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Clinton Power Station 151 Program Document No.: CLNO5.G03Clinton Power Station ISI Program Document No.: CLN05.GO3

Exelkn..Nuclear

CLINTON POWER STATION

UNIT 1

ISI PROGRAM PLANTHIRD TEN-YEAR INSPECTION INTERVAL

Commercial Service Date:

Unit 1 - 04/24/87

Clinton Power StationRR 3, Box 228

Clinton, IL 61727

Exelon Generation Company (EGC), LLC300 Exelon Way

Kennett Square, PA 19348

Prepared By:Alion Science and Technology Corporation

Engineering Programs DivisionWarrenville, Illinois

S/ALIONSCICNCE ANO TECHNOLOGY

151 Program PlanClinton Power Station Unit]1, Third Interval

REVISION APPROVAL SHEET

TITLE: ISI Program PlanThird Ten-Year Inspection IntervalClinton Power Station, Unit 1

DOCUMENT NUMBER: CLN05.G03 REVISION: 14

PREPARED TRANSMITTAL

PREPARED:

REVIEWED:

APPROVED:

St .en V. Co1 anAliorn Project 'gineer

Kevin M. Johns~nAlion. Project Engineer

Daniol W. LamoretMAlion Project Manager

EXELON ACCEPTANCE

APPROVED:Mirza Bo C iISI Plogifam Coordinato'"r"

/ I( o1

Alion Science & Technology i CLN05. GO3Revision 14

ISI Program PlanClinton Power Station Unit 1, Third Interval

REVISION APPROVAL SHEET

TITLE: ISI Program PlanThird Ten-Year Inspection IntervalClinton Power Station, Unit 1

DOCUMENT NUMBER: CLN05.G03

EXELON PREPARATION. REVIEW. AND APPROVAL

REVISION: 14

REVIEWED:

REVIEWED:

REVIEWED:

APPROVED:

CONCURRED

Mirza a (

ISI Program Coordinator

DaPrssrees ad nbPressure "Iest and Snub Program Owner

/10 - 1-. 2.010

lo'-4-20i0o

I:

Thomas ParrentContainment (IWE/IWL) Program Owner

Mike Haydon VAuthorized Nuclear Inservice Inspector (ANII)

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

Alion Science & Technology ii CLN05. G03Revision 14


REVISION CONTROL SHEET

Major changes should be outlined within the table below. Minor editorial and formattingrevisions are not required to be logged.

REVISION DATE I REVISION SUMMARY

14 08/20/10 Initial issuance. (This ISI Program Plan was developed by AlionScience and Technology Corporation as part of the Third Interval ISIProgram 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 wassubmitted to the NRC. Future revisions of this document made within the Third ISIInterval will be maintained and controlled at the Clinton Power Station; however, they arenot required to be and will not be submitted to the NRC. The exception to this is thatnew or revised Relief Requests shall be submitted to the NRC for safety evaluation andapproval.

Alion Science & Technology iii CLNO5.G03Revision 14


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

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JS! Program Plan

Clinton Power Station Unit 1, Third Interval

TABLE OF CONTENTS

SECTION DESCRIPTION PAGE

1.0 INTRODUCTION AND BACKGROUND ..................................................................... 1-11.1 Introduction1.2 Background1.3 First Interval ISI Program1.4 Second Interval ISI Program1.5 Third Interval ISI Program1.6 First Interval CISI Program1.7 Second Interval CISI Program1.8 Code of Federal Regulations 1 OCFR50.55a Requirements1.9 Code Cases1.10 Relief Requests

2.0 BASIS FOR INSERVICE INSPECTION PROGRAM ................................................... 2-12.1 ASME Section XI Examination Requirements2.2 Augmented Examination Requirements2.3 System Classifications and P&ID Drawings2.4 ISI Isometric and Component Drawings for Nonexempt ISI Class Components

and Supports2.5 Technical Approach and Positions

3.0 C O M PO N EN T ISI PLA N ................................................................................................ 3-13.1 Nonexempt ISI Class Components3.2 Risk-Informed Examination Requirements3.3 Reactor Coolant Pressure Boundary Normal Make-up Calculation

4.0 SU PPO R T ISI PL A N ....................................................................................................... 4-14.1 Nonexempt ISI Class Supports4.2 Snubber Examination and Testing Requirements

5.0 SYSTEM PRESSURE TESTING ISI PLAN .................................................................. 5-15.1 ISI Class Systems5.2 Risk-Informed Examination of Socket Welds

6.0 CON TA IN M EN T ISI PLAN ........................................................................................... 6-16.1 Nonexempt CISI Class Components6.2 Augmented Examination Areas6.3 Component Accessibility6.4 Responsible Individual and Engineer

7.0 COM PONENT SUM M ARY TABLES ........................................................................... 7-17.1 Inservice Inspection Summary Tables7.2 Snubber Inspection Summary Tables

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ISI Program Plan


TABLE OF CONTENTS (Continued)

SECTION DESCRIPTION PAGE

8.0 RELIEF REQUESTS FROM ASME SECTION XI ........................................................ 8-1

9.0 R E F E R E N C E S ................................................................................................................ 9-1

APPENDICES

A. ISI PROGRAM PLAN REQUIREMENTS

B. ISI PROGRAM PLAN COMPONENT AND PIPING EXAMINATION BOUNDARY

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1SI Program Plan


TABLE OF CONTENTS (Continued)

TABLES DESCRIPTION PAGE

1.1-1 THIRD ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3COM PONENT EXAM INATION S) ................................................................................ 1-3

1.1-2 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR CISICLASS MCCOM 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 R A W IN G S ........................................................................................................ 2-16

2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS .................................................. 2-18

2.4-2 CISI REFEREN CE DRA W IN 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 IN D EX ............................................................................................ 8-2

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ISI Program Plan


1.0 INTRODUCTION AND BACKGROUND

1.1 Introduction

This Inservice Inspection (ISI) Program Plan details the requirements for theexamination and testing of ISI Class 1, 2, 3, MC, and CC pressure retainingcomponents, 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 InserviceInspections (AUG), Snubber Visual Examination and Functional Testing (SNUB),and System Pressure Testing (SPT) requirements imposed on or committed to byCPS. At CPS, the Inservice Testing (IST) Program and the Containment InserviceInspection (CISI) Program are maintained and implemented separately from theISI Program. The IST Program Plan and the IWE/IWL Containment InspectionPlan contain all of the applicable CISI and IST program requirements. (See theIST Program Plan and the IWE/IWL Containment Inspection Plan for moredetails.)

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

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

Paragraph IWA-2430(d)(1) of ASME Section XI allows an inspection interval tobe 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 servicecontinuously for six months or more. The extension may be taken for a period oftime not to exceed the duration of the outage. See Tables 1.1-1, 1.1-2, and 1.1-3for intervals, periods, and extensions that apply to CPS's Third ISI Interval andSecond CISI Interval.

The Third ISI Interval and Second CISI Interval are divided into number ofinspection periods as determined by calendar years within the intervals. Tables

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1.1-1, 1.1-2, and 1.1-3 identify the period start and end dates for the Third ISIInterval and the Second CISI Interval as defined by Inspection Program B. Inaccordance with Paragraph IWA-2430(d)(3), the inspection periods specified inthese Tables may be decreased or extended by as much as 1 year to enableinspections to coincide with CPS's refueling outages.

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1SI Program PlanClinton 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 OutagesStart Date to Start Date to End Date Projected Outage Outage

End Date' Start Date or NumberOutage Duration

Scheduled C1R13Ist 1/12

3rd 07/1/10 to 06/30/1307/1/10 to 06/30/20

Scheduled C1R142 nd 1/14

07/1/13 to 06/30/17 Scheduled CIR151/16

Scheduled C1R163 rd 1/18

07/1/17 to 06/30/20 Scheduled CI R 171/20

Note 1: The end of the CPS Second ISI Interval was initially extended for one year fromJanuary 1, 2010 to December 31, 2010, but then a decision was made to reduce theextension to six months instead of one year from January 1, 2010 toJune 30, 2010per Paragraph IWA-2430(d) of ASME Section XI. (See Section 1.4 for details). Aspermitted by Paragraph IWA-2430(d), the interval was extended accordingly, thusaffecting the start date of the Third ISI Interval.

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TABLE 1.1-2

SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX(FOR CISI CLASS MC COMPONENT EXAMINATIONS)

Interval Period OutagesStart Date to Start Date to End Date Projected Outage Outage

End Date Start Date or NumberOutage Duration

Scheduled C1R121st 1/10

2nd 08/14/08 to 08/13/1108/14/08 to 08/13/18

Scheduled C1R132 nd 1/12

08/14/11 to 08/13/15 Scheduled CIR141/14

Scheduled CIR153 rd 1/16

08/14/15 to 08/13/18 Scheduled CIR161/18



TABLE 1.1-3

SECOND ISI INTERVAL/PERIOD/OUTAGE MATRIX(FOR ISI CLASS CC-CONCRETE COMPONENT EXAMINATIONS)

Interval 5-Year Period OutagesStart Date to Rolling Exam # - Date Projected Outage Outage

End Date (2 Year Window) Start Date or NumberOutage Duration

3P - 10/14/10 Scheduled1/10

C1R12

2 nd

09/09/08 to 09/08/18Scheduled C1R13

1/12

Scheduled C1R141/14

Scheduled CIR151/16

Scheduled C1R161/18

I A -''-1 1

Note 1: The CISI Interval for Class CC components parallels the CISI Interval for Class MCcomponents. The actual inspection schedule however is based on a rolling 5 yearfrequency (+/- 1 year) from the date of completion of the original examinations(10/14/00) performed during the initial September 9, 1996 - September 8, 2001Rulemaking implementation period. The rolling 5 year inspection schedule forcontainment concrete is in accordance with the Inservice Inspection Schedulespecified in Subarticle IWL-2400. (Note that CISI concrete examinations werescheduled prior to the Structural Integrity Test (SIT) and then every 5 yearsthereafter. The SIT was performed in refueling outage C 1R09 (February, 2004);therefore, all refueling outage C 1R07 (October, 2000) examinations were performedprior to the SIT.)



1.2 Background

The Illinois Power Company, now known commercially as Exelon GenerationCompany or Exelon, obtained Construction Permit to build CPS on February 24,1976. The Docket Number assigned to CPS is 50-461. After satisfactory plantconstruction and preoperational testing was completed, CPS was granted a fullpower operating license on April 17, 1987, NPF-62, and subsequently commencedcommercial operation on April 24, 1987.

CPS's piping systems and associated components were designed and fabricated tobe inspected and tested in accordance with the requirements of ASME Section XI.Although this plant was specifically designed to meet the inspection and testingrequirements of ASME Section XI, literal compliance may not be feasible orpractical within the limits of the current plant design. Certain limitations arelikely to occur due to conditions such as accessibility, geometric configuration,and/or metallurgical characteristics. For some inspection categories, an alternatecomponent may be selected for examination and the code statistical anddistribution requirements can still be maintained. If ASME Section XI requiredexamination criteria cannot be met, a relief request will be submitted inaccoidance 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 requirementsof Paragraph (g), Inservice inspection requirements, of that section.

Specifically, Paragraph 1OCFR50.55a(g)(4)(i) calls for the inservice inspectionrequirements of the 120 month inspection interval to comply with therequirements of the latest Edition and Addenda of ASME Section XI referenced inParagraph (b) of lOCFR50.55a on the date twelve months prior the date ofissuance of the operating license, subject to the limitations and modificationslisted in lOCFR50.55a(b).

CPS started commercial operation on April 24, 1987, which marked the beginningof the First ISI Interval. The version of 1OCFR50.55a in effect twelve monthsprior to this date referenced the 1980 Edition with Addenda through the Winter1980 (80W80) of ASME Section XI. The inservice inspection requirementsapplicable to nondestructive examination and system pressure testing for the FirstInservice Inspection Program were based on these rules.

The CPS First ISI Interval was originally effective from April 24, 1987 to April23, 1997, but CPS was shut down from September 1996 through May 1999.ASME Section XI permitted a one year interval extension to allow for outagecorrelation 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



extend the ISI Interval for an equivalent period. As such, the First ISI Intervalwas extended to December 31, 1999 based on plant out of service and/or tocoincide with the scheduling or completion of an outage.

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

1.4 Second Interval ISI Program

Pursuant to 1OCFR50.55a(g), CPS was required to update the ISI Program at theend of the First ISI Interval. The ISI Program was required to comply with thelatest Edition and Addenda of ASME Section XI incorporated by reference in1OCFR50.55a twelve months prior to the start of the Second ISI Interval per10CFR50.55a(g)(4)(ii).

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

Starting in February 2004, in the Second Period of the Second ISI Interval, CPSconverted from an 18 month fuel cycle to a 24 month fuel cycle (2 year outagecycle). A study was performed to determine the effect this change to a 24 monthfuel cycle would have on the current Second Interval ISI Program (e.g.,documents, schedule, crediting, etc.), and where needed, what changes to theprogram needed to be made to alleviate any Code compliance issues presented bythe change in fuel cycle.

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

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

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1.5 Third Interval ISI Program

Pursuant to 1OCFR50.55a(g), licensees are required to update their ISI Programsto meet the requirements of ASME Section XI once every ten years or inspectioninterval. The ISI Program is required to comply with the latest Edition andAddenda of ASME Section XI incorporated by reference in 10CFR50.55a twelvemonths prior to the start of the interval per 10CFR50.55a(g)(4)(ii). As discussedin 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 XIreferenced in 1OCFR50.55a(b)(2) twelve months prior to the start of the Third ISIInterval was the 2004 Edition, No Addenda.

The CPS Third Interval ISI Program Plan was developed in accordance with therequirements of 10CFR50.55a including all published changes , and the 2004Edition, No Addenda of ASME Section XI, subject to the limitations andmodifications contained within Paragraph (b) of the regulation. These limitationsand modifications are detailed in Table 1.8-1 of this section. This Third IntervalISI Program Plan addresses Subsections IWA, IWB, IWC, IWD, IWF, MandatoryAppendices, approved ASME Code Cases, approved alternatives through reliefrequests 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, whichwas supplemented by Code Case N-578-1, for implementing risk-informedinservice inspections during the Second ISI Interval. The RISI Program willcontinue for the Third ISI Interval. Implementation of the RISI Program is inaccordance with Relief Request 13R-01.

CPS also adopted the EPRI Topical Report TR-1006937, Rev. 0-A, methodologyfor additional guidance for adaptation of the RISI evaluation process to BreakExclusion Region (BER) piping, also referred to as the High Energy Line Break(HELB) region. This change to the BER program was made under 1OCFR50.59evaluation criteria. The RISI evaluation for BER piping remains in effect for theThird ISI Interval.

1.6 First Interval CISI Program

CISI examinations were originally invoked by amended regulations containedwithin a Final Rule issued by the United States Nuclear Regulatory Commission(NRC). The amended regulation incorporated the requirements of the 1992Edition through the 1992 Addenda of ASME Section XI, Subsections IWE andIWL, subject to specific modifications that were included in Paragraphs1OCFR50.55a(b)(2)(ix) and 1OCFR50.55a(b)(2)(x).



The final rulemaking was published in the Federal Register on August 8, 1996 andspecified an effective date of September 9, 1996. Implementation of theSubsection IWE and IWL Program from a scheduling standpoint was driven bythe five year expedited implementation period per 1 OCFR50.55a(g)(6)(ii)(B),which specified that the examinations required to be completed by the end of theFirst Period of the First CISI Interval (per Table IWE-2412-1) be completed bythe effective date (by September 9, 2001).

ASME Section XI Subsections IWE, IWL, approved ASME IWE/IWL CodeCases, and approved alternatives through related relief requests and SER's wereadded 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 throughSeptember 8, 2008.

1.7 Second Interval CISI Program

Pursuant to 1 OCFR50.55a(g), licensees are required to update their CISI Programsto meet the requirements of ASME Section XI once every ten years or inspectioninterval. The CISI Program is required to comply with the latest Edition andAddenda of ASME Section XI incorporated by reference in I OCFR50.55a twelvemonths pricr to the start of the interval per 1OCFR50.55a(g)(4)(ii). Based on thisdate, the latest Edition and Addenda of ASME Section XI referenced inI OCFR50.55a(b)(2) twelve months prior to the start of the Second CISI Intervalwas the 2001 Edition through the 2003 Addenda.

The CPS Second Interval ClSI Program Plan was developed in accordance withthe requirements of 1OCFR50.55a including all published changes throughSeptember 30, 2006, and the 2001 Edition through the 2003 Addenda of ASMESection XI, subject to the limitations and modifications contained withinParagraph (b) of the regulation. These limitations and modifications are detailedin Table 1.8-1 of this section. This Second Interval CISI Program Plan addressesSubsections IWE, IWL, approved ASME IWE/IWL Code Cases, approvedalternatives through related relief requests and SER's, and utilizes InspectionProgram B as defined therein.

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

1.8 Code of Federal Regulations 1 OCFR50.55a Requirements

There are certain Paragraphs in 1 OCFR50.55a that list the limitations,modifications, and/or clarifications to the implementation requirements of ASMESection XI. These Paragraphs in IOCFR50.55a that are applicable to CPS aredetailed in Table 1.8-1.

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TABLE 1.8-1CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS

10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications

1 OCFR50.55a(b)(2)(viii)(E) (CISI) Examination of concrete containments: For Class CCapplications, the licensee shall evaluate the acceptability ofinaccessible areas when conditions exist in accessible areasthat could indicate the presence of or result in degradation tosuch inaccessible areas. For each inaccessible area identified,the licensee shall provide the following in the ISI SummaryReport required by IWA-6000:(1) A description of the type and estimated extent of

degradation, and the conditions that led to thedegradation;

(2) An evaluation of each area, and the result of theevaluation, and;

(3) A description of necessary corrective actions.

1OCFR50.55a(b)(2)(viii)(F) (CISI) Examination of concrete containments. Personnelthat examine containment concrete surfaces and tendonhardware, wires, or strands must meet the qualificationprovisions in IWA-2300. The "owner-defined" personnelqualification provisions in IWL-23 10(d) are not approved foruse.

1OCFR50.55a(b)(2)(ix)(A) (CISI) Examination of metal containments and the liners ofconcrete containments. For Class MC applications, thelicensee shall evaluate the acceptability of inaccessible areaswhen conditions exist in accessible areas that could indicatethe presence of or result in degradation to such inaccessibleareas. For each inaccessible area identified, the licensee shallprovide the following in the ISI Summary Report as requiredby IWA-6000:(1) A description of the type and estimated extent of

degradation, and the conditions that led to thedegradation;

(2) An evaluation of each area, and the result of theevaluation, and;

1 (3) A description of necessary corrective actions.

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I 0CFR50.55a(1,)(2)(ix)(B) (CISI) Examination of metal containments and the liners ofconcrete containments: When performing remotely the visualexaminations required by Subsection IWE, the maximumdirect examination distance specified in Table IWA-2210-1may be extended and the minimum illumination requirementsspecified in Table IWA-2210-1 may be decreased providedthat the conditions or indications for which the visualexamination is performed can be detected at the chosendistance and illumination.

1OCFR50.55a(b)(2)(ix)(F) (CISI) Examination of metal containments and the liners ofconcrete containments. VT-I and VT-3 examinations mustbe conducted in accordance with IWA-2200. Personnelconducting examinations in accordance with the VT-I orVT-3 examination method shall be qualified in accordancewith IWA-2300. The "owner-defined" personnel qualificationprovisions in IWE-2330(a) for personnel that conduct VT-1and VT-3 examinations are not approved for use.

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

I examinations.

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I OCFR50.55a Paragraphs Limitations, Modifications, and Clarifications

I OCFR50.55a(b)(2)(ix)(H) (CISI) Examination of metal containments and the liners ofconcrete containments: Containment bolted connections thatare disassembled during the scheduled performance of theexaminations in Item El .11 of Table IWE-2500-1 must beexamined using the VT-3 examination method. Flaws ordegradation identified during the performance of a VT-3examination must be examined in accordance with the VT- 1examination method. The criteria in the material specificationor IWB-3517.1 must be used to evaluate containment boltingflaws or degradation. As an alternative to performing VT-3examinations of containment bolted connections that aredisassembled during the scheduled performance of ItemEl. 11, VT-3 examinations of containment bolted connectionsmay be conducted whenever containment bolted connectionsare disassembled for any reason.

IOCFR50.55a(b)(2)(ix)(I) (CISI) Examination of metal containments and the liners ofconcrete containments: The ultrasonic examinationacceptance standard specified in IWE-3511.3 for Class MCpressure-retaining components must also be applied tometallic 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 throughthe latest Edition and Addenda incorporated by reference inParagraph (b)(2) of this section, are not approved for use onirradiated material.

10CFR50.55a(b)(2)(xviii)(A) (ISI) Certification of NDEpersonnel: Level I and IInondestructive examination personnel shall be recertified on a3-year interval in lieu of the 5-year interval specified in the1997 Addenda and 1998 Edition of IWA-2314, andIWA-2314(a) and IWA-2314(b) of the 1999 Addenda throughthe latest Edition and Addenda incorporated by reference inparagraph (b)(2) of this section.

1 OCFR50.55a(b)(2)(xviii)(B) (ISI) Certification of NDE personnel: Paragraph IWA-2316of the 1998 Edition through the latest Edition and Addendaincorporated by reference in paragraph (b)(2) of this section,may only be used to qualify personnel that observe forleakage during system leakage and hydrostatic tests conductedin accordance with IWA-52 11 (a) and (b), 1998 Editionthrough the latest Edition and Addenda incorporated byreference in paragraph (b)(2) of this section.

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IS1 Program PlanClinton Power Station Unit 1, Third Interval



1 OCFR50.55a(b)(2)(xviii)(C) (ISI) Certification of NDEpersonnel: When qualifyingvisual examination personnel for VT-3 visual examinationsunder paragraph IWA-2317 of the 1998 Edition through thelatest Edition and Addenda incorporated by reference inparagraph (b)(2) of this section, the proficiency of the trainingmust be demonstrated by administering an initial qualificationexamination and administering subsequent examinations on a3-year interval.

1OCFR50.55a(b)(2)(xix) (ISI) Substitution of alternative methods: The provisions forthe substitution of alternative examination methods, acombination of methods, or newly developed techniques inthe 1997 Addenda of IWA-2240 must be applied. Theprovisions in IWA-2240, 1998 Edition through the latestEdition and Addenda incorporated by reference in paragraph(b)(2) of this section, are not approved for use. Theprovisions in IWA-4520(c), 1997 Addenda through the latestEdition and Addenda incorporated by reference in paragraph(b)(2) of this section, allowing the substitution of alternativeexamination methods, a combination of methods, or newlydeveloped techniques for the methods specified in theConstruction Code are not approved for use.

10CFR50.55a(b)(2)(xx)(B) (ISI) System leakage tests: The NDE provision inIWA-4540(a)(2) of the 2002 Addenda of Section XI must beapplied when performing system leakage tests after repair andreplacement activities performed by welding or brazing on apressure retaining boundary using the 2003 Addenda throughthe latest Edition and Addenda incorporated by reference inparagraph (b)(2) of this section.

10CFR50.55a(b)(2)(xxi)(B) (ISI) Table IWB-2500-1 examination requirements.- Theprovisions of Table IWB-2500-1, Examination CategoryB-G-2, Item B7.80, that are in the 1995 Edition are applicableonly to reused bolting when using the 1997 Addenda throughthe latest Edition and Addenda incorporated by reference inparagraph (b)(2) of this section.

IOCFR50.55a(b)(2)(xxii) (ISI) Surface Examination: The use of the provision inIWA-2220, "Surface Examination," of Section XI, 2001Edition through the latest Edition and Addenda incorporatedby reference in paragraph (b)(2) of this section, that allow useof an ultrasonic examination method is prohibited.

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TABLE 1.8-1CODE OF FEDERAL REGULATIONS 1OCFR50.55a REQUIREMENTS


1OCFR50.55a(b)(2)(xxiii) (ISI) Evaluation of Thermally Cut Surfaces: The use of theprovisions for eliminating mechanical processing of thermallycut surfaces in IWA-4461.4.2 of Section XI, 2001 Editionthrough the latest Edition and Addenda incorporated byreference in Paragraph (b)(2) of this section are prohibited.

1 OCFR50.55a(b)(2)(xxiv) (PDI) Incorporation of the Performance DemonstrationInitiative and Addition of Ultrasonic Examination Criteria.The use of Appendix VIII and the supplements to AppendixVIII and Article 1-3000 of Section XI of the ASME BPVCode, 2002 Addenda through the latest Edition and Addendaincorporated by reference in Paragraph (b)(2) of this section,is prohibited.

I OCFR50.55a(b)(2)(xxv) (ISI) Mitigation of Defects by Modification: The use of theprovisions in IWA-4340, "Mitigation of Defects byModification," Section XI, 2001 Edition through the latestEdition and Addenda incorporated by reference in Paragraph(b)(2) of this section are prohibited.

1 OCFR50.55a(b)(2)(xxvi) (SPT) Pressure Testing Class 1, 2, and 3 Mechanical Joints:The repair and replacement activity provisions inIWA-4540(c) of the 1998 Edition of Section XI for pressuretesting Class 1, 2, and 3 mechanical joints must be appliedwhen using the 2001 Edition through the latest Edition andAddenda incorporated by reference in Paragraph (b)(2) of thissection.

1OCFR50.55a(b)(2)(xxvii) (ISI) Removal of Insulation: When performing visualexaminations in accordance with IWA-5242 of Section XI,2003 Addenda through the latest Edition and Addendaincorporated by reference in paragraph (b)(2) of the section,insulation must be removed from 17-4 PH or 410 stainlesssteel studs or bolts aged at a temperature below 1100 'F orhaving a Rockwell Method C hardness value above 30, andfrom A-286 stainless steel studs or bolts preloaded to 100,000pounds per square inch or higher.

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).55a Paragraphs Limitations, Modifications, and Clarifications

10CFR50.55a(b)(3)(v) (ISI) Subsection ISTD: Article IWF-5000, "InserviceInspection Requirements for. Snubbers," of the ASME BPVCode, Section XI, provides inservice inspection requirementsfor examinations and tests of snubbers at nuclear powerplants. Licensees may use Subsection ISTD, "InserviceTesting of Dynamic Restraints (Snubbers) in Light-WaterReactor Power Plants," ASME OM Code, 1995 Editionthrough the latest Edition and Addenda incorporated byreference in paragraph (b)(3) of this section, in place of therequirements for snubbers in Section XI, IWF-5200(a) and (b)and IWF-5300(a) and (b), by making appropriate changes totheir technical specifications or licensee-controlleddocuments. Preservice and inservice examinations must beperformed using the VT-3 visual examination methoddescribed in IWA-2213.

IOCFR50.55a(b)(5) (ISI) Inservice Inspection Code Cases: Licensees may applythe ASME Boiler and Pressure Vessel Code Cases listed inRegulatory Guide 1.147 through Revision 15, without priorNRC approval subject to the following:(i) When a licensee initially applies a listed Code Case, thelicensee shall apply the most recent version of that Code Caseincorporated by reference in this paragraph.(ii) If a licensee has previously applied a Code Case and alater version of the Code Case is incorporated by reference inthis paragraph, the licensee may continue to apply, to the endof the current 120-month interval, the previous version of theCode Case as authorized or may apply the later version of theCode Case, including any NRC-specified conditions placedon its use.(iii) Application of an annulled Code Case is prohibitedunless a licensee previously applied the listed Code Case priorto it being listed as annulled in Regulatory Guide 1.147. AnyCode Case listed as annulled in any Revision of RegulatoryGuide 1. 147 which a licensee has applied prior to it beinglisted as annulled, may continue to be applied by that licenseeto the end of the 120-month interval in which the Code Case

I was implemented.

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1OCFR50.55a(b)(6) (ISl) Operation and Maintenance of Nuclear Power PlantsCode Cases: Licensees may apply the ASME Operation andMaintenance Nuclear Power Plants Code Cases listed inRegulatory Guide 1.192 without prior NRC approval subjectto the following:(i) When a licensee initially applies a listed Code Case, thelicensee shall apply the most recent version of that Code Caseincorporated by reference in this paragraph.(ii) If a licensee has previously applied a Code Case and alater version of the Code Case is incorporated by reference inthis paragraph, the licensee may continue to apply, to the endof the current 120-month interval, the previous version of theCode Case as authorized or may apply the later version of theCode Case, including any NRC-specified conditions placedon its use.(iii) Application of an annulled Code Case is prohibitedunless a licensee previously applied the listed Code Case priorto it being listed as annulled in Regulatory Guide 1.192. If alicensee has applied a listed Code Case that is later listed asannulled in Regulatory Guide 1.192, the licensee maycontinue to apply the Code Case to the end of the current 120-month interval.

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IS! Program PlanClinton Power Station Unit 1, Third Interval

1.9 Code Cases

Per I OCFR50.55a(b)(5) and (b)(6), ASME Code Cases that have been determinedto be suitable for use in ISI Program Plans by the NRC are listed in RegulatoryGuide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI,Division I ". The approved Code Cases in Regulatory Guide 1.147, which arebeing utilized by CPS, are included in Section 2.1.1. The most recent version of agiven Code Case incorporated in the revision of Regulatory Guide 1.147referenced in 1 OCFR50.55a(b)(5)(i) at the time it is applied within the ISIProgram shall be used. The latest version of Regulatory Guide 1.147 incorporatedinto this document is Revision 15. As this guide is revised, newly approved CodeCases should be assessed for plan implementation at CPS per Paragraph1WA-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 beauthorized by the Director of the office of Nuclear Reactor Regulation uponrequest pursuant to 1OCFR50.55a(a)(3). Code Cases not approved for use inRegulatory Guide 1.147, which are being utilized by CPS through associatedrelief requests, are included in Section 8.0.

This ISI Program Plan will also utilize Regulatory Guide 1.192, "Operation andMaintenance Code Case Acceptability, ASME OM Code". The approved CodeCases in Regulatory Guide 1.192, which are being utilized by CPS, are includedin Section 2.1.2. The latest version of Regulatory Guide 1.192 incorporated intothis document is Revision 0. As this guide is revised, newly approved Code Casesshould 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 1OCFR50.55a, when a licensee either proposes alternatives toASME Section XI requirements which provide an acceptable level of quality andsafety, determines compliance with ASME Section XI requirements would resultin hardship or unusual difficulty without a compensating increase in the level ofquality and safety, or determines that specific ASME Section XI requirements forinservice inspection are impractical, the licensee shall notify the NRC and submitinformation to support the determination.

The submittal of this information will be referred to in this document as a "ReliefRequest". Relief Requests for the Third ISI Interval and the Second CISI Intervalare included in Section 8.0 of this document. The text of the Relief Requestscontained in Section 8.0 will demonstrate one of the following: the proposedalternatives provide an acceptable level of quality and safety per10CFR50.55a(a)(3)(i), compliance with the specified requirements would result inhardship or unusual difficulty without a compensating increase in the level of

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qual ity and safety per 1 OCFR50.55a(a)(3)(ii), or the code requirements areconsidered impractical per 1OCFR50.55a(g)(5)(iii).

Per lOCFR5O.55a Paragraphs (a)(3) and (g)(6)(i), the Director of the Office ofNuclear Reactor Regulation will evaluate relief requests and "may grant suchrelief and may impose such alternative requirements as it determines is authorizedby law and will not endanger life or property or the common defense and securityand is otherwise in the public interest giving due consideration to the burden uponthe licensee that could result if the requirements were imposed on the facility".


ISI Program Plan


2.0 BASIS FOR INSERVICE INSPECTION PROGRAM

2.1 ASME Section XI Examination Requirements

As required by the 1OCFR50.55a, this Program was developed in accordance withthe requirements detailed in the 2004 Edition, No Addenda, of the ASME Boilerand Pressure Vessel Code, Section XI, Division 1, Subsections IWA, IWB, IWC,IWD, IWE, IWF, IWL, Mandatory Appendices, Inspection Program B ofParagraph IWA-2432, approved ASME Code Cases, and approved alternativesthrough relief requests and Safety Evaluation Reports (SER's).

Performance Demonstration Initiatives (PDI) is an organization comprised of all USnuclear utilities that was formed to provide an efficient implementation of AppendixVIII performance demonstration requirements. The Electric Power ResearchInstituWe (EPRI) NDE Center was selected as the administrator of this program. ThePD1 program is administered according to the "PDI Program Description". The ISIPi.-g.'am implements Appendix VIII "Performance Demonstration for UltrasonicExamination Systems," ASME Section XI 2001 Edition, No Addenda as requiredby 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 ofthe procedures, personnel, and equipment used to detect and size flaws in piping,bolting, and the reactor pressure vessel (RPV). Each organization (e.g., owner orvendor) will be required to have a written program to ensure compliance with therequirements. These requirements were initially implemented through thePerformance Demonstration Initiative (PDI) Program according to the scheduledefined in 10CFR50.55a(g)(6)(ii)(C). CPS does not have in-house capabilities toperform ultrasonic examinations and intends to utilize NDE contractors to performultrasonic examinations. However, CPS still has the responsibility to ensure thatAppendix VIII requirements are properly implemented.

For the Third ISI Interval, CPS's inspection program for ASME Section XIExamination 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 EPRITopical 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 forthe classification of piping structural elements under the RISI Program. The RISIProgram scope has been implemented as an alternative to the 2004 Edition, NoAddenda of the ASME Section XI examination program for ISI Class 1 B-F and B-Jwelds and ISI Class 2 C-F-1 and C-F-2 welds in accordance withIOCFR50.55a(a)(3)(i). The basis for the resulting Risk Categorizations of thenonexempt ISI Class I and 2 piping systems at CPS is defined and maintained in theFinal Report "Risk Informed Inservice Inspection Evaluation" as referenced inSection 9.0 of this document.

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

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is described in EPRI Topical Report TR-1006937, Rev. 0-A, which has been used todefine the inspection scope in lieu of the 100% volumetric examination of all pipingwelds in the previous BER augmented program. Therefore, all welds in the originalaugmented program for BER remain evaluated under the RISI program using anintegrated 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 thisdoctument.

2.1.1 ASME Section XI Code Cases

As referenced by 1 OCFR50.55a(b)(5) and allowed by NRC RegulatoryGuide 1.147, Revision 15, the following Code Cases are beingincorporated into the CPS ISI Program. Several of these Code Cases areincluded as contingencies, to ensure that they are available for futurerepair/replacement activities.

N-432-1 Repair Welding Using Automatic or Machine GasTungsten-Arc Welding (GTAW) Temper Bead Technique.Regulatory Guide 1.147, Revision 15.

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

N-504-3 Alternative Rules for Repair of Class 1, 2, and 3 AusteniticStainless Steel Piping

Code Case N-504-3 is acceptable subject to the followingcondition specified in Regulatory Guide 1.147, Revision15:

The provisions of Section XI, Nonmandatory AppendixQ, "Weld Overlay Repair of Class 1, 2, and 3 AusteniticStainless Steel Piping Weldments," must also be met.

N-516-3 Underwater Welding


Licensee must obtain NRC approval in accordance with1 OCFR50.55a(a)(3) regarding the technique to be usedin the weld repair or replacement of irradiated materialunderwater.

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N-517-1 Quality Assurance Program Requirements for Owners.Regulatory Guide 1.147, Revision 15.

N-526 Alternative Requirements for Successive Inspections ofClass I and 2 Vessels. Regulatory Guide 1.147, Revision15.

N-528-1 Purchase, Exchange, or Transfer of Material BetweenNuclear Plant Sites


The requirements of 1 OCFR Part 21, "Reporting ofDefects and Noncompliance", are to be applied to thenuclear plant site supplying the material as well as tothe nuclear plant site receiving the material that hasbeen purchased, exchanged, or transferred betweensites.

N-532-4 Alternative Requirements to Repair and ReplacementDocumentation Requirements and Inservice SummaryReport Preparation and Submission as Required byIWA-4000 and IWA-6000. Regulatory Guide 1.147,Revision 15.

N-552 Alternative Methods - Qualification for Nozzle InsideRadius Section from the Outside Surface

Code Case N-552 is acceptable subject to the followingconditions specified in Regulatory Guide 1.147, Revision15:

To achieve consistency with the 1OCFR50.55a rulechange published September 22, 1999 (64 FR 51370),incorporating Appendix VIII, "PerformanceDemonstration for Ultrasonic Examination Systems," toASME Section XI, add the following to the specimenrequirements:

"At least 50 percent of the flaws in the demonstrationtest set must be cracks and the maximum misorientationmust be demonstrated with cracks. Flaws in nozzleswith bore diameters equal to or less than 4 inches maybe notches."

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Add to detection criteria, "The number of false callsmust not exceed three."

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

N-586-1 Alternative Additional Examination Requirements forClass 1, 2, and 3 Piping, Components, and Supports.Regulatory Guide 1.147, Revision 15.

N-597-2 Requirements for Analytical Evaluation of Pipe WallThinning

Code Case N-597-2 is acceptable subject to the followingconditions specified in Regulatory Guide 1.147, Revision15:

(1) Code Case must be supplemented by the provisionsof EPRI Nuclear Safety Analysis Center Report202L-R2, April 1999, "Recommendations for anEffective Flow Accelerated Corrosion Program," fordeveloping the inspection requirements, the methodof predicting the rate of wall thickness loss, and thevalue of the predicted remaining wall thickness. Asused in NSAC-202L-R2, the term "should" is to beapplied as "shall" (i.e., requirement).

(2) Components affected by flow-accelerated corrosionto which this Code Case are applied must berepaired or replaced in accordance with theconstruction code of record and Owner'srequirements or a later NRC approved Edition ofSection III, "Rules for Construction of NuclearPlant Components," of the ASME Code prior to thevalue of tp reaching the allowable minimum wallthickness, tri,, as specified in -3622.1 (a)(1) of thisCode Case. Alternatively, use of the Code Case issubject to NRC review and approval per1OCFR50.55a(a)(3).

(3) For Class 1 piping not meeting the criteria of -3221,the use of evaluation methods and criteria is subjectto NRC review and approval per1OCFR50.55a(a)(3).

(4) For those components that do not require immediaterepair or replacement, the rate of wall thickness lossis to be used to determine a suitable inspection

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frequency so that repair or replacement occurs priorto reaching allowable minimum wall thickness, t,,,n.

(5) For corrosion phenomenon other than flowaccelerated corrosion, use of the Code Case issubject to NRC review and approval. Inspectionplans and wall thinning rates may be difficult tojustify for certain degradation mechanisms such asMIC and pitting per 1OCFR50.55a(a)(3).

N-600 Transfer of Welder, Welding Operator, Brazer, and BrazingOperator Qualifications Between Owners. RegulatoryGuide 1.147, Revision 15.

N-606-1 Similar and Dissimilar Metal Welding Using AmbientTemperature Machine GTAW Temper Bead Technique forBWR CRD Housing/Stud Tube Repairs


Prior to welding, an examination or verification must beperformed to ensure proper preparation of the basemetal, and that the surface is properly contoured so thatan acceptable weld can be produced. The surfaces to bewelded, and surfaces adjacent to the weld, are to be freefrom contaminants' such as, rust, moisture, grease, andother foreign material or any other condition that wouldprevent proper welding and adversely affect the qualityor strength of the weld. This verification is to berequired in the welding procedures.

N-613-1 Ultrasonic Examination of Full Penetration Nozzles inVessels, Examination Category B-D, Item No's. B3.10 andB3.90, 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, Revision15.

N-629 Use of Fracture Toughness Test Data to EstablishReference Temperature for Pressure Retaining Materials.Regulatory Guide 1.147, Revision 15.


IS Program PlanClinton Power Station Unit 1, Third Interval

N-63 8-1 Similar and Dissimilar Metal Welding Using AmbientTemperature Machine GTAW Temper Bead Technique


UT examinations shall be performed with personnel andprocedures qualified for the repaired volume andqualified by demonstration using representative sampleswhich contain construction type flaws. The acceptancecriteria of NB-5330 in the 1998 Edition through the2000 Addenda of Section III Edition and Addendaapproved in 10CFR50.55a apply to all flaws identifiedwithin the repaired volume.

N-639 Alternative Calibration Block Material


Chemical ranges of the calibration block may vary fromthe materials specification if (1) it is within thechemical range of the component specification to beinspected, and (2) the phase and grain shape aremaintained in the same ranges produced by the thermalprocess required by the material specification.

N-641 Alternative Pressure-Temperature Relationship and LowTemperature Overpressure Protection SystemRequirements. Regulatory Guide 1.147, Revision 15.

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

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

N-661 Alternative Requirements for Wall Thickness Restorationof Classes 2 and 3 Carbon Steel Piping for Raw WaterService




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

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

(c) When through-wall repairs are made by welding onsurfaces that are wet or exposed to water, the weldoverlay repair is only acceptable until the next refuelingoutage.

N-665 Alternative Requirements for Beam Angle MeasurementsUsing 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 withthose approved for use in the latest published revision of Regulatory Guide1.147 or 1OCFR50.55a at that time.

2.1.2 ASME OM Code Cases

As referenced by 1OCFR50.55a(b)(6) and allowed by NRC RegulatoryGuide 1.192, Revision 0, the following Code Cases are being incorporatedinto the CPS ISI Program:

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

Additional Code Cases invoked in the future shall be in accordance withthose approved for use in the latest published revision of Regulatory Guide1.192 or 1OCFR50.55a at that time.

2.2 Augmented Examination Requirements

Augmented examination requirements are those examinations that are performedabove and beyond the requirements of ASME Section XI. Below is a summary ofthose examinations performed by CPS that are not specifically addressed byASME Section XI, or the examinations that will be performed in addition to therequirements of ASME Section XI on a routine basis during the Third ISI Interval



and Second CISI Interval. Previous revisions of the CPS ISI Program categorizedsome 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 OutsideContainment", discusses protection against postulated piping failures influid systems outside containment, and includes requirements for licenseesto perform 100% volumetric examination of circumferential andlongitudinal pipe welds within the pipe break exclusion areas associatedwith high energy piping in containment penetration areas.

CPS has committed to the requirements of the NRC MEB TechnicalPosition 3-1 through letters to the NRC. Updated Safety Analysis Report(USAR) MEB [Draft Safety Evaluation Report (DSER)] Item No. 11 forISI Class 1 and USAR Section 6.6.8 for ISI Class 2, detail CPS'scompliance with NRC MEB Technical Position 3-1. Examination isrequired for all piping welds between containment isolation valves. (Forthose systems which do not have an inboard valve designated as acontainment isolation valve per CPS Technical Specification Table3.6.4-1, the first valve inside the containment shall be considered thepenetration boundary in satisfying this requirement) as follows:

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

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

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

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

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



Additional guidance for adaptation of the RISI evaluation process to BERpiping 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% ofthin wall (between 3/8" and 1/2") ISI Class 2 RHR system piping weldswhich would require only surface examinations per ASME Section XI.The 2004 Edition, No Addenda of ASME Section XI requires CPS toperform volumetric examination of thin wall ISI Class 2 system pipingwelds. Therefore, the augmented requirements of performing volumetricexamination of 10% of thin wall ISI Class 2 RHR system piping welds hasbeen met by this Third ISI Interval ISI Program Plan.


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

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

2.2.3 Boiling Water Reactor Owners' Group (BWROG) ReportGE-NE-523-A71-0594-A, Revision 1, "Alternate BWR Feedwater NozzleInspection Requirements, May 2000," as approved by NRC final SERdated March 10, 2000, Boiling Water Reactor Owners' Group (BWROG)Report GE-NE-523-A71-0594, "Alternate BWR Feedwater NozzleInspection Requirements, August 1999," as conditionally approved byNRC final SER dated June 5, 1998, and NRC NUREG 0619, BWRFeedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking,dated November 1980

These documents discuss the examination requirements for BWRFeedwater (FW) Nozzle and Control Rod Drive (CRD) Return LineNozzle Cracking. The alternate approach was developed and submitted tothe NRC by the BWROG. The NRC conditionally accepted thesealternate requirements in the BWROG, Safety Evaluation of ProposedAlternative to BWR Feedwater Nozzle Inspections, dated June 5, 1998.

The CPS requirement in the USAR requires the FW nozzles and the CRDretuin line nozzle, which is capped, to be examined using the methods,techniques, and frequency outlined in the initial examination requirements



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 byNRC SER (TAC NO. MA6787) dated March 10, 2000.


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

2.2.4 Generic Letter 88-01, "NRC Position on IGSCC in BWR AusteniticStainless Steel Piping," Revision 2 / Supplement 1 to Generic Letter88-01, NUREG 0313, "Technical Report on Material Selection andProcess Guidelines for BWR Coolant Pressure Boundary Piping,"Revision 2, EPRI Topical Report TR-1 13932, "BWR Vessel and InternalsProject, Technical Basis for Revisions to Generic Letter 88-01 InspectionSchedules (BWRVIP-75)," as conditionally approved by NRC final SER'sdated September 15, 2000 and May 14, 2002, and EPRI Topical ReportTR- 1012621, "BWR Vessel and Internals Project, Technical Basis forRevisions to Generic Letter 88-01 Inspection Schedules(BWRVIP-75-A)," as conditionally approved by NRC final SER datedMarch 16, 2006

These documents discuss the examination requirements for IntergranularStress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless SteelPiping. References to Generic Letter 88-01 (GL 88-01) within the ISIProgram refer to the comprehensive requirements to all of thesedocuments. The final SER's of BWRVIP-75 and BWRVIP-75-A revisedthe GL 88-01 inspection schedules. The BWRVIP-75 and BWRVIP-75-Arevised inspection schedules were based on consideration of inspectionresults and service experience gained by the industry since issuance of GL88-01 and NUREG-0313, and includes additional knowledge regarding thebenefits of improved BWR water chemistry.

The original CPS responses concerning Generic Letter 88-01 were sentthrough letters to the NRC. Austenitic stainless steel piping componentssusceptible to IGSCC shall be examined in accordance with CPS(AmerGen) response to Generic Letter 88-01, NRC Position on IGSCC inBWR Austenitic Stainless Steel Piping and NRC Request for AdditionalInformation - 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 ofthe ASME Section XI meets the GL 88-01 requirements.

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Since the issuance of GL 88-01, the BWR Vessels and Internals Project(BWRVIP) has been created. This BWR owners group has worked on themitigation of IGSCC for BWR internal components. As part of theiractivities, EPRI Topical Report TR-1 13932, "BWR Vessel and InternalsProject, Technical Basis for Revisions to Generic Letter 88-01 InspectionSchedules (BWRVIP-75), dated October 27, 1999" and EPRI TopicalReport TR- 1012621, "BWR Vessel and Internals Project, Technical Basisfor 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 inspectionschedules for IGSCC susceptible welds. Two different inspectionschedules 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 performancecriteria are met.

After review of BWRVIP-75 and BWRVIP-75-A, the NRC issued aSafety Evaluation Report (SER) approving the documents with minorchanges. (Letter from NRC to Carl Terry, BWRVIP Chairman, FinalSafety Evaluation of the "BWR Vessel and Internals Project, TechnicalBasis 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 andInternals Project, Technical Basis for Revisions to Generic Letter 88-01Inspection Schedules (BWRVIP-75-A)", dated March 16, 2006.)

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

RISI regulations are being invoked for CPS in this ISI Program Plan.Under these new guidelines, ISI Class I and 2 piping structural elementsare 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 TopicalReports and this Code Case, welds within the plant that are assigned toIGSCC Categories B through G will continue to meet existing IGSCCschedules, while IGSCC Category A welds have been subsumed into theRISI Program. (CPS currently has only IGSCC Category D welds.)

Alion Science & Technologv 2-11 CLN05. GO3U•

Revision 14


Implementation of the CPS program addressing these documents isincluded in Exelon Nuclear Procedure ER-AA-330-002, Section 7.0 of thisISI Program Plan, and the associated ISI Database.

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

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

Previous revisions of the CPS ISI Program Plan classified augmentedexaminations of IGSCC welds as Augmented Inspection Programs F17.2-Iand 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 weldwhere alloy 182 buttering is applied and extended into the nozzle bore areaas recommended in CPS response to General Electric Service InformationLetter (GE SIL) No. 455, Rev. 1 (Memorandum from F. A. Spangenbergto File, Y-207823, dated June 13, 1988). Also, CPS will incorporate theultrasonic testing technique and recommendations on repair processes(when required) for nozzle-to-safe end welds where alloy 182 buttering isapplied as recommended in IP response to GE SIL No. 455, Revision 1Supplement I (Memorandum from R. D. Freeman to D. L. Holtzscher,Y-92355, dated September 25, 1989).


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

2.2.6 NRC Regulatory Guide 1.150, Revision 1, Appendix A, "UltrasonicTesting of Reactor Vessel Welds During Preservice and InserviceExamination"

This Regulatory Guide includes inspection requirements for the ultrasonicexamination of RPV welds during preservice and inservice examinations.The examination criteria of Regulatory Guide 1.150 supplements therequirements of ASME Section XI. However, 1OCFR50.55a 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 withAppendix VIII, 2001 Edition, No Addenda. The prescriptive guidance in



Regulatory Guide 1.150 is not in total agreement with the Appendix VIIIrequirements; therefore, Regulatory Guide 1.150 remains applicable onlyfor the examination of the RPV components that are not examined inaccordance with Appendix VIII per 10CFR50.55a.

Therefore, CPS examinations of the RPV that are performed using bothmanual and mechanized examination techniques from the outside surfaceof the vessel shall be in accordance with Appendix VIII of the ASMESection 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 ofBWR Scram System Piping, Section 5.1

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

2.2.8 1 OCFR50.55a(g)(6)(ii)(A), Augmented Examination of Reactor PressureVessel

Effective September 8, 1992, 1OCFR50.55a(g)(6)(ii)(A) requiredimplementation 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 forCPS. Accordingly, CPS was required to satisfy this rule by the end of theFirst ISI Interval or to propose an alternative examination program forNRC approval. The rule required examination of "essentially 100%" of allvessel shell welds by the end of the First ISI Interval. The CPS ISIProgram First ISI Interval RPV shell weld examinations were conducted inaccoi dance with the NRC augmented examination requirements.

The examinations of RPV shell welds, Examination Category B-A, ItemNumber B 1.11, at CPS, will be conducted in accordance with previouslysubmitted and approved Second ISI Interval Relief Request 4215. Theapproval authorized under NRC SER dated December 30, 2009 for CPSwas 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.


ISI Program Plan


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 containmentstructures (metal and concrete). Below is a summary of the classification criteriaused within the ISI Classification Basis Document.

Each safety related, fluid system containing water, steam, air, oil, etc. included inthe CPS USAR was reviewed to determine which safety functions they performduring all modes of system and plant operation. Based on these safety functions,the systems and components were evaluated per classification documents. Thesystems 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 theinterfaces (boundary valves and boundary barriers) meet the criteria set by1OCFR50.2, 10CFR50.55a(c)(1), 1OCFR50.55a(c)(2), and Regulatory Guide 1.26,Revision 3. SRP 3.2.2 and ANSI/ANS-58.14-1993 (CPS is not committed to orlicensed in accordance with these documents) were also used for guidance inevaluating the classification boundaries where 1 OCFR and the Regulatory Guidedid not address a given situation.

Components within the reactor coolant pressure boundary, as defined byI OCFR50.2, are typically designated as ISI Class 1 while the other safety relatedcomponents are evaluated for ISI Class 2 or 3 designation in accordance with theguidelines of Regulatory Guide 1.26. Per Regulatory Guide 1.26 Paragraphs Aand B, the ISI Class 2 and 3 boundaries are limited to safety related systems andcomponents. Where sufficient classification criteria is not provided within1OCFR50 or Regulatory Guide 1.26, other industry documents such asNUREG-0800 and ANSI/ANS standards are consulted "for guidance".

According to 1OCFR50.55a, Paragraph (g)(4), the ISI requirements of ASMESection XI are assigned to these components, within the constraints of existingplant design. The CPS ISI Class 1, 2, and 3 components that are exempt fromexamination are those which meet the criteria of ASME Section XI, ParagraphsIWB-1220, IWC-1220, and IWD-1220. Supports which meet the criteria ofParagraph IWF-1230 of ASME Section XI are also exempt from examination.For Containment, Class MC components which meet Paragraph IWE- 1220 areexempt from examination, and Class CC components which meet ParagraphIWL-1220 are exempt from examination.

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

Alion Science & Technoloiv 2-14 CLN05. GO3Revision 14


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 theThird ISI Interval at CPS. Appendix B to this ISI Program Plan includes acomponent list which provides those subject to examination as well as somewhich are exempt from examination. One (1) inch or less nominal pipe size ISIClass 1 components and four (4) inches or less nominal pipe size ISI Class 2 and 3components, except as required by an augmented examination, have not beenincluded in Appendix B, since these piping lines are all exempted by ASMESection XI.



TABLE 2.3-1P&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)



2.4 ISI Isometric and Component Drawings for Nonexempt ISI Class Componentsand 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 componentlocations are identified on the ISI Isometric and Component Drawings listed inTable 2.4-1. The CISI Class MC and CC components are identified on the CISIDrawings Diagram, Specification, and Procedures listed in Table 2.4-2.

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

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

2-17 CLN05. G03Revision 14


TABLE 2.4-1ISI 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 NAA-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 1 E22-COO1

DET-2-1 B-82 RESIDUAL HEAT REMOVAL I RH87AA(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"




FIGURE NO. FIGURE NO. SYSTEM LINE NO._(OLD)

FW-02-2 A-8 FEEDWATER 1 FW-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 I FW-02KB-20"

FW-03A-4 B-4 FEEDWATER 1 FW-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"


IS! Program PlanClinton Power Station Unit]1, Third Interval


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 I LP-02E- 10"

LP-04-3 B-101B LOW PRESSURE CORE SPRAY I LP-02A- 12"

LP-04-4 B-10A LOW PRESSURE CORE SPRAY I LP-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 I MS-A-24"

MS-A-2 A-4B MAIN STEAM IMS-ASA(ASB)-10"

MS-B-1 A-5A MAIN STEAM I MS-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 I MS-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"


IS! Program Plan



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 I RH-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"




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-09-9 B-28 RESIDUAL HEAT REMOVAL 1RH-06A-16"




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"




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-14-3 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-03-3 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"




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 I SC02DA(DD)-3"I SC03DB-3"

1 SC02DE-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"


1Sf Program PlanClinton Power Station Unit 1, Third Interval


FIGURE NO. FIGURE NO. SYSTEM LINE NO.(OLD) T

SD-90-2 B-75 SCRAM DISCHARGE VOLUME 90 DEGREE SIDE-12"



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IS Program PlanClinton Power Station Unit 1, Third Interval

TABLE 2.4-2CISI DRAWINGS, DIAGRAM, SPECIFICATION, AND PROCEDURES

DRAWING NUMBER TITLE OF DRAWING/DIAGRAM/SPECIFICATION

S27-1905, Sht. 1 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 ContainmentStructures

ER-AA-330-007 Visual Examination of Section XI Class MC Surfaces and Class CCLiners

ER-AA-335-018 Detailed, General, VT-I, VT-IC, VT-3 and VT-3C, VisualExamination of ASME Class MC and CC Containment Surfacesand

Components



2.5 Technical Approach and Positions

When the requirements of ASME Section XI are not easily interpreted, CPS hasreviewed general licensing/regulatory requirements and industry practice todetermine a practical method of implementing the Code requirements. TheTechnical Approach and Position (TAP) documents contained in this section havebeen provided to clarify CPS's implementation of ASME Section XI requirements.An index which summarizes each technical approach and position is included inTable 2.5-1.

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TABLE 2.5-1TECHNICAL APPROACH AND POSITIONS INDEX

Position Revision Status' (Program) Description of Technical ApproachNumber Date2 and Position

13T-0 0 Active (SPT) System Leakage Testing of Non-Isolable05/21/10 Buried Components.

0 (SPT) Valve Seats/Disks as Pressurization05/21/10 Boundaries.

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

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



TECHNICAL APPROACH AND POSITION NUMBER 13T-01Revision 0

COMPONENT IDENTIFICATION:

Code Class: 2 and 3Reference: IWA-5244(b)(2)Examination Category: C-H, D-BItem Number: C7.10, D2.10Description: System Leakage Testing of Non-Isolable Buried ComponentsComponent Numbei: Non-Isolable Buried Pressure Retaining Components

CODE REOUIREMENT:

Paragraph IWA-5244(b)(2) requires non-isolable buried components be tested to confirm that flowduring 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 thefollowing acceptance criteria:

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

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

For open ended portions of systems where the process fluid is pneumatic, evidence ofgaseous discharge shall be considered reasonable proof of adequate flow through the openended line. Such test may include passing smoke through the line, hanging balloons orstreamers, 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 ParagraphIWA-5244(b)(2).

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



TECHNICAL APPROACH AND POSITION NUMBER 13T-02Revision 0

COMPONENT IDENTIFICATION:

Code Class:Reference:

Examination Category:Item Number:Description:Component Number:

CODE REQUIREMENT:

1, 2, and 3IWA-5221IWA-5222B-P, C-H, D-BB15.10, C7.10, D2.10Valve Seats/Disks as Pressurization BoundariesAll Pressure Testing Boundary Valves

Paragraph IWA-522 1 requires the pressurization boundary for system leakage testing extend tothose 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 valveutilized for isolation. For example, in order to pressure test the ISI Class 1 components, the valvethat provides the Class break would be utilized as the isolation point. In this case the truepressurization 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 pipingon 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 visualexamination of the entire boundary valve body and bonnet (during pressurization up to the valveseat/disk).



3.0 COMPONENT ISI PLAN

The CPS Component ISI Plan includes ASME Section XI nonexempt pressure retainingwelds, piping structural elements, pressure retaining bolting, attachment welds, pumpcasings, valve bodies, reactor pressure vessel interior, reactor pressure vessel interiorattachments, 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 areidentified on the P&ID Drawings listed in Section 2.3, Table 2.3-1. ProcedureER-AA-330-002 "Inservice Inspection of Welds and Components", implements theASME Section XI welds and components program. This Component ISI Plan alsoincludes Augmented Examination Program examination requirements specified bydocuments other than ASME Section XI. For a detailed discussion of these examinationrequirements, 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 whichare not exempted under the criteria of Paragraphs IWB- 1220, IWC- 1220, andIWD- 1220, respectively. A summary of CPS ASME Section XI nonexemptcomponents 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 ISIIsometrics 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 CPSindividual 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 riskranked as High (1, 2, and 3), Medium (4 and 5), and Low (6 and 7). Per the EPRITopical Reports TR- 112657, Rev. B-A, TR-1006937, Rev. 0-A, and Code CaseN-578-1, piping structural elements ranked as High or Medium Risk are subject toexamination while piping structural elements ranked as Low Risk are not subjectto examinations (except for pressure testing). Thin wall welds that were excludedfrom volumetric examination under ASME Section XI rules per TableIWC-2500-1 are included in the element scope that is potentially subject to RISIexamination at CPS.

Pipiag structural elements may be excluded from examination (other than pressuretesting) under the RISI Program if the only degradation mechanism present for agiven location is inspected for cause under certain other CPS programs such as theFlow Accelerated Corrosion (FAC) or Intergranular Stress Corrosion Cracking(IGSCC) Programs. These piping structural elements will remain part of the FACor IGSCC programs, which already perform "for cause" inspections to detect



these degradation mechanisms. Piping structural elements susceptible to FAC orIGSCC along with another degradation mechanism (e.g., thermal fatigue) areretained as part of the RISI scope and are included in the element selection for thepurpose of performing examinations to detect the additional degradationmechanism. The RISI Program element examinations are performed inaccordance 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 thatare connected to the reactor coolant pressure boundary may be exempted from thesurface and volumetric examination requirements of ASME Section XI, providedthey are of such a size and shape that upon a postulated pressure boundaryrupture, the resulting flow of coolant under normal operating conditions is withinthe capacity of makeup systems.

3.3.1 Makeup Calculation

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

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

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

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

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IS! Program Plan


4.0 SUPPORT ISI PLAN

The CPS Support ISI Plan includes the supports of ASME Section XI nonexempt ISIClass 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 ASMESection 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 meetthe t:xemption criteria of Paragraph IWF-1230 of ASME Section Xi. A summaryof 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 individualsupport 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 beperformed in accordance with the Operation and Maintenance of NuclearPower Plants (OM), Standard ASME/ANSI OM, Part 4. As allowed by1 0CFR50.55a(b)(3)(v), CPS will use Subsection ISTD, "Inservice Testingof Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants,"ASME OM Code, 2004 Edition, No Addenda, to meet the requirements inASM/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 performedusing the VT-3 visual examination method described in ParagraphIWA-2213. A summary of the CPS safety-related and non-safety relatedsnubbers is included in Section 7.0.

Procedure ER-AA-330-004 "Visual Examination of TechnicalSpecification Snubbers", implements the visual inspection program forsafety related and non-safety related snubbers. Corporate proceduresER-AA-330-010 "Administration of Snubber Functional Testing",ER-AA-330-011 "Snubber Service Life Monitoring Program", and stationsurveillance test procedures are used to implement the functional testingand service life monitoring requirements for safety-related and non-safetyrelated snubbers.

The ASME Section XI ISI Program uses Subsection IWF to define supportinspection requirements. The ISI Program maintains the Code Classsnubbers in the populations subject to inspection per Article IWF-2000.This is done to accommodate scheduling and inspection requirements of

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the related attachment hardware per Paragraphs IWF-5200(c) andIWF-5300(c). (See Section 4.2.2 below.)

4.2.2 ASME Section XI Paragraphs IWF-5200(c) and IWF-5300(c) requireintegral and non-integral attachments for snubbers to be examined inaccordance with Subsection IWF of ASME Section XI. This results inVT-3 visual examination of the snubber attachment hardware includingthe bolting, pins, and their interface to the clamp, but does not include thecomponent-to-clamp interface.

The ASME Section XI ISI Program uses Subsection IWF to define theinspection requirements for all Class 1, 2, and 3 supports, regardless oftype. The ISI Program maintains the Code Class snubbers in the supportpopulations subject to inspection per Article IWF-2000. This is done tofacilitate scheduling and inspection requirements of the snubberattachment 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 attachmentswill be performed in accordance with the ASME Section XI SubsectionsIWB, IWC, and IWD welded attachment examination requirements (e.g.,Examination Categories B-K, C-C, and D-A).



5.0 SYSTEM PRESSURE TESTING ISI PLAN

The CPS System Pressure Testing (SPT) ISI Plan includes pressure retaining ASMESection XI, ISI Class 1, 2, and 3 components, with the exception of those specificallyexempted by Paragraphs IWA-5 110(c), IWC-5222(b), and IWD-5222(b). RISI pipingstructural elements, regardless of risk classification, remain subject to pressure testing aspart of the current ASME Section XI program.

The SPT ISI Plan performs system pressure tests and required VT-2 visual examinationson the ISI Class 1, 2, and 3 pressure retaining components to verify system andcomponent 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-specifictest procedures, implement the ASME Section XI System Pressure Testing ISI Plan.

This SPT ISI Plan also includes Augmented Examination Program examinationrequirements specified by documents other than ASME Section XI. For detaileddiscussion 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 reactorcoolant pressure boundary, are required to be tested. Those portions of ISI Class 2and 3 systems that are required to be tested include the pressure retainingboundaries of components required to operate or support the system safetyfunctions. ISI Class 2 and 3 open ended discharge piping and components areexcluded from the examination requirements per Paragraphs IWC-5222(b) andIWD-5222(b).

5.1.1 Identification of Class 1, 2, and 3 Components

Components subject to ASME Section XI System Pressure Testing areshown on the P&ID Drawings listed in Section 2.3, Table 2.3-1.Additional information on the classification of various system boundariesis 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 thewalkdown to define which systems, or portions of systems, fall under aspecific system pressure test. Individual tests are identified andmaintained in the CPS ISI Database.

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5.2 Risk-Informed Examinations of Socket Welds

Socket welds selected for examination under the RISI program are to be inspectedwith a VT-2 visual examination each refueling outage per ASME Code CaseN-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socketwelds selected for inspection under the RISI program are pressurized eachrefueling outage during a system pressure test in accordance with ParagraphIWA-521 I(a).

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ISI Program Plan


6.0 CONTAINMENT ISI PLAN

The CPS Containment ISI Plan includes ASME Section XI CISI Class MC pressureretaining components and their integral attachments, and CISI Class CC components andstructures that meet the criteria of Subarticle IWA-1300. This Containment ISI Plan alsoincludes information related to augmented examination areas, component accessibility,and examination review.

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

6.1 Nonexempt CISI Class Components

The CPS CISI Class MC and CC components identified on the CISI ReferenceDrawings are those not exempted under the criteria of Paragraphs IWE-1220 andIWL-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 inSection 7.0.

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

Components that are classified as CISI Class MC and CC must meet therequirements of ASME Section XI in accordance with 1OCFR50.55a(g)(4).Supports of Subsection IWE components are not required to be examined inaccordance with I OCFR50.55a(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 ReferenceDrawings 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 exemptedfrom examination based on design and accessibility per the requirementsof Paragraphs IWE- 1220 and IWL- 1220.

The process for exempting CPS components from the Containment ISIPlan per Paragraphs IWE-1220 and IWL-1220 is included in thecontainment sections of the ISI Classification Basis Document. Thesesections include discussions of exempt components and the bases for thoseexemptions.

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6.2 Augmented Examinations Areas

The containment section of the ISI Classification Basis Document discusses thecontainment design and components. Metal containment surface areas subject toaccelerated degradation and aging require augmented examination perExamination Category E-C and Paragraph IWE- 1240.

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

No significant conditions were identified in the First CISI Interval and nosignificant conditions are currently identified in the Second CISI Interval asrequiring application of additional augmented examination requirements underParagraph IWE-1240 or IWL-23 10.

6.3 Component Accessibility

CISI Class MC and CC components subject to examination shall remainaccessible for either direct or remote visual examination from at least one side perthe requirements of ASME Section XI, Paragraph IWE-1230.

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

Portions of components embedded in concrete or otherwise made inaccessibleduring construction are exempted from examination, provided that therequirements of ASME Section XI, Paragraph IWE-1232 have been fullysatisfied.

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

6.4 Responsible Individual and Engineer

ASME Section XI Subsection IWE requires the Responsible Individual to beinvolved 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 beinvolved in the development, approval, and review of the CISI examinations. The



Responsible Engineer shall meet the requirements of ASME Section XI,Paragraph IWL-2320.

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ISI Program Plan


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 pressureretaining components, supports, containment structures, system pressure testing,and augmented program components for the Third ISI Interval and the SecondCISI Interval at CPS.

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

Examination Item Number Description Exam Total Number of Relief Request/ NotesCategory (with (or Risk Requirements Components by TAP NumberExamination Categ or r System

Category AugmentedDescription) Number)

(1) (2) (3) (4) (5) (6) __7

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

Provides the Examination Category and description as identified in ASMESection 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 ASMESection XI Examination Categories B-F, B-J, C-F-I, and C-F-2 to identify ISIClass I and 2 piping structural elements for the RISI program. Only thoseExamination Categories applicable to CPS are identified.

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

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

Provides the Item Number as identified in ASME Section XI, TablesIWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, IWF-2500-1, andIWL-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, and2.2.8 have been developed to identify Augmented Examination Programs.


ISI Program Plan


(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 isprovided.

For Augmented Examination Program examinations, a description of theAugmented Examination Program basis is provided.

(4) Examination Requirements:

Provides the examination methods required by ASME Section XI, TablesIWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, IWF-2500-1, andIWL-2500-1.

Provides the examination requirements for piping structural elements under RISIthat 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 alist of these systems.

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

Note that the total number of components by system are subject to change aftercompletion of plant modifications, design changes, and ISI system classificationupdates.

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

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

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ISI Program PlanClinton Power Station Unit), Third Interval

(7) Notes:

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


1Sf Program Plan


TABLE 7.1-1INSERVICE 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 4215Pressure 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: IB 1.22 Meridional Head Welds (Reactor Vessel) Volumetric AAC: 4

1 _AAM: 6B11.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric AAA: 1B 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 11Full Penetration Welds AAN: 2of Nozzles in Vessels B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric AAH: 29 13R-02

AAO: 2


I51 Program PlanClinton Power Station Unit 1, Third Interval



Description) System

B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-I AAJ: 1Pressure Retaining B6.20 Clesu-, Studs (Reactor Vessel) Volumetric AAJ: 1 12

Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric AAJ: 12 in. In Diameter B6.50 Closure Washers (Reactor Vessel) Visual, VT-i AAJ: I

B6.180 Bolts & Studs (Pumps) Volumetric TRA: 1I TRB: 1

B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-I TRA: ITRB: I

B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-I TRA: 1TRB: 1





Description) I System

B-G-2 B7.10 Bolts, Studs, & Nuts (Reactor Vessel) Visual, VT-i AAP: IPressure Retaining B7.50 Bolts, Studs, & Nuts (Piping) Visual, Vi- 1 PMS: 16

Bolting, 2 in. and Less PNB: 1PRI: 2PRR: 4

B7.70 Bolts, Studs, & Nuts (Valves) Visual, VT-I PFW: 6PHP: 3PLP: 2

PMS: 24PRH: 12PR: 3PRR: 6PRT: 2

B7.80 Bolts, Studs, & Nuts in CRD Housing (Reactor Vessel) Visual, VT-I AAE: 1 10


151 Program PlanClinton Power Station Unit 1, Third Interval



Description) SystemB-K B110.10 Welded Attachments (Pressure Vessels) Surface or AAA: I

Welded Attachments Volumetricfor Vessels, Piping, B 10.20 Welded Attachments (Piping) Surface PFW: 4Pumps, and Valves PHP: 3

PLP: 3PMS: 8PRH: 9PRI: 7PRR: 2PRT: 1PSC: I

B110.30 Welded Attachments (Pumps) Surface TRA: 1TRB: I


ISI Program Plan




Description) System _ _ _ _

B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 TRA: 1Pump Casings TRB: 1

B-M-2 B12.50 Valve Bodies (Exceeding NPS 4) (Valves) Visual, VT-3 PFW: 6Valve Bodies PHP: 3

PLP: 3PMS: 24PRH: 12

PRI: 3PRR: 6PRT: 3

B-N-nr Bo3.10 IVessel interior (Reactor Vessel) Visual, VT-3 AAK: TInterior of Reactor Vessel V V I AI

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

Support Structures and Interior B 13.30 Interior Attachments Beyond Beltline Region (Reactor Visual, VT-3 AAK: 9Attachments 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 14Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface

Control Rod Housings I S


1~>

ISI Program Plan




Description) System

B-P B15.10 System Leakage Test (IWB-5220) Visual, VT-2 FW 13T-01All Pressure HP 13T-02

Retaining Components isLPMSNBRHRI

RPVRRRTSC





Description) _ SystemC-A C 1.10 Shell Circumferential Welds (Pressure Vessels) Volumetric AAA: 1

Pressure Retaining Welds AAB: Iin Pressure Vessels C1.20 Head Circumferential Welds (Pressure Vessels) Volumetric AAA: I

AAB: IC-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: 2Nozzle Welds in _ Nominal Thickness (Pressure Vessels)

Vessels C2.22 Nozzle Inside Radius Section Without Reinforcing Plate, Volumetric AAA: 2Greater Than 1/2" Nominal Thickness (Pressure Vessels) AAB: 2

C-C C3.10 Welded Attachments (Pressure Vessels) Surface AAA: 2Welded Attachments AAB: 2for Vessels, Piping, C3.20 Welded Attachments (Piping) Surface PHP: 6Pumps, and Valves PLP: 5

PRH: 32PRI: 9PSD: 2

C3.30 Welded Attachments (Pumps) Surface TRI: I


(~N




Description) System

C-G C6.10 Pump Casing Welds (Pumps) Surface TAA: 7 13R-05Pressure Retaining Welds TAB: 7

in Pumps and Valves TAC: 7THP: 7TLP: 7TRI: 4

C-H C7.10 System Leakage Test (IWC-5220) Visual, VT-2 CC 13R-03All Pressure CY 13R-04

Retaining Components FC 13T-01FP 13T-02FWHGHPISLPMSRHRISASCSDSFSMVPVQVRwo




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

Description) SystemD-A D1.20 Welded Attachments (Piping) Visual, VT-I SX: 32

Welded Attachmentsfor 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-01Retaining Components PV 13T-02

SXVC




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 226Containment Surfaces Accessible Surface Areas

El.I I Containment Vessel Pressure Retaining Boundary- Visual, VT-3 4 6Bolted Connections, Surfaces

El.12 Containment Vessel Pressure Retaining Boundary- Visual, VT-3 13 7Wetted Surfaces of Submerged Areas

E 1.20 Containment Vessel Pressure Retaining Boundary- Visual, VT-3 1 7BWR 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 9Augmented Examination Minimum Wall Thickness Locations Thickness





Description) SystemF-A I F1.10 Class 1 Piping Supports Visual, VT-3 FW: 32

Supports IHP: 6LP: 9

MS: 42RH: 45RI: 31RR: 32RT: 34SC: 26

F 1.20 Class 2 Piping Supports Visual, VT-3 HP: 67LP: 43MS: 4

RH: 348RI: 74SD: 21

F1.30 Class 3 Piping Supports Visual, VT-3 SX: 243F 1.40 Supports Other Than Piping Supports Visual, VT-3 HP: 1 I

(Class 1, 2, and 3) LP: 1RH: 5RI: 1

RPV: 1RR: 12




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 AreasL 1.12 Concrete Surfaces - Detailed Visual --

Suspect Areas (No Suspect Areas Identified)


ISI Program PlanClinton Power Station Unit]1, Third Interval


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 2Risk-Informed Piping PLP: 2 3

Examinations PRH: 9 4PRI: 1 5

3 Risk Category 3 Elements See Notes PFW: 67 13R-0 1 2PRH: 33 3

45

4 Risk Category 4 Elements See Notes PHP: 8 13R-01 2PLP: 3 3PMS: 4 4PRH: 80 5PRI: I

PRR: 133PSC: 12

5 Risk Category 5 Elements See Notes AAP: 1 13R-01 2PFW: 4 3PLP: 3 4

PRH: 12 5PRI: 64


I1S Program PlanClinton Power Station Unit 1, Third Interval


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 5Augmented Energy Circumferential and Longitudinal Piping Welds SurfaceComponents 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 13of Thin Wall Class 2 RHR System Piping Welds WhichWould Require Only Surface Examinations per ASMESection XI (Previously 1.2-G)

2.2.3 BWR Feedwater Nozzle and Control Rod Drive Return Volumetric AAH: 4Line Nozzle Cracking Components (BWROG and AAK: 2NUREG-0619) (Previously 1.2-H) PFW: 4

2.2.4 Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category D: 42 I3R-01 4Austenitic Stainless Steel Piping Components, TR- 113932,"BWR Vessel and Internals Project, Technical Basis forRevisions to Generic Letter 88-01 Inspection Schedules(BWRVIP-75)", and TR-1012621, "BWR Vessel andInternals Project, Technical Basis for Revisions to GenericLetter 88-01 Inspection Schedules (BWRVIP-75-A)"(Previously 1.2-I). Evaluation and Repair for Any CracksDetected 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: 2PRH: 6PRR: 26




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

Description) System

NA 2.2.6 NRC Regulatory Guide 1.150, Revision 1, Appendix A, Volumetric - AAA: 5Augmented Ultrasonic Testihig of Reactor Vessel Welds Dur~ng AAB: I I

Components Preservice and Inservice Examination (Previously 1.2-N) AAC: 4(Continued) AAG: 29

AAH: 29AAJ: 2AAK: 2AAL: 2AAM: 6AAN: 2AAO: 2AAP: 4PFW: 4PHP: 2PLP: 2PMS: 4PRH: 6PRR: 26

2.2.7 NUREG 0803 Generic Safety Evaluation Report Regarding Volumetric NAIntegrity of BWR Scram System Piping, Section 5.1, page5-3 Requires Inspection of Scram Discharge Volume Pipingin Accordance With ASME Section XI (Previously 1.2-0)

2.2.8 Reactor Pressure Vessel Shell Welds Volumetric AAA: 4 4215(1OCFR50.55a(g)(6)(ii)(A), Final Rule)



TABLE 7.1-2INSERVICE INSPECTION SUMMARY TABLE PROGRAM NOTES

Note # Note Summary1 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, SubsectionISTD Program will be the mechanism for actually performing the visual examinationsand functional testing scheduled within the ISI Program. For a detaileddiscussion 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 Programmethodology 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 2C-F-i and C-F-2 welds in accordance with lOCFR50.55a(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 IGSCCCategories B through G will continue to meet existing IGSCC schedules, whileIGSCC Category A welds have been subsumed into the RISI Program. (CPScurrently 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. SeeEPRI 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. AllBER augmented welds have been evaluated under the RISI methodology and have been integrated into the RISI Program under the I OCFR50.59 changeprocess. Additional guidance for adaptation of the RISI evaluation process to BER piping is given in EPRI TR-1006937 Rev. 0-A. Thus, these welds havebeen categorized and selected for examination in accordance with the EPRI Topical Reports TR- 12657, Rev. B-A, TR-1006937, Rev. 0-A, and Code CaseN-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 1and USAR Section 6.6.8 for Class 2.

6 Bolted connections examined per ItemNumber El. 11 require a General Visual examination each period anda VT-3 visual examination once per interval andeach time the connection is disassembled during a scheduled Item Number E 1.11 examination. Additionally, a VT-I visual examination shall be performed ifdegradation or flaws are identified during the VT-3 visual examination. These modifications are required by IOCFR50.55a(b)(2)(ix)(G) and1OCFR50.55a(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 1OCFR50.55a(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 lOCFR50.55a(b)(2)(xxi)(B), Table IWB-2500-1 examination requirements, the provisions of Table IWB-2500-1, Examination Category B-G-2, Item

Number B7.80, that are in the 1995 Edition are applicable only to reused bolting when using the 1997 Addenda through the latest Edition and Addendaincorporated 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, and3 of the Code Case in lieu of the previous examination volumes of ASME Section XI, Figures IWBr2500-7(a), (b), and (c).



TABLE 7.1-2INSERVICE INSPECTION SUMMARY TABLE PROGRAM NOTES

Note # Note Summary12 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, B6.20component will represent the sixty-four RPV closure studs. For tracking purposes, this component also includes the five cattle chute studs for CPS which areroutinely 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 hasbeen 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 isfor pressure retaining welds in 10% of the peripheral CRD Housings. A total of 34out of 145 CRD Housings are classified as peripheral components and eachhas 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).


ISI Program Plan


7.2 Snubber Inspection Summary Tables

1OCFR50.55a "Codes and Standards" allows usage of ASME OM CodeSubsection ISTD in place of ASME Section XI Paragraphs IWF-5200(a) andIWF-5300(a) and (b), using VT-3 visual examination methods described inParagraph IWA-2213.

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

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

ASME OM Code IISubsection OM Article Article Number Exam Totals Frequency Relief Request/

(with Subsection Number Description Requirements TAP NumberDescription)

(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 NuclearPower Plants (OM) subsection number and a description as obtained fromISTD. Only applicable subsections to CPS are identified.

(2) OM Article Number:

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

(3) Article Number Description:

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

(4) Examination Requirements:

Provides the visual examination and functional testing methods requiredby 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 aftercompletion of plant modifications and design changes.



(6) Frequency:

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

(7) Relief Request/TAP Number:

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

(8) Notes:

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

Alion Science & TechA ology 7-22 CLN05. G03Revision 14


TABLE 7.2-1SNUBBER INSPECTION SUMMARY

ASME OM CodeSubsection OM Article Article Number Exam T Relief Request/

(with Subsection Number Description Requirements TAP NumberDescription)

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 2Snubber 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 2Type 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



TABLE 7.2-2SNUBBER INSPECTION SUMMARY TABLE PROGRAM NOTES

Note # Note SuinmaqI 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".



8.0 RELIEF REQUESTS FROM ASME SECTION XI

This section contains relief requests written per 1 OCFR50.55a(a)(3)(i) for situationswhere alternatives to ASME Section XI requirements provide an acceptable level ofquality and safety; per 1OCFR50.55a(a)(3)(ii) for situations where compliance withASME Section XI requirements results in a hardship or an unusual difficulty without acompensating 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 1 OCFR50.55aparagraph citing for CPS relief requests. 1OCFR50.55a(a)(3)(i) andI OCFR50.55a(a)(3)(ii) provide alternatives to the requirements of ASME Section XI,while 1 OCFR50.55a(g)(5)(iii) recognizes situational impracticalities.

10CFR50.55a(a)(3)(i):

1 OCFR50.55a(a)(3 )(fi):

1 0CFR50.55a(g)(5)(iii):

Cited in relief requests when alternatives to the ASMESection XI requirements which provide an acceptable levelof quality and safety are proposed. Examples are reliefrequests which propose alternative NDE methods and/orexamination frequency.

Cited in relief requests when compliance with the ASMESection XI requirements is deemed to be a hardship orunusual difficulty without a compensating increase in thelevel of quality and safety. Examples of hardship and/orunusual difficulty include, but are not limited to, excessiveradiation exposure, disassembly of components solely toprovide access for examinations, and development ofsophisticated tooling that would result in only minimalincreases in examination coverage.

Cited in relief requests when conformance with ASMESection XI requirements is deemed impractical. Examplesof impractical requirements are situations where thecomponent would have to be redesigned, or replaced toenable the required inspection to be performed.

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

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


ISI Program Plan_____________Clinton Power Station Unit 1, Third Interval

TABLE 8.0-1RELIEF REQUEST INDEX

Relief Revision Status 2 (Program) Description/Request Date3 Approval Summary'

(ISI) Alternate Risk-Informed Selection and0 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 Flange13R-03 Submitted Seal Leak Detection System. Revision 0

O5/2 1 /10 Submitted.

(SPT) Alternative to Performance of SystemPressure Tests and VT-2 Visual Examination

0 Requirements for all ISI Class 2 Instrument Air13R-04 Submitted (IA) Piping and the ISI Class 3 IA Piping

05/21/10 Supplying, all SRV's, and both FeedwaterContainment Outboard Isolation Check Valves.Revision 0 Submitted.

(ISI) Examination of the ISI Class 2 HighRS 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 ofRPV Circumferential Shell Welds. PermanentRelief Request (Second ISI Interval ReliefRequest 4215) for deferral of the RPV

0 circumferential shell weld examinations was4215 Authorized authorized by the NRC per the SER dated05/2 1/10auhrzdythNR peteSE dtd

12/30/09 and thus applies to the remainingterm of operation under the existing, initiallicense, including this Third InspectionInterval.

Note 1: The NRC grants relief requests pursuant to 1OCFR50.55a(g)(6)(i) when Code requirements cannot be metand proposed alternatives do not meet the criteria of 1OCFR50.55(a)(3). The NRC authorizes reliefrequests pursuant to 1OCFR50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level ofquality and safcty or under 1OCFR50.55a(3)(ii) if compliance with the specified requirements would resultin hardship or unusual difficulties without a compensating increase in the level of safety.



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 beenwithdrawn 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 theNRC by the station and is awaiting approval; Pending- Relief has been awaiting station and Corporatereview and submittal to the NRC.

Note 3: The revision listed is the latest revision of the subject relief request. The date this revision became effectiveis the date of the approving SER which is listed in the fourth column of the table. The date notel in thesecond column is the date of the ISI Program Plan revision when the relief request was incorporated into thedocument.



10CFR50.55a RELIEF REQUEST: 13R-01Revision 0

(Page 1 of 7)

Request for Relief for Alternate Risk-Informed Selection and Examination Criteria forExamination Category B-F, B-J, C-F-i, and C-F-2 Pressure Retaining Piping Welds

In Accordance with 1OCFR50.55a(a)(3)(i)

1.0 ASME CODE COMPONENTS AFFECTED:

Code Class:Reference:Examination Category:Item Number:

Description:

Component Number:

1 and 2Table IWB-2500-1, Table IWC-2500-1B-F, B-J, C-F-1, and C-F-2B5.10, B5.20, B9.1 1, B9.21, B9.31, B9.32, B9.40, C5.1 1,C5.51, and C5.81Alternate Risk-Informed Selection and ExaminationCriteria for Examination Category B-F, B-J, C-F-1, andC-F-2 Pressure Retaining Piping WeldsPressure Retaining Piping

2.0 APPLICABLE CODE EDITION AND ADDENDA:

The code of record for the third ten-year Inservice Inspection Program interval at ClintonPower Station (CPS) is the American Society of Mechanical Engineers (ASME) Boilerand 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 surfaceexaminations on all welds for Item Number B5.10 and surface examinations for all weldsfor Item Number B5.20.

Table IWB-2500-1, Examination Category B-J, requires volumetric and surfaceexaminations on a sample of welds for Item Numbers B9.11 and B9.31 and surfaceexaminations on a sample of welds for Item Numbers B9.21, B9.32, and B9.40. Theweld 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 othercomponents where the stress levels exceed either of the following limits underloads associated with specific seismic events and operational conditions:

Alion Science & Techuiology CLN05. GO3


1OCFR50.55a RELIEF REQUEST: 13R-01Revision 0

(Page 2 of 7)

a. primary plus secondary stress intensity range of 2 .4 Sm, for ferritic steel andaustenitic 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 thecircumferential butt welds, branch connection, or socket welds in the reactorcoolant piping system. This total does not include welds exempted by ParagraphIWB-1220.

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

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

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

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

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

Alion Science & Techaology CLN05. G03



(Page 3 of 7)

4.0 REASON FOR REQUEST:

Pursuant to 1 OCFR50.55a(a)(3)(i), relief is requested on the basis that the proposedalternative utilizing Reference I along with two enhancements from Reference 4 willprovide an acceptable level of quality and safety.

As stated in "Safety Evaluation Report Related to EPRI Risk-Informed InserviceInspection Evaluation Procedure (EPRI TR-l 12657, Revision B, July 1999)" (i.e.,Reference 2):

"The staff concludes that the proposed RISI Program as describedin EPRI TR-11265 7, Revision B, is a sound technical approachand will provide an acceptable level of quality and safety pursuantto I OCFR50. 55a for the proposed alternative to the piping ISIrequirements with regard to the number of locations, locations ofinspections, and methods of inspection."

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

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

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

The actual "'valuation and ranking" procedure including the Consequence Evaluationand Degradation Mec.hanism Assessment processes of the currently approved (Reference

Alion Science & Tech1 uAlogy CLN05. GO3



(Page 4 of 7)

5) RISI Program remain unchanged and are continually applied to maintain the RiskCategorization 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 andrevised as major revisions of the site Probabilistic Risk Assessment (PRA) occur andmodifications to phlat configuration are made. The Consequence Evaluation,Degradation Mechanism Assessment, Risk Ranking, Element Selection, and Risk ImpactAssessment steps encompass the complete living program process applied under the CPSRISI Program.

5.0 PROPOSED ALTERNATIVE AND BASIS FOR USE:

The proposed alternative originally implemented in the risk informed in-serviceinspection 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 forCPS's Second Inspection Interval as documented in Reference 5.

The Third Inspection Interval RISI Program will be a continuation of the currentapplication and will continue to be a living program as described in the Reason ForRequest section of this relief request. No changes to the evaluation methodology ascurrently implemented under EPRI TR-1 12657, Revision B-A, are required as part of thisinterval 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 therequirements of Paragraph -2430, "Additional Examinations" contained in CodeCase N-578-1 (Reference 4). The alternative criteria for additional examinationscontained in Code Case N-578-1 provide a more refined methodology forimplementing necessary additional examinations. The reason for this selection isthat the guidance discussed in EPRI TR-1 12657 includes requirements foradditional examinations at a high level, based on service conditions, degradationmechanisms, and the performance of evaluations to determine the scope ofadditional examinations, whereas ASME Code Case N-578-1 provides morespecific and clearer guidance regarding the requirements for additionalexaminations that is structured similar to the guidance provided in ASME SectionXI, Paragraphs IWB-2430 and IWC-2430. Additionally, similar to the currentrequirements, of ASME Section XI, CPS intends to perform additionalexaminations that are required due to the identification of flaws or relevantconditions exceeding the acceptance standards, during the outage the flaws areidentified.

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b. TQ supplement the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of EPRITR- 112657, CPS will utilize the provisions listed in Table 1, ExaminationCategory R-A, "Risk-Informed Piping Examinations" contained in Code CaseN-578-1 (Reference 4). To implement Note 10 of this table, paragraphs andfigures from the 2004 Edition of ASME Section XI (CPS's Code of record for theThird Interval) will be utilized which parallel those referenced in the Code Casefor the 1989 Edition. Table 1 of Code Case N-578-1 will be used as it provides adetailed breakdown for Examination Method and Categorization of Parts to beExamined. Based on these Methods and Categorization, the examination figuresspecified in Section 4 of EPRI TR- 112657 will then be used to determine theexamination volume based on the degradation mechanism and componentconfiguration.

CPS uses UT techniques for RISI volumetric examinations.

For the components addressed by the RISI Program, ASME Section XI focuses primarilyon weld examinations. Risk Informed examination volumes also include portions ofpiping and fitting base materials that are susceptible to particular degradationmechanisms.

The ASME Section XI, Mandatory Appendix I, "Ultrasonic Examinations," specifies thatUT examinalion procedures, equipment, and personnel used to detect and size flaws inpiping welds shall be qualified by performance demonstration in accordance with ASMESection XI Appendix VIII, "Performance Demonstration for Ultrasonic ExaminationSystems." The RISI Program complies with Appendix VIII for weld examinations. Incases where the examination requirements cannot be met, CPS will submit a request forrelief in accordance with I OCFR50.55a, "Codes and standards."

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

Additionally, all CPS dissimilar metals (DM) welds, as characterized in ASME SectionXI, Article IWA-9000, have been evaluated for failure potential and consequence offailure along with the other non-exempt piping. The piping segments containing the DMwelds were classified into the appropriate RISI categories, and appropriate elements wereselected per the category requirements for examination during the Second InspectionInterval.

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

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mechanism(s) are removed from the RISI Program population. They are contained in theCPS Intergranular Stress Corrosion Cracking (IGSCC) Augmented Inspection Program(2.2.4) and are subject to the inspection requirements of BWRVIP-75-A "BWR Vesseland Internals Project Technical Basis for Revisions to Generic Letter 88-01 InspectionSchedules". Furthermore, all piping welds and welds, including DM welds in vesselnozzles classified as Category A (resistant material) per BWRVIP-75-A are included inthe 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, theguidance for the examination volume for a given degradation mechanism is provided bythe EPRI TR- 112657 while the guidance for the examination method and categorizationof parts to be examined are provided by the EPRI TR-1 12657 as supplemented by CodeCase 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 thetechnical adequacy of the CPS PRA Model to support this RISI application.

In addition to this risk-informed evaluation, selection, and examination procedure, allASME Section XI piping components, regardless of risk classification, will continue toreceive 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 TestingProgram, 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 April8, 2002. The Third Inspection Interval Relief Request utilizes an identical RISImethodology as was previously approved.

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IOCFR50.55a RELIEF REQUEST: 13R-01Revision 0

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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," December1999.

2. Letter from W. H. Bateman (NRC) to G. L. Vine (EPRI) "Safety EvaluationReport 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 PowerStation, 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 VesselCodi Section XI Requirements for Class I and 2 Piping Welds Risk-InformedInservice 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 PowerStation, Unit 1 - Risk-Informed Inservice Inspection Program, Relief Request4208 (TAC No. MB5321 1)," dated April 8, 2002.

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ENCLOSURE 1CPS 2006A PRA TECHNICAL CAPABILITY ASSESSMENT FOR


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Summary Statement of CPS PRA Model Capability for Use in Risk-Informed LicensingActions

Introduction

Exelon Generation Company (EGC) employs a multi-faceted approach to establishing andmaintaining the technical adequacy and plant fidelity of the PRA models for all operating EGCnuclear generation sites. This approach includes both a proceduralized PRA maintenance andupdate pcocess, and the use of self-assessments and independent peer reviews. The followinginformation 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 accuratereflection of the as-built and as-operated plants. This process is defined in the EGC RiskManagement progiam, which consists of a governing procedure (ER-AA-600, "RiskManagement") and subordinate implementation procedures. EGC procedure ER-AA-600-1015,"FPIE PRA Model Update" delineates the responsibilities and guidelines for updating the fullpower internal events PRA models at all operating EGC nuclear generation sites. The overallEGC Risk Management program, including ER-AA-600-1015, defines the process forimplementing regularly scheduled and interim PRA model updates, for tracking issues identifiedas potentially affecting the PRA models (e.g., due to changes in the plant, errors or limitationsidentified in the model, industry operating experience), and for controlling the model andassociated computer files. To ensure that the current PRA model remains an accurate reflectionof the as-built, as-operated plants, the following activities are routinely performed:

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

* New engineering calculations and revisions to existing calculations are reviewed fortheir 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 modelare updated approximately every four years.

In addition to these activities, EGC risk management procedures provide the guidance forparticular 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) productsincluding PRA update information, PRA models, and PRA applications.

* Guidelines for updating the full power, internal events PRA models for EGC nucleargeneration 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, surveillancetests and modifications) on systems, structures, and components (SSCs) within thescope of the Maintenance Rule (1 OCFR50.65(a)(4)).

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on anapproximately 4-year cycle; longer intervals may be justified if it can be shown that the PRAcontinues to adequately represent the as-built, as-operated plant. The most recent update of theCPS 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 theCPS 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 OwnersGroup (BWROG) in 2000, following the Industry PRA Peer Review process [1]. Thispeer review included an assessment of the PRA model maintenance and update process.

* A self-assessment analysis was previously performed against Addenda B of the ASMEPRA Standard (ASME RA-Sb-2005, [4]) and the draft of Revision 1 Regulatory Guide1.200 (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 PRAcross-comparisons study performed in support of implementation of the mitigatingsystems performance indicator (MSPI) process.

* The CPS 2006 PRA self-assessment was revised in 3Q09 to address consistency withRegulatory 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 of2009.

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-twosignificance 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 PRAUpdate. 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). Potentialgaps to Capability Category II of the Standard were identified and used to plan the CPS 2006PRA Update.

The CPS 2006 PRA self-assessment was revised in 3Q09 in preparation for the CPS 2009 peerreview to address consistency with Regulatory Guide 1.200 Revision 1 [6], including the NRCpositions stated in Appendix A of [6] and the clarifications in [5], and to identify which gapswere closed folloving completion of the CPS 2006 PRA update and the CPS PRA 2009 internalflooding update. Identified gaps have been considered with particular focus on technicalelements 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, includingany new items that may be identified in the planned industry peer review, will be reviewed forconsideration during future model updates. The currently identified items are judged to have lowimpact on the PRA model or its ability to support a full range of PRA applications. These itemsare or are being documented in the PRA Updating Requirements Evaluation (URE) database sothat they can be tracked and their potential impacts accounted for in applications whereappropriate. In addition, plant changes made since the last PRA update have been reviewed anddetermined to not have a significant PRA impact. These items are also documented in UREs forconsideration in future PRA updates, as appropriate.

General Conclusion Regarding PRA Capability

The CPS PRA maintenance and update processes and technical capability evaluations describedabove provide a robust basis for concluding that the PRA is suitable for use in risk-informedlicensing actions. As specific risk-informed PRA applications are performed, remaining gaps tospecific requirements in the PRA standard will be reviewed to determine which, if any, wouldmerit 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 alternativeinservice inspection requirements. Plant-specific PRA-derived risk significance information isused 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 technicalcapability, 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 arechosen for RISI inspection as illustrated below. Broad ranges are used to define these bins sothat the impact of uncertainty is minimized and only substantial PRA changes would be expectedto 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, itdepends only on the range in which the PRA result falls. As a consequence, any PRA modelinguncertainties would be mitigated by the wide binning provided in the methodology.Additionally, conservatism in the binning process (e.g., as would typically be introduced throughPRA attributes meeting ASME PRA Standard Capability Category I versus II) will tend to resultin a larger inspection population.

Secondly, the impacts of particular PRA consequence results are further dampened by the jointconsideration of the weld failure potential via a non-PRA-dependent damage mechanismassessment. The results of the consequence assessment and the damage mechanism assessmentare 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 intoconsideration, is reproduced below.





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POTENTIAL FORPIPE RUPTURE

PER DEGRADA'ION MECHANISMSCREENING CRITERIA

CONSEQUENCES OF PIPE RUPTUREIMPACTS ON CONDITIONAL CORE DAMAGE PROBABILITY

AND LARGE EARLY RELEASE PROBABILITY

NONE LOW MEDIUM HIGH

4 4 4 4.

HIGHFLOW ACCELERATED CORROSION

LOWCategory 7

MEDIUMCategory 5

~HIGH~C~t.o

HIGH-Category 1

MEi)IUM LOW LOW MEDIUM. . HIGH01I HER 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 resultin 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) mostlikely would result in a smaller inspection population.

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

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

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

There are, however, some applications that, because of the nature of the proposed change,have a limited impact on risk, and this is reflected in the impact on the elements of the riskmodel.





(Page 6 of 17)

An example is risk-informed inservice inspection (RI-ISI). In this application, risk significancewas used as one criterion for selectingpipe segments to be periodically examined forcracking. During the staff review it became clear that a high level of emphasis on PRAtechnical acceptability was not necessary. Therefore, the staff review ofplant-specific RI-ISItypically 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 noteliminated from the ISI Program on the basis of risk information. For example, the risksignificance of a weld may fall from Medium Risk Ranking to Low Risk Ranking, resulting in itnot 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 mayagain become a candidate for inspection. If it is discovered during the RISI update process that aweld is now susceptible to flow-accelerated corrosion (FAC), inter-granular stress corrosioncracking (IGSCC), or microbiological induced cracking (MIC) in the absence of any otherdamage mechanism, then it is addressed in an "augmented" program where it is monitored forthose special damage mechanisms. That occurs no matter what the Risk Ranking of the weld isdetermined 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 isbased 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 PRAStandard Capability Category I.

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

References

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





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2. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment forNuclear Power Plant Applications, ASME RA-S-2002, New York, New York, April 2002.

3. U.S. Nuclear Regulatory Commission, An Approach for Determining the TechnicalAdequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, DraftRegulatory Guide DG-1 122, November 2002.

4. American Society of Mechanical Engineers, Standard for Probabilistic Risk Assessment forNuclear Power Plant Applications, ASME RA-Sb-2005, New York, New York, December2005.

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

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

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

8. U.S. Nuclear Regulatory Commission, An Approach for Using Probabilistic Risk Assessmentin Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, RegulatoryGuide 1.174, Revision 1, November 2002.





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TABLE 1

IMPACT OF OPEN SIGNIFICANT PRA PEER REVIEW FINDINGS FORTHE CPS PRA MODEL

Peer FACTS & OBSERVATIONS (F&Os)ReviewElement ID Priority Summary Impact Assessment

TH-8 TH-8-1 B Additional plant specific room heat-up calculations (or Non-Significant Impact Primarily a documentationenhancements to existing calculations) should be performed issue. The PRA already makes appropriate assumptionsto support modeling assumptions regarding room-cooling regarding the need for room cooling in the appropriaterequirements. 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 contributescreening estimates. For post-initiator screening HEPs with approximately 2% of CDF). Fine-tuning the HEPs forRAWs greater than 1.1, the HEPs were re-evaluated with pre-initiators would be expected to reduce the relativemore detailed calculations. For consistency sake, the pre- importance of these events. No significant impact oninitiator HEP calculations should follow the same approach. RISI application.




ENCLOSURE ICPS 2006A PRA TECHNICAL CAPABILITY ASSESSMENT FOR


(Page 9 of 17)

TABLE 2

STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY IIOF THE ASME PRA STANDARD

...... ........~ ~ ~ ~ ~ ~~~~~~~~~. ....... .,z .. • .... . ...... " • ',.* " . •"• • . :• • ,>'• '

.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 IEA rigorous explicit assessment of all the events in analysis, which includes initiating eventsNUREG-1275 has not been performed. known to be relevant to BWR-6 plants

in general and CPS in particular. Thistype of activity is known to have beenperformed for another BWR plant(review of hundreds of events INPOSENs, SOERs, SERs, and NRC SECYletters on precursors) and no newinitiating events were identified. It isexpected that future industry studies willprovide this generic assessment.





10CFR50.559 RELIEF REQUEST: 13R-01Revision 0

(Page 10 of 17)

TABLE 2


tecrit i oicablSks Current Status ,/Comment, . Imortance to RISi

2 Assumptions regarding loss of switchgear room IE-C4 Deferred: Switchgear room cooling Non-Significant Impact The PRAcooling should be supported by room cooling AS-B3 calculations have not been performed at already makes appropriatecalculation. SC-B2 this time but are being considered for a assumptions regarding the need for

Sc-c 1 future update. room cooling and explicitly modelsSC-C2 room cooling in certain areas.

SY-A 17 Modeling cooling failures forSY-A19 switchgear might make the SXpipingSY-A20 going to the SX cooler moreSY-B7 important, but this is Class 3 piping,SY-B8 which is not in the scope of the RISI

Program. SX failures already havehigh importance for DG cooling,ECCS room cooling and DHR, andmore extensive Switchgear heatremoval modeling would not change_this.



/_-\



10CFR50.55a RELIEF REQUEST: I3R-01Revision 0

(Page 11 of 17)

TABLE 2


em .Descriptiono Gap Applcbe .SRs Current 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 ainitiator 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 review2) 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 currentprocedures, and the disposition of the action methodology and documentation for(screened or evaluated), identifying pre-initiator HEPs is judged

adequate to support applications of the4) Identify T&M activities that require PRA. Any additional documentation

realignment of the system outside its normal e nhanemetwouldoturesultinopertionl o stad b staus.enhancement would not result in

operational or stand by status. increasing the number of pre-initiator

HEPs included in the model orsignificantly impact their relative_importances.






(Page 12 of 17)

'IABLE 2


.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 fordominant pre-initiator HEPs. calculations are based on representative pre-initiators would be expected to

procedures/practices for similar pre- reduce the relative importance ofinitiator HEPs. The current estimates these events.are generally higher error probabilitiesthan would be obtained if variousexplicit recovery factors and testingfrequencies were applied in specific1-EP calculations for each pre-initiator.The impact on the model is non-significant, pre-initiator HEPs contribute

_approximately 2% to the CL06C CDF.






(Page 13 of 17)

TABLE 2


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 Anydata should fulfill the requirements of DA-C10, and used directly, but a adjustment to failure data countsshould be documented appropriately. Review confirmation that the data are collected resulting from a rigorous review ofsurveillance test procedures and identify all failure exactly consistent with the requirements testing procedures is judged to havemodes that are fully tested by the procedures. in the Standard has not been a non-significant impact on CDF andInclude 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 isequipment 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 RISIon 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 Anydetermination, perform interviews of maintenance PRA will consider performance of refinements to maintenancestaff for equipment with incomplete or limited interviews of plant personnel to unavailabilities are judged to resultmaintenance 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 trainswith good maintenance information-ECCS trains, RCIC, EDGs, SX).The model is reasonably consistentwith data from the plant MRdatabase, which is adequate for RISI_application.





(Page 14 of 17)

I'ABLE 2


item Description of Gap ApplicableS Ss 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 Internaldocumentation should be updated to meet ASME IF Technical CL06c model has recently been flooding analyses do not impact RISIStandard 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 areStatus / Comment] of the PRA. not used to represent the

consequences from flooding events.Rather the impact of flooding fromthe RISI consequence analysis isevaluated by tagging appropriatebasic events from the non-floodingportions of the Internal Events PRAmodel. Therefore the fact the RISIanalysis does not use the results ofthe updated Internal Floodinganalysis is not critical to the results

._ .... ........ of the RISI analysis.


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'-N




(Page 15 of 17)

TABLE 2


o .... .:.,......1 .. •.... .. . .:...:... ,.,:.:....:,..•..•..j•. •. ,..

tem, , .. ,Description of Gap :. .. Applicable s...n C re nt sCom et :impornce to s R 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 ofare 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 quantificationsequence models. Documentation of the results, it is possible to estimate theseimportance of failures in initiating event importances. However, initiatorsfault trees in the base PRA notebooks is associated with this gap are nota documentation enhancement, directly used in the RISI analysis

Documenting the relative importance

of basic events to CDF and LERF forthese fault tree based initiators has nobearing on the conditional coredamage (and large early release)probability calculations used in the

_RISI analysis.

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(Page 16 of 17)

TABLE 2


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 thewith the documentation requirements in the base PRA and are maintained forStandard: consideration for future PRA updates.

* Provide a list of human actions and equipmentfailures (significant basic events) that causeaccidents to be non-dominant.

* Bases for the elimination of mutually exclusiveevents from the model need to be added.

* Include cutsets segregated by accident sequencein the documentation.


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TABLE 2


,,!tem •,J.• , Description of Gap Applicable SRs CV u'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 newuncertainty 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 isclarification to its endorsement of the PRA QU-F4 modeling uncertainty, however the defined such that model uncertaintiesStandard. 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 approachthese 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 thatHR-13 then from that performing selected may be pertinent to applications.DA-E3 quantitative sensitivity studies. TheseIF-F3 recently redefined SRs will be addressed

LE-G4 during a future PRA model update usinga process consistent with NUREG-1855.


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Request for Relief for the Alternative to Nozzle to Vessel Weldand Inner Radius Examinations

In Accordance with 10CFR50.55a(a)(3)(i)


Code Class: IComponent Number: Nozzles N1, N2, N3, N5, N6, N7, N8, N9, N10, and N16

(See Enclosure 1 for specific nozzle identification numbers)Examination Category: B-DItem Number: B3.90 and B3.100Description: Alternative to Nozzle to Vessel Weld and Inner Radius

Examinations (IWB-2500, Table IWB-2500-1 - InspectionProgram B)


The code of record for the third ten-year Inservice Inspection Program interval at ClintonPower Station (CPS) is the American Society of Mechanical Engineers (ASME) Boilerand Pressure Vessel (B&PV) Code, Section XI, 2004 Edition. Additionally, forultrasonic examinations, ASME Section XI, Appendix VIII, "Performance Demonstrationfor Ultrasonic Examination Systems," of the 2001 Edition is implemented as required(and modified) by 1OCFR50.55a(b)(2)(xv) and 1OCFR50.55a(b)(2)(xxiv).

3.0 APPLICABLE CODE REQUIREMENT:

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


The identified ISI Class 1 nozzles are scheduled for examination for the upcominginspection initerval at CPS. The proposed alternative provides an acceptable level ofquality and safety, and the reduction in scope could provide a dose savings of as much as25 Rem for the entire interval.

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Pursuant to 1OCFR50.55a(a)(3)(i), relief is requested from performing the requiredexaminations on 100% of the identified nozzles. Alternatively, in accordance with CodeCase N-702 (Reference 2), CPS proposes to examine a minimum of 25% of the nozzleinner radii and nozzle-to-vessel welds, including at least one nozzle from each system andnominal pipe size. For each of the identified nozzles, both the inner radius and thenozzle-to-shell weld would be examined. As a minimum, the following nozzles would beselected for examination: one of the two 20" recirculation outlet nozzles (i.e., N I); threeof the ten 10" recirculation inlet nozzles (i.e., N2); one of the four 24" main steamnozzles (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 thevibration instrumentation nozzle (i.e., N16).

Code Case N-702 proposes that visual examination (i.e., VT-1) may be used in lieu ofvolumetric examination for the nozzle inner radii (i.e., Item B3.100). Note, however, thatCPS is not currently using ASME Code Case N-648-1 on enhanced magnification visualexamination and has no plans of using this Code Case in the future. CPS will continue toperform 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 InspectionRequirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and NozzleBlend Radii," (Reference 1) provides the basis for Code Case N-702. The EPRI reportfound that failure probabilities due to a low temperature overpressure event at the nozzleblend radius region and nozzle-to-vessel shell weld are very low (i.e., < 1 x 10-6 for 40years) with or without any inservice inspection.

On December 19, 2007, the NRC issued a Safety Evaluation (SE) approving the use ofBWRVIP-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 requestrelief from the ASME Section XI requirements for RPV nozzle-to-vessel shell welds andnozzle inner radius sections may reference the BWRVIP-108 report as the technical basisfor the use of Code Case N-702 as an alternative. However, each licensee shoulddemonstrate the plant-specific applicability of the BWRVIP-108 report to their units inthe relief request by showing that the general and nozzle-specific criteria addressed beloware satisfied:

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(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate islimited 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(ro2+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(ro2+ri 2)/(ro2-ri2)]/CNozzLE< 1.15

Demonstration of how CPS meets the NRC plant-specific applicability is provided inEnclosure 2. Based upon all RPV nozzle-to-vessel shell welds and nozzle inner radiisections 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 safetypursuant to 1OCFR50.55a(a)(3)(i) for all RPV nozzle-to-vessel shell welds and nozzleinner radii sections.



7.0 PRECEDENTS:


a. A similar request was approved for use at Duane Arnold Energy Center on August29, 2008 (i.e., Reference 4).

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

8.0 REFERENCES:

1. EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and InternalsProject Technical Basis for the Reduction of Inspection Requirements for theBoiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radii,"dated October 2002.

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(Page 4 of 4)

2. ASME Boiler and Pressure Vessel Code, Code Case N-702, "AlternativeRequirements 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 InternalsProject, Technical Basis for the Reduction of Inspection Requirements for theBoiling 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 EnergyCenter), "Duane Arnold Energy Center - Safety Evaluation for Request forAlternative to Reactor Pressure Vessel Nozzle to Vessel Weld and Inner RadiusExaminations (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 I OCFR50.55a Examination Requirements forReactor Pressure Vessel Weld Inspections (TAC No. ME0218)," dated August 24,2009.

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ENCLOSURE 11OCFR5O.55a RELIEF REQUEST: 13R-02

Revision 0(Page I of 2)

Table of ASME Section XI Components AffectedCPS, Unit I

IDENTIFICATION WELD DESCRIPTION CODE ITEMNUMBER CATEGORY NUMBERNIA 20" Recirculation Outlet Nozzle N IA to Vessel B-D B3.90

WeldN IA-IRS 20" Recirculation Outlet Nozzle Ni A Inner Radius B-D B3.100NIB 20" Recirculation Outlet Nozzle N IB to Vessel B-D B3.90

WeldN I B-IRS 20" Recirculation Outlet Nozzle N IB Inner Radius B-D B3. 100N2A 10" Recirculation Inlet Nozzle N2A to Vessel Weld B-D B3.90N2A-IRS 10" Recirculation Inlet Nozzle N2A Inner Radius B-D B3.100N2B 10" Recirculation Inlet Nozzle N2B to Vessel Weld B-D B3.90N2B-IRS 10" Recirculation Inlet Nozzle N2B Inner Radius B-D B3.100N2C 10" Recirculation Inlet Nozzle N2C to Vessel Weld B-D B3.90N2C-IRS 10" Recirculation Inlet Nozzle N2C Inner Radius B-D B3.100N2D 10" Recirculation Inlet Nozzle N2D to Vessel Weld B-D B3.90N2D-IRS 10" Recirculation Inlet Nozzle N2D Inner Radius B-D B3.100N2E 10" Recirculation Inlet Nozzle N2E to Vessel Weld B-D B3.90N2E-IRS 10" Recirculation Inlet Nozzle N2E Inner Radius B-D B3. 100N2F 10" Recirculation Inlet Nozzle N2F to Vessel Weld B-D B3.90N2F-IRS 10" Recirculation Inlet Nozzle N2F Inner Radius B-D B3.100N2G 10" Recirculation Inlet Nozzle N2G to Vessel Weld B-D B3.90N2G-IRS 10" Recirculation Inlet Nozzle N2G Inner Radius B-D B3.100N2H 10" Recirculation Inlet Nozzle N2H to Vessel Weld B-D B3.90N2H-IRS 10" Recirculation Inlet Nozzle N2H Inner Radius B-D B3.100N2J 10" Recirculation Inlet Nozzle N2J to Vessel Weld B-D B3.90N2J-IRS 10" Recirculation Inlet Nozzle N2J Inner Radius B-D B3.100N2K 10" Recirculation Inlet Nozzle N2K to Vessel Weld B-D B3.90N2K-IRS 10" Recirculation Inlet Nozzle N2K Inner Radius B-D B3.100N3A 24" Main Steam Nozzle N3A to Vessel Weld B-D B3.90N3A-IRS 24" Main Steam Nozzle N3A Inner Radius B-D B3.100N3B 24" Main Steam Nozzle N3B to Vessel Weld B-D B3.90N3B-IRS 24" Mlain Steam Nozzle N3B Inner Radius B-D B3.100N3C 24" Main Steam Nozzle N3C to Vessel Weld B-D B3.90N3C-IRS 24" Main Steam Nozzle N3C Inner Radius B-D B3.100N3D 24" Main Steam Nozzle N3D to Vessel Weld B-D B3.90N3D-IRS 24" Main Steam Nozzle N3D Inner Radius B-D B3.100N5A 12" Core Spray Nozzle N5A to Vessel Weld B-D B3.90N5A-1RS 12" Core Spray Nozzle N5A Inner Radius B-D B3.100N5B 12" Core Spray Nozzle N5B to Vessel Weld B-D B3.90



ENCLOSURE 11OCFR5O.55a RELIEF REQUEST: 13R-02

Revision 0(Page 2 of 2)

Table of ASME Section XI Components AffectedCPS. Unit 1

IDENTIFICATION WELD DESCRIPTION CODE ITEMNUMBER CATEGORY NUMBERN5B-IRS 12" Core Spray Nozzle N5B Inner Radius B-D B3.100N6A 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 B3.100Radius

N6B 10" Low Pressure Core Injection Nozzle N6B to B-D B3.90Vessel Weld

N6B-IRS 10" Low Pressure Core Injection Nozzle N6B Inner B-D B3.100Radius

N6C 10" Low Pressure Core Injection Nozzle N6C to B-D B3.90Vessel Weld

N6C-IRS 10" Low Pressure Core Injection Nozzle N6C Inner B-D B3.100Radius

N7 6" Top Head Spray Nozzle N7 to Vessel Weld B-D B3.90N7-IRS 6" Top Head Spray Nozzle N7 Inner Radius B-D B3.100N8 6" Top Head Spare Nozzle N8 to Vessel Weld B-D B3.90N8-IRS 6" Top Head Spare Nozzle N8 Inner Radius B-D B3.100N9A 4" Jet Pump Instrumentation Nozzle N9A to Vessel B-D B3.90

Weld

N9A-IRS 4" Jet Pump Instrumentation Nozzle N9A Inner B-D B3.100Radius

N9B 4" Jet Pump Instrumentation Nozzle N9B to Vessel B-D B3.90Weld

N9B-IRS 4" Jet Pump Instrumentation Nozzle N9B Inner B-D B3.100Radius

N 16 Vibration Instrumentation Nozzle to Vessel Weld B-D B3.90N 16-IRS Vibration Instrumentation Nozzle Inner Radius B-D B3.100


Alion Science& Technoloff CLN05. GO3


ENCLOSURE 210CFR50.55a RELIEF REQUEST: 13R-02


Responses to NRC Plant Specific Applicability

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

This criterion is met by adherence to CPS Technical Specification 3.4.11, "ReactorCoolant System Pressure/Temperature Limits," Surveillance Requirement 3.4.11.1 whichrequires verification that the Reactor Coolant System heatup and cooldown rates arelimited 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 =

r RPV inner radius =t RPV wail thickness =

CRPv =

1025 psig110.19 inches6.1 inches19332

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:


r = RPV inner radius =t = RPV wall thickness =

CRPV =


Result: (pr/t)/CRpv = 1.14

4. For the Reactor Recirculation Inlet (N2) Nozzles [p(ro2+ri2)/(ro2-ri2)]/CNozzLE must beless than 1.15, where:


ro = nozzle outlet radiusri = nozzle inner radius =

CNOZLE =


Result: [p(ro 2+ri2)/(ro -ri )]/CNOzzLE = 1.04

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ENCLOSURE 210CFR50.55a RELIEF REQUEST: I3R-02


5. For the Reactor Recirculation Outlet (N I) Nozzles [p(ro 2+ri 2)/(ro2-ri2)]/CNozzLE must beless than 1.15, where:

p = normal RPV pressure = 1025 psigro nozzle outlet radius = 16.3125 inchesri = tozzle inner radius = 9.0 inchesCNOZZLE = 1977

Result: [p(ro 2+ri2 )/(ro2 -ri2)J/CNozzLE = 0.97





(Page 1 of 5)

Request for Relief for Inservice Inspection Impracticality of Pressure Testingthe RPV Head Flange Seal Leak Detection System

In Accordance with 10CFR50.55a(g)(5)(iii)



Examination Category:Item Number:Description:

Component Number:Drawing Number:

1, 2, and 3Table IWB-2500-1, IWB-5200Table IWC-2500-1, IWC-5200Table IWD-2500-1, IWD-5200B-P, C-H, and D-BB15.10, C7.10, and D2.10Pressure Testing the RPV Head Flange Seal Leak DetectionSystemRPV Head Flange Seal Leak Detection SystemM05-1071, Sht. 1




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

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

Table IWD-2 500-1, Examination Category D-B, Item Number D2. 10, requires all ISIClass 3 pressure retaining components be subject to a system leakage test with a VT-2visual examination in accordance with Paragraph IWD-5220. This pressure test is to beconducted once each inspection period.




(Page 2 of 5)

4.0 IMPRACTICALITY OF COMPLIANCE:

Pursuant to 1OCFR50.55a(g)(5)(iii), relief is requested on the basis that pressure testingthe RPV Flange Leak Detection Line is deemed impractical.

The Reactor Vessel Head Flange Leak Detection Line is separated from the reactorpressure boundary by one passive membrane, a silver-plated O-ring located on the vesselflange. 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 indicatefailure of the inner flange seal O-ring. Failure of the O-ring would result in a HighPressure Alarm in the Main Control Room.

The configuration of this system precludes manual testing while the vessel head isremoved. As figure 13R-03.1 portrays, the configuration of the vessel tap, combined withthe 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 isdesigned to withstand pressure in one direction only. Due to the groove that the O-ringsits in and the pin/wire clip assembly (See Figure 13R-03.2), pressurization in theopposite direction into the recessed cavity and retainer clips would likely damage theO-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 linewill only be pressurized in the event of a failure of the inner O-ring. Purposely failing theinner 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, installthe new O-rings, and then reset and retension the reactor vessel head. This is consideredto impose an undue hardship and burden on CPS.

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


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

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(Page 3 of 5)



8.0 PRECEDENTS:


Peach Bottom Atomic Power Station Fourth Interval Relief Request 14R-25 was grantedper SER dated February 26, 2009

Limerick Generating Station Third Interval Relief Request I3R-08 was granted per SERdated March 11, 2008

LaSalle County Station Third Interval Relief Request I3R-08 was granted per SER datedJanuary 30, 2008

Susquehanna Steam Electric Station Third Inspection Interval Relief Request 3RR-07was granted per SER dated September 24, 2004.

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FIGURE 13R-03.1

FLANGE SEAL LEAK DETECTION LINE DETAIL

Outer FlangeSeal Ring-

Hiqh Pressure LeakDetection Monitoring Top

Inner FlangeSeal Ring

See Detail "A"

Detail "A"

Vessel FlangeSectional View




(Page 5 of 5)

FIGURE 13R-03.2

O-RING CONFIGURATION

SECTION A-A





(Page 1 of 5)

Request for Relief for Hardship Or Unusual Difficulty Without Compensating Increase InLevel 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) Pipingand the ISI Class 3 IA Piping Supplying, all SRV's, and both Feedwater Containment

Outboard Isolation Check ValvesIn Accordance with 10CFR50.55a(a)(3)(ii)



Examination Category:Item Number:Description:

Componc nt Nuiabec:

2,3Table IWC-2500-1, IWC-5200Table IWD-2500-1, IWD-5200C-H, D-BC7.10, D2.10Alternative to Performance of System Pressure Tests andVT-2 Visual Examination Requirements for all ISI Class 2Instrument Air (IA) Piping and the ISI Class 3 IA PipingSupplying, all SRV's, and both Feedwater ContainmentOutboard Isolation Check ValvesMultiple 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 isolationvalv es 1 AO 12A/B and 1 AO 13 A/B and check valves 11A042A/B. This includes thefollowing lines, valves, and components shown on CPS Piping and InstrumentationDiagram (P&ID) M05-1040 Sht. 7 not listed above.

" Lines IlA71BA/BB-1, IIA14GA/GB-1, 1lA95A/B-1, lIA93AA/BA-3/4, and11A96AA/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 thefollowing IA lines and valves supplying all 16 safety relief valves (SRV's) and bothFeedwater 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,

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(Page 2 of 5)

1 IA88A/B- 1/2, 1 1A71 CA/CB- 1, 1 1A71 DA/EA/FA/GA- 1/2, and11A71DB/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 1 IA26FA/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, 1 MS71CE/DE- 1/2, 1MS72AE/BE- 1/2, 1MS73BE/CE- 1/2,1MS74CE- 1/2, 1 MS71CG/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, 1 MS71CF/DF-3/4, 1MS72AF/BF-3/4, 1MS73BF/CF-3/4,IMS74CF-3/4, 1 MS71CC/DC-2, 1MS72AC/BC-2, 1MS73BC/CC-2, I MS74CC-2,1 MS71CJ/CK/DJ/DK- 11/4, 1MS72AJ/AK/BJ/BK- 11/4, 1 MS73BJ/BK/CJ/CK- 11/4, and IMS74CJ/CK-1 1/4

* P&ID M 10-9002 Sht. 1 valves - 1 lA094A/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 - 1 B21 -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, 1 MS77AC/CC/DC-2, IMS78BC/CC-2, 1MS75AG/AH/BG/BH- 11/4, 1MS76CG/CH/DG/DH- 1 1/4, 1MS77AG/AH/CG/CH/DG/DH- 1 1/4, and1MS78BG/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 andI B2 l-F08 1 A/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




Table IWC-2500-1, Examination Category C-H, Item Number C7. 10, requires all ISIClass 2 pressure retaining components be subject to a system leakage test with a VT-2

Alioit Science & Technology CLN05. GO3


I OCFR50.55a RELIEF REQUEST: 13R-04Revision 0

(Page 3 of 5)

visual examination in accordance with Paragraph IWC-5220. This pressure test is to beconducted once each inspection period.

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


Pursuant to I OCFR50.55a(a)(3)(ii), relief is requested on the basis that compliance withthe specified requirements would result in hardship or unusual difficulty without acompensating increase in the level of quality and safety.

Performance of a VT-2 visual examination would require applying a leak detectionsolution to a large amount of piping and components, many of which are in elevateddose rate areas with limited access. VT-2 visual inspections would result in additionalradiation exposure (estimated 2 Rem) and industrial safety challenges without anyadded benefit in the level of quality and safety. These inspections would not beconsistent with radiation exposure practices of "As Low As Reasonably Achievable(ALARA)."

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


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

Surveillance procedure CPS 9061.11, verifies the operability of SRV actuation capabilityand check valves in the IA supply lines to all 16 SRV's and both Feedwater containmentoutboard isolation check valves. This surveillance test is performed for each individualSRV and boih Feedwater containment outboard isolation check valves as a requirementof 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 checkvalve accumulators, as well as associated piping and valves. The pressure decay test is

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performed by isolating and pressurizing these accumulators and associated piping to thenominal operating pressure. The decay in pressure is then monitored through calibratedpressure measuring instrumentation. If any pressure decay acceptance criterion (seeEnclosure 1) is exceeded, the surveillance identifies appropriate troubleshooting steps toperform, including soap-bubble application to locate leakage.

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

The VT-2 visual examination described in Tables IWC-2500-1 and IWD-2500-1 andperformed once per inspection period, would not provide an increase in safety, systemreliability, or structural integrity. In addition, performance of a VT-2 visual examinationwould require applying a leak detection solution to a large amount of piping andcomponents, many of which are in elevated dose rate areas with limited access. VT-2visual inspections would result in additional radiation exposure (estimated 2 Rem) andindustrial 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 LowAs Reasonably Achievable (ALARA)."

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



7.0 PRECEDENTS:





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

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



ENCLOSURE 110CFR50.55a RELIEF REQUEST: 13R-04


Acceptance Criteria From Procedure CPS 9061.11(For Information Only)

CPS, Unit 1Component Leakage Pressure Drop Comments

Criterion Test DurationAccumulator Headers for all < 1.5 psig > 108 minutesSRV's except IB21-F05ICand DAccumulator Headers for < 1.5 psig > 31 minutes Smaller volume than other SRV's.1B21-FO51C and DAccumulator Headers for < 1.5 psig > 26 minutes Smaller volume than SRV's.Feedwater Check Valve IADS Supply Header to < 22 psig > 60 minutes This inspection tests over 200 feetAccumulator Headers of piping and components.



10CFR50.55a RELIEF REQUEST: I3R-05Revision 0

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Request for Relief Regarding Inservice Inspection Impracticality Due to theExamination of the High Pressure Core Spray, Low Pressure Core Spray, and Residual

Heat Removal Pump Casing WeldsIn Accordance with 10CFR50.55a(g)(5)(iii)


Code Class:Reference:Examination Category:Item Number:Description:

Component Number:

Drawing Number:

2IWC-2500, Table IWC-2500-1C-GC6.10Examination of the ISI Class 2 High Pressure Core Spray,Low Pressure Core Spray, and Residual Heat RemovalPump Casing Welds1A RHR Pump Casing Welds1 B RHR Pump Casing Welds1 C RHR Pump Casing WeldsHPCS Pump Casing WeldsLPCS Pump Casing WeldsB-69, B-71, and B-73




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

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

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4.0 IMPRACTICALITY OF COMPLIANCE:

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

CPS's three Residual Heat Removal Pumps (1E 12-CO02A, 1E 12-COO2B, and1E 12-CO02C), one High Pressure Core Spray Pump (1 E22-COO1), and one Low PressureCore Spray Pump (1 E21 -COO 1) were originally designed where the pump casing weldswere encased in concrete, thus making the welds inaccessible for inservice inspection.

Therefore, it is impractical for CPS to perform the surface examination of these weldswithout destruction of the concrete resulting in unnecessary engineering and installationcosts and radiation exposure without a compensating increase in safety. Additionally, dueto the design of the subject pumps, access to the affected welds can only be achievedthrough disassembly of the pump, removal of the pump internals, and the required surfaceexaminations perfoimed from the inside surface of the welds. This effort, in the absenceof any other necessary pump maintenance, represents a significant expenditure of manhours and radiation exposure to plant personnel, without a compensating increase in plantsafety.

5.0 BURDEN CAUSED BY COMPLIANCE:

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


In the event the subject welds become accessible upon disassembly of any one (1) of thepumps, the welds will be surface examined from the inside surface or a VT-1 visualexamination will be performed for that particular pump group to the maximum extentpracticable based on the obstructions and geometric constraints detailed in theImpracticality Of Compliance section of this relief request. The examination method willbe 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 ExaminationCategory 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 theconcrete) for evidence of leakage in accordance with Paragraph IWA-5241(b). Theseexaminations will provide reasonable assurance of continued structural integrity of thepiping systems.

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8.0 PRECEDENTS:


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

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

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

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10CFR50.55a RELIEF REQUEST: 4215Revision 0

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*** NOTE ***

Second ISI Interval Relief Request 4215, Revision 0 was previously submitted and approvedunder the Second Interval ISI Program Plan. The approval authorized under NRC SER datedDecember 30, 2009 for CPS, Exelon Generation Company, LLC., was for permanent relief fordeferral of RPV shell weld examinations, and thus applies to the remaining term of operationunder the existing, initial license, including this Third Inspection Interval. All ASME Codereferences 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 revisedauthorization is required. Formatting for Relief Request 4215, Revision 0 varied from thestandard ISI Program Plan format due to the fact that it also requested relief from theAugmented Reactor Pressure Vessel examination contained in 1OCFR50a(g)(6)(ii)(A)(2).

10CFR50.55a RequestRegarding 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:Component Numbers:Examination Category:Item Number:Description:

1RPV-C 1, RPV-C2, RPV-C3, and RPV-C4B-ABI.11Reactor 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 complieswith the 1989 Edition of American Society of Mechanical Engineers (ASME) Boiler andPressure Vessel Code (Code), Section XI. Additionally, for ultrasonic examinations, Section XI,Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," of the 1995Edition, 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 1 OCFR50.55a, "Codes and standards," paragraph (a)(3)(i),Exelon Generation Company, LLC (EGC) requests permanent relief (for the remaining portion ofthe initial license period that expires on September 29, 2026) for CPS, Unit 1, from the followingrequirements:

1. Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to beexamined. Table IWB-2500-1 requires volumetric examination of all RPV shellcircumferential welds each inspection interval (i.e., Examination Category B-A, ItemNo. B1.11);

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

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

4. Reason for Request

Reference 1 provides the technical basis for permanently deferring the augmented inspections ofcircumferential welds in boiling water reactor (BWR) RPV's. In the report, the BWR Vessel andInternals Project (B WRVIP) concluded that the probabilities of failure for BWR RPVcircumferential welds are orders of magnitude lower than that of the longitudinal welds. TheNRC 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 levelof 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 I OCFR50.55a(a)(3)(i), and consistent with information contained inReference 3, EGC considers the following alternate provisions for the subject weld examinations.

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1OCFR50.55a RELIEF REQUEST: 4215Revision 0

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Inservice Inspection Scope

The failure frequency for RPV shell circumferential welds is sufficiently low to justify theirelimination from the ISI requirement of 1OCFR50.55a(g) based on the NRC Safety Evaluation(Reference 2).

The ISI and augmented examination requirements of the ASME Code Section XI, TableIWB-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 shallinclude inspection of the circumferential welds only at the intersection of these welds with thelongitudinal 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 thebest possible examination of the RPV shell longitudinal welds.

The procedures for these examinations shall be qualified such that flaws relevant to the RPVintegrity can be reliably detected and sized, and the personnel implementing these proceduresshall 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), successiveexaminations 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 nogrowth 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., Reference1);

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

3. The vessel containing the flaw is acceptable for continued service in accordance withSubarticle IWB-3600, "Analytical Evaluation of Flaws," and the flaw is demonstratedacceptable 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 consistentwith 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, RPVshell circumferential welds (i.e., at intersections with longitudinal welds), additionalrequirements per Subsubarticle IWB-2430, "Additional Examinations," are not required for flawsprovided the following conditions are met:

1. If the flaw is characterized as subsurface in accordance with BWRVIP-05, then noadditional examinations are required;

2. If the flaw is not characterized as subsurface in accordance with BWRVIP-05, then anengineering 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 thevessel, then additional examinations may be limited to those welds subject to the rootcause conditions and failure mechanisms, up to the number of examinations required byparagraph (a) of Subsubarticle IWB-2430. If the engineering evaluation determines thatthere are no additional welds subject to the same root cause conditions or no failuremechanism 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 withSubsubarticle IWB-2430. All flaws in RPV shell longitudinal welds shall require additionalweld examinations consistent with ASME Code and regulatory requirements. Examinations ofthe RPV shell circumferential welds shall be performed if RPV longitudinal welds reveal anactive, mechanistic mode of degradation.

Basis for Use

Reference 1 provides the technical basis to justify relief from the examination requirements ofRPV shell circumferential welds. The results of the NRC's evaluation of Reference 1 aredocumented in Reference 2. Reference 3 permits BWR licensees to request permanent relieffrom the ISI requirements of 1OCFR50.55a(g) (i.e., for the remaining term of operation under theexisting, 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). Thisrelief can be granted by demonstrating that:

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1 At the expiration of their license, the circumferential welds will continue to satisfy thelimiting 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 thefrequency 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 tosatisfy 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 estimatethe RPV shell weld failure probabilities. Three key assumptions of the PFMA are: (1) theneutron fluence used was the estimated end-of-life mean fluence, (2) the chemistry values aremean values based on vessel types, and (3) the potential for beyond-design-basis events isconsidered.

Table 1 provides a comparison of the limiting RPV circumferential weld parameters for CPS tothose 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 determinethe acceptable reference temperatures at CPS. Although the unirradiated reference temperaturefor CPS is higher than the NRC limit, the combination of unirradiated reference temperature andembrittlement shift yields adjusted reference temperatures considerably lower than the NRCmean analysis values.

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


1Sf Program PlanClinton Power Station Unit]1, Third Interval


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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*

#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 0.51Fluence (1019 n/cm 2)

ARTNDT (OF) 50.77 109.5

ARTNDT(U) (0F) -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 thatlimit 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 reactorpressure during all phases of operation, including cold shutdown. Use of these procedures willprevent an over-pressure event, and are reinforced through operator training. Operatingprocedures contain sufficient guidance to prevent a low temperature over-pressurization event. Areactor coolant system leakage test is performed prior to each restart after a refueling outage. Apre-job briefing is required prior to test commencement with all involved personnel. Duringpressure testing, measures are taken to limit the potential for system perturbations that could leadto 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 systemsactuations that are not required to be operable. Vessel temperature and pressure are required tobe monitored and controlled to within CPS Technical Specifications pressure and temperature



1OCFR50.55a RELIEF REQUEST: 4215Revision 0

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(P/T) limits during all portions of testing. Pre-job briefings and careful coordination ensure thatpressure transients are minimized.

The high pressure coolant sources that could inadvertently initiate and result in a low temperatureoverpressurization event are the Feedwater, Reactor Core Isolation Cooling (RCIC), and HighPressure Core Spray (HPCS) systems. During normal RPV fill prior to pressure testing, theControl Rod Drive (CRD) system is the preferred method for filling the reactor. TheCondensate/Condensate Booster systems are used as an alternative means to fill the reactor. Themotor driven reactor feedwater pump is prevented from starting by the high water level feedwaterpump trip signal, which is present due to the high reactor water levels required during pressuretesting. During the reactor coolant system leakage test, the reactor is in cold shutdown, and as aresult, there is no steam available to drive the turbine driven RCIC and turbine driven reactorfeedwater pumps.

The HPCS system is a high pressure make-up system at CPS. The HPCS pump is motoroperated, so it can be operated when the reactor is in cold shutdown. However, the HPCS systemwould require manual initiation, inadvertent initiation, or manual startup to start and inject intothe RPV. Also, there is a high RPV water level interlock for the HPCS injection valve to preventoverfilling the RPV. This high level interlock is not normally overridden. Even if the HPCSsystem is inadvertently started, it would not inject and pressurize the reactor due to the high RPVwater level interlock.

The CRD .system is a high pressure system used to operate the control rods. The CRD system isa low flow rate systemwith about 50 gpm flow rate to the reactor. During cold shutdownconditions, 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 reactorcoolant system leakage testing. During cold shutdown conditions, operators closely monitorreactor water level, pressure, and temperature. With the low CRD flowrate, the operators shouldhave 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 thereactor if the control rods fail to insert. The SLC system has no automatic start function so aspurious stait is unlikely. The SLC system must be manually initiated by the use of a keylockswitch for each pump.

During cold shutdown conditions, the condensate booster pumps of the Condensate system areshutdown. It would require direct operator action to start a main Condensate Booster systempump and inject into the reactor pressure vessel. The Condensate/Condensate Booster systemsare used as an alternate method for filling the RPV and as the primary method for initiallypressuring the RPV for pressure testing.




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These actions are taken in accordance with procedural guidance that includes verification thatRPV coolant and metal temperatures will support filling and pressurizing the RPV with theCondensate/Condensate Booster Pump systems without exceeding the Technical SpecificationP/T limits.

Low pressure coolant sources include the Emergency Core Cooling Systems (ECCS) (i.e., LowPressure Core Spray (LPCS) and Low Pressure Coolant Injection (LPCI) systems), and theCondensate system. The shutoff heads of the ECCS pumps and condensate pumps aresufficiently low to preclude a low temperature overpressurization event that would exceed theP/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/Tcurves and requirements of the Technical Specifications. Simulator sessions are conductedwhich include plant heat-up and cool-down. Additionally, in response to industry operatingexperience, the operator training program is routinely evaluated and revised, as necessary, toreduce the possibility of events such as a low temperature overpressurization event.

Based on the above, procedural and administrative controls, as reinforced in operator training, arein place to effectively limit a low temperature overpressurization event.

Summary

In summary, EGC hMs reviewed the methodology used in Reference 1, and considering CPS plantspecific 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 1 OCFR50.55a forRPV circumferential shell welds since the proposed alternative provides an acceptable level ofquality 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, includingDresden Nuclear Power Station, Units 2 and 3 (References 4 and 5), Susquehanna Steam ElectricStation, Units 1 and ,2 (References 6 and 7), and Quad Cities Nuclear Power Station, Units I and2 (References 9 and 10).

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8. References

I BWRVIP-05, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel ShellWeld 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-05Report to Request Relief from Augmented Examination Requirements on ReactorPressure Vessel Circumferential Shell Welds," dated November 10, 1998

4 Letter from J. M. Heffley (Commonwealth Edison Company) to U. S. Nuclear RegulatoryCommission, "Relief Request for Alterative Weld Examination of CircumferentialReactor 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 AlterativeReactor Pressure Vessel Circumferential Weld Examinations (TAC Nos. MA6228 andMA6229)," dated February 25, 2000

6 Letter from R. G. Byram (PPL Susquehanna, LLC) to U. S. Nuclear RegulatoryCommission, "Request for Alternative to 10CFR50.55a Examination Requirements ofCategory B 1.11 Reactor Pressure Vessel Welds for PPL Susquehanna LLC Units 1 and 2PLA-525 1," dated November 7, 2000

7 Letter from M. Gamberoni (U. S. Nuclear Regulatory Commission) to R. G. Byram (PPLSusquehanna, LLC), "Relief Request No. 22 (RR-22) from American Society ofMechanical Enginecrs Boiler and Pressure Vessel Code, Section XI, Susquehanna SteamElect.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 andInternals Project BWRVIP-05 Report (TAC NO. MA3395)," dated March 7, 2000

9 Letter fiom P. R. Simpson (Exelon), "Relief Request for Alternative Reactor PressureVessel Circumferential Weld Examinations for the Fourth Interval Inservice InspectionProgram," dated May 16, 2003

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10 Letter from A. J. Mendiola (U. S. Nuclear Regulatory Commission) to C. Crane (Exelon),"Quad Cities Nuclear Power Station, Units 1 and 2 - Authorization For ProposedAlternative Reactor Pressure Vessel Circumferential Shell Weld Examination (TAC Nos.MB8985 and MB8986)," dated April 29, 2004

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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 PressureBoundary".- Part 50, Appendix J, Primary Reactor Containment Testing for Water CooledPower Reactors.

SECY-96-080, Issuance of Final Amendment To 1 OCFR50.55a To Incorporate ByRefeience 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, "InserviceInspection 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 ForConstruction of Nuclear Power Plant Components", the 2004 Edition, NoAddenda.

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, ASMESectikn XI, Division 1".

6) Regulatory Guide 1.150, Rev. 1, "Ultrasonic Testing of Reactor Vessel WeldsDuring 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 andProcessing Guidelines for BWR Coolant Pressure Boundary Piping".



13) NRC NUREG-0578 dated July 1979, "TMI-2 Lessons Learned Task Force StatusRepcrt and Short-Term Recommendations".

14) NRC NUREG-0619, dated November 1980, "BWR Feedwater Nozzle andCotitrol 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 onIntergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic StainlessSteel Piping".

17) Generic Letter 88-01, Supplement 1, dated February 4, 1992, "NRC Position onIntergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic StainlessSteel Piping".

18) BWR Vessel and Internals Project, Technical Basis for Revisions to GenericLetter 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 Basisfor Revisions to Generic Letter 88-01 Inspection Schedules BWRVIP-75), EPRIReport TR-1 13932, October 1999", (TAC NO. MA5012), dated May 14, 2002.

20) BWR Vessel and Internals Project, Technical Basis for Revisions to GenericLetter 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 Basisfor 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 InspectionRequirements," 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 InspectionRequirements, 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 NozzleInspiection 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. GO3Revision 14


Inspcction Requirements, May 2000", (TAC No. MA6787), dated March 10,2000.

26) Branch Technical Position MEB 3-1, dated November 24, 1975, "High EnergyFluid Systems, Protection Against Postulated Piping Failures in Fluid SystemsOutside Containment".

27) Gencric Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05Report to Request Relief From Augmented Examination Requirements on ReactorPressure Vessel Circumferential Shell Welds", dated November 10, 1998.

28) NRC Final SER related to the "BWR Reactor Vessel Shell Weld InspectionRecommendations (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 EPRIRisk-Informed Inservice Inspection (RI-ISI) Methodology to Break ExclusionRegion (BER) Programs", August 2002.

33) NRC SER related to EPRI Topical Report TR-1006937, Rev. 0, "Extension of theEPRI Risk-Informed Inservice Inspection (RI-ISI) Methodology to BreakExclusion Region (BER) Programs", dated June 27, 2002.

34) Exelhn Risk-Informed Inservice Inspection Evaluation (Final Report) for ClintonPowc 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, SecondTen-Year Inspection Interval.



39) Exelon Procedures ER-AA-330, "Conduct of Inservice Inspection Activities",ER-AA-330-001, "Section XI Pressure Testing", ER-AA-330-002, "InserviceInspection of Welds and Components", ER-AA-330-003, "Visual Examination ofSection XI Component Supports", ER-AA-330-004, "Visual Examination ofTechnical Specification Snubbers", ER-AA-330-005, "Visual Examination ofSection 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, "SnubberService Life Monitoring Program".

40) Letter Y-109584 (121-09(08-24)-L), Revision 0, "Makeup Capacity Exemption ofClass I Components per ASME Section XI", dated August 24, 2009.


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Clinton, Unit 1, ISI Program Plan Third Ten-Year ... › docs › ML1126 › ML11269A018.pdf · Clinton Power Station Unit 1, Third Interval 1.0 INTRODUCTION AND BACKGROUND 1.1 Introduction