ML11269A018

ISI Program Plan Third Ten-Year Inspection Interval

ML11269A018
Person / Time
Site:	Clinton
Issue date:	10/12/2010
From:	Coleman S ALION Science & Technology Corp
To:	Office of Nuclear Reactor Regulation, Exelon Generation Co
References
CLN05.G03, Rev 14
Download: ML11269A018 (152)

Text

Program 151 Program Document No.: CLNO5.G03 Power Station ISI Clinton Power Document No.: CLN05.GO3 Exelkn..

Nuclear CLINTON POWER STATION UNIT 1 ISI PROGRAM PLAN THIRD TEN-YEAR INSPECTION INTERVAL Commercial Service Date:

Unit 1 - 04/24/87 Clinton Power Station RR 3, Box 228 Clinton, IL 61727 Exelon Generation Company (EGC), LLC 300 Exelon Way Kennett Square, PA 19348 Prepared By:

Alion Science and Technology Corporation Engineering Programs Division Warrenville, Illinois S

/ALION SCICNCE ANO TECHNOLOGY

151 Program Plan Clinton Power Station Unit]1, Third Interval REVISION APPROVAL SHEET TITLE: ISI Program Plan Third Ten-Year Inspection Interval Clinton Power Station, Unit 1 DOCUMENT NUMBER: CLN05.G03 REVISION: 14 PREPARED TRANSMITTAL PREPARED:

St V..en Co1 an Aliorn Project 'gineer REVIEWED:

Kevin M. Johns~n Alion. Project Engineer APPROVED:

Daniol W. LamoretM Alion Project Manager EXELON ACCEPTANCE APPROVED: / I( o1 Mirza Bo C i ISI Plogifam Coordinato'"r" Alion Science & Technology i CLN05. GO3 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval REVISION APPROVAL SHEET TITLE: ISI Program Plan Third Ten-Year Inspection Interval Clinton Power Station, Unit 1 DOCUMENT NUMBER: CLN05.G03 REVISION: 14 EXELON PREPARATION. REVIEW. AND APPROVAL REVIEWED: /10 . 2.010 Mirza a (

ISI Program Coordinator REVIEWED:

DaPrssrees Pressure "Iestad and Snub nb Program Owner REVIEWED: lo'-4-20i0o Thomas Parrent Containment (IWE/IWL) Program Owner APPROVED:

CONCURRED I:

Mike Haydon V Authorized Nuclear Inservice Inspector (ANII)

Each time this document is revised, the Revision Approval Sheet will be signed and the following Revision Control Sheet should be completed to provide a detailed record of the revision history. The signatures above apply only to the changes made in the revision noted. If historical signatures are required, Clinton Power Station archives should be retrieved.

Alion Science & Technology ii CLN05.G03 Revision 14

ISI ProgramPlan ClintonPower Station Unit 1, Third Interval REVISION CONTROL SHEET Major changes should be outlined within the table below. Minor editorial and formatting revisions are not required to be logged.

REVISION DATE I REVISION

SUMMARY

14 08/20/10 Initial issuance. (This ISI Program Plan was developed by Alion Science and Technology Corporation as part of the Third Interval ISI Program update.)

Prepared: S. Coleman Reviewed: K. Johnson Approved: D. Lamond Notes:

1. This ISI Program Plan (Sections 1 - 9 inclusive) is controlled by the Clinton Power Station Programs Engineering Group.

2. Revision 14 of this document was issued as the Third Interval ISI Program Plan and was submitted to the NRC. Future revisions of this document made within the Third ISI Interval will be maintained and controlled at the Clinton Power Station; however, they are not required to be and will not be submitted to the NRC. The exception to this is that new or revised Relief Requests shall be submitted to the NRC for safety evaluation and approval.

Alion Science & Technology iii CLNO5.G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval REVISION

SUMMARY

SECTION EFFECTIVE PAGES REVISION DATE Preface i to vii 14 08/20/10 1.0 1-1 to 1-18 14 08/20/10 2.0 2-1 to 2-30 14 08/20/10 3.0 3-1 to 3-2 14 08/20/10 4.0 4-1 to 4-2 14 08/20/10 5.0 5-1 to 5-2 14 08/20/10 6.0 6-1 to 6-3 14 08/20/10 7.0 7-1 to 7-24 14 08/20/10 8.0 8-1 to 8-3 14 08/20/10 9.0 9-1 to 9-4 14 08/20/10 Alion Science & Technology iv CLN05. GO3 Revision 14

JS! ProgramPlan Clinton Power Station Unit 1, ThirdInterval TABLE OF CONTENTS SECTION DESCRIPTION PAGE

1.0 INTRODUCTION AND BACKGROUND

..................................................................... 1-1 1.1 Introduction 1.2 Background 1.3 First Interval ISI Program 1.4 Second Interval ISI Program 1.5 Third Interval ISI Program 1.6 First Interval CISI Program 1.7 Second Interval CISI Program 1.8 Code of Federal Regulations 10CFR50.55a Requirements 1.9 Code Cases 1.10 Relief Requests 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM ................................................... 2-1 2.1 ASME Section XI Examination Requirements 2.2 Augmented Examination Requirements 2.3 System Classifications and P&ID Drawings 2.4 ISI Isometric and Component Drawings for Nonexempt ISI Class Components and Supports 2.5 Technical Approach and Positions 3.0 C O M PO N EN T ISI PLAN ................................................................................................ 3-1 3.1 Nonexempt ISI Class Components 3.2 Risk-Informed Examination Requirements 3.3 Reactor Coolant Pressure Boundary Normal Make-up Calculation 4.0 SU PPO R T ISI PL A N ....................................................................................................... 4-1 4.1 Nonexempt ISI Class Supports 4.2 Snubber Examination and Testing Requirements 5.0 SYSTEM PRESSURE TESTING ISI PLAN .................................................................. 5-1 5.1 ISI Class Systems 5.2 Risk-Informed Examination of Socket Welds 6.0 CON TA INM EN T ISI PLAN ........................................................................................... 6-1 6.1 Nonexempt CISI Class Components 6.2 Augmented Examination Areas 6.3 Component Accessibility 6.4 Responsible Individual and Engineer 7.0 COM PONENT SUM M ARY TABLES ........................................................................... 7-1 7.1 Inservice Inspection Summary Tables 7.2 Snubber Inspection Summary Tables Alion Science & Technology v CLNO5.G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE OF CONTENTS (Continued)

SECTION DESCRIPTION PAGE 8.0 RELIEF REQUESTS FROM ASME SECTION XI ........................................................ 8-1 9.0 RE FE RENC E S ................................................................................................................ 9-1 APPENDICES A. ISI PROGRAM PLAN REQUIREMENTS B. ISI PROGRAM PLAN COMPONENT AND PIPING EXAMINATION BOUNDARY Alion Science & Technologv vi CLN05. G03 Revision 14

1SI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE OF CONTENTS (Continued)

TABLES DESCRIPTION PAGE 1.1-1 THIRD ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3 COM PONENT EXAM INATION S) ................................................................................ 1-3 1.1-2 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISICLASS MC COM PONENT EXAM IN ATION S) ................................................................................ 1-4 1.1-3 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISI CLASS CC-CONCRETE COMPONENT EXAMINATIONS) .......................................................... 1-5 1.8-1 CODE OF FEDERAL REGULATIONS IOCFR50.55a REQUIREMENTS ................ 1-10 2.3-1 P& ID D RA WIN G S ........................................................................................................ 2-16 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS .................................................. 2-18 2.4-2 CISI REFEREN CE DRAWIN GS .................................................................................. 2-26 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX .............................................. 2-28 7.1-1 INSERVCICE INSPECTION

SUMMARY

....................................................................... 7-4 7.1-2 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES ..................... 7-19 7.2-1 SNUBBER INSPECTION

SUMMARY

TABLE .......................................................... 7-23 7.2-2 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES ....................... 7-24 8.0-1 RELIEF REQ U EST IND EX ............................................................................................ 8-2 Alion Science & Technology vii CLN05.G03 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval

1.0 INTRODUCTION AND BACKGROUND

1.1 Introduction This Inservice Inspection (ISI) Program Plan details the requirements for the examination and testing of ISI Class 1, 2, 3, MC, and CC pressure retaining components, supports, and containment structures at Clinton Power Station (CPS)

Unit 1. This ISI Program Plan also includes Containment Inservice Inspection (CISI), Risk-Informed Inservice Inspection (RISI), Augmented Inservice Inspections (AUG), Snubber Visual Examination and Functional Testing (SNUB),

and System Pressure Testing (SPT) requirements imposed on or committed to by CPS. At CPS, the Inservice Testing (IST) Program and the Containment Inservice Inspection (CISI) Program are maintained and implemented separately from the ISI Program. The IST Program Plan and the IWE/IWL Containment Inspection Plan contain all of the applicable CISI and IST program requirements. (See the IST Program Plan and the IWE/IWL Containment Inspection Plan for more details.)

The CPS Third ISI Interval is effective from July 1, 2010 through June 30, 2020 for Class 1, 2, and 3 components, including their supports. The CPS Second CISI Program is effective from September 9, 2008 through September 8, 2018 for Class MC and CC components. The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Code of Record for the Third ISI Interval is the 2004 Edition, No Addenda and the ASME Section XI Code of Record for the Second CISI Interval is the 2001 Edition through the 2003 Addenda. This ISI Program Plan is controlled and revised in accordance with the requirements of procedure ER-AA-330, "Conduct of Inservice Inspection Activities," which implements the ASME Section XI ISI Program.

Note that with the update of the ISI Program for the Third ISI Interval, Exelon Generation Company, LLC (Exelon) has not elected to synchronize intervals with the CISI Program for the Second CISI Interval to share a common interval start and end date. Also, this update will not enable the ISI and CISI Program components / elements to be based on the same effective Edition and Addenda of ASME Section XI as noted above.

Paragraph IWA-2430(d)(1) of ASME Section XI allows an inspection interval to be extended or decreased by as much as one year, and Paragraph IWA-2430(e) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The extension may be taken for a period of time not to exceed the duration of the outage. See Tables 1.1-1, 1.1-2, and 1.1-3 for intervals, periods, and extensions that apply to CPS's Third ISI Interval and Second CISI Interval.

The Third ISI Interval and Second CISI Interval are divided into number of inspection periods as determined by calendar years within the intervals. Tables Alion Science & Technology 1-1 CLNOS.G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 1.1-1, 1.1-2, and 1.1-3 identify the period start and end dates for the Third ISI Interval and the Second CISI Interval as defined by Inspection Program 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 inspections to coincide with CPS's refueling outages.

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1SI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 1.1-1 THIRD ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3 COMPONENT EXAMINATIONS)

Interval Period Outages Start Date to Start Date to End Date Projected Outage Outage End Date' Start Date or Number Outage Duration Scheduled C1R13 Ist 1/12 3 rd 07/1/10 to 06/30/13 07/1/10 to 06/30/20 Scheduled C1R14 2 nd 1/14 07/1/13 to 06/30/17 Scheduled CIR15 1/16 Scheduled C1R16 3 rd 1/18 07/1/17 to 06/30/20 Scheduled CI R 17 1/20 Note 1: The end of the CPS Second ISI Interval was initially extended for one year from January 1, 2010 to December 31, 2010, but then a decision was made to reduce the extension to six months instead of one year from January 1, 2010 toJune 30, 2010 per Paragraph IWA-2430(d) of ASME Section XI. (See Section 1.4 for details). As permitted by Paragraph IWA-2430(d), the interval was extended accordingly, thus affecting the start date of the Third ISI Interval.

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ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 1.1-2 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISI CLASS MC COMPONENT EXAMINATIONS)

Interval Period Outages Start Date to Start Date to End Date Projected Outage Outage End Date Start Date or Number Outage Duration Scheduled C1R12 1st 1/10 2 nd 08/14/08 to 08/13/11 08/14/08 to 08/13/18 Scheduled C1R13 2 nd 1/12 08/14/11 to 08/13/15 Scheduled CIR14 1/14 Scheduled CIR15 3 rd 1/16 08/14/15 to 08/13/18 Scheduled CIR16 1/18 Alion Science & Technology 1-4 CLN05.G03 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 1.1-3 SECOND ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS CC-CONCRETE COMPONENT EXAMINATIONS)

Interval 5-Year Period Outages Start Date to Rolling Exam # - Date Projected Outage Outage End Date (2 Year Window) Start Date or Number Outage Duration 3P - 10/14/10 Scheduled C1R12 1/10 2 nd Scheduled C1R13 09/09/08 to 09/08/18 1/12 Scheduled C1R14 1/14 Scheduled CIR15 I A --1 1 1/16 Scheduled C1R16 1/18 Note 1: The CISI Interval for Class CC components parallels the CISI Interval for Class MC components. The actual inspection schedule however is based on a rolling 5 year frequency (+/- 1 year) from the date of completion of the original examinations (10/14/00) performed during the initial September 9, 1996 - September 8, 2001 Rulemaking implementation period. The rolling 5 year inspection schedule for containment concrete is in accordance with the Inservice Inspection Schedule specified in Subarticle IWL-2400. (Note that CISI concrete examinations were scheduled prior to the Structural Integrity Test (SIT) and then every 5 years thereafter. The SIT was performed in refueling outage C 1R09 (February, 2004);

therefore, all refueling outage C 1R07 (October, 2000) examinations were performed prior to the SIT.)

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ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval 1.2 Background The Illinois Power Company, now known commercially as Exelon Generation Company or Exelon, obtained Construction Permit to build CPS on February 24, 1976. The Docket Number assigned to CPS is 50-461. After satisfactory plant construction and preoperational testing was completed, CPS was granted a full power operating license on April 17, 1987, NPF-62, and subsequently commenced commercial operation on April 24, 1987.

CPS's piping systems and associated components were designed and fabricated to be inspected and tested in accordance with the requirements of ASME Section XI.

Although this plant was specifically designed to meet the inspection and testing requirements of ASME Section XI, literal compliance may not be feasible or practical within the limits of the current plant design. Certain limitations are likely to occur due to conditions such as accessibility, geometric configuration, and/or metallurgical characteristics. For some inspection categories, an alternate component may be selected for examination and the code statistical and distribution requirements can still be maintained. If ASME Section XI required examination criteria cannot be met, a relief request will be submitted in accoidance with 10CFR50.55a.

1.3 First Interval ISI Program Pursuant to the Code Of Federal Regulations, Title 10, Part 50, Section 55a, Codes and standards,(1 OCFR50.55a), CPS was required to meet the requirements of Paragraph (g), Inservice inspection requirements,of that section.

Specifically, Paragraph 10CFR50.55a(g)(4)(i) calls for the inservice inspection requirements of the 120 month inspection interval to comply with the requirements of the latest Edition and Addenda of ASME Section XI referenced in Paragraph (b) of 10CFR50.55a on the date twelve months prior the date of issuance of the operating license, subject to the limitations and modifications listed in 10CFR50.55a(b).

CPS started commercial operation on April 24, 1987, which marked the beginning of the First ISI Interval. The version of 10CFR50.55a in effect twelve months prior to this date referenced the 1980 Edition with Addenda through the Winter 1980 (80W80) of ASME Section XI. The inservice inspection requirements applicable to nondestructive examination and system pressure testing for the First Inservice Inspection Program were based on these rules.

The CPS First ISI Interval was originally effective from April 24, 1987 to April 23, 1997, but CPS was shut down from September 1996 through May 1999.

ASME Section XI permitted 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 more may Alion Science & Technology 1-6 CLN05. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval extend the ISI Interval for an equivalent period. As such, the First ISI Interval was extended to December 31, 1999 based on plant out of service and/or to coincide with the scheduling or completion of an outage.

Therefore, the CPS First ISI Interval was effective from April 24, 1987 through December 31, 1999.

1.4 Second Interval ISI Program Pursuant to 10CFR50.55a(g), CPS was required to update the ISI Program at the end of the First ISI Interval. The ISI Program was required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a twelve months prior to the start of the Second ISI Interval per 10CFR50.55a(g)(4)(ii).

The CPS Second ISI Interval was developed in accordance with the requirements of the 1989 Edition, No Addenda of ASME Section XI. The CPS Second Interval ISI Program Plan addressed Subsections IWA, IWB, IWC, 1WD, IWF, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SER's, and utilized Inspection Program B as defined therein.

Starting in February 2004, in the Second Period of the Second ISI Interval, CPS converted from an 18 month fuel cycle to a 24 month fuel cycle (2 year outage cycle). A study was performed to determine the effect this change to a 24 month fuel cycle would have on the current Second Interval ISI Program (e.g.,

documents, schedule, crediting, etc.), and where needed, what changes to the program needed to be made to alleviate any Code compliance issues presented by the change in fuel cycle.

The Second ISI Interval was originally effective from January 1, 2000 through December 31, 2009. However, an extension was taken per Paragraph IWA-2430(d) of ASME Section XI, which allows an inspection interval to be extended or decreased by as much as one year. As permitted by this allowance, the CPS Second ISI Interval was initially extended by one year from January 1, 2010 to December 31, 2010, but then a decision was made to reduce the extension to six months instead of one year from January 1, 2010 to June 30, 2010. In affect, the Third ISI Interval will start following completion of the Second ISI Interval on July 1, 2010 and will end on June 30, 2020.

Therefore, the CPS Second ISI Interval was effective from January 1, 2000 through June 30, 2010.

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ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 1.5 Third Interval ISI Program Pursuant to 10CFR50.55a(g), licensees are required to update their ISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI Program is required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a twelve months prior to the start of the interval per 10CFR50.55a(g)(4)(ii). As discussed in Section 1.4 above, the start of the Third ISI Interval for CPS will be on July 1, 2010. Based on this date, the latest Edition and Addenda of ASME Section XI referenced in 10CFR50.55a(b)(2) twelve months prior to the start of the Third ISI Interval was the 2004 Edition, No Addenda.

The CPS Third Interval ISI Program Plan was developed in accordance with the requirements of 10CFR50.55a including all published changes , and the 2004 Edition, No Addenda of ASME Section XI, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.8-1 of this section. This Third Interval ISI Program Plan addresses Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices, approved ASME Code Cases, approved alternatives through relief requests and SER's, and utilizes Inspection Program B as defined therein.

The CPS Third ISI Interval is effective from July 1, 2010 through June 30, 2020.

CPS adopted the EPRI Topical Report TR- 112657, Rev. B-A methodology, which was supplemented by Code Case N-578-1, for implementing risk-informed inservice inspections during the Second ISI Interval. The RISI Program will continue for the Third ISI Interval. Implementation of the RISI Program is in accordance with Relief Request 13R-01.

CPS also adopted the EPRI Topical Report TR-1006937, Rev. 0-A, methodology for additional guidance for adaptation of the RISI evaluation process to Break Exclusion Region (BER) piping, also referred to as the High Energy Line Break (HELB) region. This change to the BER program was made under 10CFR50.59 evaluation criteria. The RISI evaluation for BER piping remains in effect for the Third ISI Interval.

1.6 First Interval CISI Program CISI examinations were originally invoked by amended regulations contained within a Final Rule issued by the United States Nuclear Regulatory Commission (NRC). The amended regulation incorporated the requirements of the 1992 Edition through the 1992 Addenda of ASME Section XI, Subsections IWE and IWL, subject to specific modifications that were included in Paragraphs 10CFR50.55a(b)(2)(ix) and 10CFR50.55a(b)(2)(x).

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ISI ProgramPlan Clinton Power Station Unit 1, Third Interval The final rulemaking was published in the Federal Register on August 8, 1996 and specified an effective date of September 9, 1996. Implementation of the Subsection IWE and IWL Program from a scheduling standpoint was driven by the five year expedited implementation period per 10CFR50.55a(g)(6)(ii)(B),

which specified that the examinations required to be completed by the end of the First Period of the First CISI Interval (per Table IWE-2412-1) be completed by the effective date (by September 9, 2001).

ASME Section XI Subsections IWE, IWL, approved ASME IWE/IWL Code Cases, and approved alternatives through related relief requests and SER's were added to the ISI Program near the end of the First ISI Interval to address CISI.

The CPS First CISI Interval was effective from September 9, 1996 through September 8, 2008.

1.7 Second Interval CISI Program Pursuant to 10CFR50.55a(g), licensees are required to update their CISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The CISI Program is required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in IOCFR50.55a twelve months pricr to the start of the interval per 10CFR50.55a(g)(4)(ii). Based on this date, the latest Edition and Addenda of ASME Section XI referenced in IOCFR50.55a(b)(2) twelve months prior to the start of the Second CISI Interval was the 2001 Edition through the 2003 Addenda.

The CPS Second Interval ClSI Program Plan was developed in accordance with the requirements of 10CFR50.55a including all published changes through September 30, 2006, and the 2001 Edition through the 2003 Addenda of ASME Section XI, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.8-1 of this section. This Second Interval CISI Program Plan addresses Subsections IWE, IWL, approved ASME IWE/IWL Code Cases, approved alternatives through related relief requests and SER's, and utilizes Inspection Program B as defined therein.

The CPS Second CISI Interval is effective from September 9, 2008 through September 8, 2018.

1.8 Code of Federal Regulations 10CFR50.55a Requirements There are certain Paragraphs in 10CFR50.55a that list the limitations, modifications, and/or clarifications to the implementation requirements of ASME Section XI. These Paragraphs in IOCFR50.55a that are applicable to CPS are detailed in Table 1.8-1.

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ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(viii)(E) (CISI) Examination of concrete containments: For Class CC 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 ISI Summary Report 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)(viii)(F) (CISI) Examination of concrete containments. Personnel that examine containment concrete surfaces and tendon hardware, wires, or strands must meet the qualification provisions in IWA-2300. The "owner-defined" personnel qualification provisions in IWL-23 10(d) are not approved for use.

10CFR50.55a(b)(2)(ix)(A) (CISI) Examinationof metal 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 ISI Summary Report as required by IWA-6000:

(1) A description of the type and estimated extent of degradation, and the conditions that led to the degradation; (2) An evaluation of each area, and the result of the evaluation, and; 1 (3) A description of necessary corrective actions.

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ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications I0CFR50.55a(1,)(2)(ix)(B) (CISI) Examination of metal containments and the liners of concrete containments: When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

10CFR50.55a(b)(2)(ix)(F) (CISI) Examination of metal containments and the liners of concrete containments. VT-I and VT-3 examinations must be conducted in accordance with IWA-2200. Personnel conducting examinations in accordance with the VT-I or VT-3 examination method shall be qualified in accordance with IWA-2300. The "owner-defined" personnel qualification provisions in IWE-2330(a) for personnel that conduct VT-1 and VT-3 examinations are not approved for use.

10CFR50.55a(b)(2)(ix)(G) (CISI) Examination of metal containments and the liners of concrete containments: The VT-3 examination method must be used to conduct the examinations in Items El. 12 and El1.20 of Table IWE-2500-1, and the VT- 1 examination method must be used to conduct the examination in Item E4. 11 of Table IWE-2500-1. An examination of the pressure-retaining bolted connections in Item El. 11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The "owner-defined" visual examination provisions in IWE-23 10(a) are not approved for use for VT-I and VT-3 I examinations.

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ISI ProgramPlan Clinton PowerStation Unit 1, ThirdInterval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS I OCFR50.55a Paragraphs Limitations, Modifications, and Clarifications IOCFR50.55a(b)(2)(ix)(H) (CISI) Examination of metal containments and the liners of concrete containments: Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item El .11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in accordance with the VT- 1 examination method. The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item El. 11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.

IOCFR50.55a(b)(2)(ix)(I) (CISI) Examination of metal containments and the liners of concrete containments: The ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must also be applied to metallic liners of Class CC pressure-retaining components.

I OCFR50.55a(b)(2)(xii) (ISI) Underwater Welding: The provisions in IWA-4660, "Underwater Welding," of Section XI, 1997 Addenda through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section, are not approved for use on irradiated material.

10CFR50.55a(b)(2)(xviii)(A) (ISI) Certificationof NDEpersonnel: Level I and II nondestructive examination personnel 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.

1 OCFR50.55a(b)(2)(xviii)(B) (ISI) Certificationof NDE 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-52 11 (a) and (b), 1998 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section.

Alion Scien.ce & Technology 1-12 CLN05. GO3 Revision 14

IS1 ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(xviii)(C) (ISI) Certificationof NDEpersonnel: When qualifying visual examination personnel for VT-3 visual examinations 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 must be demonstrated by administering an initial qualification examination and administering subsequent examinations on a 3-year interval.

10CFR50.55a(b)(2)(xix) (ISI) Substitution of alternative methods: The provisions for the substitution of alternative examination methods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 must be applied. The provisions in IWA-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) of this section, allowing the substitution of alternative examination methods, a combination of methods, or newly developed techniques for the methods specified in the Construction Code are not approved for use.

10CFR50.55a(b)(2)(xx)(B) (ISI) System leakage tests: The NDE provision in IWA-4540(a)(2) of the 2002 Addenda of Section XI must be applied when performing system leakage tests after repair and replacement activities performed by welding or brazing on a pressure retaining boundary using the 2003 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section.

10CFR50.55a(b)(2)(xxi)(B) (ISI) Table IWB-2500-1 examination requirements.- The provisions of Table IWB-2500-1, Examination Category B-G-2, Item 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.

IOCFR50.55a(b)(2)(xxii) (ISI) 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 examination method is prohibited.

Alion Science & Techuology 1-13 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(xxiii) (ISI) Evaluation of Thermally Cut Surfaces: The use of the provisions for eliminating mechanical processing of thermally cut surfaces in IWA-4461.4.2 of Section 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) (PDI) Incorporationof the PerformanceDemonstration Initiative andAddition of UltrasonicExamination Criteria.

The use of Appendix VIII and the supplements to Appendix VIII and Article 1-3000 of Section XI of the ASME BPV Code, 2002 Addenda through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section, is prohibited.

IOCFR50.55a(b)(2)(xxv) (ISI) Mitigation of Defects by Modification: The use of the provisions in IWA-4340, "Mitigation of Defects by Modification,"Section XI, 2001 Edition through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section are prohibited.

10CFR50.55a(b)(2)(xxvi) (SPT) PressureTesting Class 1, 2, and 3 MechanicalJoints:

The repair and replacement activity provisions in IWA-4540(c) of the 1998 Edition of Section XI for pressure testing Class 1, 2, and 3 mechanical joints must be applied when using the 2001 Edition through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section.

10CFR50.55a(b)(2)(xxvii) (ISI) Removal of Insulation: When performing visual examinations 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 temperature below 1100 'F or having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per square inch or higher.

Alion Science & Technology 1-14 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS

).55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(3)(v) (ISI) Subsection ISTD: Article IWF-5000, "Inservice Inspection Requirements for. Snubbers," of the ASME BPV Code,Section XI, provides inservice inspection requirements for examinations and tests of snubbers at nuclear power plants. Licensees may use Subsection ISTD, "Inservice Testing of Dynamic 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 requirements for snubbers in Section XI, IWF-5200(a) and (b) and IWF-5300(a) and (b), by making appropriate changes to their technical specifications or licensee-controlled documents. Preservice and inservice examinations must be performed using the VT-3 visual examination method described in IWA-2213.

IOCFR50.55a(b)(5) (ISI) Inservice 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 NRC 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-month interval, the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use.

(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, may continue to be applied by that licensee to the end of the 120-month interval in which the Code Case I was implemented.

Alion Science & Technologv 1-15 CLN05. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(6) (ISl) Operation andMaintenance of 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 NRC 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-month interval, the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use.

(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 may continue to apply the Code Case to the end of the current 120-month interval.

Alion Science & Techaology 1-16 CLN05. G03 Revision 14

IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 1.9 Code Cases Per IOCFR50.55a(b)(5) and (b)(6), ASME Code Cases that have been determined to be suitable for use in ISI Program Plans by the NRC are listed in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI, Division I". The approved Code Cases in Regulatory Guide 1.147, which are being utilized by CPS, 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 time it is applied within the ISI Program 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 CPS per Paragraph 1WA-2441(d) and proposed for use in revisions to the ISI Program Plan.

The use of other Code Cases (than those listed in Regulatory Guide 1.147) may be authorized by the Director of the office of Nuclear Reactor Regulation upon request pursuant to 10CFR50.55a(a)(3). Code Cases not approved for use in Regulatory Guide 1.147, which are being utilized by CPS through associated relief requests, are included in Section 8.0.

This ISI Program 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 CPS, are included in Section 2.1.2. The latest version of Regulatory Guide 1.192 incorporated into this document is Revision 0. As this guide is revised, newly approved Code Cases should be assessed for plan implementation at CPS per Paragraph IWA-2441 (d) and proposed for use in revisions to the ISI Program Plan.

1.10 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 requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, or determines that specific ASME Section XI requirements for inservice inspection are impractical, the licensee shall notify the NRC and submit information to support the determination.

The submittal of this information will be referred to in this document as a "Relief Request". Relief Requests for the Third ISI Interval and the Second CISI Interval are included in Section 8.0 of this document. The text of the Relief Requests contained in Section 8.0 will demonstrate one of the following: the proposed alternatives provide an acceptable level of quality and safety per 10CFR50.55a(a)(3)(i), compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of Alion Science & Te hnuolopi' 0*

1-17 CLN05.GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval qual ity and safety per 10CFR50.55a(a)(3)(ii), or the code requirements are considered impractical per 10CFR50.55a(g)(5)(iii).

Per 10CFR5O.55a Paragraphs (a)(3) and (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate relief requests and "may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility".

Alion Science & Technology 1-18 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2.1 ASME Section XI Examination Requirements As required by the 10CFR50.55a, this Program was developed in accordance with the requirements detailed in the 2004 Edition, No Addenda, of the ASME Boiler and Pressure Vessel Code,Section XI, Division 1, Subsections IWA, IWB, IWC, IWD, IWE, IWF, IWL, Mandatory Appendices, Inspection Program B of Paragraph IWA-2432, approved ASME Code Cases, and approved alternatives through relief requests and Safety Evaluation Reports (SER's).

Performance Demonstration Initiatives (PDI) is an organization comprised of all US nuclear utilities that was formed to provide an efficient implementation of Appendix VIII performance demonstration requirements. The Electric Power Research InstituWe (EPRI) NDE Center was selected as the administrator of this program. The PD1 program is administered according to the "PDI Program Description". The ISI Pi.-g.'am implements Appendix VIII "Performance Demonstration for Ultrasonic Examination Systems," ASME Section XI 2001 Edition, No Addenda as required by 1 CFR50.55a(b)(2)(xxiv) and with modifications as identified in OCI'R50.55a(b)(2)(xiv), (xv), and (xvi). 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) will be required to have a written program to ensure compliance with the requirements. These requirements were initially implemented through the Performance Demonstration Initiative (PDI) Program according to the schedule defined in 10CFR50.55a(g)(6)(ii)(C). CPS does not have in-house capabilities to perform ultrasonic examinations and intends to utilize NDE contractors to perform ultrasonic examinations. However, CPS still has the responsibility to ensure that Appendix VIII requirements are properly implemented.

For the Third ISI Interval, CPS's inspection program for ASME Section XI Examination Categories B-F, B-J, C-F-I, and C-F-2 will be governed by risk-informed regulations. The RISI Program methodology is described in the EPRI Topical Report TR-1 12657, Rev. B-A. To supplement the EPRI Topical Report, Code Case N-578-1 (as applicable per Relief Request 13R-01) is also being used for the classification of piping structural elements under the RISI Program. The RISI Program scope has been implemented as an alternative to the 2004 Edition, No Addenda of the ASME Section XI examination program for ISI Class 1 B-F and B-J welds and ISI Class 2 C-F-1 and C-F-2 welds in accordance with IOCFR50.55a(a)(3)(i). The basis for the resulting Risk Categorizations of the nonexempt ISI Class I and 2 piping systems at CPS is defined and maintained in the Final Report "Risk Informed Inservice Inspection Evaluation" as referenced in Section 9.0 of this document.

For the Third ISI Interval, the RISI Program scope continues to include welds in the BER piping, also referred to as the HELB region. The BER program methodology Alion Science & Technologv

.......................... Oj 2-1 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval is described in EPRI Topical Report TR-1006937, Rev. 0-A, which has been used to define the inspection scope in lieu of the 100% volumetric examination of all piping welds in the previous BER augmented program. Therefore, all welds in the original augmented program for BER remain evaluated under the RISI program using an integrated risk-informed approach.

The CISI Program per Subsection IWE and IWL is included in Section 6.0, "Containment ISI Plan". The CISI relief requests are included in Section 8.0 of this doctument.

2.1.1 ASME Section XI Code Cases As referenced by 10CFR50.55a(b)(5) and allowed by NRC Regulatory Guide 1.147, Revision 15, the following Code Cases are being incorporated into the CPS ISI Program. Several of these Code Cases are included as contingencies, to ensure that they are available for future repair/replacement activities.

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

Regulatory Guide 1.147, Revision 15.

N-460 Alternative Examination Coverage for Class 1 and Class 2 Welds. Regulatory Guide 1.147, Revision 15.

N-504-3 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, Nonmandatory Appendix Q, "Weld Overlay Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping Weldments," must also be met.

N-516-3 Underwater Welding Code Case N-516-3 is acceptable subject to the following condition specified in Regulatory Guide 1.147, Revision 15:

Licensee must obtain NRC approval in accordance with 10CFR50.55a(a)(3) regarding the technique to be used in the weld repair or replacement of irradiated material underwater.

Alion Sciepe.-tý e- ic:.. . 2-2 CLNO5.G03 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval N-517-1 Quality Assurance Program Requirements for Owners.

Regulatory Guide 1.147, Revision 15.

N-526 Alternative Requirements for Successive Inspections of Class I and 2 Vessels. Regulatory Guide 1.147, Revision 15.

N-528-1 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 requirements of 10CFR Part 21, "Reporting of Defects and Noncompliance", are to be applied to the nuclear plant site supplying the material as well as to the nuclear plant site receiving the material that has been purchased, exchanged, or transferred between sites.

N-532-4 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000. Regulatory Guide 1.147, Revision 15.

N-552 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 September 22, 1999 (64 FR 51370),

incorporating Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," to ASME Section XI, add the following to the specimen requirements:

"At least 50 percent of the flaws in the demonstration test set must be cracks and the maximum misorientation must be demonstrated with cracks. Flaws in nozzles with bore diameters equal to or less than 4 inches may be notches."

Alion Science & Technology 2-3 CLN05.GO3 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval Add to detection criteria, "The number of false calls must not exceed three."

N-566-2 Corrective Action for Leakage Identified at Bolted Connections. Regulatory Guide 1.147, Revision 15.

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

Regulatory Guide 1.147, Revision 15.

N-597-2 Requirements for Analytical Evaluation of Pipe Wall Thinning Code Case N-597-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15:

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

(2) Components affected by flow-accelerated corrosion to which this Code Case are applied must be repaired or replaced in accordance with the construction code of record and Owner's requirements or a later NRC approved Edition of Section III, "Rules for Construction of Nuclear Plant Components," of the ASME Code prior to the value of tp reaching the allowable minimum wall thickness, tri,, as specified in -3622.1 (a)(1) of this Code Case. Alternatively, use of the Code Case is subject to NRC review and approval per 10CFR50.55a(a)(3).

(3) For Class 1 piping not meeting the criteria of -3221, the use of evaluation methods and criteria is subject to NRC review and approval per 10CFR50.55a(a)(3).

(4) For those components that do not require immediate repair or replacement, the rate of wall thickness loss is to be used to determine a suitable inspection Alion Science & Technwlogy 2-4 CLN05. G03 Revision 14

IS! ProgramPlan Clinton Power Station Unit 1, Third Interval frequency so that repair or replacement occurs prior to reaching allowable minimum wall thickness, t,,,n.

(5) For corrosion phenomenon other than flow accelerated corrosion, use of the Code Case is subject to NRC review and approval. Inspection plans and wall thinning rates may be difficult to justify for certain degradation mechanisms such as MIC and pitting per 10CFR50.55a(a)(3).

N-600 Transfer of Welder, Welding Operator, Brazer, and Brazing Operator Qualifications Between Owners. Regulatory Guide 1.147, Revision 15.

N-606-1 Similar and Dissimilar Metal Welding Using Ambient Temperature 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 examination or verification must 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 from contaminants' such as, rust, moisture, grease, and other foreign material 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.

N-613-1 Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No's. B3.10 and B3.90, Reactor Nozzle-to-Vessel Welds, Figs.

IWB-2500-7(a), (b), and (c). Regulatory Guide 1.147, Revision 15.

N-624 Successive Inspections. Regulatory Guide 1.147, Revision 15.

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

Regulatory Guide 1.147, Revision 15.

Alion Science & Technology 2-5 CLN05. GO3 Revision 14

IS ProgramPlan ClintonPower Station Unit 1, Third Interval N-63 8-1 Similar and Dissimilar Metal Welding Using Ambient 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:

UT examinations shall be performed with personnel and procedures qualified for the repaired volume and qualified by demonstration using representative samples which contain construction type flaws. The acceptance criteria of NB-5330 in the 1998 Edition through the 2000 Addenda of Section III Edition and Addenda approved in 10CFR50.55a apply to all flaws identified within the repaired volume.

N-639 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 from the materials specification if (1) it is within the chemical range of the component specification to be inspected, and (2) the phase and grain shape are maintained in the same ranges produced by the thermal process required by the material specification.

N-641 Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements. Regulatory Guide 1.147, Revision 15.

N-649 Alternative Requirements for IWE-5240 Visual Examination. Regulatory Guide 1.147, Revision 15.

N-651 Ferritic and Dissimilar Metal Welding Using SMAW Temper Bead Technique Without Removing the Weld Bead Crown of the First Layer. Regulatory Guide 1.147, Revision 15.

N-661 Alternative Requirements for Wall Thickness Restoration of Classes 2 and 3 Carbon Steel Piping for Raw Water Service Alion Science & Technology 2-6 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Code Case N-661 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 15:

(a) If the root cause of the degradation has not been determined, the repair is only acceptable for one cycle.

(b) Weld overlay repair of an area can only be performed once in the same location.

(c) When through-wall repairs are made by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage.

N-665 Alternative Requirements for Beam Angle Measurements Using Refracted Longitudinal Wave Search Units.

Regulatory Guide 1.147, Revision 15.

N-686 Alternative Requirements for Visual Examinations, VT-.1, VT-2, and VT-3. 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 or 10CFR50.55a at that time.

2.1.2 ASME OM Code Cases As referenced by 10CFR50.55a(b)(6) and allowed by NRC Regulatory Guide 1.192, Revision 0, the following Code Cases are being incorporated into the CPS ISI Program:

OMN-13, Rev. 0 Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants, OM Code.

Additional Code Cases invoked in the future shall be in accordance with those approved for use in the latest published revision of Regulatory Guide 1.192 or 10CFR50.55a at that time.

2.2 Augmented Examination Requirements Augmented examination requirements are those examinations that are performed above and beyond the requirements of ASME Section XI. Below is a summary of those examinations performed by CPS that are not specifically addressed by ASME Section XI, or the examinations that will be performed in addition to the requirements of ASME Section XI on a routine basis during the Third ISI Interval Alion Science & Technology 2-7 CLN05. G03 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval and Second CISI Interval. Previous revisions of the CPS ISI Program categorized some Augmented Examination Programs by using Augmented Numbers.

2.2.1 NRC Mechanical Engineering Branch (MEB) Technical Position 3-1, dated November 1975 The NRC MEB Technical Position 3-1, "High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment", discusses protection against postulated piping failures in fluid systems outside containment, and includes requirements for licensees to perform 100% volumetric examination of circumferential and longitudinal pipe welds within the pipe break exclusion areas associated with high energy piping in containment penetration areas.

CPS has committed to the requirements of the NRC MEB Technical Position 3-1 through letters to the NRC. Updated Safety Analysis Report (USAR) MEB [Draft Safety Evaluation Report (DSER)] Item No. 11 for ISI Class 1 and USAR Section 6.6.8 for ISI Class 2, detail CPS's compliance with NRC MEB Technical Position 3-1. Examination is required for all piping welds between containment isolation valves. (For those systems which do not have an inboard valve designated as a containment isolation valve per CPS Technical Specification Table 3.6.4-1, the first valve inside the containment shall be considered the penetration boundary in satisfying this requirement) as follows:

(1) ISI Class 1 - Piping welds greater than one (1) inch nominal pipe size, including pipe to valve welds, and associated containment head fitting welds.

(2) ISI Class 2 - High energy piping welds greater than four (4) inches nominal pipe size, including pipe to valve welds, and associated containment head fitting welds as well as all socket welds.

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

Previous revisions of the CPS ISI Program Plan classified augmented examinations of ASME Examination Category B-F, B-J, C-F-1, and C-F-2 welds within high energy line break exclusion regions identified in the USAR as Augmented Inspection Program F 1.2-F. 1 and 1.2-F.2.

Note: This requirement was previously maintained in accordance with USAR MEB [Draft Safety Evaluation Report (DSER)] Item No. 11 for Class 1 and USAR Section 6.6.8 for Class 2. With the implementation of the RISI-BER Program, all BER augmented welds were evaluated under the RISI methodology and were integrated into the RISI Program.

Alion Science & Technology 2-8 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Additional guidance for adaptation of the RISI evaluation process to BER piping is given in EPRI TR-1006937 Rev. 0-A.

2.2.2 CPS USAR Section 6.6.9 The CPS USAR Section 6.6.9 requires volumetric examination of 10% of thin wall (between 3/8" and 1/2") ISI Class 2 RHR system piping welds which would require only surface examinations per ASME Section XI.

The 2004 Edition, No Addenda of ASME Section XI requires CPS to perform volumetric examination of thin wall ISI Class 2 system piping welds. Therefore, the augmented requirements of performing volumetric examination of 10% of thin wall ISI Class 2 RHR system piping welds has been met by this Third ISI Interval ISI Program Plan.

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

Previous revisions of the CPS ISI Program Plan classified augmented examinations of thin wall welds in the USAR as Augmented Inspection Program F1.2-G.

Note: The thin wall welds > 3/8" that were subject to examination under ASME Section XI rules remain in the RISI element selection scope that has been risk evaluated and is potentially subject to RISI examination at CPS.

2.2.3 Boiling Water Reactor Owners' Group (BWROG) Report GE-NE-523-A71-0594-A, Revision 1, "Alternate BWR Feedwater Nozzle Inspection Requirements, May 2000," as approved by NRC final SER dated March 10, 2000, Boiling Water Reactor Owners' Group (BWROG)

Report GE-NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements, August 1999," as conditionally approved by NRC final SER dated June 5, 1998, and NRC NUREG 0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, dated November 1980 These documents discuss the examination requirements for BWR Feedwater (FW) Nozzle and Control Rod Drive (CRD) Return Line Nozzle Cracking. The alternate approach was developed and submitted to the NRC by the BWROG. The NRC conditionally accepted these alternate requirements in the BWROG, Safety Evaluation of Proposed Alternative to BWR Feedwater Nozzle Inspections, dated June 5, 1998.

The CPS requirement in the USAR requires the FW nozzles and the CRD retuin line nozzle, which is capped, to be examined using the methods, techniques, and frequency outlined in the initial examination requirements Alion Science & Technology 2-9 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval of NRC NUREG 0619. Future inspections will comply with BWROG "Alternate BWR Feedwater Nozzle Inspection Requirements,"

GE-NE-523-A71-0594-A, Revision 1, dated May 2000 as accepted by NRC SER (TAC NO. MA6787) dated March 10, 2000.

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

Previous revisions of the CPS ISI Program Plan classified augmented examinations of the Feedwater Nozzle Inner Radii, Nozzle Bores, Nozzle Safe Ends, and Spargers as Augmented Inspection Program Fl.2-H.

2.2.4 Generic Letter 88-01, "NRC Position on IGSCC in 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 Topical Report TR-1 13932, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75)," as conditionally approved by NRC final SER's dated September 15, 2000 and May 14, 2002, and EPRI Topical Report TR- 1012621, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A)," as conditionally approved by NRC final SER dated March 16, 2006 These documents discuss the examination requirements for Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping. References to Generic Letter 88-01 (GL 88-01) within the ISI Program refer to the comprehensive requirements to all of these documents. The final SER's of BWRVIP-75 and BWRVIP-75-A revised the GL 88-01 inspection schedules. The BWRVIP-75 and BWRVIP-75-A revised inspection schedules were based on consideration of inspection results and service experience gained by the industry since issuance of GL 88-01 and NUREG-0313, and includes additional knowledge regarding the benefits of improved BWR water chemistry.

The original CPS responses concerning Generic Letter 88-01 were sent through letters to the NRC. Austenitic stainless steel piping components susceptible to IGSCC shall be examined in accordance with CPS (AmerGen) response to Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping and NRC Request for Additional Information - CPS response to Generic Letter 88-01 (letters from D. P.

Hall to U. S. Nuclear Regulatory Commission, U-601217, dated July 29, 1988, and U-601533, dated September 21, 1989, respectively).

Performing ultrasonic examination in accordance with Appendix VIII of the ASME Section XI meets the GL 88-01 requirements.

Alion Science & Technoloev o.i 2-10 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval Since the issuance of GL 88-01, the BWR Vessels and Internals Project (BWRVIP) has been created. This BWR owners group has worked on the mitigation of IGSCC for BWR internal components. As part of their activities, EPRI Topical Report TR-1 13932, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), dated October 27, 1999" and EPRI Topical Report TR- 1012621, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), dated October 2005", were submitted to the NRC.

Among other issues, this document proposed alternative inspection schedules for IGSCC susceptible welds. Two different inspection schedules were presented; one for plants on Normal Water Chemistry (NWC) and one for plants on effective Hydrogen Water Chemistry (HWC). The HWC schedule may be utilized if the applicable performance criteria are met.

After review of BWRVIP-75 and BWRVIP-75-A, the NRC issued a Safety Evaluation Report (SER) approving the documents with minor changes. (Letter from NRC to Carl Terry, BWRVIP Chairman, Final Safety Evaluation of the "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75)", dated May 14, 2002 and letter from NRC to Bill Eaton, BWRVIP Chairman, Final Safety Evaluation of the "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A)", dated March 16, 2006.)

Based upon NRC endorsement of BWRVIP-75 and BWRVIP-75-A, the CPS conformance to GL 88-01 inspection schedules was changed to BWRVIP-75 and BWRVIP-75-A for NWC plants except for Category A welds. In these documents, the inspection frequency of Category 'D' welds was changed and accepted from every other refueling outage to every six (6) years, for plants with NWC. BWRVIP-75 was adopted by CPS on August 1, 2001 for examination of the IGSCC Category 'D' welds. Subsequently, CPS implemented the requirements of BWRVIP-75-A as a modification to the existing GL 88-01 and BWRVIP-75 Program.

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

Under these new guidelines, ISI Class I and 2 piping structural elements are inspected in accordance with EPRI Topical Reports TR-1 12657, Rev.

B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1. Per these Topical Reports and this Code Case, welds within the plant that are assigned to IGSCC Categories B through G will continue to meet existing IGSCC schedules, while IGSCC Category A welds have been subsumed into the RISI Program. (CPS currently has only IGSCC Category D welds.)

Alion Science & Technologv U*

2-11 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Implementation of the CPS program addressing these documents is included in Exelon Nuclear Procedure ER-AA-330-002, Section 7.0 of this ISI Program Plan, and the associated ISI Database.

Note: The evaluation and repair for any cracks detected on piping susceptible to IGSCC shall be in conformance with Subarticle IWB-3600 of ASME Section XI 2004 Edition, No Addenda.

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

Previous revisions of the CPS ISI Program Plan classified augmented examinations of IGSCC welds as Augmented Inspection Programs F17.2-I and F1.2-J.

2.2.5 RPV Nozzle-To-Safe End Weld (GE SIL No. 455)

CPS will expand the examination area of the RPV nozzle-to-safe end weld where alloy 182 buttering is applied and extended into the nozzle bore area as recommended in CPS response to General Electric Service Information Letter (GE SIL) No. 455, Rev. 1 (Memorandum from F. A. Spangenberg to File, Y-207823, dated June 13, 1988). Also, CPS will incorporate the ultrasonic testing technique and recommendations on repair processes (when required) for nozzle-to-safe end welds where alloy 182 buttering is applied as recommended in IP response to GE SIL No. 455, Revision 1 Supplement I (Memorandum from R. D. Freeman to D. L. Holtzscher, Y-92355, dated September 25, 1989).

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

Previous revisions of the CPS ISI Program Plan classified augmented examinations of RPV nozzle-to-safe end welds (GE SIL No. 455) as Augmented Inspection Program F1.2-K.

2.2.6 NRC Regulatory Guide 1.150, Revision 1, Appendix A, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examination" This Regulatory Guide includes inspection requirements for the ultrasonic examination of RPV welds during preservice and inservice examinations.

The examination criteria of Regulatory Guide 1.150 supplements the requirements of ASME Section XI. However, 10CFR50.55a requires that, with the exception of the shell-to-flange weld and head-to-flange weld, that examination of the RPV welds be conducted in accordance with Appendix VIII, 2001 Edition, No Addenda. The prescriptive guidance in Alion Science & Technology 2-12 CLNO5.G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Regulatory Guide 1.150 is not in total agreement with the Appendix VIII requirements; therefore, Regulatory Guide 1.150 remains applicable only for the examination of the RPV components that are not examined in accordance with Appendix VIII per 10CFR50.55a.

Therefore, CPS examinations of the RPV that are performed using both manual and mechanized examination techniques from the outside surface of the vessel shall be in accordance with Appendix VIII of the ASME Section XI. Examinations of the RPV shell-to-flange and RPV head-to-flange will be in compliance with Regulatory Guide 1.150, Revision 1, Appendix A.

2.2.7 NUREG-0803, Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping, Section 5.1 NUREG-0803, Generic Safety Evaluation Report Regarding Integrity of BWR Scram System Piping, Section 5.1, Page 5-3 requires inspection of Scram Discharge Volume (SDV) system piping in accordance with ASME Section XI. The SDV piping at CPS is ISI Class 2 and is within the scope of the ISI Program. As such, SDV piping and its supports are subject to the applicable ASME Section XI ISI requirements for ISI Class 2 including any alternative examinations such as RISI; no additional Augmented Inservice Inspections are required.

2.2.8 10CFR50.55a(g)(6)(ii)(A), Augmented Examination of Reactor Pressure Vessel Effective September 8, 1992, 10CFR50.55a(g)(6)(ii)(A) required implementation of Augmented Inservice Inspections of RPV shell welds -

Item Number B 1.10 of Examination Category B-A of ASME Section XI.

The interval in effect on September 8, 1992 was the First ISI Interval for CPS. Accordingly, CPS was required to satisfy this rule by the end of the First ISI Interval or to propose an alternative examination program for NRC approval. The rule required examination of "essentially 100%" of all vessel shell welds by the end of the First ISI Interval. The CPS ISI Program First ISI Interval RPV shell weld examinations were conducted in accoi dance with the NRC augmented examination requirements.

The examinations of RPV shell welds, Examination Category B-A, Item Number B 1.11, at CPS, will be conducted in accordance with previously submitted and approved Second ISI Interval Relief Request 4215. The approval authorized under NRC SER dated December 30, 2009 for CPS was for this permanent relief for deferral of RPV shell weld examinations, and thus applies to the remaining term of operation under the existing, initial license, including this Third ISI Interval.

Alion Science & Technology 2-13 CLNO5.G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 2.3 System Classifications and P&ID Drawings The ISI Classification Basis Document details those systems that are ISI Class 1, 2, or 3 that fall within the ISI scope of examinations including the containment structures (metal and concrete). Below is a summary of the classification criteria used within the ISI Classification Basis Document.

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

When a particular group of components is identified as performing an ISI Class 1, 2, or 3 safety function, the components are further reviewed to assure the interfaces (boundary valves and boundary barriers) meet the criteria set by 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 (CPS is not committed to or licensed in accordance with these documents) were also used for guidance in evaluating 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 IOCFR50.2, are typically designated as ISI Class 1 while the other safety related components are evaluated for ISI 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 ISI Class 2 and 3 boundaries are limited to safety related systems and components. Where sufficient 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".

According to 10CFR50.55a, Paragraph (g)(4), the ISI requirements of ASME Section XI are assigned to these components, within the constraints of existing plant design. The CPS ISI Class 1, 2, and 3 components that are exempt from examination are those which meet the criteria of ASME Section XI, Paragraphs IWB-1220, IWC-1220, and IWD-1220. Supports which meet the criteria of Paragraph IWF-1230 of ASME Section XI are also exempt from examination.

For Containment, Class MC components which meet Paragraph IWE- 1220 are exempt from examination, and Class CC components which meet Paragraph IWL-1220 are exempt from examination.

The systemns and components (piping, pumps, valves, vessels, etc.), which are subject to the examinations of Articles IWB-2000, IWC-2000, IWD-2000, and IWF-2000, and pressure tests of Articles IWB-5000, IWC-5000, and IWD-5000 are identified on P&ID Drawings.

Alion Science & Technoloiv 2-14 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Table 2.3-1 provides a listing of the P&ID Drawings that depict the Class 1, 2, and 3 components subject to the requirements of ASME,Section XI, during the Third ISI Interval at CPS. Appendix B to this ISI Program Plan includes a component list which provides those subject to examination as well as some which are exempt from examination. One (1) inch or less nominal pipe size ISI Class 1 components and four (4) inches or less nominal pipe size ISI Class 2 and 3 components, except as required by an augmented examination, have not been included in Appendix B, since these piping lines are all exempted by ASME Section XI.

Alion Science & Technology 2-15 CLNO5.G03 Revision 14

IS! Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.3-1 P&ID DRAWINGS P&ID NUMBER J SYSTEM M05-1002, SHTS. 1, 2, 3, 6 Main Steam (MS)

M05-1004, SHT. 1 Feedwater (FW)

Component Cooling Water (CC) &

M05-1032, SHTS. 2, 6 Fuel Pool Cooling and Cleanup (FC)

M05-1035, SHTS. 1, 2, 3 Diesel Generator (DG)

M05-1037, SHTS. 2, 3 Fuel Pool Cooling and Cleanup (FC)

M05-1047, SHT. 6 Auxiliary Building Drains (RF)

M05-1052, SHTS. 1, 2 Shutdown Water System (SX)

M05-1063, SHT. 1 Combustible Gas Control (HG)

M05-1070, SHT. I MSIV Leakage Control (IS)

M05-1071, SHTS. 1, 2 Nuclear Boiler (NB)

M05-1072, SHTS. 1, 2 Reactor Recirculation (RR)

M05-1073, SHT. 1 Low Pressure Core Spray (LP)

M05-1074, SHT. 1 High Pressure Core Spray (HP)

M05-1075, SH'I S. 1, 2, 3, 4 Residual Heat Removal (RH)

M05-1076, SHTS. 1,4 Reactor Water Cleanup (RT)

M05-1077, SHT. 1 Standby Liquid Control (SC)

M05-1078, SHT. 2 Control Rod Drive (RD)

M05-1079, SHTS. 1, 2 Reactor Core Isolation Cooling (RI)

M05-1105, SHTS. 1, 2 Standby Gas Treatment (VG)

M05-1 110, SHT. 2 Drywell Purge (VQ)

M05-1 111, SHTS. 1, 5 Containment Building Ventilation (VR)

Alion Science & Technology 2-16 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval 2.4 ISI Isometric and Component Drawings for Nonexempt ISI Class Components and Supports 1SI Isometric and Component Drawings were developed to detail the ISI Class 1, 2, and 3 components (welds, bolting, etc.) locations at CPS. These ISI component locations are identified on the ISI Isometric and Component Drawings listed in Table 2.4-1. The CISI Class MC and CC components are identified on the CISI Drawings Diagram, Specification, and Procedures listed in Table 2.4-2.

CPS's ISI Program, including the ISI Database, ISI Classification Basis Document, and ISI Selection Document and schedule, addresses the nonexempt components, which require examination and testing.

A summary of CPS ASME Section XI nonexempt components and supports is included in Section 7.0. A summary of the examination requirements for these components is included in Appendix A.

2-17 CLN05. G03 Revision 14

IS! Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD)

A-i NA REACTOR PRESSURE VESSEL NA A-2 NA RPV-BOTTOM HEAD ASSEMBLY NA A-3A NA CLOSURE HEAD ASSEMBLY NA A-3B NA VENT & HEAD SPRAY NA "APPERTENANCE" A-40 NA MAIN RECIRCULATION PUMP B33-COOIA(B)

A-41A NA CONTROL ROD DRIVE NA A-41B NA CONTROL ROD DRIVE NA B-1A NA RESIDUAL HEAT REMOVAL HEAT EXCHANGER A(B)

B-lB NA RESIDUAL HEAT REMOVAL HEAT EXCHANGER A(B)

B-69 NA RHR PUMP PUMP A,B,C B-71 NA LPCS PUMP 1E21-C001 B-73 NA HPCS PUMP 1E22-COO1 DET-2-1 B-82 RESIDUAL HEAT REMOVAL IRH87AA(AB)-3/4" DET-2-2 B-83 RESIDUAL HEAT REMOVAL 1RHA1AA(AB)-3/4" DET-l 1-1 B-82 RESIDUAL HEAT REMOVAL 1RH87AA(AB)-3/4" DET- 11-2 B-81 RESIDUAL HEAT REMOVAL 1RH86AA(AB)-3/4" DET-32-1 B-81 RESIDUAL HEAT REMOVAL 1RH86AA(AB)-3/4" CY-18-1 B-80 RESIDUAL HEAT REMOVAL 1RH63AA(AB)-3/4" FC-12-1 B-54 RESIDUAL HEAT REMOVAL 1RH-51CA-10" FC-15-1 B-55 RESIDUAL HEAT REMOVAL 1RH-51CB-10" FW-01-1 A-11 FEEDWATER 1FW-02HB- 12" FW-01-2 A-9 FEEDWATER IFW-02GB-18" FW-01-3 A-15 FEEDWATER 1FW-02JB-18" FW-01-4 A-17 FEEDWATER 1FW-02KB-20" FW-0 1-5 A- 13 FEEDWATER 1FW-02HD- 12" FW-02-1 A-10 FEEDWATER 1FW-02HA-12" Alion Science & Technology 2-18 CLN05. GO3 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

_(OLD)

FW-02-2 A-8 FEEDWATER 1FW-02GA- 18" FW-02-3 A- 14 FEEDWATER 1FW-02JA- 18" FW-02-4 A-16 FEEDWATER 1FW-02KA-20" FW-02-5 A-12 FEEDWATER IFW-02HC- 12" FW-03A-I A-16 FEEDWATER IFW-02KA-20" FW-03A-2 B-3 FEEDWATER IFW-02FA-20" FW-03A-3 A-17 FEEDWATER IFW-02KB-20" FW-03A-4 B-4 FEEDWATER 1FW-02FB-20" FW-03A-5 B-50 RESIDUAL HEAT REMOVAL IRH-40BA-10" FW-03A-6 B-51 RESIDUAL HEAT REMOVAL 1RH-40BB-10" HP-01-1 A-25B HIGH PRESSURE CORE SPRAY 1HP-02E-12" HP-01-2 A-25A HIGH PRESSURE CORE SPRAY 1HP-02D-10" HP-01-3 A-24 HIGH PRESSURE CORE SPRAY IHP-02C- 10" HP-02-1 A-24 HIGH PRESSURE CORE SPRAY 1HP-02C-10" HP-03-1 A-24 HIGH PRESSURE CORE SPRAY 1HP-02C- 10" HP-03-2 13-13D HIGH PRESSURE CORE SPRAY 1HP-02B-10" HP-03-3 B-13C HIGH PRESSURE CORE SPRAY 1HP-02A-14" HP-03-4 B-13B HIGH PRESSURE CORE SPRAY IHP-02A-14" HP-03-5 B-13A HIGH PRESSURE CORE SPRAY 1HP-02F-16" HP-03-6 B-14 HIGH PRESSURE CORE SPRAY IHP-18A-12" HP-03-i B-16 HIGH PRESSURE CORE SPRAY IHP-18F-12" HP-03-8 B-15 HIGH PRESSURE CORE SPRAY IHP-18D-10" HP-03-9 B-17 HIGH PRESSURE CORE SPRAY 1HP-19A-10" LP-01-1 A-23C LOW PRESSURE CORE SPRAY ILP-02C- 12" LP-01-2 A-23B LOW PRESSURE CORE SPRAY ILP-02B-10" LP-01-3 A-23A LOW PRESSURE CORE SPRAY ILP-02B-10" LP-02-1 A-23A LOW PRESSURE CORE SPRAY 1LP-02B-10" Alion Science & Technology 2-19 CLN05. GO3 Revision 14

IS! Program Plan Clinton Power Station Unit]1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD) S LP-04-1 A-23A LOW PRESSURE CORE SPRAY ILP-02B-10" LP-04-2 B-10C LOW PRESSURE CORE SPRAY ILP-02E- 10" LP-04-3 B-101B LOW PRESSURE CORE SPRAY ILP-02A- 12" LP-04-4 B-10A LOW PRESSURE CORE SPRAY ILP-02A- 12" LP-04-5 B-1 LOW PRESSURE CORE SPRAY 1LP-02D-14" LP-04-6 B-12 LOW PRESSURE CORE SPRAY ILP-18A-10" MS-07-1 B-2 MAIN STEAM 1MS-OIEA(B,C,D)-24" MS-A-i A-4A MAIN STEAM IMS-A-24" MS-A-2 A-4B MAIN STEAM IMS-ASA(ASB)-10" MS-B-1 A-5A MAIN STEAM IMS-B-24" MS-B3-2 A-513 MAIN STEAM MS-BSA(BSB, BSC,BSD,BSE)-I0" MS-C-1 A-6A MAIN STEAM IMS-C-24" MS-C-2 A-6B MAIN STEAM IMS-CSA(CSB,CSC, CSD,CSE,CSF)-10" MS-D-1 A-71 MAIN STEAM IMS-24" MS-D-2 A-7B MAIN STEAM IMS-DSA(DSB,DSC)- 10" RH-01-1 A-26B RESIDUAL HEAT REMOVAL IRH-03DA-10" RH-01-2 A-26A RESIDUAL HEAT REMOVAL IRH-03CA-12" RH-02-1 A-26A RESIDUAL HEAT REMOVAL IRH-03CA-12" RH-02-2 B-24 RESIDUAL HEAT REMOVAL 1RH-03BA-12" RH-02-3 B-52 RESIDUAL HEAT REMOVAL IRH-50AA-10" RH-02-4 B-54 RESIDUAL HEAT REMOVAL IRH-51CA-10" RH-03-1 A-27C RESIDUAL HEAT REMOVAL IRH-03DB-10" RH-03-2 A-27B RESIDUAL HEAT REMOVAL IRH-03CB-12" RH-03-3 A-27A RESIDUAL HEAT REMOVAL IRH-03CB-12" RH-04-1 A-27A RESIDUAL HEAT REMOVAL IRH-03CB-12" RH-04-2 B-25 RESIDUAL HEAT REMOVAL IRH-03BB(FB)-12" Alion Science & Technology 2-20 CLN05.GO3 Revision 14

IS! Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE No.

(OLD)

RH-04-3 B-53A RESIDUAL HEAT REMOVAL 1RH-50GB-10" RH-04-4 B-53B RESIDUAL HEAT REMOVAL 1RH-50AB-10" RH-04-5 B-53C RESIDUAL HEAT REMOVAL 1RH-50AB-10" RH-04-6 B-55 RESIDUAL HEAT REMOVAL 1RH-51CB-10" RH-05-1 A-28D RESIDUAL HEAT REMOVAL 1RH-04C-10" RH-05-2 A-28C RESIDUAL HEAT REMOVAL 1RH-04B-12" RH-05-3 A-28B RESIDUAL HEAT REMOVAL 1RH-04B-12" RH-06-1 A-28B RESIDUAL HEAT REMOVAL 1RH-04B- 12" RH-06-2 A-28A RESIDUAL HEAT REMOVAL IRH-04B-12" RH-07A-1 B-36 RESIDUAL HEAT REMOVAL IRH-29BA-8" I?-H-07A-2 B-38A RESIDUAL HEAT REMOVAL 1RH-29CA(DA,FA,GA)-6" RH-07A-3 B-38B RESIDUAL HEAT REMOVAL IRH-29EA-14" RH-07A-4 B-33A RESIDUAL HEAT REMOVAL 1RH-22AA-14" RH-07A-5 B-33B RESIDUAL HEAT REMOVAL IRH-22BA- 18" RH-07A-6 B-20 RESIDUAL HEAT REMOVAL 1RH-02AA- 14" RH-07A-7 B-22A RESIDUAL HEAT REMOVAL 1RH-03AA-14" RH-07A-8 B-22C RESIDUAL HEAT REMOVAL IRH-03EA-18" RH-07A-9 B-48 RESIDUAL HEAT REMOVAL I RH-03AA- 14" RH-07A-10 B-50 RESIDUAL HEAT REMOVAL 1RH-40AA- I0" RH-07A-I 1 B-40 RESIDUAL HEAT REMOVAL 1RH-40BA-10" RH-07B-1 B-24 RESIDUAL HEAT REMOVAL 1RH-03BA-12" RH-07B-2 B-22B RESIDUAL HEAT REMOVAL 1RH-03AA-14" RH-07B-3 B-22A RESIDUAL HEAT REMOVAL 1RH-03AA-14" RH-07B-4 B-45 RESIDUAL HEAT REMOVAL IRH-37AA-12" RH-07B-5 B-46 RESIDUAL HEAT REMOVAL 1RH-38AA-14" RH-08-1 B-21 RESIDUAL HEAT REMOVAL 1RH-02AB-14" RH-08-2 B-34A RESIDUAL HEAT REMOVAL IRH-22AB- 14" Alion Science & Technology 2-21 CLN05.GO3 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD)T RH-08-3 B-34B RESIDUAL HEAT REMOVAL IRH-22BB-18" RH-08-4 B-23A RESIDUAL HEAT REMOVAL 1RH-03AB-14" RH-08-5 B-25 RESIDUAL HEAT REMOVAL IRH-03BB(FB)-12" RH-08-6 B-23B RESIDUAL HEAT REMOVAL 1R.H-03EB-I 8" RH-08-7 B-41 RESIDUAL HEAT REMOVAL 1RH-30AB-8" RH-08-8 B-39B RESIDUAL HEAT REMOVAL 1RH-29EB-14" RH-08-9 B-39A RESIDUAL HEAT REMOVAL 1RH-29CB(DB,FB,GB)-6" RH-08-10 B-37 RESIDUAL HEAT REMOVAL IRH-29BB-8" RH-08-11 B-49 RESIDUAL HEAT REMOVAL IRH-40AB-10" RH-08-12 B-51 RESIDUAL HEAT REMOVAL 11RH-40BB- 10" RH-08-13 B-47 RESIDUAL HEAT REMOVAL I RH-38AB- 14" RH-09-1 A-29A RESIDUAL HEAT REMOVAL IRH-09A-18" RH-09-2 B-32 RESIDUAL HEAT REMOVAL IRH-09B-18" RH-09-3 8-29B RESIDUAL HEAT REMOVAL 1Rtt-07B- 18" RH-09-4 B-29A RESIDUAL HEAT REMOVAL 1RH-07A-16" RH-09-5 B-18B RESIDUAL HEAT REMOVAL IRH-0IBA-20" RH-09-6 B-30 RESIDUAL HEAT REMOVAL 1RH-07C-16" RH-09-7 B-19 RESIDUAL HEAT REMOVAL 1RH-O1BB-20" RH-09-8 B-18A RESIDUAL HEAT REMOVAL 1RH-06A-16" RH-09-9 B-28 RESIDUAL HEAT REMOVAL 1RH-06A-16" RH-09-10 B-31 RESIDUAL HEAT REMOVAL 1RH-08A-14" RH-10-1 B-31 RESIDUAL HEAT REMOVAL 1RH-08A-14" RH-11-1 B-26A RESIDUAL HEAT REMOVAL 1RH-04A-14" RH-I 1-2 B-56 RESIDUAL HEAT REMOVAL IRH-97A- I0" RH-12-1 A-28A RESIDUAL HEAT REMOVAL 1RH-04B-12" RH- 12-2 B-27 RESIDUAL HEAT REMOVAL 1RH-04D-12" RH-12-3 B-26A RESIDUAL HEAT REMOVAL I RH-04A- 14" Alion Science & Technology 2-22 CLN05. G03 Revision 14

IS! Program Plan Clinton Power Station Unit]1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD) 1 RH- 12-4 B-26B RESIDUAL HEAT REMOVAL 1RH-39A-14" RH-14-1 B-58 REACTOR CORE ISOLATION COOLING 1RI-04B-8" RH-14-2 B-35 RESIDUAL HEAT REMOVAL 1RH-29A-8" RH-14-3 B-36 RESIDUAL HEAT REMOVAL 1RH-29BA-8" RIH-14-4 B-37 RESIDUAL HEAT REMOVAL 1RH-29BB-8" RH-21-1 B-52 RESIDUAL HEAT REMOVAL IRH-50AA-10" RH-22-1 B-53C RESIDUAL HEAT REMOVAL IRH-50AB-10" RH-23-1 A-39 REACTOR CORE ISOLATION COOLING IRI-29A-4" RH-23-2 A-30 RESIDUAL HEAT REMOVAL IRH-46B-4" RH-34-1 A-29B RESIDUAL HEAT REMOVAL IRH-09C-18" RH-34-2 A-29A RESIDUAL HEAT REMOVAL IRH-09A-18" RH-35-1 B-79 RESIDUAL HEAT REMOVAL 1RH41AA(AB)-I" RH-38-1 B-79 RESIDUAL HEAT REMOVAL IRH41AA(AB)-1" RI-01-1 A-38 REACTOR CORE ISOLATION COOLING IRI-04A-8" RI-02-1 A-38 REACTOR CORE ISOLATION COOLING IRI-04A-8" RI-02-2 B-58 REACTOR CORE ISOLATION COOLING 1RI-04B-8" RI-02-3 B-64 REACTOR CORE ISOLATION COOLING IRI-69A-10" RI-03-1 B-59 REACTOR CORE ISOLATION COOLING IRI-07A-12" RI-03-2 B-60 REACTOR CORE ISOLATION COOLING IRI-08A-12" RI-03-3 B-63 REACTOR CORE ISOLATION COOLING 1RI-43A-8" RI-08-1 A-37A REACTOR CORE ISOLATION COOLING 1RI-03C-4" RI-08-2 A-36 REACTOR CORE ISOLATION COOLING 1RI-03B-6" RI-08-3 B-57 REACTOR CORE ISOLATION COOLING 1RI-03A-6" RI-08-4 A-39 REACTOR CORE ISOLATION COOLING 1RI-29A-4" RI-08-5 B-78 RCIC PUMP IE51 -COOl RI-10-1 A-37B REACTOR CORE ISOLATION COOLING IRI-03C-4" RI-10-2 A-37A REACTOR CORE ISOLATION COOLING IRI-03C-4" Alion Science & Technology 2-23 CLN05. GO3 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD) STE RI-i 1-1 A-37B REACTOR CORE ISOLATION COOLING 1RI-03C-4" RI-I 1-2 A- 18 NUCLEAR BOILER I NB-01A-4" TBD RR-750 REACTOR RECIRCULATION 1RR04AA(AB)-2" TBD RR-767 REACTOR WATER CLEANUP 1RRl5A-2" RR-A-I A-21A REACTOR RECIRCULATION lRR-A-20" RR-A-2 A-19A REACTOR RECIRCULATION 1RR-AM-16" RR-A-3 A-19B REACTOR RECIRCULATION 1RR-AA(AB,AC, AD,AE)- 10" RR-A-4 B-21B REACTOR RECIRCULATION I RR-A-CRW(DRW)-4" RR-B-1 A-22A REACTOR RECIRCULATION 1RR-B-20" RR-B-2 A-20A REACTOR RECIRCULATION I RR-BM- 16" RR-B-3 A-20B REACTOR RECIRCULATION lRR-BF(BG,BH, BJ,BK)-10" RR-B-4 A-22B REACTOR RECIRCULATION 1RR-B-CRW(DRW)-4" RT-01-1 A-32B REACTOR WATER CLEANUP 1RT-O1EB(D)-4" RT-01-2 A-32A REACTOR WATER CLEANUP 1RT-01AB-4" RT-01-3 A-33A REACTOR WATER CLEANUP 1RT-01 B-6" RT-01-4 A-33B REACTOR WATER CLEANUP 1RT-01 B-6" RT-0 1-5 A-31B REACTOR WATER CLEANUP 1RT-OIEA(C)-4" RT-01-6 A-31A REACTOR WATER CLEANUP 1RT-O1AA-4" RT-06-1 A-33B REACTOR WATER CLEANUP 1RT-01B-6" TBD RT-33 REACTOR WATER CLEANUP 1RT28D-3" TBD RT-34 REACTOR WATER CLEANUP 1RT28C-3" TBD RT-37 REACTOR WATER CLEANUP 1RT28A(B,C,D)-3" TBD SC-3 STANDBY LIQUID CONTROL ISC02DA(DD)-3" I SC03DB-3" 1SC02DE-3" TBD SC-4 STANDBY LIQUID CONTROL 1SCO3DB-3" I 4 S A D SC03DB(DC)-3" SD-90-1 I B-74 I SCRAM DISCHARGE VOLUME 190 DEGREE SIDE- 10" Alion Science & Technology 2-24 CLN05. G03 Revision 14

1Sf ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS FIGURE NO. FIGURE NO. SYSTEM LINE NO.

(OLD) T SD-90-2 B-75 SCRAM DISCHARGE VOLUME 90 DEGREE SIDE-12" SD-270-1 B-76 SCRAM DISCHARGE VOLUME 270 DEGREE SIDE-10" SD-270-2 B-77 SCRAM DISCHARGE VOLUME 270 DEGREE SIDE-12" Alion Science & Technulogy 2-25 CLN05. GO3 Revision 14

IS ProgramPlan Clinton PowerStation Unit 1, Third Interval TABLE 2.4-2 CISI DRAWINGS, DIAGRAM, SPECIFICATION, AND PROCEDURES DRAWING NUMBER TITLE OF DRAWING/DIAGRAM/SPECIFICATION S27-1905, Sht. 1 Containment Liner Internal Face S27-1906, Sht. 2 Containment Liner Internal Face S27-1907, Sht. 3 Containment Liner Internal Face S27-1908, Sht. 4 Containment Liner Internal Face S27-1403, Sht. 1 Containment Exterior Face S27-1404, Sht. 2 Containment Exterior Face M03-1 101, Sheels 1, 2, & 3 Mechanical Penetrations Containment M06-1000 Sheets 1-8 Head Fitting and Guard Pipe Details E27-1310 Electrical Penetrations Containment K2816 Steel Liner Work for Reactor Containment Structures K2944 Concrete and Grout Work K2882 Piping Design K2978 Electrical Penetrations K2801-0150 Vol II Part 3 Inclined Fuel Transfer Table K2816-0001 Tab 2 Equipment Hatch K2816-0001 Tab I Personnel Airlocks K2978-0001 Electrical Penetrations DS-ME-09 Mechanical Penetrations Design DS-ME-09-CP Design Specification for Piping Penetration Assemblies DS-SD-03-CP Containment Structure Design Criteria ER-AA-1 100 Implementing and Managing Engineering Programs ER-AA-330 Conduct of Inservice Inspection Activities ER-AA-336-005 Visual Examination of Section XI Class CC Concrete Containment Structures ER-AA-330-007 Visual Examination of Section XI Class MC Surfaces and Class CC Liners ER-AA-335-018 Detailed, General, VT-I, VT-IC, VT-3 and VT-3C, Visual Examination of ASME Class MC and CC Containment Surfacesand Components Alion Science & Technology 2-26 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 2.5 Technical Approach and Positions When the requirements of ASME Section XI are not easily interpreted, CPS has reviewed general licensing/regulatory requirements and industry practice to determine a practical method of implementing the Code requirements. The Technical Approach and Position (TAP) documents contained in this section have been provided to clarify CPS's implementation of ASME Section XI requirements.

An index which summarizes each technical approach and position is included in Table 2.5-1.

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ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX Position Revision Status' (Program) Description of Technical Approach Number Date 2 and Position 13T-0 0 Active (SPT) System Leakage Testing of Non-Isolable 05/21/10 Buried Components.

0 (SPT) Valve Seats/Disks as Pressurization 05/21/10 Boundaries.

Note 1: ISI Program Technical Approach and Position Status Options: Active - Current Technical Approach and Position is being utilized at CPS; Deleted - Technical Approach and Position is no longer being utilized at CPS.

Note 2: The revision listed is the latest revision of the subject Technical Approach and Position. The date noted in the second column is the date of the ISI Program Plan revision when theTechnical Approach and Position was incorporated into the document.

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IS1 ProgramPlan Clinton Power Station Unit 1, Third Interval TECHNICAL APPROACH AND POSITION NUMBER 13T-01 Revision 0 COMPONENT IDENTIFICATION:

Code Class: 2 and 3

Reference:

IWA-5244(b)(2)

Examination Category: C-H, D-B Item Number: C7.10, D2.10

Description:

System Leakage Testing of Non-Isolable Buried Components Component Numbei: Non-Isolable Buried Pressure Retaining Components CODE REOUIREMENT:

Paragraph IWA-5244(b)(2) requires non-isolable buried components be tested to confirm that flow during operation is not impaired.

POSITION:

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

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

For lines in which the open end is accessible to visual examination while the system is in operation, visual evidence of flow discharging the line is considered as reasonable proof of adequate flow through the open ended line.

For open ended portions of systems where the process fluid is pneumatic, evidence of gaseous discharge shall be considered reasonable proof of adequate flow through the open ended line. Such test may include passing smoke through the line, hanging balloons or streamers, using a remotely operated blimp, using thermography to detect hot air, etc.

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

CPS's position is that proof of adequate flow is all that is required for testing these open ended lines and that no further visual examination is necessary. This is consistent with the requirements for buried piping, which is not subject to visual examination.

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IS! ProgramPlan Clinton Power Station Unit 1, Third Interval TECHNICAL APPROACH AND POSITION NUMBER 13T-02 Revision 0 COMPONENT IDENTIFICATION:

Code Class: 1, 2, and 3

Reference:

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

==

Description:==

Valve Seats/Disks as Pressurization Boundaries Component Number: All Pressure Testing Boundary Valves CODE REQUIREMENT:

Paragraph IWA-522 1 requires the pressurization boundary for system leakage testing extend to those pressure retaining components under operating pressures during normal system service.

POSITION:

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

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

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

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IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 3.0 COMPONENT ISI PLAN The CPS Component ISI Plan includes ASME Section XI nonexempt pressure retaining welds, piping structural elements, pressure retaining bolting, attachment welds, pump casings, valve bodies, reactor pressure vessel interior, reactor pressure vessel interior attachments, and reactor pressure vessel welded core support structures of ISI Class 1, 2, and 3 components that meet the criteria of Subarticle IWA-1300. These components are identified on the P&ID Drawings listed in Section 2.3, Table 2.3-1. Procedure ER-AA-330-002 "Inservice Inspection of Welds and Components", implements the ASME Section XI welds and components program. This Component ISI Plan also includes Augmented Examination Program examination requirements specified by documents other than ASME Section XI. For a detailed discussion of these examination requirements, see Section 2.2 of this document.

3.1 Nonexempt ISI Class Components The CPS ISI Class 1, 2, and 3 components subject to examination are those which are not exempted under the criteria of Paragraphs IWB- 1220, IWC- 1220, and IWD- 1220, respectively. A summary of CPS ASME Section XI nonexempt components is included in Section 7.0.

3.1.1 Identification of lSI Class 1, 2, and 3 Nonexempt Components ISI Class I and 2 nonexempt components are identified on the ISI Isometrics and Component Drawings listed in Section 2.4, Table 2.4-1.

Note that ISI Class 1, 2, and 3 welded attachments are identified on CPS individual support detail drawings and on M06 drawings.

3.2 Risk-Informed Examination Requirements Piping structural elements that fall under RISI Examination Category R-A are risk ranked as High (1, 2, and 3), Medium (4 and 5), and Low (6 and 7). Per the EPRI Topical Reports TR- 112657, Rev. B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1, piping structural elements ranked as High or Medium Risk are subject to examination while piping structural elements ranked as Low Risk are not subject to examinations (except for pressure testing). Thin wall welds that were excluded from volumetric examination under ASME Section XI rules per Table IWC-2500-1 are included in the element scope that is potentially subject to RISI examination at CPS.

Pipiag structural elements may be excluded from examination (other than pressure testing) under the RISI Program if the only degradation mechanism present for a given location is inspected for cause under certain other CPS programs such as the Flow Accelerated Corrosion (FAC) or Intergranular Stress Corrosion Cracking (IGSCC) Programs. These piping structural elements will remain part of the FAC or IGSCC programs, which already perform "for cause" inspections to detect Alion Science & Technology 3-1 CLNO5. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval these degradation mechanisms. Piping structural elements susceptible to FAC or IGSCC along with another degradation mechanism (e.g., thermal fatigue) are retained as part of the RISI scope and are included in the element selection for the purpose of performing examinations to detect the additional degradation mechanism. The RISI Program element examinations are performed in accordance with Relief Request I3R-01.

3.3 Reactor Coolant Pressure Boundary Normal Makeup Exemption In accordance with ASME Section XI, Paragraph IWB- 1220(a), components that are connected to the reactor coolant pressure boundary may be exempted from the surface and volumetric examination requirements of ASME Section XI, provided they are of such a size and shape that upon a postulated pressure boundary rupture, the resulting flow of coolant under normal operating conditions is within the capacity of makeup systems.

3.3.1 Makeup Calculation The basis for determining the makeup size exemption of ISI Class 1 water and steam lines is provided in Letter Y-109584, dated August 24, 2009, "Makeup Capacity Exemption of Class 1 Components Per ASME Section XI". The makeup flow rate is determined from systems which are not part of the emergency core cooling system and which are operable from on-site emergency power.

Based on Letter Y-109584, the following ISI Class 1 piping qualifies for the make-up capacity exemption of Paragraph IWB- 1220(a):

1. Steam system piping with an inside diameter (ID) of 2.752" and smaller.

2. Water system piping with an ID of 1.376" and smaller.

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IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 4.0 SUPPORT ISI PLAN The CPS Support ISI Plan includes the supports of ASME Section XI nonexempt ISI Class 1, 2, and 3 components as described in Section 3.0. Procedure ER-AA-330-003 "Visual Examination of Section XI Component Supports", implements the ASME Section XI Support ISI Plan.

4.1 Nonexempt ISI Class Supports The CPS ISI Class 1, 2, and 3 nonexempt supports are those which do not meet the t:xemption criteria of Paragraph IWF-1230 of ASME Section Xi. A summary of CPS ASME Section XI nonexempt supports is included in Section- 7.0.

4.1.1. Identification of ISI Class 1, 2, and 3 Nonexempt Supports ISI Class 1, 2, and 3 nonexempt supports are identified on CPS individual support detail drawings and on M06 drawings.

4.2 Snubber Examination and Testing Requirements 4.2.1 ASME Section XI Paragraphs IWF-5200(a) and (b) and IWF-5300(a) and (b) require VT-3 Visual Examination and Inservice Tests of snubbers to be performed in accordance with the Operation and Maintenance of Nuclear Power Plants (OM), Standard ASME/ANSI OM, Part 4. As allowed by 10CFR50.55a(b)(3)(v), CPS will use Subsection ISTD, "Inservice Testing of Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants,"

ASME OM Code, 2004 Edition, No Addenda, to meet the requirements in ASM/iE Section XI Paragraphs IWF-5200(a) and (b) and IWF-5300(a) and (b). Per 10CFR50.55a(b)(3)(v), visual examinations shall be performed using the VT-3 visual examination method described in Paragraph IWA-2213. A summary of the CPS safety-related and non-safety related snubbers is included in Section 7.0.

Procedure ER-AA-330-004 "Visual Examination of Technical Specification Snubbers", implements the visual inspection program for safety related and non-safety related snubbers. Corporate procedures ER-AA-330-010 "Administration of Snubber Functional Testing",

ER-AA-330-011 "Snubber Service Life Monitoring Program", and station 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 XI ISI Program uses Subsection IWF to define support inspection requirements. The ISI Program maintains the Code Class snubbers in the populations subject to inspection per Article IWF-2000.

This is done to accommodate scheduling and inspection requirements of Alion Science & Technologv 4-1 CLN05. G03 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval the related attachment 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 attachments for snubbers to be examined in accordance with Subsection IWF of ASME Section XI. This results in VT-3 visual examination of the snubber attachment hardware including the bolting, pins, and their interface to the clamp, but does not include the component-to-clamp interface.

The ASME Section XI ISI Program uses Subsection IWF to define the inspection requirements for all Class 1, 2, and 3 supports, regardless of type. The ISI Program maintains the Code Class snubbers in the support populations subject to inspection per Article IWF-2000. 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 fWF-5300(c).

It should be noted that the examination of snubber welded attachments will be performed in accordance with the ASME Section XI Subsections IWB, IWC, and IWD welded attachment examination requirements (e.g.,

Examination Categories B-K, C-C, and D-A).

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ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 5.0 SYSTEM PRESSURE TESTING ISI PLAN The CPS System Pressure Testing (SPT) ISI Plan includes pressure retaining ASME Section XI, ISI Class 1, 2, and 3 components, with the exception of those specifically exempted by Paragraphs IWA-5 110(c), IWC-5222(b), and IWD-5222(b). RISI piping structural elements, regardless of risk classification, remain subject to pressure testing as part of the current ASME Section XI program.

The SPT ISI Plan performs system pressure tests and required VT-2 visual examinations on the ISI Class 1, 2, and 3 pressure retaining components to verify system and component structural integrity. This program conducts both Periodic and Interval (10-Year frequency) pressure tests as defined in ASME Section XI Inspection Program B.

Procedure ER-AA-330-001, "Section XI Pressure Testing," as well as CPS site-specific test procedures, implement the ASME Section XI System Pressure Testing ISI Plan.

This SPT ISI Plan also includes Augmented Examination Program examination requirements specified by documents other than ASME Section XI. For detailed discussion of these examination requirements, see Section 2.2 of this document.

5.1 ISI Class Systems All ISI Class 1 pressure retaining components, typically defined as the reactor coolant pressure boundary, are required to be tested. Those portions of ISI Class 2 and 3 systems that are required to be tested include the pressure retaining boundaries of components required to operate or support the system safety functions. ISI Class 2 and 3 open ended discharge piping and components are excluded from the examination requirements per Paragraphs IWC-5222(b) and IWD-5222(b).

5.1.1 Identification of Class 1, 2, and 3 Components Components subject to ASME Section XI System Pressure Testing are shown on the P&ID Drawings listed in Section 2.3, Table 2.3-1.

Additional information on the classification of various system boundaries is provided in the ISI Classification Basis Document.

5.1.2 Identification of System Pressure Tests The ISI Boundary Drawings are highlighted and then utilized during the walkdown to define which systems, or portions of systems, fall under a specific system pressure test. Individual tests are identified and maintained in the CPS ISI Database.

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IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 5.2 Risk-Informed Examinations of Socket Welds Socket welds selected for examination under the RISI program are to be inspected with a VT-2 visual examination each refueling outage per ASME Code Case N-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socket welds selected for inspection under the RISI program are pressurized each refueling outage during a system pressure test in accordance with Paragraph IWA-521 I(a).

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ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 6.0 CONTAINMENT ISI PLAN The CPS Containment ISI Plan includes ASME Section XI CISI Class MC pressure retaining components and their integral attachments, and CISI Class CC components and structures that meet the criteria of Subarticle IWA-1300. This Containment ISI Plan also includes information related to augmented examination areas, component accessibility, and examination review.

The inspection of containment structures and components are performed per procedures ER-AA-330-005, "Visual Examination of Section XI Class CC Concrete Containment Structures" and ER-AA-330-007, "Visual Examination of Section XI Class MC Surfaces and Class CC Liners".

6.1 Nonexempt CISI Class Components The CPS CISI Class MC and CC components identified on the CISI Reference Drawings are those not exempted under the criteria of Paragraphs IWE-1220 and IWL-1220 in the 2001 Edition through the 2003 Addenda of ASME Section XI.

A summary of CPS ASME Section XI nonexempt CISI components is included in Section 7.0.

The process for scoping CPS components for inclusion in the Containment ISI Plan is included in the containment sections of the ISI Classification Basis Document. These sections include a listing and detailed basis for inclusion of containment components.

Components that are classified as CISI Class MC and CC must meet the requirements of ASME Section XI in accordance with 10CFR50.55a(g)(4).

Supports of Subsection IWE components are not required to be examined in accordance with IOCFR50.55a(g)(4)(v).

6.1.1 Identification of CISI Class MC and CC Nonexempt Components CISI Class MC and CC components are identified on the CISI Reference Drawings listed in Section 2.4, Table 2.4-2.

6.1.2 Identification of CISI Class MC and CC Exempt Components Certain containment components or parts of components may be exempted from examination based on design and accessibility per the requirements of Paragraphs IWE- 1220 and IWL- 1220.

The process for exempting CPS components from the Containment ISI Plan per Paragraphs IWE-1220 and IWL-1220 is included in the containment sections of the ISI Classification Basis Document. These sections include discussions of exempt components and the bases for those exemptions.

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ISI ProgramPlan ClintonPower Station Unit 1, ThirdInterval 6.2 Augmented Examinations Areas The containment section of the ISI Classification Basis Document discusses the containment design and components. Metal containment surface areas subject to accelerated degradation and aging require augmented examination per Examination Category E-C and Paragraph IWE- 1240.

Similarly, concrete surfaces may be subject to Detailed Visual examination in accordance with Item Number L 1.12 and Paragraph IWL-23 10(b), if declared to be 'Suspect Areas'.

No significant conditions were identified in the First CISI Interval and no significant conditions are currently identified in the Second CISI Interval as requiring application of additional augmented examination requirements under Paragraph IWE-1240 or IWL-23 10.

6.3 Component Accessibility CISI Class MC and CC components subject to examination shall remain accessible for either direct or remote visual examination from at least one side per the requirements of ASME Section XI, Paragraph IWE-1230.

Paragraph IWE-123 1(a)(3) requires 80% of the pressure-retaining boundary that was accessible after construction to remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant.

Portions of components embedded in concrete or otherwise made inaccessible during construction are exempted from examination, provided that the requirements of ASME Section XI, Paragraph IWE-1232 have been fully satisfied.

In addition, inaccessible surface areas exempted from examination include those surface areas where visual access by line of sight with adequate lighting from permanent vantage points is obstructed by permanent plant structures, equipment, or components; provided these surface areas do not require examination in accordance with the inspection plan, or augmented examination in accordance with Paragraph IWE-1240.

6.4 Responsible Individual and Engineer ASME Section XI Subsection IWE requires the Responsible Individual to be involved in the development, performance, and review of the CISI examinations.

The Responsible Individual shall meet the requirements of ASME Section XI, Paragraph IWE-2320.

ASME Section XI Subsection IWL requires the Responsible Engineer to be involved in the development, approval, and review of the CISI examinations. The Alion Science & Technology 6-2 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval Responsible Engineer shall meet the requirements of ASME Section XI, Paragraph IWL-2320.

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ISI ProgramPlan Clinton PowerStation Unit 1, ThirdInterval 7.0 COMPONENT

SUMMARY

TABLES 7.1 Inservice Inspection Summary Tables The following Table 7.1-1 provides a summary of the ASME Section XI pressure retaining components, supports, containment structures, system pressure testing, and augmented program components for the Third ISI Interval and the Second CISI Interval at CPS.

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

Examination Item Number Description Exam Total Number of Relief Request/ Notes Category (with (or Risk Requirements Components by TAP Number Examination Categ or r System Category Augmented Description) Number)

(1) (2) (3) (4) (5) (6) __7 (1) Examination Category (with Examination Category Description):

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

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

Examination Category "NA" is used to identify Augmented Examination Programs.

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

Provides the Item Number as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, IWF-2500-1, and IWL-2500- 1. Only those Item Numbers applicable to CPS are identified.

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

Specific abbreviations such as 2.2.1, 2.2.2, 2.2.3, 2.2.4, 2.2.5, 2.2.6, 2.2.7, and 2.2.8 have been developed to identify Augmented Examination Programs.

Alion Science & Technology 7-1 CLN05. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval (3) Item Number (or Risk Category Number or Augmented Number)

Description:

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

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

For Augmented Examination Program examinations, a description of the Augmented Examination Program basis 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, IWF-2500-1, and IWL-2500-1.

Provides the examination requirements for piping structural elements under RISI that are in accordance with the EPRI Topical Reports TR-1 12657, Rev. B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1.

Provides the examination requirements for Augmented Examination Programs.

(5) Total Number Of Components by System:

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

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

Note that the total number of components by system are subject to change after completion of plant modifications, design changes, and ISI system classification updates.

(6) Relief Request/Technical Approach & Position Number:

Provides a listing of Relief Request/ TAP Numbers applicable to specific components, the ASME Section XI Item Number, Risk Category Number, or Augmented Number. Relief Requests and TAP Numbers that generically apply to all components, or an entire class are not listed. If a Relief Request/ TAP Number is identified, see the corresponding relief request in Section 8.0 or the TAP Number in Section 2.5.

Alion Science & Technologv 7-2 CLN05.GO3 Revision 14

ISI ProgramPlan Clinton PowerStation Unit), ThirdInterval (7) Notes:

Provides a listing of program notes applicable to the ASME Section XI Item Number, Risk Category Number, or Augmented Number. If a program note number is identified, see the corresponding program note in Table 7.1-2.

Alion Science & Technology 7-3 CLN05.GO3 Revision 14

1Sf Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) I System B-A B 1.11 Circumferential Shell Welds (Reactor Vessel) Volumetric AAA: 4 4215 Pressure Retaining Welds B 1.12 Longitudinal Shell Welds (Reactor Vessel) Volumetric AAB: 11 in Reactor Vessel B 1.21 Circumferential Head Welds (Reactor Vessel) Volumetric AAL: I B 1.22 Meridional Head Welds (Reactor Vessel) Volumetric AAC: 4 1 _AAM: 6 B11.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric AAA: 1 B 1.40 Head-to-Flange Weld (Reactor Vessel) Volumetric & AAL: I Surface B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric AAG: 29 13R-02 11 Full Penetration Welds AAN: 2 of Nozzles in Vessels B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric AAH: 29 13R-02 AAO: 2 Alion Science & Technology 7-4 CLN05. GO3 Revision 14

I51 Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-I AAJ: 1 Pressure Retaining B6.20 Clesu-, Studs (Reactor Vessel) Volumetric AAJ: 1 12 Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric AAJ: 1 2 in. In Diameter B6.50 Closure Washers (Reactor Vessel) Visual, VT-i AAJ: I B6.180 Bolts & Studs (Pumps) Volumetric TRA: 1 I TRB: 1 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-I TRA: I TRB: I B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-I TRA: 1 TRB: 1 Alion Science & Technology 7-5 CLNO5. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) I System B-G-2 B7.10 Bolts, Studs, & Nuts (Reactor Vessel) Visual, VT-i AAP: I Pressure Retaining B7.50 Bolts, Studs, & Nuts (Piping) Visual, Vi- 1 PMS: 16 Bolting, 2 in. and Less PNB: 1 PRI: 2 PRR: 4 B7.70 Bolts, Studs, & Nuts (Valves) Visual, VT-I PFW: 6 PHP: 3 PLP: 2 PMS: 24 PRH: 12 PR: 3 PRR: 6 PRT: 2 B7.80 Bolts, Studs, & Nuts in CRD Housing (Reactor Vessel) Visual, VT-I AAE: 1 10 Alion Science & Technology 7-6 CLN05. GO3 Revision 14

151 ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System B-K B110.10 Welded Attachments (Pressure Vessels) Surface or AAA: I Welded Attachments Volumetric for Vessels, Piping, B 10.20 Welded Attachments (Piping) Surface PFW: 4 Pumps, and Valves PHP: 3 PLP: 3 PMS: 8 PRH: 9 PRI: 7 PRR: 2 PRT: 1 PSC: I B110.30 Welded Attachments (Pumps) Surface TRA: 1 TRB: I Alion Science & Technology 7-7 CLN05. G03 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System _ _ _ _

B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 TRA: 1 Pump Casings TRB: 1 B-M-2 B12.50 Valve Bodies (Exceeding NPS 4) (Valves) Visual, VT-3 PFW: 6 Valve Bodies PHP: 3 PLP: 3 PMS: 24 PRH: 12 PRI: 3 PRR: 6 PRT: 3 B-N-nr Interior of Reactor Vessel Bo3.10 IVessel interior (Reactor Vessel) Visual, V VT-3 V I AAK:

AI T B-N-2 B13.20 Interior Attachments Within Beltline Region (Reactor Visual, VT-I AAK: 13 Welded Core Vessel)

Support Structures and Interior B 13.30 Interior Attachments Beyond Beltline Region (Reactor Visual, VT-3 AAK: 9 Attachments to Vessel)

Reactor Vessels B 13.40 Core Support Structure (Reactor Vessel) Visual, VT-3 AAK: 4 B-O B14.10 Welds in CRD Housing (Reactor Vessel) Volumetric or AAD: 2 14 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings I S Alion Science & Technology 7-8 CLN05. G03 Revision 14

1~>

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

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 FW 13T-01 All Pressure HP 13T-02 Retaining Components is LP MS NB RH RI RPV RR RT SC Alion Science & Technology 7-9 CLN05.G03 Revision 14

IS! Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) _ System C-A C 1.10 Shell Circumferential Welds (Pressure Vessels) Volumetric AAA: 1 Pressure Retaining Welds AAB: I in Pressure Vessels C1.20 Head Circumferential Welds (Pressure Vessels) Volumetric AAA: I AAB: I C-B C2.21 Nozzle-to-Shell (Nozzle to Head or Nozzle to Nozzle) Volumetric & AAA: 2 Pressure Retaining Welds Without Reinforcing Plate, Greater Than 1/2" Surface AAB: 2 Nozzle Welds in _ Nominal Thickness (Pressure Vessels)

Vessels C2.22 Nozzle Inside Radius Section Without Reinforcing Plate, Volumetric AAA: 2 Greater Than 1/2" Nominal Thickness (Pressure Vessels) AAB: 2 C-C C3.10 Welded Attachments (Pressure Vessels) Surface AAA: 2 Welded Attachments AAB: 2 for Vessels, Piping, C3.20 Welded Attachments (Piping) Surface PHP: 6 Pumps, and Valves PLP: 5 PRH: 32 PRI: 9 PSD: 2 C3.30 Welded Attachments (Pumps) Surface TRI: I Alion Science & Technology 7-10 CLN05. GO3 Revision 14

(~N ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System C-G C6.10 Pump Casing Welds (Pumps) Surface TAA: 7 13R-05 Pressure Retaining Welds TAB: 7 in Pumps and Valves TAC: 7 THP: 7 TLP: 7 TRI: 4 C-H C7.10 System Leakage Test (IWC-5220) Visual, VT-2 CC 13R-03 All Pressure CY 13R-04 Retaining Components FC 13T-01 FP 13T-02 FW HG HP IS LP MS RH RI SA SC SD SF SM VP VQ VR wo Alion Science & Technology 7-11 CLN05. GO3 Revision 14

IS! ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ f Notes (with Examination Category Number Requirements Components by TAP Number Description) System D-A D1.20 Welded Attachments (Piping) Visual, VT-I SX: 32 Welded Attachments for Vessels, Piping, Pumps, and Valves ..

13-13 D2.1I0 System Lea*kage Test (IWYD-522 1) Visual, VT-2 " cc DIR-04 All Pressure FC 13T-01 Retaining Components PV 13T-02 SX VC Alion Science & Technology 7-12 CLNO5.G03 Revision 14

1Sf Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components TAP Number Description)

E-A 1E1.11 Containment Vessel Pressure Retaining Boundary - General Visual 226 Containment Surfaces Accessible Surface Areas El.I I Containment Vessel Pressure Retaining Boundary- Visual, VT-3 4 6 Bolted Connections, Surfaces El.12 Containment Vessel Pressure Retaining Boundary- Visual, VT-3 13 7 Wetted Surfaces of Submerged Areas E 1.20 Containment Vessel Pressure Retaining Boundary- Visual, VT-3 1 7 BWR Vent System Accessible Surface Areas E-C E4.11 Containment Surface Areas- Visible Surfaces Visual, VT-1 0 8 Containment Surfaces Requiring E4.12 Containment Surface Areas- Surface Area Grid Ultrasonic 0 9 Augmented Examination Minimum Wall Thickness Locations Thickness Alion Science & Technology 7-13 CLN05. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System F-A I F1.10 Class 1 Piping Supports Visual, VT-3 FW: 32 Supports IHP: 6 LP: 9 MS: 42 RH: 45 RI: 31 RR: 32 RT: 34 SC: 26 F 1.20 Class 2 Piping Supports Visual, VT-3 HP: 67 LP: 43 MS: 4 RH: 348 RI: 74 SD: 21 F1.30 Class 3 Piping Supports Visual, VT-3 SX: 243 F 1.40 Supports Other Than Piping Supports Visual, VT-3 HP: 1 I (Class 1, 2, and 3) LP: 1 RH: 5 RI: 1 RPV: 1 RR: 12 Alion Science & Technology 7-14 CLN05.G03 Revision 14

1Sf ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components TAP Number Description) [_

L-A L. 1I1 Concrete Surfaces - General Visual 9 Concrete Surfaces All Accessible Surface Areas L 1.12 Concrete Surfaces - Detailed Visual --

Suspect Areas (No Suspect Areas Identified)

Alion Science & Technology 7-15 CLN05.GO3 Revision 14

ISI Program Plan Clinton Power Station Unit]1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Risk Description Exam Total Number of Relief Request/ Notes (with Examination Category Category Requirements Components by TAP Number Description) Number System R-A 2 Risk Category 2 Elements See Notes PHP: 5 13R-01 2 Risk-Informed Piping PLP: 2 3 Examinations PRH: 9 4 PRI: 1 5 3 Risk Category 3 Elements See Notes PFW: 67 13R-0 1 2 PRH: 33 3 4

5 4 Risk Category 4 Elements See Notes PHP: 8 13R-01 2 PLP: 3 3 PMS: 4 4 PRH: 80 5 PRI: I PRR: 133 PSC: 12 5 Risk Category 5 Elements See Notes AAP: 1 13R-01 2 PFW: 4 3 PLP: 3 4 PRH: 12 5 PRI: 64 Alion Science & Technology 7-16 CLN05.G03 Revision 14

I1S Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Aug Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System NA 2.2.1 NRC MEB Technical Position 3-1, Examination of High Volumetric or NA 5 Augmented Energy Circumferential and Longitudinal Piping Welds Surface Components USAR MEB [Draft Safety Evaluation Report (DSER)] Item No. 11 for Class 1 and USAR Section 6.6.8 for Class 2 (Pieviously 1.2-F.1 & 1.2-F.2) 2.2.2 CPS USAR Section 6.6.9 - Volumetric Examination of 10% Volumetric NA 13 of Thin Wall Class 2 RHR System Piping Welds Which Would Require Only Surface Examinations per ASME Section XI (Previously 1.2-G) 2.2.3 BWR Feedwater Nozzle and Control Rod Drive Return Volumetric AAH: 4 Line Nozzle Cracking Components (BWROG and AAK: 2 NUREG-0619) (Previously 1.2-H) PFW: 4 2.2.4 Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category D: 42 I3R-01 4 Austenitic Stainless Steel Piping Components, TR- 113932, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75)", and TR-1012621, "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A)"

(Previously 1.2-I). Evaluation and Repair for Any Cracks Detected on Piping Susceptible to IGSCC (Previously 1.2-J) 2.2.5 RPV Nozzle-To-Safe End Weld (GE SIL No. 455) Volumetric PFW: 4 (Previously 1.2-K) PHP: 2 PLP: 2 PRH: 6 PRR: 26 Alion Science & Technology 7-17 CLN05. GO3 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-1 INSERVICE INSPECTION

SUMMARY

Examination Category Aug Description Exam Total Number of Relief Request/ Notes (with Examination Category Number Requirements Components by TAP Number Description) System Revision 1.150, Vessel 1, Appendix A, Volumetric - AAA: 5 NA Augmented 2.2.6 Ultrasonic TestihigGuide NRC Regulatory of Reactor Welds Dur~ng AAB: I I Components Preservice and Inservice Examination (Previously 1.2-N) AAC: 4 (Continued) AAG: 29 AAH: 29 AAJ: 2 AAK: 2 AAL: 2 AAM: 6 AAN: 2 AAO: 2 AAP: 4 PFW: 4 PHP: 2 PLP: 2 PMS: 4 PRH: 6 PRR: 26 2.2.7 NUREG 0803 Generic Safety Evaluation Report Regarding Volumetric NA Integrity of BWR Scram System Piping, Section 5.1, page 5-3 Requires Inspection of Scram Discharge Volume Piping in Accordance With ASME Section XI (Previously 1.2-0) 2.2.8 Reactor Pressure Vessel Shell Welds Volumetric AAA: 4 4215 (10CFR50.55a(g)(6)(ii)(A), Final Rule)

Alion Science & Technology 7-18 CLN05. GO3 Revision 14

151 Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.1-2 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 1 ISI snubber visual examinations and functional testing are performed in accordance with the ASME OM Code, Subsection ISTD Program. The number ofCPS supports identified, include snubbers for the visual examination and functional testing of the integral and nonintegral attachments per Paragraphs IWF-5200(c),

IWF-5300(c), and IWF-2500(a). The snubbers are scheduled and administratively tracked in the ISI Program; however, the ASME OMCode, Subsection ISTD Program will be the mechanism for actually performing the visual examinationsand functional testing scheduled within the ISI Program. For a detailed discussion of the snubber program, see Section 4.2.

For the Third Inspection Interval, CPS's ISI 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-1 12657, Rev. B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1. The RISI Program scope has been implemented as an alternative to the 2004 Edition, No Addenda of the ASME Section XI examination program for ISI Class 1 B-F and B-J welds and ISI Class 2 C-F-i and C-F-2 welds in accordance with 10CFR50.55a(a)(3)(i).

3 Per the EPRI Topical Reports TR-1 12657, Rev. B-A, TR-1006937, Rev. 0-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, whileIGSCC Category A welds have been subsumed into the RISI Program. (CPS currently has only IGSCC Category D welds.)

4 Examination requirements within the RISI Program are determined by the various degradation mechanisms present at each individual piping structural element. See EPRI Topical Reports TR-l 12657, Rev. B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1 for specific examination method requirements.

5 For the Third Inspection Interval, the RISI program scope has been expanded to include welds in the BER piping, also referred to as the HELB region. All BER augmented welds have been evaluated under the RISI methodology and have been integrated into the RISI Program under the IOCFR50.59 change process. Additional guidance for adaptation of the RISI evaluation process to BER piping is given in EPRI TR-1006937 Rev. 0-A. Thus, these welds have been categorized and selected for examination in accordance with the EPRI Topical Reports TR- 12657, Rev. B-A, TR-1006937, Rev. 0-A, and Code Case N-578-1 in lieu of the original commitment toNRC MEB 3-1 detailed under USAR MEB [Draft Safety Evaluation Report (DSER)] Item No. II for Class 1 and USAR Section 6.6.8 for Class 2.

6 Bolted connections examined per ItemNumber El. 11require a General Visual examination each period anda VT-3 visual examination once per interval and each time the connection is disassembled during a scheduled Item Number E 1.11 examination. Additionally, a VT-I visual examination shall be performed 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).

7 Item Numbers E1.12 and E1.20 require VT-3 visual examination in lieu of General Visual examination, as modified by 1(CFR50.55a(b)(2)(ix)(G).

8 Item Number E4.11 requires VT-I visual examination in lieu of Detailed Visual examination, as modified by 10CFR50.55a(b)(2)(ix)(G).

9 The ultrasonic examination acceptance standard specified in Paragraph IWE-3511.3 for CISI Class MC pressure-retaining components must also be applied to metallic liners of CISI Class CC pressure-retaining components, as modified by 10CFR50.55a(b)(2)(ix)(I).

10 Per 10CFR50.55a(b)(2)(xxi)(B), Table IWB-2500-1 examinationrequirements,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) of this section.

11 As allowed by Code Case N-613-1, CPS will perform a volumetric examination using a reduced examination volume (AB-C-D-E-F-G-H) of Figures 1, 2, and 3 of the Code Case in lieu of the previous examination volumes of ASME Section XI, Figures IWBr2500-7(a), (b), and (c).

Alion Science & Technology 7-19 CLN05. GO3 Revision 14

IS1 ProgramPlan Clinton Power Station Unit 1, ThirdInterval TABLE 7.1-2 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 12 Examination Category B-G-1, Item Numbers B6.20 "Closure Studs, In Place" and B6.30 "Closure Studs, When Removed" have beencombined into and renamed as Item Number B6.20 "Closure Studs", in Table IWB-2500-1 of ASME Section XI, 2004 Edition, No Addenda. Therefore, one B-G-l, B6.20 component will represent the sixty-four RPV closure studs. For tracking purposes, this component also includes the five cattle chute studs for CPS which are routinely removed each refueling outage 13 These thin wall welds > 3/8" that were included for volumetric examination under ASME Section XI rules remain in the RISI element selection scope that has been risk evaluated and is potentially subject to RISI examination at CPS.

14 Examination Category B-O (Pressure-Retaining Welds In Control Rod Housings), Item Number B14.10 (Welds in CRD Housing)- the scope of examination is for pressure retaining welds in 10% of the peripheral CRD Housings. A total of 34out of 145 CRD Housings are classified as peripheral components and each has 2 welds (lower and upper housing welds). CPS has selected the welds on 4 CRD Housings (two welds per housing) to be examined during the interval (10% of 34).

Alion Science & Technology 7-20 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 7.2 Snubber Inspection Summary Tables 10CFR50.55a "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 examination methods described in Paragraph IWA-2213.

The following Table 7.2-1 provides a summary of the ASME OM Code, Subsection ISTD, Snubber visual examinations and functional testing for the Third ISI Interval at CPS.

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

ASME OM Code Subsection OM Article Article Number II Exam Totals Frequency Relief Request/

(with Subsection Number Description Requirements TAP Number Description)

(1) (2) (3) (4) (5) I (6) (7) _1(8)_J (1) ASME OM Code Subsection:

Provides the applicable Code for Operation and Maintenance of Nuclear Power Plants (OM) subsection number and a description as obtained from ISTD. Only applicable subsections to CPS are identified.

(2) OM Article Number:

Provides the article number as identified in ISTD. Only those article numbers applicable to CPS are identified.

(3) Article Number

Description:

Provides the article description as identified in ISTD. Identifies the methods selected to be performed at CPS.

(4) Examination Requirements:

Provides the visual examination and functional testing methods required by ISTD.

(5) Totals:

Provides the total number of snubbers that pertain to that article of ISTD.

Note that the total number of snubbers are subject to change after completion of plant modifications and design changes.

Alion Science & Technology 7-21 CLN05. G03 Revision 14

1SI ProgramPlan Clinton PowerStation Unit 1, Third Interval (6) Frequency:

Provides the frequency for visual examinations and functional testing as addressed in ISTD and approved ISTD Code Cases.

(7) Relief Request/TAP Number:

Provides a listing of Relief Request/TAP Numbers to specific snubber components. Relief requests and TAP Numbers that generically apply to all components, or an entire class are not listed. If a Relief Request/TAP Number is identified, see the corresponding relief request in Section 8.0 or the TAP Number in Section 2.5.

(8) Notes:

Provides a listing of program notes applicable to the ISTD article number. If a program note number is identified, see the corresponding program note in Table 7.2-2.

Alion Science & TechA ology 7-22 CLN05. G03 Revision 14

ISI Program Plan Clinton Power Station Unit 1, Third Interval TABLE 7.2-1 SNUBBER INSPECTION

SUMMARY

ASME OM Code Subsection OM Article Article Number Exam T Relief Request/

(with Subsection Number Description Requirements TAP Number Description)

IST1.'I ISTD-4200 Accessible and Inaccessible Snubbers (1 population) Visual, VT-3 567 Once every 10 1 Snubber Years Examinations ISTD ISTD-5200 10% Functional Test Plan - Functional Testing 91 Every Outage 2 Snubber Type 1 Snubbers (PSA-1/4, PSA-1/2)

Testing 10% Functional Test Plan - Functional Testing 317 Every Outage 2 Type 2 Snubbers (PSA-1, PSA-3, PSA-10) 10% Functional Test Plan - Functional Testing 117 Every Outage 2 Type 3 Snubbers (PSA-35, PSA-100) 10% Functional Test Plan - Functional Testing 4 Every Outage 2 Type 4 Snubbers (PSB-0.05) 10% Functional Test Plan - Functional Testing 38 Every Outage 2 Type 6 Snubbers (E-System 30, 50, and 70 Series) I I III Alion Science & Technology 7-23 CLN05. G03 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval TABLE 7.2-2 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Suinmaq I Examinations performed per Code Case OMN-13, "Requirements for Extendkig Snubber Inservice Visual Examination Interval at LWR Power Plants".

2 Per ISTD 2004 Edition, No Addenda, Article ISTD-5240 "Test Frequency".

Alion Science & Technology 7-24 CLN05.G03 Revision 14

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests written 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 compliance with ASME Section XI requirements results in a hardship or an unusual difficulty without a compensating increase in the level of quality and safety; and per IOCFR50.55a(g)(5)(iii) for situations where ASME Section XI requirements are considered impractical.

The following NRC guidance was utilized to determine the correct 10CFR50.55a paragraph citing for CPS relief requests. 10CFR50.55a(a)(3)(i) and IOCFR50.55a(a)(3)(ii) provide alternatives to the requirements of ASME Section XI, while 10CFR50.55a(g)(5)(iii) recognizes situational impracticalities.

10CFR50.55a(a)(3)(i): Cited in relief requests when alternatives to the ASME Section XI requirements which provide an acceptable level of quality and safety are proposed. Examples are relief requests which propose alternative NDE methods and/or examination frequency.

10CFR50.55a(a)(3 )(fi): Cited in relief requests when compliance with the ASME Section XI requirements is deemed to be a hardship or unusual difficulty without a compensating increase in the level of quality and safety. Examples of hardship and/or unusual difficulty include, but are not limited to, excessive radiation exposure, disassembly of components solely to provide access for examinations, and development of sophisticated tooling that would result in only minimal increases in examination coverage.

10CFR50.55a(g)(5)(iii): Cited in relief requests when conformance with ASME Section XI requirements is deemed impractical. Examples of impractical requirements are situations where the component would have to be redesigned, or replaced to enable the required inspection to be performed.

An index for CPS relief requests is included in Table 8.0-1. The "13R-XX" relief requests are applicable to ISI, CISI, SPT, and PDI.

The following relief requests are subject to change throughout the inspection interval.

Alion Science & Technology 8-1 CLNO5. G03 Revision 14

ISI Program Plan

_____________Clinton Power Station Unit 1, Third Interval TABLE 8.0-1 RELIEF REQUEST INDEX Relief Revision Status 2 (Program) Description/

Request Date3 Approval Summary' (ISI) Alternate Risk-Informed Selection and 0 Examination Criteria for Examination Category 05/21/10 t B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds. Revision 0 Submitted.

0 (ISI) Alternative Requirements for Nozzle-To-13R-02 Submitted Vessel Weld and Inner Radius Examinations.

05/2 1/10 Revision 0 Submitted.

0 (SPT) Pressure Testing the RPV Head Flange 13R-03 Submitted Seal Leak Detection System. Revision 0 O5/2 1/10 Submitted.

(SPT) Alternative to Performance of System Pressure Tests and VT-2 Visual Examination 0 Requirements for all ISI Class 2 Instrument Air 13R-04 Submitted (IA) Piping and the ISI Class 3 IA Piping 05/21/10 Supplying, all SRV's, and both Feedwater Containment Outboard Isolation Check Valves.

Revision 0 Submitted.

(ISI) Examination of the ISI Class 2 High RS i Pressure Core Spray, Low Pressure Core Spray, 05/21/10 and Residual Heat Removal Pump Casing Welds.

Revision 0 Submitted.

(ISI) Alternative Volumetric Examination of RPV Circumferential Shell Welds. Permanent Relief Request (Second ISI Interval Relief Request 4215) for deferral of the RPV 0 circumferential shell weld examinations was 4215 Authorized 05/2 1/10auhrzdythNR authorized by the NRC per the SER dated peteSE dtd 12/30/09 and thus applies to the remaining term of operation under the existing, initial license, including this Third Inspection Interval.

Note 1: The NRC 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 NRC authorizes relief requests pursuant to 10CFR50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level of quality and safcty or under 10CFR50.55a(3)(ii) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety.

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ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval Note 2: This colunm represents the status of the latest revision. Relief Request Status Options: Authorized-Approved for use in an NRC SER (See Note 1); Granted - Approved for use in an NRC SER (See Note 1);

Authorized Conditionally - Approved for use in an NRC SER which imposes certain conditions; Denied -

Use denied in an NRC SER; Expired - Approval for relief has expired; Withdrawn - Relief has been withdrawn by CPS; Not Required - The NRC has deemed the relief unnecessary in an SER or RAI; Cancelled - Relief has been cancelled by CPS prior to issue; Submitted - Relief has been submitted to the NRC by the station and is awaiting approval; Pending- Relief has been awaiting station and Corporate review and submittal to the NRC.

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 notel in the second column is the date of the ISI Program Plan revision when the relief request was incorporated into the document.

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1Sf Program Plan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-01 Revision 0 (Page 1 of 7)

Request for Relief for Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-i, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 1 and 2

Reference:

Table IWB-2500-1, Table IWC-2500-1 Examination Category: B-F, B-J, C-F-1, and C-F-2 Item Number: B5.10, B5.20, B9.1 1, B9.21, B9.31, B9.32, B9.40, C5.1 1, C5.51, and C5.81

Description:

Alternate Risk-Informed Selection and Examination Criteria for Examination Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Component Number: Pressure Retaining Piping 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at Clinton Power Station (CPS) is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition.

3.0 APPLICABLE CODE REOUIREMENT:

Table IWB-2500-1, Examination Category B-F, requires volumetric and surface examinations on all welds for Item Number B5.10 and surface examinations for all welds for Item Number B5.20.

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

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

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

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a. primary plus secondary stress intensity range of 2 .4 Sm, for ferritic steel and austenitic steel.

b. cumulative usage factor U of 0.4.

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

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

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

I1. Welds selected for examination shall include 7.5%, but not less than 28 welds, of all dissimilar metal, austenitic stainless steel and high alloy welds (Examination Category C-F-1) or of all carbon and low alloy steel welds (Examination Category C-F-2) not exempted by Paragraph IWC-1220. (Some welds not exempted by Paragraph IWC- 1220 are not required to be nondestructively examined per Examination Categories C-F-I and C-F-2. These welds, however, shall be included in the total weld count to which the 7.5% sampling rate is applied.) The examinations shall be distributed as follows:

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

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

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

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1SI ProgramPlan Clinton Power Station Unit 1, ThirdInterval 10CFR50.55a RELIEF REQUEST: 13R-01 Revision 0 (Page 3 of 7) 4.0 REASON FOR REQUEST:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative utilizing Reference I along with two enhancements from Reference 4 will provide an acceptable level of quality and safety.

As stated in "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-l 12657, Revision B, July 1999)" (i.e.,

Reference 2):

"The staff concludes that the proposedRISI Programas described in EPRI TR-11265 7, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to I OCFR50.55afor the proposedalternative to the piping ISI requirements with regardto the number of locations, locationsof inspections, and methods of inspection."

The initial CPS Risk-Informed Inservice Inspection (RISI) Program was submitted during the First Period of the Second Inspection Interval. This initial RISI Program was developed in accordance with EPRI TR-1 12657, Revision B-A, as supplemented by Code Case N-578-1. The program was approved for use by the NRC via a Safety Evaluation as transmitted to Exelon (Reference 5).

The transitioa from the 1989 Edition to the 2004 Edition of ASME Section XI for CPS's Third Inspection Interval does not impact the currently approved Risk-Informed ISI evaluation methods and process used in the Second Inspection Interval, and the requirements of the new Code Edition/Addenda will be implemented as detailed in the CPS ISI Program Plan.

The Risk Impact Assessment completed as part of the original baseline RISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI program to the new RISI methodology. For the Third Interval ISI update, there is no transition occurring between two different methodologies, but rather, the currently approved RISI methodology and evaluation will be maintained for the new interval. The original methodology of the evaluation has not changed, and the change in risk was simply re-assessed using the initial 1989 ASME Section XI program prior to RISI and the new element selection for the Third Interval RISI Program. This same process has been maintained in each revision to the CPS RISI assessment that has been performed to. date.

The actual "'valuationand ranking" procedure including the Consequence Evaluation and Degradation Mec.hanism Assessment processes of the currently approved (Reference Alion Science & Tech1 uAlogy CLN05. GO3

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5) RISI Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-1 12657, Revision B-A.

These portions of the RISI Program have been and will continue to be reevaluated and revised as major revisions of the site Probabilistic Risk Assessment (PRA) occur and modifications to phlat configuration are made. The Consequence Evaluation, Degradation Mechanism Assessment, Risk Ranking, Element Selection, and Risk Impact Assessment steps encompass the complete living program process applied under the CPS RISI Program.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

The proposed alternative originally implemented in the risk informed in-service inspection plan for CPS (Reference 3), along with the two enhancements noted below, provide an acceptable level of quality and safety as required by 10CFR50.55a(a)(3)(i).

This original program along with these same two enhancements is currently approved for CPS's Second Inspection Interval as documented in Reference 5.

The Third Inspection 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 methodology as currently implemented under EPRI TR-1 12657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

a. In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RISI Selected Examinations" of EPRI TR-1 12657, CPS will utilize the requirements of Paragraph -2430, "Additional Examinations" contained in Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in Code Case N-578-1 provide a more refined methodology for implementing necessary additional examinations. The reason for this selection is that the guidance discussed in EPRI TR-1 12657 includes requirements for additional examinations at a high level, based on service conditions, degradation mechanisms, and the performance of evaluations to determine the scope of additional examinations, 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, similar to the current requirements, of ASME Section XI, CPS intends to perform additional examinations that are required due to the identification of flaws or relevant conditions exceeding the acceptance standards, during the outage the flaws are identified.

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b. TQ supplement the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of EPRI TR- 112657, CPS will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations" contained in Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 2004 Edition of ASME Section XI (CPS's Code of record for the Third 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 Examination Method and Categorization of Parts to be Examined. Based on these Methods and Categorization, the examination figures specified in Section 4 of EPRI TR- 112657 will then be used to determine the examination volume based on the degradation mechanism and component configuration.

CPS uses UT techniques for RISI volumetric examinations.

For the components addressed by the RISI Program, ASME Section XI focuses primarily on weld examinations. Risk Informed examination volumes also include portions of piping and fitting base materials that are susceptible to particular degradation mechanisms.

The ASME Section XI, Mandatory Appendix I, "Ultrasonic Examinations," specifies that UT examinalion procedures, equipment, and personnel used to detect and size flaws in piping welds shall be qualified by performance demonstration in accordance with ASME Section XI Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems." The RISI Program complies with Appendix VIII for weld examinations. In cases where the examination requirements cannot be met, CPS will submit a request for relief in accordance with IOCFR50.55a, "Codes and standards."

The examination methods are designed to be effective for specific degradation mechanisms and examination locations. The volumetric scanning will be in both axial and circumferential directions to detect the flaws in these orientations.

Additionally, all CPS dissimilar metals (DM) 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-exempt piping. The piping segments containing the DM welds were classified into the appropriate RISI categories, and appropriate elements were selected per the category requirements for examination during the Second Inspection Interval.

Piping welds, including DM welds in vessel nozzles, that are susceptible to IGSCC (i.e.,

IGSCC Categories B through G, as applicable) and not subject to other degradation Alion Scieave & Technilogy CLNO5. G03

IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-01 Revision 0 (Page 6 of 7) mechanism(s) are removed from the RISI Program population. They are contained in the CPS Intergranular Stress Corrosion Cracking (IGSCC) Augmented Inspection Program (2.2.4) and are subject to the inspection requirements of BWRVIP-75-A "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules". Furthermore, 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 Program.

The CPS RISI Program, as developed in accordance with EPRI TR-1 12657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as "High" risk (i.e.,

Risk Categ6iy 1, 2, and 3) and 10% of the elements that are categorized as "Medium" risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI TR- 112657 while the guidance for the examination method and categorization of parts to be examined are provided by the EPRI TR-1 12657 as supplemented by Code Case N-578-1.

For Staff consideration in the evaluation of this alternative Risk-Informed ISI Program, Enclosure 1 to the relief request contains a summary of the Regulatory Guide 1.200, Revision 1, evaluation performed on CPS Quantification Notebook, CPS-PSA-014, Revision 4, March 2007 (Model 2006C) and the impact of the identified gaps on the technical adequacy of the CPS PRA Model to support this RISI application.

In addition to this risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to receive Code lequired pressure testing as part of the current ASME Section XI program.

VT-2 visual examinations are scheduled in accordance with the CPS Pressure Testing Program, which remains unaffected by the RISI Program.

6.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Ten-Year Inspection Interval for CPS.

7.0 PRECEDENTS

Similar relief requests have been approved for:

CPS Second Inspection Interval Relief Request 4208 was authorized per SER dated April 8, 2002. The Third Inspection Interval Relief Request utilizes an identical RISI methodology as was previously approved.

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Peach Bottom Atomic Power Station Fourth Inspection Interval Relief Request 14R-44 was authorized per SER dated February 26, 2009.

8.0 REFERENCES

I. Electric Power Research Institute (EPRI) Topical Report (TR) 112657 Rev. B-A, "Revised Risk-Informed Inservice Inspection Evaluation Procedure," December 1999.

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

3. Initial Risk-Informed Inservice Inspection Evaluation, Revision 0 - Clinton Power Station, dated October 15, 2001 (Letter RS-01-219 from K. A. Ainger (Amergen) to the NRC, Clinton Power Station Second Interval Inservice Inspection Program

- Relief Request 4208, "Alternative to the ASME Boiler and Pressure Vessel Codi Section XI Requirements for Class I and 2 Piping Welds Risk-Informed Inservice Inspection Program," dated October 15, 2001.)

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

5. Lettu*f from A. J. Mendiola, (NRC) to J. L. Skolds (Exelon) "Clinton Power Station, Unit 1 - Risk-Informed Inservice Inspection Program, Relief Request 4208 (TAC No. MB5321 1)," dated April 8, 2002.

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IS! ProgramPlan Clinton Power Station Unit 1, ThirdInterval ENCLOSURE 1 CPS 2006A PRA TECHNICAL CAPABILITY ASSESSMENT FOR 10CFR50.55a RELIEF REQUEST: 13R-01 Revision 0 (Page 1 of 17)

Summary Statement of CPS PRA Model Capability for Use in Risk-Informed Licensing Actions Introduction Exelon Generation Company (EGC) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the PRA models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update pcocess, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the CPS PRA.

PRA Maintenance and Update The EGC risk management process ensures that the applicable PRA model remains an accurate reflection of the as-built and as-operated plants. This process is defined in the EGC Risk Management progiam, which consists of a governing procedure (ER-AA-600, "Risk Management") and subordinate implementation procedures. EGC procedure ER-AA-600-1015, "FPIE PRA Model Update" delineates the responsibilities and guidelines for updating the full power internal events PRA models at all operating EGC nuclear generation sites. The overall EGC Risk Management program, including ER-AA-600-1015, defines the process for implementing regularly scheduled and interim PRA model updates, for tracking issues identified as potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitations identified in the model, industry operating experience), and for controlling the model and associated computer files. To ensure that the current PRA model remains an accurate reflection of the as-built, as-operated plants, the following activities are routinely performed:

" Design Lchanges and procedure changes are reviewed for their impact on the PRA model.

• New engineering calculations and revisions to existing calculations are reviewed for their impact on the PRA model.

• Maintenince unavailabilities are captured, and their impact on CDF is trended.

• Plant specific initiating event frequencies, failure rates, and maintenance unavailabilities for equipment that can have a significant impact on the PRA model are updated approximately every four years.

In addition to these activities, EGC risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance includes:

• Documentation of the PRA model, PRA products, and bases documents.

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" The approach for controlling electronic storage of Risk Management (RM) products including PRA update information, PRA models, and PRA applications.

• Guidelines for updating the full power, internal events PRA models for EGC nuclear generation sites.

" Guidance for use of quantitative and qualitative risk models in support of the On-Line Work Control Process Program for risk evaluations for maintenance tasks (corrective maintenance, preventive maintenance, minor maintenance, surveillance tests and modifications) on systems, structures, and components (SSCs) within the scope of the Maintenance Rule (10CFR50.65(a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant. The most recent update of the CPS PRA model (designated the 2006C model) was completed in March 2007.

PRA Self Assessment and Peer Review Several assessments of technical capability have been made and continue to be planned for the CPS PRA model. A chronological list of the assessments performed includes the following:

• An independent PRA peer review was conducted under the auspices of the BWR Owners Group (BWROG) in 2000, following the Industry PRA Peer Review process [1]. This peer review included an assessment of the PRA model maintenance and update process.

• A self-assessment analysis was previously performed against Addenda B of the ASME PRA Standard (ASME RA-Sb-2005, [4]) and the draft of Revision 1 Regulatory Guide 1.200 (DG-1 161) to support scoping/planning for the CPS PRA 2006 update project.

• During 2005 and 2006 the CPS PRA model results were evaluated in the BWROG PRA cross-comparisons study performed in support of implementation of the mitigating systems performance indicator (MSPI) process.

• The CPS 2006 PRA self-assessment was revised in 3Q09 to address consistency with Regulatory Guide 1.200 Revision 1 [6] in preparation for the CPS 2009 PRA peer review.

• A current industry peer review of the CPS PRA is scheduled for the fourth quarter of 2009.

A summary of the disposition of the BWROG PRA Peer Review facts and observations (F&Os) for the CPS PRA models was documented as part of the statement of PRA capability for MSPI.

All of the significance level "A" F&Os have been resolved and ninety of the ninety-two significance level "B" F&Os have been resolved. The remaining 2 open significance level "B" F&Os are not significant for the current model, as noted in Table 1.

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A Self-Assessment of the 2003 CPS PRA was performed in support of the CPS 2006 PRA Update. This Gap Analysis was performed using Addenda B of the ASME PRA Standard (ASME RA-Sb-2005) and the draft of Revision 1 Regulatory Guide 1.200 (DG-1 161). Potential gaps to Capability Category II of the Standard were identified and used to plan the CPS 2006 PRA Update.

The CPS 2006 PRA self-assessment was revised in 3Q09 in preparation for the CPS 2009 peer review to address consistency with Regulatory Guide 1.200 Revision 1 [6], including the NRC positions stated in Appendix A of [6] and the clarifications in [5], and to identify which gaps were closed folloving completion of the CPS 2006 PRA update and the CPS PRA 2009 internal flooding update. Identified gaps have been considered with particular focus on technical elements important to the FHSI relief request.

A summary of this assessment of the current open items, including the partially resolved items, relative to the RISI relief request is provided in attached Table 2. The remaining gaps, including any new items that may be identified in the planned industry peer review, will be reviewed for consideration during future model updates. The currently identified items are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. These items are or are being documented in the PRA Updating Requirements Evaluation (URE) database so that they can be tracked and their potential impacts accounted for in applications where appropriate. In addition, plant changes made since the last PRA update have been reviewed and determined to not have a significant PRA impact. These items are also documented in UREs for consideration in future PRA updates, as appropriate.

General Conclusion Regarding PRA Capability The CPS PRA maintenance and update processes and technical capability evaluations described above provide a robust basis for concluding that the PRA is suitable for use in risk-informed licensing actions. As specific risk-informed PRA applications are performed, remaining gaps to specific requirements in the PRA standard will be reviewed to determine which, if any, would merit application-specific sensitivity studies in the presentation of the application results.

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Assessment of IRA Capability Needed for Risk-Informed Inservice Inspection In the RISI Program at CPS, the EPRI RISI methodology [7] is used to define alternative inservice inspection requirements. Plant-specific PRA-derived risk significance information is used during the RISI plan development to support the consequence assessment, risk ranking, element selection and risk impact steps.

The importance of PRA consequence results, and therefore the scope of PRA technical capability, is tempered by three fundamental components of the EPRI methodology.

First, PRA consequence results are binned into one of three conditional core damage probability (CCDP) and conditional large early release probability (CLERP) ranges before any welds are chosen for RISI inspection as illustrated below. Broad ranges are used to define these bins so that the impact of uncertainty is minimized and only substantial PRA changes would be expected to have an impact on the consequence ranking results.

Consequence Results Binning Groups Consequence Category CCDP Range CLERP Range High CCDP > 1E-4 CLERP > 1E-5 Medimn 1E-6 < CCDP < 1E-4 IE-7 < CLERP < IE-5 Low CCDP < IE-6 CLERP < 1E-7 The risk importance of a weld is therefore not tied directly to a specific PRA result. Instead, it depends only on the range in which the PRA result falls. As a consequence, any PRA modeling uncertainties would be mitigated by the wide binning provided in the methodology.

Additionally, conservatism in the binning process (e.g., as would typically be introduced through PRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to result in a larger inspection population.

Secondly, the impacts of particular PRA consequence results are further dampened by the joint consideration of the weld failure potential via a non-PRA-dependent damage mechanism assessment. The results of the consequence assessment and the damage mechanism assessment are combined to determine the risk ranking of each pipe segment (and ultimately each element) according to the EPRI Risk Matrix. The Risk Matrix, which equally takes both assessments into consideration, is reproduced below.

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CONSEQUENCES OF PIPE RUPTURE POTENTIAL FOR IMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY PIPE RUPTURE AND LARGE EARLY RELEASE PROBABILITY PER DEGRADA'ION MECHANISM SCREENING CRITERIA NONE LOW MEDIUM HIGH 4 4 4 4.

HIGH LOW MEDIUM ~HIGH~ HIGH FLOW ACCELERATED CORROSION Category 7 Category 5 C~t.o -Category 1 MEi)IUM LOW LOW MEDIUM. . HIGH 01IHER DEGRADATION MECHANISMS Category 7 Category 6 Category 5 "a"egor' 2 LOW LOW LOW LOW MEDIUM, NO DEGRADATION MECHANISMS Category 7 Category 7 Category 6 Category 4..

Thirdly, the EPRI RISI methodology uses an absolute risk ranking approach. As such, conservatism in either the consequence assessment or the failure potential assessment will result in a larger inspectioii population rather than masking other important components. That is, providing more realism into the PRA model (e.g., by meeting higher capability categories) most likely would result in a smaller inspection population.

These three facets of the methodology reduce the importance and influence of PRA on the final list of candidate welds.

The limited manner of PRA involvement in the RISI process is also reflected in the risk-informed license application guidance provided in Regulatory Guide 1.174 [8].

Section 2.2.6 of Regulatory Guide 1.174 provides the following insight into PRA capability requirements for this type of application:

There are, however, some applicationsthat, because of the nature of the proposedchange, have a limited impact on risk, and this is reflected in the impact on the elements of the risk model.

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An example is risk-informed inservice inspection (RI-ISI). In this application,risk significance was used as one criterionfor selectingpipe segments to be periodicallyexaminedfor cracking. Duringthe staff review it became clear that a high level of emphasis on PRA technical acceptability was not necessary. Therefore, the staff review ofplant-specific RI-ISI typically will include only a limited scope review of PRA technical acceptability.

In addition to the above, it is noted that welds determined to be low risk significant are not eliminated from the ISI Program on the basis of risk information. For example, the risk significance of a weld may fall from Medium Risk Ranking to Low Risk Ranking, resulting in it not being a candidate for inspection. However, it remains in the program, and if, in the future, the assessment of its ranking changes (either by damage mechanism or PRA risk) then it may again become a candidate for inspection. If it is discovered during the RISI update process that a weld is now susceptible to flow-accelerated corrosion (FAC), inter-granular stress corrosion cracking (IGSCC), or microbiological induced cracking (MIC) in the absence of any other damage mechanism, then it is addressed in an "augmented" program where it is monitored for those special damage mechanisms. That occurs no matter what the Risk Ranking of the weld is determined to be.

Conclusion Regarding PRA Capability for Risk-Informed Inservice Inspection The CPS PRA model continues to be suitable for use in the RISI application. This conclusion is based on:

• PRA maintenance and update processes in place,

• PRA technical capability evaluations that have been performed and are being planned, and

• RISI process considerations, as noted above, that demonstrate the relatively limited sensitivity of the EPRI RISI process to PRA attribute capability beyond ASME PRA Standard Capability Category I.

In support of the PRA analyses for the CPS Third Ten-Year Inspection Interval evaluation using the CL06C PRA model, the remaining gaps to the PRA standard have been reviewed to determine which, if any, would merit RISI-specific sensitivity studies in the presentation of the application results. The result of this assessment concluded that no additional sensitivity studies are merited.

References

1. Boiling Water Reactors Owners' Group, BWROG PSA Peer Review Certification Implementation Guidelines,Revision 3, January 1997.

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2. American Society of Mechanical Engineers, Standardfor ProbabilisticRisk Assessment for Nuclear Power PlantApplications, ASME RA-S-2002, New York, New York, April 2002.

3. U.S. Nuclear Regulatory Commission, An Approachfor Determiningthe Technical Adequacy of ProbabilisticRisk Assessment Results for Risk-Informed Activities, Draft Regulatory Guide DG-1 122, November 2002.

4. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-Sb-2005, New York, New York, December 2005.

5. U.S. Nuclear Regulatory Commission Memorandum to Michael T. Lesar from Farouk Eltawila, "Notice of Clarification to Revision 1 of Regulatory Guide 1.200," for publication as a Federal Register Notice, July 27, 2007.

6. U.S. Nuclear Regulatory Commission, Regulatory Guide 1.200, Revision 1, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," January 2007

7. Revised Risk-Informed Inservice Inspection EvaluationProcedure,EPRI TR- 112657, Revision B-A, December 1999.

8. U.S. Nuclear Regulatory Commission, An Approachfor Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Regulatory Guide 1.174, Revision 1, November 2002.

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TABLE 1 IMPACT OF OPEN SIGNIFICANT PRA PEER REVIEW FINDINGS FOR THE CPS PRA MODEL Peer FACTS & OBSERVATIONS (F&Os)

Review Element ID Priority Summary Impact Assessment TH-8 TH-8-1 B Additional plant specific room heat-up calculations (or Non-Significant Impact Primarily a documentation enhancements to existing calculations) should be performed issue. The PRA already makes appropriate assumptions to support modeling assumptions regarding room-cooling regarding the need for room cooling in the appropriate requirements. Areas specifically identified are Control areas. No impact on RISI application.

Room, RCIC, LPCS, LPCI, and SWGR rooms.

HR-6 HR-6-1 B All pre-initiator HEPs in the CPS PSA model are based on Non-significant impact Pre-initiator HEPs contribute screening estimates. For post-initiator screening HEPs with approximately 2% of CDF). Fine-tuning the HEPs for RAWs greater than 1.1, the HEPs were re-evaluated with pre-initiators would be expected to reduce the relative more detailed calculations. For consistency sake, the pre- importance of these events. No significant impact on initiator HEP calculations should follow the same approach. RISI application.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II

~~........

...... ~~~ ~~~~~~~~~.

.,z OF THE ASME PRA STANDARD * . ...... " * ',.* " . *"* * . :* * ,>'* '

.t...DescriptioofGap.....

Applicable :SRs C urrent Status / Comment. Importance to RISI I Review initiating event precursors in identifying the IE-A7 Deferred: Explicit analysis of event No Impact: Documentation item.

initiating events to be modeled. precursors is judged not to provide significant insights to the CPS IE A rigorous explicit assessment of all the events in analysis, which includes initiating events NUREG-1275 has not been performed. known to be relevant to BWR-6 plants in general and CPS in particular. This type of activity is known to have been performed for another BWR plant (review of hundreds of events INPO SENs, SOERs, SERs, and NRC SECY letters on precursors) and no new initiating events were identified. It is expected that future industry studies will provide this generic assessment.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD tecriti oicablSks Current Status ,/Comment, . Imortance to RISi 2 Assumptions regarding loss of switchgear room IE-C4 Deferred: Switchgear room cooling Non-Significant Impact The PRA cooling should be supported by room cooling AS-B3 calculations have not been performed at already makes appropriate calculation. SC-B2 this time but are being considered for a assumptions regarding the need for Sc-c 1 future update. room cooling and explicitly models SC-C2 room cooling in certain areas.

SY-A 17 Modeling cooling failures for SY-A19 switchgear might make the SXpiping SY-A20 going to the SX cooler more SY-B7 important, but this is Class 3 piping, SY-B8 which is not in the scope of the RISI Program. SX failures already have high importance for DG cooling, ECCS room cooling and DHR, and more extensive Switchgear heat removal modeling would not change

_this.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD em .Descriptiono Gap Applcbe .SRsCurrent Statusb Com.ment ........... Imporance toRSI.

3 The following should be considered in the pre- HR-A1, HR-A2, Deferred: The CPS PRA includes over No Impact: This is primarily a initiator HEP evaluation: HR-A3, HR-C2, 100 pre-initiator HEPs in the model, and Documentation item.

1) A list of the PRA systems to consider for test HR-C3 the approach is believed to meet the

1) d maistnfthne Pacstes tintent of the identified SRs. Performing this task with a more rigorous review

2) Rules for identifying and screening test and and documentation of test and maintenance actions from the PRA maintenance procedures is judged not to

3) A list of procedures reviewed, the potential test have significant impact on the PRA and maintenance actions associated with the model and results. The current procedures, and the disposition of the action methodology and documentation for (screened or evaluated), identifying pre-initiator HEPs is judged

4) Identify T&M activities that require adequate PRA. Any to additional support applications of the documentation realignment of the system outside its normal e nhanemetwouldoturesultin o stad b staus.enhancement opertionl would not result in operational or stand by status. increasing the number of pre-initiator HEPs included in the model or significantly impact their relative

_importances.

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'IABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD

.t...Description of Gap, Appicable SRs Current Status I Comment J., mpo nce to RISI.

4 Pre-initiator HEPs in the CPS PRA model are based HR-B I, HR-B2, Deferred: Future updates of the CPS Non-significant impact Pre-initiator on screening estimates (URE 2001-084, peer review HR-D 1, HR-D2, PRA will consider explicit/specific pre- HEPs contribute approximately 2%

F&O HR-6-1), should not use screening values for HR-D3, HR-D4 initiator HEP calculations. The current of CDF). Fine-tuning the HEPs for dominant pre-initiator HEPs. calculations are based on representative pre-initiators would be expected to procedures/practices for similar pre- reduce the relative importance of initiator HEPs. The current estimates these events.

are generally higher error probabilities than would be obtained if various explicit recovery factors and testing frequencies were applied in specific 1-EP calculations for each pre-initiator.

The impact on the model is non-significant, pre-initiator HEPs contribute

_approximately 2% to the CL06C CDF.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD Item, ý,:Description ofýap- - Applcable SRs Current Status W Comment. ImprtanetoRIS i 5 Failure data development using surveillance test DA-C10 Deferred: The maintenance rule data is Non-significant impact Any data should fulfill the requirements of DA-C10, and used directly, but a adjustment to failure data counts should be documented appropriately. Review confirmation that the data are collected resulting from a rigorous review of surveillance test procedures and identify all failure exactly consistent with the requirements testing procedures is judged to have modes that are fully tested by the procedures. in the Standard has not been a non-significant impact on CDF and Include data for the failure modes that are fully performed. Future updates of the CPS LERF values.

tested. The results of unplanned demands on PRA will consider enhancement to the Not significant The model is equipment should also be accounted for. documentation and investigation of the reasonably consistent with data plant failure data implied by this SR. from the plant MR database, This is judged to have a minimal impact which is adequate for RISI on the unavailabilities and failure application.

probabilities used in the model.

6 As needed in maintenance unavailability DA-C12 Deferred: Future updates of the CPS Non-significant impact Any determination, perform interviews of maintenance PRA will consider performance of refinements to maintenance staff for equipment with incomplete or limited interviews of plant personnel to unavailabilities are judged to result maintenance information and document supplement maintenance unavailability in a negligible impact on CDF (i.e.,

appropriately, estimates for equipment with limited the dominant maintenance terms, by maintenance information. far, with respect to CDF are trains with good maintenance information-ECCS trains, RCIC, EDGs, SX).

The model is reasonably consistent with data from the plant MR database, which is adequate for RISI

_application.

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I'ABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD item Description of Gap ApplicableSSs Current statu*s /Comment . *.Iempol rtanceto .RII 7 The CPS internal flooding analysis and SC-A6, SY-A4, An internal flooding update to the CPS Non-significant impact Internal documentation should be updated to meet ASME IF Technical CL06c model has recently been flooding analyses do not impact RISI Standard expectations. Element completed and will be available in the calculations. For the RISI analysis

[This has been recently addressed - See Current short-term for use in future applications the Internal Flooding initiators are Status / Comment] of the PRA. not used to represent the consequences from flooding events.

Rather the impact of flooding from the RISI consequence analysis is evaluated by tagging appropriate basic events from the non-flooding portions of the Internal Events PRA model. Therefore the fact the RISI analysis does not use the results of the updated Internal Flooding analysis is not critical to the results

.... of the RISI analysis.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD o... . .:.,......1 .. *.... .. . ,.,:.:....:,..*..*..j*. *. ,.. .:...:...

tem, , .. ,Description of Gap :. .. Applicable s...n C rent sCom et :impornce to R s I:I 8 Document significant basic events that contribute to QU-D5a Deferred: This documentation aspect No Impact: Documentation item.

the significant initiating events whose frequencies has not been incorporated into the CPS Although the overall importance of are quantified using fault tree methods. PRA notebooks. Initiating event fault some basic events may not be trees are not linked into the accident directly obtained in the quantification sequence models. Documentation of the results, it is possible to estimate these importance of failures in initiating event importances. However, initiators fault trees in the base PRA notebooks is associated with this gap are not a documentation enhancement, directly used in the RISI analysis Documenting the relative importance of basic events to CDF and LERF for these fault tree based initiators has no bearing on the conditional core damage (and large early release) probability calculations used in the

_RISI analysis.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD te.....m  :.* D...

escriptionl of.Gap... A ble SRs Current' tatus .CornmentIp 9 The following enhancements to the documentation QU-F2 Deferred: These recommendations are No Impact: Documentation item.

of the CPS PRA should be considered to comply documentation enhancements for the with the documentation requirements in the base PRA and are maintained for Standard: consideration for future PRA updates.

• Provide a list of human actions and equipment failures (significant basic events) that cause accidents to be non-dominant.

• Bases for the elimination of mutually exclusive events from the model need to be added.

• Include cutsets segregated by accident sequence in the documentation.

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TABLE 2 STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD

,,!tem *,J.* , Description of Gap Applicable SRs CVu'reft Sttus/Comment .I.portan to.RISI 10 Several SRs associated with treatment of model QU-E1 The CPS 2006 PRA includes CDF and To be determined once the new uncertainty and related model assumptions have QU-E2 LERF parametric uncertainty analysis; NRC/EPRI guidance is implemented.

been recently redefined. NRC has issued [6] a QU-E4 consideration has been given to However, the EPRI RISI process is clarification to its endorsement of the PRA QU-F4 modeling uncertainty, however the defined such that model uncertainties Standard. NRC and EPRI are currently preparing IE-D3 approach used pre-dates NUREG- 1855. will not unduly influence results, guidance on an acceptable process for meeting AS-C3 It involves documenting how and, further, the current approach these requirements. SC-C3 assumptions for the technical elements provides appropriate insights into SY-C3 of a PRA can impact the risk results, and important modeling assumptions that HR-13 then from that performing selected may be pertinent to applications.

DA-E3 quantitative sensitivity studies. These IF-F3 recently redefined SRs will be addressed LE-G4 during a future PRA model update using a process consistent with NUREG-1855.

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Request for Relief for the Alternative to Nozzle to Vessel Weld and Inner Radius Examinations In Accordance with 10CFR50.55a(a)(3)(i) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: I Component Number: Nozzles N1, N2, N3, N5, N6, N7, N8, N9, N10, and N16 (See Enclosure 1 for specific nozzle identification numbers)

Examination Category: B-D Item Number: B3.90 and B3.100

Description:

Alternative to Nozzle to Vessel Weld and Inner Radius Examinations (IWB-2500, Table IWB-2500 Inspection Program B) 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at Clinton Power Station (CPS) is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition. Additionally, for ultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 2001 Edition is implemented as required (and modified) by 10CFR50.55a(b)(2)(xv) and 10CFR50.55a(b)(2)(xxiv).

3.0 APPLICABLE CODE REQUIREMENT:

Class I nozzle-to-vessel weld and nozzle inner radii examination requirements are given in Subsection IWB, Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzle in Vessels - Inspection Program B," Item Numbers B3.90 and B3.100, respectively. The method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles must be examined each interval. All of the nozzles identified in Enclosure 1 are full penetration welds.

4.0 REASON FOR REQUEST:

The identified ISI Class 1 nozzles are scheduled for examination for the upcoming inspection initerval at CPS. The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 25 Rem for the entire interval.

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Pursuant to 10CFR50.55a(a)(3)(i), relief is requested from performing the required examinations on 100% of the identified nozzles. Alternatively, in accordance with Code Case N-702 (Reference 2), CPS proposes to examine a minimum of 25% of the nozzle inner radii and nozzle-to-vessel welds, including at least one nozzle from each system and nominal pipe size. For each of the identified nozzles, both the inner radius and the nozzle-to-shell weld would be examined. As a minimum, the following nozzles would be selected for examination: one of the two 20" recirculation outlet nozzles (i.e., N I); three of the ten 10" recirculation inlet nozzles (i.e., N2); one of the four 24" main steam nozzles (i.e., N3); one of the two 12" core spray nozzles (i.e., N5); one of the three 10" low pressure coolant injection nozzles (i.e., N6); one of the two 6" head spray nozzles (i.e., N7 and N8); one of the two 4" jet pump instrumentation nozzles (i.e., N9); and the vibration instrumentation nozzle (i.e., N16).

Code Case N-702 proposes that visual examination (i.e., VT-1) may be used in lieu of volumetric examination for the nozzle inner radii (i.e., Item B3.100). Note, however, that CPS is not currently using ASME Code Case N-648-1 on enhanced magnification visual examination and has no plans of using this Code Case in the future. CPS will continue to perform volumetric examinations of all required nozzle inner radii.

Basis for Use.

The Electric Power Research Institute (EPRI) Technical Report 1003557, "BWRVIP-108, BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii," (Reference 1) provides the basis for Code Case N-702. The EPRI report found that failure probabilities due to a low temperature overpressure event at the nozzle blend radius region and nozzle-to-vessel shell weld are very low (i.e., < 1 x 10-6 for 40 years) with or without any inservice inspection.

On December 19, 2007, the NRC issued a Safety Evaluation (SE) approving the use of BWRVIP-108 as a basis for using Code Case N-702 (Reference 3). In Reference 3, Section 5.0, "Plant Specific Applicability," it states that licensees who plan to request relief from the ASME Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-108 report as the technical basis for the use of Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability of the BWRVIP-108 report to their units in the relief request by showing that the general and nozzle-specific criteria addressed below are satisfied:

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(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115 'F per hour.

(2) For the Recirculation Inlet Nozzles, the following criteria must be met:

a. (pr/t)/Cppv< 1.15

b. [p(ro 2 +ri 2)/(ro2-ri 2)]/CNozzLE<l. 15 (3) For the Recirculation Outlet Nozzles, the following criteria must be met:

a. (pr/t)/CRPv< 1.15

b. [p(ro 2+ri 2)/(ro 2-ri 2)]/CNozzLE< 1.15 Demonstration of how CPS meets the NRC plant-specific applicability is provided in Enclosure 2. Based upon all RPV nozzle-to-vessel shell welds and nozzle inner radii sections meeting the NRC plant-specific criteria, Code Case N-702 is applicable to CPS.

Therefore, use of Code Case N-702 provides an acceptable level of quality and safety pursuant to 10CFR50.55a(a)(3)(i) for all RPV nozzle-to-vessel shell welds and nozzle inner radii sections.

6.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Ten-Year Inspection Interval for CPS.

7.0 PRECEDENTS

Similar relief requests have been approved for:

a. A similar request was approved for use at Duane Arnold Energy Center on August 29, 2008 (i.e., Reference 4).

b. An identical request was approved for use at CPS during the stations' Second Inservice Inspection Interval on August 29, 2009 (i.e., Reference 5).

8.0 REFERENCES

1. EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,"

dated October 2002.

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2. ASME Boiler and Pressure Vessel Code, Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds,Section XI, Division 1," dated February 20, 2004.

3. Letter from Matthew A. Mitchell (NRR), to Rick Libra, BWRRVIP Chairman, "Safety Evaluation of Proprietary EPRI Report, 'BWR Vessel and Internals Project, Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Inner Radius (BWRVIP-108),"' dated December 19, 2007.

4. Letter from Lois James (NRR) to Richard L. Anderson (Duane Arnold Energy Center), "Duane Arnold Energy Center - Safety Evaluation for Request for Alternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner Radius Examinations (TAC NO. MD8193)," dated August 29, 2008.

5. Letter from S. J. Campbell (NRR) to C. G. Pardee (EGC) "Clinton Power Station, Unit No. 1 - Proposed Alternative to IOCFR50.55a Examination Requirements for Reactor Pressure Vessel Weld Inspections (TAC No. ME0218)," dated August 24, 2009.

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Table of ASME Section XI Components Affected CPS, Unit I IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER NIA 20" Recirculation Outlet Nozzle N IA to Vessel B-D B3.90 Weld N IA-IRS 20" Recirculation Outlet Nozzle Ni A Inner Radius B-D B3.100 NIB 20" Recirculation Outlet Nozzle N IB to Vessel B-D B3.90 Weld N I B-IRS 20" Recirculation Outlet Nozzle N IB Inner Radius B-D B3. 100 N2A 10" Recirculation Inlet Nozzle N2A to Vessel Weld B-D B3.90 N2A-IRS 10" Recirculation Inlet Nozzle N2A Inner Radius B-D B3.100 N2B 10" Recirculation Inlet Nozzle N2B to Vessel Weld B-D B3.90 N2B-IRS 10" Recirculation Inlet Nozzle N2B Inner Radius B-D B3.100 N2C 10" Recirculation Inlet Nozzle N2C to Vessel Weld B-D B3.90 N2C-IRS 10" Recirculation Inlet Nozzle N2C Inner Radius B-D B3.100 N2D 10" Recirculation Inlet Nozzle N2D to Vessel Weld B-D B3.90 N2D-IRS 10" Recirculation Inlet Nozzle N2D Inner Radius B-D B3.100 N2E 10" Recirculation Inlet Nozzle N2E to Vessel Weld B-D B3.90 N2E-IRS 10" Recirculation Inlet Nozzle N2E Inner Radius B-D B3. 100 N2F 10" Recirculation Inlet Nozzle N2F to Vessel Weld B-D B3.90 N2F-IRS 10" Recirculation Inlet Nozzle N2F Inner Radius B-D B3.100 N2G 10" Recirculation Inlet Nozzle N2G to Vessel Weld B-D B3.90 N2G-IRS 10" Recirculation Inlet Nozzle N2G Inner Radius B-D B3.100 N2H 10" Recirculation Inlet Nozzle N2H to Vessel Weld B-D B3.90 N2H-IRS 10" Recirculation Inlet Nozzle N2H Inner Radius B-D B3.100 N2J 10" Recirculation Inlet Nozzle N2J to Vessel Weld B-D B3.90 N2J-IRS 10" Recirculation Inlet Nozzle N2J Inner Radius B-D B3.100 N2K 10" Recirculation Inlet Nozzle N2K to Vessel Weld B-D B3.90 N2K-IRS 10" Recirculation Inlet Nozzle N2K Inner Radius B-D B3.100 N3A 24" Main Steam Nozzle N3A to Vessel Weld B-D B3.90 N3A-IRS 24" Main Steam Nozzle N3A Inner Radius B-D B3.100 N3B 24" Main Steam Nozzle N3B to Vessel Weld B-D B3.90 N3B-IRS 24" Mlain Steam Nozzle N3B Inner Radius B-D B3.100 N3C 24" Main Steam Nozzle N3C to Vessel Weld B-D B3.90 N3C-IRS 24" Main Steam Nozzle N3C Inner Radius B-D B3.100 N3D 24" Main Steam Nozzle N3D to Vessel Weld B-D B3.90 N3D-IRS 24" Main Steam Nozzle N3D Inner Radius B-D B3.100 N5A 12" Core Spray Nozzle N5A to Vessel Weld B-D B3.90 N5A-1RS 12" Core Spray Nozzle N5A Inner Radius B-D B3.100 N5B 12" Core Spray Nozzle N5B to Vessel Weld B-D B3.90 Alion Science & Technology CLN05. G03

151 Program Plan Clinton Power Station Unit 1, Third Interval ENCLOSURE 1 10CFR5O.55a RELIEF REQUEST: 13R-02 Revision 0 (Page 2 of 2)

Table of ASME Section XI Components Affected CPS. Unit 1 IDENTIFICATION WELD DESCRIPTION CODE ITEM NUMBER CATEGORY NUMBER N5B-IRS 12" Core Spray Nozzle N5B Inner Radius B-D B3.100 N6A 10" Low Pressure Core Injection Nozzle N6A to B-D B3.90 Vessel Weld N6A-IRS 10" Low Pressure Core Injection Nozzle N6A Inner B-D B3.100 Radius N6B 10" Low Pressure Core Injection Nozzle N6B to B-D B3.90 Vessel Weld N6B-IRS 10" Low Pressure Core Injection Nozzle N6B Inner B-D B3.100 Radius N6C 10" Low Pressure Core Injection Nozzle N6C to B-D B3.90 Vessel Weld N6C-IRS 10" Low Pressure Core Injection Nozzle N6C Inner B-D B3.100 Radius N7 6" Top Head Spray Nozzle N7 to Vessel Weld B-D B3.90 N7-IRS 6" Top Head Spray Nozzle N7 Inner Radius B-D B3.100 N8 6" Top Head Spare Nozzle N8 to Vessel Weld B-D B3.90 N8-IRS 6" Top Head Spare Nozzle N8 Inner Radius B-D B3.100 N9A 4" Jet Pump Instrumentation Nozzle N9A to Vessel B-D B3.90 Weld N9A-IRS 4" Jet Pump Instrumentation Nozzle N9A Inner B-D B3.100 Radius N9B 4" Jet Pump Instrumentation Nozzle N9B to Vessel B-D B3.90 Weld N9B-IRS 4" Jet Pump Instrumentation Nozzle N9B Inner B-D B3.100 Radius N 16 Vibration Instrumentation Nozzle to Vessel Weld B-D B3.90 N 16-IRS Vibration Instrumentation Nozzle Inner Radius B-D B3.100 CLNO5. G03 A lion Science Alion Science&& Technology Technoloff CLN05. GO3

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval ENCLOSURE 2 10CFR50.55a RELIEF REQUEST: 13R-02 Revision 0 (Page 1 of 2)

Responses to NRC Plant Specific Applicability

1. The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 115 'F/hour.

This criterion is met by adherence to CPS Technical Specification 3.4.11, "Reactor Coolant System Pressure/Temperature Limits," Surveillance Requirement 3.4.11.1 which requires verification that the Reactor Coolant System heatup and cooldown rates are limited to less than or equal to 100 'F in any one hour period and, less than or equal to 20

'F in any one hour period during RPV pressure testing.

2. For the Reactor Recirculation Inlet (N2) Nozzles, (pr/t)/CRpv must be less than 1.15, where:

p = normal RPV pressure = 1025 psig r RPV inner radius = 110.19 inches t RPV wail thickness = 6.1 inches CRPv = 19332 Result: (pr/t)/CRIpv = 0.96

3. For the Reactor Recirculation Outlet (N 1) Nozzles, (pr/t)/CRPv must be less than 1.15, where:

p = normal RPV pressure = 1025 psig r = RPV inner radius = 110.19 inches t = RPV wall thickness = 6.1 inches CRPV = 16171 Result: (pr/t)/CRpv = 1.14

4. For the Reactor Recirculation Inlet (N2) Nozzles [p(ro 2 +ri2)/(ro 2-ri 2)]/CNozzLE must be less than 1.15, where:

p = normal RPV pressure = 1025 psig ro = nozzle outlet radius 11.69 inches ri = nozzle inner radius = 5.81 inches CNOZLE = 1637 Result: [p(ro 2 +ri2 )/(ro -ri )]/CNOzzLE = 1.04 CLNO5. G03 A

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ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval ENCLOSURE 2 10CFR50.55a RELIEF REQUEST: I3R-02 Revision 0 (Page 2 of 2)

5. For the Reactor Recirculation Outlet (N I) Nozzles [p(ro 2+ri 2)/(ro 2-ri 2)]/CNozzLE must be less than 1.15, where:

p = normal RPV pressure = 1025 psig ro nozzle outlet radius = 16.3125 inches ri = tozzle inner radius = 9.0 inches CNOZZLE = 1977 Result: [p(ro 2+ri2 )/(ro2-ri2)J/CNozzLE = 0.97 CLNO5. G03 A

Alion & Technology Science &

lion Science Technology CLN05. GO3

ISI Program Plan Clinton Power Station Unit]1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-03 Revision 0 (Page 1 of 5)

Request for Relief for Inservice Inspection Impracticality of Pressure Testing the RPV Head Flange Seal Leak Detection System In Accordance with 10CFR50.55a(g)(5)(iii) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 1, 2, and 3

Reference:

Table IWB-2500-1, IWB-5200 Table IWC-2500-1, IWC-5200 Table IWD-2500-1, IWD-5200 Examination Category: B-P, C-H, and D-B Item Number: B15.10, C7.10, and D2.10

Description:

Pressure Testing the RPV Head Flange Seal Leak Detection System Component Number: RPV Head Flange Seal Leak Detection System Drawing Number: M05-1071, Sht. 1 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at Clinton Power Station (CPS) is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition.

3.0 APPLICABLE CODE REOUIREMENT:

Table IWB-2500-1, Examination Category B-P, Item Number B 15.10, requires all ISI Class 1 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWB-5220. This pressure test is to be conducted prior to plant startup following each reactor refueling outage.

Table IWC-2500-1, Examination Category C-H, Item Number C7.10, requires all ISI Class 2 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWC-5220. This pressure test is to be conducted once each inspection period.

Table IWD-2 500-1, Examination Category D-B, Item Number D2. 10, requires all ISI Class 3 pressure retaining components be subject to a system 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.

Alion Science & Technology CLN05. G03

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-03 Revision 0 (Page 2 of 5) 4.0 IMPRACTICALITY OF COMPLIANCE:

Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that pressure testing the RPV Flange Leak Detection Line is deemed impractical.

The Reactor Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver-plated O-ring located on the vessel flange. A second O-ring is located on the opposite side of the tap in the vessel flange (See Figure 13R-03.1). This line is required during plant operation and will indicate failure of the inner flange seal O-ring. Failure of the O-ring would result in a High Pressure Alarm in the Main Control Room.

The configuration of this system precludes manual testing while the vessel head is removed. As figure 13R-03.1 portrays, the configuration of the vessel tap, combined with the small size of the tap and the high test pressure requirement (approximately 1025 psig),

prevents the tap from being temporarily plugged. Also, when the vessel head is installed, an adequate pressure test cannot be performed due to the fact that the inner O-ring is designed to withstand pressure in one direction only. Due to the groove that the O-ring sits in and the pin/wire clip assembly (See Figure 13R-03.2), pressurization in the opposite direction into the recessed cavity and retainer clips would likely damage the O-ring and thus result in further damage to the O-ring.

5.0 BURDEN CAUSED BY COMPLIANCE:

Pressure testing of this line during the System Leakage Test is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. Purposely failing the inner O-ring to perform the Code Required test would require purchasing a new set of 0-rings, additional time and radiation exposure to detension the reactor vessel head, install the new O-rings, and then reset and retension the reactor vessel head. This is considered to impose an undue hardship and burden on CPS.

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

6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

A VT-2 visual examination on the RPV Flange Leak Detection Line will be performed during each refueling outage when the RPV head is off and the head cavity is flooded above the vessel flange. The static head developed with the leak detection line filled with water will allow for the detection of any gross indications in the line. This examination will be performed each refueling outage as per the frequency specified by Tables IWB-2500-1, IWC-2500-1, and IW-D-2500-1.

Alion Science & Tech -iology CLN05. GO3

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-03 Revision 0 (Page 3 of 5) 7.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Ten-Year Inspection Interval for CPS.

8.0 PRECEDENTS

Similar relief requests have been approved for:

Peach Bottom Atomic Power Station Fourth Interval Relief Request 14R-25 was granted per SER dated February 26, 2009 Limerick Generating Station Third Interval Relief Request I3R-08 was granted per SER dated March 11, 2008 LaSalle County Station Third Interval Relief Request I3R-08 was granted per SER dated January 30, 2008 Susquehanna Steam Electric Station Third Inspection Interval Relief Request 3RR-07 was granted per SER dated September 24, 2004.

CLNO5. G03 A lion Science Alion & Tech~wlogy Science & Techitlogy CLN05. G03

IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-03 Revision 0 (Page 4 of 5)

FIGURE 13R-03.1 FLANGE SEAL LEAK DETECTION LINE DETAIL Outer Flange Seal Ring-Hiqh Pressure Leak Detection Monitoring Top Inner Flange Seal Ring See Detail "A" Detail "A" Vessel Flange Sectional View Alion Science & Technology CLN05. GO3

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-03 Revision 0 (Page 5 of 5)

FIGURE 13R-03.2 O-RING CONFIGURATION SECTION A-A CLNOS. G03 A lion Science Alion & Technology Science & Technology CLN05. G03

ISI Program Plan Clinton Power Station Unit]1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 1 of 5)

Request for Relief for Hardship Or Unusual Difficulty Without Compensating Increase In Level Of Quality Or Safety for the Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all ISI Class 2 Instrument Air (IA) Piping and the ISI Class 3 IA Piping Supplying, all SRV's, and both Feedwater Containment Outboard Isolation Check Valves In Accordance with 10CFR50.55a(a)(3)(ii) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 2,3

Reference:

Table IWC-2500-1, IWC-5200 Table IWD-2500-1, IWD-5200 Examination Category: C-H, D-B Item Number: C7.10, D2.10

Description:

Alternative to Performance of System Pressure Tests and VT-2 Visual Examination Requirements for all ISI Class 2 Instrument Air (IA) Piping and the ISI Class 3 IA Piping Supplying, all SRV's, and both Feedwater Containment Outboard Isolation Check Valves Componc nt Nuiabec: Multiple lines (See Note 1 below)

Note 1: A more detailed description of the pressure testing boundary is identified below.

ISI Class 2 Instrument Air (1A) piping and components between containment isolation valv es 1 AO 12A/B and 1 AO 13 A/B and check valves 11A042A/B. This includes the following lines, valves, and components shown on CPS Piping and Instrumentation Diagram (P&ID) M05-1040 Sht. 7 not listed above.

" Lines IlA71BA/BB-1, IIA14GA/GB-1, 1lA95A/B-1, lIA93AA/BA-3/4, and 11A96AA/BA-3/4

• Valves 11A131 A/B, 11A129A/B, and the blind flanges on lines IIA95A/B-1 ISI Class 3 IA system piping and components requiring inspection. This includes the following IA lines and valves supplying all 16 safety relief valves (SRV's) and both Feedwater containment outboard isolation check valves.

• P&ID M05-1040 Sht. 7 lines - 11A79CA/CB-1, 1lA92AA/BA-3/4, IlA102BA-1/2, IIA103BA-1/2, IlA71AA/AB-1, llA87A/B-1/2, llA125A/B-1/2, llA122A/B-1, Alion Science & 1echnology CLN05. GO3

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 2 of 5) 1IA88A/B- 1/2, 11A71 CA/CB- 1, 11A71 DA/EA/FA/GA- 1/2, and 11A71DB/EB/FB/GB/FC- 1/2

" P&ID M05-1040 Sht. 7 valves - lIA075A/B, llA076A/B, IIA13OA/B, IA1 170A/B, OIAI8MA/B, 11A044A/B, 11A 1171 A/B, 1Al 172A/B, 11A096C/D, and 11A097A/B.

NOTE - Strainers 1IA26FA/FB are not Code components

" P&ID M10-9002 Sht. 1 lines - IlA71DA/DB/EA/EB/FA/FB/FC-1/2, 11A85A/B/C/D/E/F/G- 1/2, 1MS71CE/DE- 1/2, 1MS72AE/BE- 1/2, 1MS73BE/CE- 1/2, 1MS74CE- 1/2, 1MS71CG/DG-3/4, 1MS72AG/BG-3/4, 1MS73BG/CG-3/4, 1MS74CG-3/4, 1MS71CH/DH-1/2, 1MS72AH/BH-1/2, 1MS73BH/CH-1/2, 1MS74CH- 1/2, 1MS71CF/DF-3/4, 1MS72AF/BF-3/4, 1MS73BF/CF-3/4, IMS74CF-3/4, 1MS71CC/DC-2, 1MS72AC/BC-2, 1MS73BC/CC-2, IMS74CC-2, 1MS71CJ/CK/DJ/DK- 11/4, 1MS72AJ/AK/BJ/BK- 11/4, 1MS73BJ/BK/CJ/CK- 1 1/4, and IMS74CJ/CK-1 1/4

• P&ID M 10-9002 Sht. 1 valves - 1lA094A/B/C/D/E/F/G, 1B21-F039B/C/D/E/H/K/S, 1B21-F331C/D, 1B21-F332A/B, 1B21-F3336/C, 1B21-F334, valves 'G' on M1O-9002 Sht. 1 and 1B21-F082B/C/D/E/H/K/S

• P&ID M 10-9002 Sht. 1 accumulators - 1B21 -A003 B/C/D/E/H/K/S

• P&ID M1O-9002 Sht. 2 lines - IlA71GA/GB-1/2, llA86C/E-1/2, 1MS75AE/BE-1/2, 1MS76CE/DE- 1/2, 1MS77AE/CE/DE- 1/2, 1MS78BE/CE- 1/2, 1MS75AC/BC-2, 1MS76CC/DC-2, 1MS77AC/CC/DC-2, IMS78BC/CC-2, 1MS75AG/AH/BG/BH- 1 1/4, 1MS76CG/CH/DG/DH- 1 1/4, 1MS77AG/AH/CG/CH/DG/DH- 1 1/4, and 1MS78BG/BH/CG/CH-1 1/4

• P&ID M10-9002 Sht. 2 valves - llA095C/E, 1B21-F036A/F/G/J/L/M/N/P/R and IB2 l-F08 1A/F/G/J/L/M/N/P/R

• P&ID M10-9002 Sht. 2 accumulators - 1B21-A004A/F/G/J/L/M/N/P/R

• P&ID M1O-9004 Sht. 8 lines - 1FW26BA/BB-1/2, 1FW27BA/BB-1/2, 1FW26CA/CB-2, and 1FW28AA/AB-3/4

• P&ID M1O-9004 Sht. 8 valves - 1B21-F433A/B and 1B21-F492A/B

• P&ID M10-9004 Sht. 8 accumulators - 1B21-A300A/B 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at Clinton Power Station (CPS) is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition.

3.0 APPLICABLE CODE REOUIREMENT:

Table IWC-2500-1, Examination Category C-H, Item Number C7. 10, requires all ISI Class 2 pressure retaining components be subject to a system leakage test with a VT-2 Alioit Science & Technology CLN05. GO3

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval I OCFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 3 of 5) visual examination in accordance with Paragraph IWC-5220. This pressure test is to be conducted once each inspection period.

Table IWD-2500- 1, Examination Category D-B, Item Number D2. 10, requires all ISI Class 3 pressure retaining components be subject to a system leakage test with a VT-2 visual examination in accordance with Paragraph IWD-5220. This pressure test is to be conducled once each inspection period.

4.0 REASON FOR REQUEST:

Pursuant to IOCFR50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

Performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, many of which are in elevated dose rate areas with limited access. VT-2 visual inspections would result in additional radiation exposure (estimated 2 Rem) and industrial safety challenges without any added benefit in the level of quality and safety. These inspections would not be consistent with radiation exposure practices of "As Low As Reasonably Achievable (ALARA)."

Relief is reqaested from the performance of system pressure tests and VT-2 visual examination requirements specified in Tables IWC-2500-1 and IWD-2500-1 for all ISI Class 2 IA piping and the ISI Class 3 IA piping supplying all SRV's and both Feedwater containment outboard isolation check valves.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

As an alternative to the examination requirements of Tables IWC-2500-1 and IWD-2500-1, CPS will perform pressure decay testing on the ISI Class 2 and 3 IA piping supplying all 16 SRV's and both Feedwater containment outboard isolation check valves as required in surveillance procedure CPS 9061.11, "Instrument Air Check Valve Operability and Pipe Pressure Test."

Surveillance procedure CPS 9061.11, verifies the operability of SRV actuation capability and check valves in the IA supply lines to all 16 SRV's and both Feedwater containment outboard isolation check valves. This surveillance test is performed for each individual SRV and boih Feedwater containment outboard isolation check valves as a requirement of the CPS Inservice Testing (IST) Program. One specific test this surveillance performs, is a pressure decay test of the SRV and Feedwater containment outboard isolation check valve accumulators, as well as associated piping and valves. The pressure decay test is Alioa Science & Technology CLN05. G03

ISI ProgramPlan Clinton PowerStation Unit 1, ThirdInterval 10CFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 4 of 5) performed by isolating and pressurizing these accumulators and associated piping to the nominal operating pressure. The decay in pressure is then monitored through calibrated pressure measuring instrumentation. If any pressure decay acceptance criterion (see Enclosure 1) is exceeded, the surveillance identifies appropriate troubleshooting steps to perform, including soap-bubble application to locate leakage.

The pressure decay test performed as part of CPS 9061.11 identifies any degradation of the ISI Class 2 and 3 ADS supply piping and the SRV and Feedwater containment outboard isolation check valve accumulators and associated piping. The volume tested by this surveillance encompasses all piping and components requiring testing under ASME Section XI for these portions of the IA system. This surveillance is performed on a greater frequency than that required in Tables IWC-2500-1 or IWD-2500-1 and the test pressure is consistent with the pressure requirements of both tables. Thus, the testing performed during this surveillance will provide the same level of quality and safety as the pressure testing and VT-2 visual examination requirements of Tables IWC-2500-1 and IWD-2500-1.

The VT-2 visual examination described in Tables IWC-2500-1 and IWD-2500-1 and performed once per inspection period, would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to a large amount of piping and components, many of which are in elevated dose rate areas with limited access. VT-2 visual inspections would result in additional radiation exposure (estimated 2 Rem) and industrial safety challenges without any added benefit in the level of quality and safety.

These inspections would not be consistent with radiation exposure practices of "As Low As Reasonably Achievable (ALARA)."

In summary, relief is requested from the performance of system pressure tests and VT-2 visual examination requirements specified in Tables IWC-2500-1 and IWD-2500-1 for the ISI Class 2 and 3 IA system piping and components identified in this request on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

6.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Ten-Year Inspection Interval for CPS.

7.0 PRECEDENTS

Similar relief requests have been approved for:

Alion Science & Technology CLN05. G03

ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 5 of 5)

CPS Second Inspection Interval Relief Request 4212, Rev. 1 was authorized per SER dated December 13, 2007. The Third Inspection Interval Relief Request utilizes an identical approach as was previously approved.

LaSalle County Station Second Inspection Interval Relief Requests PR-08 and PR-10 were authorized per SER dated June 28, 2002.

Alion Science & Technology CLN05. GO3

ISI Program Plan Clinton Power Station Unit 1, Third Interval ENCLOSURE 1 10CFR50.55a RELIEF REQUEST: 13R-04 Revision 0 (Page 1 of 1)

Acceptance Criteria From Procedure CPS 9061.11 (For Information Only)

CPS, Unit 1 Component Leakage Pressure Drop Comments Criterion Test Duration Accumulator Headers for all < 1.5 psig > 108 minutes SRV's except IB21-F05IC and D Accumulator Headers for < 1.5 psig > 31 minutes Smaller volume than other SRV's.

1B21-FO51C and D Accumulator Headers for < 1.5 psig > 26 minutes Smaller volume than SRV's.

Feedwater Check Valve I ADS Supply Header to < 22 psig > 60 minutes This inspection tests over 200 feet Accumulator Headers of piping and components.

Alion Science & Technology CLN05. GO3

1Sf Program Plan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: I3R-05 Revision 0 (Page 1 of 3)

Request for Relief Regarding Inservice Inspection Impracticality Due to the Examination of the High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal Pump Casing Welds In Accordance with 10CFR50.55a(g)(5)(iii) 1.0 ASME CODE COMPONENTS AFFECTED:

Code Class: 2

Reference:

IWC-2500, Table IWC-2500-1 Examination Category: C-G Item Number: C6.10

Description:

Examination of the ISI Class 2 High Pressure Core Spray, Low Pressure Core Spray, and Residual Heat Removal Pump Casing Welds Component Number: 1A RHR Pump Casing Welds 1B RHR Pump Casing Welds 1C RHR Pump Casing Welds HPCS Pump Casing Welds LPCS Pump Casing Welds Drawing Number: B-69, B-71, and B-73 2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at Clinton Power Station (CPS) is the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code,Section XI, 2004 Edition.

3.0 APPLICABLE CODE REOUIREMENT:

Table IWC-2500-1 states that the pump casing welds require a surface examination in accordance with the examination requirements illustrated in Figure IWC-2500-8.

Per Table IWC-2500-1, the multiple-component concept applies, and examinations are limited to either 100% of the welds of one of three Residual Heat Removal Pumps, one High Pressure Core Spray Pump, and one Low Pressure Core Spray Pump, or distributed among any of the pumps of that same group with similar design, size, function, and service in the systenm. The examination may be performed from either the inside or outside surface of the component.

CLNO5. G03 A

Alion Science &

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& Technology CLN05. G03

ISI ProgramPlan Clinton PowerStation Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-05 Revision 0 (Page 2 of 3) 4.0 IMPRACTICALITY OF COMPLIANCE:

Pursuant to IOCFR50.55a(g)(5)(iii), relief is requested on the basis that conformance with these code requirements is impractical as conformance would require extensive structural modifications to these pumps.

CPS's three Residual Heat Removal Pumps (1E 12-CO02A, 1E 12-COO2B, and 1E 12-CO02C), one High Pressure Core Spray Pump (1E22-COO1), and one Low Pressure Core Spray Pump (1E21 -COO 1) were originally designed where the pump casing welds were encased in concrete, thus making the welds inaccessible for inservice inspection.

Therefore, it is impractical for CPS to perform the surface examination of these welds without destruction of the concrete resulting in unnecessary engineering and installation costs and radiation exposure without a compensating increase in safety. Additionally, due to the design of the subject pumps, access to the affected welds can only be achieved through disassembly of the pump, removal of the pump internals, and the required surface examinations perfoimed from the inside surface of the welds. This effort, in the absence of any other necessary pump maintenance, represents a significant expenditure of man hours and radiation exposure to plant personnel, without a compensating increase in plant safety.

5.0 BURDEN CAUSED BY COMPLIANCE:

Compliance with the applicable Code requirements can only be accomplished by redesigning and refabricating the subject pumps. Based on this, the Code requirements are deemed impractical in accordance with 10CFR50.55a(g)(5)(iii).

6.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

In the event the subject welds become accessible upon disassembly of any one (1) of the pumps, the welds will be surface examined from the inside surface or a VT-1 visual examination will be performed for that particular pump group to the maximum extent practicable based on the obstructions and geometric constraints detailed in the Impracticality Of Compliance section of this relief request. The examination method will be determined by CPS based on radiation environment data at the time access is enabled.

Additionally, a VT-2 visual examination during system pressure testing per Examination Category C-H will be performed once each period by examining the surrounding area (exposed areas around these components where the pump casing join/merge with the concrete) for evidence of leakage in accordance with Paragraph IWA-5241(b). These examinations will provide reasonable assurance of continued structural integrity of the piping systems.

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ISI ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 13R-05 Revision 0 (Page 3 of 3) 7.0 DURATION OF PROPOSED ALTERNATIVE:

Relief is requested for the Third Ten-Year Inspection Interval for CPS.

8.0 PRECEDENTS

Similar relief requests have been approved for:

" LaSalle County Station Third Inspection Interval Relief Request 13R-03 was granted per SER dated January 30, 2008.

• Limerick Generating Station Third Inspection Interval Relief Request 13R-07 was granted per SER dated March 11, 2008.

• Susquehanna Steam Electric Station Third Inspection Interval Relief Request 3RR-02 was granted per SER dated February 1, 2005.

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• • • NOTE ***

Second ISI Interval Relief Request 4215, Revision 0 was previously submitted and approved under the Second Interval ISI Program Plan. The approval authorized under NRC SER dated December 30, 2009 for CPS, Exelon Generation Company, LLC., was for permanent relief for deferral of RPV shell weld examinations, and thus applies to the remaining term of operation under the existing, initial license, including this Third Inspection Interval. All ASME Code references were made in accordance with the 1989 Edition, No Addenda of ASME Section XI.

No changes to the actual approved relief request have been made and no further or revised authorization is required. Formatting for Relief Request 4215, Revision 0 varied from the standard ISI Program Plan format due to the fact that it also requested relief from the Augmented Reactor Pressure Vessel examination contained in 10CFR50a(g)(6)(ii)(A)(2).

10CFR50.55a Request Regarding Alternative Provides Acceptable Level Of Quality And Safety (10CFR50.55a(a)(3)(i))

10CFR50.55a Request Number 4215

1. ASME Code Component(s) Affected Code Class: 1 Component Numbers: RPV-C 1, RPV-C2, RPV-C3, and RPV-C4 Examination Category: B-A Item Number: BI.11

==

Description:==

Reactor Pressure Vessel (RPV) Shell Circumferential Welds

2. Applicable Code Edition and Addenda

Clinton Power Station (CPS) is currently in its second 10-year inspection interval and complies with the 1989 Edition of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code),Section XI. Additionally, for ultrasonic examinations,Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995 Edition, with the 1996 Addenda, is implemented as required (and modified) by IOCFR50.55a.

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3. Applicable Code Requirement

In accordance with the provisions of 10CFR50.55a, "Codes and standards," paragraph (a)(3)(i),

Exelon Generation Company, LLC (EGC) requests permanent relief (for the remaining portion of the initial license period that expires on September 29, 2026) for CPS, Unit 1, from the following requirements:

1. Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires volumetric examination of all RPV shell circumferential welds each inspection interval (i.e., Examination Category B-A, Item No. B1.11);

2. Subsubarticle IWB-2420 requires the sequence of component examinations which was established during the first inspection interval to be repeated during each successive inspection interval, to the extent practical. Therefore, performance of successive examinations of RPV shell circumferential welds is required by Subsubarticle 1\VB-2420; and

3. Subsubarticle IWB-2430 requires examinations performed in accordance with Table IWB-2500-1 that reveal flaws or relevant conditions exceeding the acceptance standards of Table IWB-3410-1 to be extended to include additional examinations during the current outage.

4. Reason for Request

Reference 1 provides the technical basis for permanently deferring the augmented inspections of circumferential welds in boiling water reactor (BWR) RPV's. In the report, the BWR Vessel and Internals Project (B WRVIP) concluded that the probabilities of failure for BWR RPV circumferential welds are orders of magnitude lower than that of the longitudinal welds. The NRC conducted an independent risk-informed, probabilistic fracture mechanics assessment (PFMA) of the anal) sis presented in Reference 1, and the results are documented in Reference 2.

EGC has determined that the proposed alternative described below provides an acceptable level of quality and safety and satisfies the requirements of 10CFR50.55a(a)(3)(i).

5. Proposed Alternative and Basis for Use Proposed Alternative In accordance with IOCFR50.55a(a)(3)(i), and consistent with information contained in Reference 3, EGC considers the following alternate provisions for the subject weld examinations.

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Inservice Inspection Scope The failure frequency for RPV shell circumferential welds is sufficiently low to justify their elimination from the ISI requirement of 10CFR50.55a(g) based on the NRC Safety Evaluation (Reference 2).

The ISI and augmented examination requirements of the ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B1.12, RPV shell longitudinal welds (i.e.,

also known as vertical or axial welds) shall be performed, to the extent possible, and shall include inspection of the circumferential welds only at the intersection of these welds with the longitudinal welds, or approximately 2 to 3 percent of the RPV shell circumferential welds.

When this examination is performed, an automated ultrasonic inspection system will provide the best possible examination of the RPV shell longitudinal welds.

The procedures for these examinations shall be qualified such that flaws relevant to the RPV integrity can be reliably detected and sized, and the personnel implementing these procedures shall be qualified in the use of these procedures.

Successive Examination of Flaws For ASME Code Section XI, Table IWB-2500- 1, Examination Category B-A, Item No. B 1.11, RPV shell circumferential welds (i .e., at intersections with longitudinal welds), successive examinations per Subsubarticle IWB-2420 are not required for nonthreatening flaws (i.e.,

original vessel material or fabrication flaws such as inclusions which exhibit negligible or no growth during the life of the vessel), provided that the following conditions are met:

1. The flaw is characterized as subsurface in accordance with BWRVIP-05 (i.e., Reference 1);

2. The non-destructive examination technique and evaluation that detected and characterized the flaw as originating from material manufacture or vessel fabrication is documented in a flaw evaluation report; and

3. The vessel containing the flaw is acceptable for continued service in accordance with Subarticle IWB-3600, "Analytical Evaluation of Flaws," and the flaw is demonstrated acceptable for the intended service life of the vessel.

For ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B 1.12, RPV shell longitudinal welds, all flaws shall be reinspected at successive intervals consistent with ASME Code and regulatory requirements.

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Additional Examinations of Flaws For ASME Section XI, Table IWB-2500- 1, Examination Category B-A, Item No. B 1.11, RPV shell circumferential welds (i.e., at intersections with longitudinal welds), additional requirements per Subsubarticle IWB-2430, "Additional Examinations," are not required for flaws provided the following conditions are met:

1. If the flaw is characterized as subsurface in accordance with BWRVIP-05, then no additional examinations are required;

2. If the flaw is not characterized as subsurface in accordance with BWRVIP-05, then an engineering evaluation shall be performed, addressing the following as a minimum:

" A determination of the root cause of the flaw,

" An evaluation of any potential failure mechanisms,

" An evaluation of service conditions which could cause subsequent failure, and

" An evaluation per Subarticle IWB-3600 demonstrating that the vessel is acceptable for continued service; and

3. If the flaw meets the criteria of Subarticle IWB-3600 for the intended service life of the vessel, then additional examinations may be limited to those welds subject to the root cause conditions and failure mechanisms, up to the number of examinations required by paragraph (a) of Subsubarticle IWB-2430. If the engineering evaluation determines that there are no additional welds subject to the same root cause conditions or no failure mechanism exists, then no additional examinations are required.

For ASME Code Section XI, Table IWB-2500-1, Examination Category B-A, Item No. B 1.12, RPV shell longitudinal welds, additional examination for flaws shall be in accordance with Subsubarticle IWB-2430. All flaws in RPV shell longitudinal welds shall require additional weld examinations consistent with ASME Code and regulatory requirements. Examinations of the RPV shell circumferential welds shall be performed if RPV longitudinal welds reveal an active, mechanistic mode of degradation.

Basis for Use Reference 1 provides the technical basis to justify relief from the examination requirements of RPV shell circumferential welds. The results of the NRC's evaluation of Reference 1 are documented in Reference 2. Reference 3 permits BWR licensees to request permanent relief from the ISI requirements of 10CFR50.55a(g) (i.e., for the remaining term of operation under the existing, initial license) for the volumetric examination of RPV shell circumferential welds (i.e.,

ASME Code Section XI, Table IWB2500-1, Examination Category B-A, Item No. Bl. 11). This relief can be granted by demonstrating that:

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IS! ProgramPlan Clinton Power Station Unit 1, Third Interval 10CFR50.55a RELIEF REQUEST: 4215 Revision 0 (Page 5 of 10) 1 At the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 28, 1998, safety evaluation, and 2 Licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staffs July 28, 1998, safety evaluation.

Reference 3 also states that licensees will still need to perform the required inspections of "essentially 100 pelcent" of all axial welds.

Generic Letter 98-05, Criterion 1 Demonstrate that at the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the NRC's July 28, 1998, safety evaluation.

Response

The NRC evaluation of BWRVIP-05 utilized the FAVOR code to perform a PFMA to estimate the RPV shell weld failure probabilities. Three key assumptions of the PFMA are: (1) the neutron fluence used was the estimated end-of-life mean fluence, (2) the chemistry values are mean values based on vessel types, and (3) the potential for beyond-design-basis events is considered.

Table 1 provides a comparison of the limiting RPV circumferential weld parameters for CPS to those found in Table 2.6-4 of the NRC final safety evaluation of BWRVIP-05 (i.e., Reference 2) for a Chicago Bridge and Iron (CB&I) vessel. The material composition and chemistry factors, and the inside diameter fluences at 32 effective full power years (EFPYs) were used to determine the acceptable reference temperatures at CPS. Although the unirradiated reference temperature for CPS is higher than the NRC limit, the combination of unirradiated reference temperature and embrittlement shift yields adjusted reference temperatures considerably lower than the NRC mean analysis values.

As a result, the shift in reference temperature is lower than the 32 EFPY shift from the NRC analysis. Therefore, the RPV shell weld embrittlement due to fluence is calculated to be less than the NRC's limiting case, and the RPV shell circumferential weld failure probabilities are bounded by the conditional failure probability in the NRC's limiting plant specific analysis (32 EFPY) through the projected end of license. For these reasons, the limiting conditional failure probability for CPS RPV circumferential welds is bounded by Reference 2.

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Table 1 Effects of Irradiation on RPV Circumferential Weld Properties Parameter Description CPS Parameters at 32 EFPY NRC Limiting Plant (Weld Wire Heat/Flux Lot Specific Analysis*

1. 76492/L430B27AE)

Copper (weight %) 0.10 0.10 Nickel (weight %) 1.08 0.99 Chemistry Factor 135 134.9 End of Life Inside Diameter 0.081 0.51 Fluence (1019 n/cm 2)

ARTNDT (OF) 50.77 109.5 ARTNDT(U) ( 0 F) -30 -65 Mean RTNDT (7F) 20.77 44.5 Lable L.0-4, Summary o0 liesults o0 INKLC Stai and t wrvlr Lmitmg Flani-Speciic Analyses kj5 Err 1),

corrected pe,' Reference 8.

Generic Letter 98-05, Criterion 2 Demonstrate that licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC's July 28, 1998 safety evaluation (Reference 2).

Response

Procedures are in place for CPS that guide operators in controlling and monitoring reactor pressure during all phases of operation, including cold shutdown. Use of these procedures will prevent an over-pressure event, and are reinforced through operator training. Operating procedures contain sufficient guidance to prevent a low temperature over-pressurization event. A reactor coolant system leakage test is performed prior to each restart after a refueling outage. A pre-job briefing is required prior to test commencement with all involved personnel. During pressure testing, measures are taken to limit the potential for system perturbations that could lead to pressure transients. These measures include both administrative and/or hardware controls, such as limiting testing or work activities, or installing jumpers or simulators, to defeat systems actuations that are not required to be operable. Vessel temperature and pressure are required to be monitored and controlled to within CPS Technical Specifications pressure and temperature Alion Science & Technology CLN05. GO3

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(P/T) limits during all portions of testing. Pre-job briefings and careful coordination ensure that pressure transients are minimized.

The high pressure coolant sources that could inadvertently initiate and result in a low temperature overpressurization event are the Feedwater, Reactor Core Isolation Cooling (RCIC), and High Pressure Core Spray (HPCS) systems. During normal RPV fill prior to pressure testing, the Control Rod Drive (CRD) system is the preferred method for filling the reactor. The Condensate/Condensate Booster systems are used as an alternative means to fill the reactor. The motor driven reactor feedwater pump is prevented from starting by the high water level feedwater pump trip signal, which is present due to the high reactor water levels required during pressure testing. During the reactor coolant system leakage test, the reactor is in cold shutdown, and as a result, there is no steam available to drive the turbine driven RCIC and turbine driven reactor feedwater pumps.

The HPCS system is a high pressure make-up system at CPS. The HPCS pump is motor operated, so it can be operated when the reactor is in cold shutdown. However, the HPCS system would require manual initiation, inadvertent initiation, or manual startup to start and inject into the RPV. Also, there is a high RPV water level interlock for the HPCS injection valve to prevent overfilling the RPV. This high level interlock is not normally overridden. Even if the HPCS system is inadvertently started, it would not inject and pressurize the reactor due to the high RPV water level interlock.

The CRD .system is a high pressure system used to operate the control rods. The CRD system is a low flow rate systemwith about 50 gpm flow rate to the reactor. During cold shutdown conditions, reactor water level is maintained with CRD and the Reactor Water Cleanup System (RWCU). These systems are also used to raise and maintain reactor test pressure for the reactor coolant system leakage testing. During cold shutdown conditions, operators closely monitor reactor water level, pressure, and temperature. With the low CRD flowrate, the operators should have sufficient time to react to unanticipated level changes and regain control of reactor pressure, should any abnormalities occur.

The Standby Liquid Control (SLC) System is a high pressure system used to shut down the reactor if the control rods fail to insert. The SLC system has no automatic start function so a spurious stait is unlikely. The SLC system must be manually initiated by the use of a keylock switch for each pump.

During cold shutdown conditions, the condensate booster pumps of the Condensate system are shutdown. It would require direct operator action to start a main Condensate Booster system pump and inject into the reactor pressure vessel. The Condensate/Condensate Booster systems are used as an alternate method for filling the RPV and as the primary method for initially pressuring the RPV for pressure testing.

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These actions are taken in accordance with procedural guidance that includes verification that RPV coolant and metal temperatures will support filling and pressurizing the RPV with the Condensate/Condensate Booster Pump systems without exceeding the Technical Specification P/T limits.

Low pressure coolant sources include the Emergency Core Cooling Systems (ECCS) (i.e., Low Pressure Core Spray (LPCS) and Low Pressure Coolant Injection (LPCI) systems), and the Condensate system. The shutoff heads of the ECCS pumps and condensate pumps are sufficiently low to preclude a low temperature overpressurization event that would exceed the P/T curve limits and an inadvertent low pressure ECCS injection.

In addition to the procedural barriers, licensed operators are provided specific training on the P/T curves and requirements of the Technical Specifications. Simulator sessions are conducted which include plant heat-up and cool-down. Additionally, in response to industry operating experience, the operator training program is routinely evaluated and revised, as necessary, to reduce the possibility of events such as a low temperature overpressurization event.

Based on the above, procedural and administrative controls, as reinforced in operator training, are in place to effectively limit a low temperature overpressurization event.

Summary In summary, EGC hMs reviewed the methodology used in Reference 1, and considering CPS plant specific materials properties, fluence, operational practices, and the provisions of Reference 2, the criteria established in Generic Letter 98-05 (i.e., Reference 3) are satisfied.

Therefore, permanent relief is requested from the examination requirements of 10CFR50.55a for RPV circumferential shell welds since the proposed alternative provides an acceptable level of quality and safety.

6. Duration of Proposed Alternative Permanent relief is requested for the remainder of the existing operating license for CPS.

7. Precedents The NRC has previously approved similar relief for several nuclear power plants, including Dresden Nuclear Power Station, Units 2 and 3 (References 4 and 5), Susquehanna Steam Electric Station, Units 1 and ,2 (References 6 and 7), and Quad Cities Nuclear Power Station, Units I and 2 (References 9 and 10).

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8. References I BWRVIP-05, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05)," dated September 28, 1995 2 Letter from G. C. Lainas (U. S. Nuclear Regulatory Commission) to C. Terry (BWRVIP),

"Final Safcty, Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925)," dated July 28, 1998 3 NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10, 1998 4 Letter from J. M. Heffley (Commonwealth Edison Company) to U. S. Nuclear Regulatory Commission, "Relief Request for Alterative Weld Examination of Circumferential Reactor Pressure Vessel Shell Welds," dated July 26, 1999 5 Letter from A. J. Mendiola (U. S. Nuclear Regulatory Commission) to 0. D. Kingsley (Commonwealth Edison Company), "Dresden -Authorization for Proposed Alterative Reactor Pressure Vessel Circumferential Weld Examinations (TAC Nos. MA6228 and MA6229)," dated February 25, 2000 6 Letter from R. G. Byram (PPL Susquehanna, LLC) to U. S. Nuclear Regulatory Commission, "Request for Alternative to 10CFR50.55a Examination Requirements of Category B 1.11 Reactor Pressure Vessel Welds for PPL Susquehanna LLC Units 1 and 2 PLA-525 1," dated November 7, 2000 7 Letter from M. Gamberoni (U. S. Nuclear Regulatory Commission) to R. G. Byram (PPL Susquehanna, LLC), "Relief Request No. 22 (RR-22) from American Society of Mechanical Enginecrs Boiler and Pressure Vessel Code,Section XI, Susquehanna Steam Elect.ic Station Units I and 2 (TAC Nos. MB0484 and MB0485)," dated February 28, 2001 8 Letter from J. R. Strosnider (U. S. Nuclear Regulatory Commission) to C. Terry (BWRVIP Chairman), "Supplement to Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC NO. MA3395)," dated March 7, 2000 9 Letter fiom P. R. Simpson (Exelon), "Relief Request for Alternative Reactor Pressure Vessel Circumferential Weld Examinations for the Fourth Interval Inservice Inspection Program," dated May 16, 2003 CLNOS.G03 Science &

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"Quad Cities Nuclear Power Station, Units 1 and 2 - Authorization For Proposed Alternative Reactor Pressure Vessel Circumferential Shell Weld Examination (TAC Nos.

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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 J, Primary Reactor Containment Testing for Water Cooled Power Reactors.

SECY-96-080, Issuance of Final Amendment To 10CFR50.55a To Incorporate By Refeience The ASME Boiler And Pressure Vessel Code,Section XI, Division 1, Subsection IWE and IWL.

2) ASME Boiler and Pressure Vessel Code,Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components."

- 2004 Edition, No Addenda. ( 3rd ISI Interval).

- 2001 Edition through the 2003 Addenda. ( 2 nd CISI Interval).

3) ASME Boiler and Pressure Vessel Code,Section III, Division 1, "Rules For Construction of Nuclear Power Plant Components", the 2004 Edition, No Addenda.

4) ASME OM Code, Code For Operation and Maintenance of Nuclear Power Plants, 2004 Edition, No Addenda.

5) Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability, ASME Sectikn XI, Division 1".

6) Regulatory Guide 1.150, Rev. 1, "Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examination".

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) Clinton Power Station Unit 1 Updated Safety Analysis Report (USAR).

10) Clinton Power Station Unit I Operational Requirements Manual (ORM).

11) Clinton Power Station Unit 1 Technical Specifications (TS).

12) NRC NUREG-0313, Revision 2, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping".

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13) NRC NUREG-0578 dated July 1979, "TMI-2 Lessons Learned Task Force Status Repcrt and Short-Term Recommendations".

14) NRC NUREG-0619, dated November 1980, "BWR Feedwater Nozzle and Cotitrol Rod Drive Return Line Nozzle Cracking".

15) NRC NUREG-0737, dated November 1980, "TMI Action Plan Requirements".

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

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

18) BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), EPRI Report TR-1 13932, October 1999.

19) NRC Final SER related to "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules BWRVIP-75), EPRI Report TR-1 13932, October 1999", (TAC NO. MA5012), dated May 14, 2002.

20) BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), EPRI Report TR-1012621, October 2005

21) NRC Final SER related to "BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A),

EPRI Report TR- 1012621, October 2005", dated March 16, 2006

22) Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A71-0594, "Alternate BW'R Feedwater Nozzle Inspection Requirements," dated August 1999.

23) NRC Final SER related to the Boiling Water Reactor Owners' Group (BWROG)

Report, GE-NE-523-A71-0594, "Alternate BWR Feedwater Nozzle Inspection Requirements, August 1999", (TAC No. M94090), dated June 5, 1998.

24) Boiling Water Reactor Owners' Group (BWROG) Report, GE-NE-523-A71-0594-A, Revision 1, "Alternate BWR Feedwater Nozzle Inspiection Requirements," dated May 2000.

25) NRC Final SER related to the Boiling Water Reactor Owners' Group (BWROG)

Report, GE-NE-523-A71-0594-A, Revision 1, "Alternate BWR Feedwater Nozzle Alion Science & Technoloev 9-2 CLN05. GO3 Revision 14

ISI ProgramPlan Clinton Power Station Unit 1, ThirdInterval Inspcction Requirements, May 2000", (TAC No. MA6787), dated March 10, 2000.

26) Branch Technical Position MEB 3-1, dated November 24, 1975, "High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment".

27) Gencric Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds", dated November 10, 1998.

28) NRC Final SER related to the "BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05), EPRI Report EPRI Report TR- 105697, September, 1995", dated July 28, 1998.

29) BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05),

EPRI Report TR- 105697, September, 1995.

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

31) NRC SER related to EPRI Topical Report TR- 112657, Rev. B, Final Report, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 1999",

dated October 28, 1999.

32) EPRI Topical Report TR-1006937, Rev. 0-A, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs", August 2002.

33) NRC SER related to EPRI Topical Report TR-1006937, Rev. 0, "Extension of the EPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to Break Exclusion Region (BER) Programs", dated June 27, 2002.

34) Exelhn Risk-Informed Inservice Inspection Evaluation (Final Report) for Clinton Powc r Station Unit 1.

35) Clinton Power Station Unit 1, ISI Classification Basis Document (CLN05.G04),

Third Ten-Year Inspection Interval.

36) Clinton Power Station Unit 1, ISI Selection Document (CLN05.G05), Third Ten-Year Inspection Interval.

37) Clinton Power Station Unit 1, ISI In-Vessel Inspection Program.

38) Clinton Power Station Unit 1, IWE/IWL Containment Inspection Plan, Second Ten-Year Inspection Interval.

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39) Exelon Procedures ER-AA-330, "Conduct of Inservice Inspection Activities",

ER-AA-330-001, "Section XI Pressure Testing", ER-AA-330-002, "Inservice Inspection of Welds and Components", ER-AA-330-003, "Visual Examination of Section XI Component Supports", ER-AA-330-004, "Visual Examination of Technical Specification Snubbers", ER-AA-330-005, "Visual Examination of Section XI Class CC Concrete Containment Structures", ER-AA-330-007, "Visual Examination of Section XI Class MC Surfaces and Class CC Liners",

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

ER-AA-330-010, "Snubber Functional Testing", and ER-AA-330-0 11, "Snubber Service Life Monitoring Program".

40) Letter Y-109584 (121-09(08-24)-L), Revision 0, "Makeup Capacity Exemption of Class I Components per ASME Section XI", dated August 24, 2009.

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