Axis

WTR-1-31 02 STEAM GENERATOR#1 00 03-081 Primary 03-060 Primary 03-078 Primary Manway Bolting @ Nozzle to Transition Nozzle to Transition Piece Butt Weld (Cold Piece Butt Weld (Cold Leg) C 315° Leg) @ 45°03-077 Primary Nozzle Inside Radius03-079 Primary (Cold Leg) @ 45° Nozzle Inside Radius (Cold Leg) C 315 2700 03-082 Primary Manway Bolting @ 112.5 03-086 Primary Nozzle Inside Radius03-085 Primary Nozzle to Transition Piece Butt Weld (Hot Leg)

BOTTOM VIEW

-13201 STEAM GENERATOR #2 Key Bracket to < Upper Shell to Eiliptical Upper Shell Weld Head Weld CL-068 04-071 SG#2 Upper Key Support @ 0 Axis04-072 SG#2 Upper Key Support @ 180 Axis N

Snubber LuO to Upper Shell Weld 04-069 SG#2 Snubber Lug @ 90 04-070 SG#2 Snubber Lug @ 270

/C

---04-073 Feedwater Nozzle to Shell Weld

/ 074 Feedwater Nozzle Inside Radius 7725 04 075 Tubesheet to ower Shell Weld Primary Nozzle Insi e Radius Hot Leg)04-086 ---- 04-076 Channel Head to Primary Nozzle to It nsition

/ -

ubesheet Girth Weld

ç Primary Nozzle Inside Radius (Cold Legs)

Piece Butt Weld (Hot Leg)- 04-077 @ 45 Axis04-08D 7 04-079 315 Axis I j 7 / I I Support Pedestal to Channel *--

/ N Primary Nozzle to Transition Piece Butt Weld Head Weld 04-083 / (Cold Legs)

Primary Manway Bolbng / 04-078 @ 5 Axis04-080 315 Axis04-081 @0Axis04-082 1 12.5w Axis

WTR-1 -3202 STEAM GENERATOR #2 0°04-081 Primary Manway Bolting 04-080 Primary 04-078 Primary @ 0 Nozzle to Transition Nozzle to Transition Piece Butt Weld (Cold Piece Butt Weld (Cold Leg) @315° Leg) @ 45°04-077 Primary Nozzle Inside Radius04-079 Primary (Cold Leg) @ 45° Nozzle Inside Radius (Cold Leg) @315° 2700 04-082 Primary Manway Bolting @ 112.5 04-086 Primary Nozzle Inside Radius i Q no 04-085 Primary Nozzle to Transition Piece Butt Weld (Hot Leg)

BOTTOM VDEW

CD CD CD CD CD CD 2

0 F1 C

z 0 CD CD CD CD

-D 9-CD F

CD CD Ui CD CD

-H C

C C

C C

-u NJ I

C CD rJ CD

WiR11Oi REACTOR COOLANT DIP]NC LOOP 16 SECT I 6N 03066A 1i001 4It 10013 D014CB 0-002/

/1 09009A 00OO7LB r- 19019L0 , 1OO1OLA

/

/

RC30RL61 0007 / 100.5 TE/0 EI.

00O8J 0-O10-IE 10001 09-004A 09-O2OLA 10-011LB PEE. 14104 10OO3LA 10000 icG 4 iRC313J STEAM GENE R A T0R 10-004L6 PEEi -504 09017 REP 14109 EIENALE

WI11m.dgr

WTR1 1 1 02 SHUTDOWN COOLING FROM LOOP 1 CLASS 1

_ol 21 ols 01?

024 21 71 REF. DRAWiNGS:

81691 5 846929 0167 SN- 2 0172 F1LENAME WTRl4102.dqn

WTRi 1*] 03 SAFETY INJECTiON -o LOOP 1A, CLASS 1 1Ep I7 7 1 -

7_oo

/

51 11 REF. DRAWINGS:

8S91 2 9

BR925 C-i67 SH. 2 Ci F liEIIAFIE: WTRiI1O3.dg

WTP1 11 O1 SAFETY NJECIJCN TO LOOP 16 CLASS 1 11S S..

SS 050 REF. ORAWjC5:

RS91 0 1 8 422 8692Q 43051894 G 67 SH. 2 01 72 r:LLN4i.c: W1H4IO4.dgr

TR1 41 05 H.P.S. 1. LINE TO LOOP 1 HOT LEG, CLASS 1 A

122q I2l1 5IfP 51 51PP-j2l1 21 21 21 21 2l 21 L

_i2 _051 21 REF. DRAWINGS: 21°

\

84R930 G1G7 SH. 2 A FILENAME: WTR14105.dçn

WIR1 41 O HOT LEG DRAIN LOOP 1

_o°2 J

006 0i REF. DRAW]NGS:

E3029 LW3 RC19 0172 FILENAME: WTR11O6.dgr

WTR1 1 1 07 COLD LEG DRAIN LOOP 1A REF. _;4100 REE. DRAWINGS:

E3029LW3RC 15 G1 72 FLENME: WIR14iO?.gn

\NTR1 Li 1 08 CHARGING LI TO LOOP 1A 4_02 IcfP 34 34 34 3

_000 34 REF. DRAWINGS:

E3O29LW3RC E3029LW3RC1 7 0168 SH. 1 0172 F}LNAME: WTRt4108.clqn

WTR1 Lii 09 COLD LEO DRAIN LOOP 19

_41°

.Oo1 o2 REF. DRAWINGS:

E 3029LW3 RC 15 C---1 72 FlLNAME: WTRl1O.dqn

m CD

-H D3 0

cm 0

F U

0 z 0

(1 cm C

0 0

cm N) 1 L 0

-H 0

m C) m Ci 0

i) cm 0 Is P N)

F 0 2 0

WfR1 1201 REACTOR COOLANT PIE INCLOOP 26 & HOTLEC SECTION 4012It 13015LA 1401 3TE 13000LA 1 1011 3-0 4 B 14001 14 -010 N 13009L0 C 3010LA SDOYLE 20 Ii 13 --003L B 3 O2DLA M

P

/NI 4OC8LA -

REF. 12206 3CO2CA 14OOBLE i3019L0 4OCSLA

/

STEAM 14OO4LB GENERATOR 1 3002X1-1301 7 13016 13013-LEIIAIE WTR1 4201 - 04

WTRi 4202 SHUTDOWN COOLING FROM LOOP 2 CLASS 1 22 2,020 _2 22 22...

011 22...

009

_001 q

22 REF. DRAWINGS:

8691 1 8692R G167 SH. 2 01 72 FILENMAE: WTR14202.dVn

pr)

/

0 f 0 4/

0 g

a

/

NJ g -

/ )

/_ J

,0!:; \ d / /

- V IL 0

0

-J 0

U H

U r LU z

\

H-LU U

C Lu CD CL o Dr CL CL Co U)

CD CL CL I IC-CL CL Co CO CO (j) N-U U-CL U-CD U-CC CD CO CD CD CD

C 0

0 0

0 CD

-J CD 0

H U

CD

-D

-7 H-CD CD

/ I / 1 CD r I

fL 1 cc CD C\J cc o -C\iCJ I I Ir-r\

C O cc cC I U- CCcDUD UJ Jcr cC cc

VTR1 4205 LINE TO LOOP 2 HOT LEO, CLASS 1 7Z

91 118 Ill I

04 REF. DRAWINGS:

846916 G167 SH. 2 FILENAME: WTRl4205dVfl

VVIRi 2O6 COLD LEG DRAIN LOOP 2A RF. 14?OO OS.

4 REF. DRAVINGS:

E3029 LW3 RC15 Ci 72 F1LEN4ME: WTR12OO.dqn

\TR1 i2O7 LETDOYN LINE FROM LOOP 2B

-1

_0zl 30

_001 30 50 lc 011 00q

_00 30

_o06 30

_008 30

_00 30

_0i4. REF. DRAWINGS:

30 E 3029 L W3 CH 32 E3029LW3RC1 4 C18 SH. 1 Ci 72 EILENML: WR112D7.dg

WTR1 4208 CHARGING LINE TO LOOP 2A R.

o1

_o5 2

L

_ool 2

02a lB REF. DRAW[NGS:

E3029LW3CH37 E3029LW3RC2 E3029--LW3RC1 3 0168 SH. 1 01 72 FiLENAME: WTRi4208.dVn

iTR1 1500 PRESSURIZER SURGE

_Q1 S

ç;,c Cc

_o02 l

6_OO1 16 PEr:.

REFiA!NN Ci 72 V9.11002 1 I 5641012 O2 70Pt--i 40 FILENAME: WTR14500.dgn

WTR1 5O1 PRESSLRIZER SAFETY VALVE PIPING 14 31 REF. DRAWINGS:

8469 724 8469728 Ci 72 F J LENAME: WTR1 45O dQn

VVTRJ 5O2 COMBINED PRESSURIZER SPRAY 1 2° 00 ols 1.l50s 01

_o12 2-006 REF. DRAWINGS:

869726 01 72 FILENAME: WTR145O.dcn

VqTR1 5O3 PRESSURIZER SPRAY FROM LOOP 1A Ii oil 02

.q-D1.

118 I

liO REF. DRAWINGS:

8l69 727 0172 FILENAME WTR145O.dgn

WTR1 5O PRESSURIZER SPRAY FROM LOOP lB O2

_41O pEF REF. DRAWINGS:

81S9725 Ci 72 FILENAME: W1R15O4.dgr

WTR1 4505 PRESSURIZER A]XILIARY SPRAY ISO 1i3 21 107 REF. DRAWINGS:

E3029LW3RC1 S G1 72 F1LENME: WTR14505.d0n

WTR1 150b PRESSURIZER A]XILIARY SPRAY 2

_002

.-Qlo

çQQ2 P33 2

I4 REF. DRAWINGS:

E3029LW3CH37 E 3029 L W3 R C C E3029LW3RC1 S ESSERCl C58 0168 SH. 1 01 72 F ICENAME: WTPt 4606. dOn

WTR1 5100 REACTOR COOLANT PUMP 1A Li___

37-006 37007 FLANGE STUDS NUTS LIGAMENTS 37SCl 37NCl 37L01 37S02 37N02 37LD2 37S03 37N03 37L03 31S0 37ND 37L0 37S05 37N05 37L05 37S06 37N06 37L06 LLLI 37 SC? 37NC? 37L07 37S08 37N08 37LOB 37SOS 37NOS 37L09 37Sb 37NiC 37LbO 37Sli 37Nli 37Lil 37001 37Sb2 37N12 37Li2

. I 7c13 7tI 1 31-S-14 37-N-14 37-L-14 37-S-15 37-N-iS 37-L 15 37S16 37N16 37L16 37005 37 003 37O0 37002

/ HX. BOLTING 37COB F ANGE S0CTNc.

SEAL BOLTING 37009 F ILENAME W1H-51CO.dQfl

WIR1 5200 REACTOR COOLANT PUMP 16 38006 38-007 FLANGE STUDS NUTS LIGAMENTS 38501 38NOi 38LO1 3BS02 38N02 38L--02 38S03 38N03 38L03 38S01 38NO 38L01 38SOS 38N05 38LO5 38506 38NOb 38LO6 38507 38N07 38L07 38508 38NO8 38L08 38SOS 38NOS 38L09 38S1O 38NiC 38L1O 38Si 1 38Ni 1 38Li 1 38001 38S12 38N12 38L12 38S13 38N13 38L13 38S1i 38N14 38L14 38Si5 38NiS 38L15 38516 38NlB 38L16 38005 38 003 33002 38004 HX BOLTING 38008 FLANGE BOLTING SEAL BOLTING 38009 FILENAE WIRi52QD.dn

WTR1 5300 REACTOR COOLANT PUMP 2A 39006 L__.1 1 I I 39007 FLANGE STUDS NUTS LI GAMENTS 39501 39N01 39L01 39S02 39N02 39L02 39503 39N03 39LO3 39501 39N04 39L01 39505 39N05 39L05 39S06 39N06 39L06 39507 39N 07 39CaT 39SOS 39 NOS 39L0B 39509 39NOS 39L09 395 10 39 N 10 39L1O 39Sli 39Nli 39L1l 39001 39912 39N12 39L12

.9 N 39914 39N 1 39L14 39515 39N15 39L15 39916 39NiS 39Cla 39005 N 39003 39002 39-001 HX.6OCTING EL ANCE DOc 7 INC tL SEAL BOLTING 39 009 FILENAME: W1R1 5300 dn

WFR1 _5L100 REACTOR COOLANT PUMP 2B 40006 40007 FLANGE STUDS NUTS LIGAMENTS 40SCl 40N0l 40L0l 40S02 40N02 40L02 40S03 40N03 40L03 40S04 40N04 40L04 40S05 40N05 40L05 40506 40N06 40L06

\HH/

40S07 40NC? 40L07 10508 40N08 40L08 40509 40N09 40L09 40S-b 40NiC 40L10 40Sil 40Nil 40Li 1 40001 40S12 40N12 40L12 40313 40N 13 lU L b 40S14 40N14 40L14 40SiS 4ON15 40Li5 40S16 40N16 40L16 40005 10003 10002 40004 HX. BOLTING 40008 FLAIJOt BOT1NG SEAL BOLTING 40009 FIENAIE: WTR1540D.dg-

WIR21 100 SHUTDOWN COOLING HEAl EXCHANGER A 5308 53086 53088 53080 53090 53087 53091 53079 FILENAME: WTH2l100.dçn

WIR21 200 SHUTDOWN COOLINO HEAT EXCHANGER B 54084 51081 54083 54075 54085 54082 54086 54074 F)LENANIE: WTR2 1200.dgn

WIR-2-31 01 STEAM GENERATOR # 1 00 03-073 Feedwater Nozzle to Shell Weld 03-074 Feedwater Nozzle Inside Radius03-071 Key Bracket to Upper Shell Weld @ 0 03-070 Snubber Lug to Upper 03-069 Snubber Shell Weld Lug to Upper Shell Weld

@ 270°

@ 90° 2700 900 Secondary Secondary Manway Manway 03-072 Key Bracket to Upper Shell Weld @ 180 180° TOP VIEW

WTR-2-320 1 STEAM GENERATOR #2 0°04-073 Feedwater Nozzle to Shell Weld 04-074 Feedwater Nozzle Inside Radius04-071 Key Bracket to Upper Shell Weld @ 0 04-070 Snubber 04-069 Snubber Lug to Upper Lug to Upper Shell Weld Shell Weld

@ 2700 @900 270° 90° Secondary Secondary Manway Manway Key Bracket to Upper Shell Weld @ 180 180° iop V)EW

IN STEAM HEADER A WTR21 00 IHSIDE COHTAINMFNT I

..O1 4

) °os H

19 i1 a,

L4

i. 10 p01/i A

4 Si 01?

REP. DRAWING 43054931 0151 SN.

01 FILENAME: WTR-24100.dAfl

WTR21 101 MAIN STEAM HEADER A OUTSIDE CONTAINMENT 410° 13 j1

_o61 43 43 43 o02 43 4)

( q3 13 03c °° A

43

°20 A

REF. DRAW[NGS:

43O5431 3O5855O G151 SH. 1 FiLENAME: WTR24101.dgn

I UJuJ

-oQ cn ui cn LI LflLNJ G XcDU >

.u1  :

N

/_*

, 7 2  :

/

/

N

/_

/

/

if-- -

7

/I E

A

., /

/

0 t

/

rr I L3

/\

tJ. I 2 [Ti

\\ 2 o

m N / -

f\Y 7 c I N

?

/ A N m

/

S * / 7

)

/:/ .

-5 /7 SE 0

rj

WIR2-4 103 CONTAINMENT SPRAY AND SAEEIY INJECTION SUCIION HEADER A A

SI 1131 os 61_026 61 l_0368 9

A 61 REE. DRAWINGS:

43051884 43051888 G167 SH. 1 .3 FILENAME: WTR2--1O3-dgn

WTR24 1 Ol LOW PRESSURE SAFETY INJECTION PUMP A SUCTION 11

-o1l

_ols 49 49 1S 49 REF. DRAWINGS:

13O51 659 3O5i 668 G167 SH. 3 FiLENAME: NTR211O.dn

WTR24 105 SHUTDOWN COOLING FROM LOOP 2. CLASS 2 OUTSIDE CONTAINMENT t6 111 oO 9

5 REF _Z

-1 Si

_0213 oio SI.. Dq

_Q13 REF. DRAWINGS: _7,2 3O51 861 L13Q51696 9_DO A

0167 SF1. 3 NOTE: WECO 5104G 0167 SH. 2 PIPE COUNTER BORE CHANGES SCHEDULE FILENAME: WTR21lO5.dQfl

SHUTDOWN COOLING FROM LOOP 2 WTR24 106 CLASS 2INSIDE CONTAINMENT

_oO1 Si 151 5I5P-I

_oO9 10 REE. DRAY/IN 86 26 NOTC:V/ELO 51 002 G167 SH.2 PIPE COUNTER 3ORE CHANGES SCHEDULE FILLNAME: WTR2410E.dgn

HIGH PRESSURE SAFETY INJECTION PUMP A SUCTION WTR2 1O7

_0 61 (o-.

0 61 61 3° (C

61 REF. DRAWING 3O51 870 13051 87 0167 SN. 1.3 FILENAAE: WTR24107.dgn

WIR24 JOB CONTAINMENT SPRAY PUMP A SUCTION 381 C51F o7 REP. DRAWING 3O56 43051884 01 S3 0167 SH.3

LPSI PUMP A AND CS PUMP A DISCHARGE WTR24 109 TO SDCHX A REF. DRAWINGS:

L3Q54 13O5J BGG 43054701 01 03 0167 SH.

SDCHX A DISCHARGE WTR2l 110 57

.010 57 o1l 1105 S7

@5 i

033 REF. DRAWINGS:

3O5I 3O52 ci 63 FILENAME: WTR2411Qdn

(N EI 4 c_-\

/

(N (\

I\

f)

-J _:

(N r -

0 0

0

-J j//

0 U

w z

F w

U cJ

\

LI (3

V) (J LOll c1 LI LI LI LI -

LI LI LI LI UI LI 0 w tq I LI OLD

COMBINED SAFETY WTR211 12 INJECTION TO LOOP 2 CLASS 2 SI 5s

_001 REF. DRAWINGS:

43051 862 43054699 G-167 SH. 3 FILENAME WIR24112.dr

o C) - CD I I CD CD U] UD

-Ni C UDU]CDN) -

CD CDCDLO C)

JLJ cm

, /

I I

1cm m

0 0

z b

0

-U rj LU 0

r Ln

VTR_2L1 11 SAFElY INJECTIGN FROM SI TANK 1A1 CLASS 2 REF. DRAWING 6167 SH.

FILENAME: WTR24114.dQr

VITR2_tl 115 SAFETY INJECTION FROM SI TANK 1B CLASS 2 q1o REF. DRAR1NS:

86S2O G-1b7 SH. 4 FCENAME TR24115.dQn

WTR241 16 HIGH PRESSURE SAFETY INJECT[QN PUMP A DISCHARGE lzp REF. 24215, RFRRENCE_DRAW1NCS 3Q-1 888 4O51 888 G167 SH.1 WTR24116.dn

CONTAINMENT SPRAY A DISCHARGE PIPING WTR2 118 Os,

_94 54 YOsT OS1

_0s0 1

046 A4

_0 REF. DRAWING 3O52 G1 63 F]LtNAE: WIR2411a.don

WTR241 1 CONTAINMENT SPRAY A DISCHARGE PIPING

_o13

-J

-I RE DRAWINCS:

1i 16 430521 6163 FILENAME: WTR249qn

CONTAINMENT SPRAY DISCHARGE PIPING WIR21 120 Cs I-41 H

REF. DRAWING 43052 7 0163 FILENAME: WTfl2412O.dn

cJ nJ Er F

cI] 9/

\ cm cm 0

H-C)

CL 1

2:

CL

-J H

0 \

cm

<C cm

\

_)

\\

cCj 1

(I-)

CL

=

Cr Ir Crc w

w Cr - C-)

HPS1 A AND B TO RCS LOOP 2A (COLD LEG) WTR21 122 191 ill is.

REF.__DRAWING E302DLV/3S I 7 E3029LWSS] 8 Gi 7 FtLENAE: WTR2412?.d,n

rl U

z C)

LU

-J C)

-J 0

C)

(\J 0

0

-J &

0 C)

O9 0

H UI U

L,-os ZI Lu

• a UrO Lu

HPSI A AND WTR_2_tl 1 2 B TO RCS LOOP 1A (COLD LEG

.01 5

60 0

Si, gO REF. DRAWING E3029LW3S 12 E3029LW3S 14 SO 43051883 846925 F1LENAME WTR24124.dn

HPSI A AND A/B RECIRC. WTR2425

-5;_ c,_ -

I -5 21°;ii 4

0 6;3)

I (o

L 6 < ( - I 4

26

_50s - 3Q 0 -

N Th-.

6o26

ØA

çD

,fl5 I REF. DRAWING E3O2LW3SI24 E3029LVL3S 13 E3029LW3S 121 43O5494 3O51b95 G167 SH.3 FILENAME: WTfl2125.don

C) -

-0 00 0,_a

\,

S / / -_

rn C-)

0 z

7 if I

-H 0

)

[5) 0 0

WTR24201 MAiN STEAM HEADER B OUTSIDE CONTAINMENT 420 44 Ii 10

_fl 44 REF. DRAWINGS:

3O593 3o5935 4O58551 G151 SH. 1 F )LENAAE: WTR247fl1 .drn

N 0

cJ 5

PE2 3:

---n -

\ \ . /

• 2 \_

/ \ -

/ 7 7 *- S

/

S .

l F /

0 / \ F S

0 0 I 0

I fr\

0 0

0 0

z 0

0 0 7 0 .7 0

, c s-,-,

?

/

/7

\

/ 2

_N,

, L* G 0,

rnOco Lfl N- O I I C\i NJ LI) L) 0 i q N)

IN-)

Lfl U LIJ Lfl oos JJ rflrfl)

I

WTR24203 CONTAINMENT SPRAY AND SAFETY INJECTION SUCTION HEADER B

_33

-oo1 1o S

REF. DRAWINGS:

3O51 87 6167 SH. 1 ,3 F ILENAME: WTR242Q3.dn

WTR24204 LOW PRESSURE SAFETY INJECTION PUMP B SUCTION so-3

_22 036

.so078 so Cs REF. DRAWINGS:

_070 43051 864 so 43051868 0167 SH. 3 F ILENAME: WIR24204.dçr.

C, Li Li a: V 0

U, C\J C, a, a: V r i:

Li Li

.J 2 Lu 0 Li Li a:

C-Li a:

Li Liz 4

2 Li H U.

z w

z H

z 0

U w

0 F

0 cJ Lfl

-J C) a-0 0

-j 0

a LL U

z c-a

__j o .

o U

z 0 0 0

0 a -a-H 0 ED 0 i-a: T

,CD r) Ci a

3:

U-) CDII 0)0) U) LI)

ID U-) CD CD U 00-w rDrDJI

0) -ar r CD CD

\jTR22OB SHUTDOVN COOLING FROM LOOP i CLASS 2 INSIDE CONTAINMENT REF -

o02 4 0

_4

_oo9 55 1

_ol0 I

S Iç

°0 REF. DRAV/1NCS 8929 0167 SH. 2 NOTE: WELD 52002 PIPE COUNTER BORE CHANGES SCHEDULE FILENAME: WTR242DB.dgn

HIGH PRESSURE SAFETY INJECTION PUMP B SUCTION WIR24207 62 Os3 K I3-62 REF. DRAWINGS:

3o5J 860 4301863 0167 SH. 1

WTR_2_t120S CONTAINMENT SPRAY PUMP B SUCTION Ep 03 REF. DRAWINGS:

13O56 4301 879 CI 63 G167 SH.

FLENMt WIF242O3.rj0r,

7e

-U Cl

-U

-U 0

0 0 z

• _ c 0

Cl

/T 0

cn Cl 6 6 0 I

-U 0

m 0

Cl 0

0 I

0 C) C) 1 -

C)

I ILJLJUJ PT C) C) Ui Ui Ui 4IIj

---U]

Ui C)Co I oiUi >

z L C)

Ui 0

C 0

C CD

WTR2121 0 SDCHX 6 DISCHARGE REE. DRAWINGS:

43053 0200 5H2

WTR2 21i SAFETY INJECTION 0 LOOP 1A CLASS 2 65, a6 a QI s6

_Q2 g31 iPI S

c19 a- .

PEE. DRAWNO5:

8469 9 43051865 016] SN. 3 o 15] SH. 4

WTR221 2 SAFETY INJECTON 0 LOOP 1B CLASS 2 260 76

_/I pE1 5601

)

5V 56 iS r6 pL 2

6100 053 1

/

O5 L

/

REE. RAWNS:

Fit

WTR2---421 3 HIGH PRESSURE SAFETY INJECTION PUMP A/B SUCTION 5I 6

IL t36 E

10?

REF. DRA[NGS:

13O51 871 L3Q1 890 G17 SH.

WTR221 SAFETY INJECTION FROM SI TANK 2A CLASS 2 19 REF. DRAViNGS:

846928 6167 SH. 2 FILENAME: WIR21214.dqn

WIR2121 S SAFElY INJECTION FROM SI TANK 2B CLASS 2 013 REF. DRAV1NCS 8169 22 G167 SH. 2 FILENAME: WTR24215dVr

HIGH PRESSURE SAFETY WTR2.12 6 INJECTION UMP B DISCHARGE EERENCE DAW1NGS 43o5 856 4305i 581 HPS I 0187 S-i. I DUMP B FjLENM4E (IR2?i1.dQn

HIGH PRESS]RE SAFETr INJECTCN PUMP A/B DISCHARGE WIR221 7 rERENCE_ORAWINGS os; i 0167 SH.1 FlLENA4F WTR22l7dgr

WTR2121 B CONTAINMENT SPRAY B DISCHARGE PIPING 0A0 2.

A 0A 0 6s

_Q37 2

F.

6 REF. DRAWINGS; 66 6-163 43063 FILNAM: WTR2-4218A.dgn

1TR2121 9 CONTAINMENT SPRAY B DISCHARGE PIPING S

I 1

Q116 A

I REF. DRAWINGS:

6163 430523

VTR2L122O CONTAINMENT SPRAY B DISCHARGE PIPING 21 REF. DRAWINGS; G1 63 130523 430521 ALL INSIDE CONTAINMENT FILENAME: WTR2422OA.dgr

HPSI HOT LEO INJECTION WTR24221 REP. DRAWtNG 43056833 846916 E3029LW3S 131 0167 SH.1 0167 SH.2

C 0

cJ cJ S ID c0 0:

I 0

10 2

H z

C-)

LU U

m 0

-J ZCJ LI) 0 I

_.I In In cD In LI)

_J_JI- C 0

C m rI Ir CO In 010 L r) In OLD w I IrI)tI C Ui w to ()

a F

3 z

LI LI LI E3t *Lfl

<J _J I LI LI I I tIN ornm a IJCJN I

• OO tO LI W LI LI C WI I In LI Ui Ui LIV

8/1 WTR2122L HPS I AND A

S4

(N (N

Qr 9

o9 C)

U w

o9 03 N

0 0

U a

I:i:

a a-, v, cJ Or LL U

a W 0

V1R251 DO CONTAINMENT SPRAY PUMP A SUPPORT 113i C

C C C 3 C

C 3 P C 3 3

C 3 C 3 C 3 C

C C )

C 1=1 FILENAME; flR251OO.dgr

Vd I R 25200 CONTAINMENT SPRAY PUMP B SUPPORT 62104 C

C J P C

C ]

C L

j C :i C

C C ]

r HLENAME: WTR2 5200dQn

NT R 25300 HIGH PRESSURE SAFETY INJECTION PUMP A 61-141 I 1

]

]

H ,

L_

St-151

]

i

i

]

[Jo[J flo[J I 61-141 SUPPORT 61132 WELDED LUGS (FEET) Siill CASING WELD 6ii51 F ILENAME: WTR23OO. dgn

WTR25lOO HIGH PRESSURE SAFETY INJECTION PUMP B 61-143 I

j 3

3 3

3 3

ii LDB SUPPORT 62i 05 WELDED LUGS (FEET) 6ii3 CASING 6;iS3 L P H M R Ii I H F ILErMME: WTR?5OO. dçn

W JR 25500 HIGH PRESSURE SAFETY INJECTION PUMP A/B i-I42 1

I goB]

]

]

S1-152 cD ]

]

]

]

]

t 61-142 SUPPORT 61 133 WELDED LOGS (FEET) 61-12 CASING WELD 61152 FILENAME: Wifl255OO.dc

WTR25E00 LO1 PRESSURE SAFETY INJECTIflN PUMP A SUPPORT 119 080 9O67 - -

- 49 -0S5 FILtNA?.E: WIH25600.

\VTR257OO LOI PRESSURE SAFETY NJECIION PUMP B SUPPORT 50 096 50064 50082 50083 FlLENAMt WTF251OO.cign

Munich RE Lisa Sire November 30, 2017 ISI Engineer Inservice Inspection Program Owner David Sherer ANII WF3 Station IS-983 I ATL Region Cell -225-450-7306 Offic-504-739-6543 david_sherer@hsb.com Risk Solutions Dear Ms. Site The Hartford Steam Boiler Inspection and Insurance Co.

200 Ashford Center North, Suite 200 Waterford 3 Steam Electric Station Nuclear Power Plant Atlanta, GA 30338 Revision of SEP-ISI-f 04 Revision 7 www hsbcom The purpose of this letter is to document the HSB Code Services review of the Revision listed above to the Waterford 3 Steam Electric Station Nuclear Power Plant, ASME Section XI, Division 1, WF3 INSERVICE INSPECTION PROGRAM, as required by the American Society of Mechanical Engineers (ASME) Code Section XI, Division 1,2007/2008 Edition and Addenda, paragraph IWA-2110 (a). David Sherer is the Authorized Nuclear Inservice Inspector who performed this review.

The review was based on the requirements set forth in the ASME Code Section XI, Division 1, 2007 Edition, with 2008 Addenda. The revision was reviewed for Code compliance with Section IWA-2000.ltems listed in this revision are as follows, inspection of Class 1, 2 or 3 Components, Piping, Welds, Pump and Valve, Snubber Inservice Examination, were reviewed for code compliance In our opinion, SEP-lSl-104 Rev 7 was found acceptable as meeting the minimum requirements of ASME Code Section Xl, Division 1, 2007 Edition, with 2008 Addenda.

Included in this review was verification of the availability of the applicable Code Books and Code Cases. No Code deficiencies were noted during this review.

REVISION DESCRIPTION

. 1866 Hartford Steam Boiler

Page 2 November 30, 2017 Proposed Changes; To incorporate ASME Boiler and Pressure Vessel Code David ShererDavid ShererDavid Section XI, 2007 Edition with 2008 Addenda for Fourth Ten Year Interval and AN reformatted for consistency with Entergy Inservice Inspection Program. IS-983 fAIL Region Cell -225-450-7306225-450-7306 Oftic-504-739-6543 david_sherer@hsb.com REQUESTS FOR RELIEF Requests for Relief (RFR) approved by the Nuclear Regulatory Commission (NRC) are incorporated into the ASME Section Xl R&R Program upon approval. The status of Requests for Relief (RFRs) submitted to the NRC but not yet approved are tracked through the list in the Relief Request Section of the Program and by this letter. A brief description of each RFR is as follows:

No Requests for Relief issued at this time that are applicable to this PDCN.

Relief Request Description Status Number Thank you for the cooperation and assistance provided to me during my review of this revision to the ASME Section Xl ISI Program.

If you have any questions concerning this review or if I may be of any assistance please do not hesitate to contact me.

I erely David Sherer Authorized Nuclear Inservice Inspector The Hartford Steam Boiler Inspection & Insurance Co.

Attachment 2 to W3F1-2017-0082 Waterford 3 Fourth 10-Year Interval Containment Inservice Inspection Plan (63 pages)

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 1 of 50 PROGRAM SECTION FOR ASME SECTION XI, DIVISION 1 WF3 CONTAINMENT INSERVICE INSPECTION PROGRAM ENTERGY NUCLEAR ENGINEERING PROGRAMS APPLICABLE SITES ALL SITES:

AN1 CNS GGN IP3 PAL PNP VTY Specific Sites:

AN2 HQN IP2 JAF PLP RBS WF3 Yes Safety Related:

No Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-1 04 Revision No.: 2 Entergy Page No.: Page 2 of 50 REVIEW AND CONCURRENCE SHEET Program Section

Title:

ASME SECTION Xl, DIVISION 1, WF3 CONTAINMENT INSERVICE INSPECTION PROGRAM Prepared By: Date: 7i7,7 Checked By: Date:

ANII:

C421j Date:

Reviewed By (or NA)

Concurred: Date: t(-3G..-/7 Responsibl upervisor Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 3 of 50 REVISION STATUS SHEET REVISION

SUMMARY

REVISION ISSUE DATE DESCRIPTION 0 Initial Revision Issued For Use. CEP-CISI-104 was used as the bases for this revision which converts CEP-CISI-104 to SEP-CISI-104.

1 1) Updated Table 3.3-1 and 3.4-1 to reflect that Code Case N-753 is approved per RG 1.147 Rev.16 and Relief Request CEP-ISI-012 is no longer applicable. 2) Added W3-CISI-001 and W3-CISI-002 to Table 3.4-1 and the associated reference documents.

2 Updated SEP-CISI-104 to incorporate ASME Boiler and Pressure Vessel Code Section XI 2007 Edition 2008 Addenda for Fourth Ten-Year Interval and reformatted for consistency with Entergy Containment Inservice Inspection Programs standard format.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 4 of 50 Table of Contents Section Description Page Number NA Cover Page 1 NA Review and Concurrence Sheet 2 NA Revision Status Sheet 3 NA Table of Contents 4 1.0 CONTAINMENT INSERVICE INSPECTION PROGRAM 7 1.1 Purpose 7 ASME Section XI Code of Record for the Fourth Ten-Year 1.2 7 CISI Interval 1.3 ASME Section XI CISI Program Description 8 1.4 Inspection Intervals 9 1.5 Administrative Controls 10 1.6 Responsibilities 11 2.0 ASME CODE CASE APPLICABILITY 13 2.1 Adoption of Code Cases 13 Regulatory Guide 1.147. Revision 17 Approved Code 2.2 16 Cases 2.3 Code Cases Approved Through Request for Alternatives 16 2.4 Code Cases Mandated Via 10 CFR 50.55a 17 3.0 RELIEF REQUESTS 17 Requests for Relief Required Due to Impracticality or 3.1 17 Limited Examinations Requests for Alternatives to Code and Standards 10 CFR 3.2 18 50.55a(z) Acceptable Level of Quality and Safety 3.3 Table of Requests 19 Requests to use Later Edition and Addenda of ASME 3.4 19 Section XI Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 5 of 50 Table of Contents Section Description Page Number 4.0 20 DEVELPOMENTAL REFERENCES 5.0 ASME EXAMINATION BOUNDARIES 21 5.1 Code Classes 21 5.2 Exemptions 21 6.0 APPLICATION CRITERIA AND CODE COMPLIANCE 23 6.1 Examination Methods 23 6.2 Examination Categories 27 6.3 Evaluations 38 7.0 LISTING OF DEFINITIONS 42 8.0 REPAIR/REPLACEMENTS 45 9.0 PRESSURE TESTING 46 10.0 RECORDS AND REPORTS 46 INSERVICE INSPECTION DATA MANAGEMENT 11.0 48 SOFTWARE 12.0 COMPONENT DRAWINGS 49 TABLES Table Code Cases Adopted from Regulatory Guide 1.147 2.2-1 16 Table Code Cases Adopted Via NRC Approved Requests 2.3-1 16 Table Code Cases Adopted Via 10 CFR 50.55a 2.4-1 17 Table Fourth Ten-Year CISI Interval Relief Requests 3.3-1 19 FIGURES Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 6 of 50 Table of Contents Section Description Page Number Figure 1.4-1 WF3 4th IWE Interval Schedule 10 APPENDICES Appendix WF3 CISI Drawing List A 50 Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 7 of 50 1.0 CONTAINMENT INSERVICE INSPECTION PROGRAM 1.1 Purpose The Waterford Unit No. 3 Steam Electric Station (WF3) containment system is a large dry containment consisting of a free standing steel cylindrical containment vessel with a hemispherical head and an ellipsoidal bottom. The containment building is surrounded by a reinforced concrete shield building which provides missile protection to the containment.

Subsection IWL of ASME Boiler and Pressure Vessel Code Section XI does not apply to Waterford Unit No. 3 since the containment vessel does not rely on the detached concrete shell for structural support or pressure retention.

This Program Section contains the details of the ASME Section XI, Division 1, Containment Inservice Inspection (CISI) Program for the Waterford Unit No. 3 Steam Electric Station (WF3). Implementation of a Containment Inservice Inspection (CISI) Program is in accordance with the requirements of ASME Section XI, Division 1, as mandated by the United States Code of Federal Regulations, Title 10, Part 50, Section 55a (10 CFR 50.55a).

The scope of this Program Section includes the examination and testing of ASME Section XI Class CC and Class MC Components and their integral attachments for the Fourth Ten-Year Inservice Inspection Interval.

1.2 ASME Section XI Code of Record for the Fourth Ten-Year CISI Interval 10 CFR 50.55a(g)(5)(i) for the ISI program update, which is also applicable to the CISI Program, states that the program must be revised to meet the requirements of 10 CFR 50.55a(g)(4),(g)(4)(i) and (g)(4)(ii). This requires that inservice inspection of components and system pressure tests conducted during successive 120 month inspection intervals must comply with the requirements of the latest edition and addenda of the ASME Section XI Code incorporated by reference 12 months prior to the start of the 120 month inspection interval.

The initial Containment lSI Program commenced July 1 , 1997 and continued through June 30, 2007. The interval was extended until May 31, 2008, as permitted under IWA-2430(d). The initial Containment lSI Program Plan was included in Program Section No.CEP-CII-002 and the Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 8 of 50 Code of Record for the initial interval was the 1992 Edition with the 1992 Addenda of ASME Section XI as modified by 10 CFR 50.55a.

Accordingly, based on lock in date of May 31, 2008, the 2001 Edition with the 2003 Addenda of ASME Section XI is the version that was used for the previous interval. The Third Ten-Year Interval dates were from May 31, 2008 concluding on November 30, 2017 thus reclaiming 6 months of the 12 month extension that occurred in the 2nd Interval.

In accordance with 10 CFR 50.55a(g), Entergy is required to update the ASME Section XI (the Code) Containment ISI Program once every ten years. The updated Containment ISI Program is required to comply with the latest edition and addenda of the Code incorporated by reference in 10 CFR 50.55a(a)(1)(ii), one year prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii).

Because the fourth interval begins on 12/1/2017, the 10 CFR 50.55a in effect one year prior (December 1, 2016), requires the Containment ISI Program meet the requirements of ASME Section XI 2007 Edition through the 2008 Addenda.

The CISI Program Section for ASME Class MC components for the Fourth Ten-Year CISI Interval has been developed using the ASME Boiler and Pressure Vessel Code, Section XI, 2007 Edition through the 2008 Addenda; except where specific written alternatives from Code requirements have been requested by Entergy and authorized by the NRC or as amended by the NRC in 10 CFR 50.55a.

1.3 ASME Section XI CISI Program Description The WF3 CISI Program Section details compliance with ASME Code, 2007 Edition through the 2008 Addenda, Article IWE for examination of Class MC components. This document defines the Class MC components and the Code required examinations for each ASME Section XI examination category, and the augmented inspection scope, as applicable.

The components subject to the requirements of this CISI Program are those which make up the containment structure, its leak tight barrier (including integral attachments) and those which contribute to its structural integrity. Specifically, these components consist of the Class MC pressure retaining components. The CISI Program examinations shall apply to surface areas, including welds, base metal, and pressure Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 9 of 50 retaining bolting. This CISI Program Section has been developed to be consistent with the requirements of the plant technical specifications The administrative procedures and inspection schedule described in the CISI Program, combined with applicable Entergy procedures, and approved vendor procedures, constitute the CISI portion of the Fourth Ten-Year ISI Program.

The CISI Program schedule of examinations is contained in the IDDEAL Software Suite computer database. This information is retained at the plant site and is available for review.

1.4 Inspection Intervals 1.4.1 IWE (Class MC) Inspection Interval and Periods The Fourth Ten-Year Interval Program for performance of Containment ISI complies with IWE-2411 and will commence on December 1, 2017 and end on November 30, 2027. These dates were determined from the regulatory requirements of 10 CFR 50.55a. The three periods, within the interval, are defined in Section XI and are as follows:

First Period 12/1/2017 through 11/30/2020 (3 years)

Second Period 12/1/2020 through 11/30/2024 (4 years)

Third Period 12/1/2024 through 11/30/2027 (3 years)

Per IWA-2430(c), each inspection interval may be reduced or extended by as much as one year. Adjustments shall not cause successive intervals to be altered by more than one year from the original pattern of intervals. If an inspection interval is extended, neither the start and end dates nor the inservice inspection program for the successive interval needs to be revised.

Containment ISI during the fourth interval shall be performed in accordance with the 2007 Edition and 2008 Addenda of ASME Section XI identified in Section 1.2, the Code Cases identified in Section 2.0 and Relief Requests identified in Section 3.0.

Note: Within the ASME Section XI, Code of Record, Table IWE-2411-1 INSPECTION PROGRAM allows and requires percentages and limits of Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 10 of 50 examinations to be performed each inspection period, but based on the Examination Requirements of Table IWE-2500-1, Examination Categories E-A, E-C and E-G, where all required examinations within these Examination Categories are to be completed either 100% each Inspection Period or 100% each Inspection Interval NO partial percentages are applicable and thus the Table is not used in this CISI Program.

Figure 1.4-1 WF3 4th IWE Interval Schedule 1RF22 1RF23 1RF24 1RF25 1RF26 1RF27 S 2019 F 2020 S 2022 F 2023 S 2025 F 2026 1/2017 1/2018 1/2019 1/2020 1/2021 1/2022 1/2023 1/2024 1/2025 1/2026 1/2027 1/2028 st nd rd 1 Period 2 Period 3 Period 36 Months 48 Months 36 Months 12/01/2017 12/01/2020 12/01/2024 11/30/2027 TH TH Start 4 IWE End 4 IWE Interval Interval 1.5 Administrative Controls The ASME Section XI CISI Program Plan is developed and maintained by the Program Owner assigned to the site Engineering Programs. As specified in 10 CFR 50.55a(g)(6)(ii)(B), Licensees do not have to submit to the NRC for approval of their containment inservice inspection programs that were developed to satisfy the requirements of Subsection IWE with specified conditions. The program elements and the required documentation must be maintained on site for audit.

Changes to the contents of this program section shall be requested in accordance with EN-DC-174, Nuclear Management Manual Procedure.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 11 of 50 Repair/Replacement requirements are implemented by Entergy Program Section CEP-RR-001, ASME Section XI Repair/Replacement Program.

Refer to Section 8.0 for additional details.

Scheduling and component information for this CISI Program is controlled by the IDDEAL databases. The IDDEAL database is controlled in accordance with CEP-COS-0110.

The CISI Program is mandated by 10 CFR 50.55a(g)(4). As a result, in accordance with EN-LI-100 Revision 11, Process Applicability Determination is not required for the initial issue but a Process Applicability Determination will be performed on all initial documents for Process Applicability Exclusion.

1.6 Responsibilities 1.6.1 Site Engineering Programs 1.6.1.1 The Engineering Programs Supervisor, or designee, is responsible for implementation, control, and interpretations of this program, as well as generating Code documents following each outage and designating a Responsible Individual (RI) or Responsible Engineer (RE).

1.6.1.2 Develop a CISI Program in accordance with applicable ASME Section XI Code requirements and applicable conditions specified in 10 CFR 50.55a.

1.6.1.3 Ensure that the Inspection Plan(s) are correct for implementation at the site and provide applicable Inspection Plan(s) to the Site Containment Inservice Inspection (CISI) Program Engineer in a timely manner.

1.6.1.4 Ensure examinations are properly scheduled to comply with the applicable examination frequency and extent requirements outlined within ASME Section XI Code and/or Regulatory conditions.

1.6.1.5 Ensure the proper disposition of components or items requiring acceptance by engineering evaluation or corrective action.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 12 of 50 1.6.1.6 Ensure the proper examination expansion based upon engineering evaluation and Code requirements.

1.6.1.7 Implement the Inspection Plan provided by the Site Engineering Programs.

1.6.1.8 Ensure that the Inspection Plan and Containment Inservice Inspection (CISI) Summary Reports for IWE have been properly prepared and/or submitted to the enforcement and/or regulatory authorities having jurisdiction at the plant site.

1.6.2 Responsible Individual (RI) for Subsection IWE-2320 1.6.2.1 Develop the inspection plans, procedures, and schedules for the examination of ASME Code Class MC components. The inspection plan will be turned over to the Site Engineering Inservice Inspection (ISI) Program Manager to maintain.

1.6.2.2 Coordinate with the NDE Corporate Level III, or designee, to provide instruction, training, and approval of visual examination personnel performing ASME Code Class MC component examinations 1.6.2.3 Evaluate examination results pertaining to ASME Code Class MC components.

1.6.2.4 Submit the Summary Report to the enforcement and/or regulatory authorities having jurisdiction at the plant site.

1.6.2.5 Provide screening criteria for the relevancy of indications noted in the CISI examinations. This duty will normally reside jointly with a designated on-site RE from the engineering group retaining design authority for the containment.

1.6.2.6 Perform an acceptance review of CISI examination results to determine if the examination results are acceptable by examination in accordance with IWE-3122.1. This duty will normally reside jointly with a designated on-site RE from the engineering group retaining design authority for the containment.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 13 of 50 1.6.2.7 Specifying supplementary examinations as necessary to support the acceptance review. This duty will normally reside jointly with a designated on-site RE from the engineering group retaining design authority for the containment and with the RE responsible for maintaining this program plan.

1.6.2.8 Specifying and recommending corrective action for items that are not acceptable by examination. This duty will normally reside with a designated on-site RE from the engineering group retaining design authority for the containment.

1.6.2.9 Review of engineering evaluations for items accepted in accordance with IWE-3122.3. This duty will normally reside jointly with a designated on-site RE from the engineering group retaining design authority for the containment and with the RE responsible for maintaining this program plan.

1.6.2.10 Approval of training requirements for General Visual Examination personnel.

1.6.2.11 Maintenance of the CISI Program Plan.

1.6.3 The Site Repair/Replacement Engineer, or designee, is responsible for review of work orders relating to IWA-4000 and for preparation of required Repair/Replacement reports.

1.6.4 Authorized Nuclear Inservice Inspector (ANII) 1.6.4.1 Duties as specified in IWA-2110, including Repair/Replacement activities for Class MC components.

2.0 ASME CODE CASE APPLICABILITY This section contains ASME Code Cases applicable to the WF3 Fourth Ten-Year CISI Interval.

2.1 Adoption of Code Cases Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 14 of 50 Code Cases adopted for ASME Section XI activities for use during the Fourth Ten-Year CISI Interval are listed in Table 2.2-1. The use of Code Cases is in accordance with ASME Section XI, IWA-2440, 10 CFR 50.55a, and Regulatory Guide 1.147. As permitted by ASME Section XI and Regulatory Guide 1.147 or 10 CFR 50.55a, ASME Section XI Code Cases may be adopted and used as described below:

2.1.1 Adoption of Code Cases Listed for Generic Use in Regulatory Guide 1.147 Code Cases that are listed for generic use in the latest revision of Regulatory Guide 1.147 may be included in the CISI Program provided any additional conditions specified in the Regulatory Guide are also incorporated. Table 2.2-1 identifies those code cases approved for generic use and adopted for the fourth ten-year interval.

2.1.2 Adoption of Code Cases Not Approved in Regulatory Guide 1.147 Certain Code Cases that have been approved by the ASME Board of Nuclear Codes and Standards may not have been reviewed and approved by the NRC Staff for generic use and listed in Regulatory Guide 1.147. Use of such Code Cases may be requested in the form of a Request for Alternative in accordance with 10 CFR 50.55a(z). Once approved, these Requests for Alternatives will be available for use until such time as the Code Cases are adopted into Regulatory Guide 1.147, at which time compliance with the conditions contained in the Regulatory Guide are required.

Table 2.3-1 identifies those Code Cases that have been requested through Requests for Alternatives, as applicable. For convenience to the user of this CISI Program, the appropriate internal correspondence number will be provided to assist in retrieval from Document Control. All other Requests for Alternatives and Relief Requests (those not associated with NRC approval of Code Cases) are addressed in Section 3.0.

2.1.3 Adoption of Code Cases Mandated by 10 CFR 50.55a The NRC may require the licensee to follow an augmented inservice inspection program for systems and components for which the Commission deems that added assurance of structural Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 15 of 50 reliability is necessary. Many times these Augmented Requirements will be contained in Code Cases that ASME has approved. The NRC may mandate their use and add conditions it believes are necessary via 10 CFR 50.55a(g)(6)(ii). Table 2.4-1 will be available to identify Code Cases Mandated by NRC in the regulation that would affect the CISI Program.

2.1.4 Use of Annulled Code Cases As permitted by Regulatory Guide 1.147(B), Code Cases that have been adopted for use in the current inspection interval that are subsequently annulled by ASME, may be used for the remainder of the interval.

2.1.5 Code Case Revisions Initial adoption of a Code Case requires use of the latest revision of that Code Case listed in Regulatory Guide 1.147. However, if an adopted Code Case is later revised and approved by the NRC, then either the earlier or later revision may be used as permitted by Regulatory Guide 1.147(B). An exception to this provision would be the inclusion of any conditions on the later revision necessary to enhance safety. In this situation, the condition imposed on the later revision must be incorporated into the program.

2.1.6 Adoption of Code Cases Issued Subsequent to Filing the Inservice Inspection Plan Code Cases issued by ASME subsequent to filing the Inservice Inspection Plan with the NRC may be incorporated within the provisions of paragraphs 2.1.1 or 2.1.2 by revision to this CISI Plan.

Any subsequent Code Cases should be added to this program and identified in either Table 2.2-1 or Table 2.3-1, as applicable, prior to their use.

2.1.7 Non Containment Inservice Inspection Code Cases Only Code Cases applicable to CISI for Class MC and Class CC are included in Table 2.2-1, 2.3-1 and 2.4-1. Code Cases applicable to Inservice Inspection, System Pressure Testing and Repair/Replacement are addressed in their associated Program Sections.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 16 of 50 2.1.8 Code Cases not approved for use by the NRC Code Cases that have been approved by the ASME Board of Nuclear Codes and Standards may not be approved by the NRC Staff for generic use and listed in Regulatory Guide 1.193, ASME Code Cases Not Approved for Use. However, the NRC may approve their use in specific cases. Code Cases listed in the Regulatory Guide will not be used at WF3 without an approved Request for Alternative in accordance with 10 CFR 50.55a(z).

2.2 Regulatory Guide 1.147, Revision 17 Approved Code Cases Regulatory Guide (RG) 1.147, Revision 17 is approved for use in 10 CFR 50.55a(a)(3)(ii) and subject to the conditions in 10 CFR 50.55a(b)(5),(b)(5)(i), (b)(5)(ii) and (b)(5)(iii).

Table 2.2-1 Code Cases Adopted from Regulatory Guide 1.147 Code Case Title NRC Conditions Number N-532-5 Repair/Replacement Activity Documentation None Requirements and Inservice Inspection Summary Report Preparation and Submission N-765 Alternative to Inspection Interval Scheduling None Requirements of IWA-2430 2.3 Code Cases Approved Through Request for Alternatives The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 17 and require a request for alternative prior to implementation. Refer to Section 3.0 of this inspection plan for the applicable requests.

Table 2.3-1 Code Cases Adopted Via NRC Approved Requests Code Case Request for Title Number Alternative No.

NA NA NA Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 17 of 50 2.4 Code Cases Mandated Via 10 CFR 50.55a The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 17, but are mandated in 10 CFR 50.55a(g)(6)(ii) as augmented requirements.

Table 2.4-1 Code Cases Mandated Via 10 CFR 50.55a Code Case Title Notes Number NA NA NA 3.0 RELIEF REQUESTS The 2007 Edition through the 2008 Addenda of ASME Section XI provides the rules for the inservice inspection of nuclear power plants. However, not all requirements of ASME Section XI are applicable, or possible to be performed, at every plant. Therefore, Entergy has reviewed the requirements contained in the 2007 Edition through the 2008 Addenda of Section XI and determined where those requirements are not viable at WF3. 10 CFR 50.55a provides two options for submission of such determinations to the Nuclear Regulatory Commission staff for review and approval.

3.1 Requests for Relief Required Due to Impracticality or Limited Examinations 10 CFR 50.55a(a)(g)(5)(iii) and (iv) allows relief to be requested in instances when a Code requirements is deemed impractical with (iv) being specific to examination requirements that are determined to be impractical. The provisions of these two paragraphs are typically used to address impracticalities like limited examination coverage. Under 10 CFR 50.55(a)(g) (5)(iv), relief requests for impractically determinations must be submitted for NRC review and approval not later than 12 months after expiration of the initial or subsequent 120-month inspection interval for which relief is sought.

In cases where conformance with ASME Section XI requirements for inservice inspection are determined to be impractical, Entergy will notify the NRC and submit information to support the determination, as required by 10 CFR 50.55a(g)(5)(iii). The submittal of this information will be Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 18 of 50 referred to as a Request for Relief. Per 10 CFR 50.55a paragraph (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate Requests for Relief per Paragraph (g)(5) and "may grant such relief and may impose such alternative requirements as it determines are authorized by law, will not endanger life or property or the common defense and security, and are 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".

10 CFR 50.55a(g)(5)(iv) ISI program update: Schedule for completing impracticality determinations. Where the licensee determines that an examination required by the Code edition or addenda is impractical, the basis for this determination must be submitted for NRC review and approval not later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.

3.2 Requests for Alternatives to Codes and Standards 10 CFR 50.55a(z)

Acceptable Level of Quality and Safety When seeking an alternative to the rules contained in 10 CFR 50.55a(b) through (h) the request is submitted under the provision of 10 CFR 50.55a(z). Once approved by the Director, Office of Nuclear Reactor Regulation, the alternative may be incorporated into the ISI program.

These types of requests are typically used to request use of Code Cases, Code Edition, or Addenda not yet approved by the NRC. Request for Alternatives must be approved by the NRC prior to their implementation or use. Within the provisions of 10 CFR 50.55a(z) there are two specific methods of submittal:

10 CFR 50.55(z)(1): Acceptable level of quality and safety: The proposed alternative would provide an acceptable level of quality and safety.

10 CR 50.55a(z)(2): Hardship without a compensating increase in quality and safety. Compliance with the specified requirements of this section would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

In cases where Entergy proposes alternatives to ASME Section XI when compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, a Request for Alternative as allowed by 10 CFR 50.55a(z), will be submitted to the NRC.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 19 of 50 3.3 Table of Requests Table 3.3-1 contains an index of Requests For Alternatives and Requests For Relief written in accordance with 10 CFR 50.55a(z) and 10 CFR 50.55a(g)(5)(iii). The applicable Entergy submittal and NRC Safety Evaluation Report (SER) correspondence numbers are also included in Table 3.3-1 for each request for alternative and request for relief.

Note: Only Requests for Alternatives or Requests for Relief applicable to the requirements for Class MC components are addressed in Table 3.3-1.

Table 3.3-1 Fourth Ten-Year CISI Interval Relief Requests Entergy NRC SER Relief Request Relief Request Description Correspondence Correspondence W3-CISI-001* Request for Alternative to ASME IWE- W3F1-2010- ILN10-0066 (ML 5221 Regarding Post Repair Testing of 0002 1003000122)

Waterford 3s Steel Containment (ML100470056) ILN10-0119 (ML Vessel Opening (RF17) 102990371)

W3-CISI-002* Request for Alternative to ASME IWE- W3F1-2011- ILN12-006 (ML 5221 Regarding Post Repair Testing of 0041 113330137)

Waterford 3s Steel Containment (ML112150195) ILN12-0045 (ML Vessel Opening (RF18) 120580756) -

corrections

• The requests for alternatives from the Third Ten Year Interval are retained for historical data only.

3.4 Requests to use Later Edition and Addenda of ASME Section XI On July 28, 2004, the NRC published Regulatory Issue Summary (RIS) 2004-12, Clarification on Use of Later Editions and Addenda to ASME OM Code and Section XI. This RIS clarifies the NRC position on using Editions and Addenda of Section XI, in whole or in part, later than those specified in the ISI program. If the desired Edition or Addenda are referenced in 10 CFR 50.55a(b)(2), the request is submitted following the guidance of the RIS.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 20 of 50 These types of request are not required to demonstrate hardship, difficulty, or provide evidence of quality and safety. They do need to ensure that all related requirements are also used. Requests to use edition and/or addenda of ASME Section XI that are referenced in 10 CFR 50.55a(b)(2) that are later than the initial Code of Record established for the ISI program shall be submitted under the provisions of 10 CFR 50.55a(g)(4)(iv).

4.0 DEVELOPMENTAL REFERENCES 4.1 U.S Nuclear Regulatory Commission Code of Federal Regulations; Title 10, Energy; Part 50, Domestic Licensing of Production and Utilization Facilities; Section 50.55a, Codes and Standards 4.2 U.S. Nuclear Regulatory Commission Code of Federal Regulations; Title 10, Energy; Part 50, Domestic Licensing of Production and Utilization Facilities; Appendix J, Primary Containment Leakage Testing for Water-Cooled Power Reactors 4.3 ASME Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components, 2007 Edition through the 2008 Addenda 4.4 ASME Section III, Rules for Construction of Nuclear Power Plant Components 4.5 NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 4.6 NRC Regulatory Information Summary 2016-07, Containment Shell or Liner Moisture Barrier Inspection 4.7 U.S. Nuclear Regulatory Commission Regulatory Guide 1.193, ASME Code Cases Not Approved for Use.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 21 of 50 5.0 ASME EXAMINATION BOUNDARIES 5.1 ASME Class MC This section defines those systems that are designated as ASME Section XI Class MC and provides justification for their inclusion or exclusion within the Fourth Ten-Year CISI Program. The 2007 Edition through the 2008 Addenda of the ASME Code, Section XI, defines the inspection requirements for each of the ASME Code Classes. The Containment ISI Drawings listed in Section 12.0 identify the IWE examination boundaries of the primary containment structure.

5.1.1 ASME Class 1, 2 and 3 ASME Class 1, 2 and 3 piping penetrating the containment vessel are attached to penetration sleeves. The piping is not within the scope of Subsections IWE. The interface between the Class 1, 2, and 3 piping and the Class MC containment is the weld joining the piping and the penetration sleeve assembly. Class 1, 2 and 3 components and the interfacing welds are inspected in accordance with Subsections IWB, IWC or IWD as applicable.

5.1.2 ASME Class MC The entire steel liner of the containment structure and connecting penetrations, appurtenances, and parts which form the containment leak tight boundary are classified as MC components. All items within the IWE program boundary are considered "Class MC" regardless of the construction or design code applicable to the component. This class shall be inspected per the requirements of ASME Section XI Subsection IWE Table IWE-2500-1 as modified by 10 CFR 50.55a, Code Cases, or Relief Requests approved by the NRC.

5.2 Exemptions 5.2.1 IWE The following components (or parts of components) are exempted from the examination requirements of IWE-2000, as specified in IWE-1220:

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 22 of 50 5.2.1.1 Vessels, parts and appurtenances outside the boundaries of the containment as defined in the Design Specification in accordance with IWE-1220(a). Essentially, only the metallic portions of the containment structure that are either pressure-retaining, provide for a leak-tight membrane, or are load-bearing are considered within the boundary of IWE. All other parts and appurtenances, such as non-pressure-retaining portions of electrical and mechanical penetrations, are exempt.

5.2.1.2 Embedded or inaccessible portions of containment vessels, parts and appurtenances that met the requirements of the original Construction Code in accordance with IWE-1220(b). Essentially, only the components that are accessible by visual line of site with adequate lighting from permanent vantage points without being obstructed by permanent plant structures, equipment, or components are required to be examined per IWE-1232(c). Embedded parts include those that have been covered with concrete since the construction of the containment.

5.2.1.3 Portions of containment vessels, parts and appurtenances that become embedded or inaccessible as a result of vessel repair/replacement activities if the conditions of IWE-1232(a) and (b) and IWE-5220 are met.

IWE-1232(a)(1) no openings or penetrations are embedded in the concrete; IWE-1232(a)(2) all welded joints that are inaccessible for examination are double butt welded and are fully radiographed; and, prior to being covered, are tested for leak tightness using a gas medium test, such as Halide Leak Detector Test; IWE-1232(a)(3) the vessel is leak rate tested after completion of construction or repair/replacement activities to the leak rate requirements of the Design Specifications.

5.2.1.4 Piping, pumps and valves that are part of the containment system, or which penetrate or are attached to the containment vessel are specifically exempted from the Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 23 of 50 examination requirements of IWE in accordance with IWE-1220(d). These components shall be examined in accordance with the requirements of Subsections IWB or IWC as appropriate to the classification defined by the Design Specifications.

6.0 APPLICATION CRITERIA AND CODE COMPLIANCE 6.1 Examination Methods Examination methods which will be used to satisfy Code examination requirements for nonexempt Class MC components are provided below.

The examinations conducted under the CISI Program are performed to meet the requirements of ASME Section XI Subsections IWE as modified by 10 CFR 50.55a. The following modifications apply to the 2007 Edition through 2008 Addenda.

In accordance with 10 CFR 50.55a(b)(2)(ix), Metal containment examinations. Applicants or Licensees applying Subsection IWE, 2007 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, must satisfy the requirements of paragraphs (b)(2)(ix)(A)(2) and (b)(2)(ix)(B) and (J) of this section.

In accordance with 10 CFR 50.55a(b)(2)(ix)(A)(2), (A)(2)(i), A(2)(ii), and (A)(2)(iii), Metal containment examinations: First provision. The applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified for evaluation, the applicant or licensee must provide the following in the ISI Summary Report as required by IWA-6000:

(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; (3) A description of necessary corrective actions.

Note: WF3 will provide the required information above in 10 CFR 50.55a(b)(2)(ix)(A)(2) for the Metallic Liner as part of the Owners Activity Report, Form OAR-1 in accordance with NRC approved Code Case N-Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 24 of 50 532-5 in lieu of the ISI Summary Report required in IWA-6000.

In accordance with 10 CFR 50.55a(b)(2)(ix)(B) Metal containment examinations: Second provision. When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination. (Note that the Table IWA-2210-1 referenced in the 10CFR50.55a section above is only in 1992 edition of the ASME Code; this table has been updated to a different reference number in the 2007 edition of the code being implemented in this plan and would be IWA-2211-1.)

In accordance with 10 CFR 50.55a(b)(2)(ix)(J) Metal containment examinations: Tenth provision. In general, a repair/replacement activity such as replacing a large containment penetration, cutting a large construction opening in the containment pressure boundary to replace steam generators, reactor vessel heads, pressurizers, or other major equipment; or other similar modification is considered a major containment modification. When applying IWE-5000 to Class MC pressure-retaining components, any major containment modification or repair/replacement, must be followed by a Type A test to provide assurance of both containment structural integrity and leaktight integrity prior to returning to service, in accordance with 10 CFR part 50 Appendix J, Option A or Option B on which the applicants or licensees Containment Leak-Rate Testing Program is based. When applying IWE-5000, if a Type A, B, or C Test is performed, the test pressure and acceptance standard for the test must be in accordance with 10 CFR part 50, Appendix J.

Entergy will apply the station 10 CFR 50 Appendix B Quality Assurance Program to Section XI activities. Therefore the ASME Section XI condition in 10 CFR 50.55a(b)(2)(x) will not apply at WF3, In accordance with IWA-2240 and IWE-2500(a), alternative examination methods may be used provided the ANll is satisfied that the results are demonstrated to be equivalent or superior to the results of the method specified by Subsection IWE. The 2008 Addenda of IWA-2240 must be used in accordance with 10 CFR 50.55a(b)(2)(xix) which states:

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 25 of 50 10 CFR 50.55a(b)(2)(xix) Selection XI condition: Substitution of alternative methods. The provisions in IWA-4520(b)(2) and IWA-4521 of the 2008 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, allowing the substitution of ultrasonic examination for radiographic examination specified in the Construction Code, are not approved for use.

Approved alternative examinations shall be documented by incorporation into this program section by a Program Change Notice (PCN).

6.1.1 IWE Examinations 6.1.1.1 Personnel performing IWE examinations shall be qualified in accordance with written procedures prepared as required by IWE-2300, Entergys Written Practice CEP-NDE-100, Administration and Control of ENS NDE or approved vendor written practice for certification and qualification of NDE personnel.

6.1.1.2 General visual examinations per IWE-2311 are conducted to assess the general condition of containment surfaces.

General visual examination shall be performed with adequate illumination to detect evidence of degradation.

General Visual Examinations for Subsection IWE shall be performed in accordance with CEP-CII-003, Visual Examinations of Class MC Components.

6.1.1.3 VT-3 Visual Examinations per IWE-2312 are conducted to access the condition of wetted surfaces of submerged areas and to access the condition of vent system surfaces of BWR containments.

6.1.1.4 VT-1 Visual Examinations per IWE-2313 are conducted:

(1) to access the initial condition of surfaces requiring augmented examinations in accordance with IWE-1241 and to determine the magnitude and extent of any deterioration and distress of these surfaces during subsequent augmented examinations (2) to determine the condition of inaccessible areas [IWE-1232(c)] when conditions are initially detected in Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 26 of 50 accessible areas that could indicate the presence of, or result in, degradation to such inaccessible areas; and (3) In accordance with IWE-2500 and Table IWE-2500-1, Examination Category E-G, to access the condition of containment pressure retaining bolting.

6.1.1.5 If a volumetric examination is performed to detect discontinuities in the volume of a material and material thickness, the Ultrasonic (UT) examinations shall be conducted as required using ultrasonic thickness measurement method specified in ASME Boiler and Pressure Vessel Code Section XI Appendix I.

6.1.1.5.1 When ultrasonic thickness measurements are performed, grids not exceeding one foot square shall be used. The number and location of the grids shall be determined by the CISI Program Owner.

6.1.1.5.2 Ultrasonic thickness measurements shall be used to determine the minimum wall thickness within each grid. The location of the minimum wall thickness within each grid shall be marked or recorded such that periodic reexamination can be performed in accordance with the requirements of Table IWE-2500-1, Examination Category E-C. A sampling plan may be used to determine the number and location of ultrasonic thickness measurement grids within each contiguous examination area provided. The UT examinations shall be performed utilizing either manual or mechanized UT techniques.

6.1.1.6 When access or other conditions prevent direct examination, remote visual examination can be substituted for direct examination provided that the requirements of IWA-2211(g) and IW A-2213(g) are met.

IWA-2211(g) and IW A-2213(g) requires that the selected test characters of Table IWA-2211-1 can be resolved as a part of the remote examination procedure Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 27 of 50 demonstration. Additionally, the remote examination system shall have the capability of distinguishing and differentiating between the colors applicable to the component examination being conducted. Remote visual examination aids include but are not limited to mirrors, telescopes, periscopes, borescopes, fiberoptics, and Closed Circuit Television (CCTV) systems with or without permanent recording capabilities. Refer to Section 6.1 addressing 10 CFR 50.55a(b)(2)(ix)(B), when performing remotely the visual examinations required by Subsection IWE for use of maximum direct examination distance specified in Table IWA-2210-1 and the minimum illumination requirements specified in Table IWA-2210-1.

(Note that the Table IWA-2210-1 referenced in the 10CFR50.55a section above is only in 1992 edition of the ASME Code; this table has been updated to a different reference number in the 2007 edition of the code being implemented in this plan and would be IWA-2211-1.)

6.2 Examination Category IWE The following provides a summary of the application of ASME Code, Section XI, 2007 Edition through the 2008 Addenda, for the WF3 CISI Program, Unit 3, Ten-Year Program for the Fourth Inspection Interval.

The results of this application are summarized by ASME Section XI Category and Item Number and are contained with Table 6-2. Table 6-2 only contains those ASME Item Numbers that are relevant to WF3.

6.2.1 IWE Program Boundary The IWE program boundary, covering inservice inspection and repair and replacement, is defined in accordance with IWE-1100 as determined by Class MC pressure retaining components and their integral attachments required by the applicable provisions of 10 CFR 50.55a(g)(4)(v).

10 CFR 50.55a(g)(4)(v)(A) expands this boundary to include:

Metal containment pressure retaining components and their integral attachments must meet the inservice inspection, repair, and replacement requirements applicable to components that are Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 28 of 50 classified as ASME Code Class MC.

6.2.2 Inspection Intervals - Subsection IWE IWA-2430(a), (b), (d) and (e) shall be used as required for IWE Inspections in accordance with ASME Section XI, 2007 Edition through the 2008 Addenda. The applicable alternative requirements of NRC approved Code Case N-765 will be used in lieu of the requirements of IWA-2430(c). All Subsection IWE examinations shall be completed during each of the inspection intervals for the service lifetime of the power unit. The inspection program shall conform to IWA-2431 with the alternatives to IWA-2430(c) from Code Case N-765.

IWA-2430 (a) The inservice examinations and system pressure tests required by IWB, IWC, IWD, IWE, and inservice examinations and tests of IWF shall be completed during each of the inspection intervals for the service lifetime of the plant. The inspections shall be performed in accordance with the schedule of the Inspection Program of IWA-2431.

(b) The inspection interval shall be determined by calendar years following placement of the plant into commercial service.

(d) In addition to IWA-2430(c), for plants that are out of service continuously for 6 months or more, the inspection interval during which the outage occurred may be extended for a period equivalent to the outage and the original pattern of intervals extended accordingly for successive intervals.

(e) The inspection intervals for items installed by repair /

replacement activities shall coincide with remaining intervals, as determined by the calendar years of plant service at the time of the repair /replacement activities.

Code Case N-765 Alternatives for IWA-2430(c)

(a) Each Inspection interval may be extended by as much as one year, and may be reduced without restriction, provided the examinations required for the interval have been completed.

Successive intervals shall not extend more than one year Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 29 of 50 beyond the original pattern of ten-year Intervals, and shall not exceed eleven years in length. For extended intervals, neither the start and end dates nor the inservice inspection program for the successive interval need be revised.

(b) Examinations may be performed to satisfy the requirements of an extended interval in conjunction with examinations performed to satisfy the requirements of the successive interval.

However, an examination performed to satisfy requirements of either the extended interval or the successive interval shall not be credited to both intervals.

(c) That portion of an Inspection interval described as an inspection period may be extended by as much as one year, and may be reduced without restriction, provided the examinations required for that period have been completed. This adjustment shall not alter the requirements for scheduling Inspection Intervals.

(d) The Inspection Interval for which an examination was performed shall be identified on examination records.

6.2.3 Inspection Schedule - Subsection IWE Per IWA-2420, inspection plans and schedules shall be prepared for the first inservice inspection interval and subsequent inservice inspection intervals. Per IWA-2420(b), an implementation schedule for performance of examinations and tests shall be prepared for each inspection plan.

Per N-765, each Inspection interval may be extended by as much as one year, and may be reduced without restriction, provided the examinations required for the interval have been completed. Successive intervals shall not extend more than one year beyond the original pattern of ten-year intervals, and shall not exceed eleven years in length. For extended intervals, neither the start and end dates nor the inservice inspection program for the successive interval need be revised.

Per IWA-2430(d), In addition to N-765, for plants that are out of service continuously for 6 months or more, the inspection interval during which the outage occurred may be extended for a period equivalent to the outage and the original pattern of intervals extended accordingly for successive intervals.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 30 of 50 Subarticle IWE-2400 includes the requirements for the scheduling of examination and tests for Class MC components and Metallic Liners of Class CC components.

Specific scheduling criteria are included in IWE-2411. This paragraph references Table IWE-2411-1, which includes minimum and maximum percentages of examinations required to be completed by each inspection period. Table 1.4-1 summarizes this information.

Per IWE-2411(a), examinations listed in Table IWE-2500-1 as deferrable to the end of the inspection interval are not required to meet the criteria in Table IWE-2411-1.

Per IWE-2420(a), the sequence of component examinations established during the first inspection interval shall be repeated during each successive inspection interval, to the extent practical. The sequence of component examinations may be modified in a manner that optimizes scaffolding, radiological, insulation removal, or other considerations, provided that the percentages of Table IWE-2411-1 are maintained.

6.2.4 IWE EXAMINATION CATEGORY E-A - CONTAINMENT SURFACES Examination Category E-A of the ASME Code Section XI, 2007 Edition through the 2008 Addenda, requires examination of Class MC "Metallic Containment" pressure-retaining components and their integral attachments, as well as, the metallic shell and penetration liners of Class CC "Concrete Containment" pressure-retaining components and their integral attachments.

The Containment Inspection Engineer selects examinations for a given interval and period in accordance with the requirements of ASME Section XI and 10 CFR 50.55a.

Accessible containment surfaces, pressure-retaining bolted connections and moisture barriers shall be examined in accordance with Table IWE-2500-1 Examination Category E-A.

The accessible surface areas, pressure retaining bolted connections and moisture barriers are identified in the Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 31 of 50 Containment ISI drawings listed in Appendix A. Identifiers and Examination Selections for these IWE components are included in the IDDEAL database.

WF3 Moisture Barrier examinations were scheduled previously based on CRs (Reference CR-W3-2000-1275 and CR-W3-2000-1375). Degradation of the Moisture Barrier was previously identified during the first period of the second interval (RF10). In addition to the requirements of Table IWE-2500-1, Owner Elected Examinations were performed every outage from RF10 through RF16. Due to the degradation remaining essentially unchanged over this time period, these areas no longer required augmented examinations in accordance with IWE-2420(c).

Note: Category E-A, Item No. E1.12 refers to Containment Vessel Wetted Surfaces of Submerged Areas and is not applicable to W3.

Note: Category E-A, Item No. E1.20 refers to Vent Systems Accessible Surface Areas of BWR containments and is not applicable to WF3.

6.2.4.1 10 CFR 50.55a Limitations and Modifications The requirements of Table IWE-2500-1 Examination Category E-A, are modified by 10 CFR 50.55a as follows:

Per 10 CFR 50.55a(b)(2)(ix)(B), Metal containment examinations: Second provision. When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

(Note that the Table IWA-2210-1 referenced in the 10CFR50.55a section above is only in 1992 edition of the ASME Code; this table has been updated to a different reference number in the 2007 edition of the code being implemented in this plan and would be IWA-2211-1.)

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 32 of 50 6.2.4.2 Acceptance Criteria - IWE-3510 Acceptance of general visual examinations is accomplished by an acceptance review by the Responsible Individual (RI) in accordance with IWE-3511 for Coated and Non-coated areas, and per the requirements of IWE-3512, for Moisture Barriers and IWE-3513 for Visual Examination, VT-3.

6.2.5 IWE EXAMINATION CATEGORY E-C - CONTAINMENT SURFACES REQUIRING AUGMENTED EXAMINATION Containment surface areas subject to accelerated degradation and aging require the augmented examinations identified in Table IWE-2500-1, Examination Category E-C. In accordance with IWE-2420(b), examinations accepted by evaluation per IWE-3000 shall be examined under Category E-C in the next inspection period. Whenever WF3 has an area(s) requiring examination under Category E- C, Identifiers and Examination Selections for these IWE components will be included in the IDDEAL database as Category E-C Item No. E4.11 and/or Item No. E4.12 as follows:

For surfaces where the side requiring augmented examination is not accessible for visual examination, ultrasonic thickness measurements shall be performed in accordance with Examination Category E-C, Item No. E4.12, and in accordance with IWE-2500(b)(2), IWE-2500(b)(3) and IWE-2500(b)(4).

The examination(s) must be performed once per period until the areas examined remain essentially unchanged for the next inspection period. In accordance with Table IWE-2500-1, Examination Category E-C, Note 2 and IWE-2420(c), if the areas examined remain essentially unchanged, they no longer require Examination Category E-C examination.

6.2.5.1 Identification of IWE Augmented Examination of Containment Surface Areas- Examination Category E-C.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 33 of 50 Whenever WF3 has an area(s) requiring examination under Category E- C, the area shall be identified in the IDDEAL database.

6.2.5.2 Tracking of IWE Augmented Examination Areas IWE-1241 requires augmented examination of surface areas subject to accelerated degradation and aging identified in Table IWE-2500-1, Examination Category E-C, or surface areas accepted by evaluation as specified in IWE-2420(b).

When it is determined that a given surface area requires augmented examination, the area shall be added to the IDDEAL database. Areas added to the augmented examination table due to the provisions of IWE-2420(b) may be removed from the IDDEAL database when the provisions of IWE- 2420(c) have been met.

Augmented examination areas added to the IDDEAL database table due to the provisions of IWE-1241(a), (b) or (c) may be removed from the IDDEAL database only after determination that the area is no longer subject to accelerated degradation and aging as described in IWE-1241(a), (b) or (c) as applicable.

Removal and/or deactivation of areas from the IDDEAL database shall be documented as required by CEP-COS-0110.

6.2.5.3 Acceptance Criteria - IWE-3520 Acceptance for augmented areas is accomplished by an acceptance review by the RI per the requirements of IWE-3521, for Visual Examination, VT-1 and IWE-3522, for Ultrasonic Examination.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 34 of 50 NOTE SEP-CISI-104 does not authorize changes to plant configuration. All removal and replacement of moisture barrier and coating material associated with the supplementary examination must be performed in accordance with applicable work and configuration controls.

6.2.6 IWE EXAMINATION CATEGORY E-G - Pressure Retaining Bolting Examinations shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes. Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled.

6.2.6.1 Acceptance Criteria - IWE-3530 Acceptance of VT-1 visual examinations is accomplished by an acceptance review by the RI per the requirements of IWE-3531, for Visual Examination, VT-1.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 35 of 50 Table 6-2 Waterford Unit 3 Code Category Summary Number Number to Number to Number to Required to Number of Examination Examined or be be be Item be Examined Category Description Exam Method Components Percentage Scheduled Examined Examined Examined Number During in Item No. Required During the in First in Second in Third Interval Interval Period Period Period E-A Containment Surfaces Containment Vessel Each Pressure Retaining General (3)

E-A E1.11 24 72 100% Inspection 24 24 24 Boundary Accessible Visual Period Surface Areas(1)

Each General E-A E1.30 Moisture Barriers(2) 15 45(3) 100% Inspection 15 15 15 Visual Period Category Total 39 117 39 39 39 Notes for Note 1. Examinations shall include all accessible interior and exterior surfaces of Class MC components, parts, and appurtenances, and Cat. E-A metallic shell and penetration liners of Class CC components.

Note 2: Examination shall include moisture barrier materials intended to prevent intrusion of moisture against inaccessible areas of the pressure retaining metal containment shell or liner at concrete-to metal interfaces and at metal-to-metal interfaces which are not seal welded. Containment moisture barrier materials include caulking, flashing, and other sealants used for this application.

Note 3: Examination of this item number is required each period. Therefore the number required during the interval is three times the total number of components. This is also reflected in the category total.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 36 of 50 Table 6-2 Waterford Unit 3 Code Category Summary Number Number to Number to Required to Number of Examination Examined or be be Number to be Item be Examined Category Description Exam Method Components Percentage Scheduled Examined Examined Examined in Number During in Item No. Required During the in First in Second Third Period Interval Interval Period Period E-C Containment Surfaces Requiring Augmented Examination 100% of Containment Surface surface areas E-C E4.11 VT-1 0 0 100% 0 0 0 Areas Visible Surfaces identified IWE-1242(1) 100% of minimum wall Containment Surface thickness Area Grid Minimum Ultrasonic E-C E4.12 0 0 100% locations 0 0 0 Wall Thickness Thickness during each Location inspection period(2)

Category Total 0 0 0 0 0 Notes for Cat. E-C Note 1: Containment surface areas requiring augmented examinations are those identified in IWE-1240.

Note 2: The extent of examination shall be 100% for each inspection period until the areas examined remain essentially unchanged for the next inspection period. Such areas no longer require augmented examination in accordance with IWE-2420(c).

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 37 of 50 Table 6-2 Waterford Unit 3 Code Category Summary Number Number to Number to Required to Number of Examination Examined or be be Number to be Item be Examined Category Description Exam Method Components Percentage Scheduled Examined Examined Examined in Number During in Item No. Required During the in First in Second Third Period Interval Interval Period Period E-G Pressure Retaining Bolting 100% of each E-G E8.10 Bolted Connections(1)(2) VT-1 27 27 100% bolted 0 0 27 connection Category Total 27 27 0 0 27 Notes for Cat. E-G Note 1: Examination shall include bolts, studs, nuts, bushings, washers, and threads in base material and flange ligaments between fastener holes.

Note 2: Examination may be performed with the connection assembled and bolting in place under tension, provided the connection is not disassembled during the interval. If the bolted connection is disassembled for any reason during the interval, the examination shall be performed with the connection disassembled.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 38 of 50 6.3 Evaluations If examination results require evaluation, the evaluation and associated report shall be performed in accordance with ASME Section XI 2007 Edition 2008 Addenda Article IWE-3000 and the regulatory amendments in 10 CFR 50.55a.

Evaluations shall be performed by the Responsible Individual (RI).

Acceptance of components for continued service shall be subject to the rules of Articles IWE-3000.

6.3.1 Acceptance Reviews Unlike Subsections IWB, IWC and IWD; Subsections IWE does not provide detailed acceptance standards for many of the required examinations. Instead, IWE-3500 specifies that the owner shall define the acceptance criteria for many of the examinations.

With the exception of wall thickness criteria mentioned in IWE-3122.3(a) and IWE-3522, no numerical acceptance standards are provided for Class MC components by IWE.

Entergy will ensure that this standard is met by having a Responsible Engineer (RE) conduct acceptance reviews of examination results.

The requirement to conduct an acceptance review does not prohibit the RE from personally performing the examination.

In lieu of detailed acceptance criteria, IWE w i l l rely on the expertise and engineering judgment of the Responsible Engineer (RE) to detect conditions which could affect the leak tightness or structural integrity of the containment or prevent an inspected component from performing its intended function to protect containment integrity. The acceptance review criteria are as follows:

6.3.2 Screening Criteria The RE may designate screening criteria for a particular examination method and/or a particular component. Screening criteria provide the examiner with RE guidance on indications that are not relevant to the acceptance review. Unless specified within the screening criteria, indications that are less severe than the screening criteria are not required to be reported on the examination record and do not require RE acceptance review.

6.3.3 Conduct of Acceptance Reviews for Class MC Components 6.3.3.1 The RI reviews the examination data for:

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 39 of 50 a) Conditions which could affect the leak tightness or structural integrity of the containment or prevent an inspected component from performing its intended function to protect containment integrity; b) Conditions which would violate the design basis of the containment; c) Conditions in accessible areas which could indicate the presence of or result in degradation of inaccessible areas.

6.3.3.2 If the RI determines the leak tightness or structural integrity of containment could be compromised by the indicated condition or that a non-structural component (such as a seal, gasket, or moisture barrier) may not carry out its intended containment function, then:

a) The item is not acceptable for continued service without further evaluation.

b) The RI or RE (or designee) shall prepare a condition report.

As a minimum the corrective actions shall include an evaluation of:

1) The acceptability of the item for continued service;

2) The nature and extent of any required repairs or replacements;

3) Whether additional component examinations are required 6.3.3.3 RE Input to OAR-1 The RE (or designee) shall provide inputs to the OAR-1 for each flaw or area of degradation in accordance with the rules outlined in Section 10.0 of this Program Section.

6.3.3.4 RE Examination Category E-C Determination For surface areas, if the flaw or area of degradation fails to meet the acceptance criteria of IWE-3000, t h e n the RE shall add the item to the Examination Category E-C Component List in the IDDEAL Database through a PDCN.

6.3.3.5 Degradation in Inaccessible Areas Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 40 of 50 If the RE determines the examination reveals conditions which could indicate the presence of or result in degradation of inaccessible areas then:

(a) The RE (or designee) shall prepare a Condition Report to evaluate the acceptability of the inaccessible area in question.

(b) The RE (or designee) shall provide inputs to the OAR-1 for each inaccessible area identified above to include:

(1) A description of the type and estimated extent of degradation and the conditions that led to the degradation; (2) An evaluation of each area, and the results of the evaluation, and; (3) A description of necessary corrective (c) If the RE (or designee) determines the examination reveals surface areas likely to experience accelerated degradation as described in IWE-1241 then the RE shall add the item to the Examination Category E-C Component List through a PDCN.

6.3.3.6 Conduct of Acceptance Reviews for Concrete Surfaces The RE reviews the examination data for:

(a) Conditions which could affect the leak tightness or structural integrity of the concrete containment; (b) Conditions which would violate the design basis of the containment; (c) Conditions in accessible areas which could indicate the presence of or result in degradation of inaccessible areas.

6.3.3.7 RE Evaluation If the RE determines that the item cannot be accepted by examination in accordance with IWL-3211, then:

(a) The item is not acceptable for continued service without further evaluation.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 41 of 50 (b) The RE (or designee) shall prepare a condition report. As a minimum the corrective actions shall include completion of the Engineering Evaluation report required by IWL-3300.

This report documents requirements of IWL-3310:

(1) The cause of the condition which does not meet the acceptance standards; (2) The applicability of the condition to any other plants at the same site; (3) The acceptability of the concrete containment without repair of the item; (4) Whether or not repair/replacement is required and, if required, the extent, method, and completion date for the repair/replacement activity; (5) Extent, nature, and frequency of additional examinations.

(c) If the RE determines the examination reveals conditions which could indicate the presence of or result in degradation of inaccessible areas, then:

(1) The RE (or designee) shall prepare a Condition Report to evaluate the acceptability of the inaccessible area in question.

(2) The RE (or designee) shall provide inputs to the OAR-1 or each inaccessible area identified above to include:

(i) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (ii) An evaluation of each area, and the results of the evaluation, and; (iii) A description of necessary corrective actions.

6.3.4 Successive Inspections Successive Inspections will be performed in accordance with IWE-2420 when a Class MC component is accepted for continued service per IWE-3122.3.

These Class MC components will be examined in accordance with Table IWE-2500-1, Category E-C.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 42 of 50 Successive Inspections are not performed on Class CC concrete components.

Plant specific successive inspections are included in the IDDEAL database Scheduleworks module for each plant.

6.3.5 Supplemental Examinations Supplemental Examinations will be performed in accordance with IWE-3200 when a Class MC component is accepted for continued service per IWE-3122.3. These Class MC components will be examined in accordance with Table IWE-2500-1, Category E-C.

Successive Inspections are not performed on Class CC concrete components.

Plant specific successive inspections are included in the IDDEAL database module for each plant.

7.0 LISTING OF DEFINITIONS 7.1 Accessible Areas - Those areas of the containment pressure retaining surface, including integral attachments, that can be examined directly or remotely (line of sight) without installation of temporary means (i.e. scaffolding or ladder) to accomplish the examination.

7.2 Augmented Examination - Subsection IWE uses the term to refer specifically to Examination Category E-C examinations.

7.3 Authorized Nuclear Inservice Inspector - A person who is employed by an Authorized Inspection Agency and who will perform the duties of the inspector in accordance with the requirements of ASME Section XI.

7.4 Coating - A surface applied material (including paint) that are used to protect an ASME Code Class MC component against corrosion.

7.5 Component - A vessel, concrete containment, pump, valve, storage tank piping system, or core support structure.

7.6 Construction Code - Nationally recognized Codes, Standards, and Specifications (e.g., ASME, ASTM, USAS, ANSI, API, AWWA, AISC, MSS, AWS) including designated Cases, providing construction requirements for an item.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 43 of 50 7.7 Engineering Evaluation - An evaluation of indications that exceed allowable acceptance standards to determine if the margins required by the Design Specifications and Construction Code are maintained.

7.8 Evaluation - The process of determining the significance of examination or test results, including the comparison of examination or test results with applicable acceptance criteria or previous results.

7.9 Examination - The performance of a nondestructive test or observation to determine the condition of a component.

7.10 Examination Category - A grouping of similar items to be examined or tested.

7.11 Fabrication - Actions by Repair/Replacement Organizations such as forming, machining, assembling, welding, brazing, heat treating, examination, testing, and inspection, but excluding design, required to manufacture, parts, appurtenances, piping subassemblies, or supports.

7.12 General Visual Examination - A visual examination performed either directly or remotely (line of sight) with sufficient illumination and resolution suitable for the local environment to assess general conditions that may affect either the structural integrity or leak tightness of the pressure-retaining component.

7.13 Inaccessible Areas - Those areas of the containment pressure retaining surface, including integral attachments, that cannot be examined directly or remotely due to permanent obstruction (i.e. embedment in concrete, interference of plant equipment or structures), or requiring the installation of a temporary means of access (i.e.: scaffolding) to accomplish the examination.

7.14 Inservice Inspection - Methods and actions for assuring the structural and pressure-retaining integrity of safety-related nuclear power plant components in accordance with the rules of Section XI of the ASME Code.

7.15 Inservice Inspection Program Owner - The individual responsible for development and oversight of station activities required to implement the inservice Inspection program.

7.16 Inspection - Verification of the performance of examinations and tests by an Inspector.

7.17 Inspection Program - The plan and schedule for performing examinations or tests as required by Section XI of the ASME Code.

7.18 Inspector - An Authorized Nuclear Inservice Inspector, except for those instances where so designated as an Authorized Nuclear Inspector.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 44 of 50 7.19 Integral Attachment - Items connected to the pressure retaining surface of containment that serve a pressure retaining function, or are a structural attachment.

7.20 Leak Tight - The ability of a component to maintain a prescribed maximum leakage rate under service conditions.

7.21 Liner - A permanent metal membrane attached to the surface of a concrete containment or penetration in order to form a leak-tight enclosure.

7.22 Nondestructive Examination - An examination by the visual, surface, or volumetric method.

7.23 Owner - The organization legally responsible for the construction and/or operation of a nuclear facility including but not limited to one who has applied for, or who has been granted, a construction permit or operating license by the regulatory authority having lawful jurisdiction.(i.e., Entergy).

7.24 Piping System - A fluid containing nuclear power plant system with boundaries defined by a plant Design Specification, and/or System Flow Diagrams.

7.25 Pressure-Retaining - Components that are necessary to maintain a pressure difference across its boundary. (Items such as seals, packing, gaskets, wear plates, and materials used for electrical conduction and insulating functions in electrical penetration assemblies are NOT considered pressure-retaining).

7.26 Primary Containment - Those parts of the containment structure (e.g.,

containment vessel, access closures, penetration sleeves, appurtenances, etc.)

that serve as a leak-tight barrier to prevent the uncontrolled release of radioactivity to the environment under normal or postulated accident conditions.

7.27 Primary Containment Component - Metallic or reinforced concrete parts (i.e.:

metallic liner, metallic penetration liners, concrete containment shell, bolting) that have been classified as ASME Code Class MC or Class CC for the purpose of performing Section XI ISI.

7.28 Regulatory Authority - A federal government agency, such as the United States Nuclear Regulatory Commission, that is empowered to issue and enforce regulations affecting the design, construction, and operation of nuclear power plants.

7.29 Structural Attachment - Items that have a pressure retaining function or are in the containment support load path and are welded, cast, or forged integrally to the inside or outside surface of the containment pressure retaining component.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 45 of 50 7.30 Structural Integrity - The ability of a structure or component to withstand prescribed loads.

7.31 Visual Examination VT A visual examination conducted in accordance with ASME Section XI, IWA-2211 to detect discontinuities and imperfections on the surface of components, including such conditions as cracks, wear, corrosion, or erosion.

7.32 Visual Examination VT A visual examination conducted in accordance with ASME Section XI, IWA-2213 to determine the general mechanical and structural condition of components and their supports by verifying parameters such as clearances, settings, and physical displacements; and to detect discontinuities and imperfections, such as loss of integrity at bolted or welded connections, loose or missing parts, debris, corrosion, wear, or erosion.

8.0 REPAIR/REPLACEMENT 8.1 Program requirements for repairs and replacements of containment items are controlled in accordance with the ASME Section XI Repair and Replacement Program and IWA-4000. Requirements are implemented by Entergy Program CEP-RR-001, ASME Section XI Repair/Replacement Program. 10 CFR 50.55a(b)(2)(ix)(J) contains specific requirements for modifications replacing penetrations or removing portions of the steel liner.

8.2 When a vessel, liner, or a portion thereof is subjected to a repair/replacement activity during the service lifetime of a plant, the preservice examination requirements for the portion of the vessel affected by the repair/replacement activity shall be met.

8.2.1 For IWE, preservice examination of repair/replacement activities shall be performed upon completion of the activity. If the repair/replacement activity is performed while the plant is not in service, the preservice examination shall be performed prior to resumption of service.

8.2.2 When a pressure test is required by IWE-5220, the preservice examination may be performed during or following the test.

8.3 Welds made as part of repair/replacement activities shall be examined in accordance with the requirements of IWA-4000, except that for welds joining Class MC components to items designed, constructed, and installed to the requirements of Section III, Class 1, 2, and 3, the examination requirements of IWB-2000, IWC-2000, or IWD-2000, as applicable, shall also apply.

8.4 Preservice examination for a repair/replacement activity may be conducted prior to installation provided:

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 46 of 50 8.4.1 The examination is performed after the pressure test required by the Construction Code has been completed; 8.4.2 The examination is conducted under conditions and with equipment and techniques equivalent to those that are expected to be employed for subsequent inservice examinations; and 8.4.3 The shop or field examination records are, or can be, documented and identified in a form consistent with that required by IWA-6000.

9.0 PRESSURE TESTING Program requirements for pressure testing of containment items, addressed in IWE-5000, are controlled in accordance with the requirements contained in 10 CFR 50 Appendix J. Pressure test following repair/replacement activities shall be performed in accordance with IWE-5220.

10.0 RECORDS AND REPORTS 10.1 Examination and test records and documented evaluation reports provide the basis for comparison with previous results and subsequent inspections. In accordance with Section XI, IWA-6000, these records and reports shall be maintained for the service lifetime of the component or system. Records and reports for the CISI Program, outage examination schedules, examination results, procedures, certifications, test, repairs, and replacements are maintained in accordance with Entergy procedures, and meet the requirements of ASME, Section XI, Article IWA-6000 and Code Case N-532-5.

10.2 IWE inspection and testing information shall be included in the ISI Summary Report required by Article IWA-6000 and the regulatory amendments in 10 CFR 50.55a. IWE repair and replacement information shall be included in the Owner's Report for Repair/Replacement Activities required by Subarticle IWA-6000 and as modified by Code Case N-532-5. Form OAR-1 is to include ASME Activities performed during the outage and the previous operating cycle. Code Case N-532-5 requires completion of the OAR-1 within 90 calendar days after completion of each refueling outage. Form OAR-1 will be prepared, maintained, and submitted in accordance with Program Section CEP-RR-001, ASME Section XI Repair/Replacement Program.

Form NIS-2A documents Repair/Replacement activities performed during the outage and the previous operating cycle. Form NIS-2A will be prepared and maintained in accordance with Program Section CEP-RR-001, ASME Section XI Repair/Replacement Program.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 47 of 50 10.3 IWE and IWL reports should be provided to the CISI Engineer within 45 days of startup following an outage.

10.4 10 CFR 50 Appendix J reports should be provided to the CISI Engineer within 60 days of startup following an outage.

10.5 Resolution and tracking of incomplete or missed examinations shall be in accordance with the Program Change Notice (PCN) process as described in EN-DC-174, Engineering Program Sections:

10.5.1 PCN Requirements A PCN to the program shall be developed in a timely manner. The PCN should contain the following:

a) The affected component.

b) The examination requirement that has not been, or will not be, met.

c) The reason for incomplete or missed examination.

d) Alternate examinations performed or recommended, (i.e. best effort vs. VT-1, etc.) to include reasons why this is acceptable.

e) For partial examinations, the extent of coverage the component received or will receive using approved methods.

10.6 Requests for Relief If a Request for Relief is required for an incomplete or missing examination, the following steps shall be completed:

a) The site should develop and submit the necessary documentation to the NRC.

b) If the Request for Relief is approved by the NRC, the approval should be incorporated into the program plan by a PCN.

c) If the Request for Relief is disapproved, the site should initiate a Condition Report (CR).

10.7 Missed Examinations Which Cannot Be Made Up If it is discovered that a required examination has been missed and cannot be performed before the end of the inspection period in accordance with the Program Plan, a Condition Report (CR) should be initiated.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 48 of 50 NOTE The period includes any extensions made within the ASME Section XI and 10 CFR 50.55a provisions. The end of the period does not occur until any applicable extensions are completed.

11.0 INSERVICE INSPECTION DATA MANAGEMENT SOFTWARE 11.1 The Inservice Inspection Data Management software is comprised of a series of program modules assembled in a comprehensive software package entitled the Iddeal Software Suite. This software suite is NETWORK based within the Entergy system and is accessible by any computer within Entergy.

11.2 The Iddeal Software Suite is a quality level B software. It is the Controlled copy of the CISI Schedule. The software has limited access to protect the contents of the databases. Access to the software is limited through the use of a user defined password controlled via Central Engineering Procedure CEP-COS-100, Control and Use of Iddeal Concepts Software.

11.3 Access to the Iddeal Software Suite is obtained through the Windows Citrix' icon entitled the Iddeal Software Suite. There are two versions of the Iddeal Software Suite. The software has limited access to protect the contents of the databases. Access to the software is limited through the use of a user defined password and Active Directory membership. The following programs are being used for the version 8 IDDEAL SOFTWARE SUITE:

IDDEAL - IDDEAL is used to track and progress completion of outage examinations. IDDEAL generates various reports to status examinations.

CERTWORKS - Maintains and tracks the personnel certifications of inspectors.

EQUIPWORKS - Maintains and tracks the NDE equipment certifications.

The following modules are being used for the version 9 IDDEAL SOFTWARE SUITE:

COMPONENTS - is used to identify all of the components that exist in the plant and being considered for the ISI Program.

LIBRARIES - is used to identify all of the requirements that could be applied to the components.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 49 of 50 PROGRAMS - is used to create and maintain program manuals, scheduling, and completion of examinations for credit to those programs.

ACTIVITIES - is used to maintain and track examinations within an outage or other activity.

EQUIPMENT - is used to maintain and track the equipment used in performing the examinations (i.e. calibration blocks).

11.4 The Programs Module contains the GGNS information necessary to perform CISI Programs. Maintenance of the Module is controlled via Central Engineering Procedure CEP-COS-0110 Control and Use of the ScheduleWorks Module of Iddeal Software. All changes to data in the GGNS ISI Database must be made in accordance with CEP-COS-0110.

The controlled Iddeal Software Suite (both versions) houses the information below which meets the requirements of ASME Section XI, IWA-2420(b).

Identification of the components selected for examination and test, including successive exams from prior periods; The Code requirements by examination category and item number for each component and the examination or test to be performed and the extent of the examination or test; Identification of drawings showing items that require examination; List of examination procedures; Description of alternative examinations and identification of components to be examined using alternative methods; Identification of calibration blocks used for ultrasonic examination of components.

12.0 COMPONENT DRAWINGS 12.1 The Component Drawings Index Appendix A identifies the components required to be examined under the rules of ASME Section XI Code, 2007 Edition through the 2008 Addenda. Other Plant Controlled drawings, marked-up sketches, and the Iddeal Software Suite database will be used when additional information is required.

Prepared for Entergy Nuclear by

Program Section No.: SEP-CISI-104 Revision No.: 2 Page No.: Page 50 of 50 WF3 CISI Drawing List - Appendix A CISI DWG. No. WF3 CISI Drawing Description IWE-01 Typical Type I Mechanical Penetration IWE-02 Typical Type II Mechanical Penetration IWE-03 Typical Type III Mechanical Penetration IWE-04 Fuel Transfer Tube (MPEN 25)

IWE-05 IWE Boundaries For Typical Electrical Penetrations IWE-06 Personnel Lock Interior View IWE-07 Personnel Lock Exterior View IWE-08 Escape Lock Interior View IWE-09 Escape Lock Exterior View IWE-10 Maintenance Hatch IWE-11 Reactor Containment Surface Inspection Layout Inside Surface IWE-12 Reactor Containment Surface Inspection Layout Outside Surface IWE-13 ASME Containment Inspection Location of Moisture Barrier Segments Prepared for Entergy Nuclear by