Text

Fifth Interval Inservice Inspection Program Plan and Relief Requests
	ML12275A070
Person / Time
Site:	Quad Cities
Issue date:	09/28/2012
From:	Gullott D; Exelon Generation Co
To:	; Office of Nuclear Reactor Regulation, Document Control Desk
References
	RS-12-150
	Download: ML12275A070 (180)
	v • d • e

September 28, 2012 10 CFR 50.55a U. S. Nuclear Regulatory Commission ATIN: Document Control Desk Washington, D.C. 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 EGC is submitting the Fifth Interval Ten-Year ISI Program Plan (see Attachment A) in In accordance with 10 CFR 50.55a, "Codes and standards," paragraph (g), "Inservice Renewed Facility Operating License Nos. DPR-29 and DPR-30 NRC Docket Nos. 50-254 and 50-265 Quad Cities Nuclear Power Station, Units 1 and 2, Fifth Interval Inservice inspection requirements," Exelon Generation Company, LLC (EGC) is required to update

Subject:

Quad Cities Nuclear Power Station, Units 1 and 2, Fifth Intervallnservice 10 CFR 50.55a Inspection Program Plan and Relief Requests Quad Cities Nuclear Power Station, Units 1 and 2 accordance with IWA-1400, "Owners Responsibility," paragraph (c) of the Code. Please the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, Inservice Inspection (ISI)

In accordance with 10 CFR 50.55a, "Codes and standards, paragraph (g), "Inservice It inspection requirements," Exelon Generation Company, LLC (EGG) is required to update Program as described in the American Society of Mechanical Engineers (ASME) Section XI the Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, Inservice Inspection (lSI)

Program as described in the American Society of Mechanical Engineers (ASME) Section XI note that EGC is requesting NRC approval of only the relief requests contained in Section 8, (the Code) once every 120-month inspection interval. Specifically, the lSI program is (the Code) once every 120-month inspection interval. Specifically, the ISI program is required to comply with the latest edition and addenda of the Code incorporated by Renewed Facility Operating License Nos. DPR-29 and DPR-30 reference in 10 CFR 50.55a 12 months prior to the start of the interval in accordance with Inspection Program Plan and Relief Requests 10 CFR 50.55a(g)(4)(ii). The next interval for QCNPS Units 1 and 2 is currently scheduled to commence on April 2, 2013, and scheduled to be completed by April 1, 2023.

required to comply with the latest edition and addenda of the Code incorporated by Accordingly, the 2007 Edition through the 2008 Addenda of ASME Section XI is the Code "Relief Requests from ASME Section Xi." The remaining sections of the ISI Program Plan that QCNPS will implement for the Fifth Interval Ten-Year lSI Program.

Introduction and Background reference in 10 CFR 50.55a 12 months prior to the start of the interval in accordance with EGC is submitting the Fifth Interval Ten-Year lSI Program Plan (see Attachment A) in U. S. Nuclear Regulatory Commission accordance with IWA-1400, "Owners Responsibility," paragraph (c) of the Code. Please NRC Docket Nos. 50-254 and 50-265 note that EGC is requesting NRC approval of only the relief requests contained in Section 8, 10 CFR 50.55a(g)(4)(ii). The next interval for QCNPS Units 1 and 2 is currently scheduled are provided for information only. The QCNPS ISI Program Plan contains the following "Relief Requests from ASME Section XI." The remaining sections of the lSI Program Plan are provided for information only. The QCNPS lSI Program Plan contains the following sections:

ATTN: Document Control Desk to commence on April 2, 2013, and scheduled to be completed by April 1, 2023.

Section 1: Introduction and Background September 28, 2012 Section 2: Basis for Inservice Inspection Program sections:

Accordingly, the 2007 Edition through the 2008 Addenda of ASME Section XI is the Code Section 3: Component lSI Plan Section 4: Support lSI Plan Washington, D.C. 20555-0001 Section 1:

Section 5: System Pressure Testing lSI Plan RS-12-150 that QCNPS will implement for the Fifth Interval Ten-Year ISI Program.

Section 6: Containment lSI Plan

Subject:

Section 7: Component Summary Tables Section 8: Relief Requests from ASME Section XI Section 9: References

September 28 2012 U. S. Nuclear Regulatory Commission Page 2 cc:

Where alternatives to Code requirements are being proposed (i.e., in accordance with 10 CFR 50.55a(a)(3)(i) and (ii)), or the implementation of certain Code requirements has Where alternatives to Code requirements are being proposed (Le., in accordance with been determined to be impractical (i.e., in accordance with 10 CFR 50.55a(g)(5)(iii)), specific Attachment A: Respectfully, September 28 2012 relief requests have been included in Section 8 of the Fifth Interval ISI Program Plan. Note Manager - Licensing that Relief Requests 15R-01, 02, 03, 04 and 06 are similar to those previously approved for 10 CFR SO.SSa(a)(3)(i) and (ii>>, or the implementation of certain Code requirements has the Fourth Ten-Year Inspection Interval ISI Program. Relief Requests 15R-05 and 15R-08 have been previously authorized for the Fifth Ten-Year Inspection Interval and are included The subject relief requests deal with the implementation of certain lSI program requirements in this ISI Program Plan for completeness only. No further action by the NRC is requested U. S. Nuclear Regulatory Commission for these two relief requests. Relief Requests 15R-09 and 15R-10 are new relief requests for been determined to be impractical (Le., in accordance with 10 CFR SO.sSa(g)(S)(iii>>, specific the upcoming 10-year interval. Relief Request 15R-07 is reserved for future use.

Exelon Generation Company, LLC The subject relief requests deal with the implementation of certain ISI program requirements relief requests have been included in Section 8 of the Fifth Interval lSI Program Plan. Note which affect components to be examined during the Unit 2 April 2014 refueling outage and Page 2 the Unit 1 March 2015 refueling outage. Approval of these relief requests in the customary Regional Administrator - NRC Region III one-year period (i.e., by September 28, 2013) will be sufficient to support these outages. which affect components to be examined during the Unit 2 April 2014 refueling outage and that Relief Requests ISR-01 , 02, 03, 04 and 06 are similar to those previously approved for Should you have any questions concerning this letter, please contact Joseph A. Bauer at 630-Should you have any questions concerning this letter, please contact Joseph A. Bauer at 630-657-2804.

Quad Cities Nuclear Power Station, Units 1 and 2, Inservice Inspection the Fourth Ten-Year Inspection Interval lSI Program. Relief Requests ISR-OS and ISR-08 Respectfully, have been previously authorized for the Fifth Ten-Year Inspection Interval and are included the Unit 1 March 201S refueling outage. Approval of these relief requests in the customary vid M. Gullott Manager - Licensing Exelon Generation Company, LLC in this lSI Program Plan for completeness only. No further action by the NRC is requested NRC Senior Resident Inspector - Quad Cities Nuclear Power Station Attachment A: Quad Cities Nuclear Power Station, Units 1 and 2, Inservice Inspection Program Plan, Fifth 10-Year Inspection Interval for these two relief requests. Relief Requests ISR-09 and ISR-10 are new relief requests for cc: Regional Administrator - NRC Region III Program Plan, Fifth 10-Year Inspection Interval NRC Senior Resident Inspector - Quad Cities Nuclear Power Station one-year period (Le., by September 28,2013) will be sufficient to support these outages.

the upcoming 10-year interval. Relief Request ISR-07 is reserved for future use.

6S7-2804.

ATTACHMENT A Quad Cities Nuclear Power Station, Units 1 and 2 Inservice Inspection Program Plan Fifth Ten-Year Interval Quad Cities Nuclear Power Station, Units 1 and 2 ATTACHMENT A Inservice Inspection Program Plan Fifth Ten-Year Interval

Exelon Generation Company Quad Cities Nuclear Power Station Units 1 & 2 ISI Program Plan Fifth Ten-Year Inspection Interval Report: QC-481565-RP03 Rev: 0 Prepared By Prepared For AUTOMATED ENGINEERING SERVICES CORP.

Automated Engineering Services Exelon Generation Company One Energy Center Quad Cities Nuclear Power Station 40 Shuman Blvd., Suite 340 22710 206th Avenue North Naperville, IL 60563 Cordova, IL 61242

QC-481565-RP03, Rev. 0 Page i ISI Program Plan - 5th Interval REVISION APPROVAL SHEET TITLE: ISI Program Plan REVISION: 0 Fifth Ten-Year Inspection Interval Quad Cities Nuclear Power Station, Units 1 & 2 PREPARED TRANSMITTAL PREPARED: ______________________________/_________

Stephen J. Coleman AES Programs Engineer PREPARED: ______________________________/_________

Matthew R. King AES Programs Engineer REVIEWED: ______________________________/_________

Kevin M. Johnson AES Programs Engineer APPROVED: ______________________________/_________

Daniel W. Lamond AES Programs Manager EXELON ACCEPTANCE APPROVED: ______________________________/_________

Nicholas Johnson Quad Cities Inservice Inspection Program Owner Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page ii ISI Program Plan - 5th Interval REVISION APPROVAL SHEET TITLE: ISI Program Plan REVISION: 0 Fifth Ten-Year Inspection Interval Quad Cities Nuclear Power Station, Units 1 & 2 PROGRAM ACCEPTANCE REVIEWED: ______________________________/_________

Nicholas Johnson Quad Cities Inservice Inspection Program Owner APPROVED: ______________________________/_________

Mathew Rice Quad Cities Engineering Programs Supervisor Each time this document is revised, the Revision Control Sheet should be completed to provide a detailed record of the revision history. The major changes should be outlined within the table.

Editorial and formatting revisions are not required to be logged. The signatures above apply only to the changes made in the revision noted. If historical signatures are required, Quad Cities Nuclear Power Station archives will need to be retrieved.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page iii ISI Program Plan - 5th Interval REVISION CONTROL SHEET Major changes should be outlined within the table below. Minor editorial and formatting revisions are not required to be logged.

Passport Date Revision Summary Revision 0 9/26/12 Initial Issue. (Developed by Automated Engineering Services Corporation as part of the Quad Cities Nuclear Power Station Fifth Interval ISI Program Update.)

Prepared: S. Coleman/M. King Reviewed: K. Johnson Approved: D. Lamond Note: This ISI Program Plan is (Sections 1 - 9 inclusive) controlled by the Quad Cities Nuclear Power Station Engineering Programs Group.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page iv ISI Program Plan - 5th Interval REVISION

SUMMARY

Section Effective Pages Section Date Revision Preface i to vii 0 9/26/12 1.0 1-1 to 1-17 0 9/26/12 2.0 2-1 to 2-33 0 9/26/12 3.0 3-1 to 3-4 0 9/26/12 4.0 4-1 to 4-2 0 9/26/12 5.0 5-1 to 5-2 0 9/26/12 6.0 6-1 to 6-5 0 9/26/12 7.0 7-1 to 7-38 0 9/26/12 8.0 8-1 to 8-3 0 9/26/12 9.0 9-1 to 9-4 0 9/26/12 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page v ISI Program Plan - 5th Interval TABLE OF CONTENTS SECTION DESCRIPTION PAGE

1.0 INTRODUCTION AND BACKGROUND

..................................................................... 1-1 1.1 Introduction 1.2 Background 1.3 Third Interval ISI Program 1.4 Fourth Interval ISI Program 1.5 Fifth Interval ISI Program 1.6 First Interval CISI Program 1.7 Second Interval CISI Program 1.8 Code of Federal Regulations 10CFR50.55a Requirements 1.9 Code Cases 1.10 Relief Requests 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM................................................... 2-1 2.1 ASME Section XI Examination Requirements 2.2 Augmented Inspection Requirements 2.3 System Classifications and P&ID Boundary Drawings 2.4 ISI Isometric Drawings for Nonexempt ISI Class Components/Supports and Calibration Standards 2.5 Technical Approach and Positions 3.0 COMPONENT ISI PLAN ............................................................................................... 3-1 3.1 Nonexempt ISI Class Components 3.2 Risk-Informed Examination Requirements 3.3 Reactor Coolant Pressure Boundary Normal Makeup Calculation 3.4 Reactor Coolant Pressure Boundary Normal Makeup Calculation For Peripheral CRD Housing Welds 4.0 SUPPORT ISI PLAN ....................................................................................................... 4-1 4.1 Nonexempt ISI Class Supports 4.2 Snubber Examination and Testing Requirements 5.0 SYSTEM PRESSURE TESTING ISI PLAN .................................................................. 5-1 5.1 ISI Class Systems 5.2 Risk-Informed Examination of Socket Welds Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page vi ISI Program Plan - 5th Interval TABLE OF CONTENTS (Continued)

SECTION DESCRIPTION PAGE 6.0 CONTAINMENT ISI PLAN ........................................................................................... 6-1 6.1 Nonexempt ISI Class Components 6.2 Augmented Examination Areas 6.3 Component Accessibility 6.4 Responsible Individual and Engineer 6.5 CISI Reference Drawing Descriptions 7.0 COMPONENT

SUMMARY

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

9.0 REFERENCES

................................................................................................................ 9-1 9.1 NRC References 9.2 Industry References 9.3 Licensee References 9.4 License Renewal References Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page vii ISI Program Plan - 5th Interval TABLE OF CONTENTS (Continued)

TABLES DESCRIPTION PAGE 1.1-1 UNITS 1 & 2 FIFTH ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3 COMPONENT EXAMINATIONS) .............................................. 1-3 1.1-2 UNITS 1 & 2 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS MC COMPONENT EXAMINATIONS) ........................................................... 1-4 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS ............... 1-10 2.3-1 P&ID BOUNDARY DRAWINGS (ISI) ....................................................................... 2-16 2.3-2 P&ID BOUNDARY DRAWINGS (CISI)..................................................................... 2-17 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS .................................................. 2-19 2.4-2 UNIT 1 CISI REFERENCE DRAWINGS .................................................................... 2-22 2.4-3 UNIT 2 CISI REFERENCE DRAWINGS .................................................................... 2-24 2.4-4 CALIBRATION STANDARDS MATRIX ................................................................... 2-26 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX .............................................. 2-31 7.1-1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

SUMMARY

TABLE ......................................... 7-18 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES .................... 7-32 7.2-1 UNIT 1 & COMMON SNUBBER INSPECTION

SUMMARY

TABLE .................... 7-36 7.2-2 UNIT 2 SNUBBER INSPECTION

SUMMARY

TABLE ............................................ 7-37 7.2-3 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES ....................... 7-38 8.0-1 RELIEF REQUEST INDEX............................................................................................ 8-2 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-1 ISI Program Plan - 5th Interval

1.0 INTRODUCTION AND BACKGROUND

1.1 Introduction This Inservice Inspection (ISI) Program Plan details the requirements for the examination and testing of ISI Class 1, 2, 3, and MC pressure retaining components, supports, and containment structures at Quad Cities Nuclear Power Station (QCNPS), Units 1, 2, and Common (1/2). Unit Common components are included in the Unit 1 sections, reports, and tables. This ISI Program Plan also includes Containment Inservice Inspection (CISI), Risk-Informed Inservice Inspections (RISI), Augmented Inspections (AUG), and System Pressure Testing (SPT) requirements imposed on or committed to by QCNPS. This ISI Program Plan is controlled and revised in accordance with the requirements of procedure ER-AA-330, Conduct of Inservice Inspection Activities, which implements the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI ISI Program. At QCNPS, the Inservice Testing (IST) Program is maintained and implemented separately from the ISI Program. The IST Basis Document and IST Program Plan contain all applicable inservice testing requirements. Procedure ER-AA-321, Administrative Requirements for Inservice Testing, implements the IST Program. The Snubber Program is also maintained and implemented separately from the ISI Program at QCNPS. The Snubber Program documents contain all of the applicable snubber visual examination, functional testing, and service life monitoring requirements. The ISI Program Plan is credited as the existing program for QCNPS License Renewal Aging Management Programs B.1.1 ASME Section XI Inservice Inspection IWA, IWB, IWC, & IWD (AT 101562.01), B.1.3 Reactor Head Closure Studs (AT 101562.03), B.1.6 BWR Control Rod Drive Return Line Nozzle (AT 101562.06),

B.1.26 ASME Code, Section XI, Subsection IWE (101562.35), and B.1.27 ASME Code, Section XI, Subsection IWF (AT 101562.37).

The Fifth ISI Interval is effective from April 2, 2013, through April 1, 2023, for QCNPS Units 1 and 2. The Second CISI Interval is effective from September 9, 2008, through September 8, 2018, for QCNPS Units 1 and 2. These effective interval dates are based on the fact that QCNPS has been approved to extend plant operation under the license renewal application. The ASME Section XI Code of Record for the Fifth ISI Interval is the 2007 Edition through the 2008 Addenda, and the ASME Section XI Code of Record for the Second CISI Interval is the 2001 Edition through the 2003 Addenda.

Paragraph IWA-2430(c)(1) of ASME Section XI allows an inspection interval to be extended or decreased by as much as one year, and Paragraph IWA-2430(d) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The extension may be taken for a period of time not to exceed the duration of the outage. See Tables 1.1-1 and 1.1-2 for Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-2 ISI Program Plan - 5th Interval intervals, periods, and extensions that apply to QCNPSs Fifth ISI Interval and Second CISI Interval.

The Fifth ISI Interval and Second CISI Interval are divided into three successive inspection periods as determined by calendar years of plant service within the inspection interval. Tables 1.1-1 and 1.1-2 identify the period start and end dates for the Fifth ISI Interval and Second CISI Interval as defined by Inspection Program. In accordance with Paragraph IWA-2430(c)(3), the inspection periods specified in these Tables may be decreased or extended by as much as 1 year to enable inspection to coincide with QCNPSs refueling outages.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-3 ISI Program Plan - 5th Interval TABLE 1.1-1 UNIT 1 & 2 FIFTH ISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS 1, 2, AND 3 COMPONENT EXAMINATIONS)

UNIT 1 Period Interval Period Unit 2 Outage Projected Outage Start Date to End Date Start Date to Start Date to End Date Projected Outage Outage Number Start Date or End Date Start Date or Number Outage Duration Outage Duration Q1R23 Scheduled 1st 1st Scheduled Q2R22 3/15 4/2/13 to 4/1/16 4/2/13 to 4/1/16 3/14 Q1R24 Scheduled 2nd Scheduled Q2R23 3/17 4/2/16 to 2/28/203 5th (Unit 1) 3/16 Q1R25 Scheduled 4/2/131 to 4/1/23 2nd Scheduled Q2R24 3/19 th 4/2/16 to 2/28/203 3/18 5 (Unit 2)

Q1R26 Scheduled 3rd 4/2/132 to 4/1/23 3rd Scheduled Q2R25 3/21 3/1/20 to 4/1/23 3/1/20 to 4/1/23 3/20 Q1R27 Scheduled Scheduled Q2R26 3/23 3/22 Note 1: The Unit 1 Third and Fourth ISI Intervals were extended by 20 and 23 days, respectively, as permitted by Paragraph IWA-2430(d). These extensions are being carried forward to the Fifth ISI Interval to accommodate both Units 1 and 2 having the same interval start date. No interval overlap is being implemented, and thus the start dates for the Fifth ISI Interval are being adjusted the same. As required by Paragraph IWA-2430(c)(1), successive intervals shall not be altered by more than one year from the original pattern. This means that for the remainder of the Fifth ISI Interval, only 322 days are available to use under the Paragraph IWA-2430(c) extension.

Note 2: The Unit 2 Fourth ISI Interval was extended by 22 days as permitted by Paragraph IWA-2430(d). This extension is being carried forward to the Fifth ISI Interval to accommodate both Units 1 and 2 having the same interval start date. No interval overlap is being implemented, and thus the start dates for the Fifth ISI Interval are being adjusted the same. As required by Paragraph IWA-2430(c)(1), successive intervals shall not be altered by more than one year from the original pattern. This means that for the remainder of the Fifth ISI Interval, only 343 days are available to use under the Paragraph IWA-2430(c) extension.

Note 3: The Unit 2 second period of the Fifth ISI Interval was reduced by 32 days as permitted by Paragraph IWA-2430(c)(3) to include Q2R25 in the third period since it was scheduled to start in March 2020. (Note that the Unit 1 second period of the Fifth ISI Interval was also extended by 32 days as permitted by Paragraph IWA-2430(c)(3) for unit consistency.)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-4 ISI Program Plan - 5th Interval TABLE 1.1-2 UNITS 1 & 2 SECOND CISI INTERVAL/PERIOD/OUTAGE MATRIX (FOR ISI CLASS MC COMPONENT EXAMINATIONS)

Unit 1 Period Interval Period Unit 2 Outage Projected Outage Start Date to End Date Start Date to Start Date to End Date Projected Outage Outage Number Start Date or End Date Start Date or Number Outage Duration Outage Duration Q1R20 Scheduled 1st 1st Scheduled Q2R20 5/09 9/9/08 to 9/8/11 9/9/08 to 9/8/11 3/10 Q1R21 Scheduled 2nd Scheduled Q2R21 2nd (Unit 1) 5/11 9/9/08 to 9/8/18 9/9/11 to 9/8/15 3/12 Q1R22 Scheduled 2nd Scheduled Q2R22 3/13 9/9/11 to 9/8/15 2nd (Unit 2) 3/14 Q1R23 Scheduled 9/9/08 to 9/8/18 3rd Scheduled Q2R23 3/15 9/9/15 to 9/8/18 3/16 Q1R24 Scheduled 3rd Scheduled Q2R24 3/17 9/9/15 to 9/8/18 3/18 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-5 ISI Program Plan - 5th Interval 1.2 Background The Commonwealth Edison Company, now known commercially as Exelon Generation Company (Exelon), obtained construction permits to build QCNPS on February 15, 1967, for Unit 1, CPPR-23, and for Unit 2, CPPR-24. The docket numbers assigned to QCNPS are 50-254 for Unit 1 and 50-265 for Unit 2. After satisfactory plant construction and preoperational testing was completed, QCNPS was granted a full power operating license for Unit 1, DPR-29, and subsequently commenced commercial operation on February 18, 1973; the full power operating license for Unit 2, DPR-30, was granted and commercial operation commenced on March 10, 1973.

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

1.3 Third Interval ISI Program Pursuant to the Code Of Federal Regulations, Title 10, Part 50, Section 55a, Codes and standards, (10CFR50.55a), Paragraph (g), Inservice inspection requirements, QCNPS was required to update the ISI Program to meet the requirements of ASME Section XI once every ten years or inspection interval.

The ISI Program was required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a 12 months prior to the start of the interval per 10CFR50.55a(g)(4)(ii). Accordingly, the Inservice Inspection requirements applicable to the Third Interval ISI Program should have been based on the rules set forth in the 1986 Edition, No Addenda of ASME Section XI.

However, ComEd by letter dated June 3, 1992, and as supplemented on December 3, 1992, requested United States Nuclear Regulatory Commission (NRC) approval to meet the requirements set forth in the 1989 Edition, No Addenda of ASME Section XI prior to its incorporation by reference into 10CFR50.55a(b)(2).

NRC approval was received under the letter from J. E. Dyer to D. L. Farrar dated April 19, 1993, Inservice Inspection Program Update - Quad Cities, Units 1 and

2. Therefore, the 1989 Edition, No Addenda of ASME Section XI is the Code Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-6 ISI Program Plan - 5th Interval that QCNPS Units 1 and 2 met for the Third ISI Interval. The ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI.

Per Relief Request CR-33 Alternate Selection Criteria for Class 1 Pressure Retaining Dissimilar Metal Welds and Pressure Retaining Welds in Piping and Class 2 Pressure Retaining Welds in Austenitic Stainless or High Alloy Steel and Carbon or Low Alloy Steel Piping, as approved by NRC SE dated February 5, 2002 (TAC Nos. MB0721 and MB0722), QCNPS adopted the EPRI Topical Report TR-112657, Rev. B-A methodology, which was supplemented by ASME Code Case N-578-1, for implementing risk-informed inservice inspections. The RISI Program was in effect from the middle of the third period through the end of the Third ISI Interval. This approach replaced the categorization, selection, and examination volume requirements of ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 applicable to QCNPS with Examination Category R-A as defined in ASME Code Case N-578-1.

QCNPS Unit 1 was shut down from December 19, 1997, to May 31, 1998, for an Appendix R outage. QCNPS Unit 1 received approval for schedular exemption in a letter from Carl F. Lyon to John L. Skolds dated September 16, 2002, for several Examination Category B-D, Item Number B3.90 and B3.100 components. These inspections were repeated during the first period of the Fourth ISI Interval in accordance with the Fourth Interval ISI Program.

The Unit 1 Third ISI Interval was extended by 20 days as permitted by Paragraph IWA-2430(d). This extension was carried forward to the Fourth ISI Interval to accommodate both Units 1 and 2 having the same interval start date.

Therefore, the QCNPS Third Interval ISI Program was effective from February 18, 1993, through March 9, 2003, for Units 1 and 2, respectively.

1.4 Fourth Interval ISI Program Pursuant to 10CFR50.55a(g), QCNPS was required to update the ISI Program to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI Program was required to comply with the latest Edition and Addenda of the Code incorporated by reference in 10CFR50.55a twelve months prior to the start of the Fourth ISI Interval per 10CFR50.55a(g)(4)(ii).

The QCNPS Fourth Interval started on March 10, 2003, for Units 1 and 2 using the requirements of 10CFR50.55a, and the 1995 Edition through the 1996 Addenda of ASME Section XI. The QCNPS Fourth Interval ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices of ASME Section XI, approved Code Cases, approved alternatives through relief requests and Safety Evaluations (SEs).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-7 ISI Program Plan - 5th Interval The Unit 1 Third and Fourth ISI Intervals were extended by 20 and 23 days as permitted by Paragraph IWA-2430(d). The Unit 2 Fourth ISI Interval was extended by 22 days as permitted by Paragraph IWA-2430(d). These extensions are being carried forward to the Fifth ISI Interval to accommodate both Units 1 and 2 having the same interval start date. No interval overlap is being implemented, and thus the start dates for the Fifth ISI Interval are being adjusted the same.

The QCNPS Fourth ISI Interval was effective from March 10, 2003, through April 1, 2013, for Units 1 and 2, respectively.

1.5 Fifth Interval ISI Program Pursuant to 10CFR50.55a(g), licensees are required to update their ISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The ISI Program is required to comply with the latest Edition and Addenda of the Code incorporated by reference in 10CFR50.55a twelve months prior to the start of the Fifth ISI Interval per 10CFR50.55a(g)(4)(ii). As discussed in Section 1.4 above, the start of the Fifth ISI Interval will be on April 2, 2013, for QCNPS Units 1 and 2. Based on this date, the latest Edition and Addenda of the Code referenced in 10CFR50.55a(b)(2) twelve months prior to the start of the Fifth ISI Interval was the 2007 Edition through the 2008 Addenda.

The QCNPS Fifth Interval ISI Program Plan was developed in accordance with the requirements of 10CFR50.55a, and the 2007 Edition through the 2008 Addenda of ASME Section XI, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.8-1 of this section. This ISI Program Plan addresses Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices of ASME Section XI, approved Code Cases, approved alternatives through relief requests and SEs, and utilizes the Inspection Program.

QCNPS adopted the EPRI Topical Report TR-112657, Rev. B-A methodology, which was supplemented by ASME Code Case N-578-1, for implementing risk-informed inservice inspections during the Third ISI Interval. The RISI Program will continue for the Fifth ISI Interval. Implementation of the RISI Program is in accordance with Relief Request I5R-02.

The QCNPS Fifth ISI Interval is effective from April 2, 2013, through April 1, 2023, for Units 1 and 2, respectively.

1.6 First Interval CISI Program CISI examinations were originally invoked by amended regulations contained within a Final Rule issued by the NRC. The amended regulation incorporated the requirements of the 1992 Edition through the 1992 Addenda of the ASME Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-8 ISI Program Plan - 5th Interval Section XI, Subsection IWE, subject to specific modifications that were included in Paragraph 10CFR50.55a(b)(2)(x).

The final rulemaking was published in the Federal Register on August 8, 1996, and specified an effective date of September 9, 1996. Implementation of the Subsection IWE Program from a scheduling standpoint was driven by the five year expedited implementation period per 10CFR50.55a(g)(6)(ii)(B), which specified that the examinations required to be completed by the end of the first period of the First CISI Interval (per Table IWE-2412-1) be completed by the effective date (by September 9, 2001).

ASME Section XI Subsections IWE, Mandatory Appendices of ASME Section XI, approved IWE Code Cases, and approved alternatives through relief requests and SEs were developed to implement these requirements.

The QCNPS First CISI Interval was effective from September 30, 1998, through September 8, 2008, for Units 1 and 2, respectively.

1.7 Second Interval CISI Program Pursuant to 10CFR50.55a(g), licensees are required to update their CISI Programs to meet the requirements of ASME Section XI once every ten years or inspection interval. The CISI Program is required to comply with the latest Edition and Addenda of ASME Section XI incorporated by reference in 10CFR50.55a twelve months prior to the start of the interval per 10CFR50.55a(g)(4)(ii). The start of the Second CISI Interval was September 9, 2008. Based on this date, the latest Edition and Addenda of ASME Section XI referenced in 10CFR50.55a(b)(2) twelve months prior to the start of the Second CISI Interval was the 2001 Edition through the 2003 Addenda.

The QCNPS Second Interval CISI Program Plan was developed in accordance with the requirements of 10CFR50.55a and the 2001 Edition through the 2003 Addenda of ASME Section XI, subject to the limitations and modifications contained within Paragraph (b) of the regulation. These limitations and modifications are detailed in Table 1.8-1 of this section. This Second Interval CISI Program Plan addresses Subsections IWE, Mandatory Appendices of ASME Section XI, approved IWE Code Cases, approved alternatives through relief requests and SEs, and utilizes Inspection Program B.

The QCNPS Second CISI Interval is effective from September 9, 2008, through September 8, 2018, for Units 1 and 2, respectively.

1.8 Code of Federal Regulations 10CFR50.55a Requirements There are certain Paragraphs in 10CFR50.55a that list the limitations, modifications, and/or clarifications to the implementation requirements of ASME Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-9 ISI Program Plan - 5th Interval Section XI. These Paragraphs in 10CFR50.55a that are applicable to QCNPS are detailed in Table 1.8-1.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-10 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(ix)(A) (CISI) Examination of metal containments and the liners of concrete containments: For Class MC applications, the licensee shall evaluate the acceptability of inaccessible areas when conditions exist in accessible areas that could indicate the presence of or result in degradation to such inaccessible areas. For each inaccessible area identified, the licensee shall provide the following in the ISI Summary Report as required by IWA-6000:

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

10CFR50.55a(b)(2)(ix)(B) (CISI) Examination of metal containments and the liners of concrete containments: When performing remotely the visual examinations required by Subsection IWE, the maximum direct examination distance specified in Table IWA-2210-1 may be extended and the minimum illumination requirements specified in Table IWA-2210-1 may be decreased, provided that the conditions or indications for which the visual examination is performed can be detected at the chosen distance and illumination.

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-11 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(ix)(G) (CISI) Examination of metal containments and the liners of concrete containments: The VT-3 examination method must be used to conduct the examinations in Items E1.12 and E1.20 of Table IWE-2500-1, and the VT-1 examination method must be used to conduct the examination in Item E4.11 of Table IWE-2500-1. An examination of the pressure-retaining bolted connections in Item E1.11 of Table IWE-2500-1 using the VT-3 examination method must be conducted once each interval. The owner-defined visual examination provisions in IWE-2310(a) are not approved for use for VT-1 and VT-3 examinations.

10CFR50.55a(b)(2)(ix)(H) (CISI) Examination of metal containments and the liners of concrete containments: Containment bolted connections that are disassembled during the scheduled performance of the examinations in Item E1.11 of Table IWE-2500-1 must be examined using the VT-3 examination method. Flaws or degradation identified during the performance of a VT-3 examination must be examined in accordance with the VT-1 examination method. The criteria in the material specification or IWB-3517.1 must be used to evaluate containment bolting flaws or degradation. As an alternative to performing VT-3 examinations of containment bolted connections that are disassembled during the scheduled performance of Item E1.11, VT-3 examinations of containment bolted connections may be conducted whenever containment bolted connections are disassembled for any reason.

10CFR50.55a(b)(2)(ix)(I) (CISI) Examination of metal containments and the liners of concrete containments: The ultrasonic examination acceptance standard specified in IWE-3511.3 for Class MC pressure-retaining components must also be applied to metallic liners of Class CC pressure-retaining components.

10CFR50.55a(b)(2)(xii) (ISI) Underwater welding: The provisions in IWA-4660, Underwater Welding, of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, are not approved for use on irradiated material.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-12 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(xviii)(A) (ISI) Certification of NDE personnel: Level I and II nondestructive examination personnel shall be recertified on a 3-year interval in lieu of the 5-year interval specified in the 1997 Addenda and 1998 Edition of IWA-2314, and IWA-2314(a) and IWA-2314(b) of the 1999 Addenda through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section.

10CFR50.55a(b)(2)(xix) (ISI) Substitution of alternative methods: The provisions for substituting alternative examination methods, a combination of methods, or newly developed techniques in the 1997 Addenda of IWA-2240 must be applied when using the 1998 Edition through the 2004 Edition of Section XI of the ASME B&PV Code. The provisions in IWA-4520(c), 1997 Addenda through the 2004 Edition, allowing the substitution of alternative methods, a combination of methods, or newly developed techniques for the methods specified in the Construction Code are not approved for use. The provisions in IWA-4520(b)(2) and IWA-4521 of the 2008 Addenda through the latest edition and addenda approved in paragraph (b)(2) of this section, allowing the substitution of ultrasonic examination for radiographic examination specified in the Construction Code are not approved for use.

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

10CFR50.55a(b)(2)(xxii) (ISI) Surface Examination: The use of the provision in IWA-2220, Surface Examination, of Section XI, 2001 Edition through the latest Edition and Addenda incorporated by reference in paragraph (b)(2) of this section, that allow use of an ultrasonic examination method is prohibited.

10CFR50.55a(b)(2)(xxiii) (ISI) Evaluation of Thermally Cut Surfaces: The use of the provisions for eliminating mechanical processing of thermally cut surfaces in IWA-4461.4.2 of Section XI, 2001 Edition through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section are prohibited.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-13 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(2)(xxiv) (PDI) Incorporation of the performance demonstration initiative and addition of ultrasonic examination criteria:

The use of Appendix VIII and the supplements to Appendix VIII and Article I-3000 of Section XI of the ASME B&PV Code, 2002 Addenda through the 2006 Addenda is prohibited.

10CFR50.55a(b)(2)(xxv) (ISI) Mitigation of Defects by Modification: The use of the provisions in IWA-4340, Mitigation of Defects by Modification, Section XI, 2001 Edition through the latest Edition and Addenda incorporated by reference in Paragraph (b)(2) of this section are prohibited.

10CFR50.55a(b)(2)(xxvi) (SPT) Pressure Testing Class 1, 2, and 3 Mechanical Joints:

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

10CFR50.55a(b)(2)(xxvii) (ISI) Removal of Insulation: When performing visual examinations in accordance with IWA-5242 of Section XI of the ASME B&PV Code, 2003 Addenda through the 2006 Addenda, or IWA-5241 of the 2007 Edition through the latest Edition and Addenda incorporated in Paragraph (b)(2) of this section, insulation must be removed from 17-4 PH or 410 stainless steel studs or bolts aged at a temperature below 1100

°F or having a Rockwell Method C hardness value above 30, and from A-286 stainless steel studs or bolts preloaded to 100,000 pounds per square inch or higher.

10CFR50.55a(b)(2)(xxix) (ISI) Nonmandatory Appendix R: Nonmandatory Appendix R, Risk-Informed Inspection Requirements for Piping, of Section XI, 2005 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, may not be implemented without prior NRC authorization of the proposed alternative in accordance with paragraph (a)(3)(i) of this section.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-14 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(3)(v)(B) (ISI) Subsection ISTD: Article IWF-5000, Inservice Inspection Requirements for Snubbers, of the ASME BP&V Code, Section XI, must be used when performing inservice inspection examinations and tests of snubbers at nuclear power plants, except as conditioned in Paragraph (b)(3)(v)(B) of this section. (B) Licensees shall comply with the provisions for examining and testing snubbers in Subsection ISTD of the ASME OM Code and make appropriate changes to their technical specifications or licensee-controlled documents when using the 2006 Addenda and later editions and addenda of Section XI of the ASME B&PV Code.

10CFR50.55a(b)(5) (ISI) Inservice Inspection Code Cases: Licensees may apply the ASME Boiler and Pressure Vessel Code Cases listed in Regulatory Guide 1.147, Revision 16, without prior NRC approval subject to the following:

(i) When a licensee initially applies a listed Code Case, the licensee shall apply the most recent version of that Code Case incorporated by reference in this paragraph.

(ii) If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in this paragraph, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use.

(iii) Application of an annulled Code Case is prohibited unless a licensee previously applied the listed Code Case prior to it being listed as annulled in Regulatory Guide 1.147. Any Code Case listed as annulled in any Revision of Regulatory Guide 1.147 which a licensee has applied prior to it being listed as annulled, may continue to be applied by that licensee to the end of the 120-month interval in which the Code Case was implemented.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-15 ISI Program Plan - 5th Interval TABLE 1.8-1 CODE OF FEDERAL REGULATIONS 10CFR50.55a REQUIREMENTS 10CFR50.55a Paragraphs Limitations, Modifications, and Clarifications 10CFR50.55a(b)(6) (ISI) Operation and Maintenance of Nuclear Power Plants Code Cases: Licensees may apply the ASME Operation and Maintenance Nuclear Power Plants Code Cases listed in Regulatory Guide 1.192 without prior NRC approval subject to the following:

(i) When a licensee initially applies a listed Code Case, the licensee shall apply the most recent version of that Code Case incorporated by reference in this paragraph.

(ii) If a licensee has previously applied a Code Case and a later version of the Code Case is incorporated by reference in this paragraph, the licensee may continue to apply, to the end of the current 120-month interval, the previous version of the Code Case as authorized or may apply the later version of the Code Case, including any NRC-specified conditions placed on its use.

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-16 ISI Program Plan - 5th Interval 1.9 Code Cases Per 10CFR50.55a(b)(5), Code Cases that have been determined to be suitable for use in ISI Program Plans by the NRC are listed in Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability-ASME Section XI, Division 1.

The approved Code Cases in Regulatory Guide 1.147 being utilized by QCNPS are included in Section 2.1.1 of this document. The most recent version of a given Code Case incorporated in the revision of Regulatory Guide 1.147, referenced in 10CFR50.55a(b)(5)(i), at the time it is applied within the ISI Program shall be used. The latest version of Regulatory Guide 1.147 incorporated into this document is Revision 16. As this guide is revised, newly approved Code Cases will be assessed for plan implementation at QCNPS per Paragraph IWA-2441(e) and proposed for use in revisions to the ISI Program Plan.

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

Per 10CFR50.55a(b)(6), this ISI Program Plan will also utilize Regulatory Guide 1.192, Operation and Maintenance Code Case Acceptability, ASME OM Code.

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

1.10 Relief Requests In accordance with 10CFR50.55a, when a licensee either proposes alternatives to ASME Section XI requirements which provide an acceptable level of quality and safety, determines compliance with ASME Section XI requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, or determines that specific ASME Section XI requirements for inservice inspection are impractical, the licensee shall notify the NRC and submit information to support the determination.

The submittal of this information will be referred to in this document as a relief request. Relief requests for the Fifth ISI Interval and the Second CISI Interval are included in Section 8.0 of this document. The text of the relief requests contained in Section 8.0 will demonstrate one of the following: the proposed alternatives provide an acceptable level of quality and safety per 10CFR50.55a(a)(3)(i), compliance with the specified requirements would result Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 1-17 ISI Program Plan - 5th Interval in hardship or unusual difficulty without a compensating increase in the level of quality and safety per 10CFR50.55a(a)(3)(ii), or the code requirements are considered impractical per 10CFR50.55a(g)(5)(iii).

Per 10CFR50.55a Paragraphs (a)(3) and (g)(6)(i), the Director of the Office of Nuclear Reactor Regulation will evaluate relief requests and may grant such relief and may impose such alternative requirements as it determines is authorized by law and will not endanger life or property or the common defense and security and is otherwise in the public interest giving due consideration to the burden upon the licensee that could result if the requirements were imposed on the facility.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-1 ISI Program Plan - 5th Interval 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2.1 ASME Section XI Examination Requirements As required by 10CFR50.55a, this program was developed in accordance with the requirements detailed in the 2007 Edition through the 2008 Addenda of the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Subsections IWA, IWB, IWC, IWD, IWE, IWF, Mandatory Appendices, Inspection Program of Paragraph IWA-2431, and approved alternatives through relief requests and SEs.

The Performance Demonstration Initiative (PDI) is an organization comprised of all US nuclear utilities that was formed to provide an efficient implementation of Appendix VIII performance demonstration requirements. The Electric Power Research Institute (EPRI) NDE Center was selected as the administrator of this program. The PDI program is administered according to the PDI Program Description. The ISI Program implements Appendix VIII, Performance Demonstration for Ultrasonic Examination Systems, ASME Section XI 2007 Edition through the 2008 Addenda as required by 10CFR50.55a(b)(2)(xxiv) and with modifications as identified in 10CFR50.55a(b)(2)(xiv), (xv), and (xvi).

Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in piping, bolting, and the reactor pressure vessel (RPV). Each organization (e.g., owner or vendor) will be required to have a written program to ensure compliance with the requirements. QCNPS maintains the responsibility to ensure that Appendix VIII requirements are properly implemented.

For the Fifth ISI Interval, QCNPSs inspection program for ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 will be governed by risk-informed requirements. The RISI Program methodology is described in the EPRI Topical Report TR-112657, Rev. B-A. To supplement the EPRI Topical Report, ASME Code Case N-578-1 (as applicable per Relief Request I5R-02) is also being used for the classification of piping structural elements under the RISI Program. The RISI Program scope has been implemented as an alternative to the 2007 Edition through the 2008 Addenda, ASME Section XI examination program for ISI Class 1 B-F and B-J welds and ISI Class 2 C-F-1 and C-F-2 welds in accordance with 10CFR50.55a(a)(3)(i). The basis for the resulting risk categorizations of the nonexempt ISI Class 1 and 2 piping systems at QCNPS is defined and maintained in the Final Report, Risk Informed Inservice Inspection Evaluation, as referenced in Section 9.0 of this document. References to ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 have been replaced with Examination Category R-A to identify them as part of the RISI Program.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-2 ISI Program Plan - 5th Interval The CISI Program Plan per Subsection IWE has been incorporated into Section 6.0 Containment ISI Plan of this ISI Program Plan. The CISI relief requests are included in Section 8.0 of this document.

2.1.1 ASME Section XI Code Cases As referenced by 10CFR50.55a(b)(5) and allowed by NRC Regulatory Guide 1.147, Revision 16 the following Code Cases are being incorporated into the QCNPS ISI Program:

N-62-7 Internal and External Valve Items, Section III, Division 1, Classes 1, 2, and 3 Code Case N-62-7 is acceptable subject to the following conditions in Regulatory Guide 1.84, Revision 35:

The Code requires that ISI Class 1 and 2 valve manufacturers meet the provisions of NCA-4000, Quality Assurance. ISI Class 3 valve manufacturers must also meet the provisions of NCA-4000 because all Code Class valve items are subject to the licensees 10CFR, Part 50, Appendix B approved QA program.

Note: Code Case N-62-7 (approved for use per Regulatory Guide 1.84, Revision 35) was utilized for guidance in defining pressure retaining bolting for valves.

N-405-1 Socket Welds, Section III, Division 1 Note: Code Case N-405-1 (approved for use per Regulatory Guide 1.84, Revision 35) was utilized for guidance for appurtenances (with outside diameter equal to that of 2 in. standard pipe size and less) that connect to nozzles of a Section III, Class 1 vessels. These appurtenances may be designed and constructed using weld joints in accordance with Fig. 1 and provided that requirements (a through e) listed in the Code Case are met.

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

Regulatory Guide 1.147, Revision 16.

N-513-3 Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-3 ISI Program Plan - 5th Interval Code Case N-513-3 is acceptable subject to the following condition specified in Regulatory Guide 1.147, Revision 16:

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

N-516-3 Underwater Welding Code Case N-516-3 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16.

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

N-526 Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels. Regulatory Guide 1.147, Revision 16.

N-532-4 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000. Regulatory Guide 1.147, Revision 16.

Note: Limited to the 2005 Addenda by references in Table 3 of the Code Case. Code Case N-532-4 to be used in accordance with Relief Request I5R-09.

N-578-1 Risk-Informed Requirements for Class 1, 2, or 3 Piping, Method B.

Note: Code Case N-578-1 to be used in accordance with Relief Request I5R-02.

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

Regulatory Guide 1.147, Revision 16.

Note: RISI Program Relief Request I5R-02 requires that scope expansion for RISI elements will be determined using Paragraph -2430 of ASME Code Case N-578-1.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-4 ISI Program Plan - 5th Interval N-597-2 Requirements for Analytical Evaluation of Pipe Wall Thinning Code Case N-597-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16.

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

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

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

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

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-5 ISI Program Plan - 5th Interval N-600 Transfer of Welder, Welding Operator, Brazer, and Brazing Operator Qualifications Between Owners. Regulatory Guide 1.147, Revision 16.

N-606-1 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stud Tube Repairs Code Case N-606-1 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16:

Prior to welding, an examination or verification must be performed to ensure proper preparation of the base metal, and that the surface is properly contoured so that an acceptable weld can be produced. The surfaces to be welded, and surfaces adjacent to the weld, are to be free from contaminants, such as, rust, moisture, grease, and other foreign material or any other condition that would prevent proper welding and adversely affect the quality or strength of the weld. This verification is to be required in the welding procedures.

N-613-1 Ultrasonic Examination of Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Nozzle-to-Vessel Welds, Figs. IWB-2500-7(a), (b),

and (c). Regulatory Guide 1.147, Revision 16.

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

Regulatory Guide 1.147, Revision 16.

N-638-4 Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique Code Case N-638-4 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16:

(1) Demonstration for ultrasonic examination of the repaired volume is required using representative samples which contain construction type flaws.

(2) The provisions of 3(e)(2) or 3(e)(3) may only be used when it is impractical to use the interpass temperature measurement methods described in 3(e)(1), such as in situations where the weldment Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-6 ISI Program Plan - 5th Interval area is inaccessible (e.g., internal bore welding) or when there are extenuating radiological conditions Note: Limited to the 2004 Edition by references in Table 1 of the Code Case.

N-639 Alternative Calibration Block Material Code Case N-639 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16:

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

N-641 Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements. Regulatory Guide 1.147, Revision 16.

N-649 Alternative Requirements for IWE-5240 Visual Examination. Regulatory Guide 1.147, Revision 16.

Note: Limited to the 2000 Addenda by references in the Code Case.

N-651 Ferritic and Dissimilar Metal Welding Using SMAW Temper Bead Technique Without Removing the Weld Bead Crown of the First Layer. Regulatory Guide 1.147, Revision 16.

N-661-1 Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service Code Case N-661-1 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16:

(1) If the cause of the degradation has not been determined, the repair is only acceptable until the next refueling outage.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-7 ISI Program Plan - 5th Interval (2) When through-wall repairs are made by welding on surfaces that are wet or exposed to water, the weld overlay repair is only acceptable until the next refueling outage.

Note: Limited to the 2005 Addenda in ASME Code Case Applicability Index. Code Case N-661-1 to be used in accordance with Relief Request I5R-09.

N-666 Weld Overlay of Class 1, 2, and 3 Socket Welded Connections. Regulatory Guide 1.147, Revision 16.

Note: Limited to the 2004 Edition by references in Table 1 of the Code Case.

N-705 Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks. Regulatory Guide 1.147, Revision 16.

N-730 Roll Expansion of Class 1 Control Rod Drive Bottom Head Penetrations in Boiling Water Reactors (BWR).

Regulatory Guide 1.147, Revision 16.

N-733 Mitigation of Flaws in NPS 2 (DN 50) and Smaller Nozzles and Nozzle Partial Penetration Welds in Vessels and Piping by Use of a Mechanical Connection Modification.

Regulatory Guide 1.147, Revision 16.

Note: Limited to the 2006 Addenda by references to IWA-5242(a) in (h)(2)(d) of the Code Case.

N-735 Successive Inspection of Class 1 and 2 Piping Welds.

Regulatory Guide 1.147, Revision 16.

N-751 Pressure Testing of Containment Penetration Piping.

Regulatory Guide 1.147, Revision 16.

Code Case N-751 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 16:

When a 10CFR Part 50, Appendix J, Type C test is performed as an alternative to the requirements of IWA-4540 (IWA-4700 in the 1989 Edition through the 1995 Edition) during repair and replacement activities, nondestructive examination must be performed in Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-8 ISI Program Plan - 5th Interval accordance with IWA-4540(a)(2) of the 2002 Addenda of Section XI.

Additional Code Cases invoked in the future shall be in accordance with those approved for use in the latest published revision of Regulatory Guide 1.147 or 10CFR50.55a at that time.

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

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

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

2.2 Augmented Inspection Program Requirements Augmented inspection program requirements are those inspections that are performed above and beyond the requirements of ASME Section XI. Below is a summary of those inspections performed by QCNPS that are not specifically addressed by ASME Section XI, or the inspections that will be performed in addition to the requirements of ASME Section XI on a routine basis during the Fifth ISI Interval (Second CISI Interval).

2.2.1 Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping, Revision 2 / Supplement 1 to Generic Letter 88-01, NUREG-0313, Technical Report on Material Selection and Process Guidelines for BWR Coolant Pressure Boundary Piping, Revision 2, EPRI Topical Report TR-113932, BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), as conditionally approved by NRC final SEs dated September 15, 2000, and May 14, 2002, and EPRI Topical Report TR-1012621, BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), as conditionally approved by NRC final SE dated March 16, 2006 These documents discuss the examination requirements for Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping. References to Generic Letter 88-01 (GL 88-01) within the ISI Program refer to the comprehensive commitments to all of these Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-9 ISI Program Plan - 5th Interval documents. The final SEs of BWRVIP-75 and BWRVIP-75-A revised the GL 88-01 inspection schedules. The BWRVIP-75 and BWRVIP-75-A revised inspection schedules were based on consideration of inspection results and service experience gained by the industry since issuance of GL 88-01 and NUREG-0313, and includes additional knowledge regarding the benefits of improved BWR water chemistry.

QCNPS has committed to the requirements of these documents as discussed in Updated Final Safety Analysis Report (UFSAR) Section 5.2.3.5. The original QCNPS commitment concerning Generic Letter 88-01 was sent to the NRC in a letter from W. E. Morgan (CECo) to the NRC dated July 29, 1988. The NRC reviewed this commitment in letters from T. M. Ross (NRC) to T. J. Kovach (CECo) dated May 22, 1989, and from L N. Olshan (NRC) to T. J. Kovach (CECo) August 21, 1990. Since the issuance of GL 88-01, the BWR Vessel and Internals Project (BWRVIP) has been created. This BWR owners group has worked on the mitigation of IGSCC for BWR internal components. As part of their activities, EPRI Topical Report TR-113932, BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), dated October 27, 1999, and EPRI Topical Report TR-1012621, BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), dated October 2005, were submitted to the NRC.

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

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

Based upon NRC endorsement of BWRVIP-75 and BWRVIP-75-A, the QCNPS conformance to GL 88-01 inspection schedules was changed to BWRVIP-75 and BWRVIP-75-A for NWC plants except for Category A welds. (See Risk-Informed Inservice Inspection discussion below and BWRVIP discussion in Section 2.2.4) The BWRVIP-75 and Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-10 ISI Program Plan - 5th Interval BWRVIP-75-A interval began on January 1, 2003, for both QCNPS Units 1 and 2.

The outboard RWCU piping has been excluded from Generic Letter 88-01. The basis for this exclusion is documented in a letter from P. L.

Piet (CECo) to T. E. Murley (NRC) dated August 20, 1993, and in a letter from the NRC from R. M. Pulsifer (NRC) to D. L. Farrar (ComEd) dated September 22, 1994.

RISI regulations have been invoked for QCNPS in this ISI Program Plan.

Under these new guidelines, ISI Class 1 and 2 piping structural elements are inspected in accordance with EPRI Topical Report TR-112657, Rev.

B-A and ASME Code Case N-578-1. Per this Topical Report and Code Case, welds within the plant that are assigned to IGSCC Categories B through G will continue to meet existing IGSCC schedules, while IGSCC Category A welds have been subsumed into the RISI Program.

The First Ten-Year GL 88-01/BWRVIP-75-A inspection schedule started in 2000 and ended in 2010. The Second GL 88-01/BWRVIP-75-A inspection schedule started in 2011 and will end in 2020. Hydrogen Water Chemistry is credited through IR 1263859.

Implementation of the QCNPS program addressing these documents is included in Section 7.0 of this ISI Program Plan and the associated ISI Database.

This section describes inspections credited for License Renewal Aging Management Plan B.1.7 BWR Stress Corrosion Cracking (AT 101562.07).

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

Report GE-NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements, August 1999, as conditionally approved by NRC final SE dated June 5, 1998, and NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking, dated November 1980 These documents discuss the current and initial examination requirements for BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking. The alternate approach was developed and submitted to the NRC by the BWROG. The NRC accepted these alternate requirements in a final SE dated March 10, 2000.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-11 ISI Program Plan - 5th Interval QCNPS initially committed to the requirements of NUREG-0619 as stated in the Third Interval ISI Program Plan. QCNPS revised this commitment to utilize the BWROG alternate inspections in a letter from J. P. Dimmette (ComEd) to the NRC dated October 15, 1998. The NRC sent a letter from R. M. Pulsifer (NRC) to O. D. Kingsley (ComEd) dated April 30, 1999, to confirm the discussions of an April 1, 1999, conference call in that ComEd will use the more recent fatigue curves that address environmental effects as approved by ASME Section XI. Future inspections will comply with BWROG Alternate BWR Feedwater Nozzle Inspection Requirements, GE-NE-523-A71-0594-A, Revision 1, dated May 2000, as accepted by NRC SE (TAC No. MA6787) dated March 10, 2000.

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

This section describes inspections credited for License Renewal Aging Management Plan B.1.5 BWR Feedwater Nozzles (NUREG-0619)

(AT 101562.05).

2.2.3 BWR Vessel and Internals Project (BWRVIP)

Increased awareness of the presence of in-vessel component degradation has led to the formation of the BWRVIP. BWRVIP is an association of BWR utilities focused on the common purpose of investigating and developing effective and acceptable approaches for addressing in-vessel component degradation through improved detection, mitigation, and/or repair techniques. In accordance with the BWRVIP charter, the organization is tasked with providing generic resolution to BWR issues and representing the member utilities in negotiating with the NRC for approval of the groups recommended actions. Exelon, as a member utility of the BWRVIP, has endorsed the objectives prescribed by the BWRVIP.

The BWRVIP is comprised of a series of Inspection & Evaluation Guidelines and documents that discuss RPV internals. The BWRVIP encompasses pertinent information and requirements presented in I.E.

Bulletins (IEBs), General Electric (GE) Service Information Letters (SILs) and Rapid Information Communication Services Information Letters (RICSILs). The BWRVIP guidelines are intended to be followed in lieu of GE SILs and RICSILs issued prior to issuance of the BWRVIP guidelines.

Exelons commitments to the BWRVIP are discussed in BWRVIP letters to the NRC dated May 30, 1997, and October 30, 1997. The NRCs response to the discussion of BWRVIP utility commitments is discussed in an NRC letter to the BWRVIP dated July 29, 1997.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-12 ISI Program Plan - 5th Interval Examinations of RPV internals, as required by QCNPS commitments to the BWRVIP and other related documents, are performed in accordance with procedure ER-AB-331.

Implementation of BWRVIP inspection requirements in lieu of ASME Section XI inspection requirements for Examination Categories B-N-1 and B-N-2 was requested in Relief Request I5R-06.

2.2.4 NUREG-0737, dated November 1980 This document discusses TMI Action Plan Requirements, and includes requirements in Item III.D.1.1 for leak testing and periodic visual examinations of systems outside of primary containment which could contain highly radioactive fluids during a serious transient or accident.

QCNPS has committed to the requirements of this document item as discussed in Technical Specification Section 5.5.2. Commitments made concerning NUREG-0737 are required to be maintained per the QCNPS Operating Licenses.

Implementation of the QCNPS program addressing these requirements is included in Procedure QCTP 0820-08 and Section 7.0 of this ISI Program Plan.

2.2.5 NUREG-1796 Portions Regarding Nonexempt Class MC Piping and Class MC Component Supports, dated October 2004 This commitment results from resolution of Open Item (OI) 3.5.2.3.2-1 in NUREG-1796 Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 and 2, and the Exelon response letter dated April 22, 2004.

Nonexempt Class MC Piping Supports The details of the augmented Piping Support inspections are contained in License Renewal Application (LRA), Appendix B, Item B.1.30, and in Commitment #30 in the associated SE, as well as in UFSAR Supplement Appendix A (see ATI # 101562.49.48).

The commitment modifies the stations Structural Monitoring Program (SMP) to perform augmented Class MC piping support inspections. These augmented inspections are to include a 15% sample size of nonexempt piping (> 4" NPS) supports, distributed across systems with such supports, and performed by certified VT-3visual inspectors. The total number of nonexempt Class MC piping supports under this commitment includes 26 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-13 ISI Program Plan - 5th Interval Unit 1 and 37 Unit 2 supports from the Reactor Building Closed Cooling Water and Primary Containment Vent and Purge systems.

Class MC Component Supports The details of the augmented Component Support inspections are contained in License Renewal Application (LRA), Appendix B, Item B.1.27 and the associated SE (see ATI # 101562.37).

The commitment is to perform VT-3 visual examinations on the Drywell, Suppression Chamber, and Vent System supports. These examinations will be similar to those required by ASME Section XI, Subsection IWF (see ATI # 101562.37).

Finally, the NUREG-1796 Piping and Component Support commitment includes performing a baseline inspection of both the sample Class MC piping supports and the Class MC component supports prior to entering the period of extended operation (for Unit 1 - midnight on 12/14/12; and for Unit 2 - midnight on 12/14/12). After these initial baseline inspections have been completed, the 15% piping support sample population and the component supports will be scheduled and inspected during the Fifth ISI Interval and successive intervals thereafter.

Implementation of the examination commitments is included in Sections 6.0 and 7.0 of this ISI Program Plan and the associated ISI Database.

2.2.6 License Renewal Requirements for Cast Austenitic Stainless Steel Components The QCNPS License Renewal process developed an Aging Management Program (AMP) to address cast austenitic stainless steel (CASS) components in the reactor vessel that are susceptible to a loss of fracture toughness due to thermal aging/neutron embrittlement. This AMP was developed in accordance with NUREG-1801, Generic Aging Lessons Learned. Procedure ER-AB-331-101, Evaluation for Thermal Aging Neutron Embrittlement of Reactor Internals Components, identified the following CASS components as susceptible to thermal aging/neutron embrittlement at QCNPS: the fuel support piece, the control rod guide tube base, the Jet Pump (JP) mixer flange, JP mixer flare, JP mixer ring, JP inlet-mixer nozzles and JP inlet-mixer elbows.

The evaluation of the CASS components concluded that periodic inspection was required for each of the identified components. The inspections consist of enhanced VT-l visual examinations on a sample of the identified components each refuel outage. As part of the license renewal commitment documented in A/R 101562-49-01, these inspections Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-14 ISI Program Plan - 5th Interval are to be included in the ISI Program. The evaluation and inspection requirements are documented in SVP-10-080, Revised License Renewal Commitment for Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Reactor Internal Components.

Note: The components and schedule of examinations are maintained and implemented in the QCNPS Reactor Internals Program documents.

This section describes inspections credited for License Renewal Aging Management Plan B.1.10 Thermal Aging of CASS (AT 101562.13).

2.3 System Classifications and P&ID Boundary Drawings The ISI Classification Basis Document details those systems that are ISI Class 1, 2, or 3 that fall within the inservice inspection scope of examinations including the containment structure (metal). Below is a summary of the classification criteria used within the Basis Document.

Each safety related fluid system containing water, steam, air, oil, etc. included in the QCNPS UFSAR was reviewed to determine which safety functions they perform during all modes of system and plant operation. Based on these safety functions, the systems and components were evaluated per classification documents. The systems were then designated as ISI Class 1, 2, 3, or non-classed accordingly. This evaluation followed the guidelines of UFSAR Section 5.2.4 for ISI Class 1 and UFSAR Section 6.6 for ISI Classes 2 and 3. Safety related portions of systems are defined by the Piping and Instrumentation Diagrams (P&IDs) with an S flag.

When a particular group of components is identified as performing a ISI Class 1, 2, or 3 safety function, the components are further reviewed to assure the interfaces (boundary valves and boundary barriers) meet the criteria set by 10CFR50.2, 10CFR50.55a(c)(1), 10CFR50.55a(c)(2), and Regulatory Guide 1.26.

Although QCNPS is not committed to or licensed in accordance with these documents, Standard Review Plan (SRP) 3.2.2 System Quality Group Classification, and American National Standards Institute/American Nuclear Society (ANSI/ANS)-58.14-1993 Safety and Pressure Integrity Classification Criteria for Light Water Reactors, were also used for guidance in evaluating the classification boundaries when 10CFR and Regulatory Guide 1.26 did not address a given situation. The valve positions shown on the system flow diagrams are assumed to be the normal positions during system operation unless otherwise noted.

At the time the construction permits for QCNPS Units 1 and 2 were issued, ASME Section III covered only pressure vessels, primarily nuclear reactor vessels. The majority of piping, pumps, and valves were designed and installed according to the rules of USAS B31.1.0-1967 Edition, Power Piping.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-15 ISI Program Plan - 5th Interval Consequently, the QCNPS ISI Program has essentially no ASME Section III Class 1, 2, or 3 piping systems.

ISI classification boundaries are defined by the P&IDs with a classification flag.

A summary of the coding system used on the P&IDs to identify the safety related systems or portions of systems subject to examination is included on Drawing M-12 SH. 3. The coding designators 1, 2, 3, and MC, respectively, were used for classifying nonexempt ASME Section XI components. The remaining codings shown on M-12 SH. 3 (Coding Designators 1C, 1F, 1S, 1V, 2E, 2P, 2V, 3G, and 3P) were used to identify exempt ASME Section XI components.

The systems and components (piping, pumps, valves, vessels, etc.), which are subject to examinations of Articles IWB-2000, IWC-2000, IWD-2000, and IWF-2000, and pressure tests of Articles IWB-5000, IWC-5000 and IWD-5000 are identified on the QCNPS P&IDs as detailed in Table 2.3-1. The systems and components, which are subject to the examinations of Article IWE-2000 are identified on the P&IDs as detailed in Table 2.3-2.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-16 ISI Program Plan - 5th Interval TABLE 2.3-1 P&ID BOUNDARY DRAWINGS (ISI)

UNIT 1 & COMMON UNIT 2 TITLE M-12, SH. 3 M-12, SH. 3 Piping & Instrumentation Diagram Symbols M-13, SH. 1 & 2 M-60, SH. 1 & 2 Diagram of Main Steam Piping (MS)

M-15, SH. 1 M-62 SH. 1 Diagram of Reactor Feed Piping (FW)

M-22, SH. 1, 3 & 5 M-69, SH. 1, 3 & 5 Diagram of Service Water Piping - Diesel Generator Cooling Water (DGCW)

M-34, SH. 1 M-76, SH. 1 Diagram of Pressure Suppression Piping (PS)

M-35, SH. 1, 2, & 5 M-77, SH. 1, 2, & 5 Diagram of Nuclear Boiler & Reactor Recirculation Piping (RX & RR)

M-36 M-78 Diagram of Core Spray Piping (CS)

M-37 M-79 Diagram of RHR Service Water Piping (RHR & RHRSW)

M-39, SH. 1, 2, 3 & 4 M-81, SH. 1, 2 & 3 Diagram of Residual Heat Removal Piping (RHR &

RHRSW)

M-40 M-82 Diagram of Standby Liquid Control Piping (SBLC)

M-41, SH. 1 & 3 M-83, SH. 1 & 3 Diagram of Control Rod Drive Hydraulic Piping (CRD)

M-46, SH. 1, 2, & 3 M-87, SH. 1, 2, & 3 Diagram of H.P. Coolant Injection Piping (HPCI)

M-47 SH. 1 M-88 SH. 1 Diagram of Reactor Water Clean-Up Piping (RWCU)

M-50 SH. 1 M-89 SH. 1 Diagram of Reactor Core Isolation Cooling Piping (RCIC)

M-70 M-70 Diagram of Safe Shutdown Make-Up Pump System (SSMP)

M-725, SH. 3 M-725, SH. 3 Diagram of Control Room HVAC M-1056, SH. 1 M-1061, SH. 1 Diagram of High Radiation Sampling System Piping (HRSS)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-17 ISI Program Plan - 5th Interval TABLE 2.3-2 P&ID BOUNDARY DRAWINGS (CISI)

UNIT 1 & COMMON UNIT 2 TITLE M-24 SH. 13 M-71 SH. 8 Containment - Diagram of Instrument Air Piping M-25 SH. 1 M-72 SH. 1 Containment - Diagram of Service Air Piping M-33 SH. 2 M-75 SH. 2 Containment - Diagram of Reactor Building Closed Cooling Water Piping (RBCCW)

M-34 SH. 1 M-76 SH. 1 Containment - Diagram of Pressure Suppression Piping (PS)

M-35 SH. 2 M-77 SH. 2 Containment - Diagram of Nuclear Boiler &

Recirculation Piping M-43 M-85 Containment - Diagram of Reactor Building Equipment Drains M-46 SH. 2 M-87 SH. 2 Containment - Diagram of High Pressure Coolant Injection Piping (HPCI)

Containment - Diagram of Reactor Core Isolation M-50 SH. 1 M-89 SH. 1 Cooling Piping (RCIC)

Containment - Diagram of Clean & Contaminated M-58 SH. 4 M-58 SH. 4 Condensate Piping Containment - Diagram of Process Sampling Part 3 -

M-461 SH. 1 M-463 SH. 1 Primary Containment Sampling Systems Containment - Diagram of Transversing In-Core Probe M-584 SH. 1 M-584 SH. 2 System (TIP)

M-641 SH. 1 M-641 SH. 2 Containment - Diagram of Containment Atmosphere Monitor System (CAM)

M-642 SH. 1 M-642 SH. 2 Containment - Diagram of Atmospheric Containment Atmosphere Dilution System (ACAD)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-18 ISI Program Plan - 5th Interval 2.4 ISI Isometric Drawings for Nonexempt ISI Class Components/Supports and Calibration Standards ISI isometric and component drawings were developed to identify the ISI Class 1, 2, and 3 components (welds, bolting, etc.) and support locations at QCNPS. SPT isometric drawings were also developed to show those components subject to pressure testing. These ISI components and supports are identified on the ISI isometric and component drawings listed in Table 2.4-1. The ISI Class MC components are identified on the CISI Reference Drawings listed in Tables 2.4-2 and 2.4-3 and described in Section 6.5 of the ISI Selection Document. (Note:

These CISI Reference Drawings have been prepared from construction drawings and may not in some cases show the as built configurations. Drawings discrepancies found during inspections shall be corrected as per QCNPS procedures.) Calibration Standards approved for use at QCNPS are listed in Table 2.4-4.

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

A summary of QCNPS Units 1 and 2 ASME Section XI nonexempt components and supports is included in Section 7.0.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-19 ISI Program Plan - 5th Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS UNIT 1 & COMMON UNIT 2 TITLE M-3101, SH. 1, 2 & 4 M-3111, SH. 1, 2 & 4 ISI Class 1 Main Steam System M-3102, SH. 1 & 2 M-3112, SH. 1 & 2 ISI Class 1 Reactor Feedwater System M-3103, SH. 1-5 M-3113, SH. 1-5 ISI Class 1 Reactor Recirculation System and Jet Pump Instrument M-3104, SH. 1 & 2 M-3114, SH. 1 & 2 ISI Class 1 Core Spray System M-3105, SH. 1-3 M-3115, SH. 1-3 ISI Class 1 Residual Heat Removal System M-3106, SH. 1 M-3116, SH. 1 ISI Class 1 Standby Liquid Control System M-3107, SH. 1 M-3117, SH. 1 ISI Class 1 Control Rod Drive System M-3108, SH. 1 M-3118, SH. 1 ISI Class 1 High Pressure Coolant Injection System M-3109, SH. 1 & 2 M-3119, SH. 1 & 2 ISI Class 1 Reactor Water Cleanup System M-3110, SH. 1 M-3120, SH. 1 ISI Class 1 Reactor Core Isolation Cooling System M-3121 SH. 1 M-3121, SH. 2 ISI Class 1 Reactor Vessel M-3130 SH. 1-3 M-3135 SH. 1-4 ISI Class 2 Core Spray System M-3131 SH. 1-15 M-3136 SH. 1-13 ISI Class 2 Residual Heat Removal System M-3132, SH. 1-4 M-3137, SH. 1-4 ISI Class 2 High Pressure Coolant Injection System M-3134, SH. 1 & 2 M-3139, SH. 1 & 2 ISI Class 2 Control Rod Drive System M-3140, SH. 1 M-3141, SH. 1 ISI Class 2 ECCS Ring Header and RCIC Suction Line N/A M-1042G ISI Class 2 Reactor Core Isolation Cooling System M-3143, SH. 1-6 M-3145, SH. 1-6 ISI Class 3 Residual Heat Removal System M-3144, SH. 1-6 & SH. 11 M-3144, SH. 7-10 ISI Class 3 Diesel Generator Service Water System M-3202-1 M-3219-1 System Pressure Test Walkdown Isometric Reactor Head Cavity, EL 665'-0" M-3202-2 M-3219-2 System Pressure Test Walkdown Isometric Drywell Fourth Level, EL-651'-0" M-3202-3 M-3219-3 System Pressure Test Walkdown Isometric Drywell Third Level, EL 640'-0" M-3202-4 M-3219-4 System Pressure Test Walkdown Isometric Drywell Second Level, EL 614'-0" Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-20 ISI Program Plan - 5th Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS UNIT 1 & COMMON UNIT 2 TITLE M-3202-5 M-3219-5 System Pressure Test Walkdown Isometric Drywell First Level, EL 592'-0" M-3202-6 M-3219-6 System Pressure Test Walkdown Isometric Drywell Basement, EL 579'-0" M-3202-7 M-3219-7 System Pressure Test Walkdown Isometric Lower Head CRD Area, EL 588'-0" M-3202-8 M-3219-8 System Pressure Test Walkdown Isometric Instrumentation M-3203 M-3220 System Pressure Test Walkdown Isometric Head Flange Seal Leak Detection M-3204, SH. 1 & 2 M-3221, SH. 1 & 2 System Pressure Test Walkdown Isometric Control Rod Drive Hydraulic Piping M-3205 M-3222 System Pressure Test Walkdown Isometric Standby Liquid Control Piping M-3206 M-3223 System Pressure Test Walkdown Isometric Standby Liquid Control Piping M-3207 M-3224 System Pressure Test Walkdown Isometric Standby Liquid Control Piping M-3208 --------- System Pressure Test Walkdown Isometric Diesel Generator Cooling Water Piping M-3209 M-3225 System Pressure Test Walkdown Isometric Diesel Generator Cooling Water Piping M-3210 M-3226 System Pressure Test Walkdown Isometric High Pressure Coolant Injection System M-3211, SH. 1 M-3227, SH. 1 System Pressure Test Walkdown Isometric ECCS Ring Header & RCIC Suction Line M-3211, SH. 2 M-3227, SH. 2 System Pressure Test Walkdown Isometric Core Spray Piping M-3211, SH. 3 & 4 M-3227, SH. 3 & 4 System Pressure Test Walkdown Isometric Residual Heat Removal Piping M-3211, SH. 5 M-3227, SH. 5 System Pressure Test Walkdown Isometric High Pressure Coolant Injection System M-3212 M-3228 System Pressure Test Walkdown Isometric ECCS Keepfill Pump and Piping M-3213 M-3229 System Pressure Test Walkdown Isometric Diesel Generator Service Water Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-21 ISI Program Plan - 5th Interval TABLE 2.4-1 ISI ISOMETRIC AND COMPONENT DRAWINGS UNIT 1 & COMMON UNIT 2 TITLE M-3214, SH. 1, 2, 3, 4, &5 M-3230, SH. 1, 2, 3, 4, &5 System Pressure Test Walkdown Isometric Residual Heat Removal System M-3215, SH. 1, 2, 3, 4, &5 M-3231, SH. 1, 2, 3, 4, &5 System Pressure Test Walkdown Isometric Residual Heat Removal Service Water System M-3216, SH. 1, 2 & 3 M-3232, SH. 1, 2 & 3 System Pressure Test Walkdown Isometric High Pressure Coolant Injection System M-3217 M-3233 System Pressure Test Walkdown Isometric High Pressure Coolant Injection System M-3218 M-3234 System Pressure Test Walkdown Isometric Core Spray Piping Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-22 ISI Program Plan - 5th Interval TABLE 2.4-2 UNIT 1 CISI REFERENCE DRAWINGS DRAWING TITLE NUMBER 1-CISI-1000 SH. 1 IWE Component Drawing: Primary Containment General Arrangement 1-CISI-1000 SH. 2 IWE Component Rollout: Containment Shell (Drywell) View Looking Out, 0° To 360° Azimuth 1-CISI-1000 SH. 3 IWE Component Drawing: Containment Shell (Torus), 0° To 360° Azimuth 1-CISI-1000 SH. 4 IWE Component Drawing: Containment Shell - Vent Line Connection 1-CISI-1000 SH. 5A IWE Component Detail: Drywell Head 1-CISI-1000 SH. 5B IWE Component Detail: Drywell Head 1-CISI-1000 SH. 5C IWE Component Detail: Drywell Head Penetration 1-CISI-1000 SH. 5D IWE Component Detail: Drywell Head Access Hatch Penetration X-4 1-CISI-1000 SH. 6 IWE Component Detail: Equipment Hatch Penetration X-1 1-CISI-1000 SH. 7 IWE Component Detail: CRD Removal Hatch Penetration X-6 1-CISI-1000 SH. 8 IWE Component Detail: Personnel Access Hatch Pen. X-200a & X-200b 1-CISI-1000 SH. 9A IWE Component Detail: Personnel Air Lock Penetration X-2 1-CISI-1000 SH. 9B IWE Component Detail: Personnel Air Lock Penetration X-2 1-CISI-1000 SH. 9C IWE Component Detail: Personnel Air Lock Penetration X-2 1-CISI-1000 SH. 10 IWE Component Detail: Reactor Stabilizer Support Shear Lug Access Hatch 1-CISI-1000 SH. 11A IWE Component Detail: Torus Vent System Vent Header Connection 1-CISI-1000 SH. 11B IWE Component Detail: Torus Vent System Vent Header Connection 1-CISI-1000 SH. 11C IWE Component Detail: Torus Vent System Vent Header Connection 1-CISI-1000 SH. 11D IWE Component Detail: Torus Vent System Vent Header Connection 1-CISI-1000 SH. 12 IWE Component Detail: Torus Vent System Vacuum Breaker Valve 1-CISI-1000 SH. 13 Typical IWE Component Surface And Attachment Details 1-CISI-1001 SH. 1 Piping And Instrument Penetration Details Configuration No. 1 1-CISI-1001 SH. 2 Piping And Instrument Penetration Details Configuration No. 2 1-CISI-1001 SH. 3 Piping And Instrument Penetration Details Configuration No. 3 1-CISI-1001 SH. 4 Piping And Instrument Penetration Details Configuration No. 4 1-CISI-1001 SH. 5 Piping And Instrument Penetration Details Configuration No. 5 1-CISI-1001 SH. 6 Piping And Instrument Penetration Details Configuration No. 6 1-CISI-1001 SH. 7 Piping And Instrument Penetration Details Configuration No. 7 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-23 ISI Program Plan - 5th Interval TABLE 2.4-2 UNIT 1 CISI REFERENCE DRAWINGS DRAWING TITLE NUMBER 1-CISI-1001 SH. 8 Piping And Instrument Penetration Details Configuration No. 8 1-CISI-1001 SH. 9 Piping And Instrument Penetration Details Configuration No. 9 1-CISI-1001 SH. 10 Piping And Instrument Penetration Details Configuration No. 10 1-CISI-1001 SH. 11 Piping And Instrument Penetration Details Configuration No. 11 1-CISI-1001 SH. 12 Piping And Instrument Penetration Details Configuration No. 12 1-CISI-1001 SH. 13 Piping And Instrument Penetration Details Configuration No. 13 1-CISI-1001 SH. 14 Piping And Instrument Penetration Details Configuration No. 14 1-CISI-1001 SH. 15 Piping And Instrument Penetration Details Configuration No. 15 1-CISI-1001 SH. 16 Piping And Instrument Penetration Details Configuration No. 16 1-CISI-1001 SH. 17 Piping And Instrument Penetration Details Configuration No. 17 (Drawing Cancelled) 1-CISI-1001 SH. 18 Piping And Instrument Penetration Details Configuration No. 18 1-CISI-1001 SH. 19 Piping And Instrument Penetration Details Configuration No. 19 1-CISI-1001 SH. 20 Piping And Instrument Penetration Details Configuration No. 20 1-CISI-1001 SH. 21 Piping And Instrument Penetration Details Configuration No. 21 1-CISI-1001 SH. 22 Piping And Instrument Penetration Details Configuration No. 22 1-CISI-1002 SH. 1A Electrical Penetration Details Configuration No. 1 1-CISI-1002 SH. 1B Electrical Penetration Section Configuration No. 1 1-CISI-1002 SH. 2 Electrical Penetration Details Configuration No. 2 1-CISI-1002 SH. 3 Electrical Penetration Details Configuration No. 3 1-CISI-1002 SH. 4A Electrical Penetration Details Configuration No. 4 1-CISI-1002 SH. 4B Electrical Penetration Details Configuration No. 4 1-CISI-1002 SH. 5A Electrical Penetration Details Configuration No. 5 1-CISI-1002 SH. 5B Electrical Penetration Details Configuration No. 5 1-CISI-1004 SHTS. IWE Component Information Table Piping And Instrument Penetrations 1 THROUGH 9 1-CISI-1005 SHTS. IWE Component Information Table Electrical Penetrations 1 THROUGH 2 1-CISI-1006 SHTS. IWE Component Information Table Miscellaneous Components 1 THROUGH 7 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-24 ISI Program Plan - 5th Interval TABLE 2.4-3 UNIT 2 CISI REFERENCE DRAWINGS DRAWING TITLE NUMBER 2-CISI-1000 SH. 1 IWE Component Drawing: Primary Containment General Arrangement 2-CISI-1000 SH .2 IWE Component Rollout Containment Shell (Drywell) View Looking Out - 0° To 360° Azimuth 2-CISI-1000 SH. 3 IWE Component Drawing: Containment Shell (Torus) 0° To 360° Azimuth 2-CISI-1000 SH. 4 IWE Component Drawing: Containment Shell Vent Line Connection 2-CISI-1000 SH. 5A IWE Component Detail: Drywell Head 2-CISI-1000 SH. 5B IWE Component Detail: Drywell Head 2-CISI-1000 SH. 5C IWE Component Detail: Drywell Head 2-CISI-1000 SH. 5D IWE Component Detail: Drywell Head Access Hatch Penetration X-4 2-CISI-1000 SH. 6 IWE Component Detail: Equipment Hatch Penetration X-1 2-CISI-1000 SH. 7 IWE Component Detail: CRD Removal Hatch Penetration X-6 2-CISI-1000 SH. 8 IWE Component Detail: Personnel Access Hatch Pen. X-200a & X-200b 2-CISI-1000 SH. 9A IWE Component Detail: Personnel Air Lock Penetration X-2 2-CISI-1000 SH. 9B IWE Component Detail: Personnel Air Lock Penetration X-2 2-CISI-1000 SH. 9C IWE Component Detail: Personnel Air Lock Penetration X-2 2-CISI-1000 SH. 10 IWE Component Detail: Reactor Stabilizer Support Shear Lug Access Hatch 2-CISI-1000 SH. 11A IWE Component Detail: Torus Vent System Vent Header Connection 2-CISI-1000 SH. 11B IWE Component Detail: Torus Vent System Vent Header Connection 2-CISI-1000 SH. 11C IWE Component Detail: Torus Vent System Vent Header Connection 2-CISI-1000 SH. 11D IWE Component Detail: Torus Vent System Vent Header Connection 2-CISI-1000 SH. 12 IWE Component Detail: Torus Vent System Vacuum Breaker Valve 2-CISI-1000 SH. 13 Typical IWE Component Surface And Attachment Details 2-CISI-1001 SH. 1 Piping And Instrument Penetration Details Configuration No. 1 2-CISI-1001 SH. 2 Piping And Instrument Penetration Details Configuration No. 2 2-CISI-1001 SH. 3 Piping And Instrument Penetration Details Configuration No. 3 2-CISI-1001 SH. 4 Piping And Instrument Penetration Details Configuration No. 4 2-CISI-1001 SH. 5 Piping And Instrument Penetration Details Configuration No. 5 2-CISI-1001 SH. 6 Piping And Instrument Penetration Details Configuration No. 6 2-CISI-1001 SH. 7 Piping And Instrument Penetration Details Configuration No. 7 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-25 ISI Program Plan - 5th Interval TABLE 2.4-3 UNIT 2 CISI REFERENCE DRAWINGS DRAWING TITLE NUMBER 2-CISI-1001 SH. 8 Piping And Instrument Penetration Details Configuration No. 8 2-CISI-1001 SH. 9 Piping And Instrument Penetration Details Configuration No. 9 2-CISI-1001 SH. 10 Piping And Instrument Penetration Details Configuration No. 10 2-CISI-1001 SH. 11 Piping And Instrument Penetration Details Configuration No. 11 2-CISI-1001 SH. 12 Piping And Instrument Penetration Details Configuration No. 12 2-CISI-1001 SH. 13 Piping And Instrument Penetration Details Configuration No. 13 2-CISI-1001 SH. 14 Piping And Instrument Penetration Details Configuration No. 14 2-CISI-1001 SH. 15 Piping And Instrument Penetration Details Configuration No. 15 2-CISI-1001 SH. 16 Piping And Instrument Penetration Details Configuration No. 16 2-CISI-1001 SH. 17 Piping And Instrument Penetration Details Configuration No. 17 2-CISI-1001 SH. 18 Piping And Instrument Penetration Details Configuration No. 18 2-CISI-1001 SH. 19 Piping And Instrument Penetration Details Configuration No. 19 2-CISI-1001 SH. 20 Piping And Instrument Penetration Details Configuration No. 20 2-CISI-1002 SH. 1A Electrical Penetration Details Configuration No. 1 2-CISI-1002 SH. 1B Electrical Penetration Section Configuration No. 1 2-CISI-1002 SH. 2 Electrical Penetration Details Configuration No. 2 2-CISI-1002 SH. 3 Electrical Penetration Details Configuration No. 3 2-CISI-1002 SH. 4A Electrical Penetration Details Configuration No. 4 2-CISI-1002 SH. 4B Electrical Penetration Details Configuration No. 4 2-CISI-1002 SH. 5A Electrical Penetration Details Configuration No. 5 2-CISI-1002 SH. 5B Electrical Penetration Details Configuration No. 5 2-CISI-1004 SHTS. IWE Component Information Table Piping And Instrument Penetrations 1 THROUGH 8 2-CISI-1005 SHTS. IWE Component Information Table Electrical Penetrations 1 THROUGH 2 2-CISI-1006 SHTS. IWE Component Information Table Miscellaneous Components 1 THROUGH 7 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-26 ISI Program Plan - 5th Interval TABLE 2.4-4 CALIBRATION STANDARDS MATRIX STANDARD TYPE NOTES NUMBER SAFE END-NOZZLE MOCK-UP 99902QC NONE.

SEGMENT 99903QC OVERLAY MOCK-UP SEGMENT NONE.

99905QC OVERLAY SEGMENT NONE.

99906QC OVERLAY SEGMENT NONE.

99908QC SEGMENT NONE.

99922QC SPOOL NONE.

99923QC SEGMENT NONE.

99924QC SEGMENT NONE.

99925QC SEGMENT NONE.

99926QC SEGMENT ARCHIVED - 11/26/96.

99926AQC SEGMENT REPLACED 99926QC.

99927QC STUD NONE.

99933QC SEGMENT NONE.

99937QC SPOOL NONE.

99940QC SEGMENT NONE.

99941QC BOLT NONE.

99942QC SPOOL NONE.

99943QC SEGMENT NONE.

99944QC SEGMENT NONE.

99953QC SPOOL NONE.

99955QC SPOOL NONE.

99958QC SPOOL NONE.

99958AQC SPOOL NONE.

99960QC SPOOL NONE.

99963QC SEGMENT NONE.

99964QC FLAT NONE.

99970QC SEGMENT NONE.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-27 ISI Program Plan - 5th Interval TABLE 2.4-4 CALIBRATION STANDARDS MATRIX STANDARD TYPE NOTES NUMBER 99973QC FLAT ARCHIVED - 07/14/97.

99973AQC SEGMENT REPLACED 99973QC.

99974QC FLAT NONE.

99975QC FLAT NONE.

99976QC SEGMENT ARCHIVED - 07/14/97.

99976AQC SEGMENT REPLACED 99976QC.

99978AQC FLAT VESSEL NONE.

99978BQC FLAT VESSEL NONE.

99979QC SEGMENT NONE.

99980QC SEGMENT ARCHIVED - 07/14/97.

99980AQC SEGMENT REPLACED 99980QC.

99982QC SEGMENT NONE.

99983QC SEGMENT NONE.

99988QC FLAT NONE.

99997QC BOLT NONE.

99997AQC STUD NONE.

99999QC SPOOL NONE.

99997BQC STEP STUD NONE.

99997CQC STUD NONE 99997DQC STEP STUD NONE CRDMU NOZZLE MOCK-UP NONE.

FWNUNCLAD NOZZLE MOCK-UP E308L S/S CLAD REMOVED.

FWNWCLAD NOZZLE MOCK-UP NONE.

HT3119059 FLAT VESSEL NONE.

SMAD9409SS BAR S/S NOTCH AND HOLE SIZING BLOCK.

Q89-3RD-001 SPOOL NONE.

Q89-3RD-002 SEGMENT NONE.

Q89-3RD-003 SEGMENT NONE.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-28 ISI Program Plan - 5th Interval TABLE 2.4-4 CALIBRATION STANDARDS MATRIX STANDARD TYPE NOTES NUMBER Q89-3RD-004 SEGMENT NONE.

Q89-3RD-005 SEGMENT NONE.

Q89-3RD-006 SEGMENT NONE.

USE ID AXIAL NOTCH FOR DAC SENSITIVITY Q89-3RD-007 SPOOL ON CIRC. SCANS.

Q89-3RD-009 SPOOL NONE.

Q89-3RD-011 SPOOL NONE.

Q89-3RD-012 SPOOL NONE.

Q89-3RD-013 SEGMENT NONE.

Q89-3RD-016 SPOOL NONE.

Q89-3RD-017 SPOOL NONE.

Q89-3RD-018 SPOOL DISSIMILAR ONLY, NOT FOR FERRITIC USE.

Q89-3RD-019 SEGMENT NONE.

Q89-3RD-020 SEGMENT NONE.

Q89-3RD-021 SEGMENT DISSIMILAR ONLY, NOT FOR FERRITIC USE.

Q89-3RD-022 SEGMENT NONE.

Q89-3RD-023 OVERLAY SEGMENT NONE.

Q89-3RD-024 ID CLAD MOCK-UP SEGMENT NONE.

Q89-3RD-025A STEP BLOCK NONE.

Q89-3RD-025B STEP BLOCK NONE.

USE ID AXIAL NOTCH FOR DAC SENSITIVITY Q89-3RD-026 SPOOL ON CIRC. SCANS.

Q89-3RD-027 SPOOL MAIN STEAM FORGING.

Q89-3RD-028 STEP BLOCK PDI ALTERNATIVE CALIBRATION BLOCK Q89-3RD-029 STEP BLOCK PDI ALTERNATIVE CALIBRATION BLOCK Q89-3RD-030 STEP BLOCK PDI ALTERNATIVE CALIBRATION BLOCK Q89-3RD-031 STEP BLOCK PDI ALTERNATIVE CALIBRATION BLOCK Q89-3RD-032 STEP BLOCK PDI ALTERNATIVE CALIBRATION BLOCK Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-29 ISI Program Plan - 5th Interval TABLE 2.4-4 CALIBRATION STANDARDS MATRIX STANDARD TYPE NOTES NUMBER WOR PDI ALTERNATIVE CALIBRATION QC-WOR STEP BLOCK BLOCK D89-3RD-121 SEGMENT FEED WATER NOZZLES.

D89-3RD-122 SEGMENT MAIN STEAM NOZZLES.

D89-3RD-123 SPOOL JPI & HEAD SPRAY NOZZLES.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-30 ISI Program Plan - 5th Interval 2.5 Technical Approach and Positions When the requirements of ASME Section XI are not easily interpreted, QCNPS has reviewed general licensing/regulatory requirements and industry practice to determine a practical method of implementing the Code requirements. The Technical Approach and Position (TAP) documents contained in this section have been provided to clarify QCNPSs implementation of ASME Section XI requirements. An index which summarizes each TAP is included in Table 2.5-1.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-31 ISI Program Plan - 5th Interval TABLE 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX Position Revision Status1 (Program) Description of Technical Approach Number Date2 and Position I5T-01 0 Active (SPT) System Leakage Testing of Non-Isolable Buried 9/26/12 Components.

I5T-02 0 Active (SPT) Valve Seats/Discs as Pressurization Boundaries.

9/26/12 Note 1: ISI Program Technical Approach and Position Status Options: Active - Current Technical Approach and Position is being utilized at QCNPS; Deleted - Technical Approach and Position is no longer being utilized at QCNPS.

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-32 ISI Program Plan - 5th Interval TECHNICAL APPROACH AND POSITION I5T-01 Revision 0 (Page 1 of 1)

COMPONENT IDENTIFICATION:

Code Class: 3

Reference:

IWA-5244(b)(2)

Examination Category: NA Item Number: NA

==

Description:==

System Leakage Testing of Non-Isolable Buried Components.

Component Number: Non-Isolable Buried Pressure Retaining Components CODE REQUIREMENT:

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

POSITION:

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

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

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

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 2-33 ISI Program Plan - 5th Interval TECHNICAL APPROACH AND POSITION I5T-02 Revision 0 (Page 1 of 1)

COMPONENT IDENTIFICATION:

Code Class: 1, 2, and 3

Reference:

IWA-5221, IWA-5222 Examination Category: B-P, C-H, D-B Item Number: B15.10, B15.20, C7.10, D2.10

==

Description:==

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

IWA-5221 requires the pressurization boundary for system leakage testing extend to those pressure retaining components under operating pressures during normal system service.

POSITION:

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

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

The extension of the pressurization boundary during an operational test would require an abnormal valve line-up. Extending the boundary for a hydrostatic test would require the over pressurization of low pressure piping at systems that have a high/low pressure interface (such as RHR and Core Spray).

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 3-1 ISI Program Plan - 5th Interval 3.0 COMPONENT ISI PLAN The QCNPS Component ISI Plan includes ASME Section XI nonexempt pressure retaining welds, piping structural elements, pressure retaining bolting, attachment welds, pump casings, and valve bodies of ISI Class 1, 2, and 3 components that meet the criteria of Subarticle IWA-1300. These components are identified on the P&IDs listed in Section 2.3, Table 2.3-1. Procedure ER-AA-330-002, Inservice Inspection of Section XI Welds and Components, implements the ASME Section XI Component ISI Plan.

This Component ISI Plan also includes component augmented inspection program requirements specified by documents other than ASME Section XI as referenced in Section 2.2.

The RPV interior, interior attachments, and welded core support structures are inspected in accordance with the Reactor Internals Program per Relief Request I5R-06.

3.1 Nonexempt ISI Class Components The ISI Class 1, 2, and 3 nonexempt components subject to examination are those which are not exempted under the criteria of Paragraphs IWB-1220, IWC-1220, and IWD-1220, respectively. (Note: those systems which provide reactor coolant makeup and application of the exemptions are documented in Section 3.3 below.)

A summary of QCNPS Units 1, 2, and Common ASME Section XI nonexempt components is included in Section 7.0.

3.1.1 Identification of ISI Class 1, 2, and 3 Nonexempt Components ISI Class 1, 2, and 3 nonexempt components are identified on the isometric and component drawings listed in Section 2.4, Table 2.4-1.

Welded attachments are also identified by controlled QCNPS individual support detail drawings.

3.2 Risk-Informed Examination Requirements Inspections of ASME Examination Categories B-F, B-J, C-F-1 and C-F-2 components have been exempted from ASME Section XI required inspections by Relief Request I5R-02. This Relief Request allows for the implementation of a RISI Program. Piping structural elements that fall under RISI Examination Category R-A are risk ranked as High (1, 2, and 3), Medium (4 and 5), and Low (6 and 7). Per the EPRI Topical Report TR-112657, Rev. B-A and ASME Code Case N-578-1, piping structural elements ranked as High or Medium Risk are subject to examination while piping structural elements ranked as Low Risk are not subject to examinations (except for pressure testing). Thin wall welds that were excluded from volumetric examination under ASME Section XI rules per Table IWC-2500-1 are included in the element scope that is potentially subject to RISI examination at QCNPS.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 3-2 ISI Program Plan - 5th Interval Piping structural elements may be excluded from examination (other than pressure testing) under the RISI Program if the only degradation mechanism present for a given location is inspected for cause under certain other QCNPS programs such as the Flow Accelerated Corrosion (FAC) or Intergranular Stress Corrosion Cracking (IGSCC) Programs. These piping structural elements will remain part of the assigned programs which already perform for cause inspections to detect these degradation mechanisms. Piping structural elements susceptible to FAC or IGSCC along with another degradation mechanism (e.g.,

thermal fatigue) are retained as part of the RISI scope and are included in the element selection for the purpose of performing examinations to detect the additional degradation mechanism.

3.3 Reactor Coolant Pressure Boundary Normal Makeup Calculation Exelon has determined through the criteria of Paragraph IWB-1220(a) that ISI Class 1 components which are (1) 1.57" ID and smaller for Liquid filled components or (2) 3.14" ID and smaller for Steam filled components are exempt from the volumetric and surface examinations.

The basis for determining the size of ISI Class 1 water and steam lines exempted from the volumetric and surface examination requirements of Subarticle IWB-1200 are provided in the following calculation.

Calculation No. (XCE.040.0202):

In determining the size of the liquid and steam lines exempt from surface and volumetric examination per Paragraph IWB-1220(a), liquid lines were defined as those which penetrate the RPV below the normal water level and steam lines as those which penetrate the RPV above the normal water level.

The reactor coolant makeup system at QCNPS consists of the following systems:

System Pump Maximum Emergency Flow Rate Fluid Temp. Power Safe Shutdown - 400 GPM 140o F Yes, UFSAR, Section 5.4.6.5. On-site RCIC - 400 GPM 140o F Yes, UFSAR, Section 5.4.6 On-site Water flow rates from a liquid line break are taken as 8000 lbs/sec/ft2 at 1000 psi.

Steam flow rates from a steam line are taken as 2000 lbs/sec/ft2 at 1000 psi.

Makeup water weighs 8.33 lbs per gallon at 70o F. On this basis, the exclusion diameters based on reactor coolant makeup system capacity are as follows:

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 3-3 ISI Program Plan - 5th Interval

[General Electric Boiling Water Reactor System Department, Doc No. 22A2750, pg. 7]

V 70 M 70 V 140 Dw =

17.8 Ds = 2Dw where:

Dw = exemption diameter for water in inches of inside pipe diameter.

Ds = exemption diameter for steam in inches of inside pipe diameter.

M70 = Volumetric flow rate of makeup water at 70o F in gal/min.

V70 = Specific volume of water at 70o F in ft3/lbm V140 = Specific volume of water at 140o F in ft3/lbm 0.01605 800 0.01629 Dw = = 1.57" I.D.

17.8 Ds = 2 x 1.57" = 3.14" I.D.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 3-4 ISI Program Plan - 5th Interval 3.4 Reactor Coolant Pressure Boundary Normal Makeup Calculation For Peripheral CRD Housing Welds Scope of Examination - Pressure-retaining welds in 10% of the peripheral CRD Housings (ASME Section XI Examination Category B-O, Item Number B14.10)

QCNPS has chosen not to utilize the results of Design Analysis No.

QDC-0200-M-1279, therefore, the welds in the peripheral CRD housings will not be exempted from surface and volumetric examination at this time.

Note: QCNPS Design Analysis No. QDC-0200-M-1279, demonstrates that the makeup capacity of 109 lb/sec (800 gpm) of the RCIC and SSMP systems is greater than the potential leakage of 75 lb/sec due to a weld failure in the peripheral CRD housings. This may allow welds in the peripheral CRD housings to be exempted from surface and volumetric examination due to meeting the make-up flow capacity exemption criteria of ASME Section XI Paragraph IWB-1220(a). See Reference No. 43 for calculation/justification in Section 9.0 of this document.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 4-1 ISI Program Plan - 5th Interval 4.0 SUPPORT ISI PLAN The QCNPS Support ISI Plan includes the supports of ASME Section XI nonexempt ISI Class 1, 2, and 3 components as described in Section 3.0. Procedure ER-AA-330-003, Inservice Inspection of Section XI Component Supports, implements the ASME Section XI Support ISI Plan.

4.1 Nonexempt ISI Class Supports The QCNPS ISI Class 1, 2, and 3 nonexempt supports are those which do not meet the exemption criteria of Paragraph IWF-1230. A summary of QCNPS Units 1, 2, and Common 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 supports are identified on the ISI isometric drawings listed in Section 2.4, Table 2.4-1. Supports are also identified by controlled QCNPS individual support detail drawings.

4.2 Snubber Examination and Testing Requirements 4.2.1 As allowed by 10CFR50.55a(b)(3), (b)(3)(v), and (b)(3)(v)(B), QCNPS will use Subsection ISTD, Inservice Testing of Dynamic Restraints (Snubbers) In Light Water Reactor Power Plants, ASME Operation and Maintenance of Nuclear Power Plants Code (ASME OM Code), 2004 Edition through the 2006 Addenda, to meet the visual examination, functional testing, and service life monitoring requirements for safety-related snubbers. This approach is consistent with ASME Section XI, Paragraph IWF-1220, which excludes inservice inspection of snubbers and defers to the ASME OM Code for examination, testing, and monitoring requirements. A summary of the QCNPS snubber program is included in Section 7.2.

Procedure ER-AA-330-004, Visual Examination of Snubbers, implements the visual examination of snubbers. Procedure ER-AA-330-010, Snubber Functional Testing, and ER-AA-330-011 Snubber Service Life Monitoring Program, implement the functional testing and service life monitoring of snubbers.

4.2.2 The ASME Section XI visual examination boundary of a support containing a snubber is defined in Figure IWF-1300-1(f). This boundary does not include the snubber pin-to-pin and does not include the connections to the snubber assembly (pins) per Paragraph IWF-1300(h).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 4-2 ISI Program Plan - 5th Interval This results in the remaining ASME Section XI requirements for VT-3 visual examination of the snubber attachment hardware including bolting and clamps. The ASME Section XI ISI Program uses Subsection IWF to define the inspection requirements for all ISI Class 1, 2, and 3 supports, regardless of type. The ISI Program maintains the Code Class snubbers in the support populations subject to inspection per Subsection IWF. This is done to facilitate scheduling, preparation including insulation removal, and inspection requirements of the snubber attachment hardware (e.g.,

bolting and clamps).

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

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 5-1 ISI Program Plan - 5th Interval 5.0 SYSTEM PRESSURE TESTING ISI PLAN The QCNPS System Pressure Testing ISI Plan includes pressure retaining ASME Section XI, ISI Class 1, 2, and 3 components, with the exception of those specifically exempted by Paragraphs IWA-5110(c), IWC-5222(b) and IWD-5222(b). RISI piping structural elements, regardless of risk classification, remain subject to pressure testing as part of the current ASME Section XI program.

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

Procedure ER-AA-330-001, Section XI Pressure Testing, implements the ASME Section XI System Pressure Testing ISI Plan. Additional requirements regarding the QCNPS SPT ISI Plan are discussed in Section 2.2 of this document.

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

5.1.1 Identification of ISI Class 1, 2, and 3 Components Components subject to ASME Section XI System Pressure Testing are shown on the P&IDs listed in Section 2.3, Table 2.3-1. Additional information on the classification of various system code boundaries is provided in the ISI Classification Basis Document.

5.1.2 Identification of System Pressure Tests System Pressure Test Walkdown Isometric Drawings are listed in Section 2.4, Table 2.4-1. Individual tests are identified and maintained in the QCNPS ISI Database.

All tested components are set up within the work control process and are automatically scheduled every refuel outage or by the Interval period.

Test pressures used are the systems nominal pressure range used during performance of the Operability Test. Test medium is that which is normal for the system. Exception is taken for the RHR Spray header system. A Pneumatic Test is performed beyond the normal RHR Spray water filled volume. Open-ended portions of lines are verified to have open flow Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 5-2 ISI Program Plan - 5th Interval paths by the successful system operation. Buried non-isolable components are confirmed to have un-impaired flow by the successful system operation (See Technical Approach and Position I5T-01).

5.2 Risk-Informed Examinations of Socket Welds Socket welds selected for examination under the RISI Program are to be inspected with a VT-2 visual examination each refueling outage per Relief Request I5R-02 and ASME Code Case N-578-1 (see footnote 12 in Table 1 of the ASME Code Case). To facilitate this, socket welds selected for inspection under the RISI Program are pressurized each refueling outage during a system pressure test in accordance with Paragraph IWA-5211(a).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 6-1 ISI Program Plan - 5th Interval 6.0 CONTAINMENT ISI PLAN The QCNPS Containment ISI Plan includes ASME Section XI ISI Class MC pressure retaining components and their integral attachments that meet the criteria of Subarticle IWA-1300. This Containment ISI Plan also includes information related to augmented examination areas, component accessibility, and examination review.

QCNPS has no ISI Class CC components which meet the criteria of Subarticle IWL-1100, therefore, no requirements to perform examinations in accordance with Subsection IWL are incorporated into this Containment ISI Plan.

The inspection of containment structures and components are performed per procedures ER-AA-330-007, Visual Examination of Section XI Class MC Surfaces and Class CC Liners, implements the Containment ISI Plan, ER-AA-335-004, Manual Ultrasonic Measurement of Material Thickness, and ER-AA-335-018, Detailed, General, VT-1, VT-1C, VT-3, and VT-3C, Visual Examination of ASME Class MC and CC Containment Surfaces and Components.

6.1 Nonexempt ISI Class Components The QCNPS ISI Class MC components identified on the CISI Reference Drawings are those not exempted under the criteria of Paragraph IWE-1220 in the 2001 Edition through the 2003 Addenda of ASME Section XI. A summary of QCNPS Units 1 and 2 ASME Section XI nonexempt CISI components is included in Section 7.0.

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

Components that are classified as ISI Class MC must meet the requirements of ASME Section XI in accordance with 10CFR50.55a(g)(4). Class MC Supports of Subsection IWE components are not required to be examined in accordance with 10CFR50.55a(g)(4)(v). (Note that per NUREG-1796, QCNPS will perform a VT-3 visual examination on nonexempt Class MC piping supports, which were added to the augmented inspection program in accordance with the QCNPS commitment for license renewal.) (See Section 2.2.5 of the ISI Program Plan.)

6.1.1 Identification of ISI Class MC Nonexempt Components ISI Class MC components are identified on the CISI Reference Drawings listed in Section 2.4, Tables 2.4-2 and 2.4-3 and described in Section 6.5.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 6-2 ISI Program Plan - 5th Interval 6.1.2 Identification of ISI Class MC Exempt Components Certain containment components or parts of components may be exempted from examination based on design and accessibility per the requirements of Paragraph IWE-1220.

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

6.2 Augmented Examination Areas The containment sections of the ISI Classification Basis Document discuss the containment design and components. Metal containment surface areas subject to accelerated degradation and aging require augmented examination per Examination Category E-C and Paragraph IWE-1240.

A significant condition is a condition that is identified as requiring application of additional augmented examination requirements under Paragraph IWE-1240.

In the First CISI Interval, during the QCNPS Unit 2 Outage Q2R19 Torus underwater IWE examinations, recordable indications were identified on the surface areas in the Torus Shell at Bays 3, 6, and 16. Portions of the Torus surface area near these Bays have been identified as augmented surface areas requiring examination in accordance with Paragraph IWE-1240. These surface areas have been categorized in accordance with Table IWE-2500-1, Examination Category E-C, Item Number E4.11, requiring visual examination of 100% of the surface areas identified during each inspection period until the areas examined remain essentially unchanged for the next three inspection periods. In the Second CISI Interval, augmented surface areas require visual examination of 100% of the surface areas identified during each inspection period until the areas examined remain essentially unchanged for the next inspection period. Once an augmented area remains unchanged for one full period, the areas fall back to the normal Examination Category E-A examination schedule.

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

Paragraph IWE-1231(a)(3) requires 80% of the pressure-retaining boundary that was accessible after construction to remain accessible for either direct or remote visual examination, from at least one side of the vessel, for the life of the plant.

QCNPS Calculation QDC-1600-M-1617 addresses compliance with this Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 6-3 ISI Program Plan - 5th Interval requirement by calculating the containment pressure boundary surface area that was accessible for examination at the beginning of the CISI Program and determining the limit for surface area which may be made inaccessible for the balance of plant life.

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

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

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

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

6.5 CISI Reference Drawing Descriptions The following series format was utilized for the development of these drawings:

Note: the CISI Reference Drawings have been prepared from construction drawings and may not in some cases show the as built configurations.

1(2)-CISI-1000 SERIES: IWE General Arrangement, Rollout and Component Drawings, and Component Detail Drawings

The General Arrangement drawings provide an elevation view of the major components included in the IWE program. References to other 1000 Series detail drawings are also included.

A Rollout Drawing was developed to show the entire surface of the drywell containment shell. All penetrations are included on this drawing as well as obstructions, which make portions of the liner surface inaccessible such as walls and floors, etc.

A Component Drawing was developed to show a plan view and elevation sectional view of the Torus.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 6-4 ISI Program Plan - 5th Interval

Component Detail Drawings were developed for the following IWE Components:

- Drywell Head and Drywell Head Penetrations

- Containment Vent Line Connections

- Drywell Head Access Hatch

- CRD Removal Hatch

- Personnel Access Hatch

- Personnel Air Lock

- Equipment Hatch

- Reactor Stabilizer Support Shear Lug Access Hatch

- Torus Vent System Vent Header Piping

- Torus Vent System Vacuum Breaker Piping

- Typical IWE Component Surface and Attachment Details 1(2)-CISI-1001 SERIES: Piping and Instrument Penetration Detail Drawings The Piping and Instrument Penetration Detail Drawings provide the IWE examination boundaries for each penetration configuration, and also identify and detail any moisture barriers or bolting subject to examination.

1(2)-CISI-1002 SERIES: Electrical Penetration Detail Drawings The Electrical Penetration Detail Drawings provide the IWE examination boundaries for each penetration configuration, and also identify and detail any moisture barriers or bolting subject to examination.

1(2)-CISI-1004 SERIES: IWE Component Information Tables-Piping Penetrations The IWE Component Information Tables-Piping Penetrations are a summary of all piping penetrations subject to examination per Subsection IWE. These tables provide a cross-reference of penetration number to CISI Reference Drawings, P&ID No, associated piping line number, location, penetration sleeve detail drawing reference, penetration head detail drawing reference, and penetration bolting subject to examination. The left hand column of the tables lists all surfaces subject to examination. The right hand columns under the heading IWE COMPONENTS list all pressure retaining bolting subject to examination.

1(2)-CISI-1005 SERIES: IWE Component Information Tables-Electrical Penetrations The IWE Component Information Tables-Electrical Penetrations are a summary of all electrical penetrations subject to examination per Subsection IWE. These tables provide a cross-reference of penetration number to CISI Reference Drawings, service description, location, penetration detail drawing number, and Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 6-5 ISI Program Plan - 5th Interval penetration bolting subject to examination. The left hand column of the tables lists all surfaces subject to examination; the right hand columns under the heading IWE COMPONENTS list all pressure retaining bolting subject to examination.

1(2)-CISI-1006 SERIES: IWE Component Information Tables-Miscellaneous Components The IWE Component Information Tables-Miscellaneous Components are a summary of all miscellaneous components (MC piping, hatches, airlocks, containment shell and vent system surfaces, etc.) subject to examination per Subsection IWE. These tables provide a cross-reference of component to CISI Reference Drawings, component description, P&ID No, associated piping line number, location, and associated IWE components: bolting, moisture barriers, bellows and MC piping subject to examination. The left hand column of the tables lists all surfaces subject to examination, and the right hand columns under the heading ASSOCIATED IWE COMPONENTS lists all moisture barriers and pressure retaining bolting subject to examination.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-1 ISI Program Plan - 5th Interval 7.0 COMPONENT

SUMMARY

TABLES 7.1 Inservice Inspection Summary Tables The following Tables 7.1-1 and 7.1-2 provide a summary of the ASME Section XI pressure retaining components, supports, containment structures, system pressure testing, and augmented inspection program components for the Fifth ISI Interval and the Second CISI Interval at QCNPS Units 1, 2, and Common.

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

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

(1) (2) (3) (4) (5) (6) (7)

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

Provides the examination category and description as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1. Only those examination categories applicable to QCNPS are identified.

Examination Category R-A from ASME Code Case N-578-1 is used in lieu of ASME Section XI Examination Categories B-F, B-J, C-F-1, and C-F-2 to identify ISI Class 1 and 2 piping structural elements for the RISI Program.

Examination Category NA is used to identify augmented inspection programs and other QCNPS requirements.

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

Provides the item number as identified in ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1. Only those item numbers applicable to QCNPS are identified.

For piping structural elements under the RISI Program, the Risk Category Number (e.g., 1 through 5) is used in place of the Item Number.

Specific abbreviations such as BWROG, BWRVIP, IGSCC, 0737, 1796.27, and 1796.30 have been developed to identify augmented inspection programs and other QCNPS requirements.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-2 ISI Program Plan - 5th Interval (3) Item Number (or Risk Category or Augmented)

Description:

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

For Risk-Informed piping structural elements, a statement of the Risk Category is provided.

For augmented inspection programs, a description of the augmented basis is provided.

(4) Examination Requirements:

Provides the examination methods required by ASME Section XI, Tables IWB-2500-1, IWC-2500-1, IWD-2500-1, IWE-2500-1, and IWF-2500-1.

Provides the examination requirements for piping structural elements under the RISI Program that are in accordance with the EPRI Topical Report TR-112657, Rev. B-A and ASME Code Case N-578-1.

Provides the examination requirements for augmented inspection program components.

(5) Total Number Of Components by System:

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

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

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

(6) Approved Relief Request/TAP Number:

Provides a listing of Approved Relief Request/TAP Numbers applicable to specific components, the ASME Section XI Item Number, Risk Category Number or Augmented Number. Relief Requests and TAP Numbers that generically apply to all components, or an entire class are not listed. If a Relief Request/

TAP Number is identified, see the corresponding relief request in Section 8.0 or the TAP Number in Section 2.5.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-3 ISI Program Plan - 5th Interval (7) Notes:

Provides a listing of program notes applicable to the ASME Section XI Item Number, Risk Category Number, or Augmented Number. If a program note number is identified, see the corresponding program note at the end of the Table 7.1-3.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-4 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-A B1.11 Circumferential Shell Welds (Reactor Vessel) Volumetric RPV: 4 I5R-05 Pressure Retaining B1.12 Longitudinal Shell Welds (Reactor Vessel) Volumetric RPV: 15 I5R-05 Welds in Reactor Vessel B1.21 Circumferential Head Welds (Reactor Vessel) Volumetric RPV: 3 B1.22 Meridional Head Welds (Reactor Vessel) Volumetric RPV: 16 B1.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric RPV: 1 B1.40 Head-to-Flange Weld (Reactor Vessel) Volumetric & RPV: 1 Surface B1.51 Beltline Region Repair Weld (Reactor Vessel) Volumetric RPV: 5 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 29 14 Full Penetration Welds B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 29 I5R-01 of Nozzles in Vessels Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-5 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-1 RPV: 1 (92 Nuts)

Pressure Retaining B6.20 Closure Studs (Reactor Vessel) Volumetric RPV: 1 (92 Studs) 9 Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 1 (92 Threads) 2 in. In Diameter B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-1 RPV: 2 (92 Washers, Bushings)

B6.180 Bolts and Studs (Pumps) Volumetric RR: 2 (32 Bolts, Studs)

B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-1 RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-1 RR: 2 (32 Nuts, Bushings, Washers)

B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-1 MS: 12 Pressure Retaining RPV: 3 Bolting, 2 in. and Less RR: 2 In Diameter RWCU: 1 B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-1 CSA: 2 CSB: 2 MS: 21 RHRA: 1 RHRB: 1 RR: 4 RWCU: 1 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-6 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-K B10.10 Welded Attachments (Pressure Vessels) Surface or RPV: 9 Welded Attachments for Volumetric Vessels, Piping, Pumps, B10.20 Welded Attachments (Piping) Surface CSA: 2 and Valves CSB: 1 FWA: 6 FWB: 6 HPCI: 2 MS: 18 RHRA: 4 RHRB: 4 RR: 9 RWCU: 2 SDC: 1 B10.30 Welded Attachments (Pumps) Surface RR: 6 B10.40 Welded Attachments (Valves) Surface RR: 4 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-7 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 RR: 2 Pump Casings B-M-2 B12.50 Valve Bodies (NPS 4 or Larger) (Valves) Visual, VT-3 CSA: 3 Valve Bodies CSB: 3 FWA: 3 FWB: 3 HPCI: 2 MS: 21 RHRA: 3 RHRB: 3 RR: 6 RWCU: 3 SDC: 2 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-8 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-N-1 B13.10 Vessel Interior (Reactor Vessel) Visual, VT-3 RPV: 1 I5R-06 15 Interior of Reactor Vessel B-N-2 B13.20 Interior Attachments Within Beltline Region (Reactor Vessel) Visual, VT-1 RPV: 26 I5R-06 15 Welded Core Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region (Reactor Visual, VT-3 RPV: 40 I5R-06 15 Attachments to Vessel)

Reactor Vessels B13.40 Core Support Structure (Reactor Vessel) Visual, VT-3 RPV: 1 I5R-06 15 B-O B14.10 Welds in CRD Housing (Reactor Vessel) Volumetric or RPV: 32 13 Pressure Retaining Welds in (10% of Peripheral CRD Housings to be inspected. 32 of Surface (32 CRD Housings Control Rod Housings the 177 CRD Housings are identified as peripheral) with 2 Welds Each)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-9 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-P B15.10 Pressure Retaining Components [IWB-5222(a)] Visual, VT-2 CRD I5T-02 All Pressure System Leakage Test (IWB-5220) CS Retaining Components FW (Periodic) HPCI MS RCIC RHR RPV RR RWCU SBLC B-P B15.20 Pressure Retaining Components [IWB-5222(b)] Visual, VT-2 CRD I5T-02 All Pressure System Leakage Test (IWB-5220) CS Retaining Components FW (Interval) HPCI MS RCIC RHR RPV RR RWCU SBLC Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-10 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number C-A C1.10 Shell Circumferential Welds (Pressure Vessels) Volumetric RHRA: 3 Pressure Retaining Welds RHRB: 3 in Pressure Vessels C1.20 Head Circumferential Welds (Pressure Vessels) Volumetric RHRA: 1 RHRB: 1 C-B C2.31 Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface ECCS: 4 Pressure Retaining with Reinforcing Plates in Vessels, Greater than 1/2" RHRA: 4 Nozzle Welds in Nominal Thickness (Pressure Vessels) RHRB: 4 Vessels C2.33 Nozzle-to-Shell (or Head or Nozzle) Welds with Visual, VT-2 ECCS: 4 Reinforcing Plates when Inside of Vessel is Inaccessible RHRA: 2 for Vessels, Greater than 1/2" Nominal Thickness RHRB: 2 (Pressure Vessels)

C-C C3.10 Welded Attachments (Pressure Vessels) Surface RHRA: 4 Welded Attachments RHRB: 4 for Vessels, Piping, C3.20 Welded Attachments (Piping) Surface CRD: 2 Pumps, and Valve CSA: 6 CSB: 8 ECCS: 6 HPCI: 10 RHR: 1 RHRA: 19 RHRB: 15 C3.30 Welded Attachments (Pumps) Surface CSA: 1 CSB: 1 HPCI: 1 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-11 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number C-H C7.10 Pressure Retaining Components Visual, VT-2 CRD I5R-03 All Pressure System Leakage Test (IWC-5220) CS I5R-04 Retaining Components ECCS I5T-02 (Periodic) FW HPCI RCIC RHR RPV Head Flange RR SBLC Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-12 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number D-A D1.20 Welded Attachments (Piping) Visual, VT-1 DGCW: 2+2 2 Welded Attachments for RHRSW: 30+1 Vessels, Piping, Pumps, and Valves D-B D2.10 Pressure Retaining Components Visual, VT-2 DGCW I5T-01 All Pressure System Leakage Test (IWD-5220) HVAC I5T-02 Retaining Components PS (Periodic) RHRSW Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-13 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) TAP Number E-A E1.11 Containment Vessel Pressure Retaining Boundary - General Visual 231 Containment Accessible Surface Areas Surfaces E1.11 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 54 10 Bolted Connections, Surfaces E1.12 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 16 11 Wetted Surfaces of Submerged Areas E1.20 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 121 11 BWR Vent System Accessible Surface Areas E1.30 Moisture Barriers General Visual 6 E-C E4.11 Containment Surface Areas - Visible Surfaces Visual, VT-1 0 12 Containment Surfaces Requiring E4.12 Containment Surface Areas; Surface Area Grid Volumetric 0 Augmented Examination Minimum Wall Thickness Location (UT Thickness)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-14 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number F-A F1.10 Class 1 Piping Supports Visual, VT-3 CSA: 5 1 Supports CSB: 5 FWA: 7 FWB: 7 HPCI: 5 MS: 40 RHRA: 6 RHRB: 6 RR: 37 RWCU: 14 SDC: 6 F1.20 Class 2 Piping Supports Visual, VT-3 CRD: 24 1 CSA: 13 CSB: 22 ECCS: 30 FWB: 1 HPCI: 51 RHR: 13 RHRA: 36 RHRB: 39 F1.30 Class 3 Piping Supports Visual, VT-3 DGCW: 96+63 1 RHRSW: 126+4 2 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-15 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number F-A F1.40 Supports Other Than Piping Supports Visual, VT-3 CSA: 1 1 Supports (Class 1, 2, 3, and MC) CSB: 1 2 (Continued) DGCW: 1+1 HPCI: 2 JPI: 2 RHRA: 8 RHRB: 8 RHRSW: 8 RPV: 9 RR: 12 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-16 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

TABLE Examination Category Risk Description Exam Total Number of Approved Notes (with Examination Category Category Requirements Components by Relief Request/

Description) Number System TAP Number R-A 1 Risk Category 1 Elements See Notes FWA: 12 I5R-02 3 Risk-Informed Piping FWB: 8 5 Examinations 2 Risk Category 2 Elements See Notes CSA: 12 I5R-02 3 CSB: 9 5 RHR: 4 RHRA: 68 RHRB: 73 SDC: 5 3 Risk Category 3 Elements See Notes FWA: 1 I5R-02 3 5

4 Risk Category 4 Elements See Notes CSA: 11 I5R-02 3 CSB: 11 5 ECCS: 54 HPCI: 37 MS: 146 RCIC: 8 RHRA: 33 RHRB: 42 RPV: 2 RR: 36 RWCU: 26 5 Risk Category 5 Elements See Notes HPCI: 18 I5R-02 3 RHR: 19 5 RHRA: 9 RHRB: 8 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-17 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 1 & COMMON INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number NA BWROG BWROG, BWR Feedwater Nozzle and Control Rod Drive Volumetric RPV: 16 Augmented Return Line Nozzle Cracking Components Components BWRVIP IGSCC Management Program BWR Vessel Internals and Various In accordance with Piping Components (GE SILs and RICSILs) BWRVIP Program IGSCC Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category C: 123 4 Austenitic Stainless Steel Piping Components, TR-113932, Category D: 4 6 BWR Vessel and Internals Project, Technical Basis for Category E: 43 Revisions to Generic Letter 88-01 Inspection Schedules Category F: 1 (BWRVIP-75), and TR-1012621, BWR Vessel and Category G: 1 7 Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A) 0737 Leak testing and periodic visual examinations of systems Leakage Test CS, HPCI, outside of primary containment which could contain highly MS, RCIC, radioactive fluids during a serious transient or accident RHR, RR (NUREG-0737) RWCU 1796.27 Class MC Component Supports Visual, VT-3 BS: 1 16 (NUREG-1796, Section B.1.27) DW: 3 SC: 6 TV: 1 1796.30 Nonexempt Class MC Piping Supports Visual, VT-3 PS: 23 17 (NUREG-1796, Section B.1.30) RBCCW: 3 1801 Cast Austenitic Stainless Steel Components (NUREG-1801) Enhanced In accordance with 18 Visual, EVT-1 BWRVIP Program Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-18 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-A B1.11 Circumferential Shell Welds (Reactor Vessel) Volumetric RPV: 4 I5R-05 Pressure Retaining B1.12 Longitudinal Shell Welds (Reactor Vessel) Volumetric RPV: 13 I5R-05 Welds in Reactor Vessel B1.21 Circumferential Head Welds (Reactor Vessel) Volumetric RPV: 3 B1.22 Meridional Head Welds (Reactor Vessel) Volumetric RPV: 16 B1.30 Shell-to-Flange Weld (Reactor Vessel) Volumetric RPV: 1 B1.40 Head-to-Flange Weld (Reactor Vessel) Volumetric & RPV: 1 19 Surface B1.51 Beltline Region Repair Weld (Reactor Vessel) Volumetric RPV: 1 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 29 14 Full Penetration Welds B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 29 I5R-01 of Nozzles in Vessels Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-19 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-1 RPV: 1 (92 Nuts)

Pressure Retaining B6.20 Closure Studs (Reactor Vessel) Volumetric RPV: 1 (92Studs) 9 Bolting, Greater Than B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 1 (92Threads) 2 in. in Diameter B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-1 RPV: 2 (92Washers, Bushings)

B6.180 Bolts & Studs (Pumps) Volumetric RR: 2 (32 Bolts, Studs)

B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-1 RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-1 RR: 2 (32 Nuts, Bushings, Washers)

B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-1 MS: 12 Pressure Retaining RPV: 3 Bolting, 2 in. and Less RR: 2 In Diameter RWCU: 1 B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-1 CSA: 2 CSB: 2 MS: 21 RHRA: 1 RHRB: 1 RR: 4 RWCU: 1 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-20 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-K B10.10 Welded Attachments (Pressure Vessels) Surface or RPV: 9 Welded Attachments for Volumetric Vessels, Piping, Pumps, B10.20 Welded Attachments (Piping) Surface CSA: 2 and Valves CSB: 2 FWA: 5 FWB: 5 HPCI: 2 MS: 16 RHRA: 3 RHRB: 4 RR: 4 SDC: 1 B10.30 Welded Attachments (Pumps) Surface RR: 6 B10.40 Welded Attachments (Valves) Surface RR: 4 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-21 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-L-2 B12.20 Pump Casings (Pumps) Visual, VT-3 RR: 2 Pump Casings B-M-2 B12.50 Valve Bodies, (NPS 4 or Larger) (Valves) Visual, VT-3 CSA: 3 Valve Bodies CSB: 3 FWA: 3 FWB: 3 HPCI: 2 MS: 21 RHRA: 3 RHRB: 3 RR: 6 RWCU: 3 SDC: 2 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-22 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-N-1 B13.10 Vessel Interior (Reactor Vessel) Visual, VT-3 RPV: 1 I5R-06 15 Interior of Reactor Vessel B-N-2 B13.20 Interior Attachments Within Beltline Region (Reactor Vessel) Visual, VT-1 RPV: 26 I5R-06 15 Welded Core Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region (Reactor Visual, VT-3 RPV: 40 I5R-06 15 Attachments to Vessel)

Reactor Vessels B13.40 Core Support Structure (Reactor Vessel) Visual, VT-3 RPV: 1 I5R-06 15 B-O B14.10 Welds in CRD Housing (Reactor Vessel) Volumetric or RPV: 32 13 Pressure Retaining Welds in (10% of Peripheral CRD Housings to be inspected. 32 of Surface (32 CRD Housings Control Rod Housings the 177 CRD Housings are identified as peripheral) with 2 Welds Each)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-23 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number B-P B15.10 Pressure Retaining Components [IWB-5222(a)] Visual, VT-2 CRD I5T-02 All Pressure System Leakage Test (IWB-5220) CS Retaining Components FW (Periodic) HPCI MS RCIC RHR RPV RR RWCU SBLC B-P B15.20 Pressure Retaining Components [IWB-5222(b)] Visual, VT-2 CRD I5T-02 All Pressure System Leakage Test (IWB-5220) CS Retaining Components FW (Interval) HPCI MS RCIC RHR RPV RR RWCU SBLC Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-24 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number C-A C1.10 Shell Circumferential Welds (Pressure Vessels) Volumetric RHRA: 3 Pressure Retaining Welds RHRB: 3 in Pressure Vessels C1.20 Head Circumferential Welds (Pressure Vessels) Volumetric RHRA: 1 RHRB: 1 C-B C2.31 Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface ECCS: 4 Pressure Retaining with Reinforcing Plates in Vessels, Greater than 1/2 RHRA: 4 Nozzle Welds Nominal Thickness (Pressure Vessels) RHRB: 4 in Vessels C2.33 Nozzle-to-Shell (or Head or Nozzle) Welds with Visual, VT-2 ECCS: 4 Reinforcing Plates when Inside of Vessel is Inaccessible RHRA: 2 for Vessels, Greater than 1/2 Nominal Thickness RHRB: 2 (Pressure Vessels)

C-C C3.10 Welded Attachments (Pressure Vessels) Surface RHRA: 4 Welded Attachments RHRB: 4 for Vessels, Piping, C3.20 Welded Attachments (Piping) Surface CRD: 2 Pumps, and Valves CSA: 6 CSB: 8 ECCS: 6 HPCI: 9 RHR: 3 RHRA: 14 RHRB: 18 C3.30 Welded Attachments (Pumps) Surface CSA: 1 CSB: 1 HPCI: 1 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-25 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number C-H C7.10 Pressure Retaining Components Visual, VT-2 CRD I5R-03 All Pressure System Leakage Test (IWC-5220) CS I5R-04 Retaining Components ECCS I5T-02 (Periodic) FW HPCI RCIC RHR RPV Head Flange RR SBLC Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-26 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number D-A D1.20 Welded Attachments (Piping) Visual, VT-1 DGCW: 3 Welded Attachments for RHRSW: 27 Vessels, Piping, Pumps, and Valves D-B D2.10 Pressure Retaining Components Visual, VT-2 DGCW I5T-01 All Pressure System Leakage Test (IWD-5220) HVAC I5T-02 Retaining Components PS (Periodic) RHRSW Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-27 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) TAP Number E-A E1.11 Containment Vessel Pressure Retaining Boundary - General Visual 232 Containment Accessible Surface Areas Surfaces E1.11 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 55 10 Bolted Connections, Surfaces E1.12 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 16 11 Wetted Surface and Submerged Areas E1.20 Containment Vessel Pressure Retaining Boundary - Visual, VT-3 121 11 BWR Vent System Accessible Surface Areas E1.30 Moisture Barriers General Visual 6 E-C E4.11 Containment Surface Areas - Visible Surfaces Visual, VT-1 3 12 Containment Surfaces Requiring E4.12 Containment Surface Areas; Surface Area Grid Volumetric 0 Augmented Examination Minimum Wall Thickness Location (UT Thickness)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-28 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number F-A F1.10 Class 1 Piping Supports Visual, VT-3 CSA: 5 1 Supports CSB: 6 FWA: 7 FWB: 6 HPCI: 7 MS: 37 RHRA: 5 RHRB: 6 RR: 27 RWCU: 13 SDC: 5 F1.20 Class 2 Piping Supports Visual, VT-3 CRD: 24 1 CSA: 18 CSB: 27 ECCS: 30 FWB: 1 HPCI: 42 RHR: 15 RHRA: 31 RHRB: 39 F1.30 Class 3 Piping Supports Visual, VT-3 DGCW: 111 1 RHRSW: 115 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-29 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number F-A F1.40 Supports Other Than Piping Supports Visual, VT-3 CSA: 1 1 Supports (Class 1, 2, 3, and MC) CSB: 1 (Continued) DGCW: 1 HPCI: 2 JPI: 2 RHRA: 10 RHRB: 10 RHRSW: 8 RPV: 9 RR: 12 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-30 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number R-A 1 Risk Category 1 Elements See Notes FWA: 12 I5R-02 3 Risk-Informed Piping FWB: 8 5 Examinations 2 Risk Category 2 Elements See Notes CSA: 9 I5R-02 3 CSB: 10 5 RHR: 4 RHRA: 60 RHRB: 79 SDC: 5 3 Risk Category 3 Elements See Notes FWA: 1 I5R-02 3 5

4 Risk Category 4 Elements See Notes CSA: 11 I5R-02 3 CSB: 12 5 ECCS: 55 HPCI: 20 MS: 142 RCIC: 7 RHRA: 36 RHRB: 32 RPV: 2 RR: 37 RWCU: 28 5 Risk Category 5 Elements See Notes HPCI: 16 I5R-02 3 RHR: 19 5 RHRA: 10 RHRB: 12 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-31 ISI Program Plan - 5th Interval TABLE 7.1 UNIT 2 INSERVICE INSPECTION

SUMMARY

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

Description) System TAP Number NA BWROG BWROG, BWR Feedwater Nozzle and Control Rod Drive Volumetric RPV: 16 Augmented Return Line Nozzle Cracking Components Components BWRVIP IGSCC Management Program BWR Vessel Internals and Various In accordance with Piping Components (GE SILs and RICSILs) BWRVIP Program IGSCC Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric Category C: 129 4 Austenitic Stainless Steel Piping Components, TR-113932, Category D: 1 8 BWR Vessel and Internals Project, Technical Basis for Category D/E: 1 Revisions to Generic Letter 88-01 Inspection Schedules Category E: 39 (BWRVIP-75), and TR-1012621, BWR Vessel and Category F/E: 1 Internals Project, Technical Basis for Revisions to Generic Category G: 1 7 Letter 88-01 Inspection Schedules (BWRVIP-75-A) 0737 Leak testing and periodic visual examinations of systems Leakage Test CS, HPCI, outside of primary containment which could contain highly MS, RCIC, radioactive fluids during a serious transient or accident RHR, RR (NUREG-0737) RWCU 1796.27 Class MC Component Supports Visual, VT-3 BS: 1 16 (NUREG-1796, Section B.1.27) DW: 3 SC: 6 TV: 1 1796.30 Nonexempt Class MC Piping Supports Visual, VT-3 PS: 31 17 (NUREG-1796, Section B.1.30) RBCCW: 6 1801 Cast Austenitic Stainless Steel Components (NUREG-1801) Enhanced In accordance with 18 Visual, EVT-1 BWRVIP Program Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-32 ISI Program Plan - 5th Interval TABLE 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary Snubber visual examinations, functional testing, and service life monitoring are performed in accordance with the ASME OM Code, Subsection ISTD. For a 1

detailed discussion of the QCNPS Snubber Program, refer to Section 4.2 of this document.

The Unit 1 population counts include those components that are common to both units (typically designated as 1/2) and are listed in Table following a +

2 symbol.

For the Fifth ISI Interval, QCNPSs ISI Class 1 and 2 piping inspection program will be governed by risk-informed regulations. The RISI Program methodology is described in the EPRI Topical Report TR-112657, Rev. B-A and ASME Code Case N-578-1. The RISI Program scope has been implemented as an alternative to the 3

2007 Edition through the 2008 Addenda of the ASME Section XI Code examination program for ISI Class 1 B-F and B-J welds and ISI Class 2 C-F-1 and C-F-2 welds in accordance with 10CFR50.55a(a)(3)(i).

Per the EPRI Topical Report TR-112657, Rev. B-A and ASME Code Case N-578-1, welds within the plant that are assigned to IGSCC Categories B through 4

G will continue to meet existing IGSCC schedules, while IGSCC Category A welds have been subsumed into the RISI Program.

Examination requirements within the RISI Program are determined by the various degradation mechanisms present at each individual piping structural 5

element. See EPRI TR-112657, Rev. B-A and ASME Code Case N-578-1 for specific examination method requirements.

Weld 02BS-F4 was overlaid during Q1R18. The IGSCC Category will change from F to E and be placed into the IGSCC weld population following 6

inspection within three (3) outages (no later than following Q1R21 in 2011).

7 Welds are 100% obstructed.

Welds 02B-S7 and 02BD-F9 were overlaid during Q2R17. The IGSCC weld population will increase by two (2) to 41 following inservice examinations 8

within three (3) outages (no later than Q2R20 in 2010).

9 Examination Category B-G-1, Item Numbers B6.20 Closure Studs, In Place and B6.30 Closure Studs, When Removed have been combined into and renamed as Item Number B6.20 Closure Studs in Table IWB-2500-1 of ASME Section XI, 2007 Edition through the 2008 Addenda.

10 Bolted connections examined per Item Number E1.11 require a General Visual examination each period and a VT-3 visual examination once per interval and each time the connection is disassembled during a scheduled Item Number E1.11 examination. Additionally, a VT-1 visual examination shall be performed if degradation or flaws are identified during the VT-3 visual examination. These modifications are required by 10CFR50.55a(b)(2)(ix)(G) and 10CFR50.55a(b)(2)(ix)(H).

11 Item Numbers E1.12 and E1.20 require VT-3 visual examination in lieu of General Visual examination, as modified by 10CFR50.55a(b)(2)(ix)(G).

12 Item Number E4.11 requires VT-1 visual examination in lieu of Detailed Visual examination, as modified by 10CFR50.55a(b)(2)(ix)(G).

13 Examination Category B-O (Pressure-Retaining Welds In Control Rod Housings), Item Number B14.10 (Welds in CRD Housing) - the scope of examination is for pressure retaining welds in 10% of the peripheral CRD Housings. A total of 32 out of the 177 CRD Housings are classified as peripheral components.

QCNPS has selected the welds on 4 CRD Housings (two welds per housing) to be examined during the interval (10% of 32).

14 As allowed by ASME Code Case N-613-1, QCNPS will perform a volumetric examination using a reduced examination volume (A-B-C-D-E-F-G-H) of Figures 1, 2, and 3 of the Code Case in lieu of the previous examination volumes of ASME Section XI, Figures IWB-2500-7(a), (b), and (c).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-33 ISI Program Plan - 5th Interval TABLE 7.1-3 INSERVICE INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 15 The RPV interior requires examination per the BWRVIP in lieu of ASME Section XI Examination Categories B-N-1 and B-N-2 per Relief Request I5R-06.

Augmented inspection programs associated with the BWRVIP and the QCNPS Vessel Internals Program are maintained independently. The QCNPS BWRVIP procedure ER-AB-331, BWR Internals Program Management, describes these requirements and the BWRVIP in more detail.

16 Per NUREG-1796, QCNPS will perform a VT-3 visual examination on Class MC component supports. These component supports were added to the augmented inspection program in accordance with the QCNPS commitment for license renewal. (NUREG-1796, Section B.1.27) 17 Per NUREG-1796, QCNPS will perform a VT-3 visual examination on nonexempt Class MC piping supports. These piping supports were added to the augmented inspection program in accordance with the QCNPS commitment for license renewal. (NUREG-1796, Section B.1.30) 18 Per NUREG-1801, QCNPS will perform a EVT-1 visual examination on a sample of the identified CASS components every refueling outage when accessible due to Control Rod Blade (CRB) exchange. If there is no CRB exchange, than at least one fuel support piece will be inspected every other refueling outage.

This requirement was added in response to the License Renewal Requirement and implemented starting in Fall 2011. (Ref. NUREG-1801 and SVP-10-080.)

Note: the components and schedule of examinations is maintained in the QCNPS Reactor Internals Program Documents.

19 License Renewal AT 00101562-49 (B.1.1). Perform QCNPS Unit 2 Reactor Head Crack Inspection. This assignment tracks the License Renewal commitment to perform one additional UT Examination of the QCNPS Unit 2 Reactor Head cracks. The inspection is to be completed in 2018 (plus or minus 2 years).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-34 ISI Program Plan - 5th Interval 7.2 Snubber Inspection Summary Tables 10CFR50.55a Codes and Standards allows usage of ASME OM Code, Subsection ISTD, using VT-3 visual examination methods described in Paragraph IWA-2213.

The following Tables 7.2-1 and 7.2-2 provide a summary of the ASME OM Code, Subsection ISTD, Snubber visual examinations and functional testing for the Fifth ISI Interval at QCNPS Units 1, 2, and Common.

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

ASME OM Code Approved Subsection OM Article Article Number Exam Totals Frequency Relief Request/ Notes (with Subsection Number Description Requirements TAP Number Description)

(1) (2) (3) (4) (5) (6) (7) (8)

(1) ASME OM Code Subsection:

Provides the applicable ASME OM Code subsection number and a description as obtained from Subsection ISTD. Only applicable subsections to QCNPS are identified.

(2) OM Article Number:

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

(3) Article Number

Description:

Provides the article description as identified in Subsection ISTD.

Identifies the methods selected to be performed at QCNPS.

(4) Examination Requirements:

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

(5) Totals:

Provides the total number of snubbers that pertain to that article of Subsection ISTD. Note that the total number of snubbers are subject to change after completion of plant modifications and design changes.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-35 ISI Program Plan - 5th Interval (6) Frequency:

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

(7) Approved Relief Request/TAP Number:

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

(8) Notes:

Provides a listing of program notes applicable to the Subsection ISTD article number. If a program note number is identified, see the corresponding program note in Table 7.2-3.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-36 ISI Program Plan - 5th Interval TABLE 7.2 UNIT 1 SNUBBER INSPECTION

SUMMARY

TABLE ASME OM Code Approved Subsection OM Article Exam Article Number Description Totals Frequency Relief Request/ Notes (with Subsection Number Requirements TAP Number Description)

ISTD ISTD-4200 Accessible and Inaccessible Snubbers (1 population) Visual, VT-3 118 Various 1 Snubber 3 Examinations ISTD ISTD-5200 10% Functional Test Plan - Functional 34 Every Outage 2 Snubber Type 1 Snubbers (LISEGA 30 Series-1, 4, 5, & 6) Testing Testing 10% Functional Test Plan - Functional 40 Every Outage 2 Type 2 Snubbers (PSA-35, PSA-100) Testing 10% Functional Test Plan - Functional 44 Every Outage 2 Type 3 Snubbers (PSA-1, PSA-3, PSA-10) Testing Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-37 ISI Program Plan - 5th Interval TABLE 7.2 UNIT 2 SNUBBER INSPECTION

SUMMARY

TABLE ASME OM Code Approved Subsection OM Article Exam Article Number Description Totals Frequency Relief Request/ Notes (with Subsection Number Requirements TAP Number Description)

ISTD ISTD-4200 Accessible and Inaccessible Snubbers (1 population) Visual, VT-3 121 Various 1 Snubber 3 Examinations ISTD ISTD-5200 10% Functional Test Plan - Functional 41 Every Outage 2 Snubber Type 1 Snubbers (LISEGA 30 Series-1, 4, 5, & 6) Testing Testing 10% Functional Test Plan - Functional 34 Every Outage 2 Type 2 Snubbers (PSA-35, PSA-100) Testing 10% Functional Test Plan - Functional 46 Every Outage 2 Type 3 Snubbers (PSA-1, PSA-3, PSA-10) Testing Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 7-38 ISI Program Plan - 5th Interval TABLE 7.2-3 SNUBBER INSPECTION

SUMMARY

TABLE PROGRAM NOTES Note # Note Summary 1 Examinations performed per ASME Code Case OMN-13, Requirements for Extending Snubber Inservice Visual Examination Interval at LWR Power Plants.

2 Per the ASME OM Code, Subsection ISTD, 2004 Edition through the 2006 Addenda, Paragraph ISTD-5240 Test Frequency.

3 Per the ASME OM Code, Subsection ISTD, 2004 Edition through the 2006 Addenda, Paragraph ISTD-4250 Inservice Examination Intervals.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 8-1 ISI Program Plan - 5th Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests written per 10CFR50.55a(a)(3)(i) for situations where alternatives to ASME Section XI requirements provide an acceptable level of quality and safety; per 10CFR50.55a(a)(3)(ii) for situations where compliance with ASME Section XI requirements results in a hardship or an unusual difficulty without a compensating increase in the level of quality and safety; and per 10CFR50.55a(g)(5)(iii) for situations where ASME Section XI requirements are considered impractical.

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

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

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

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

An index for QCNPS relief requests is included in Table 8.0-1. The I5R-XX relief request is applicable to ISI, CISI, SPT, and PDI.

The following relief requests are subject to change throughout the inspection interval (e.g., NRC approval, withdrawal). Changes to NRC approved alternatives (other than withdrawal) require NRC approval.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 8-2 ISI Program Plan - 5th Interval TABLE 8.0-1 RELIEF REQUEST INDEX Relief Revision Status2 (Program) Description/ Approval Summary1 Request Date3 I5R-01 0 Submitted (ISI) Inspection of Standby Liquid Control 9/26/12 Nozzle Inner Radius. Revision 0 Submitted.

I5R-02 0 Submitted (ISI) Alternate Risk-Informed Selection and 9/26/12 Examination Criteria for Examination Categories B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds. Revision 0 Submitted.

I5R-03 0 Submitted (SPT) Exemption From Pressure Testing Reactor 9/26/12 Pressure Vessel Head Flange Seal Leak Detection System. Revision 0 Submitted.

I5R-04 0 Submitted (SPT) Continuous Pressure Monitoring of the 9/26/12 Control Rod Drive (CRD) System Accumulators.

Revision 0 Submitted.

I5R-05 1 Authorized (ISI) Extension of Relief for Alternative Reactor (I4R-10, 9/26/12 Pressure Vessel Circumferential Weld Rev. 1) Examinations for Additional License Operating Period. Revision 1 authorized per NRC SE dated 3/23/05 (TAC Nos. MC2190, MC2191, MC2192, and MC2193). Permanent relief was authorized by NRC SE and thus applies to the 20 year extended period of operation of the renewed operating license, including this Fifth Inspection Interval.

I5R-06 0 Submitted (ISI) Use of BWRVIP Guidelines in Lieu of 9/26/12 Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection. Revision 0 Submitted.

I5R-07 Reserved Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 8-3 ISI Program Plan - 5th Interval TABLE 8.0-1 RELIEF REQUEST INDEX Relief Revision Status2 (Program) Description/ Approval Summary1 Request Date3 I5R-08 0 Authorized (ISI) Use of ASME Code Case N-789, (I4R-18) 9/26/12 Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate Energy Carbon Steel Exelon Piping for Raw Water Service. Revision 0 Fleet Relief authorized per NRC SE dated 5/10/12, for both Request the Fourth and Fifth Intervals. (See Note 2 of the relief request for details of the NRC SE authorized change.)

I5R-09 0 Submitted (ISI) Expanded Applicability for Use of ASME 9/26/12 Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission.

Revision 0 Submitted.

I5R-10 0 Submitted (ISI) Expanded Applicability for Use of ASME 9/26/12 Code Case N-661-1, Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service.

Revision 0 Submitted.

Note 1: The NRC grants relief requests pursuant to 10CFR50.55a(g)(6)(i) when Code requirements cannot be met and proposed alternatives do not meet the criteria of 10CFR50.55(a)(3). The NRC authorizes relief requests pursuant to 10CFR50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level of quality and safety or under 10CFR50.55a(a)(3)(ii) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety.

Note 2: This column represents the status of the latest revision. Relief Request Status Options: Authorized -

Approved for use in an NRC SE (See Note 1); Granted - Approved for use in an NRC SE (See Note 1);

Authorized Conditionally - Approved for use in an NRC SE which imposes certain conditions; Granted Conditionally - Approved for use in an NRC SE which imposes certain conditions; Denied - Use denied in an NRC SE; Expired - Approval for relief has expired; Withdrawn - Relief has been withdrawn by the station; Not Required - The NRC has deemed the relief unnecessary in an SE or RAI; Cancelled - Relief has been cancelled by the station prior to issue; and Submitted - Relief has been submitted to the NRC by the station and is awaiting approval.

Note 3: The revision listed is the latest revision of the subject relief request. The date this revision became effective is the date of the approving SE which is listed in the fourth column of the table. The date noted in the second column is the date of the ISI Program Plan revision when the relief request was incorporated into the document.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-01 Revision 0 (Page 1 of 3)

Request for Relief for Inspection of Standby Liquid Control Nozzle Inner Radius In Accordance with 10CFR50.55a(g)(5)(iii)

Inservice Inspection Impracticality

1. ASME Code Component(s) Affected:

Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-D Item Number: B3.100

Description:

Inspection of Standby Liquid Control Nozzle Inner Radius Component Number: Unit 1: N10 Unit 2: N10 Drawing Number: Unit 1: M-3106 Sh. 1 Unit 2: M-3116 Sh. 1

2. Applicable Code Edition and Addenda:

The Inservice Inspection Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

IWB-2500 states that components shall be examined and tested as specified in Table IWB-2500-1.

Table IWB-2500-1 requires a volumetric examination to be performed on the inner radius section of all reactor pressure vessel nozzles each inspection interval.

4. Impracticality of Compliance:

Pursuant to 10CFR50.55a(g)(5)(iii), relief is requested on the basis that compliance with the specified Code requirement has been determined to be impractical.

The Standby Liquid Control (SBLC) nozzle, as shown in Figure I5R-01.1, is designed with an integral socket to which the boron injection piping is fillet welded. The SBLC nozzle is located near the bottom of the vessel in an area which is inaccessible for ultrasonic examinations from the inside surface of the RPV. Therefore, ultrasonic examinations would need to be performed from the outside diameter of the RPV. As shown in Figure I5R-01.1, Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-01 Revision 0 (Page 2 of 3) the ultrasonic beam would need to travel through the full thickness of the vessel into a complex cladding/socket configuration. These geometric and material reflectors inherent in the design prevent a meaningful examination from being performed on the inner radius of the SBLC nozzle.

In addition, the inner radius socket attaches to the piping which injects boron at locations far removed from the nozzle. Therefore, the SBLC nozzle inner radius is not subjected to turbulent mixing conditions that are a concern at other nozzles.

5. Burden Caused By Compliance:

Compliance with the applicable Code requirements would require an ultrasonic examination to be performed on the outside diameter of the RPV. Geometric and material reflectors would prevent a meaningful examination, resulting in inaccurate data. Based on this, the Code requirements are deemed impractical in accordance with 10CFR50.55a(g)(5)(iii).

6. Proposed Alternative and Basis for Use:

As an alternate examination, Quad Cities Nuclear Power Station Units 1 and 2 will perform a VT-2 visual examination of the subject nozzles each refueling outage in conjunction with the Class 1 System Leakage Test.

7. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval Quad Cities Nuclear Power Station Units 1 and 2.

8. Precedents:

Quad Cities Nuclear Power Station Units 1 and 2 Fourth Inspection Interval Relief Request I4R-01 was authorized per NRC SE dated January 28, 2004. The Fifth Inspection Interval Relief Request utilizes a similar approach that was previously approved.

Dresden Nuclear Power Station Units 2 and 3 Fourth Inspection Interval Relief Request I4R-01 was authorized per NRC SE dated September 4, 2003.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-01 Revision 0 (Page 3 of 3)

FIGURE I5R-01.1 2 INCH STANDBY LIQUID CONTROL NOZZLE Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 1 of 15)

Request for Relief for Alternate Risk-Informed Selection and Examination Criteria for Examination Categories B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 1 and 2 Examination Category: B-F, B-J, C-F-1, and C-F-2 Item Number: B5.10, B5.20, B9.11, B9.21, B9.31, B9.32, B9.40, C5.11, C5.51, C5.70, and C5.81

Description:

Alternate Risk-Informed Selection and Examination Criteria for Examination Categories B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds Component Number: Pressure Retaining Piping

2. Applicable Code Edition and Addenda:

The Inservice Inspection Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

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

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

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

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 2 of 15)

a. primary plus secondary stress intensity range of 2.4Sm for ferritic steel and austenitic steel; and

b. cumulative usage factor U of 0.4.

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

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

Table IWC-2500-1, Examination Categories C-F-1 and C-F-2 require volumetric and surface examinations on a sample of welds for Item Numbers C5.11 and C5.51, and surface examinations on a sample of welds for Item Numbers C5.70 and C5.81. The weld population selected for inspection includes the following:

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

a. the examinations shall be distributed among the Class 2 systems prorated, to the degree practicable, on the number of nonexempt dissimilar metal, austenitic stainless steel and high alloy welds (Examination Category C-F-1) or carbon and low alloy welds (Examination Category C-F-2) in each system;

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

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

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 3 of 15)

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative utilizing Reference 1 along with two enhancements from Reference 4 will provide an acceptable level of quality and safety.

As stated in Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999)

(Reference 2):

The staff concludes that the proposed RI-ISI Program as described in EPRI TR-112657, Revision B, is a sound technical approach and will provide an acceptable level of quality and safety pursuant to 10CFR50.55a for the proposed alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection.

The initial Quad Cities Nuclear Power Station Units 1 and 2 Risk-Informed Inservice Inspection (RISI) Program was submitted during the third period of the Third ISI Interval for both Units 1 and 2. This initial RISI Program was developed in accordance with EPRI TR-112657, Revision B-A, as supplemented by ASME Code Case N-578-1. The program was approved for use by the NRC via Safety Evaluation as transmitted to Exelon on February 5, 2002 (Reference 5).

The Quad Cities Nuclear Power Station RISI Program was resubmitted using the same approach during the Fourth ISI Interval for both Units 1 and 2. The program was approved for use by the NRC via Safety Evaluation as transmitted to Exelon on January 28, 2004 (Reference 6).

The transition from the 1995 Edition through the 1996 Addenda to the 2007 Edition through the 2008 Addenda of ASME Section XI for Quad Cities Nuclear Power Stations Fifth ISI Interval does not impact the currently approved Risk-Informed ISI evaluation methods and process used in the Fourth ISI Interval, and the requirements of the new Code Edition/Addenda will be implemented as detailed in the Quad Cities Nuclear Power Station ISI Program Plan.

The Risk Impact Assessment completed as part of the original baseline RISI Program was an implementation/transition check on the initial impact of converting from a traditional ASME Section XI program to the new RISI methodology. For the Fifth Interval ISI update, there is no transition occurring between two different methodologies, but rather, the currently approved RISI methodology and evaluation will be maintained for the new interval. The original methodology of the evaluation has not changed, and Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 4 of 15) the change in risk was simply re-assessed using the initial 1989 Edition, No Addenda ASME Section XI program prior to RISI and the new element selection for the Fifth Interval RISI Program. This same process has been maintained in each revision to the Quad Cities Nuclear Power Station RISI assessment that has been performed to date.

The actual evaluation and ranking procedure including the Consequence Evaluation and Degradation Mechanism Assessment processes of the currently approved (Reference

6) RISI Program remain unchanged and are continually applied to maintain the Risk Categorization and Element Selection methods of EPRI TR-112657, Revision B-A.

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

5. Proposed Alternative and Basis for Use:

The proposed alternative originally implemented in the Risk Informed Inservice Inspection Plan, Quad Cities Units 1 and 2 (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 same program along with these enhancements was resubmitted and is currently approved for Quad Cities Nuclear Power Stations Fourth ISI Interval as documented in Reference 6.

The Fifth Inspection Interval RISI Program will be a continuation of the current application and will continue to be a living program as described in the Reason for Request section of this relief request. No changes to the evaluation methodology as currently implemented under EPRI TR-112657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

a. In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, RI-ISI Selected Examinations of EPRI TR-112657, Quad Cities Nuclear Power Station will utilize the requirements of Paragraph -2430, Additional Examinations contained in ASME Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in ASME Code Case N-578-1 provide a more refined methodology for implementing necessary additional examinations. The reason for this selection is that the guidance discussed in EPRI TR-112657 includes requirements for additional examinations at a high level, based on service conditions, degradation mechanisms, and the performance of evaluations to determine the scope of additional examinations, whereas ASME Code Case N-578-1 provides more specific and clearer guidance Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 5 of 15) regarding the requirements for additional examinations that is structured similar to the guidance provided in ASME Section XI, IWB-2430 and IWC-2430.

Additionally, similar to the current requirements of ASME Section XI, Quad Cities Nuclear Power Station intends to perform additional examinations that are required due to the identification of flaws or relevant conditions exceeding the acceptance standards, during the outage the flaws are identified.

b. To supplement the requirements listed in Table 4-1, Summary of Degradation-Specific Inspection Requirements and Examination Methods, of EPRI TR-112657, Quad Cities Nuclear Power Station will utilize the provisions listed in Table 1, Examination Category R-A, Risk-Informed Piping Examinations, contained in ASME Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from the 2007 Edition through the 2008 Addenda of ASME Section XI (Quad Cities Nuclear Power Stations Code of Record for the Fifth Interval) will be utilized which parallel those referenced in the Code Case for the 1989 Edition, No Addenda. Table 1 of ASME Code Case N-578-1 will be used as it provides a detailed breakdown for Examination Method and Categorization of Parts to be Examined. Based on these Methods and Categorization, the examination figures specified in Section 4 of EPRI TR-112657 will then be used to determine the examination volume/area based on the degradation mechanism and component configuration.

The Quad Cities Nuclear Power Station RISI Program, as developed in accordance with EPRI TR-112657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as High risk (i.e., Risk Categories 1, 2, and 3) and 10% of the elements that are categorized as Medium risk (i.e., Risk Categories 4 and 5) be selected for inspection. For this application, the guidance for the examination volume for a given degradation mechanism is provided by the EPRI TR-112657 while the guidance for the examination method and categorization of parts to be examined are provided by the EPRI TR-112657 as supplemented by ASME Code Case N-578-1.

For NRC staff consideration in the evaluation of this alternative Risk-Informed ISI Program, Attachment 1 to the relief request contains a summary of the Regulatory Guide 1.200, Revision 2 (Reference 7), evaluation performed on Quad Cities Nuclear Station Quantification Notebook, QC PSA-014, Revision 2, January 2011 (PRA Model QC110A) (Reference 8), and the impact of the identified gaps on the technical adequacy of the Quad Cities Nuclear Power Station PRA Model to support this RISI application (see Attachment 1, Table 1).

In addition to this risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of risk classification, will continue to Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 6 of 15) receive Code required pressure testing as part of the current ASME Section XI program.

VT-2 visual examinations are scheduled in accordance with the Quad Cities Nuclear Power Station pressure testing program, which remains unaffected by the RISI Program.

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

Quad Cities Nuclear Power Station Units 1 and 2 Fourth Inspection Interval Relief Request I4R-02 was authorized per NRC SE dated January 28, 2004. The Fifth Inspection Interval Relief Request utilizes a similar RISI methodology that was previously approved.

Three Mile Island Station Fourth Inspection Interval Relief Request I4R-02 was authorized per NRC SE dated July 20, 2011.

Clinton Power Station Third Inspection Interval Relief Request I3R-01 was authorized per NRC SE dated December 22, 2010.

8.

References:

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

2) Letter from W. H. Bateman (NRC) to G. L. Vine (EPRI) Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR-112657, Revision B, July 1999), dated October 28, 1999

3) Initial Risk-Informed Inservice Inspection Evaluation - Quad Cities Nuclear Power Station Units 1 and 2, dated August 2000

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

5) Letter from A. J. Mendiola (NRC) to O. D. Kingsley (Exelon), Safety Evaluation of Third Interval Risk-Informed Inservice Inspection Program Relief Request, dated February 5, 2002 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 7 of 15)

6) Safety Evaluation from Anthony J. Mendiola (NRC) to John L. Skolds (Exelon),

Quad Cities Nuclear Power Station, Units 1 and 2 - Fourth 10-Year Interval Inservice Inspection Relief Requests Nos. I4R-01 through I4R-09 (TAC Nos.

MB7695 through MB7712), dated January 28, 2004

7) Regulatory Guide 1.200, Revision 2, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, dated March 2009

8) Quad Cities Nuclear Power Station Quantification Notebook, QC PSA-014, Revision 2, January 2011, PRA Model QC110A Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 8 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION Summary Statement of Quad Cities Nuclear Power Station PRA Model Capability for Use in Risk-Informed Inservice Inspection Applications Introduction Exelon Generation Company (EGC) employs a multi-faceted approach to establishing and maintaining the technical adequacy and plant fidelity of the Probabilistic Risk Assessment (PRA) models for all operating EGC nuclear generation sites. This approach includes both a proceduralized PRA maintenance and update process, and the use of self-assessments and independent peer reviews. The following information describes this approach as it applies to the Quad Cities Nuclear Power Station PRA.

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

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

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

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

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

In addition to these activities, EGC risk management procedures provide the guidance for particular risk management and PRA quality and maintenance activities. This guidance Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 9 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION includes:

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

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

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

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

In accordance with this guidance, regularly scheduled PRA model updates nominally occur on an approximately 4-year cycle; longer intervals may be justified if it can be shown that the PRA continues to adequately represent the as-built, as-operated plant.

The most recent update of the Quad Cities Nuclear Power Station PRA model (designated the QC110A model) (Reference 9) was completed in 2010 as a result of a regularly scheduled update to the previous 2005B PRA model. This model is the most recent evaluation of the risk profile at Quad Cities Nuclear Power Station for internal event challenges, including internal flooding.

The Quad Cities Nuclear Power Station PRA modeling is highly detailed, including a wide variety of initiating events, modeled systems, operator actions, and common cause events. The PRA model quantification process used for the Quad Cities Nuclear Power Station PRA is based on the event tree / fault tree methodology, which is a well-known methodology in the industry.

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

An independent PRA peer review was conducted under the auspices of the BWR Owners Group in February 2000 (Reference 2), following the Industry PRA Peer Review process (Reference 1). This peer review included an assessment of the PRA model maintenance and update process.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 10 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION

In 2004, prior to the 2005 PRA update, a self-assessment analysis was performed against the available version of the ASME PRA Standard, Addendum A (Reference 3).

During 2005 and 2006, the Quad Cities Nuclear Power Station PRA model results were evaluated in the BWR Owners Group PRA cross-comparisons study performed in support of implementation of the Mitigating Systems Performance Indicator (MSPI) process.

In 2009, an update of the self-assessment analysis was performed against ASME PRA Standard, Addendum B (Reference 3). The 2009 self-assessment also addresses the updated Supporting Requirements associated with PRA Model Uncertainty as provided in the Combined PRA Standard (Reference

4) and Regulatory Guide 1.200, Rev. 2 (Reference 5).

In May 2010, a focused PRA Peer Review was performed on the internal flooding element of the Quad Cities Nuclear Power Station QC105B full power internal events PRA. The results of the focused internal flood PRA Peer Review indicated a small number of the Internal Flood Supporting Requirements (SRs) were not met. The associated findings were added to the Updating Requirements Evaluation (URE) database to ensure resolution and have been included in the 2010 PRA self-assessment.

In 2010, an update of the PRA self-assessment analysis was performed against the Combined PRA Standard (Reference 4) following completion of the Quad Cities Nuclear Power Station 2010 PRA (QC110A) update. The self-assessment considered all of the findings from the 2010 focused PRA Peer Review on Internal Flooding. The self-assessment was performed in 2010, but signed off in January 2011.

A summary of the disposition of the 2000 Industry PRA Peer Review Facts and Observations (F&Os) for the Quad Cities Nuclear Power Station PRA models was documented as part of the statement of PRA capability for MSPI in the Quad Cities Nuclear Power Station MSPI Basis Document (Reference 7). As noted in that document, there were no significance level A F&Os from the peer review, and all significance level B F&Os were addressed and closed out with the completion of the current model of record (2005B). Also noted in that submittal was the fact that, after allowing for plant-specific features, there are no MSPI cross-comparison outliers for Quad Cities Nuclear Power Station (refer to the third bulleted item above).

A Self-Assessment Analysis (Gap Analysis) for the Quad Cities Nuclear Power Station PRA model was completed in 2004. This Gap Analysis was performed against the ASME PRA Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 11 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION Standard, Addendum A (Reference 3). The 2004 gap analysis defined a list of 85 supporting requirements from the Standard for which potential gaps to Capability Category II of the Standard were identified. For each such potential gap, a PRA updating requirements evaluation (URE) (EGC model update tracking database) was documented for resolution.

A previous PRA model update was completed in 2005. In updating the PRA, changes were made to the PRA to address most of the identified gaps, as well as to address other open UREs.

Following the update, an assessment of the status of the gap analysis relative to the new model and the updated requirements in Addendum B of the ASME PRA Standard concluded that 69 of the gaps were fully resolved (i.e., are no longer gaps), and another one (1) was partially resolved.

After accounting for the number of SRs added or deleted as part of Addendum B, the Quad Cities Nuclear Power Station PRA contained 21 potential gaps to Capability Category II of the Standard.

As indicated above, a recent PRA model update was completed in 2010, resulting in the QC110A updated PRA model. In updating the PRA, changes were made to address most of the previously identified gaps as well as to address other open UREs. Following the update, an assessment of the status of the Gap Analysis relative to the new PRA model and the requirements in Addendum B of the ASME PRA Standard (Reference 3) concluded that 19 of the gaps were fully resolved (i.e., are no longer gaps) (Reference 10). The Quad Cities Nuclear Power Station PRA (QC110A) contains 2 potential gaps to Capability Category II of the Standard (Reference 4).

A summary of this assessment of the current open items relative to the RISI relief request is provided in Table 1. All remaining gaps will be reviewed for consideration during the next Quad Cities Nuclear Power Station model update (anticipated to be 2014) but are judged to have low impact on the PRA model or its ability to support a full range of PRA applications. These items are documented in the PRA URE database so that they can be tracked and their potential impacts accounted for in applications where appropriate. In addition, plant changes made since the last PRA update have been reviewed and determined to not have a significant PRA impact.

These items are also documented in UREs for consideration in future PRA updates, as appropriate.

Guidance from EPRI Report on PRA Technical Adequacy for RISI EPRI report TR-1021467-A (Reference 8) provides guidance on the PRA Standard Capability Category necessary to support RISI. This report received a Safety Evaluation (SE) from the Nuclear Regulatory (NRC) in January 2012. Reg. Guide 1.200 considers it a good practice to Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 12 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION have, in general, SRs meet Capability Category II for applications. However, according to the EPRI report not all SRs require Capability Category II to adequately support RISI applications.

According to the EPRI report the Quad Cities Nuclear Power Station gaps listed in Table 1 (SC-A5 and DA-C12) do not require Capability Category II, but instead only require Capability Category I, the most basic level. Therefore according to EPRI TR-1021467-A and the associated NRC SE, the Quad Cities Nuclear Power Station PRA model QC110A is adequate for use in the RISI application.

Sensitivity Studies Two gaps to the PRA standard were identified in Table 1 (SC-A5 and DA-C12). These gaps have been reviewed to determine which, if any, would merit RISI-specific sensitivity studies in the presentation of the application results. The result of this assessment concluded that no additional sensitivity studies are merited.

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

Conclusion Regarding PRA Capability for Risk-Informed ISI The Quad Cities Nuclear Power Station PRA model continues to be suitable for use in the risk-informed inservice inspection application. This conclusion is based on:

PRA maintenance and update processes in place.

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

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

In support of the PRA analyses for the Quad Cities Nuclear Power Station 10-year interval evaluation using the QC110A PRA model, the remaining gaps to the PRA standard have been Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 13 of 15)

ATTACHMENT 1 QUAD CITIES 2010 PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION reviewed to determine which, if any, would merit RISI-specific sensitivity studies in the presentation of the application results. The result of this assessment concluded that no additional sensitivity studies are merited.

References

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

2. Quad Cities Nuclear Power Station PRA Peer Review, February 2000

3. Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, (ASME RA-S-2002), Addenda RA-Sa-2003, and Addenda RA-Sb-2005, December 2005

4. ASME/American Nuclear Society, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME/ANS RA-Sa-2009, March 2009

5. An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment results for Risk-Informed Activities, Regulatory Guide 1.200, U.S. Nuclear Regulatory Commission, March 2009 Revision 2

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

7. Reactor Oversight Program MSPI Bases Document, Quad Cities Nuclear Power Station, Revision 4, February 2008

8. Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-informed Inservice Inspection Programs, EPRI TR-1021467-A, June 2012

9. Quad Cities Nuclear Power Station Quantification Notebook, QC PSA-014, Revision 2, January 2011

10. Self-Assessment of the Quad Cities Nuclear Power Station PRA Against the ASME PRA Standard Requirements, QC PSA-016, Revision 2, January 2011 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 14 of 15)

ATTACHMENT 1 QUAD CITIES PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION TABLE 1 - STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD TITLE DESCRIPTION OF GAP APPLICABLE CURRENT STATUS COMMENT SRS Gap #1 Reaching a safe stable end state defines the SC-A5 Open. Enhance documentation to The current approach is judged to success of a sequence and therefore the mission justify why extending FTR be reasonable for long term time of the sequence to achieve the Level 1 end mission times beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months scenarios (e.g., long term loss of state. The mission times for failure to run for loss of DHR sequences is not DHR) and is adequate for RISI calculations are assessed at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or less if necessary. The considerations application.

specifically justified. that support the choice of the mission time are as follows:

Extending the FTR mission time beyond 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months for loss of DHR sequences is considered to be an Equipment failure rates unnecessary complication and does not affect (failures/hour) are judged to be PRA insights nor does it significantly affect its too conservative for times greater quantitative evaluation. than a few hours of operation.

For times greater than a few hours, the ability to repair and recover equipment can compete with the failure rate such that there can be considered to be a steady state equilibrium condition reached.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-02 Revision 0 (Page 15 of 15)

ATTACHMENT 1 QUAD CITIES PRA (QC110A) TECHNICAL CAPABILITY ASSESSMENT FOR RISK-INFORMED INSERVICE INSPECTION TABLE 1 - STATUS OF IDENTIFIED GAPS TO CAPABILITY CATEGORY II OF THE ASME PRA STANDARD TITLE DESCRIPTION OF GAP APPLICABLE CURRENT STATUS COMMENT SRS Gap #2 No interviews of plant staff were performed to DA-C12 Open. This deviation from the The model is consistent with data generate uncertainty estimates of unavailability SR is not considered to from the plant MR database, per maintenance act. significantly alter the PRA which is adequate for RISI qualitative or quantitative results. application.

An exception is taken to DA-C12. The plant staff does not have reasonable insights applicable to the level of uncertainty associated with the maintenance durations. Most plant staff have rotated positions and do not have sufficient longevity to provide this insight.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 1 of 6)

Request for Relief for Exemption from Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System In Accordance with 10CFR50.55a(a)(3)(ii)

Hardship or Unusual Difficulty Without Compensating Increase in Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 2

Reference:

IWC-2500, Table IWC-2500-1 Examination Category: C-H Item Number: C7.10

Description:

Exemption from Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection System Component Number: Flange Seal Leak Detection Line Pressure Retaining Components Drawing Number: Unit 1: M-35 Sh. 1 Unit 2: M-77 Sh. 1

2. Applicable Code Edition and Addenda:

The Inservice Inspection Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirements:

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

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(ii), relief is requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

The Reactor Pressure Vessel Head Flange Leak Detection Line is separated from the reactor pressure boundary by one passive membrane, a silver plated O-ring located on the vessel Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 2 of 6) flange. A second O-ring is located on the opposite side of the tap in the vessel flange (see Figure I5R-03.2). This line is required during plant operation in order to indicate failure of the inner flange seal O-ring. Failure of the O-ring would result in the annunciation of a High Level Alarm in the control room. On this annunciation, control room operators would quantify the leakage rate from the O-ring and then isolate the leak detection line from the drywell sump by closing the AO 1(2)-220-51 valve (see Figure I5R-03.1). This action is taken in order to prevent steam cutting of the O-ring and the vessel flange. Failure of the inner O-ring is the only condition under which this line is pressurized.

The configuration of this system precludes manual testing while the vessel head is removed because the odd configuration of the vessel tap (see I5R-03.2), combined with the small size of the tap and the high test pressure requirement (1000 psig minimum), prevents the tap in the flange from being temporarily plugged. The opening in the flange is only 3/16 of an inch in diameter and is smooth walled making a high pressure temporary seal very difficult.

Failure of this seal could possibly cause ejection of the device used for plugging into the vessel.

A pneumatic test performed with the head installed is precluded due to the configuration of the top head. The top head of the vessel contains two grooves that hold the O-rings. The O-rings are held in place by a series of retainer clips spaced 15o apart. The retainer clips are contained in a recessed cavity in the top head (see Figure I5R-03.3). If a pressure test was performed with the head on, the inner O-ring would be pressurized in a direction opposite to what it would see in normal operation. This test pressure would result in a net inward force on the O-ring that would tend to push it into the recessed cavity that houses the retainer clips.

The O-ring material is only .050" thick with a silver plating thickness of .004" to .006" and could very likely be damaged by this deformation into the recessed areas on the top head.

In addition to the problems associated with the O-ring design that preclude this testing it is also questionable whether a pneumatic test is appropriate for this line. Although the line will initially contain steam if the inner O-ring leaks, the system actually detects leakage rate by measuring the level of condensate in a collection chamber. This would make the system medium water at the level switch. Finally, the use of a pneumatic test performed at a minimum of 1000 psig would represent an unnecessary risk in safety for the inspectors and test engineers in the unlikely event of a test failure, due to the large amount of stored energy contained in air pressurized to 1000 psig.

System leakage testing of this line is precluded because the line will only be pressurized in the event of a failure of the inner O-ring. It is extremely impractical to purposely fail the inner O-ring in order to perform a test.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 3 of 6)

Based on the above, Quad Cities Nuclear Power Station Units 1 and 2 requests relief from the ASME Section XI requirements for system leakage testing of the Reactor Pressure Vessel Head Flange Seal Leak Detection System.

5. Proposed Alternate and Basis for Use:

A VT-2 visual examination will be performed on the line during vessel flood-up during a refueling outage. The static head developed due to the water above the vessel flange during flood-up will allow for the detection of any gross indications in the line. This examination will be performed with the frequency specified by Table IWC-2500-1 for a System Leakage Test (once each inspection period).

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

Quad Cities Nuclear Power Station Units 1 and 2 Fourth Inspection Interval Relief Request I4R-05 was authorized per NRC SE dated January 28, 2004. The Fifth Inspection Interval Relief Request utilizes a similar approach that was previously approved.

Three Mile Island Station Fourth Inspection Interval Relief Request I4R-03 was granted per NRC SE dated July 20, 2011.

Clinton Power Station Third Inspection Interval Relief Request I3R-03 was authorized per NRC SE dated December 22, 2010.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 4 of 6)

FIGURE I5R-03.1 HEAD FLANGE SEAL LEAK DETECTION SCHEMATIC Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 5 of 6)

FIGURE I5R-03.2 FLANGE SEAL LEAK DETECTION LINE DETAIL Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-03 Revision 0 (Page 6 of 6)

FIGURE I5R-03.3 O-RING CONFIGURATION Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-04 Revision 0 (Page 1 of 3)

Request for Relief for Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 2

Reference:

IWC-2500, Table IWC-2500-1 Examination Category: C-H Item Number: C7.10

Description:

Continuous Pressure Monitoring of the Control Rod Drive (CRD) System Accumulators Component Number: CRD Accumulators and Associate Piping Drawing Number: Unit 1: M-41 Shts. 1 & 3 Unit 2: M-83 Shts. 1 & 3

2. Applicable Code Edition and Addenda:

The Inservice Inspection Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

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

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative provides an acceptable level of quality and safety.

As required by Quad Cities Nuclear Power Station Units 1 and 2 Technical Specifications, the CRD System Accumulator pressure must be greater than or equal to 940 psig to be considered operable. The accumulator pressure is continuously monitored by system instrumentation. Since the accumulators are isolated from the source of make-up nitrogen, the continuous monitoring of the CRD accumulators functions as a pressure Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-04 Revision 0 (Page 2 of 3) decay type test. Should accumulator pressure fall below 1000 psig, an alarm is received in the control room. The pressure drop for the associated accumulator is then recorded, and the accumulator is recharged in accordance with Quad Cities Nuclear Power Station procedures. If an accumulator requires charging more than twice in a thirty day period, then corrective actions are taken. When leakage is detected, corrective actions are taken to repair the leaking component as required by Quad Cities Nuclear Power Station procedures.

Since monitoring the nitrogen side of the accumulators is continuous, any leakage from the accumulator would be detected by normal system instrumentation. An additional VT-2 visual examination performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a VT-2 visual examination would require applying a leak detection solution to 177 accumulators per unit resulting in additional radiation exposure without any added benefit in safety. This inspection would not be consistent with As Low As Reasonably Achievable (ALARA) practices.

Relief is requested from the VT-2 visual examination requirements specified in Table IWC-2500-1 for the nitrogen side of the CRD system accumulators on the basis that Quad Cities Nuclear Power Station Technical Specification Surveillance requirements exceed the code requirement for a VT-2 visual examination.

5. Proposed Alternate and Basis for Use:

As an alternate to the VT-2 visual examination requirements of Table IWC-2500-1, Quad Cities Nuclear Power Station will perform continuous pressure decay monitoring in conjunction with Technical Specifications for the nitrogen side of the CRD accumulators including attached piping.

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

Quad Cities Nuclear Power Station Units 1 and 2 Fourth Inspection Interval Relief Request I4R-06 was authorized per NRC SE dated January 28, 2004. The Fifth Inspection Interval Relief Request utilizes a similar approach that was previously approved.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-04 Revision 0 (Page 3 of 3)

LaSalle County Nuclear Power Station Units 1 and 2 Third Inspection Interval Relief Request I3R-09 was authorized per NRC SE dated January 30, 2008. Revision 1 submitted under Supplement RA-07-036a, dated July 20, 2007.

Dresden Nuclear Power Station Units 2 and 3 Fourth Inspection Interval Relief Request I4R-07 was authorized per NRC SE dated September 4, 2003.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 1 of 10)

Request for Relief for Extension of Relief for Alternative Reactor Pressure Vessel Circumferential Weld Examinations for Additional License Operating Period In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

- - NOTE ***

Quad Cities Nuclear Power Station Units 1 and 2 Fifth Inspection Interval Relief Request I5R-05, Revision 1 is provided for completeness and has no technical changes from the previously approved relief request. This relief request was previously submitted to address both Quad Cities Nuclear Power Station and Dresden Nuclear Power Station on February 23, 2004, and approved under the Fourth Inspection Interval ISI Program Plan as Relief Request I4R-10, Revision 1. The approval authorized under NRC SE dated March 23, 2005, for Quad Cities Nuclear Power Station, was for permanent relief and thus applies to the 20 year extended period of operation of the renewed operating license, including this Fifth Inspection Interval.

The relief request is carried here and renumbered as I5R-05, Revision 1. All ASME Code references were made in accordance with the 1995 Edition through the 1996 Addenda of ASME Section XI. No changes to the actual approved relief request have been made and no further or revised authorization is required.

1. ASME Component(s) Affected:

Components affected are American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code (Code), Section XI, Class 1 pressure retaining reactor pressure vessel (RPV) shell circumferential welds, Examination Category B-A, Item Number B1.11.

2. Applicable Code Edition and Addenda:

The applicable ASME Code, Section XI, for Dresden Nuclear Power Station (DNPS),

Units 2 and 3, and Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2, is the 1995 Edition through the 1996 Addenda.

3. Applicable Code Requirement:

In accordance with the provisions of 10CFR50.55a, Codes and standards, paragraph (a)(3)(i), Exelon Generation Company, LLC (EGC) requests permanent relief for the Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 2 of 10) additional license operating period requested in Reference 1 for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2, from the requirement of ASME Code, Section XI, Subarticle IWB-2500, Table IWB-2500-1, Examination Category B-A, Item Number B1.11.

Subarticle IWB-2500 requires components specified in Table IWB-2500-1 to be examined. Table IWB-2500-1 requires volumetric examination of all RPV shell circumferential welds each inspection interval (i.e., Examination Category B-A, Item Number B1.11).

4. Reason for Request:

Reference 2 provides the technical basis for permanently deferring the augmented inspections of circumferential welds in boiling water reactor (BWR) RPVs. In the report, the BWR Vessel and Internals Project (BWRVIP) concluded that the probabilities of failure for BWR RPV circumferential welds are orders of magnitude lower than that of the longitudinal welds. The NRC conducted an independent risk-informed, probabilistic fracture mechanics assessment (PFMA) of the analysis presented in Reference 2, and the results are documented in Reference 3. EGC has determined that the proposed alternative described below provides an acceptable level of quality and safety and satisfies the requirements of 10CFR50.55a(a)(3)(i).

5. Proposed Alternative and Basis for Use:

Proposed Alternative In accordance with 10CFR50.55a(a)(3)(i), and consistent with information contained in Reference 4, EGC proposes the following alternate provisions for the subject weld examinations since the proposed alternative provides an acceptable level of quality and safety.

The failure frequency for RPV shell circumferential welds is sufficiently low to justify their elimination from the Inservice Inspection (ISI) requirement of ASME Code, Section XI, Table IWB-2500-1, Examination Category B-A, Item Number B1.11.

The ISI examination requirements of the ASME Code, Section XI, Table IWB-2500-1, Examination Category B-A, Item Number B1.12, RPV shell longitudinal welds (i.e., also known as vertical or axial welds) shall be performed, to the extent possible, and shall include inspection of the circumferential welds only at the intersection of these welds with the longitudinal welds, or approximately 2 to 3 percent of the RPV shell circumferential welds. When this examination is performed, an automated ultrasonic inspection system will provide the best possible examination of the RPV shell longitudinal welds. These welds are generally only accessible from inside surfaces of the Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 3 of 10)

RPV using an automated ultrasonic inspection system, which provides the best possible examination of the RPV shell longitudinal welds. Inspections from the outside surfaces have limited access due to the close proximity of the biological shield to the RPV. Also, the reflective insulation that occupies this space is not designed for removal.

Basis for Use Reference 2 provides the technical basis to justify relief from the examination requirements of RPV shell circumferential welds. The results of the NRCs evaluation of Reference 2 are documented in Reference 3. Reference 4 permits BWR licensees to request permanent (i.e., for the remaining term of operation under the existing, initial, license) relief from the ISI requirements of 10CFR50.55a(g) for the volumetric examination of RPV shell circumferential welds (i.e., ASME Code, Section XI, Table IWB-2500-1, Examination Category B-A, Item Number B1.11). This relief can be granted by demonstrating that:

1. at the expiration of their license, the circumferential welds will continue to satisfy the limiting conditional failure probability for circumferential welds in the staffs July 30, 1998, safety evaluation, and

2. licensees have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the staffs July 30, 1998, safety evaluation.

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

Response

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

Tables 1 and 2 provide a comparison of the limiting RPV circumferential weld parameters for each DNPS and QCNPS unit to those found in Table 2.6-5 of the NRC final safety evaluation of BWRVIP-05 (i.e., Reference 3) for a Babcock and Wilcox vessel. Although the chemistry composition and chemistry factor for DNPS Unit 3 are Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 4 of 10) higher than the limits of the NRC analysis, the shifts in reference temperature for both units are lower than the shift from the NRC limiting analysis. In addition, the unirradiated reference temperatures for both DNPS units are lower. The combination of unirradiated reference temperature and embrittlement shift yields adjusted reference temperatures considerably lower than the NRC mean analysis values.

The chemistry composition and chemistry factor for QCNPS Unit 1 are less than or equal to the limits of the NRC analysis. While the nickel content for Unit 2 is higher than the value utilized in the NRC analysis, the Unit 2 copper content and the chemistry factor are considerably lower than the values utilized in the NRC analysis. Additionally, the unirradiated reference temperatures for both QCNPS units are lower than the NRC limits.

The combination of unirradiated reference temperature and embrittlement shift yields adjusted reference temperatures considerably lower than the NRC mean analysis values.

The end of life (i.e., 54 effective full power year (EFPY)) inside diameter fluences for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2, are considerably lower than the NRC estimated 54 EFPY fluence. The 54 EFPY fluence estimates were calculated using the fluence methodology of General Electric Nuclear Energy licensing topical report NEDC-32983P (i.e., Reference 5), which was approved by the NRC in Reference 6, and adheres to the guidance of Regulatory Guide 1.190 (i.e., Reference 7). The end of extended license operating time of 54 EFPY includes the extended power uprate approved by the NRC in References 8 and 9. There are no additional uprates planned for DNPS or QCNPS.

The shifts in reference temperature for all four units are lower than the 54 EFPY shift from the NRC analysis. Therefore, for each unit, the RPV shell weld embrittlement due to fluence is calculated to be less than the NRCs limiting case, and each units RPV shell circumferential weld failure probabilities are bounded by the conditional failure probability, P(FIE), in the NRCs limiting plant specific analysis (54 EFPY) through the projected additional license operating period. For these reasons, the DNPS, Units 2 and 3, and QCNPS, Units 1 and 2, RPVs are bounded by Reference 3.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 5 of 10)

Table 1: Effects of Irradiation on RPV Circumferential Weld Properties - DNPS DNPS Unit 3 DNPS Unit 2 Parameters at Parameters at NRC Limiting Plant Parameter 54 EFPY (Weld Wire 54 EFPY (Weld Wire Specific Analysis Description Heat/Flux Lot #

Heat/Flux Lot # (54 EFPY) 299L44/8650 71249/8504)

(WF-19/WF-25))

Copper (weight %) 0.23 0.34 0.31 Nickel (weight %) 0.59 0.68 0.59 Chemistry Factor 168 221 196.7 End of Life Inside 0.042 0.041 0.19 Diameter Fluence (1019 n/cm2)

RTNDT (°F) 44 58 109.4 RTNDT(U) (°F) 10 -5 20 Mean RTNDT (°F) 54 53 129.4 Table 2: Effects of Irradiation on RPV Circumferential Weld Properties - QCNPS QCNPS Unit 2 QCNPS Unit 1 Parameters at Parameters at NRC Limiting Plant Parameter 54 EFPY (Weld Wire 54 EFPY (Weld Wire Specific Analysis Description Heat/Flux Lot #

Heat/Flux Lot # (54 EFPY)

S3986/3870) 406L44/8688)

Linde 124 Copper (weight %) 0.27 0.05 0.31 Nickel (weight %) 0.59 0.96 0.59 Chemistry Factor 183 68 196.7 End of Life Inside 0.041 0.041 0.19 Diameter Fluence (1019 n/cm2)

RTNDT (°F) 48 18 109.4 RTNDT(U) (°F) -5 -32 20 Mean RTNDT (°F) 43 -14 129.4 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 6 of 10)

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

Response

EGC has procedures in place for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2, that guide operators in controlling and monitoring reactor pressure during all phases of operation, including cold shutdown. Use of these procedures minimizes the potential for low temperature over-pressurization (LTOP) events, and is reinforced through operator training. A Primary System Leakage test is performed prior to each restart after a refueling outage. The associated station test procedure has sufficient guidance to minimize the likelihood of an LTOP event, and requires a pre-job briefing prior to test commencement with all involved personnel. During pressure testing, measures are taken to limit the potential for system perturbations that could lead to pressure transients.

These measures include both administrative and/or hardware controls, such as limiting testing or work activities, or installing jumpers to defeat system actuations that are not required operable. RPV temperature and pressure are required to be monitored and controlled to within the Technical Specifications pressure and temperature (P/T) limits curve during all portions of the testing. The normal and contingency methods to enact pressure control are specified in the test procedure.

A designated Test Coordinator is responsible for the coordination of the test (i.e., from initiation to conclusion) and maintains cognizance of test status. A controlled rate of pressure increase is administratively limited in the test procedure to no greater than 30 pounds per square inch (psi) per minute at DNPS, and not greater than 50 psi per minute at QCNPS. If the rate of pressurization exceeds this limit, a contingency sequence portion of the testing procedures provides directions to reduce the rate of pressure increase by depressurizing through the Reactor Water Cleanup System, securing Control Rod Drive (CRD) pumps, and opening the main steam drain lines.

Other than the CRD system, the other high pressure coolant sources that could inadvertently initiate and result in an LTOP event are the Condensate/Feedwater, the Safe Shutdown Makeup Pump (SSMP) at QCNPS, Reactor Core Isolation Cooling (RCIC) at QCNPS, and High Pressure Coolant Injection (HPCI) Systems.

During a normal RPV fill sequence prior to pressure testing, the Condensate System is used to fill the reactor. This evolution is carefully controlled per the test procedure to minimize the potential for an LTOP. The feedwater pump motors are prevented from starting by the reactor water level high feedwater pump trip signal, which is present due Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 7 of 10) to the high reactor water levels required during pressure testing. The SSMP is a manually operated system that has no automatic initiation signals. Initiation of the SSMP is strictly governed by station procedures. During pressure testing, the reactor is in cold shutdown, and as a result, there is no steam available to drive the turbine driven RCIC or HPCI pumps. In addition, the HPCI and RCIC steam supply and pump discharge valves are closed and their associated motor operator breakers are opened in accordance with the test procedures.

The Standby Liquid Control (SLC) system is also a high pressure water source to the RPV. Similar to the SSMP, there are no automatic initiation signals associated with this system. Operation of the SLC system is strictly governed by station emergency operating procedures, and requires an operator to manually start the system from the main control room via a keylock switch manipulation.

The low pressure coolant sources include the Emergency Core Cooling Systems (ECCS)

(i.e., Core Spray and Residual Heat Removal) and the Condensate System. Operation of the ECCS systems is also governed by station emergency operating procedures.

Although certain automatic initiation signals are required operable during pressure testing, an ECCS actuation would occur only when reactor conditions warranted RPV injection (for example, during a low water level condition). In addition, the shutoff head of the ECCS pumps is relatively low and the injection valves are interlocked closed at pressures greater than approximately 300 psig. For these reasons, an LTOP event that would exceed the P/T curve limits due to an inadvertent ECCS injection is considered unlikely. As mentioned above, the Condensate System is normally used for RPV fill and is carefully governed by the test procedure.

During cold shutdown when the reactor head is tensioned, an LTOP event is prevented by the normal unit shutdown procedure, which requires the operator to place the RPV head vent valves in an open position when reactor coolant temperatures are below 190°F.

In addition to the procedural barriers, licensed operators are provided specific training on the P/T curves and requirements of the Technical Specifications. Simulator sessions are conducted which include plant heat-up and cool-down. Additionally, in response to industry operating experience, the operating training program is routinely evaluated and revised, as necessary, to reduce the possibility of events such as an LTOP.

Summary In summary, EGC has reviewed the methodology used in Reference 2, and considering DNPS and QCNPS plant specific materials properties, fluence, operational practices, and the provisions of Reference 3, the criteria established in Generic Letter 98-05 (i.e.,

Reference 4) are satisfied. Therefore, permanent relief is requested from the examination Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 8 of 10) requirements of the ASME Code, Section XI, Subarticle IWB-2500, Table IWB-2500-1, Examination Category B-A, Item Number B1.11 for RPV circumferential shell welds since the proposed alternative provides an acceptable level of quality and safety in accordance with 10CFR50.55a(a)(3)(i).

6. Duration of Proposed Alternative:

Permanent relief is requested for the additional license operating period requested in Reference 1 for DNPS, Units 2 and 3, and QCNPS, Units 1 and 2. Although Reference 4 permits BWR licensees to request permanent relief for the remaining term of the existing initial operating license, EGC has demonstrated that the criteria specified in Reference 4 will continue to be met for the entire additional operating period requested in Reference 1. Therefore, the requested duration of the proposed alternative is justified.

7. Precedents:

The NRC has previously approved similar relief for several nuclear power plants, including Dresden Nuclear Power Station Units 2 and 3 (i.e., Docket Numbers 50-237 and 50-249, TAC Nos. MA6228 and MA6229), and Susquehanna Steam Electric Station Units 1 and 2 (i.e., Docket Numbers 50-387 and 50-388, TAC Nos. MB0484 and MB0485). The relief request for Dresden Nuclear Power Station Units 2 and 3 was submitted to the NRC in Reference 10, and the NRC granted the relief in Reference 11.

The relief request for Susquehanna Steam Electric Station Units 1 and 2 was submitted to the NRC in Reference 12, and the NRC granted the relief in Reference 13.

8.

References:

1. Letter from J. A. Benjamin (Exelon Generation Company, LLC) to U. S. NRC, Application for Renewed Operating Licenses, dated January 3, 2003

2. BWRVIP-05, BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations (BWRVIP-05), dated September 28, 1995

3. Letter from G. C. Lainas (U. S. Nuclear Regulatory Commission) to C. Terry (BWRVIP), Final Safety Evaluation of the BWR Vessel and Internals Project BWRVIP-05 Report (TAC No. M93925), dated July 28, 1998 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 9 of 10)

4. NRC Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, dated November 10, 1998

5. NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluations, dated August 2000

6. Letter from S. A. Richards (U. S. NRC) to J. F. Klapproth (GE Nuclear Energy),

Safety Evaluation for NEDC-32983P, General Electric Methodology for Reactor Pressure Vessel Fast Neutron Flux Evaluation (TAC No. MA9891),

dated September 14, 2001

7. NRC Regulatory Guide 1.190, Calculational and Dosimetry Methods for Determining Pressure Vessel Neutron Fluence, dated March 2001

8. Letter from L. W. Rossbach (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), Dresden Nuclear Power Station, Units 2 and 3 - Issuance of Amendments for Extended Power Uprate (TAC Nos. MB0844 and MB0845),

dated December 21, 2001

9. Letter from S. N. Bailey (U. S. NRC) to O. D. Kingsley (Exelon Generation Company, LLC), Quad Cities Nuclear Power Station, Units 1 and 2 - Issuance of Amendments for Extended Power Uprate (TAC Nos. MB0842 and MB0843),

dated December 21, 2001

10. Letter from J. M. Heffley (Commonwealth Edison Company) to U. S. Nuclear Regulatory Commission, Relief Request for Alternative Weld Examination of Circumferential Reactor Pressure Vessel Shell Welds, dated July 26, 1999

11. Letter from A. J. Mendiola (U. S. Nuclear Regulatory Commission) to O. D.

Kingsley (Commonwealth Edison Company), Dresden - Authorization for Proposed Alternative Reactor Pressure Vessel Circumferential Weld Examinations (TAC Nos. MA6228 and MA6229), dated February 25, 2000

12. Letter from R. G. Byram (PPL Susquehanna, LLC) to U. S. Nuclear Regulatory Commission, Request for Alternative to 10CFR50.55a Examination Requirements of Examination Category B1.11 Reactor Pressure Vessel Welds for PPL Susquehanna LLC Units 1 and 2 PLA-5251, dated November 7, 2000 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-05 Revision 1 (Page 10 of 10)

13. Letter from M. Gamberoni (U. S. Nuclear Regulatory Commission) to R. G.

Byram (PPL Susquehanna, LLC), Relief Request No. 22 (RR-22) from American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section XI, Susquehanna Steam Electric Station Units 1 and 2 (TAC Nos.

MB0484 and MB0485), dated February 28, 2001 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 1 of 15)

Request for Relief for the Use of BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 1

Reference:

IWB-2500, Table IWB-2500-1 Examination Category: B-N-1 and B-N-2 Item Number: B13.10, B13.20, B13.30, and B13.40

Description:

Use of BWRVIP Guidelines in Lieu of Specific ASME Code Requirements on Reactor Pressure Vessel Internals and Components Inspection Component Numbers: Vessel Interior, Interior Attachments within Beltline Region, Interior Attachments beyond Beltline Region, and Core Support Structure

2. Applicable Code Edition and Addenda:

The Inservice Inspection Program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirements:

ASME Section XI requires the examination of components within the Reactor Pressure Vessel. These examinations are included in Table IWB-2500-1, Examination Categories B-N-1 and B-N-2, and identified with the following item numbers:

B13.10 Examine accessible areas of the reactor vessel interior each period by the VT-3 method (B-N-1).

B13.20 Examine interior attachment welds within the beltline region each interval by the VT-1 method (B-N-2).

B13.30 Examine interior attachment welds beyond the beltline region each interval by the VT-3 method (B-N-2).

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 2 of 15)

B13.40 Examine surfaces of the welded core support structure each interval by the VT-3 method (B-N-2).

These examinations are performed to assess the structural integrity of components within the boiling water reactor pressure vessel.

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested for the proposed alternative to the Code requirements provided above on the basis that the use of the BWRVIP guidelines discussed below will provide an acceptable level of quality and safety.

The BWRVIP Inspection and Evaluation (I&E) guidelines have recommended aggressive specific inspection by BWR operators to completely identify material condition issues with BWR components. A wealth of inspection data has been gathered during these inspections across the BWR industry. I&E guidelines focus on specific and susceptible components, specify appropriate inspection methods capable of identifying real anticipated degradation mechanisms, and require re-examination at conservative intervals. In contrast, the code inspection requirements were prepared before the BWRVIP initiative and have not evolved with BWR inspection experience.

Use of this proposed alternative will maintain an adequate level of quality and safety and avoid unnecessary inspections.

5. Proposed Alternative and Basis for Use:

In lieu of the requirements of ASME Section XI, the proposed alternative is detailed in attached Table 1 for Examination Category B-N-1 and B-N-2.

Quad Cities Nuclear Power Station Units 1 and 2 will satisfy the Examination Category B-N-1 and B-N-2 requirements as described in Table 1 in accordance with BWRVIP guideline requirements. This relief request proposes to utilize the identified BWRVIP guidelines in lieu of the associated Code requirements, including examination method, examination volume, frequency, training, successive and additional examinations, flaw evaluations, and reporting.

Not all the components addressed by these guidelines are code components. The following guidelines are applicable to this Relief Request:

- BWRVIP-03, BWR Vessel and Internals Project, Reactor Pressure Vessel and Internals Examination Guidelines Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 3 of 15)

- BWRVIP-18, Rev. 1, BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines

- BWRVIP-25, BWR Core Plate Inspection and Flaw Evaluation Guidelines

- BWRVIP-26-A, BWR Top Guide Inspection and Flaw Evaluation Guidelines

- BWRVIP-27-A, BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines

- BWRVIP-38, BWR Shroud Support Inspection and Flaw Evaluation Guidelines

- BWRVIP-47-A, BWR Lower Plenum Inspection and Flaw Evaluation Guidelines

- BWRVIP-48-A, Vessel ID Attachment Weld Inspection and Flaw Evaluation Guidelines

- BWRVIP-76, Rev. 1, BWR Core Shroud Inspection and Flaw Evaluation Guidelines

- BWRVIP-94, Rev. 2, BWR Vessel and Internals Project Program Implementation Guide

- BWRVIP-138, Rev. 1, Updated Jet Pump Beam Inspection and Flaw Evaluation

- BWRVIP-183, BWR Vessel and Internals Project, Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines Inspection services, by an Authorized Inspection Agency, will be applied to the proposed alternative actions of this relief request.

BWRs now examine reactor internals in accordance with BWRVIP guidelines. These guidelines have been written to address the safety significant vessel internal components and to examine and evaluate the examination results for these components using appropriate methods and reexamination frequencies. The BWRVIP has established a reporting protocol for examination results and deviations. The NRC has agreed with the BWRVIP approach as documented in References 1 through 10. Therefore, use of these guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

As additional justification, Attachment 1 (Comparison of Code Examination Requirements to BWRVIP Examination Requirements) provides specific examples which compare the inspection requirements of ASME Code Item Numbers B13.10, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the inspection requirements in the BWRVIP documents. Specific BWRVIP documents are provided as examples. This comparison also includes a discussion of the inspection methods. These comparisons demonstrate that use of these guidelines, as an alternative to the subject Code requirements, provides an acceptable level of quality and safety and will not adversely impact the health and safety of the public.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 4 of 15)

From May 2006 (Unit 1), and April 2006 (Unit 2), and since the conditional authorization of the Fourth Interval BWRVIP Relief Request, Revision 1 on April 30, 2008, no indications or flaws have been found on the BWRVIP subject welds in subsequent examinations.

Table 1 compares present ASME Examination Category B-N-1 and B-N-2 requirements with the above current BWRVIP guideline requirements, as applicable, to Quad Cities Nuclear Power Station. Therefore, Table 1 only represents a current comparison. Any deviations from the BWRVIP guidelines referenced within this relief request for the duration of the proposed alternative will be appropriately documented and communicated to the NRC, per the BWRVIP Deviation Disposition Process.

Also, the reactor vessel internals inspection program at Quad Cities Nuclear Power Station has been developed and implemented to satisfy the requirements of BWRVIP-94.

It is recognized that the BWRVIP executive committee periodically revises the BWRVIP guidelines to include enhancements in inspection techniques and flaw evaluation methodologies. Where the revised version of a BWRVIP inspection guideline continues to also meet the requirements of the version of the BWRVIP inspection guideline that forms the safety basis for an NRC-authorized proposed alternative to the requirements of 10CFR50.55a, it may be implemented. Otherwise, the revised guidelines will only be implemented after NRC approval of the revised BWRVIP guidelines or a plant-specific request for relief has been approved.

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

Exelon/Amergen BWR fleet as discussed in Reference 11. The Exelon/Amergen Fleetwide Relief Request for BWRVIP was authorized conditionally per SE dated April 30, 2008. The Fifth Inspection Interval Relief Request utilizes a similar approach that was previously approved.

Perry Nuclear Power Plant, Unit No. 1 as discussed in Reference 12. Perry Nuclear Power Plant, Unit No. 1 was authorized per NRC SE dated January 31, 2012.

Fermi 2 as discussed in Reference 13. Fermi 2 was authorized per NRC SE dated February 17, 2012.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 5 of 15)

8.

References:

1. Letter NRC to BWRVIP, Final Safety Evaluation for Electric Power Research Institute Boiling Water Reactor Vessel and Internals Project Technical Report 1016568, BWRVIP-18, Revision 1: BWR Core Spray Internals Inspection and Flaw Evaluation Guidelines (TAC No. ME2189), dated January 30, 2012 (ML113620684)

2. Letter NRC to BWRVIP, NRC Approval Letter of BWRVIP-26-A, BWR Vessel and Internals Project Boiling Water Reactor Top Guide Inspection and Flaw Evaluation Guidelines, dated September 9, 2005

3. Letter NRC to BWRVIP, Proprietary Version of NRC Staff Review of BWRVIP-27-A, BWR Standby Liquid Control System/Core Plate P Inspection and Flaw Evaluation Guidelines, dated June 10, 2004

4. Letter NRC to BWRVIP, Final Safety Evaluation of the BWR Vessel and Internals Project, BWR Shroud Support Inspection and Flaw Evaluation Guidelines (BWRVIP-38), EPRI Report TR-108823 (TAC No. M99638), dated July 24, 2000

5. Letter NRC to BWRVIP, NRC Approval Letter of BWRVIP-47-A, BWR Vessel and Internals Project Boiling Water Reactor Lower Plenum Inspection and Flaw Evaluation Guidelines, dated September 9, 2005

6. Letter NRC to BWRVIP, NRC Approval Letter of BWRVIP-48-A, BWR Vessel and Internals Project Vessel ID Attachment Weld Inspection and Flaw Evaluation Guideline, dated July 25, 2005

7. BWRVIP-76NP, Rev. 1: BWR Vessel and Internals Project BWR Core Shroud Inspection and Flaw Evaluation Guidelines, dated May 2011 (ML11195A182)

8. Letter from Chairman, BWR Vessel and Internals Project to NRC, Project No.

704 - BWRVIP Program Implementation Guide (BWRVIP-94NP, Revision 2),

dated September 22, 2011 (ML11271A058)

9. BWRVIP-138NP, Revision 1: BWR Vessel and Internals Project, Updated Jet Pump Beam Inspection and Flaw Evaluation Guidelines, dated January 2009 (ML090760986)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 6 of 15)

10. BWRVIP-183NP: BWR Vessel and Internals Project Top Guide Grid Beam Inspection and Flaw Evaluation Guidelines, dated December 7, 2007 (ML080220433)

11. Letter from NRC to Exelon/Amergen, Clinton Power Station, Unit No. 1; Dresden Nuclear Power Station, Units 2 and 3; LaSalle County Station, Units 1 and 2; Limerick Generating Station, Units 1 and 2; Oyster Creek Nuclear Generating Station; Peach Bottom Atomic Power Station, Units 2 and 3; and Quad Cities Nuclear Power Station, Units 1 and 2 - Relief Request to Use Boiling Water Reactor Vessel and Internals Project Guidelines in Lieu of Specific ASME Code Requirements (TAC Nos. MD5352 through MD5363), dated April 30, 2008 (ML080980311)

12. Letter from NRC to FirstEnergy Nuclear Operating Company (Perry Nuclear Power Plant, Unit No. 1), Perry Nuclear Power Plant, Unit No. 1, RE: Safety Evaluation in Support of 10CFR50.55a Requests for the Third 10-Year In-Service Inspection Interval (TAC Nos. ME5373, ME5376, ME5377, ME5379, and ME5380), dated January 31, 2012 (ML120180372)

13. Letter from NRC to Detroit Edison Company (Fermi 2), Fermi 2- Evaluation of Applicable 10-Year Interval Inservice Inspection Relief Request - Use of Boiling Water Reactor Vessel and Internals Project (BWRVIP) Guidelines in Lieu of Specific ASME Code Requirements (TAC No. ME6765), dated February 17, 2012 (ML120370286)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 7 of 15)

TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements(1)

ASME Item BWRVIP Number, ASME Exam ASME ASME Authorized BWRVIP BWRVIP Component Exam Table IWB- Scope Exam Frequency Alternative Exam Frequency Scope 2500-1 B13.10 Accessible VT-3 Each BWRVIP Overview examinations of components during Reactor Areas period R1, 25, 26-A, BWRVIP examinations satisfy Code VT-3 visual Vessel 27-A, 38, 47-A, inspection requirements.

Interior 48-A, 76-R1, and 138-R1 B13.20 Jet Pump Riser Braces Accessible VT-1 Each BWRVIP-48-A, Riser Brace EVT-1 100% in first 12 Interior Welds 10-year Table 3-2 Attachment years (with 50% to Attachments Interval be inspected in the Within first 6 years); 25%

Beltline during each Region subsequent 6 years Lower Surveillance BWRVIP-48-A, Bracket VT-1 Each 10-year Specimen Holder Table 3-2 Attachment Interval Brackets B13.30 Guide Rod Brackets Accessible VT-3 Each BWRVIP-48-A, Bracket VT-3 Each 10-year Interior Welds 10-year Table 3-2 Attachment Interval Attachments Steam Dryer Support Interval BWRVIP-48-A, Bracket VT-3(5) Each 10-year Beyond Brackets Table 3-2 Attachment Interval Beltline Feedwater Sparger BWRVIP-48-A, Bracket VT-3(5) Each 10-year Brackets Table 3-2 Attachment Interval Core Spray Piping BWRVIP-48-A, Bracket EVT-1 Every 4 Refueling Brackets Table 3-2 Attachment Cycles Upper Surveillance BWRVIP-48-A, Bracket VT-3 Each 10-year Specimen Holder Table 3-2 Attachment Interval Brackets Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 8 of 15)

TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements(1)

ASME Item BWRVIP Number, ASME Exam ASME ASME Authorized BWRVIP BWRVIP Component Exam Table IWB- Scope Exam Frequency Alternative Exam Frequency Scope 2500-1 Shroud Support (Weld BWRVIP-38, Weld H9(2) EVT-1 or Maximum of 6 H9) 3.1.3.2, UT years for one-sided Figures 3-2 and EVT-1, Maximum 3-5 of 10 years for UT Shroud Support Legs (Rarely BWRVIP-38, Weld H12 Per When accessible (Weld H12) Accessible) 3.2.3 BWRVIP-38 NRC SE (7-24-2000),

inspect with appropriate method(4)

B13.40 Shroud Support (Weld Accessible VT-3 Each BWRVIP-38, Shroud EVT-1 or Based on as found Welded Core H10) Surfaces 10-year 3.1.3.2, Support UT conditions, to a Support Interval Figure 3-2 and (Weld H10) maximum 6 years Structure 3-5 and Leg for one-sided Welds EVT-1, 10 years for UT where accessible Shroud Vertical Welds BWRVIP Vertical and EVT-1 or Maximum 6 years R1, Ring UT for one-sided 3.3 Segment EVT-1, 10 years for Figure 3-1 and Welds as UT 3-3 applicable Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 9 of 15)

TABLE 1 Comparison of ASME Examination Category B-N-1 and B-N-2 Requirements With BWRVIP Guidance Requirements(1)

ASME Item BWRVIP Number, ASME Exam ASME ASME Authorized BWRVIP BWRVIP Component Exam Table IWB- Scope Exam Frequency Alternative Exam Frequency Scope 2500-1 Shroud Repairs(3) BWRVIP Tie-Rod VT-3 Per designer R1, Repair recommendations Section 3.5 per BWRVIP R1 NOTES:

1) This Table provides only an overview of the requirements. For more details, refer to ASME Section XI, Table IWB-2500-1, and the appropriate BWRVIP document.

2) In accordance with Appendix A of BWRVIP-38, a site specific evaluation will determine the minimum required weld length to be examined.

3) Shroud repairs are currently installed on both units at Quad Cities Nuclear Power Station.

4) When inspection tooling and methodologies are available, they will be utilized to establish a baseline inspection of these welds.

5) Quad Cities Nuclear Power Station does not have furnace-sensitized stainless steel (E308/309 or E308L/309L) or Alloy-182 material, therefore the Code required VT-3 visual examination will be performed in accordance with BWRVIP-48-A, Table 3-2.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 10 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements The following discussion provides a comparison of the examination requirements provided in ASME Code Item Numbers B13.10, B13.20, B13.30, and B13.40 in Table IWB-2500-1, to the examination requirements in the BWRVIP guidelines. Specific BWRVIP guidelines are provided as examples for comparisons. This comparison also includes a discussion of the examination methods.

1. Code Requirement - B13.10 - Reactor Vessel Interior Accessible Areas (B-N-1)

The ASME Section XI Code requires a VT-3 visual examination of reactor vessel accessible areas, which are defined as the spaces above and below the core made accessible during normal refueling outages. The frequency of these examinations is specified as the first refueling outage, and at intervals of approximately 3 years, during the First Inspection Interval, and each period during each successive 10-year Inspection Interval. Typically, these examinations are performed every other refueling outage of the Inspection Interval. This examination requirement is a non-specific requirement that is a departure from the traditional ASME Section XI examinations of welds and surfaces. As such, this requirement has been interpreted and satisfied differently across the domestic fleet. The purpose of the examination is to identify relevant conditions such as distortion or displacement of parts; loose, missing, or fractured fasteners; foreign material, corrosion, erosion, or accumulation of corrosion products; wear; and structural degradation.

Portions of the various examinations required by the applicable BWRVIP guidelines require access to accessible areas of the reactor vessel during each refueling outage.

Examination of core spray piping and spargers (BWRVIP-18-R1), top guide (BWRVIP-26-A), jet pump welds and components (BWRVIP-138-R1), interior attachments (BWRVIP-48-A), core shroud welds (BWRVIP-76-R1), shroud support (BWRVIP-38), and lower plenum components (BWRVIP-47-A) provides such access.

Locating and examining specific welds and components within the reactor vessel areas above, below (if accessible), and surrounding the core (annulus area) entails access by remote camera systems that essentially perform equivalent VT-3 visual examination of these areas or spaces as the specific weld or component examinations are performed.

This provides an equivalent method of visual examination on a more frequent basis than that required by the ASME Section XI Code. Evidence of wear, structural degradation, loose, missing, or displaced parts, foreign materials, and corrosion product buildup can be, and has been observed during the course of implementing these BWRVIP examination requirements. Therefore, the specified BWRVIP Guideline requirements meet or exceed the subject Code requirements for examination method and frequency of Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 11 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements the interior of the reactor vessel. Accordingly, these BWRVIP examination requirements provide an acceptable level of quality and safety as compared to the subject Code requirements.

2. Code Requirement - B13.20 - Interior Attachments Within the Beltline (B-N-2)

The ASME Section XI Code requires a VT-1 visual examination of accessible reactor interior surface attachment welds within the beltline each 10-year interval. In the model 3 boiling water reactor, this includes the jet pump riser brace welds-to-vessel wall and the lower surveillance specimen support bracket welds-to-vessel wall. In comparison, the BWRVIP requires the same examination method and frequency for the lower surveillance specimen support bracket welds, and requires an EVT-1 visual examination on the remaining attachment welds in the beltline region in the first 12 years, and then 25% during each subsequent 6 years.

The jet pump riser brace examination requirements are provided below to show a comparison between the Code and the BWRVIP examination requirements.

Comparison to BWRVIP Requirements - Jet Pump Riser Braces (BWRVIP-138, Rev. 1 and BWRVIP-48-A)

The ASME Code requires a 100% VT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds each 10-year interval.

The BWRVIP requires an EVT-1 visual examination of the jet pump riser brace-to-reactor vessel wall pad welds the first 12 years and then 25% during each subsequent 6 years.

BWRVIP-48-A specifically defines the susceptible regions of the attachment that are to be examined.

The Code VT-1 visual examination is conducted to detect discontinuities and imperfections on the surfaces of components, including such conditions as cracks, wear, corrosion, or erosion. The BWRVIP enhanced VT-1 (EVT-1) visual examination is conducted to detect discontinuities and imperfections on the surface of components and is additionally specified to detect potentially very tight cracks characteristic of fatigue and inter-granular stress corrosion cracking (IGSCC), the relevant degradation mechanisms for these components. General wear, corrosion, or erosion although generally not a concern for inherently tough, corrosion resistant stainless steel material, would also be detected during the process of performing a BWRVIP EVT-1 visual examination.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 12 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements The Code VT-1 visual examination method requires that at a maximum distance of 2 feet or a letter character with a height of 0.044 inches can be read. The BWRVIP EVT-1 visual examination method requires resolution of 0.044 inch characters on the examination surface. BWRVIP-48-A includes a diagram and prescribes examination for the configuration of this plant.

The calibration standards used for BWRVIP EVT-1 visual examinations utilize the Code characters, thus assuring at least equivalent resolution compared to the ASME Code.

Although the BWRVIP examination may be less frequent, it is a more comprehensive method. Therefore, the enhanced flaw detection capability of an EVT-1 visual examination, with a less frequent examination schedule provides an acceptable level of quality and safety to that provided by the ASME Code.

3. Code Requirement - B13.30 - Interior Attachment Beyond the Beltline Region (B-N-2)

The ASME Section XI Code requires a VT-3 visual examination of accessible reactor interior surface attachment welds beyond the beltline each 10-year interval. In the BWR/3 model, this includes the core spray piping primary and supplemental support bracket welds-to-vessel wall, the upper surveillance specimen support bracket welds-to-vessel wall, the feedwater sparger support bracket welds-to-reactor vessel wall, the steam dryer support welds-to-reactor vessel wall, the guide rod support bracket weld-to-reactor vessel wall, and the shroud support plate-to-vessel weld. BWRVIP-48-A requires as a minimum the same VT-3 visual examination method as the Code for some of the interior attachment welds beyond the beltline region, and in some cases specifies an enhanced visual examination technique EVT-1 visual examination for these welds. For those interior attachment welds that have the same VT-3 method of visual examination, the same scope of examination (accessible welds), the same examination frequency (each 10-year interval) and ASME Section XI flaw evaluation criteria, the level of quality and safety provided by the BWRVIP requirements are equivalent to that provided by the ASME Code.

For the core spray primary and secondary support bracket attachment welds, the steam dryer support bracket attachment welds, the feedwater sparger support bracket attachment welds, and the shroud support plate-to-vessel welds, as applicable, the BWRVIP guidelines require an EVT-1 visual examination at the same frequency as the Code. Therefore, the BWRVIP requirements provide the same level of quality and safety to that provided by the ASME Code.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 13 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements The core spray piping bracket-to-vessel attachment weld is used as an example for comparison between the Code and BWRVIP examination requirements as discussed below.

Comparison to BWRVIP Requirements - Core Spray piping Bracket Welds (BWRVIP-48-A)

The Code examination requirement is a VT-3 visual examination of each weld every 10 years.

The BWRVIP examination requirement is an EVT-1 visual examination for the Core Spray piping bracket attachment welds with each weld examined every four cycles (8 years for units with a two year fuel cycle).

The BWRVIP examination method EVT-1 visual examination has superior flaw detection and sizing capability, and the same flaw evaluation criteria are used.

The Code VT-3 visual examination is conducted to detect component structural integrity by ensuring the components general condition is acceptable. An enhanced EVT-1 visual examination is conducted to detect discontinuities and imperfections on the examination surfaces, including such conditions as tight cracks caused by IGSCC or fatigue, the relevant degradation mechanisms for BWR internal attachments.

Therefore, with the EVT-1 visual examination method, the same examination scope (accessible welds), the same examination frequency, the same flaw evaluation criteria (Section XI), the level of quality and safety required by the BWRVIP criteria is superior than that required by the Code.

4. Code Requirement - B13.40 - Welded Core Support Structures (B-N-2)

The ASME Code requires a VT-3 visual examination of accessible surfaces of the welded core support structure each 10-year interval. In the boiling water reactor, the welded core support structure has primarily been considered the shroud support structure, including the shroud. In later designs, the shroud itself is considered part of the welded core support structure. Historically, this requirement has been interpreted and satisfied differently across the industry. The proposed alternate examination replaces this ASME requirement with specific BWRVIP guidelines that examine susceptible locations for known relevant degradation mechanisms.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 14 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements Comparison to BWRVIP Requirements - Shroud Supports (BWRVIP-38)

The Code requires a VT-3 visual examination of accessible surfaces each 10-year interval.

The BWRVIP requires either an enhanced visual examination technique (EVT-1) every 6 years or volumetric examination (UT) every 10 years as compared to the Code requirement (VT-3). (Only 10% of the weld is required to be examined.)

BWRVIP recommended examinations of welded core support structures are focused on the known susceptible areas of this structure, including the welds and associated weld heat affected zones. In many locations, the BWRVIP guidelines require a volumetric examination of the susceptible welds at a frequency identical to the Code requirement.

Shroud repair tie-rods have been installed at Quad Cities Nuclear Power Station; therefore, the BWRVIP referenced examinations are the same as the Code requirements.

Shroud repair tie-rod examinations are recommended in BWRVIP-76-R1, and have the same basic VT-3 method of visual examination, the same scope of examination (accessible surfaces), the same examination frequency (each 10-year interval) and the same flaw evaluation criteria. Therefore, the BWRVIP requirements provide a level of quality and safety equivalent to that provided by the ASME Code.

For other welded core support structure components, the BWRVIP requires an EVT-1 visual examination or UT of core support structures. The core shroud is used as an example for comparison between the Code and BWRVIP examination requirements as shown below.

Comparison to BWRVIP Requirements - BWR Core Shroud Examination and Flaw Evaluation Guideline (BWRVIP-76-A, Rev. 1)

The Code requires a VT-3 visual examination of accessible surfaces each 10-year interval.

The BWRVIP requires an EVT-1 visual examination from the inside and outside surface where accessible or ultrasonic examination of each core shroud circumferential weld that has not been structurally replaced with a shroud repair at a calculated end of interval (EOI) that will vary depending upon the amount of flaws present, but not to exceed ten years.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-06 Revision 0 (Page 15 of 15)

ATTACHMENT 1 Comparison of Code Examination Requirements to BWRVIP Examination Requirements The BWRVIP recommended examinations specify locations that are known to be vulnerable to BWR relevant degradation mechanisms rather than accessible surfaces.

The BWRVIP examination methods (EVT-1 or UT) are superior to the Code required VT-3 visual examination for flaw detection and characterization. The superior flaw detection and characterization capability and the comparable flaw evaluation criteria, results in the BWRVIP criteria providing a level of quality and safety equivalent to or superior to that required by the Code requirements.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-07 (Page 1 of 1)

(RESERVED)

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-08 Revision 0 (Page 1 of 4)

Request for Relief for Use of ASME Code Case N-789, Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping for Raw Water Service In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

- - NOTE 1 ***

Quad Cities Nuclear Power Station Units 1 and 2 Fifth Interval Relief Request I5R-08 is provided for completeness and has no technical changes from the previously approved Relief Request I4R-18. This relief request was submitted as an Exelon Fleet request under cover letter RS-11-182 on October 7, 2011, with supplemental letters dated November 10, 2011, and February 13, 2012, and was authorized during the Fourth Interval under NRC SE dated May 10, 2012, which for Quad Cities Nuclear Power Station covered both the Fourth and Fifth Intervals as specified in Table 1 of the NRC SE.

No technical changes to the actual approved relief request have been made in the Fifth Interval ISI Program Plan and no further or revised authorization is required.

- - NOTE 2 ***

Per NRC SE dated May 10, 2012, the NRC staff concluded that the proposed alternative failed to meet the regulatory standard of 10CFR50.55a(a)(3)(i). However, the NRC staff further concluded that the proposed alternative provided reasonable assurance of structural integrity and leak tightness of the ASME Section XI, Class 2 and 3, moderate energy carbon steel raw water piping and that complying with the specified requirement would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. Accordingly, the NRC staff concluded that the licensee had adequately addressed all of the regulatory requirements set forth in 10CFR50.55a(a)(3)(ii), and as such, authorized the relief request.

1. ASME Code Component(s) Affected:

All Class 2 and 3 moderate energy carbon steel raw water piping systems. Raw water is defined as water such as from a river, lake, or well or brackish/salt water - used in plant equipment, area coolers, and heat exchangers. In many plants it is referred to as Service Water. This Code Case applies to Class 2 and 3 moderate energy (i.e., less than or equal to 200°F (93°C) and less than or equal to 275 psig (1.9 MPa) maximum operating conditions) carbon steel raw water piping.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-08 Revision 0 (Page 2 of 4)

2. Applicable Code Edition and Addenda:

PLANT INTERVAL EDITION START END Braidwood Station, Units 1 Third 2001 Edition, through July 29, 2008 July 28, 2018 and 2 the 2003 Addenda October 17, 2008 October 16, 2018 Byron Station, Units 1 and 2 Third 2001 Edition, through January 16, 2006 July 15, 2016 the 2003 Addenda Clinton Power Station Third 2004 Edition July 1, 2010 June 30, 2020 Dresden Nuclear Power Fourth 1995 Edition, through January 20, 2003 January 19, 2013 Station, Units 2 and 3 the 1996 Addenda Dresden Nuclear Power Fifth 2007 Edition, through January 20, 2013 January 19, 2023 Station, Units 2 and 3 the 2008 Addenda LaSalle County Station, Third 2001 Edition, through October 1, 2007 September 30, Units 1 and 2 the 2003 Addenda 2017 Limerick Generating Third 2001 Edition, through February 1, 2007 January 31, 2017 Station, Units 1 and 2 the 2003 Addenda Oyster Creek Nuclear Fourth 1995 Edition, through October 15, 2002 October 14, 2012 Generating Station the 1996 Addenda Oyster Creek Nuclear Fifth 2007 Edition, through October 15, 2012 October 14, 2022 Generating Station the 2008 Addenda Peach Bottom Atomic Fourth 2001 Edition, through November 5, November 4, 2018 Power Station, Units 2 and 3 the 2003 Addenda 2008 Quad Cities Nuclear Power Fourth 1995 Edition, through March 10, 2003 April 1, 2013 Station, Units 1 and 2 the 1996 Addenda Quad Cities Nuclear Power Fifth 2007 Edition, through April 2, 2013 April 1, 2023 Station, Units 1 and 2 the 2008 Addenda Three Mile Island Nuclear Fifth 2004 Edition April 20, 2011 April 19, 2022 Station, Unit 1

3. Applicable Code Requirement:

ASME Code, Section Xl, IWA-4400 of the 1995 Edition through the 1996 Addenda, 2001 Edition through the 2003 Addenda, 2004 Edition, and 2007 Edition through the 2008 Addenda provides requirements for welding, brazing, metal removal, and installation of repair/replacement activities.

4. Reason for Request:

In accordance with 10CFR50.55a(a)(3)(i), Exelon Generation Company, LLC (Exelon) is requesting a proposed alternative from the requirement for replacement or internal weld repair of wall thinning conditions resulting from degradation in Class 2 and 3 moderate energy carbon steel raw water piping systems in accordance with IWA-4000. Such degradation may be the result of mechanisms such as erosion, corrosion, cavitation, and Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-08 Revision 0 (Page 3 of 4) pitting - but excluded are conditions involving flow-accelerated corrosion (FAC),

corrosion-assisted cracking, or any other form of cracking. IWA-4000 requires repair or replacement in accordance with the Owners Requirements and the original or later Construction Code. Other alternative repair or evaluation methods are not always practicable because of wall thinness and/or moisture issues.

The primary reason for this request is to permit installation of a technically sound temporary repair to provide adequate time for evaluation, design, material procurement, planning and scheduling of appropriate permanent repair or replacement of the defective piping, considering the impact on system availability, maintenance rule applicability, and availability of replacement materials.

5. Proposed Alternative and Basis for Use:

In accordance with 10CFR50.55a(a)(3)(i), Exelon proposes to implement the requirements of ASME Code Case N-789 (Alternative Requirements for Pad Reinforcement of Class 2 and 3 Moderate-Energy Carbon Steel Piping for Raw Water Service, Section XI, Division 1) as a temporary repair of degradation in Class 2 and 3 moderate energy raw water piping systems resulting from mechanisms such as erosion, corrosion, cavitation, or pitting, but excluding conditions involving flow-accelerated corrosion (FAC), corrosion-assisted cracking, or any other form of cracking. These types of defects are typically identified by small leaks in the piping system or by pre-emptive non-code required examinations performed to monitor the degradation mechanisms.

ASME Code Case N-789, which is included as part of this relief request, is attached.

(Note: ASME Code Case N-789 is not attached to this document)

The alternative repair technique described in ASME Code Case N-789 involves the application of a metal reinforcing pad welded to the exterior of the piping system, which reinforces the weakened area and restores pressure integrity. This repair technique will be utilized when it is determined that this temporary repair method is suitable for the particular defect or degradation being resolved.

The Code Case requires that the cause of the degradation be determined, and that the extent and rate of degradation in the piping be evaluated to ensure that there are no other unacceptable locations within the surrounding area that could affect the integrity of the repaired piping. The area of evaluation will be dependent on the degradation mechanism present. A baseline thickness examination will be performed for a completed structural pad, attachment welds, and surrounding area, followed by monthly thickness monitoring for the first three months, with subsequent frequency based on the results of this monitoring, but at a minimum of quarterly. Areas containing pressure pads shall be visually observed at least once per month to monitor for evidence of leakage. If the areas Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-08 Revision 0 (Page 4 of 4) containing pressure pads are not accessible for direct observation, then monitoring will be accomplished by visual assessment of surrounding areas or ground surface areas above pressure pads on buried piping, or monitoring of leakage collection systems, if available.

The repair will be considered to have a maximum service life of the time until the next refueling outage, when a permanent repair or replacement must be performed.

Additional requirements for design of reinforcement pads, installation, examination, pressure testing, and inservice monitoring are provided in ASME Code Case N-789.

Based on the above justification, the use of ASME Code Case N-789 as a proposed alternative to the requirements of ASME Section XI will provide an acceptable level of quality and safety.

All other ASME Section XI requirements for which relief was not specifically requested and authorized by the NRC staff will remain applicable including third party review by the Authorized Nuclear Inservice Inspector.

ASME Code Case N-789 was approved by the ASME Board on Nuclear Codes and Standards on June 25, 2011; however, it has not been incorporated into NRC Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1, and thus is not available for application at nuclear power plants without specific NRC approval. Therefore, Exelon requests use of this alternative repair technique described in the Code Case via this relief request.

6. Duration of Proposed Alternative:

The proposed alternative is for use of the Code Case for the remainder of each plants ten (10) year inspection interval as specified in Section 2.

Any reinforcing pads installed before the end of the ten-year inservice inspection interval will be removed during the next refueling outage, even if that refueling outage occurs after the end of the ten-year interval.

7. Precedents:

A similar repair relief request (RR-3-43) was approved for Indian Point Nuclear Generating Unit No. 3 per NRC SE dated February 22, 2008.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-09 Revision 0 (Page 1 of 2)

Request for Relief for Expanded Applicability for Use of ASME Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 1, 2, 3, and MC

Reference:

IWA-2441(b) and ASME Code Case N-532-4 Examination Category: NA Item Number: NA

Description:

Expanded Applicability for Use of ASME Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission Component Number: NA

2. Applicable Code Edition and Addenda:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

IWA-2441(b) requires Code Cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission, provides requirements that may be used to document repair/replacement activities.

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

On April 2, 2013, Quad Cities Nuclear Power Station Units 1 and 2 will start its Fifth Ten-Year Interval ISI Program under the requirements of the 2007 Edition through the Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-09 Revision 0 (Page 2 of 2) 2008 Addenda of ASME Section XI. When implementing this edition of ASME Section XI, Paragraph IWA-2441(b) requires code cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-532-4 has an applicability limited up to the 2004 Edition through the 2005 Addenda, which is identified in the latest applicability index for ASME Section XI Code Cases. Since ASME Code Case N-532-4 only applies up to the 2004 Edition through the 2005 Addenda, Paragraph IWA-2441(b) does not allow the use of ASME Code Case N-532-4 for the Quad Cities Nuclear Power Station Fifth Ten-Year Interval ISI Program.

5. Proposed Alternative and Basis for Use:

Quad Cities Nuclear Power Station requests the applicability of ASME Code Case N-532-4 be extended to the 2007 Edition through the 2008 Addenda for use in the plants Fifth Interval ISI Program. The NRC has accepted the use of ASME Code Case N-532-4 as an acceptable method for repair/replacement activity documentation requirements and inservice summary report preparation and submission in the latest revision of Regulatory Guide 1.147, Revision 16.

No technical changes to ASME Code Case N-532-4 are being proposed in this relief request. This relief request is being submitted to correct a timing situation, which has resulted from the application of the 2007 Edition through the 2008 Addenda of ASME Section XI for Quad Cities Nuclear Power Station. Since no technical change is proposed in this relief request, Quad Cities Nuclear Power Station considers that this alternative provides an acceptable level of quality and safety, and is consistent with provisions of 10CFR50.55a(a)(3)(i).

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

None.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-10 Revision 0 (Page 1 of 2)

Request for Relief for Expanded Applicability for Use of ASME Code Case N-661-1, Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service In Accordance with 10CFR50.55a(a)(3)(i)

Alternative Provides Acceptable Level of Quality and Safety

1. ASME Code Component(s) Affected:

Code Class: 2 and 3

Reference:

IWA-2441(b) and ASME Code Case N-661-1 Examination Category: NA Item Number: NA

Description:

Expanded Applicability for Use of ASME Code Case N-661-1, Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service Component Number: Class 2 and 3 Piping

2. Applicable Code Edition and Addenda:

The Inservice Inspection program is based on the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel (B&PV) Code, Section XI, 2007 Edition through the 2008 Addenda.

3. Applicable Code Requirement:

IWA-2441(b) requires Code Cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-661-1, Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service, provides requirements that may be used to restore wall thickness for raw water piping systems that have experienced internal wall thinning.

4. Reason for Request:

Pursuant to 10CFR50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative will provide an acceptable level of quality and safety.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 ISI Program Plan - 5th Interval 10CFR50.55a RELIEF REQUEST I5R-10 Revision 0 (Page 2 of 2)

On April 2, 2013, Quad Cities Nuclear Power Station Units 1 and 2 will start its Fifth Ten-Year Interval ISI Program under the requirements of the 2007 Edition through the 2008 Addenda of ASME Section XI. When implementing this edition of ASME Section XI, Paragraph IWA-2441(b) requires code cases be applicable to the Edition and Addenda specified in the Inspection Plan.

ASME Code Case N-661-1 has an applicability limited up to the 2004 Edition through the 2005 Addenda, which is identified in the latest applicability index for ASME Section XI Code Cases. Since ASME Code Case N-661-1 only applies up to the 2004 Edition through the 2005 Addenda, Paragraph IWA-2441(b) does not allow the use of ASME Code Case N-661-1 for the Quad Cities Nuclear Power Station Fifth Ten-Year Interval ISI Program.

5. Proposed Alternative and Basis for Use:

Quad Cities Nuclear Power Station requests the applicability of ASME Code Case N-661-1 be extended to the 2007 Edition through the 2008 Addenda for use in the plants Fifth Interval ISI Program. The NRC has accepted the use of ASME Code Case N-661-1 as an acceptable method for restoring wall thickness for raw water piping systems that have experienced internal wall thinning in the latest revision of Regulatory Guide 1.147, Revision 16.

No technical changes to ASME Code Case N-661-1 are being proposed in this relief request. This relief request is being submitted to correct a timing situation, which has resulted from the application of the 2007 Edition through the 2008 Addenda of ASME Section XI for Quad Cities Nuclear Power Station. Since no technical change is proposed in this relief request, Quad Cities Nuclear Power Station considers that this alternative provides an acceptable level of quality and safety, and is consistent with provisions of 10CFR50.55a(a)(3)(i).

6. Duration of Proposed Alternative:

Relief is requested for the Fifth Ten-Year Inspection Interval for Quad Cities Nuclear Power Station Units 1 and 2.

7. Precedents:

None.

Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 9-1 ISI Program Plan - 5th Interval 9.0 References The references used to develop this Inservice Inspection Program Plan include:

9.1 NRC References 9.1.1 Code of Federal Regulations, Title 10, Energy

a. Part 50, Paragraph 50.55a, Codes and Standards

b. Part 50, Paragraph 2, Definitions, the definition of Reactor Coolant Pressure Boundary

c. Part 50, Appendix J, Primary Reactor Containment Testing for Water Cooled Power Reactors 9.1.2 NRC Regulatory Guide 1.147, Revision 16, Inservice Inspection Code Case Acceptability, ASME Section XI, Division 1 9.1.3 NRC Regulatory Guide 1.150, Revision 1, Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examination 9.1.4 NRC Regulatory Guide 1.192, Revision 0, Operation and Maintenance Code Case Acceptability, ASME OM Code 9.1.5 NRC Regulatory Guide 1.193, Revision 3, ASME Code Cases Not Approved for Use 9.1.6 NRC Regulatory Guide 1.84, Revision 35 Design, Fabrication, and Materials Code Case Acceptability, ASME Section III, Division 1 9.1.7 NRC Regulatory Guide 1.26, Revision 3, Quality Group Classifications and Standards for Water-, Steam-, and Radioactive Waste- Containing Components of Nuclear Power Plants 9.1.8 USAS B31.1.0-1967, Power Piping 9.1.9 NRC Final SE related to the Boiling Water Reactor Owners Group (BWROG) Report, GE-NE-523-A71-0594-A, Revision 1, Alternate Boiling Water Reactor (BWR) Feedwater Nozzle Inspection Requirements, May 2000, (TAC No. MA6787), dated March 10, 2000 9.1.10 NRC Final SE related to the Boiling Water Reactor Owners Group (BWROG) Report, GE-NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements, August 1999, (TAC No. M94090),

dated June 5, 1998 9.1.11 NRC Final SE related to BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), EPRI Report TR-1012621, October 2005, dated March 16, 2006 9.1.12 NRC Final SE related to the BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), EPRI Report TR-113932, October, 1999 (TAC NO. MA5012), dated May 14, 2002 9.1.13 NRC Final SE related to the BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05), EPRI Report TR-105697, September, 1995, dated July 28, 1998 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 9-2 ISI Program Plan - 5th Interval 9.1.14 NRC SE related to EPRI Topical Report TR-112657, Rev. B, Final Report, Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 1999, dated October 28, 1999 9.1.15 NRC Final SE related to the License Renewal Safety Evaluation Report for the DNPS Units 2 and 3, and QCNPS Units 1 and 2, dated July 23, 2004 9.2 Industry References 9.2.1 ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Inservice Inspection of Nuclear Power Plant Components

a. 2007 Edition through the 2008 Addenda (including Appendix VIII) (5th ISI Interval)

b. 2001 Edition through the 2003 Addenda (Subsection IWE) (2nd CISI Interval)

c. 1995 Edition through the 1996 Addenda (4th ISI Interval)

d. 1992 Edition through the 1992 Addenda (1st CISI Interval)

e. 1989 Edition, No Addenda (3rd ISI Interval) 9.2.2 ASME Boiler and Pressure Vessel Code, Section V, Nondestructive Examination, the 2007 Edition through the 2008 Addenda [The Edition and Addenda for ASME Section V are the same as the Edition and Addenda of ASME Section XI used for the inspection interval for both ISI and Non-ISI NDE examinations. Reference ASME Interpretation XI 89-02]

9.2.3 ASME Boiler and Pressure Vessel Code, Section III, Division 1, Rules For Construction of Nuclear Power Plant Components, the 2007 Edition through the 2008 Addenda 9.2.4 ASME OM Code, Code for Operation and Maintenance of Nuclear Power Plants, the 2004 Edition through the 2006 Addenda (Subsection ISTD) 9.2.5 NUREG-0313, Revision 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping 9.2.6 NUREG-0619, dated November 1980, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking 9.2.7 NUREG-0737, TMI Action Plan Requirements, dated November 1980 9.2.8 NUREG-1796, Safety Evaluation Report Related to the License Renewal of the Dresden Nuclear Power Station, Units 2 and 3 and Quad Cities Nuclear Power Station, Units 1 And 2 9.2.9 NUREG-1801, Generic Aging Lessons Learned 9.2.10 Boiling Water Reactor Owners Group (BWROG) Report GE-NE-523-A71-0594-A, Revision 1, Alternate BWR Feedwater Nozzle Inspection Requirements, dated May 2000 9.2.11 Boiling Water Reactor Owners Group (BWROG) Report GE-NE-523-A71-0594, Alternate BWR Feedwater Nozzle Inspection Requirements, dated August 1999 Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 9-3 ISI Program Plan - 5th Interval 9.2.12 BWR Vessel and Internals Project ,Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75-A), EPRI Report TR-1012621, October 11, 2005 9.2.13 BWR Vessel and Internals Project, Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules (BWRVIP-75), EPRI Report TR-113932, October 1999 9.2.14 Generic Letter 88-01, Revision 2, dated January 25, 1988, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping 9.2.15 Generic Letter 88-01, Supplement 1, dated February 4, 1992, NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping 9.2.16 Generic Letter 98-05, Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief From Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds, dated November 10, 1998 9.2.17 BWR Reactor Vessel Shell Weld Inspection Recommendations (BWRVIP-05), EPRI Report TR-105697, September, 1995 9.2.18 BWR Vessel and Internals Project Program Implementation Guide (BWRVIP-94), EPRI Report TR-1011702, dated December 2005 9.2.19 EPRI Topical Report TR-112657, Rev. B-A, Final Report, Revised Risk-Informed Inservice Inspection Evaluation Procedure, December 1999 9.2.20 EPRI Containment Inspection Program Guide (TR-110698-R1) 9.2.21 INPO Engineering Program Guide EPG-11, Inservice Inspection Program 9.3 Licensee References 9.3.1 QCNPS Units 1 and 2 Updated Final Safety Analysis Report (UFSAR) 9.3.2 QCNPS Units 1 and 2 Technical Specification (TS) 9.3.3 QCNPS Units 1 and 2 Technical Requirements Manual (TRM) 9.3.4 QCNPS Units 1 and 2 ISI Classification Basis Document (QC-481565-RP02), Fifth Ten-Year Inspection Interval 9.3.5 QCNPS Units 1 and 2 ISI Selection Document (QC-481565-RP04), Fifth Ten-Year Inspection Interval 9.3.6 Exelon Risk-Informed Inservice Inspection Evaluation (Final Report) for QCNPS Units 1 and 2 9.3.7 DNPS Units 2 and 3, and QCNPS Units 1 and 2, Operating License Renewal Application, January 3, 2003 9.3.8 General Electric Boiling Water Reactor System Department, Document No. 22A2750, Revision 6 9.3.9 QCNPS Reactor Coolant Pressure Boundary Normal Makeup Calculation, XCE.040.0202 9.3.10 QCNPS Design Analysis No. QDC-0200-M-1279, ISI/RCPB Normal Makeup for CRD Housing Welds Automated Engineering Services Corp

QC-481565-RP03, Rev. 0 Page 9-4 ISI Program Plan - 5th Interval 9.3.11 Calculation to Determine 80% of Primary Containment Remains Accessible for Examination, QDC-1600-M-1617 (EC366634) for Quad Cities Nuclear Power Station, Units 1 and 2 9.3.12 QCNPS Appendix J Leak Rate Testing Program QCTP 0130-01 9.3.13 Procedures ER-AA-330, Conduct of Inservice Inspection Activities, ER-AA-330-001, Section XI Pressure Testing, ER-AA-330-002, Inservice Inspection of Welds and Components, ER-AA-330-003, Visual Examination of Section XI Component Supports, ER-AA-330-004, Visual Examination of Snubbers, 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, ER-AA-330-011, Snubber Service Life Monitoring Program, ER-AA-335-004, Manual Ultrasonic Measurement of Material Thickness, ER-AA-335-018, Detailed, General, VT-1, VT-1C, VT-3, and VT-3C, Visual Examination of ASME Class MC and CC Containment Surfaces and Components, ER-AB-331, BWR Internals Program Management, ER-AB-331-101, Evaluation for Thermal Aging/Neutron Embrittlement of Reactor Internals Components, and ER-QC-330-100, Primary Containment and Coating Inspections 9.4 License Renewal References 9.4.1 AT 101562.49.49 - (A.3.1.5), Revised P-T Curves 9.4.2 AT 101562.49.60 - (A.3.1.6), RPV Circumferential Weld Relief 9.4.3 AT 00101562-49 (B.1.1), Perform QCNPS Unit 2 Reactor Head Crack Inspection 9.4.4 AT 101562.04 - (B.1.4), BWR Vessel Id Attachment Welds 9.4.5 AT 101562.08 - (B.1.8), BWR Penetrations 9.4.6 AT 101562.49.11, AT 101562.59, and AT 101562.60 - (B.1.23J), One-Time Inspections Automated Engineering Services Corp

v t e Letter Sequence Request
TAC:ME9668, (Approved, Closed)
Initiation Request, Request, Request Acceptance, Acceptance Supplement, Supplement, Supplement, Supplement, Supplement, Supplement Administration Questions 21 December 2012 9 April 2013 19 April 2013 24 April 2013 2 May 2013 Answers Results Approval, Approval, Approval Other:
MONTHYEAR2013-01-24 [Table View]

RS-12-150, Fifth Interval Inservice Inspection Program Plan and Relief Requests

Text

Subject:

Subject:

SUMMARY

1.0 INTRODUCTION AND BACKGROUND

SUMMARY

9.0 REFERENCES

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

1.0 INTRODUCTION AND BACKGROUND

Reference:

Reference:

SUMMARY

Description:

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

SUMMARY

Description:

SUMMARY

SUMMARY

SUMMARY

Reference:

Description:

Description:

References:

Reference:

Description:

Reference:

Description:

Response

Response

References:

Reference:

Description:

References:

Reference:

Description:

Reference:

Description:

Navigation menu

Search