ML16300A036

Fourth Ten-Year Interval Inservice Inspection Program

ML16300A036
Person / Time
Site:	Byron
Issue date:	10/26/2016
From:	Kanavos M Exelon Generation Co
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML16300A036 (131)

Text

Byron Generating Station 4450 North German Church Rd Exe[on Generation wwwexeloncorp.com October 26, 2016 10 CFR 50.55a LTR:

BYRON 2016-0095 File:

1.10.0101 (JD.101, 3A.132)

United States Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Byron Station, Units 1 and 2 Renewed Facility Operating License Nos. NPF-37 and NPF-66 NRC Docket Nos. STN 50-454 and STN 50-455

Subject:

Byron Station, Units 1 and 2, Transmittal of Inservice Inspection Program for the Fourth Ten-Year Interval Enclosed is the Byron Station, Units 1 and 2, Fourth Ten-Year Interval Inservice Inspection Program. The enclosed plan replaces the Third Ten-Year Interval Inservice Inspection Program in its entirety. The Fourth Interval began July 16, 2016 and will end July 15, 2025.

Section 8 of the enclosed plan contains the fourth interval proposed alternatives to the American Society of Mechanical Engineers,Section XI, Rules for Inspection and Testing of Components of Light Water Cooled Plants, (ASME Code), 2007 Edition with the 2008 Addenda.

In accordance with 10CFR5O.55a, Codes and Standards, paragraphs 10CFR5O.55a(z)1 and 10CFR5O.55a(z)2 as applicable, Byron Relief Requests 14R-01, 14R-05, 14R-06 and 14R-08 were previously submitted for NRC review by letter dated April 15, 2016, Relief Requests Associated with the Fourth Inservice Inspection Interval (ML16106A116). Byron Relief Request 14R-09 was previously submitted for NRC review by letters dated January 28, 2016 (ML16029A003) and supplemented by letter dated June 14, 2016 (ML16167A015). The submittal and review of submitted relief requests are addressed separately from this enclosed plan. No additional review of these relief requests is required under this submittal.

Should you have any questions concerning this mailer, please contact Mr. Douglas Spitzer, Regulatory Assurance Manager, at (815) 406-2800.

Respectfully, Mark E. Kanavos Site Vice President Byron Generating Station MEK/GC/sg

Enclosure:

Byron Nuclear Power Station, Units 1 & 2, Inservice Inspection Program Fourth Ten-Year Interval

Exelon Generation Company Byron Nuclear Power Station Units 1 & 2 Prepared By Prepared For amec foster wheeler Exeton.

Amec Foster Wheeler, Inc.

ISI Program Plan Fourth Ten-Year Inservice Inspection Interval Document: BYR-525537-O1-RPO4, Rev 0 One Energy Center 40 Shuman Blvd., Suite 340 Naperville, IL 60563 Exelon Generation Company Byron Nuclear Power Station 4450 North German Church Rd.

Byron, IL 61010

IS! Program Plan Units 7 & 2, Fourth Interval TITLE:

REVISION APPROVAL SHEET ISI Program Plan Fourth Ten-Year Inspection Interval Byron Nuclear Power Station, Units 1 & 2 DOCUMENT:

BYR-525537-01 -RPO4 REVISION:

0 PROGRAM ACCEPTANCE PREPARED:

REVIEWED:

REVIEWED:

4C;/

/

72 )7//42 Inservice Inspection Program e\\

/

7/27/20/6 CISI ResØonsible mdlvi ual 7/2?//

CISI Responsible Engineer O

Pressure Testing Program CAtctrnoiJ SupporflSnubber Progran L2o]i Engineering Programs Supe isor Each time this document is revised, the Revision Approval Sheet will be signed and the following Revision Control Sheet should be completed to provide a detailed record of the revision history. The signatures above apply only to the changes made in the revision noted.

If historical signatures are required, Byron Station archives should be retrieved.

Individual programs may not be applicable to the current revision and may have NA entered.

• signature does not constitute ANII acceptance of document, it indicates acknowledgment of a pending revision. ANII acceptance will be in a formal report to the Owner documenting review per ASME Section Xl, 2007 Edition with the 2008 Addenda, Paragraph IWA-21 10(a) and (b).

/______

REVIEWED:

REVIEWED:

• REVI EW ED:

APPROVED:

Nic

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.-JL.Me, W !Je 1L_-)a t(L1nrç

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HSB ANII Acknowledgement

/_____

Exelon Byron Station Revision 0

IS) Program Plan Units 7 & 2, Fourth Interval REVISION CONTROL SHEET Major changes should be outlined within the table below. Minor editorial and formatting revisions are not requited to be logged.

REVISION [

DATE

[

REVISION

SUMMARY

0 7/29/16 Initial issuance. Developed byAmec Foster Wheeler, Inc as part of the Byron Station Fourth Interval ISI Program Update.

Prepared: S. Coleman Reviewed: M. King Approved: D. Lamond Notes:

1.

This ISI Program Plan (Sections 1

- 9 inclusive) is controlled by the Byron Nuclear Power Station Engineering Programs Group.

2.

Revision 0 of this document was issued as the Fourth Interval ISI Program Plan and was submitted to the NRC for review, including approval of the initial Fourth Interval Relief Requests. Future revisions of this document made within the Fourth Interval will be maintained and controlled at Byron Station; however, they are not required to be and will not be submitted to the NRC for approval. The exception to this is that new or revised Relief Requests shall be submitted to the NRC for safety evaluation and approval.

Exelon Byron Station Revision 0

1.31 Program Plan Units 1 & 2, Fourth Interval REVISION

SUMMARY

SECTION EFFECTIVE PAGES REVISION DATE Preface i to vi 0

7/29/16 1.0 1-ito 1-33 0

7/29/16 2.0 2-1 to 2-45 0

7/29/16 3.0 3-1 to 3-2 0

7/29/16 4.0 4-1 to 4-3 0

7/29/16 5.0 5-1 0

7/29/16 6.0 6-lto6-2 0

7/29/16 7.0 7-1 to 7-31 0

7/29/16 6.0 8-1 to 8-3 0

7/29/16 9.0 9-1 to 9-3 0

7/29/16 Exelon Byron Station i1 Revision 0

151 Program Plan Units 1 & 2, Fourth Interval TABLE OF CONTENTS SECTION DESCRIPTION PAGE

1.0 INTRODUCTION AND BACKGROUND

1-1 1.1 Introduction

1.2 Background

1.3 First Interval SI Program 1.4 Second Interval 151 Program 1.5 Third Interval ISI Program 1.6 Fourth Interval ISI Program 1.7 First Interval CISI Program 1.8 Second Interval CISI Program 1.9 Third Interval CISI Program 1.10 Code of Federal Regulations 10 CFR 50.55a Requirements 1.11 Code Cases 1.12 Relief Requests 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2-1 2.1 ASME Section Xl Examination Requirements 2.2 Augmented Examination Requirements 2.3 System Classifications and P&ID Boundary Drawings 2.4 lSl Isometric and Component Drawings for Nonexempt ISI Class Components and Supports 2.5 Technical Approach and Positions 3.0 COMPONENT 151 PLAN 3-1 3.1 Nonexempt 151 Class Components 3.2 Risk-Informed Examination Requirements 3.3 Weld Numbering 4.0 SUPPORT 151 PLAN 4-1 4.1 Nonexempt IS! Class Supports 4.2 Snubber Examination and Testing Requirements 5.0 SYSTEM PRESSURE TESTING ISI PLAN 5-1 5.1 ISI Class Systems 5.2 Risk-Informed Examination of Socket Welds 6.0 CONTAINMENT 151 PLAN 6-1 6.1 Nonexempt ISI Class Components 6.2 Augmented Examination Areas 6.3 Component Accessibility 6.4 Responsible Individual and Engineer 7.0 COMPONENT

SUMMARY

TABLES 7-1 7.1 Inservice Inspection Summary Tables Exelon Byron Station iv Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TABLE OF CONTENTS (Continued)

SECTION DESCRIPTION PAGE 8.0 RELIEF REQUESTS FROM ASME SECTION XI 8-1

9.0 REFERENCES

9-1 9.1 NRC References 9.2 Industry References 9.3 Licensee References 9.4 License Renewal References / Commitments Exelon Byron Station v

Revision 0

IS! Program Plan Units I & 2, Fourth Interval TABLE OF CONTENTS (Continued)

TABLES DESCRIPTION PAGE 1.1-1 UNIT 1 AND 2 FOURTH ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR 151 CLASS 1, 2, AND 3 COMPONENT EXAMINATIONS) 1-3 1.1-2 UNIT 1 AND 2 THIRD CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS MC COMPONENT EXAMINATIONS) 1-4 1.1-3 UNIT 1 AND 2 THIRD CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS CC-CONCRETE COMPONENT EXAMINATIONS) 1-5 1.1-4 UNIT 1 AND 2 THIRD CISI INTERVALIPERIOD/OUTAGE MATRIX (FOR ISI CLASS CC-TENDON COMPONENT EXAMINATIONS) 1-7 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS 1-15 2.2.9-1 N-722-1 TABLE 1 EXAMINATION ITEMS 2-8 2.3-1 COLOR CODED ISI P&ID BOUNDARY DRAWINGS 2-12 2.3-2 COLOR CODED ISI C&ID BOUNDARY DRAWINGS 2-13 2.4-1 UNIT 1 & COMMON 151 ISOMETRIC AND COMPONENT DRAWINGS 2-15 2.4-2 UNIT 2 ISI ISOMETRIC AND COMPONENT DRAWINGS 2-19 2.4-3 UNIT 1 CONTAINMENT ISI DRAWINGS 2-23 2.4-4 UNIT 2 CONTAINMENT ISI DRAWINGS 2-24 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX 2-26 7.1-1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE 7-3 7.1-2 UNIT 2 INSERVICE INSPECTION

SUMMARY

TABLE 7-14 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES 7-24 8.0-1 RELIEF REQUEST INDEX 8-2 Exelon Byron Station vi Revision 0

IS! Program Plan Units 1 & 2, Focidh Interval

1.1 INTRODUCTION

This Inservice Inspection (lSl) Program Plan details the requirements for the examination and testing of 1St Class 1, 2, 3, MC, and CC pressure retaining components, supports, containment structures, metal liners, and post-tensioning systems at Byron Nuclear Power Station (Byron Station), Units 1, 2, and Common. Unit Common components are included in the Unit 1 sections, reports, and tables. This ISI Program Plan also includes Containment Inservice Inspection (CISI), Risk-Informed Inservice Inspection (RI-ISI), Augmented Examinations (AUG), and System Pressure Testing (S PT) requirements imposed on or committed to by Byron Station. This ISI Program Plan is controlled and revised in accordance with the requirements of procedure ER-AA-330, Conduct of Inservice Inspection Activities, which implements the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl ISI Program. At Byron Station, the Inservice Testing (1ST) Program is maintained and implemented separately from the ISI Program. The 1ST Basis Document and 1ST Program Plan contain all applicable inservice testing requirements. Procedure ER-AA-321, Administrative Requirements for Inservice Testing, implements the 1ST Program. The Snubber Program is maintained and implemented separately from the ISI Program at Byron Station. The Snubber Program Document contains all of the applicable snubber visual examination, functional testing, and service life monitoring requirements. The 151 Program Plan is also credited as the existing program for Byron Station License Renewal Aging Management Programs (Reference Section 9.4).

The Steam Generator Inservice Inspection Plan is not included in this document except for applicable Code Cases and relief requests. A program addressing inspection requirements is maintained in separate documents and procedures. Eddy current examination of steam generator tubing is controlled and maintained under Byron Station Technical Specifications.

The ASME Section Xl RepairlReplacement Program is not included in this document except for referenced Code Cases and relief requests. The program addressing code and regulatory requirements are maintained in separate documents and procedures.

The Byron Station Flow Accelerated Corrosion (FAC) Program is not included in this document except for referenced Code Cases and relief requests. The program addressing code and regulatory requirements are maintained in separate documents and procedures.

The Byron Station Turbine Disk and Rotor Integrity Program is not included in this document except for minor references. The program addressing regulatory requirements are maintained in separate documents and procedures.

The Fourth ISI and Third CISI Intervals are effective from July 16, 2016 through July 15, 2025 for Byron Station. (See Tables 1.1-1, 1.1-2, 1.1-3, and 1.1-4 for detailed notes regarding current extensions being taken.) With the update to the ISI Program for the Fourth 151 Interval for 151 Class 1, 2, and 3 components, including their supports, the CISI Program is also being updated to its Third CISI Interval for ISI Class MC and CC components. This update will enable all of the 151 and CISI Program components I piping structural elements (elements) to be based on the same effective Edition and Addenda of ASME Section Xl, as well as share a common interval start and end date. The common ASME Code of Record for the Fourth ISI Interval and the Third CISI Interval is the 2007 Edition with the 2008 Addenda. (Note that the 1ST Program is in the Fourth 1ST Interval that is applicable from July 1, 2016 through June 30, 2026. See the 1ST Basis Document and 1ST Program Plan for further details.)

Paragraph IWA-2430(c)(1) of ASME Section Xl allows an inspection interval to be extended or decreased by as much as one year, and Paragraph IWA-2430(d) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The Exelon

- Byron Station 1-1 Revision 0

151 Pro.ç;ram Plan Units I & 2, Fourth Interval extension may be taken for a period of time not to exceed the duration of the outage. See Tables 1.1-1, 1.1-2, 1.1-3, and 1.1-4 for intervals, periods, and extensions that apply to Byron Stations Fourth IS! Interval and Third CISI Interval.

The Fourth 151 Interval and the Third CISI Interval are divided into two or three inspection periods as determined by calendaryearswithin the intervals. Tables 1.1-1, 1.1-2, 1.1-3, and 1.1-4 identify the period start and end dates for the Fourth 151 Interval and the Third CISI Interval as defined by the Inspection Program.

In accordance with Paragraph IWA-2430(c)(3),

the inspection periods specified in these Tables may be decreased or extended by as much as 1 year to enable inspections to coincide with Byron Stations refueling outages.

The inspection of lSl Class CC Components and Surfaces for the Third CISI Interval shall be performed in accordance with Paragraphs IWL-241 0 and IWL-2420. Tables 1.1-3 and 1.1-4 identify the inspection schedule.

Exelon

- Byron Station 1-2 Revision 0

151 Program Plan Units I & 2, Fourth Interval TABLE 1.1-1 UNIT I AND 2 FOURTH ISI INTERVALIPERIODIOUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3 COMPONENT EXAMINATIONS)

Unit I Period Interval Period Unit 2 Outage Projected Outage Start Date to End Date Start Date to End Date Start Date to End Date Projected Outage Outage Number Start Date Start Date Number B1R21 Spring 2017 Fall 2017 B2R20 (Start 4th ISI Interval) 1s 1st (Start 4th ISI Interval) 07/16/16 to 07/15/19 07/16/16 to 07/1 5/19 B1R22 Fal12018 Spring2Ol9 62R21 4th (Unit 1)

B1R23 Spring 2020 07/16/16 to 07/15/251 Fall 2020 62R22 2

07/16/19 to 07/15/22 07/16/19 to 07/1 5/22 B1R24 Fall 2021 4th (Unit 2)

Spring 2022 B2R23 07/16/16 to 07/15/251 B1R25 Spring 2023 Fall 2023 B2R24 3rd 3rd Fall 2024 07/16/22 to 07/15/25 07/16/22 to 07/15/25 Spring 2025 B1R26 (End 4th ISI Interval)

(End 4th ISI Interval)

B2R25 Note 1 The Byron Station Units 1 and 2 Fourth ISI Interval was reduced by one year as permitted by Paragraph IWA-2430(c)(1) in order to coincide with the plant refueling outage schedule. From the Byron Station Unit 1 ISI interval history for extensions/reductions from the commercial service date (September 16, 1985) and prior ISI intervals, the July 15, 2025 planned end date will result in Unit 1 being two months prior to the rolling ten-year ISI interval date, whereas the Unit 2 ISI interval end date will be six months early from the established sequence of intervals based on the Relief Request l3R-01 ISI interval start date (January 16, 2006).

Exelon

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151 Program Plan Units 7 & 2, Fourth Inteival TABLE 1.1-2 UNIT I AND 2 THIRD CISI INTERVALIPERIODIOUTAGE MATRIX (FOR 151 CLASS MC COMPONENT EXAMINATIONS)

Unit I Period Interval Period Unit 2 Start Date to Outage Projected Outage Start Date to End Date Start Date to End Date Projected Outage Outage Number Start Date End Date Start Date Number Spring 2017 Fall 2017 B2R20 B1R21 (Start 3td CISI Interval) 1st 1st (Start 3rd CISI Interval) 07/16/16 to 07/15/19 07/16/16 to 07/15/19 B1R22 Fall 2018 Spring 2019 B2R21 2 (Unit 1)

B1R23 Spring 2020 07/16/16 to 07/15/251 Fall 2020 B2R22 2n 2nd 07/16/19 to 07/15/22 07/16/19 to 07/15/22 B1R24 Fall 2021 2nd (Unit 2)

Spring 2022 B2R23 07/16/16 to 07/15/251 B1R25 Spring 2023 Fall 2023 B2R24 3rd 3rd Fall 2024 07/16/22 to 07/15/25 07/16/22 to 07/15/25 Spring 2025 B1R26 (End 3rd CISI Interval)

(End 3rd CISI Interval)

B2R25 Note 1: The Byron Station Units 1 and 2 Third CISI Interval was reduced by one year as permitted by Paragraph IWA-2430(c)f 1) in order to coincide with the plant refueling outage schedule.

For the Byron Station Units 1 and 2 CISI interval, the July 15, 2025 planned end date will result in the Units 1 and 2 CISI interval end date being six months early from the established sequence of intervals based on the Relief Request 13R-01 CISI interval start date (January 16, 2006).

Exelon

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IS! Program Plan Units 7 & 2, Fourth lntetval TABLE 1.1-3 UNIT I AND 2 THIRD CISI INTERVAL/PERIODIOUTAGE MATRIX (FOR 1St CLASS CC-CONCRETE COMPONENT EXAMINATIONS)

Unit 1 5-Year Period Interval 5-Year Period Unit 2 Outage Projected Outage Start Exam #

- Date Start Date to Exam #

- Date Projected Outage Start Outage Number Date or (2 Year Window)

End Date (2 Year Window)

Date or Number Outage Duration Outage Duration B1R18 FalI2Ol2 NoSectionXlExams 2d(unit1)

NoSectionXlExams Sprng2013 B2R17 Bi R19 Spring 2014 No Section Xl Exams 07/16/06 to 10/15/16 No Section Xl Exams Fall 2014 B2R18 2 (Unit 2)

B1R20 Fa112015 401/16/16 07/16/O6to 10/15/16 406/12/16 Spring2Dl6 B2R19 (End 2 CISI Interval)

(01)16/15 to 01)15/17)123 (06/12/1 5 to 06/1 1/17)123 (End 2 CISI Interval)

(Start 3td cs Interval)

(Start 3 CISI Interval)

B1 R21 Spring 2017 No Section XI Exams No Section Xl Exams Fall 2017 B2R20 B1R22 Fall 2018 No Section XI Exams No Section Xl Exams Spring 2019 B2R21 B1R23 Spring 2020 No Section Xl Exams 07/16/16 to 07/15/25 No Section XI Exams FaIl 2020 B2R22 B1R24 Fal12021 5tn01/16/21 3td(Unjt2) 5-06)12)21 Spring2O22 B2R23 (01/16/20 to 01/15/22)13 07/16/16k to 07/15/25 (06/12/20 to 06/1 1/22)13 Bi R25 Spring 2023 No Section Xl Exams No Section XI Exams FaIl 2023 B2R24_

Bi R26 Fall 2024 No Section XI Exams No Section XI Exams Spring 2025 B2R25 (End 3td CISI Interval)

(End 3td CISI Interval)

B1R27 (Start 4th CISI Interval) 6tn

- 01/16/26 h

6

- 06/12/26 (Start 4th CISI Interval)

B2I2O Spring 2026 (01/16/25 to 01/15/27)123 41 (Unit 1)

(06/12/25 to 06/11/27)1.13 Fall 2026 07/16/25 to 07/15/35 Bi R28 Fall 2027 No Section XI Exams 4th (Unit 2)

No Section XI Exams Spring 2028 B2R27 B1R29 Spring 2029 No Section Xl Exams 07/16/25 to 07/15/35 No Section Xl Exams Fall 2029 B2R28 Note 1: The Subsection IWL examination schedule for lSl Class CC concrete surfaces meets the requirements of Subarticle IWL-2400. Paragraph IWL-251 0 examinations will be performed once every 5 years. They will begin not more than 1 year prior to the specified date and will be completed not more than 1 year after such date. The initial Subsection IWL concrete examinations for each unit were required to be completed between September 9, 1996 and September 8, 2001 by 10 CFR 50.55a. The rolling 5 year examination date and associated 2 year window for each unit is determined from these first examination dates (01/16/01 and 06/12/01 for Units 1 and 2, respectively). Therefore, the schedule of the concrete surface examinations is relative to the Exelon

- Byron Station 1-5 Revision 0

IS! Program Plan Units I & 2, Fourth Interval TABLE 1.1-3 UNIT I AND 2 THIRD CISI INTERVALIPERIODIOUTAGE MATRIX (FOR ISI CLASS CC-CONCRETE COMPONENT EXAMINATIONS) baseline (1st 5-Year Period) concrete surface examinations that were completed when the use of the requirements of Subsection IWL of ASME Section Xl was initially mandated.

Note 2: The ISI Class CC concrete surface examination 2 Year Window will straddle the 2nd and 3rd CISI Intervals, as well as, the 3rd and 4th CISI Intervals.

Therefore, any examinations performed before or after the interval start date should coincide with the Subsection IWL requirements of the approved Code of Record for that given interval. Any outage required ISI Class CC concrete surface examinations should be performed in Bi R20 (Unit 1) and B2R1 9 (Unit 2), and B1R27 (Unit 1) and B2R26 (Unit 2) to fall within the 2 Year Window.

Note 3 All Byron Station ISI Class CC concrete surfaces should be accessible for examination during operational periods and completion of examinations should not be outage dependent.

Note 4: The Byron Station Units 1 and 2 Second CISI Interval for IWL-concrete was extended by six months as permitted by Paragraph IWA-2430(d)(1) in order to coincide with the plant refueling outage schedule. (Note that the Byron Station Units 1 and 2 Second CISI Interval was extended three extra months to coincide with the IWL-concrete examination schedule; however, the extra three months will not roll-over to the next Third CISI Interval for IWL-concrete.)

Exe/on

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IS! Program Plan Units 7 & 2, Fourth Inteival TABLE 1.1-4 UNIT I AND 2 THIRD CISI INTERVALIPERIODIOUTAGE MATRIX (FOR ISI CLASS CC-TENDON COMPONENT EXAMINATIONS)

Unit 1 5-Year Period Interval 5-Year Period Unit 2 Outage Projected Outage Start Exam # - Date Start Date to Exam #

- Date Projected Outage Start Outage Number Date or (2 Year Window)

End Date (2 Year Window)

Date or Number Outage Duration Outage Duration B1R21 Spring 2017 No Section Xl Exams (Start 3rd CISI Interval)

(Start 3a CISI Interval)

No Section XI Exams Fall 2017 B2R20 81R22 FaIl 2018 35td

- 09/11/18 No Section Xl Exams Spring 2019 B2R21 (09/11/17 to_09/10/19)1 3td (Unit 1) 35

- 05/27/20 FaIl 2020 B2R22 81 R23 Spring 2020 No Section XI Exams 07/1 6/16 to 07/1 5/25 (05/27/19 to 05/26/21 )12J B1 R24 Fall 2021 No Section XI Exams 3rd (Unit 2) 07/16/16 to 07/15/25 No Section Xl Exams Spring 2022 B2R23 81 R25 Spring 2023 4o

- 09/11/23 No Section XI Exams FaIl 2023 B2R24 (09/11/22 to_09/10/24)12 (End 3rd CISI Interval) 4Qth

- 05/27/25 Spring 2025 B2R25 81 R26 FaIl 2024 No Section Xl Exams (05/27/24 to 05/26/26)12 (End 3rd CISI Interval) 81 R27 Spring 2026 No Section XI Exams (Start 4th CISI Interval) 82R26 (Start 4tfl CISI Interval)

No Section Xl Exams FaH 2026 B1R28 Fail 2027 45-09/11/28 4

(Unit 1)

No SectionXl Exams Spring 2028 B2R27 (09/11/27 to 09/10/29)12 07/16/25 to 07/15/35 81 R29 Spring 2029 4th (Unit 2) 45th

- 05/27/30 Fall 2029 B2R28 No Section XI Exams 07/16/25 to Q7/j5/354 (05/27/29 to 05/26/31)12J B1R3O Fall 2030 Spring 2031 82R29 No Section XI Exams No_Section_Xl_Exams Note 1 The Subsection IWL examination schedule for ISI Class CC post-tensioning system meets the requirements of Subarticle IWL-2400. Paragraph IWL-2520 examinations will be performed once every 5 years based on a roiling 5 year frequency (÷1-1 year) from the date of completion of the previous examinations (09/07-11/83 and 05/23-27/85 for Units 1 and 2, respectively) under the Byron Station Tendon Surveillance program. These original dates were based on the initial Structural Integrity Tests (SITs).

Note 2: ASME Section XI Item Number L2.10 and L2.20 physical tests and examinations are performed during this surveillance. These tests and examinations are performed every other 5-year period for each individual Unit such that the two Units alternate every five years. Byron Station meets the requirements of ASME Section Xl, Paragraph IWL-2421, Sites with Multiple Plants. The Byron Station containments utilize the same pre-stressing system, are essentially identical in design, were constructed within two years, and are similarly exposed to and protected from the outside environment.

Exelon

- Byron Station 1-7 Revision 0

IS! Program Plan Units I & 2, Fourth Intetial TABLE 1.1-4 UNIT 1 AND 2 THIRD CISI INTERVAL/PERIODIOUTAGE MATRIX (FOR PSI CLASS CC-TENDON COMPONENT EXAMINATIONS)

Note 3: With the exception of some dome tendon anchorages, which are considered not accessible due to safety hazards, all Byron Station ISI Class CC post-tensioning systems should be accessible for examination during operational periods.

In the event one or more of the inaccessible dome tendons anchorages are selected under Paragraph IWL-2521, the requirements of Paragraph IWL-2521.1, Exemptions, shall be applied. Completion of lSl Class CC post-tensioning system examinations for Byron Station should not be outage dependent.

Note 4: The requirements of 10 CFR 50.55a(b)(2)(viii) Paragraph (E) shall be applied to examinations and tests performed in accordance with ASME Section Xl, Subsection IWL.

Exelon

- Byron Station 1-8 Revision 0

IS! Program Plan Units 1 & 2, Fourth Interval

1.2 BACKGROUND

The Commonwealth Edison Company, now known commercially as Exelon Generation Company (EGC), LLC

, obtained Construction Permits to build Byron Station Units I and 2 on December 31, 1975, for Unit 1, CPPR-1 30, and for Unit 2, CPPR-1 31. The Docket Numbers assigned to Byron Station are 50-454 for Unit 1 and 50-455 for Unit 2. After satisfactory plant construction and pre-operational testing was completed, Byron Station was granted a full-power operating license for Unit 1, NPF-37, and subsequently commenced commercial operation on September 16, 1985; the full-power operating license for Unit 2, NPF-66, was granted and commercial operation commenced on August 21, 1987.

Byron Stations piping systems and associated components were designed and fabricated to be inspected and tested in accordance with the requirements of ASME Section Xl. Although this plant was specifically designed to meet the inspection and testing requirements of ASME Section XI, literal compliance may not be feasible or practical within the limits of the current plant design. Certain limitations are likely to occur due to conditions such as accessibility, geometric configuration, and/or metallurgical characteristics. For some inspection categories, an alternate component may be selected for examination and the code statistical and distribution requirements can still be maintained.

If ASME Section XI required examination criteria cannot be met, a relief request will be submitted in accordance with Code Of Federal Regulations, Title 10, Part 50, Section 55a, Codes and standards, (10 CFR 50.55a).

1.3 FIRST INTERVAL ISI PROGRAM Pursuant to 10 CFR 50.55a, licensees were required to meet the requirements of Paragraph (g), Inservice inspection requirements, of that section.

Specifically, Paragraph 10 CFR 50.55a(g)(4)(i) called for the inservice inspection requirements of the 120-month inspection interval to comply with the requirements of the latest Edition and addenda of ASME Section XI referenced in Paragraph (b) of 10 CFR 50.55a on the date twelve months prior the date of issuance of the operating license, subject to the limitations and modifications listed in 10 CFR 50.55a(b).

The version of 10 CFR 50.55a in effect twelve months prior to the issuance of the Byron Station Unit 1 operating license referenced ASME Section XI, 1980 Edition including Addenda through the Winter 1981 (8OAN81) in Paragraph (b)(2). Similarly, the version of 10 CFR 50.55a in effect twelve months prior to the issuance of the Byron Station Unit 2 operating license referenced ASME Section XI, 1983 Edition including Addenda through the Summer 1983 Addenda (83/583) in Paragraph (b)(2). The extent of the application of ASME Section Xl 81NV81 and 831S83 is limited by Paragraph (2)(iv)(A) such that ASME Section XI, 1974 Edition including Addenda through the Summer 1975 Addenda (74/S75) must be utilized for ISI Class 2 pressure retaining welds in Residual Heat Removal Systems, Emergency Core Cooling Systems, and Containment Heat Removal Systems. Optionally, per Paragraph (2)(iv)(B), plants with Construction Permits docketed prior to July 1, 1978, such as Byron Station, may use ASME Section XI 74/575 to examine ISI Class 2 pressure retaining welds in systems other than those in Paragraph (2)(iv)(A).

Based on these 10 CFR 50.55a mandatory and optional requirements, the Byron Station 151 Program Plan for the First 151 Interval was developed by Ebasco Services Incorporated. As allowed by ASME Section XI, IWA-2400(c) the First 151 Interval at Byron Station Unit 1 was extended from September 15, 1995 to June 30, 1996 to include Refueling Outage BIRO7.

Accordingly, the Second 151 Interval at Byron Station Unit 1 commenced July 1, 1996. The First ISI Interval at Byron Station Unit 2 was also extended from August 21, 1997 to August 15, 1998 to include Refueling Outage B2R07. Accordingly, the Second 151 Interval at Byron Station Unit 2 started on August 16, 1998.

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IS! Program Plan Units I & 2, Fourth Interval The Byron Station First Interval ISI Program started on September 16, 1985 and ended on June 29, 1996 for Unit 1, and started on August 22, 1987 and ended on August 15, 1998 for Unit 2.

Augmented 151 of Byron Station Unit 1 Reactor Vessel shell welds as mandated by 10 CFR 50.55a(g)(6)(ii)(A), was completed during the last period of First ISI Interval.

Volumetric examination of greater than 90% of the weld volume was completed, except as detailed in Relief Request NR-20 of the First Interval 151 Program Plan.

Augmented 151 of Byron Station Unit 2 Reactor Vessel shell welds as mandated by 10 CFR 50.55afg)(6)(ii)(A), was completed during the last period of First ISI Interval.

Volumetric examination of greater than 90% of the weld volume was completed, except as detailed in Relief Request NR-27 of the First Interval ISI Program Plan.

1.4 SECOND INTERVAL 151 PROGRAM Pursuant to 10 CFR 50.55a(g), licensees were required to update their ISI Programs at the end of the First 151 Interval. The ISI Program was required to comply with the latest Edition and Addenda of ASME Section Xl incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the Second ISI Interval per 10 CFR 50.55a(g)(4)(ii).

The Byron Station Second Interval ISI Program Plan was initially developed in accordance with the requirements of 10 CFR 50.55a including all published changes through June 30, 1995 and September 15, 1997 for Units I and 2 respectively, and the 1989 Edition, No Addenda of ASME Section XI. This Second Interval ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, IWF, and Mandatory Appendices of ASME Section Xl, approved ASME Code Cases, approved alternatives through relief requests and Safety Evaluation Reports (SERs), and utilized Inspection Program B.

As an alternative to the full ten-year interval duration requirements of Paragraphs IWA-2430(b) and (d) and Paragraph IWA-2432 for the Unit 2 Second 151 Interval and for the Units 1 and 2 First CISI Intervals, Byron Station proposed Relief Request 13R-01 to modify the interval dates of the Unit 2 Second ISI Interval and of the Units 1 and 2 First CISI Intervals.

This permitted the subsequent 151 and CISI Programs to share a common inspection interval start and end date and implemented common Code Editions for 151 Class 1, 2, 3, MC, and CC components. As such, the Second ISI Interval was effective from June 30, 1996 through January 15, 2006 for Byron Station Unit 1 and effective from August 16, 1998 through January 15, 2006 for Byron Station Unit 2.

1.5 THIRD INTERVAL 151 PROGRAM Pursuant to 10 CFR 50.55a(g), licensees were required to update their IS! Programs at the end of the Second 151 Interval. The 151 Program was required to comply with the latest Edition and Addenda of ASME Section Xl incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the Third ISI Interval per 10 CFR 50.55a(g)(4)(ii). As discussed in Section 1.4 above, the start of the Third IS! Interval was on January 16, 2006 for Byron Station Units 1 and 2. Based on this date, the latest Edition and Addenda of ASME Section XI referenced in 10 CFR 50.55a twelve months prior to the start of the Third 151 Interval per was the 2001 Edition through the 2003 Addenda.

The Byron Station Third Interval 151 Program Plan was developed in accordance with the requirements of 10 CFR 50.55a including all published changes through November 1, 2004 for Units 1 and 2 respectively, and the 2001 Edition through the 2003 Addenda of ASME Section Xl, subject to the limitations and modifications contained within Paragraph (b) of the regulation. This Third Interval ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, Exelon

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IS! Program Plan Units I & 2, FoLIflh Interval IWF, Mandatory Appendices of ASME Section Xl, approved ASME Code Cases, approved alternatives through relief requests and SERs, and utilized Inspection Program B.

Byron Station adopted the EPRI Topical Report TR-1 12657, Rev. B-A methodology, which was supplemented by ASME Code Case N-578-i (N-578-1), for implementing risk-informed inservice inspections during the Third ISI Interval. The RI-ISI Program continued for the Third 151 Interval. This approach replaced the categorization, selection, and examination volume requirements of ASME Section Xl Examination Categories B-F, B-J, C-F-i, and C-F-2 applicable to Byron Station with the associated requirements of TR-i12657, Rev. B-A for the associated requirements of TR-1 12657, Rev. B-A for Examination Category R-A as defined in N-578-1. Implementation of the RISI Program was in accordance with Relief Request 13R-02.

Byron Station also adopted the EPRI Topical Report TR-1 006937, Rev. 0-A, methodology for additional guidance for adaptation of the RI-ISI evaluation process to Break Exclusion Region (BER) piping, also referred to as the High Energy Line Break (HELB) region. This change to the BER program was made under 10 CFR 50.59 evaluation criteria. The BER Program continued for the Third (SI Interval.

The Byron Station Third (SI Interval was originally effective from January 16, 2006 through January 15, 2016 for Units I and 2, respectively. The Byron Station Units 1 and 2 First Period was extended by nine months as permitted by Paragraph IWA-2430(d)(3) in order to coincide with the plant refueling outage schedule. The Byron Station Units 1 and 2 Third 151 Interval was also extended by six months as permitted by Paragraph IWA-2430(d)(i) in order to coincide with the plant refueling outage schedule.

Therefore, the Byron Station Third ISI Interval was effective from January 16, 2006 through July 15, 2016 for Units 1 and 2, respectively.

1.6 FOURTH INTERVAL ISI PROGRAM Pursuant to 10 CFR 50.55a(g), licensees are required to update their ISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI Program is required to comply with the latest Edition and Addenda of the Code incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the Fourth 151 Interval per 10 CFR 50.55a(g)(4)(ii). As discussed in Section 1.4 above, the start of the Fourth lSl Interval will be on July 16, 2016, for Byron Station Units 1 and 2. Based on this date, the latest Edition and Addenda of the Code referenced in 10 CFR 50.55a(b)(2) twelve months prior to the start of the Fourth ISI Interval was the 2007 Edition with the 2008 Addenda.

The Byron Station Fourth Interval 151 Program Plan was developed in accordance with the requirements of 10 CFR 50.55a, and the 2007 Edition with the 2008 Addenda of ASME Section Xl, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.10-i of this section.

This ISI Program Plan addresses Subsections IWA, IWB, IWO, IWD, IWF, Mandatory Appendices of ASME Section XI, approved Code Cases, approved alternatives through relief requests and SEs, and utilizes the Inspection Program as defined therein.

Byron Station adopted the EPRI Topical Report TR-1 12657, Rev. B-A methodology, which was supplemented by N-578-i, for implementing risk-informed inservice inspections during the Third 151 Interval. The RISI Program will continue for the Fourth ISI Interval.

Implementation of the RISI Program is in accordance with Relief Request 14R-0i.

Byron Station also adopted the EPRI Topical Report TR-1006937, Rev. 0-A, methodology for additional guidance for adaptation of the RI-ISI evaluation process to BER piping, also referred to as the HELB region. The BER Program will continue for the Fourth ISI Interval.

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IS! Program Plan Urits 7 & 2, Fourth Interval The Byron Station Fourth 151 Interval is effective from July 16, 2016, through July 15, 2025, for Units 1 and 2, respectively.

[Note that the start and end dates for the Third 151 Interval and Second CISI Interval were aligned, as well as subsequent intervals per the wording in previous Third ISI Interval and Second CISI Interval Relief Request I3R-01 that was authorized by the NRC per SER dated September 7, 2006. Therefore, a Fourth ISI Interval and Third CISI Interval relief request is not requited. Previous Relief Request I3R-01 stated Relief is requested to modify the end dates of the Byron Station Unit 2 Second SI Interval and of the Byron Station Units 1 and 2 First 0151 Intervals and the start and end dates of all subsequent ISI and CISI Intervals for Byron Station Units I and 2. 13R-01 also stated that All inspection periods for Class 1, 2, 3, and MC components will commence for the next interval based on the modified common interval start date. Any examination methods unique to and specifically required in the third period under the previous interval, that will likewise be required in the next interval, will be scheduled and completed in the first period of the subsequent interval. The examinations will be conducted and credited under the rules of the new Code of Record (i.e., 2001 Edition through the 2003 Addenda of ASME Section XI). These examinations originally unique to the third period of the previous interval will henceforth be conducted in the first period of all subsequent ISI intervals, and deferral to the end of future intervals will not be available.

In addition, the roIling five-year IWL frequency applicable to Class CC components that are subject to Subsection IWL requirements will be maintained as currently scheduled.J Thus, the Byron Station Unit 2 end of interval examinations will be conducted at the end of the first period of the Fourth ISI Interval using the 2007 Edition with the 2008 Addenda of ASME Section Xl.

1.7 FIRST INTERVAL 0151 PROGRAM CISI examinations were originally invoked by amended regulations contained within a Final Rule issued by the Nuclear Regulatory Commission (NRC). The amended regulation incorporated the requirements of the 1992 Edition with the 1992 Addenda of the ASME Section Xl, Subsections IWE and IWL, subject to specific modifications that were included in Paragraphs 10 CFR 50.55a(b)(2)(ix) and 10 CFR 50.55a(b)(2)(x). Relief from the examination requirements of Subsections IWE and IWL of the 1992 Edition with the 1992 Addenda of ASME Section XI was granted by the NRC to allow Byron Station to use the 1998 Edition, No Addenda of Subsections IWE and IWL of ASME Section Xl for inspection of containment components.

The final rulemaking was published in the Federal Register on August 8, 1996 and specified an effective date of September 9, 1996. Implementation of the Subsection IWE and IWL Program from a scheduling standpoint was driven by the five year expedited implementation period per 10 CFR 50.55a(g)(6)(ii)(B), which specified that the examinations required to be completed by the end of the First Period of the First 0151 Interval (per Table IWE-241 2-1) be completed by the effective date (by September 9, 2001).

ASME Section Xl Subsections IWE, IWL, Mandatory Appendices, approved ASME Code Cases, and approved alternatives through relief requests and SERs were added to the ISI Program midway through the Second 0151 Interval to address CISI. The 0151 Program Plan was developed and implemented prior to the required date, and examinations for the first and second periods were performed in accordance with the First CISI Interval schedule.

As an alternative to the full ten-year interval duration requirements of Paragraphs IWA-2430(b) and (d) and Paragraph IWA-2432 for the Unit 2 Second 151 Interval and for the Units I and 2 First CISI Intervals, Byron Station proposed Relief Request l3R-01 to modify the interval dates of the Unit 2 Second SI Interval and of the Units 1 and 2 First CISI Intervals.

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IS! Program Plan Units 1 & 2, Fourth Interval start and end date and implemented common Code Editions for IS! Class 1, 2, 3, MC, and CC components. As such, the First CISI Interval occurred approximately three years early and was effective from September 9, 1996 through January 15, 2006 for Byron Station Units I and 2.

1.8 SECOND INTERVAL CISI PROGRAM Pursuant to 10 CFR 50.55a(g), licensees were required to update their 0151 Programs at the end of the First CISI Interval. The CISI Program was required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the Second CISI Interval per 10 CFR 50.55a(g)(4)(ii). As discussed in Section 1.7 above, the start of the Second 0151 Interval was on January 16, 2006 for Byron Station Units 1 and 2. Based on this date, the latest Edition and Addenda of the referenced Code twelve months prior to the start of the Second 0151 Interval was the 2001 Edition through the 2003 Addenda.

The Byron Station Second Interval CISI Program Plan was developed in accordance with the requirements of 10 CFR 50.55a including all published changes through November 1,2004, and the 2001 Edition through the 2003 Addenda of ASME Section Xl, subject to the limitations and modifications contained within Paragraph (b) of the regulation. This Second Interval CISI Program Plan addressed Subsections IWE, IWL, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SERs, and utilized Inspection Program B.

The Byron Station Second CISI Interval was originally effective from January 16, 2006 through January 15, 2016 for Units 1 and 2, respectively. The Byron Station Units 1 and 2 Second Period was extended by nine months as permitted by Paragraph IWA-2430(d)(3) in order to coincide with the plant refueling outage schedule. The Byron Station Units 1 and 2 Second CISI Interval was also extended by six months as permitted by Paragraph IWA-2430(d)(1) in order to coincide with the plant refueling outage schedule.

Therefore, the Byron Station Second 0151 Interval was effective from January 16, 2006 through July 15, 2016 for Units 1 and 2, respectively. (Note that the Byron Station Second 0151 Interval was extended three extra months to coincide with the IWL-concrete examination schedule effective from January 16, 2006 through October 15, 2016 for Units 1 and 2, respectively. However, the extra three months will not roll-over to the next Third 0151 Interval for IWL-concrete.)

1.9 THIRD INTERVAL CISI PROGRAM Pursuant to 10 CFR 50.55a(g), licensees were required to update their CISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The CISI Program is required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10 CFR 50.55a twelve months prior to the start of the Third 0151 Interval per 10 CFR 50.55a(g)(4)(ii). As discussed in Section 1.8 above, the start of the Third CISI Interval will be on July 16, 2016 for Byron Station Units 1 and 2. Based on this date, the latest Edition and Addenda of the referenced Code twelve months prior to the start of the Third 0151 Interval was the 2007 Edition with the 2008 Addenda.

The Byron Station Third Interval 0151 Program Plan was developed in accordance with the requirements of 10 CFR 50.55a including all published changes through December11, 2014, and the 2007 Edition with the 2008 Addenda of ASME Section XI, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.10-1 of this section. This Third Interval CISI Program Plan addresses Subsections IWE, IWL, Mandatory Appendices, approved ASME Code Cases, Exelon

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IS!.rogram Plan Units I & 2, Fourth Interval approved alternatives through relief requests and SEs, and utilizes Inspection Program as defined therein.

The Byron Station Third 0151 Interval is effective from July 16, 2016 through July 15, 2025 for Units I and 2, respectively. (Note that the Byron Station Units I and 2 Second CISI Interval end dates for IWL-concrete were modified by the extensions shown in Section 1.8; however, the Third CISI Interval start date remains unchanged.)

[Note that the start and end dates for the Third ISI Interval and Second CISI Interval were aligned, as well as subsequent intervals per the wording in previous Third 151 Interval and Second CISI Interval Relief Request 13R-01 that was authorized by the NRC per SER dated September 7, 2006. Therefore, a Fourth ISI Interval and Third CISI Interval relief request is not required. Previous Relief Request I3R-01 stated Relief is requested to modify the end dates of the Byron Station Unit 2 Second ISI Interval and of the Byron Station Units 1 and 2 First CISI Intervals and the start and end dates of all subsequent ISI and CISI Intervals for Byron Station Units I and 2. I3R-01 also stated that All inspection periods for Class 1, 2, 3, and MC components will commence for the next interval based on the modified common interval start date. Any examination methods unique to and specifically requited in the third period under the previous interval, that will likewise be requited in the next interval, will be scheduled and completed in the first period of the subsequent interval. The examinations will be conducted and credited under the rules of the new code of record (i.e., 2001 Edition through the 2003 Addenda of ASME Section XI). These examinations originally unique to the third period of the previous interval will henceforth be conducted in the first period of all subsequent ISI intervals, and deferral to the end of future intervals will not be available.

In addition, the roIling five-year IWL frequency applicable to Class CC components that are subject to Subsection IWL requirements will be maintained as currently scheduled.] Thus, the Byron Station Unit 2 end of interval 0151 examinations will be conducted at the end of the first period of the Fourth 151 Interval using the 2007 Edition with the 2008 Addenda of ASME Section XI.

NOTE: This document is credited for meeting commitments associated with the Byron Station License Renewal Project for Aging Management Programs (AMP).

CM-f: AMP XI.S1

- ASME Section XI, Subsection IWE.

CM-2: AMP XI.S2

- ASME Section Xl, Subsection IWL.

CM-3: AMP XI.S3

- ASME Section XI, Subsection IWF.

1.10 CODE OF FEDERAL REGULATIONS 10 CFR 50.55A REQUIREMENTS There are certain Paragraphs in 10 CFR 50.55a that list the limitations, modifications, and/or clarifications to the implementation requirements of ASME Section XI. These Paragraphs in 10 CFR 50.55a, including all published changes through December 11,2014(79 FR 73462),

that are applicable to Byron Station are detailed in Table 1.10-1.

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IS! Program Plan Units 7 & 2, Fourth Inteival TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS a) Documents approved for incorporation by reference. The standards listed in this paragraph have been approved for incorporation by reference by the Director of the Federal Register pursuant to 5 u.s.c. 552(a) and 1 CER part 51. The standards are available for inspection at the NRC Technical Library, 11545 Rockville Pike, Rockville, Maryland 20852; telephone: 301-415-6239; or at the National Archives and Records Administration fNARA). For information on the availability of this material at NARA, call 202-741-6030 or go to httpilwww.archives.gov/federal-register/cfr/ibr Iocations.html.

(1) American Society of Mechanical Engineers (ASME),

fi) ASME Boiler and Pressure Vessel Code,Section III. The editions and addenda for Section III of the ASME Boiler and Pressure Vessel Code are listed below, but limited to those provisions identified in paragraph (b)(1) of this section.

(A) Rules for Construction of Nuclear Vessels:

(7) 1963 Edition, (2) Summer 1964 Addenda,

] (3) Winter 1964 Addenda, (4) 1965 Edition, (5) 1965 Summer Addenda,

] (6) 1965 Winter Addenda, (7) 1966 Summer Addenda, (8) 1966 Winter Addenda, j (9) 1967 Summer Addenda, (10) 1967 Winter Addenda, (77) 1968 Edition, f (12) 1968 Summer Addenda, (13) 1968 Winter Addenda, (14) 1969 Summer Addenda, j(75) 1 969 Winter Addenda, (16) 1970 SummerAddenda, and (17) 1970 WinterAddenda.

f (B) Rules for Construction of Nuclear Power Plant Components:

(7) 1971 Edition, (2) 1971 Summer Addenda, (3) 1971 Winter Addenda, (4) 1972 Summer Addenda, (5) 1972 Winter Addenda, (6) 1973 Summer Addenda, and (7) 1973 Winter Addenda.

(C) Division 1 Rules for Construction of Nuclear Power Plant Components:

(7) 1974 Edition, (2) 1974 Summer Addenda, (3) 1974 Winter Addenda, (4) 1975 Summer Addenda, (5) 1975 Winter Addenda, J- (6) 1976 Summer Addenda, and (7) 1976 Winter Addenda; (D) Rules for Construction of Nuclear Power Plant ComponentsDivision 1; (1) 1977 Edition, (2) 1977 Summer Addenda, (3) 1977 Winter Addenda, (4) 1978 SummerAddenda, (5) 1978 WinterAddenda, (6) 1979 SummerAddenda, (7) 1979 Winter Addenda, (8) 1980 Edition, (9) 1980 Summer Addenda, (70) 1980 Winter Addenda, (71) 1981 Summer Addenda, (72) 1981 Winter Addenda, (73) 1982 Summer Addenda, (14) 1982 Winter Addenda, (75) 1983 Edition, (76) 1983 Summer Addenda, (17) 1983 Winter Addenda, (78) 1984 Summer Addenda, (19) 1984 Winter Addenda, (20) 1985 Summer Addenda, (21) 1985 Winter Addenda, (22) 1986 Edition, (23) 1986 Addenda, (24) 1987 Addenda, (25) 1988 Addenda, (26) 1989 Edition, (27) 1989 Addenda, (28) 1990 Addenda, (29) 1991 Addenda, (30) 1992 Edition, (31) 1992 Addenda, (32) 1993 Addenda, (33) 1994 Addenda, (34) 1995 Edition, (35) 1995 Addenda, (36) 1996 Addenda, and (37) 1997 Addenda.

(E) Rules for Construction of Nuclear Facility ComponentsDivision 1 (7) 1998 Edition, (2) 1998 Addenda, (3) 1999 Addenda, (4) 2000 Addenda, (5) 2001 Edition, (6) 2001 Addenda, (7) 2002 Addenda, (8) 2003 Addenda, (9) 2004 Edition, (70) 2005 Addenda, f II) 2006 Addenda, (12) 2007 Edition, and (13) 2008 Addenda.

(ii) ASME Boiler and Pressure Vessel Code, Section Xl. The editions and addenda for Section Xl of the ASME Boiler and Pressure Vessel Code are listed below, but limited to those provisions identified in paragraph (b)(2) of this section.

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ISI Program Plan Units 1 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (A) Rules for Inservice Inspection of Nuclear Reactor Coolant Systems:

(7) 1970 Edition, J (2) 1971 Edition, j_ç3) 1971 SummerAddenda, (4) 1971 Winter Addenda, (5) 1972 Summer Addenda, (6) 1972 Winter Addenda, (7) 1973 Summer Addenda, and (8) 1973 Winter Addenda.

(B) Rules for Inservice Inspection of Nuclear Power Plant Components:

(7) 1974 Edition, (2) 1974 Summer Addenda, (3) 1974 Winter Addenda, and (4) 1975 Summer Addenda.

(5) 1975 Winter Addenda, (6) 1976 Summer Addenda, and (7) 1976 Winter Addenda, (C) Rules for Inservice Inspection of Nuclear Power Plant ComponentsDivision 1 (7) 1977 Edition, (2) 1977 Summer Addenda, (3) 1977 Winter Addenda, (4) 1978 Summer Addenda, (5) 1978 Winter Addenda, (6) 1979 Summer Addenda, (7) 1979 WinterAddenda, (8) 1980 Edition, (9) 1980 Winter Addenda, (70) 1981 Summer Addenda, (17) 1981 Winter Addenda, (12) 1982 Summer Addenda, (73) 1982 WinterAddenda, (74) 1983 Edition, (75) 1983 Summer Addenda, (16) 1983 Winter Addenda, (17) 1984 Summer Addenda,

( 78) 1 984 Winter Addenda, (79) 1985 Summer Addenda, (20) 1985 Winter Addenda, (27) 1986 Edition, (22) 1986 Addenda, (23) 1987 Addenda, (24) 1988 Addenda, (25) 1989 Edition, (26) 1989 Addenda, (27) 1990 Addenda, (28) 1991 Addenda, (29) 1992 Edition, (30) 1992 Addenda, (31) 1993 Addenda, (32) 1994 Addenda, (33) 1995 Edition, (34) 1995 Addenda, (35) 1996 Addenda, (36) 1997 Addenda, (37) 1998 Edition, (38) 1998 Addenda, (39) 1999 Addenda, (40) 2000 Addenda, (41) 2001 Edition, (42) 2001 Addenda, (43) 2002 Addenda, (44) 2003 Addenda, (45) 2004 Edition, (46) 2005 Addenda, (47) 2006 Addenda, (48) 2007 Edition, and (49) 2008 Addenda.

(Ui) ASME Code Cases. Nuclear Components-(A) ASME Code Case N-722-1. ASME Code Case N-722-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/1 82 Materials, Section Xl, Division 1 (Approval Date: January 26, 2009), with the conditions in paragraph (g)(6)(ii)(E) of this section.

(B) ASME Code Case N-729-1. ASME Code Case N-729-1, Alternative Examination Requirements for PWR Reactor Vessel Upper Heads With Nozzles Having Pressure-Retaining Partial-Penetration Welds, Section Xl, Division 1 (Approval Date: March 28, 2006), with the conditions in paragraph (g)(6)(ii)(D) otthis section.

(C) ASME Code Case N-770-7. ASME Code Case N-770-1, Additional Examinations for PWR Pressure Retaining Welds in Class 1 Components Fabricated with Alloy 600/82/1 82 Materials, Section Xl, Division 1 (Approval Date:

December 25, 2009), with the conditions in paragraph (g)(6)(ii)(F) of this section, (iv) ASME Operation and Maintenance Code. The editions and addenda for the ASME Code for Operation and Maintenance of Nuclear Power Plants are listed below, but limited to those provisions identified in paragraph (b)(3) of this section.

(A) Code for Operation and Maintenance ot Nuclear Power Plants:

(7) 1995 Edition, (2) 1996 Addenda, (3) 1997 Addenda, (4) 1998 Edition, (5) 1999 Addenda, (6) 2000 Addenda, (7) 2001 Edition, (8) 2002 Addenda, (9) 2003 Addenda, (10) 2004 Edition, (77) 2005 Addenda, and (72) 2006 Addenda.

(B) [Reserved]

(2) (I) (ii) (iU)

(3)

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IS! Prociram Plan Units I & 2. Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (i)NRCRegulatoiyGuide 1.84, Revision 36. NRC Regulatory Guide 1.84, Revision 36, Design, Fabrication, and Materials Code Case Acceptability. ASME Section III, dated August 2014, with the requirements in paragraph (b)f4) of this section.

(ii) NRC Regulatory Guide 1.747, Revision 17. NRC Regulatory Guide 1.147, Revision 17, Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1, dated August 2014, which lists ASME Code Cases that the NRC has approved in accordance with the requirements in paragraph (b)(5) of this section.

(iii) NRC Regulatory Guide 1.192, Revision 1. NRC Regulatory Guide 1.192, Revision 1, Operation and Maintenance Code Case Acceptability, ASME OM Code, dated August 2014, which lists ASME Code Cases that the NRC has approved in accordance with the requirements in paragraph (b)f6) of this section.

(b) Use and conditions on the use of standards. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code (BPV Code) and the ASME Code for Operation and Maintenance of Nuclear Power Plants (OM Code) as specified in this paragraph. Each combined license for a utilization facility is subject to the following conditions.

(1) Conditions on ASME BPV Code Section Ill. Each manufacturing license, standard design approval, and design certification under part 52 of this chapter is subject to the following conditions. As used in this section, references to Section Ill refer to Section III of the ASME Boiler and Pressure Vessel Code and include the 1963 Edition through 1973 Winter Addenda and the 1974 Edition (Division 1) through the 2008 Addenda (Division 1), subject to the following conditions:

(i)Section III condition: Section III materials. When applying the 1992 Edition of Section III, applicants or licensees must apply the 1992 Edition with the 1992 Addenda of Section II of the ASME Boiler and Pressure Vessel Code.

(ii)Section III condition: Weld leg dimensions, When applying the 1989 Addenda through the latest edition and addenda, applicants or licensees may not apply subparagraphs NB-3683.4(c)f1) and NB-3683.4(c)(2) or Footnote 11 from the 1989 Addenda through the 2003 Addenda, or Footnote 13 from the 2004 Edition through the 2008 Addenda to Figures NC-3673.2(b)-i and ND-3673.2(b)-1 for welds with leg size less than 1.09 t,.

(iU) Section IP condition: Seismic design ofpiping. Applicants or licensees may use Subarticles NB-3200, NB-3600, NC-3600, and ND-3600 for seismic design of piping, up to and including the 1993 Addenda, subject to the condition specified in paragraph (b)(1 )(ii) of this section. Applicants or licensees may not use these subarticles for seismic design of piping in the 1994 Addenda through the 2005 Addenda incorporated by reference in paragraph (a)(1) of this section, except that Subarticle NB-3200 in the 2004 Edition through the 2008 Addenda may be used by applicants and licensees, subject to the condition in paragraph (b)(i )(iii)(A) of this section. Applicants or licensees may use Subarticles NB-3600, NC-3600, and ND-3600 for the seismic design of piping in the 2006 Addenda through the 2008 Addenda, subject to the conditions of this paragraph corresponding to those subarticles.

(A) Seismic design of piping: First provision. When applying Note (1) of Figure NB-3222-1 for Level B service limits, the calculation of Pb stresses must include reversing dynamic loads (including inertia earthquake effects) if evaluation of these loads is required by NB-3223(b).

(B) Seismic design of piping: Second provision. For Class 1 piping, the material and D0/t requirements of NB-3656(b) must be met for all Service Limits when the Service Limits include reversing dynamic loads, and the alternative rules for reversing dynamic loads are used.

(iv) Section Ill condition: Quality assurance. When applying editions and addenda later than the 1989 Edition of Section III, the requirements of NQA-1, Quality Assurance Requirements for Nuclear Facilities, 1986 Edition through the 1994 Edition, are acceptable for use, provided that the edition and addenda of NQA-1 specified in NCA-4000 is used in conjunction with the administrative, quality, and technical provisions contained in the edition and addenda of Section Ill being used.

(v) Section Ill condition: Independence of inspection. Applicants or licensees may not apply NCA4134.10(a) of Section Ill, 1995 Edition through the latestedition and addenda incorporated by reference in paragraph (a)(1) of this section.

(vi)Section III condition: Subsection NH. The provisions in Subsection NH, Class 1 Components in Elevated Temperature Service, 1995 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1) of this section, may only be used for the design and construction of Type 316 stainless steel pressurizer heater sleeves where service conditions do not cause the components to reach temperatures exceeding 900 F.

(vii)Section III condition: Capacity certification and demonstration of function of incompressible-fluid pressure-relief valves.

When applying the 2006 Addenda through the 2007 Edition up to and including the 2008 Addenda, applicants and licensees may use paragraph NB-7742, except that paragraph NB-7742(a)(2) may not be used. For a valve design of a single size to be certified over a range of set pressures, the demonstration of function tests under paragraph NB-7742 must be conducted as prescribed in NB-7732.2 on two valves covering the minimum set pressure for the design and the maximum set pressure that can be accommodated at the demonstration facility selected for the test.

(2) Conditions on ASME BPV Code Section Xl. As used in this section, references to Section XI refer to Section Xl, Division 1, of the ASME Boiler and Pressure Vessel Code, and include the 1970 Edition through the 1976 Winter Addenda and the 1977 Edition through the 2007 Edition with the 2008 Addenda, subject to the following conditions:

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IS! Program Plan Units 1 & 2, Fourth interval TABLE 1.10-f CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS fi) [Reserved]

(ii) Section Xl condition: Pressure-retaining welds in ASME Code Class I piping (applies to Table IWB-2500 and IWB-2500-I and Category B-]). If the facilitys application for a construction permit was docketed prior to July 1, 1978, the extent of examination for Code Class 1 pipe welds may be determined by the requirements of Table IWB-2500 and Table IWB-2600 Category B-J of Section XI of the ASME BPV Code in the 1974 Edition and Addenda through the Summer 1975 Addenda or other requirements the NRC may adopt.

(Hi) [Reservedj iv) [Reserved)

(v) [Reserved]

(vi) Section Xl condition: Effective edition and addenda of Subsection IWE and Subsection IWL. Applicants or licensees may use either the 1992 Edition with the 1992 Addenda or the 1995 Edition with the 1996 Addenda of Subsection IWE and Subsection IWL, as conditioned by the requirements in paragraphs (b)(2)(viii) and (ix) of this section, when implementing the initial 120-month inspection interval for the containment inservice inspection requirements of this section. Successive 120-month interval updates must be implemented in accordance with paragraph (g)f4)(ii) of this section.

(vii) Section Xl condition: Section Xl references to CM Part 4, CM Part 6, and CM Part 10 (Table IWA-f 600-7). When using Table IWA-1600-1, Referenced Standards and Specifications, in the Section Xl, Division 1, 1987 Addenda, 1988 Addenda, or 1989 Edition, the specified Revision Date or Indicator for AS ME/ANSI CM part 4, ASME/ANSI part 6, and ASME/ANSI part 10 must be the OMa-1988 Addenda to the CM-i 987 Edition. These requirements have been incorporated into the CM Code, which is incorporated by reference in paragraph (a)(1)(iv) of this section.

(viii) Section Xl condition: Concrete containment examinations.

... Applicants or licensees applying Subsection IWL, 2007 Edition through the latest edition and addenda incorporated by reference in paragraph fa)(1 )(ii) of this section, must apply paragraph (b)(2)(viii)(E) of this section. (A) (B) (C) (D) (1) (2) (3)

(E) Concrete containment examinations: Fifth provision. For Class CC applications, the applicant or licensee must evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or the result in degradation to such inaccessible areas. For each inaccessible area identified, the applicant or licensee must provide the following in the ISI Summary Report required by IWA-6000:

(I) 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.

(F)(G) fix) Section Xl condition: Metal containment examinations.

... Applicants or licensees applying Subsection IWE, 2007 Edition through the latest addenda incorporated by reference in paragraph fa)fl)(ii) of this section, must satisfy the requirements of paragraphs (b)(2)Ox)(A)(2) and (b)(2)(ix)(B) and f]) of this section. (A) Metal containment examinations:

First provision. For Class MC applications, the following apply to inaccessible areas.

(1)

(2) For each inaccessible area identified for evaluation, the applicant or licensee must provide the following in the IS)

Summary Report as required by IWA-6000:

(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 result of the evaluation; and (Hi) A description of necessary corrective actions.

(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-i may be extended and the minimum illumination requirements specified in Table IWA-2210-i 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.

(C) (D) (1) fi) (ii) (iii) (2) (E) (F) (G) (H) (I)

(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 leak-tight 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. (x) Section Xl condition:

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IS! Program Plan Units I & 2, Fotirt5 Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS Quality assurance. When applying Section Xl editions and addenda later than the 1989 Edition, the requirements of NQA-1, Quality Assurance Requirements for Nuclear Facilities, 1979 Addenda through the 1989 Edition, are acceptable as permitted by IWA-1400 of Section Xl, if the licensee uses its 10 CFR part 50, Appendix B, quality assurance program, in conjunction with Section Xl requirements. Commitments contained in the licensees quality assurance program description that are more stringent than those contained in NQA-1 must govern Section XI activities. Further, where NQA-1 and Section XI do not address the commitments contained in the licensees Appendix B quality assurance program description, the commitments must be applied to Section Xl activities.

(xi) [Reserved]

(xii) Section Xl condition: Underwater welding. The provisions in IWA-4660, Underwater Welding, of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, are not approved for use on irradiated material.

(xiii) [Reserved]

(xiv) Section Xl condition: Appendix VIII personnel qualification. All personnel qualified for performing ultrasonic examinations in accordance with Appendix VIII must receive 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of annual hands-on training on specimens that contain cracks, Licensees applying the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1 )Øi) of this section may use the annual practice requirements in VIl-4240 of Appendix VII of Section Xl in place of the 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> of annual hands-on training provided that the supplemental practice is performed on material or welds that contain cracks, or by analyzing prerecorded data from material or welds that contain cracks. In either case, training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensees facility.

(xv)Section XI condition: Appendix VIII specimen set and qualification requirements. Licensees using Appendix VIII in the 1995 Edition through the 2001 Edition of the ASME Boiler and Pressure Vessel Code may elect to comply with all of the provisions in paragraphs (b)(2)fxv)(A) through (M) of this section, except for paragraph (b)(2)(xv)(F) of this section, which may be used at the licensees option. Licensees using editions and addenda after 2001 Edition through the 2006 Addenda must use the 2001 Edition of Appendix VIII and may elect to comply with all of the provisions in paragraphs fb)(2)(xv)(A) through (M) of this section, except for paragraph (b)(2)(xv)(F) of this section, which may be used at the licensees option.

(A) Specimen set and qualification: First provision. When applying Supplements 2, 3, and 10 to Appendix VIII, the following examination coverage criteria requirements must be used:

(1) Piping must be examined in two axial directions, and when examination in the circumferential direction is required, the circumferential examination must be performed in two directions, provided access is available. Dissimilar metal welds must be examined axially and circumferentially.

(2) Where examination from both sides is not possible, full coverage credit may be claimed from a single side for ferritic welds. Where examination from both sides is not possible on austenitic welds or dissimilar metal welds, full coverage credit from a single side may be claimed only after completing a successful single-sided Appendix VIII demonstration using flaws on the opposite side of the weld. Dissimilar metal weld qualifications must be demonstrated from the austenitic side of the weld, and the qualification may be expanded for austenitic welds with no austenitic sides using a separate add-on performance demonstration. Dissimilar metal welds may be examined from either side of the weld.

(B) Specimen set and qualification: Second provision. The following conditions must be used in addition to the requirements of Supplement 4 to Appendix VIII:

(7) Paragraph 3.1, Detection acceptance criteriaPersonnel are qualified for detection if the results of the performance demonstration satisfy the detection requirements of ASME Section Xl, Appendix VIII, Table VIlI-S4-1, and no flaw greater than 0.25 inch through-wall dimension is missed.

(2) Paragraph 1.1(c), Detection test matrixFlaws smaller than the 50 percent of allowable flaw size, as defined in IWS 3500, need not be included as detection flaws. For procedures applied from the inside surface, use the minimum thickness specified in the scope of the procedure to calculate alt. For procedures applied from the outside surface, the actual thickness of the test specimen is to be used to calculate a/t.

(C) Specimen set and qualification: Third provision. When applying Supplement 4 to Appendix VIII, the following conditions must be used:

(1) A depth sizing requirement of 0.15 inch RMS must be used in lieu of the requirements in Subparagraphs 3.2(a) and 3.2(c), and a length sizing requirement of 0.75 inch RMS must be used in lieu of the requirement in Subparagraph 3.2(b).

(2) In lieu of the location acceptance criteria requirements of Subparagraph 2.1(b), a flaw will be considered detected when reported within 1.0 inch or 10 percent of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions.

(3) In lieu of the flaw type requirements of Subparagraph 1.1 (e)(1), a minimum of 70 percent of the flaws in the detection and sizing tests must be cracks. Notches, if used, must be limited by the following:

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IS! Program Plan Units 1 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (I) Notches must be limited to the case where examinations are performed from the clad surface.

(ii) Notches must be semielliptical with a tip width of less than or equal to 0.010 inches.

(iii Notches must be perpendicular to the surface within +/-2 degrees.

(4) In lieu of the detection test matrix requirements in paragraphs 1.1(e)(2) and 1.1(e)(3), personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations.

(D) Specimen set and qualification: Fourth provision. The following conditions must be used in addition to the requirements of Supplement 6 to Appendix VIII:

(7) Paragraph 3.1, Detection Acceptance CriteriaPersonnel are qualified for detection if:

(i) No surface connected flaw greater than 0.25 inch through-wall has been missed.

(ii) No embedded flaw greater than 0.50 inch through-wall has been missed.

(2) Paragraph 3.1, Detection Acceptance CriteriaFor procedure qualification, all flaws within the scope of the procedure are detected.

(3) Paragraph 1.1(b) for detection and sizing test flaws and locationsFlaws smaller than the 50 percent of allowable flaw size, as defined in IWB-3500, need not be included as detection flaws. Flaws that are less than the allowable flaw size, as defined in IWB-3500, may be used as detection and sizing flaws.

(4) Notches are not permitted.

(E) Specimen set and qualification: Fifth provision. When applying Supplement 6 to Appendix VIII, the following conditions must be used:

(1) A depth sizing requirement of 0.25 inch RMS must be used in lieu of the requirements of subparagraphs 3.2(a),

3.2(c)(2), and 3.2(c)(3).

(2) In lieu of the location acceptance criteria requirements in Subparagraph 2.1(b), a flaw will be considered detected when reported within 1.0 inch or 10 percent of the metal path to the flaw, whichever is greater, of its true location in the X and Y directions, (3) In lieu of the length sizing criteria requirements of Subparagraph 3.2(b), a length sizing acceptance criteria of 0.75 inch RMS must be used.

(4) In lieu of the detection specimen requirements in Subparagraph 1.1 (e)(1), a minimum of 55 percent of the flaws must be cracks. The remaining flaws may be cracks or fabrication type flaws, such as slag and lack of fusion. The use of notches is not allowed.

(5) In lieu of paragraphs 1.1(e)(2) and 1.1(e)(3) detection test matrix, personnel demonstration test sets must contain a representative distribution of flaw orientations, sizes, and locations.

(F) Specimen set and qualification: Sixth provision. The following conditions may be used for personnel qualification for combined Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII qualification. Licensees choosing to apply this combined qualification must apply all ot the provisions of Supplements 4 and 6 including the following conditions:

(1) For detection and sizing, the total number of flaws must be at least 10. A minimum of 5 flaws must be from Supplement 4, and a minimum of 50 percent of the flaws must be from Supplement 6. At least 50 percent of the flaws in any sizing must be cracks. Notches are not acceptable for Supplement 6.

(2) Examination personnel are qualified for detection and length sizing when the results of any combined performance demonstration satisfy the acceptance criteria of Supplement 4 to Appendix VIII.

(3) Examination personnel are qualified for depth sizing when Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII flaws are sized within the respective acceptance criteria of those supplements.

(G) Specimen set and qualification: Seventh provision. When applying Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or combined Supplement 4 and Supplement 6 qualification, the following additional conditions must be used, and examination coverage must include:

(1) The clad-to-base-metal-interface, including a minimum of 15 percent T (measured from the clad-to-base-metal-interface), must be examined from four orthogonal directions using procedures and personnel qualified in accordance with Supplement 4 to Appendix VIII.

(2) If the clad-to-base-metal-interface procedure demonstrates detectability of flaws with a tilt angle relative to the weld centerline of at least 45 degrees, the remainder of the examination volume is considered fully examined if coverage is obtained in one parallel and one perpendicular direction. This must be accomplished using a procedure and personnel qualified for single-side examination in accordance with Supplement 6. Subsequent examinations of this volume may be performed using examination techniques qualified for a tilt angle of at least 10 degrees.

(3) The examination volume not addressed by paragraph (b)(2)(xv)(G)(1) of this section is considered fully examined if Exelon

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ISI Program Plan Units 1 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS coverage is obtained in one pacallel and one perpendicular direction, using a procedure and personnel qualified for single sided examination when the conditions in paragraph (b)(2)(xv)(G)(2) are met.

(H) Specimen set and qualification: Eighth provision. When applying Supplement 5 to Appendix VIII, at least 50 percent of the flaws in the demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks.

Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches.

(I) Specimen set and qualification: Ninth provision. When applying Supplement 5, Paragraph (a), to Appendix VIII, the number of false calls allowed must be D/1 0, with a maximum of 3, where D is the diameter of the nozzle.

(1) [Reserved]

(K) Specimen set and qualification: Eleventh provision. When performing nozzle-to-vessel weld examinations, the following conditions must be used when the requirements contained in Supplement 7 to Appendix VIII are applied for nozzle-to-vessel welds in conjunction with Supplement 4 to Appendix VIII, Supplement 6 to Appendix VIII, or combined Supplement 4 and Supplement 6 qualification.

(1) For examination of nozzle-to-vessel welds conducted from the bore, the following conditions are required to quality the procedures, equipment, and personnel:

(I) For detection, a minimum of four flaws in one or more full-scale nozzle mock-ups must be added to the test set. The specimens must comply with Supplement 6, paragraph 1.1, to Appendix VIII, except for flaw locations specified in Table VIII S6-1. Flaws may be notches, fabrication flaws, or cracks. Seventy-five (75) percent of the flaws must be cracks or fabrication flaws. Flaw locations and orientations must be selected from the choices shown in paragraph (b)(2)(xv)(K)(4) of this section. Table VIII-S7-1 Modified, with the exception that flaws in the outer eighty-five (85) percent of the weld need not be perpendicular to the weld. There may be no more than two flaws from each category, and at least one subsurface flaw must be included.

(ii) For length sizing, a minimum of four flaws as in paragraph (b)(2)(xv)fK)(1)(,) of this section must be included in the test set. The length sizing results must be added to the results of combined Supplement 4 to Appendix VIII and Supplement 6 to Appendix VIII. The combined results must meet the acceptance standards contained in paragraph (b)(2)(xv)(E)(3) of this section.

(iii) For depth sizing, a minimum of four flaws as in paragraph (b)(2)fxv)(K)(f)fi of this section must be included in the test set. Their depths must be distributed over the ranges of Supplement 4, Paragraph 1.1. to Appendix VIII, for the inner 15 percent of the wall thickness and Supplement 6, Paragraph 1.1, to Appendix VIII, for the remainder of the wall thickness.

The depth sizing results must be combined with the sizing results from Supplement 4 to Appendix VIII for the inner 15 percent and to Supplement 6 to Appendix VIII for the remainder of the wall thickness. The combined results must meet the depth sizing acceptance criteria contained in paragraphs fb)(2)fxv)fC)(1), fb)(2)(xv)fE)(1), and fb)(2)fxv)(F)(3) of this section.

(2) For examination of reactor pressure vessel nozzle-to-vessel welds conducted from the inside of the vessel, the following conditions are required:

(i) The clad-to-base-metal-interface and the adjacent examination volume to a minimum depth of 15 percent I (measured from the clad-to-base-metal-interface) must be examined from four orthogonal directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII as conditioned by paragraphs (b)(2)(xv)(B) and (C) of this section.

(ii) When the examination volume defined in paragraph (b)(2)(xv)(K)(2)(iJ of this section cannot be effectively examined in all four directions, the examination must be augmented by examination from the nozzle bore using a procedure and personnel qualified in accordance with paragraph (b)(2)(xv)(K)(7) of this section.

(II,) The remainder of the examination volume not covered by paragraph (b)(2)(xv)(K)(2)(ii) of this section or a combination of paragraphs (b)(2)(xv)(K)(2)(i and (ii) of this section, must be examined from the nozzle bore using a procedure and personnel qualified in accordance with paragraph (b)f2)(xv)(K)(f) of this section, or from the vessel shell using a procedure and personnel qualified for single sided examination in accordance with Supplement 6 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.

(3) For examination of reactor pressure vessel nozzle-to-shell welds conducted from the outside of the vessel, the following conditions are required:

(i The clad-to-base-metal-interface and the adjacent metal to a depth of 15 percent T (measured from the clad-to-base metal-interface) must be examined from one radial and two opposing circumferential directions using a procedure and personnel qualified in accordance with Supplement 4 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(B) and (C) of this section, for examinations performed in the radial direction, and Supplement 5 to Appendix VIII, as conditioned by paragraph (b)(2)(xv)fJ) of this section, for examinations performed in the circumferential direction.

(ii) The examination volume not addressed by paragraph (b)(2)fxv)fK)(3)(i of this section must be examined in a minimum Exelon

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IS! Program Plan Units 7 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS of one radial direction using a procedure and personnel qualified for single sided examination in accordance with Supplement 6 to Appendix VIII, as conditioned by paragraphs (b)(2)(xv)(D) through (G) of this section.

(4) Table VlII-S7-1 Flaw Locations and Orientations, Supplement 7 to Appendix VIII, is conditioned as follows:

Table VlIl-S7-f -Modified Flaw Locations and Orientations Parallel to weld Perpendicular to weld

Inner 15 percent X

X t-------,--.--

ODSurface X

Subsurface X

(L) Specimen set and qualification: Twelfth provision. As a condition to the requirements of Supplement 8, Subparagraph 1.1(c), to Appendix VIII, notches may be located within one diameter of each end otthe bolt or stud.

(M) Specimen set and qualification: Thirteenth provision. When implementing Supplement 12 to Appendix VIII, only the provisions related to the coordinated implementation of Supplement 3 to Supplement 2 performance demonstrations are to be applied.

xvi) Section Xl condition: Appendix VIII single side ferritic vessel and piping and stainless steel piping examinations. When applying editions and addenda prior to the 2007 Edition of Section XI, the following conditions apply.

(A) Ferritic and stainless steel piping examinations: First provision. Examinations performed from one side of a ferritic vessel weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as conditioned by this paragraph and paragraphs (b)(2)(xv)(B) through (G) of this section, on specimens containing flaws with non-optimum sound energy reflecting characteristics or flaws similar to those in the vessel being examined.

(B) Ferritic and stainless steel piping examinations: Second provision. Examinations performed from one side of a ferritic or stainless steel pipe weld must be conducted with equipment, procedures, and personnel that have demonstrated proficiency with single side examinations. To demonstrate equivalency to two sided examinations, the demonstration must be performed to the requirements of Appendix VIII, as conditioned by this paragraph and paragraph (b)(2)(xv)(A) of this section.

(xvii) Section Xl condition: Reconciliation of quality requirements. When purchasing replacement items, in addition to the reconciliation provisions of IWA-4200, 1995 Addenda through 1998 Edition, the replacement items must be purchased, to the extent necessary, in accordance with the licensees quality assurance program description required by 10 CFR 50.34(b)(6)(ii).

(xviii) Section Xl condition: NDE personnel certification. (A) NDE personnel certification: First provision. Level I and II nondestructive examination personnel must be recertified on a 3-year interval in lieu of the 5-year interval specified in the 1997 Addenda and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)(U) of this section.

(B) NDE personnel certification: Second provision. When applying editions and addenda prior to the 2007 Edition of Section XI, paragraph IWA-2316 may only be used to qualify personnel that observe leakage during system leakage and hydrostatic tests conducted in accordance with WA 5211(a) and (b).

(C) NDE personnel certification: Third provision. When applying editions and addenda prior to the 2005 Addenda of Section XI, licensees qualifying visual examination personnel for VT-3 visual examination under paragraph IWA-2317 of Section XI must demonstrate the proficiency of the training by administering an initial qualification examination and administering subsequent examinations on a 3-year interval.

(xix)Section XI condition: Substitution of alternative methods. 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 when using the 1998 Edition through the 2004 Edition of Section Xl of the ASME BPV Code. The provisions in IWA 4520(c), 1997 Addenda through the 2004 Edition, allowing the substitution of alternative methods, a combination of methods, or newly developed techniques for the methods specified in the Construction Code, are not approved for use. 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 )Oi) of this section, allowing the substitution of ultrasonic examination for radiographic examination specified in the Construction Code, are not approved for use.

(xx)Section XI condition: System leakage tests-(A) System leakage tests: First provision. When performing system leakage tests in accordance with IWA-5213(a), 1997 through 2002 Addenda, the licensee must maintain a 10-minute hold time after Exelon

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151 Program Plan Units I & 2, Fourth Interval TABLE 1.10-f CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS test pressure has been reached for Class 2 and Class 3 components that ate not in use during normal operating conditions.

No hold time is required for the remaining Class 2 and Class 3 components provided that the system has been in operation for at least 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for insulated components or 10 minutes for uninsulated components.

(B) System leakage tests: Second provision. The NDE provision in IWA-4540(a)(2) of the 2002 Addenda of Section Xl must be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1)Øi) of this section.

(xxi) Section Xl condition: Table IWB-2500-7 examination requirements. (A) Table IWB-2500-1 examination requirements:

First pro vision. The provisions of Table IWB 2500-1, Examination Category B-D, Full Penetration Welded Nozzles in Vessels, Items B3.40 and B3.60 (Inspection Program A) and Items B3.120 and B3.140 (Inspection Program B) 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., all 0.5), may be performed instead of an ultrasonic examination.

(B) (ReservedJ (xxii) Section Xl condition: Surface examination. 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 ot an ultrasonic examination method is prohibited.

(xxiii) Section Xl condition: Evaluation of thermally cut surfaces. The use of the provisions for eliminating mechanical processing of thermally cut surfaces in IWA-4461.4.2 of Section Xl, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section, is prohibited.

(xxiv) Section Xl condition: Incorporation of the performance demonstration initiative and addition of ultrasonic examination criteria. The use of Appendix VIII and the supplements to Appendix VIII and Article 1-3000 of Section XI of the ASME BPV Code, 2002 Addenda through the 2006 Addenda, is prohibited.

(xxv) Section Xl condition: Mitigation of defects by modification. The use of the provisions in IWA-4340, Mitigation of Defects by Modification,Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)fii) of this section are prohibited.

xxvi) Section Xl condition: Pressure testing Class 1, 2 and 3 mechanicaljoints. The repair and replacement activity provisions in IWA-4540(c) of the 1998 Edition of Section Xl for pressure testing Class 1, 2, and 3 mechanical joints must be applied when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii) of this section.

(xxvii) Section Xl condition: Removal of insulation. When performing visual examination in accordance with IWA-5242 of Section XI of the ASME BPV Code, 2003 Addenda through the 2006 Addenda, or IWA-5241 of the 2007 Edition through the latest edition and addenda incorporated by reference in paragraph fa)(1)(ii) of this section, insulation must be removed from 17-4 PH or 410 stainless steel studs or bolts aged at a temperature below 1100 F or having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per square inch or higher.

(xxviii) Section Xl condition: 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)fl):

For R <0, K1 depends on the crack depth (a), and the flow stress (Or). The flow stress is defined by at l/2(a +

where ays is the yield strength and Out is the ultimate tensile strength in units ksi (MPa) and (a) is in units in. (mm). For

- 2 R

0 and Krnax 0.8 x 1.12 o(na), S = 1 and K1= Kmax. For R <-2 and Kmax Kmin0.8 x 1.12 ati(ua), S = 1 and

= (1

- R) Kmaxl3. For R < 0 and Kmax

- Kmin> 0.8 X 1.12 a(iTa), S = 1 and K1 Kmax - Kmjn.

(xxix) Section Xl condition: Nonmandatory Appendix R. Nonmandatory Appendix R, Risk-Informed Inspection Requirements for Piping, of Section XI, 2005 Addenda through the latest edition and addenda incorporated by reference in paragraph (a)(1 )(ii) of this section, may not be implemented without prior NRC authorization of the proposed alternative in accordance with paragraph (z) of this section.

(3) Conditions on ASME OM Code. As used in this section, references to the OM Code refer to the ASME Code for Operation and Maintenance of Nuclear Power Plants, Subsections ISTA, ISTB, ISTC, ISTD, Mandatory Appendices I and II, and Nonmandatory Appendices A through H and J, including the 1995 Edition through the 2006 Addenda, subject to the following conditions:

fi) CM condition: Quality assurance. When applying editions and addenda of the CM Code, the requirements of NQA-1, Quality Assurance Requirements for Nuclear Facilities, 1979 Addenda, are acceptable as permitted by ISTA 1.4 of the 1995 Edition through 1997 Addenda or ISTA-1500 of the 1998 Edition through the latest edition and addenda incotporated Exelon

- Byron Station 1-23 Revision 0

IS! Program Plan Units I & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS by reference in paragraph (a)(1)Ov) of this section, provided the licensee uses its 10 CFR part 50, Appendix B, quality assurance program in conjunction with the CM Code requirements. Commitments contained in the licensees quality assurance program description that are more stringent than those contained in NQA-1 govern CM Code activities. If NQA-1 and the CM Code do not address the commitments contained in the licensees Appendix B quality assurance program description, the commitments must be applied to CM Code activities.

ii) CM condition: Motor-Operated Valve (MOV) testing. Licensees must comply with the provisions for MCV testing in CM Code ISTC 4.2, 1995 Edition with the 1996 and 1997 Addenda, or ISTC-3500, 1998 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)fiv) of this section, and must establish a program to ensure that motor-operated valves continue to be capable of performing their design basis safety functions.

(iii) IReserved]

(iv) CM condition: Check valves (Appendix II). Licensees applying Appendix II, Check Valve Condition Monitoring Program, of the CM Code, 1995 Edition with the 1996 and 1997 Addenda, must satisfy the requirements of(b)(3)fiv)(A) through (C) of this section. Licensees applying Appendix II, 1998 Edition through the 2002 Addenda, must satisfy the requirements of (b)(3)(iv)fA), (B), and (D) of this section.

(A) Check valves: First provision. Valve opening and closing functions must be demonstrated when flow testing or examination methods (nonintrusive, or disassembly and inspection) are used; (B) Check valves: Second provision. The initial interval for tests and associated examinations may not exceed two fuel cycles or 3 years, whichever is longer; any extension of this interval may not exceed one fuel cycle per extension with the maximum interval not to exceed 10 years. Trending and evaluation of existing data must be used to reduce or extend the time interval between tests.

(C) Check valves: Third provision. If the Appendix II condition monitoring program is discontinued, then the requirements of ISTC 4.5,1 through 4.5.4 must be implemented.

(D) Check valves: Fourth provision. The applicable provisions of subsection ISTC must be implemented if the Appendix II condition monitoring program is discontinued.

(v) CM condition: Snubbers ISTD. Article IWF-5000, Inservice Inspection Requirements for Snubbers, of the ASME BPV Code, Section Xl, must be used when performing inservice inspection examinations and tests of snubbers at nuclear power plants, except as conditioned in paragraphs (b)(3)(v)(A) and f B) of this section.

(A) Snubbers: First provision. Licensees may use Subsection ISTD, Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Power Plants, ASME CM Code, 1995 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(iv) of this section, in place of the requirements for snubbers in the editions and addenda up to the 2005 Addenda of the ASME BPV Code,Section XI, IWF-5200(a) and (b) and IWF-5300(a) and (b), by making appropriate changes to their technical specifications or licensee-controlled documents.

Preservice and inservice examinations must be performed using the VT-3 visual examination method described in (WA 22 13.

(B) Snubbers: Second provision. Licensees must comply with the provisions for examining and testing snubbers in Subsection ISTD of the ASME CM Code and make appropriate changes to their technical specifications or licensee-controlled documents when using the 2006 Addenda and later editions and addenda of Section Xl of the ASME BPV Code.

(vi) OM condition: Exercise interval for manual valves. Manual valves must be exercised on a 2-year interval rather than the 5-year interval specified in paragraph ISTC-3540 of the 1999 through the 2005 Addenda of the ASME CM Code, provided that adverse conditions do not require more frequent testing.

(4) Conditions on Design, Fabrication, and Materials Code Cases. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter is subject to the following conditions. Licensees may apply the ASME BPV Code Cases listed in NRC Regulatory Guide 1.84, Revision 36, without prior NRC approval, subject to the following conditions:

fi) Design, Fabrication, and Materials Code Case condition: Applying Code Cases. When an applicant or licensee initially applies a listed Code Case, the applicant or licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.

(ii) Design, Fabrication, and Materials Code Case condition: Applying different revisions of Code Cases. If an applicant or licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the applicant or licensee may continue to apply the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use, until it updates its Code of Record for the component being constructed.

(iii) Design, Fabrication, and Materials Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless an applicant or licensee applied the listed Code Case prior to it being listed as annulled in Exelon

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151 Program Plan Units 1 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS Regulatory Guide 184. If an applicant or licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.84, the applicant or licensee may continue to apply the Code Case until it updates its Code of Record for the component being constructed.

(5) Conditions on inseivice inspection Code Cases. Licensees may apply the ASME BPV Code Cases listed in Regulatory Guide 1.147, Revision 17, without prior NRC approval, subject to the following:

(i) 151 Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.

(ii) 151 Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month IS) program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of Regulatory Guide 1.147, Revision 17.

(iii) 151 Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.147. If a licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.147, the licensee may continue to apply the Code Case to the end of the current 120-month interval.

(6) Conditions on Operation and Maintenance of Nuclear Power Plants Code Cases. Licensees may apply the ASME Operation and Maintenance Code Cases listed in Regulatory Guide 1.192, Revision 1, without prior NRC approval, subject to the following:

(i) OM Code Case condition: Applying Code Cases. When a licensee initially applies a listed Code Case, the licensee must apply the most recent version of that Code Case incorporated by reference in paragraph (a) of this section.

(ii) OM Code Case condition: Applying different revisions of Code Cases. If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in paragraph (a) of this section, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case, as authorized, or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use. Licensees who choose to continue use of the Code Case during subsequent 120-month 151 program intervals will be required to implement the latest version incorporated by reference into 10 CFR 50.55a as listed in Tables 1 and 2 of Regulatory Guide 1.192, Revision 1.

(iii) OM Code Case condition: Applying annulled Code Cases. Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.192. It a licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.192, the licensee may continue to apply the Code Case to the end of the current 120-month interval.

(c) Reactor coolant pressure boundary. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter and each combined license for a utilization facility is subject to the following conditions:

(1) Standards requirement for reactor coolant pressure boundary components. Components that are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section III of the ASME BPV Code, except as provided in paragraphs (c)(2) through (4) of this section.

(2) Exceptions to reactor coolant pressure boundary standards requirement. Components that are connected to the reactor coolant system and are part of the reactor coolant pressure boundary as defined in § 50.2 need not meet the requirements of paragraph (c)(1) of this section, provided that:

(i) Exceptions: Shutdown and cooling capability. 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, or (ii) Exceptions: Isolation capability. The component is or can be isolated from the reactor coolant system by o 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.

(3) Applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to components of the reactor coolant pressure boundary must be determined by the provisions of paragraph NCA-1 140, Subsection NCA of Section III of the ASME BPV Code, subject to the following conditions:

Exelon

- Byron Station 1-25 Revision 0

ISI Program Plan Units I & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (i) Reactor coolant pressure boundary condition; Code edition and addenda. The edition and addenda applied to a component must be those that are incorporated by reference in paragraph (a)(1 )(i) of this section; ii) Reactor coolant pressure boundary condition; Earliest edition and addenda for pressure vessel. The ASME Code provisions applied to the pressure vessel may be dated no earlier than the summer 1972 Addenda of the 1971 Edition; iii) Reactor coolant pressure boundary condition; Earliest edition and addenda for piping, pumps, and valves. The ASME Code provisions applied to piping, pumps, and valves may be dated no earlier than the Winter 1972 Addenda of the 1971 Edition; and (iv) Reactor coolant pressure boundary condition; Use of Code Cases. The optional Code Cases applied to a component must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.

(4) Standards requirement for components in older plants. For a nuclear power plant whose construction permit was issued prior to May 14, 1984, the applicable Code edition and addenda for a component of the reactor coolant pressure boundary continue to be that Code edition and addenda that were required by Commission regulations for such a component at the time of issuance of the construction permit.

fd) Quality Group B components. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter, and each combined license for a utilization facility is subject to the following conditions; (1) Standards requirement for Quality Group B components. For a nuclear power plant whose application for a construction permit under this part, or a combined license or manufacturing license under part 52 of this chapter, docketed after May 14, 1964, or for an application for a standard design approval or a standard design certification docketed after May 14, 1984, components classified Quality Group B Z must meet the requirements for Class 2 Components in Section III of the ASME BPV Code.

(2) Quality Group B; Applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to the systems and components identified in paragraph (d)(1) of this section must be determined by the rules of paragraph NCA-1 140, Subsection NCA of Section III of the ASME BPV Code, subject to the following conditions; (i) Quality Group B condition; Code edition and addenda. The edition and addenda must be those that are incorporated by reference in paragraph (a)(1)(i) of this section; ii) Quality Group B condition; Earliest edition and addenda for components. The ASME Code provisions applied to the systems and components may be dated no earlier than the 1980 Edition; and (iii) Quality Group B condition; Use of Code Cases. The optional Code Cases must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.

fe) Quality Group C components. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each manufacturing license, standard design approval, and design certification application under part 52 of this chapter and each combined license for a utilization facility is subject to the following conditions.

(1) Standards requirement for Quality Group C components. For a nuclear power plant whose application for a construction permit under this part, or a combined license or manufacturing license under part 52 of this chapter, docketed after May 14, 1984, or for an application for a standard design approval or a standard design certification docketed after May 14, 1984, components classified Quality Group C must meet the requirements for Class 3 components in Section III of the ASME BPV Code.

(2) Quality Group C applicable Code and Code Cases and conditions on their use. The Code edition, addenda, and optional ASME Code Cases to be applied to the systems and components identified in paragraph (e)(1) of this section must be determined by the rules of paragraph NCA-1 140, subsection NCA of Section III of the ASME BPV Code, subject to the following conditions; fi) Quality Group C condition; Code edition and addenda. The edition and addenda must be those incorporated by reference in paragraph (a)(1)(i) of this section; (ii) Quality Group C condition; Earliest edition and addenda for components. The ASME Code provisions applied to the systems and components may be dated no earlier than the 1980 Edition; and (iii) Quality Group C condition; Use of Code Cases. The optional Code Cases must be those listed in NRC Regulatory Guide 1.84 that is incorporated by reference in paragraph (a)(3)(i) of this section.

(f) lnse,vice testing requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code and ASME Code for Operation and Maintenance of Nuclear Power Plants as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is Exelon

- Byron Station 1-26 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS subject to the following conditions. Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (f)(4) through (6) of this section must be met only after the Commission makes the finding under § 52.103(g) of this chapter. Requirements for inservice inspection of Class 1, Class 2, Class 3, Class MC, and Class CC components (including their supports) are located in § 50.55a(g).

(1) lnse,vice testing requirements for older plants (pre-1971 CPs). For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued prior to January 1 1971, pumps and valves must meet the test requirements of paragraphs (fif4) and (5) of this section to the extent practical. Pumps and valves that are part of the reactor coolant pressure boundary must meet the requirements applicable to components that are classified as ASME Code Class 1. Other pumps and valves that perform a function to shut down the reactor or maintain the reactor in a safe shutdown condition, mitigate the consequences of an accident, or provide overpressure protection for safety-related systems (in meeting the requirements of the 1986 Edition, or later, of the BPV or OM Code) must meet the test requirements applicable to components that are classified as ASME Code Class 2 or Class 3.

(2) Design and accessibility requirements for performing inservice testing in plants with CPs issued between 1971 and 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, pumps and valves that are classified as ASME Code Class 1 and Class 2 must be designed and provided with access to enable the performance of inservice tests for operational readiness set forth in editions and addenda of Section Xl of the ASME BPV incorporated by reference in paragraph (a)(1)(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively) in effect 6 months before the date of issuance of the construction permit. The pumps and valves may meet the inservice test requirements set forth in subsequent editions of this Code and addenda that are incorporated by reference in paragraph (a)(1)(U) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17; or Regulatory Guide 1.192, Revision 1, that are incorporated by reference in paragraphs (a)(3)Øi) and (hi) of this section, respectively), subject to the applicable conditions listed therein.

(3) Design and accessibility requirements for performing inservice testing in plants with CPs issued after 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit under this part or design approval, design certification, combined license, or manufacturing license under part 52 of this chapter was issued on or after July 1, 1974:

(i)-(ii) [Reserved]

(iii) (A) (B)

(iv) (A) (B)

(v)

(4) (i) (ii)

(iii) [Reserved]

(iv)

(5) (i) (ii) (iii) (iv)

(6) (i) (ii)

(g) Inset-vice inspection requirements. Systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME BPV Code as specified in this paragraph. Each operating license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions. Each combined license for a boiling or pressurized water-cooled nuclear facility is subject to the following conditions, but the conditions in paragraphs (g)(4) through (6) of this section must be met only after the Commission makes the finding under § 52.103(g) of this chapter.

Requirements for inservice testing of Class 1, Class 2, and Class 3 pumps and valves are located in § 50.55a(f).

(1) Inservice inspection requirements for older plants (pre-1971 CPs). For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued before January 1, 1971, components (including supports) must meet the requirements of paragraphs (g)(4) and (g)(5) of this section to the extent practical. Components that are part of the reactor coolant pressure boundary and their supports must meet the requirements applicable to components that are classified as ASME Code Class 1. Other safety-related pressure vessels, piping, pumps and valves, and their supports must meet the requirements applicable to components that are classified as ASME Code Class 2 or Class 3.

(2) Design and accessibility requirements for performing inseivice inspection in plants with CPs issued between 1977 and 1974. For a boiling or pressurized water-cooled nuclear power facility whose construction permit was issued on or after January 1, 1971, but before July 1, 1974, components (including supports) that are classified as ASME Code Class 1 and Class 2 must be designed and be provided with access to enable the performance of inservice examination of such components (including supports) and must meet the preservice examination requirements set forth in editions and addenda of Section II or Section XI of the ASME BPV Code incorporated by reference in paragraph fa)(1) of this section (or the Exelon

- Byron Station 1-27 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) in effect 6 months before the date of issuance of the construction permit. The components (including supports) may meet the requirements set forth in subsequent editions and addenda of this Code that are incorporated by reference in paragraph (a) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section), subject to the applicable limitations and modifications.

(3) Design and accessibility requirements for performing insevice inspection in plants with CPs issued after 7974. For a boiling or pressurized water-cooled nuclear power facility, whose construction permit under this part, or design certification, design approval, combined license, or manufacturing license under part 52 of this chapter, was issued on or after July 1, 1974, the following are required:

(i) IS! design and accessibility requirements: Class I components and supports. Components (including supports) that are classified as ASME Code Class 1 must be designed and be provided with access to enable the performance of inservice examination of these components and must meet the preservice examination requirements set forth in the editions and addenda of Section III or Section Xl of the ASME BPV Code incorporated by reference in paragraph (a)(1) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph fa)(3)(ii) of this section) applied to the construction of the particular component.

(ii) 151 design and accessibility requirements: Class 2 and 3 components and supports. Components that are classified as ASME Code Class 2 and Class 3 and supports for components that are classified as ASME Code Class 1, Class 2, and Class 3 must be designed and provided with access to enable the performance of inservice examination of these components and must meet the preservice examination requirements set forth in the editions and addenda of Section Xl of the ASME BPV Code incorporated by reference in paragraph (a)(1 )(ii) of this section (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, that are incorporated by reference in paragraph (a)(3)(ii) of this section) applied to the construction of the particular component.

(iii)-(iv) [Reserved]

(v) IS! design and accessibility requirements: Meeting later 151 requirements. All components (including supports) may meet the requirements set forth in subsequent editions of codes and addenda or portions thereof that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed therein.

(4) Inservice inspection standards requirement for operating plants. Throughout the service life of a boiling or pressurized water-cooled nuclear power facility, components (including supports) that are classified as ASME Code Class 1, Class 2, and Class 3 must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section Xl of editions and addenda of the ASME BPV Code (or ASME CM Code for snubber examination and testing) that become effective subsequent to editions specified in paragraphs (g)(2) and (3) of this section and that are incorporated by reference in paragraph (a)(1 )(ii) or (iv) for snubber examination and testing of this section, to the extent practical within the limitations of design, geometry, and materials of construction of the components. Components that are classified as Class MC pressure retaining components and their integral attachments, and components that are classified as Class CC pressure retaining components and their integral attachments, must meet the requirements, except design and access provisions and preservice examination requirements, set forth in Section Xl of the ASME BPV Code and addenda that are incorporated by reference in paragraph (a)(1)çi) of this section, subject to the condition listed in paragraph (b)(2)(vi) of this section and the conditions listed in paragraphs (b)(2)(viii) and (ix) of this section, to the extent practical within the limitation of design, geometry, and materials of constwction of the components.

(i) Applicable IS! Code: Initial 120-month interval. Inservice examination of components and system pressure tests conducted during the initial 120-month inspection interval must comply with the requirements in the latest edition and addenda of the Code incorporated by reference in paragraph (a) of this section on the date 12 months before the date of issuance of the operating license under this part, or 12 months before the date scheduled for initial loading of fuel under a combined license under part 52 of this chapter (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using Section Xl, or Regulatory Guide 1.192, Revision 1, when using the CM Code, that are incorporated by reference in paragraphs (a)(3)(ii) and (iii) of this section, respectively), subject to the conditions listed in paragraph (b) of this section.

(ii) Applicable IS! Code: Successive 120-month intervals. Inservice examination 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 Code incorporated by reference in paragraph (a) of this section 12 months before the start of the 120-month inspection interval (or the optional ASME Code Cases listed in NRC Regulatory Guide 1.147, Revision 17, when using Section XI, or Regulatory Guide 1.192, Revision 1, when using the CM Code, that are incorporated by reference in paragraphs (a)(3)Øi) and (iii) of this section), subject to the conditions listed in paragraph (b) of this section. However, a licensee whose inservice inspection interval commences during the 12 through 18-month period after July 21, 2011, may delay the update of their Appendix VIII program by up to 18 months after July 21, 2011.

Exelon

- Byron Station 1-28 Revision 0

IS! Program Plan Units 7 & 2, Fotirth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (iii) Applicable 151 Code: Optional surface examination requirement. When applying editions and addenda prior to the 2003 Addenda of Section XI of the ASME BPV Code, licensees may, but are not required to, perform the surface examinations of high-pressure safety injection systems specified in Table IWB-2500-1, Examination Category BJ: Item Numbers 89.20, B9.21, and 89.22.

(iv) Applicable 151 Code: Use of subsequent Code editions and addenda. Inservice examination of components and system pressure tests may meet the requirements set forth in subsequent editions and addenda that are incorporated by reference in paragraph (a) of this section, subject to the conditions listed in paragraph (b) of this section, and subject to Commission approval. Portions of editions or addenda may be used, provided that all related requirements of the respective editions or addenda are met.

(v) Applicable 151 Code: Metal and concrete containments. For a boiling or pressurized water-cooled nuclear power facility whose construction permit under this part or combined license under part 52 of this chapter was issued after January 1, 1956, the following are required:

(A) Metal and concrete containments: First provision. Metal containment pressure retaining components and their integral attachments must meet the inservice inspection, repair, and replacement requirements applicable to components that are classified as ASME Code Class MC; (B) Metal and concrete containments: Second provision. Metallic shell and penetration liners that are pressure retaining components and their integral attachments in concrete containments must meet the inservice inspection, repair, and replacement requirements applicable to components that are classified as ASME Code Class MC; and (C) Metal and concrete containments: Third provision. Concrete containment pressure retaining components and their integral attachments, and the post-tensioning systems of concrete containments, must meet the inservice inspections, repair, and replacement requirements applicable to components that are classified as ASME Code Class CC.

(5) Requirements for updating ISI programs(i) ISI program update: Applicable lSl Code editions and addenda, The inservice inspection program for a boiling or pressurized water-cooled nuclear power facility must be revised by the licensee, as necessary, to meet the requirements of paragraph (g)f4) of this section.

(ii) ISI program update: Conflicting ISI Code requirements with technical specifications. If a revised inservice inspection program for a facility conflicts with the technical specifications for the facility, the licensee must apply to the Commission for amendment of the technical specifications to conform the technical specifications to the revised program. The licensee must submit this application, as specified in § 50.4, at least six months before the start of the period during which the provisions become applicable, as determined by paragraph (g)(4) of this section.

(iii) ISI program update: Notification of impractical 151 Code requirements. If the licensee has determined that conformance with a Code requirement is impractical for its facility the licensee must notify the NRC and submit, as specified in § 50.4, information to support the determinations. Determinations of impracticality in accordance with this section must be based on the demonstrated limitations experienced when attempting to comply with the Code requirements during the inservice inspection interval for which the request is being submitted. Requests for relief made in accordance with this section must be submitted to the NRC no later than 12 months after the expiration of the initial or subsequent 120-month inspection interval for which relief is sought.

(iv) IS! program update: Schedule for completing impracticality determinations. Where the licensee determines that an examination required by 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.

(6) Actions by the Commission for evaluating impractical and augmented ISI Code requirementsO) Impractical 151 requirements: Granting of relief. The Commission will evaluate determinations under paragraph (g)(5) of this section that code requirements are impractical. The Commission 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.

(ii) Augmented 151 program. The Commission 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 reliability is necessary.

(A) [Reserved]

(B) Augmented 151 requirements: Submitting containment IS! programs. 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 WE and Subsection IWL with specified conditions. The program elements and the required documentation must be maintained on site for audit.

(C) Augmented IS! requirements: Implementation of Appendix VIII to Section Xl. (1) Appendix VIII and the supplements to Exelon

- Byron Station 1-29 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS Appendix VIII to Section Xl, Division 1 1995 Edition with the 1996 Addenda of the ASME BPV Code must be implemented in accordance with the following schedule: Appendix VIII and Supplements 1 2, 3, and 8May 22, 2000; Supplements 4 and 6November22, 2000; Supplement 11November22, 2001; and Supplements 5,7, and 10November22, 2002.

(2) Licensees implementing the 1989 Edition and earlier editions and addenda of IWA-2232 of Section Xl, Division 1, of the ASME BPV Code must implement the 1995 Edition with the 1996 Addenda of Appendix VIII and the supplements to Appendix VIII of Section Xl, Division 1, of the ASME BPV Code.

(D) Augmented 151 requirements: Reactor vessel head inspections(7) All licensees of pressurized water reactors must augment their inservice inspection program with ASME Code Case N-729-1, subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section. Licensees of existing operating reactors as of September 10, 2008, must implement their augmented inservice inspection program by December 31, 2008. Once a licensee implements this requirement, the First Revised NRC Order EA-03-009 no longer applies to that licensee and shall be deemed to be withdrawn.

(2) Note 9 of ASME Code Case N-729-1 must not be implemented.

(3) Instead of the specified examination method requirements for volumetric and surface examinations in Note 6 of Table 1 ot Code Case N-729-1, the licensee must perform volumetric and/or surface examination of essentially 100 percent of the required volume or equivalent surfaces of the nozzle tube, as identified by Figure 2 otASME Code Case N-729-1. A demonstrated volumetric or surface leak path assessment through all J-groove welds must be performed. If a surface examination is being substituted for a volumetric examination on a portion of a penetration nozzle that is below the toe of the J-groove weld [Point Eon Figure 2 of ASME Code Case N-729-1], the surface examination must be of the inside and outside wetted surface of the penetration nozzle not examined volumetrically.

(4) By September 1, 2009, ultrasonic examinations must be performed using personnel, procedures, and equipment that have been qualified by blind demonstration on representative mockups using a methodology that meets the conditions specified in paragraphs (g)(6)(ii)(D)(4)(i) through (iv), instead of the qualification requirements of Paragraph -2500 of ASME Code Case N-729-1, References herein to Section XI, Appendix VIII, must be to the 2004 Edition with no addenda of the ASME BPV Code.

(i) The specimen set must have an applicable thickness qualification range of +25 percent to -40 percent for nominal depth through-wall thickness. The specimen set must include geometric and material conditions that normally require discrimination from primary water stress corrosion cracking (PWSCC) flaws.

(ii) The specimen set must have a minimum often (10) flaws that provide an acoustic response similar to PWSCC indications. All flaws must be greater than 10 percent of the nominal pipe wall thickness. A minimum of 20 percent of the total flaws must initiate from the inside surface and 20 percent from the outside surface. At least 20 percent of the flaws must be in the depth ranges of 10-30 percent through-wall thickness and at least 20 percent within a depth range of 31-50 percent through-wall thickness. At least 20 percent and no more than 60 percent of the flaws must be oriented axially.

(II,) Procedures must identify the equipment and essential variables and settings used for the qualification, in accordance with Subarticle VIII-2100 of Section Xl, Appendix VIII. The procedure must be requalified when an essential variable is changed outside the demonstration range as defined by Subarticle VllI-3130 of Section XI, Appendix VIII, and as allowed by Articles VIII-4100, VllI-4200, and VllI-4300 of Section Xl, Appendix VIII. Procedure qualification must include the equivalent of at least three personnel performance demonstration test sets. Procedure qualification requires at least one successful personnel performance demonstration.

(iv) Personnel performance demonstration test acceptance criteria must meet the personnel performance demonstration detection test acceptance criteria of Table VIIIS10-1 of Section Xl, Appendix VIII, Supplement 10. Examination procedures, equipment, and personnel are qualified for depth sizing and length sizing when the RMS error, as defined by Subarticle VlIl-3120 of Section XI, Appendix VIII, of the flaw depth measurements, as compared to the true flaw depths, do not exceed 1/8 inch (3 mm) and the root mean square (RMS) error of the flaw length measurements, as compared to the true flaw lengths, do not exceed 3/8 inch (10 mm), respectively.

(5) If flaws attributed to PWSCC have been identified, whether acceptable or not for continued service under Paragraphs 3130 or -3140 of ASME Code Case N-729-1, the re-inspection interval must be each refueling outage instead of the re inspection intervals required by Table 1, Note (8), otASME Code Case N-729-1.

(6) Appendix I of ASME Code Case N-729-1 must not be implemented without prior NRC approval.

(2) Augmented IS! requirements: Reactor coolant pressure bounda,y visual inspections i(7) All licensees of pressurized water reactors must augment their inservice inspection program by implementing ASME Code Case N-722-1, subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through (4) of this section. The inspection requirements of ASME Code Case N-722-1 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/1 82 materials that have been mitigated by weld overlay or stress improvement.

(2) If a visual examination determines that leakage is occurring from a specific item listed in Table 1 of ASME Code Case N Exelon

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IS! Program P1n Units I & 2. Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS 722-1 that is not exempted by the ASME Code, Section Xl, IWB-1220(b)(1), additional actions must be performed to characterize the location, orientation, and length of a crack or cracks in Alloy 600 nozzle wrought material and location, orientation, and length of a crack or cracks in Alloy 82/1 82 butt welds. Alternatively, licensees may replace the Alloy 600/82/182 materials in all the components under the item number of the leaking component.

(3) If the actions in paragraph fg)(6)fii)(E)(2) of this section determine that a flaw is circumferentially oriented and potentially a result of primary water stress corrosion cracking, licensees must perform non-visual NDE inspections of components that fall under that ASME Code Case N-722-i item number. The number of components inspected must equal or exceed the number of components found to be leaking under that item number. If circumferential cracking is identified in the sample, non-visual NDE must be performed in the remaining components under that item number.

(4) If ultrasonic examinations of butt welds are used to meet the NDE requirements in paragraphs (g)(6)(ii)(E)(2) or (3) of this section, they must be performed using the appropriate supplement of Section XI, Appendix VIII, of the ASME BPV Code.

(F) Augmented IS! requirements: Examination requirements for Class 7 piping and nozzle dissimilar-metal butt weldsti)

Licensees of existing, operating pressurized-water reactors as of July 21, 2011, must implement the requirements ofASME Code Case N-770-1, subject to the conditions specified in paragraphs (g)(6)çi)(F)(2) through (70) of this section, by the first refueling outage after August 22, 2011.

(2) Full structural weld overlays authorized by the NRC staff may be categorized as Inspection Items C or F, as appropriate.

Welds that have been mitigated by the Mechanical Stress Improvement Process (MSIPTM) may be categorized as Inspection Items D orE, as appropriate, provided the criteria in Appendix I of the Code Case have been met. For ISI frequencies, all other butt welds that rely on Alloy 82/1 82 for structural integrity must be categorized as Inspection Items A 1, A-2 or B until the NRC staff has reviewed the mitigation and authorized an alternative Code Case Inspection Item for the mitigated weld, or until an alternative Code Case Inspection Item is used based on conformance with an ASME mitigation Code Case endorsed in Regulatory Guide 1.147 with conditions, if applicable, and incorporated by reference in this section.

(3) Baseline examinations for welds in Table 1, Inspection Items A-i, A-2, and B, must be completed by the end of the next refueling outage after January 20, 2012. Previous examinations of these welds can be credited for baseline examinations if they were performed within the re-inspection period for the weld item in Table 1 using Section XI, Appendix VIII, requirements and met the Code required examination volume of essentially 100 percent. Other previous examinations that do not meet these requirements can be used to meet the baseline examination requirement, provided NRC approval of alternative inspection requirements in accordance with paragraphs (z)(i) or (2) of this section is granted prior to the end of the next refueling outage after January 20, 2012.

(4) The axial examination coverage requirements of Paragraph2500(c) may not be considered to be satisfied unless essentially 100 percent coverage is achieved.

(5) All hot-leg operating temperature welds in Inspection Items G, H, J, and K must be inspected each inspection interval. A 25 percent sample of Inspection Items G, H, J, and K cold-leg operating temperature welds must be inspected whenever the core barrel is removed (unless it has already been inspected within the past 10 years) or 20 years, whichever is less.

(6) For any mitigated weld whose volumetric examination detects growth of existing flaws in the required examination volume that exceed the previous IWB-3600 flaw evaluations or new flaws, a report summarizing the evaluation, along with inputs, methodologies, assumptions, and causes of the new flaw or flaw growth is to be provided to the NRC prior to the weld being placed in service other than modes 5 or 6.

(7) For Inspection Items G, H, J, and K, when applying the acceptance standards of ASME BPV Code, Section Xl, IWB 3514, for planar flaws contained within the inlay or onlay, the thickness V in IWB-3514 is the thickness of the inlay or onlay.

For planar flaws in the balance of the dissimilar metal weld examination volume, the thickness t in IWB-3514 is the combined thickness of the inlay or onlay and the dissimilar metal weld.

(8) Welds mitigated by optimized weld overlays in Inspection Items D and E are not permitted to be placed into a population to be examined on a sample basis and must be examined once each inspection interval.

(9) Replace the first two sentences of Extent and Frequency of Examination for Inspection Item D in Table 1 of Code Case N-770-i with, Examine all welds no sooner than the third refueling outage and no later than 10 years following stress improvement application. Replace the first two sentences of Note (11 )(b)(2) in Code Case N-770-1 with, The first examination following weld inlay, onlay, weld overlay, or stress improvement for Inspection Items D through K must be performed as specified.

(10) General Note (b) to Figure 5(a) of Code Case N-770-1 pertaining to alternative examination volume for optimized weld overlays may not be applied unless NRC approval is authorized under paragraphs (z)(i) or (2) of this section.

(h) Protection and safety systems. Protection systems of nuclear power reactors of all types must meet the requirements specified in this paragraph. Each combined license for a utilization facility is subject to the following conditions.

(1) [Reserved}

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IS! Prog-am Plan Units I & 2, Fourth Interval TABLE 1.10-1 CODE OF FEDERAL REGULATIONS 10 CFR 50.55a REQUIREMENTS LIMITATIONS, MODIFICATIONS, AND CLARIFICATIONS (2)

(3) fi)-fy) [Reserved]

(z) Alternatives to codes and standards requirements. Alternatives to the requirements of paragraphs (b) through (h) of this section or portions thereof may be used when authorized by the Director, Office of Nuclear Reactor Regulation, or Director, Office of New Reactors, as appropriate. A proposed alternative must be submitted and authorized prior to implementation.

The applicant or licensee must demonstrate that:

(1) Acceptable level of quality and safety. The proposed alternative would provide an acceptable level of quality and safety; or (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.

Footnotes to § 50.55a:

USAS and ASME Code addenda issued prior to the winter 1977 Addenda are considered to be in effect or effective 6 months after their date of issuance and after they are incorporated by reference in paragraph (a) of this section. Addenda to the ASME Code issued after the summer 1977 Addenda are considered to be in effect or effective after the date of publication of the addenda and after they are incorporated by reference in paragraph (a) of this section.

2-3 [Reserved].

For ASME Code editions and addenda issued prior to the winter 1977 Addenda, the Code edition and addenda applicable to the component is governed by the order or contract date for the component, not the contract date for the nuclear energy system. For the winter 1977 Addenda and subsequent editions and addenda the method for determining the applicable Code editions and addenda is contained in Paragraph NCA 1140 of Section lii of the ASME Code.

5-6 [Reserved].

Z Guidance for quality group classifications of components that are to be included in the safety analysis reports pursuant to

§ 50.34(a) and § 50.34(b) may be found in Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-,

Steam-, and Radiological-Waste-Containing Components of Nuclear Power Plants, and in Section 3.2.2 of NUREG-0800, Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants.

[Reserved].

For inspections to be conducted once per interval, the inspections must be performed in accordance with the schedule in Section Xl, paragraph IWB-2400, except for plants with inservice inspection programs based on a Section Xl edition or addenda prior to the 1994 Addenda. For plants with inservice inspection programs based on a Section Xl edition or addenda prior to the 1994 Addenda, the inspection must be performed in accordance with the schedule in Section Xl, paragraph IWB-2400, of the 1994 Addenda.

[80 FR 45843, Aug. 3,2015] Page Last Reviewed/Updated Wednesday, December 02, 2015 Exelon

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IS! Program Plan Units I & 2, Fourth Interval 1.11 CODECASES Per 10 CFR 50.55a(b)(5), ASME Code Cases that have been determined to be suitable for use in ISI Program Plans by the NRC are listed in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability-ASME Section Xl, Division 1. The approved Code Cases in Regulatory Guide 1.147 being utilized by Byron Station, are included in Section 2.1.2 of this document. The most recent version of a given Code Case incorporated in the revision of Regulatory Guide 1.147 referenced in 10 CFR 50.55a(b)(5)(i) at the time it is applied within the ISl Program shall be used. As this guide is revised, newly approved Code Cases may be assessed for plan implementation at Byron Station per Paragraph IWA-2441(e) and proposed for use in revisions to the ISI Program Plan.

Per the latest revision of Regulatory Guide 1.147, if a Code Case is implemented by a licensee and a later version of the Code Case is incorporated by reference into 10 CFR 50.55a and listed in Tables 1 and 2 during the licensees present 120-month 151 program interval, that licensee may use either the later version or the previous version. An exception to this provision would be the inclusion of a limitation or condition on the use of the Code Case that is necessary, for example, to enhance safety.

The use of Code Cases (other than those listed in Regulatory Guide 1.147) may be authorized by the Director of the Office of Nuclear Reactor Regulation upon request pursuant to 10 CFR 50.55a(z). Code Cases not approved for use in Regulatory Guide 1.147, which are being utilized by Byron Station through associated Relief Requests, are included in Section 8.0.

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

The submittal of this information will be referred to in this document as a Relief Request.

Relief Requests for the Fourth ISI Interval and the Second CISI Interval are included in Section 8.0 of this document. The text of the Relief Requests contained in Section 8.0 will demonstrate one of the following: the proposed alternatives provide an acceptable level of quality and safety per 10 CFR 50.55a(z)f 1); or compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety per 10 CFR 50.55a(z)(2), or the Code requirements are considered impractical per 10 CFR 50.55a(g)(5)(iii).

Per 10 CFR 50.55a Paragraphs (z) and (g)(6)(i), the NRC will evaluate relief requests and may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

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IS! PrcraIT Plan Units 7 & 2, Fourth Interval 2.0 BASIS FOR INSERVJCE INSPECTION PROGRAM 2.1 ASME SECTION Xl EXAMINATION REQUIREMENTS 2.1.1 Welds and Components, Supports, and Pressure Tests As required by 10 CFR 50.55a, this program was developed in accordance with the requirements detailed in the 2007 Edition with the 2008 Addenda, of the ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, Subsections IWA, IWB, IWC, IWD, IWE, IWF, IWL, Mandatory Appendices, Inspection Program of Paragraph IWA-2431, approved ASME Code Cases, and approved alternatives through Relief Requests and Safety Evaluations (SEs).

The Performance Demonstration Initiative (PDI) is an organization comprised of all United States (US) nuclear utilities that was formed to provide an efficient implementation of Appendix VIII performance demonstration requirements. The Electric Power Research Institute (EPRI) NDE Center was selected as the administrator of this program. The PDI program is administered according to the PDI Program Description. The ISI Program implements Appendix VIII Performance Demonstration for Ultrasonic Examination Systems, ASME Section XI 2007 Edition with the 2008 Addenda with modifications as identified in 10 CFR 50.55a(b)(2)(xiv). Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in piping, bolting, and the reactor pressure vessel fRPV). Each organization (e.g., owner or vendor) is required to have a written program to ensure compliance with the requirements. Byron Station maintains the responsibility to ensure that Appendix VIII requirements are properly implemented. In accordance with ASME Section Xl, ultrasonic examinations performed during the Fourth ISI Interval, as required by Paragraph IWA-2232, shall use the requirements identified in Mandatory Appendix I, Ultrasonic Examinations. Per 1-2600, Appendix VIII of ASME Section Xl may be applied to components for which it is not applicable, provided such components, materials, sizes, and shapes are within the scope of the qualified examination procedure.

For the Fourth ISI Interval, Byron Stations inspection program for ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 will be governed by risk-informed regulations. The RI-ISI Program methodology is described in the EPRI Topical Report TR-1 12657, Rev. B-A. To supplement the EPRI Topical Report, N-578-1 (as applicable per Relief Request I4R-01) is also being used for the classification of piping structural elements under the RI-ISI Program. The RI ISI Program scope has been implemented as an alternative to the 2007 Edition with the 2008 Addenda ASME Section Xl examination program for ISI Class 1 B-F and B-J welds and ISI Class 2 C-F-i and C-F-2 welds in accordance with 10 CFR 50.55a(z)(i). The basis for the resulting Risk Categorizations of the nonexempt ISI Class 1 and 2 piping systems at Byron Station is defined and maintained in the Final Report Risk Informed Inservice Inspection Evaluation as referenced in Section 9.0 of this document. References to ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 have been replaced with Examination Category R-A to identify them as part of the RI-ISI program.

For the Fourth Inspection Interval, the RI-ISI Program scope continues to include welds in the BER piping, also referred to as the HELB region, which includes several Non-Class welds that fall within the BER augmented inspection program. The BER program methodology is described in EPRI Topical Report TR-i006937, Rev. 0-A, which has been used to define the inspection scope in lieu of the 100% volumetric examination of all piping welds in the previous BER augmented program. Therefore, all welds in the original augmented program for BER remains evaluated under the RI-lSl Program using an integrated risk-informed approach.

The CISI Program Plan per Subsections WE and IWL has been incorporated into Section 6.0 Containment ISI Plan of this ISI Program Plan. The CISI relief requests are included in Section 8.0 of this document.

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IS! Program Plan Units I & 2. Fourth Interval 21.2 ASME Section Xl Code Cases As referenced by 10 CFR 50.55a(b)(5) and allowed by NRC Regulatory Guide 1.147, Revision 17, being the latest incorporated into this ISI Program Plan, the following Code Cases are incorporated into the Byron Station lSl Program. These Code Cases have been determined by the NRC to be acceptable alternatives to applicable parts of ASME Section Xl. These Code Cases may be used by Byron Station without a Relief Request from the NRC, provided that they are used with any identified limitations or modifications. Code Cases implemented through the Relief Request process are included in Section 8.0 of this document. Some of the Code Cases listed below are acceptable to the NRC for application at Byron Station within the limitations imposed by the NRC staff. Unless otherwise stated, limitations imposed by the NRC are in addition to the conditions specified in the Code Case. Several of these Code Cases are included as contingencies, to ensure that they are available for future repair/replacement activities.

N-432-1 Repair Welding Using Automatic or Machine Gas Tungsten-Arc Welding (GTAW)

Temper Bead Technique.

N-508-4 Rotation of Serviced Snubbers and Pressure Retaining Items for the Purpose of Testing.

ASME Code Case N-508-4 is acceptable subject to the following condition:

When Section Xl requirements are used to govern the examination and testing of snubbers and the 151 Code of Record is earlier than Section Xl, 2006 Addenda, Footnote 1 shall not be applied.

N-5 13-3 Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping.

ASME Code Case N-513-3 is acceptable subject to the following condition:

The repair or replacement activity temporarily deferred under the provisions of this Code Case shall be performed during the next scheduled outage.

N-526 Alternative Requirements for Successive Inspections of Class 7 and 2 Vessels.

N-532-5 Alternative Requirements to Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission.

N-561-2 Alternative Requirements for Wall Thickness Restoration of Class 2 and High Energy Class 3 Carbon Steel Piping.

ASME Code Case N-561-2 is acceptable subject to the following conditions:

(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.

(2) Paragraph 7(c): if the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage.

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like defects exist.

(4) Piping with wall thickness less than the diameter of the electrode shall be depressurized before welding.

N-562-2 Alternative Requirements for Wall Thickness Restoration of Class 3 Moderate Energy Carbon Steel Piping.

ASME Code Case N-562-2 is acceptable subject to the following conditions:

(1) Paragraph 5(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.

(2) Paragraph 7(c): if the cause of the degradation has not been determined, the Exelon

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iS! Program Plan Units 7 & 2, Fourth Interval repair is only acceptable until the next refueling outage.

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like defects exist.

(4) Piping with wall thickness less than the diameter of the electrode shall be depressurized before welding.

N-569-1 Alternative Rules for Repair by Electrochemical Deposition of Classes I and 2 Steam Generator Tubing.

ASME Code Case N-569-1 is acceptable subject to the following conditions:

NOTES: Steam generator tube repair methods requite prior NRC approval through the Technical Specifications. This Code Case does not address certain aspects of this repair, e.g., the qualification of the inspection and plugging criteria necessary for staff approval of the repair method.

In addition, if the user plans to reconcile, as described in Footnote 2, the reconciliation is to be performed in accordance with IWA-4200 in the 1995 Edition, 1996 Addenda of ASME Section Xl.

N-586-1 Alternative Additional Examination Requirements for Classes 1, 2, and 3 Piping, Components, and Supports.

Note: This Code Case is implemented for Examination Categories other than R-A.

RI-ISI Program Relief Request 14R-01 requires that scope expansion for RI-ISI piping structural elements will be determined using Paragraph -2430 of N-578-1.

N-597-2 Requirements for Analytical Evaluation of Pipe Wall Thinning.

ASME Code Case N-597-2 is acceptable subject to the following conditions:

(1) Code Case must be supplemented by the provisions of EPRI Nuclear Safety Analysis Center Report 202L-R2, Recommendations for an Effective Flow Accelerated Corrosion Program (Ref. 6), April 1999, for developing the inspection requirements, the method of predicting the rate of wall thickness loss, and the value of the predicted remaining wall thickness. As used in NSAC-202L-R2, the term should is to be applied as shall (i.e., a requirement).

(2) Components affected by flow-accelerated corrosion to which this Code Case are applied must be repaired or replaced in accordance with the construction code of record and Owners requirements or a later NRC approved edition of Section III, Rules for Construction of Nuclear Power Plant Components, of the ASME Code (Ref. 7) prior to the value of tp reaching the allowable minimum wall thickness, tmin, as specified in -3622.1(a)(1) of this Code Case. Alternatively, use of the Code Case is subject to NRC review and approval per 10 CFR 50.55a(z).

(3) For Class 1 piping not meeting the criteria of -3221, the use of evaluation methods and criteria is subject to NRC review and approval per 10 CFR 50.55a(z).

(4) For those components that do not require immediate repair or replacement, the rate of wall thickness loss is to be used to determine a suitable inspection frequency so that repair or replacement occurs prior to reaching allowable minimum wall thickness, tmin.

(5) For corrosion phenomenon other than flow accelerated corrosion, use of the Code Case is subject to NRC review and approval. Inspection plans and wall thinning rates may be difficult to justify for certain degradation mechanisms such as MIC and pitting.

(6) The evaluation criteria in Code Case N-513-2 may be applied to Code Case N 597-2 for the temporary acceptance of wall thinning (until the next refueling outage) for moderate-energy Class 2 and 3 piping. Moderate energy piping is defined as Class 2 and 3 piping whose maximum operating temperature does not exceed 200°F (93°C) and whose maximum operating pressure does not exceed Exelon

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(SI Prcm Plan Units 1 & 2, Fourth Interval 275 psig (1.9 MPa). Code Case N-597-2 shall not be used to evaluate through wall leakage conditions.

N-600 Transfer of Welder, Welding Operator, Brazer, and Brazing Operator Qualifications Between Owners.

N-6 13-1 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. 83. 70 and 83.90, Reactor Nozzle-to-Vessel Welds, Figs.

IWB-2500-7(a), (b), and (c).

N-629 Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials.

N-639 Alternative Calibration Block Material.

ASME Code Case N-639 is acceptable subject to the following conditions:

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 requited by the material specification.

N-641 Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements.

N-643-2 Fatigue Crack Growth Rate Curves for Ferritic Steels in PWR Water Environment.

N-648-i Alternative Requirements for Inner Radius Examinations of Class 7 Reactor Vessel Nozzles.

ASME Code Case N-648-1 is acceptable subject to the following condition:

In lieu of a UT examination, licensees may perform a VT-i 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-651 Ferritic and Dissimilar Metal Welding Using SMAW Temper Bead Technique Without Removing the Weld Bead Crown for the First Layer.

N-660 Risk-Informed Safety Classification for Use in Risk-Informed Repair/Replacement Activities.

The Code Case must be applied only to ASME Code Classes 2 and 3, and non-Code Class pressure retaining components and their associated supports.

N-661 -2 Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service.

ASME Code Case N-661-2 is acceptable subject to the following conditions:

(1) Paragraph 4(b): for repairs performed on a wet surface, the overlay is only acceptable until the next refueling outage.

(2) Paragraph 7(c):

if the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage.

(3) The area where the weld overlay is to be applied must be examined using ultrasonic methods to demonstrate that no crack-like defects exist.

(4) Piping with wall thickness less than the diameter of the electrode shall be depressurized before welding.

N-705 Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks.

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IS! Program Plan Units 7 & 2, Fourth Interval N-706-1 Alternative Examination Requirements of Table IWB-2500-1 and Table IWC-2500-I for PWR Stainless Steel Residual and Regenerative Heat Exchangers. See TAP 14T-07 for details.

N-73 1 Alternative Class I System Leakage Test Pressure Requirements.

N-735 Successive Inspections of Class 7 and 2 Piping Welds.

N-747 Reactor Vessel Head-to-Flange Weld Examinations.

N-751 Pressure Testing of Containment Penetration Piping.

When a 10 CFR 50, Appendix J, Type C test is performed as an alternative to the requirements of IWA-4540 (IWA-4700 in the 1989 edition through the 1995 edition) during repair and replacement activities, nondestructive examination must be performed in accordance with IWA-4540(a)(2) of the 2002 Addenda of Section Xl.

N-762 Temper Bead Procedure Qualification Requirements for Repair/Replacement Activities Without Postweld Heat Treatment.

N-765 Alternative to Inspection Interval Scheduling Requirements of IWA-2430.

N-773 Alternative Qualification Criteria for Eddy Current Examinations of Piping Inside Surfaces.

Additional Code Cases invoked in the future shall be in accordance with those approved for use in the latest published revision of Regulatory Guide 1.147 or 10 CFR 50.55a at that time.

2.2 AUGMENTED EXAMINATION REQUIREMENTS Augmented examination requirements are those examinations that are performed in addition to the requirements of ASME Section Xl. Below is a summary of those examinations performed by Byron Station that are not specifically prescribed by ASME Section Xl, or the examinations that will be performed in addition to the requirements of ASME Section Xl on a routine basis during the Fourth 161 Interval and Third CISI Interval. Previous revisions of the Byron Station ISI Program categorized some Augmented Examination Programs by using Augmented Numbers.

2.2.1 NRC Mechanical Engineering Branch (MEB) Technical Position 3-1 (MEB 3-1), High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment, dated November 24, 1975.

UFSAR Sections 3.6.1 and 3.6.2 detail Byron Station compliance with NRC Branch Technical Position MEB 3-1, which includes requirements for licensees to perform a 100% volumetric examination each interval of circumferential and longitudinal pipe welds within the pipe break exclusion regions associated with high energy piping in containment penetration areas.

Implementation of the examination requirements is included in Section 7.0 of this 151 Program Plan and the associated ISI Database.

Note: This requirement was previously maintained in accordance with UFSAR Section 3.6.1 and 3.6.2. With the implementation of the Rl-lSl-BER Program, all BER augmented welds were evaluated under the RI-ISI methodology and were integrated into the RI-ISI Program.

The RI-ISI Program will also include several Non-Class welds that fall within the BER augmented examination program. Additional guidance for adaptation of the RI-ISI evaluation process to BER piping is given in EPRI TR-1 006937, Rev. 0-A.

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IS! Program Plan Units 7 & 2, Fotiflh Interval 2.2.2 NRC Regulatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity, as modified by the requirements of Byron Station License Amendment #118 and Technical Requirements Manual Appendix G.

The requirement to Regulatory Guide 1.14 has been modified by Byron Station License Amendment #118 and Technical Requirements Manual Appendix G.

In lieu of Regulatory Position c4.b.(1) and c4.b.(2), a qualified in-place ultrasonic testing (UT) examination over the volume from the inner bore of the flywheel to the circle of one-half the outer radius or a surface examination (MT and/or PT) of exposed surfaces of the removed flywheel may be conducted at approximately 10 year intervals coinciding with the Inservice Inspection schedule as required by ASME Section XI.

Implementation of the examination requirements is included in Section 7.0 of this ISl Program Plan and the associated ISI Database.

2.2.3 Byron Station UFSAR Section 10.2.3, Turbine Disk and Rotor Integrity.

This details Byron Stations requirement to perform visual and magnetic particle examination of the accessible areas of the high-pressure turbine rotor, low-pressure turbine blades, and low-pressure disks. In addition, visual examinations of the turbine coupling and coupling bolts are performed.

This program has been removed from the Engineering Group and is maintained by the Turbine Maintenance organization.

2.2.4 NRC Bulletin 88-08, Thermal Stresses in Piping Connected to Reactor Coolant Systems, including supplements 1, 2, and 3.

This details Byron Stations requirement to examine welds susceptible to thermal stratification.

To address NRC Bulletin 88-08, Byron Station had committed to inspecting critical locations on the suction lines to the Residual Heat Removal (RHR) pumps from RCS Loops 1 and 3, and the Auxiliary and Main Pressurizer Spray lines every other refueling outage.

With the implementation of the RI-ISI Program, the NRC Bulletin 88-08 augmented inspection requirement will no longer be required at Byron Station. The RI-ISI Program completely subsumes this requirement because the Degradation Mechanism assessment and Risk Categorization involve full assessment for Thermal Transients and Thermal Stratification, Cycling, and Striping. Thus, these piping structural elements have been categorized and selected for examination in accordance with the EPRI Topical Report TR-1 12657, Rev. B-A and N-578-1 in lieu of the original requirement to NRC Bulletin 88-08. The evaluation of susceptible lines is also addressed under MRP-146 (see below).

2.2.5 Information Notice 79-19, Pipe Cracks in Stagnant Borated Water Systems at PWR Plants.

Volumetric examinations will be performed on ISI Class 2 ECCS systems (or portions of systems) that are currently not subject to evaluation under the RI-lSl Program. The inspections include 7.5% sampling of the total population of circumferential piping welds (greater than 4 inches nominal pipe size) that contain stagnant borated water.

For the current inspection interval, the areas subject to augmented examination are limited to the 10 Safety Injection piping from the SI Accumulators (1/2SIO4TA, B, C, and D) to the class boundary second check valve (1/2SI8956A, B, C, and D). These lines are exempted from ASME Section Xl surface or volumetric examination by Paragraph IWC-1221(c).

The components selected for these examinations are to be examined before the end of the inspection interval.

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ISI Prc:ram P12n Units 7 & 2, FoUrth Interval 2.2.6 NRC NUREG 0737, TMlAction Plan Requirements, Section IlI.D.1.1, dated November 1980.

Requires applicants to implement a program to reduce leakage from systems outside containment that would or could contain highly radioactive fluids or gases during a serious transient or accident to as low as practical levels.

In response to this NUREG requirement, Section E.77, Primary Coolant Sources Outside Containment, was included in the Byron/Braidwood Station UFSAR. This UFSAR Section along with Technical Specifications Section 5.5.2 require performance of integrated leak tests at refueling cycle intervals or less on each system or portions of systems, which could potentially contain highly radioactive liquids or gases.

Implementation of the Byron Station program addressing these requirements is included in site procedure BVP 200-7.

2.2.7 Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Pressure Boundary Components in PWR Plants.

The NRC issued Generic Letter (GL) 88-05 to all licensees of operating Pressurized Water Reactors (PWR) in March, 1988. This Generic Letter deals with boric acid corrosion of carbon steel reactor coolant pressure boundary components in PWR plants. Specifically, GL 88-05 requested information to assess safe operation of PWRs when reactor coolant leaks below Tech Spec limits develop and the coolant containing Boric Acid comes in contact with and degrades low alloy carbon steel components. Byron Stations response to GL 88-05 requirements are incorporated through the completion of normal station operator walkdowns, heightened Maintenance and Tech Staff (now System Engineering) training, the normal Inservice Inspection Program, and the ASME Section XI System Pressure Testing Program.

To ensure compliance with this augmented examination requirement, the Reactor Coolant Pressure Boundary (RCPB), as defined by UFSAR Section 5.2, shall have a system inspection performed by certified VT-2 visual examiners every refueling outage consisting of a pre-outage visual examination as well as a visual examination conducted prior to startup.

These examinations shall be conducted to identify evidence of boric acid crystallization and residue accumulations.

Implementation of the Byron Station program addressing these requirements is included in site procedure BVP-200-7.

2.2.8 N-722-1 Augmented Examination Program.

Per 10 CFR 50.55a(g)(6)(ii)(E)(7), all licensees of pressurized water reactors shall augment their inservice inspection program by implementing ASME Code Case N-722-1 (N-722-1) subject to the conditions specified in paragraphs (g)(6)(ii)(E)(2) through (4). The inspection requirements of N-722-1 do not apply to components with pressure retaining welds fabricated with Alloy 600/82/1 82 materials that have been mitigated by weld overlay or stress improvement. This requirement is implemented with the Second Period of the Third ISI Interval.

Implementation of the examination requirements is included in Section 7.0 of this ISI Program Plan and the associated SI Database.

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(SI Program Plan Units I & 2, Fotiflh Interval TABLE 2.2.9-1 N-722-1 TABLE I EXAMINATION ITEMS ITEM NO.

DESCRIPTION BYRON STATION APPLICABILITY Reactor Vessel B15.80 RPV bottom-mounted instrument penetrations 58 nozzles per unit.

B1 5.90 Hot leg nozzle-to-pipe connections Not Applicable. Note 1.

B15.95 Cold leg nozzle-to-pipe connections Not Applicable. Note 1.

B15.100 Instrument connections Not Applicable.

Steam Generators B15.110 Hot leg nozzle-to-pipe connections Not Applicable. Non-A600 material.

B15.115 Cold leg nozzle-to-pipe connections Not Applicable. Non-A600 material.

615.120 Bottom channel head drain tube penetration Not Applicable. Note 2.

615.130 Primary side hot leg instrument connections Not Applicable. Non-A600 material.

B15.135 Primary side cold leg instrument connection Not Applicable. Non-A600 material.

Pressurizer B15.140 Heater penetrations Not Applicable. Non-A600 material.

Bi 5.150 Spray nozzle-to-pipe connections Not Applicable. Note 3.

815.160 Safety and relief nozzle-to-pipe connections Not Applicable. Note 3.

815.170 Surge nozzle-to-pipe connections Not Applicable. Note 3.

815.180 Instrument connections Not Applicable. Non-A600 material.

815.190 Drain nozzle-to-pipe connections Not Applicable.

Piping 615.200 Hot leg instrument connections Not Applicable. Non-A600 material.

615.205 Cold leg instrument connections Not Applicable. Non-A600 material.

815.210 Hot leg full penetration welds Not Applicable.

61 5.215 Cold leg full penetration welds Not Applicable.

Note 1: Safe-end welds have been mitigated by stress improvement are exempted from N-722-1 requirements per 10 CFR 50.55a(g)(6)(ii)(E)(1).

Note 2: The Unit 2 Bottom channel head drain tube penetrations have been replaced with resistant materials and are exempted from N-722-1 requirements.

Note 3: Safe-end welds have been mitigated by weld overlay with resistant material are exempted from N-722-1 requirements per 10 CFR 50.55a(g)(6)(ii)(E)(1).

2.2.9 N-729-1 Augmented Examination Program.

Pet 10 CFR 50.55a(g)(6)(ii)(D)(I), all licensees of pressurized water reactors shall augment their inservice inspection program with ASME Code Case N-729-1 (N-729-1) subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6). N-729-1 governs the visual and volumetric/surface examinations of the reactor vessel closure head penetrations and surrounding exterior surface. This requirement was implemented with the First Period of the Third 151 Interval.

Implementation of the examination requirements is included in Section 7.0 of this SI Program Plan and the associated ISI Database.

2.2.10 N-770-1 Augmented Examination Program.

Per 10 CFR 50.55a(g)f6)(ii)(F)(l), licensees of existing, operating pressurized-water reactors as of July 21, 2011 shall implement the requirements of ASME Code Case N-770-1 (N-770-1),

subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(1 0) of this Exelon

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IS! Procram Plan Units 7 & 2, Focirth lntenial section, by the first refueling outage after August 22, 2011. Tnis requirement replaced the pervious industry standard Material Reliability Program: Primary System Piping Butt Weld Inspection and Evaluation Guideline (MRP-139). N-770-1 governs the visual and volumetric/surface examinations of the Alloy 600 butt welds. This requirement was implemented with the Second Period of the Third ISI Interval.

Implementation of the examination requirements is included in Section 7.0 of this 161 Program Plan and the associated ISI Database.

2.2.11 MRP-146 Augmented Examination Program.

This guideline is for the screening, evaluation, and inspection requirements for potential thermal fatigue cracking that may occur in normally stagnant non-isolable piping systems attached to pressurized water reactor (PWR) main reactor coolant system (RCS) piping. The objective of this guideline is to provide a common industry approach to use in effectively reducing the probability of cracking in and leakage from piping potentially susceptible to thermal fatigue. Some of the piping that is covered by this guideline was previously identified as being susceptible to thermal fatigue with the issuance of NRC Bulletin 88-08. The scope of this guideline applies only to ISI Class 1 piping.

Currently, the examinations are limited to the following locations:

Cold leg charging lines 1/2RC28A-3 and 1/2RC37A-3.

Loop Drain lines J/2RC14AX-2 (Loops A, B, C, and D)

Implementation of the examination requirements is included in Section 7.0 of this ISI Program Plan and the associated ISI Database.

2.2.12 MRP-192 Augmented Examination Program.

MRP-192 was issued as a good practice industry guideline per NEI 03-08 classification, because of a thermal fatigue leakage event in piping downstream of a RHR heat exchanger.

Current operating conditions at Byron Station for temperature differentials of the mixing flows and cumulative operating time in mixing mode, are within the limits specified in MRP-192 (<

144°F), therefore, no examinations are currently scheduled.

2.2.13 Appendix Q Program.

Non-Alloy 600/82/182 butt welds with full-structural overlays are removed from the RI-ISI Program and treated solely under the requirements of ASME Section Xl, 2007 Edition with the 2008 Addenda, Nonmandatory Appendix Q. These locations may include repaired welds or welds adjacent to Alloy 600/82/182 welds where both welds were overlaid due to the proximity of the welds.

Implementation of the examination requirements is included in Section 7.0 of this 161 Program Plan and the associated ISI Database.

2.2.14 MRP-227 Augmented Examination Program PWR Internals Program augmented examinations are implemented in accordance with MRP 227. The specific scope and requirements of this program are provided in BB-PBD-AMP Xl.M16A Reactor Vessel Internals and the applicable program documents ER-AP-333 Pressurized Water Reactor Internals Management Program and ER-AP-333-1 001 Pressurized Water Reactor (PWR) Internals Program. These examinations should be performed simultaneously with the 10-Year ISI Reactor Vessel Internals examinations on a 10-year frequency starting during refueling outage B1R26 in Fall 2024 for Unit 1 (Fourth ISI Interval), and during refueling outage B2R26 in Fall 2026 for Unit 2 (Fifth ISI Interval).

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IS! Projram Plan Units 7 & 2, Focirth Intewal Implementation of the examination requirements is included in Section 7.0 of this ISI Program Plan and the associated 151 Database. These components are Augmented Category, Item Number MRP-227.xx in the ISI Database, where xx is used to identify the type of component.

(Note that the NRC has recently endorsed MRP-227, so specific examination scope and schedule is currently being developed by EGC. Byron Station has committed to submitting a reactor vessel internals inspection plan to the NRC no later than 24 months prior to entering the period of extended operation. This inspection plan will provide detailed information on components subject to examination and the schedule for examination.)

2.2.15 License Renewal

- (CM-3)

The license renewal supplement to the UFSAR, Appendix F, describes enhancements to the 151 Programs beyond the requirements of ASME Section Xl.

The ASME Section Xl, Subsection IWF aging management program commitments include the following:

1.

Examinations of the MC supports, at both ends of the fuel transfer tube (in the refueling cavity in the Containment Structure and in the refueling canal in the Fuel Handling Building), are added to the scope of the program.

2.

Periodic visual examinations of all (100%) high strength bolts (ASME SA 540 and ASTM A490 materials), greater than 1 in diameter, used on the steam generator, reactor coolant pump, and pressurizer supports, are added to the scope of the program. The periodic visual examinations, to detect a corrosive environment that supports the potential for stress corrosion cracking, are to be performed prior to the period of extended operation, and then each inspection interval of ten years thereafter.

3.

VT-3 visual examinations, of the control rod drive mechanism (CRDM) seismic support assembly, are added to the scope of the program, for Class I component supports, during every ten (10) year 161 interval. The reactor head lifting lugs, which also provide restraint for the bottom of the (CRDM) seismic support assembly, are included as part of the CRDM seismic support assembly examinations.

Implementation of the commitments is included in Sections 4 and 7 of this 151 Program Plan and the associated ISI Database.

2.3 SYSTEM CLASSIFICATIONS AND P&ID BOUNDARY DRAWINGS The lSl Classification Basis Document details those systems that are lSl Class 1, 2, 3, or MC that fall within the 151 scope of examinations including the containment structures (metal and concrete) and post-tensioning system, which are shown on the containment roll-out drawings.

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

Each safety related, fluid system containing water, steam, air, oil, etc. included in the Byron Station UFSAR was reviewed to determine which safety functions they perform during all modes of system and plant operation. Based on these safety functions, the systems and components were evaluated per classification documents. The systems were then designated as 151 Class 1, 2, 3, MC, or Non-Classed accordingly. This evaluation followed the guidelines of UFSAR Section 5.2.4 for Class 1 and 6.6 for Classes 2 and 3. Safety related portions of systems are defined on the Piping and Instrument Diagrams (P&lDs) and Control and Instrumentation Diagrams (C&lDs).

When a particular group of components is identified as performing a 161 Class 1, 2, or 3 safety function, the components are further reviewed to assure the interfaces (boundary valves and boundary barriers) meet the criteria set by 10 CFR 50.2, 10 CFR 50.55a(c)(1), 10 CFR Exelon

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ISI Program Plan Units I & 2, Fotnih Interval 50.55a(c)(2), Regulatory Guide 1.26, and ANSI N 18.2-1973. Although Byron Station is not committed to or licensed in accordance with these documents, Standard Review Plan (SRP) 3.2.2 System Quality Group Classification, and other American National Standards Institute/American Nuclear Society (ANSI/ANS) standards were also used for guidance in evaluating the classification boundaries when 10 CFR and Regulatory Guide 1.26 did not address a given situation. The valve positions shown on the system flow diagrams are assumed to be the normal positions during system operation unless otherwise noted.

SI classification boundaries are defined by the 151 Code Boundary Drawings (lSl CBDs) with classification line codes. A summary of the line coding system used on the ISI CBDs to identify safety related systems or portions of systems subject to examination is included on drawing ISl-CBD-LEGEND. Typically, unhatched, solid coding (blue, yellow and green, Coding Designators 1A, 2A, and 3A, respectively) was used for nonexempt ASME Section Xl components. Some hatched codings, (Coding Designators 2HPSI, 2F, and 3C) were also used to identify nonexempt ASME Section Xl components. The remaining codings shown on lSl-CBD-LEGEND (Coding Designators IB, JC, 1D, 2B, 2C, 2D, 2E, 3B, and 3D) were used to identify exempt ASME Section Xl components.

In addition to the line coding system shown on lSl-CBD-LEGEND, codings used to develop Byron Station Units 1 and 2 System Pressure Testing Program are shown on drawing SPT-TBD-LEGEND, Sheet 1.

The systems and components (piping, pumps, valves, vessels, etc.), which are subject to the examinations of Articles IWB-2000, IWC-2000, IWD-2000, and IWF-2000, and pressure tests of Articles IWB-5000, IWC-5000, and, IWD-5000 are identified on the ISI CBDs as detailed in Tables 2.3-1 and 2.3-2. Containment components subject to examination of Articles IWE 2000 and IWL-2000 are identified on the CISI Drawings shown in Tables 2.4-3 and 2.4-4.

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IS! Proaram Plan Units I & 2, Fourth Interval TABLE 2.3-1 COLOR CODED ISI P&ID BOUNDARY DRAWINGS UNIT I & COMMON

[

UNIT 2 SYSTEM OR DESCRIPTION M-34-1 2, 3, 4, 5 M-34-1 2, 3, 4, 5 P&ID Index & Symbols M-35-1, 2 M-120-1, 2A, 2B Main Steam (MS)

M-36-1A 18, 1C, 1D M-121-1A 13, 1C, 1D M-152-45 M-152-45 Feedwater (FW)

M-37 M-122 Auxiliary Feedwater (AF)

M-42-1A lB 2A 2B 3 M-42-1A 18, 2A 28 4 5A 5B 6 7 M-126-1 2 3 EssentIal Service Water (SX)

M-46-IA, 1B, 1C M-129-1A, 18, JC Containment Spray(CS)

M-47-2 M-150-2 Off Gas Hydrogen Recombiners fOG)

M-48-5A M-48-5B Waste Disposal

- Steam Generator Slowdown (SD)

M-48-6A, 6B M-48-6A, 68 Waste Disposal Aux. Building Floor Drains (RF)

M-48-18 Waste Disposal Resin Removal (WX)

M-49-1A M-49-1B Make-Up Demineralizer (WM)

M-50-1A, 18, 1C, JD, 3 M-130-1A, 18, 2 Diesel Fuel Oil (DO)

M-52-1 Fire Protection (FP)

M-54-2, 4A M-54-2, 48 Service Air (SA)

M-55-4, 9 M-55-5, 7D Instrument Air (IA)

M-59-1A, lB M-149 Nitrogen (NT)

M-60-JA lB 2 3 4 5 M-135-1A lB 2, 3 4, 5 6 8 6 8 Reactor Coolant (RC & RY)

M-61-1A, 18, 2,3,4, 5, M-136-1, 2,3,4, 5,6 Safety Injection (SI)

M-62 M-137 Residual Heat Removal (RH)

M-63-1A, 1B, 1C M-63-1A, 13, 1C Fuel Pool Cooling and Clean-Up (FC)

M-64-1 2 3A 38 4A M-138-1 2 3A 3B 4 4B 5

5A 58 ChemIcal and Volume Control (CV)

M-64-6 7 M-1 38-6 7 Chemical and Volume Control I Boron Thermal Regeneration (CV & BR)

M-65-1B, 2A, 3, 5A, 58, M-65-1B, 5A, 58,6 Boric Acid (AB)

M-66-IA 18 2 3A 3B M-66-3A, 38, 43, 4A 4C 4D 4C, 4D, Component Cooling (CC)

M-139-1, 2 M-68-1A, 18, 6, 7, 8 M-140-1, 5, 6 Process Sampling (PS)

M-69-1, 2, 3 Radioactive Waste Gas (GW)

M 70 1 2

M 141-1 2 Reactor Building Equipment Drains & Vents to Radwaste

(RE)

M-76-6, 10 Process Radiation Monitoring (PR)

M-82-l, 2, 3, 5, 14, 15 M-82-1, 2, 3, 5, 6 Auxiliary Building & Containment Equipment Drains (WE)

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is! P-cqram Plan Units 1 & 2, Fourth Interval TABLE 2.3-1 COLOR CODED ISI P&ID BOUNDARY DRAWINGS UNIT 1 & COMMON UNIT 2 SYSTEM OR DESCRIPTION M-105-1 M-106-1 Containment Purge! Pressure & Vacuum Relief Systems (VQ & VP)

M-105-3 M-105-3 Integrated Leak Rate System (VO)

M-118-1, 5,14 M-118-7 Control Room Chilled Water (WO)

M-152-9 M-152-9, 10 Diesel Generator Lube Oil (DG & DO)

M-152-14 M-152-14 Diesel Generator Jacket Water (DG)

M-152-19 M-152-19 Diesel Generator Cooling Water (DG)

M-152-20 M-152-20 Diesel Generator (DG)

TABLE 2.3-2 COLOR CODED ISI C&ID BOUNDARY DRAWINGS UNIT 1 & COMMON UNIT 2 TITLE M-2060-6, 7,8, 17,_18_[

M-2135-6, 7,8, 17,18 C&ID Reactor Coolant System (RC)

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IS! Proriram Plan Units 1 & 2. Fourth IntenaI 2.4 ISP ISOMETRIC AND COMPONENT DRAWINGS FOR NONEXEMPT IS! CLASS COMPONENTS AND SUPPORTS PSI Isometric and Component drawings were developed to identify the ISP Class 1, 2, 3 components (welds, bolting, etc.) and support locations at Byron Station. These ISP component and support locations are identified on the ISI Isometric and Component drawings listed in Tables 2.4-1 and 2.4-2. The ISI Class MC and CC components are identified on the CISI Component Drawings listed in Tables 2.4-3 and 2.4-4.

Byron Stations ISI Program, including the ISP Database, 151 Classification Basis Document, and ISI Selection Document, addresses the non-exempt components, which require examination and testing.

A summary of Byron Station ASME Section Xl nonexempt components and supports is included in Section 7.0.

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/5/ Program Plan Units I & 2, Fociflh Interval TABLE 2.4-1 UNIT I & COMMON ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DR W TI NUMBER NUMBER A

ING TLE 1AF-1-ISI 1

Auxiliary Feedwater Lines 1AFO2DA-4, 1AF02DE-4, and 1AFO2EA-4 1AF-1-ISI 2

Auxiliary Feedwater Lines JFW06M-4 and 1FW87BA-3 [[::JAF-1-lSI|JAF-1-lSI]] 3

Auxiliary Feedwater Lines 1AFO2DB-4, 1AF02DF-4, and 1AF02EB-4 1AF-1-ISI 4

Auxiliary Feedwater Lines 1FW06AB-4 and 1FW87BB-3 1AF-1-lSl 5

Auxiliary Feedwater Lines 1AF02DC-4. 1AF02DG4, and 1AFQ2EC-4 1AF-1-ISI 6

Auxiliary Feedwater Lines JFW06AC-4 and 1FW87BC-3 1AF-1-ISI 7

Auxiliary Feedwater Lines JAF02DD-4, 1AFO2DH-4, and 1AF02ED-4 1AF-1-lSl 8

Auxiliary Feedwater Lines 1FWO6AD-4 and 1FW87BD-3 1CS-1-ISI 1

Containment Spray Line 1CS02M-10 1CS-1-ISI 2

Containment Spray Line 1CS1OAA-6 1CS-1-ISI 3

Containment Spray Lines 1CSO1AA-16, 1CS23AA-14, and 1CS06AA-6 1CS-1-ISI 4

Containment Spray Lines 1CSO1AB-16, ICS23AB-14, and 1CSO6AB-6 1CS-1-ISI 5

Containment Spray Line 1CS02AB-10 1CS-1-lSl 6

Containment Spray Lines 1CSO2AB-10 and 1CS1OAB-6 1CS-1-lSI 7

Containment Spray Line 1CS02M-10 1CV-1-ISI 1

Chemical & Volume Control Line 1CVB7A-3 1CV-1-ISI 2

Chemical & Volume Control Lines 1RY18A-2 and 1CV45B-2 1CV-1-ISl 3

Chemical & Volume Control Lines 1CV14FB-2 and 1CV14GB-1%

1CV-1-lSl 4

Chemical & Volume Control Lines 1CVA5AB-2 and 1CVA6AB-2 1CV-1-lSI 5

Chemical & Volume Control Line 1CVA3B-2 1CV-1 -ISI 6

Chemical & Volume Control Lines 1CV14FA-2 and JCVI4FD-2 1CV-1-lSl 7

Chemical & Volume Control Line 1CVA3B-2 1CV-1-ISI 8

Chemical & Volume Control Line 1CVA5AA-2 1CV-1-ISI 9

Chemical & Volume Control Lines 1CVA3B-2, JCVA3AB-2, and 1CVA7AB-2 iCy-i -ISI 10 Chemical & Volume Control Line 1CVA3AB-2 1CV-1-ISI 11 Chemical & Volume Control Lines 1CVA3B-2 and ICVA6AA-2 1CV-1-ISI 12 Chemical & Volume Control Line 1CV45B-2 1CV-1-ISl 13 Chemical & Volume Control Line 1CVA3B-2 1CV-1-ISl 14 Chemical & Volume Control Line 1CVA3B-2 1CV-1-lSl 15 Chemical & Volume Control Line JCVA3B-2 1CV-1-ISI 16 Chemical & Volume Control Lines 1CV14FC-2 and 1CV14GC-11/2 1CV-1-ISI 17 Chemical & Volume Control Lines 1CV99A-8, 1CVO5B-8, and 1CVA1A-6 1CV-1-ISI 18 Chemical & Volume Control Lines 1CVO5B-8, 1CVO5CA-6, 1CV98BA-8, 1CV98BB-8, and 1 CV98BC-8 1CV-1-ISI 19 Chemical & Volume Control Line 1CVO5CB-6 1CV-1-ISI 20 Chemical & Volume Control Lines 1CVO8AB-4, 1CV12AA-3, and 1CV42M-2 1CV-1-ISI 21 Chemical & Volume Control Lines JCV]4A-4, 1CVO9A-4, and 1CVO8BA-4 1FW-1 -ISI 1

Feedwater Lines 1FWO3DD-16 and IFW86AD-16 1FW-1-ISI 2

Feedwater Lines 1FWO3DA-16 and 1FW86AA-16 1FW-1 -151 3

Feedwater Lines 1FW86AB-16 and 1FWO3DB-16 1FW-1-ISI 4

Feedwater Lines 1FWO3DC-16 and 1FW86AC-16 1FW-1-ISI 5

Feedwater Lines 1FW81AB-6, 1flN81BB-6. and 1FW87CB-6 iFw-i-IsI 10 Feedwater Lines 1FW81AC-6, 1FW81BC-6, and 1FW87CC-6 1FW-1-ISI 11 Feedwater Lines 1FW81AA-6, 1FW81BA-6, and 1FW87CA-6 1FW-1-ISI 12 Feedwater Lines 1FW81AD-6, 1AN81BD-6, and 1FW87CD-6 1MS-1-ISI 1

Main Steam Line 1MSO1AD-30 1/4 (Loop 4)

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IS! Prcgram Plan Units 7 & 2. Focirth Interval TABLE 2.4-1 UNIT I & COMMON ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DRAWING 77TLE NUMBER NUMBER 1MS-1-ISI 2

Main Steam Lines 1MSO1BD-30 1/4, 1MSO7AD-28, 1MS13AD-8, 1MSO7BD-28, and 1MS143AD-1 2 (Loop 4) 1MS-1-ISI 3

Main Steam Line 1MS0JM-30 1/4 (Loop 1) 1MS-1lSl 4

Main Steam Lines 1MS01BA-30 1/4, 1MSO7AA-28, IMS13M-8, 1MSO7BA-28, and 1MS143AA-12 (Loop 1) 1MS-1-ISI 5

Main Steam Line 1MSO1AB-32 3/4 (Loop 2) 1MS-1-ISI 6

Main Steam Lines 1MS01BB-32 3/4, 1MSO7AB-28, IMS13AB-8, 1MSO7BB-28, and 1MS143AB-12 (Loop 2) 1MS-1-ISI 7

Main Steam Line 1MSO1AC-32 3/4 (Loop 3) 1MS-1-lSI 8

Main Steam Lines 1MSO1BC-32 3/4, 1MSO7AC-28, 1MS13AC-8, 1MS143AC-12, and 1MSO7BC-28 (Loop 3)

JRC-1-lSI 1

Primary Coolant System Loop 1 To Steam Gen. No. 1RC-01-BA IRC-1-lSl 2

Primary Coolant System Loop 2 To Steam Gen. No. 1RC-01-BB 1RC-1-lSl 3

Primary Coolant System Loop 3 To Steam Gen. No. 1RC-01-BC 1 RC-1 -ISI 4

Primary Coolant System Loop 4 To Steam Gen. No. 1 RC-0 1-SD 1RC-1-lSI 5

Reactor Coolant Surge Line 1RY11A-14 1RC-1-ISI 6

Reactor Coolant Lines 1RC21AA-8 and 1RC21BA-8 1RC-1-lSl 7

Reactor Coolant Lines 1RC28A-3, 1CV10DA-3, 1RC37A-3, 1CV1ODB-3, and IRC36A-3 1RC-1-lSl 9

Reactor Coolant Line 1RC21AB-8 1RC-1-lSl 11 Reactor Coolant Lines 1RC04AB-12 and 1RC05AB-6; Residual Heat Removal Line 1RH01AB-12 1RC-1-ISI 12 Reactor Coolant Lines 1RC21AC-8 and 1RC21BC-8 JRC-1-ISI 14 Reactor Coolant Lines 1RC24AB-4 and 1RY01AB-4 1RC-1-ISI 15 Reactor Coolant Lines 1RC21AD-8 and 1RC21SD-8 JRC-1-ISI 16 Reactor Coolant Lines JRYO1B-6 and 1RYO1C-4 1RC-1-ISI 17 Reactor Coolant Lines 1RC24M-4 1RYO1AA-4, 1RYO1AB-4, and 1RYOIB-6 1RC-1-ISI 19 Reactor Coolant Lines 1RC22AB-1% and 1RC46AB-3 1RC-1-ISI 20 Reactor Coolant Lines 1RC22AD-IW and 1RC46AD-3 RC-1-ISI 21 Reactor Coolant Line 1RC22AB-1Y2 1 RC-1-ISI 22 Reactor Coolant Lines 1 RCO5AA-6 (Loop 2) and 1 RC35AB-6 (Loop 4)

JRC-1-ISI 23 Reactor Coolant Lines 1RC22AA-1% and 1RC46M-3 1RC-1-ISI 24 Reactor Coolant Lines 1RC22AC-1% and JRC46AC-3 1RC-1-ISI 27 Reactor Coolant Lines 1RC22M-11,4 and 1RC22AC-1%

1RC-1-ISI 29 Reactor Coolant Lines 1RC16AC-2 (Loop 3) and 1RC16AD-2 (Loop 4) 1RC-1-ISI 30 Reactor Coolant Lines 1RC13AA-2, 1RC13AB-2, 1RC13AC-2, and 1RC13AD-2 1 RC-1-lSI 31 Reactor Coolant Lines 1 RCI4AB-2 (Loop 2) and 1 RC26A-2 (Loop 4) 1RC-1-lSI 32 Reactor Coolant Lines 1RYO3M-6, 1RYO3AB-6, 1RYO3AC-6, 1RYO3BA-6, 1RYO3BB-6, and 1 RYO3BC-6 JRC-1-ISI 35 Reactor Coolant Lines 1RYO2A-6, 1RYO6A-3, and 1RYO2B-3 1RC-1-ISI 36 Reactor Coolant Lines 1RCI4AA-2 and ICVA3M-2 1RC-1-ISI 37 Reactor Coolant Lines 1RC14AD-2 and 1CVA7M-2 1RC-1-ISI 41 Reactor Coolant Lines 1RC16AA-2 (Loop 1) and 1RC16AB-2 (Loop 2) 1 RC-1 -ISI 42 Reactor Coolant Line 1 RC1 4AC-2 1PZR-1 -151 Pressurizer No. 1RY-01-S 1RCP-1-ISI Reactor Coolant Pumps 1RC-01-PA, JRC-01-PB, 1RC-01-PC, and IRC-01-PD 1RPV-1-lSI Reactor Pressure Vessel No. 1RC-01-R 1SG-1-ISI 5

Replacement Steam Generator No. 1RC-01-BA Exelon

- Byron Station 2-16 Revision 0

IS! Program P/an Units 7 & 2, Fourth Interval TABLE 2.4-1 UNIT I & COMMON 151 ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET I

R WING TILE LNUMBER NUMBER]

D A I

1SG-1-ISI 6

Replacement Steam Generator No. 1RC-01-BB 1SG-1-ISI 7

Replacement Steam Generator No. 1RC-01-BC JSG-1-ISl 8

Replacement Steam Generator No. 1RC-01-BD 1RH-1-ISl 1

Residual Heat Removal Line 1RHO1AB-1 2 1RH-1-ISI 2

Residual Heat Removal Line IRHO1M-12 1RH-1-lSI 3

Residual Heat Removal Lines 1RH03M-8 and 1RH12A-8 JRH-1-ISI 4

Residual Heat Removal Lines 1RHOJBA-12 and 1RH01CA16u 1RH-1-lSl 5

Residual Heat Removal Lines 1RHO2M-8 and 1RH09M-8 1RH-1-ISI 6

Residual Heat Removal Lines 1RHO2AB-8, 1RHD3AB-8, and 1RHO9AB-8 1RH-1-ISI 7

Residual Heat Removal Lines 1RHO3AB-8, 1RH14A-8, and 1RHO3AA-8 1RH-1-ISI 8

Residual Heat Removal Lines 1RHO1BB-12. 1RHOJCB-16, and 1SI82BB-12 1RH-1-ISI 9

Residual Heat Removal Line 1RHO2AB-8 1RHP-1-ISl Residual Heat Removal Pumps 1RH-01-PA-1-1A and 1RH-01-PB-2-1B 1RHX-1-lSI Residual Heat Exchanger Nos. 1RH-02-AA and 1RH-02-AB 1SD-1-ISI 1

nservice Inspection Isometric Cont. Bldg. & Safety Valve Rm.

Loop 1 1SD-1-lSl 2

Inservice Inspection Isometric Cont. Bldg. & Safety Valve Rm.

- Loop 2 1SD-1-ISI 3

Inservice Inspection Isometric Cont. Bldg. & Safety Valve Rm.

- Loop 3 1SD-1-ISI 4

Inservice Inspection Isometric Cont. Bldg. & Safety Valve Rm.

- Loop 4 1SI-1-ISI 1

Safety Injection Lines IRC29AA-10 and 1SIO9BA-10 1SI-1-lSI 2

Safety Injection Lines ISIA4B-8, 1SIO3FA-2, IRCQ4M-12, and IRC35M-6 1SI-1-ISI 3

Safety Injection Line 1SIO5DA-6 lSl-1-ISI 4

Safety Injection Lines ISIO5BA-8, 1SIO5CA-8, and 1SIO5CD-8 SI-1-ISI 5

Safety Injection Lines JRC29AB-10 and 1SI09BB-10 151-1-151 6

Safety Injection Lines ISIO5DB-6 and 1SI18FB-2 1SI-1-ISI 7

Safety Injection Lines ISIOBFA-3, JSIO8FB-3, and 1SIO8E-3 1SI-1-ISI 8

Safety Injection Line 1SIO8FA-3 151-1-ISI 9

Safety Injection Lines 1RC29AC-10 and 1SIO9BC-10 1SI-1-ISI 10 Safety Injection Lines 1SIO5DC-6 and 1SI18FC-2 151-1-151 11 Safety Injection Lines 1SIQ4D-8 and 1SIO3DB-2 1SI-1-ISI 12 Safety Injection Lines 1SIO4A-12, 1SIO4B-12, 1SIO4C-8, and 1SIA4A-8 lSl-1-ISI 13 Safety Injection Lines 1RC29AD-10 and 1SIO9BD-10 151-1-ISI 14 Safety Injection Line 1SID5DD-6 1SI-1-ISI 15 Safety Injection Lines 1SIO8JC-1W 1RC45AC-3, and 1RC3QAC-1%

151-1-151 16 Safety Injection Lines 1SIO8JD-1%, 1RC45AD-3, and 1RC3OAD-1W 1SI-1-ISI 17 Safety Injection Lines 1S108]B-11/2, 1RC45AB-3, and 1RC3OAB-1Y2 1S1-1-ISI 18 Safety Injection Lines 1SIO8HB-2, 15108GB-lW, and 1SIO8JB-1W 151-1-151 19 Safety Injection Lines 1SIO8GA-1W, 1SIO8HA-2, and 1SIO8JA-11,4 1SI-1-ISI 20 Safety Injection Lines 1SIO8GC-1Y2, 1SIO8HC-2, 1SIO8JC-1 W, 1SIO8GD-11/2, 1SIO8HD-2, and 1 SIO8JD-1 %

1SI-1-ISI 21 Safety Injection Line 1SIO3DA-2 1SI-1-lSI 22 Safety Injection Line 1SIO3FB-2 1SI-1-ISl 23 Safety Injection Lines 1SI18FA-2 and 1SI18FD-2, and Reactor Coolant Line 1RY76A-2 1SI-1-ISI 24 Safety Injection Lines 1SIO6BA-24 and 1SIO6BB-24 1SI-1-ISI 25 Safety Injection Line 1SIO5AA-8 1SI-1-ISI 26 Safety Injection Lines 1SIO5BB-8, 1SIO5CB-8, and 1SIO5CC-8 1S1-1-ISI 27 Safety Injection Line 1SIO8JD-1W Exelon

- Byron Station 2-17 Revision 0

ISI Program Pkn Units I & 2, Fourth Interval TABLE 2.4-1 UNIT I & COMMON ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DRAWING TITLE NUMBER NUMBER 1SI-1-ISI 28 Safety Injection Lines 1SIO9AB-10 & 1SIO9AC-l0 1SI-1-ISI 29 Safety Injection Line 1SIO8JC-l%

1SI-l-ISI 31 Safety Injection Lines lSlO8]A-1Y2, 1RC3OAA-1%, and 1RC45AA-3 1SI-1-ISI 32 Safety Injection Line 1SIO5AB-8 1SI-1-ISI 33 Safety Injection Line 1S134A-8 1SI-l-ISI 34 Safety Injection Lines JSIO2A-8, 1SIO1B-24, and 1SI82AB-12 1SI-1-ISI 35 Safety Injection Lines 1S182M-12, 1SIO1A-8, JSI53AA-14, and 1SIO1B-24 lSl-1-lSI 36 Safety Injection Lines 1SIO2BB-6, 1SIF9A-8, and 1SIO2BA-6 1SI-l-ISI 37 Safety Injection Lines 1SII3A-6, 1SI13BA-6, and 1SI13BB-6 lSl-1-ISI 38 Safety Injection Lines 1SIO8D-3, 1SIO8B-4, 1SIOBCA-4, and 1SIO8CB-4 1SX-1-ISI 1

Essential Service Water Lines 1SXO6EA-10, 1SXO6CA-14, and 1SXO6BA-16 1SX-1-ISI 2

Essential Service Water Lines 1SXO6DC-10, 1SXO6EC-10, ISXO8AC-10, and 1SXO8BC-1O 1SX-1-ISI 3

Essential Service Water Lines 1SXO6EA-10, 1SXO6FA-10, 1SXOBAA-10, and 1SXD8BA-10 1SX-1-ISI 4

Essential Service Water Lines 1SXO6EB-10, 1SXO6CB-14. and 1SXO6BB-16 1SX-1-ISI 5

Essential Service Water Lines 1SXO6DD-10, 1SXO6ED-10, ISX08AD-10, and 1SXO8BD-1O 15X-1-I5l 6

Essential Service Water Lines 1SXO6EB-10, 1SXO6FB-10, 1SXO8AB-10, and 1SXO8BB-10 1 SX-1 -ISI 7

Essential Service Water Lines 1 SXO7CB-1 0, 1 SXO7EB-1 4, and 1 SXO7FB-1 6 1SX-1-ISI 8

Essential Service Water Lines 1SXO7BB-10, 1SXO7CB-10, ISXO9CB-10, and 1SXO9BB-10 1SX-1-ISI 9

Essential Service Water Lines JSXO7BD-10, 1SXO7CD-10. 1SXO9BD-10, and 1SXO9CD-10 1SX-1-ISl 10 Essential Service Water Lines 1SXO7CA-10, 1SXO7EA-14, and 1SXO7FA-16 JSX-1-IS!

11 Essential Service Water Lines 1SXO7BA-10, 1SX07CA-10, 1SXO9CA-10, and 1SXO9BA-10 1SX-1-IS!

12 Essential Service Water Lines 1SXO7CC-1O, JSXO7BC-10, 1SXO9CC-10, and 1SXO9BC-10 1VCT-1-ISI Containment Spray Pumps 1CS-01-PA-1 and 1CS-01-PB-2 1VQ-1-ISI 1

Primary Containment Purge Lines 1VQO3A-8, 1VQO4A-8, 1VQO5A-8, IVQO1A-48, and 1 VQO2A-48 Exelon

- Byron Station 2-18 Revision 0

IS! Prciram Plan Units I & 2, Fourth Intenval TABLE 2.4-2 UNIT 2 ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DRAWING TITLE NUMBER NUMBER 2AF-1-ISI 1

Auxiliary Feedwater Lines 2AFO2DA-4, 2AFO2DE-4, and 2AFO2EA-4 2AF-1-lSl 2

Auxiliary Feedwater Lines 2FWO6AA-4 and 2FW87BA-3 2AF-1-lSl 3

Auxiliary Feedwater Lines 2AFO2DB-4, 2AF02DF-4, and 2AFO2EB-4 2AF-1-lSI 4

Auxiliary Feedwater Lines 2FWO6AB-4 and 2FW87BB-3 2AF-1-lSI 5

Auxiliary Feedwater Lines 2AFO2DC-4, 2AFO2DG-4, and 2AFO2EC-4 2AF-1 -151 6

- Auxiliary Feedwater Lines 2FWO6AC-4 and 2FW87BC-3 2AF-1-lSl 7

Auxiliary Feedwater Lines 2AFO2DD-4, 2AFO2DH-4, and 2AFO2ED-4 2AF-1-lSl 8

Auxiliary Feedwater Lines 2FWO6AD-4 and 2FW87BD-3 205-1-151 1

Containment Spray Line 2CS02AA-1 0 205-1-151 2

Containment Spray Line 2CS1OAA-6 2CS-1-ISI 3

Containment Spray Lines 2CSO1AA-16 and 2CS23AA-14 2CS-1-ISI 4

Containment Spray Lines 2CSO1AB-16 and 2CS23AB-14 205-1-151 5

Containment Spray Line 2CSO2AB-10 2CS-1-ISI 6

Containment Spray Lines 2CSO2AB-10 and 2CS1OAB-6 2CS-1-lSI 7

Containment Spray Line 2CSO2M-1O 2CS-1-lSI 8

Containment Spray Lines 2CSO6AA-6 and 2CSO6AB-6 2CV-1-ISI 1

Chemical & Volume Control Line 2CVB7A-3 2CV-1-lSI 2

Chemical & Volume Control Lines 2RY18A-2 and 2CV45B-2 2CV-1-lSI 3

Chemical & Volume Control Lines 2CV14FB-2 and 2CV14GB-1%

2CV-1-ISI 4

Chemical & Volume Control Lines 2CVA5AB-2 and 2CVA6AB-2 2CV-1-lSI 5

Chemical & Volume Control Line 2CVA3B-2 2CV-1-ISI 6

Chemical & Volume Control Lines 2CV14FA-2, 2CV14FD-2, and 2CV14GB-1%

2CV-1-ISI 7

Chemical & Volume Control Line 2CVA3B-2 2CV-1-lSI 8

Chemical & Volume Control Line 2CVA5AA-2 2CV-1-ISI 9

Chemical & Volume Control Lines 2CVA3B-2, 2CVA3AB-2, and 2CVA7AB-2 2CV-1-lSI 10 Chemical & Volume Control Line 2CVA3AB-2 2CV-1-ISI 11 Chemical & Volume Control Lines 2CVA3B-2 and 2CVA6M-2 2CV-1-ISI 12 Chemical & Volume Control Line 2CV45B-2 2CV-1-ISI 13 Chemical & Volume Control Line 2CVA3B-2 2CV-1-ISI 14 Chemical & Volume Control Line 2CVA3B-2 2CV-1 -ISI 15 Chemical & Volume Control Line 2CVA3B-2 2CV-1-ISI 16 Chemical & Volume Control Lines 2CV14FC-2 and 2CV14GC-1%

2CV-1-ISI 17 Chemical & Volume Control Lines 2CV99A-8, 2CVO5B-8, and 2CVA1A-6 2CV-1-ISI 18 Chemical & Volume Control Lines 2CV058-8, 2CVO5CA-6, 2CV98BA-8, 2CV98BB-8, and 2CV98BC-8 2CV-1-ISI 19 Chemical & Volume Control Line 2CVC5CB-6 2CV-1-lSl 20 Chemical & Volume Control Lines 2CVO8AB-4, 2CV12AA-3, and 2CV42M-2 2CV-1-lSl 21 Chemical & Volume Control Lines 2CV]4A-4, 2CVO9A-4, and 2CVO8BA-4 2FW-1 -(SI 1

Feedwater Lines 2FWO3DD-16 and 2FW86AD-16 2FW-1 -IS!

2 Feedwater Lines 2FWO3DA-16 and 2FW86M-16 2FW-1 -(SI 3

Feedwater Lines 2FW86A8-16 and 2FWO3DB-16 2FW-1 -(SI 4

Feedwater Lines 2FWO3DC-1 6 and 2FW86AC-16 2FW-1 -(SI 5

Feedwater Lines 2FW81AB-6, 2FW81BB-6, and 2FWB7CB-6 2FW-1 -(SI 6

Feedwater Line 2FW87CB-6 2FW-1-lSl 7

Feedwater Line 2FW87CC-6 2FW-1 -151 8

Feedwater Line 2FW87CD-6 Exelon

- Byron Station 2-19 Revision 0

IS! Program Plan Units 1 & 2, Fourth Interval TABLE 2.4-2 UNIT 2 ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DRAWING TITLE NUMBER NUMBER 2FW-l-ISl 9

Feedwater Line 2FW87CA-6 2FW-1-ISI 10 Feedwater Lines 2FW81AC-6, 2FW81BC-6, and 2FW87CC-6 2FW-1-lSl 11 Feedwater Lines 2FW81AA-6. 2FW81BA-6, and 2FW87CA-6 2FW-1-ISI 12 Feedwater Lines 2FW81AD-6, 2FW81BD-6, and 2FW87CD-6 2MS-1-lSI 1

Main Steam Line 2MSOJ AD-3D 1/4 (Loop 4) 2MS-1-ISI 2

Main Steam Lines 2MSO1 SD-3D 1/4, 2MSO7AD-28, 2MS13AD-8, 2MSO7BD-28,_and_2MS143AD-12_(Loop_4) 2MS-1-ISl 3

Main Steam Line 2MSO1AA-30 1/4 (Loop 1) 2MS-1-ISI 4

Main Steam Lines 2MSO1BA-30 1/4, 2MSO7AA-28, 2MS13AA-8, 2MSO7BA-28,_and_2MS143M-12_(Loop_1) 2MS-1-ISl 5

Main Steam Line 2MSO1AB-32 3/4 (Loop 2) 2MS-1-ISI 6

Main Steam Lines 2MSO1BB-32 3/4, 2MSO7AB-28, 2MS13AB-8, and 2MS143AB-12 (Loop 2) 2MS-1-ISI 7

Main Steam Line 2MSO1AC-32 3/4 (Loop 3) 2MS-7-l51 8

Main Steam Lines 2MSO1BC-32 3/4, 2MSO7AC-28, 2MS13AC-8, and 2MS143AC-12 2RC-1 -ISI 1

Primary Coolant System Loop 1 To Steam Gen. No, 2RC-Q 1-BA 2RC-1 -ISI 2

Primary Coolant System Loop 2 To Steam Gen. No. 2RC-Q 1-SB 2RC-1-ISI 3

Primary Coolant System Loop 3 To Steam Gen. No. 2RC-0 1-BC 2RC-1 -ISI 4

Primary Coolant System Loop 4 To Steam Gen. No. 2RC-01-BD 2RC-1-lSl 5

Reactor Coolant Surge Line 2RYI 1A-14 2RC-1-ISI 6

Reactor Coolant Lines 2RC21AA-8 and 2RC21BA-8 2RC-1 -151 7

Reactor Coolant Lines 2RC28A-3, 2CV1ODA-3, 2RC37A-3, 2CV1ODB-3, and 2RC36A-3 2RC-1-ISI 9

Reactor Coolant Lines 2RC21AB-8 and 2RC2JBB-8 2RC-1-ISI 11 Reactor Coolant Lines 2RCO4AB-1 2 and 2RCO5AB-6; Residual Heat Removal Line 2RHO lAB-12 2RC-1-ISI 12 Reactor Coolant Lines 2RC21AC-8 and 2RC2JBC-8 2RC-1-ISI 14 Reactor Coolant Lines 2RC24AB-4 and 2RYO1AB-4 2RC-1-lSl 15 Reactor Coolant Lines 2RC21AD-8 and 2RC21BD-8 2RC-1-lSl 16 Reactor Coolant Lines 2RYO1B-6° and 2RYO1C-4 2RC-1-lSI 17 Reactor Coolant Lines 2RC24AA-4 2RYO1AA-4, 2RYO1AB-4, and 2RYO1B-6 2RC-1-ISI 19 Reactor Coolant Lines 2RC22AB-1% and 2RC46AB-3 2RC-1-ISI 20 Reactor Coolant Lines 2RC22AD-1% and 2RC46AD-3 2RC-1 -ISI 21 Reactor Coolant Line 2RC22AB-1%

2RC-1 -ISI 22 Reactor Coolant Lines 2RCO5AA-6 (Loop 2) and 2RC35AB-6 (Loop 4) 2RC-1-ISI 23 Reactor Coolant Lines 2RC22AA-11/2 and 2RC46AA-3 2RC-1-ISI 24 Reactor Coolant Lines 2RC22AC-1W and 2RC46AC-3 2RC-1-lSl 27 Reactor Coolant Lines 2RC22M-1% and 2RC22AC-1%

2RC-1-ISI 29 Reactor Coolant Lines 2RCJ6AC-2 (Loop 3) and 2RC16AD-2 (Loop 4) 2RC-1-lSl 30 Reactor Coolant Lines 2RC13AA-2, 2RC13AB-2, 2RC13AC-2, and 2RC 1 3AD-2 2RC-1-ISI 31 Reactor Coolant Lines 2RC14AB-2 (Loop 2) and 2RC26A-2 (Loop 4) 2RC-1 -ISI 32 Reactor Coolant Lines 2RYO3M-6, 2RYO3AB-6, 2RYO3AC-6, 2RYO3BA-6, 2RYO3BB-6, and 2RYO3BC-6 2RC-1-IS(

35 Reactor Coolant Lines 2RYO2A-6, 2RYO6A-3, and 2RYO2B-3 2RC-1-ISI 36 Reactor Coolant Lines 2RC14M-2 and 2CVA3AA-2 2RC-1-lSl 37 Reactor Coolant Lines 2RC14AD-2 and 2CVA7AA-2 2RC-1-ISl 41 Reactor Coolant Lines 2RC16M-2 (Loop 1) and 2RC16AB-2 (Loop 2) 2RC-1-ISl 42 Reactor Coolant Line 2RC14AC-2 Exelon

- Byron Station 2-20 Revision 0

IS! Program Plan Units I & 2, Fourth Interval TABLE 2.4-2 UNIT 2 151 ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET DRAWING TITLE NUMBER NUMBER 2PZR-1-ISI Pressurizer No. 2RY-01-S 2RCP-1-ISI

Reactor Coolant Pumps 2RC-01-PA, 2RC-01-PB, 2RC-01-PC, and 2RC-01-PD 2RPV-1-ISI Reactor Pressure Vessel No. 2RC-01-R 2SG-1-ISI 1

Steam Generator No. 2RC-01-BA 25G-1 -IS!

2 Steam Generator No. 2RC-01-BB 2SG-1-ISI 3

Steam Generator No. 2RC-0 1-BC 2SG-1-ISI 4

Steam Generator No. 2RC-01-BD 2RH-1-ISI 1

Residual Heat Removal Line 2RHO1AB-12 2RH-1-ISI 2

Residual Heat Removal Line 2RHO1M-12 2RH-1-ISI 3

Residual Heat Removal Line 2RHO3AA-8 2RH-1 -IS!

4 Residual Heat Removal Line 2RHO1CA-16 2RH-1 -IS!

5 Residual Heat Removal Lines 2RHO2AA-8 and 2RHO9AA-8 2RH-1 -151 6

Residual Heat Removal Line 2RHO3AB-8 2RH-1 -IS!

7 Residual Heat Removal Lines 2RHO3AB-8, 2RH14A-8, and 2RH03AA-8 2RH-1-ISI 8

Residual Heat Removal Line 2RHO1CB-16 2RH-1-ISI 9

Residual Heat Removal Line 2RHO2AB-8 2RH-1-ISI 10 Residual Heat Removal Lines 2RHO3AA-8 and 2RH12A-8 2RH-1-ISI 11 Residual Heat Removal Line 2RHO1BC-12 and 2S182BB-12 2RH-1-ISI 12 Residual Heat Removal Line 2RHO1BA-12 2RH-1-lSI 13 Residual Heat Removal Line 2RHO2AB-8 and 2RHO9AB-8 2RHP-1-ISI Residual Heat Removal Pumps 2RHO1 PA-i-lA and 2RHO1PB-2-1B 2RHX-1-ISI Residual Heat Exchanger Nos. 2RHO2AA and 2RHO2AB 2SI-1-ISI 1

Safety Injection Lines 2RC29M-1 0 and 2SIO9BA-i0 2SI-i-ISI 2

Safety Injection Lines 251A4B-8, 2SIO3FA-2, 2RCO4AA-12, and 2RC35AA-6 251-1-ISI 3

Safety Injection Line 2SIO5DA-6 251-1-ISI 4

Safety Injection Lines 2SIO5BA-8, 2SIO5CA-8, and 2SIO5CD-8 251-1-151 5

Safety Injection Lines 2RC29AB-1 0 and 2SIO9BB-i0 2S1-1-ISI 6

Safety Injection Lines 2SIO5DB-6 and 2SI1BFB-2 2S1-1-ISI 7

Safety Injection Lines 2SIO8FA-3, 2SIO8FB-3, and 2S108E-3 2S1-1-ISI 8

Safety Injection Line 2SIO8FA-3 2S1-1-ISI 9

Safety Injection Lines 2RC29AC-10 and 2S1098C-10 2SI-1-ISI 10 Safety Injection Lines 2SIO5DC-6 and 2SI18FC-2 2S1-1-ISI 11 Safety Injection Lines 2SIO4D-8 and 2SIO3DB-2 251-1-IS!

12 Safety Injection Lines 2S104A-1 2, 2SIO4B-12, 2S104C-8, and 2SIA4A-8 251-1-IS) 13 Safety Injection Lines 2RC29AD-10 and 2SIO9BD-10 2S1-1-ISI 14 Safety Injection Line 2SIO5DD-6 251-1-IS!

15 Safety Injection Lines 2S)O8JC-1% 2RC45AC-3, and 2RC3OAC-11/2 2S1-1-ISI 16 Safety Injection Lines 2SIO8JD-1%, 2RC45AD-3, and 2RC3OAD-i%

2S1-1-ISI 17 Safety Injection Lines 2SIO8JB-1%, 2RC45AB-3, and 2RC3OAB-11/2 251-1-ISI 18 Safety Injection Lines 2SIO8HB-2, 25108GB-lW, and 2SIO8JB-1%

2S1-1-ISI 19 Safety Injection Lines 2SIO8GA-1W, 2SIO8HA-2, and 2SIO8JA-JW 2S1-1-ISI 20 Safety Injection Lines 2SIO8GC-1W, 2SIO8HC-2, 2SIO8JC-1%, 2SIO8GD-1V2, 2SIO8HD-2, and 2SIO8JD-1 1/20 251-1-151 21 Safety Injection Line 2SIO3DA-2 2S1-i-ISI 22 Safety Injection Line 2SIO3FB-2 2S1-1-ISI 23 Safety Injection Lines 2SI18FA-2 and 2SI18FD-2, and Reactor Coolant Line 2RY76A-2 Exelon

- Byron Station 2-21 Revision 0

IS! Program Plan Units 1 & 2, Fourth litarval TABLE 2.4-2 UNIT 2 ISI ISOMETRIC AND COMPONENT DRAWINGS DRAWING SHEET D AWING T TLE NUMBER NUMBER R

I 281-1-151 24 Safety Injection Lines 2SIO6BA-24 and 2SIO6BB-24 251-i -ISI 25 Safety Injection Line 2SIO5AA-8 251-1-151 26 Safety Injection Lines 2SIO5BB-8, 2SIO5CB-8, and 2SIO5CC-8 251-1-151 27 Safety Injection Line 25108JD-1%

2S1-1 -IS!

29 Safety Injection Line 25108JC-1%

251-i -IS!

31 Safety Injection Lines 2S108]A-1W, 2RC3OAA-1Y2, and 2RC45AA-3 251-1-IS!

32 Safety Injection Line 2SIQ5AB-8 2S1-1 -IS!

33 Safety Injection Line 2S134A-8 251-1-IS!

34 Safety Injection Line 2SIO5CB-8 2SX-1-ISI 1

Essential Service Water Lines 2SXO6BA-16, 2SX06CA-14, 2SX06DC-i0, 2SXO6EA-i 0, 2SXO8M-10, and 2SXO8AC-l0 25X-1-ISI 2

Essential Service Water Lines 2SXO6DC-1O, 2SXO6EC-i0, 2SXO8AC-10, and 2SXO8BC-10 2SX-1-ISI 3

Essential Service Water Lines 2SXO6EA-10, 2SXO6FA-i0, 2SXO8AA-i0, and 2SXO8BA-10 2SX-1-ISI 4

Essential Service Water Lines 2SXO6BB-16, 2SXO6CB-14, 2SXO6EB-10, 2SXO8AB-10, and 2SXO8AD-1 0 2SX-1-ISl 5

Essential Service Water Lines 2SXO6DD-10, 2SXO6ED-10, 2SXO8AD-10, and 2SXO8BD-10 25X-1-ISI 6

Essential Service Water Lines 2SXO6EB-10, 2SXO6FB-10, 2SXO8AB-10, and 2SXO8BB-i0 2SX-1-ISI 7

Essential Service Water Lines 2SXO7CB-10, 2SXO7EB-14, 2SXO7FB-16, 2SXO9CB-i0, and 2SXO9CD-1 0 2SX-1-ISI 8

Essential Service Water Lines 2SX07BB-10, 2SX07C8-10, 2SXO9BB-10, and 2SXO9CB-10 2SX-1-ISI 9

Essential Service Water Lines 2SXO7BD-10, 2SXO7CD-10, 2SXO9BD-10, and 2SXO9CD-i0 2SX-i-!SI 10 Essential Service Water Lines 2SXO7CA-i0, 2SXO7EA-14, and 2SXO7FA-16 2SX-1-!SI 11 Essential Service Water Lines 2SX07BA-10, 2SXO7CA-10, 2SXO9BA-10, and 2SXO9CA-10 25X-1-lSI 12 Essential Service Water Lines 2SXO7BC-10, 2SXO7CC-10, 2SXO9BC-10, and 2SXO9CC-i0 2VCT-1 -IS!

Containment Spray Pumps 2CS-0 1-PA-i and 2CS-01-PB-2 2VQ-1 -IS!

1 Primary Containment Purge Lines 2VQO3A-8, 2VQQ4A-8, 2VQO5A-8, 2VQO1A-48, and 2VQO2A-48 Exelon

- Byron Station 2-22 Revision 0

IS! Program Plan Units I & 2, Fourth Interval TABLE 2.4-3 UNIT I CONTAINMENT ISI DRAWINGS

[

CISI DWG. NO.

CISI DRAWING TITLE 1-CISI-l000 SH.1 WE Component Rollout Inside Containment Liner View Looking Out Q0 To 1800 Azimuth 1-CISI-1000 SN 2 WE Component Rollout Inside Containment Liner View Looking Out 1800 To 3600 Azimuth 1-CISI-l000 SH. 3 IWE Component Drawing Inside Containment Mat Plan View-EL. 377- 0 1-CISI-l000 SH. 4 IWE Component Drawing Containment Dome Liner View Looking UP 1-CISI-1 000 SH. 5 IWE Component Detail Recirc. Sump A & B Guard Pipe & Bellows Assembly 1-CISI-1 000 SH. 7A WE Component Detail Fuel Transfer Tube Pen. (P-98) Reactor Pool Area 1-CISI-1000 SH. 7B WE Component Sections Fuel Transfer Tube Pen. (P-98) Reactor Pool Area 1-CISI-l000 SH. 9A IWE Component Detail Equipment Hatch/Personnel Air Lock 1-CISI-1 000 SH. 9B IWE Component Detail Equipment Hatch/Personnel Air Lock 1-CISI-1000 SH. 9C WE Component Detail Equipment Hatch/Personnel Air Lock 1-CISI-1000 SH. 9D IWE Component Detail Equipment Hatch/Personnel Air Lock 1-CISI-1 000 SH. iDA IWE Component Detail Emergency Personnel Air Lock 1-CISI-1 000 SH. lOB IWE Component Detail Emergency Personnel Air Lock 1-CISI-l000 SH. 1OC WE Component Detail Emergency Personnel Air Lock 1-CISI-1000 SH. 1OD WE Component Detail Emergency Personnel Air Lock 1-CISI-l000 SH. 11 Typical IWE Component Surface and Attachment Details 1-CISI-1000 SN. 12 Typical Penetration Details Inside Containment Configuration Nos 1,2 & 3 1-CISI-1000 SN. 13 Typical Penetration Details Inside Containment Configuration Nos 4 & 5 1-CISI-lOOl, SH. Al SI Identifier Format and Explanation 1-CISI-lOOl SN. 1A THRU iF IWE Component Information Table Piping Penetrations 1-CISI-lOOl SN. 1G THRU 1]

IWE Component Information Table Electrical Penetrations 1-CISI-lOOl SN. 1K IWE Component Information Table Instrument Penetrations 1-CISI-lOOl SN. 1LTHRU lR IWE Component Information Table Miscellaneous Components 1-CISI-lOOl SN 2A Electrical Penetration Details Outside Containment Configuration No. 1 1-CISI-100i SN 2B Electrical Penetration Section Outside Containment Configuration No. 1 1-CISI-lOOl SN 3A Electrical Penetration Details Outside Containment Configuration No. 2 1-CISI-lOOl SN 3B Electrical Penetration Sections Outside Containment Configuration No. 2 1-CISI-lOOl SN 4A Electrical Penetration Details Outside Containment Configuration No. 3 1-CISI-lOOl SN 4B Electrical Penetration Sections Outside Containment Configuration No. 3 1-CISI-lOOl SN 5A Electrical Penetration Details Outside Containment Configuration No. 4 1-CISI-lOOl SN 5B Electrical Penetration Section Outside Containment Configuration No. 4 l-CISI-lOOi SH 6A Electrical Penetration Details Personnel Air Locks Configuration No. 5 1-CISI-lOOl SN 6B Electrical Penetration Section Outside Containment Configuration No. 5 1-CISI-lOol SH. 7 Instrument Penetration Details Outside Containment Configuration Nos 1, 2 & 3 1-CISI-lOol SH. 8 Piping Penetration Details Outside Containment Configuration Nos 1 & 2 l-CISI-1001 SH. 9 Piping Penetration Details Outside Containment Configuration Nos 3 & 4 1-CISI-lOOl SH. 10 Piping Penetration Detail Outside Containment Configuration No. 5 1-CISI-lOOl SH. 11 Piping Penetration Detail Outside Containment Configuration No. 6 1-CISI-1001 SN. 12 Piping Penetration Detail Outside Containment Configuration No. 7 1-CISI-2000 SN. 1 IWL/IWE Component Rollout Outside Containment QO To 1800 Azimuth 1-CISI-2000 SN. 2 IWL/IWE Component Rollout Outside Containment 1800 To 3600 Azimuth 1-CISI-2000 SN. 3 IWL Component Drawing Containment Dome Exterior Plan View 1-CISI-2000 SN. 4 WL Component Drawing Tendon Gallery Plan View 1-CISI-2000 SH. 5 IWL Component Detail Tendon Anchorage Assembly 1-CISI-2000 SH. 6 IWL Component Drawing Dome Tendon Layout Exelon

- Byron Station 2-23 Revision 0

15/ Program Plan Units 7 & 2, Fourth Interval TABLE 2.4-4 UNIT 2 CONTAINMENT ISI DRAWINGS CISI DWG. NO.

CISI DRAWING TITLE 2-CISI-1 000 SH.1 WE Component Rollout Inside Containment Liner View Looking Out Q0 To 1800 Azimuth 2-CIS(-l000 SH 2 WE Component Rollout Inside Containment Liner View Looking Out 1800 To 3600 Azimuth 2-CISI-1 000 SH. 3 IWE Component Drawing Inside Containment Mat Plan View - EL. 377

- 0 2-CISI-1000 SH. 4 WE Component Drawing Containment Dome Liner View Looking UP 2-CISI-1000 SH. 5 IWE Component Detail Reciro. Sump A & B Guard Pipe & Bellows Assembly 2-CISI-1000 SH. 7A IWE Component Detail Fuel Transfer Tube Penetration (P-98) Reactor Pool Area 2-CISI-1000 SN. 78 IWE Component Sections Fuel Transfer Tube Penetration (P-98) Reactor Pool Area 2-CISI-1 000 SH. 9A WE Component Detail Equipment Hatch/Personnel Air Lock 2-CISI-1 000 SH. 9B WE Component Detail Equipment Hatch/Personnel Air Lock 2-CISI-1000 SH. 9C IWE Component Detail Equipment Hatch/Personnel Air Lock 2-CISI-1 000 SH. 9D IWE Component Detail Equipment Hatch/Personnel Air Lock 2-CISI-1 000 SH. iDA IWE Component Detail Emergency Personnel Air Lock 2-ClSl-1000 SH. 108 IWE Component Detail Emergency Personnel Air Lock 2-CISI-i 000 SH. 1OC IWE Component Detail Emergency Personnel Air Lock 2-CISI-l000 SN. 10D WE Component Detail Emergency Personnel Air Lock 2-CISI-1000 SN. 11 Typical IWE Component Surface and Attachment Details 2-CISI-l 000 SN. 12 Typical Penetration Details Inside Containment Configuration Nos 1, 2 & 3 2-CISI-1 000 SH. 13 Typical Penetration Details Inside Containment Configuration Nos 4 & 5 2-CISI-1 001, SH. Al 151 Identifier Format and Explanation 2-CISI-J 001 SH. JA THRU iF WE Component Information Table Piping Penetrations 2-CISI-lOOl SN. 1G THRU 1P IWE Component Information Table Electrical Penetrations 2-CISI-1 001 SN. 1Q IWE Component Information Table Instrument Penetrations 2-CISI-lOOl SN. 1R THRU 1W IWE Component Information Table Miscellaneous Components 2-CISI-lOOl SN 2A Electrical Penetration Details Outside Containment Configuration No. 1 2-CISI-lOOl SH 28 Elect. Penetration Sections Outside Containment Configuration No. 1 2-CISI-lOOl SH 3A Electrical Penetration Details Outside Containment Configuration No. 2 2-CISI-1 001 SH 3B Elect. Penetration Sections Outside Containment Configuration No. 2 2-CISI-1 001 SN 4 Electrical Penetration Details Personnel Air Locks Configuration No. 3 2-CISI-lOOl SN 5A Electrical Penetration Details Outside Containment Configuration No. 4 2-CISI-lOOl SN 58 Elect. Penetration Sections Outside Containment Configuration No. 4 2-CISI-lOOl SN. 6 Instrument Penetration Details Outside Containment Configuration Nos 1,2 & 3 2-CISI-lOOl SN. 7 Piping Penetration Details Outside Containment Configuration Nos 1 & 2 2-CISI-1 001 SH. 8 Piping Penetration Details Outside Containment Configuration Nos 3 & 4 2-CISI-lool SH. 9 Piping Penetration Detail Outside Containment Configuration No.5 2-CISI-lOOl SH. 10 Piping Penetration Detail Outside Containment Configuration No. 6 2-CISI-lOOl SN. 11 Piping Penetration Detail Outside Containment Configuration No. 7 2-CISI-2000 SN. 1 IWUIWE Component Rollout Outside Containment 00 To 1800 Azimuth 2-CISI-2000 SN. 2 IWUIWE Component Rollout Outside Containment 1800 To 3600 Azimuth 2-CISI-2000 SN. 3 IWL Component Drawing Containment Dome Exterior Plan View 2-CISI-2000 SN, 4 IWL Component Drawing Tendon Gallery Plan View 2-CISI-2000 SN. 5 IWL Component Detail Tendon Anchorage Assembly 2-CISI-2000 SN. 6 IWL Component Drawing Dome Tendon Layout Exelon

- Byron Station 2-24 Revision 0

TsIrnPIan Units 7 & 2, Fourth Interval 2.5 TECHNICAL APPROACH AND PosITIONs When the requirements of ASME Section Xl are not easily interpreted, Byron Station has reviewed general licensing/regulatory requirements and industry practice to determine a practical method of implementing ASME Section Xl requirements. The Technical Approach and Position (TAP) documents contained in this section have been provided to clarify Byron Stations implementation of ASME Section Xl requirements. An index which summarizes each TAP is included in Table 2.5-1.

Exelon

- Byron Station 2-25 Revision 0

ISI Program Plan Units 7 & 2, Fourth Interval TABLE 2.5-f TECHNICAL APPROACH AND POSITIONS INDEX POSITION REVISION!

STATUS (PROGRAM) DESCRIPTION OF TECHNICAL APPROACH AND POSITION NUMBER DATE2 0

Active (151) RI-ISI Examination Volumes and Methods.

14T-01 7/29/16 141 02 0

(151) Determination of Additional Examinations per ASME Code Case Active 7/29/16 N-578-1, Paragraph -2430.

0 Active (SPI) System Leakage Testing of Non-Isolable Buried Components.

14T-03 7/29/16 0

Active (SPT) Valve Seats/Discs as Pressurization Boundaries.

14T-04 7/29/16 14T05 0

Active (151) Unit 1 Steam Generator Lower Shell-to-Transition Cone Welds.

7/29/16 14T 06 0

fiSt) Preservice (PSI) Requirements Under the RI-ISI Program for Butt Active 7/29/16 and Branch Connection Welds.

141 07 0

(ISI) Implementation of ASME Code Case N-706-1 for RHR Heat Active 7/29/16 Exchangers.

0 (151) Repair Requirements for 1St Class 1 Repairs in Piping > 3/8 141-08 Active 7/29/16 Nominal Pipe Size and Tubing Size> 1/2 in Diameter.

0 (1Sf) Repair/Replacement Requirements for New ISI Class 2 and 3 141-09 Active 7/29/16 Branch Connections 1 and Smaller.

0 (ISI) Examination Requirements for ASME Section Xl, IWA-5244, 141-10 Active 7/29/16 Buried Components.

Note 1 1St Program Technical Approach and Position Status Options: Active

- Current Technical Approach and Position is being utilized at Byron Station; Deleted

- Technical Approach and Position is no longer being utilized at Byron Station Note 2: The revision listed is the latest revision of the subject Technical Approach and Position. The date noted in the second column is the date of the ISI Program Plan revision when the Technical Approach and Position was incorporated into the document.

Exelon

- Byron Station 2-26 Revision 0

IS! Program Plan Units 1 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 141-01 Revision 0 Component Identification Code Class:

1 and 2

Reference:

Byron Station Request for Relief 4R-O1, Alternative to the ASME Section XI Requirements for Class I and Class 2 Piping Welds Executive Summary, Risk Informed Inservice Inspection Program Plan Byron Nuclear Power Station Units I and 2 ASME Code Case N-578-1: Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B Section Xl, Division I Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev.

B-A, Revised Risk-Informed Inseivice Inspection Evaluation Procedure Examination Category:

Previously B-F, B-J, C-F-i, and C-F-2 now incorporated into R-A

==

Description:==

Rl-ISI Examination Volumes and Methods Code Requirement The requirements for examination methods and areas/volumes are assembled from several sources other than the stations base edition of the ASME Code.

Relief Request l4R-O1:

For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI Topical Report while the guidance for the examination method is provided by ASME Code Case N-578-1 (N-578-i).

Executive Summary, Section 3.5 Inspection Location Selection and NDE Selection:

N-578-1 Table 1, Examination Category R-A, Risk-Informed Piping Examinations will also be used in conjunction with Table 4-1 of EPRI TR-i12657 to categorize the parts examined under the RI-ISI Program. N-578-l Table 1 provides examination requirements, examination method, acceptance standards, examination extent and frequency for piping structural elements not subject to a damage mechanism.

N-578-1, Section 1-5.2 Examination Volumes and Methods:

Examination programs developed in accordance with this Case shall use NDE techniques suitable for specific degradation mechanisms and examination locations. The examination volumes and methods that are appropriate for each degradation mechanism are provided in Table 1 of this Case. The methods and procedures used for the examinations shall be qualified to reliably detect and size the relevant degradation mechanisms identified for each elements.

TR-112657, Section 4 Mechanism Specific Examination Volumes and Methods:

Application of RI-ISI uses NDE techniques that are designed to be effective for specific degradation mechanisms and examination locations. This inspection for cause approach involves identification of specific damage mechanisms that are likely to be operative, the location where they may be operative, and the appropriate examination methods and volumes specific to address the damage mechanism.

Exelon

- Byron Station 2-27 Revision 0

ISI Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-Of Revision 0 Position Table 14T-O1-1: Degradation Mechanisms with Examination Methods and Volumes DEGRADATION MECHANISM (DM)

N-578-1 TABLE I TR-112657 TABLE 4-1 COMMENTS OR COMPONENT TYPE EXAM METHOD EXAM VOLUME OR AREA Includes expanded examination Thermal Fatigue Volumetric Figure 4-1 thru 4-4 5

volume for piping. See Note High Cycle Mechanical Visual VT-2 Not Applicable Note1 None currently identified at station.

Fatigue Erosion Cavitation Volumetric Figure 4-16 thru 4-22 None currently identified at station.

Crevice Corrosion Cracking Volumetric Figure 4-6 and 4-7 None currently identified at station.

Primary Water Stress See Note2 See Note2 See Note2 Corrosion Cracking Effected components not subject to lntergranular or Transgranular an additional DM. Only SCC type Stress Corrosion Cracking Volumetric Figure 4-10 thru 4-14 examinations requited for components.

Microbiologically Corrosion Volumetric Figure 4-15 See Note3 Flow Accelerated Corrosion Volumetric Figure 4-16 thru 4-22 In accordance w/ FAC Program External Chloride Stress Surface Affected Surface None currently identified at station.

Corrosion Cracking Figure 4-1 IWB-2500-8(c)

Includes expanded examination No Damage Mechanism Volumetric IWB-2500-9, 10, 1 1 4,5,6 IWC-2500-7(a) volume for piping. See Notes See Notes456 Socket Welds (All DM)

Visual, VT-2 Not Applicable, Note1 See Note1 Note 1:

VT-2 visual examinations are performed during each refueling cycle. V-2 visual examination area is not identified in N-578-1 orTR-112657 (TR-Rl-ISI). Socket welds are not specifically addressed in TR-Rl-lSl with the exception of FAC examinations. N-578-1 Table 1 Note 12 specifies that socket welds require only a VT-2 visual examination.

Note 2:

N-578-1 requires a Vf-2 visual examination for this DM while TR-1 12657 requires a volumetric or visual method. Recent industry events necessitated the change to volumetric examination techniques (where qualified examination techniques are available) for detection prior to through-wall leakage. TR-Rl-lSl identifies Figures 4-8 and 4-9 for the required examination volumes based on component configuration. Figure 4-8 would not be applicable to components incorporated into Rl-lSl. At Byron Station, all components previously subject to PWSCC (8 in each unit) ate now classified DM None and as Medium-Risk Group, Risk Category 4.

Note 3:

DM currently limited to SX system components. These components have been removed from the RI-IS I inspection population and default by incorporation into the Service Water Inspection program.

Exelon

- Byron Station 2-28 Revision 0

ISI Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 141-01 Revision 0 Note 4:

Examination of components without an identified DM is not addressed in TR-Rl-ISI. N-578-1 requires that these components receive the same examination as components subject to thermal fatigue. For no-DM components, the examination requirements of N-578-1 will be used.

Note 5:

For piping butt welds with no DM, the length for the examination volume shall be increased to include % beyond each side of the detectable base metal thickness transition or counterbore. For components without a detectable base metal thickness transition or counterbore, the basic examination volume specified in TR-Rl-)Sl Figure 4-1 shall be used.

The figure applicable for use shall be based on the detectable presence of a counterbore regardless of the pipe size.

Note 6:

For branch connection piping without a DM, the examination volume shall be determined using the figures specified in N-578-1 (Figures IWB-2500-9, 10, 1 1 of the 2007 Edition with the 2008 Addenda).

Exelon

- Byron Station 2-29 Revision 0

IS! Program Plan Units 1 & 2, Fotirth Intenial TECHNICAL APPROACH AND POSITION: 14T-02 Revision 0 Component Identification Code Class:

1, 2, and 3

Reference:

Byron Station Request for Relief 14R-01 ASME Code Case N-578-1: Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B Section Xl, Division 1 Examination Category:

Previously B-F, B-], C-F-i, and C-F-2 now incorporated into R-A

==

Description:==

Determination of Additional Examinations per ASME Code Case N-578-1, Paragraph -2430 Code Requirement

-2430 Additional Examinations (a) Examinations performed in accordance with -2500 that reveal flaws or relevant conditions exceeding the acceptance standards of -3000 shall be extended to include additional examinations.

The additional examinations shall include piping structural elements described in Table 1 with the same postulated failure mode and the same or higher failure potential.

(1) The number of additional elements shall be the number of piping structural elements with the same postulated failure mode originally scheduled for that fuel cycle.

(2) The scope of the additional examinations may be limited to those High-Safety-Significant (HSS) piping structural elements within systems, whose materials and service conditions are determined by an evaluation to have the same postulated failure mode as the piping structural element that contained the original flaw or relevant condition.

(b) lithe additional examinations required by -2430(a) reveal flaws or relevant conditions exceeding the acceptance standards of -3000, the examination shall be further extended to include additional examinations.

(1) These examinations shall include all remaining piping elements within Table 1 whose postulated failure modes are the same as the piping structural elements originally examined in -2430(a)

(2) An evaluation shall be performed to establish when those examinations are to be conducted.

The evaluation must consider failure mode and potential.

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

Underlined portions of the requirements of the code case identify issues addressed in this technical approach.

TAP 14T-02 Basis Sections of N-578-1 A. -9000 Glossary failure mode - a condition or degradation mechanism that can cause a failure B.

Figure -1 Risk Evaluation Process Failure Mode Assessment:

Design & Operating Conditions Service Experience Degradation Mechanisms C. Appendix I: Requirements for Risk-Informed Selection Process Exelon

- Byron Station 2-30 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 141-02 Revision 0 1-3.1 Failure Potential Assessment 1-3.1.1 Identification of Degradation Mechanisms. Potential active degradation mechanisms for each pipe segment within the selected system boundaries shall be identified. The following conditions shall be considered.

(a) Design characteristics, including material, pipe size and schedule, component type (e.g., fitting type or ANSI standard) and other attributes related to the system configuration.

(b) Fabrication practices, including welding and heat treatment.

(c) Operating conditions, including temperatures and pressures, fluid conditions (e.g., stagnant, laminar flow, and turbulent flow), fluid quality (e.g., primary water, raw water, dry steam, and chemical control), and service environment (e.g., humidity and radiation).

(U) Industry-wide service experience with the systems being evaluated.

(e) Results of preservice, inservice, and augmented examinations, and the presence of prior repairs in the system.

(1 Deqradation mechanisms identified in Table I-i.

Underlined portions of the requirements of the code case identify issues addressed in this technical approach.

Exelon

- Byron Station 2-31 Revision 0

IS! Program Plan Units I & 2, Fotiflh Interval TECHNICAL APPROACH AND POSITION: 141-02 Revision 0 Position

-2430 (a)(1) Additional Examination Selection Criteria by Failure Mode of Additional Examinations The criteria of additional selection are based upon the failure mode of the initially rejected element. The following aspects may restrict the potential population of the additional elements.

Failure Mode Evaluation (determining element attributes)

(a) Design characteristics:

material pipe size and schedule component type (joint configuration)

(b) Operating conditions:

temperature pressure fluid quality (c) Degradation mechanisms:

shown in N-578-1 Table I-I This evaluation may be performed prior to the outage in which the initial rejection occurs. The station population of RI-ISI elements may be organized into predetermined groups.

-2430 (a)(2) High-Safety-Significance and Failure Potential of Additional Examinations Additional selections are not restricted by the Risk Category of the rejected element. High-Safety-Significant piping structural elements are identified as those components included in Risk Categories 1, 2, 3, 4, and 5. The additional examinations include elements with the same or higher failure potential.

Because consequence is not considered, selections along the horizontal axis are not restricted by N-578-1. The Failure Potential of the rejected element restricts selections along the vertical axis to the same or higher position.

Table 14T-02-1: Unit I System Distribution in N-578-1 Risk Matrix Categories Subject to Examination1 N-578-1 CONSEQUENCE CATEGORY TABLE 1-8 LOW MEDIUM HIGH CATEGORY 5(H)

CATEGORY 3 CATEGORY 1 w

HIGH None FW None D

MEDIUM CATEGORY 5(M)

CATEGORY 2 AF, CV, SI RC2

°-

LOW CATEGORY 4 CS, CV, RH, RC2, SI LRISKGROUPS L MEDIUM-CAT4&5 HIGH-CATJ,2,&3 NOTES:

(1)

Table does not include elements subsumed into other station programs. Table does include BER elements incorporated into the RI-ISI Program.

(2)

The RC System includes both the RC and RY System elements.

Exelon

- Byron Station 2-32 Revision 0

IS! Program Plan Units I & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-02 Revision 0 Table 14T-02-2: Unit 2 System Distribution in N-578-1 Risk Matrix Categories Subject to Examination1 N-578-1 CONSEQUENCE CATEGORY TABLE 1-8 LOW MEDIUM I

HIGH CATEGORY 5(H)

CATEGORY 3 CATEGORY 1 HIGH None FW None Z

MEDIUM CATEGORY 5(M)

CATEGORY 2 AF,CV,SI RC2 uO°-

LOW CS, CV, MS, RC2, RH, SI R[SKGROUPS MEDIUM-CAT4&5 I

HIGH-CATJ,2,&3 NOTES:

(1)

Table does not include elements subsumed into other station programs. Table does include BER elements incorporated into the RI-ISI Program.

(2)

The RC System includes both the RC and RY System elements.

In addition to the restrictions identified in -2430(a)(1), the potential population for additional examinations is limited by the following factors from the N-578-1 Table 1-8 Risk Matrix for Failure Potential.

CATEGORY 3: Selections remain within Category 3.

CATEGORY 5: Selections may be taken from Categories 3, 5(M) and 2.

CATEGORY 2: Selections may be taken from Categories 3, 5(M) and 2.

CATEGORY 4: Selections may be taken from Categories 3, 5(M), 2 and 4.

-2430(b)(1): Second Expansion Scope of Additional Examinations The second expansion scope includes the remaining elements in the original group determined under the -2430(a) criteria.

-2430(b)(2): Scheduling of the Second Expansion Scope Per the response to the second l4R-01 RAI, the second expansion selections will be examined in the current refueling cycle.

-2430(c): Return to Original Schedule of Component Selection and Examination In the initial expansion population, credit may be taken for examinations performed on components scheduled later in the same Inspection Period (i.e., the initial expansion may include components scheduled for the next refueling outage). The scheduling of components with other postulated failure modes are not affected by the additional examination scope(s).

Exelon

- Byron Station 2-33 Revision 0

lSI Program Plan Units I & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-03 Revision 0 Component Identification Code Class:

2 and 3

Reference:

IWA-5244(b)(2)

Examination Category:

C-H, D-B Item Number:

07.10, D2.10

==

Description:==

System Leakage Testing of Non-Isolable Buried Components Component Number:

Non-Isolable Buried Pressure Retaining Components Code Requirement IWA-5244(b)(2) requires non-isolable buried components be tested to confirm that flow during operation is not impaired.

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

For open ended lines on systems that require Inservice Testing (1ST) or performance testing of pumps, adherence to 1ST or performance testing acceptance criteria is considered as reasonable proof of adequate flow through the lines.

For lines in which the open end is accessible to visual examination while the system is in operation, visual evidence of flow discharging the line is considered as reasonable proof of adequate flow through the open ended line.

For open ended portions of systems where the process fluid is pneumatic, evidence of gaseous discharge shall be considered reasonable proof of adequate flow through the open ended line. Such test may include passing smoke through the line, hanging balloons or streamers, using a remotely operated blimp, using thermography to detect hot air, etc.

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

Byron Stations position is that proof of adequate flow is all that is required for testing the buried pipe segments of these open ended lines and that no further visual examination is necessary. This is consistent with the requirements for buried piping, which is not subject to visual examination.

Exelon

- Byron Station 2-34 Revision 0

151 Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-04 Revision 0 Component Identification Code Class:

1, 2, and 3

Reference:

IWA-522 1 IWA-5222 Examination Category:

B-P, C-H, D-B Item Number:

B15.1O, B15.20, C7.1O, D2.1O

==

Description:==

Valve Seats/Discs as Pressurization Boundaries Component Number:

All Pressure Testing Boundary Valves Code Requirement IWA-5221 requires the pressurization boundary for system leakage testing extend to those pressure retaining components under operating pressures during normal system service.

Position Byron Stations position is that the test pressurization boundary extends up to the valve seat/disc of the valve utilized for isolation. For example, in order to pressure test the ISI Class 1 components, the valve that provides the Class break would be utilized as the isolation point. In this case the true pressurization boundary, and Class break, is actually at the valve seat/disc.

Any requirement to test beyond the valve seat/disc is dependent only on whether or not the piping on the other side of the valve seat/disc is 151 Class 1, 2, or 3.

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

Exelon

- Byron Station 2-35 Revision 0

IS! Program Plan Units 7 & 2, Fotir[h Interval TECHNICAL APPROACH AND POSITION: 14T-05 Revision 0 Component Identification Code Class:

2

Reference:

IWB-2500-1 Examination Category:

C-A Item Number:

C1.1O

==

Description:==

Unit I Steam Generator Lower Shell-to-Transition Cone Welds Component Number:

1RC-OI-BA, SGW-05, 1RC-O1-BB, SGW-05, 1RC-O1-BC, SGW-05 and IRC-O1-BD, SGW-05 Code Requirement Table IWC-2500-1 Examination Category C-A, Pressure Retaining Welds in Pressure Vessels, Item Number C1.1O, Shell Circumferential Welds (Pressure Vessels) requires a volumetric examination of vessel cylindrical-shell-to-conical-shell-junction welds and shell (or head)-to-flange welds.

Position The weld configuration show in Figure IWC-2500-1 (c) was typically used in older steam generator designs.

The weld joining the lower shell to the transition cone is positioned at the junction where the vessel diameter begins to increase towards the diameter of the larger steam drum. The joint configuration of the SGW-05 welds is of the type normally found at intermediate shell joints that consist of butted, parallel plates. Intermediate shell welds do not require examination. In previous ISI Intervals, the intermediate shell welds were exempted from examination due to the weld joint being categorized as a non-gross structural discontinuity location. Steam generator welds where the joints are shell-to-tube sheet, shell-to-head, and shell-to-transition cone (per Figure IWC-2500-1 (c)) were categorized as being located at gross structural discontinuities and required examination.

For the Second and Third ISI Intervals, Byron Station exempted the SGW-05 weld based on an evaluation of the weld location. Babcock & Wilcox, Canada (BWI) Engineering Evaluation CM9015189 - B2 Exelon Generation Company RSG

- Shell Circumferential Weld Evaluation with Respect to Section XI IS!

Rules was used to determine the classification of SGW-05. The evaluation determined that this weld should be not classified as being located at a gross structural discontinuity as defined in ASME Section III, NB-3213.2 and therefore is removed from the Item Number C1.1O population. EGC Owners Review of this evaluation was performed under EC 354211 Owner Review of B&W CALC # CM9075189-82 RSG Shell Circumferential Weld Evaluation with Respect to Section Xl IS! Rules. This evaluation showed the stresses of SGW-05 resemble the intermediate shell weld locations and not the location of the base of the transition cone.

Since the joint configuration is a butted-plate type rather than the configuration seen in IWC-2500-1 (c) and the stresses evaluated in CM9015189-B2 show the location is like those found at remote intermediate shell welds, this weld will be exempted from examination in this ISI Interval by removal from the C1.1O population.

Exelon

- Byron Station 2-36 Revision 0

IS! Program Plan Units I & 2. Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-06 Revision 0 Component Identification Code Class:

1, 2, and 3

Reference:

Byron Station Relief Request l4R-02 ASME Code Case N-578-1: Risk-Informed Requirements for Class 1, 2, or3 Piping, Method B Section Xl, Division I Examination Category:

Previously B-F, B-J, C-F-i, and C-F-2 now incorporated into R-A

==

Description:==

Preservice Inspection (PSI) Requirements Under the Rl-lSl Program for Buff and Branch Connection Welds Code Requirement IWB-2200 PRESERVICE EXAMINATION (a) Examinations required by this Article (with the exception of Examination Category B-F, and the VT-3 visual examination of the internal surfaces of Examination Categories B-L-2 and B-M-2, of Table IWB-2500-1) shall be completed prior to initial plant startup.

In addition, these preservice examinations shall be extended to include essentially 100% of the pressure retaining welds in all Class I components, except in those components exempted from examination by IWB-1220(a), (b), or (c).

However, in the case of Examination Category B-O (Table IWB-2500-1), the examination shall be extended to include essentially 100% of the welds in the installed peripheral control rod drive housings only.

(b) Shop and field examinations may serve in lieu of the on-site preservice examinations provided:

(2) such examinations ate conducted under conditions and with equipment and techniques equivalent to those that are expected to be employed for subsequent inservice examinations; IWC-2200 PRESERVICE EXAMINATION (a) All examinations required by this Article (with the exception of Examination Category C-H of Table IWC-2500-1) for those components initially selected for examination in accordance with Inspection Program and not exempt from inservice examinations by IWC-1 220 shall be completed prior to initial plant startup.

(b) Shop and field examinations may serve in lieu of the on-site preservice examinations, provided:

(2) such examinations are conducted under conditions and with equipment and techniques equivalent to those which are expected to be employed for subsequent inservice examinations; IWD-2200 PRESERVICE EXAMINATION All examinations required by this Article (with the exception of Examination Category D-B of Table IWD-2500-1) shall be performed completely, once, as a preservice examination requirement prior to initial plant startup.

ASME CODE CASE N-578-1 (N-578-1)

PSI issue is not addressed.

EPRI TR-lf 2657 PSI issue is not addressed.

RI-ISI EXECUTIVE

SUMMARY

PSI issue is not addressed.

Exelon

- Byron Station 2-37 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-06 Revision 0 Position The replacement of Table IWB-2500-1 (Examination Categories B-F and B-J) and Table IWC-2500-i (Examination Categories C-F-i and C-F-2) by the Rl-lSl Program does not include the PSI examination requirements. The requirements for PSI are now determined by the requirements specified by the RI-I SI Program for lSl examinations. Since surface examinations of welds are excluded from Rl-ISI, no PSI surface examination is required. The following guidance meets the intent of ISI Class 1, 2, 3, and Non-Class components and incorporates the examination aspects of the RI-ISI Program.

151 Class I RI-ISI Categories 1 through 5:

RI-ISI Item Number requirements for 100% of welds. Examinations that result in limitations of 10% or more will require a relief request.

151 Class 1 RI-ISI Categories 6 and 7:

RI-ISI required examination volume for 100% of welds.

If no item number is assigned to the weld, then the examination requirements of Item Number Ri.11 will be used (based on possible change in risk category at later date). Examinations that result in limitations of 10% or more will require a relief request.

IWC-1220 Non-Exempt ISI Class 2 RI-ISI Categories I through 5:

Required per RI-ISI Item Number for 100% of welds. Examinations on those components selected for examination by RI-ISI that result in limitations of 10% or more will require a relief request.

Examinations on those components not selected for examination by RI-ISI that result in limitations of 10% or more will not require a relief request.

IWC-1 220 Non-Exempt ISI Class 2 RI-ISI Categories 6 and 7:

Required per RI-ISI Item Number for 100% of welds.

151 Class 2 does not require 100% examination for PSI. However, augmented examinations may be required if applicable. The recommendation is to perform the PSI examination due to possible changes in the initially assigned examination categories.

If no item number is assigned to the weld, then the examination requirements of Item Number R1.11 will be used (based on possible change in risk category at later date).

IWC-1220 Exempt 161 Class 2 RI-ISI Categories 1 through 7:

Not required.

151 Class 2 does not require 100% examination for PSI. However, augmented examinations may be required if applicable.

ISI Class 3 or Non-Class RI-ISI Categories 1 through 5:

Required per Rl-ISI Item Number for 100% of welds per corporate mandate.

151 Class 3 or Non-Class RI-ISI Categories 6 and 7:

Not required.

ISI Class 3 or Non-Class does not require surface or volumetric examination for PSI or 151. However, augmented examinations may be required if applicable.

Exelon

- Byron Station 2-38 Revision 0

IS! Program Plan Units 7 & 2, Fotiflh Interval TECHNICAL APPROACH AND POSITION: 14T-07 Revision 0 Component Identification Code Class:

2

Reference:

ASME Code Case N-706-1: Alternative Examination Requirements of Table IWB-2500-1 and Table IWC-2500-1 for PWR Stainless Steel Residual and Regenerative Heat Exchangers Section Xl, Division I Examination Category:

C-A and C-B

==

Description:==

Implementation of ASME Code Case N-706-i for RHR Heat Exchangers Code Requirement The 151 Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section Xl, 2007 Edition with the 2008 Addenda. Volumetric and surface examinations are specified for these components per IWC-2500 for Examination Categories C-A and C-B.

The applicable Residual Heat Exchanger welds are classified per Table IWC-2500-1. These welds are shown below.

Table 14T-07-i: RHR Heat Exchanger Welds VESSEL WELD DESCRIPTION CATEGORY ITEM NO.

EXAMINATION J

COMMENTS RHEC-Ol Shell to Flange C-A C1.10 Volumetric Note (1)

RHEC-02 Shell to Head C-A Cl.20 Volumetric RHXN-01 Inlet Nozzle C-B C2.21 Volumetric & Surface Note (1)

RHXN-0i-NIR Nozzle Inner Radius C-B C2.22 Volumetric Note (2)

RHXN-02 Outlet Nozzle C-B C2.21 Volumetric & Surface Note (1)

RHXN-02-NIR Nozzle Inner Radius C-B C2.22 Volumetric Note (2)

NOTES:

(1)

Limited volumetric (UT) examination due to component configuration. Single-sided scanning access only.

(2)

Configuration prevents examination to Item Number C2.22 requirements. The nozzle inner radius section as depicted in the IWC-2500 examination figures is not present in the Byron Station RHR Heat Exchanger design.

Position ASME Code Case N-706-1 (N-706-1) lnqufty: What alternative to the requirements of Table IWB-2500-1, Examination Categories B-B, B-D, and B-J, and Table IWC-2500-i, Examination Categories C-A, C-B, and C-F-i, may be used for PWR stainless steel regenerative and residual heat exchangers?

Reply: It is the opinion of the Committee that the requirements of Table 1 may be used for PWR stainless steel regenerative and residual heat exchangers, in lieu of the requirements of Table IWB-2500-i, Examination Categories B-B, B-D, and B-J, and Table IWC-2500-1, Examination Categories C-A, C-B, and C-F-i.

This Code Case may not be applied to any heat exchanger nor to any heat exchanger design or configuration that has experienced a through-wall leak, such as heat exchangers with an inner shell (inner barrel). The Owner shall evaluate industry experience to determine which heat exchanger designs or configurations have leaked.

If any leakage is detected, it shall be corrected in accordance with IWA-4000 or (IWA-7000 prior to the 1991 Addenda). Use of this Code Case shall be discontinued Exelon

- Byron Station 2-39 Revision 0

IS! Program Pthn Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-07 Revision 0 for that heat exchanger design or configuration. The affected heat exchanger and others of the same design or configuration shall be examined in accordance with IWB-2500 or IWC-2500, as applicable.

Table 14T-07-2: N-706-1 Table 1 and Byron Station Applicability EXTENT AND FREQUENCY OF EXAMINATION z

ITEM PARTS EXAMINED NO.

[NOTES (1) (2)]

o FIRST SUCCESSIVE Z

z INSPECTION INSPECTION

,< w INTERVAL INTERVALS w x o w

<co 1.10 Residual and regenerative heat exchangers 1.11 Not Applicable to Byron Station 1.12 Not Applicable to Byron Station 1.13 Not Applicable to Byron Station 1.14 Examination Category C-A welds VT-2 IWC-All welds Same as for Not permissible

[Note (4)]

3516 1st interval 1.15 Examination Category C-B welds VT-2 IWC-All welds Same as for Not permissible

[Note (4)]

3516 1st interval 1.16 Not Applicable to Byron Station Specified Conditions for Use of N-706-1 a.

This Code Case may not be applied to any heat exchanger nor to any heat exchanger design or configuration that has experienced a through-wall leak, such as heat exchangers with an inner shell (inner barrel). The Owner shall evaluate industry experience to determine which heat exchanger designs or configurations have leaked.

The industry experience was evaluated in the following documents.

1.

Westinghouse Owners Group (WOG) project MUHP 5093, Working Group Inservice Inspection Optimization Action 97-01, lSl-03-06, BCO3-338, Technical Basis for Revision of Inspection Requirements for Regenerative and Residual Heat Exchangers, August, 2003.

2.

Pacific Northwest National Laboratory PVP2005-71633 Assessment of ASME Code Examinations on Regenerative, Letdown, and Residual Heat Removal Heat Exchangers.

These studies did not identify leakage events for the Byron Station RHR Heat Exchanger design.

PRESSURE RETAINING WELDS IN PWR STAINLESS STEEL RESIDUAL AND REGENERATIVE HEAT EXCHANGERS NOTES:

(1)

Application of the requirements of this table is limited to those welds that are part of the as-received heat exchanger assembly. The regenerative heat exchanger assembly may be formed from multiple smaller heat exchanger subcomponents connected by sections of piping. All of the smaller heat exchanger subcomponents and the connecting piping are within the boundary of the heat exchanger assembly.

(3)

Not Applicable to Byron Station.

(2)

All welds, other than reinforcing plate welds, shati have received at least one volumetric examination; the preservice or Construction Code volumetric examination may be used to meet this requirement.

Reinforcing plate welds shall have received at least one surface examination. This does not apply to nozzle inside-radius sections.

(4)

Component shall be examined for evidence of leakage while undergoing the system leakage test (IWC-5220) as required by Examination Category C-H, to be performed every inspection period.

Exelon

- Byron Station 2-40 Revision 0

151 Program Plan Units 1 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 141-07 Revson 0 b.

Note (2): AU welds shall have received at least one volumetric examination; the preservice or Construction Code volumetric examination may be used to meet this requirement.

Table 14T-07-3; Most Recent RHR Heat Exchanger Volumetric Examination EXAMNATON COMPONENT STAGE DATE METHOD REPORT #

Vessel: IRH-02-AA RHEC-Ol Fabrication 11/26/1975 Radiographic Job# J-2267-1A RHEC-02 Fabrication 1 1/26/1 975 Radiographic Job# J-2267-1A RHXN-01 B1R08 IS!

12/23/1997 Ultrasonic 97BY1-UTD-148 RHXN-02 B1RO5 ISI 02/03/1 993 Ultrasonic 93BY1-UT-029 Vessel: I RH-02-AB RHEC-Ol B1R14 ISI 08/29/2006 Ultrasonic B1R14-UT-010 RHEC-02 B1RJ4ISI 08/29/2006 Ultrasonic B1R14-UT-011 RHXN-01 B1RO5 ISI 02/02/1 993 Ultrasonic 938Y1-UT-026 RHXN-02 B1R05 1St 02/02/1993 Ultrasonic 93BY1-UT-027 Vessel: 2RH-02-AA RHEC-Ol B2R12 IS!

]

09/22/2005 Ultrasonic B2R12-UT-008 RHEC-02 B2R12 IS!

09/22/2005 Ultrasonic B2R12-UT-008 RHXN-01 B2R05 151 02/20/2095 Ultrasonic 95BY2-UTD-088 RHXN-02 B2R09 ISI 04/04/2001 Ultrasonic B2R09-UT-081 Vessel: 2RH-02-AB RHEC-Ol

[ Preservice 07/26

- 07/31/1986 Ultrasonic UT-A-Ui, UT-C-Ui RHEC-02 Preservice 08/04/1986 Ultrasonic UT-A/C-02 RHXN-01

[2R05 IS!

02/28/1995 Ultrasonic 95BY2-UTD-090 RHXN-02

[ B2R05 ISI 02/28/1995 Ultrasonic 955Y2-UTD-091 c.

Note (4): Component shall be examined for evidence of leakage while undergoing the system leakage test (IWC-5220) as required by Examination Category C-H, to be performed every inspection period.

Table l4T-07-4: System Pressure Test Implementation PRESSURE TEST EXAMINA11ON INFORMATION Vessel P&ID TEST BLOCK PMID 1RH-02-M M-62 RH-2-J 122551 Performed with 1RHO1PA pump run.

1RH-02-AB M-62 RH-2-2 122552 Performed with 1RHO1PB pump run.

2RH-02-M M-137 RH-2-1 124070 Performed with 2RHO1PA pump run.

2RH-02-AB M-137 RH-2-2 124071 Performed with 2RHO1PB pump run.

Exelon

- Byron Station 2-41 Revision 0

IS! Program Plan Units 1 & 2, Fotirth Interval TECHNICAL APPROACH AND POSITION: 14T-08 Revision 0 Component Identification Code Class:

I

References:

IWA-41 31 (a)(2)

Examination Category:

NA Item Number:

NA

==

Description:==

Repair Requirements for SI Class I Repairs in Piping > 3/8 Nominal Pipe Size and Tubing Size> 1/2 in Diameter Component Number:

Not Applicable Code Requirement IWA-4131.1(a)(2) requires that the size and design such that, in the event of postulated failure during normal plant operating conditions, the reactor can be shut down and cooled in an orderly manner, assuming makeup is provided by normal reactor coolant makeup systems operable from on-site emergency power.

Position ISI Class 1 Repairs and Replacements on NPS 3/8 (3/8T nominal pipe size) and less diameter, and tubing W diameter and less may apply the small items alternative requirements of IWA-4131.1(a)(2).

Reference Byron/Braidwood UFSAR Section 3.9.1.1 Exelon

- Byron Station 2-42 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: 14T-09 Revision 0 Component Identification Code Class:

2 and 3

References:

IWA-4000 ASME Section Xl Code Interpretation Examination Category:

NA Item Number:

NA

==

Description:==

Repair/Replacement Requirements for New ISP Class 2 and 3 Branch Connections 1 and Smaller Component Number:

N/A Code Requirement IWA-4131 provides for alternate Repair/Replacement requirements for IS) Class 2 and 3 piping less than or equal to 1 inch. The alternate requirements remove the need for a repair/replacement plan, NIS-2A, and system pressure test (VT-2 visual examination).

Position The installation of branch connections 1 inch and smaller onto any other pipe diameter in a IS) Class 2 or 3 system is governed by alternate requirements of IWA-41 31. This includes the original pipe (greater than 1 inch diameter) to branch connection weld.

Exelon

- Byron Station 2-43 Revision 0

IS! Program Plan Units I & 2, Eocidh Interval TECHNICAL APPROACH AND POSITION: l4T-f 0 Revision 0 Component Identification Code Class:

3

Reference:

IWA-5244 ASME Interpretation Xl-1-07-37 Examination Category:

D-B Item Number:

D2.10

==

Description:==

Alternative Examination Requirements of ASME Section Xl, IWA-5244, Buried Components Component Number:

Supply Lines: OSX01AA-48, 0SX01AB-48, 0SXA8AA-3/4 OSXA8AB-3I4 1SXOJ BA-36 1 SXO 1 BB-36, 2SX0 1 BA-36, 2SX0 1 BB-36, Return Lines: OSXO3CA-48, OSX03CB-48, OSX79AA-6, 0SX79AB-6 0SX97AA-24 OSX97AB-24 05X97AC-24 OSX97AD-24 OSX97AE-24 OSX97AF-24 0SX97AG-24 OSX97AH-24 OSX98AA-24 0SX98AB-24, OSX98AC-24, and 0SX98AD-24 Drawing Number:

M-42-1A, M-42-1B, M-42-2A, M-42-2B, M-42-6, and M-42-7 Code Requirement IWA-5244(b)(1) requires buried components that are isolable by means of valves be tested to determine the rate of pressure loss. Alternatively, the test may determine the change in flow between the ends of the buried components and the Owner shall establish the acceptable rate of pressure loss or flow.

ASME Interpretation Xl-1-07-37, approved by the NRC in October, 2008, states the following

- Is it the intent of IWA-5244(b)(1) that the configuration of isolable by means of valves applies to buried components with butterfly valves that are not designed to be leak tight? No The justification for ASME Interpretation Xl-1-07-37, also supported a ASME Section Xl, 2009 Addenda Code revision to IWA-5244(b)(1) and (2) for buried components not isolable by valves that are required to be essentially leak tight.

Position The buried piping in question consists of two 48 common (i.e., Unit 0) supply headers and two 48 Unit 0 return headers between the Essential Service Water Cooling Towers (SXCT) and the Auxiliary Building. Each 48 supply header with 314 sampling lines, branches into two 36 pump supply lines (i.e., each unit). Each 48 return header branches into four 24 risers and two 24 hot water bypass lines, and has a 6 blowdown line connection. Both 48 return headers and each of the 24 risers have a line-stop fitting that was previously installed for maintenance of the system. These components are all buried between the SXCT and the Turbine Building (TB) or encased in the TB foundation. There is no access to the buried sections without excavation. In addition, no annulus was provided during original construction that would allow for examination of these buried sections of piping.

IWA-5244(b)(1) requires a test that isolates the buried sections of piping to conduct a pressure decay test or to perform a test that determines the change in flow between the buried ends.

In order to perform a pressure decay test, it would be necessary to close three large butterfly valves1 to isolate the buried portion of each supply header. For the return header piping, it would be necessary to close 1 A-Train: OSX138A, 1SXOOJA, and 2SXOO1A.

B-Train: OSXI38B, 1SXOO1B, and 2SXOO1B.

Exelon

- Byron Station 2-44 Revision 0

ISI Program Plan Units I & 2, Fourth Interval TECHNICAL APPROACH AND POSITION: I4T-1O Revision 0 several large butterfly valves2 to isolate the buried portions. This would also result in the isolation of an entire return train of SX, which is a configuration not allowed by the Byron Station Technical Specifications (IS). These butterfly valves on both the supply and return headers are not designed or required to be leak tight.

Consistent with ASME Interpretation Xl-1-07-37 and the latest editions and addenda of ASME Section Xl, for the buried piping sections required to provide flow from the SXCT through the 48 headers and each of the 36 supply lines to the SX Pumps (i.e., Supply Headers) and the buried piping sections required to return flow from the SX System through the 48 headers to the 24 branch lines into the SXCT (i.e., Return Header), a test will be conducted to confirm unimpaired flow in accordance with IWA-5244(b)(2). These requirements call for a test that confirms flow is unimpaired in nonisolable buried components. To confirm that flow is unimpaired in these buried pipes, Byron Station 1ST Program will be used to ensure adequate flow. Byron Station will use the Owner established minimum flow rate specified in the site 151 surveillances, currently specified at 24,000 gallons per minute (gpm) for all SX pumps, as the acceptance criteria for IWA-5244 pressure testing of SX System buried piping.

2 A-Train: 1SXO1O, 2SX010, 1SXO15A, 2SXO15A, 1SXO57A, 2SX057A, 1SXJ14A, 2SX114A, 1SX147A, 2SX147A, OSX162A, OSX162C, OSX163A, 0SX163B, 0SX163C, and OSX163D.

B-Train: 1SXO15B, 2SXO15B, 1SX057B, 2SX057B, 1SX1J4B, 2SX1148, 1SX136, 2SX136, 2SX1478, ISXJ47B, 0SX1628, OSXI62D, OSX163E, 0SX163F, OSX163G, and OSX163H.

Exelon

- Byron Station 2-45 Revision 0

ISI PrO cvin P1(11?

Units I & 2. Fourth Interval 3.0 COMPONENT IS) PLAN The Byron Station Component 151 Plan includes ASME Section Xl nonexempt pressure retaining welds, piping structural elements, pressure retaining bolting, welded attachments, pump casings, valve bodies, reactor vessel interior, reactor vessel interior attachments, reactor vessel removable core support structures, and steam generator tubing of IS) Class 1, 2, and 3 components that meet the criteria of Subarticle IWA-1300. These components are identified on the ISI CBDs listed in Section 2.3, Tables 2.3-1 and 2.3-2. Procedure ER-AA 330-002, Inservice Inspection of Section Xl Welds and Components, implements the ASME Section Xl Component 151 Plan. This Component 151 Plan also includes augmented examination program requirements specified by documents other than ASME Section Xl as referenced in Section 2.2 of this document.

3.1 NONEXEMPT ISI CLASS COMPONENTS The Byron Station IS) Class 1, 2, and 3 nonexempt components subject to examination identified on ISI CBDs are those which are not exempted under the criteria of Paragraphs IWB-1220, IWC-1220, and IWD-1220 of ASME Section Xl, respectively. A summary of Byron Station ASME Section Xl nonexempt components is included in Section 7.0.

The process for scoping Byron Station components for inclusion in the Component ISI Plan is included in the applicable sections of the ISI Classification Basis Document.

3.1.1 Identification of ISI Class 1, 2, and 3 Nonexempt Components ISI Class 1, 2, and 3 nonexempt components are identified on the ISI Isometric and Component Drawings listed in Section 2.4, Tables 2.4-1 and 2.4-2. Welded attachments are also identified by controlled individual support detail drawings.

3.1.2 Components Exempt From Examination Certain components or parts of components may be exempted from examination based on design and accessibility per the requirements of Paragraphs IWB-1220, IWC-1220, and IWD-J 220.

The process for exempting Byron Station components from the Component ISI Plan per Paragraphs IWB-1 220, IWC-1 220, and IWD-1 220 is included in the applicable sections of the IS) Classification Basis Document. These sections include discussions of exempt components and the bases for those exemptions.

3.2 RISK-INFORMED EXAMINATION REQUIREMENTS The Rl-lSl Program element examinations are performed in accordance with Relief Request l4R-01.

3.2.1 Piping structural elements that fall under RI-ISI Examination Category R-A are risk ranked as High (1, 2, and 3), Medium (4 and 5), and Low (6 and 7). Per the EPRI Topical Reports TR 112657, Rev. B-A, TR-1 006937, Rev. 0-A, and N-578-1, piping structural elements ranked as High or Medium Risk are subject to examination while piping structural elements ranked as Low Risk are not subject to examinations (except for pressure testing). Thin-wall welds that were excluded from volumetric examination under ASME Section Xl rules per Table IWC-2500-1 are included in the element scope that is potentially subject to RI-IS) examination at Byron Station.

3.2.2 Piping structural elements may be excluded from examination (other than pressure testing) under the RI-IS) Program if the only degradation mechanism present for a given location is inspected for cause under certain other Byron Station programs such as the Flow Accelerated Corrosion (FAC) or Microbiologically Induced Corrosion (MIC) Programs. These Exelon Byron Station 3-1 Revision 0

181 Pioiain P1cm Units 1 & 2. fourth Intcrvcil piping structural elements will remain part of the assigned programs that already perform tot cause inspections to detect these degradation mechanisms. Piping structural elements susceptible to FAC or MIC and pitting along with another degradation mechanism (e.g.,

thermal fatigue) are retained as part of the Rl-lSl scope and are included in the element selection for the purpose of performing examinations to detect the additional degradation mechanism.

3.2.3 Weld Locations with Full-Structural Overlays a.

Alloy 600/82/182 locations with applied full-structural weld overlays where the degradation mechanism assessment of the overlaid weld identified PWSCC, or PWSCC and another degradation mechanism as determined by the RI-lSl Program, will be removed from the Rl-ISI Program and administered solely under the Byron Station Alloy 600 Augmented Examination Program. These locations will receive examinations as specified under N-770-x separate from the RI-ISI Program in order to maintain compliance with 10 CFR 50.55a(g)(6)(ii)(F).

b.

Non Alloy 600/82/1 82 locations with applied full-structural weld overlays will be removed from the RI-ISI Program and treated solely under the requirements of the ASME Section Xl, 2007 Edition with the 2008 Addenda, Nonmandatory Appendix Q.

3.2.4 Weld Locations Mitigated with Mechanical Stress Improvement Process For Alloy 600/82/1 82 locations where the PWSCC degradation mechanism has been mitigated with a Mechanical Stress Improvement Process, the elements will remain in the RI-ISI Program and are subject to the normal RI-ISI element selection process. These welds will also be governed by the Byron Station Alloy 600 Augmented Examination Program under N-770-x. The selection and examination of these welds will comply with both Rl-ISl and the N-770-x requirements. The examinations in the Fourth Inspection Interval may be credited to both programs.

3.3.

Weld Numbering 3.3.1.

Upon request for a repair/modification weld number, obtain the applicable information (line number and location) from the requestor.

3.3.2.

Original welds are nominally numbered sequentially in the direction of flow. The numbering scheme shall be followed as closely as possible for welds added to a line.

3.3.3.

When an existing weld is repaired/modified, the new weld will be determined by using the original weld number followed by a decimal point and a two-digit numeral. The first two digit numeral assigned will be 01.

If the same weld requires additional repair later, the two-digit numeral shall be incremented by one to reflect the repair.

3.3.4.

When a repair/modification requires a new weld to be added to a line, the new weld number will be determined as follows:

1.

Locate the nearest upstream weld (NUW) from the new weld.

2.

Add an alphabetic designator to the original weld number of the NUW. The first alphabetic designator will be A, and sequential lettering will be used for multiple new welds. For example, two welds added between C2.01 and C3 will be identified as C2A and C2B.

3.

If weld which has no ISI weld number is to be incorporated into the ISI program (i.e.

lines previously exempt from 131), the welds in that line will be sequentially numbered beginning with 1.

Exelon Byron Station 3-2 Revision 0

IS? Program Plan Units 7 & 2, Fotiflh Interval 4.0 SUPPORT ISI PLAN The Byron Station Support 151 Plan includes the supports of ASME Section Xl nonexempt lSl Class 1, 2, 3, and MC components as described in Section 3.0. Procedure ER-AA-330-003, Inservice Inspection of Section Xl Component Supports, implements the ASME Section Xl Support lSI Plan. (CM-3) 4.1 NONEXEMPT ISI CLASS SUPPORTS The Byron Station ISI Class 1, 2, 3, and MC nonexempt supports are those which do not meet the exemption criteria of Paragraph IWF-1230 of ASME Section Xl. A summary of the ASME Section XI nonexempt supports is included in Section 7.0. (CM-3) 4.1.1 Identification of 151 Class 1, 2, 3, and MC Nonexempt Supports ISI Class 1, 2, 3, and MC nonexempt supports are identified on the ISI Isometrics and Component Drawings listed in Section 2.4, Tables 2.4-1 and 2.4-2. Supports are also identified by controlled individual support detail drawings. (CM-3) 4.2 SNUBBER EXAMINATION AND TESTING REQUIREMENTS 4.2.1 As allowed by 10 CFR 50.55a(b)(3)(v)(B), Byron Station will use Subsection ISTD, Inservice Testing of Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants, ASME Operation and Maintenance of Nuclear Power Plants Code (ASME OM Code), 2004 Edition through the 2006 Addenda, to meet the visual examination, functional testing, and service life monitoring requirements for safety related and important to safety snubbers. This approach is consistent with ASME Section Xl, Paragraph IWF-1220, which excludes inservice inspection of snubbers and defers to the ASME OM Code for visual examination (ISTD-4000), functional testing (ISTD-5000), and service life monitoring (ISTD-6000) requirements. For a detailed discussion of the ASME OM Code Snubber Program, see the Snubber Program Document.

4.2.2 The ASME Section XI visual examination boundary of a support containing a snubber is defined in Figure IWF-1300-1(f). This boundary does not include the snubber pin-to-pin and does not include the connections to the snubber assembly (pins) per Paragraph IWF-1 300(h).

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

It should be noted that the examination of snubber welded attachments will be performed in accordance with the ASME Section XI Subsections IWB, IWC, and IWD welded attachment examination requirements (e.g., Examination Categories B-K, C-C, and D-A).

4.3 HIGH STRENGTH BOLTS - CLASS 1 COMPONENT SUPPORTS - (CM-3) 4.3.1 In order to meet commitments associated with License Renewal, high strength bolts (ASME SA 540 and ASTM A490 materials), greater than 1 in diameter, used on the ISI Class 1 component supports, steam generator, reactor coolant pump, and pressurizer supports, require an Owner augmented, periodic visual examination as follows:

1.

Perform periodic visual examinations to detect a corrosive environment that supports SCC potential for all (100%) of high strength bolting greater than one-inch nominal Exelon Byron Station 4-1 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval diameter used for ISI Class I component supports, prior to the period of extended operation, and then each inspection interval of 10 years thereafter.

2.

These examinations of high strength bolting, greater than one-inch nominal diameter used for ISI Class 1 component supports, shall not be considered as completed examination credit for the current inspection period.

4.3.2 The periodic visual examinations will include criteria to identify if the bolting has been exposed to moisture or other contaminants by evidence of moisture, residue, foreign substance, or corrosion.

4.3.3 Adverse conditions identified during the examinations should be evaluated by engineering to determine if the bolt has been exposed to a corrosive environment with the potential to cause Stress Corrosion Cracking (SCC).

The additional parameters and criteria to address the qualitative and quantitative acceptance criteria are described below:

A.

If moisture is present at or near a bolt or stud, factors considered by engineering include, but will not be limited to:

1.

The source of leakage or condensation that supplied the moisture.

2.

The proximity of the moisture to the bolt or stud.

3.

The probable or analyzed chemical characteristics of the moisture, including the presence of contaminants.

4.

The visible or likely pathway, if any, that the liquid traversed to arrive at or near the bolt or stud.

5.

The amount of any corrosion on or near the bolt or stud.

6.

The material condition of the coatings on the bolt or stud, and associated support.

7.

The characteristics of any corrosion on or near the bolt or stud.

8.

The proximity to the bolt or stud of any nearby evidence of corrosion.

B. The material condition of accessible concrete or grout near the bolt or stud.

If there is evidence that moisture had been present at or near a bolt or stud, but no moisture is present at or near a bolt or stud, factors considered by engineering include, but will not be limited to:

1.

The probable sources of past leakage or condensation that could have supplied the moisture.

2.

The proximity to the bolt or stud to the evidence that moisture had been present.

3.

The probable or analyzed chemical characteristics of any moisture residue, including the presence of contaminants.

4.

The visible or likely pathway, if any, that the liquid may have traversed to arrive at or near the bolt or stud.

5.

The amount of any corrosion on or near the bolt or stud.

6.

The material condition of any coatings on the bolt or stud, and associated support.

7.

The characteristics of any corrosion on or near the bolt or stud.

Exelon Byron Station 4-2 Revision 0

ISI Program Plan Units I & 2, Fourth Interval 8.

The proximity to the bolt or stud of any nearby evidence of corrosion.

9.

The material condition of concrete or grout near the bolt or stud.

2.

The extent to which each of the above environmental indicators will be considered and weighed in the engineering evaluation will be determined by the conditions that are observed during the initial visual examinations of the bolting locations and during any follow up visual examination or analysis. Some of the listed environmental indicators may not be present, e.g., moisture. Some of the factors that are observed may have minimal impact on the outcome of the evaluation. Environmental indicators, which are present at each evaluated high strength bolt, will be evaluated together to provide the most accurate characterization of the environment, lithe engineering evaluation concludes that the bolting material had been subjected to an environment with the potential to cause SCC, then the affected bolts will be included in the sample population subject to supplemental volumetric examinations.

4.3.4 The bolts determined to have been exposed to corrosive environment with the potential to cause SCC should be included in a sample population for each specific bolt material where SOC is a concern. A sample size equal to 20 percent (rounded up to the neatest whole number) of the bolts in the sample population, with a maximum sample size of 25 bolts will be subject to supplemental volumetric examination to determine if SOC is present. The selection of the samples will consider susceptibility to stress corrosion cracking (e.g., actual measured yield strength) and ALARA principles.

4.3.5 Volumetric examinations should be performed in accordance with the requirements of ASME Section Xl, Appendix VIII, Supplement 8. The results of the volumetric examinations should be evaluated by engineering to determine if additional actions are warranted such as expansion of sample size, scope, and frequency of any additional supplemental visual or volumetric examinations, as well as any ASME Section XI requirements.

Exelon Byron Station 4-3 Revision 0

IS! Program Plan Units I & 2, Fourth Interval 5.0 SYSTEM PRESSURE TESTING 151 PLAN The Byron Station System Pressure Testing ISI Plan includes pressure retaining ASME Section Xl, ISI Class 1, 2, and 3 components, with the exception of those specifically exempted by Paragraphs IWA-5110(c), IWC-5222(b), and IWD-5222(c). RI-ISI piping structural elements, regardless of risk classification, remain subject to pressure testing as part of the current ASME Section XI program.

The SPT SI Plan performs system pressure tests and required VT-2 visual examinations on the ISI Class 1, 2, and 3 pressure retaining components to verify system and component structural integrity. This program conducts both Periodic and Interval (JO-year frequency) pressure tests as defined in ASME Section XI Inspection Program. Procedure ER-AA-330-001,Section XI Pressure Testing, as well as Byron Station site-specific test procedures, implement the ASME Section XI System Pressure Testing ISI Plan. This System Pressure Testing ISI Plan also includes augmented examination program requirements specified by documents other than ASME Section XI as referenced in Section 2.2 of this document.

5.1 lSI CLASS SYSTEMS All ISI Class 1 pressure retaining components, typically defined as the reactor coolant pressure boundary, are required to be tested. Those portions of ISI Class 2 and 3 systems that are required to be tested include the pressure retaining boundaries of components required to operate or support the system safety functions.

ISI Class 2 open ended discharge piping and components are excluded from the examination requirements per Paragraph IWC-5222(b). ISI Class 3 open ended discharge piping and components are subject to the examination requirements per Paragraph IWD-5222(b). Also, Paragraph IWA 5244 defines buried component pressure testing methods. ASME Interpretation XI-1-10-06 clarifies that Paragraph IWA-5244 only applies to buried components that fall within the boundaries of Paragraphs IWC-5222 and IWD-5222, and thus buried component testing is not required for open ended discharge piping when the buried section is beyond the last shutoff valve.

5.1.1 Identification of ISI Class 1, 2, and 3 Components Components subject to ASME Section Xl System Pressure Testing and augmented pressure testing are shown on the color coded ISI CBDs listed in Section 2.3, Tables 2.3-1 and 2.3-2.

Additional information on the classification of various system boundaries is provided in the ISI Classification Basis Document.

5.1.2 Identification of System Pressure Tests The System Pressure Test Boundary Drawings used to define which systems, or portions of systems, fall under a specific test are also shown on the color coded ISI CBDs listed in Section 2.3, Tables 2.3-1 and 2.3-2. Individual tests are identified and maintained in the Byron Station ISI Database.

5.2 RISK-INFORMED EXAMINATIONS OF SOCKET WELDS Socket welds selected for examination under the RI-ISI Program are to be inspected with a VT-2 visual examination each refueling outage per N-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socket welds selected for inspection under the RI-ISI Program are pressurized each refueling outage during a system pressure test in accordance with Paragraph IWA-521 1(a).

Exelon Byron Station 5-1 Revision 0

IS! Program Plan Units 7 & 2, Fotiflh Interval 6.0 CONTAINMENT ISI PLAN The Byron Station Containment 151 Plan includes ASME Section Xl 151 Class MC pressure retaining components and their integral attachments (including the SI Class CC metal liner),

and (SI Class CC components and structures, and post-tensioning systems that meet the criteria of Subarticle IWA-1300. This Containment 181 Plan also includes information related to augmented examination areas, component accessibility, and examination review.

The inspection of containment structures, components, and post-tensioning systems are performed per procedures ER-AA-330-005, Visual Examination of Section XI Class CC Concrete Containment Structures, ER-AA-330-006, Inservice Inspection and Testing of the Pre-Stressed Concrete Containment Post Tensioning Systems, and ER-AA-330-007, Visual Examination of Section XI Class MC Surfaces and Class CC Liners. In addition, vendor procedures are used to complete more complex surveillances such as tendon testing.

6.1 NONEXEMPT ISI CLASS COMPONENTS The Byron Station lSI Class MC and CC components identified on the Containment 151 Drawings are those not exempted under the criteria of Paragraphs WE-I 220 and IWL-1 220 in the 2007 Edition with the 2008 Addenda of ASME Section XI. A summary of Byron Station ASME Section XI nonexempt CISI components is included in Section 7.0.

The process for scoping Byron Station components for inclusion in the Containment ISI Plan is included in the containment sections of the 181 Classification Basis Document. These sections include a listing and detailed basis for inclusion of containment components.

Components that are classified as ISI Class MC and CC must meet the requirements of ASME Section XI in accordance with 10 CFR 50.55a(g)(4).

ISI Class MC supports of Subsection (WE components are not required to be examined in accordance with 10 CFR 50.55a(g)(4)(v).

6.1.1 Identification of 181 Class MC and CC Nonexempt Components ISI Class MC and CC components are identified on the Containment ISI Drawings listed in Section 2.4, Tables 2.4-3 and 2.4-4.

6.1.2 Identification of ISI Class MC and CC Exempt Components Certain containment components or parts of components may be exempted from examination based on design and accessibility per the requirements of Paragraphs IWE-1220 and IWL-1 220.

The process for exempting Byron Station components from the Containment ISI Plan per Paragraphs IWE-1220 and IWL-1220 is included in the containment sections of the ISI Classification Basis Document. These sections include discussions of exempt components and the bases for those exemptions.

6.2 AUGMENTED EXAMINATIONS AREAS The containment sections of the 181 Classification Basis Document discuss the containment design and components. Metal containment surface areas subject to accelerated degradation and aging require augmented examination per Examination Category E-C and Paragraph WE-i 240.

Similarly, concrete surfaces may be subject to Detailed Visual examination in accordance with Paragraph IWL-2310(b), if declared to be Suspect Areas.

A significant condition is a condition that is identified as requiring application of additional augmented examination requirements under Paragraphs IWE-1 240 or IWL-2310.

Exelon Byron Station 6-1 Revision 0

IS! Program Plan Units I & 2, Fourth Interval No significant conditions were identified in the First CISI Interval; however, significant conditions were identified in the Second CISI Interval as requiring application of additional augmented examination requirements under Paragraph IWE-2420 or IWL-2310.

In the Second CISI Interval, metal loss in excess of 10% of the Byron Station Unit 1 containment liner at two locations below the moisture barrier at elevation 377 have been identified as augmented surface areas requiring successive examinations in accordance with Paragraph IWE-2420(b). The indications were evaluated and accepted but re-examination of these two locations in the next inspection period (First Period, Third CISI Interval) is required.

These surface areas have been categorized in accordance with Table IWE-2500-J, Examination Category E-C, Item Number E4.11, requiring detailed visual examinations (i.e.,

VT-i) of 100% of the identified surface area each inspection period until the areas examined remain essentially unchanged for the next inspection period. When/If such areas no longer require augmented examination in accordance with Paragraph IWE-2420(d), the examination requirements and associated extent and frequency of Examination Category E-A apply for the remainder of the interval.

6.3 COMPONENT ACCESSIBILITY ISI Class MC and CC components subject to examination shall remain accessible for either direct or remote visual examination from at least one side per the requirements of ASME Section Xl, Paragraph IWE-1230.

Paragraph IWE-1231(a)(3) requires 80% of the pressure-retaining boundary that was accessible after construction to remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant.

Byron Station Calculation BYR2000-181 addresses compliance with this requirement by calculating the containment pressure boundary surface area that was accessible for examination at the beginning of the CISI Program and determining the limit for surface area which may be made inaccessible for the balance of plant life.

Portions of components embedded in concrete or otherwise made inaccessible during construction are exempted from examination, provided that the requirements of ASME Section XI, Paragraph IWE-1232 have been fully satisfied.

In addition, inaccessible surface areas exempted from examination include those surface areas where visual access by line of sight with adequate lighting from permanent vantage points is obstructed by permanent plant structures, equipment, or components; provided these surface areas do not require examination in accordance with the inspection plan, or augmented examination in accordance with Paragraph IWE-1240.

6.4 RESPONSIBLE INDIVIDUAL AND ENGINEER ASME Section XI Subsection IWE requires the Responsible Individual to be involved in the development, performance, and review of the CISI examinations. The Responsible Individual shall meet the requirements of ASME Section Xl, Paragraph IWE-2320.

ASME Section Xl Subsection IWL requires the Responsible Engineer to be involved in the development, approval, and review of the CISI examinations. The Responsible Engineer shall meet the requirements of ASME Section Xl, Paragraph IWL-2330.

Exelon Byron Station 6-2 Revision 0

IS! Program Plan Units I & 2, Fotidh Interval 7.0 COMPONENT

SUMMARY

TABLES 7.1 INSERVICE INSPECTION

SUMMARY

TABLES Tables 7.1-1 and 7.1-2 provide a summary of the ASME Section XI pressure retaining components, supports, containment structures, metal liners, post-tensioning systems, system pressure testing, and augmented examination program components for the Fourth 151 Interval and the Third CISI Interval at Byron Station.

If a particular Examination Category and Item Number do not apply to Byron Station, they are not included in these tables.

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

EXAMINATION ITEM NUMBER (OR CATEGORY (WITH RISK CATEGORY TOTAL NUMBER OF RELIEF REQUEST!

EXAMINATION NUMBER OR DESCRIPTION EXAM COMPONENTS BY NOTES REQUIREMENTS TAP NUMBER CATEGORY AUGMENTED SYSTEM DESCRIPTION)

NUMBER)

(1)

(2)

(3)

(4)

(5)

(6)

(7)

(1)

Examination Category (with Examination Category Description):

Provides the Examination Category and description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, IWF-2500-i, and IWL-2500-1. Only those examination categories applicable to Byron Station are identified.

Examination Category R-A from N-578-1 is used in lieu of ASME Section XI Examination Categories B-F, B-J, C-F-i, and C-F-2 to identify ISI Class 1 and 2 piping structural elements for the Rl-lSl program.

Examination Category NA is used to identify Augmented Examination Programs and other Byron Station requirements.

(2)

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

Provides the Item Number as identified in ASME Section XI, Tables IWB-2500-i, IWC-2500-i, IWD-2500-i, IWE-2500-i, IWF-2500-i, and IWL-2500-i. Only those Item Numbers applicable to Byron Station are identified.

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

Specific abbreviations such as RG1.14, ECCS, 0737, GL8805, APP Q, MRP-i46, MRP-i92, MRP-227, N-722-i, N-729-i, and N-770-i are used to identify Augmented Examination Programs and other Byron Station requirements.

(3)

Item Number (or Risk Category Number or Augmented Number)

Description:

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

For Risk-Informed piping structural elements, a description of the Risk Category Number is provided.

For Augmented Examination Programs, a description of the augmented requirement is provided.

Exelon Byron Station 7-1 Revision 0

iS! Pro ciram Plan Units I & 2, Fourth Interval (4)

Examination Requirements:

Provides the examination methods required by ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, IWF-2500-1, and IWL-2500-l.

Provides the examination requirements for piping structural elements under the RI-ISI Program that are in accordance with the EPRI Topical Reports TR-1 12657, Rev. B-A, TR-1006937, Rev. 0-A, and N-578-1.

Provides the examination requirements for Augmented Examination Program components.

The examination requirements described in this document are reflected in the ISI Database code pages.

(5)

Total Number of Components by System:

Provides the system designator (abbreviations). See Section 2.3, Tables 2.3-1 and 2.3-2 for a list of these systems.

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

Note that the total number of components by system are subject to change after completion of plant modifications, design changes, and 151 system classification updates and will be maintained within the ISI Database.

(6)

Relief Request/Technical Approach and Position Number:

Provides a listing of Relief Request! TAP Numbers applicable to specific components, the ASME Section Xl Item Number, Risk Category Number, or Augmented Number.

Relief Requests and TAP Numbers that generically apply to all components, or an entire class are not listed.

If a Relief Request! TAP Number is identified, see the corresponding relief request in Section 8.0 or the TAP Number in Section 2.5.

(7)

Notes:

Provides a listing of program notes applicable to the ASME Section Xl Item Number, Risk Category Number, or Augmented Number.

If a program note number is identified, see the corresponding program note in Table 7.1-3.

Exelon Byron Station 7-2 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Unit I & Common Inservice Inspection Summary Table 7.1-1 EXAMINATION ITEM c

p o EXAM TOTAL NUMBER RELIEF REQ.I NOTES CATEGORY NUMBER DES RI TI N

REQUIREMENTS BY SYSTEM TAP NUMBER B-A Bill Circumferential Shell Welds (Reactor Vessel)

Volumetric RPV: 3 Pressure Retaining 61.21 Circumferential Head Welds (Reactor Vessel)

Volumetric RPV: 2 Welds in Reactor 61.30 Shell-to-Flange Weld (Reactor Vessel)

Volumetric RPV: 1 Vessel 61.40 Head-to-Flange Weld (Reactor Vessel)

Volumetric &

RPV: 1 25 Surface B-B B2.l 1 Circumferential Shell-To-Head Welds (Pressurizer)

Volumetric PZR: 2 Pressure Retaining Welds in Vessels 62.12 Longitudinal Shell-To-Head Welds (Pressurizer)

Volumetric PZR: 2 Other Than Reactor 62.40 Tube Sheet-To-Head Weld (Steam Generator)

Volumetric SG: 4 Vessels B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel)

Volumetric RPV: 8 11 Full Penetration 63.100 Nozzle Inside Radius Section (Reactor Vessel)

See Note RPV: 8 12 Welds of Nozzles in 63.110 Nozzle-to-Vessel Welds (Pressurizer)

Volumetric PZR: 6 Vessels 63.1 20 Nozzle Inside Radius Section (Pressurizer)

See Note PZR: 6 14 63.140 Nozzle Inside Radius Section (Steam Generator)

See Note SG: 8 14 Exelon Byron Station 7-3 Revision 0

IS! Program Plan Units I

& 2, Fourth Interval Unit I & Common Inservice Inspection Summary Table 7.1-1 EXAMINATION ITEM DESCRIPTION EXAM rT0TAL NUMBER RELIEF REQ.!

N CATEGORY NUMBER REQUIREMENTS [

BY SYSTEM TAP NUMBER OTES B-G-1 B610 Closure Head Nuts (Reactor Vessel)

Visual, VT-i RPV: 3 3

Pressure Retaining B6.20 Closure Studs (Reactor Vessel)

Volumetric RPV: 3 3

Bolting, Greater Than 4

2 In Diameter 86.40 Threads in Flange (Reactor Vessel)

Volumetric RPV: 1 3

86.50 Closure Washers, Bushings (Reactor Vessel)

Visual, VT-I RPV: 3 3

86.90 Bolts & Studs (Steam Generator)

Volumetric SG: 8 3

86.100 Flange Surface, When Connection Disassembled (Steam Generator)

Visual, VT-i SG: 8 3

86.1 10 Nuts, Bushings, and Washers (Steam Generator)

Visual, VT-I SG: 8 3

86.180 Bolts & Studs (Pumps)

Volumetric RC: 4 3

34 86.190 Flange Surface, When Connection Disassembled (Pumps)

Visual, VT-i RC: 4 3

34 86.200 Nuts, Bushings, and Washers (Pumps)

Visual, VT-i RC: 1 3

21 34 86.210 Bolts & Studs (Valves)

Volumetric RC: 8 3

34 86.220 Flange Surface, When Connection Disassembled (Valves)

Visual, VT-i RC: 8 3

34 B6.230 Nuts, Bushings, and Washers (Valves)

Visual, VT-i RC: 8 3

34 Exelon Byron Station 7-4 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Unit I & Common Inservice Inspection Summary Table 7:1-1 EXAMINATION ITEM DESCRIPTION EXAM TOTALNUMBER RELIEFREQ.I CATEGORY NUMBER REQUIREMENTS BY SYSTEM TAP NUMBER I

OT b

B-G-2 B7.10 Bolts, Studs, & Nuts (Reactor Vessel)

Visual, VT-i RPV: 1 3

Pressure Retaining B7.20 Bolts, Studs, & Nuts (Pressurizer)

Visual, \\IT-J PZR: 1 3

Bolting, 2 and Less B7.50 Bolts, Studs, & Nuts (Piping)

Visual, VT-i CV: 4 3

In Diameter RC: 4 RY: 4 SI:_8 B7.60 Bolts, Studs, & Nuts (Pumps)

Visual, VT-i RC: 4 3

35 B7.70 Bolts, Studs, & Nuts (Valves)

Visual, VT-i RC: 4 3

RH:4 35 SI:_18 B-K BlOb Welded Attachments (Pressure Vessels)

Surface or PZR: 2 15 Welded Attachments Volumetric for Vessels, Piping, Bi0.20 Welded Attachments (Piping)

Surface CV: 1 16 Pumps, and Valves SI: 6 B-L-2 B 12.20 Pump Casing (Pumps)

Visual, VT-3 RC: 4 34 Pump Casings 35 B-M-2 Bi2.50 Valve Body, Exceeding NPS 4 (Valves)

Visual, VT-3 RC: 12 34 Valve Bodies RH: 4 35 RY: 3 SI:_18 Exelon Byron Station 7-5 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Unit I & Common Inservice Inspection Summary Table 7.7-i EXAMINATION ITEM D

CR TI EXAM TOTAL NUMBERI RELIEF REQ.!

CATEGORY NUMBER ES IP ON REQUIREMENTS BY SYSTEM

]

TAP NUMBER NOTES B-N-i B13.iO Vessel Interior (Reactor Vessel)

Visual, VT-3 RPV: I Interior of Reactor Vessel B-N-2 Bi3.60 Interior Attachments Beyond Beltline Region (Reactor Vessel)

Visual, VT-3 RPV: 1 Welded Core Support Structures and Interior Attachments to Reactor Vessels B-N-3 B13.70 rCore Support Structure (Reactor Vessel)

Visual, VT-3 RPV: I Removable Core Support Structures 6-0 B14.20 Welds in CRD Housing (Reactor Vessel)

Volumetric or RPV: 45 27 Pressure Retaining (10% of Peripheral CRD Housing welds to be inspected. 45 of 78 Surface Welds in Control Rod welds are identified as peripheral. Examine 5.)

Housings B-P 615.10 Pressure Retaining Components [IWB-5222(a)]

Visual, VT-2 RC: 2 14T-04 22 All Pressure System Leakage lest (IWB-5220) (Outage)

(Includes CV, 32 Retaining RH, RY, and Components SI) 615.20 Pressure Retaining Components [IWB-5222(b)]

Visual, VT-2 RC: 1 14T-04 22 System Leakage lest (IWB-5220) (Interval)

(Includes CV, 32 RH, RY, and SI)

B-Q 616.20 Steam Generator Tubing in U-Tube Design (Steam Generator)

Volumetric Per SG: 4 Steam Generator Tech Specs Tubing Exelon Byron Station 7-6 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Unit I & Common Inservice Inspection Summary Table 7.1-i EXAMINATION ITEM EXAM TOTAL NUMBER RELIEF REQ.I N

T CATEGORY NUMBER DESCRIPTION REQUIREMENTS f BY SYSTEM TAP NUMBER 0

ES C-A CilO Shell Circumferential Welds (Pressure Vessels)

Volumetric RH: 2 14T-07 13 Pressure Retaining SG: 4 14T-05 Welds in Pressure Cl 20 Head Circumferential Welds (Pressure Vessels)

Volumetric RH: 2 14T-07 13 Vessels SG: 4 C1.30 Tubesheet-to-Shell-Welds (Pressure Vessels)

Volumetric SG: 4 C-B C2.21 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Welds Without Volumetric &

RH: 4 141-07 13 Pressure Retaining Reinforcing Plate, Greater Than 1/2 Nominal Thickness (Pressure Surface SG: 8 Nozzle Welds in Vessels)

Vessels C2.22 Nozzle Inside Radius Section Without Reinforcing Plate, Greater Than Volumetric RH: 0 141-07 6

1/2 Nominal Thickness (Pressure Vessels)

SG: 4 13 C-C C3.10 Welded Attachments (Pressure Vessels)

Surface RH: 2 Welded Attachments C320 Welded Attachments (Piping)

Surface CS: 2 for Vessels, Piping, CV: 2 Pumps, and Valves MS: 22 RH: 8 SI: 10 SX: 12 VQ: 4 C3.30 Welded Attachments (Pumps)

Surface CS: 6 CV: 8 RH:_6 Exelon Byron Station 7-7 Revision 0

IS! Program Plan Units I & 2, Fourth Inte,val Unit I & Common lnservice Inspection Summary Table 7.1-1 EXAMINATION ITEM C

o EXAM J TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER DES RIPTI N

REQUIREMENTS BY SYSTEM TAP NUMBER NOTES C-H C710 Pressure Retaining Components Visual, VT-2 AF: 2 141-03 22 All Pressure System Leakage Test (IWC-5220)

CS: 5 14T-04 Retaining CV: 10 Components (Periodic)

FP: 1 FW: 2 IA: 1 MS: 2 OG: 1 PC: 1 PR: 1 PS: 6 RC: 1 RE: 3 RF: 1 RH: 6 RY: 4 SA: 1 SD: 1 SI: 9 SX: 2 VQ: 6 WM: 1 WO: 2 Exelon Byron Station 7-8 Revision 0

151 Program Plan Units I & 2, Fourth Inteival Unit I & Common Inservice Inspection Summary Table 7.1-1 EXAMINATION ITEM ESC o

EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER 0

RI TI N

REQUIREMENTS BY SYSTEM TAP NUMBER NOTES D-A D1.10 Welded Attachments (Pressure Vessels)

Visual, VT-I CC: 4÷2 2

Welded Attachments DG: 12 for Vessels, Piping, RH: 2 Pumps, and Valves sx: 8 D1.20 Welded Attachments (Piping)

Visual, VT-i AF: 3 2

CC: 60+4 SX:_43+10 Di.30 Welded Attachments (Pumps)

Visual, VT-i AF: 8 2

SX:_0+4 D-B D2.10 Pressure Retaining Components Visual, VT-2 AB: 2 141-03 22 All Pressure System Leakage Test (IWD-5220)

AF: 3 141-04 Retaining CC: 7 141-10 Components DG: 10 (Periodic)

DO: 4 FC: 2 RY: 2 SX: 17 WO: 2 Exelon Byron Station 7-9 Revision 0

IS! Program Plan Units I

& 2, Fourth Inteival Unit I

& Common Inservice Inspection Summary Table 7.1-I EXAMINATION ITEM ESCR TION EXAM TOTAL NUMBER RELIEF REQ.!

NOTES CATEGORY NUMBER 0

IP REQUIREMENTS BY SYSTEM TAP NUMBER E-A El.11 Containment Vessel Pressure Retaining Boundary General Visual CC: 5 7

Containment Accessible Surface Area CS: 6 Surfaces CV: 6 FC: 2 FP: 1 FW: 8 IA: I MS: 4 NT: 56 OG: 3 PR: 1 PS: 5 RE: 2 RE: 1 RH: 2 RY: 3 SA: 1 SD: 8 SI: 8 SX: 4 VQ: 5 WM: 1 WO: 4 XX: 47 E1.30 Containment Vessel Pressure Retaining Boundary General Visual 1

7 Moisture Barriers E-C E4.1 1 Containment Surface Areas Visual, VT-i 2

7 Containment Visible Surfaces Surfaces Requiring E4.12 Containment Surface Areas Volumetric 0

7 Augmented Surface Area Grid, Minimum Wall Thickness Location (Ultrasonic Examination Thickness)

Exelon Byron Station 7-10 Revision 0

IS! Program Plan Units I & 2, Fourth InIervI Unit I & Common Inservice Inspection Summary Table 7.1 -

EXAMINATION ITEM

[

EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER DESCRIPTION IREQUIREMENTS BY SYSTEM TAP NUMBER NOTES E-G

[8.10 Containment Vessel Pressure Retaining Boundary-Visual, VT-i [

NT: 47 7

Bolted Connections Bolted Connections, Surfaces j

XX: 28 F-A Fib Class 1 Piping Supports Visual, VT-3 CV: 136 1

Supports RC: 89 RH: 20 RY: 34 SI:_190 Fl.20 Class 2 Piping Supports Visual, VT-3 AF: 34 1

CS: 52 CV:66 FW: 39 MS: 27 RH: 65 SI: 157 SX: 157 VQ:_5 Fl.30 Class 3 Piping Supports Visual, VT-3 AF: 34 1

CC: 326+26 2

SX: 374+248 Fi.40 Supports Other Than Piping Supports Visual, VT-3 AF: 2 1

(Class 1, 2, 3, and MC)

CC: 4+2 2

CS: 6 CV: 8 DG: 2 RC: 29 RH: 10 RY: 5 SI: 4 SX:_12 Exelon Byron Station 7-11 Revision 0

IS! Program Plan Units I & 2, Fourth Inteival Unit I & Common Inservice Inspection Summary Table 7.1-i EXAMINATION ITEM ES RI 110 EXAM TOTAL NUMBER RELIEF REQ.I CATEGORY NUMBER D

C P

N REQUIREMENTS BY SYSTEM TAP NUMBER NOTES L-A Liii Concrete Surfaces General Visual 42 7

Concrete All Accessible Surface Areas Li.i2 Concrete Surfaces Detailed Visual 7

Suspect Areas (No Suspect Areas Identified)

L-B L2.10 Tendon IWL-2522 483 7

Unbonded Post-L2.20 Tendon IWL-2523.2 483 7

Tensioning System Wire or Strand L2.30 Tendon Detailed Visual 966 7

Anchorage Hardware and Surrounding Concrete L2.40 Tendon IWL-2525.2(a),

7 Corrosion Protection Medium IWL-2526 L2.50 Tendon IWL-2525.2(b) 7 Free Water R-A 2

Risk Category 2 Elements See Notes RC: 199 14R-Oi 8

Risk-Informed Piping 3

Risk Category 3 Elements See Notes FW: 128 14T-Oi 9

Examinations 4

Risk Category 4 Elements See Notes CS: 172 141-02 10 CV: 197 l4T06 17 RC:488 18 RH: 200 20 SI:_249 5

Risk Category 5 Elements See Notes AF: 20 CV: 131 SI:_170 Exelon Byron Station 7-12 Revision 0

NA Augmented Components IS! Program Plan Units I & 2, Fourth Interval Unit I & Common lnservice Inspection Summary Table 7.1-1 EXAMINATION ITEM ES TI EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER D

CRIP ON REQUIREMENTS BY SYSTEM TAP NUMBER NOTES 3.6.2 Examination of High Energy Circumferential and Longitudinal Piping Volumetric or NA 5

Welds (MEB 3-1, UFSAR 3.6.1 and 3.6.2).

Surface 10 RG1.14 Augmented Examination of Reactor Coolant Pump Flywheel Per Volumetric or RC: 4 28 Regulatory Guide 1.14 as modified by Surface Byron Station License Surface Amendment #1 18 and Technical Requirements Manual Appendix G.

ECCS Information Notice 79-19, Pipe Cracks in Stagnant Borated Water Volumetric SI: 94 29 Systems at PWR Plants.

0737 Leak Testing and Periodic Visual Examinations of Systems Outside of Visual, VT-2 CS 30 Primary Containment Which Could Contain Highly Radioactive Fluids CV During a Serious Transient or Accident (NUREG 0737).

FC GW OG PS RH SI GL8805 Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Visual, VT-2 RC 31 Pressure Boundary Components in PWR Plants.

APP Q Weld Overlay of PZR Nozzle to Safe-Ends and First Piping Welds.

Volumetric RY: 6 17 MRP-146 EPRI Materials Reliability Program, Management of Thermal Fatigue Visual &

RC: 6 23 in Normally Stagnant Non-Iso/able Reactor Coolant Branch Lines Volumetric fMRP-146).

MRP-1 92 EPRI Materials Reliability Program, Assessment of RHR Mixing Tee Volumetric NA 24 Thermal Fatigue in PWR Plants (MRP-1 92).

MRP-227 Augmented Examination of Pressurized Water Reactor Internals Visual &

RC: 1 26 fMRP-227).

Volumetric N-722-1 B15.80 RPV Bottom-Mounted Instrument Penetrations Visual RC: 1 18 N-729-1 B4.10 Reactor Head Exterior Surface Visual RPV: 1 19 B4.20 Reactor Head Penetrations Volumetric &

RPV: 1 Surface N-770-1 Examination of Class 1 piping and nozzle dissimilar-metal butt welds

Visual, PZR: 6 141-08 20 per 10 CFR 50.55a(g)(6)(ii)(F).

Volumetric &

RPV: 8 Surface Exelon Byron Station 7-13 Revision 0

IS! Program Plan Units I & 2, Fourth InteIv,EI Unit 2 nservice Inspection Summary Table 7.1-2 EXAMINATION ITEM DESCRIPTION EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER REQUIREMENTS BY SYSTEM TAP NUMBER NOTES B-A 81.11 Circumferential Shell Welds (Reactor Vessel)

Volumetric RPV: 3 14R-08 Pressure Retaining 81.21 Circumferential Head Welds (Reactor Vessel)

Volumetric RPV: 2 l4R-08 Welds in Reactor Bi.30 Shell-to-Flange Weld (Reactor Vessel)

Volumetric RPV: 1 14R-08 Vessel Bi.40 Head-to-Flange Weld (Reactor Vessel)

Volumetric &

RPV: 1 14R-08 25 Surface B-B 82.1 1 Circumferential Shell-To-Head Welds (Pressurizer)

Volumetric PZR: 2 Pressure Retaining 82.12 Longitudinal Shell-To-Head Welds (Pressurizer)

Volumetric PZR: 2 Welds in Vessels B2.40 Tube Sheet-To-Head Weld (Steam Generator)

Volumetric SG: 4 Other Than Reactor Vessels B-D 83.90 Nozzle-to-Vessel Welds (Reactor Vessel)

Volumetric RPV: 8 14R-08 11 Full Penetration B3.100 Nozzle Inside Radius Section (Reactor Vessel)

See Note RPV: 8 l4R-08 12 Welds of Nozzles in 83.110 Nozzle-to-Vessel Welds (Pressurizer)

Volumetric PZR: 6 Vessels 83.120 Nozzle Inside Radius Section (Pressurizer)

See Note PZR: 6 14 83.140 Nozzle Inside Radius Section (Steam Generator)

See Note SG: 8 14 B-G-1 B6.10 Closure Head Nuts (Reactor Vessel)

Visual, VT-i RPV: 3 3

Pressure Retaining B6.20 Closure Studs (Reactor Vessel)

Volumetric RPV: 3 3

Bolting, Greater 4

Than 2 In Diameter 86.40 Threads in Flange (Reactor Vessel)

Volumetric RPV: 1 3

86.50 Closure Washers, Bushings (Reactor Vessel)

Visual, VT-I RPV: 3 3

86.180 Bolts & Studs (Pumps)

Volumetric RC: 4 3

34 86.190 Flange Surface, When Connection Disassembled (Pumps)

Visual, VT-i RC: 4 3

34 86.210 Bolts & Studs (Valves)

Volumetric RC: 8 3

34 86.220 Flange Surface, When Connection Disassembled (Valves)

Visual, VT-i RC: 8 3

34 86.230 Nuts, Bushings, and Washers (Valves)

Visual, VT-i RC: 8 3

34 Exelon Byron Station 7-14 Revision 0

IS! Program Plan Units I & 2, Fourth lnte,v&

Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY J_NUMBER DESCRIPTION REQUIREMENTS BY SYSTEM TAP NUMBER NOTES B-G-2 B7.1O Bolts, Studs, & Nuts (Reactor Vessel)

Visual, VT-i RPV: 2 3

Pressure Retaining B7.20 Bolts, Studs, & Nuts (Pressurizer)

Visual, VT-i PZR: 1 3

Bolting, 2 and Less B7.30 Bolts, Studs, & Nuts (Steam Generator)

Visual, VT-i SG: 4 3

In Diameter B7.50 Bolts, Studs, & Nuts (Piping)

Visual, VT-i CV: 4 3

RC: 4 RY: 4 SI:8 B7.60 Bolts, Studs, & Nuts (Pumps)

Visual, VT-i RC: 4 3

35 B7.70 Bolts, Studs, & Nuts (Valves)

Visual, VT-i RC: 4 3

RH:4 35 SI:_18 B-K BlOb Welded Attachments (Pressure Vessels)

Surface or PZR: 2 15 Welded Attachments Volumetric for Vessels, Piping, BiO.20 Welded Attachments (Piping)

Surface CV: 1 16 Pumps, and Valves SI: 7 B-L-2 B12.20 Pump Casing (Pumps)

Visual, VT-3 RC: 4 34 Pump Casings 35 B-M-2 Bi2.50 Valve Body, Exceeding NPS 4 (Valves)

Visual, VT-3 RC: 12 34 Valve Bodies RH: 4 35 RY: 3 SI: 18 Exelon Byron Station 7-15 Revision 0

IS! Program Plan Units I & 2, Fourth Inte,val Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM EXAM TOTAL NUMBER RELIEF REQ.I CATEGORY NUMBER DESCRIPTION REQUIREMENTS BY SYSTEM TAP NUMBER NOTES t

B-N-i B13.10 Vessel Interior (Reactor Vessel)

Visual, VT-3 RPV: 1 Interior of Reactor Vessel B-N-2 B13.60 Interior Attachments Beyond Beitline Region (Reactor Vessel)

Visual, VT-3 RPV: 1 Welded Core Support Structures and Interior Attachments to Reactor Vessels B-N-3 Si 3.70 Core Support Structure (Reactor Vessel)

Visual, VT-3 RPV: 1 Removable Core Support Structures B-C 814.20 Welds in CRD Housing (Reactor Vessel)

Volumetric or RPV: 45 27 Pressure Retaining (10% of Peripheral CRD Housing welds to be inspected. 45 0178 Surface Welds in Control Rod welds are identified as peripheral. Examine 5.)

Housings B-P 815.10 Pressure Retaining Components [IWB-5222(a)]

Visual, VT-2 RC: 2 14T-04 22 All Pressure System Leakage Test (IWB-5220) (Outage)

(Includes CV, 32 Retaining RH, RY, and Components SI) 815.20 Pressure Retaining Components [IWB-5222(b)]

Visual, VT-2 RC: 1 141-04 22 System Leakage Test (IWB-5220) (Interval)

(Includes CV, 32 RH, RY, and SI)

B-Q 816.20 Steam Generator Tubing in U-Tube Design (Steam Generator)

Volumetric Per SG: 4 Steam Gen. Tubing Tech Specs Exelon Byron Station 7-16 Revision 0

IS! Program Plan Units I & 2, Fourth Inteival Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM c

o EXAM TOTAL NUMBER RELIEF REQ.i T

CATEGORY NUMBER DES R

TI N

REQUIREMENTS BY SYSTEM TAP NUMBER NO ES C-A C1.10 Shell Circumferential Welds (Pressure Vessels)

Volumetric J

RH: 2 41-07 13 Pressure Retaining I

I SG: 12 Welds in Pressure Cl.20 Head Circumferential Welds (Pressure Vessels)

I Volumetric I RH: 2 14T-07 13 Vessels J

SG:4 Cl.30 Tubesheet-to-Shell-Welds (Pressure Vessels)

Volumetric SG: 4 C-B C2.21 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Welds Without Volumetric &

RH: 4 l4T-07 13 Pressure Retaining Reinforcing Plate, Greater Than 1/2 Nominal Thickness (Pressure Surface SG: 12 Nozzle Welds in Vessels)

Vessels C2.22 Nozzle Inside Radius Section Without Reinforcing Plate, Greater Volumetric RH: 0 I4T-07 6

Than 1/2 Nominal Thickness (Pressure Vessels)

SG: 4 13 C-C C3.10 Welded Attachments (Pressure Vessels)

Surface RH: 2 Welded Attachments C3.20 Welded Attachments (Piping)

Surface CS: 3 for Vessels, Piping, CV: 2 Pumps, and Valves FW: 4 MS: 32 RH: 6 SI: 12 SX: 13 VQ: 4 C3.30 Welded Attachments (Pumps)

Surface CS: 6 CV: 8 RH:_6 Exelon Byron Station 7-17 Revision 0

IS! Program Plan Units I & 2, Fourth lnte,v&

Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM ES RI 110 EXAM TOTAL NUMBER RELIEF REQ.I CATEGORY NUMBER D

C P

N REQUIREMENTS BY SYSTEM TAP NUMBER NOTES C-H C7.10 Pressure Retaining Components Visual, VT-2 AF: 2 141-03 22 All Pressure System Leakage Test (IWC-5220)

CS: 5 14T04 Retaining cv: io Components (Periodic)

FP: 1 FW: 2 IA: 1 MS: 2 OG: 1 PC: 1 PR: 1 PS: 6 RC: 1 RE: 3 RF: 1 RH: 6 RY: 4 SA: 1 SD: 1 SI: 9 SX: 2 VQ: 6 WM: 1 WO: 2 Exelon Byron Station 7-18 Revision 0

IS! Program Plan Units I

& 2, Fourth lntenial Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM Sc IP o

I EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER DE R

TI N

IREQUIREMENTS BY SYSTEM TAP NUMBER NOTES D-A Di.10 Welded Attachments (Pressure Vessels)

Visual, VT-i CC: 4 Welded Attachments DG: 12 for Vessels, Piping, RH: 2 Pumps, and Valves SX: 8 Dl.20 Welded Attachments (Piping)

Visual, VT-i AF: 2 CC: 4 SX:_15 Dl.30 Welded Attachments (Pumps)

Visual, VT-i AF: 8 D-B D2.10 Pressure Retaining Components Visual, VT-2 AB: 1 14T-03 22 All Pressure System Leakage Test (IWD-5220)

AF: 3 141-04 Retaining CC:6 14T-10 Components DG: 10 (Periodic)

DO: 3 FC: I RY: 2 SX:_5 Exelon Byron Station 7-19 Revision 0

IS! Program Plan Units I & 2, Fourth Inteiv&

Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION F

ITEM E

R i

EXAM TOTAL NUMBER RELIEF REQ.!

NO ES CATEGORY j

NUMBER D Sc IPT ON REQUIREMENTS BY SYSTEM TAP NUMBER T

E-A El.11 Containment Vessel Pressure Retaining Boundary General Visual CC: 5 7

Containment Accessible Surface Area CS: 6 Surfaces CV: 6 FC: 2 FP: 1 FW: 8 IA: 1 MS: 4 NT: 56 OG: 3 PR: 1 PS: 5 RE: 2 RF: 1 RH: 2 RY: 3 SA: 1 SD: 8 SI: 8 SX: 4 VQ: 5 WM: 1 WO: 4 XX: 47 El.30 Containment Vessel Pressure Retaining Boundary General Visual 1

7 Moisture Barriers E-C E4.i 1 Containment Surface Areas Visual, VT-i 0

7 Containment Visible Surfaces Surfaces Requiring E4.12 Containment Surface Areas Volumetric 0

7 Augmented Surface Area Grid, Minimum Wall Thickness Location (Ultrasonic Examination Thickness)

Exelon Byron Station 7-20 Revision 0

IS! Program Plan Units 7 & 2, Fourth lnte,val Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM DESCRIPTION EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER REQUIREMENTS BY SYSTEM TAP NUMBER NOTES E-G ES. 10 Containment Vessel Pressure Retaining Boundary -

Visual, VT-i NT: 92 7

Bolted Connections Bolted Connections, Surfaces XX: 28 F-A Fib Class 1 Piping Supports Visual, VT-3 CV: 143 Supports RC: 86 RH: 26 RY: 37 SI:_175 Fl.20 Class 2 Piping Supports Visual, VT-3 AF: 36 CS: 55 CV: 57 FW: 97 MS: 32 RH: 78 SI: 147 SX: 155 VQ:_5 F1.30 Class 3 Piping Supports Visual, VT-3 AF: 41 CC: 52 SX:_277 Fl.40 Supports Other Than Piping Supports Visual, VT-3 AF: 2 (Class 1, 2, 3, and MC)

CC: 4 CS: 6 CV: 8 DG: 2 RC: 29 RH: 10 RY: 5 SI: 4 SX: 4 Exelon Byron Station 7-21 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Unit 2 lnservice Inspection Summary Table Z.1-:.

EXAMINATION ITEM EXAM TOTAL NUMBER RELIEF REQ.!

CATEGORY NUMBER DESCRIPTION REQUIREMENTS BY SYSTEM TAP NUMBER NOTES L-A Li.11 Concrete Surfaces -

General Visual 42 7

Concrete All Accessible Surface Areas L1.12 Concrete Surfaces Detailed Visual 7

Suspect Areas (No Suspect Areas Identified)

L-B L2.10 Tendon IWL-2522 483 7

Unbonded Post-L2.20 Tendon IWL-2523.2 483 7

Tensioning System Wire or Strand L2.30 Tendon Detailed Visual 966 7

Anchorage Hardware and Surrounding Concrete L2.40 Tendon IWL-2525.2(a),

7 Corrosion Protection Medium IWL-2526 L2.50 Tendon IWL-2525.2(b) 7 Free Water R-A 2

Risk Category 2 Elements See Notes RC: 186 I4R-01 8

Risk-Informed Piping 3

Risk Category 3 Elements See Notes FW: 242 14T-0i 9

Examinations 4

Risk Category 4 Elements See Notes CS: 164 I4T-02 10 CV: 203 I4T-06 17 RC:465 18 RH: 215 20 SI:_241 5

Risk Category 5 Elements See Notes AF: 20 CV: 127 SI: 159 Exelon Byron Station 7-22 Revision 0

Volumetric or Surface

Visual, Volumetric &

Surface NA PZR: 6 RPV: 8 NA Augmented Components 3.6.2 IS! Program Plan Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.1-2 EXAMINATION ITEM OESCRIPTIO EXAM TOTAL NUMBER 1 RELIEF REQ.I CATEGORY NUMBER N

REQUIREMENTS BY SYSTEM LTAP NUMBER NOTES Examination of High Energy Circumferential and Longitudinal Piping Welds (MEB 3-1, UFSAR 3.6.1 and 3,6.2).

5 10 RG1.14 Augmented Examination of Reactor Coolant Pump Flywheel Per Volumetric or RC: 4 28 Regulatory Guide 1.14 as modified by Surface Byron Station License Surface Amendment #118 and Technical Requirements Manual Appendix G.

ECCS Information Notice 79-19, Pipe Cracks in Stagnant Borated Water Volumetric SI: 98 29 Systems at PWR Plants.

0737 Leak Testing and Periodic Visual Examinations of Systems Outside of Visual, VT-2 CS 30 Primary Containment Which Could Contain Highly Radioactive Fluids CV During a Serious Transient or Accident (NUREG 0737).

FC GW OG PS RH SI GL8805 Generic Letter 88-05, Boric Acid Corrosion of Carbon Steel Reactor Visual, VT-2 RC 31 Pressure Boundary Components in PWR Plants.

APP Q Weld Overlay of PZR Nozzle to Safe-Ends and First Piping Welds.

Volumetric RY: 6 17 MRP-146 EPRI Materials Reliability Program, Management of Thermal Fatigue Visual &

RC: 6 23 in Normally Stagnant Non-Isolable Reactor Coolant Branch Lines Volumetric (MRP-146).

MRP-192 EPRI Materials Reliability Program, AssessmentofRHR Mixing Tee Volumetric NA 24 Thermal Fatigue in PWR Plants (MRP-1 92).

MRP-227 Augmented Examination of Pressurized Water Reactor Internals Visual &

RC: 1 26 (MRP-227).

Volumetric N-722-1 B15.80 RPV Bottom-Mounted Instrument Penetrations Visual RC: 1 18 N-729-J 84.10 Reactor Head Exterior Surface Visual RPV: 1 19 B4.20 Reactor Head Penetrations Volumetric &

RPV: 1 Surface N-770-1 Examination of Class 1 piping and nozzle dissimilar-metal butt welds per 10 CFR 50.55a(g)(6)(ii)(F).

141-08 20 Exelon Byron Station 7-23 Revision 0

151 Program Plan Units I & 2, Fourth Interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-3 NOTE#

NOTE

SUMMARY

1 Snubber visual examinations, functional testing, and service life monitoring are performed in accordance with ASME CM Code, Subsections ISTA and ISTD. For a detailed discussion of the Byron Station Snubber Program, refer to the Snubber Program Document and Section 4.2 of this document.

2 The Byron Station Unit 1 population counts include those components that are common to both units (typically designated as Common. These Common components are referenced in Table 7.1-1 following a + symbol to designate the Common Unit.

3 Bolting is characterized by one entry per valve, pump, piping flanges, or vessel manways not by the actual total number of bolts or studs. When the examination is requited for a given items bolting, all bolts shall be inspected. The reactor vessel closure head studs, nuts, and washers (54 total for each item) are examined during more than one Inspection Period. The number of separate examinations for each item identifies the population of these components.

4 Examination Category B-G-1, Item Numbers B6.20 Closure Studs, In Place and B6.30 Closure Studs, When Removed have been combined into and renamed as Item Number B6.20 Closure Studs, in Table IWB-2500-1 of ASME Section Xl, 2007 Edition with the 2008 Addenda.

5 The population counts reported represents the number of non-exempt circumferential welds. Longitudinal welds are also subject to examination, but actual counts are not reported here. Byron Station examines the portion of the longitudinal weld that falls within the intersecting circumferential weld examination volume.

6 Subsection IWC, Table IWC-2500-1, Examination Category C-B, Item Number C2.22 requires volumetric examination of the nozzle inner radii of nozzles without reinforcing plates in vessels with nominal thickness> 1/2 in.

a.

The Main Steam nozzle was designed with an internal multiple venture-type flow restrictor with an equivalent throat diameter of 16 inches. This design is used to limit the flow in the event of a postulated steam line break.

b.

The Residual Heat Removal Heat Exchanger nozzle unique configuration with the reinforcing pads being on the internal surface, the nozzle inner radius section is inaccessible for examination.

These designs do not utilize a radius nozzle as described in Figures IWC-2500-4(a) and (b), and therefore is not considered as an Examination Category C-B, Item Number C2.22 component. However, these nozzles will receive a system leakage test during each inspection period to verify structural integrity.

Exelon Byron Station 7-24 Revision 0

151 Program Plan Units I & 2, Fourth interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-NOTE #

NOTE

SUMMARY

7 Examination requirements of Examination Category E-A components Includes all unique identified inspectable surface areas, i.e., Each penetration is one component (total 158).

Examination requirements of Examination Category E-G components Bolted Connections: Each connection bolt group is counted as I item (i.e., 20 bolt flange connection equals 1 item).

Examination requirements of Examination Category L-A components Counted three main Areas (Exterior wall, Exterior Dome, and Tendon gallery ceiling)

Examination requirements of Examination Category L-B components Equals total number of bearing plates (each bearing plate includes Anchorage hardware and surrounding concrete)

Includes (4) Distinct Areas:

Horizontal Wall Tendons 402 bearing plates Dome Tendons 240 bearing plates Upper Vertical Tendons 162 bearing plates Lower Vertical Tendons 162 bearing plates (Total)

(966 bearing plates) 8 For the Fourth 1St Interval, Byron Stations 1St Class 1 and 2 piping inspection program will be governed by risk-informed regulations. The RI-ISI Program methodology is described in the EPRI Topical Reports TR-112657, Rev. B-A, TR-1006937, Rev. 0-A, and N-578-1. The RI-ISI Program scope has been implemented as an alternative to the 2007 Edition with the 2008 Addenda of the ASME Section XI examination program for ISI Class 1 B-F and B-J welds and 151 Class 2 C-F-i and C-F-2 welds in accordance with 10 CFR 50.55a(z).

9 Examination requirements for 1St Class 1 and 2 piping structural elements within the RI-ISI Program are determined by the various degradation mechanisms present at each individual piping structural element. See EPRI Topical Reports TR-i 12657, Rev. B-A, TR-1 006937, Rev. 0-A, and N-578-1 for specific examination method requirements.

10 For the Fourth ISI Interval, the RI-ISI Program scope continues to include welds in the BER piping, also referred to as the HELB region, which includes several non-class welds that fall within the BER augmented inspection program. All BER augmented welds have been evaluated under the Rl-lSl methodology and have been integrated into the RI-ISI Program. Additional guidance for adaptation of the RI-ISI evaluation process to BER piping is given in EPRI TR-1 006937, Rev. 0-A. Thus, these welds have been categorized and selected for examination in accordance with the EPRI Topical Reports TR-1 12657, Rev. B-A, TR-i 006937, Rev. 0-A, and N-578-i in lieu of the original requirement to NUREG 0800 in UFSAR Section 3.6.2. The populations are identified with the applicable systems within the Rt-ISI section of the tables.

ii As allowed by ASME Code Case N-61 3-I (N-61 3-1), Byron Station will perform a volumetric examination using a reduced examination volume (A-B-C D-E-F-G-H) of Figures 1, 2, and 3 of the Code Case in lieu of the previous examination volumes of ASME Section Xl, Figures IWB-2500-7(a), (b), and (c).

12 As allowed by ASME Code Case N-648-1, Byron Station will perform a VT-i visual examination in lieu of a volumetric examination in ASME Section Xl utilizing the allowable flaw length criteria of Table IWB-3512-i with limiting assumptions on the flaw aspect ratio. For Item B3.100, a VT-i visual examination may be performed per conditionally approved ASME Code Case N-648-1 (N-648-i). (See Note 12 above for details on N-648-1.)

Exelon Byron Station 7-25 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-3 NOTE #_[

NOTE

SUMMARY

13 As allowed by N-706-1, Byron Station will perform a VT-2 visual examination each period in lieu of the volumetric and/or volumetric and surface examinations of ASME Section Xl. Note that the alternative requirements detailed in Table I of the Code Case apply 2jjjy to the PWR stainless steel regenerative and residual heat exchanger components. (See TAP 14T-07 in Section 2.5 for details on this Code Case.)

14 Per 10 CFR 50.55a(b)(2)(xxi)(A), Table IWB-2500-7 examination requirements, the provisions of Table IWB-2500-1, Examination Category B-D, Item Numbers 83.120 and 83.140 in the 1998 Edition must be applied when using the 1999 Addenda through the latest Edition and Addenda, and requires that a visual examination with magnification may be performed on the inside radius section instead of an ultrasonic examination.

15 The examination requirements for the pressurizer support skirt to vessel weld is identified as Figure IWB-2500-13. The examination surfaces are A-B for the outer portion or C-D for the inner portion of the weld. As shown on the fabrication records for this weld, the inner portion is not accessible for tile surface examination of C-D. A backing ring was used during the welding of the support weld. This design does not utilize an exposed inner surface as shown in Figure lWB-2500-1 3 and therefore is not considered subject to the inner surface examination.

16 For lSl Class I penetration closure plate welds (Item Number 810.20), the examination surfaces are A-B for the outer portion and C-D for the inner portion of the weld. As shown on the fabrication records, the inner portion is not accessible for the surface examination of C-D. A backing ring was used during the welding of a single-V support weld. This design does not utilize an exposed inner surface as shown in Figure IWB-2500-1 5 and therefore is not considered subject to the inner surface examination.

17 Butt welds fabricated with Alloy 600 materials and overlayed with PWSCC resistant materials are incorporated into programs required by: ASME Section Xl Nonmandatory Appendix Q for full structural weld overlays.

18 Per 10 CFR 50.55a(g)(6)(ii)(E)(7), all licensees of pressurized water reactors shall augment their inservice inspection program by implementing N-722-1 subject to the conditions specified in paragraphs (g)(6)(ii)fE)(2) through (4) of this section. The inspection requirements do not apply to components with pressure retaining welds that have been mitigated by weld overlay or stress improvement.

19 Per 10 CFR 50.55a(g)(6)(ii)(D)(1), all licensees of pressurized water reactors shall augment their inservice inspection program with N-729-1 subject to the conditions specified in paragraphs (g)(6)(ii)(D)(2) through (6) of this section.

20 Per 10 CFR 50.55a(g)(6)(ii)(F), all licensees of pressurized water reactors shall augment their inservice inspection program by implementing N-770-1 subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (10) of this section.

21 The replacement of the Reactor Coolant Pump 1RCO1PB main flange bolting in B1R17 with studs, nuts, and washers added Item Number 86.200 to the examination tables.

22 The Total Number By System represents the total number by system pressure test blocks for Examination Category B-P, C-H, and D-B components.

23 MRP-146 Augmented Examination Program for the screening, evaluation, and inspection requirements for potential thermal fatigue cracking that may occur in normally stagnant non-isolable piping systems attached to pressurized water reactor (PWR) main reactor coolant system (RCS) piping.

Currently, the examinations are limited to the following locations:

Cold leg charging lines 1/2RC28A-3 and 1/2RC37A-3.

Loop Drain lines 1/2RCJ4AX-2 (Loops A, B, C, and D).

Exelon Byron Station 7-26 Revision 0

151 Program Plan Units I & 2, Fourth Interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-3 NOTE # ]

NOTE

SUMMARY

24 MRP-192 Augmented Examination Program was issued as a good practice industry guideline per NEI 03-08 classification. Current operating conditions for temperature differentials of the mixing flows and cumulative operating time in mixing mode, are within the limits specified in MRP-i 92 (<

144°F), therefore, no examinations are currently scheduled.

25 Byron Station will utilize the alternative requirements of ASME Code Case N-747 to provide the reactor vessel head-to-flange weld to be inspected by surface examination once each ten-year inspection interval, using the current surface examination area shown in Figure IWB-2500-5. This alternative requirement may only be implemented after the weld has received at least one inservice volumetric examination, which may be performed as part of the preservice inspection, with no service-induced flaws having been identified. Hence, there have been no defects detected at Byron Station on this weld during pre-service or inservice examinations. It is therefore concluded that the concurrent volumetric and surface examination requirement may be eliminated for the reactor vessel head-to-flange weld, and that the outer surface examination discussed above will be performed.

26 The RPV interior requires examination per ASME Section XI requirements for Examination Categories B-N-i, B-N-2, and B-N-3. In addition to the ASME Section XI requirements, the Byron Station PWR Internals Program for the inspection, repair, replacement, degradation evaluation, and mitigation of the PWR Reactor Internals will ensure that Materials Reliability Program (MRP) and PWR Owners Group (PWROG) Guidelines are met.

Augmented examination requirements associated with the Byron Station PWR Internals Program are maintained and controlled in procedures ER-AP 333, Pressurized Water Reactor Internals Management Program and ER-AP-333-1 001, Pressurized Water Reactor Internals Program.

27 Examination Category B-O (Pressure-Retaining Welds In Control Rod Housings), Item Number B14.20 (Welds in CRD Housing)

- the scope of examination is for pressure retaining welds in 10% of the peripheral CRD Housings. A total of 45 out of the 78 CRD Housings are classified as peripheral components. Byron Station is required to select the welds on 5 Peripheral CRD Housings to be examined during the interval (10% of 45).

28 The Byron Station requirement to NRC Regulatory Guide 1.14, Reactor Coolant Pump Flywheel Integrity, has been modified by Byron Station License Amendment #118 and Technical Requirements Manual Appendix G. (See Section 2.2.2 of the ISI Program Plan for details on this Augmented Examination Program.)

29 The NRC has expressed a concern in lines that contain stagnant borated water. Byron Station will perform augmented volumetric examinations on lSl Class 2 ECCS system; Safety Injection (SI) that is not currently subject to volumetric examination as required by ASME Section XI. The inspections shall include seven and one-half percent (7.5%) sampling of the total population of circumferential welds > 4 nominal pipe size which contain stagnant borated water. Nominal pipe wall thickness and pressure/temperature exemptions do not apply.

Exelon Byron Station 7-27 Revision 0

IS! Program Plan Units I & 2, Fourth Interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-3 NOTE #

NOTE

SUMMARY

30 Byron Station UFSAR Section E.77 addresses the requirement to implement a station leakage inspection program. The program is required as part of Byron Stations commitment to NUREG-0737, TMI Action Plan, Section Ill.D.1.1, Integrity of Systems Outside Containment Likely to Contain Radioactive Material for Pressurized Water Reactors and Boiling Water Reactors.

UFSAR Section E.77 outlines the specific inspection requirements of the leakage inspection program and the systems, or portions of systems, which the program shall inspect. This documentation provides the licensing basis for the scope of Byron Stations NUREG-0737 program. Systems inspected are limited to those listed in E.77. To define a system, Byron Station component numbers and piping line numbers, as shown on the P&lDs, shall detail which system an individual component or line is a part of for the purpose of applying the inspection requirements of UFSAR Section E.77 and NUREG-0737.

Byron Station Unit 1 Systems or portions of systems subject to the augmented testing of NUREG-0737 are included on P&IDs M-46 SheetsiA, 1 B, and 1C; M-47 Sheet 2; M-48 Sheet 18; M-61 Sheets 1A, 18, 2, 3 and 4; M-62; M-63 Sheets 1A, 1B, and 1C; M-64 Sheets 1, 2, 3A, 38, 4A, 48, 5, 6, and 7; N-65 Sheets 18 and 2A; M-68 Sheets JA, 18, and 6; M-69 Sheets 1, 2, and 3; M-70 Sheet 1; M-82 Sheet 1, 2, 3, 5, and 14; and M-152 Sheet 6.

Byron Station Unit 2 Systems or portions of systems subject to the augmented testing of NUREG-0737 are included on P&lDs M-82 Sheet 6; M-129 Sheet slA and 1C; M-136 Sheets 1,2,3, and 4; M-137; M-138 Sheets 1,2, 3, 3A, 3B, 4, 5A, 6, and 7; M-140 Sheets 1 and 5; M-141 Sheet 1; M-150 Sheet 2; and M-1 52 Sheet 6.

31 Generic Letter 88-05 dated March 17, 1988 addresses boric acid corrosion of carbon steel reactor pressure boundary components in pressurized water reactors. Per the Byron Station response to Generic Letter 88-05, Commonwealth Edison letter from W. E. Morgan to A. B. Davis dated May 31, 1988 (NTS Item #456-104-88-00500), a pre-outage VT-2 visual examination shall be performed to locate evidence of boric acid leakage from the reactor coolant pressure boundary. ASME Section XI also requires a VT-2 visual examination of all ISI Class 1 components prior to reactor start-up. For those portions of the reactor coolant pressure boundary that are ASME Class 2 as defined by UFSAR Section 5.2, a post-outage VT-2 visual examination shall also be performed to locate evidence of boric acid leakage.

Leakage from systems containing boric acid results in residue and crystallization accumulations. By performing a VT-2 visual inspection prior to entering an outage (a station augmented inspection commitment as outlined above), any evidence of boric acid accumulations will be found and investigated to determine the source of leakage before normal outage maintenance activities can clean off the crystals and residue. These visual examinations are performed using certified VT-2 visual examiners in accordance with the Exelon VT-2 procedure. This procedure utilizes ASME Section XI VT-2 certified visual examiners and has proven successful in locating evidence of boric acid leakage in the past. Since this augmented inspection is performed solely for the purpose of detecting evidence of boric acid leakage, the system is not required to be pressurized provided any evidence of boric acid accumulations is evaluated and the source of the leakage is determined. This action may require pressurizing the system as part of the evaluation process.

Byron Station Unit 1 systems or portions of systems subject to the augmented testing of Generic Letter 88-05 are included on P&IDs M-60 Sheets 1A, 18, 2, 3, 4, and 5; M-61 Sheets 2, 3, 4, 5, and 6; M-62; M-64 Sheets 1, 2, 38, and 5; M-68 Sheets 1A, 18, and 7; and M-2060 Sheets 6, 7, 8, 17, and 18.

Byron Station Unit 2 systems or portions of systems subject to the augmented testing of Generic Letter 88-05 are included on P&IDs M-135 Sheets 1A, 18, 2, 3, 4, and 5; M-136 Sheets 2, 3, 4, 5, and 6; M-137; M-138 Sheets 1, 2, 38, 5A, and 5B; M-140 Sheet 1; and M-2135 Sheets 6, 7, 8, 17, and 18.

Exelon Byron Station 7-28 Revision 0

IS! Program Plan Units I & 2, Fourth Inteiva!

Inservice Inspection Program Notes Inservice Inspection Summary Table 7.1-3 NOTE #

NOTE

SUMMARY

32 Paragraph IWB-5221(a) requires that for the 10-year interval system pressure test, the system shall be pressurized to a pressure not less than the pressure corresponding to 100% rated reactor power. Paragraph IWB-5222(b) requires that the boundary subject to test pressurization during this test shall extend to all ISI Class I pressure retaining components within the system boundary.

Certain portions of the ISI Class 1 boundary are normally isolated during the periodic system leakage test. Per Paragraph IWB-5222(b), these portions of the system are required to be pressurized and inspected once per 10-year interval. Typically, these portions of the Class 1 boundary are emergency core cooling system (ECCS) injection lines which are isolated from the reactor coolant pressure boundary by two check valves or valves whose logic is linked together. During the ISI Class I system leakage test, the inboard valve is closed and these segments of lines are not pressurized. Also, the ECCS systems are not typically tested during an injection to the reactor vessel but rather during a test mode line up, and thus the lSl Class 1 isolated line segments are not pressurized during the ECCS system test either. The pressure retaining boundary during the system leakage test shall correspond to the reactor coolant system boundary, with all valves in the normal position, which is required for normal reactor operation startup. The VT-2 visual examination shall, however, extend to and include the second closed valve at the boundary extremity.

Through the use of jumper hoses, an external test rig, or an abnormal valve line-up, the isolated portions of the 151 Class 1 boundary shall be pressurized once every 1 0-year inspection interval. The test pressure shall be the same test pressure as that of the lSl Class 1 system leakage test per Paragraph IWB-5222(b). These portions of lines are ISI Class 1 due to their reactor coolant isolation function. They are not made ISI Class 1 due to the connected ECCS system safety function.

They are also designed and installed to withstand the pressures associated with isolating the Reactor Coolant System. Taking into account the IS!

Class 1 boundary isolation function of these lines, the nominal operating pressure of these segments when performing their 151 Class I safety function would be the normal reactor coolant pressure as required by the Code Case. Therefore, any special tests performed on these lines shall be performed at minimum pressure associated with normal reactor coolant pressure (approx. 2235 psig).

33 Footnote 1 of table IWC-2500-1, Examination Category C-H, requires the system parts to be VT-2 visually examined be those other than open ended portions of systems. ASME Interpretation XI-1-89-30 supports the approach of excluding open ended portions of Class 2 systems from normal periodic VT-2 visual examination.

The definition of open ended in ASME Section XI, Subsection IWA-9000 is a condition of piping which permits free discharge. Therefore, open ended piping up to the first isolation valve on ISI Class 2 systems is excluded from the system pressure testing program per Footnote 1 as discussed above.

Since check valves in discharge lines still permit free discharge flow, these valves do not qualify as an isolation valve for the purpose of applying Footnote 1 of Table IWC-2500-1, Examination Category C-H. The excluded ISI Class 2 piping thus extends up to the next isolation point in discharge lines when the last valve in the piping is a check valve (Reference the Containment Spray discharge piping and ring headers for an example of this configuration).

Exelon Byron Station 7-29 Revision 0

151 Program Plan Units I & 2, Fourth Interval Inservice Inspection Program Notes Inservice Inspection Summary Table 7.7-3 NOTE#

NOTE

SUMMARY

34 Table IWB-2500-1, Examination Category B-G-1, Note 4 allows limiting pressure retaining bolting examinations to those components selected for examination under Examination Categories B-B, B-L-2 and B-M-2.

Pumps and valves shall be grouped in accordance with Examination Categories B-L-2 and B-M-2 as follows:

EXAMINATION CATEGORY B-L-2 EXAMINATION CATEGORY B-M-2 JRCO1PA 2RCO1PA 1RC800JA 2RC800IA 1RCOIPB 2RCO1PB 1RC8001B 2RC8001B 1 RCO1 PC 2RCO1 PC 1 RC800I C 2RC8001 C 1RCO1PD 2RCO1PD 1RC800ID 2RC80010 1RC8002A 2RC8002A IRC8002B 2RC8002B 1RC8002C 2RC8002C IRC8002D 2RC8002D ExelonByron Station 7-30 RevisionO

151 Program Plan Units I & 2, Fourth lnterv&

Inservice Inspection Program Notes Inservice Inspection Summary Table 71-3 NOTE #

NOTE

SUMMARY

35 Table IWB-2500-1, Examination Category B-G-2, Note 2 allows limiting pressure retaining bolting examinations to those components selected for examination under Examination Categories B-B, B-L-2 and B-M-2.

Pumps and valves shall be grouped in accordance with Examination Categories B-L-2 and B-M-2 as follows:

EXAMINATION CATEGORY B-L-2 EXAMINATION CATEGORY B-M-2 1RCO1PA 2RCO1PA 1RC8003A 2RC8003A 1S18841A 2S18841A 1RCO1PB 2RCOJPB 1RC8003B 2RC8003B 1S18841B 2S18841B 1RCOIPC 2RCO1PC 1RC8003C 2RC8003C JRCO1PD 2RCO1PD 1RC8003D 2RC8003D 1S18948A 2S18948A 1SI89488 2S18948B 1 RH87O1A 2RH8701A 1 S18948C 2S18948C 1 RH8701 B 2RH8701 B 1 S18948D 2S18948D 1 RH8702B 2RH8702B 1 S18956A 2S18956A 1 RH8702A 2RH8702A 1 S18956B 2S18956B 1S18956C 2S18956C 1 RY8O1 OA 2RY801 OA 1 S18956D 2S18956D 1RY8O1OB 2RY8O1OB 1 RY8O1 OC 2RY801 OC 1 S)8949A 2S18949A JS18949B 2S189498 161881 8A 2S1881 8A 1 SI8949C 2S18949C 1 Sl881 8B 2S1881 8B 1 S18949D 2S18949D 1S18818C 2S18818C 1S18818D 2SI8818D Exelon Byron Station 7-31 Revision 0

IS! Program Plan Units I & 2, Fourth Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests written per 10 CFR 50.55a(z)(1) for situations where alternatives to ASME Section Xl requirements provide an acceptable level of quality and safety; per 10 CFR 50.55a(z)(2) for situations where compliance with ASME Section Xl requirements results in a hardship or an unusual difficulty without a compensating increase in the level of quality and safety; per 10 CFR 50.55a(g)(5)(iii) for situations where ASME Section Xl requirements are considered impractical; and for situations where use of a subsequent approved ASME Section Xl Edition and Addenda is requested.

The following NRC guidance was utilized to determine the correct 10 CFR 50.55a Paragraph citing for Byron Station relief requests.

10 CFR 50.55a(z)(1):

Cited in relief requests when alternatives to the ASME Section Xl requirements which provide an acceptable level of quality and safety are proposed. Examples are relief requests which propose alternative NDE methods andlor examination frequency.

10 CFR 50.55afz)f2):

Cited in relief requests when compliance with the ASME Section Xl requirements is deemed to be a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to:

Having to enter multiple Technical Specifications Limiting Conditions for Operation, As low as reasonably achievable (ALARA) concerns such as excessive radiation exposure, Replacing equipment or in-line components, Creating significant hazards to plant personnel, Disassembly of components solely to provide access for examinations, and Development of sophisticated tooling that would result in only minimal increases in examination coverage.

10 CFR 50.55a(g)(5)(iii):

Cited En relief requests when conformance with ASME Section Xl requirements is deemed impractical. Examples of impractical requirements are situations where the component would have to be redesigned, or replaced to enable the required inspection to be performed.

10 CFR 50.55a(g)(4)(iv):

Cited in relief requests to use subsequent editions and addenda of ASME Section Xl. These editions and addenda are incorporated by reference in 10 CFR 50.55a(b), subject to the limitations and modifications listed in paragraph 10 CFR 50.55a(b), and subject to Commission approval. Portions of editions or addenda may be used provided that all related requirements of the respective editions or addenda are met.

An index for Byron Station relief requests is included in Table 8.0-1. The l4R-XX relief requests are applicable to ISI, CISI, SPT, and PDI.

The following relief requests are subject to change throughout the inspection interval (e.g.,

NRC approval, withdrawal). Changes to NRC approved alternatives (other than withdrawal) require NRC approval.

Exelon Byron Station 8-1 Revision 0

IS! Program Plan Units I & 2, Fourth Intetval TABLE 8.0-1 RELIEF REQUEST INDEX RELIEF REVISION 2

(PROGRAM) DESCRIPTION OF RELIEF REQUEST!

REQUEST DATE1 STATUS NRC APPROVAL

SUMMARY

3 13R-01 1

Authorized (ISI & CISI) Synchronization of Ten-Year ISI Intervals between Unit 1 09/12/05 and Unit2for Class 1,2, and 3.

In addition, alignment of Containment Inservice Inspection (CISI) len-Year Intervals for Class MC and CC with the synchronized Unit 1 and 2 Ten-Year 151 Interval.

Authorized per NRC SER dated 09107106. [Note that the start and end dates for the Third ISI Interval and Second CISI Interval were aligned, as well as subsequent intervals per the wording in previous Third ISI Interval and Second CISI Interval Relief Request 13R-01 that was authorized by the NRC per SER dated September 7, 2006.

Therefore, a Fourth ISI Interval and Third CISI Interval relief request is not required. Previous Relief Request l3R-01 stated Relief is requested to modify the end dates of the Byron Station Unit 2 Second 151 Interval and of the Byron Station Units 1 and 2 First CISI Intervals and the start and end dates of all subsequent ISI and CISI Intervals for Byron Station Units 1 and 2. 13R-01 also stated that All inspection periods for Class 1, 2, 3, and MC components will commence for the next interval based on the modified common interval start date. Any examination methods unique to and specifically required in the third period under the previous interval, that will likewise be required in the next interval, will be scheduled and completed in the first period of the subsequent interval. The examinations will be conducted and credited under the rules of the new Code of Record (i.e., 2001 Edition through the 2003 Addenda of ASME Section XI). These examinations originally unique to the third period of the previous interval will henceforth be conducted in the first period of all subsequent ISI intervals, and deferral to the end of future intervals will not be available.

In addition, the rolling five-year IWL frequency applicable to Class CC components that are subject to Subsection IWL requirements will be maintained as currently scheduled.] Thus, the Byron Station Unit 2 end of interval 151 and CISI examinations will be conducted at the end of the first period of the Fourth ISI Interval using the 2007 Edition with the 2008 Addenda of ASME Section XI.

14R-01 0

Submitted (151) Alternate Risk-Informed Selection and Examination Criteria for 7/29/16 04/15/16 Examination Category B-F, B-J, C-F-i, and C-F-2 Pressure Retaining Piping Welds.

14R-02 0

On Hold (151) Repair of Control Rod Drive Mechanism (CRDM) Canopy Seal 7/29/16 Welds in Accordance with IWA-4000.

l4R-03 0

Cancelled (SPT) Alternative Examination Requirements of ASME Section XI, 7/29/16 IWA-5244, Buried Components. Cancelled by Byron Station.

l4R-04 0

On Hold (151) Alternative Requirements for Limited Examination of the Reactor 7/29/16 Vessel Head Penetration Welds.

14R-05 0

Submitted (lSl) Use of ASME Code Case N-789, Alternative Requirements for 7/29/16 04/i 5/16 Pad Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping for Raw Water Service.

14R-06 0

Submitted (ISI) Use of ASME Code Case N-786, Alternative Requirements for 7/29/16 04/15/16 Sleeve Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping.

Exelon Byron Station 8-2 Revision 0

IS! Program Plan Units 1 & 2, Fourth Interval TABLE 8.0-1 RELIEF REQUEST INDEX RELIEF REVISION 2

(PROGRAM) DESCRIPTION OF RELIEF REQUEST!

REQUEST DATE1 STATUS NRC APPROVAL

SUMMARY

3 14R-07 0

On Hold (1St) Request for Relief for Alternative Requirements for Pressure 7/29/16 Retaining Boundary During System Leakage Test.

14R-08 0

Submitted fISt) Alternative Requirements to Extend the Reactor Vessel Inservice 7/29/16 04/15/16 Inspection Interval.

14R-09 0

Submitted fISt) Use of ASME Code Case N-51 3-4, Evaluation Criteria for 7/29/16 01/28/16 Temporary Acceptance Flaws in Moderate Energy Class 2 or 3 Piping.

Submitted as an EGC Fleet relief request under RS-16-041.

14R-1 0 0

Drafted fiSt) Alternative Requirements for the Repair of the Reactor Vessel Head Penetrations In Accordance with 10 CFR 50.55a(a)(3)(i).

Note 1: The revision listed is the latest revision of the subject relief request. The date this revision became effective is the date of the approving SE that is listed in the fourth column of the table. The date noted in the second column is the date of the ISI Program Plan revision when the relief request was incorporated into the document.

Note 2: This column represents the status of the latest revision. Relief Request Status Options:

Authorized

- Approved for use in an NRC SE (See Note 3);

Granted

- Approved for use in an NRC SE (See Note 3);

Authorized Conditionally

- Approved for use in a NRC SE that imposes certain conditions; Granted Conditionally

- Approved for use in a NRC SE that imposes certain conditions; Denied

- Use denied in a NRC SE; Expired

- Approval for relief request has expired; Withdrawn

- Relief request has been withdrawn by the Byron Station; Not Required

- The NRC has deemed the relief request unnecessary in an SE or RAI; Cancelled

- Relief request has been cancelled by the Byron Station prior to issue; Drafted

- Drafted relief request awaiting submittal and/or pending approval; and Submitted

- Relief request has been submitted to the NRC by the Byron Station and is awaiting approval.

Note 3: The NRC grants relief requests pursuant to 10 CFR 50.55a(g)(6)(i) under paragraph 10 CFR 50.55a(g)(5)(iii) when ASME Section Xl requirements cannot be met and proposed alternatives do not meet the criteria of JO CFR 50.55a(z). The NRC authorizes relief requests pursuant to 10 CFR 50.55a(z)(1) if the proposed alternatives would provide an acceptable level of quality and safety or under 10 CFR 50.55a(z)(2) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety. Relief requests under 10 CFR 50.55a(g)(5)(iii) are not to be submitted to the NRC for evaluation prior to the licensee performing the ASME Section Xl-required examination. The NRC may also impose alternative requirements as it determines.

Exelon Byron Station 8-3 Revision 0

IS! Program Plan Units 1 & 2, Fourth Interval

9.0 REFERENCES

The references used to develop this ISI Program Plan include:

9.1 NRC References 9.1.1 Code of Federal Regulations, Title 10, Energy.

Part 50, Paragraph 50.55a, Codes and Standards.

Part 50, Paragraph 2, Definitions, the definition of Reactor Coolant Pressure Boundary.

Part 50, Appendix J, Primary Reactor Containment Testing for Water Cooled Power Reactors.

SECY-96-080, Issuance of Final Amendment to 70 CFR 50. 55a to Incorporate by Reference the ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, Subsections IWE and IWL.

9.1.2 NRC Mechanical Engineering Branch (MEB) Technical Position 3-1 (MEB 3-1), dated November 24, 1975, High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment.

9.1.3 Regulatory Guide 1.14, Revision 1, Reactor Coolant Pump Flywheel Integrity, as modified by the requirements of Byron Station License Amendment #118 and Technical Requirements Manual Appendix G.

9.1.4 NRC Regulatory Guide 1.26, Revision 3, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive Waste-Containing Components of Nuclear Power Plants.

9.1.5 NRC Regulatory Guide 1.137, Revision 1, Fuel-Oil Systems for Standby Diesel Generators.

9.1.6 NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section Xl, Division 1 (See NRC.gov Reading Room for the most current revision).

9.1.7 NRC Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code (See NRC.gov Reading Room for the most current revision).

9.1.8 NRC Regulatory Guide 1.193, ASME Code Cases Not Approved For Use (See NRC.gov Reading Room for the most current revision).

9.1.9 NRC NUREG 0737, dated November 1980, Clarification of IMI Action Plan Requirements.

9.1.10 NRC SER related to EPRI Topical Report TR-1 12657, Rev. B, Final Report, Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 7999, dated October 28, 1999.

9.1.11 NRC SER related to EPRI Topical Report TR-1 006937, Rev. 0, Extension of the EPRI Risk Informed Insetvice Inspection (Rl-lSl) Methodology to Break Exclusion Region (BER)

Programs, dated June 27, 2002.

9.2 INDUSTRY REFERENCES 9.2.1 ASME Boiler and Pressure Vessel Code, Section Xl, Division 1, lnservice Inspection of Nuclear Power Plant Components,

2007 Edition with the 2008 Addenda (including Appendix VIII) (4th 151 Interval and 3rd ClSl Interval),

2004 Edition with No Addenda (Nonmandatory Appendix Q only),

2001 Edition through the 2003 Addenda (3rd 151 Interval and 2 CISI Interval),

2001 Edition with No Addenda,

1998 Edition with No Addenda (Ist CISI Interval),

1995 Edition through the 1997 Addenda,

1995 Edition with the 1995 Addenda, Exelon Byron Station 9-1 Revision 0

IS! Firqram Plan Units I & 2, Fourth Interval

1992 Edition with the 1992 Addenda (iSt CISI Interval),

1989 Edition with No Addenda (2r1d ISI Interval),

1983 Edition with Addenda through the Summer 1983 Addenda (83/S83) (1st 151 Interval) Byron Station Unit 2, and

1980 Edition with Addenda through the Winter 1981 Addenda (801W81) (1st ISI Interval)

Byron Station Unit 1.

9.2.2 ASME Boiler and Pressure Vessel Code,Section V, Nondestructive Examination,

- 2007 Edition with the 2008 Addenda [The Edition and Addenda for ASME Section V are the same as the Edition and Addenda of ASME Section XI used for the inspection interval for both ISI and Non-ISI NDE examinations. Reference ASME Interpretation XI-1-89-02].

9.2.3 ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants,

- 2004 Edition through the 2006 Addenda (Subsections ISTA and ISTD). (Fourth Snubber Interval).

9.2.4 MRP-146, EPRI Materials Reliability Program, Management of Thermal Fatigue in Normally Stagnant Non-Isolable Reactor Coolant System Branch Lines, Revision 1, Report 1022564, June2011.

9.2.5 MRP-192, EPRI Materials Reliability Program, Assessment of RHR Mixing Tee Thermal Fatigue in PWR Plants, Revision 1, Report 1018395, November 2008.

9.2.6 EPRI Topical Report TR-1 12657, Rev. B-A, Final Report, Revised Risk-Informed Inseivice Inspection Evaluation Procedure, December 1999.

9.2.7 EPRI Topical Report TR-1006937, Rev. 0-A, Extension of the EPRI Risk-Informed Insenjice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs, August 2002.

9.2.8 EPRI Containment Inspection Program Guide (TR-110698-R1).

9.2.9 INPO Engineering Program Guide EPG-11, Inservice Inspection Program.

9.3 LICENSEE REFERENCES 9.3.1 Byron Station Units 1 and 2, Updated Final Safety Analysis Report (UFSAR).

9.3.2 Byron Station Units I and 2, Technical Specifications, Limiting Conditions for Operation and Suiveillance Requirements.

9.3.3 Byron Station Units 1 and 2, Technical Specifications (TS), Bases.

9.3.4 Byron Station Units 1 and 2, Technical Requirements Manual (TRM).

9.3.5 Byron Station Units 1 and 2, ISI Classification Basis Document (BYR-525537-RPO3), Fourth Ten-Year Inspection Interval.

9.3.6 Byron Station Units 1 and 2, 151 Selection Document (BYR-525537-RPQ5), Fourth Ten-Year Inspection Interval.

9.3.7 Byron Station Units 1 and 2, Exelon Risk-Informed Inservice Inspection Evaluation (Final Report) (BYR-525537-RPO6), Fourth Ten-Year Inspection Interval.

9.3.8 Byron Station Units 1 and 2, Snubber Program Document (BYR-525537-RPO7), Fourth Ten-Year Inspection Interval.

9.3.9 Procedures ER-AA-330, Conduct of Insenjice Inspection Activities, ER-AA-330-001, Section Xl Pressure Testing, ER-AA-330-002, Inservice Inspection of Section Xl Welds and Components, ER-AA-330-003, Inservice Inspection of Section Xl Component Supports, ER AA-330-004, Visual Examination of Snubbers, ER-AA-330-005, Visual Examination of Section Xl Class CC Concrete Containment Structures, ER-AA-330-006, !nseivice Exelon Byron Station 9-2 Revision 0

IS! Program Plan Units 7 & 2, Fourth Interval Inspection and Testing of The Pre-Stressed Concrete Containment Post Tensioning Systems, ER-AA-330-007, Visual Examination of Section Xl Class MC Surfaces and Class CC Liners, ER-AA-330-009, ASME Section Xl Repair/Replacement Program, ER-AA-330-010, Snubber Functional Testing, and ER-AA-330-01 1, Snubber Service Life Monitoring Program.

9.3.10 NRC letter dated May 17, 1990, Stephen P. Sands, NRC to Thomas J. Kovach, Commonwealth Edison Company - Safety Evaluation of Containment Leak Chase Channels, Byron Station Unit Nos. 1 and 2, Braidwood Station Unit Nos. 1 and 2.

9.3.11 Calculation to Determine 80% of Primary Containment IWE/MC Surface Area Remains Accessible for Examination, BYR2000-181, for Byron Station, Units 1 and 2.

9.3.12 Unit 1 Steam Generator Lower Shell-to-Transition Cone Weld Exemption. Babcock &

Wilcox, Canada Engineering Evaluation CM9015189

- B2, Exelon Generation Company RSG

- Shell Circumferential Weld Evaluation With Respect To Section Xl Rules. Exelon Generation Company Owners Review of this evaluation was performed under EC 354211.

9.3.13 Curtiss-Wright Correspondence CPS-14-014: Byron Penetrant Test Evaluation (Re. IR 02393595).

9.3.14 Byron Station Units 1 and 2, License Renewal Application, July 2015.

9.3.15 NRC SE related to Relief Request 13R-01, Synchronization of Ten-Year ISI Intervals between Unit 1 and Unit 2 for Class 1, 2, and 3.

In addition, alignment of Containment Inservice Inspection (CISI) Ten-Year Intervals for Class MC and CC with the synchronized Unit 1 and 2 Ten-Year 151 Interval, authorized per NRC SER dated 09/07/06.

9.3.16 BY-PRA-031, Rev. 0, Byron Nuclear Generating Station Units 1 and 2, PRA Capability Assessment for Rl-lSl, (Summary: PRA Capability Assessment for Risk-Informed Inservice Inspection Applications), dated December 2015.

9.3.17 Reactor Flange Leak Off Lines, BYRON-98-5030, dated 02/04/98.

9.3.18 Acceptance Criteria for Containment Liner Reduced Thickness, NDIT No. BYR-88-226 dated 10/27/99.

9.4 LICENSE RENEWAL REFERENCES / COMMITMENTS 9.4.1 CM-I AR 01367499-29-05, License Renewal Aging Management-ASME Section Xl, Subsection IWE program. (Section 1.9) 9.4.2 CM-2 AR 01 367499-30-01, License Renewal Aging Management-ASME Section XI, Subsection IWL program. (Section 1.9) 9.4.3 CM-3 AR 01367499-31-07, License Renewal Aging Management-ASME Section XI, Subsection IWF program. (Sections 1.9, 2.2.15, 4.0, 4.1, 4.1.1, and 4.3)

Exelon Byron Station 9-3 Revision 0