ML092330858

Submittal of Fourth Ten-Year Interval Inservice Inspection Program Plan

ML092330858
Person / Time
Site:	Peach Bottom
Issue date:	08/19/2009
From:	David Helker Exelon Generation Co, Exelon Nuclear
To:	Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML092330858 (168)

Text

{{#Wiki_filter:Exelon Nuclear 200 Exelon Way Kennett Square, PA 19348 August 19,2009 www.exeloncorp.com Nuclear 10CFR50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Peach Bottom Atomic Power Station, Units 2 and 3 Renewed Facility Operating License Nos. DPR-44 and DPR-56 NRC Docket Nos. 50-277 and 50-278

Subject:

Submittal of Fourth Ten-Year Intervallnservice Inspection Program Plan In accordance with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, IWA-1400(c) ("Owner's Responsibility"), attached for your information is a copy of the Fourth Ten-Year Intervallnservice Inspection (lSI) Program Plan for Peach Bottom Atomic Power Station, Units 2 and 3. The new interval began on November 5, 2008, and will conclude on November 4, 2018. This plan complies with the ASME Boiler and Pressure Vessel Code, Section XI, 2001 Edition through 2003 Addenda. There are no regulatory commitments contained within this letter. If you have any questions or require additional information, please call Tom Loomis (610-765-5510). Respectfully, David P. Helker Manager - Licensing & Regulatory Affairs Exelon Generation Company, LLC

Attachment:

lSI Program Plan cc: USNRC Region I, Regional Administrator USNRC Senior Resident Inspector, PBAPS USNRC Project Manager, PBAPS R. R. Janati, Bureau of Radiation Protection S. T. Gray, State of Maryland

ATTACHMENT lSI Program Plan

Peach Bottom Atomic Power Station lSI Program Exelon" Nuclear Document No.: PBT05.G03 PEACH BOTTOM ATOMIC POWER STATION UNITS 2 &3 lSI PROGRAM PLAN FOURTH TEN-YEAR INSPECTION INTERVAL Commercial Service Dates: Unit 2 - 07/05/74 Unit 3 - 12/23/74 Peach Bottom Atomic Power Station 1848 Lay Road Delta, PA. 17314 Exelon Generation Company (EGC), LLC 300 Exelon Way Kennett Square, PA. 19348 Prepared By: Alion Science and Technology Corporation Engineering and Technical Programs Division Warrenville, Illinois* A L ION SCllN(E AND HCHNOt.OGY

lSI Program Plan Pelleh Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval REVISION APPROVAL SHEET TITLE: DOCUMENT: lSI Program Plan FOUlih Ten-Year Inspection Interval Peach Bottom Atomic Power Station, Units 2 & 3 PBT05.G03 REVISION: o PREPARED TRANSMITTAL PREPARED: REVIEWED:'---~:5:=;;J / t(I../lr ~ W. Windhorst tl 4 Alion Project Engineer APPROVED: ~~~ Alion Project Manager EXELON ACCEPTANCE APPROVED: 1i'::1A4/:Y,Jq1L: lSI Program Engineer APPROVED: ~L-r~ /~o~ Samantha L. Stahl CISI Program Engineer Alioll Sciellce & Tecllllology i PBTOS.G03 Revisioll 0

lSI Program Pltm Peach Bol1om Atomic Power Station Units 2 & 3, FOllrth Interwll REVISION APPROVAL SHEET TITLE: DOCUMENT: lSI Program Plan Fourth Ten-Year Inspection Interval Peach Bottoln AtOlnic Power Station, Units 2 & 3 PBT05.G03 REVISION: o EXELON PROGRAM ACCEPTANCE REVIEWED: REVIEWED: APPROVED: REVIEWED: ~L~/~\\)~ Samantha L. Stahl CISI Program Engineer ~!~.,;,~/~ Engineering Programs Manager --I\\~~""-=,--,,,,-=-.:::,--- I~ 0~ e Fuhrman T Authorized Nuclear Inservice Inspector Each tin1e this docllInent is revised, the Revision Approval Sheet will be signed and the following Revision Control Sheet should be con1pleted to provide a detailed record of the revision history. The signatures above apply only to the changes n1ade in the revision noted. If historical signatures are required, Peach Bottom Aton1ic Power Station archives should be retrieved. Alion Science & Tecllll%gy ii PBT05.G03 Revisioll ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval REVISION CONTROL SHEET Major changes to this document should be outlined within the table below. Editorial and formatting revisions are not required to be logged. Revision Date Revision Summary 0 11/05/08 Initial issuance. (This ISI Program Plan was developed by Alion Science and Technology Corporation as part of the Fourth Interval lSI Program update.) Prepared: S. Coleman Reviewed: D. Windhorst Approved: D. Lamond Notes: 1. This lSI Program Plan (Sections 1 - 9 inclusive) is controlled by the Peach Bottom Atomic Power Station, Engineering Programs Group. 2. Revision 0 of this document was issued as the Fourth Interval lSI Program Plan and was submitted to the USNRC for review, including approval of the initial Fourth lSI Interval and Second CISI Interval Relief Requests. Future revisions of this document made within the Fourth 1St Interval will be maintained and controlled at Peach Bottom Atomic Power Station; however, they are not required to be and will not be submitted to the USNRC for review. The exception to this is that new or revised Relief Requests shall be submitted to the USNRC for safety evaluation and approval. 3. This Augmented Inspection Plan is controlled by the Peach Bottom Atomic Power Station Engineering Programs Group as a Nuclear Record in accordance with procedure RM-AA-101-1008, Processing and Storage of Records. 4. The SRRS record ill is 3A.117, and the Title is "Fourth Interval lSI Program Plan" 5. Passport IR 873610 has been initiated to track "Pen and Ink" changes to this document as permitted by procedure ER-AA-330, Conduct of In-Service Inspection Activities. 6. Passport assignment 873610-01 has been initiated to document the yearly revision assessment of this document (reference ER-AA-330, section 4.4.3). Alion Science & Technology iii PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Stiltioll Units 2 & 3, Fourth Illtervlli REVISION

SUMMARY

Section Effective Pages Revision Date Preface i to vii 0 11/05/08 1.0 1-1 to 1-17 0 11/05/08 2.0 2-1 to 2-37 0 11/05/08 3.0 3-1 to 3-3 0 11/05/08 4.0 4-1 to 4-2 0 11/05/08 5.0 5-1 to 5-2 0 11/05/08 6.0 6-1 to 6-5 0 11/05/08 7.0 7-1 to 7-36 0 11/05/08 8.0 8-1 to 8-3 0 11/05/08 9.0 9-1 to 9-3 0 11/05/08 Alioll Science & Technology iv PBT05.G03 Revision ()

lSI Program Plan Pellch Bottom Atomic Power Statiou Uuits 2 & 3, 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 lSI Progranl 1.4 Second Interval lSI Program 1.5 Third Interval lSI Progratn 1.6 Fourth Interval lSI Progratn 1.7 First Interval CISI Progranl 1.8 Second Interval CISI Progranl 1.9 Code of Federal Regulations 10CFR50.55a Reguirelnents 1.1a Code Cases 1.11 ReIief Reguests 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2-1 2.1 ASME Section XI Exatnination Reguirenlents 2.1.1 ASME Section XI Code Cases 2.1.2 OM Code Cases 2.2 Augmented Inspection Plan Requirements 2.3 System Classifications and P&ID Boundary Drawings 2.4 lSI Isonletric and COlnponent Drawings for Nonexelnpt lSI Class Conlponents and Supports 2.5 Technical Approach and Positions 3.0 COMPONENT lSI PLAN 3-1 3.1 Nonexenlpt lSI Class Components 3.2 Risk-Infornled Exalnination Reguiren1ents 3.3 Reactor Coolant Pressure Boundary Normal Makeup Exemption 4.0 SUPPORT lSI I>LAN 4-1 4.1 Nonexempt lSI Class Supports 4.2 Snubber Examination and Testing Requiren1ents 5.0 SYSTEM PRESSURE TESTING lSI PLAN 5-1 5.1 lSI Class Systems 5.2 Risk-Informed Exalnination of Socket Welds 6.0 CONTAINMENT lSI PLAN 6-1 6.1 Nonexen1pt CISI Class COlnponents 6.2 Augn1ented Examination Areas 6.3 COlllponent Accessibility 6.4 Responsible Individual Alion Science & Technology v PBT05.G03 Revision ()

lSI Program Plall Peach Bottom Atomic Power Slatioll Units 2 & 3, Fourth Interval 7.0 COMPONENT

SUMMARY

TABLES 7-1 7.1 Inservice Inspection Sumn1ary Tables 7.2 Snubber Inspection Sumnlary Tables 8.0 RELIEF REQUESTS FROM ASME SECTION XI. 8-1

9.0 REFERENCES

9-1 Alioll Sciellce & Tecll1lology vi PBT05.G03 Revisioll ()

TABLES lSI Program PIa" Pellch Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE OF CONTENTS (Continued) DESCRIPTION PAGE 1.1-1 UNITS 2 & 3 FOURTH lSI INTERVAL/PERIOD/OUTAGE MATRIX (FOR lSI CLASS 1,2, AND 3 COMPONENT EXAMINATIONS) 1-2 1.1-2 UNITS 2 & 3 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISI CLASS MC COMPONENT EXAMINATIONS) 1-3 1.9-1 CODE OF FEDERAL REGULATIONS IOCFR50.55a REQUIREMENTS 1-9 2.3-1 SYSTEM DESIGNATORS 2-12 2.3-2 INSERVICE INSPECTION BOUNDARY DIAGRAMS 2-13 2.3-3 CONTAINMENT INSERVICE INSPECTION BOUNDARY DIAGRAl\\I1S......... 2-14 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC DRAWINGS 2-16 2.4-2 UNITS 2 AND 3 lSI COMPONENT DRAWINGS 2-21 2.4-3 ASME SECTION XI, lSI CALIBRATION BLOCK DRAWINGS 2-22 2.4-4 LIMERICK GENERATING STATION lSI CALIBRATIONBLOCKS USED AT PBAPS 2-25 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS 2-26 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX 2-35 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE 7-4 7.1-2 UNIT 3 INSERVICE INSPECTION SUl\\IIMARY TABLE 7-17 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES 7-30 7.2-1 UNIT 2 & COIVIMON SNUBBER INSPECTION SUMMARy 7-34 7.2-2 UNIT 3 SNUBBER INSPECTION

SUMMARY

TABLE 7-35 7.2-3 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES 7-36 8.0-1 RELIEF REQUEST INDEX 8-2 Alion Science & Teclmology vii PBT05.G03 Revision ()

lSI Program Piau Peach Bottom Atomic Power Station Units 2 & 3, FOllrth Interval

1.0 INTRODUCTION AND BACKGROUND

1.1 Introduction This Inservice Inspection (lSI) Program Plan details the requirenlents for the exanlination and testing of lSI Class 1, 2, 3, and MC pressure retaining components and supports at Peach Bottom Atomic Power Station (PBAPS) Units 2, 3, and 2/3 (COlnmon). Unit Comlnon conlponents are included in the Unit 2 sections, reports, and tables. This lSI Program Plan also includes Containnlent Inservice Inspection (CISI), Risk-Informed Inservice Inspections (RISI), Auglnented Inservice Inspections (AUG), and System Pressure Testing (SPT) requiretnents inlposed on or committed to by PBAPS. At PBAPS, the Inservice Testing (1ST) Progrmn is maintained and implemented separately fronl the lSI Program. The 1ST Basis Document and Progrmn Plan contain all of the applicable inservice testing requirements. The Fourth lSI Interval for PBAPS Units 2 and 3 is effective frOnl Novenlber 5, 2008 through Novenlber 4,2018. With the update to the lSI Program for the Fourth lSI Interval for lSI Class 1, 2, and 3 components, including their supports, Exelon Generation Company, LLC (Exelon) has also updated the CISI Program to its Second CISI Interval for CISI Class MC conlponents at the same tinle. This update will enable all of the lSI and CISI Progratn conlponents / elenlents to be based on the saIne effective Edition and Addenda of the Alnerican Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, as well as share a conlffion interval start and end date. The common ASME Code of Record for the Fourth lSI Interval and the Second CISI Interval is the 2001 Edition through the 2003 Addenda. This lSI Progratn Plan is controlled and revised in accordance with the requiretnents of procedure ER-AA-330, "Conduct of Inservice Inspection Activities," which implenlents the ASME Section XI lSI Program. Paragraph IWA-2430(d)(l) of ASME Section XI allows an inspection interval to be extended or decreased by as tnuch as one year, and Paragraph IWA-2430(e) allows an inspection interval to be extended when a unit is out of service continuously for six months or nlore. The extension nlay be taken for a period of time not to exceed the duration ofthe olltage. See Tables 1.1-1 and 1.1-2 for intervals, periods, and extensions that apply to PBAPS's Fourth lSI Interval and Second CIS I Interval. The Fourth lSI Interval and the Second CISI Interval are divided into three inspection periods as detennined by calendar years within the intervals. Tables 1.1-1 and 1. I-2 identify the period start and end dates for the Fourth lSI Interval and the Second CISI Interval as defined by Inspection Prograln B. In accordance with Paragraph IWA-2430(d)(3), the inspection periods specified in these Tables may be decreased or extended by as much as 1 year to enable inspection to coincide with PBAPS's refueling outages. Alion Science & Technology 1-1 PBT05.G03 Revision 0

lSI Program PlalZ Peach Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval TABLE 1.1-1 UNITS 2 & 3 FOURTH lSI INTERVAL/PERIODlOUTAGE MATRIX (FOR lSI CLASS 1,2, AND 3 COMPONENT EXAMINATIONS) Unit 2 Period Interval Period Unit 3 Outage Projected Outage Start Date to End Date Start Date to Start Date to End Date Projected Outage Outage Number Start Date or End Date Start Date or Number Outage Duration Outage Duration P2RI8 Scheduled 1st 1st Scheduled P3RI7 9/10 11/5/08 to 11/4/12 11/5/08 to 5/5/12 9/09 P2RI9 Scheduled 4th (Unit 2) Scheduled P3RI8 9/12 11/5/08 to 11/4/18 9/11 P2R20 Scheduled 2nd 4th (Unit 3) 2nd Scheduled P3RI9 9/14 11/5/12 to 11/4/152 1115/08 to 11/4/18' 5/6/12 to 5/5/152 9/13 P2R21 Scheduled 3ed 3ed Scheduled P3R20 9/16 11/5/15 to 11/4/18 5/6/15 to 11/4/18 9/15 P2R22 Scheduled Scheduled P3R21 9/18 9/17 = Note I: The PBAPS Unit 3 Third Period and Third lSI Interval were extended by 82 days per Third lSI Interval Relief Request I3R-45. Per this relief request, the extension is being carried forward to the Fourth lSI Interval to facilitate both PBAPS Units 2 and 3 having the same lSI Interval start and end dates, and codes of record, as well as this common interval date maching the CISI Program dates. As required by ASME Section XI, the intervals will be scheduled in ten-year increments from this point forward with the modificaions similar to Paragraph IWA-2430 fully available to future intervals and periods including a one-year extension allowance based on the new synchronized unit interval date Note 2: The PBAPS Units 2 and 3 Second Period was reduced by one year and the First Period was extended by one year as permitted by Paragraph IWA-2430(d)(3) in order to coincide with the plant refueling outage schedule. Alio" Science & Technology 1-2 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 1.1-2 UNITS 2 & 3 SECOND CISI INTERVALIPERIOD/OUTAGE MATRIX (FOR CISI CLASS MC COMPONENT EXAMINATIONS) Unit 2 Period Interval Period Unit 3 Outage Projected Outage Start Date to End Date Start Date to Start Date to End Date Projected Outage Outage Number Start Date or End Date Start Date or Number Outage Duration Outage Duration P2RI8 Scheduled 1st 1st Scheduled P3RI7 9/10 II/5/08 to II/4/12 11/5/08 to 5/5/12 9/09 P2RI9 Scheduled 2nd (Unit 2) Scheduled P3RI8 9/12 II/5/08 to 11/4/18 9/11 P2R20 Scheduled 2nd 2nd (Unit 3) 2nd Scheduled P3RI9 9/14 II/5/12 to lI/4/I5 2 11/5/08 to 11/4/181 5/6/12 to 5/5/152 9/13 P2R21 Scheduled 3rd 3rd Scheduled P3R20 9/16 IllS/IS to lI/4/18 5/6/15 to 11/4/18 9/15 P2R22 Scheduled Scheduled P3R21 9/18 9/17 Note I: The PBAPS Unit 3 Third Period and the First CISI Interval were extended by 365 days per First CISI Interval Relief Request CRR-12. Per this First CISI Interval relief request, the Second CISI Interval for PBAPS Unit 3 will overlap the duration of the First CISI Interval for one year in order to start the Second CISI Interval on time and keep it aligned with the Unit 2Second CISI Interval while still finishing examinations the First CISI Interval. It will also facilitate both PBAPS Units 2 and 3 having the sameCISI Interval start and end dates, and codes of record, as well as this common interval date matching the lSI Program dates. As required by ASME Section Xl, the intervals will be scheduled in ten-year increments from this point forward with the modifications similar to Paragraph IWA-2430 fully available to future intervals and periods including a one-year extension allowance based on the new synchronized unit interval date Note 2: The PBAPS Units 2 and 3 Second Period was reduced by one year and the First Period was extended by one year as permitted by Paragraph IWA-2430(d)(3) in order to coincide with the plant refueling outage schedule. Alion Science & Tee/urology 1-3 PBT05.G03 Revbiion 0

lSI Program Plan Peach Bottom Atomic Power Station Unit.s 2 & 3, Fourth Interval

1.2 Background

The Philadelphia Electric Con1pany (PECO) and forn1erly PECO Energy Con1pany, LLC., novv known con1mercially as Exelon Generation Conlpany, LLC. (Exelon), obtained a construction pern1it to build PBAPS Units 2 and 3 on January 31, 1968. The docket numbers assigned to PBAPS Units 2 and 3 are 50-277 and 50-278, respectively. After satisfactory plant construction and preoperational testing was cOlupleted, PBAPS was granted a full power operating license on October 25, 1973, DPR-44 for Unit 2 and on July 2, 1974, DPR-56 for Unit 3. PBAPS started con1nlercial operation on July 5, 1974 for Unit 2 and on December 23, 1974 for Unit 3. PBAPS's piping systen1s and associated components were designed and fabricated before the exan1ination requirenlents of ASME Section XI were formalized and published. Since this plant was not specifically designed to meet the requirements of ASME Section XI, literal compliance is not feasible or practical within the limits of the current plant design. Certain linlitations are likely to occur due to conditions such as accessibility, geometric configuration, and/or metallurgical characteristics. For son1e inspection categories, an alternate cOluponent n1ay be selected for exan1ination and the code statistical and distribution requiren1ents can still be ll1aintained. If ASME Section XI required exanlination criteria cannot be met, a relief request will be subn1itted in accordance with 10CFR50.55a. 1.3 First Interval lSI Progranl Pursuant to the Code Of Federal Regulations (CFR), Title 10, Part 50, Section 55a, Codes and standards, (l OCFR50.55a), PBAPS was required to meet the requireluents of Paragraph (g), lnservice inspection requirements, of that section. Specifically, Paragraph 10CFR50.55a(g)(4)(i) called for the inservice inspection requireluents of the initial 120 month inspection interval to conlply with the requireluents of the latest Edition and Addenda of ASME Section XI referenced in Paragraph (b) of 10CFR50.55a on the date twelve months prior the date of issuance of the operating license, subject to the liluitations and luodifications listed in 10CFR50.55a(b). PBAPS started conl111ercial operation on July 5, 1974 for Unit 2 and on Decen1ber 23, 1974 for Unit 3, which marked the beginning of the First lSI Interval. The version of 10CFR50.55a in effect twelve lTIonths prior to this date referenced the 1970 Edition with Addenda through the Winter 1970 (70W70) of the ASNIE Section XI. The inservice inspection requirements applicable to nondestructive exanlination and systenl pressure testing for the First Interval lSI Program were based on these rules. The First Period of the First lSI Interval was based on 70W70. Selection of components subject to examination for the First Period was based on an owner Alion Science & Tecllllo[ogy 1-4 PBT05.G03 Revision ()

IS} Program Plall Peach Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval established "focused approach" program. 10CFR50.55a Federal regulations later required an update of ASME Section XI, lSI Progranl to cOluply with the 1974 Edition with Addenda through the Smuluer 1975 (74S75) for the remaining two periods of the First lSI Interval for PBAPS Units 2 and 3, respectively. The PBAPS First lSI Interval was originally effective fronl July 5, 1974 to July 4, 1984 for Unit 2 and fron1 Decen1ber 23, 1974 to Decen1ber 22, 1984 for Unit 3. ASME Section XI pernlitted a one year interval extension to allow for outage correlation and an additional extension equivalent to the length of the outage. Also, plants which are out of service continuously for one year or luorc luay extend the lSI Interval for an equivalent period. As such, the First lSI Interval was extended to Septenlber 18, 1986 for Unit 2 and to Decenlber 22, 1985 for Unit 3 based on plant out of service and/or to coincide with the scheduling or cOlupletion of outages. Therefore, the PBAPS First lSI Interval was effective frOln July 5, 1974 to Septenlber 18, 1986 for Unit 2 and from December 23, 1974 to Decen1ber 22, 1985 for Unit 3. 1.4 Second Interval lSI Program Pursuant to 10CFR50.55a(g), PBAPS was required to update the lSI Program at the end of the First lSI Interval. The lSI Program was required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.5 5a twelve months prior to the start of the Second Interval per 10CFR50.55a(g)(4)(ii). The Second Interval lSI Progran1 Plan was developed in accordance with the requirements of 10CFR50.55a and the 1980 Edition through the Winter 1981 Addenda (80W81) ofASME Section XI. The Second Interval lSI Prograrn Plan addressed Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI, and utilized Inspection Prograln B as defined therein. The PBAPS Second lSI Interval started on Septeluber 19, 1986 for Unit 2 and on December 23, 1985 for Unit 3. Due to extended outages and plant out of service, the PBAPS Units 2 and 3 Second lSI Interval end dates were extended to Novembel' 4, 1998 and August 14, 1998, in accordance with 80W81 of ASME Section XI, Paragraph IWA-2400(c). Therefore, the PBAPS Second lSI Interval was effective frOlu Septen1ber 19, 1986 to Novenlber 4, 1998 for Unit 2 and frorn Decernber 23, 1985 to August 14, 1998 for Unit 3. Alion Science & Technology 1-5 PBT05.G()3 Revisioll ()

IS} ProgNlI1l Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval 1.5 Third Interval lSI Progran1 Pursuant to 10CFR50.55a(g), PBAPS was required to update the lSI Program to n1eet the requirements of ASME Section XI once every ten years or inspection interval. The lSI Progran1 was required to COl11ply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a twelve n10nths prior to the start of the Third lSI Interval per 10CFR50.55a(g)(4)(ii). The PBAPS Third lSI Interval started on Novel11ber 5, 1998 for Unit 2 and on September 15, 1998 for Unit 3 using the 1989 Edition, No Addenda of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI. l'he PBAPS Third Interval lSI Progran1 Plan addressed Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SER's, and utilized Inspection Program B as defined therein. At the end ofthe Third lSI Interval and in preparation for the Fourth lSI Interval, PBAPS sent Relief Request 13R-45 to the United States Nuclear Regulatory Con1mission (USNRC) to request an extension in order to synchronize the Fourth lSI Interval between PBAPS Units 2 and 3 for lSI Classes 1, 2, and 3, and the Second CISI Interval for CISI Class MC. This relief request was approved by the USNRC and the changes assured that both the lSI and CISI Programs would use the sanle Code Edition and Addenda for the next and successive intervals and would likewise establish comluon in1pleluenting procedures for both units. Therefore, the PBAPS Third lSI Interval was effective frolu Novelnber 5, 1998 through Novenlber 4, 2008 for Unit 2 and frol11 August 15, 1998 through November 4,2008 for Unit 3. 1.6 Fourth Interval lSI Progranl Pursuant to 10CFR50.55a(g), licensees are required to update their lSI Progratus to meet the requirelnents of ASME Section XI once every ten years or inspection interval. The lSI Progratu is required to con1ply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a twelve n10nths prior to the sta11 of the interval per 10CFR50.55a(g)(4)(ii). As discussed in Section 1.5 above, the start of the Fourth lSI Interval will be on Noven1ber 5, 2008 for PBAPS Units 2 and 3, respectively. Based on this date, the latest Edition and Addenda of ASME Section XI referenced in 10CFR50.55a(b)(2) twelve n10nths prior was the 2001 Edition through the 2003 Addenda. The PBAPS Fourth Interval lSI Progran1 Plan was developed in accordance with the requirements of 10CFRSO.S5a including all published changes through February 19. 2006, and the 2001 Edition through the 2003 Addenda of ASl\\1E Section XI, subject to the linlitations and 1110diftcations contained within Paragraph (b) of the regulation. These liluitations and n10diftcations are detailed in Table 1.9-1 of this section. This lSI PrograI11 Plan addresses Subsections IWA, Alion Science & Technology 1-6 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, FOllrth Interval IWB, IWC, IWD, IWF, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SER's, and utilizes Inspection Progratn B as defined therein. PBAPS adopted the EPRI Topical Report TR-112657, Rev. B-A n1ethodology, which was supplemented by Code Case N-578-1, for implementing risk-infonned inservice inspections. The RISI Program will be in effect for the entire Fourth Inspection Interval. This approach replaces the categorization, selection, and examination volUlne requirements of ASME Section XI Exanlination Categories B-F, B-J, C-F-l, and C-F-2 applicable to PBAPS with Examination Category R-A as defined in Code Case N-578-1. Implelnentation of RISI Progranl is in accordance with Relief Request 14R-44. The PBAPS Fourth lSI Interval is effective from Novenlber 5, 2008 through November 4, 2018 for Units 2 and 3, respectively. 1.7 First Interval CISI Program CISI examinations were originally invoked by anlended regulations contained within a Final Rule issued by the USNRC. The amended regulation incorporated the requirell1ents of the 1992 Edition through the 1992 Addenda of the ASME Section XI, Subsection IWE, subject to specific n10difications that were included in Paragraphs 10CFR50.55a(b)(2)(ix) and 10CFR50.55a(b)(2)(x). The final rulemaking was published in the Federal Register on August 8, 1996 and specified an effective date of Septenlber 9, 1996. Implelnentation of the Subsection 1WE Program from a scheduling standpoint was driven by the five year expedited implementation period per 10CFR50.55a(g)(6)(ii)(B), \\vhich specified that the examinations required to be c0111pleted by the end of the first period of the first inspection interval (per Table IWE-2412-1) be completed by the effective date (by September 9, 2001). ASME Section XI Subsections IWE, Mandatory Appendices, approved ASME Code Cases, and approved alternatives through relief requests and SER's \\vere added to the lSI Prograln at beginning of the Third lSI Interval to address CISI. The First CISI Interval for the PBAPS Units 2 and 3 CISI Progranl was aligned with the PBAPS Units 2 and 3 lSI Program and was effective fron1 Noven1ber 5, 1998 through Noven1ber 4, 2008, respectively. In accordance \\vith Relief Request CRR-12, the First CISI Interval for Unit 3 was extended by 365 days (one-year) to enCOlnpass Outage P3R17 and allow tinle to conlplete the remaining VT-3 visual exanlinations for the sublnerged portion of the Unit 3 Torus vent header downcOlners and pressure boundary structural attachment welds. These exanlinations were not completed in Outage P3R16 due to poor water clarity and will be perfornled in Outage P3R17 and credited under the requiretnents of ASME Section XI, 1992 Edition through the 1992 Addenda. Per this First CISI Interval relief request, the Second CISI Interval for PBAPS Unit 3 overlapped the duration A!iol1 Science & TecllJlology 1-7 PBT05.G03 Revision ()

lSI Progrmll Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval of the First CISI Interval for one year in order to start the Second CISI Interval on time and keep it aligned with the Unit 2 Second CISI Interval while still finishing exanlinations the First CISI Interval. These sanle exm11inations were scheduled again at the end of the Second CISI Interval per the Second CISI Inten,al Code of record to 1uaintain the original sequence and frequency of examination. As such, all of the PBAPS Units 2 and 3 First CISI Interval examinations were scheduled to be completed prior to November 5, 2009. As detailed in the submittal of the Fourth Interval lSI Progrmn, the transition from the First CISI Interval to the Second CISI Interval coincides with the transition fronl the Third lSI Interval to the Fourth lSI Interval to provide a common interval start and end date and Code of record between the lSI and CISI Programs. 1.8 Second Interval CISI Progrmu Pursuant to 10CFR50.55a(g), licensees are required to update their CISI Progranls to meet the requirements of ASME Section XI once every ten years or inspection interval. The CISI Progran1 is required to comply with the latest Edition and Addenda of the ASME Section XI incorporated by reference in 10CFR50.55a twelve n10nths prior to the start ofthe interval per 10CFR50.55a(g)(4)(ii). As discussed in Section 1.7 above, the start of the Second CISI Interval will be on November 5, 2008 for PBAPS Units 2 and 3. Based on this date, the latest Edition and Addenda of the ASME Section XI referenced in 10CFR50.55a(b)(2) twelve 1110nths prior was the 2001 Edition through the 2003 Addenda. The PBAPS Second CISI Interval is effective from Novenlber 5, 2008 through Novenlber 4,2018 for Units 2 and 3, respectively. The PBAPS Second Interval CISI Program Plan was developed in accordance with the requirements of 10CFR50.55a and the 2001 Edition through the 2003 Addenda of ASME Section XI, subject to the li1nitations and 111odifications contained within Paragraph (b) of the regulation. These linlitations and nl0difications are detailed in Table 1.9-1 of this section. This CI SI Program Plan addresses Subsections IWE, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SER's, and utilizes Inspection Progranl B as defined therein. 1.9 Code of Federal Regulations 10CFR50.55a Requirenlents There are certain paragraphs in 10CFR50.55a that list the linlitations, nl0difications, and/or clarifications to the implenlentation requirenlents of ASME Section XI. These paragraphs in IOCFR50.55a that are applicable to the PBAPS scheduled lSI and CISI examination progranls are detailed in Table 1.9-1. Alion Science & Technology 1-8 PBT05.G03 Revision 0

lSI Program Plan Peacl, Bot/om Atomic Power Station Vnits 2 & 3, FOllrth !nterl'{ll TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFRSO.SSa REQUIRElVlENTS 10CFRSO.S5a Paragraphs Limitations, Modifications, and Clarifications (Note: The words "this section" ill this column refer to the IOCFR paragraph that is the source of the Limitation, Modith'lltioll, or Clarification) 10CFR50.55a(b)(2)(ix)(A) (CISI) Examination ofmetal containments and the liners of concrete containments: For Class MC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the lSI Stllnlnary Report as required by IWA-6000: (1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; (3) A description of necessary corrective actions. 10CFR50.55a(b)(2)(ix)(B) (CISI) Examination ofmetal containments and the liners of concrete containments: When perfornling remotely the visual exanlinations required by Subsection IWE, the tnaxin1lun direct exatnination distance specified in Table IWA-221 0-1 may be extended and the nlininlum illumination requirements specified in Table IWA-221 0-1 may be decreased provided that the conditions or indications for which the visual exanlination is perfornled can be detected at the chosen distance and illl1l11ination. 10CFR50.55a(b)(2)(ix)(F) (CISI) Examination ofmetal containments and the liners of concrete containments: VT-l and VT-3 examinations nlust be conducted in accordance with IWA-2200. Personnel conducting examinations in accordance with the VT-I or VT-3 exalnination method shall be qualified in accordance with IWA-2300, The "owner-defined" personnel qualification provisions in IWE-2330(a) for personnel that conduct VT-l and VT-3 exanlinations are not approved for use. Alion Science & Technology 1-9 PBT05.G03 Revision 0

lSI Progrfl111 Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications (Note: The wOl'ds "this scction" in this column rcfcr to the IOCFR paragraph that is the SOlJl"CC of the Limitation, Modification, or Clarification) 10CFR50.55a(b)(2)(ix)(G) (CISI) Examination ofmetal containments and the liners of concrete containments: The VT-3 exanlination method lnust be used to conduct the exalninations in HeIns E 1.12 and E1.20 of Table IWE-2500-1, and the VT-l examination nlethod must be used to conduct the exanlination in Iteln £4.11 of Table IWE-2500-1. An exanlination of the pressure-retaining bolted connections in Itenl E1.11 of Table IWE-2500-1 using the VT-3 examination nlethod must be conducted once each intervaL The "owner-defined" visual examination provisions in IWE-23 IO(a) are not approved for use for VT-I and VT-3 exanlinations. 10CFR50.55a(b)(2)(ix)(H) (CISI) Examination ofmetal containments and the liners of concrete containments: Containment bolted connections that are disassembled during the scheduled perfofll1ance of the eXaininations in Item E 1.11 of Table IWE-2500-1 nlust be examined using the VT-3 exmnination method. Flaws or degradation identified during the perfonnance of a VT-3 exanlination nlust be examined in accordance with the VT-l examination nlethod. The criteria in the nlaterial specification or IWB-3517.1 must be used to evaluate containnlent bolting flaws or degradation. As an alternative to perfornling VT-3 exaluinations of containnlent bolted connections that are disassenlbled during the scheduled perfornlance of Itenl E1.11, VT-3 examinations of containnlent bolted connections may be conducted whenever containnlent bolted connections are disassenlbled for any reason. 10CFR50.55a(b)(2)(xii) (lSI) Underwater vVelding: The provisions in IWA-4660, "Underwater Welding," of Section XI, 1997 Addenda through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section, are not approved for use on irradiated nlateriaL 10CFR50.55a(b)(2)(xviii)(A) (lSI) Certification ofNDE personnel: Level I and II nondestructive exalnination persolulel shall 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 (b)(2) of this section. Alion Science & Technology /-/0 PBT05.G03 Revision ()

lSI Progrllll1 Plan Peach Bol1oll1 Atomic Power Station Units 2 & 3, FOllrth Interval TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFR50.SSa REQUIRElVIENTS 10CFRSO.55a Paragraphs LiInitafions, Modifications, and Clarifications (Note: The words "this sectioll" in this columll refer to the IUCFR paragraph that is the source of the Limitation, Modification, or Clarification) 10CFR50.55a(b)(2)(xviii)(B) (lSI) Certification ofNDE personnel: Paragraph IWA-2316 of the 1998 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section, may only be used to qualify personnel that observe for leakage during system leakage and hydrostatic tests conducted in accordance with IWA-5211 (a) and (b), 1998 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section. 10CFR50.55a(b)(2)(xviii)(C) (lSI) Certification ofNDE personnel: When qualifying visual exanlination personnel for VT-3 visual exalninations under paragraph IWA-2317 of the 1998 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section, the proficiency of the training nlust be delnonstrated by adnlinistering an initial qualification exanlination and administering subsequent exalninations on a 3-year interval. 10CFR50.55a(b)(2)(xix) (lSI) Substitution ofalternative methods: The provisions for the substitution of alternative examination nlethods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 nlust be applied. The provisions in IWA-2240, 1998 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section, are not approved for use. The provisions in IWA-4520(c), 1997 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) ofthis section, allowing the substitution of alternative exatnination tnethods, a combination of methods, or newly developed techniques for the tnethods specified in the Construction Code are not approved for use. 10CFR50.55a(b)(2)(xx)(B) (lSI) System leakage tests: The NDE provision in IWA-4540(a)(2) of the 2002 Addenda of Section XI lnust be applied when perfornling systeln leakage tests after repair and replacement activities perfornled by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section. Alion Science & Technology 1-11 PBT05.G03 Rel';sioll ()

lSI Program Pla/l Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIRElVlENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications (Note: The words "this section" in this column refer to the IOCFR paragraph that is the source of the Limitation, Modification, or Clarification) 10CFR50.55a(b)(2)(xxi)(B) (lSI) Table I~VB-2500-1 examination requirements: The provisions of Table IWB-2500-1, Exanlination Category B-G-2, Itenl B7.80, that are in the 1995 Edition are applicable only to reused bolting when using the 1997 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section. 10CFR50.55a(b)(2)(xxii) (lSI) 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 (b)(2) of this section, that allow use of an ultrasonic exanlination nlethod is prohibited. 10CFR50.55a(b)(2)(xxiii) (lSI) Evaluation ofThermally Cut Sw1aces: The use ofthe provisions for elilninating nlechanical processing of thennally cut surfaces in IWA 4461.4.2 of Section XI, 2001 Edition through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section are prohibited. 10CFR50.55a(b)(2)(xxiv) (lSI) Incorporation ofthe Performance Demonstration Initiative and Addition ofUltrasonic Examination Criteria. The use of Appendix VIII and the supplements to Appendix VIII and Article 1-3000 of Section XI of the ASME BPY Code, 2002 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, is prohibited. 10CFR50.55a(b)(2)(xxv) (lSI) Mitigation ofDefects by A10dijication: The use of the provisions in IWA-4340, "Mitigation of Defects by Modification," Section XL 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section are prohibited. 10CFR50.55a(b)(2)(xxvi) (lSI) Pressure Testing Class 1, 2, and 3 Mechanical Joints. The repair and replacenlent activity provisions in IWA-4540(c) of the 1998 Edition of Section XI for pressure testing Class 1, 2, and 3 lnechanical joints nlllst be applied when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section. A!iO/l Science & Technology 1-12 PBT05.G03 Revision 0

lSI Prognl111 Plan Peach Bottom Atomic Power Statioll Units 2 & 3, FOllrth Interval TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications (Note: The words "this section" in this column refer to the IOCFI{ paragraph that is the soul'ce of the Limitation, Modification, or Clarification) 10CFR50.55a(b)(2)(xxvii) (lSI) Removal ofInsulation: When performing visual exalninations in accordance with IWA-5242 of Section XI, 2003 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of the section, insulation must be removed from 17-4 PH or 410 stainless steel studs or bolts aged at a telnperature below 1100 OF 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. 10CFR50.55a(b)(3)(v) (lSI) Subsection ISTD. Article IWF-5000, "Inservice Inspection Requirenlents for Snubbers," of the ASME BPV Code, Section XI, provides inservice inspection requirenlents for examinations and tests of snubbers at nuclear power plants. Licensees may use Subsection ISTD, "Inservice Testing of Dynanlic Restraints (Snubbers) in Light-Water Reactor Power Plants," ASME OM Code, 1995 Edition through the latest edition and addenda incorporated by reference in paragraph (b)(3) of this section, in place of the requirenlents for snubbers in Section XI, IWF-5200(a) and (b) and IWF-5300(a) and (b), by nlaking appropriate changes to their technical specifications or licensee-controlled documents. Preservice and inservice exanlinations must be perfonned using the VT-3 visual exmnination lnethod described in IWA-2213. Alion Science & Technology 1-13 PBT05.G03 Revisioll 0

lSI Program Plan Peach Botton, Atomic Power Station Units 2 & 3, Fourth Interval TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFRSO.SSa REQUIREMENTS 10CFR50.S5a Paragraphs Limitations, Modifications, and Clarifications (Note: The wOI'ds "this section" in this colullln refer to the IOCFR paragraph that is the source of the Limitation, Modification, or Clarification) 10CFR50,55a(b)(5) (lSI) fnservice Inspection Code Cases: Licensees may apply the ASME Boiler and Pressure Vessel Code Cases listed in Regulatory Guide 1.147 through Revision 15, without prior USNRC approval subject to the following: (i) When a licensee initially applies a listed Code Case, the licensee shall apply the n10st recent version of that Code Case incorporated by reference in this paragraph. (ii) If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in this paragraph, 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 l11ay apply the later version of the Code Case, including any USNRC-specified conditions placed on its use. (iii) 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. Any Code Case listed as annulled in any Revision of Regulatory Guide 1.147 which a licensee has applied prior to it being listed as annulled, l11ay continue to be applied by that licensee to the end of the 120-n10nth interval in which the Code Case was inlplenlented. Alioll Science & Technology 1-14 PBT05.G03 Revision 0

[SI Progrtll1l PItt" Peach Bottol1l Atomic Power Statio/l Units 2 & 3, Fourth [Ilten'al TABLE 1.9-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREl"lENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications (Note: The words "this section" in this column refer to the IOCFR paragraph that is the source of the Limitation, M()dific~lti()n, or Clarification) 10CFR50.55a(b)(6) (lSI) Operation and Jvlaintenance o/Nuclear Power Plants Code Cases. Licensees may apply the ASME Operation and Maintenance Nuclear Power Plants Code Cases listed in Regulatory Guide 1.192 without prior USNRC approval subject to the following: (i) When a licensee initially applies a listed Code Case, the licensee shall apply the most recent version of that Code Case incorporated by reference in this paragraph. (ii) If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in this paragraph, the licensee may continue to apply, to the end of the current 120-n1011th interval, the previous version of the Code Case as authorized or may apply the later version ofthe Code Case, including any USNRC-specified conditions placed on its use. (iii) 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. If a licensee has applied a listed Code Case that is later listed as annulled in Regulatory Guide 1.192, the licensee lTIay continue to apply the Code Case to the end of the current 120-month interval. Alioll Sciellce & Technology 1-15 PBT05.G03 Revision 0

lSI Program Plan Peach Bollom Atomic Power Statioll Vllils 2 & 3, FOllrth Illterl',,1 1.10 Code Cases Per 10CFR50.55a(b)(5) and (b)(6), ASME Code Cases that have been determined to be suitable for use in lSI Program Plans by the USNRC are listed in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1". The approved Code Cases in Regulatory Guide 1.147, which are being utilized by PBAPS, are included in Section 2.1.1. The most recent version of a given Code Case incorporated in the revision of Regulatory Guide 1.147 referenced in 10CFR50.55a(b)(5)(i) at the tinle it is applied within the lSI Progratn shall be used. The latest version of Regulatory Guide 1.147 incorporated into this document is Revision 15. As this guide is revised, newly approved Code Cases should be assessed for plan implementation at PBAPS per Paragraph IWA-2441 (d) and proposed for use in mnendnlents to the lSI Progratn Plan. The use of Code Cases, other than those listed in Regulatory Guide 1.147 luay be authorized by the Director of the office ofNuclear Reactor Regulation upon request pursuant to 10CFR50.55a(a)(3). Code Cases not generically approved for use in Regulatory Guide 1.147, which are being utilized by PBAPS through associated relief requests, are included in Section 8.0. This lSI Progratu Plan will also utilize Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code". The approved Code Cases in Regulatory Guide 1.192, which are being utilized by PBAPS, are included in Section 2.1.2. The latest version of Regulatory Guide 1.192 incorporated into this docmnent is Revision O. As this guide is revised, newly approved Code Cases should be assessed for plan implementation at PBAPS per Paragraph IWA-2441 (d) and proposed for use in amendments to the lSI Progranl Plan. 1.11 Relief Requests In accordance with 10CFR50.55a, when a licensee either proposes alternatives to ASME Section XI requirements which provide an acceptable level of quality and safety, determines compliance with ASME Section XI requirenlents would result in hardship or unusual difficulty without a cOlnpensating increase in the level of quality and safety, or determines that specific ASME Section XI requirenlents for inservice inspection are iInpractical, the licensee shall notify the USNRC and SUblUit information to support the determination. The subluittal of this information will be referred to in this docunlent as a "Relief Request." Relief requests for the Fourth lSI Interval and 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 10CFR50.55a(a)(3)(i), compliance with the specified requireluents would result in hardship or unusual difficulty without a cOlupensating increase in the level of quality and safety per Alion Science & Techllology 1-16 PBT05.G03 Revisioll 0

lSI Program Plan Pellc" Bottom Atomic Power Station Units 2 & 3, Fourth Interval 10CFR50.5Sa(a)(3)(ii), or the code requirenlents are considered inlpractical per 10CFR50.5Sa(g)(5)(iii). Per 10CFRSO.5Sa Paragraphs (a)(3) and (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate relief requests and "n1ay grant such relief and may impose such alternative requireluents as it deternlines is authorized by law and will not endanger life or property or the COlnmon defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requireluents were imposed on the facility." Alion Science & Technology 1-17 PBT05.G03 Revision 0

lSI Progrttm Plan Peach Bottom Atomic Power Statioll Ullits 2 & 3, Fourth Illterval 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2.1 ASME Section XI Examination Requirements As required by the 10CFR50.55a, this Progratll was developed in accordance \\vith the requiren1ents detailed in the 2001 Edition through the 2003 Addenda, of the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Subsections IWA, IWB, IWC, IWD, IWE, IWF, Mandatory Appendices, Inspection Progranl B of Paragraph IWA-2432, approved ASME Code Cases, and approved alternatives through relief requests and SER's. The lSI Progratll impletnents Appendix VIII "Perfonnance Detllonstration for Ultrasonic Exan1ination Systems," ASME Section XI 2001 Edition, No Addenda as required by 10CFR50.55a(b)(2)(xxiv). Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in piping, bolting, and the reactor pressure vessel (RPV). Each organization (e.g., owner or vendor) is required to have a written progratll to ensure compliance with the requirelnents. PBAPS initially in1plemented these requirements by invoking the Perf0f111anCe Demonstration Initiative (PDI) Progran1 according to the schedule defined in 10CFR50.55a(g)(6)(ii)(C). For the Fourth lSI Interval, PBAPS's inspection progranl for ASME Section XI Exanlination Categories B-F, B-J, C-F-l, and C-F-2 will be governed by risk-infonned regulations. The RISI progranltnethodology is described in the EPRI Topical Report TR-112657, Rev. B-A. To supplenlent the EPRl Topical Report, Code Case N-578-1 (as applicable per Relief Request 14R-44) is also being used for the classification ofpiping structural elenlents under the RISI program. The RlSI progranl scope has been inlplemented as an alternative to the 2001 Edition through the 2003 Addenda of the ASME Section XI examination progratll for lSI Class 1 B-F and B-1 welds and lSI Class 2 C-F-1 and C-F-2 welds in accordance with 10CFR50.55a(a)(3)(i). The basis for the resulting Risk Categorizations of the nonexempt lSI Class 1 and 2 piping systetllS at PBAPS is defined and maintained in the Final RepOli "Risk Informed Inservice Inspection Evaluation" as referenced in Section 9.0 of this document. References to ASME Section XI Exatllination Categories B-F, B-1, C-F-l, and C-F-2 have been replaced with Exanlination Category R-A to identitY thenl as part of the RISI progratll. The CISI Program per Subsection IWE is included in Section 6.0, "Containment lSI Plan". The CISI relief requests are included in Section 8.0 of this docull1ent. 2.1.1 ASI\\/IE Section XI Code Cases As referenced by 10CFR50.55a(b)(5) and allowed by USNRC Regulatory Guide 1.147, Revision 15, the following Code Cases are being incorporated into the PBAPS lSI Program. These Code Cases have been detennined by the USNRC to be acceptable alternatives to applicable parts Alioll Sciellce & Technology 2-1 PBT05.G03 Revisioll 0

lSI Progrtll1l Plan Peach Bottom Atomic Power Statio/l U/lits 2 & 3, FOllrth Illterval of ASME Section XI. These Code Cases 111ay be used by PBAPS without a relief request from the USNRC, provided that they are used with any identified linlitations or 1110difications. Some of the Code Cases listed below are acceptable to the USNRC for application at PBAPS within the limitations ilnposed by the USNRC staff. Unless otherwise stated, lilllitations imposed by the USNRC are in addition to the conditions specified in the Code Case. N-432-1 N-460 N-504-3 N-513-2 N-516-3 N-517-1 N-526 Repair Welding Using Autolnatic or Machine Gas Tungsten-Arc Welding (GTAW) Telnper Bead Technique. Regulatory Guide 1.147, Revision 15. Alternative Exanlination Coverage for Class 1 and Class 2 Welds. Regulatory Guide 1.147, Revision 15. Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping Code Case N-504-3 is acceptable subject to the following condition specified in Regulatory Guide 1.147, Revision 15: The provisions of Section XI, Nomnandatory Appendix Q, "Weld Overlay Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping Weldnlents," lnust also be luet. Evaluation Criteria for Tenlporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping. Regulatory Guide 1.147, Revision 15. Underwater Welding Code Case N-516-3 is acceptable subject to the following condition specified in Regulatory Guide 1.147, Revision 15: Licensee lnust obtain USNRC approval in accordance with 10CFR50.55a(a)(3) regarding the technique to be used in the weld repair or replacelnent of irradiated material underwater. Quality Assurance Progranl Requirenlents for Owners. Regulatory Guide 1.147, Revision 15. Alternative Requirelllents for Successive Inspections of Class 1 and 2 Vessels. Regulatory Guide 1.147, Revision 15. AliO/l Scie/lce & Technology 2-2 PBT05.G03 Rel/isio/l 0

lSI Program Plan Petlcll Bottom Atomic Power Station Units 2 & 3, FOllrtlt Imerval N-528-1 N-532-4 N-552 N-566-2 Purchase, Exchange, or Transfer of Material Between Nuclear Plant Sites Code Case N-528-1 is acceptable subject to the following condition specified in Regulatory Guide 1.147, Revision 15: The requirelnents of 10CFR Part 21, "Reporting of Defects and Noncolnpliance", are to be applied to the nuclear plant site supplying the material as,"vell as to the nuclear plant site receiving the material that has been purchased, exchanged, or transferred between sites. Alternative RequirelTIents to Repair and Replacement Documentation Requirements and Inservice Sun1mary Report Preparation and Subn1ission as Required by IWA-4000 and IWA-6000. Regulatory Guide 1.147, Revision 15. Alternative Methods - Qualification for Nozzle Inside Radius Section from the Outside Surface Code Case N-552 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: To achieve consistency with the 10CFR50.55a rule change published Septen1ber 22, 1999 (64 FR 51370), incorporating Appendix VIII, "Performance Denl0nstration for Ultrasonic Examination Systelns," to ASME Section XI, add the following to the specilnen requirements: "At least 50 percent of the flaws in the demonstration test set must be cracks and the InaxinlUlTI misorientation must be delllonstrated with cracks. Flaws in nozzles with bore diatneters equal to or less than 4 inches lTIay be notches." Add to detection criteria, "The nUlnber of false calls must not exceed three." Corrective Action for Leakage Identified at Bolted Connections. Regulatory Guide 1.147, Revision 15. Alion Science & Technology 2-3 PBT05. G03 Revision 0

lSI Program PIa" Petleh Bottom Atomic.: Power Stat/on Units 2 & 3, Fourtlt Interval N-586-1 N-597-2 Alternative Additional Exmuination Requirements for Class 1, 2, and 3 Piping, Components, and Supports. Regulatory Guide 1.147, Revision 15. Requirements for Analytical Evaluation of Pipe Wall 111inning Code Case N-597-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: (l) Code Case must be supplet11ented by the provisions of EPRI Nuclear Safety Analysis Center Report 202L-R2, April 1999, "Recomnlendations for an Effective Flow Accelerated Corrosion Progranl," for developing the inspection requiretuents, the 111ethod of predicting the rate of wall thickness loss, and the value of the predicted remaining wall thickness. As used in NSAC-202L-R2, the ternl "should" is to be applied as "shall" (i.e., requirenlent). (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 Owner's requirenlents or a later USNRC approved Edition of Section III, "Rules for Construction of Nuclear Plant Conlponents," of the ASME Code prior to the value of fp reaching the allowable tuininl111u wall thickness, fnun, as specified in -3622.1 (a)(1) of this Code Case. Alternatively, use of the Code Case is subject to USNRC review and approval per 10CFR50.55a(a)(3). (3) For Class 1 piping not tueeting the criteria of -3221, the use of evaluation Iuethods and criteria is subject to USNRC review and approval per 10CFR50.55a(a)(3). (4) For those conlponents that do not require immediate repair or replacenlent, the rate of wall thickness loss is to be used to determine a suitable inspection frequency so that repair or replacenlent occurs prior to reaching allowable mininlum wall thickness, '/11111' (5) For corrosion phenomenon other than flow accelerated corrosion, use of the Code Case is subject to USNRC review and approval per 10CFR50.55a(a)(3). Inspection plans and wall Alion Science & Technology 2-4 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 &. 3, Fourth Interval thinning rates nlay be difficult to justify for certain degradation nlechanisnls such as MIC and pitting. N-600 N-606-1 N-613-1 N-624 N-629 N-638-1 Transfer of Welder, Welding Operator, Brazer, and Brazing Operator Qualifications Between Owners. Regulatory Guide 1.147, Revision 15. Sitnilar and Dissilnilar Metal Welding Using Ambient Telnperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stud Tube Repairs Code Case N-606-1 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: Prior to welding, an exanlination or verification luust be performed to ensure proper preparation of the base metal, and that the surface is properly contoured so that an acceptable weld can be produced. The surfaces to be welded, and surfaces adjacent to the weld, are to be free fron1 contanlinants, such as, rust, lnoisture, grease, and other foreign lnaterial or any other condition that would prevent proper welding and adversely affect the quality or strength of the weld. This verification is to be required in the welding procedures. Ultrasonic Exanlination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.1°and B3.90, Reactor Nozzle-to-Vessel Welds, Figs. IWB-2500-7(a), (b), and (c). Regulatory Guide 1.147, Revision 15. Successive Inspections. Regulatory Guide 1.147, Revision 15. Use of Fracture Toughness Test Data to Establish Reference Temperature for Pressure Retaining Materials. Regulatory Guide 1.147, Revision 15. Sinlilar and Dissimilar Metal Welding Using Alnbicnt Temperature Machine GTAW Temper Bead Technique Code Case N-638-1 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: Alioll Science & Technology 2-5 PBT05.G03 Revision 0

lSI Program Pia" Peach Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval UT examinations shall be perforn1ed with personnel and procedures qualified for the repaired volume and qualified by demonstration using representative sanlples which contain construction type flaws. The acceptance criteria ofNB-5330 in the 1998 Edition through the 2000 Addenda of Section III apply to all flaws identified within the repaired volun1e. N-639 N-641 N-649 N-651 N-652-1 N-66l Alternative Calibration Block Material Code Case N-639 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: Chemical ranges of the calibration block may vary fr01n the materials specification if (l) it is within the chemical range of the conlponent specification to be inspected, and (2) the phase and grain shape are n1aintained in the san1e ranges produced by the thennal process required by the material specification. Alternative Pressure-Telllperature Relationship and Low Ten1perature Overpressure Protection Systeln Requirenlents. Regulatory Guide 1.147, Revision 15. Alternative Requirelnents for IWE-5240 Visual Examination. Regulatory Guide 1.147, Revision 15. Ferritic and Dissilnilar Metal Welding Using SMAW Ten1per Bead Technique Without Removing the Weld Bead Crown of the First Layer. Regulatory Guide 1.147, Revision 15. Alternative Requirelnents to Exan1ination Categories B-G-l, B-G-2, and C-D Bolting Examination Methods and Selection Criteria. Regulatory Guide 1.147, Revision 15. Alternative Requirelnents for Wall Thickness Restoration of Classes 2 and 3 Carbon Steel Piping for Raw Water Service Code Case N-661 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15: Alio/1 Science & Tecllllology 2-6 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, FOllrth IlI1erval (a) If the root cause of the degradation has not been detern1ined, the repair is only acceptable for one cycle. (b) Weld overlay repair of an area can only be perfonned once in the same location. (c) When through-wall repairs are rnade by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage. N-664 N-665 N-686 N-695 N-696 N-700 Perfonnance Den10nstration Requiren1ents for Examination of Unclad Reactor Pressure Vessel Welds, Excluding Flange Welds. Regulatory Guide 1.147, Revision 15. Alternative Requirements for Bean1 Angle Measuren1ents Using Refracted Longitudinal Wave Search Units. Regulatory Guide 1.147, Revision 15. Alternative Requirernents for Visual Examinations, VT-1, VT-2, and VT-3. Regulatory Guide 1.147, Revision 15. Qualification Requirements for Dissimilar Metal Piping Welds. Regulatory Guide 1.147, Revision 15. Qualification Requiren1ents for Appendix VIn Piping Examinations Conducted from the Inside Surface. Regulatory Guide 1.147, Revision 15. Alternative Rules for Selection of Classes 1, 2, and 3 Vessel Welded Attachn1ents for Exan1ination. Regulatory Guide 1.147, Revision 15. 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 at that tin1e. 2.1.2 ASME OM Code Cases As referenced by 10CFR50.55a(b)(6) and allowed by USNRC Regulatory Guide 1.192, Revision 0, the following Code Cases are being incorporated into the PBAPS lSI Program. OMN-13, Rev. 0 Requirements for Extending Snubber Inservice Visual Exmnination Interval at LWR Power Plants. Additional Code Cases invoked in the future shall be in accordance with those approved for use in the latest published revision of Regulatory Guide 1.1 92 at that time. ,4li0/1 Science & Tech/1ology 2-7 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, FOllrth Interval 2.2 Auglnented Inspection Plan Requirements Auglnented Inspection Plan requirelnents are those inspections that are perfonned above and beyond the requirelnents of ASME Section XI. Below is a sunlnlary of those examinations perfornled by PBAPS that are not specifically addressed by ASME Section XI, or the inspections that will be performed in addition to the requirements of ASME Section XI on a routine basis during the Fourth lSI Interval and Second CISI Interval. IInplelnentation of the exanlination comlnitments is included in Section 7.0 of this lSI Prograrn Plan and the associated lSI Database. See the Auglnented Inspection Plan for details on the PBAPS lSI and CISI mandatory and non-Inandatory augnlented progranls, exarnination requirements, and references. 2.2.1 (AUG-OI), IGSCC Inspection Program; Generic Letter 88-01; USNRC Position on rGSCC in Boiling Water Reactor (BWR) Austenitic Stainless Steel Piping, Revision 2 / Supplement 1 to Generic Letter 88-01, NUREG 0313, Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2, EPRI Report TR-113932, BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), and EPRI Report TR-1012621, BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A) 2.2.2 (AUG-02), BWR Feedwater Nozzle Inspection Requirelnents; NUREG 0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking and Boiling Water Reactor Owners' Group (BWROG) Reports GE-NE-523-A71-0594 and GE-NE-523-A71-0594-A, Revision 1, Alternate BWR Feedwater Nozzle Inspection Requirements Note: AugInented examination of the Feedwater Spargers has been transferred to procedure ER-PB-331-1 001, "RPV & Internals Program Basis and Implenlentation Docwnent", whereas augInented examination of the Feedwater Nozzles relnains with this section. 2.2.3 (AUG-05), Snubber Visual EXaIllination and Functional Testing Progranl 2.2.4 (AUG-12), 10CFR50.55a and BWRVIP-05, IOCFR50 AugInented Requirements for Reactor Pressure Vessel Shell Weld Examinations 2.2.5 (AUG-I 5), Standby Liquid Control Nozzle-to-Safe-End 2.2.6 (AUG-D), RPV Closure Flange O-ring Sealing Surfaces 2.2.7 (AUG-C2), BPCI, RHR and Core Spray Suction Strainers Alion Science & Technology 2-8 PBT05.G03 Revision ()

lSI ProgNI111 Plan Peach Bottom Atomic Power Station Units 2 & 3, FOllrth Interval 2.2.8 (AUG-C3), Sludge Acctunulation on the Torus Floor 2.2.9 (AUG-CA), Exanlination of Class MC Supports 2.2.10 (AUG -CB), Exanlination of Drywell External COJ1lpOnents Located Outside Stabilizer Access Hatches 2.2.11 (AUG -CC), Exanlination of Drywell Airgap Drain Lines 2.2.12 (AUG -CD), Exanlination of Bolting in ECCS Suction Strainers 2.2.13 Augnlented inspect ion progranls associated with the BWRVIP and the PBAPS Vessel Internals Progranl have been transferred to procedure ER-PB-331-1001, "RPV & Internals Progrmn Basis and Itnplementation Document" The PBAPS Vessel Internals Progrmn procedure now includes the following: AUG-02 AUG-03 AUG-04 AUG-06 AUG-07 AUG-08 AUG-09 AUG-IO AUG-II AUG-I4 AUG-I9 BWR Feedwater Nozzle Inspection Requirements (Feedwater Sparger Only); NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking and BWR Feedwater Nozzle Inspection Requirenlents; Boiling Water Reactor Owners' Group (BWROG) Reports GE-NE-523-A71-0594 and GE-NE-523-A71-0594-A, Revision I, Alternate BWR Feedwater Nozzle Inspection Requirenlents. (See Note in Section 2.2.2 for augnlented exanlination details for Feedwater Spargers) IE Bulletin No. 80-13, Cracking in Core Spray Spargers, and BWRVIP-18-A, BWR Core Spray Internals BWRVIP-41 and BWRVIP-138, Jet Punlp Assembly BWRVIP-26, Top Guide BWRVIP-38 and BWRVIP-I04, Shroud Support SIL-462 Rev. 1, Shroud Access Hole Covers BWRVIP-76, Core Shroud BWRVIP-25, Core Plate BWRVIP-47, Lo\\ver Plenunl Region BWRVIP-48, Vessel ID Attachnlent Welds BWRVIP-139, Steam Dryers 2.2.14 USNRC NUREG -0737, dated Novetnber 1980 This document discusses TMI Action Plan Requirements and includes requirenlents in Item III.D.l.1 for leak testing and periodic visual exanlinations of systenls outside of primary containnlent which could contain highly radioactive fluids during a serious transient or accident. Conlmitnlents Illade concel11ing NUREG-0737 are maintained outside of AliO/l Science & Tecllllology 2-9 PBT05.G03 Revision 0

lSI Program PIau Pead, Bol1ol1l Atomic Power Station Units 2 & 3, Fourth Interval the ISI and Pressure Testing Progran1s and are in1plen1ented per the PBAPS Technical Specifications (TS) Section 5.5.2 and Technical Requiretnents Manual (TRM) Section 5.5.2. I111plementation of the PBAPS progral1l addressing these requirements is included in a series of station surveillance tests (not part of the SPT progrmn) that complete these exan1inations every 24 n10nths. 2.3 System Classifications and P&ID Boundary Drawings The lSI Classification Basis Document details those systetns that are lSI Class 1, 2, 3, or MC that fall within the lSI scope of exmninations. Below is a sun1nlary of the classification criteria used within the lSI Classification Basis Document. Each safety related, t1uid system containing water, steanl, air, oil, etc. included in the PBAPS UFSAR was reviewed to determine which safety functions they perfornl during all modes of systeln and plant operation. Based on these safety functions, the systen1S and cOlnponents were evaluated per classification documents. The syste111S were then designated as lSI Class 1, 2, 3, MC, or non.. classed accordingly. This evaluation followed the guidelines ofUFSAR Section 1.4 and Table 1.4.2 When a particular group of cotnponents is identified as perfornling an lSI Class 1, 2, or 3 safety function, the conlponents are further reviewed to assure the interfaces (boundary valves and boundary barriers) tneet the criteria set by 10CFR50.2, 10CFR50.55a(c)(1), 10CFR50.55a(c)(2), and Regulatory Guide 1.26, Revision 3. SRP 3.2.2 and ANSI!ANS-58.14-1993 (PBAPS is not conln1itted to or licensed in accordance with the last t\\VO dOCU111ents) were also used for guidance in detennining the classification boundaries where 10CFR and the Regulatory Guide did not address a given situation. Components within the reactor coolant pressure boundary, as defined by 10CFR50.2, are typically designated as lSI Class 1 while the other safety related con1ponents are evaluated for lSI Class 2 or 3 designation in accordance with the guidelines of Regulatory Guide 1.26. Per Regulatory Guide 1.26 Paragraphs A and B, the lSI Class 2 and 3 boundaries are limited to safety related systems and components (reference UFSAR Table 1.4.2). Where sut1icient classification criteria is not provided within 10CFR50 or Regulatory Guide 1.26, other industry documents such as NUREG-0800 and ANSI!ANS standards are consulted "for guidance". PBAPS is only cOlnmitted to and licensed in accordance with the rules and regulations of 10CFR50 and Regulatory Guide 1.26. According to 10CFR50.55a, Paragraph (g)(4), the lSI requirements of ASME Section XI are assigned to these conlponents, within the constraints of existing plant design. The PBAPS lSI Class 1, 2, and 3 components that are eXetllpt frOlTI exanlination are those which meet the criteria of ASME Section XI, Paragraphs IWB-1220, IWC-1220, and IWD-1220. Supports which nleet the criteria of Alioll Sciellce & Technology 2-10 PBT05.G03 Revisioll 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval Paragraph IWF-1230 ofASME Section XI are also exenlpt fronl examination. For Containment, CISI Class MC conlponents which l11eet Paragraph IWE-1220 are exempt from exal11ination. The systel11s and conlponents (piping, pumps, valves, vessels, etc.), which are subject to the eXaIl1inations 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 Inservice Inspection Boundary Drawings with lSI classification color coding as detailed in Table 2.3-2. Table 2.3-1 provides a listing of the System Designations used for the piping systems and components that are subject to lSI at PBAPS, Units 2 and 3. Table 2.3-2 provides a listing of the lSI Boundary Drawings that depict the Class 1, 2, and 3 conlponents subject to the requirements of ASME, Section XI, during the Fourth lSI Interval at PBAPS. These drawings are derived from the station piping and instrumentation diagrams (P&IDs). Table 2.3-3 provides a listing of the CISI Boundary Drawings that depict the Class MC components subject to the requirenlents ofASME Section XI during the Second CISI Interval at PBAPS. In addition to the general boundary drawings listed below, specific boundaries for the Containment Atmosphere Control System are shown on drawings ISI-367, Sht. 1 (Unit 2) and Sht. 2 (Unit 3). Alion Science & Technology 2-11 PBT05.G03 Revisioll 0

lSI Program PIlIn Peach Bottom Atomic Power Station Ullits 2 & 3, FOllrth Interval TABLE 2.3-1 SYSTEM DESIGNATORS Designator System 01 Main Steatn (MS) 02 Main Recirculation (MR) 03 Control Rod Hydraulic (CRD) 04 Nuclear Boiler Vessel Instnunentation (NBVI) 06 Feedwater (FW) 10 Residual Heat Removal (RHR) 11 Standby Liquid Control (SLC) 12 Reactor Water Clean-Up (RWCU) 13 Reactor Core Isolation Cooling (RCIC) 14 Core Spray (CS) 23 High Pressure Coolant Injection (HPCI) 32 High Pressure Service Water (HPSW) 33 En1ergency Service Water (ESW) 48 En1ergency Cooling Water (ECW) AliO/1 Science & Tech/1ology 2-12 PJJT05.G03 Revision (J

lSI Program Plan Peacll Bottolll Atomic Power Station Units 2 & 3, Fourtll IntervlIl TABLE 2.3-2 INSERVICE INSPECTION BOUNDARY DRAWINGS Drawing Sheet Number Title Number ISI-303 1, 3 Main Steam, Bypass, and Crossaround ISI-308 1,2,3,4 Feedwater and Feed Pumps ISI-309 1,2 Condensate Storage ISI-315 1, 2, 3, 4, 5 Elllergency Service Water and High Pressure Service Water ISI-330 1 Elllergency Cooling Systeln ISI-331 1,3 Off-Gas Reconlbiner Systelll ISI-351 1,2,3,4 Nuclear Boiler ISI-352 1,2,3,4 Nuclear Boiler Vessel Instn.llnentation ISI-353 1,2,3,4 Reactor Recirculation Punlp SystelTI IS1-354 1, 2 Reactor Water Cleanup System ISI-356 1,2 Control Rod Drive (CRD) Hydraulic Systenl Part A IS1-357 1,2 eRD Hydraulic System Part B IS1-358 1,2 Standby Liquid Control System IS1-359 1, 2 Reactor Core Isolation Cooling (RCIC) Systenl IS1-360 1, 2 RCIC Pump Turbine Details ISI-361 1,2,3,4 Residual Heat Removal Systeln IS1-362 1, 2 Core Spray Cooling System IS1-365 1,2 High Pressure Coolant Injection (HPCI) SystelTI ISI-366 1, 4 HPCI PUlnp Turbine Details ISI-367 1,2 Containment Atnl0spheric Control Systenl IS1-372 1,2 Containment Atnl0spheric Dilution Systenl Alion Science & Tecllnology 2-13 PBT05.G03 Revision ()

lSI Progrllm Pla/l Peach Bottom Atomic Power S/atio/l U/lits 2 & 3, FOllr/h Interval TABLE 2.3-3 CONTAINlVlENT INSERVICE INSPECTION BOUNDARY DRA'VINGS Drawing Sheet Unit Title Number Number ISI-400 1 2 Containment - General Arrangen1ent IS1-400 2 2 Contaimnent Details ISI-400 3 2 Contai11lnent Details IS1-400 4 3 Contai11lnent - General Arrangen1ent IS1-400 5 3 Contai11lnent Details 1S1-400 6 3 Contaimnent Details IS1-401 1 2 Drywell 1.D. Roll-Out and Penetrations 1S1-401 2 2 Drywell 1.0. Roll-Out and Penetrations IS1-401 3 3 Orywell 1.0. Roll-Out and Penetrations IS1-401 4 3 Orywell 1.0. Roll-Out and Penetrations IS1-402 1 2 Suppression Chamber Penetrations 1S1-402 2 3 Suppression Chamber Penetrations Alio/l Scie/lce & Technology 2-14 PBT05.G03 Revisio/l 0

lSI Progrtll1l Plall Peach Bol1o111 Atomic Power Station Units 2 & 3, Fourth Interval 2.4 lSI Isometric and Component Drawings for Nonexelnpt lSI Class Con1ponents and Supports lSI Isometric and Component Drawings were developed to detail the ISI Class 1, 2, and 3 con1ponents (\\velds, bolting, etc.) and support locations at PBAPS. These lSI con1ponent and suppol110cations are identified on the lSI Piping Ison1etric Dra\\vings listed in Tables 2.4-1 and 2.4-2. These drawings identify piping welds, valves, pumps, and supports that are within the non-exempt piping boundaries. In addition, systen1 identifications, pipe sizes, containn1ent penetrations, and piping configurations are identified. Calibration Standards approved for use at PBAPS are listed in Table 2.4-3. Limerick Generating Station (LGS) Calibration Standards approved for use at PBAPS are listed in Table 2.4-4. The Prilnary Containment (Drywell) Penetrations penetrated by lnultiple Class 1, 2, 3, and 11011-classed piping lines, and numerous electrical and instnllnentation penetrations are listed in Table 2.4-5. PBAPS's lSI ProgralTI, including the lSI Database, lSI Classification Basis Document, lSI Selection Document and schedule, addresses the nonexempt components, which require exan1ination and testing. A SUmll1ary ofPBAPS ASME Section XI nonexempt con1ponents and supports is included in Section 7.0. AUon Science & Technology 2-15 PBTOS.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC DRAWINGS System Unit 2 and COlumon Unit 3 Isometric Drawings Isometric Drawings Core Spray DCN-14-MI-203-5-A DCN-14-MI-303-4-A DCN-14-MI-203-5-B DCN-14-MI-303-4-B GB-14-MI-202-2-A GB-14-MI-302-2-A GB-14-MI-202-2-B GB-14-MI-302-2-B GB-14-MI-202-2-C GB-14-MI-302-2-C GB-14-MI-202-2-D GB-14-MI-302-2-D GB-14-MI-203-3-A GB-14-MI-303-3-A GB-14-MI-203-3-B GB-14-MI-303-3-B GB-14-MI-203-4-A HB-14-MI-30 l-I-A GB-14-MI-203-4-B HB-14-MI-30 I-1-B HB-14-MI-20 l-I-A HB-I4-MI-30 l-I-C HB-14-MI-20 l-I-B HB-I4-MI-30 I-1-D HB-14-MI-20 l-I-C HCR-27-MI-301-1 HB-14-MI-201-1-D HCR-27-MI-301-2 HC-27-MI-201-1 Emergency Cooling Water GB-48-MI-OO 1-2 GB-48-MI-OO1-2 GB-48-MI-OO 1-3 GB-48-MI-OO 1-3 HB-48-MI-OOl-1 HB-48-MI-OO 1-1 HB-48-MI-OO 1-3 HB-48-MI-OO 1-3 HB-48-MI-OO 1-4 HB-48-MI-OO 1-4 HB-48-MI-OO 1-5 HB-48-MI-OO 1-5 .4liO/l Scie/lce & Tech/lology 2-16 PBT05.G03 Revision 0

lSI Program Plan Peac" Bollom Atomic Power Statioll Uults 2 & 3, FOllrth Illterval TABLE 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC DRAWINGS System Unit 2 and COInmon Unit 3 Isolnetric Drawings Isometric Drawings El11ergency Service \\Vater HB-33-MI-201-1 HB-33-MI-30I-I HB-33-MI-20 1-2 HB-33-MI-30 1-2 HB-33-MI-20 1-3 HB-33-MI-30 1-3 HB-33-MI-20 1-4 HB-33-MI-30 1-4 HB-33-MI-20 1-5 HB-33-MI-30 1-5 HB-33-MI-20 1-6 HB-33-MI-30 1-6 HB-33-MI-20 1-7 HB-33-MI-30 1-7 HB-33-MI-20 1-8 HBC-33-MI-30 1-8 HB-33-MI-20 1-9 HBC-33-MI-30 1-9 HB-33-MI-20 1-1 0 HBC-33-MI-301-10 HB-33-MI-201-12 HBC-33-MI-30 1-11 HBC-33-MI-20 1-13 HBC-33-MI-30 1-12 HBC-33-MI-201-14 HBC-33-MI-30 1-13 HBC-33-MI-20 1-15 HBC-33-MI-30 1-14 HBC-33-MI-201-16 HBC-33-MI-301-15 HBC-33-MI-20 1-17 HBC-33-MI-302-16 HBC-33-MI-20 1-18 HBC-33-MI-302-17 HBC-33-MI-201-19 HBC-33-MI-302-18 HBC-33-MI-201-20 HBC-33-MI-302-19 HBC-33-MI-20 1-21 HBC-33-MI-302-20 HBC-33-MI-201-22 HBC-33-MI-302-21 HBC-33-MI-20 1-23 HBC-33-MI-302-22 HBC-33-MI-201-24 HBC-33-MI-302-23 HBC-33-MI-201-25 HBC-33-MI-302-24 HBC-33-MI-201-26 HBC-33-MI-201-27 HBC-33-MI-20 1-28 HBC-33-MI-201-29 HBC-33-MI-20 1-30 HBC-33-MI-201-31 HBC-33-MI-201-32 Feedwater DDN-06-MI-201-2-A DDN-06-MI-30 1-2-A DDN-06-MI-201-2-B DDN-06-MI-30 1-2-B DD-06-MI-20 1-1 DD-06-MI-30 1-1 Alioll Science & Technology 2-/7 PBT05.G03 Revision 0

lSI Progrtl111 Plan Peach Bott011l Atomic Power Slation Units 2 & 3, Fourth Interval TABLE 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC DRAWINGS System Unit 2 and COlumon Unit 3 Isometric Drawings Isometric Drawings High Pressure Coolant DBN-23-MI-203-6 DBN-23-MI-303-5 Injection BN-23-MI-203-7 DBN-23-MI-303-6 DDN-23-MI-202-2 DBN-23-MI-303-7 DDN-23-MI-202-3 DDN-23-MI-302-2 DDN-23-MI-202-4 DDN-23-MI-302-3 DDN-23-MI-202-5 DDN-23-MI-302-4 HB-23-MI-201-1 HB-23-MI-30 1-1 HB-23-MI-204-8 HB-23-MI-304-8 High Pressure Service Water GB-32-MI-20I-l GB-32-MI-30 1-1 GB-32-MI-20I-2 GB-32-MI-30 1-2 GB-32-MI-20 1-3 GB-32-MI-301-3 GB-32-MI-20 I-4-A GB-32-MI-30 1-4 GB-32-MI-201-4-B GB-32-MI-30 1-5-A GB-32-MI-20 1-5-A GB-32-MI-30 1-5-B GB-32-MI-20 I-5-B GB-32-MI-30 1-6-A GB-32-MI-202-6-A GB-32-MI-30 1-6-B GB-32-MI-202-6-B GB-32-MI-302-7-A GB-32-MI-202-7 GB-32-MI-302-7-B GB-32-MI-302-8 Main Recirculation RCS-02-MI-20 I-I-A RCS-02-MI-30 l-1-A RCS-02-MI-20 I-1-B RCS-02-MI-30 I-1-B Main Stean1 OBN-O I-MI-20 l-1-A DBN-O I-MI-30 I-I-A DBN-O I-MI-20 I-I-B DBN-O I-MI-30 I-I-B DBN-O I-MI-20 I-I-C DBN-O I-MI-30 I-I-C DBN-O I-MI-201-1-D DBN-O I-MI-301-1-D DB-O I-MI-20 1-2-A DB-O I-MI-30 1-2-A DB-OI-MI-20 I-2-B DB-O I-MI-30 1-2-B DB-OI-MI-20 1-2-C DB-O I-MI-301-2-C DB-OI-MI-201-2-0 DB-O I-MI-30 1-2-D DB-OI-MI-221-3 DB-O I-MI-321-3 DB-OI-MI-221-4 DB-O I-MI-321-4 OB-OI-MI-222-5-A DB-O I-MI-322-5-A DB-OI-MI-222-5-B DB-O I-MI-322-5-B A!ion Science & Technology 2-/8 PBT05.G03 Revision ()

IS] Progrtllll Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC DRAWINGS System Unit 2 and COlnmon Unit 3 Isometric Drawings Isometric Drawings Main Stean1 Relief Valve 00-01-MI-271-A 00-01-MI-371-A 00-01-MI-271-B 00-01-MI-371-B 00-01-MI-271-C 00-01-MI-371-C 00-01-MI-271-D 00-01-MI-371-D 00-01-MI-271-E 00-01-MI-371-E 00-01-MI-271-F 00-01-MI-371-F 00-01-MI-271-0 00-0 I-MI-371-0 OG-O I-MI-271-H 00-01-MI-371-H 00-01-MI-271-J 00-01-MI-371-J OG-OI-MI-271-K 00-01-MI-371-K OG-OI-MI-271-L 00-01-MI-371-L Reactor Core Isolation DDN-13-MI-20 1-1 DDN-13-MI-30 1-1 Cooling Reactor Drain DCN-04-MI-201-2 DCN-04-MI-301-2 DDN-04-MI-201-1 DDN-04-MI-30 1-1 Reactor Water Cleanup DCA-12-MI-201-1 DCA-12-MI-301-1 DCA-12-MI-203-3 DCA-12-MI-303-3 DE-12-MI-202-2 DE-12-MI-302-2 DE-12-MI-203-4 DE-12-MI-303-4 DE-12-MI-203-5 DE-12-MI-303-5 DE-12-MI-203-6 DE-12-MI-303-6 Alioll Science & Technology 2-19 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, FOllrth Illterval TABLE 2.4-1 UNITS 2 AND 3 lSI PIPING ISOMETRIC ORAWINGS System Unit 2 and COlnmon Unit 3 Isometric Drawings Isometric Drawin2s Residual Heat Rellloval DCN-I0-MI-206-15 DCA-I0-MI-303-9-A DeN-I0-MI-207-16 DCA-IO-MI-303-9-B DDN-I0-MI-203-8-A DCA-I0-MI-306-13 DDN-I0-MI-203-8-B DDN-I0-MI-303-8-A DE-I0-MI-203-9-A DDN-I0-MI-303-8-B DE-I0-MI-203-9-B GB-I0-MI-302-3-A GB-I0-MI-202-3-A GB-I0-MI-302-3-B GB-I0-MI-202-3-B GB-I0-MI-302-3-C GB-I0-MI-202-3-C GB-I0-MI-302-3-D GB-l O-MI-202-3-D GB-I0-MI-302-4-A GB-I0-MI-202-4-A GB-I0-MI-302-4-B GB-I0-MI-202-4-B GB-I0-MI-302-4-C GB-I0-Ml-202-4-C GB-I0-MI-302-4-D GB-I0-MI-202-4-D GB-I0-MI-303-5-A GB-I0-MI-203-5-B GB-I0-MI-303-5-B GB-I0-MI-203-6-B GB-I0-MI-303-6-A GB-I0-MI-203-7-A OB-l 0-MI-303-7-A GB-I0-MI-203-7-B GB-I0-MI-303-7-B OB-l 0-MI-204-1 O-A GB-l 0-MI-304-1 O-A GB-l 0-MI-204-10-B GB-l 0-MI-304-1 O-B GB-I0-MI-204-11-A GB-I0-MI-304-11-A GB-I0-MI-204-l1-B OB-I0-MI-304-11-B GB-l O-MI-205-l2-A GB-I0-MI-305-12-A GB-I0-MI-205-13-A GB-I0-MI-305-12-B GB-I0-MI-205-13-B HB-l 0-MI-30 l-1-A GB-I0-MI-206-14 HB-l 0-MI-30 I-1-B HB-l 0-MI-201-I-A HB-l 0-MI-30 l-I-C HB-l 0-MI-20 I-1-B HB-l 0-MI-30 I-I-D HB-l 0-MI-20 l-I-C HB-l 0-MI-30 1-2-A HB-l 0-MI-20 1-1-0 HB-l 0-MI-30 1-2-C HB-I0-MI-201-2-A HB-l O-MI-30 I-I-D HB-l O-Ml-20 1-2-C HB-I0-MI-306-14 HB-l 0-MI-20 1-2-0 HB-I0-MI-207-17 Scran1 0 ischarge Vol U111e CS-03-MI-201-I-A CS-03-Ml-30 l-1-A CS-03-MI-201-1-B CS-03-MI-30 I-1-B CS-03-MI-20 1-2-A AUolI Science & Technology 2-20 PBT05,G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Statioll Ullits 2 & 3, Fourth Illterval TABLE 2.4-2 UNITS 2 AND 3 COMPONENT DRAWINGS Component Unit 2 Component Drawings Unit 3 Component Drawings Main Recirculation PU111PS ISI-2-02-1 ISI-3-02-1 ISI-2-02-2 ISI-3-02-2 Reactor Pressure Vessel ISI-2-RV-01 ISI-3-RV-Ol ISI-203-RV-02 ISI-203-RV-02 ISI-203-RV-03 ISI-203-RV-03 ISI-203-RV-04 ISI-203-RV-04 ISI-2-RV-05 ISI-3-RV-05 ISI-203-RV-06 ISI-203-RV-06 ISI-203-RV-07 ISI-203-RV-07 ISI-203-RV-08 ISI-203-RV-08 ISI-203-RV-09 ISI-203-RV-09 IS1-203-RV-I 0 ISI-203-RV-10 ISI-203-RV-11 ISI-203-RV-11 ISI-203-RV-12 ISI-203-RV-12 ISI..203-RV-13 ISI-203-RV-13 ISI-203-RV-14 ISI-203-RV-14 ISI-203-RV-15 ISI-203-RV-15 ISI-203-RV-16 ISI-203-RV-16 ISI-203-RV-17 ISI-203-RV-17 ISI-203-RV-18 ISI-203-RV-18 ISI-203-RV-19 ISI-203-RV-19 ISI-203-RV-20 ISI-203-RV-20 ISI-203-RV-21 ISI-203-RV-21 ISI-203-RV-22 ISI-203-RV-22 I81-203-RV-23 ISI-203-RV-23 ISI-203-RV-24 ISI-203-RV-24 ISI-3-RV-50 ISI-3-RV-51 ISI-3-RV-52 Core Spray / RHR PU111P 181-2-10-1 ISI-3-10-1 ISI-2-10-2 ISI-3-10-2 _"Ilioll Sciellce & Techllology 2-21 PBT05.G03 Revisioll 0

lSI Program PIau Peach Bottom Atomic Power Slatioll Units 2 & 3, Fourth Illlerl'{ll TABLE 2.4-3 ASME SECTION XI, lSI CALIBRATION BLOCI( DRAWINGS Drawing Title Number CBO-l ASME, Section XI, UT Calibration Block for PBAPS Units No.2 & 3 CBO-IA ASME, Section XI, UT Calibration Block for 26" Main Steam CBO-2A ASME, Section XI, UT Calibration Block for 24" Pipe CBO-3 ASME, Section XI, UT Calibration Block for 12" Feedwater Riser CBO-4 ASME, Section XI, UT Calibration Block for 6" Head Spray CBO-5A ASME, Section XI, UT Calibration Block for 6" Reactor Water Cleanup CBO-6A ASME, Section XI, UT Calibration Block for 6" Main Steam Safety and Relief CBO-7A ASME. Section XI, UT Calibration Block for 10" High Pressure Coolant Injection CBO-8 ASME, Section XI, UT Calibration Block for 14" High Pressure Coolant Injection CBO-8A ASME. Section XI, UT Calibration Block for 14" High Pressure Coolant Injection CBD-9A ASME, Section XI, UT Calibration Block for 28" Main Recire. Suction & Discharge CBO-IO ASME, Section XI, UT Calibration Block for 4" RWCD Main Recirculation Bypass, CRO CBO-I0A ASME, Section XI, UT Calibration Block for 4" RWCD Main Recirculation Bypass, CRO CBO-12 ASME. Section XI, UT Calibration Block for 12" Core Spray CBO-13A ASME, Section XI, UT Calibration Block for 20" Feedwater (block not used) CBO-13C ASME, Section XI, UT Calibration Block for 20" Feedwater CBD-I4A ASME, Section XI, UT Calibration Block for 22" Main Recirculation Manifold CBO-I5 ASME, Section XI, UT Calibration Block for 14" Feedwater Riser CBD-16A ASME, Section XI, UT Calibration Block for 12" Feedwater Riser CBO-I7 ASME, Section XI, UT Calibration Block for 4.5" Pipe CBO-18A ASME, Section XI, UT Calibration Block for 12" Main Recire. Nozzle-To-Safe End CBO-19A ASME. Section XI, UT Calibration Block for 12" Main Recire. Safe End-To-Nozzle CBO-20A ASME, Section XI, UT Calibration Block for 3PI-I/2" Flat Pipe CBO-2I ASME, Section XI, UT Calibration Block for 5.125 CRD Safe End CBO-22 ASME, Section XI, UT Calibration Block for RPV Nut Alfoll Science & Technology 2-22 PIJT05.G03 Revisioll 0

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval TABLE 2.4-3 ASME SECTION XI, lSI CALIBRATION BLOCK DRAWINGS Drawing Title Number CBO-24 ASME, Section XI, UT Calibration Block for Closure Head Thickness CBO-25 ASME, Section XI, UT Calibration Block for 6" Pipe CBO-26 ASME, Section XI, UT Calibration Block for 12" Pipe CBO-27 ASME, Section XI, UT Calibration Block for 4" Pipe CBD-28 ASME, Section XI, UT Calibration Block for 26" Pipe CBD-29 ASME, Section XI, UT Calibration Block for 20" Pipe CBD-30 ASME, Section XI, UT Calibration Block for 24" Pipe CBD-31 ASME, Section XI, UT Calibration Block for 24" Pipe CBD-32 ASME, Section XI, UT Calibration Block for 24" Pipe CBD-33A ASME, Section XI, UT Calibration Block for 20" Pipe CBD-34 ASME, Section XI, UT Calibration Block for 10" Pipe CBO-35 ASME, Section XI, UT Calibration Block for 6" SS Pipe CBO-36 ASME, Section XI, UT Calibration Block for 12" Pipe CBO-38 ASME, Section XI, UT Calibration Block for OD Inner Radius CBD-39 ASME, Section XI, UT Calibration Block for Clad Vessel CBD-41 ASME, Section XI, UT Calibration Block for Nozzle Cap CBD-42 ASME, Section XI, UT Calibration Block for Pmnp Stud CBD-43 ASME, Section XI, UT Calibration Block for Stabilizer Bracket CBO-44 ASME, Section XI,UT Calibration Block for Pun1p Nut CBD-45 ASME, Section XI, UT Calibration Block for Nut CBD-46 ASME, Section XI, UT Calibration Block for Stud CBD-47 ASME, Section XI, UT Calibration Block for RPV Closure Head CBD-48 ASME, Section XI, UT Calibration Block for RPV Stud CBD-49 ASME, Section XI, UT Calibration Block for 10" Pipe CBD-50 ASME, Section XI, UT Calibration Block for 6" SS Pipe CBO-51 ASME, Section XI, UT Calibration Block for 6" CS Pipe CBD-52 ASME, Section XI, UT Calibration Block for 20" SS Pipe CBO-53 ASME, Section XI, UT Calibration Block for 20" CS Pipe CBD-54 ASME, Section XI, UT Calibration Block for 24" SS Pipe Alioll Science & Technology 2-23 PBT05.G03 Revision 0

lSI Program Plan Petleh Bot/om Atomic Power Statioll Ullits 2 & 3, Fourth Illterval TABLE 2.4-3 ASlVIE SECTION XI, lSI CALIBRATION BLOCI( DRAWINGS Drawing Title Number CBD-55 ASME, Section XI, UT Calibration Block for 24" CS Pipe CBD-56 ASME, Section XI, UT Calibration Block for 12" SCH 100 Pipe CBD-57 ASME, Section XI, UT Calibration Block for 22" Pipe CBD-58 ASME, Section XI, UT Calibration Block for 28" Pipe CBD-59 ASME, Section XI, UT Calibration Block for 28" Pipe CBD-60 ASME, Section XI, UT Calibration Block for 30" Pipe CBD-61 ASME, Section XI, UT Calibration Block for 24" SCH 120 CS Pipe CBD-62 ASME, Section XI, UT Calibration Block for %" CS Plate CBD-62A ASME, Section XI, UT Calibration Block for SS 12" Clad Overlay (02-19-60) CBD-64 ASME, Section XI, UT Calibration Block for Jet Pun1p Seal CBD-65 ASME, Section XI, UT Calibration Block for 20" Pipe CBD-65A ASME, Section XI, UT Calibration Block for 4" Pipe CBD-66 ASME, Section XI, UT Calibration Block for 4" RWCU Pipe CBD-67 ASME, Section XI, UT Calibration Block for Clad Core Spray (3-453) CBD-67A ASME, Section XI, UT Calibration Block for Core Spray Nozzle-To-Safe End CBD-68 ASME, Section XI, UT Calibration Block for Vessel (12-CS-5) CBD-69 ASME, Section XI, UT Calibration Block for SS 22" Overlay (12-08-46) (block not used) CBD-70 ASME, Section XI, UT Calibration Block for SS 20" Overlay (03-13-46) (block not used) CBD-71 ASME, Section XI, UT Calibration Block for Jet PUlnp Overlay (09-08-47) CBD-72 Alternative ASME Calibration Block (Type 304 SS) CBD-73 Alternative ASME Calibration Block (Type 316 SS) CBD-74 Alternative ASME Calibration Block (A516-70) "ilioll Science & TechIIology 2-24 PBT05.G03 Revisioll 0

lSI Program Plall Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-4 LIMERICI( GENERATING STATION lSI CALIBRATION BLOCI(S FOR USED AT PBAPS CALIBRATION BLOCK NUMBER LIM-4-.337-SS304 LIM-6-.432-CS LIM-6-.432-SS LIM-6-.432-SS316 LIM-10-.593-CS-R LIM-12-.843-CS LIM-12-.688-SS LIM-12-.843-SS316 LIM-18-.938-CS LIM-20-1.031-CS LIM-22-1.009-SS LIM-26-1.013-CS LIM-28-1.285-SS316 LIM-U2-NI-NOZ T-1.972 LIM-20-.500-CS LIM-12-.688-SS316 LIM-U2-N1-SE T-2.209 LIM-16-.375-CS LIM-12-.688-CS-R Closure Stud-L/8, L/4, L/2, 3L/4, & L (Set) SIS NO.4 Nut No. 1361 Alion Sciellce & Technology 2-25 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-l 12'-0" N/A Equipment Access detail C M-832 N/A - No Pipe N-2 12'-0" N/A Equipment Access detail B M-832 N/A - No Pipe With Personnel Lock N-3 24" N/A Construction Manway detail J M-832 N/A - No Pipe N-4 24" N/A Head Access detail J M-832 N/A - No Pipe N-5 A to H 8'-0" 6'-9" Vent Line detail G M-832 N/A - No Pipe N-6 36" N/A CRD Removal Hatch detail D M-832 N/A - No Pipe N-7 A to D 44" 26" Steam to Turbine 2 M-851 ASMECL 1 N-8 8" ,",II Condensate Drain, M-851 ASMECL 1 EX .J .J Main Steam N-9 A,B 42" 24" RPV Feedwater 2 M-851 ASME CL 1 N-10 12" Steam to RCIC Turbine M-859 ASME CL 1 EX _1 N-Il 26" 10" Steam to HPCI Turbine 2 M-865 ASME CL 1 N-12 36" 20" RHR Punlp Supply 2 M-861 ASME CL 1 N-13 A,B 42" 24" RHR Pump Discharge 2 M-861 ASMECL 1 N-14 22" 6" RWC Pump Supply 2 M-854 ASME CL 1 N-16 A,B 28" 12" Core Spray Pump 2 M-862 ASMECL 1 Discharge N-17 22" 6" RPV Head Spray 2 M-861 Unclassified Abandoned N-18 '"'II ,",II Floor Drains M-868 Unclassified .) N-19 '"'II Equipment Drains M-868 Unclassified .) N-20 8" N/A Spare None N/A - No Pipe N-21 ,",II 1" Breathing Service Air M-820, Unclassified Sh21 Alion Science & Technology 2-26 PBT05.G03 Revision ()

lSI Program Plan Pellc" Bottom Atomic Power Stlltion Units 2 & 3, Fourth Interval TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-22 1" Instrument Air "l M-833, Unclassified .J .J Sh 1 N-23 8" 8" RB Closed Cooling M-8I6 Unclassified .J Water Supply N-24 8" 8" RB Closed Cooling "l M-816 Unclassified .J Water Return N-25 18" 18" Vent to Drywell "l M-867 MC .J N-26 18" 18" Vent from Drywell "l M-867 MC N-26A,B 1" 1" Vesse! Instrumentation M-852 ASMECL 1 EX N-27 A to D N/A Spares M-853 N/A - No Pipe N-27 E & F 12" 1" Core Plate Delta P 8 M-852 ASMECL 1 EX N-28-A,B,C,F 12" 1" Instrumentation 8 M-852 ASMECL 1 EX N-28 D 1" RPV Flange Leak-Off M-851 ASMECL2 EX N-28 E 1" Spare M-852 N/A - No Pipe N-29 A to F 12" 1" InstrUluentation 8 M-852 ASME CL 1 EX U2: B, Care spares. U3: A-E are spares N-30 A to F 12" 1" Instrumentation 8 M-851 ASMECL 1 EX N-31 A to F 12" 1" Instrumentation 8 M-853 ASME CL 1 EX N-31E,F 12" N/A Spare None N/A - No Pipe N-32A & B 12" 1" Instrun1entation 8 M-853 ASMECL 1 EX N-32C & D 12" 1" Instrumentation 8 M-832 Unclassified N-32E & F 12" 1" Instrumentation 8 M-862 ASMECL 1 EX N-33A&B 12" 1" Instrumentation 8 M-853 ASMECL 1 EX N-33C & D 12" 1" Instrulnentation M-853 ASMECL 1 EX Alion Science & Technology 2-27 PBT05. G03 Revision (J

lSI Program Plall Peach Bottom Atomic Power Statio/l Units 2 & 3, Fourtlt l/lten'al TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-33E 12" N/A Spare None N/A - No Pipe N-33F 12" 1" Instrumentation 8 M-867 Unclassified ~ N-34 A to F 12" 1" Instru111entation 8 M-851 ASMECL 1 EX N-35 A 1-112" N/A Spare 1 M-8321 N/A - No Pipe (Unit 2) M-876 N-35 B,D 1-1/2" N/A Spare 1 M-8321 N/A - No Pipe (Unit 3) M-876 N-35 B,C,D 1-112" 3/8" TIP Purge 1 M-8321 Unclassified (Unit 2) M-876 N-35 A,C 1-1/2" 3/8" TIP Purge 1 M-832/ Unclassified (Unit 3) M-876 N-35 E to G 1-1/2" 3/8" TIP Drives 1 M-876 Unclassified N-36 12" 4" CRD Hydraulic Return None N/A - No Pipe Abandoned N-37 A to D 80 1" CRD Insert M-857 ASME CL 1 EX N-38 A to D 80 1" CRD Withdraw M-857 ASMECL 1 EX N-39 A,B 14" 12" Drywell Spray System M-861 ASMECL2 EX N-40 A,B,C,D 12" 1" Jet Pump Instr. 8 M-852 ASMECL 1 EX N-41 6" 1" Recirc. Loop Sanlple M-853 ASME CL 1 EX N-42 4" 1-1/2" Standby Liq. Control M-851 ASMECL2EX N-43 30" N/A Spare None N/A - No Pipe Unit 2 Only N-44 26" N/A Spare None N/A - No Pipe N-45 34" N/A Spare None N/A - No Pipe Unit 2 Only N-46 12" NIA Spare None N/A - No Pipe (Unit 2) N-46 A,B 12" 1" Instrumentation M-861 ASMECL2EX (Unit 3) Alio/l Science & Techllology 2-28 PBT05.G03 Revisiol1 (J

lSI Program Plan Peach Bottom Atomic Power Sttttion Units 2 & 3, Fourth Interval TABLE 2.4-5 UNITS 2 AND 3 DRY\\VELL PENETRATIONS Number Nom Process Senrices Penetration QAD Piping Notes Diameter Line Size Type Classification N-46 C to F 12" N/A Spare 3 None N/A - No Pipe (Unit 3) N-47 6" 1" ADS Pneumatic Supply M-833 Unclassified .) N-48 6" N/A Spare 3 None N/A - No Pipe U2 (N-49E,F); 12" 1" Containment 8 M-861 ASMECL2 EX U3 (N-49B,C) Instrumentation U2 (N-49A-D); 12" N/A Spare 8 None NIA - No Pipe U3 (N-49A,D-F) N-50 A to E 12" 1" Instrumentation 8 M-853, ASMECL2EX M-859, M-854 N-50F 12" N/A Spare 8 None N/A - No Pipe N-51 A to D 12" 1" CAC Sample 8 M-867 Unclassified N-51E 12" 1" Recirc Suction Press 8 M-853 ASMECL 1 EX N-51F 12" N/A Spare 8 None N-52 A to D 12" N/A Spare 8 M-853 Unclassified N-52 E 12" 1" Instrumentation 8 M-852 ASMECL 1 EX N-52 F 12" 1" Instrumentation 8 M-833 Unclassified N-53 12" 8" Chilled Water Return A 3 M-827, Unclassified Sh2 N-54 12" 8" Chilled Water Return B M-827, Unclassified .) Sh 2 N-55 12" 8" Chilled Water Supply M-827, Unclassified .) B Sh2 Alio" Science & Technology 2-29 PBT05.G03 Revision (J

Peach Bottom Atomic Power Station Units 2 & 3, Fourtll Interval TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-56 12" 8" Chilled Water Supply "'I M-827, Unclassified A Sh2 N-57 6" I" Main Steam Sample "'I M-851 ASMECL I EX .) N-IOO A-F 12" N/A Neutron Monitoring B is type 3 M-832 Unclassified Electrical N-IOI A-F 12" N/A Recirc. Pump Power "'I M-832 Unclassified Electrical .) N-I02 A 12" N/A Spare "'I M-833/ Unclassified .) M-861 N-I02 BA,BB 12" I" U2 Instrumentation; U3 "'I M-833 ASMECL2 EX Spare N-I02 BC 12" 1" U2 Instruulentation; U3 "'I M-873 Unclassified Spare N-I02 BD 12" "'Ill U2 Spare, U3 "'I M-832 Unclassified .) .) Breathing Air N-I03 A,B 12" N/A A Spare/B A is type 3 M-832 Unclassified Thern10couples N-I04 A-H 12" N/A CRD Rod Position Ind 3 M-853, Unclassified Electrical M-859, M-854 N-I05 A,B,C,D 12" N/A PWR., Lights, Fans, "'I M-832 Unclassified Electrical N-I06 A,B,C,D 12" N/A Indication and Control M-832 Unclassified Electrical N-I07 12" N/A Thermocouples M-832 Unclassified Electrical N-I08 12" N/A Spare "'I None N/A - No Pipe U2 only N-I09 A.B 3" N/A Communication and B is type 3 None Unclassified N-I09B Spare; Lights Thern10couples in personnel air lock Alion Science & Technology 2-30 PBT05.G03 Revision (J

lSI Program PlalZ Peach Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-110 A thru H 16" N/A Stabilizer Inspection ""l M-832 N/A - No Pipe 16" dia opening j Manway in stabilizer insert plate N-58 26" N/A Spare 2 None N/A - No Pipe Unit 3 Only N-150 8" N/A Spare M-832 Unclassified N-200 A,B 54"10 N/A Access Hatch detail E M-832 n/a - no pipe N-201 A to H 6'-9" 6'-9" Drywell to Torus Vents M-832 n/a - no pipe N-202 A to M 18" 18" Drywell Vacuum 4 nJa - no pipe offvent lines, Breakers inside Torus N-203 1" 1" Oxygen Analyzer 4 M-867 unclassified N-205 A,B 20" 20" A Torus Vacuum 4 M-867 Me Breaker B-Drywell and Torus Purge Supply and Vacuum Relief N-206 A,B 2" 2" Level and Pressure 4 M-865 ASMECL 2 EX Instrunlentation N-207 A to H 1" 1" Vent Line Drain 4 - detail G offvent lines, inside Torus N-208 A to M 12" 12" Electromatic Relief 5 through vent Valve Discharge lines inside Torus N-209 A to 0 1 1" Air and Water Temp. 6 M-861 unclassified A&B (C&D) Abandoned N-210 A,B 18" 18" RHR Torus Cooling 5 M-861 ASMECL2EX and Pump Test Line Alion Science & Techn%gy 2-31 PBT()5.G03 Revision (J

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-5 UNITS 2 AND 3 DRYWELL PENETRATIONS Number Nom Process Serviees Penetration QAD Piping Notes Diameter Line Size Type Classification N-211 A,B 6" 6" Containment Cooling 4 M-861 ASMECL2 EX to Spray Header (RHR) N-2I2 12" 12" RCIC Turbine Exhaust 5 M-859 ASMECL2 EX N-2I3 A,B 8" 8" Const. Drain 7 M-832 unclassified N-214 24" 24" HPCI Turbine Exhaust 5 M-865 ASMECL 2 EX N-215 4" Open U2 Instrumentation; U3 4 M-865 ASMECL2EX Ended Spare N-216 4" 4" HPCI Min. Recirc. 5 M-865 ASMECL2 EX N-217 A,B 2" 2" N-217 A Spare; N-217 4 M-865 B ASMECL2EX B HPCI and RCIC VaCUUlTI Relief N-2I8 A,B,C 1" 1" lnstr. Air (A), Spare 4 M-832 C, unclassified (C), 02 Analyzer (B) M-867 A B N-219 18" 18" Purge Exhaust 4 M-867 unclassified N-220 10" N/A ILRT Sensors 4 M-832 unclassified Abandoned during ILRT N-221 2" 2" RCIC Vacuum Pump 5 M-859 ASMECL2 EX Discharge N-223 2" 2" Condensate from HPCI 5 M-865 ASMECL2 EX Turbine Drain Pot N-224 10" 10" Core Spray Test and 5 M-862 ASMECL2 EX Unit 2 Only Flush N-225 6" 6" RCTC PUlnp Suction 5 M-859 ASMECL 2 EX N-226 A to D 24" 24" RHR PUlTIP Suction 5 M-861 ASMECL2 N-227 16" 16" HPCI PUlTIP Suction 5 M-865 ASMECL2 A/ioll Science & Technology 2-32 PBT05.G03 Revision ()

lSI Program Plan Pellch Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.4-5 UNITS 2 AND 3 nRYWELL PENETRATIONS Number Nom Process Services Penetration QAD Piping Notes Diameter Line Size Type Classification N-228 A to D 16" 16" Core Spray Pump 5 M-862 ASMECL2 Suction N-229 M-862 ASMECL2 EX Unit 2 Only I 6" 6" Core Spray Min.Flow 5 N-230 4" 4" RCIC Punlp Min. 5 M-859 ASMECL2 EX Recirc N-23l A,B N/A Spares 4 M-832 unclassified Electrical N-232 10" N/A Spare 5 None n/a - no pipe N-250 8" Open U2 - Spare / U3-Level 4 M-832/ ASMECL2 EX Ended Instrument M-865 N-233 10" 10" HPCI & RCIC Test 5 M-865 ASMECL2EX Unit 2 Only Flush N-234 10" 10" Core Spray Test and 5 M-862 ASMECL2 EX Flush N-235 4" 4" HPCI and RCIC Test 5 M-865 ASMECL2 EX Unit 3 Only and Flush N-236 A,B 4" 4" Core Spray Min. Flow 5 M-862 ASMECL2 EX Unit 3 Only Alion Science & Technology 2-33 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval 2.5 Technical Approach and Positions When the requiretnents of ASME Section XI are not easily interpreted, PBAPS has reviewed general licensing/regulatory requirements and industry practice to determine a practical method ofin1pletnenting the Code requiretl1ents. The Technical Approach and Positions (TAPs) contained in this section have been provided to clarify PBAPS's itnplementation ofASME Section XI requiren1ents. An index which sUlnmarizes each technical approach and position is included in Table 2.5-1. AliO/1 Science & Tecllll%gy 2-34 P/JT05.G03 Revision ()

lSI Program Plan Peac" Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX Position Revision Status l (Prograln) Description of Technical Approach Number Date2 and Position 0 (SPT) Systen1 Leakage Testing of Non-IsolabIe 14T-Ol Active 11105/08 Buried Components. 0 14T-02 Active (SPT) Valve Seats as Pressurization Boundaries. 11/05/08 Note 1: lSI Program Technical Approach and Position Status Options: Active - Current Technical Approach and Position is being utilized at PBAPS; Deleted - Technical Approach and Position is no longer being utilized at PBAPS. 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 lSI Program Plan revision when theTechnical Approach and Position was incorporated into the document. Alion Science & Tecll/lology 2-35 PBT05.G03 Revision 0

lSI Progrtll1l Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER I4T-Ol Revision 0 COlVIl>ONENT IDENTIFICATION: Code Class:

Reference:

Exmnination Category: !tenl Nlunber:

== Description:== Conlponent NUlnber: CODE REQUIREMENT: 2 and 3 IWA-5244(b)(2) C-H,D-B C7.10, D2.10 Systenl Leakage Testing ofNon-IsolabIe Buried Conlponents Non-Isolable Buried Pressure Retaining COlnponents Paragraph IWA-5244(b)(2) requires non-isolable buried conlponents be tested to confirm that flow during operation is not itnpaired. ]>OSITION: Aliicle 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, PBAPS has established the following acceptance criteria: For opened ended lines on systems that require Inservice Testing (1ST) ofplunps, adherence to 1ST acceptance criteria is considered as reasonable proof of adequate flow through the lines. For lines in \\vhich the open end is accessible to visual examination while the systenl 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 systenls where the process fluid is pneumatic, evidence of gaseous discharge shall be considered reasonable proofof adequate flow through the open ended line. Such test nlay include passing snl0ke through the line, hanging balloons or streanlers, using a remotely operated blullp, using thennography to detect hot air, etc. This acceptance criteria \\vill be utilized in order to nleet the requirenlents of Paragraph IWA-5244(b)(2). PBAPS 's position is that proof of adequate flow is all that is required for testing these open ended Iines and that no further visual examination is necessary. This is consistent with the requireillents for buried piping, which is not subject to visual eXaIllination. Alion Science & Technology 2-36 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Statio" Units 2 & 3, Fourth I"terval TECHNICAL APPROACH AND POSITION NUMBER I4T-02 Revision 0 COMPONENT IDENTIFICATION: Code Class:

Reference:

EXaIllination Category: Item Nmnber:

== Description:== Conlponent Nunlber: CODE REQUIREMENT: 1, 2, and 3 IWA-5221 IWA-5222 B-P, C-H, D-B 815.10, C7.10, D2.10 Valve Seats as Pressurization Boundaries All Pressure Testing Boundary Valves Paragraph IWA-5221 requires the pressurization boundary for system leakage testing extend to those pressure retaining conlponents under operating pressures during normal systeln service. POSITION: PBAPS's position is that the pressurization bOlmdary extends up to the valve seat of the valve utilized for isolation. For example, in order to pressure test the lSI Class 1 C0111pOnents, the valve that provides the Class break would be utilized as the isolation point. In this case the true pressurization boundary, and Class break, is actually at the valve seat. Any requirelnent to test beyond the valve seat is dependent only on whether or not the piping on the other side of the valve seat is lSI Class 1,2, or 3. In order to silnplif)r exanlination of classed cOlnponents, PBAPS will perfOlm a VT-2 visual eXaIllination of the entire boundary valve body and bonnet (during pressurization up to the valve seat). Alioll Science & Tecllllology 2-37 PIJT05.G'03 Revisioll 0

lSI Progr(l/11 Plan Peach Bottom Atomic Power Statloll Units 2 & 3, Fourtlt Interval 3.0 COl\\lPONENT lSI PLAN The PBAPS COluponent lSI Plan includes ASME Section XI nonexenlpt pressure retaining welds, piping structural eleillents, pressure retaining bolting, attachnlent welds, pump casings, and valve bodies of lSI Class 1, 2, and 3 components that Ineet the criteria of Subarticle IWA-1300. These components are identified on the lSI Boundary Drawings listed in Section 2.3, Table 2.3-2. Procedure ER-AA-330-002, "Inservice Inspection of Welds and COlnponents", inlplenlents the ASME Section XI Welds and Con1ponents lSI Plan. This COInponent lSI Plan also includes conlponent augn1ented inservice inspection program inforn1ation specified by docun1ents other than ASME Section XI as referenced in Section 2.2 of this document. See the Auglnented Inspection Plan for details on these additional requirements and references. The RPV interior, core support structures, and interior attachnlents have been transferred to procedure ER-PB-331-100I, "RPV & Internals Program Basis and Itnplementation Document" per Relief Request 14R-49. 3.1 Nonexen1pt lSI Class Components The lSI Class 1,2, and 3 con1ponents subject to exanlination are those that are not exenlpted under the criteria of Paragraphs IWB-1220, IWC-1220, and IWD-1220, respectively. A SUlnlnary of PBAPS Units 2, 3, and COlnn10n ASME Section XI nonexenlpt conlponents is included in Section 7.0. 3.1.1 Identification of lSI Class 1, 2, and 3 Nonexelnpt COIllponents lSI Class 1, 2, and 3 nonexen1pt components are identified on the lSI Isometric and C0111ponent Drawings listed in Section 2.4, Table 2.4-1. Welded attac1m1ents are also identified by controlled PBAPS individual support detail drawings. 3.2 Risk-Informed Examination Requirements Piping structural elelllents that fall under RISI Examination Category R-A are risk ranked as High (1, 2, and 3), Mediunl (4 and 5), and Low (6 and 7). Per the EPRI Topical Reports TR-112657, Rev. B-A, and Code Case N-578-1, piping structural elen1ents ranked as High or Medium Risk are subject to exmnination while piping structural elements ranked as Low Risk are not subject to exatninations (except for pressure testing). Thin wall welds that were excluded from volUlnetric examination under ASME Section XI rules per Table IWC-2500-1 are included in the elelnent scope that is potentially subject to RISI examination at PBAPS. (PBAPS has "thin wall" welds, but 1110St of them were evaluated and ranked as Low Risk Category "6 or 7". Therefore, none of theln were selected under RISI.) Piping structural elelnents may be excluded fr01n examination (other than pressure testing) under the RISI Progratn if the only degradation n1echanisl11 present for a Alioll Science & Tecllllology 3-/ PBT05.G()3 Revisloll ()

lSI Program Pilln Pellch Bot/om Atomic Power Stlltioll Units 2 & 3, FOllrth Interval given location is inspected for cause under certain other PBAPS progranls such as the Flow Accelerated Corrosion (FAC) or IGSCC Progrmus. These piping structural elements will ren1ain part of the FAC or IGSCC Progratus, which already perfornl "for cause" inspections to detect these degradation mechanisn1s. Piping structural eleillents susceptible to FAC or IGSCC along with another degradation mechanisn1 (e.g., thermal fatigue) are retained as part of the RISI scope and are included in the elelnent selection for the purpose of performing exanlinations to detect the additional degradation Inechanislu. The RISI Progran1 element exanlinations are performed in accordance with Relief Request 14R-44. 3.3 Reactor Coolant Pressure Boundary Normal Makeup EXetllption In accordance with ASME Section XI, Paragraph IWB-1220(a), con1ponents that are connected to the reactor coolant pressure boundary may be exen1pted from the surface and volmuetric examination requirenlents of ASME Section Xl, provided they are of such a size and shape that upon a postulated pressure boundary rupture, the resulting flow of coolant under nonual operating conditions is within the capacity of luakeup systenls. 3.3.1 Makeup Calculation The basis for determining the n1akeup size exetuption of lSI Class 1 water and stemu lines is provided in PBAPS Calculation No. ME-34, Revision 2, dated 9/15/95. The makeup flow rate is deter111ined fro111 systelus which are not part ofthe emergency core cooling system and which are operable from on-site emergency power. The reactor coolant makeup systenls at PBAPS consist of the following: System Pump Maximum Emergency Flow Rate Fluid Temp. Power RCIC 600 GPM 95° F

Yes, UFSAR, Section 4.7 On-site CRD 176 GPM 95° F

Yes, UFSAR, Section 3.4 On-site Note that RCIC does not perfornl an ECCS function, therefore, this systenl meets the ASME Section XI criteria for inclusion as a nlakeup source under IWB-1220(a). This same approach applies to the alternative repair and replacement requirenlents of IWA-4131.1 (a)(2).

Based on Calculation ME-34, the following lSI Class 1 piping qualifies for the Inake-up capacity exenlption of Paragraph IWB-1220(a): Alioll Science & Technology 3-2 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Statloll UllitS 2 & 3, Fourt" Interval 1. Steanl system piping with an inside dimneter (ID) of2.9974" and smaller. 2. Water system piping subject to recirculation pU111p discharge pressure with an ID of 1.4831" and s111aller. 3. Water systeln piping not subject to recirculation punlp discharge pressure with an ID of 1.5488" and slnaller. 3.3.2 Application to Peripheral CRD Housing Welds ASME Section Xl, Table rWB-2500-1, Exan1ination Category 8-0, Itenl Nunlber B14.10 addresses pressure-retaining welds in the peripheral CRD Housings. For those C0111pOnents which fall within this category that are not exenlpt, examination is required for ten percent of the peripheral housings. However, Stone and Webster Calculation No. PM-0945, Revision 1, dated 11/30/95, addresses the exelnption of these conlponents. This calculation demonstrates that the water makeup capability of the RCIC and CRD systems is greater than the potential leakage due to a failure of a CRD housing. As discussed in Section 3.3.1 above, the reactor coolant makeup systems for PBAPS are RCIC and CRD. At the power rerate dome pressure, the total makeup capacity of these systems is approxin1ately 698 gpm. Taking into account the design and geonletry of the control rod drive units, the maximum calculated CRD leakage is approximately 604 gpm. Based on adequate tnakeup systenl capability, the welds in the peripheral CRD housings are exelnpted frOITI exanlination per IWB-1220(a), and the requirenlents of Exanlination Category 8-0 do not apply. Alioll Sciellce & Tecllllology 3-3 PBT05.G03 Revisioll 0

lSI Program Piau Peach Bottom Atomic Power Statioll Units 2 & 3, Fourth Illterval 4.0 SUI>PORT lSI PLAN The PBAPS Support lSI Plan includes the supports of ASME Section XI nonexempt lSI Class 1, 2, and 3 cOlnponents as described in Section 3.0; and CISI Class MC conlponents as described in Section 6.0. Procedure ER-AA-330-003 "Visual Exmnination of Section XI COlnponent Supports", ilnplelnents the ASME Section XI Support lSI Plan. 4.1 Nonexenlpt lSI Class Supports The PBAPS lSI Class 1, 2, 3, and MC nonexelnpt supports are those \\vhich do not nleet the exemption criteria of Paragraph IWF-1230 of ASME Section XI. A summary of PBAPS ASME Section XI nonexempt supports is included in Section 7.0. 4.1.1 Identification of lSI Class 1, 2, and 3 Nonexen1pt Supports lSI Class 1, 2, and 3 supports are identified on the lSI Isonletrics and Conlponent Drawings listed in Section 2.4, Table 2.4-1. Supports are identified by controlled PBAPS individual support detail drawings. CISI Class Me supports are identified on the CISI Boundary Drawings listed in Section 2.3, Table 2.3-3. 4.2 Snubber Examination and Testing Requirenlents 4.2.1 ASME Section XI Paragraphs IWF-5200(a) and (b) and IWF-5300(a) and (b) require VT-3 Visual Exanlination and Inservice Tests of snubbers to be perforn1ed in accordance with the Operation and Maintenance of Nuclear Power Plants Standard (ASME OM Code). As allowed by 10CFR50.55a(b)(3)(v), PBAPS will use Subsection ISTD, "Inservice Testing of Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants," ASME OM Code, 2001 Edition through the 2003 Addenda, to nleet the requirenlents in ASME Section XI Paragraphs IWF-5200(a) and (b) and IWF-5300(a) and (b). Per 10CFR50.55a(b)(3)(v), visual exanlinations shall be perfonned using the VT-3 visual examination nlethod described in Paragraph IWA-2213. A SU1111nary of the PBAPS Units 2, 3, and Comnl0n safety-related and non-safety related snubbers is included in Section 7.0. The details of the PBAPS snubber prograrn and exalllination requirements are located in Appendix AUG-05 of the Auglnented Inspection Plan. Procedure ER-AA-330-004, "Visual Exalnination of Technical Specification Snubbers", implenlents the visual inspection progrmn for safety-related and non safety-related snubbers. Procedures ER-AA-330-010, "Administration of Snubber Functional Testing", ER-AA-330-011, "Snubber Service Life Monitoring Progran1", and station Alioll Science & Tecllllology 4-1 PBT05.G03 Revisioll ()

lSI Program Pla11 Peach Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval surveillance test procedures are used to implement the functional testing and service life monitoring requirements for safety-related and non safety-related snubbers. The ASME Section Xl lSI PrograITI uses Subsection IWF to define support inspection requirements. The lSI PrograITI lnaintains the Code Class snubbers in the populations subject to inspection per Subsection I\\VF. This is done to acconlluodate scheduling and inspection requirenlents of the related attachillent hardware per Paragraphs IWF-5200(c) and IWF-5300(c). (See Section 4.2.2 below.) 4.2.2 ASME Section XI Paragraphs IWF-5200(c) and IWF-5300(c) require integral and non-integral attachnlents for snubbers to be examined in accordance with Subsection IWF of ASME Section Xl. This results in VT-3 visual exanlination of the snubber attachlnent hardware including the bolting, pins, and their interface to the claInp, but does not include the component-to-clalTIp interface. The ASME Section XI lSI Program uses Subsection IWF to define the inspection requireluents for allISl Class 1, 2, and 3 supports, regardless of type. The lSI Progran1 nlaintains the Code Class snubbers in the suppot1 populations subject to inspection per Subsection IWF. This is done to facilitate scheduling and inspection requirements of the snubber attachment hardware (e.g., bolting and pins) per Paragraphs IWF-5200(c) and IWF-5300(c). It should be noted that the eXaIllination of snubber welded attachments will be performed in accordance with the ASME Section XI Subsections IWB, IWC, and IWD welded attachment exanlination requirenlents (e.g., EXaIllination Categories B-K, C-C, and D-A). Alioll Science & Technology 4-1 PBT05.G03 Revisioll ()

lSI Progrtll1l Plan Peach Bot/om Atomic Power Statioll Units 2 & 3, FOllrth Illterval 5.0 SYSTEM PRESSURE TESTING lSI PLAN The PBAPS Systenl Pressure Testing (SPT) lSI Plan includes all pressure retaining ASME Section XI, lSI Class 1, 2, and 3 conlponents, with the exception of those specifically excluded by Paragraphs IWA-511 O(c), IWC-5222(b), and IWD-5240(b). All RISI piping structlu'al elements, regardless of risk classification, reluain subject to pressure testing as part ofthe cutTent ASME Section XI progranl. The SPT lSI Plan details systetu pressure tests and visual inspections on the lSI Class 1, 2, and 3 pressure retaining components to verify systenl and component structural integrity. This progranl conducts both Periodic and Interval (1 0-Year frequency) pressure tests as defined in ASME Section XI Inspection Program B. Procedure ER-AA-330-00 1, "Section XI Pressure Testing," as well as several PBAPS site-specific test procedures, implement the ASME Section XI Systelll Pressure Testing lSI Plan. Additional COlll111itments regarding the PBAPS SPT lSI Plan are discussed in Section 2.2 of this docunlent. 5.1 lSI Class Systen1s All lSI Class 1 pressure retaining conlponents, typically defined as the reactor coolant pressure boundary, are required to be tested. Those portions of lSI Class 2 and 3 systenls that are required to be tested include the pressure retaining boundaries of cOlllponents required to operate or support the system safety functions. lSI Class 2 and 3 open ended discharge piping and cOlllponents are excluded from the exanlination requirenlents per Paragraphs IWC-5222(b) and IWD-5240(b). 5.1.1 Identification oflSI Class 1, 2, and 3 Conlponents All conlponents subject to ASME Section XI System Pressure Testing are shown \\vithin the lSI classification color coding on the PBAPS lSI Boundary Drawings listed in Section 2.3, Tables 2.3-1. Additional information on the classification of various systelll boundaries is provided in the lSI Classification Basis Document. 5.1.2 Identification of Systelu Pressure Tests The System Pressure Test Boundary Dra\\vings used to define which systems, or portions of systenls, fall under a specific test are idcntified in attachments to the site-specific test procedures. Individual tests arc identified and maintained in the PBAPS lSI Database. 5.2 Risk-Informed Exanlinations of Socket Welds Socket welds selected for exaluination under the RISI program are to be inspected with a VT-2 visual exalllination each refueling outage per ASME Code Case Alion Science & Techllology 5-1 PBT05. G03 Revisioll ()

lSI Progrtllll Pla/l Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval N-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socket welds selected for inspection under the RISI progrmll shall be pressurized each refueling outage during a system pressure test in accordance \\vith Paragraph IWA-5211 (a). Alion Science & Technology 5-1 PB T05. G03 Revision 0

lSI Progrtll1l Pia" Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interl'lll 6.0 CONTAINMENT lSI PLAN The PBAPS Containment lSI Plan includes ASME Section XI CISI Class MC pressure retaining conlponents and their integral attachments that Ineet the criteria of Subarticle IWA-1300. This Containment lSI Plan also includes infornlation related to augnlented exanlination areas, conlponent accessibility, and examination review. PBAPS has no CISI Class CC components, which nleet the criteria of Subarticle IWL-I100, therefore, no requirenlents to perfonn examinations in accordance with Subsection IWL are incorporated into this Containnlent lSI Plan. Since both PBAPS unit's containnlent vessels are free-standing structural steel contaimnent vessel, only Subsection IWE is applied. The examination of containnlent cOlnponents are perfornled per procedures ER-AA-330-007, "Visual Exmnination of Section XI Class MC Surfaces and Class CC Liners", ER-AA-330-008, "Exelon Service Levelland Safety Related (Service Level 3) Protective Coatings", ER-AA-335-004, "Manual Ultrasonic Measurement of Material Thickness", and ER-AA-335-018, "Detailed, General, VT-I, VT-1C, VT-3, and VT-3C, Visual Exmnination of ASME Class MC and CC Containment Surfaces and COlnponents" 6.1 Nonexenlpt CISI Class Components The PBAPS CISI Class MC conlponents identified on the CISI Boundary Drawings are those not exempted under the criteria of Paragraph IWE-I220 in the 2001 Edition through the 2003 Addenda of ASME Section XI. A summary of PBAPS ASME Section XI nonexempt CISI conlponents is included in Section 7.0. The process for scoping PBAPS c0111ponents for inclusion in the Contaimnent lSI Plan is included in the containment sections of the lSI Classification Basis Docunlent. These sections include a listing and detailed basis for inclusion of containment components. Conlponents that are classified as CISI Class MC nlust Ineet the requirements of ASME Section XI in accordance with 10CFR50.55a(g)(4). Although CISI Class MC supports of Subsection IWE CO!11ponents are not strictly required to be exanlined in accordance with 10CFR50.55a(g)(4)(v), PBAPS has elected to perfoflll these exanlinations in accordance with Augmented Inspection Prograln (AUG-CA). AUG-CA describes the inspection program for CISI Class Me supports at PBAPS. For guidance, this progrmn is based on the support exmnination criteria of ASME Section XI, Subsection IWF. Alion Science & Technology 6-1 PBT05.G03 Revision 0

lSI Program Pia" Peach Bol1om Atomic Power Station Units 2 & 3, Fourth Interval 6.1.1 Identification ofCISI Class MC Nonexenlpt COlnponents CISI Class :NIC cOlnponents are identified on the PBAPS CISI Boundary Dra\\vings listed in Section 2.3, Table 2.3-3. The PBAPS Drywell Penetrations are identified in Section 2.4, Table 2.4-5. 6.1.2 Identification of CISI Class MC Exempt C0111pOnents Certain containment components or pm1s of cotnponents Inay be exenlpted fron1 examination based on design and accessibility per the requiren1ents of Paragraph IWE-1220. The process for exempting PBAPS cotnponents from the Contail1lnent lSI Plan per Paragraph IWE-1220 is included in the containment sections of the lSI Classification Basis Document. These sections include discussions of exenlpt cOluponents and the bases for those exen1ptions. 6.2 Auglnented Exan1ination Areas Metal contaim11ent components potentially subject to auglnented exanlination per Paragraph IWE-1240 have been evaluated in the containment sections of the lSI Classification Basis Document. These sections define the areas that are subjected to augn1ented examination. A significant condition is a condition that is identified as requiring application of additional augn1ented exmnination requirelnents under Paragraph IWE-1240. 6.2.1 In the First CISI Interval, the Augn1ented Inspection Progranl AUG-C I addressed wetted and sublnerged Suppression Chan1ber (Torus) Interior Surfaces for both PBAPS Units 2 and 3. However, this Augnlented Inspection Progran1 is no longer active since the First CISI Interval Relief Requests CRR-I0 and CRR-II were not resublnitted for the Second CISI Interval. In accordance with ASME Section XI, Paragraph IWE-1241, augnlented exanlinations were to be performed on those surface areas, which were likely to experience accelerated degradation and aging. At PBAPS, this included the wetied (i.e., itnn1ersion zone) and subtuerged portions of the suppression chamber. These areas have undergone exmninations in the past to quantify and evaluate coating problen1s and pitting. Results of exalninations perforn1ed in both Units in 1991, again in PBAPS Unit 3 in 1997, and in PBAPS Unit 2 in 1998, revealed nU111erOUS pits and degraded coatings. In conjunction with these results, an evaluation was performed in ABB In1pell Report No. 03-0670-1360. This report concluded that the structural integrity of the suppression chan1ber was maintained, and continued operation was justified. The report also established an Alion Science & Technology 6-2 PBT05.G03 Revision 0

lSI Program Plan Pellcll Bottom Atomic Power Statioll Units 2 & 3, Fourtll Interval evaluation methodology, acceptance criteria, and a suggested reexamination schedule. In 2005, MPR Associates perforn1ed an evaluation of the differences between PBAPS Units 2 and 3 water chen1istry and corrosion rates. In 2006, 1000/0 of the wetted surface of the Unit 2 torus was inspected and pit depth IneaSUre111ents were taken. The torus was detennined to be structurally sound until 2010. The results of this inspection are documented in PIMS AR A 1545908 Eval 38. In 2007, 1000/0 of the wetted surface of the Unit 3 torus was inspected and pit depth measuretnents were taken. The torus was detern1ined to be structurally sound until 2011. The results of this inspection are docu111ented in PIMS AR A1554416 Eva132. Based upon the examination results and discussions, the in1n1ersion zone and submerged surfaces in the suppression chan1ber in both units were classified as augn1ented exan1ination areas subject to Exan1ination Category E-C. However, in lieu of the exan1inations stated in Table IWE-2500-1, Examination Category E-C, PBAPS followed approved Relief Request CRR-II that established alternative exatnination criteria. Relief Request CRR-II utilized results from previous exmninations and evaluations to implement an alternative examination progratn that provided an acceptable level of quality and safety in accordance with IOCFR50.55a (a)(3)(i). There were no other Exan1ination Category E-C augmented exmnination surfaces identified during the developtnent of this lSI Program Plan, as of its issuance. This conclusion was based on a review of design docUlnents, and satisfactory results from thorough, docUlnented exmninatiolls of dry, accessible contail1lnent surfaces. In 1985, three typical drywell surfaces were selected as representative examination areas. These three areas have been visually exan1ined on a regular basis since 1985. In 1992 (Unit 2) and 1993 (Unit 3), the exalninations were expanded to include a general area visual exatnination of the drywell and suppression chamber as well as the specific visual exatninations on the three representative areas. 6.2.2 In the First CISI Interval, these wetted and subn1erged suppression chat11ber (Torus) surfaces were identified as augtnented surface areas requiring examination in accordance with Table IWE-I240. These surface areas had been categorized in accordance with ASME Section XI, Table IWE-2500-1, Examination Category E-C, Iten1 Number E4.II, requiring visual examination of 100% of the surface areas identified during each inspection period until the areas exan1ined remain essentially unchanged for the next three inspection periods. In the Second CISI Interval, these wetted and submerged suppression chamber (Torus) augn1ented surface areas will require visual exan1ination of 1000/0 of the surface areas identified during each inspection period until Alioll Science & Technology 6-3 PBT05.G03 Revision ()

lSI Program PIa" Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval the areas examined remain essentially unchanged for the next inspection period. Once an augn1ented area remains unchanged for one full period, the areas fall back to the nonnal EXaJllination Category E-A exan1ination schedule. 6.3 Component Accessibility CISI Class ~1C pressure retaining components subject to exan1ination shall relnain accessible for either direct or ren10te visual eXaJnination fi:on1 at least one side for the life of the plant per the requirements of ASME Section XI, Paragraph IWE-1230. Paragraph I\\VE-1231(a)(3) requires 800/0 of the pressure-retaining boundary that was accessible after construction to ren1ain accessible for either direct or rernote visual exan1ination, froln at least one side of the vessel, for the life of the plant. Portions of con1ponents elnbedded in concrete or otherwise made inaccessible during construction are exen1pted from exan1ination, provided that the requirements of ASME Section XI, Paragraph IWE-1232 have been fully satisfied. (Note that Relief Request 14R-48 (CRR-13) has been approved to take credit for the integrated leak rate testing in accordance with the PBAPS Appendix J progran1 in lieu of performing the required exan1inations for Penetration N-3 which does not meet the exen1ption criteria of Paragraph IWE-1232(a)(2).) In addition, inaccessible surface areas exempted from examination include those surface areas where visual access by line of sight with adequate lighting from pernlanent vantage points is obstructed by pern1anent plant structures, equipn1ent, or components; provided these sUlface areas do not require examination in accordance with the inspection plan, or auglllented exan1ination in accordance with Paragraph IWE-1240. 6.4 Responsible Individual ASME Section XI Subsection IWE requires a Responsible Individual to be involved in the development perfonnance, and review of the CISI examinations. The Responsible Individual assigned to perfonn these duties shall n1ect the requirernents of ASME Section XI, Paragraph IWE-2320. 6.5 Structural Attachlnents Surfaces of attachn1ent welds between structural attachll1ents and the containn1ent pressure boundary or reinforcing structure will be subject to exanlination per Subsection IWE. To establish which containn1cnt attachnlents are subject to this exanlination requirell1ent, guidance has been taken froln ASME Section HI, Subsection NE. In accordance with ASME Section III, Subsection NE, Paragraph NE-l132.1 (d), structural attachn1ents are those attachll1ents that perfornl a Alion Science & Technology 6-4 PBT05.G03 Revision 0

lSI Progrtllll Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval pressure retaining function or are in the containn1ent vessel support load path. Therefore, exmninations will be required on \\velded attachments associated with the containnlent vessel supports, such as the suppression chmnber column supports. However, examinations will not be required on welded attachments associated with con1ponents that are not pressure retaining or are not in the containment vessel support load path, such as pipe supports, stairways and structural steel. Alion Science & Technology PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Inter}/al 7.0 COMPONENT

SUMMARY

TABLES 7.1 Inservice Inspection SUmlTIary Tables The following Table 7.1-1 and 7.1-2 provide a sunl1nary of the ASME Section XI pressure retaining components, supports, systenl pressure testing, and augnlented progranl C0111pOnents for the Fourth lSI Interval and Second C1S1 Interval at PBAPS. The fonnat of the Inservice Inspection Sunlnlary Tables is as depicted below and provides the following information: Examination Item Number (or Description Exam Total Number Relief Request/ Notes Category (with Risk Category Requirements of Com ponents TAP Number Examination Number or by System Category Augmented Description) Number) (1) (2) (3) (4) (5) (6) (7) (l) Exan1ination Category (with Exmnination Category Description): Provides the Examination Category and description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1. and IWF-2500-1. Only those Exmnination Categories applicable to PBAPS are identified. Exan1ination Category "R-A" frOln Code Case N-578-1 is used in lieu of ASME Section XI Exanlination Categories B-F, B-J, C-F-l, and C-F-2 to identify lSI Class 1 and 2 piping structural elements for the RISI Progrmn. Exan1ination Category "AUG" is used to identify Auglnented lSI cOlnponents and other PBAPS conl1nitments. (2) Iteln NUlnber (or Risk Category Nunlber or Auglnented Nunlber): Provides the Item Nunlber as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-I. Only those ItelTI Nun1bers applicable to PBAPS are identified. For piping structural elements under the RISI Progran1, the Risk Category NWllber (e.g., 1-5) is used in place of the Itenl Number. Specific abbreviations such as reference paragraph nunlbers (AUG-Ol, AUG-02, AUO-05, AUG-12, AUO-15, AUG-D, AUO-C2, AUG-C3, AUG-CA, AUO-CB, AUO-CC, and AUG-CD) have been developed to identify Augn1ented lSI exanlinations and other PBAPS COnl111itments. Alion Science & Tecllllology 7-1 PBT05. G03 Revision 0

lSI Program Pla1l Pellch Bottom Atomic Power Slation Units 2 & 3, FOllrth Interval (3) henl Nllnlber (or Risk Category Nunlber or Allglnented Nunlber)

Description:

Provides the description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, lWE-2500-1, and IWF-2500-1. For Risk-Infonned piping exmninations, a description of the Risk Category Nunlber is provided. For augmented inspection conlmitments, a description of the auglnented requiretnent is provided. (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, and IWF-2500-1. Provides the examination requirelnents for piping structural elelnents under the RISI Progrmn that are in accordance with the EPRI Topical Reports TR-112657, Rev. B-A, and Code Case N-578-1. Provides the examination requirelnents for augmented cOlnponents from PBAPS cOlnnlilnlents or relief requests. (5) Total Number Of Components by Systetn Provides the system designator (abbreviations). See Section 2.3, Table 2.3-1 for a list of these systems. This column also provides the nmnber of components within a particular system for that ASME Section XI Item Nunlber, Risk Category Nunlber, or Augnlented Nunlber. Note that the total nlllnber of C0111pOnents by systenl are subject to change after cOlnpletion of plant n10dificatiol1s, design changes, and lSI systenl classification updates, and will be Inaintained only within the lSI Database. (6) Relief Request/Technical Approach & Position Number Provides a listing of Relief Request/TAP Nunlbers applicable to specific components, the ASME Section XI Itenl Number, Risk Category NU111ber, or Auglnented Nunlber. Relief Requests and TAP Numbers that generically apply to all components, or an entire class are not listed. If a Relief Request/TAP Nunlber is identified, see the corresponding relief request in Section 8.0 or the TAP NU111ber in Section 2.5. /{lion Science & TecllJlology 7-2 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Statioll Ullits 2 & 3, FOlirthIllterl'al (7) Notes Provides a listing of program notes applicable to the ASME Section XI !tenl Nun1ber, Risk Category Number, or Augmented NUll1ber. If a progran1 note nunlber is identified, see the corresponding progran1 note in Table 7.1-3. Alion Science & Technology 7-3 PBT05.G03 Revisioll 0

lSI Program PlalZ Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-A B1.11 Circumferential Shell Welds (Reactor Vessel) Volumetric RPV:5 I4R-41 Pressure Retaining B1.12 Longitudinal Shell Welds (Reactor Vessel) Volumetric RPV: 15 14R-41 WeIds in Reactor VesseI Bl.21 Circumferential Head Welds (Reactor Vessel) Volumetric RPV: 3 81.22 Meridional Head Welds (Reactor Vessel) Volumetric RPV: 16 81.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric RPV: 1 B1.40 Head-to-Flange Weld (Reactor Vessel) Volumetric & RPV: 1 Surface Bl.51 Beltline Region Repair Weld (Reactor Vessel) Volumetric RPV: 1 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 31 11 Full Penetration Welds ofNozzles in Vessels B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 31 Alion Science & Technology 7-4 PBT05.G03 Rel'i.fiiion (J

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief RequestJ Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-G-I B6.IO Closure Head Nuts (Reactor Vessel) Visual, VT-I RPV: 2 (92 nuts) Pressure Retaining B6.20 Closure Studs (Reactor Vessel) Volumetric RPV: 2 (92 studs) 16 Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 2 (92 threads) 2 in. In Diameter B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-I RPV: 6 (276 washers, bushings) B6.180 Bolts and Studs (Pumps) Volumetric MR:2 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-l MR:2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-l MR:2 B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-l MS: 12 Pressure Retaining RPV: 3 Bolting, 2 in. and Less B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-I CS:2 In Diameter FW:4 HPCI: 1 MR:4 MS: 21 RCIC: 2 RHR:2 RWCU:2 B7.80 Bolts, Studs, & Nuts in CRD Housing (Reactor Vessel) Visual, VT-l RPV: 1 8 (185 housings) Alion Science & Technology 7-5 PBT05. G03 Revisioll 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Com ponents by TAP Number Description) System B-K BIO.IO Welded Attachments (Pressure Vessels) Surface or RPV:9 12 Welded Attachments for Volumetric Vessels, Piping, Pumps, BI0.20 Welded Attachments (Piping) Surface CS: 6 and Valves FW: II HPCI: 2 MR:4 MS: 34 RWCU:2 BI0.30 Welded Attachments (Pumps) Surface MR:7 B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 MR:2 Pump Casings B-M-2 B12.50 Valve Bodies (Exceeding NPS 4) (Valves) Visual, VT-3 CS: 7 Valve Bodies FW:8 HPCI: 4 MR:4 MS: 21 RCIC: 2 RHR:9 RWCU: 6 Alion Science & Technology 7-6 PBT05. G03 Revision ()

lSI Program Plan Peaclt Bottom Atomic Power Stlltion Units 2 & 3, Fourtlt Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-N-I B13.1O Vessel Interior (Reactor Vessel) Visual, VT-3 RPV: 2 I4R-49 13 Interior of Reactor Vessel B-N-2 BI3.20 Interior Attachments Within Beltline Region (Reactor Visual, VT-I RPV: 16 I4R-49 13 Welded Core Vessel) Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region (Reactor Visual, VT-3 RPV: 38 14R-49 13 Attachments to Vessel) Reactor Vessels 813.40 Core Support Structure (Reactor Vessel) Visual, VT-3 RPV:8 I4R-49 13 B-O B14.10 Welds in eRD Housing (Reactor Vessel) Volumetric or NA 9 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings Alion Science & Technology 7-7 PBT05. G03 Revision 0

lSI Program Plan Pellclt Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-P B15.10 System Leakage Test (IWB-5220) Visual, VT-2 RPV:3 14R-47 All Pressure I4T-02 Retaining Components Alion Science & Technology 7-8 PBT05.G03 Revisiol1 (J

lSI Program PlalZ Peaclt Bottom Atomic Power Statioll Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System C-A CLIO Shell Circumferential Welds (Pressure Vessels) Volumetric RHR: 8 Pressure Retaining Welds in Pressure Vessels C-B C2.3l Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface RHR:8 Pressure Retaining Nozzle with Reinforcing Plates in Vessels, Greater than 1/2" Welds in Vessels Nominal Thickness (Pressure Vessels) C2.33 Nozzle-to-Shell (or Head) Welds with Reinforcing Plates Visual, VT-2 RHR: 8 when Inside of Vessel is Inaccessible, Greater than 1/2" Nominal Thickness (Pressure Vessels) C-C C3.10 Welded Attachments (Pressure Vessels) Surface RHR: 12 12 Welded Attachments C3.20 Welded Attachments (Piping) Surface CS: 31 for Vessels, Piping, HPCI: 19 Pumps, and Valve MS: 14 RHR: 60 SDV: I C3.30 Welded Attachments (Pumps) Surface CS:4 RHR:4 Alion Science & Technology 7-9 PBT05. G03 Revisioll 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System C-H C7.10 System Leakage Test (IWC-5220) Visual, VT-2 CAC: 2 I4R-25 AII Pressure CAO:4 I4T-Ol Retaining Components CS: 2 14T-02 HPCI: 1 MS: I RCIC: I RHR: 5 RPV: 1 SLC: 2 SOY: 1 Alion Science & Technology 7-10 PBT05.G03 Rel'i...iol1 ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System O-A 01.20 Welded Attachments (Piping) Visual, VT-l ECW: 63 + 10 Welded Attachments for ESW: 45 Vessels, Piping, Pumps, HPSW: 37 and Valves MSRV: 96 O-B 02.10 System Leakage Test (IWD-5221) Visual, VT-2 ECW: I + 14R-46 15 All Pressure ESW: 1 + I4T-0 I Retaining Components HPSW: I I4T-02 Alion Science & Teclllwlogy 7-/ I PBT05.G03 Revisioll 0

lSI Program PfalZ Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Cateoorv Number Requirements Components TAP Number b Description) E-A E1.11 Containment Vessel Pressure Retaining Boundary-General Visual 19 14R-48 Containment Surfaces Accessible Surface Areas £ 1.11 Containment Vessel Pressure Retaining Boundary-Visual, VT-3 38 14R-48 5 Bolted Connections, Surfaces EI.12 Containment Vessel Pressure Retaining Boundary-Visual, VT-3 1 14R-48 6 Wetted Surfaces of Submerged Areas El.20 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 1 6 BWR Vent System Accessible Surface Areas £1.30 Containment Vessel Pressure Retaining Boundary-General Visual I Moisture Barriers E-C E4.11 Containment Surface Areas - Visible Surfaces Visual, VT-I 1 7 Containment Sw"faces Requiring E4.12 Containment Surface Areas - Surface Area Grid Ultrasonic 1 Augmented Examination Minimum Wall Thickness Location Thickness Alion Science & Tecllllology 7-12 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TA&P Number Description) System F-A Fl.I0 Class 1 Piping Supports Visual, VT-3 CS: 14 4 Supports FW:40 HPCI: 9 MR: 14 MS: 38 RCIC: 1 RD: 10 RHR: 11 RWCU: 10 F1.20 Class 2 Piping Supports Visual, VT-3 CS: 68 4 HPCI: 52 MS: 63 RHR: 158 SOY: 48 Fl.30 Class 3 Piping Supports Visual, VT-3 ECW: 88 + 4 ESW: 117 10 HPSW: 98 MSSW: 170 FIAO Supports Other Than Piping Supports Visual, VT-3 CS:4 4 (Class 1,2, and 3) ECW: 3 + 10 ESW:2 HPSW: 4 MR: 16 RHR: 16 RPV:9 Alion Science & Technology 7-13 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Illterval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Risk Description Exam Total Number of Relief Request/ Notes (with Examination Category Categor:y Requirements Components by TA&P Number Description) Number System R-A 1 Risk Category I Elements See Notes FW: 51 14R-44 1 Risk-InfOlmed Piping 2 Examinations 3 2 Risk Category 2 Elements See Notes CS: 4 14R-44 I MR: 10 2 RHR: 28 3 RPV:2 3 Risk Category 3 Elements See Notes FW: I I4R-44 1 HPCl: 20 2 RWCU: 1 3 4 Risk Category 4 Elements See Notes CS: 53 14R-44 1 HPCI: 25 2 MR:62 MS: 275 RD:27 RHR: 195 RPV: 29 RWCU:9 5 Risk Category 5 Elements See Notes HPCI: 10 14R-44 1 RCIC: 2 2 3 Alion Sciellce & Techllology 7-14 PBT05.G03 Rel1isio11 ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Aug Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components TAP Number Description) AUG AUG-Ol lntergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category D: 15 2 Augmented Austenitic Stainless Steel Piping Components, TR-ll3932, Category E: 2 Components "BWR Vessel and Internals Project, Technical Basis for Category S: 106 Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75)", and TR-lOl2621, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-0 I Inspection Schedules (BWRVIP-75-A)" AUG-02 BWROG, BWR Feedwater Nozzle and Control Rod Drive Volumetric FW:6 ]4 Return Line Nozzle Cracking Components RPV: 18 AUG-05 Snubber Visual Examination and Functional Testing Functional Test COND: 1 Program & Visual, VT-3 CRD:8 CS: 6 ESW: 5 FW: 10 HPCl: 16 IN: 1 MR: 10 MS: 20 MSRV: 92 RCIC: 5 RHR: 26 RPVINST: 2 RWCU:4 AUG-12 10CFRSO Augmented Requirements for Reactor Pressure Volumetric RPV: 20 I4R-41 Vessel Shell Weld Examinations AUG-15 Standby Liquid Control Nozzle-to-Safe-End Surface RPV: ] AUG-D RPV Closure Flange O-Ring Sealing Surfaces Visual-RPVCF: 3 Measurement AUG-C2 HPCI, RHR, and Core Spray Suction Strainers Visual 13 Alion Science & 'Technology 7-15 PBT05.G03 Revi.'liol1 ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-1 UNIT 2 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Aug Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components TAP Number Description) AUG AUG-C3 Sludge Accumulation on the Torus Floor Visual - Sludge 1 Augmented Measurement Components AUG-CA Examination of Class MC Supports Visual, VT-3 50 (Continued) AUG-CB Examination of Drywell External Components Located Visual, VT-3 16 Outside Stabilizer Access Hatches AUG-CC Examination of Drywell Airgap Drain Lines Functional Test 4 & Visual AUG-CD Examination of Bolting in ECCS Suction Strainers Visual 9 Alioll Sciellce & Technology 7-16 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval ~x~ ,~,~:W~ TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-A B1.11 Circumferential Shell Welds (Reactor Vessel) Volumetric RPV: 5 I4R-41 Pressure Retaining B1.12 Longitudinal Shell Welds (Reactor Vessel) Volumetric RPV: 15 14R-41 Welds in Reactor Vessel BI.21 Circumferential Head Welds (Reactor Vessel) Volumetric RPV: 3 BI.22 Meridional Head Welds (Reactor Vessel) Volumetric RPV: 16 BI.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric RPV: I BI.40 Head-to-Flange Weld (Reactor Vessel) Volumetric & RPV: 1 Surface Bl.51 Beltline Region Repair Weld (Reactor Vessel) Volumetric RPV: 1 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 31 11 Full Penetration Welds of Nozzles in Vessels B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 31 Alion Science & Tecll1lology 7-/7 PBT05.Gf)3 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-G-l B6.1O Closure Head Nuts (Reactor Vessel) Visual, VT-I RPV: 2 (92 nuts) Pressure Retaining B6.20 Closure Studs (Reactor Vessel) Volumetric RPV: 2 (92 studs) 16 Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 2 (92 threads) 2 in. In Diameter B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-l RPV: 6 (276 washers, bushings) B6.180 Bolts and Studs (Pumps) Volumetric MR:2 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-l MR:2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-l MR:2 B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-I MR:4 Pressure Retaining MS: 12 Bolting, 2 in. and Less RPV: 3 In Diameter B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-I CS:2 FW:4 HPCI: 1 MR:4 MS: 21 RCIC: 2 RHR:2 RWCU: 1 B7.80 Bolts, Studs, & Nuts in CRD Housing (Reactor Vessel) Visual, VT-l RPV: I 8 (185 housings) Alion Science & Technology 7-18 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-K B10.10 Welded Attachments (Pressure Vessels) Surface or RPV:9 12 Welded Attachments for Volumetric Vessels, Piping, Pumps, B10.20 Welded Attachments (Piping) Surface CS:6 and Valves FW:24 HPCI: 1 MR:6 MS: 34 RWCU:2 B10.30 Welded Attachments (Pumps) Surface MR:7 B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 MR:2 Pump Casings 8-M-2 812.50 Valve Bodies (Exceeding NPS 4) (Valves) Visual, VT-3 CS: 6 Valve Bodies FW:8 HPCI: 4 MR:4 MS:21 RCIC:2 RHR: 9 RWCU: 5 Alion Science & Teclmology 7-/9 PBT05.G03 Revi...ion (J

lSI Program Plan Peacll Bottom Atomic Power Station Units 2 & 3, Fourtll Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System 8-N-l 813.10 Vessel Interior (Reactor Vessel) Visual, VT-3 RPV: 1 I4R-49 13 interior of Reactor Vessel 8-N-2 813.20 Interior Attachments Within 8eltline Region (Reactor Visual, VT-l RPV: 16 I4R-49 13 Welded Core Vessel) Support Structures and Interior 813.30 Interior Attachments 8eyond 8eltline Region (Reactor Visual, VT-3 RPV: 34 I4R-49 13 Attachments to Vessel) Reactor Vessels 813.40 Core Support Structure (Reactor Vessel) Visual, VT-3 RPV: 7 14R-49 B-0 B14.10 Welds in CRD Housing (Reactor Vessel) Volumetric or NA 9 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings Alio" Science & Tee/molog)' 7-20 PBT05.G03 Revi.~ion 0

lSI Program Plall Peach Bottom Atomic Power Station lInits 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TAP Number Description) System 8-P 815.10 System Leakage Test (IWB-5220) Visual, VT-2 RPV: 3 14R-47 All Pressure 14T-OI Retaining Components 14T-02 Alion Science & Technology 7-21 PBT05. G03 Revision 0

lSI Program Plall Peach Bottom Atomic Power Statioll Ullits 2 & 3, Fourth Illterval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Requestl Notes (with Examination Category Number Requirements Components by TAP Number Description) System C-A Cl.IO Shell Circumferential Welds (Pressure Vessels) Volumetric RHR: 8 Pressure Retaining Welds in Pressure Vessels C-B C2.31 Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface RHR:8 Pressure Retaining Nozzle with Reinforcing Plates in Vessels, Greater than 1/2" Welds in Vessels Nominal Thickness (Pressure Vessels) C2.33 Nozzle-to-Shell (or Head) Welds with Reinforcing Plates Visual, VT-2 RHR: 8 when Inside of Vessel is Inaccessible, Greater than 1/2" Nominal Thickness (Pressure Vessels) C-C C3.10 Welded Attachments (Pressure Vessels) Surface RHR: 12 12 Welded Attachments C3.20 Welded Attachments (Piping) Surface CS: 38 for Vessels, Piping, HPCI: 25 Pumps, and Valve MS: 11 RHR: 58 SDV:4 C3.30 Welded Attachments (Pumps) Surface CS: 4 RHR:4 Alioll Science & Technology 7-22 PBT05.G03 Revision (J

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Requestl Notes (with Examination Category Number Requirements Components by TAP Number Description) System C-H C7.l0 System Leakage Test (IWC-5220) Visual, VT-2 CAC:2 14R-25 All Pressure CAO:4 14T-O I Retaining Components CS: 2 14T-02 HPCI: I MS: I RCIC: I RHR: 5 RPV: 1 SLC: 2 SOV: 1 Alion Science & Technology 7-23 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief RequestJ Notes (with Examination Category Number Requirements Components by TAP Number Description) System O-A 01.20 Welded Attachments (Piping) Visual, VT-l ESW: 44 Welded Attachments for HPSW: 38 Vessels, Piping, Pumps, MSRV: 97 and Valves O-B 02.10 System Leakage Test (IWD-5221) Visual, VT-2 HPSW: I 14R-46 AII Pressure 14T-OI Retaining Components 14T-02 Alion Science & Technology 7-24 PBT05.G03 Revision 0

lSI Program Plan Peach Boltom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components TAP Number Description) E-A E1.11 Containment Vessel Pressure Retaining Boundary-General Visual 19 14R-48 Containment Surfaces Accessible Surface Areas E1.11 Containment Vessel Pressure Retaining Boundary-Visual, VT-3 38 14R-48 5 Bolted Connections, Surfaces El.I2 Containment Vessel Pressure Retaining Boundary-Visual, VT-3 I 14R-48 6 Wetted Surfaces of Submerged Areas EI.20 Containment Vessel Pressure Retaining Boundary-Visual, VT-3 I 6 BWR Vent System Accessible Surface Areas E1.30 Containment Vessel Pressure Retaining Boundary-General Visual 1 Moisture Barriers E-C E4.ll Containment Surface Areas - Visible Surfaces Visual, VT-l 1 7 Containment Surfaces Requiring E4.12 Containment Surface Areas - Surface Area Grid Ultrasonic I Augmented Examination Minimum Wall Thickness ~ocation Thickness Alion Science & Technology 7-25 PBT05.G03 Revision ()

lSI Program Plall Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Item Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components by TA&P Number Description) System F-A F1.IO Class I Piping Supports Visual, VT-3 CS: 14 4 Supports FW:40 HPCl: 8 MR: 14 MS: 38 RCIC: I RD:9 RHR: 11 RWCU: 10 Fl.20 Class 2 Piping Supports Visual, VT-3 CS: 64 4 HPCI: 64 MS: 63 RHR: 151 SDV: 48 Fl.30 Class 3 Piping Supports Visual, VT-3 ESW: liS 4 HPSW: 97 MSRV: 167 FlAG Supports Other Than Piping SUPP0l1S Visual, VT-3 CS: 4 4 (Class 1,2, and 3) HPCI: 1 HPSW: 4 MR: 16 RHR: 16 RPV:9 Alion Science & Tee/urology 7-26 PBT05.G03 Revision 0

lSI Program Plan Pellch Bottom Atomic Power Station Units 2 & 3, Fourth IntenJll1 TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Risk Description Exam Total Number of Relief RequestJ Notes (with Examination Category Category Requirements Components by TA&P Number Description) Number System R-A 1 Risk Category 1 Elements See Notes FW: 54 14R-44 1 Risk-Infonned Piping 2 Examinations 3 2 Risk Category 2 Elements See Notes CS: 2 14R-44 1 MR: 10 2 RHR: 23 3 RPV:2 3 Risk Category 3 Elements See Notes FW: 1 I4R-44 1 HPCI:21 2 RWCU: 1 3 4 Risk Category 4 Elements See Notes CS: 56 14R-44 1 HPCI: 24 2 MR: 74 MS: 300 3 RD:25 RHR: 191 RPV: 30 RWCU: 8 5 Risk Category 5 Elements See Notes HPCI: 8 14R-44 1 RCIC: 2 2 3 Alion Science & Technology 7-27 PBT05.G03 Rel'i.fiion ()

lSIProgram Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Aug Description Exam Total Number of Relief Request! Notes (with Examination Category Number Requirements Components TAP Number Description) AUG AUG-Ol Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category C: 5 2 Augmented Austenitic Stainless Steel Piping Components, TR-113932, Category E: I Components "BWR Vessel and Internals Project, Technical Basis for Category S: 97 Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75)", and TR-IOI2621, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A)" AUG-02 BWROG, BWR Feedwater Nozzle and Control Rod Drive Volumetric FW:6 14 Return Line Nozzle Cracking Components RPV: 18 AUG-05 Snubber Visual Examination and Functional Testing Functional Test CRD:8 Program & Visual, VT-3 CS: 10 ESW:5 FW: 10 HPCI: 14 IN: 2 MR: 10 MS:20 MSRV: 89 RCIC: 1 RHR: 23 RWCU:4 AUG-12 IOCFR50 Augmented Requirements for Reactor Pressure Volumetric RPV: 20 14R-41 Vessel Shell Weld Examinations AUG-I5 Standby Liquid Control Nozzle-to-Safe-End Surface RPV: I AUG-D RPV Closure Flange O-Ring Sealing Surfaces Visual-RPVCF: 3 Measurement AUG-C2 HPCI, RHR, and Core Spray Suction Strainers Visual 11 AUG-C3 Sludge Accumulation on the Torus Floor Visual - Sludge I Measurement Alion Science & Technology 7-28 PIlT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Inten'al TABLE 7.1-2 UNIT 3 INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Aug Description Exam Total Number of Relief Requestl Notes (with Examination Category Number Requirements Components TAP Number Description) AUG AUG-CA Examination of Class MC Supports Visual, VT-3 50 Augmented Components AUG-CB Examination of Dryweil External Components Located Visual, VT-3 16 (Continued) Outside Stabilizer Access Hatches AUG-CC Examination of Drywell Airgap Drain Lines Functional Test 4 & Visual AUG-CD Examination of Bolting in ECCS Suction Strainers Visual 9 Alion Science & Technology 7-29 PBT05. G03 Revision (J

lSI Program PlalZ Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 1 For the Fourth lSI Interval, PBAPS's lSI Class 1 and 2 piping inspection program will be governed by risk-informed regulations. The RISI Program methodology is described in the EPRI Topical Reports TR-112657, Rev. B-A and Code Case N-578-1. The RISI Program scope has been implemented as an alternative to the 2001 Edition through the 2003 Addenda ofthe ASME Section XI examination program for lSI Class 1 B-F and B-1 welds and lSI Class 2 C-F-l and C-F-2 welds in accordance with IOCFR50.55a(a)(3)(i). 2 Per the EPRI Topical Reports TR-112657, Rev. B-A and Code Case N-578-1, welds within the plant that are assigned to IGSCC Categories B through G will continue to meet existing IGSCC schedules, whilelGSCC Category A welds have been subsumed into the RlSI Program. 3 Examination requirements within the RlSI Program are determined by the various degradation mechanisms present at each individual piping structural element. See EPRl Topical Reports TR-112657, Rev. B-A and Code Case N-578-1 for specific examination method requirements. 4 lSI snubber visual examinations and functional testing are performed in accordance with the ASME OM Code, Subsection ISTD Program. The number of PBAPS Unit 2, 3, and Common supports identified, include snubbers for the visual examination and functional testing ofthe integral and nonintegra attachments per Paragraphs IWF-5200(c), IWF-5300(c), and IWF-2500(a). The snubbers are scheduled and administratively tracked in the lSI Program; however, the ASME OM Code, Subsection ISTD Program will be the mechanism for actually perfonning the visualexaminations and functional testing scheduled within the lSI Program. For a detailed discussion ofthe snubber program, see Section 4.2. 5 Bolted connections examined per Item Number El.Il require a General Visual examination each period and a VT-3 visual examination once per interval and each time the connection is disassembled during a scheduled Item Number El.ll examination. Additionally, a VT-I visual examination shall be perfonned if degradation or flaws are identified during the VT-3 visual examination. These modifications are required by IOCFR50.55a(b)(2)(ix)(G) and 10CFR50.55a(b)(2)(ix)(H). 6 Item Numbers E1.12 and E1.20 require VT-3 visual examination in lieu of General Visual examination, as modified by 10CFR50.55a(b)(2)(ix)(G). 7 Item Number E4.ll requires VT-I visual examination in lieu of Detailed Visual examination, as modified by IOCFRSO.55a(b)(2)(ix)(G). 8 Per IOCFR50.55a(b)(2)(xxi)(B), Table lWB-2500-1 examination requirements, the provisions of Table IWB-2500-1, Examination Category B-G-2, Item Number B7.80, that are in the 1995 Edition are applicable only to reused bolting when using the 1997 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) ofthis section. 9 Per Stone and Webster Calculation No. PM-0945, Rev. I, dated 11/30/95, welds in peripheral CRD housings are exempt from the surface and volumetric examination requirements of Table IWB-2500-1, Examination Category B-O, based on the IWB-1220(a) normal makeup exemption. 10 The Unit 2 population counts include those components that are common to both units (typically designated as "Common" or "Unit2/3"). These Common components are referenced in Table 7.1-1 following a "+" symbol to designate Unit Common. II As allowed by Code Case N-613-1, PBAPS will perform a volumetric examination using a reduced examination volume (AB-C-D-E-F-G-H) of Figures 1,2, and 3?fthe Code Case in lieu of the previous examination volumes of ASME Section XI, Figures IWB2500-7(a), (b), and (c). 12 PBAPS will utilize the alternative requirements of Code Case N-700 for Class 1,2, and 3 welded attachments, which clarifies that for single vessels, only one welded attachment is required to be selected for examination. The welded attachment selected shall be urder continuous load during normal operation, or if one does not exist, shall be one subject to potential intermittent loads (e.g., seismic, water hammer, etc.) Ali01l Science & Teclt1lo!ogy 7-30 PBT05.G03 Revision ()

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 13 The RPV interior requires examination per the BWRVIP in lieu of ASME Section XI Examination Categories B-N-I and B-N-2 per Relief Request 14R-49. Augmented Inspection Programs associated with the BWRVIP and thePBAPS Vessel Internals Program are maintained independently in procedure ER-PB-331-I00 I, "RPV & Internals Program Basis and Implementation Document" 14 Augmented examination ofthe Feedwater Spargers has been transferred to procedure ER:PB-33 1-1001, "RPV & Internals Program Basis and Implementation Document" for the BWRVIP, whereas augmented examinationofthe Feedwater Nozzles remains with the lSI Program. 15 The system pressure tests required for Class 3 systems ECW and ESW utilize PBAPS Unit 2 system pressure test procedures that are common to both PBAPS Units 2 and 3. 16 Examination Category B-G-I, 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 XI, 2001 Edition through the 2003 Addenda Two components (Studs 1-46) and (Studs 47-92) represent the ninety-two RPV closure studs per unit. For tracking purposes, one ofthe two B-G-l, B6.20 components (Studs 1-46) will also represent the six cattle chute studs (21, 22, 23, 24, 25, and 26) for PBAPS Unit 2 and one ofthe two B-G-l, B6.20 components (Studs 47-92) will also represent the six cattle chute studs (67,68,69,70,71, and 72) for PBAPS Unit 3 which are removed each refueling outage Alioll Science & Technology 7-31 PBT05.G03 Revision 0

lSI Progrtll11 Pla11 Peach !Jottom Atomic Power Statioll Units 2 & 3, Fourth Illterval 7.2 Snubber Inspection Sun1n1ary Tables 1OCfR50.55a "Codes and Standards" allows usage of ASME OM Code Subsection ISTD in place of ASME Section XI Paragraphs IWF-5200(a) and IWF-5300(a) and (b), using VT-3 visual exan1ination n1ethods described in Paragraph I\\VA-2213. The following Tables 7.2-1 and 7.2-2 provide a sun1mary of the ASME OM Code, Subsection ISTD, Snubber exan1inations and testing for the Fourth lSI Interval at PBAPS Units 2, 3, and Comn10n. The forn1at of the Snubber Inspection Sun1mary Tables is as depicted below and provides the following inforn1ation: ASME OM Code OM Article Article Nurn bel' Exam Relief Request! Subsection Totals Frequency Notes (with Subsection Number Description Requirements TAP Number Description) (1) (2) (3) (4) (5) (6) (7) (8) (1) ASME OM Code Subsection: Provides the applicable Code for Operation and Maintenance of Nuclear Power Plants (OM Code) subsection nun1ber and a description as obtained from ISTD. Only applicable subsections to PBAPS are identified. (2) OM Article Number: Provides the article number as identified in ISTD. Only those article nun1bers applicable to PBAPS are identified. (3) Article NUlnber

Description:

Provides the article description as identified in IS'rD. Identifies the methods selected to be performed at PBAPS. (4) EXaInination Requirelnents: Provides the visual examination and functional testing Inethods required by ISTD. .'Ilion Science & Technology 7-32 PIJ T05. G03 Revision 0

lSI Program PIau Peach Bottom Atomic Power Station Vnits 2 & 3, Fourth Interval (5) Totals: Provides the totalnunlber of snubbers that pertain to that article of ISTO. Note that the total number of snubbers are subject to change after completion of plant modifications and design changes. (6) Frequency: Provides the frequency for examinations and testing as addressed in ISTO and approved ISTO Code Cases. (7) Relief Request/TAP Nunlber: Provides a listing of Relief Request/TAP Numbers to specific snubber components. Relief requests and TAP Nunlhers that generically apply to all components, or an entire class are not listed. If a Relief Request/TAP NU111ber is identified, see the corresponding relief request in Section 8.0 or the TAP NU111her in Section 2.5. (8) Notes: Provides a listing of progran1 notes applicable to the ISTO article number. If a program note nlU11ber is identified, see the corresponding program note in Table 7.2-3. Alioll Science & Tec!l/lology 7-33 PIJT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourth Interval TABLE 7.2-1 UNIT 2 & COMMON SNUBBER INSPECTION

SUMMARY

ASME OM Code Subsection OM Article Article Number Exam Totals Frequency Relief Request! Notes (with Subsection Number Description Requirements TAP Number Description) ISTD ISTD-4200 Accessible and Inaccessible Snubbers (I population) Visual, VT-3 206 Once every 10 I Snubber Years Examinations ISTD ISTD-5200 10% Functional Test Plan-Functional Testing 2 Every Outage 2 Snubber Type I Snubbers (PSA-l/4, PSA-l/2) Testing 10% Functional Test Plan - Functional Testing 65 Every Outage 2 Type 2 Snubbers (PSA-l, PSA-3, PSA-I0) 10% Functional Test Plan-Functional Testing 8 Every Outage 2 Type 3 Snubbers (PSA-35) 10% Functional Test Plan - Functional Testing 131 Every Outage 2 Type 4 Snubbers (Anvil International Hydraulic) Alion Science & Tecllllology 7-34 PBT05.G03 Revision 0

lSI Program Plan Peach Bottom Atomic Power Statio" U"its 2 & 3, Fourth I"terval TABLE 7.2-2 UNIT 3 SNUBBER INSPECTION

SUMMARY

ASME OM Code Subsection OM Article Article Number Exam Totals Frequency Relief Req uest/ Notes (with Subsection Number Description Requirements TAP Number Description) ISTD ISTD-4200 Accessible and Inaccessible Snubbers (I population) Visual, VT-3 196 Once every 10 I Snubber Years Examinations ISTD ISTD-5200 10% Functional Test Plan - Functional Testing 0 Every Outage 2 Snubber Type I Snubbers (PSA-l/4, PSA-l/2) Testing 10% Functional Test Plan - Functional Testing 74 Every Outage 2 Type 2 Snubbers (PSA-I, PSA-3, PSA-IO) 10% Functional Test Plan - Functional Testing 6 Every Outage 2 Type 3 Snubbers (PSA-35) 10% Functional Test Plan - Functional Testing 116 Every Outage 2 Type 4 Snubbers (Anvil International Hydraulic) Alioll Sciellce & Technology 7-35 PBT05.G03 Revisioll 0

lSI Program Plan Peach Bottom Atomic Power Station Units 2 & 3, Fourtlt Interval ___________________________~~~~5~'{~~1 ,;:;~ ~:~>~yK~i§~_:>&'_' TABLE 7.2-3 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary I Examinations performed per Code Case OMN-13, "Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants". 2 Per ISTO 2001 Edition through the 2003 Addenda, Article ISTD-5240 "Test Frequency". Alioll Science & Technology 7-36 PBT05.G03 Revision ()

IS1 PI'Ogrt1l11 Plall Peach Bottom Atomic Power Station Vllits 2 alld 3, FOllrth Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests \\vritten per 10CFR50.55a(a)(3)(i) for situations where alternatives to ASME Section XI requirements provide an acceptable level of quality and safety; per 10CFR50.55a(a)(3)(ii) for situations where cotnpliance with ASME Section XI requirements results in a hardship or an unusual difficulty without a cotnpensating increase in the level of quality and safety; and per IOCFR50.55a(g)(5)(iii) for situations where ASME Section XI requirenlents are considered impractical. The following USNRC guidance was utilized to detennine the correct 10CFR50.55a paragraph citing for PBAPS relief requests. 10CFR50.55a(a)(3)(i) and 10CFR50.55a(a)(3)(ii) provide alternatives to the requiretnents of ASME Section XI, while 10CFR50.55a(g)(5)(iii) recognizes situational impracticalities. 10CFll50.55a(a)(3)(i): 10CFR50.55a(a)(3)(ii): 10CFR50.55a(g)(5)(iii): Cited in relief requests when alternatives to the ASIvlE Section XI requirements which provide an acceptable level of quality and safety are proposed. Exanlples are relief requests which propose alternative NDE methods and/or exanlination frequency. Cited in relief requests when compliance with the ASME Section XI requirements is deen1ed to be a hardship or unusual difficulty without a con1pensating increase in the level of quality and safety. Exatnples of hardship and/or unusual difficulty include, but are not lin1ited to, excessive radiation exposure, disassembly of con1ponents solely to provide access for exatninations, and development of sophisticated tooling that would result in only n1inimal increases in examination coverage. Cited in relief requests when conforll1ance \\vith ASME Section XI requirctnents is deelned inlpractical. Exmnples of itnpractical requirements are situations where the component would have to be redesigned, or replaced to enable the required inspection to be performed. An index for PBAPS relief requests is included in Table 8.0-1. The "I4R-XX" relief requests are applicable to lSI, CISI, SPT, and POI. The following relief requests are subject to change throughout the inspection interval. Alion Science & Technology 8-1 PBT05.G03 Revisioll ()

lSI Program PI(ln Peach Bottom Atomic Power Station Units 2 (I11t1 3, FOllrth Interval TABLE 8.0-1 RELIEF REQUEST INDEX Relief Revision I Status2 (Program) Description/ Request Date3 I Approval SUlnnulry I (lSI) Lin1ited Volun1etric EXaIl1ination of RHR 0 Heat Exchanger Pressure Retaining Shell I4R-08 Withdrawn 11/0S/08 CircUlnferential Welds (Shell-to-Flange Welds). Revision 0 Withdrawn by PBAPS. 0 (SPT) Pressure Testing the RPV Head Flange I4R-2S Granted Seal Weld Leak Detection System. Revision 0 11/0S/08 Granted per SER dated 02/26/09. (lSI) Alternative Volu111etric Exan1ination of RPV Circlm1ferential Shell Welds. Pennanent 0 relief was authorized per SER dated 06/1500 I4R-41 Authorized (CCN 00-00069) and thus applies to the 11/0S/08 remaining tenn of operation under the existing, initial operating license, including this Fourth Inspection Interval. (lSI) Alternate Risk-Inforn1ed Selection and 0 Examination Criteria for Examination Category 14R-44 Authorized B-F, B-1, C-F-I, and C-F-2 Pressure Retaining 11/0S108 Piping Welds. Revision 0 Authorized per SER dated 02/26/09. (SPT) Alternative Exan1ination Requiren1ents of 0 ASME Section XI, Paragraph IWA-S244, "Buried 14R-46 Not Required 11/0S/08 Components". Revision 0 Not Required per SER dated 02/26/09. 1 (SilT) Testing of Control Rod Drive Boundaries. 14R-47 I Authorized 11/0S108 I Revision 0 Authorized per SEH. dated 02/26/09. (CIS)) Alternative Examination Requiren1ents of 14R-48 0 ASME Section XI, Paragraph IWE-1232, Granted (CRR-13) 11/0S/08 "Inaccessible Surface Areas". Revision 0 Granted per SER dated 02/26/09. Alion Science & Tecllllology 8-2 PBT05.G03 Revisiol1 ()

lSI Program Plan Peach Boltom Atomic Power Station Vnits 2 all(/ 3, Fourth Illterval TABLE 8.0-1 RELIEF REQUEST INDEX Relief Revision Status2 (Program) Description/ Request Date3 Approval SUlnmaryl (lSI) Use of BWRVIP Guidelines in Lieu of Specific ASME Code Requiren1ents. Revision 0 Authorized Conditionally per SER dated 1 Authorized 04/30/08. Note that previous Third Interval 14R-49 Conditionally I3H.-45 relief subsequently was approved to 11/05/08 modify the Unit 3 Interval date to be consistent with the Unit 2 date. The duration of this I4R-19 relief is thus modified as such to expit*e on 11/04/18 for both units. Note 1: The USNRC grants relief requests pursuant to 10CFR50.55a(g)(6)(i) when Code requirements cannot be met and proposed alternatives do not meet the criteria of 10CFR50.55(a)(3). The USNRC authorizes relief requests pursuant to 10CFR50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level of quality and safety or under 10CFR50.55a(a)(3)(ii) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety. Note 2: This column represents the status of the latest revision. Relief Request Status Options: Authorized-Approved for use in an USNRC SER (See Note 1); Granted - Approved for use in an USNRC SER (See Note I); Authorized Conditionally - Approved for use in an USNRC SER which imposes certain conditions; Granted Conditionally - Approved for use in an USNRC SER which imposes certain conditions; Denied - Use denied in an USNRC SER; Expired - Approval for relief has expired; Withdrawn - Reliefhas been withdrawn by PBAPS; Not Required - The USNRC has deemed the relief unnecessary in an SER or RAI; Cancelled - Reliefhas been cancelled by PBAPS prior to issue; and Submitted - Reliefhas been submitted to the USNRC by the station and is awaiting approval. Note 3: The revision listed is the latest revision of the subject relief request. The date this revision became effective is the date of the approving SER, which is listed in the fourth column of the table. The date noted in the second column is the date of the lSI Program Plan revision when the relief request was incorporated into the document. Alioll Science & Tecllllology 8-3 PIJ TOJ. G03 Revision 0

lSI Program Plan Peac" Bottom Atomic Power Sta/ion Units 2 and 3, FOllrthlnlerval 10CFR50.55a RELIEF REQUEST: 14R-25 Revision 0 (Page 1 of 5) Relief Request for Pressure Testing the IUlV Head Flange Seal Weld Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii) 1.0 ASME CODE COMPONENTS AFFECTED: Drawing Number: C0111pOnent Nunlber: Code Class:

Reference:

Exanlination Category: Item NU111ber:

== Description:== 2 Table IWC-2500-1 IWC-5200 C-H C7.10 Pressure Testing the RPV Head Flange Seal Weld Leak Detection System Class 2 RPV Head Flange Seal Weld Leak Detection System Unit 2: Piping 4DCN-l" Unit 3: Piping 4DCN-l" ISI-351, Sht. 1 (Unit 2) ISI-351, Sh1. 3 (Unit 3) 2.0 APPLICABLE CODE EDITION AND ADDENDA: The Il1service Inspection progranl is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2001 Edition through the 2003 Addenda. 3.0 APPLICABLE CODE REQUIREMENT: Table IWC-2500-1, Exanlination Category C-H, Ite111 Nmuber C7.1 0, requires all Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWC-5220 (to operate or support the safety function up to the first normally closed valve, or valve capable of autoluatic closure). This pressure test is to be conducted once each inspection period. 4.0 IlVIPRACTICALITY OF COMPLIANCE: Pursuant to 10CFR50.55a(g)(5)(iii), reliefis requested on the basis that c0111pliance with AS:NIE Section XI requireluents is inlpractical. Alion Science & Technology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 and 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-25 Revision 0 (Page 2 of 5) The Reactor Vessel Head Flange Leak Detection Piping is separated fron1 the reactor pressure boundary by one passive seal, an O-ring located on the vessel flange. A second O-ring is located on the outer side of the vessel flange~ on the opposite side of the tap in the vessel flange (See Figures I4R-25-1 and 14R-25-2). This piping is required during plant operation to indicate failure of the inner flange O-ring seal. Failure of the O-ring results in a High Pressure Alann in the Main Control ROOln. Failure of the inner O-ring is the only condition under which this piping is pressurized. The configuration of this systeln precludes 111anual pressure testing while the vessel head is ren10ved because the orifice in the vessel flange and the associated 1 inch pi ping and cOlnponents do not incorporate a means for pressure testing. As sho\\vl1 in Figure 14R-25-2, the configuration of the vessel tap, con1bined with the slnall size of the tap and the high-test pressure requirement (approxin1ately 1045 psig), precludes the tap fr01n being temporarily plugged or connected to other piping. The opening of the flange is only 3/16 of an inch in dian1eter and is sn100th walled, n1aking the effectiveness of a temporary seal at the required test pressure very lilnited. Failure of a ten1porary pressure seal could possibly cause ejection of the device. The configuration also precludes pressure testing with the vessel head installed, because the seal prevents con1plete water filling of the piping due to the absence of a vent path. Additionally, a pneumatic test perforn1ed with the head installed is not recon11nended due to the configuration of the top head. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are in tum held in place within these grooves by a series of retainer clips that are housed in recessed cavities in the flange face. If a pressure test were performed with the head on, the inner O-ring would be pressurized in a direction opposite to \\vhat it would see in normal operation. This test pressure would result in a net inward force on the inner O-ring that would tend to push it into the recessed cavities that house the retainer clips. This inward force would very likely damage the thin O-ring 111aterial. In addition to the problems associated with the O-ring design that preclude this testing, it is also questionable whether a pneumatic test is appropriate for this piping. The use of a pneumatic test performed at RPV non1inal operating pressure would represent an unnecessary safety risk to personnel in the unlikely event of a test failure, due to the large amount of stored energy contained in pressurized air. 5.0 BURDEN CAUSED BY COMPLIANCE: Pressure testing of this Class 2 piping during the Class 1 System Leakage Test is prevented because the piping is not connected to the source of pressure (the leak-off piping is only pressurized in the event of a failure of the inner O-ring). As discussed in section 4.0 above~ the extensive modifications required to accomlllodate testing~ without affecting the integrity of the O-ring seal, is impractical. Alion Science & Teclll1%gy PBT05.G03

lSI Program Plall Peach Bottom Atomic Power Statioll Units 2 ([lid 3, FOllrth Interval 10CFR50.55a IffiLIEF REQUEST: I4R-25 Revision 0 (Page 3 of 5) Based on the above~ Peach BottOln Atonlic Power Station requests the following alternative examination be perfonned on the Reactor Vessel Head Flange Seal Leak Detection Systen1. 6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: A VT-2 visual exan1ination on the Class 2 portion of the RPV Flange Leak Detection Piping will be performed during each refueling outage when the RPV head is ofland the head cavity is nooded above the vessel flange. The static head developed with the leak detection piping filled with water will allow for the detection of any leaks in the piping. This exan1ination will be performed per the frequency specified by Table IWC-2500-1. 7.0 DURATION OF PROPOSED ALTERNATIVE: Relief is requested for the fourth ten-year inspection interval for Peach Botton1 Aton1ic Power Station, Units 2 and 3. The Fourth lSI Interval begins on Novenlber 5, 2008 for both Units as proposed in Relief Request 13R-45 (Third lSI Interval Relief Request), and will conclude Novelnber 4, 2018 for Peach Bottonl AtOlnic Power Station, Units 2 and 3.

8.0 PRECEDENTS

Similar relief requests have been approved for: 1. Peach Bottom AtOlnic Power Station Third Inspection Interval Relief Request RR-25 was authorized per SER on July 31, 2000. 2. LaSalle County Station Second Inspection Interval Relief Request PR-04 \\vas authorized per SER dated July 3, 1996. 3. Susquehanna Steam Electric Station Third Inspection Interval Relief Request 3RR-07 was authorized per SER dated September 24, 2004. Alion Sciellce & Techllology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 alld 3, Fourth Interval 10CFR50.55n RELIEF REQUEST: 14R-25 Revision 0 (Page 4 of 5) FIGURE I4R-25-1 REACTOR PRESSURE VESSEL HEAD FLANGE LEAI(-OFF PIPING DETAILS (Ditl1ensions are typical for 13WR) Alion Science & Technology PB TOS. G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 and 3, FOllrth Interval 10CFR50.55a RELIEF REQUEST: 14R-25 Revision 0 (Page 5 of 5) FIGURE I4R-25-2 REACTOR PRESSURE VESSEL HEAD FLANGE LEAI<:-OFF PIPING DETAILS Outer Flonge Seal Ring \\ Vessel Flange Sectional View Alion Science & Technology High Pressure leak Detection Monitoring Tap See Detail "A" Detail "A" 3/16' PBT05.G03

lSI Program Plall Pellclt Bottom Atomic Power Station Unit!!J' 2 lIlld 3, Fourtlt Interval 10CFR50.55a RELIEF REQUEST: I4R-41 Revision 0 (Page 1 of 6)

• • • NOTE ***

Peach Bottoln AtOlllic Power Station Fourth Inspection Interval Relief Request 14R-41, Revision 0 is simply an administrative placeholder. This relief request was previously sublnitted on February 7, 2000 and approved under the Third Inspection Interval lSI Progranl Plan as Relief Request RR-41, Revision O. The approval authorized under USNRC SER dated June 15, 2000 (CNN 00-00069) for Peach Bottom Aton1ic Power Station, PECO, was for pennanent relief and thus applies to the ren1aining tenn of operation under the existing, initial operating license, including this Fourth Inspection Interval. Fonnatting for Relief Request RR-41, Revision 0 varied froln the standard lSI Progrmn Plan forn1at due to the fact that it also requested relief from the Augmented Vessel exan1ination contained in 10CFR50a(g)(6)(ii)(A)(2). The relief request is carried here and renun1bered as I4R-41, Revision 0 purely for adnlinistrative purposes. All ASME Code references were n1ade in accordance with the 1989 Edition, No Addenda of ASME Section XI. No changes to the actual approved relief request have been Blade and no further or revised authorization is required. REQUEST NUMBER: RR-41 REVISION 0 Request for Permanent Relief from Circumferential Shell Weld Inspection Requirements Peach Bottom Atomic Power Station, Units 2 and 3

References:

1. 2. }lroposed Relief Letter froln R. A. Capra (U.S. Nuclear Regulatory COlnlnission (USUSNRC)) to G. D. Edwards (PECO Energy Conlpany), dated December 2.1998 Letter frOln J. F. Stolz (USUSNRC) to G. A. Hunger, J1'. (PECO Energy COl11pany), dated October 7, 1997 As discussed in USNRC Generic Letter 98-05 ("Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Auglnented Exanlination Requirelnents on Reactor Pressure Vessel Circumferential Shell Welds"), Boiling Water Reactor (BWR) licensees may request pernlanent relief from the inservice inspection requirements of 10CFR50.55a(g) for the volulnetric exanlination of circumferential reactor pressure vessel welds. Accordingly, PECO Energy Company (PECO Energy) is requesting relief fr01n the following requireillents: 1) eXaInination of the RPV circunlferential shell "velds (ASME Section XI Examination Category B-A. Itenl NUlnber B1.11) as required by 10CFR50.55a(g)(6)(ii)(A)(2), 2) inservice inspection AliO/l Scie/lce & Tecllllology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Statio/1 Units 2 (Iud 3, FOllrl/l/nterval 10CFR50.55a RELIEF REQUEST: I4R-41 Revision 0 (Page 2 of 6) requirelnents for circtllnferential welds contained in the Anlerican Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 1980 Edition through Winter 1981 Addenda (a two year delay was granted in the Reference 1 and 2 letters fr01n performing this lSI inspection as required by the ASME Code), 3) inservice inspection requirenlents for circUlnferential welds contained in the current third ten-year interval ASME Boiler and Pressure Vessel Code, Section XI, 1989 Edition, and 4) inservice inspection requirenlents for circumferential welds contained in all future versions of the ASME Code through the end of the current operating licenses. Basis for Proposed Relief The basis for this request is documented in the report "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Rec01nlnendations (BWRVIP-05)", that was translnitted to the USNRC in Septenlber 1995. As discussed in Generic Letter 98-05, the staff has conlpleted its final review of the information sublnitted by the BWRVIP and the staff s safety evaluation was translnitted to Carl Terry, Chairnlan of the BWRVIP, in a letter dated July, 1998. The staff concluded that the BWRVIP-05 proposal, as nl0dified, to elinlinate BWR vessel circmnferential weld exanlinations, is acceptable. As discussed in this Generic Letter, licensees may request permanent (i.e., for the renlaining tenn of operation under the existing, initial license) relief by denl0nstrating that: 1) at the expiration of their license, the circwnferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July, 1998 safety evaluation, and 2) licensees have inlplenlented operator training and established procedures that linlit the frequency of cold over-pressure events to the amount specified in the staffs July, 1998 safety evaluation. BWRVIP-05 provides the technical basis for elilninating inspection of Boiling Water Reactor (BWR) RPV circumferential shell welds. The BWRVIP-05 report concludes that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. The USNRC staff has conducted an independent risk-infonned asseSSlllent of the analysis contained in BWRVIP-05. This assessnlent also concluded that the probability of failure of the BWR RPV circmnferential welds is orders of nlagnitude lower than that of the axial shell welds. The independent USNRC assessnlent utilized the FAVOR code to perform a probabilistic fracture nlechanics (PFM) analysis to estilnate RPV failure probabilities. Three key assumptions in the PFM analysis are: the neutron fluence was that estitnated to be end-of-license ll1ean fluence, the chemistry values are mean values based on vessel types, and the potential for beyond design basis events is considered. Although BWRVIP-05 provides the technical basis supporting an alternative, the following infonnation is provided to show the conservatislns of the USNRC analysis relative to the projected, end-of-license conditions for the PBAPS, Units 2 and 3 vessels. Alion Science & Technology PBT05.G03

lSI Progrtll11 Pia" Peach Bottol11 Atomic Power Station Units 2 awl 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-41 Revision 0 (Page 3 of 6) Basis for PBAPS, Unit 2 - Technical Evaluation During refueling outage 2R 12 at PBAPS, Unit 2, which occurred in the t~lll of 1998, the PBAPS, Unit 2 reactor vessel was exan1ined utilizing the General Electric GERIS-2000 systen1. During this examination, the C6 circun1ferential weld was exan1ined. A cun1ldative code volume of approximately 75.80/0 was exan1ined with no indications. Additionally, while perforn1ing the vertical weld exmninations, an average incidental cun1ulative code volunle of approximately 7.90/0 was exalnined for the four (4) circumferential welds. No reportable indications were found. The exmnination perforn1ed during 2R12 was an alternative approved by the USNRC for PBAPS, Unit 2 in a safety evaluation report dated December 2, 1998 (Reference 1). The following table illustrates that the PBAPS, Unit 2 plant has additional margin in c0111parison to the BWRVIP-05 Fracture Analysis liIniting case (that is, B&W SN 2 in Table 7-7). The chenlistry factor, i1RTNDT, lnargin term, n1ean ART, and upper bound ART are calculated consistent with the guidelines of Reg. Guide 1.99, Rev. 2. Table 1 - Conlparison of Peach Bottom 2 Fracture Analysis Pnrarueters to the BWI~Vn)-05 Litniting Parameters PBAPS, Unit 2 USNRC Independent Parameter Companltive Assessment Limiting Desc ri ption Parameters at 32 EFPY Fracture Analysis I Parameters Fluence. n/cm2 8.8 x 10 17 1.25 X 10 18 Initial RTNOT, OF -32 -5 Chemistry Factor 76.4 190 Cu% 0.056 0.287 Ni% 0.96 0.60 ~RT\\lOT 24.8 87.9 I Margin Term 24.8 62.2 Mean ART -7.2 82.9 Upper Bound ART 17.6 145.1 As shown above, every parameter used in the lin1iting USNRC independent asseSSlllent report (excluding Ni 0/0) bounds the circun1ferential shell weld infonnation for PBAPS, Unit 2 at 32 EFPY. 32 EFPY represents the end of the requested deferral period. The combination of the Ni Alioll Science & Tecllllology PBTOS.G03

lSI Program Plall Peach Bottom Atomic Power Station Units 2 {lJI(I 3, FOllrth Illterval 10CFR50.55a RELIEF REQUEST: 14R-41 Revision 0 (Page 4 of 6) 0/0 and Cu % determines the chemistry factor which is itself bounded by the USNRC independent assessn1ent. Basis for PBAPS, Unit 3 - Tcchnical Evaluation During refueling outage 3R 11 at PBAPS, Unit 3, which occurred in the fall of 1997, the PBAPS, Unit 3 reactor vessel,vas exan1ined utilizing the General Electric GERIS-2000 systen1. During this examination, the C6 circun1ferential weld was exmnined. A cumulative code voltllne of approximately 69.3~1o was exan1ined with no indications. Additionally, 'while perforn1ing the vertical weld exmninations, incidental coverage of approximately 2-3% was obtained for the four (4) circun1ferential welds. No reportable indications were found. The examination perfonned during 3Rll was an alternative approved by the USNRC for PBAPS, Unit 2 in a safety evaluation report dated October 7, 1997 (Reference 2). The following table illustrates that the PBAPS, Unit 3 plant has additional n1m"gin in con1parison to the BWRVIP-05 Fracture Analysis lilniting case (that is, B&W SN 2 in Table 7-7). The chen1istry factor, ~RTNOT, Inargin term, Inean ART, and upper hound ART are calculated consistent with the guidelines of Reg. Guide 1.99, Rev. 2. Table 2 -Conlparison of PBAPS, Unit 3 Fracturc Analysis Paranlcters to thc BWRVIP-05 Limiting panuucters Parameter PBAPS, Unit 3 BWRVIP-05 Limiting Description Comparative Fracture Analysis Parameters at 32 EFPY Parameters FllIcnce, n/cm2 7.9 x 10 17 1.25 X 10 14 Initial RT,'.:DT' OF -50 -5 Chem istry Factor 136.9 190 ClI% I 0.102 0.287 Ni% 0.942 0.60 ,.1RTN1H 42.2 87.9 Margin Term 42.2 62.2 Mean ART -7.8 82.9 Upper Bound ART 34.4 145.1 As shown above, every parameter used in the lilniting USNRC independent assessn1ent repoli (excluding Ni 0/0) bounds the circun1ferential shell weld infol'lnation for PBAPS, Unit 3 at 32 EFPY. 32 EFPY represents the end of the requested deferral period. The combination of the Ni 0/0 and Cu % detern1ines the chemistry factor which is itself bounded by the USNRC independent aSSeSS111ent. Alioll Sciellce & Techuology PB T05. G03

lSI Progrtll1l Plan Peach Bottom Atomic Power Station UllitS 2 lIlId 3, FOllrth Interval 10CFRSO.5Sa RELIEF REQUEST: 14R-41 Revision 0 (Page 5 of 6) PBAPS, Units 2 and 3 - Training and Procedures The following information provides justification that PECO Energy has ilnplelnented operator training and established procedures at PBAPS, Units 2 and 3 that limit the frequency of cold over-pressure events to the anl01lnt specified in the staffs July, 1998 safety evaluation. PEeo Energy has in place procedures which monitor and control reactor pressure, telnperature, and water inventory during all aspects of cold shutdown which would minilnize the likelihood of a Low Temperature Over-Pressurization (LTOP) event from occurring. Additionally, these procedures are reinforced through operator training. The code Leakage Pressure Test and the code Hydrostatic Pressure Test procedures which have been used at PBAPS, have sufficient procedural guidance to prevent a cold, over-pressurization event. The Leakage Pressure Test is performed at the conclusion of each refueling outage, while the Hydrostatic Pressure Test is performed once every ten years. Other pressurizations required for infornlationalleakage inspections are perfornled in accordance with a procedure sinlilar to the ASNIE Code test procedures. These pressurizations are infrequently-perforn1ed, complex tasks, and the test procedures are considered Plant Evolution / Special Tests. As such, a requirenlent is included in them for Operation's Section n1anagement to perfornl a "pre-job briefing" with all essential personnel. This briefing details the anticipated testing evolution with special elnphasis on: conservative decision 111aking, plant safety awareness, lessons learned fronl similar in-house or industry operating experiences, the inlportance of open conlnlunications, and, finally, the process in which the test would be aborted if plant systenls responded in an adverse n1anner. Vessel ten1perature and pressure are required to be monitored throughout these tests to ensure conlpliance with the Technical Specification pressure-ten1perature curve. Also, the procedures require the designation of a Test Coordinator for the duration of the test who is a single point of accountability, responsible for the coordination of testing frOln initiation to closure, and n1aintaining Shift Managen1ent and line lnanagelnent cognizant of the status of the test. Additionally, to ensure a controlled, deliberate pressure increase, the rate of pressure increase is adnlinistratively lin1ited throughout the perfornlance of the test. If the pressurization rate exceeds this lin1it, direction is provided to relnove the CRD pun1ps, which are used for pressurization, fron1 service. \\Vith regard to inadvertent systetn injection resulting in an LTOP condition, the high pressure nlake-up systems (High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) systenls, as well as the norn1al feedwater supply (via the Reactor Feedwater Pumps>> at PBAPS are all steam driven. During reactor cold shutdown conditions, no reactor steam is available for the operation of these systems. Therefore, it is not possible for these systems to contribute to an over-pressure event while the unit is in cold shutdown. Although auxiliary steatn is used to test the associated turbines while the plant is shutdown, the plln1p is uncoupled fr0111 the turbine during the actual test which would prevent an LTOP condition. Alion Science & Technology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 alld 3, FOllr/I, Interval 10CFR50.55a RELIEF REQUEST: 14R-41 Revision 0 (Page 6 of 6) Procedural control is also in place to respond to an unexpected or unexplained rise in reactor \\vater level \\vhich could result froin a spurious actuation of an injection system. Actions specified in this procedure include preventing condensate plllnp injection, securing ECCS systenl injection, tripping CRD pUlnps, tenninating all other injection sources, and lowering RPV level via the RWCU system. In addition to procedural barriers, Licensed Operator Training is in place which further reduces the possibility of the occurrence of LTGP events. During initial Licensed Operator Training the following topics are covered: Brittle fracture and vessel thernlal stress; Operational Transient (OT) procedures, including the OT on reactor high level; Technical Specification training, including Section 3.4.9, "ReS Pressure and Tenlperature (PIT) Liinits"; and, Sinlulator Training of plant heatup and cooldown including performance of surveillance tests which ensure pressure-telnperature curve con1pliance. In addition, operator training has been provided on the expectations for procedural cOlnpliance, as provided for in the Stations' Operations Nlanual. In addition to the above, ongoing review of industry operating plant experiences is conducted to ensure that the PECO Energy procedures consider the impact of actual events, including LTOP events. Appropriate adjusttnents to the procedures and associated training are then inlplen1ented, to preclude similar situations from occurring at PBAPS. Condusion Based on the docunlentation in BWRVIP-05, the guidance provided in GL 98-05, the risk-informed independent assessment perfonned by the USNRC staff: and the additional infonnation provided above, PEcn Energy believes that pennanent relief fr01n the RPV circunlferential shell \\velds examinations at PBAPS, Units 2 and 3 is justified. Alioll Science & Technology PBT05.G03

lSI Progrtl11l Phlll Petlch Bottom Atomic Power Station Vnits 2 aud 3, Fourtll Il1terlJlll 10CFR50.55a RELIEF REQUEST: I4R-44 Revision 0 (Page 1 of 8) Rclief Request for Alternate Risk-Informed Sclection and Exaluination Criteria for Examination Catcgory B-F, B-J, C-F-l, and C-F-2 Pressure Retaining Piping \\-Velds In Accordance with 10CFR50.55a(a)(3)(i) 1.0 ASME CODE COMPONENTS AFFECTED: Code Class:

Reference:

Exan1ination Category: Item NUlnber:

== Description:== Component Nllnlber: 1 and 2 Table IWB-2500-1, Table IWC-2500-1 B-F, B-J, C-F-l, and C-F-2 B5.10, B5.20, B9.l1, 89.21, B9.31, B9.32, B9.40, C5.11, C5.51, and C5.81 Alternate Risk-lnforn1ed Selection and Exan1ination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Unit 2 and Unit 3 Pressure Retaining Piping 2.0 APPLICABLE CODE EDITION AND ADDENDA: The Inservice Inspection progranl is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2001 Edition through the 2003 Addenda. 3.0 APPLICABLE CODE REQUIREMENT: Table IWB-2500-1, Exmnination Category B-F, requires volUlnetric and surface examinations on all welds for Iteln Number B5.1 0 and surface exanlinations for all welds for Iteln NU111ber B5.20. Table IWB-2500-1, Exmninatiol1 Category 8-J, requires voh.ll11etric and surface exan1inatiol1s on a sanlple of welds for Item Numbers 89.11 and B9.31 and surface exan1inations on a san1ple of welds for Item Numbers B9.21, B9.32, and 89.40. The weld population selected for inspection includes the following: 1. All tenninal ends in each pipe or branch run connected to vessels. 2. All tenninal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the foIlo\\ving lilnits under loads associated \\vith specific seismic events and operational conditions: Alion Science & Technology PBT05.G03

lSI Program Plall Peach Bottom Atomic Power Station UllitS 2 alld 3, FOlirthlllterval 10CFR50.55a RELIEF REQUEST: I4R-44 Revision 0 (Page 2 of 8) a. prinlary plus secondary stress intensity range of 2.4Sm for ferritic steel and austenitic steel. b. cUluulative usage factor U of 0.4. 3. All dissilnilar nletal welds not covered under Exmuination Category B-F. 4. Additional piping welds so that the total number of circunlferential butt welds, branch connections, or socket welds selected for exanlination equals 250/0 of the circumferential butt welds, branch connection, or socket \\velds in the reactor coolant piping systenl. This total does not include welds exenlpted by Paragraph IWB-I220. Table IWC-2500-1, Examination Categories C-F-I and C-F-2 require volUluetric and surface eXaIllinations on a sample of welds for Henl NUlllbers C5.11 and C5.51 and surface examinations on a sample of welds for Itenl NUlllber C5.81. The weld population selected for inspection inch;~des the following: 1. Welds selected for exanlination shall include 7.5%, but not less than 28 welds, of all dissinlilar nletal, austenitic stainless steel and high alloy welds (Exanlination Category C-F-I) or of all carbon and lo\\v alloy steel welds (Exmnination Category C-F-2) not exenlpted by Paragraph IWC-I220. (Sonle welds not exempted by Paragraph I\\VC-1220 are not required to be nondestructively exanlined per Exanlination Categories C-F-l and C-F-2. These welds, however, shall be included in the total weld count to which the 7.50/0 sanlpling rate is applied.) The exanlinations shall be distributed as follows: a. the exanlinations shall be distributed mnong the Class 2 systems prorated, to the degree practicable, on the nunlber of nonexelnpt dissinlilar metal, austenitic stainless steel and high alloy \\velds (Exmnination Category C-F-

1) or carbon and low alloy welds (Exmuination Category C-F-2) in each system; b.

within a system, the eXaIl1inations shall be distributed aIllong ternlinal ends, dissimilar metal welds, and structural discontinuities prorated, to the degree practicable, on the number of nonexempt terminal ends, dissimilar nletal welds, and structural discontinuities in the system; and c. within each systenl, exanlinations shall be distributed bet\\veen piping sizes prorated to the degree practicable. Alion Science & Teclll1ology PB T05. G03

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 amI 3, FOllrtl, Interval 10CFR50.55a RELIEF REQUEST: 14R-44 Revision 0 (Page 3 of 8) 4.0 REASON FOR REQUEST: Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative utilizing Reference 1 along with two enhancen1ents from Reference 4 will provide an acceptable level of quality and safety. As stated in "Safety Evaluation Report Related to EPRI Risk-Inforn1ed Inservice Inspection EV8luation Procedure (EPRI TR-112657, Revision B, July 1999)" (Reference 2): "The staff concludes that the proposed RISI progrmn as described in EPRI TR-112657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10CFR50.55a for the proposed alternative to the piping lSI requiren1ents with regard to the nun1ber of locations, locations of inspections, and n1ethods of inspection." The initial Peach BOtt0111 Atomic Power Station Risk-Informed Inservice Inspection (RISI) progran1 was submitted during the first period of the third interval for Unit 2 and during the second period of the third interval for Unit 3. This initial RISI program was developed in accordance with EPRI TR-112657, Revision B-A, as supplen1ented by Code Case N-578-1. The program was approved for use by the USNRC via a Safety Evaluation as transn1itted to Exelon (Reference 5). The transition from the 1989 Edition to the 2001 Edition through the 2003 Addenda of ASME Section XI for Peach Bottom Atolnic Power Station's fourth interval does not impact the currently approved Risk-Infonned lSI evaluation methods and process used in the third interval, and the requiren1ents of the new Code Edition!Addenda will be implclnented as detailed in the Peach Bottom Atomic Power Station lSI Progratn Plan. Therefore, with the exception of specific weld locations that may have changed due to maintenance or modification activities, the proposed alternative RISI program for the fourth interval is the same prograrn as approved in Reference 5 for the third interval. The Risk In1pact Assessn1ent cOlnpleted as part of the original baseline RISI Program was an implementationJtransition check on the initial impact of converting from a traditional ASME Section XI progranl to the new RISI methodology. For the fourth interval lSI update, there is no transition occurring between two different n1ethodologies, but rather, the currently approved RISI methodology and evaluation will be maintained for the new interval. The originallnethodology of the calculation has not changed, and the change in risk \\vas sin1ply re-assessed using the initial 1989 ASME Section XI progrmu prior to RISI and the new elen1ent selection for the fourth IO-year interval RISI program. This Alion Science & Tecllllology PB TOS. G03

lSI Program Plan Peach Bottom Atomic Power Station Uuits 2 al1d 3, Fourtlt Interval 10CFR50.55a RELIEF REQUEST: I4R-44 Revision 0 (Page 4 of 8) Saine process has been maintained in each revision to the Peach Bottonl Atomic Power Station RISI asseSSlnent that has been perfornled to date. The actual '"evaluation and ranking" procedure including the Consequence Evaluation and Degradation Mechanisnl AsseSSlnent processes of the currently approved (Reference

5) RISI Program relnain unchanged and are continually applied to nlaintain the Risk Categorization and Elelnent Selection nlethods of EPRI TR-112657, Revision B-A.

These portions of the RISI Program have been and will continue to be reevaluated and revised as major revisions of the site Probabilistic Risk Assessnlent (PM) occur and modifications to plant configuration are made. The Consequence Evaluation, Degradation Mechanisn1 Assesslnent, Risk Ranking, and Elenlent Selection steps encompass the cOl11plete program process applied under the Peach Bottonl Atonlic Power Station RISI PrograIn. 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: The proposed alternative originally inlplemented in the risk infonned in-service inspection plan for Peach Bottonl Atomic Povver Station Units 2 and 3 (Reference 3), along with the two enhancenlents noted below, provide an acceptable level of quality and safety as required by IOCFR50.55a(a)(3)(i). This original program along with these san1e two enhancements is currently approved for Peach BOttOI11 Atolnic Power Station's third inservice inspection interval as docunlented in Reference 5. The fourth interval RISI program will be a continuation of the current application and will continue to be a living program as described in the Reason For Request section of this relief request. No changes to the evaluation 111ethodology as currently il11plemented under EPRI TR-112657, Revision B-A, are required as part of this interval update. The following two enhancenlents will continue to be inlplell1ented. a. In lieu of the evaluation and sample expansion requirelnents in Section 3.6.6.2, "RISI Selected Examinations" of EPRI TR-112657, Peach Bottonl Atomic Power Station will utilize the requirelnents of Paragraph -2430, '"Additional Exanlinations" contained in Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in Code Case N-578-1 provide a 1110re refined methodology for inlplementing necessary additional exanlinations. The reason (or this selection is that the guidance discussed in EPRI TR-112657 includes req uirenlents for additional examinations at a high level. based on service conditions, degradation 111echanisms, and the perfornlance of evaluations to deternline the scope of additional eXaIninations, whereas ASME Code Case N-578-1 provides more specific and clearer guidance regarding the requirements for additional examinations that is structured similar to the guidance provided in ASME Section XI, Paragraphs IWB-2430 and IWC-2430. Additionally, silnilar Aliol1 Science & Techuology PBT05.G03

lSI Program Plall Peach Bottom Atomic Power Statioll Units 2 and 3, Fourth Interval 10CFRSO.S5a RELIEF REQUEST: I4R-44 Revision 0 (Page 5 of 8) to the current requirements of ASME Section XI, Peach Botton1 Atomic Power Station intends to perforn1 additional exan1inations that are required due to the identification of flaws or relevant conditions exceeding the acceptance standards, during the outage the flaws are identified. b. To supplement the requiren1ents listed in Table 4-1, "Sum1nary of Degradation-Specific Inspection Requiren1ents and Exanlination Methods" of EPRI TR-112657, Peach Bottom Atomic Power Station will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Infonned Piping Examinations" contained in Code Case N-578-1 (Reference 4). To in1plement Note 10 oftl1is table, paragraphs and figures fro1n the 2001 Edition through the 2003 Addenda of ASME Section XI (Peach Botton1 Atomic Power Station's Code of record for the Fourth Interval) will be utilized which parallel those referenced in the Code Case for the 1989 Edition. Table 1 of Code Case N-578-1 will be used as it provides a detailed breakdown for exan1ination nlethod and categorization of parts to be exanlined. Additionally, Section 4 of EPRI TR-112657 states "Application of RISI uses NDE techniques that are designed to be effective for specific degradation Inechanisms and exmnination locations." Section 4 also identifies n1ethods of exan1ination for each degradation n1echanism with the prin1ary n1ethod being ultrasonic testing (UT) techniques. However, EPRI TR-112657 does not identify the exan1ination volumes for components without a degradation n1echanisn1. In addition, EPRI TR-112657 does not specify examination volUlnes and Inethods for socket welds. Peach Bottom Aton1ic Power Station has requested to use the exanlination methods fron1 Code Case N-578-1 instead of the nlethocls froln EPRI TR-112657. The exan1ination figures specified in Section 4 of EPRI TR-112657 will be used to determine the exan1ination volun1e based on the degradation mechanisln and con1ponent configuration. Peach Bottoln Atonlic Power Station uses UT techniques for RISI volunletric exanlinations. For the components addressed by the RISI prograln, ASME Section XI focuses prilnarily on weld examinations. Risk Informed examination voh.1l11eS also include portions of piping and fitting base materials that are susceptible to particular degradation mechanisn1S. The ASME Section XI, Mandatory Appendix I, "Ultrasonic Exan1inations," specifies that UT exan1ination procedures, equipment, and personnel used to detect and size flaws in piping welds shall be qualified by perfonnance denlonstration in accordance with ASME Section XI Appendix VIII, "Performance Den10nstration for Ultrasonic Examination Systen1s." The RISI progran1 con1plies vvith Appendix VIII for weld exan1inations. In cases where the examination requiren1ents cannot be n1et, Peach BottOln Atomic Power Alion Science & Technology PB T05. G03

lSI Pmgram Plan Peach Bottom Atomic Power Station Ullit~' 2 alld 3, Fourth Interval 10CFR50.55n RELIEF REQUEST: 14R-44 Reyision 0 (Page 6 of 8) Station will subn1it a request for relief in accordance with 10CFR50.55a, "Codes and standards." The exan1ination Inethods are designed to be effective for specific degradation Inechanislns and examination locations. The volun1ctric scanning will be in both axial and circU1nferential directions to detect the flaws in these orientations. Additionally, all Peach Bottom Ato111ic Power Station dissinlilar n1etals (OM) welds, as characterized in ASME Section XI, Article IWA-9000, have been evaluated for failure potential and consequence of failure along with the other non-exen1pt piping. The piping seglnents containing the OM welds were classified into the appropriate RISI categories, and appropriate elelnents were selected per the category requirements for exmnination during the third inspection interval. Piping welds, including OM welds in vessel nozzles, that are susceptible to IGSCC (i.e., IGSCC Categories B through G, as applicable) and not subject to other degradation Inechanism(s) are removed fr01n the RISI progran1 population. They are contained in the Peach Bott01n Aton1ic Power Station lSI Auglnented Program (AUG-O I), "USNRC Generic Letter 88-0 I, Intergranular Stress Corrosion Cracking" and are subject to the inspection requiren1ents ofBWRVIP-75-A "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules". Furthennore, all piping welds and welds, including DM welds in vessel nozzles classified as Category A (resistant material) per BWRVIP-75-A are included in the RISI progrmn. The Peach Bottom At01nic Power Station RISI Program, as developed in accordance with EPRI TR-112657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as "High" risk (i.e., Risk Category 1, 2, and 3) and 100/0 of the elements that are categorized as "Medium" risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the exan1ination volume for a given degradation mechanisn1 is provided by the EPRI TR-112657 while the guidance for the exan1ination method and categorization of parts to be exmnined are provided by the EPRI TR-112657 as supplen1ented by Code Case N-578-1. In addition to this risk-inforn1ed evaluation, selection. and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to receive Code required pressure testing as part of the current ASME Section XI progrmn. VT-2 visual exan1inations are within the AS!vlE pressure boundary and are exan1ined each refueling outage as part of the systeln leakage test required by ASME Section XI. These examinations are scheduled in accordance with the Peach Bott01TI Atomic Power Station pressure-testing program, which ren1ains unaffected by the RISI prognun. AUoll Science & Techllology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Stlltioll Ullit~' 2 lI/Ill 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: I4R-44 Revision 0 (Page 7 of 8) 6.0 DURATION OF PROPOSED ALTERNATIVE: Relief is requested for the Fourth Ten-Year Inspection Interval for Peach Bottonl Atomic Power Station Units 2 and 3. The Fourth lSI Interval begins on Noven1ber 5, 2008 for both Units as proposed in Relief Request 13R-45 (Third lSI Interval Relief Request), and will conclude Novelllber 4,2018 for Peach Bottonl Atonlic Power Station, Units 2 and 3.

7.0 PRECEDENTS

Silnilar relief requests have been approved for: 1. Peach BoUOln Atomic Power Station Third Inspection Interval Relief Request RR-44 was approved per SER dated August 27, 2003. The Fourth Inspection Interval Relief Request will utilize the same RISI n1ethodology that was previously approved in the Third Inspection Interval. 2. Lilnerick Generating Station Second Inspection Interval Relief Request RR-32 was approved per SER dated March 3, 2003. 3. Susquehanna Stean1 Electric Station Third Inspection Interval Relief Request I3R-0 1 was approved per SER dated July 28, 2005. 4. Dresden Station Fourth Inspection Interval Relief Request I4R-02 was approved per SER dated Septelnber 4, 2003. 5. Quad Cities Station Fourth Inspection Interval Relief Request I4R-02 was approved per SER dated January 28, 2004.

8.0 REFERENCES

1. Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Inforl11ed Inservice Inspection Evaluation Procedure," Deceluber 1999. 2. W. H. Bateillan (USNRC) to G. L. Vine (EPR!) letter dated October 28, 1999 transn1itting "Safety Evaluation Report Related to EPRI Risk-Infonned Inservice Inspection Evaluation Procedure (EPRI TR-l12657, Revision 13, July 1999)." 3. Initial Risk-Infonned Inservice Inspection Evaluation - Peach BOtt0111 At0111ic Power Station Units 2 and 3, dated May 16, 2002. 4. Alnerican Society of Mechanical Engineers (ASME) Code Case N-578-1, "Risk-Inforn1ed Requirenlents for Class 1, 2, or 3 Piping, Method B." /.lion Sciellce & Tecllllology PBT05.G03

lSI Program PIlIn Peach Bottom Atomic Power Station Units 2 a/ld 3, Fourth I/lterval 10CFR50.55a RELIEF REQUEST: 14R-44 RevisiQD 0 (Page 8 of 8) 5. James W. Clifford (USNRC) to John L. Skolds (Exelon) letter dated August 27, 2003 transnlitting the "Peach Bottonl Atonlic Povver Station, Units 2 and 3 - Anlerican Society of Mechanical Engineers Boiler and Pressure Vessel Code-Relief for Risk-Infonned Inservice Inspection of Piping (TAC Nos. MB5512 and MB5SI3)." AliO/l Science & Tecllllology PBT05.G03

lSI Program Plall Peach Bottom Atomic Power Stll1ion Ullit!t* 2 alld 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: I4R-46 Revision 0 (Page 1 of 4) (Note: Revision 0 Not Required per SER dated 02/26/09) Relief Request for Alternative Exalnination Requirements of ASME Section XI, I>aragraph IWA-5244, "Buried Components" In Accordance with 10CFR50.55a(a)(3)(ii) 1.0 ASME CODE COMPONENTS AFFECTED: Code Class:

Reference:

Examination Category: Item Number:

== Description:== Component Nun1ber: Drawing Number: 3 IWA-5244 D-B D2.10 Alternative Exmuination Requirements of ASME Section XI, Paragraph IWA-5244, "Buried Con1ponents" Return Piping: 2-32HF-24", 3-32HF-24", 0-33HF-20" ISI-330, Sheet 1; ISI-315, Sheet 1; ISI-315, Sheet 3 2.0 APPLICABLE CODE EDITION AND ADDENDA: The Inservice Inspection progran1 is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2001 Edition through the 2003 Addenda. 3.0 APPLICABLE CODE REQUIREMENT: Table IWD-2500-1, Exan1ination Category D-B, Iten1 Number 02.10, requires all Class 3 pressure retaining con1ponents be subject to a systelu leakage test with a VT-2 visual examination in accordance with Paragraph IWD-5220. This pressure test is to be conducted once each inspection period. Paragraph IWA-5244(b)(1) requires buried cOluponents that are isolable by n1eans of valves be tested to determine the rate of pressure loss. Alternatively, the test lllay determine the change in now between the ends of the buried components and the Owner shall establish the acceptable rate of pressure loss or now. 4.0 REASON FOR REQUEST: Pursuant to 10CFR50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty vvithout a compensating increase in the level of quality and safety. Aliol1 Science & Technology PBT05.G03

lSI Program Plan Pellclt Bottom A.tomic Power Station Units 2 aud 3, FOllrtlt Illterval 10CFR50.55a RELIEF REQUEST: I4Rw46 Revision 0 (Page 2 of 4) The buried piping in question consists of one 20" dimneter COlnnlon (i.e., Unit 0) return header for En1ergency Service Water (ESW) downstremn of MO-O-33-0498 to the discharge pond; and two 24" dialneter return headers for High Pressure Service \\Vater (HPSW) downstremn of MO-2-32-2486 and MO-3-32-3486 to the discharge pond (one for Unit 2 and one for Unit 3) (See Figure I4R-46-1). These components are all buried between the Motor Operated Valves and the discharge pond with the exception of a valve pit that includes a nlanually operated gate valve and small amount of the associated piping on each of the HPSW and ESW returns to the discharge pond. There is no access to the buried sections \\vithout excavation. In addition, no annulus was provided during original construction that would allo\\v for examination of these buried sections of piping. Paragraph IWA-5244(b)(1) requires that the buried sections of piping be exanlined by a pressure decay test or a test that detennines the change in flow between the buried ends. In order to perform a pressure decay test, it would be necessary to close the associated large gate valves' to isolate the buried portion of each return header. This \\vould also result in the isolation of portions of the Elnergency Core Cooling Systems (ECCS), which would place Technical Specification limitations on the plant. These gate valves in the return headers are not expected to provide the leak tight capability \\vhich would be necessary to perfonn a pressure decay test due to the age of the valves. In order to perfonn a pressure decay test, it would be necessary to either replace these gate valves \\vith valves that possess better leakage characteristics or to install blind flanges on the pIpIng. The other potential test would be a change in flow test. However, the buried ECCS return headers were not designed with the plant instrU111entation and flow orifices that would be required to detennine the flow rates. Installation of flow measureinent devices would result in plant modifications. Conlpliance with the specified requirements is a hardship without a cOlnpensating increase in the level of quality and safety as discussed above. 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: For the HPSW and ESW buried piping sections Peach BOtt01l1 Atomic Power Station proposes to uti lize the requiren1ents of Paragraph I\\VA-5244(b)(2) along with additional data obtained during quarterly Inservice Testing (1ST) trending to provide an adequate level of quality and safety. The Paragraph IWA-5244(b)(2) requirelnents call for a test that confinns now is unilnpaired in nonisolable buried cOll1ponents. To confirm that flow I Unit 2: Unit 3: Common: HPSW: HV-2-032-22000 HPSW: HV-3-032-32200 ESW: HV-O-033-11200 Alioll Science & Tecllllology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 (llld 3, Fourth I"terval 10CFRSO.SSa RELIEF REQUEST: I4R-46 Revision 0 (Page 3 of 4) is uninlpaired in these buried pipes, Peach Bottom AtOlnic Power Station Inservice Testing procedures will be used to ensure adequate flow during operation. Peach BottOln Atonlic Power Station will use the Owner established Inininlunl flow rate contained in the site 1ST surveillances as the acceptance criteria for Paragraph IWA-5244 pressure testing of IIPSW and ESW system buried piping. If the Inininlunl flow could not be achieved during the course of an 1ST survei!lance, and the cause of the deviation was not attributed to the test instruments being used, the associated systenl would be declared inoperable as required under the 1ST surveillance and an Issue Report (IR) would be generated in accordance with the Exelon Corrective Action Progrmn. Further corrective actions (i.e., Inaintenance on the punlp, system walk downs, etc.) would be initiated as required to restore the punlp and/or the systenl back to an operable condition. 6.0 DURATION OF PROPOSED ALTERNATIVE: Relief is requested for the fourth ten-year inspection interval for Peach Bottonl Atomic Power Station, Units 2 and 3. The Fourth lSI Interval is scheduled to begin on November 5,2008 as proposed in relief request I3R-45 (Third lSI Interval Relief Request) and will conclude Novenlber 4, 2018 for Peach Bottonl Atonlic Power Station, Units 2 and 3.

7.0 PRECEDENTS

A silnilar relief request was previously approved for Byron and Braidwood Stations eM. L. Marshall, Jr. (NRR) to C.M. Crane (Exelon), "Byron Station, Unit Nos 1 and 2, and Braidwood Station, Unit Nos. 1 and 2 Evaluation of Inservice Inspection Progranl Relief Requests I3R-07 and I2R-46 pertaining to Essential Service Water Buried Piping (TAC NOS, MD1757, MD1758, MD1760)," dated January 16,2007). AUolI Science & Techllology PB T05. G03

lSI Progrtllli Plan Pelleh Bottom Atomic Power Station Units 2 alld 3, FOllrth Interval 10CFR50.55a RELIEF REQUEST: 14R-46 Revision 0 (Page 4 of 4) FIGURE 14R-46-1 BURIED COMPONENTS Alion Science & Technology PBT05.G03

lSI Progrtlm Pltl1l Peach Bottom Atomic Power Station Units 2 alld 3, FOllrth Interval 10CFR50.55a RELIEF REQUEST: I4R-47 Revision 1 (Page 1 of 4) Relief Request for Testing of Control Rod Drive Boundaries In Accordance with 10CFR50.55a(a)(3)(i) 1.0 ASlVIE CODE COMPONENTS AFFECTED: Code Class:

Reference:

Exan1ination Category: Item NutTIber:

== Description:== Con1ponentNun1ber: Drawing Nun1ber: 1 IWB-5222(b), Table IWB-2500-1 B-P B15.10 Testing of Control Rod Drive Pressure Boundaries Class 1 piping between CV-2(3)-03A-13-127 (valves AA through HC inclusive, total of 185 valves) and HV-2(3-03A-13112 (valves AA through HC inclusive, total of 185 valves) ISI-357, Sheets 1 & 2 2.0 APPLICABLE CODE EDITION AND ADDENDA: The Inservice 1nspection progratTI is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2001 Edition through the 2003 Addenda. 3.0 APPLICABLE CODE REQUIREMENT: IWB-5222(b) requires the pressure retaining boundary during the systetTI leakage test conducted at or near the end of each inspection interval shall be extended to all Class I pressure retaining components within the systen1 boundary. Table IWB-2500-1, Examination Category B-P, Item NUlTIber B15.1 0, requires all Class 1 pressure retaining con1ponents be subject to a systelTI leakage test with a VT-2 visual exarrlination in accordance with Paragraph IWB-5220. This pressure test is to be conducted prior to plant startup follo\\ving each reactor refueling outage. 4.0 REASON FOR REQUEST: Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative provides an acceptable level of quality and safety. The piping in question is the Class 1 piping between the CV-2(3)-03A-13-127 valve (valves AA through HC inclusive, total of 185 valves) and the HV-2(3 )-()3A-13112 valve Afioll Science & Technology PB T05. G03

lSI Progrt'l11 Pia" Peac" Bottom Atomic Power Station Units 2 and 3, FOllrt" Interval 10CFR50.55a RELIEF REQUEST: 14R-47 Revision 1 (Page 2 of 4) (valves AA through HC inclusive, total of 185 valves) for each of the 185 Control Rod Drive Mechanisms (See Figure I4R-47-1). During nonnal systeln lineup required for startup, the CV-2(3)-03A-13-127 valves are in the closed position. The HV-2(3)-03A-13112 valves are in the open position. The only tin1e the CV-2(3)-03A-13-127 valves are open is during a plant scram or during CRD Scran1 Tillle testing. During the performance of the systen1 leakage test prior to plant startup following a refueling outage, the test boundary is the reactor coolant boundary vvith all valves in the position required for normal reactor operation startup (the CV-2(3)-03A-13-127 valves are closed). This test is conducted in accordance with Table IWB-2500-1, Exan1ination Category B-P, Item NUlnber B15.10, and Paragraph IWB-5222(a). As required by Paragraph IWB-5222(a), the visual exan1ination is to extend to the second closed Class 1 valve at the extren1ity of the boundary. This test is conducted in accordance with the code. However, during the systen1 leakage test conducted at or near the end of the Ten-Year lSI Interval in accordance with Paragraph IWB-5222(b), the pressure boundary extends to all Class 1 pressure retaining components which includes the piping between CV-2(3)-03A-13-127 and the Scram Discharge VolUlne (HV-2(3)-03A-13112). In order to pressurize the piping between the CV-2(3)-03A-13-127 valves and the HV-2(3)-03A-13112 valves and hold it for inspection, all 185 HV-2(3)-03A-13 112 valves would have to be manually closed prior to inserting a Scratll which is a manpower intensive activity. Alternately, the piping could be pressurized for testing by isolating each of the 185 segments of piping, and pressurizing with a manual hydro pUlnp. This approach would involve filling and venting of the subject piping, and lnanipulating 4 valves and installing a fill hose at a threaded connection for each ofthe 185 piping segillents. This activity is also n1anpower intensive. This section of piping does not pressurize to nominal reactor operating pressure except for a very brief tin1c during plant Scrams and Scram Time testing silnilar to normal operating conditions. If this piping were to develop a leak, it would be identified during the SCratll Time testing by the personnel perforn1ing the testing. The piping between the HV-2(3)-3A-13 112 valves and the CHK-2(3)-3A-13114 valves is pressurized and tested during the cOlllpletion of the Scralll Discharge VolUlne Systen1 leakage tests. Table IWB-2500-1, Exalnination Category B-P, Iten1 NUlnber BI5.10, Note 2, states that the systen1 leakage test shall be conducted prior to plant startup following each reactor refueling outage. SCratll Tin1e testing is required to be 100% con1plete for each outage Alion Science & Tecllll%gy PBT05.G03

lSI Program Pltlll Peach Bottom Atomic Power Station Units 2 and 3, FOllrth Interval 10CFR50.55a RELIEF REQUEST: I4R-47 Revision 1 (Page 3 of 4) prior to exceeding 40% Reactor Thermal Power per Peach BOttOll1 Technical Specifications 3.1.4, which is after plant startup following a refueling outage. The piping in question is approximately 24 inches of 3;4 inch non1inal OD schedule 80 stainless steel socket \\velded piping for each control rod drive. This piping is not susceptible to a corrosive environment nor is it susceptible to vibrations that would induce cracking. There have been no known leaks in this piping at Peach Bottom Atomic Power Station. Units 2 and 3. 5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: As a proposed alternative in accordance with 10CFR50.55a(a)(3)(i), for the portion of the piping between the CV-2(3)-03A-13-127 valves and the HV-2(3)-03A-13112 valves for each of the 185 Control Rod Drives, Peach Bottom Aton1ic Power Station proposes to use the Scram Tin1e Testing that is performed for each Control Rod Drive after plant startup but prior to achieving 400/0 power at the conclusion of the outage at or near the end of the interval as a means to pressurize this piping. A VT-2 visual qualified individual will be present during the Scratn Tin1e Testing to perforn1 each visual examination. Additionally, as part of this test, procedures will be revised to ensure that the VT-2 visual exan1iner confirn1s with the control 1'00111 that the exmnination is c0111plete prior to the test switch being returned to norn1al. 6.0 DURATION OF }>ROPOSED ALTERNATIVE: Relief is requested for the fourth ten-year inspection interval for Peach Botton1 Atomic Power Station, Units 2 and 3. The Fourth lSI Interval is scheduled to begin on Noven1ber 5, 2008 as proposed in relief request 13R-45 (Third lSI Interval Relief Request) and will conclude on Novelnber 4, 2018 for Peach Bottom AtOlnic Power Station, Units 2 and 3.

7.0 PRECEDENTS

None Alioll Sciellce & Technology PBT05.G03

lSI Program Plan Peaclt Bottom Atomic Power Statioll Units 2 and 3, Fourtlt Interval 10CFR50.55a RELIEF REQUEST: 14R-47 Revision 1 (Page 4 of 4) FIGURE I4R-47-1 PRESSURE RETAINING BOUNDARY Aliol1 Science & Tecllllology P8 T05. G03

lSI Program Plt/n Peach Bottom Atomic Power Station Units 2 and 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: I4R-48 (CRR-13) Revision 0 (Page 1 of 5) Relief Request for Alternative Examination Requirements of ASME Section XI, Paragraph IWE-1232, "In~lccessible Surface Areas" In Accordance with IOCFR50.55a(g)(5)(iii) 1.0 ASME CODE COMPONENTS AFFECTED: Code Class:

Reference:

Examination Category: ItetTI NU111ber:

== Description:== Component Nun1ber: Drawing Number: MC IWE-1232 E-A E1.11, E1.12 Alternative Examination Requireillents of ASME Section XI, Paragraph IWE-1232, "Inaccessible Surface Areas" Penetration N-3 6280-S-53, 6280-C2-1 03-6, and 6280-C2-341-3 2.0 APPLICABLE CODE EDITION AND ADDENDA: The first Containnlent Inservice Inspection (CISl) interval began on Noveillber 5, 1998 for PBAPS, Units 2 and 3, and complied with the ASME B&PV Code, Section Xl, 1992 Edition through 1992 Edition Addenda. The Second CISI Interval will begin on November 5,2008, and will con1ply vvith the ASME B&PV Code, Section XI, 2001 Edition through 2003 Addenda. The original construction code for the Drywell is Section III, Type B, 1965 Edition through SU111mer 1966 Addenda. 3.0 APPLICABLE CODE REQUIREMENT: Paragraph IWE-1232(a) ("Inaccessible Surface Areas") states that portions of Class MC containnlent vessels, parts, and appurtenances that are enlbedded in concrete or othenvise 111ade inaccessible during construction of the vessel or as a result of vessel repair-lTIoditication, or replacement are exelnpted fron1 examination, provided: (1) no openings or penetrations are elnbedded in the concrete; (2) all welded joints that are inaccessible for exan1ination are double butt welded and are fully radiographed and, prior to being covered, are tested for leak tightness using a gas nlediunl test, such as Halide Leak Detector Test; and Alion Science & Technology PBT05.G03

IS] Program Plan Peach Bottom Atomic Power Stalion Units 2 aud 3, FOllrth Interval 10CFR50.55a RELIEF REQUEST: I4R-48 (CRR-13) Revision 0 (Page 2 of 5) (3) the vessel is leak rate tested after cOlnpletion of construction or repair/replacement activities to the leak rate requiren1ents of the Design Specifications. 4.0 IMPRACTICALITY OF COMPLIANCE: Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requiren1ents is iInpractical as conformance would require extensive modifications to the prin1ary containn1ent. When the dryvvell was being constructed, a 24-inch 111anhole \\vas placed in the botton1 head of the drywell. During constnlction, when the manhole vvas no longer needed, the penetration was seal welded, inspected, and elnbedded in concrete. Based on the original construction drawings the n1anhole is a bolted, gasket connection that was seal welded, the handles were ground smooth and either a magnetic particle test or dye penetrant examination was perfonned. The N-3 n1anhole was seal welded and cannot meet the Paragraph IWE-1232(a)(2) code requiren1ent for a double butt weld. See Figures I4R-48-1 (CRR-13-1) and I4R-48-2 (CRR-13-2) for n10re details. 5.0 BURDEN CAUSED BY COMPLIANCE: Adding a double butt weld would involve a modification to the drywell that would require excavation of the concrete around the bottOll1 head of the drywell or removal of the drywell floor thus n1aking the code requiren1ent impractical. 6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE: Integrated Leak Rate Testing will be perforn1ed in accordance with the station Appendix J Program, which is n1aintained independent of the ASME Section XI progrmn. 7.0 DURATION OF PR()POSED ALTERNATIVE: This relief is being requested for the first (previous) and second (upcon1ing) containment inservice inspection (CISI) interval. The First CISI Interval began on November 5, 1998 for PBAPS, Units 2 and 3, and cOlnplied with the ASME B&PV Code, Section XI, 1992 Edition through 1992 Edition Addenda. The Second CISI Interval will begin on Noven1ber 5, 2008 for PBAPS, Units 2 and 3, and will con1ply with the ASME B&PV Code, Section XI, 2001 Edition through 2003 Addenda. Refer to CRR-12 (First CISI Interval Relief Request) for the associated start and end dates of the intervals. Alion Science & Tecllllology PJJT05.G03

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 alit! 3, Fourth Interval 10CFH.50.55a RELIEF REQUEST: 14R-48 (CRR-13) Revision 0 (Page 3 of 5)

8.0 PIlECEDENTS

None Alion Sciellce & Tecll1lology PBT05.G03

lSI Program Pllln Peach Bottom Atomic Power Station Units 2 alld 3, Fourth Interval IOCFR50.55a RELIEF REQUEST: 14R-48 (CRR-13) Revisi1>D 0 (Page 4 of 5) FIGURE I4R-48-1 (CRR-13-I) INACCESSIBLE SURFACE AREAS If-i. "l.10-~

• • ---------------:-f:L 1~7~'

/ - - --... _..._.(~~ E:l.. I Cj\\,p-3' ~ C:.:, ~. r--, (d~ ~ r-:\\ (.;"~ ~~ e ~ \\~~9 0::tJ \\~ e l ~ --=---.::-~ ~_"__'_"__L L."",. """N"',,"" (;~:J_{!-",; 5?" --===---==~ -------- -'.-- i-~~-~~~E1--*----- ~~- _~.2~,.6 ,-.-----{~}_{~rc;-.i-J~ir<;;4----O 'i_' 'CJ ~, f ~~ Alioll Science & Technology PIJT05.G03

lSI Program Plan Pellclt Bottom Atomic Power Station Units 2 (lnd 3, FOllrtlt Interval 10CFR50.55n RELIEF REQUEST: J4R-48 (CRR-13) Revision 0 (Page 5 of 5) FIGURE I4R-48-2 (CRR-13-2) INACCESSIBLE SURFACE AREAS (DETAILS) Alion Science & Technology PB T05. G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 lind 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 1 of 17) Use of B\\VRVIP Guidelines in Lieu of Specific ASME Code Requirements Proposed Alter 4native in Accordance with 10CFR50.55a(a)(3)(i) Clinton Power Station Dresden Nuclear Power Station, Units 2 and 3 LaSalle County Station, Units 1 and 2 Lin1erick Generating Station, Units 1 and 2 Oyster Creek Generating Station Peach BOttOlli Atomic Power Station. Units 2 and 3 Quad Cities Nuclear Power Station, Units 1 and 2 Alioll Science & Technology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 lind 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: I4R-49 Revision 1 (Page 2 of 17) Use of B\\VRVIP Guidelines in Lieu of Specific ASME Code Requirements Proposed Alternative in Accordance,vith 10CFR50.55a(a)(3)(i) 1.0 ASME CODE COMPONENT(S) AFFECTED ASME Section XI, Class 1, Exanlination Categories B-N-1 (Interior of Reactor Vessel) and B-N-2 (Welded Core Support Structures and Interior Attachnlents to Reactor Vessels), Code Itenl NU111bers B13.10 - Vessel Interior, B13.20 - Interior Attachnlents within Beltline Region, B13.30 - Interior Attac1unents beyond Beltline Region, and B13.40 - Core Support Structure. 2.0 APPLICABLE CODE EDITION AND ADDENDA PLANT INTERVAL EDITION START END Clinton Power Station Second ]989 Edition, No January I, 2000 December 3 I, 2010 Addenda Dresden Nuclear Power Fourth 1995 Edition, through January 20, 2003 January 19,2013 Station, Units 2 and 3 ]996 Addenda LaSalle County Stations, Third 2001 Edition, through October 1, 2007 September 30, 2017 Units 1and 2 2003 Addenda Limerick Generating Station, Third 2001 Edition, through February 1,2007 January 3 I, 2017 Units 1and 2 2003 Addenda Oyster Creek Generating Fourth 1995 Edition, through October 15, 2002 October 14,2012 Station 1996 Addenda Peach Bottom Atomic Power Fourth 200 I Edition, through November 5, 2008 November 4, 2018 Station, Units 2 2003 Addenda Peach Bottom Atomic Power Fourth 2001 Edition, through August 15, 2008 August 14, 2018 Station, Units 3 2003 Addenda Quad Cities Nuclear Power Fourth 1995 Edition through March 10, 2003 March 9, 2013 Station, Units 1and 2 1996 Addenda 3.0 API>LICABLE CODE REQUIREMENTS ASME Section XI requires the examination of C0111pOnents within the Reactor Pressure Vessel. These exa111inations are included in Table IWB-2500-1 Categories B-N-I and B-N-2 and identified with the following itenl nunlbers: Alioll Science & Technology PBT05.G03

lSI Program PlaJl Peach Bottom Atom;c Power Station Units 2 and 3, FOllrtll Illterval 10CFR50.55a RELIEF REQ UEST: I4R-49 Revision 1 (Page 3 of 17) B13.10 B13.20 B13.30 B13.40 ExaJnine accessible areas of the reactor vessel interior each period by the VT-3 visual exanlination luethod (B-N-1). Exanline interior attachnlent welds within the beltline region each interval by the VT-l visual examination luethod (B-N-2). EXaJnine interior attachnlent welds beyond the beltline region each interval by the VT-3 visual examination tuethod (B-N-2). EXaJlline surfaces of the welded core support structure each interval by the VT-3 visual examination Juethod. These exanlinations are performed to assess the structural integrity of conlponents within the boiling water reactor pressure vessel. 4.0 REASON FOR REQUEST In accordance 10CFR50.55a(a)(3)(i), Exelon Generation Company, LLC and AmerGen Energy Company, LLC are requesting a proposed alternative to the Code requirenlents provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety. The BWRVIP Inspection and Evaluation (I&E) guidelines have recOlumended aggressive specific inspection by BWR operators to cOJupletely identify luaterial condition issues with BWR conlponents. A wealth of inspection data has been gathered during these inspections across the BWR industry. I&E guidelines focus on specific and sllsceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanis111S, and require re-exaJl1ination at conservative intervals. In contrast, the code inspection requirenlents were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience. Use of this proposed alternative willillaintain an adequate level of quality and safety and avoid unnecessary inspections, while conserving radiological dose. 5.0 PROPOSED ALTERNATIVE In lieu of the requirements of ASME Section Xl, the proposed alternative is detailed in attached Table 1 for EXaJllination Category B-N-l and B-N-2. Exelon and ArnerGen will satisfy the Examination Category B-N-I and B-N-2 requirements as described in Table 1 in accordance \\\\'ith BWRVIP guideline requirements. This relief request proposes to utilize the associated BWRVIP guidelines in lieu of the associated Code requireluents including but not liluited to eXaJllination Alion Science & Technology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 ami 3, Fourth Illterval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 4 of 17) luethod, volume, frequency, training, successive and additional exalllinations, flaw evaluations, and reporting. Not all the C0111pOnents addressed by these guidelines are code cOluponents. The following guidelines are applicable to this Relief Request: - BWRVIP-03, "BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Exalllinatiol1 Guidelines" - BWRVIP-18-A, "BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines" - BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation Guidelines" - BWRVIP-26-A, "BWR Top Guide Inspection and Fla\\v Evaluation Guidelines" - BWRVIP-27-A, "BWR Standby Liquid Control Systenl/Core Plate ~P Inspection and Flaw Evaluation Guidelines" - BWRVIP-38, "BWR Shroud Support Inspection and Flaw Evaluation Guidelines" - BWRVIP-41, Revision 1 "BWR Jet PtUUP Assenlbly Inspection and Flaw Evaluation Guidelines" - BWRVIP-42-A, "LPCI Coupling Inspection and Flaw Evaluation Guidelines" - BWRVIP-47-A, "BWR Lower Plenunl Inspection and Flaw Evaluation Guidelines" - BWRVIP-48-A, "Vessel 10 Attachlllent Weld Inspection and Flaw Evaluation Guidelines" - BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines'! (replaced BWRVIP-Ol, -07, and -63) - BWRVIP-138, "Updated Jet PUlllP Beam Inspection and Flaw Evaluation" The attached Table (Table 1) con1pares present ASl'vfE Exmnination Category B-N-1 and B-N-2 requirenlents with the above current BWRVIP guideline requirements, as applicable, to BWR/2 through BWRl6 units. In addition, where guidance in existing BWRVIP documents has been sUpple111ented or revised by subsequent correspondence approved by the BWRVIP executive committee, the n10st current approved guidance will be i111plelnented. Therefore, the attached Table only represents a current conlparison. Any deviations fron1 the referenced BWRVIP Guidelines for the duration of the proposed alternative will be appropriately docUlnented and C0111111lmicated to the USNRC. per the BWRVIP Deviation Disposition Process. Current Exelon/AmerGen deviations froln the subject guidelines above are sumlnarized in Table 2. Inspection services, by an Authorized Inspection Agency, will Ix applied to the proposed alternative actions of this relief request. Alioll Science & TecllJlology PBT05.G03

lSI Program PIau Peach Bottom Atomic Power Statioll Units 2 and 3, Fourth Illterval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 5 of 17) Conditions: Guidelines BWRVIP-41, Revision 1, "BWR Jet Pwup Assenlbly Inspection and Flaw Evaluation Guidelines" and BWRVIP-138, "Updated Jet PU111P Beanl Inspection and Fla\\v Evaluation" had been under review by the USNRC staft~ but were withdrawn fron1 USNRC review on July 18, 2007, to allow for revision. The proposed alternatives based on topical reports BWRVIP-41, Revision 1, and BWRVIP-138, are, therefore, not authorized. As a result, PBAPS must continue to impleluent ASME Section Xl lSI requiren1ents for the jet pUlUp asselubly con1ponents for \\vhich this proposed alternative was requested. 6.0 BASIS FOR USE BWRs now exan1ine reactor internals in accordance with BWRVlP guidelines. These guidelines have been written to address the safety significant vessel internal con1ponents and to examine and evaluate the exaluination results for these cOluponents using appropriate n1ethods and reexamination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The USNRC has agreed with the BWRVIP approach in principal and has issued Safety Evaluations for these guidelines (see References 2 - 12 below). Therefore, use of these guidelines, as an alternative to the subject Code requiren1ents, provides an acceptable level of quality and safety and \\vill not adversely ilupact the health and safety of the public. As additional justification, Attachment 4 ("C0111parison of Code Exmuination Requiren1ents to B\\VRVIP Exanlination Requirements") provides specific exan1ples which con1pare the inspection requirements of ASME Code Item Nmubers B 13.1 0, BI3.20, BI3.30, and B13.40 in Table IWB-2500-1, to the inspection requirenlents in the BWRVIP docun1ents. Specific BWRVIP docUluents are provided as exaluples. This cOluparison also includes a discussion of the inspection luethods. These comparisons demonstrate that use of these guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely ilupact the health and safety of the public. 7.0 DURATION OF I~ROPOSEDALTERNATIVE The duration of the alternative is for the remainder of the interval specified above for each affected uni1. 8.0 PRECEDENCE A sinlilar relief reqL1est was approved for Vermont Yankee Nuclear Power Station as discussed in Reference 1. Alioll Sciellce & Technology PBT05.G03

lSI Progflllll Plan Peach Boltom Atomic Power Statioll Units 2 ami 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 6 of 17)

9.0 REFERENCES

1. Letter frOlll U. S. Nuclear Regulatory Conll11ission (USNRC) to Entergy Nuclear Operations, "Safety Evaluation of Relief Request RI-O 1, Vernl0nt Yankee Nuclear Power Station (TAC NO. MC0690)", dated Septetllber 19,2005 2. Letter USNRC to BWRVIP, dated April 27~ 1998~ "Final Supplenlent to the Safety Evaluation of the Boiling Water Reactor Vessel Internals Project BWRVIP-07 Report (TAC NO. M94959)" 3. Letter USNRC to BWRVIP, dated October 6~ 1999~ "Staff Reevaluation of Table 1 in the BWRVIP-07 Report (TAC NO. M94959)" 4. Letter USNRC to BWRVIP, dated September 6, 2005, "USNRC Approval Letter ofBWRVIP-18-A, "BWR Vessel and Internals Project Boiling Water Reactor Core Spray Internals Inspection and Flaw Evaluation Guideline" " 5. Letter USNRC to BWRVIP, dated September 9,2005, "USNRC Approval Letter of BWRVIP-26-A, "BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines" " 6. Letter USNRC to BWRVIP, dated July 24, 2000, "Final Safety Evaluation of the '"BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38)," EPRI Report TR-1 08823 (TAC NO. M99638)" 7. Letter USNRC to BWRVIP, dated February 4~ 2001, "Final Safety Evaluation of the "BWR Vessel and Internals Project, BWR Jet Pun1p Assembly Inspection and Flaw Evaluation Guidelines (BWRVIP-41 )~" (TAC NO. M99870)" 8. Letter USNRC to BWRVIP, dated September 9,2005, USNRC approval letter of BWRVIP.;.42A, "BWR Vessel and Internals Project Boiling Water Reactor Low Pressure Coolant Injection and Flaw Evaluation Guidelines" 9. Letter USNRC to BWRVIP, dated Septenlber 9,2005, "USNRC Approval Letter of BWRVIP-47-A, "BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines" " 1O. Letter USNRC to BWRVIP, dated July 25,2005, "USNRC Approval Letter of BWRVIP-48-A~"BWR Vessel and Internals Project Vessel ID Attachtnent Weld Inspection and Flaw Evaluation Guideline" " Alion Science & Technology PBT05.G03

lSI Program Plall Peach Botto/ll Atomic Power Station Units 2 ((lid 3, FOllrtlt Illterval 10CFR50.55a RELIEF REQUEST: I4R-49 Revision 1 (Page 7 of 17) 11. Letter USNRC to BWRVIP, dated August 20, 2001, "Final Safety Evaluation of the "BWR Vessel and Internals Project, Shroud Vertical Weld Inspection and Evaluation Guidelines (BWRVIP-63)," (TAC NO. MA6015)" 12. Letter USNRC to BWRVIP, dated June 10,2004, Propriety Version of USNRC Staff Review of BWRVIP-27-A, "BWR Standby Liquid Control System/Core Plate ~p Inspection and Flaw Evaluation Guidelines" Alioll Science & Technology PB T05. G03

/' ,.) I~;~~,~~ LiW'~~(;) ~------------------------~ lSI Program Plan Peach Bottom Atomic Power Statioll Ullits 2 alld 3, Fourth Interval 10CFRSO.SSa RELIEF REQUEST: I4R-49 Revision 1 (Page 8 of 17) TABLE 1 Comparison of ASME Examination Category B-N-l and B-N-2 Requirements With BWRVIP Guidance Requirements (I) ASME Item Applicable BWRVIP

Number, ASME Exam ASME ASME BWRVIP BWRVIP Table IWB-Component Scope Exam Frequency BWRVIP Exam Exam Frequency 2500-1 Document Scope B13.lO Reactor Vessel Interior Accessible VT-3 Each I BWRVIP-18-A, Overview examinations of components during Areas period 25, 26-A, 27-A, BWRVIP examinations satisfy Code VT-3 visual (Non-specific) 38, 42-A, 47-A, inspection requirements.

--c---' 48-A, and 76 813.20 Interior Attachments Accessible VT-l Each BWRVIP-48-A, Riser Brace EVT-l }00% in first 12 Within Beltline - Jet Welds 10-year Table 3-2 Attachment years (with 50% to Pump Riser Braces Interval be inspected in the first 6 years); 25% during each f------------- subsequent 6 years Lower Surveillance BWRVIP-48-A, Bracket VT-l Each 10-year Specimen Holder Table 3-2 Attachment Interval Brackets 1------- B13.30 Interior Attachments Accessible VT-3 Each BWRVIP-48-A, Bracket VT-3 Each 10-year Beyond Beltline - Welds 10-year Table 3-2 Attachment Interval Steam Dryer Hold-Interval down Brackets Guide Rod Brackets BWRVIP-48-A, Bracket VT-3 Each 10-year Table 3-2 Attachment Interval Steam Dryer Support BWRVIP-48-A, Bracket EVT-I Each 10-year Brackets Table 3-2 Attachment Interval Feedwater Sparger BWRVIP-48-A, Bracket EVT-I Each IO-year Brackets Table 3-2 Attachment Interval ~-------- ._~- Alion Science & Technology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Statio/l Units 2 and 3, Fourth Interval ~ 'i:~~\\'i~~;'i? ';';';4.- I!..- - = = ~ 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 9 of 17) TABLE 1 Comparison of ASME Examination Category B-N-l and B-N-2 Requirements With BWRVIP Guidance Requirements (1) ASME Item Applicable BWRVIP

Number, ASME Exam ASME ASME BWRVIP BWRVIP Table lWB-Component Scope Exam Frequency B\\VRVIP Exam Exam Frequency 2500-1 Document Scope Core Spray Piping BWRVIP-48-A, Bracket EVT-I Every 4 Refueling Brackets Table 3-2 Attachment Cycles Upper and Middle BWRVIP-48-A, Bracket VT-3 Each lO-year

. Surveillance Specimen Table 3-2 Attachment Interval Holder Brackets Shroud Support (Weld BWRVIP-38, Weld H9(2) EVT-lor Maximum of6 H9) including gussets 3.1.3.2, including UT years for one sided where applicable Figures 3-2 and gussets EVT-1, Maximum 3-5 (where of 10 years for UT f---. applicable) Shroud Support Legs (Rarely BWRVIP-38, Weld H12 Per When accessible (Weld H12) Accessible) 3.2.3 BWRVlP-38 USNRC SER (7-24-2000), inspect with appropriate method (4) B13.40 Integrally Welded Core Accessible VT-3 Each BWRVIP-38, Shroud EVT-l or Based on as found Support Structure - Surfaces lO-year 3.1.3.2, Support and UT conditions, to a Shroud Support Interval Figure 3-2 and Leg Welds maximum 6 years 3-5 including for one sided gussets as EVT-l, 10 years for applicable UT where accessible Aliol1 Sciel1ce & Technology PBT05.G03

lSI Program Plall Peach Bottom Atomic Power Station Units 2 anti 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: I4R-49 Revision 1 (Page 10 of 17) TABLE 1 Comparison of ASME Examination Category B-N-l and B-N-2 Requirements With BWRVIP Guidance Requirements (I) ASME Item Applicable BWRVIP

Number, ASME Exam ASME ASME BWRVIP BWRVIP Table IWB-Component Scope Exam Frequency BWRVIP Exam Exam frequency 2500-1 Document Scope Shroud Horizontal BWRVIP-76, Welds HI-EVT-Ior Based on as found Welds 2.2.1 H7 as UT conditions, to a applicable maximum 6 years for one sided EVT-l, 10 years for UT where accessible Shroud Vertical Welds BWRVIP-76, Vertical and EVT-l or Maximum 6 years 2.3, Ring UT for one-sided Figure 3-3 Segment EVT-I, 10 years for Welds as UT applicable Shroud Repairs (3)

BWRVIP-76, Tie-Rod VT-3 Per designer Section 3.5 Repair recommendations per BWRVIP-76 NOTES: I) This Table provides only an overview of the requirements For more details, refer to ASME Section Xl, Table IWB-2500-1, and the appropriate BWRVIP document. 2) In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined. 3) Shroud repairs are currently installed on both units at Dresden and Quad Cities, and on the single units at Oyster Creek and Clinton. 4) When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection ofthese welds. Alion Science & Technology PB T05. G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 and 3, Fourth Interval tOCFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 11 of 17) TABLE 2 I BWRVIP Deviations PLANT BWRVIP LETTER DATE TO USNRC DEVIATION APPLICABILITY DOCUMENT LaSalle County BWRVIP-76 Letter from S. R. Landahl 100% of the accessible areas of the LaSalle County Station This deviation does Station, Unit I (Exelon Generation Company, Unit I core shroud was not examined in the L1R11 outage not impact the basis LLC) to U. S. nuclear in February 2006. However, sufficient coverage ofthe for use of this relief Regulatory Commission, dated accessible areas were examined, an engineering evaluation request. April 27, 2006 was performed and the results indicate the Unit 1 shroud retained sufficient structural margin and should be re-examined in six years. Oyster Creek BWRVIP-76 Letter from T. S. Rausch Per BWRVIP-76 re-inspection guidelines, a one-sided This deviation does Generating Station (AmerGen Energy Company) to visual examination technique requires reexamination in 6 not impact the basis

u. S. Nuclear Regulatory years as compared to a nvo-sided visual or volumetric for use of this relief Commission, dated April 27, technique, which can achieve a lo-year reexamination request.

2007 frequency. Contrary to this requirement, the fOUf vertical welds were not reexamined in I R21 (2006) as required by the BWRVIP-76 guidance for single-sided examination. Alion Science & Technology PBT05.G03

lSI Progrtll11 Plan Peach Bottom Atomic Power Statioll Units 2 and 3, FOllrtl' Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 12 of 17) COlnparison Of Code Exalnin~ltionRequirements To BWRVIP Exanlination Requirenlcnts The following discussion provides a comparison of the exaillination requirements provided in ASME Code Item NUlnbers B13.10, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the exan1ination requirenlents in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This cOlnparison also includes a discussion of the exanlination methods. 1.0 Codc Rcguirclnent - B13.10 - Reactor Vessel Intcrior Accessible A.'eas (B-N-l) The ASME Section XI Code requires a VT-3 visual exanlination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during norn1al refueling outages. The frequency of these exanlinations is specified as the first refueling outage, and at intervals of approxiluately 3 years, during the first inspection interval, and each period during each successive 1O-year Inspection Interval. Typically, these examinations are performed every other refueling outage of the Inspection Interval. This exanlination requirement is a non-specific requirenlent that is a departure from the traditional ASME Section XI examinations of welds and surt~lces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the exmnination is to identify relevant conditions such as distortion or displacen1ent of parts; loose, n1issing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products; wear; and structural degradation. Portions of the various exanlinations required by the applicable BWRVIP Guidelines require access to accessible areas of the reactor vessel during each refueling outage. Exanlination of core spray piping and spargers (BWRVIP-18-A), top guide (BWRVIP-26-A), jet punlp welds and components (BWRVIP-41, Rev. 1), interior attaclunents (BWRVIP-48-A), core shroud welds (BWRVIP-76), shroud support (BWRVIP-38), LPCI couplings (BWRVIP-42-A), and lowerplenum conlponents (BWRVIP-47-A) provides such access. Locating and exanlining specific welds and cOluponents within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by renlote canlera systenls that essentially perform equivalent VT-3 visual eXaIllination of these areas or spaces as the specific weld or component exan1inations are perfonned. This provides an equivalent nlethod of visual examination on a 1110re frequent basis than that required by the ASME Section XI Code. Evidence of wear, structural degradation, loose. nlissing, or displaced parts, foreign 111aterials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP exanlination requirements. Therefore, the speci fled BWRVIP Guideline requirelnents lneet or exceed the subject Code requirenlents for Alioll Sciellce & Tecltnology PBT05.G03

lSI fJfogrtl11l Piau Peach Bottom Atomic Power Station Units 2 a/1{13, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 13 of 17) exan1ination method and frequency of the interior of the reactor vessel. Accordingly, these BWRVIP exmnination requiren1ents provide an acceptable level of quality and safety as compared to the subject Code requirenlents. 2.0 Code Rc<)uirelncnt - B13.20 - Interior Attaclunents Within the Beltline (B-N-2) The ASME Section XI Code requires a VT-1 visual eXaIllination of accessible reactor interior surface attachnlent \\velds within the beltline each la-year interval. In the boiling water reactor, this includes the jet ptullp riser brace welds-to-vessel wall and the lower surveillance specinlen support bracket welds-to-vessel wall. In comparison, the BWRVIP requires the same exanlination method and frequency for the lower surveillance specilnen support bracket welds, and requires an EVT-1 visual examination on the ren1aining attachment \\velds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years. The jet pUlnp riser brace examination requiren1ents are provided below to show a comparison between the Code and the BWRVIP eXaIllination requirements. Comparison to BWRVTP Requirements - Jet PU111P Riser Braces (BWRVIP-41, Rev. 1 and BWRVIP-48-A) The ASME Code requires a 1000/0 VI'-1 visual exmnination of the jet pump riser brace-to-reactor vessel wall pad welds each la-year interval. The B\\VRVIP requires an EVT-1 visual exanlination of the jet pump riser brace-to-reactor vessel wall pad welds the first 12 years and then 25% during each subsequent 6 years. BWRVIP-48-A specifically defines the susceptible regions of the attachnlent that are to be exmnined. The Code VI'-1 visual exanlination is conducted to detect discontinuities and in1perfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP enhanced VT-1 (EVT-l) visual exan1ination is conducted to detect discontinuities and ilnperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and inter-granular stress corrosion cracking (IGSCC), the relevant degradation l11echanisms for these con1ponents. General \\vear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel l11aterial, would also be detected during the process ofperforn1ing a BWRVIP EVT-1 visual exanlination. Alion Science &: Tecllllology PBT05.G03

lSI Progrtl/11 Plait Peacll Bott0111 Atomic Power Station Ullit!i' 2 a/1(/ 3, Fourtll Interval 10CFR50.55a RELIEF REQUEST: I4R-49 Revision 1 (Page 14 of 17) The Code VT-l visual eXaInination n1ethod requires (depending on applicable Edition) that at a n1axinlum distance of 2 feet, a 1/32" black line can be resolved or a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-l visual exan1ination method requires resolution of a 1/2 n1il (0.0005 inch) \\vire on the examination surface. The jet pun1p riser brace configuration for each plant varies \\vith vessel manufacturer (B&W, CB&I, etc.) and generation (BWR/3-BWR/6). BWRVIP-48-A includes diagran1s for each configuration and prescribes exanlination for each configuration. The calibration standards used for BWRVIP EVT-l visual exan1inations utilize the Code characters and the 0.0005" \\vire, thus assuring at least equivalent resolution compared to the Code. Although the BWRVIP exan1ination may be less frequent, it is a more comprehensive nlethod. Therefore, the enhanced flaw detection capability of an EVT-l visual examination, with a less frequent exalnination schedule provides an acceptable level of quality and safety to that provided by the ASME Code. 3.0 Code Requirement - B13.30 - Interior Attachment Bevond the Beltline Region (B-N-2) The ASME Section XI Code requires a VT-3 visual eXaInination of accessible reactor interior surface attachnlent welds beyond the beltline each 10-year interval. In the boiling water reactor, this includes the core spray piping prinlary and supplelnental support bracket welds-to-vessel wall, the upper surveillance specin1en support bracket welds-to-vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steanl dryer support and hold down bracket welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, the shroud support plate-to-vessel wall, and the shroud support gussets. BWRVIP-48-A requires as a nlinilnunl the SaIne VT-3 visual exanlination method as the Code for sonle of the interior attachlnent welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique EVT-1 for these welds. For those interior attachlnent welds that have the same VT-3 method of visual examination, the SaIne scope of examination (accessible welds), the SaIne exmnination frequency (each 10 year interval) and ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provide by the ASrv1E Code. For the core spray prinlary and secondary SUppOlt bracket attachnlent welds, the steanl dryer support bracket attachlnent welds, the feedwater sparger support bracket attachn1ent \\velds, and the shroud support plate-to-vessel welds, as applicable, the BWRVLP Guidelines require an EVT-l visual eXaIninatiol1 at the sanle frequency as the Code, or at a lnore frequent rate. Therefore, the BWRVIP requirements provide the saIne level of quality and safety to that provided by the ASME Code. Alion Science & Technology PBTrJS.G03

lSI Program Plan Peach Bottom Atomic Power Station Units 2 lIlId 3, Fourth Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 15 of 17) The core spray piping bracket-to-vessel attachn1ent weld is used as an exan1ple for cotnparison between the Code and BWRVIP exanlination requirelnents as discussed belo\\v. Comparison to BWRVIP Requirements - Core Spray piping Bracket Welds (BWRVIP-48-A) The Code exan1ination requirement is a VT-3 visual examination of each weld every 1a years. The BWRVIP visual examination requiretnent is an EVT-l for the core spray piping bracket attachtnent welds with each weld examined every four cycles (8 years for units with a two year fuel cycle). The BWRVIP visual examination method EVT-l has superior £1aw detection and sizing capability, the exanlination frequency is greater than the Code requirenlents, and the same £1aw evaluation criteria are used. The Code VT-3 visual exanlination is conducted to detect conlponent structural integrity by ensuring the cOlnponents general condition is acceptable. An enhanced EVT-I visual examination is conducted to detect discontinuities and inlperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or f~ltigue, the relevant degradation nlechanistns for BWR internal attachtnents. Therefore, with the EVT-1 visual exmnination nlethod, the satne exatnination scope (accessible welds), an increased exanlination frequency (8 years instead of 1a years) in some cases, the sanle f1aw evaluation criteria (ASME Section XI), the level of quality and safety provided by the BWRVIP criteria is superior than that provided by the Code. 4.0 Code Requirement - B13.40 - Integrally Welded Core Support Structures (B-N-2) The ASME Code requires a VT-3 visual exanlination of accessible surfaces of the welded core support structure each 1a-year interval. In the boiling water reactor, the welded core support structure has pritnarily been considered the shroud support structure, including the shroud support plate (annulus £1001'), the shroud support ring, the shroud sllpport welds, the shroud sllpport gussets, and the shroud support legs (if accessible). In later designs, the shroud itself is considered part of the welded core support structure. Historically, this requirenlent has been interpreted and satisfied differently across the industry. The proposed alternate exatnination replaces this ASME requirenlent with specific BWRVIP guidelines that exmnine susceptible locations for kno\\vn relevant degradation mechanisms. Alioll Science & TedlJ1%gy PIJT05.G03

lSI Program Plall Peaclt Bottom Atomic Power Station Units 2 ant! 3, Fourtlt Interval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 16 of 17) The Code requires a VT-3 visual exan1ination of accessible surfaces each 10-year interval. The BWRVIP requires as a nlininluln the same visual exanlination nlethod (VT-3) as the Code for integrally welded Core Support Structures, and for specific areas, requires either an enhanced visual examination technique (EVT-1) or volUllletric exan1ination (UT). BWRVIP recomnlended examinations of integrally welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. As a ll1iniInunl, the sanle or superior visual exa111ination technique is required for examination at the SaIne frequency as the code exan1ination requirements. In many locations, the BWRVIP guidelines require a volunletric exanlination of the susceptible welds at a frequency identical to the Code requirenlent. Where shroud repair tie-rods have been installed (Dresden, Quad Cities, Oyster Creek and Clinton), the BWRVIP referenced examinations are the same as the Code requirernents. Shroud repair tie-rod examinations are recommended in BWRVIP-76 and have the sanle basic VT-3 method of visual exanlination, the same scope of eXaInination (accessible surfaces), the same exanlination frequency (each 10 year interval) and the sanle flaw evaluation criteria. Therefore, the BWRVIP requirenlents provide a level of quality and safety equivalent to that provided by the ASME Code. Additionally, the repair vendor has provided site-specific eXaInination recornmendations to address the unique features of each repair. For other integrally welded core support structure components, the BWRVIP requires an EVT-1 visual eXaI11ination or UT of core support structures. The core shroud is used as an example for comparison between the Code and BWRVIP exanlination requirements as shown below. Comparison to BWRVIP Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline (BWRVIP-76) The Code requires a VT-3 visual exanlination of accessible surfaces every 10 years. The BWRVIP requires an EVT-1 visual examination frOI11 the inside and outside surface where accessible or ultrasonic examination of each core shroud circumferential weld that has not been structurally replaced with a shroud repair at a calculated end of interval" (EOI) that will vary depending upon the anlount of flaws present, but not to exceed ten years. ,,'!ion Science & Tecllllology PBT05.G03

lSI Program Plan Peach Bottom Atomic Power Statioll Units 2 and 3, Fourth I/lterval 10CFR50.55a RELIEF REQUEST: 14R-49 Revision 1 (Page 17 of 17) The BWRVIP recomnlended exmninations specify locations that are known to be vulnerable to 13WR relevant degradation tuechanisnls rather than "all surfaces". The BWRVIP exml1ination methods (EVT-l or UT) are superior to the Code required VT-3 visual examination for flaw detection and characterization. The BWRVIP exanlination frequency is equivalent to or more frequent than the exmnination frequency required by the Code. The superior flaw detection and characterization capability, with an equivalent or more frequent examination frequency and the c0111parible flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that provided by the Code requirenlents. Alio/l Scie/lce & Technology PIJT05.G03

lSI Pl'Ogrlll1l Plan Peach Bottom Atomic Power Station Ullits 2 & 3, Fourth Interval

9.0 REFERENCES

The references used to develop this Inservice Inspection Program Plan include: 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.1, "Prilnary Reactor Contaimnent Testing for Water Cooled Power Reactors". 2) ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Rules for Inservice Inspection of Nuclear Power Plant Con1ponents," 2001 Edition through the 2003 Addenda. (4th lSI and 2nd CISI Intervals) 3) ASME Boiler and Pressure Vessel Code, Section III, Division 1, "Rules For Construction of Nuclear Power Plant Con1ponents", 2001 Edition through the 2003 Addenda. 4) ASME OM Code, "Code for Operation and Maintenance of Nuc1ear Power Plants," 2001 Edition through the 2003 Addenda. 5) Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1". 6) Regulatory Guide 1.150, Rev. 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Exan1ination". 7) Regulatory Guide 1.192, "Operation and Maintenance Code Case Acceptability, ASME OM Code", 8) Regulatory Guide 1.193, "ASME Code Cases Not Approved For Use". 9) Peach Bottol11 Atomic Power Station Units 2 and 3, Updated Final Safety Analysis Report (UFSAR). 10) Peach Bott0l11 Aton1ic Power Station Units 2 and 3, Technical Specifications (TS). 11) Peach BottOln AtOlnic Power Station Units 2 and 3, Technical Requirements Manual (TRM). 12) USNRC NUREG-0737, dated November 1980, "TMI Action Plan Requireluents". 13) Procedures ER-AA-330, "Conduct of Inservice Inspection Activities", ER-AA-330-00 1, "Section XI Pressure Testing", ER-AA-330-002, "Inservice Inspection of Welds and COlnponents", ER-AA-330-003, "Visual Examination of Aliol1 Science & Technology 9-1 PBT05.G03 Revision 0

lSI Progra11l Plan Peach Bott011l Atomic Power Statioll Units 2 & 3, Fourth Illterval Section XI Con1ponent Supports", ER-AA-330-004, "Visual Exalnination of Technical Specification Snubbers", ER-AA-330-007, "Visual EXaInination of Section XI Class MC Surfaces and Class CC Liners", ER-AA-330-008, "Exelon Service Level 1 and Safety Related (Service Level 3) Protective Coatings", ER-AA-330-009, "ASME Section XI Repair / Replacement Program", ER-AA-330-010, "Snubber Functional Testing", ER-AA-330-0 11, "Snubber Service Life Monitoring PrograIn", ER-AA-335-004, "Manual Ultrasonic Measurement of Material Thickness", ER-AA-335-018, "Detailed, General, VT-l, VT-IC, VT-3, VT-3C, Visual Examination of ASME Class MC and CC Containment Surfaces and COlnponents", and ER-PB-331-1 00 1, "RPV & Internals Program Basis and Implementation Document". 14) Peach Bottom Atomic Power Station lSI Classification Basis Document (PBT05.G04), Fourth Ten-Year Inservice Inspection Interval. 15) Peach Bottoln Atomic Power Station lSI Selection Document (PBT05.G05), Fourth Ten-Year Inservice Inspection Interval. 16) Peach Bottoln Atomic Power Station Augn1ented Inspection Plan (PBT05.G06), Fourth Ten-Year Inservice Inspection Interval. 17) Exelon Risk-Informed Inservice Inspection Evaluation (Final Report) for PBAPS Units 2 and 3. 18) EPR! Topical Report TR-112657, Rev. B-A, Final Report, "Revised Risk-Informed Inservice Inspection Evaluation Procedure", December 1999. 19) USNRC SER related to EPRI Topical Report TR-112657, Rev. B, Final Report, "Revised Risk-Infonned Inservice Inspection Evaluation Procedure, July 1999", dated October 28, 1999. 20) Exelon Generation Company calculation ME-34, "Class 1 Exelnption Sizes". 21) Stone & Webster calculation PM-945," CRD Housing Weld Exclusion Evaluation". 22) Safety Evaluation Report fron1 USNRC to J. A. Hutton (PECO), "Request for Relief from Performing Augn1ented Inspections of the Circunlferential Reactor Vessel Shell Welds, Peach Bott01n At01nic Power Station, Units 2 and 3 (1'AC Nos. MA8195, MA8196)", June 15, 2000. (Under the Fourth Interval lSI Program this Third lSI Interval pennanent Relief Request (RR-41 ) has been renUlnbered as 14R-41.) 23) Safety Evaluation Report fron1 USNRC to C. G. Pardee (Exelon), "Clinton Power Station, Unit No.1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Lilnerick Generating Station, Units 1 and 2; Oyster AIiO/l Science & Techllology 9-2 PBT05.G03 Revisioll 0

lSI Progrtlf11 Plan Peach Bottom Atomic Power Station Ullits 2 & 3, FOllrth Interval Creek Nuclear Generating Station; Peach Bottonl Atomic Po\\ver Station, Units 2, and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief Request to use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (TAC Nos. MD5352 thru MD5363)", April 30, 2008. (Under the Fourth Interval lSI Program this Relief Request has been numbered as 14R-49.) 24) Safety Evaluation Report frol11 USNRC to C. G. Pardee (Exelon), "Peach Bottonl AtOlnic Power Station, Units 2 and 3 - Requests for Relief associated with the Third and Fourth Inservice Inspection Intervals and the First and Second Containnlent Inservice Intervals (TAC Nos. MD8294, MD8295, MD8296, MD8297, MD8298, MD8299, MD8300, MD8301, MD8302, MD8303, NID8304, MD8305, MD8306, MD8307, MD8308 AND MD8309)", February 26,2009. Alioll Science & Technology 9-3 PJJT05.G03 Revision 0}}