Text

Fourth Interval Inservice Inspection Program Plan
	ML030430116
Person / Time
Site:	Quad Cities
Issue date:	02/07/2003
From:	Tulon T; Exelon Generation Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	SVP-03-019
	Download: ML030430116 (104)
	v • d • e

Exelon Generation Company, LLC www exeloncoTp corn Exelkn.. Nuclear Quad Cities Nuclear Power Station 22710 206th Avenue North Cordova, IL61242-9740 SVP-03-019 February 7, 2003 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Quad Cities Nuclear Power Station, Units 1 and 2 Facility Operating License No. DPR-29 and DPR 30 NRC Docket No. 50-254 and 50-265

Subject:

Quad Cities Nuclear Power Station, Units 1 and 2, Fourth Interval Inservice Inspection Program Plan

Reference:

Letter from T.J. Tulon (Exelon Generation Company, LLC) to U. S. NRC, "Quad Cities Nuclear Power Station, Units 1 and 2, Fourth Interval Inservice Inspection Program Plan," dated January 17, 2003.

On January 17, 2003, in accordance with 10 CFR 50.55a(g)(4)(ii), Exelon Generation Company, LLC (EGC) submitted the fourth interval inservice inspection (ISI) program for Quad Cities Nuclear Power Station (QCNPS), Units 1 and 2 (Reference). During the review of the submittal, the NRC discovered an administrative error in the ISI program document. A corrected program document is attached.

If you have any questions or require additional information, please contact Mr. Wally Beck at (309) 227-2800.

Respectfully, T' thy J. Tulon te Vice President Quad Cities Nuclear Power Station

Attachment:

Inservice Inspection Program Plan cc: Regional Administrator - NRC Region III (w/attachments)

NRC Senior Resident Inspector - Quad Cities Nuclear Power Station (w/o attachments)

QUAD CITIES NUCLEAR POWER-STATION Iý UNITS1 & 2 ISI PROGRAM PLAN FOURTH TEN-YEAR INSPECTION INTERVAL Commercial Service Dates:

Unit 1 - 2/18/73 Unit 2 - 3/10/73 Quad Cities Nuclear Power Station 22710 2 0 6th Avenue North Cordova, Illinois 61242 Exelon Generation Company, LLC (EGC) 200 Exelon Way Kennett Square, PA 19348 Prepared By:

ITS Corporation Naperville, Illinois

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth 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, and 3 pressur6 retaining components and supports at Quad Cities Nuclear Power Station (QCNPS), Units 1, 2, and 1/2.

Common (Unit 1/2) components are included'in the Unit 1 sections, reports, and tables. 'This ISI Program Plan also includes Risk-Informed Inservice Inspections (RISI), augmented inservice inspections, and pressure testing requirements imposed on or committed to by QCNPS. Procedure ER-AA-330, "Conduct of Inservice Inspection Activities," implements the ASME Section XI ISI Program.

The Fourth Inservice Inspection Interval is effective from March 10, 2003, through March 9, 2013, for-QCNPS Unit 1 and March 10, 2003, through March 9, 2013, for QCNPS Unit 2. These effective interval dates are based on the assumption that QCNPS will be approved to extend plant operation under the license renewal application. Paragraph IWA-2430(d)(1) of ASME Section XI allows an inspection interval to be extended or decreased by as much as one year, and Paragraph IWA-2430(e) allows an inspection interval to be extended when a unit is out of service continuously for six months or more. The extension may be t6ken for a periodof time not to exceed the duration of the outage. See Table 1.1-1 at the end of this section for extensions that apply to QCNPS's Fourth Interval.

The Fourth Inservice Inspection Interval is divided into three successive inspection periods as determined by calendar years of plant service within the inspection interval. Table 1.1-1 identifies the period dates for the Fourth Inservice Inspection Interval as defined by Inspection Program B. In accordance with Paragraph IWA-2430(d)(3), the inspection periods specified in Table 1.1-1 may be decreased or extended by as much as 1 year to enable inspection to coincide with QCNPS's refueling outages.

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, Exelon 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.

1 QCNPS's piping systems and associated components were designed and fabricated before the examination requirements of American Society of Mechanical Engineers 1-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval (ASME) Boiler and Pressure Vessel Code, Section XI were formalized and published. Since this plant was not specifically designed to meet the reqiuirements of ASME Section XI, literal compliance is not feasible or practical within the limits of the current plant design. 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 Code required examination selection criteria cannot be met, a relief request will be submitted in accordance with 10 CFR 50.55a.

1.3 Third Interval ISI Program Pursuant to the Code Of Federal Regulations, Title 10, Part 50, Section 55a, Codes and standards,(10 CFR 50.55a), Paragraph (g), Inservice inspection requirements, licensees were required to update their ISI programs 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 10 CFR 50.55a 12 months prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii). Accordingly, the Inservice Inspection requirements applicable to the Third Inservice Inspection Program should have been based on the rules set forth in the 1986 Edition 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 10 CFR 50.55a(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 that QCNPS Units 1 and 2 met for the Third Ten Year Inservice Inspection Interval.

The ISI Program Plan addressed Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI, and utilized Inspection Program B.

Subsection IWE was added to the ISI Program midway through the Third Interval to address Containment Inservice Inspections (CISI). These requirements were mandated by the Federal Register in 1996 and marked the beginning of the First Interval for CISI inspections. Implementation of the CISI Program is discussed in Section 6.0.

QCNPS adopted the EPRI Topical Report TR-1 12657, Rev. B-A methodology, which was supplemented by 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 Interval. This approach replaced the categorization, selection, and examination volume requirements of ASME Section XI Categories B-F, B-J, C-F-I, and C-F-2 applicable to QCNPS with Category R-A as defined in Code Case N-578-1.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval QCNPS Unii 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 Category B-D, Item Number B3.90 & B3.100 components. These inspections will be repeated during the Fourth Interval in accordance with the Fourth Interval ISI Program.

1.4 Fourth Interval ISI Program Per 10 CFR 50.55a(g), licensees are required to update their ISI programs to meet the requirements of ASME Section XI once every ten years or inspection interval.

The ISI program is required to comply with the latest edition and addenda of the Code incorporated by ieference in 10 CFR 50.55a twelve (12) months prior to the start of the interval per 10 CFR 50.55a(g)(4)(ii).

The QCNPS Fourth Interval ISI Program Plan was developed in accordance with the requirements' of 10 CFR 50.55a, and the 1995 Edition with the 1996 Addenda of ASME Section XI. This ISI Program Plan addresses Subsections IWA, IWB, IWC, IWD, and IWF of ASME Section XI, and utilizes Inspection Program B as defined therein. Implementation of Subsection IWE is discussed in Section 6.0.

QCNPS has 'adopted the EPRI Topical Report TR-1 12657, Rev. B-A methodology, which was supplemented by Code Case N-578-1, for implementing risk-informed inservice inspections. The RISI program will be in effect for the entire Fourth Inspection Interval. This approach replaces the categorization, selection, and examination volume requirements of ASME Section XI Categories B-F, B-J, C-F-1, and C-F-2 applicable to QCNPS with Category R-A as defined in Code Case N-578-1.

Implementation of RISI program is in accordance with relief request 14R-02.

1.5 Code Cases Per Footnote 6 of 10 CFR 50.55a, ASME Code Cases that have been determined to be suitable for use in ISI Program Plans by the NRC are listed in Regulatory Guide 1.147, "Inservice Inspection Code Case Acceptability-ASME Section XI, Division 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 latest version of Regulatory Guide 1.147 incorporated into this document is Revision 13. As this guide is revised, newly approved Code Cases will be assessed for plan implementation at QCNPS.

Footnote 6 also states that 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 10 CFR 50.55a(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.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval This ISI Program Plan will utilize the Draft Regulatory Guide DG-1091 (Proposed Revision 13 of Regulatory Guide 1.147) with the anticipation that theFinal Revision 13 of Regulatory Guide 1.147 will be approved prior to the start of the Fourth Inspection Interval. QCNPS will review the Final Revision 13 of Regulatory Guide 1.147 for ISI program impact at, wYhich time it is published.

1.6 Relief Requests iI In accordance with 10 CFR 50.55a, 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 Fourth Interval are included in Section 8.0 of this document. The text of the relief requests contained in Section 8.0 will demonstrate that one of the following: the proposed alternatives provide an acceptable level of quality and safety per 10 CFR 50.55a(a)(3)(i), compliance with a

the specified requirements would result in hardship or unusual difficulty'without compensating increase in the level of quality and safety per 10 CFR 50:55a(a)(3)(ii), or the code requirements are considered impractical per 10 CFR 50.55a(g)(5)(iii).

Per 10 CFR 50.55a 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."

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151 Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Table 1.1-1 QCNPS Unit I and Unit 2 IST Interval/Period/Outage Matrix Period Interval Period Unit 2 Unit 1 Start Date to Start Date to End Date Projected Outage Outage Outage Projected Outage Start Date to End Date End Date Start Date or Number Number Start Date or I__ Outage Duration Outage Duration 1st ist Scheduled Q2R17 Q1R18 Scheduled 3/10103 to 3/9/06- 3/10/03 to 3/9/06 2/04

- 1/05 2nd th (Unit 1) Schiduled Q2Rl8 QIR19 Scheduled 4 3/10/06 to 3/9/10 3/10/03 to 3/9/13' 1/06 2/07 e(nt2 2 nd - - Scheduled Q2R19 QIR20 Scheduled 4 (Unit 2) 3/10/06 to 3/9/10 - 3/08 2109 -d 3/10/032 to 3/9/13 3 Scheduled Q2R20

-lr QIR21 Scheduled 3 3/10/10 to 3/9/13 3/10 2/11 3/10/10 to 3/9/13 Scheduled Q2R21 Q1R22 Scheduled 2/13 _ _3/12 1______

This extension is being carried forward to the Fourth Note 1: The Unit 1 Third Inspection Interval was extended by 20 days as permitted by IWA-2430(d). successive intervals shall not be Interval to accommodate both Units 1 and 2 having the same interval start date. As required by IWA-2430(d)(1),

the remainder of the Fourth Interval, only 345 days are available to use altered by more than one year from the original pattern. This means that for by 1WA-2430(d). This extension does not affect the start and end dates of Note 2: under The Unit Third Inspection ns the 2IWA-2430(d) Interval was extended by 365 days as-permitted tension.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval 2.0 BASIS FOR INSERVICE INSPECTION PROGRAM 2.1 ASME Section XI Examination Requirements As required by the 10 CFR 50.55a, this program was developed in accordance with the requirements detailed in the 1995 Edition,, 1996 Addenda, of the ASME Boiler and Pressure Vessel Code, Section XI, Division 1, Subsections IWA, IWB, IWC, IWD, IWF, Mandatory Appendices, Inspection Program B of IWA-2432,

,I and approved alternatives through relief requests and safety evaluation reports (SERs).

The ISI program implements Appendix VIII "Performance Demonstration for Ultrasonic Examination Systems," ASME Section XI 1995 Edition with the 1996 Addenda as required by 10 CFR 50.55a(g)(6)(ii)(C). Appendix VIII requires qualification of the procedures, personnel, and equipment used to detect and size flaws in piping, bolting, and the reactor pressure vessel. Each organization (e.g.,

owner or vendor) will be required to have a written program to insure compliance with the requirements. These requirements are implemented through the Performance Demonstration Initiative (PDI) Program according to the schedule defined in 10 CFR 50.55a(g)(6)(ii)(C).

For the Fourth Inspection Interval, QCNPS's inspection program for ASME Section XI Categories B-F, B-J, C-F-i, and C-F-2 will be governed by risk-informed requirements. The RISI program methodology is described in the EPRI Topical Report TR-1 12657, Rev. B-A. To supplement the EPRI Topical Report, Code Case N-578-1 (as applicable per Relief Request 14R-02) is also being used for the classification of piping structural elements under the RISI program. The RISI program scope will be implemented as an alternative to the 1995 Edition with the 1996 Addenda, ASME Section XI examination program for Class 1 B-F and B-J welds and Class 2 C-F-1 and C-F-2 welds in accordance with 10 CFR 50.55a(a)(3)(i). The basis for the resulting risk categorizations of the non-exempt 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 10.0 of this document.

The CISI Program Plan per Subsection IWE has been incorporated into Section 6.0 of this ISI Program Plan and is not discussed in this section.

2.1.1 ASME Section XI Code Cases As referenced by 10 CFR 50.55a Footnote 6 and allowed, by NRC Regulatory Guide 1.147, Revision 13, the following Code Cases are being incorporated into the QCNPS ISI Program:

N-307-2 Revised Ultrasonic Examination Volume for Class 1 Bolting, Table IWB-2500-1, Examination Category B-G-1, 2-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval When the Examinations Are Conducted From the End of the Bolt or Stud or From the Center-Drilled Hole N-416-2 Alterhative Pressure Test Requirements for Welded Repairs, Fabrication Welds for Replacement Parts and Piping Subassemblies, or Installation of Replacement Items by Welding, Class 1, 2, and 3.

Code Case N-416-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

(1) Additional surface examinations should be performed on the root (pass) layer of butt and socket welds of the pressure retaining boundary of Class 3 components when the surface examination method is used in accordance with Section III.

(2) A 4-hour hold time must be maintained prior to the VT-2 visual examination.

(See Technical Approach and Position number 14T-03).

, N-458-1 Magnetic Particle Examination of Coated Materials N-460 Alternative Examination Coverage for Class 1 and'Class 2 Welds N-498-1 Alternative Rules for 10-Year Hydrostatic Pressure Testing for Class 1, 2, and 3 Systems Code Case N-498-1 is only being implemented as it pertains to Class 3 systems. (The portions of the Case that address Class 1 and 2 systems have been incorporated into the ASME Section XI code of record, 1995 Edition with 1996 Addenda, applicable to the QCNPS Fourth Interval.)

N-504-2 Alternative Rules for Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping N-516-2 Underwater Welding Code Case N-516-2 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval When welding is to be performed on high neutron fluence Class 1 material, then a mock-up, using material with similar fluence levels, should be welded to verify that adequate crack prevention measures were used.

N-523-2 Mechanical C1lmping Devices for Class 2 and 3 Piping N-526 Alternative Requirements for Successive Inspections of Class I and 2 Vessels N-532 Alternative Requirements to Repair and Replacement Documentation Requirements and Inservice Summary Report Preparation and Submission as Required by IWA-4000 and IWA-6000 Code Case N-532 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

An Owner's Activity Report Form OAR-1 is required to be prepared and certified upon completion of each refueling outage. The Code Case does not designate a time frame for submission to the regulatory authority.

Thus, the OAR-1 must besubmitted within 90 days.

Applicable IWA-4000 and IWA-6000 references from the 1995 Edition, with the 1996 Addenda of ASME Section XI, will be utilized in place of the code references specified in Code Case N-532. A matrix of those reference paragraphs was added in Code Case N-532-1 for various Code years.

N-546 Alternative Requirements for Qualification of VT-2 Examination Personnel Code Case N-546 is acceptable subject to the following conditions specified in Regulatory Guide 1.147, Revision 13.

(1) Qualify examination personnel by test to demonstrate knowledge of Section XI and plant specific procedures for VT-2 visual examination.

(2) Requalify examination personnel by examination every three years.

(3) This Code Case is applicable only to the performance of VT-2 examinations 2-3 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval N-573 Transfer of Procedure Qualification Records Between Owners N-588 Alternative to Reference Flaw Orientation of Appendix G for Circumferential Welds in Reactor Vessels N-598 Alternative Requirements to Required Percentages of Examinations N-623 Deferral of Inspections of Shell-to-Flange and Head-to Flange Welds of a Reactor Vessel N-624 Successive Inspections N-640 Alternative Reference Fracture Toughness for Development of P-T Limit Curves Plan Additional Code Cases may be invoked in the future based on new Any Code Cases invoked in requirements or revisions to Regulatory Guide 1.147.

the latest published the future shall be in accordance with those approved for use in revision of Regulatory Guide 1.147 at that time.

'/2.2 Augmented Examination Requirements are performed Augmented'examination requirements are those examinations that is a summary of above and beyond the requirements of ASME Section XI. Below addressed by those examinations performed by QCNPS that are not specifically in addition to the ASME Section XI, or the examinations that will be performed Inspection Interval.

requirements of the Code on a routine basis during the Fourth 2.2.1 Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Letter 88-01, Stainless Steel Piping," Revision 2 / Supplement 1 to Generic NUREG 0313, "Technical Report on Material Selection and Process 2, and Guidelines for BWR Coolant Pressure Boundary Piping," Revision Project Technical EPRI Report TR- 113932 "IBWR Vessel and Internals Basis for Revisions to Generic Letter 88-01 Inspection Schedules dated May (BWRVIP-75)," as conditionally approved by NRC final SER 14, 2002.

These documents discuss the examination requirements for Intergranular Steel Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless the ISI Piping. References to Generic Letter 88-01 (GL 88-01) within program refer to the comprehensive commitments to all of these 88-01 documents. The final SER of BWRVIP-75 revised the GL were inspection schedules. The BWRVIP-75 revised inspection schedules gained based on consideration of inspection results and service experience 2-4 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval t

by the industry since issuance of GL 88-01, 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 6ommitment 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 a*nd fromL N. Olshan (NRC) to T. J. Kovach (CECo) August 21, 1990.

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 the'T. E. Murley (NRC) dated August 20, 1993 and in a letter from the NRC from R. M. Pulsifer (NRC) to D. L. Farrar (CornEd) dated September 22, 1994.

RISI regulati 6 ns are being invoked for QCNPS in this ISI Program Plan.

Under these new guidelines, Class 1 and 2 piping structural elements are inspected in accordance with EPRI Topical Report TR-1 12657, Rev. B-A and 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 will be subsumed into the RISI program.

2.2.2 Alternate BWR Feedwater Nozzle Inspection Requirements, dated October 1995 This document discusses BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking. The alternate approach was developed and submitted to the NRC by the Boiling Water Reactor Owners' Group (BWROG). The NRC conditionally accepted these alternate requirements in an SER dated June 5, 1998.

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 0. D. Kingsley (ComEd) dated April 30, 1999 to confirm the discussions of an April 1, 1999, conference call in that CornEd will use the more recent fatigue curves that address environmental effects as approved by ASME Section XI.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval 2.2.3 BWR Vessel and Internals Project (BWRVIP)

The BWRVIP is comprised of a series of Inspection & Evaluation Guidelines and documents that discuss reactor vessel internals.

The BWRVIP encompasses pertinent information and requirements presented in General Electric Service Information Letters (SILs) and Rapid Information Communication Services Information Letters (RICSILs).

2.2.4 NRC 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. Coimmitments made concerning NUREG-0737 are required to be maintained per the QCNPS Operating Licenses.

/2.3 System Classifications, and P&ID Boundary Drawings The ISI Classification Basis Document details those systems that are ISI Class 1, Class 2, or Class 3 that fall within the inservice inspection scope of examinations.

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, Class 2; Class 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, Class 2, or Class 3 safety function, the components are further reviewed to assure the interfaces (boundary valves and boundary barriers) meet the criteria set by 10 CFR 50.2, 10 CFR 50.55a(c)(1), 10 CFR 50.55a(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 "Nuclear Safety Criteria for the Design of 2-6 Revision 0

ISI Proiram Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Stationary Boiling Water Reactor Plants," were also used for guidance in determining the classification boundaries when 10 CFR and Regulatory Guide 1.26 did not address a given situation. The valve positions shown on the system flow diagrams are assumed to be the normal positions during system operation unless otherwise noted.

At the time the construction permits for QCNPS Units I and 2 were issued, ASME only pressure vessels, primarily nuclear reactor vessels. The Section III coyered majority of piping, 'pumps, and valves were designed and installed according to the rules of USAS B31.1.0-1967 Edition, "Power'Piping." Consequently, the QCNPS ISI Program his'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 codinig system used on the P&IDs to identify the safety related systems or p6rtions 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 IC, IF, IS, IV, 2E, 2F, 2P, 2S, 2V, 3G, and 3P) were used to identify exempt ASME Section XI components.

The systems and components 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.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TABLE 2.3-1 QCNPS ISI CLASSIFICATION BOUNDARY DRAWINGS S

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

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

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

M-34, SB. 1 M-76, SH. 1 Diagram of Pressure Suppression Piping M-35, SR. 1, 2, & 5 M-77, SR. 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 (RHRR & 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)

M241, SR 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 (HiPCI)

M-47 SIL 1 M-88 SI-I Diagram of Reactor Water Clean-Up Piping (RWCU)

M-50 SR. I M-89 SH. I 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, SR. 1 M-1061, SH. 1 Diagram of High Radiation Sampling System Piping (HRSS) 2-8 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval J

2.4 ISI Isometric Drawings for Nonexempt ISI Class Components and Supports Code ISI isometric and component drawings were developed to detail the ISI locations at Class 1, 2, and 3 components (welds, bolting, etc.) and support QCNPS. SPT isometric drawings were also developed to show those components the subject to pressure testing. These comiponents and supports are identified on ISI isometric and component drawings listed in Table 2.4-1.

QCNPS's ISI program, including the database, basis document, and schedule, testing.

addresses the non-exempt components which require examination and A summary of QCNPS Units 1 and 2 ASME Section XI nonexempt components and supports is included in Section 7.0.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TABLE 2.4-1 QCNPS ISI ISOMETRIC AND COMPONENT DRAWINGS Sheet I of 3 UNIT 1 & 1/2 UNIT 2 TITLE M-3101, SH. 1,2 &4 M-3111, SHL 1,2 &4 ISI Class 1 Main Steam System

,1A-3102, SH. 1 & 2 M-3112, SR- I & 2 ISI Class 1 Reactor Feedwater System M-31 13, SR 1-5 ISI Class 1 Reactor Recirculation System and Jet M-3103, SR. 1-5 Pump Instrument M-3i04, SH. 1 & 2 M-3114, SU 1 & 2 ISI Class 1Core Spray System M-3105, SH. 1-3 M-3115, SH. 1-3 ISI Class 1Residual Heat Removal System M-3106, SH. 1 M-3116, SH. 1 ISI Class I Standby Liquid Control System M-3107, SH. 1 M-3117, SH. I 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, SRL 1 &2 M-3119, SH. 1 &2 ISI Class I Reactor Water Cleanup System M-31 10, SR. 1 M-3120, SH. I ISI Class I Reactor Core Isolation Cooling System M-3121 SH. I 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 Systerfi M-3132, SH. 1-4 M-3137, SL 1-4 ISI Class 2 High Pressure Coolant Injection System M-3134, SI. 1 & 2 M-3139, S. 1& 2 ISI Class 2 Control Rod Drive System M-3140, SHR I M-3141, SH. I 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, SR. 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'-U" M-3202-2 M-3219-2 System Pressure Test Walkdown Isometric Drywell Fourth Level, EL-651'-O" M-3202-3 M-3219-3 System Pressure Test Walkdowd Isometric Drywell Third Level, EL 640'-U" M-3202-4 M-3219-4 System Pressure Test Walkdown Isometric Drywell Second Level, EL 614'-U" 2-10 Revision 0

ISI Program Plan Quad Cities' Nuclear Power Station Units 1 & 2, Fourth Interval TABLE 2.4-1 QCNPS ISi ISOMETRIC AND COMPONENT DRAWINGS Sheet 2 of 3 UNIT 2 TITLE UNIT I & 1/2 M-3202-5 M-3219-5

. I System Pressure Test Walkdown Isometric Drywell First Level, EL 592'-0" System Pressure Test Walkdown Isometric Drywell M-3202-6 I M-3219-6 Basement, EM579'-0"'

M-3202-7 I M-3219'7 System Pressure Test Walkdown Isometric Lower Head CRD Area, EL 588'-0" System Pressure Test Walkdown Isometric M-3202-8 I M-3219-8 Instrumentation System Pressure Test Walkdown Isometric Head M-3203 M-3220 Flange Seal Leak Detection System Pressure Test Walkdown Isometric Control M-3204, SH. 1 & 2 M-3221, SH. 1 & 2 Rod Drive Hydraulic Piping System Pressure Test Walkdown Isometric Standby, M-3205 M-3222 Liquid Control Piping System Pressure Test Walkdown Isometric Standby M-3206 , M-3223 Liquid Control Piping System Pressure Test Walkdown Isometric Standby M-3207 M-3224 Liquid Control Piping System Pressure Test Walkdown Isometric Diesel M-3208 Generator Cooling Water Piping System Pressure Test Walkdown Isometric Diesel M-3209 M-3225 Generator Cooling Water Piping System Pressure Test Walkdown Isometric High M-3226 M-3210 Pressure Coolant Injection System System Pressure Test Walkdown Isometric ECCS M-321 1, Si 1 M-3227, SH. 1 Ring Header & RCIC Suction Line System Pressure Test Walkdown Isometric Core M-3211, SH. 2 1 M-3227, SIH 2 Spray Piping System Pressure Test Walkdown Isometric Residual M-3211, SH.3 & 4 i M-3227, SH. 3 & 4 Heat Removal Piping System Pressure Test Walkdown Isometric High M-3211, SH. 5 M-3227, SH. 5 Pressure Coolant Injection System System Pressure Test Walkdown Isometric ECCS M-3212 M-3228 Keepfill Pump and Piping 2-11 Revision 0

ISI Program Plan Q-1 *.A riftl Nuclear Pnwer Station Units 1 & 2. Fourth Interval TABLE 2.4-1 QCNPS ISI ISOMETRIC AND COMPONENT DRAWINGS Sheet 3 of 3 UNIT1&1/2 UNIT 2 TITLE M-3213 M-3229 System Pressure Test Walkdown Isometric Diesel Generator Service Water M-*14, 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 Piessure Test Walkdown Isometric Residual Heat Removal Service Water System M-3216, SR 1,2 & 3 M-3232, SHR 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 2-12 Revision 0

Plan ISI Program ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval 2.5 Technical Approach and Positions When the requirements of ASME Section XI are not easily interpreted, QCNPS has reviewed general licensing/regulatory requiremiients and industry practice to determine a practical method of implementing the Code requirements. The technical approach and position documents contained in this section have been provided to clarify QCNPS's implementation of ASME Section XI requirements. An index which summarizes each, technical approach/position is included in Table 2.5-1.

2-13 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TABLE 2.5-1 TECHNICAL APPROACH AND POSITIONS INDEX/SUMMARIES Revision Status1 (Program) Description Position Nnimher Date I I II I 0 Active (SPT) System Leakage Testing of Non I4T-01 Isolable Buried Conmponents.

1/17/03 0 Active (SIT) Valve Seats as Pressurization I4T-02 Boundaries.

1/17/03 0 Active (SPF) Alternative Pressure Test I4T-03 Requirements following Repair and 1/17/03 Replacenent Activities.

I I I I - -

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

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2-14 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Unit6 1 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-01 (Page 1of1)

COMPONENT IDENTIFICATION Code Class: 3

Reference:

1WA-5244(b)(2)

Examination Category: N1/A Item Number: N/A

==

Description:==

System Leakage Testing of Non-Isolable Buried Components.

Component Number: Noh-Isolable Buried Pressure Retaining Components CODE REOULIREMENT that flow during IWA-5244(b)(2) requires non-isolable buried components be tested to confirm operation is not impaired.

POSITON can be considered Article 1WA-5000 provides no guidance in setting acceptance criteria for what QCNPS has established the "adequate flow." In lieu of any formal guidance provided by the Code, following acceptance criteria; pumps, For opened ended lines on systems that require Inservice Testing (IST) of adherence to IST acceptance criteria is considered as reasonable proof of adequate flow through the lines.

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

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-02 (Page 1 of 1)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3 I

Reference:

IWA-5221 IWA-5222 Examination Category: B-P, C-H, D-B Item Number: B15.10, B15.30, B15.50, B15.70, C7.10, C7.30, C7.50, C7.70, D2.10, D2.30, D2.50, D2370

Description:

Valve Seats as Pressurization Boundaries Component Number. All Pressure Testing Boundary Valves CODEI REQUIREMENT IWA-5221 requires the pressurization boundary for system leakage testing extend to those pressure retaining components under operating pressures during normal system service.

I at or System leakage testing is perfornrd in lieu of hydrostatic pressure testing (Paragraph IWA-5222) 3 near the end of each inspection interval in accordance Code Case N-498-1 for Class 1, 2, and systems. Code Case N-498-1 require the pressurization boundary extend to all Class 1 components during'the system leakage test, and extend to all Class 2 and 3 components included in those portions of systems required to operate or support the system safety function up to and including the first normally closed valve.

POSITION utilized QCNPS's position is that the pressurization boundary extends up to the valve seat of the valve provides for isolation. For example, in order to pressure test the Class 1 components, the valve that 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.

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

The extension of the pressurization boundary during an operational test would require an abnormal of valve line-up. Extending the boundary for a hydrostatic test would require the over pressurization and Core Spray).

low pressure piping at systems that have a high/low pressure interface (such as RHR In order to simplify examination of classed components, QCNPS will perform a VT-2 visual seat).

examination of the entire boundary valve body and bonnet (during pressurization up to the valve

ISI Program Plan Quad Cities Nuclear Power Station Unit4 1 & 2, Fourth Interval I

14T-03 TECHNICAL APPROACH AND POSITION NUMBER:

(Page I of 2)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3

Reference:

IWA-4540 Examination Category: N/A I Item Number. N/A

==

Description:==

Alternative Pressure Test Requirements following Repair and Replacement Activities Components Number: All Class 1, 2, and 3 Pressure Retaining Components CODE REQUIREMENT pressure retaining IWA-4540(a) requires a system hydrostatic test be performed after welding on the as exempted by IWA-4540(b).

boundary, or installation of an item by,welding or brazing, except POSITION to perform and often represent a true Hydrostatic tests conducted at elevated pressures are difficult the difficulties associated with a hardship without any compensating increase in plant safety. Some of or removing relief valves, hydrostatic test include complicated or abnormal valve line-ups, gagging isolation, and substantially increased additional maintenance on valve internals not normally used for radiation exposure.

as an alternative to the ASME Section For this reason, QCNPS will utilize ASME Code Case N416-2 Code Case N-416-2 is conditionally XI repair/replacement hydrostatic pressure testing requirement.

1.147. QCNPS will implement this approved for use by the NRC in Revision 13 of Regulatory Guide as'follows:

Case in accordance with the Regulatory Guide imposed conditions layer of Class 3 butt and (1) Additional surface exams will be performed on the root (pass) with Section III.

socket welds when the surface exam method is used in accordance exam. [Note: This (2) A 4-hour hold time must be maintained prior to the VT-2 visual condition is consistent with established regulatory position.]

time condition stated in (2) above. The The Technical Approach established here is to clarify the hold activities and requirements included in regulatory position referenced affects 'several pressure testing is stated in both related Safety future Code editions and other ASME Code Cases. The position of establishing this position, QCNPS Evaluation Reports and in Federal Rulemaking. For the purpose 10 CFR 50 as published in the will take guidance from the latest NRC Proposed Rulemaking affecting 2001.

Federal Register, Volume 66, Number 150, dated August 3,

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TECHNICAL APPROACH AND POSITION NUMBER: 14T-03 (Page 2 of 2)

POSITION (Continued) state the NRC position Section 2.2.7, System Leakage Tests, and Section 3, Paragraph (b)(2)(xx) both time for non regarding hold times. The position as stated in Section 3 reads "a 10-minute hold and components will be insulated systems and components or a 4-hour hold fime for insulated systems requieed after attaining system operating pressure."

Guide 1.147, Code QCNPS will utilize this regulatory position for the purpose of clarifying Regulatory referenced in Condition Case N-416-2, Condition (2) as stated above. As such, the 4-hour hold time activities on (2) will only be implemented for those pressure tests conducted after repair/replacement if the insulation is insulated components. If the system or component is not normally insulated, or time will be used removed for the purpose of conducting the system leakage test, a 10-minute hold after attaining test pressure.

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval 3.0 COMPONENT ISI PLAN The QCNPS Component ISI Plan includes ASME Section XI nonexempt pressure.

retaining welds, piping structural elemrnts, pressure retatining 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 tie P&IDs listed in Section 2.3, Table 2.3-1. Procedure ER-AA-330-002, "Inservice Inspection of Welds and Components," implements the ASME Section XI Cohfiponent ISI Plan. This Component ISI Plan also ificludes component augmented inservice inspection examinations specified by documents other than ASME Section XI as referenced in Section 2.2 of this document.

3.1 QCNPS Nonexempt ISI Class Components I

The QCNPS ISI Class 1 components subject to examination are those which are not exempted under the criteria of Subarticle IWB-1220 in the 1989 Edition, No Addenda of ASME Section XI (see Section 3.1.2 below). The QCNPS SI Class 2 and 3 components identified in P&IDs are those not exempted under the criteria of Subarticles 1WC-1220 and Subarticle IWD-1220 in the 1995 Edition, 1996 Addenda of ASME Section XI. A summary of QCNPS Units 1, 2, and 1/2 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 components are identified on the ISI Isometrics and ISI Component Drawings listed in Section 2.4, Table 2.4-1. Welded attachments are also identified by controlled QCNPS support drawings.

3.1.2 10 CFR 50.55a(b)(2)(xi) specifies that the 1989 Edition, No Addenda of ASME Section XI, Subarticle IWB-1220 shall be used in lieu of the 1995 Edition, 1996 Addenda of ASME Section XI, Subarticle IWB-1220.

IWB-1220, Components Exempt from Examination (1989 Edition, No Addenda) - The following components (or parts of components) are exempted from the volumetric and surface examination requirements of IWB-2500 per the Code paragraph referenced:

(a) Components that are connected to the Reactor Coolant System and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting flow of coolant from the Reactor Coolant System, under normal plant operating conditions, is within the capacity of makeup systems which are operable from on-site emergency power; Reactor Coolant Makeup Calculation - Exelon has determined through the criteria of Subarticle IWB-1220(a) that Class 1 components which are (1) 1.57" ID and smaller for Liquid filled 3-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval components or (2) 3.14" ID and smaller for Steam filled, components are exempt from the volumetric and surface examinations. The QCNPS Reactor Coolant Pressure Boundary Normal Makeup Calculation (XCE.040.0202) is referenced in Section 3.3 of this document.

(b)(1) piping of Nominal Pipe Size (NPS) 1 and smaller; (b)(2) components and their connections in piping of NPS 1 and smaller; (c) reactor vessel head connections and associated piping, NPS 2 and smaller, made inaccessible by control rod drive penetrations.

3.2 Risk-Informed Examination Requirements Piping structural elements that fall under RISI 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-1 12657, Rev. B-A and Code Case N-578-1, piping structural elements ranked as High or Medium Risk are subject to examination while piping structural elements ranked as Low Risk are not subject to examinations (except for pressure testing). Thin wall welds that were excluded from volumetric examination under ASME Section XI rules per Table IWC-2500-1 are included in the element scope that is potentially subject to RISI examination at QCNPS.

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 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 FAC or IGSCC 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 exams to detect the additional degradation mechanism.

3.3 Reactor Coolant Pressure Boundary Normal Makeup Calculation The basis for determining the size of ISI Class 1 water and steam lines exempted for the volumetric and surface examination requirements of 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 IWB-1220(a), liquid lines were defined as those which 3-2 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval 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 cbnsists of the following system(s):

Pump Maximum Emergency System Flow Rate Fluid Temp. Power Safe Shutdown - 400 GPM 1400 F Yes,'

UFSAR, Section 5.4.6.5. On-site RCIC - 400 GPM 140CF Yes, UFSAR, Section 5.4.6 On-site 2

Water flow rates from a liquid line break are taken as 8000 lbs/sec/ft at 1000 psi.

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

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

[General Electric Boiling Water Reactor System Department, Doc No. 22A2750, pg. 71 D.=rM7 17.8 D. = 2D.

where:

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

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

M7 0 = Volumetric flow rate of makeup water at 700 F in gal/min.

3 V7o = Specific volume of water at 700 F in ft /lbm.

3 V 140 = Specific volume of water at 1400 F in ft /lb..

180 [0.016051 800 L.01629] = 1.57" l.

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

3-3 Revision 0

ISI Prograin Plan Quad Cities Nuclear Power Station Units I & 2, Fourth 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 XU Examination Category B-O, Item Number B]14.10)

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 Subsubarticle 1WB-1220(a). See Reference No. 49 for calculation/justification in Section 10.0 of this document.

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3-4 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval 4.0 SUPPORT ISI PLAN XI nonexempt ISI The QCNPS Support Program includes the supports of ASME Section ER-AA-330-003, Class 1, 2, and 3 components as described in Sectiori 3.0.' Procedure the ASME Section "Visual Examination of Section XI Component Supports," implements XI Support ISI Plan.

4.1 QCNPS .Nonexempt ISI Class Supports do not The QCNPS ISI Class 1, 2, and 3 nonexempt supports are those which Units 1, 2, and meet the criteria of Subarticle IWF-1230. A summary of QCNPS 7.0.

1/2 ASME Section XI1 nonexempt supports is included in Section 4.1.1 Identification of ISI Class 1, 2, and 3 Nonexempt Supports drawings ISI Class 1, 2, and 3 supports are identified on the ISI isometric listed in Section 2.4, Table 2.4-1. Supports are also identified by controlled QCNPS support drawings.

4.2 Snubber Examination and Testing Requirements XI Paragraphs IWF-5200(a)

ASME SectionVisual & (b) and ofIWF-5300(a)to &be(b) 4.2.1 Examinations and Inservice Tests snubbers "require VT-3 performed in accordance with the Opeiation and Maintenance of Nuclear by Power Plants (OM), Standard ASME/A14SI OM, Part 4. As allowed 10 CFR 50.55a(b)(3)(v), QCNPS will use Subsection ISTD, "Inservice Power Testing of Dynamic Restraints (Snubbers) In Light Water Reactor in lieu of Plants," ASME OM Code, 1995 Edition with the 1996 Addenda, the requirements for snubbers in ASME Section XI, Paragraphs IWF-5200(a) & (b) and IWF-5300(a) & (b). Procedure ER-AA-330-004, "Visual Examination of Technical Specification Snubbers," implements visual examination of snubbers. Procedure ER-AA-330-010, "Snubber Functional Testing," implements functional testing of snubbers.

support The ASME Section XI ISI Program uses Subsection IWF to define Class inspection requirements. The ISI Program maintains the Code IWF. This snubbers in the populations subject to inspection per Subsection and is done to address the related requirements of Paragraphs IWF-5200(c)

IWF-5300(c). (See Section 4.2.2 below.)

4.2.2 ASME Section XI Paragraphs IWF-5200(c) and IWF-5300(c) require in integral and nonintegral attachments for snubbers to be examined in VT-3 visual accordance with Subsection IWF of the Code. This results examination of the snubber attachment hardware including lugs, bolting, pins, and clamps.

4-1 Revision 0

lSI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval The ASME Section XI ISI Program uses Subsection IWF to define the inspection requirements for all Class 1, Class 2, and Class 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., lugs, bolting, pins, and clamps) per Paragraphs IWF-5200(c) and IWF-5300(c).

1 4-2 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units'l & 2, Fourth Interval 5.0 SYSTEM PRESSURE TESTING ISI PLAN The QCNPS System Pressure Testing'(SPT) Programn icludes all pressure retaining ASME Section XI, ISI Class 1, 2, and 3 components, with the exception of those specifically exempted by Paragraphs IWC-5222(b) and IWD-5240(b). All 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 visual inspections on the ISI Class 1, 2, and 3 pressure retaining components to verify system and component structural integrity. This program conducts both Periodic and Interval (10-year frequency) pressure tests as defined in ASME Section XI Inspection Program B. Procedure ER-AA-330-001, "Section XI Pressure Testing," implements the ASME Section XI System Pressure Testing ISI Plan. I I All components subject to,ASME Section XI System Pressure Testing are shown on the P&IDs listed in Section 2.3, Table 2.3-1, and System Pressure Test Walkdown Isometric Drawings listed in Table 2.4-1.

5.1 Risk-Informed Examinations of Socket Welds Socket welds selected for examination under the RISI program are to be inspected with a VT-2 exam each refueling outage per'ASME, Code Case N-578-1 (see footnote 12 in Table 1 of the Code Case). To facilitate this, socket welds selected for inspection under the RISI program shall be pressurized each refueling outage.

5-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval 7.0 INSERVICE INSPECTION

SUMMARY

TABLES The following Tables 7.0-1 and 7.0-2 provide a summary of the ASME Section XI component, support, system pressure testing, and augmented examinations and tests for the Fourth Interval at QCNPS Units 1, 2, and 1/2.

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

II Examination Item Number Description Exam Total Number of Relief Notes Category (with (or Augmented Requirements Components by Request/

Category Number or System TAP Description) Risk Category Number Number)

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

(1) Examination Category and 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 "NIA" is used to identify Augmented ISI examinations and other QCNPS commitments.

Examination Category "R-A" from Code Case N-578-1 is used in lieu of ASME Section XI Examination Categories B-F, B-J, C-F-I, and C-F-2 to identify Class 1 and 2 piping structural elements for the RISI programrL (2) Item Number (or Augmented Number or Risk Category 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.

Specific abbreviations such as BWROG, BWRVIP, BWRVIP-75, and 0737 have been developed to identify Augmented ISI examinations and other QCNPS commitments.

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

7-1 Revision 0

ISI Progjam Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval (3)

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 Augmented inspection commitments, a description of the Augmented requirement is provided.

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

(4) Exam Requirements:

Provides the examination method(s) 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 augmented components from QCNPS commitments or'Relief Requests.

Provides the examination requirements for piping structural elements under the RISI program are in accordance with the EPRI Topical Report TR-1 12657, Rev.

B-A and Code Case N-578-1.

(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, Augmented Number, or Risk Category Number.

(6) Relief Request/TAP Number Provides a listing of Relief Request/Technical Approach & Position (TAP) numbers applicable to specific components, the ASME Section XI Item Number, Augmented Number, or Risk Category Number. Relief Requests 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 technical approach and position in Section 2.5.

(7) Notes Provides a listing of program notes applicable to the ASME Section XI Item Number, Augmented Number, or Risk Category Number. If a program note number is identified, see the corresponding program note at the end of the Table 7.0-2.

7-2 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Item Description Exam Total Number of Relief Request/ Notes Examination Category (with Category Description) Number Requirements Components by TAP Number System B-A B1.11 Circumferential Shell Welds Volumetric RPV: 4 14R-04 Pressure Retaining B1.12 Longitudinal Shell Welds Volumetric RPV: 15 14R-04 Welds in Reactor Vessel B11.21 Circumferential Head Welds Volumetric RPV: 3 14R-04 B1.22 Meridional Head Welds Volumetric RPV: 16 I4R-04 B11.30 Shell-to-Flange Weld Volumetric RPV: 1 B11.40 Head-to-Flange Weld Volumetric & RPV: I Surface B 1.51 Beltline Region Repair Weld Volumetric RPV: 5 I4R-04 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 29 Full Penetration Welds B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV:29 I4R-01 of Nozzles in Vessels 7-3 Revision 0

IST Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number

_____I__ System B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-1 RR: 92 ,

Pressure Retaining B6.20 Closure Studs, in place (Reactor Vessel) Volumetric RR: 92 Bolting, Greater Than B6.30 Closure Studs, when removed (Reactor Vessel) Volumetric & RR: 92 2 in. In Diameter Surface B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 92 B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-i RPV: 92 B6.180 Bolts and Studs (Pumps) -Volumetric RR: 32 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-I RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-1 RR: 32 B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-1 - MS: 13 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: 3 CSB: 3 FWA: 3 FWB: 3 HPCI: 2 MS: 8 RHRA: 3 RHRB: 3 RR: 6 RWCU: 3 SDC: 2 7-4 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit I & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-K B10.10 Welded Attachments to Pressure Vessels Surface RPV: 9 Welded Attachments for B10.20 Welded Attachments to Piping Surface FWA: 1 Vessels, Piping, Pumps, FWB: 1 and Valves HPCI: 1 MS: 12 RHRA: 1 RHRB: 1 RR: 9 SDC: I B10.30 Welded Attachments to Pumps Surface RR: 6 B10.40 Welded Attachments to Valves Surface RR: 2 B-L-2 B12.20 Pump Casings Visual, VT-3 - RR: 2 Pump Casings I _ _

B-M-1 B12.40 Valve Body Welds (NPS 4 or Larger) Volumetric MS: 4 Pressure Retaining Welds Valve Body B-M-2 B12.50 Valve Bodies (NPS 4 or Larger) 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 7-5 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes Number Requirements Components by TAP Number (with Category Description) System...

B-N-I B13.10 Vessel Interior Visual, VT-3 RPV: 1 Interior of Reactor Vessel I I I B-N-2 B13.20 Interior Attachments Within Beltline Region Visual, VT-i RPV: 26 Welded Core _

Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region Visual, VT-3 RPV: 40 Attachments to Reactor Vessels B13.40 Core Support Structure -Visual, VT-3 RPV: I B-O B14.10 Welds in CRD Housing Volumetric or RPV: 32 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings 7-6 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 &-2, Fourth Interval Unit I & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-P B15.10 Reactor Vessel - System Leakage Test Visual, VT-2 CRD 14R-07 All Pressure B15.50 Piping - System Leakage Test Visual, VT-2 CS 14T-02 Retaining Components B15.60 Pumps - System Leakage Test Visual, VT-2 FW I4T-03 (Periodic) B15.70 Valves - System Leakage Test Visual, VT-2 HPCI MS RCIC RHR RR RWCU RX SBLC 7-7 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit I & 1/2 Inservice Inspection Summary Table 7.0-1 Exam Total Number of Relief Request/ Notes Examination Category Item Description Requirements Components by TAP Number (with Category Description) Number

_System Shell Circumferential Welds Volumetric RHRA: 3 C-A CI.10 RHRB: 3 Pressure Retaining Welds RHRA: I C1.20 Head Circumferential Welds Volumetric in Pressure Vessels RHRB:1I C-B C2.31 Reinforcing Plate Welds to Nozzle & Vessel for Nozzles Surface ECCS: 4

-RHRA: 4 Pressure Retaining with Reinforcing Plates in Vessels, Greater than 1/2" RHRB: 4 Nozzle Welds in Nominal Thickness Nozzle-to-Shell (or Head) Welds with Reinforcing Plates Visual, VT-2 ECCS: 4 Vessels C2.33 RHRA: 2 when Inside of Vessel is Inaccessible for Vessels, Greater than 1/2" Nominal Thickness _____ - RHRB: 2 Surface RJRA: 4 C-C C3.10 Welded Attachments to Pressure Vessels RHRB: 4 Welded Attachments CRD: 2 C3.20 Welded Attachments to Piping Surface for Vessels, Piping, CSA: 5 Pumps, and Valve CSB: 5 ECCS: 2 HPCI: 3 RHR: 1 RHRA: 5 RHRB: 4 Surface CSA: 1 C3.30 Welded Attachments to Pumps CSB: 1 7-8 Revision 0

IST Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Suimmary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number C-H C7.10 Pressure Vessels - System Leakage Test Visual, VT-2 CRD 14R-05 All Pressure C7.30 Piping - System Leakage Test Visual, VT-2 CS I4R-06 Retaining Components C7.50 Pumps - System Leakage Test Visual, VT-2 ECCS 14R-07 (Periodic) C7.70 Valves - System Leakage Test Visual, VT-2 FW I4T-02 HPCI 14T-03 RCIC RHR RPV Head Flange RR SBLC 7-9 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System D-A D1.20 Welded Attachments to Piping Visual, VT-1 DGCW: 2+2 Welded Attachments for RHRSW: 30+1 Vessels, Piping, Pumps, and Valves -

D-B D2.10 Pressure Vessels - System Leakage Test Visual, VT-2 - DGCW- 14R-07 All Pressure D2.30 Piping - System Leakage Test Visual, VT-2 - HVAC 14T-01 Retaining Components D2.50 Pumps - System Leakage Test -Visual,-VT-2 PS 14T-02 (Periodic) D2.70 Valves - System Leakage Test Visual, VT-2 RHRSW 14T-03 D-B D2.20 Pressure Vessels - System Hydrostatic Test Visual, VT-2 _ DGCW 14R-07 All Pressure D2.40 System Hydrostatic Test - Piping - System Hydrostatic Test Visual, VT-2 HVAC 14T-01 Retaining Components D2.60 Pumps - System Hydrostatic Test Visual, VT-2 PS 14T-02 (Interval) D2.80 Valves - System Hydrostatic Test Visual, VT-2 RHRSW 14T-03 7-10 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 &-2, Fourth Interval Unit 1 & 1/2 Inservice inspection Sufmmary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System E-A El.11 Containment Vessel; Accessible Surface Areas General Visual 67 Containment El.12 Containment Vessel; Accessible Surface Areas Visual, VT-3 228 Surfaces El.20 Vent System Accessible Surface Areas Visual, VT-3 140 E-C E4.11 Containment Surface Areas Visible Surfaces Visual, VT-l None Containment Surfaces Requiring E4.12 Containment Surface Areas; Surface Area Grid, Minimum Volumetric None Augmented Examination Wall Thickness Location ....

E-D E5.10 Seals Visual, VT-3 N/A CR-21 Seals, Gaskets & E5.20 Gaskets Visual, VT-3 N/A CR-21 Moisture Barriers E5.30 Moisture Barriers Visual, VT-3 4 E-G E8.10 Bolted Connections; Surfaces General Visual 54 CR-30 Pressure Retaining Bolting E8.20 Bolted Connections- Bolts and Nuts TorguedTension N/A CR-24 E-P E9.10 Containment Vessel; Pressure Retaining Boundary Appendix J Torus, DW, Vents All Pressure E9.20 Containment Vessel; Containment Penetration Bellows Appendix J In accordance Retaining with Procedure Components QCTP 0130-01 E9.30 Containment Vessel; Airlocks Appendix J In accordance Swith Procedure QCTP 0130-01 E9.40 Containment Vessel; Seals and Gaskets Appendix J In accordance with Procedure

__________________________________ _________ QCTP 0130-01 _____

7-11 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0.1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System F-A FL.10 Class I Piping Supports Visual, VT-3 CSA: 5 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 CSA: 13 CSB: 22 ECCS: 30 FWB:1 HPCI: 51 RHR: 13 RfIRA: 34 RHRB: 37 F1.30 Class 3 Piping Supports Visual, VT-3 DGCW: 96+63 1 RHRSW: 126+4 -2 7-12 Revision 0

- ISI Program Plan _

Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number SEsitem t m F-A F.40 Supports Other Than Piping Supports Visual, VT-3 - CSA: 1 1 Supports (Class 1, 2, 3, and MC) CSB: 1 2 (Continued) DGCW: I+1 HPCI: 2 JPI: 2 RHRA: 8 RBRB: 8 RHRSW: 8 RPV: 9 RR: 12 N/A BWROG BWR Feedwater Nozzle and Control Rod Drive Return Line Volumetric FWA: 2 Augmented Nozzle Cracking Components (BWROG) - FWB: 2 Components BWRVIP IGSCC Management Program BWR Vessel Internals and Various In accordance with

-Piping Components (GE SIL's and RICSIL's) BWRVIP Program BWRVIP Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric CSA: 15 4

-75 Austenitic Stainless Steel Piping Components and CSB: 11 BWRVIP-75 "Vessel and Internals Project Technical Basis JPI: 10 for Revisions to Generic Letter 88-01 Inspection Schedules"- RHRA: 11 RHRB: 715 RePV:

RR: 91 SDC: 12 0737 Leak testing and periodic visual examinations of systems VT-2 CS, IHPCI, outside of primary containment which could contain highly MS, RHR, radioactive fluids during a serious transient or accident RCIC, RR (NUREG 0737) RWCU 7-13 Revision 0

IST Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 1 & 1/2 Inservice Inspection Summary Table 7.0-1 Examination Category Risk Description Exam Total Number of Relief Request/ Notes (with Category Description) Category Requirements Components by TAP Number Number System R-A I Risk Category 1 Elements See Notes FWA: 12 I4R-02 3 Risk-Informed Piping FWB: 8 5 Examinations 2 Risk Category 2 Elements See Notes CSA: 12 I4R-02 3 CSB: 9 - 5 RHR: 5 RHRA: 68

- RHRB: 73 SDC: 5 3 Risk Categjory 3 Elements See Notes FWA: 1 I4R-02 3, 5 4 Risk Category 4 Elements See Notes CSA: 14 I4R-02 3 CSB: 13 5 ECCS: 54 HPCI: 31 MS: 122 RCIC: 8 RHRA: 33 RHRB: 42 RPV: 2 RR: 36 RWCU: 26 5 Risk Category 5 Elements See Notes RHR: 19 I4R-02 3 RHRA: 13 - 5 I_ RHRB: 12 7-14 Revision 0

ISI Program Plan A.,..A ("ifg.a W/nilou. Pnwrnr .ltntlnn!lintq 1 &:'. Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-A B1.11 Circumferential Shell Welds Volumetric RPV: 4 14R-04 Pressure Retaining B1.12 Longitudinal Shell Welds Volumetric RPV: 13 14R-04 Welds in Reactor Vessel B1.21 Circumferential Head Welds Volumetric RPV: 3 14R-04 B11.22 Meridional Head Welds Volumetric RPV: 16 M4R-04 B1.30 Shell-to-Flange Weld Volumetric RPV: I B1.40 Head-to-Flange Weld Volumetric & RPV: I Surface B1.51 Beltline Region Repair Weld Volumetric RPV: 1 I4R-04 B-D B3.90 Nozzle-to-Vessel Welds (Reactor Vessel) Volumetric RPV: 29 Full Penetration Welds of Nozzles in Vessels B3.100 Nozzle Inside Radius Section (Reactor Vessel) Volumetric RPV: 29 I4R-01 7-15 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-G-1 B6.10 Closure Head Nuts (Reactor Vessel) Visual, VT-1 RPV: 92 Pressure Retaining B6.20 Closure Studs, in place (Reactor Vessel) Volumetric RPV: 92 Bolting, Greater Than B6.30 Closure Studs, when removed (Reactor Vessel) Volumetric & RPV: 92 2 in. in Diameter Surface B6.40 Threads in Flange (Reactor Vessel) Volumetric RPV: 92.

B6.50 Closure Washers, Bushings (Reactor Vessel) Visual, VT-I RPV: 92 B6.180 Bolts & Studs (Pumps) - Volumetric RR: 32 B6.190 Flange Surface, when connection disassembled (Pumps) Visual, VT-I RR: 2 B6.200 Nuts, Bushings, and Washers (Pumps) Visual, VT-i RR: 32 B-G-2 B7.50 Bolts, Studs, and Nuts (Piping) Visual, VT-I MS: 13 Pressure Retaining RPV: 3 Bolting, 2 in. and Less RR: 2 In Diameter RWCU: 1 B7.70 Bolts, Studs, and Nuts (Valves) Visual, VT-i CSA: 3 CSB: 3 FWA: 3 FWB: 3 HPPCI: 2 MS: 8 RHRA: 3 RHRB: 3 RR: 6 RWCU: 3 SDC: 2 7-16 Revision 0

ISI Program Plan _

Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-K B10.10 Welded Attachments to Pressure Vessels Surface - RPV: 9 Welded Attachments for B10.20 Welded Attachments to Piping Surface CSA: 2 Vessels, Piping, Pumps, CSB: 1 and Valves FWA: 1 FWB: I IHPCI: 1 MS: 13 RHRA: 1 RHRB: 1 RR: 4 SDC: 1 B10.30 Welded Attachments to Pumps Surface - RR: 6 B10.40 Welded Attachments to Valves Surface RR: 2 B-L-2 B12.20 Pump Casings Visual, VT-3 RR: 2 Pump Casings B-M-2 B12.50 Valve Bodies, (NPS 4 or Larger) Visual, VT-3 CSA: 3 Valve Bodies CSB: 3 FWA: 3 FWB: 3 HPCI: 2 MS: 21 RHRA: 3 RI-RB: 3 RR: 6 RWCU: 3 SDC: 2 7-17 Revision 0

IST Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System B-N-1 B13.10 IVessel Interior Visual, VT-3 RPV: 1 Interior of Reactor Vessel I B-N-2 B 13.20 Interior Attachments Within Beltline Region Visual, VT-i RPV: 26 Welded Core Support Structures and Interior B13.30 Interior Attachments Beyond Beltline Region Visual, VT-3 RPV: 40 Attachments to _- _

Reactor Vessels B13.40 Core Support Structure -Visual,-VT-3 RPV: 1 B-O B14.10 Welds in CRD Housing Volumetric br RPV: 32 Pressure Retaining Welds in (10% of Peripheral CRD Housings) Surface Control Rod Housings 7-18 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number (Nst M B-P B 15.10 Reactor Vessel - System Leakage Test Visual, VT-2 CRD 14R-07 All Pressure B15.50 Piping - System Leakage Test Visual, VT-2 CS 14T-02 Retaining Components B15.60 Pumps - System Leakage Test Visual, VT-2 FW I4T-03 (Periodic) B15.70 Valves - System Leakage Test Visual, VT-2 HPCI MS RCIC RHR RR RWCU RX SBLC 7-19 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes Number Requirements Components by TAP Number (with Category Description)

System C-A CL.IO Shell Circumferential Welds Volumetric RHRA: 3 Pressure Retaining Welds RHRB: 3 in Pressure Vessels C1.20 Head Circumferential Welds Volumetric . RHRA: I RHRB: I 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 VIMHRB: 4 in Vessels C2.33 Nozzle-to-Shell (or Head) Welds with Reinforcing Plates - Visual, VT-2 ECCS: 4 when Inside of Vessel is Inaccessible for Vessels, Greater RHRA: 2 than 1/2" Nominal Thickness _ RHRB: 2 C-C C3.10 Welded Attachments to Pressure Vessels Surface RHRA: 4 Welded Attachments RBRB:4 -_

for Vessels, Piping, C3.20 Welded Attachments to Piping Surface CRD: 2 Pumps, and Valves CSA: 4 CSB'. 2 ECCS: 2 HPCI: I RHR: I RHRA: 3 RHRB:4 -4 C3.30 Welded Attachments to Pumps Surface CSA: I CSB: I 7 -20 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System C-H C7.10 Pressure Vessels - System Leakage Test Visual, VT-2 CRD I4R-05 All Pressure C7.30 Piping - System Leakage Test Visual, VT-2 CS 14R-06 Retaining Components C7.50 Pumps - System Leakage Test Visual, VT-2 ECCS I4R-07 (Periodic) C7.70 Valves - System Leakage Test Visual, VT-2 FW I4T-02 HPCI 14T-03 RCIC RHR RPV Head Flange RR SBLC 7-21 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Item Description Exam Total Number of Relief Request/ Notes Examination Category (with Category Description) Number Requirements Components by TAP Number System 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 Vessels - System Leakage Test Visual, VT-2 DGCW_ -14R-07 All Pressure D2.30 Piping - System Leakage Test Visual, VT-2 .HVAC 14T-01 Retaining Components D2.50 Pumps - System Leakage Test -Visual, VT-2 PS - 4T-02 (Periodic) D2.70 Valves - System Leakage Test Visual, VT-2 RHRSW 14T-03 D-B D2.20 Pressure Vessels - System Hydrostatic Test Visual, VT-2 DGCW 14R-07 All Pressure D2.40 Piping - System Hydrostatic Test Visual, VT-2 HVAC I4T-01 Retaining Components D2.60 Pumps - System Hydrostatic Test Visual, VT-2 PS I4T-02 (Interval) D2.80 Valves - System Hydrostatic Test Visual, VT-2 RHRSW 14T-03 7-22 Revision 0

IS1 Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0.2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number System E-A El.11 Containment Vessel; Accessible Surface Areas General Visual 67 Containment El.12 Containment Vessel; Accessible Surface Areas Visual, VT-3 229 Surfaces El.20 Vent System Accessible Surface Areas Visual, VT-3 140 E-C E4.11 Containment Surface Areas Visible Surfaces Visual, VT-1 None Containment Surfaces Requiring E4.12 Containment Surface Areas; Surface Area Grid, Minimum Volumetric None Augmented Examination Wall Thickness Location E-D E5.10 Seals Visual, VT-3 N/A CR-21 Seals, Gaskets & E5.20 Gaskets Visual, VT-3 N/A CR-21 Moisture Barriers E5.30 Moisture Barriers Visual, VT-3 4 E-G E8.10 Bolted Connections; Surfaces General Visual 55 CR-30 Pressure Retaining Bolting E8.20 Bolted Connections; Bolts and Nuts Torque/Tension N/A CR-24 E-P E9.10 Containment Vessel; Pressure Retaining Boundary Appendix J Torus, DW, Vents All Pressure E9.20 Containment Vessel; Containment Penetration Bellows Appendix J In accordance Retaining with Procedure Components ,,_QCTP 0130-01 E9.30 Containment Vessel; Airlocks Appendix J In accordance with Procedure QCTP 0130-01 E9.40 Containment Vessel; Seals and Gaskets Appendix J In accordance with Procedure QCTP 0130-01 7-23 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0.2 Item Description Exam Total Number of Relief Request/ Notes Examination Category (with Category Description) Number Requirements Components by TAP Number System F-A F1.10 Class 1 Piping Supports Visual, VT-3 CSA: 5 Supports CSB: 6 FWA: T 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 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 7-24 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval Unit 2 Inservice Inspection Sumimary Table 7.0-2 Examination Category Item Description Exam Total Number of Relief Request/ Notes (with Category Description) Number Requirements Components by TAP Number Systemn F-A F1.40 Supports Other Than Piping Supports Visual, VT-3 CSA: 1 Supports (Class 1, 2, 3, and MC) CSB: I (Continued) DGCW: 1 HPCI: 2 JPI: 2 RHRA: 10 RHRB: 10 RHRSW: 8 RPV: 9 RR: 12 N/A BWROG BWR Feedwater Nozzle and Control Rod Drive Return Line Volumetric FWA: 2 Augmented Nozzle Cracking Components (BWROG) .FWB: 2 Components BWRVIP IGSCC Management Program BWR Vessel Internals and Various In accordance with Piping Components (GE SIL's and RICSIL's) BWRVIP Program BWRVIP Intergranular Stress Corrosion Cracking (IGSCC) in BWR Volumetric CSA: 11 4

-75 Austenitic Stainless Steel Piping Components and CSB: 10 BWRVIP-75 "Vessel and Internals Project Technical Basis JPI: 10 for Revisions to Generic Letter 88-01 Inspection Schedules" RHRA: 11 RHRB: 15 RPV: 7 RR: 95 SDC: 13 0737 Leak testing and periodic visual examinations of systems VT-2 CS, HPCI, outside of primary containment which could contain highly MS, RHR, radioactive fluids during a serious transient or accident RCIC, RR

_(NUREG 0737) RWCU 7-25 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval Unit 2 Inservice Inspection Summary Table 7.0-2 Item Description Exam Total Number of Relief Request/ Notes Examination Category (with Category Description) Number Requirements Components by TAP Number System R-A 1 Risk Category 1 Elements See Notes FWA: 12 14R-02 3 Risk-Informed Piping FWB: 8 5 Examinations 2 Risk Category 2 Elements See Notes CSA: 9 14R-02 3 CSB: 10-- - 5 RHR: 4 RHRA: 62 RHRB: 79 SDC: 5 3 Risk Category 3 Elements See Notes FWA: 1 14R-02 3, 5 4 Risk Category 4 Elements See Notes - CSA: 14 14R-02 3 CSB: 14 5 ECCS: 53 HPCI: 17 MS: 118 RCIC: 7 RHRA: 36 RTRB: 32 RPV: 2 RR: 37 RWCU: 28 5 Risk Category 5 Elements See Notes RHR: 19 14R-02 3 RHRA: 14 -5 RHRB: 16 7-26 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 &-2, Fourth Interval Inservice Inspection Summary Table Program Notes Note # Note Summary The number of QCNPS Unit 1, 2, and 1 ISI snubber visual examinations are performed in accordance with the ASME OM Code, Subsection ISTID Program.

IWF-5200(c) and 1/2 supports identified includes snubbers for the visual examination of the integral and nonintegral attachments per Paragraphs ISTD Program will IWF-5300(c). The snubbers are scheduled and administratively tracked in the ISI Program; however, the ASME OM Code, Subsection be the mechanism for actually performing the visual examinations scheduled within the ISI Program.

listed in Table following a "+"

2 The Unit I population counts include those components that are common to both units (typically designated as "1/2") and are symbol.

The RISI program methodology 3 For the Fourth Inspection Interval, QCNPS's Class 1 and 2 piping inspection program will be governed by risk-informed regulations.

program scope will be implemented as an alternative to the 1995 is described in the EPRI Topical Report TR-l 12657, Rev. B-A and Code Case N-578-1. The RISI B-J welds and Class 2 C-F-1 and C-F-2 welds in accordance Edition with the 1996 Addenda of the ASME Section XI Code examination program for Class I B-F and with 10 CFR 50.55a(a)(3)(i).

4 IGSCC Category A welds subsumed into the RISI program.

individual piping structural element. See 5 Examination requirements within the RISI program are determined by the various degradation mechanisms present at each EPRI TR-112657, Rev. B-A and Code Case N-578-1 for specific exam method requirements.

7-27 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval 8.0 RELIEF REQUESTS FROM ASME SECTION XI This section contains relief requests written per 10 CFR 50.55a(a)(3)(i) for situations where alternatives to ASME Section XI requirements provide an acceptable level of quality and safety; per 10 CFR 50.55a(a)(3)(ii) for situations where compliance with a

ASME Section XI requirements results in a hardship or an unusual difficulty without compensating increase in the level of quality and 9afety; and per 10 CFR 50.55a(g)(5)(iii) for situations where ASME Section XI requirements are considered impractical.

The following NRC guidance was utilized to determin6 the correct 10 CFR 50.55a Paragraph citing'for QCNPS relief requests. 10 CFR 50.55a(a)(3)(i) and 10 CFR 50.55a(a)(3)(ii) provide alternatives to the requirements of ASME Section XI, while 10 CFR 50.55a(g)(5)(iii)I recognizes situational impracticalities.

10 CFR 50.55a(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.

10 CFR 50.55a(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.

10 CFR 50.55a(2)(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 "14R-XX" relief request is applicable to ISI and SPT.

8-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval TABLE 8.0-1 INSERVICE INSPECTION PROGRAM RELIEF REQUEST INDEX.

Sheet 1 of 2 Relief Revision Status2 (Program) Description/

Request Date3 Approval Summary1 I4R-01 0 Submitted (ISI) Inspection of Standby Liquid Control nozzle 1I 1/17/03 inner radius. Revision 0 Submitted.

I4R0-0 0 Submitted (ISI) Alternate Risk-Informed Selection and 1/17/03 Examination Criteria for Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping Welds. Revision 0 Submitted.

I4R-03 0 Submitted (1SI) Alternative Requirements to ASME Section 1/17/03 XI, Appendix VII, Subsubarticle VII-240, "Annual Training." Revision 0 Submitted.

I4R-04 0 Submitted (1SI) Alternative Requirements to Appendix VIII, 1/17/03 Supplement 4, "Qualification Requirements for the Clad/Base Metal Interface of Reactor Pressure Vessel." Revision 0 Submitted.

I4R-05 0 Submitted (SPT) Exemption from Pressure Testing Reactor 1/17/03 Pressure Vessel Head Flange Seal Leak Detection System. Revision 0 Submitted.

I4R-06 0 Submitted (SPIT) Continuous Pressure Monitoring of the 1/17/03 Control Rod Drive (CRD) System Accumulators.

Revision 0 Submitted.

14R-07 0 Submitted (SPT) Alternative Rules for Corrective Measures 1/17/03 if Leakage Occurs at Bolted Connections.

Revision 0 Submitted.

I4R-08 0 Submitted (ISI) Evaluation Criteria for Temporary 1 1/17/03 Acceptance of Flaws. Revision 0 Submitted.

8-2 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval TABLE 8.0-1 INSERVICE INSPECTION PROGRAM RELIEF REQUEST INDEX Sheet 2 of 2 Relief Revision Status2 (Program) Description/

Request Date3 Approval Summary1 14R-09 0 Submitted (ISI) Pressure Retaining Welds in Piping, Subject to 1/17/03 Appendix VIII, Supplenent 11. Revision 0 Submitted.

Note 1: The NRC grants relief requests pursuant to 10 CFR 50.55a(g)(6)(i) when Code requirements cannot be met and proposed alternatives do not meet the criteria of 10 CFR 50.55(a)(3). The NRC authorizes relief requests pursuant to 10 CFR 50.55a(a)(3)(i) if the proposed alternatives would provide an acceptable level of quality and safety or under (3)(ii) if compliance with the specified requirements would result in hardship or unusual difficulties without a compensating increase in the level of safety.

I Note 2: This column represents the status'of the latest revision. Relief Request Status Options: Authorized Approved for use in an NRC SER (See Note 1); Granted - Approved for use in an NRC SER (See Note 1);

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

8-3 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST 14R-01 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: 1

'Refe'ence: 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: N1O Unit 2:N1O CODE REOUIREMENT in Table IWB-2500-1.

IWB-2500 states that components shall be examined and tested as specified radius section of all Table IWB-2500-1 requires a volumetric examination to be perforied on the inner reactor pressure vessel nozzles each inspection interval.

BASIS , /

FOR RELIEF with the Pursuant to 10 CFR 50.55a(g)(5)(iii), relief is requested on the basis that compliance specified Code requirement has been determined to be impractical.

with an integral The Standby Liquid Control (SBLC) nozzle, as shown in Figure 14R-01.1, is designed located near the socket to which the boron injection piping is fillet welded. The SBLC nozzle is from the inside surface bottom of the vessel in an area which is inaccessible for ultrasonic examinations from the outside of the RPV. Therefore, ultrasonic examinations would need to be performed to travel through diameter of the RPV. As shown in Figure I4R-01.1, the ultrasonic beam would need These geometric and the full thickness of the vessel into a complex cladding/socket configuration.

performed on material reflectors inherent in the design prevent a meaningful examination from being the inner radius of the SBLC nozzle.

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

PROPOSED ALTERNATE EXAMINATION subject nozzles As an alternate examination, QCNPS will perform a VT-2 visual examination of the each refueling outage in conjunction with the Class 1 System Leakage Test.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST 14R-01 (Page 2 of 3)

APPLICABLE TIME PERIOD Program for Relief is requested for the fourth ten-year inspection interval of the Inservicý Inspection QCNPS Units 1 and 2.,

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ISI Program Plan ¶ Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-01 (Page 3 of 3)

FIGURE 14R-01.1 2 INCH STANDBY LIQUID CONTROL NOZZLE Vessel-- *- 6 1/4"---"J 3/4" R Min 6 3/8" 3 3.

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q ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page I of 5)

COMPONENT IDENTIFICATION Code Class: 1 and22 Examination Category: B-F, B-J, C-F-i, and C-F'2 I Item Number: B5.10, B5.20, B9.11, B9.21, B9.31, B9.32, B9.40, C5.1 1, C5.51, C5.70, and C5.81

==

Description:==

Alternate Risk-Informed Selection and Examination Criteria for Category B-F, B-J, C-F-1, and C-F-2 Pressure Retaining Piping

'Welds I Component Number: Pressure Retaining Piping

Reference:

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

2) W. H. Bateman (NRC) to G. L. Vine (EPRI) letter dated October 28, 1999 transmitting "Safety Evaluation Report Related to EPRI Risk-Informed Inservice Inspection Evaluation Procedure (EPRI TR- 112657, Revision B, July 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) A. J. Mendiola (NRC) to 0. D. Kingsley (Exelon) letter dated February 5, 2002 transmitting "Safety Evaluation of Third Interval Risk-Informed Inservice Inspection Program Relief Request" CODE REQUIREMENT Table IWB-2500-1, Examination Category B-F, requires volumetric and/or surface examinations, on all welds for Items B5.10 and B5.20.

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

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: I4R-02 (Page 2 of 5)

CODE REQUIREMENT (Continued)

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 cofiditions:

a. primary plus secondary stress intensity range of 2.4S, for ferritic steel and austenitic steel.

b. cumulative usage factor U of 0.4.

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

4. Additional piping welds so that the total number of circumferential butt welds, branch connections, or socket welds selected for examinati6n 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-i and C-F-2 require volumetric and/br surface examinations on a sample of welds for Items C5. 11, C5.5 1, C5.70, and C5.8 1. 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 (Category C-F-i) or of all carbon and low alloy steel welds (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 (Category C-F-i) or carbon and low alloy welds (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.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 3 of 5)

BASIS FOR RELIEF alternative Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that'the proposed provide an acceptable utilizing Reference 1 along 'with two enhancements from Reference 4 will level of quality and safety.,

Inspection As stated in "Safety Evaluation Report Related to EPRI Risk-Informed Inservice 2):

Evaluation Procedure (EPRI TR-1 12657, Revision B, July 1999)" (Reference "The staff concludes that the proposed RI-ISI program as described in EPRI an TR-1 12657, Revision B, is a'sound technical approach and will provide proposed acceptable level of quality and safety pursuant to 10 CFR 50.55a for the alternative to the piping ISI requirements with regard to the number of locations, locations of inspections, and methods of inspection."

of the Third Interval for The initial QCNPS RISI Program was submitted during the Third Period with EPRI both Units 1 and 2. This initial RISI program was developed in accordance program was approved TR-1 12657, Revision B-A, as supplemented by Code Case N-578-1. The on February 5, 2002 for use by the NRC via Safety Evaluation as transmitted to Exelon (Reference 5).

of ASME The transition from the 1989 Edition to the 1995 Edition with the 1996 Addenda currently approved Risk-Informed Section XI for QCNPS's Fourth Interval does not impact the of the new Code ISI evaluation process used in the Third Interval, and the requirements Plan.

edition/addenda will be implemented as detailed in the QCNPS ISI Program RISI Program was an The Risk Impact Assessment completed as part of the original baseline ASME implementation/transition check on the initial impact of converting from a traditional Interval ISI update, there is Section XI program to the new RISI methodology. For the Fourth currently approved no transition occurring between two different methodologies, but rather, the new interval. As such, the initial RISI methodology and evaluation will be maintained for the process and is not screening of the risk impact assessment is not a part of the living program required to be continually updated.

Evaluation and The actual evaluation and ranking procedure including the Consequence (Reference 5) RISI Degradation Mechanism Assessment processes of the currently approved Risk Categorization and Program remain unchanged and are continually applied to maintain the portions of the RISI Element Selection methods of EPRI TR-1 12657, Revision B-A. These to plant Program are reevaluated as major revisions of the site PRA occur and modifications Assessment, Risk configuration are made. The Consequence Evaluation, Degradation Mechanism applicable to the RISI Ranking, and Element S61ection steps define the living program process Program.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 4 of 5)

PROPOSED ALTERNATE PROVISIONS 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 10 CFR 50.55a(a)(3)(i). This original program along with these same two enhancements is currently approved for QCNPS's Third Inspection Interval as documented in Reference 5.'

The Fourth Interval RISI Program will be a continuation of the current application and will continue to be a living program as described in the Basis For Relief above. No changes to the evaluation methodology as currently implemented under EPRI TR-1 12657, Revision B-A, are required as part of this interval update. The following two enhancements will continue to be implemented.

In lieu of the evaluation and sample expansion requirements in Section 3.6.6.2, "RI-ISI Selected Examinations" of EPRI TR-1 12657, QCNPS will utilize the requirements of Subarticle -2430, "Additional Examinations" contained in Code Case N-578-1 (Reference 4). The alternative criteria for additional examinations contained in Code Case N-578-1 provides a more refined methodology for implementing necessary additional examinations.

To supplement the requirements listed in Table 4-1, "Summary of Degradation-Specific Inspection Requirements and Examination Methods" of EPRI TR-1 12657, QCNPS will utilize the provisions listed in Table 1, Examination Category R-A, "Risk-Informed Piping Examinations" contained in Code Case N-578-1 (Reference 4). To implement Note 10 of this table, paragraphs and figures from'the 1995 Edition with the 1996 Addenda of ASME Section XI (QCNPS's code of record for the Fourth Interval) will be utilized which parallel those referenced in the Code Case for the 1989 Edition. Table 1 of Code Case N 578-1 will be used as it provides a detailed breakdown for examination method and categorization of parts to be examined.

The QCNPS RISI Program, as developed in accordance with EPRI TR-1 12657, Rev. B-A (Reference 1), requires that 25% of the elements that are categorized as "High" risk (i.e., Risk Category 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 Code Case N-578-1.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-02 (Page 5 of 5)

PROPOSED ALTERNATE PROVISIONS '(Continued)'

In addition to this risk-informed evaluation, selection, and examination procedure, all ASME Section XI piping components, regardless of fisk classification, will continue to receive Code required pressure testing aý part of the current ASME Section)XI program. VT-2 visual examinations are scheduled in accordance with the QCNPS pressure testing program, which remains unaffected by the RISI program.

APPLICABLE TIME PERIOD I Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for QCNPS Units 1 and 2. . 1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: I4R-03 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: AlI

Reference:

ASME Section XI, Appendix VII, Subsubarticle VIH-4240, "Annual Training" Exarination Category: All categories for components subject to Ultrasonic Examinatiorn Item Number: All item numbers for components subject to Ultrasonic Examination

==

Description:==

Alternative Requirements to ASME Section XI, Appendix VII, I Subsubarticle VII-240, "Annual Training" Component Number: All Components Subject to Ultrasonic Examination CODE REQUIREMENT 10 CFR 50.55a, "Codes and Standards," Paragraph (b)(2) incorporates by reference, the 1995 Edition and Addenda through 1996 of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code for use in preparing inservice inspection programs. Subsubarticle VII-4240, "Annual Training," of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII, requires a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months annual training.

10 (7R50.55a(b)(2)(xiv), "Appendix VIII personnel qualification," requires that all plrsonnel qualified to perform ultrasonic examinations in accordance with ASME Section XI, Appendix VIII, shall receive 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of annual hands-on training on specimens that contain cracks. This training must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from the training provision of Subsubarticle VII-4240 of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII, that requires a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months annual training. The basis of the relief request is that the proposed alternative would provide an acceptable level of quality and safety.

On September 22, 1999, the NRC published a final rule in the Federal Register (64 FR 51370) to amend 10 CFR 50.55a(b)(2), to incorporate by reference the 1995 Edition and addenda through the 1996 Addenda, of ASME Section XI. The change included the requirement to have a minimum of 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of annual training contained in Subsubarticle VII-4240 of ASME Section XI.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-03 (Page 2 of 3)

BASIS FOR RELIEF (Continued)

Additionally, the September 22, 1999, Federal Register notice amended requires that all personnel 10 CFR 50.55a(b)(2)(xiv). The amended 10 CFR 50.55a(b)(2)(xiv)

VIII shall receive 8 qualified to perform ultras9nic examinations in accordance with Appendix This training must be taken hours of annual hands'on training on specimens that contain cracks.

facility. Paragraph no earlier than 6 months prior to performing examinations at Alicensee's statement which includes a' 2.4.1.1.1 in the Federal Register'notice contained the following Demonstration Initiative discussion of the Electric Power Research Institute (EPRI) Performance (PDI) program.

VII-4240) was inadequate for "The NRC had determined that this requirement (i.e., Subsubarticle laboratory work and two reasons. The first reason was that the training does not require and, as detailed in the examination of flawed specimens. Signals can be difficult to interpret that the examiner must regulatory analysis for this rulemaking, experience and studies indicate interpretation. The second practice on a frequent basis to maintain the capability for proper have shown that an reason is related to the length of training and its frequency. Studies 6 months if skills are not examiner's capability begins to diminish within approximately training is not sufficient maintained. Thus, the NRC had determined that 10 hours1.157407e-4 days 0.00278 hours 1.653439e-5 weeks 3.805e-6 months of annual a more frequent basis to practice to maintain skills, band that an examiner must practice on for 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months of training, maintain proper skihl level... The PDI program has adopted 'a requirement must be taken no earlier than 6 but it is required to be hands-on practice. In addition, the training PDI believes that 8 hours9.259259e-5 days 0.00222 hours 1.322751e-5 weeks 3.044e-6 months will be months prior to performing examinations at a licensee's facility.

skill area because personnel acceptable relative to an examiner's abilities in this highly specialized and other pertinent technical can gain knowledge of new developments, material failure modes, in the Final Rule the PDI topics through other means. Thus, the NRC has decided to adopt 50.55a(b)(2)(xiv) of the final position on this matter. These changes are reflected in 10 CFR rule."

VII-4240 of ASME Implementation of the training requirements contained in Subsubarticle and 10 CFR 50.55a(b)(2)(xiv)

Section XI, 1995 Edition with the 1996 Addenda, Appendix VII Relief Request, to qualify our will result in redundant training programs. The approval of this 10 CFR 50.55a(b)(2)(xiv), will personnel to perform ultrasonic examinations in accordance with provide an acceptable level of simplify record keeping, satisfy the need to maintain skills, and quality and safety.

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ISI Program Plan I Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-03 (Page 3 of 3)

PROPOSED ALTERNATIVE PROVISIONS Annual ultrasonic training shall be conducted in accordance with 10 CFR 50.55a(b)(2)(xiv) in lieu of Subsubarticle VII-4240 of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VII. The annual ultrasonic training shall require that all personnel qualified for performing ultrasonic examinations in accordance with ASME Section XI, Appendix VIII, shall receive 8' hours of annual hands-on training on specimens that contain cracks. This training must be completed po earlier than 6 months prior to performing ultrasonic examinations at a licensee's facility.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection Program for QCNPS Units 1 and 2.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-04 (Page 1of2)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

ASME Section XI, Table IWB-25004l ASME Section XI, Appendix VIII, Supplement 4, Subparagraph 3.2(c)

Examination Category: B-A Item Number: B1?I0, 1B1.11, Bl.12, B1.20, B1.21,B11.22, B1.50, B1.51 4,

==

Description:==

Alternative Requirements to Appendix VIII, Supplement of "Qualification Requirements for the Clad/Base Metal Interface Reactoi Pressure Vessel" Component Numbers: All Components Subject to Ultrasonic Examination CODE REQUIREMENT with the 1996 Addenda of 10 CFR 50.55a(b)(2) incorporates by reference, the 1995 Edition programs.

ASME Section XI for use in preparing inservice inspection Supplement 4, requires that the Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, be plotted on a two dimensional plot ultrasonic testing (UT) perf6rmance demonstration results the true ,depth plotted along the with the measured depth plotted along the ordinate axis and parameters identified in abscissa axis. For qualification, the plot must satisfy the statistical Subparagraph 3.2(c).

BASIS FOR RELIEF the statistical parameters identified in Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested from Supplement 4. The basis of the relief Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, acceptable level of quality and safety.

request is that the proposed alternatives would provide an Federal Register (64 FR 51378) to On September 22, 1999, the NRC published a final rule in the the 1995 Edition and addenda through amend 10 CFR 50.55a(b)(2), to incorporate by reference included the provisions of Subparagraph the 1996 Addenda, of ASME Section XI. The change with the 1996 Addenda, Appendix 3.2(a), 3.2(b) and 3.2(c) of ASME Section XI, 1995 Edition VIII, Supplement 4.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-04 (Page 2 of 2)

BASIS FOR RELIEF (Continued)

Additionally, the September 22, 1999, Federal Register amended 10 I CFR 50.55a(b)(2)(xv)(C)(1).

The amended 10 CFR 50.55a(b)(2)(xv)(C)(1), requires a depth sizing acceptance criterion of 0.15 inch Root Mean Square (RMS) to be used in lieu of the requirements of Subparagraph 3.2(a) and 3.2(W of ASME Section XI, Appendix VIII, Supplement 4.

On March 26, 2001, the NRC published a correction to the September 22, 1999, final rule in the Federal Register (66 FR 16390). The NRC identified that an error had occurred in the published wording of 10 CFR 50.55a(b)(2)(xv)(C)(1). The corrected 10 CFR 50.55a(b)(2)(xv)(C)(1),

requires a depth sizing acceptance criterion of 0.15 inch Root Mean Square (RMS) to be used in lieu of the requirements of Subparagraph 3.2(a) and a length'sizing requirement of 0.75 inch RMS to be used in lieu of the requirements 3.2(b) of ASME Section XI, Appendix VIII, Supplement 4.

The statistical parameters to be used in flaw sizing specified in Subparagraph 3.2(c) of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VIII, Supplement 4, rely upon the depth'sizing acceptance criteria used in Subparagraph 3.2(a) and the length sizing acceptance criteria used in Subparagraph 3.2(b). For Supplement 4 UT performance demonstrations, the linear regression line of the data required by Subparagraph 3.2(c) is not applicable because the performance demonstrations are performed on test specimens with flaws located on th6 inner 15%

through-wall. Additionally, the Subparagraph 3.2(c) specified value for evaluating the mean deviation of flaw depth is not restrictive enough for evaluating flaw depths within the inner 15%

of wall thickness. We propose to use the 10 CFR 50.55a(b)(2)(xv)(C)(1) RMS calculations of Subparagraph 3.2(a), which utilizes an RMS value of 0.15 inch depth, and the RMS calculations of Subparagraph 3.2(b), which utilizes an RMS value of 0.75 inch length, in lieu of the statistical parameters of 3.2(c).

PROPOSED ALTERNATIVE PROVISIONS The RMS calculations of Subparagraph 3.2(a) of ASME Section XI, Appendix VIII, Supplement 4, which utilize an RMS value of 0.15 depth and the RMS calculations of Subparagraph 3.2(b),

which utilizes an RMS value of 0.75 length shall be used in lieu of the statistical parameters of Subparagraph 3.2(c) of ASME Section XI, Appendix VIII, Supplement 4.

APPLICABLE TIME PERIOD Relief is requested for the fourth ten-year inspection interval of the Inservice Inspectipn Program for QCNPS Units 1 and 2.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 1 of 5)

COMPONENT IDENTIFICATION Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H

==

Description:==

Exemption From Pressure Testing Reactor Pressure Vessel Head Flange Seal Leak Detection Systen.

Component Number: flange Seal Leak Detection Line Pressure Retaining Components.

CODE REQUIREMENTS I during a system leakage Table IWC-2500-1 requires a Visual VT-2 examination to be performed test.

BASIS FOR RELIEF the proposed alternatives Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that provide an acceptable level of quality and'safety.

from the reactor pressure The Reactor Pressure Vessel Head Flange Leak Detection Line is separated on the vessel flange. A second 0 boundary by one passive mrnembrane, a silver plated 0-ring located Figure 14R-05.2). This line is ring is located on the opposite side of the tap in the vessel flange (See seal O-ring. Failure of the required during plant operation in order to indicate failure of the inner flange room. On this O-ring would result in the annunciation of a High Level Alarm in the control from the O-ring and then isolate annunciation, control room operators would quantify the leakage rate valve (see Figure the leak detection line from the drywell sump by closing the AO 1(2)-220-51 O-ring and the vessel flange.

14R-05.1). This action is taken in order to prevent steam cutting of the Failure of the inner O-ring is the only condition under which this line is pressurized.

head is removed because the The configuration of this system precludes manual testing while the vessel small size of the tap and the high odd configuration of the vessel tap (See M4R-05.2), combined with the from being temporarily test pressure requirement (1000 psig minimum), prevents the tap in the flange and is smooth walled making a plugged. The opening in the flange is only 3/16 of an inch in diameter cause ejection of the high pressure temporary seal very difficult. Failure of this seal could possibly device used for plugging into the vessel.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 2 of 5)

BASIS FOR RELIEF (Continued) of the top A pneumatic test performed with the head installed is precluded due to the configuration O-rings are held in head. The top head of the vessel contains two grooves that hold the O-rings. The in a recessed cavity placý by a series of retainer clips spaced 150 apart. The retainer clips are contained the head on, the inner in th4 top head (see Figure 14R-05.3). If a pressure test was performed with operation. This test O-ring would be pressurized in a direction opposite to what it would see in normal into the recessed pressure wquld result in a net inward force on the O-ring that would tend to push it a silver plating cavity that houses the retainer clips. The O-ring material is only .050" thick with this deformation into the recessed thickness of .004" to .006" and could very likely be damaged by areas on the top head.

it is also In addition to the problems associated with the O-ring design that preclude this testing the line will initially questionable whether a pneumatic test is appropriate for this line. Although rate by measuring the level contain steam if the inner O-ring leaks, the system actually detects leakage at the level switch.

of condensate in a collection chamber. This would make the system mredium water represent an Finally, the use of a pneumatic test performed at a minimum of 1000 psig would of a test failure, due unnecessary risk in safety for the inspectors and test engineers in the unlikely event to the large amount of stored energy contained in air pressurized to 1000 psig.

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

for system Based on the above, QCNPS requests relief from the ASME Section XI requirements System.

leakage testing of the Reactor Pressure Vessel Head Flange Seal Leak Detection PROPOSED ALTERNATE EXAMINATION during a refueling A VT-2 visual examination will be performed on the line during vessel flood-up flood-up will allow outage. The static head developed due to the water above the vessel flange during with the for the detection of any gross indications in the line. This examination will be performed inspection period).

frequency specified by Table IWC-2500-1 for a System Leakage Test (once each APPLICABLE TIME PERIOD Program for Relief is requested for the fourth ten-year inspection interval of the Inservice Inspection QCNPS Units 1 and 2.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 3 of 5)

FIGURE 14R-05.1' HEAD FLANGE SEAL LEAK DETECTION SCHEMATIC S..~r 14R-C5.2 AO C2)-22D-5%-.c LI?

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ISI Program Plan Fourth Interval Quad Cities Nuclear Power Station Units I & 2, RELIEF REQUEST NUMBER: 14R-05 (Page 4 of 5)

FIGURE 14R-05.2 FLANGE SEAL LEAK DETECTION LINE DETAIL II erRing flange DetHt,,h antor' g re*lreLeak Top

Inner Flange S1eat Ring

-See Detoa A Detail "A" Vessel Sectional Flanl e View Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-05 (Page 5 of 5) ,

FIGURE 14R-05.3' O-RING CONFIGURATION f SECTION A-A Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-06 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 2

References:

Table IWC-2500-1 Examination Category: C-H ItemiNumber: C7.10, C7.30, C7.50, C7.70

==

Description:==

Continuous Pressure Monitoring of the Control Rod Drive (CRD)

I System Accumulators.

Componbnt Number: CRD Accumulators and associate piping CODE REQUIREMENT FROM WHICH RELIEF IS REQUESTED Table IWC-2500-1 requires a Visual VT-2 examination to be performed during a system leakage test.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternatives provide an acceptable level of quality and safety.

As required by QCNPS 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 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 QCNPS procedures. If an accumulator requires charging more than twice in a thirty day period, then a leak check is performed to determine the cause of the pressure loss. When leakage is detected, corrective actions are taken to repair the leaking component as required by QCNPS 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 Visual VT-2 examination performed once per inspection period would not provide an increase in safety, system reliability, or structural integrity. In addition, performance of a Visual VT-2 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.

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1SI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-06 (Page 2 of 2)

BASIS FOR RELIEF (Continued) iequirements specified in Table Relief is requested from the Visual VT-2 examination System Accumulators on the basis that QCNPS IWC-2500-1 for the nitrogen side of the CRD VT-2 exceed the code requirement for a Visual Technical Specification ,Su-veillance requirements Examination.

PROPOSED ALTERNATE EXAMINATIONS will requirements of Table IWC-2500-1, QCNPS As an alternate to the Visual VT-2 examination for in conjunction with Technical Specifications perform continuous pressure decay monitoring including attached piping.

the nitrogen side of the CRD Accumfiulators APPLICABLE TIME PERIOD' Program inspection interval of the Inservice Inspection Relief is requested for the fourth ten-year for QCNPS Units 1 and 2.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-07 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: 1, 2, and 3

References:

IWA-5250(a)(2)

Examination Category: N/A Item Number: N/A

Description:

Alternative Rules for Corrective Measures if Leakage Occurs at Bolted Connections Component Number: All Pressure Retaining Bolted Connections CODE REQUIREMENT FOR WHICH RELIEF IS REQUESTED IWA-5250(a)(2) states that if leakage occurs at a bolted connection, one of the bolts shall be removed, VT-3 examined, and evaluated in accordance with IWA-3 100. The bolt selected shall be the one closest to the source of leakage. When the removed bolt has evidence of degradation, all remaining bolting in the connection shall be removed, VT-3 examined, and evaluated in accordance with IWA-3100.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Removal of pressure retaining bolting at mechanical connections for VT-3 visual examination and subsequent evaluation in locations where leakage has been identified is not always the most prudent course of action to determine condition of the bolting and/or the root cause of the leak. The requirement to remove, examine and evaluate bolting in this situation does not allow consideration of other factors which may indicate the condition of mechanical joint bolting. Other factors which should be considered in an evaluation of bolting condition when leakage has been identified at a mechanical joint include, but should not be limited to:'

"* Bolting materials

"* Corrosiveness of process fluid

"* Service age of joint bolting materials

"* Leakage location

"* Leakage history at connection

"* Visual evidence of corrosion at connection (connection assembled)

"* Plant / Industry studies of similar bolting materials in a similar environment

"* Condition and leakage history of adjacent components Revision 0

ISI Progiam Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-07 (Page 2 of 3)

BASIS FOR RELIEF (Continued) materials ýe.g., studs, bolts, An example at QCNPS is the complete replacement of bolting cases, when the nuts, washers, etc.) at mechanical joints during plant outages. In some leakage is identified at associated system process piping is pressurized during plant start-up, of the piping and these joints. The cause' of this leakage is often due to thermal expansion at the joint gasket. In most bolting materials at the joint and subsequent process fluid seepage Removal of any of of these cases, proper re-torquing of the joint bolting stops the leakage.

this situation. ASME the joint bolting to evaluate for corrosion wvould be unwarranted in this situation exists, and has Section XI Code Interpretation XI-1-92-01 has recognized that clarified that the requirements of IWA-5250(a)(2) do not apply.

PROPOSED ALTERNATE PROVISIONS with the methodology of Code Case QCNPS proposes the following' alteinative, consistent provide an equivalent level of N-566-2, to the requirements of IWA-5250(a)(2), which will condition at Class 1, 2, and 3 quality and safety when evaluating leakage and bolting material bolted connections.

one of the As an alternative to the to the requirements of Subparagraph IWA-5250(a)(2),

following requirements will be met for leakage at bolted conriections:

will be reviewed (a) The leakage will be stopped, and the bolting and component material for joint integrity as described in (c) below.

integrity and (b) If the leakage is not stopped, the QCNPS will evaluate the structural operability of consequences of continuing operation, and the effect on the system listed in (c) continued leakage. This engineering evaluation will include the considerations below.

of the bolting to (c) The evaluation of (a) and (b) above is to determine the susceptibility corrosion and failure. This evaluation will include the following:

(1) the number and service age of the bolts; (2) bolt and component material; (3) corrosiveness of process fluid; (4) leak location and system function; (5) leakage history at the connection or other system components; (6) visual evidence of corrosion at the assembled connection.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-07 (Page 3 of 3)

PROPOSED ALTERNATE PROVISIONS (Continued)

If any of the above parameters indicates a need for further examination, the corrective Action will be taken in accordance with IWA-5250(a)(2).

APPIICABLE TIME PERIOD Relief is reqtiested for the fourth ten-year inspection interval of the Inservice Inspection Program for QCNPS Units 1 and 2.

, /

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ISl Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST: 14R-08 (Page 1 of 2)

COMPONENT IDENTIFICATION Code Class: 2 and 3

Reference:

IWC-3122.3

,IWC-3132.3 IWC-3600 IWD-3000 Examination Category: N/A Item Number: N/A

==

Description:==

Evaluation Criteria for Temporary Acceptance of Flaws Component Number: Modeiate Energy Class 2 and 3 Piping CODE REQUIREMENTS detects flaws may IWC-3122.3 states that a componefit whose volumetric or surface examination analytical be acceptable for continued service without a repair/replacement activity if an Similar requirements for visual evaluation is performed in accordance with IWC-3600.

examinations are contained in IWC-3132.3.

Analytical In the 1995 Edition with the 1996 Addenda of ASME Section XI, IWC-3600, are in the course of preparation and Evaluation of Flaws, and IWD-3000, Acceptance Standards, state that the requirements of IWB may be used.

BASIS FOR RELIEF proposed alternatives Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the would provide an acceptable level of quality and safety.

Revision 13 of ASME Section XI Code Case N-513 is conditionally approved for use in which is Regulatory Guide 1.147; however, this Case is not applicable to the 1996 Addenda has since been QCNPS's code of record for the Fourth Inspection Interval. Code Case N-513-1 the 2001 Edition.

issued in Supplement 11 of the 1998 Edition and is currently applicable through This revision of the Code Case is not yet approved for use by the NRC.

methodology from Code Case N-513-1 revises the base case to expand the temporary acceptance Both cases provide Class 3 moderate energy piping to Class 2 and 3 moderate energy piping.

requirements which may be followed for temporary acceptance of flaws in ASME Section III, is limited to ANSI B31.1, and ANSI B31.7 piping designated as Class 2 or 3. This acceptance does not exceed moderate energy piping defined as piping whose maximum operating temperature The provisions of the 200'F and whose maximum operating pressure does not exceed 275 psig.

period of time until case demonstrate the integrity of the item containing the flaw for a limited appropriate repair/replacement or additional examination activities can be performed.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST: 14R-08 (Page 2 of 2)

PROPOSED ALTERNATE PROVISIONS for flaws in moderate energy Class When using analytical evaluation as the method of acceptance Case N-513-1 without performing a 2 or 3 piping, QCNPS will follow the provisions of Code and will remain in affect for a repair/replacement activity. This acceptance will be temporary limited time, not exceeding the time to the next scheduled outage.

contained in ASME Section XI QCNPS may implement this method or one of the other methods evaluation process be applied to to accept detected flaws; however, in no case will the temporary (a) components other than pipe or tube, (b) leakage through a gasket, prevention, or (c) threaded connections with nonstructural seal welds for leakage (d) degraded socket welds.

safety factors contained in Whenapplying the methods of Code Case N-513-1, the specific are consistent with those contained Paragraph 4.0 of the Case will be satisfied. These conditions in 10 CFR 50.55a(b)(2)(xiii) regarding the use of Code Case N-513.

APPLICABLE TIME PERIOD the Inservice Inspection Program Relief is requested for the' fourth ten-year inspection interval of for QCNPS Units 1 and 2.

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ISI Program Plan Quad Cities Nuclear Power Station Unit 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 1 of 14)

COMPONENT IDENTIFICATION Code Class: 1

Reference:

ASME Section XI, Appendix VIE, Supplement 11, "Qualification Requirements For Full Structural Overlaid Wrought Austenitic Piping Welds" B-J Examination Category:

Item Number: B9.11

==

Description:==

Pressure Retaining Welds in Piping, Subject to Appendix VIII, Supplemrent 11 (Note: Also Identified in NRC Generic Letter 88-01 as Category E)

Component Numbers: Weld Overlay Components Subject to Ultrasonic Examination CODE REQUIREMENT ,

within Appendix VIII, The Code requirements for which relief is requested are all contained metal flaws be cracks.

Supplement 11. For example, paragraph 1.1(d)(1), requires that all base shall be oriented Paragraph 1.1(e)(1) requires that at least 20% but less than 40% of the flaws requires that the rules of within +20 degrees of the pipe axial direction. Paragraph 1.1(e)(1) also be treated as single or IWA-3300 shall be used to determine whether closely spaced flaws should shall include at least 3 multiple flaws. Paragraph 1.1(e)(2)(a)(1) requires that a base grading unit weld and base metal in. of the length of the overlaid weld and the outer 25 percent of the overlaid grading units, at least 1 on both sides. Paragraph 1.1(e)(2)(a)(3) requires that for unflawed base of the base grading unit.

inch of unflawed overlaid weld and base metal shall exist on either side the overlay material Paragraph 1.l(e)(2)(b)(1) requires that an overlay grading unit shall include unit shall be and the base metal-to-overlay interface of at least 6 sq. in. The overlay grading that all extensions of rectangular, with minimum dimensions of 2 in. Paragraph 3.2(b) requires being intrusions into base metal cracking into the overlay material by at least 0.1 in. be reported as the overlay material.

right hand column of Specific Code requirements for which relief is requested are identified in the Table 14R-09.1.

BASIS FOR RELIEF proposed alternatives Pursuant to 10 CFR 50.55a(a)(3)(i), relief is requested on the basis that the provide an acceptable level of quality and safety.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 2 of 14)

BASIS FOR RELIEF (Continued) be cracks. 'As illustrated below, implanting Paragraph 1.1(d)(1), requires that all'base metal flaws flaw. While this may a crack requires excavation of the base material on at least one side of the in austenitic materials be satisfactory for ferritic materials, it does not produce a usable axial flaw must now travel becaile the sound beam, which normally passes only through base material, response. To resolve this through weld material on at least one side, producing an unrealistic flaw flaw mechanisms under issue, the PDI program revised this paragraph to allow use of alternative implantation of controlled conditions. For example, alternative flaws shall be limited to when semi-elliptical with a tip cracks precludes obtaining an effective ultrasonic response, flaws shall be in the detection and width of less than or equal to 0.002 inches, and at least 70% of the flaws sizing test shall be cracks and the remainder shall be alternative'flaws.

(.....,..... Mechanical fatigue crack in Base mnaterial interfere with detection Relief is requested to allow closer spacing of flaws provided they do not or disdrimination. The existing specimens used to date for qualification to the Tri-party than allowed by the (NRC/B WROG/EPRI) agreement have a flaw population density greater for all previous current Code requirements. These samples have been used successfully their use and provide qualifications under the Tri-party agreement program. To facilitate PDI Program has merged continuity from the Tri-party agreement program to Supplement 11, the the requirement for the Tri-party test specimens into their weld overlay program. For example:

excluded, instead using IWA-3300 for proximity flaw evaluation in paragraph 1.1(e)(1) was includes the indications will be sized based on their individual merits; paragraph 1.1(d)(1) detection or statement that intentional overlay fabrication flaws shall not interfere with ultrasonic modified to require that a characterization of the base metal flaws; paragraph 1.1 (e)(2)(a)(1) was weld, rather than 3 base metal grading unit include at least 1 in. of the length of the overlaid overlaid weld and inches; paragraph 1.1(e)(2)(a)(3) was modified to require sufficient unflawed reflections from adjacent base metal to exist on all sides of the grading unit to preclude interfering was modified flaws, rather than the 1 in. requirement of Supplement 11; paragraph 1.1 (e)(2)(b)(l) and the base metal to define an overlay fabrication grading unit as including the overlay material requirement of to-overlay interface for a length of at least 1 in., rather than the 6 sq. in.

grading units Supplement 11; and paragraph 1.1 (e)(2)(b)(2) states that overlay fabrication and unflawed base metal designed to be unflawed shall be separated by unflawed overlay material perimeter.

to-overlay interface for at least 1 in. at both ends, rather than around its entire Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 3 of 14)

BASIS FOR RELIEF (Continued)

Additionally, the requirement for axially oriented overlay fabrication flaws in paragraph 1.1(e)(1) was excluded from the PDI Program as an improbable scenario. Weld overlays are typically in applied using automated gas tungsten arc welding techniques with the filler metal being applied would also be a circumferential direction. Because resultant fabrication induced discontinuities expected to have major .dimensions oriented in the circumferential direction axial overlay fabrication flaws are unrealistic.'

base The PDI Program revised paragraph ,2.0 to permit the overlay fabrication flaw test and the metal flaw tests be performed separately.

The requirement in paragraph 3.2(b) for reporting all extensions of cracking into the overlay is in omitted from the PDI Program because it is redundant to the RMS calculations performed paragraph 3.2(c) and its presence adds confusion and ambiguity to depth sizing as required by paragraph 3.2(c). This also makes the weld overlay program consistent with the Supplement 2 depth sizing criteria.

There are, however, some additional changes that were inadvertently omitted from the Code Case. The most important change is paragraph 1.1(e)(2)(a)(1) where the phrase "and base metal on both sides," was inadvertently included in the description of a base metal grading unit. The PDI program intentionally excludes this requirement because some of the qualification samples include flaws on both sides of the weld. To avoid confusion several instances of the term "cracks" or "cracking" were changed to the term "flaws" because of the use of alternative flaw mechanisms. Additionally, to avoid confusion, the overlay thickness tolerance contained in paragraph 1.1 (b) last sentence, was reworded and the phrase "and the remainder shall be changes alternative flaws" was added to the next to last sentence in paragraph 1.1(d)(1). These are identified by bold print in the third column of Table II4R-09. 1.

PDI has submitted these changes as a Code Case and they have been approved, but the Code Case will not be published until later in 2002. A detailed comparison matrix (Table I4R-09.1) for between Supplement 11, the proposed ASME Section XI Code Case N-654 (provided information only), and the PDI Program provides supporting documentation. The first column identifies the current requirements in the 95 Edition and 96 Addenda of Supplement 11, while the second (middle) column identifies the changes made by the Code Case.

PROPOSED ALTERNATE EXAMINATIONS In lieu of the requirements of ASME Section XI, 1995 Edition with the 1996 Addenda, Appendix VIII, Supplement 11, QCNPS will use the PDI Program.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 4 of 14)

APPLICABLE TIME PERIOD Program Relief is requested for the fourth ten-year inspection interval of the Inservice Inspectio'n for QCNPS Units 1 and 2.

II Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 5 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTEN1TIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only) _

1.0 SPECIMEN REQUIREMENTS No Change No Change Qualification test specimens shall meet the requirements listed herein, unless a set of specimens is designed to accommodate specific limitations stated in the scope of the examination procedure (e.g., pipe size, weld joint configuration, access limitations). The same specimens may be used to demonstrate both detection and sizing qualification.

1.1 General. The specimen set shall conform to No Chanre No Change the following requirements.

(a) Specimens shall have sufficient volume to No Chanie No Chanze minimize spurious reflections that may interfere with the interpretation process.

(b) The specimen set shall consist of at least three No Change - (b) The specimen set shall consist of at least specimens having different nominal pipe diameters three specimens having different nominal pine and overlay thicknesses. They shall include the diameters and overlay thicknesses. They shall minimum and maximum nominal pipe diameters include the minimum and maximum nominal for which the examination procedure is applicable, pipe diameters for which the examination Pipe diameters within a range of 0.9 to 1.5 times a procedure is applicable. Pipe diameters within nominal diameter shall be considered equivalent. If a range of 0.9 to 1.5 times a nominal diameter the procedure is applicable to pipe diameters of 24 shall be considered equivalent. If the procedure in. or larger, the specimen set must include at least is applicable to pipe diameters of 24 in. or one specimen 24 in. or larger but need not include larger, the specimen set must include at least the maximum diameter. The specimen set must one specimen 24 in. or larger but need not include at least one specimen with overlay include the maximum diameter. The snecimen set thickness within -0.1 in. to +0.25 in. of the shall Include specimens with overlay thickness within maximum nominal overlay thickness for which the +0.1 In.of the minimum nominal overlay thickness and within -0.25 in.of the maxinum nominal overlay procedure is applicable. th____e____'orwih__heprocedu___sanlicable Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R.09 (Page 6 of 14)

TABLE I4R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(c) The surface condition of at least two specimens No Change No Change shall approximate the roughest surface condition for which the examination procedure is applicable.

(d) Flaw Conditions (1) Base metal flaws. All flaws must be in or near (1) Base metal flaws. All flaws must be in or (1) Base metal flaws. All flaws must be cracks in the butt weld heat-affected zone, open to the inside near the butt weld heat-affected zone, open to or near the butt weld heat-affected zone, open to the surface, and extending at least 75% through the the inside surface, and extending at least 75%

inside surface, and extending at least 75% through base metal wall. Intentional overlay fabrication through the base metal wall. Intentional the base metal wall. Flaws may extend 100% flaws shall not interfere with ultrasonic detection or overlay fabrication flaws shall not interfere with through the base metal and into the overlay characterization of the cracking. Specimens ultrasonic detection or characterization of the material; in this case, intentional overlay containing IGSCC shall be used when available. base metal flaws. Specimens containing fabrication flaws shall not interfere with ultrasonic At least 70 percent of the flaws in the detection and IGSCC shall be used when available. At least detection or characterization of the cracking. sizing tests shall be cracks. Alternative flaw 70 percent of the flaws in the detection and Specimens containing IGSCC shall be used when mechanisms, if used, shall provide crack-like sizing tests shall be cracks and the remainder available. reflective characteristics and shall be limited by the shall be alternative flaws. Alternative flaw following: mechanisms, if used, shall provide crack-like reflective characteristics and shall be limited by the following:

(a) Flaws shall be limited to when implantation of (a) Flaws shall be limited to when implantation cracks precludes obtaining a realistic ultrasonic of cracks precludes obtaining an effective response. ultrasonic response.

(b) Flaws shall be semi-elliptical with a tip width of (b) Flaws shall be semi-elliptical with a tip less than or equal to 0.002 inches. width of less than or equal to 0.002 inches."

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: I4R-09 (Page 7 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(2) Overlay fabrication flaws. At least 40% of the No Change No Change flaws shall be non-crack fabrication flaws (e.g.,

sidewall lack of fusion or laminar lack of bond) in the overlay or the pipe-to-overlay interface. At least 20% of the flaws shall be cracks. The balance of the flaws shall be of either type.

(e) Detection Specimens (1) At least 20% but less than 40% of the base (1) At least 20% but less than 40% of the base (1) At least 20% but less than 40% of the flaws metal flaws shall be oriented within-+20 deg. of the metal flaws shall be oriented within +20 deg. of shall be oriented within +20 deg. of the pipe axial pipe axial direction. The remainder shall be the pipe axial direction. The remainder shall be direction. The remainder shall be oriented oriented circumferentially. Flaws shall not be open oriented circumferentially. Flaws shall not be circumferentially. Flaws shall not be open to any to any surface to which the candidate has physical open to any surface to which the candidate has surface to which the candidate has physical or or visual access. physical or visual access.

visual access. The rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws.

(2) Specimens shall be divided into base and over- (2) Specimens shall be divided into base metal and (2) Specimens shall be divided into base metal lay grading units. Each specimen shall contain one overlay fabrication grading units. Each specimen and overlay fabrication grading units. Eich or both types of grading units. shall contain one or both types of grading units. specimen shall contain one or both types of Flaws shall not interfere with ultrasonic detection grading units. Flaws shall not interfere with or characterization of other flaws. ultrasonic detection or characterization of other flaws.

(a)(1) A base grading unit shall include at least 3 (a)(1) A base metal grading unit shall include at (a)(1) A base metal grading unit shall include in. of the length of the overlaid weld. The base least 1 in. of the length of the overlaid weld. The at least 1 in. of the length of the overlaid grading unit includes the outer 25% of the overlaid base metal grading unit includes the outer 25% of weld. The base metal grading unit includes the weld and base metal on both sides. The base the overlaid weld and base metal on both sides. outer 25% of the overlaid weld. The base grading unit shall not include the inner 75% of the The base metal grading unit shall not include the metal grading unit shall not include the inner overlaid weld and base metal overlay material, or inner 75% of the overlaid weld and base metal 75% of the overlaid weld and base metal base metal-to-overlay interface, overlay material, or base metal-to-overlay interface, overlay material, or base metal-to-overlay interface.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 8 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(a)(2) When base metal cracking penetrates into the (a)(2) When base metal cracking penetrates into the (a)(2) When base metal flaws penetrate into overlay material, the base grading unit shall overlay material, the base metal grading unit shall the overlay material, the base metal grading include the overlay metal within 1 in. of the crack not be used as part of any overlay fabrication unit shall not be used as part of any overlay location. This portion of the overlay material shall grading unit. fabrication grading unit.

not be used as part of any overlay grading unit.

(a)(3) When a base grading unit is designed to be (a)(3) Sufficient unflawed overlaid weld and base (a)(3) Sufficient unflawed overlaid weld and unflawed, at least I in. of unflawed overlaid weld metal shall exist on all sides of the grading unit to base metal shall exist on all sides of the grading and base metal shall exist on either side of the base preclude interfering reflections from adjacent flaws. unit to preclude interfering reflections from grading unit. The segment of weld length used in adjacent flaws.

one base grading unit shall not be used in another base grading unit. Base grading units need not be uniformly spaced around the specimen.

(b)(1) An overlay grading unit shall include the (b)(l) An overlay fabrication grading unit shall (b)(l) An overlay fabrication grading unit shall overlay material and the base metal-to-overlay include the overlay material and the base metal-to- include the overlay material and the base metal interface of at least 6 sq. in. The overlay grading overlay interface for a length of at least 1 in. to-overlay interface for a length of at least 1 in.

unit shall be rectangular, with minimum dimensions of 2 in.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 9 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.bollercode.orgfPDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only) _

(b)(2) An overlay grading unit designed to be (b)(2) Overlay fabrication grading units designed to (b)(2) Overlay fabrication grading units unflawed shall be surrounded by unflawed overlay be unflawed shall be separated by unflawed overlay designed to be unflawed shall be separated by material and unflawed base metal-to-overlay material and unflawed base metal-to-overlay unflawed overlay material and unflawed base interface for at least I in. around its entire interface for at least 1 in. at both ends. Sufficient metal-to-overlay interface for at least 1 in. at perimeter. The specific area used in one overlay unflawed overlaid weld and base metal shall exist both ends. Sufficient unflawed overlaid weld grading unit shall not be used in another overlay on both sides of the overlay fabrication grading unit and base-metal shall exist on both sides of the grading unit. Overlay grading units need not be to preclude interfering reflections from adjacent overlay fabrication grading unit to preclude spaced uniformly about the specimen. flaws. The specific area used in one overlay interfering reflections from adjacent flaws. The fabrication grading unit shall not be used in specific area used in one overlay fabrication another overlay fabrication grading unit. Overlay grading unit shall not be used in another overlay fabrication grading units need not be spaced fabrication grading unit. Overlay fabrication uniformly about the specimen. grading units need not be spaced uniformly about the specimen.

(b)(3) Detection sets shall be selected from Table (b)(3) Detection sets shall be selected from Table (b)(3) Detection sets shall be selected from Table VIII-S2-1. The minimum detection sample set is VIII-S2-1. The minimum detection sample set is VIII-S2-1. The minimum detection sample set five flawed base grading units, ten unflawed base five flawed base metal grading units, ten unflawed is five flawed base metal grading units, ten grading units, five flawed overlay grading units, base metal grading units, five flawed overlay unflawed base metal grading units, five flawed fabrication grading units, and ten unflawed overlay overlay fabrication grading ubits, and ten and ten unflawed overlay grading units. For each fabrication grading units. For each type of grading unflawed overlay fabrication grading units. For type of grading unit, the set shall contain at least unit, the set shall contain at least twice as many each type of grading unit, the set shall contain at twice as many unflawed as flawed grading units.

unflawed as flawed grading units. For initial least twice as many unflawed as flawed grading procedure qualification, detection sets shall include units. For initial procedure qualification, the equivalent of three personnel qualification sets. detection sets shall include the equivalent of To qualify new values of essential variables, at least three personnel qualification sets. To qualify one personnel qualification set is required. new values of essential variables, at least one personnel qualification set is required.

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ISI Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 10 of 14)

TABLE M4R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N.654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(f) Sizing Specimen (1) The minimum number of flaws shall be ten. At (1)The minimum number of flaws shall be ten.

(1) The minimum number of flaws shall be ten. At least 30% of the flaws shall be overlay fabrication At least 30% of the flaws shall be overlay least 30% of the flaws shall be overlay fabrication flaws. At least 40% of the flaws shall be cracks fabrication flaws. At least 40% of the flaws flaws. At least 40% of the flaws shall be cracks open to the inside surface. For initial procedure shall be open to the inside surface. For initial open to the inside surface. qualification, sizing sets shall include the procedure qualification, sizing sets shall include equivalent of three personnel qualification sets. To the equivalent of three personnel qualification qualify new values of essential variables, at least sets. To qualify new values of essential one nersonnel qualification set is required. variables, at least one personnel qualification set is required.

(2) At least 20% but less than 40% of the flaws No Change No Chanue shall be oriented axially. The remainder shall be oriented circumferentially. Flaws shall not be open to any surface to which the candidate has physical or visual access.

(3) Base metal cracking used for length sizing No Change (3) Base metal flaws used for length sizing demonstrations shall be oriented circumferentially. demonstrations shall be oriented circumferentially.

(4) Depth sizing specimen sets shall include at least No Change (4) Depth sizing specimen sets shall include at two distinct locations where cracking in the base least two distinct locations where flaws in the metal extends into the overlay material by at least base metal extend into the overlay material by at 0.1 in. in the through-wall direction. least 0.1 in. in the through-wall direction.

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IS1 Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 11 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only) 2.0 CONDUCT OF PERFORMANCE The specimen inside surface and identification shall The specimen inside surface and identification DEMONSTRATION be concealed from the candidate. All examinations shall be concealed from the candidate. All The specimen inside surface and identification shall be completed prior to grading the results and examinations shall be completed prior to shall be concealed from the candidate. All presenting the results to the candidate. Divulgence grading the results and presenting the results to examinations shall be completed prior to grading of particular specimen results or candidate viewing the candidate. Divulgence of particular the results and presenting the results to the of unmasked specimens after the performance specimer results or candidate viewing of candidate. Divulgence of particular specimen demonstration is prohibited. The overlay unmasked specimens after the performance results or candidate viewing of unmasked fabrication flaw test and the base metal flaw test demonstration is prohibited. The overlay specimens after the performance demonstration is may be performed separately. fabrication flaw test and the base metal flaw test prohibited. may be performed separately.

2.1 Detection Test. Flawed and unflawed grading units shall be Flawed and unflawed grading units shall be Flawed and unflawed grading units shall be randomly mixed. Although the boundaries of randomly mixed. Although the boundaries of randomly mixed. Although the boundaries of _ specific grading units shall not be revealed to the specific grading units shall not be revealed to specific grading units shall not be revealed to the candidate, the candidate shall be made aware of the the candidate, the candidate shiall be made aware candidate, the candidate shall be made aware of the type or types of grading units (base metal or overlay of the type or types of grading units (base metal type or types of grading units (base or overlay) that fabrication) that ara present for each specimen, or overlay fabrication) that are present for each are present for each specimen. specimen.

2.2 Length Sizing Test. No Chang~e (a) The length sizing test may be conducted separately or in conjunction with the detection test.

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TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(b) When the length sizing test is conducted in No Change No Change conjunction with the detection test and the detected flaws do not.satisfy the requirements of 1.1(0, additional specimens shall be provided to the candidate. The regions containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the length of the flaw in each region.

(c) For a separate length sizing test, the regions of No Change No Change each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the length of the flaw in each region.

(d) For flaws in base grading units, the candidate (d) For flaws in base metal grading units, the (d) For flaws in base metal grading units, the shall estimate the length of that part of the flaw that candidate shall estimate the length of that part of candidate shall estimate the length of that part is in the outer 25% of the base wall thickness. the flaw that is in the outer 25% of the base metal of the flaw that is in the outer 25% of the base wall thickness. metal wall thickness.

2.3 Depth Sizing Test. The candidate shall determine the depth of the flaw The candidate shall determine the depth of the For the depth sizing test, 80% of the flaws shall be in each region. flaw in each region.

sized at a specific location on the surface of the specimen identified to the candidate. For the remaining flaws, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region.

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ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 13 of 14)

TABLE 14R-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.orgtPDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only) 3.0 ACCEPTANCE CRITERIA Examination procedures are qualified for detection Examination procedures are qualified for 3.1 Detection Acceptance Criteria. when all flaws within the scope of the procedure are detection wheh all flaws Within the scope of the Examination procedures, equipment, and personnel detected and the results of the performance _ procedure are detected and the results of the are qualified for detection when the results of the demonstration satisfy the acceptance criteria of performance demonstration satisfy the acceptance performance demonstration satisfy the acceptance Table VIII-S2-l for false calls. Examination - - criteria of Table VHI-S2-1 for false calls.

criteria of Table VIII-S2-1 for both detection and equipment and personnel are qualified for detection Examination equipment and personnel are false calls. The criteria shall be satisfied separately when the results of the performance demonstration qualified for detection when the results of the by the demonstration results for base grading units satisfy the acceptance criteria of Table VIII-S2-1 for performance demonstration satisfy the acceptance and for overlay grading units. both detection and false calls. The criteria shall be criteria of Table VIII-S2-1 for both detection and satisfied separately by the demonstration results for false calls. The criteria shall be satisfied base metal grading units and for overlay fabrication separately by the demonstration results for base grading units, metal grading units and for overlay fabrication grading units.

3.2 Sizing Acceptance Criteria. No Change No Chanae Examination procedures, equipment, and personnel are qualified for sizing when the results of the performance demonstration satisfy the following criteria.

(a) The RMS error of the flaw length No Change (a) The RMS error of the flaw length measurements, as compared to the true flaw measurements, as compared to the true flaw lengths, is less than or equal to 0.75 inch. The lengths, is less than or equal to 0.75 inch. The length of base metal cracking is measured at the length of base metal flaws is measured at the 75% through-base-metal position. 75% through-base-metal position.

(b) All extensions of base metal cracking into the This requirement is omitted. This requirement is omitted.

overlay material by at least 0.1 in. are reported as being intrusions into the overlay material.

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IST Program Plan Quad Cities Nuclear Power Station Units I & 2, Fourth Interval RELIEF REQUEST NUMBER: 14R-09 (Page 14 of 14)

TABLE I41-09.1 SUPPLEMENT 11 - QUALIFICATION PROPOSED CODE CASE N-654 PDI PROGRAM:

REQUIREMENTS FOR FULL STRUCTURAL Extracted from: The Proposed Alternative to OVERLAID WROUGHT AUSTENITIC http://www.boilercode.org/PDF/bcOO-756R.pdf Supplement 11 Requirements PIPING WELDS (Provided for Information Only)

(c) The RMS error of the flaw depth measurements, (b) The RMS error of the flaw depth measurements, (b) The RMS error of the flaw depth as compared to the true flaw depths, is less than or as compared to the true flaw depths, is less than or measurements, as compared to the true flaw equal to 0.125 in. equal to 0.125 in. depths, is less than or equal to 0.125 in.

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151 Program Plan ISI Progr'am Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval

10.0 REFERENCES

The references used to develop this Inservice Inspection Program Plan include:

1) Code of Federal Regulations, Title 10, Part 50, Paragraph 50.55a, "Codes and Standards"

2) Code of Federal Regulations, Title 10, Part 50, Paragraph 2, "Definitions," the definition of "Reactor Coolant Pressure Boundary"

3) Code Of Federal Regulations, Title 10, Part 50, Appendix J, Primary Reactor Containment Testing for Water Cooled Power Reactors 1

4) ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Inservice Inspection of Nucledr Power Plant Components," the 1989 Edition with No Addenda

5) ASME Boiler and Pressure Vessel Code Section XI, Division 1, Subsections IWE and IWL, 1992 Edition with the 1992 Addenda

6) ASME Boiler and Pressure Vessel Code, Section XI, Division 1, "Inservice Inspection of Nuclear Power Plant Components," the 1995 Edition with the 1996 Addqnda /

7) ASME OM Code, 1995 Edition with the 1996 Addenda, Subsection ISTD, "Preservice and Inservice Examination and Testing of Dynamic Restraints (Snubbers) in Light-Water Reactor Nuclear Power Plants"

8) USAS B31.1.0-1967, "Power Piping"

9) SECY-96-080, Issuance of Final Amendment To 10 CFR 50.55a To Incorporate By Reference The ASME Boiler And Pressure Vessel Code (ASME Code),

Section XI, Division 1, Subsection IWE and IWL

10) Regulatory Guide 1.26, Revision 3, "Quality Group Classifications and Standards for Water-, Steam-, and Radioactive Waste- Containing Components of Nuclear Power Plants"

11) Regulatory Guide 1.147, Revision 13 "Inservice Inspection Code Case Acceptability, ASME Section XI Division 1"

12) Quad Cities Station Units 1 and 2 Updated Final Safety Analysis Report (UFSAR)

13) Quad Cities Station Units 1 and 2 Technical Specification (TS)

14) Quad Cities Station Units 1 and 2 Technical Requirements Manual (TRM) 10-1 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval

15) BWRVIP-75 "BWR Vessel and Internals Project Technical Basis for Revisions to Generic Letter 88-01 Inspection Schedules" as conditionally appro',ed by NRC SER (TAC NO. MA5012), dated September 15, 2000

16) NRC NUREG 0313, Revision 2, "Technical Reporton Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping"

'17) Generic Letter 88-01, Revision 2, dated January 25, 1988, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping"

18) Generic Letter 88-01, Supplement 1, dated February 4, 1992, "NRC Position on Intergranular Stress Corrosion Cracking (IGSCC) in BWR Austenitic Stainless Steel Piping"

19) BWROG "Alternate BWR Feedwater Nozzle Inspection Requirements," dated October 1995

20) NRC NUREG 0737, dated November 1980, TMI Action Plan Requirements"

21) NEP-17-09, Revision 0; "IGSCC Management Program For BWR Reactor Internals", dated July 21, 1998

22) Assessment of Quad Cities Station IGSCC Management Program for BWR Reactor Internals and Piping per the requirements of NEP-17-09, dated December 22, 1998

23) IGSCC Management Program Manual For BWR Vessel Internals and Piping, Revision 0, dated August 28, 1998

24) Quad Cities Appendix J Leak Rate Testing Program QCTP 0130-01

25) EPRI Containment Inspection Program Guide (TR-1 10698-Ri)

26) Policy for Implementation of ASME IWE-5240 Visual Examination dated August 14, 1998 File 2-1.9

27) Maintenance Inspection Of Existing Level I Coatings Systems, SPP CI-1

28) EPRI Topical Report TR-1 12657, Rev. B-A, Final Report, "Revised Risk Informed Inservice Inspection Evaluation Procedure," July 1999

29) NRC SER related to EPRI Topical Report TR- 112657, Rev. B, Final Report, "Revised Risk-Informed Inservice Inspection Evaluation Procedure, July 1999,"

dated October 28, 1999 10-2 Revision 0

ISI Program Plan Quad Cities Nuclear Power Station Units 1 & 2, Fourth Interval

30) CornEd Risk-Informed Inservice Inspection Project "Definition ofRISI Scope for QCNPS Units 1 and 2, Revision 1," dated April 17, 2000

31) CornEd Risk-Informed Inservice Inspection Evaluation (Final Report) for QCNPS Units 1 and 2

32) Quad Cities Station Units 1 and 2 ISI Classification Basis Document, Fourth Ten Year Inispection Interval

33) QCNPS'Units 1 and 2 ISI Selection Document, Fourth Ten-Year Inspection Interval

34) ER-AA-330, "Conduct of Inservice Inspection Activities"

35) ER-AA-330-001, "Section XI Pressure Testing"

36) ER-AA-330-002, "Ihservice Inspection of Welds and Components"

37) ER-AA-330-003, "Visual Examination of Section XI Component Supports"

38) ER-AA-330-004, "Visual Examination of Technical Specification Snubbers"

39) ER-AA-330-007, "Visual Examination of Section XI Class MC Surfaces and Class CC Liners"

40) ER-AA-330-009, "ASME Section XI Repair/Replacement Program"

41) ER-AA-330-010; "Snubber Functional Testing"

42) ER-AA-335-018, "General, VT-1, VT-IC, VT-3, and VT-3C, Visual Examination of ASME Class MC and CC Containment Surfaces and Components"

43) General Electric Boiling Water Reactor System Department, Document No.

22A2750, Revision 6

44) QCNPS Reactor Coolant Pressure Boundary Normal Makeup Calculation, XCE.040.0202

45) QCNPS Design Analysis No. QDC-0200-M-1279, "ISI/RCPB Normal Makeup for CRD Housing Welds" 10-3 Revision 0

SVP-03-019, Fourth Interval Inservice Inspection Program Plan

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1.0 INTRODUCTION AND BACKGROUND

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