Monticello, Fifth Ten-Year lnservice Inspection Plan. · 2012-07-19 · Subject: Fifth Ten-Year lnservice Inspection Plan Pursuant to 10 CFR 50.55a(g)(5)(i), Northern States Power

@ Xcel EnergyB

February 28,2012

Monticello Nuclear Generating Plant 2807 W County Rd 75 Monticello, MN 55362

L-MT-12-018 10 CFR 50.55a(g)

U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001

Monticello Nuclear Generating Plant Docket 50-263 Renewed License No. DPR-22

Subject: Fifth Ten-Year lnservice Inspection Plan

Pursuant to 10 CFR 50.55a(g)(5)(i), Northern States Power Company, a Minnesota corporation, d/b/a Xcel Energy, the licensee for the Monticello Nuclear Generating Plant (MNGP), submits its fifth ten-year interval lnservice lnspection (ISI) Program Plan (enclosure). This IS1 Plan for the fifth ten-year interval begins September 1, 2012, and, pursuant to 10 CFR 50.55a(g)(4)(ii) and 10 CFR 50,55a(b)(2), will comply with the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section XI, 2007 Edition with the 2008 Addenda and 10 CFR 50.55a.

Should you have questions regarding this letter, please contact Mr. Randy Rippy at (61 2) 330-691 1.

Summarv of Commitments /' ew commitments and no revisions to existing commitments.

Monticello Nuclear Generating Plant Northern States Power Company-Minnesota

Enclosure

Document Control Desk Page 2

cc: Administrator, Region Ill, USNRC Project Manager, Monticello Nuclear Generating Plant, USNRC Resident Inspector, Monticello Nuclear Generating Plant, USNRC

ENCLOSURE

MONTICELLO NUCLEAR GENERATING PLANT

INSERVICE INSPECTION (ISI) PLAN REVISION 0

FIFTH TEN-YEAR INSPECTION INTERVAL

76 pages follow

Monticello Nuclear Generating Plant sth Interval lnservice Inspection Plan

XCEL Energy, Inc.

NSP-Minnesota

414 Nicollet Mall

Minneapolis, M N 55401

Monticello Nuclear Generating Plant

lnservice lnspection (ISI) Plan

Revision 0

Fifth Ten-Year lnspection Interval

September 1,2012 through May 31,2022

Commercial Service Date

June 30,1971

Monticello Nuclear Generating Plant

2807 West Highway 75

Monticello, Minnesota 55362

Rev. 0 Page i See PCR-01319580 for Approvals

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan

Rev. 0 Page ii See PCR‐01319580 for Approvals

RECORD OF REVISIONS Page Rev.* ‐ 0


Rev. 0 Page iii See PCR‐01319580 for Approvals

RECORD OF REVISIONS Summary of Changes, Plan Revision 0

This is the initial issue of the Fifth Interval Inservice Inspection (ISI) Plan


Rev. 0 Page iv See PCR‐01319580 for Approvals

Contents 1.0 INTRODUCTION................................................................................................................................. 1

2.0 Background for Plan/Schedule Development................................................................................... 6

3.0 Application Criteria and Code Compliance ....................................................................................... 6

4.0 Examination Personnel/Procedures ...............................................................................................16

5.0 Reporting of Associated Section XI Programs.................................................................................16

6.0 Augmented and Owner Programs ..................................................................................................16

7.0 License Renewal Aging Management Plans and Commitments.....................................................20

8.0 Source Documents ..........................................................................................................................22


Rev. 0 Page 1 See PCR‐01319580 for Approvals

1.0 INTRODUCTION

Background: 10 CFR 50.55a requires that an inservice inspection (ISI) program be developed at 10 year (120 month) intervals. The ISI Program is prepared and maintained by Xcel Energy Inc. for the Monticello Nuclear Generating Plant. This program has been developed to comply with the American Society of Mechanical Engineers (ASME) Section XI, Rules for Inservice Inspection of Nuclear Power Plant Components and implements the requirements of Updated Safety Analysis Report (USAR) 13.4.6 “10 CFR 50.55a Inservice Inspection and Testing Programs.” The Inservice Testing (IST) Program is maintained separately from this program and is submitted under separate cover. The Containment Inspection Program, as allowed by 10 CFR 50.55a(g)(6)(ii)(B), is not submitted, it is available at the plant site for audit and review. The Snubber Program and Boiling Water Reactor Internals Project (BWRVIP) Program are also maintained separately from this plan. This plan is developed to ensure the following:

1) Conformance to Title 10, Section 50.55a of the Code of Federal Regulations (10 CFR 50.55a) 2) Conformance to the 2007 Edition with the 2008 Addenda of Section XI of the American Society

of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code 3) Conformance to the Xcel Energy Inc. Corporate Policies and Procedures 4) The proper ASME Section XI Code required examinations, tests, and administrative procedures

are implemented 5) The proper ASME Code request for alternatives and relief requests are submitted to and

approved by the regulatory authority 6) The proper examination, test and repair/replacement records and reports are maintained and

submitted The following descriptions provide the location of the following programs:

• Repair/Replacement Program is contained in Administrative Work Instruction (AWI) 4 AWI‐09.04.03 “ASME Section XI Repair/Replacement Program.”

• Containment Inservice Inspection Program is contained in a separate document titled “Containment Inservice Inspection Plan (IWE Plan)”

• System Pressure Testing Program is contained in 4 AWI‐09.04.02 “System and Component Pressure Testing Program”

• Snubber Program is contained in EWI‐08.02.01 “Snubber Program”

• Boiling Water Reactor Vessel Internals Project (BWRVIP) Program is contained in EWI‐08.01.01 “Boiling Water Reactor Vessel Internals Project Administrative Manual”



The ASME Section XI Inservice Inspection Program is comprised of seven parts: ISI Plan (Introduction and Source Documents), Appendix A “Code Cases”, Appendix B “Relief Requests”, Appendix C.1 “List of ISI Boundary Drawings (These drawings outline the Quality Group Classifications, (A, B, and C)), Appendix C.2 “List of ISI Isometrics” (These drawings delineate the ASME Section XI components or items that are included in the examination program), Appendix D “Examination Schedule Tables”, and Appendix E “Risk‐Informed Living Program Updates”. 5th Ten‐Year Interval The Monticello 5th Ten‐Year Inservice Inspection Interval is the first interval into the extended license period. The regulations in 10 CFR 50.55a(g)(4) establish the effective ASME Code edition and addenda to be used by licensees for performing inservice inspections of components (including supports). Paragraph 50.55a(g)(4)(ii) requires the use of the latest edition and addenda that has been incorporated by 10 CFR 50.55a(b), one year prior to the beginning of each 120‐month ISI interval. This is considered the Code of Record. The Code of Federal Regulation in effect one year prior to the beginning of the Fifth Interval was 76 FR 36232 with an effective date of July 21, 2011. This CFR incorporated, by reference, the ASME Section XI, 2007 Edition with the 2008 Addenda in paragraph (b)(2) with conditions. As stated in this CFR, the 5th Ten‐Year ISI Program is based on the 2007 Edition with the 2008 Addenda of the ASME, Boiler and Pressure Vessel Code, Section XI. However, the following conditions are also required to be met along with Section XI:

1) 10 CFR 50.55a(b)(2)(xii), the provisions in IWA‐4660, “Underwater Welding” of Section XI, 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section, are not approved for use on irradiated material.

2) 10 CFR 50.55a(b)(2)(xiv), licensees applying the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section may use the annual practice requirements in VII‐4240 of Appendix VII of Section XI in place of the 8 hours of annual hands‐on training provided that the supplemental practice is performed on material or welds that contain cracks, or by analyzing pre‐recorded data from material or welds that contain cracks. In either case, training used must be completed no earlier than 6 months prior to performing ultrasonic examinations at a licensee’s facility.

3) 10 CFR 50.55a(b)(2)((xviii), requires that Level I and II nondestructive examination personnel be recertified on a 3‐year interval in lieu of the 5‐year interval specified in IWA‐2314(a) & (b) of the 1999 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section.

4) 10 CFR 50.55a(b)(2)(xix), does not approve the provisions in IWA‐4520(b)(2) and IWA‐4521 allowing the use of ultrasonic examination for radiographic examination specified in the Construction Code.



5) 10 CFR 50.55a(b)(2)(xx)(B), requires that the implementation of IWA‐4540(a)(2) of the 2002

Addenda when performing system leakage tests after repair/replacement activities performed by welding or brazing on a pressure retaining boundary. IWA‐4520(a)(2) of the 2002 Addenda states: “The following requirements shall be met (a) the nondestructive examination methodology and acceptance criteria of the 1992 Edition or later of Section III shall be met prior to return to service, (b) the Owner’s Requirements shall be met prior to return to service, and (c) a system leakage test shall be performed in accordance with IWA‐5000”

6) 10 CFR 50.55a(b)(2)(xxii), prohibits the use of IWA‐2220 allowing the use of ultrasonic examination as a surface examination.

7) 10 CFR 50.55a(b)(2)(xxiii), prohibits the use of the provisions found in IWA‐4461.4.2 for eliminating mechanical processing of thermally cut surfaces.

8) 10 CFR 50.55a(b)(2)(xxv), prohibits the use of IWA‐4340 “Mitigation of Defects by Modification”

9) 10 CFR 50.55a(b)(2)(xxvi), requires the use of the 1998 Edition, IWA‐4540(c) for pressure testing of Class 1, 2, & 3 mechanical joints after a repair/replacement activity. IWA‐4540(c) of the 1998 Edition states: “Mechanical joints made in installation of pressure retaining items shall be pressure tested in accordance with IWA‐5211(a). Mechanical joints for component connections, piping, tubing (except heat exchanger tubing), valves, and fittings, NPS‐1 and smaller, are exempt from the pressure test.”

10) 10 CFR 50.55a(b)(2)(xxvii), requires insulation removal from 17‐4 PH or 410 stainless steel studs or bolts aged at a temperature below 1100°F or having a Rockwell Method C hardness value above 30, and from A‐286 stainless steel studs or bolts preloaded to 100,000 psi or higher on those systems borated for controlling reactivity when conducting pressure tests. For Monticello this would only be applicable to the Standby Liquid Control System.

11) 10 CFR 50.55a(b)(2)(xxviii), requires the following when using Nonmandatory Appendix A, A‐4300(b)(1) Equation (2):

For R<0, KI depends on the crack depth (a), and the flow stress (σf). the flow stress is defined by σf = ½ (σys + σult), where σys is the yield strength and σult is the ultimate tensile strength in units ksi (MPa) and a is in units in. (mm). For ‐2< R < 0 and Kmax – Kmin < 0.8 X 1.12 σf √(πa), S=1 and KI = Kmax. For R < ‐2 and Kmax – Kmin < 0.8 X 1.12 σf √(πa), S=1 and

KI = (1‐R) Kmax/3. For R < 0 and Kmax – Kmin > 0.8 X 1.12 σf √(πa), S=1 and KI = Kmax – Kmin.



12) 10 CFR 50.55a(b)(xxix), prohibits the use of Nonmandatory Appendix R “Risk‐Informed Inspection Requirements for Piping” without prior approval by the NRC.

The ISI Interval begins on September 1, 2012 and is scheduled to end on May 31, 2022. The inspection periods are scheduled as follows:

1st Period: From September 1, 2012 to August 31, 2015 (3 years) 2nd Period: From September 1, 2015 to August 31, 2019 (4 years) 3rd Period: From September 1, 2019 to May 31, 2022 (2 years 9 months)

In accordance with ASME Section XI, IWB‐2430(c)(3), that portion of an inspection interval described as an inspection period may be reduced or extended by as much as 1 year. This adjustment shall not alter the requirements for scheduling inspection intervals. 4th Ten‐Year Interval The Monticello 4th Ten‐Year Inservice Inspection Interval was slightly less than 120 months to regain a portion of the time period associated with an extension of the 3rd Interval, which had been extended through May 31, 2003 (L‐MT‐03‐004). The 3rd Interval overlapped the 4th Interval as permitted by IWA‐2430(d)(1), (2), (3), and (4). The 4th Interval start date was May 1, 2003 and ended August 31, 2012 (the 3rd period was extended as allowed by IWA‐2430(d)(1)). Five refueling outages were scheduled during the 4th Interval. The Code of Record for the 4th Interval was the 1995 Edition with the 1996 Addenda of ASME Section XI. 3rd Ten‐Year Interval The Monticello 3rd Ten‐Year Inservice Inspection Interval covered the time period between June 1, 1992 through May 31, 2003. The interval was extended 12 months per IWA‐2430 (Letters to the NRC in May 2002 and January 2003 providing notification of 3rd Interval extension initially through March 8, 2003 (M2002057). The Code or Record for the 3rd Interval was the 1986 Edition of ASME Section XI. 2nd Ten‐Year Interval The Monticello 2nd Ten‐Year Inservice Inspection Interval covered the time period between June 30, 1981 through May 30, 1992. The 2nd Interval was extended in accordance with IWA‐2400 due to the Recirculation Pipe replacement project that resulted in an eleven month shutdown in 1984. The Code of Record for the 2nd Interval was the 1977 Edition with the Summer 1978 Addenda of ASME Section XI. 1st Ten‐Year Interval The Monticello 1st Ten‐Year Inspection Interval covered the period between June 30, 1971 through June 29, 1981. At the beginning of the 1st Interval, the rules specified in 10 CFR 50.55a addressed only those components within the reactor coolant pressure boundary (RCPB) and required licensees to update the Edition and Addenda of ASME Code, Section XI each Inspection Period (i.e., every 40 months). In 1976, 10 CFR 50.55a was amended to require plants with operating licenses to examine and test components



that were classified as ASME Code Class 2 and 3. This provision of the regulation was to apply at the start of the regular 40‐month period. In 1979, 10 CFR 50.55a was again amended to endorse the 1978 Edition with the Summer 1978 Addenda of ASME Code, Section XI. This amendment changed the requirement for updating of the ISI Program to once every 120 months. Because of the rules in affect during the 1st Interval, MNGP was committed to three editions of the ASME Code, Section XI. These editions were the (1) 1971 Edition with the Summer 1972 Addenda, (2) 1971 Edition with the Summer 1973 Addenda, and (3) 1974 Edition with the Summer 1975 Addenda. Component Selection: With the exception of Class 1 and 2 piping welds, components were selected and scheduled using criteria in the 2007 Edition with the 2008 Addenda of ASME Section XI and 10 CFR 50.55a(g), except where relief has been granted by the Nuclear Regulatory Commission (NRC). Selection of Class 1 and 2 piping welds in ASME Categories B‐F, B‐J, C‐F‐1 and C‐F‐2 are based on Code Case N‐716 “Alternative Piping Classification and Examination Requirements.” Code Edition Summary: The code editions implemented in the 5th Interval ISI Program are summarized below: Class 1 (Quality Group A) 2007 Edition with the 2008 Addenda Class 1 Piping Welds(Quality Group A) Code Case N‐716 (Relief Request RR‐003) Class 2 (Quality Group B) 2007 Edition with the 2008 Addenda Class 2 Piping Welds (Quality Group B) Code Case N‐716 (Relief Request RR‐003) Class 3 (Quality Group C) 2007 Edition with the 2008 Addenda MC (Metal Containment) 2001 Edition with the 2003 Addenda Mandatory Appendix VIII 2001 Edition until January 21, 2013 at which time

the 2007 Edition with the 2008 Addenda will be implemented 10 CFR50.55a(g)(4)(ii) allows up to 18 months delay to update the Appendix VIII program

Pressure Testing Program for Class 1, 2, and 3 2007 Edition with the 2008 Addenda Pressure Testing Program for Class MC 2001 Edition with the 2003 Addenda Pressure Testing Program for Repair/Replacements 2007 Edition with the 2008 Addenda

10 CFR 50.55a(b)(2)(xx)(B) specifies the use of the 2002 Addenda, IWA‐4540(a)(2) after repair/replacement activities performed by welding or brazing on a pressure boundary

Pressure Testing Class 1, 2, & 3 Mechanical Joints 10 CFR 50.55a(b)(2)(xxvi) specifies the use of the 1998 Edition, IWA‐4540(c) after a repair/replacement activity

Repair/Replacement Program 2007 Edition with the 2008 Addenda (Ref. RR‐007 for Class MC Components)



2.0 Background for Plan/Schedule Development

The examination plan and schedule were developed from ASME Code requirements, Risk‐Informed Methodology, individual component examination history and plant scheduling needs such as optimizing insulation removal and scaffolding needs. During the 2nd Interval, a substantial number of component replacements and alterations were made (e.g. the recirculation piping replacement). The intent of the 5th Interval is to be consistent with the previous intervals, to the extent practical. Since Class 1 (Category B‐F and B‐J) and Class 2 (Category C‐F‐1 and C‐F‐2) piping welds were examined per the RI‐ISI Plan in the 4th Interval, the correlation is to the 4th Interval only.

3.0 Application Criteria and Code Compliance

ASME Section XI The following provides a summary of the application of ASME Code, Section XI, 2007 Edition with the 2008 Addenda to the Monticello Nuclear Generating Plant, Ten‐Year Program for the Fifth Inspection Interval. The application and distribution of examinations for this interval is based upon the requirements as defined in IWB‐2411, IWC‐2411, IWD‐2411, and IWF‐2410 of Section XI. Appendix D contains the examination schedule for the Fifth Interval and is summarized by ASME Category and Item Number. EXAMINATION CATEGORY B‐A, PRESSURE RETAINING WELDS IN REACTOR VESSEL Reactor vessel examinations are scheduled on the reactor pressure vessel to meet the alternative requirements of relief request RR‐001 “Extension of Permanent Relief from Volumetric Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term.” This alternative, previously approved for permanent use under 4th Interval 10 CFR 50.55a Request No. 17, is based on the Boiling Water Reactor Vessel Internals Project (BWRVIP) report BWRVIP‐05 “BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations,” and BWRVIP‐74, “BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Evaluation Guidelines for License Renewal.” This alternative provision is summarized below:

• The examination requirements of ASME Code Section XI, Table IWB‐2500‐1, Examination Category B‐A, Item No. B1.12, for the RPV longitudinal shell welds will be performed as required to the extent possible.

• The examination requirements for Item No. B1.11, RPV circumferential shell welds will be limited to the segment of the weld that intersects with the longitudinal weld.

• The examination requirement for circumferential weld VCBB‐1, in lieu of the intersection with the longitudinal weld approximately 2 to 3 percent of the weld at an accessible location will be completed.

• The examination may be performed from either the internal inside diameter surface or the external outside diameter surface.

• Examination of all remaining portions of the RPV circumferential welds will be deferred through the renewed operating license term as approved by Request for Alternative RR‐001.



• Examinations will be completed in accordance with Appendix VIII of the 2007 Edition with the 2008 Addenda.

EXAMINATION CATEGORY B‐B, PRESSURE RETAINING WELDS IN VESSELS OTHER THAN REACTOR VESSELS This examination category does not apply to the Monticello Nuclear Generating Plant EXAMINATION CATEGORY B‐D, FULL PENETRATION WELDED NOZZLES IN VESSELS The category applies to the reactor pressure vessel. The full penetration welded nozzles are scheduled in accordance with relief request RR‐002 “Alternative to Nozzle‐to‐Vessel Weld and Inner Radius Examinations.” The alternative requested the use of Code Case N‐702 which reduces the number of Nozzle‐to‐Vessel Welds and Inner Radii to 25%, including at least one nozzle from each system and nominal pipe size. This code case excludes the recirculation suction, feedwater, and control rod drive return line nozzles. The Bottom Head Drain Nozzle is exempt per IWB‐1220(c). The examination volume required is per Code Case N‐613‐1 which reduces the volume from t/2 to 1/2 inch. EXAMINATION CATEGORY B‐F, PRESSURE RETAINING DISSIMILAR METAL WELDS IN VESSEL NOZZLES This category addresses Nozzle‐to‐Safe End Welds and Piping Welds. Monticello has developed a Code Case N‐716 RI‐ISI Program. All Examination Category B‐F welds have been re‐categorized as R‐A welds in accordance with Code Case N‐716. Use of Code Case N‐716 will be submitted to the NRC via Request for Alternative RR‐003. Therefore, no examinations are initially scheduled to be performed per Examination Category B‐F. There are 2 refueling outages in the first period of the 5th Interval; either the risk based scope, contingent on approval by the NRC, or the inspection scope required by IWB‐2500 will be complete prior to completing the second outage of the first period. EXAMINATION CATEGORY B‐G‐1, PRESSURE RETAINING BOLTING, GREATER THAN 2” IN DIAMETER The examination of the Reactor Vessel Bolting will be deferred to the end of the interval and performed in the 3rd Period. For volumetric examination of Recirculation Pump Studs, one of two sets of reactor recirculation pump studs is selected for volumetric examination. For the visual examination of the flange surface and nuts, bushings and washers, one of the reactor recirculation pumps will be examined only if disassembled and examined under Examination Category B‐L‐2. This meets the Examination Category B‐G‐1 examination requirements in the 2007 Edition with the 2008 Addenda of Section XI. EXAMINATION CATEGORY B‐G‐2, PRESSURE RETAINING BOLTING, 2” AND LESS IN DIAMETER This category includes the reactor vessel head cooling flange bolts, reactor vessel top head flange bolts, flange bolts on the Main Steam, Recirculation Drain, RPV Head Vent, Bottom Head Drain, and Residual Heat Removal systems. In addition this category includes valve bolting of the Core Spray, Main Steam, Feedwater, Residual Heat Removal, and Recirculation Systems. Examinations will be conducted as required in Table IWB‐2500‐1 in the 2007 Edition with the 2008 Addenda of Section XI. This bolting will



only be examined when the associated connections are disassembled. For bolting other than piping, bolting examinations will be required only when the associated component is examined under Examination Category B‐L‐2 or B‐M‐2. For bolting on piping systems, the examination will be performed on only one of the bolted connection among a group of bolted connections that are similar in design, size, function, and service. Examination will be performed only when the flange is disassembled. EXAMINATION CATEGORY B‐J, PRESSURE RETAINING WELDS IN PIPING This category addresses piping welds. Monticello has developed a Code Case N‐716 RI‐ISI program. All Examination Category B‐J Welds have been re‐categorized as R‐A welds in accordance with Code Case N‐716. Use of Code Case N‐716 will be submitted to the NRC via Request for Alternative RR‐003. Therefore, no examinations are initially scheduled to be performed per Examination Category B‐J. There are 2 refueling outages in the first period of the 5th Interval; either the risk based scope, contingent on approval by the NRC, or the inspection scope required by IWB‐2500 will be complete prior to completing the second outage of the first period. EXAMINATION CATEGORY B‐K, WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category B‐K of the ASME Code, Section XI, 2007 Edition with the 2008 Addenda requires examination of welded attachments. For the reactor pressure vessel welded attachments, footnote 4 allows for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one attachment weld shall be selected. The attachment selected shall be an attachment under continuous load during the normal system operation. The RPV has five welded attachments with the vessel skirt being selected for examination. For piping, pumps, and valves, inspection of 10% of the total population of welded attachments associated with the component supports selected for examination under IWF‐2510 shall be examined. 10% of all piping and pump welded attachments associated with the component supports selected for examination under IWF‐2510 was selected for examination. There are a total of 144 supports; 39 of those supports have been selected for examination under IWF‐2510. Of those supports selected for examination only 12 have welded attachments; 10% or 2 would require examination. There are 2 reactor recirculation pumps which have 3 supports each. Three supports on one pump are selected for examination, and all 3 supports have welded attachments; 10% or 1 welded attachment is required to be examined. The ‘A’ reactor recirculation pump is selected for examination. EXAMINATION CATEGORY B‐L‐2, PUMP CASINGS This category involves only the Reactor Recirculation Pumps and requires a visual examination on the pump casing internal surfaces when the pump is disassembled. The examination is limited to one pump. No pump is scheduled for disassembly so no pumps have been selected; however the requirement will be met during the repair/replacement and/or maintenance activity that is performed on either of the pumps.



EXAMINATION CATEGORY B‐M‐2, VALVE BODIES This category involves the valves that exceed NPS 4 in the Core Spray, Main Steam, High Pressure Coolant Injection, Feedwater, Residual Heat Removal, and Recirculation Systems. This category requires examination of one valve in each group of valves that are of the same size, constructural design, and manufacturing method, and that performs similar functions in the system. The valves have been divided into 14 Groups with the scheduling notes of BB, EE, FF, GG, HH, II, JJ, KK, LL, NN, OO, PP, QQ, and RR. However, this examination is only required when a valve is disassembled for maintenance, repair, or volumetric examination. No valve body internal surfaces have been selected. This requirement will be met during the maintenance or repair/replacement activity.

ASME Section XI Category B‐M‐2 Class 1 Valve Groups

Valve Group

Valves (drawing, valve ID)

SYS Size Constructural Design

Manufacturing Method

Function

1 (BB) ISI‐13142‐33A, V‐4 ISI‐13142‐34A, V‐4 ISI‐13142‐35A, V‐4 ISI‐13142‐36A, V‐4

MST 18” Anchor Darling

Gate Valve

Cast Isolation

Containment (Outboard) (OB)

2 (EE) ISI‐13142‐33A, V‐3 ISI‐13142‐34A, V‐3 ISI‐13142‐35A, V‐3 ISI‐13142‐36A, V‐3

MST 18” Atwood & Morrill

Wye Pattern Globe Valve

Cast Isolation

Containment (Inboard) (IB)

3 (FF) ISI‐13142‐33‐A, V‐1 ISI‐13142‐33‐A, V‐2 ISI‐13142‐34‐A, V‐1 ISI‐13142‐34‐A, V‐2 ISI‐13142‐35‐A, V‐1 ISI‐13142‐35‐A, V‐2 ISI‐13142‐36‐A, V‐1 ISI‐13142‐36‐A, V‐2

MST 6” Target Rock Corp

SRV Model 67F Safety Relief

Valve

Cast

Overpressure Protection

Safety Relief Valve

4 (GG) ISI‐13142‐52‐A, V‐2 ISI‐13142‐52‐A, V‐3 ISI‐13142‐53‐A, V‐2 ISI‐13142‐53‐A, V‐3

CFW 14” Anchor Darling

Swing Check Valve

Cast Isolation

Containment (IB / OB)

5 (HH) ISI‐13142‐52‐A, V‐1 ISI‐13142‐53‐A, V‐1

CFW 14” Anchor Darling

Gate Valve

Cast

Isolation

Loop 6 (II) ISI‐97005‐A, V‐1

ISI‐97005‐A, V‐2 ISI‐97006‐A, V‐1 ISI‐97006‐A, V‐2

REC 28” Crane

Gate Valve

Cast Isolation

Loop

7 (JJ) ISI‐13142‐42‐A, V‐1

HPC 8” Velan

Gate Valve

Forged Isolation

Containment (IB)



ASME Section XI Category B‐M‐2 Class 1 Valve Groups

Valve Group

Valves (drawing, valve ID)

SYS Size Constructural Design

Manufacturing Method

Function

8 (KK) ISI‐13142‐42‐A, V‐2 HPC 8” Flowserve Gate Valve

Cast Isolation

Containment (OB) 9 (LL) ISI‐13142‐31‐A, V‐1

ISI‐13142‐31‐A, V‐3 ISI‐13142‐26‐A, V‐1 ISI‐13142‐26‐A, V‐3

CSP 8” Anchor Darling

Gate Valve

Cast Isolation

Loop Containment (OB)

10 (NN) ISI‐13142‐31‐A, V‐2 ISI‐13142‐26‐A, V‐2

CSP 8” Atwood & Morrill

Testable Check

Valve

Cast Isolation

Containment (IB)

11 (OO) ISI‐97003‐A, V‐1 ISI‐97003‐A, V‐3

RHR 18” Anchor Darling

Gate Valve

Cast Isolation


12 (PP) ISI‐97004‐A, V‐1

ISI‐97004‐A, V‐3 RHR 16” Anchor Darling

Gate Valve

Cast Isolation


13 (QQ) ISI‐97003‐A, V‐2

ISI‐97004‐A, V‐2

RHR 16” Atwood & Morrill

Testable Check

Valve

Cast Isolation

Containment (IB)

14 (RR) ISI‐97003‐B, V‐1 ISI‐97003‐B, V‐2 ISI‐97003‐B, V‐3

RHR 18” Anchor Darling

Gate Valve

Cast Isolation

Loop Containment (IB / OB)



EXAMINATION CATEGORIES B‐N‐1, B‐N‐2 AND B‐N‐3 To better define what these examination categories consist of it is important to review the basis for the development of the category. A paper titled “Development of In‐Service Inspection Safety Philosophy for U.S.A Nuclear Power Plant” by S.H. Bush and R.R. MacCary was reviewed which defined the philosophy behind the development of the Inspection Category N in the 1971 Edition of the ASME Section XI Code. This paper stated:

“The special examination category N covers the examination of the interior surfaces and internal components of the reactor vessel; it is considered one of the most critical examination requirements in the A.S.M.E Section XI Code. Among the considerations contributing to the development of this examination category were the reported experiences and difficulties encountered in the operating facilities. These interior examination areas should assure: a. Inspection of all internal support attachments welded to the reactor vessel whose

failure could result in reactor core disarrangement. b. Discovery of any loose parts which might have accumulated at the bottom of the

reactor vessel during service. c. Detection of undue wear as a result of flow‐induced vibrations of components of the

reactor core structure. d. Verification of the overall structural integrity of the core structure, including

supplementary internal components such as moisture separators, material surveillance capsules, instrumentation, and reactor control rod assembly guides.”

The 2007 Edition with the 2008 Addenda does not have a Category N but now has three categories applicable to a Boiling Water Reactor (BWR) such as Monticello, Category B‐N‐1, B‐N‐2 and B‐N‐3. The above basis for Category N in the 1971 Edition of Section XI is interpreted as follows: Item (a) above is interpreted to be addressed by the VT‐1 and VT‐3 visual examinations for Examination Category B‐N‐2, Item Numbers B13.20 and B13.30 “Interior Attachments within and beyond the beltline region”. Item (b) above is interpreted to be addressed by the VT‐3 visual examination for Examination Category B‐N‐1, Item Number B13.10, “Vessel Interior” covering spaces above and below the reactor core. Item (c) above is interpreted to be addressed by the VT‐3 visual examination for Examination Category B‐N‐2, Item Number B13.40 “Core Support Structure”, except that the supplementary components were not included when the Examination Categories B‐N‐1, B‐N‐2, B‐N‐3 or B‐I‐1 replaced the former Examination Category N in the 1974 Edition. The reactor vessel interior surfaces referred to in the former Examination Category N were addressed by Examination Category B‐I‐1 in the 1974 Edition as sample “clad patches” which was eliminated in the Summer 1976 Addenda.



EXAMINATION CATEGORY B‐N‐1, INTERIOR OF REACTOR VESSEL This category involves the examination of spaces above and below the reactor core that are made accessible by removal of components during normal refueling outages. These examinations will be conducted each inspection period. Based on the explanation above this would only be performing a VT‐3 visual examination looking for loose parts or FME. This meets the Examination Category B‐N‐1 examination requirements in the 2007 Edition with the 2008 Addenda of Section XI. EXAMINATION CATEGORY B‐N‐2, WELDED CORE SUPPORT STRUCTURES AND INTERIOR ATTACHMENTS TO REACTOR VESSELS These examinations will be deferred until the third period as allowed by Table IWB‐2500‐1. These examinations include the interior welded attachments for vessel internal components as well as those welded core support structures. The welded attachments within the beltline region will receive a Visual, VT‐1 examination and the welded attachments outside the beltline region will receive a Visual, VT‐3 examination. This meets the Examination Category B‐N‐2 examination requirements in the 2007 Edition with the 2008 Addenda of Section XI. EXAMINATION CATEGORY B‐N‐3, REMOVALBLE CORE SUPPORT STRUCTURES This examination category is not applicable to the Monticello Nuclear Generating Plant. EXAMINATION CATEGORY B‐O, PRESSURE RETAINING WELDS IN CONTROL ROD DRIVE AND INSTRUMENT NOZZLE HOUSINGS This examination category requires a volumetric or surface examination of 10% of peripheral control drive housings during the inspection interval. There are 24 peripheral control rod drives on the reactor vessel bottom head. To meet the category examination requirements 2 welds (upper and lower weld) on 3 of the peripheral control rod drive housings will be selected for examination. This meets the Examination Category B‐O examination requirements in the 2007 Edition with the 2008 Addenda. EXAMINATION CATEGORY B‐P, ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Monticello meets the requirements of ASME Code, Section XI, 2007 Edition with the 2008 Addenda for Class 1 Systems. Details of the component listing are contained in the surveillance procedure for the Class 1 System Leakage Test. EXAMINATION CATEGORY B‐Q, STEAM GENERATOR TUBING This examination category is not applicable to the Monticello Nuclear Generating Plant.



EXAMINATION CATEGORY C‐A, PRESSURE RETAINING WELDS IN PRESSURE VESSELS This category applies to the Residual Heat Removal Heat Exchangers (A and B). Note 3 in Table IWC‐2500‐1, Examination Category C‐A states “In the case of multiple vessels of similar design, size, and service (such as steam generators, heat exchangers), the required examinations may be limited to one vessel or distributed among vessels.” Both of the head welds and the shell welds were selected on one of the Residual Heat Removal Heat Exchangers. Both RHR Heat Exchangers have a Shell Cover Floating Head, therefore Item Number C1.30 is not applicable to MGNP. This meets the Examination Category C‐A examination requirements in the 2007 Edition with the 2008 Addenda. EXAMINATION CATEGORY C‐B, PRESSURE RETAIING NOZZLE WELDS IN VESSELS This category applies to the Residual Heat Removal Heat Exchangers. Note 1 in Table IWC‐2500‐1, Category C‐B, excludes manways and handholes. Note 3 require that nozzles selected initially for examination shall be reexamined over the service life of the component to the extent practical. Note 4 allow that in the case of multiple vessels of similar design, size, and service the required examinations may be limited to one vessel or distributed among the vessels. Two nozzles were selected on one of the Residual Heat Removal Heat Exchangers. This meets the Examination Category C‐B examination requirements in the 2007 Edition with the 2008 Addenda. EXAMINATION CATEGORY C‐C, WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category C‐C of the ASME Code Section XI, 2007 Edition with the 2008 Addenda requires examination of Welded Attachments. For vessel attachments Note 4 allows for multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. The welded attachment selected shall be an attachment under continuous load during normal system operation or an attachment subject to a potential intermittent load during normal system operation if an attachment under continuous load does not exist. A welded attachment on one of the Residual Heat Removal Heat Exchangers was selected for examination. For piping pumps and valves, inspection of 10% of the total population of integral welded attachments associated with the component supports selected for examination under IWF‐2510 is required. 10% of all piping and pump welded attachments associated with the component supports selected for examination under IWF‐2510 was selected for examination. There are two piping welded attachments that do not require examination (Ref. IWC‐2500‐1, Examination Category C‐C, Note (b)) as they do not provide support. There are a total of 268 piping supports of which 42 piping supports (15%) have been selected for examination under IWF‐2510. Of those 42 piping supports selected for examination under IWF‐2510, 14 have welded attachments, therefore 10% or 2 welded attachments on piping supports require examination. There are a total of six pumps that have welded attachments (4 RHR and 2 CS



pumps); only two of the associated pump supports will be selected for examination under IWF‐2510. Both of these pump supports have welded attachment, therefore 10% of 2 is 1 pump welded attachment requires examination. This meets the Examination Category C‐C examination requirements in the 2007 Edition with the 2008 Addenda of Section XI. EXAMINATION CATEGORY C‐D, PRESSURE RETAINING BOLTING GREATER THAN 2” IN DIAMETER This examination category does not apply to the Monticello Nuclear Generating Plant. EXAMINATION CATEGORIES C‐F‐1, PRESSURE RETAINING WELDS IN AUSTENITIC STAINLESS STEEL OR HIGH ALLOY PIPING This category addresses Class 2 stainless steel piping welds. Monticello has developed a Code Case N‐716 RI‐ISI program. All Examination Category C‐F‐1 welds have been re‐categorized as R‐A welds in accordance with Code Case N‐716. Use of Code Case N‐716 will be submitted to the NRC via Request for Alternative RR‐003. Therefore, no examinations are initially scheduled to be performed per Examination Category C‐F‐1. There are 2 refueling outages in the first period of the 5th Interval; either the risk based scope, contingent on approval by the NRC, or the inspection scope required by IWB‐2500 will be complete prior to completing the second outage of the first period. EXAMINATION CATEGORY C‐F‐2, PRESSURE RETAINING WELDS IN CARBON OR LOW ALLOY STEEL PIPING This category addresses Class 2 carbon steel piping welds. Monticello has developed a Code Case N‐716 RI‐ISI program. All Examination Category C‐F‐2 welds have been re‐categorized as R‐A welds in accordance with Code Case N‐716. Use of Code Case N‐716 will be submitted to the NRC via Request for Alternative RR‐003. Therefore, no examinations are initially scheduled to be performed per Examination Category C‐F‐1. There are 2 refueling outages in the first period of the 5th Interval; either the risk based scope, contingent on approval by the NRC, or the inspection scope required by IWB‐2500 will be complete prior to completing the second outage of the first period. EXAMINATION CATEGORY C‐H, ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Monticello meets the requirements of ASME Code, Section XI, 2007 Edition with the 2008 Addenda for Class 2 Systems. Details of the component listing are contained in the individual surveillance procedures for the Class 2 System Leakage Tests. EXAMINATION CATEGORY D‐A, WELDED ATTACHMENTS FOR VESSELS, PIPING, PUMPS, AND VALVES Examination Category D‐A of the ASME Code Section XI, 2007 Edition with the 2008 Addenda requires examination of welded attachments on those systems that are determined to be most subject to corrosion, such as the welded attachments of the Service Water or Emergency Service Water systems. For welded attachments of piping, pumps, and valves, a 10% sample shall be selected for examination. This percentage sample shall be proportional to the total number of nonexempt welded attachments connected to the piping, pumps, and valves in each system subject to these examinations. The only



Class 3 system at Monticello that is not exempt from examination and considered subject to corrosion is the Residual Heat Removal Service Water System. There are a total of 99 supports, of which 14 have attachments. 10% of those 14 or 2 welded attachment are required to be examined. EXAMINATION CATEGORY D‐B, ALL PRESSURE RETAINING COMPONENTS The pressure testing program at Monticello meets the requirements of ASME Code, Section XI, 2007 Edition with the 2008 Addenda for Class 3 Systems. Details of the component listing are contained in the individual surveillance procedures for the Class 3 System Leakage Tests. EXAMINATION CATEGORY F‐A, SUPPORTS Examination Category F‐A of the ASME Code Section XI, 2007 Edition through the 2008 Addenda requires 25% of Class 1 Piping Supports, 15% of Class 2 Piping Supports, and 10% of Class 3 Piping Supports to be examined during the inspection interval. For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. The supports have been separated by type as defined in Note (1) to Examination Category F‐A. A letter designation has been added to the Item Number to clearly identify each support by type. Twenty‐five percent (25%) of the Class 1 supports have been selected and are prorated by type and system. Fifteen percent (15%) of the Class 2 supports have been selected and are prorated by type and system. Ten percent (10%) of the Class 3 supports have been selected and are prorated by type and system.

For supports, other than piping supports, the components have been scheduled as follows: Reactor Vessel Support Skirt High Pressure Coolant Injection Pump One Core Spray Pump One Recirculation Pump One Residual Heat Removal Heat Exchanger One Residual Heat Removal Pump EXAMINATION CATEGORY R‐A The alternative Code Case N‐716, RI‐ISI Program for piping, as described in the forthcoming Request for Alternative RR‐003, will be utilized. The RI‐ISI Program will be substituted for the Examination Categories B‐F, B‐J, C‐F‐1, and C‐F‐2 in accordance with 10 CFR 50.55a(3)(i) by alternatively providing an acceptable level of quality and safety. The welds are selected in accordance with the final calculation provided by Structural Integrity Associates, Inc. (SI) Calculation 1000515.302, “N‐716 Evaluation for Monticello”. The calculation requires 67 welds to be examined volumetrically.

The N‐716 program contains 96 welds that are susceptible to FAC. Code Case N‐716 refers to the utility FAC Program for welds susceptible to FAC. Therefore all of the welds that have a FAC only mechanism are assigned R1.20. If it has FAC along with another degradation mechanism then the item number for



the non FAC degradation mechanism is assigned. It should be noted that 3 RWCU welds were selected as the RWCU system was determined to be High Safety Significant (HSS) which requires 10% of the total population to be selected. The 3 welds chosen are also susceptible to FAC as a degradation mechanism; however these welds were selected with no degradation mechanism and assigned Code Item No. R1.20.

4.0 Examination Personnel/Procedures

Inservice Inspection examination procedures and personnel certifications will meet the requirements specified in the 2007 Edition with the 2008 Addenda on January 21, 2013. The implementation of Appendix VIII to the 2001 Edition will continue until January 21, 2013 which is allowed by 10 CFR 50.55a(g)(4)(ii).

5.0 Reporting of Associated Section XI Programs

The Section XI Repair/Replacement Program, Containment Inservice Inspection Program, and System Pressure Testing Program are administered under separate program documents. Although these programs are administered separately, the activities by the Repair/Replacement Program, Containment Inservice Inspection Program, and System Pressure Testing Program, are reported in the “Inservice Inspection Summary Report” or “Owners Activity Report” following each refueling outage.

6.0 Augmented and Owner Programs

6.1 Augmented Programs

The ISI Plan also contains certain non‐code items to be examined or examination beyond that described in ASME Section XI. These augmented items include NRC required or license‐initiated examinations on the following components: 6.1.1 NUREG‐0619 “BWR Control Rod Drive Return Line Nozzle Cracking Post Modification”

Source Document: RAC00988

Associated Document: Commitment M83165A, Letter dated 04/27/1995 (M1995092, JRR03336) to NRR informing them of missed inspections in 1993/1994.

Purpose: The CRD Nozzle was cut and capped in response to NUREG‐0619 issues. The CRD Return Line was re‐routed to the Reactor Water Cleanup System. This inspection is to assure integrity of the re‐routed piping at the tee on Isometric NX‐13142‐49A, Line REW6‐3”‐DC.



Scope: The welds inspected are W‐11, W‐12 and W‐13 on ISI ISO NC‐ISI‐51.

Method: Volumetric examination methods are required and shall include base metal to a distance of one‐pipe‐wall thickness or ½” whichever is greater on both sides of the weld.

Industry Code or Standards: ASME Section XI to the extent practical.

Frequency: Every refueling outage.

Acceptance Criteria or Standard: ASME Section XI, IWC‐3500.

Regulatory Basis: NRC Commitment M83165A.

6.1.2 Modification 79Z018 “RWCU Return Line Modification”

Source Document: NUREG‐0619 “BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking: Resolution of Generic Technical Activity A‐10 (Technical Report)”

Associated Document: Structural Integrity Associates, Inc Calculation Package “Risk‐Informed ISI Code Case N‐578 Application to NMC Plants” File No. NMC‐01‐301.

Purpose: The RWCU return line was rerouted to be distributed equally between the two feedwater lines. The modification assures that feedwater to all four nozzles will be warmed at low flow rates by the higher temperature RWCU return water. The modification reduced the feedwater nozzle usage factors by 15% ‐ 30%. During startup and shutdown it is possible for the alternating stress to exceed the endurance strength of carbon steel. Because of the small amount of time at these conditions it is believed that thermal fatigue will not be a problem at the RWCU connection to the feedwater lines however it was recommended that the feedwater lines downstream of the HPCI and RCIC connections and the HPCI and RCIC lines downstream of the RWCU line be inspected by UT during each refueling outage. Since the implementation of the Modification, the Risked Informed ISI Program for the 4th 10 year interval evaluated the HPCI to Feedwater connection and found that it was not susceptible to thermal fatigue. In addition, the Feedwater lines themselves are evaluated under the RI‐ISI Program and are selected in accordance with that program and not as an Augmented Program. Therefore, the only welds now examined under Modification 79Z018 are the RCIC lines shown below.

Scope: The welds inspected are W‐1, W‐2, W‐3, W‐4, W‐12, and W‐12A on ISI ISO NC‐ISI‐37.

Method: Volumetric examination.

Industry Code or Standards: ASME Section XI to the extent practical.

Frequency: Every refueling outage.




Regulatory Basis: None.

6.1.3 High Energy Line Break

Source Document: MEB 3‐1 as amended by Generic Letter (GL) 87‐11 “Relaxation in Arbitrary Intermediate Pipe Rupture Requirements

Associated Document: NUREG 3.6.2 “Determination of Rupture Locations and Dynamic Effects Associated with the Postulated Rupture of Piping”, NUREG‐0800, 5.2.4 “Reactor Coolant Pressure Boundary Inservice Inspection and Testing”, and UFSAR Appendix I “Evaluation of High Energy Line Breaks Outside Containment”

Purpose: In September 1987, the NRC issued GL 87‐11 which would allow utilities to eliminate the consideration of the environmental and dynamic effects of arbitrary intermediate pipe breaks provided that the requirements of Branch Technical Position 3‐4 (formerly BTP MEB 3‐1), Revision 2 were met.

Scope: The welds inspected are:

W‐1, W‐2, W‐3, on ISI ISO ISI‐13142‐31‐A, W‐6, W‐7, W‐8 on ISI ISO ISI‐13142‐26‐A, W‐16, W‐17 on ISI ISO ISI‐13142‐42‐A, W‐29, W‐29A, W‐30, W‐31, W‐32 on ISI ISO ISI‐13142‐33‐A, W‐12, W‐13, W‐14 on ISI ISO ISI‐768A, W‐29, W‐30, W‐31, W‐32, W‐33 on ISI ISO ISI‐13142‐34‐A, W‐29, W‐30, W‐31, W‐32, W‐33, W‐34 on ISI ISO ISI‐13142‐35‐A, W‐31, W‐32, W‐33, W‐34, W‐35 on ISI ISO ISI‐13142‐36‐A, W‐13, W‐14 on ISI ISO‐13142‐43‐A, W‐24, W‐25 on ISI ISO ISI‐97003B, W‐1, W‐2 on ISI ISO ISI‐97004‐A, W‐1, W‐2 on ISI ISO ISI‐97003‐A W‐25, W‐26 on ISI ISO ISI‐73880‐A. Method: Volumetric examination.

Industry Code or Standards: ASME Section XI IWA‐2400

Frequency: Once per interval.

Acceptance Criteria or Standard: ASME Section XI, IWB‐3500



Regulatory Basis: GL 87‐11.

6.2 Owner Elected Examinations

In addition to the above identified Augmented Examinations, there are examinations performed due to internal commitments. These items are examined to the extent practical in accordance with the Section XI Code, 2007 Edition with the 2008 Addenda, not as part of the RI‐ISI Program. Relief Requests will not be submitted for these non‐code exams if Section XI Code requirements cannot be met. Non‐code exams are also subject to change without prior notification to the NRC. These are identified as Owner Elected Examination. The following Owner Elected Examinations are performed:

6.2.1 OE23699 “Standby Liquid Control Tank”

Source Document: OE23699 “Initiation of Reactor Shutdown required by Technical Specification 3.1.7 Action C.1”

Associated Document: AR 01064168

Purpose: The external operating experience from Quad Cities Station indicated that the Standby Liquid Control Tank had a pinhole leak that was not determined to make the tank inoperable per ASME Code requirements. The Quad Cities Standby Liquid Control Tank is constructed with Type 304 stainless steel. For Type 304 stainless steel tank material, the most likely cause of external stress corrosion cracking is an exposure to chlorides or other halogen containing solution, which can lead to transgranular stress corrosion cracking (TGSCC). The Monticello SBLC Tank is also made of stainless steel and is susceptible to the same mechanism. This inspection is to verify the structural integrity of the tank.

Scope: The SBLC Tank inside surfaces.

Method: Visual VT‐3 examination of the tank interior surfaces.

Industry Code or Standards: None.

Frequency: Every interval.


Regulatory Basis: None.



7.0 License Renewal Aging Management Plans and Commitments

This document supports the implementation of the following License Renewal Aging Management Programs and Commitments:

7.1 Programs

• PBD/AMP‐004, Thermal Aging and Neutron Irradiation Embrittlement of Cast Austenitic Stainless Steel (CASS) Program

• PBD/AMP‐022, Primary Containment In‐Service Inspection Program

• PBD/AMP‐024, ASME Section XI, Subsection IWF

• PBD/AMP‐033, ASME Section XI Inservice Inspection, Subsection IWB, IWC, and IWD

• PBD/AMP‐034, Reactor Head Closure Studs

• PBD/AMP‐035, BWR Vessel ID Attachment Welds Program

• PBD/AMP‐036, BWR Feedwater Nozzle

• PBD/AMP‐037, BWR Control Rod Drive Return Nozzle

• PBD/AMP‐038, BWR Stress Corrosion Cracking Program

• PBD/AMP‐039, BWR Penetrations Program

• NRC Commitments M05008A, M05009A, M05010A, M05011A, M05020A, M05021A, and M05022A

7.2 License Renewal Commitments

7.2.1 Class MC Supports

Source Document: USAR Appendix K, K2.1.3 Associated Document: PBD/AMP‐024, Commitment M05011A Purpose: As required by License Renewal Commitment M05011A, Class MC Supports will be examined per the requirements of Subsection IWF. Scope: 100% of the following Class MC Supports are included in an augmented program and will be examined each inspection interval:

• Torus/Ring Header Seismic Restraints

• Drywell Male and Female Stabilizers

• Shield Stabilizers

• Torus Columns

• Torus Saddles

• Vent System Supports



• Downcomer Bracing Method: Visual VT‐3 Examination Industry Code or Standards: ASME Section XI, Subsection IWF Frequency: 100% each interval Acceptance Criteria or Standard: ASME Section XI, IWF‐3410 Regulatory Basis: NRC Commitment M05011A and USAR Appendix K

7.2.2 Small Bore Class 1 Piping

Source Document: USAR, Appendix K, K2.1.2 Associated Document: PBD/AMP‐033, NRC Letter dated May 10, 2011 “Monticello Nuclear Generating Plant (MNGP), Examination of Class 1 Small‐Bore Piping Butt Welds” Purpose: As required by License Renewal to manage aging effects, examination of Small Bore Piping (Ref. USAR Appendix K, Section K2.1.2) has been added as an augmented program to the ISI Plan. Scope: The weld population includes W‐2 through W‐7 on ISI Drawing ISI‐786A and W‐32 through W‐34 on ISI Drawing ISI‐74215A. The exams are performed in support of License Renewal and SHALL be performed through the Renewed License period of extended operation. The base scope of approximately 10% of the population will be examined during each ISI interval.

• Method: Augmented volumetric examinations of welds are performed on Class 1 stainless steel small bore piping butt welds > NPS 2 to < NPS 4. The weld volume applicable to Category B‐F or B‐J will be used for the examination. Welds will be examined to the extent practical. If limitations are encountered that do not permit coverage of essentially 100%, a 10 CFR 50.55a request (relief request) is not required. Industry Code or Standard: Examination personnel SHALL be qualified to ASME Section XI, Appendix VIII as specified in the ISI Plan. Frequency: 10% each interval



Acceptance Standard: Welds will be evaluated in accordance with IWB‐3000 requirements applicable to Category B‐F or B‐J Regulatory Basis: License Renewal Commitment per USAR Appendix K

7.2.3 Feedwater Nozzles

Source Document: USAR, Appendix K, K2.1.8 Associated Document: PBD/AMP‐036 Purpose: As required by License Renewal Commitments, M05021A, to manage aging effects, Reactor Vessel Feedwater Nozzle examinations will be performed in accordance with the requirements of General Electric Report NE‐523‐A71‐0594A, Rev. 1 (Ref. USAR Appendix K, Section K2.1.8). Scope: Feedwater Nozzles Method: Volumetric examinations of specified zones in accordance with ASME Section XI Appendix VIII. Industry Code or Standard: ASME Section XI Frequency: Each interval Acceptance Criteria or Standard: ASME Section XI, IWB‐3512 Regulatory Basis: USAR Appendix K, K2.1.8

8.0 Source Documents

The following referenced source documents described and listed below are basis documents used and applicable to the Monticello 5th Interval ISI Plan.

• ASME BPV Code Section XI, 2007 Edition with the 2008 Addenda

• 10 CFR 50.55a, Industry Codes and Standards (76 FR 36232)

• Regulatory Guide 1.147, Revision 16, October 2010

• Monticello Inservice Inspection Licensee Control Program, 4 AWI‐09.04.00

• Monticello ASME Section XI Inservice Inspection Program, EWI‐09.04‐00



• GE Nuclear Services Information Letter, SIL. No. 483R2 “CRD Cap Screw Crack Indications,” September 5, 1992

• Generic Letter 88‐01 & NUREG 0313, Rev 2 (IGSCC (M88080A, M88082A) Note: All Monticello welds meet NUREG 0313, Rev 2 Category A

• Monticello Notification Letter to NRC, “Notification of Extension of 3rd Ten‐Year Inservice Testing and Inservice Inspection Intervals,” May 30, 2002

• NRC Commitment M97025A, “Response to LER 97‐004, Failure to Submit Relief Requests for Limited Inservice Inspection Examinations, dated March 24, 1997.” Limited Examination Relief Requests submitted within 12 months

• NUREG‐1865 “Safety Evaluation Report, Related to the License Renewal of the Monticello Nuclear Generating Plant, Docket No. 50‐263”

• NRC Commitment M05008A (Passport AR 00829849) – MNGP site‐specific administrative work instructions will be applicable to both safety and non‐safety related systems, structures and components that are subject to an aging management review consistent with the current licensing basis during the period of extended operation.

• NRC Commitment M05009A (Passport AR 00829851) – Site documents that implement aging management activities for license renewal will be enhanced to ensure that an AR is prepared in accordance with plant procedures whenever non‐conforming conditions are found (i.e., the acceptance criteria is not met)

• NRC Commitment M05010A (Passport AR 00829853) – Revisions will be made to procedures and instructions that implement or administer aging management programs and/or activities for the purpose of managing the associated aging effects for the duration of extended operation

• NRC Commitment M05011A (Passport AR 00829856‐01) – The MNGP ASME Section XI, Subsection IWF Program will be enhanced to provide inspections of Class MC components consistent with NUREG‐1801, Chapter III, Section B1.3

• NRC Commitment M05021A (Passport AR 00829893‐01) – The BWR Feedwater Nozzle Program will be enhanced so the regions being inspected; examination techniques, personnel qualifications, and inspection schedule are consistent with the recommendations of GE NE‐523‐A71‐0594 Revision 1

• USAR Appendix K, Renewed Operating License – USAR Supplement, Items (K2.1.33, K2.1.3, K2.1.26, K2.1.1, K2.1.28, K2.1.11, K2.1.8, K2.1.7, K2.1.10, K2.1.9, and K5)

• NRC Letter dated 05/10/11 “Examination of Class 1 Small‐Bore Piping Butt Welds”

• NUREG‐1865, Safety Evaluation Report Related to the License Renewal of the Monticello Nuclear Generating Plant; dated October 2006 (SER sections 3.0.3.2.2, 3.0.3.2.3, 3.0.3.2.6, 3.0.3.2.7, 3.0.3.2.8, 3.0.3.2.9, 3.0.3.2.10, 3.0.3.1.6, 3.0.3.1.7, 3.0.3.1.8)

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan APPENDIX A

CODE CASES


Adoption of Code Cases ASME Section XI Code Cases adopted for ISI and related NDE activities for the Fifth Interval are listed in Tables A.1‐1, A.1‐2, and A.1‐3. Code Cases related to Repair/Replacement, Pressure Testing, and/or Containment Program may be listed in Tables A.1‐1, A.1‐2, or A.1‐3. The use of Code Cases is in accordance with ASME Section XI, (IWA‐2440 or IWA‐4190), 10 CFR 50.55a, and Regulatory Guide 1.147. As permitted by ASME Section XI and Regulatory Guide 1.147 or 10 CFR 50.55a, ASME Section XI Code Cases may be adopted and used as described below: Adoption of Code Cases Listed for Generic Use in Regulatory Guide 1.147 Code Cases that are listed for generic use in the latest revision of Regulatory Guide 1.147 may be included in the ISI Program provided any additional conditions specified in the Regulatory Guide are also incorporated. Table A.1‐1 identifies the Code Cases approved for generic use and adopted for the Fifth Interval. Adoption of Code Cases Not Approved in Regulatory Guide 1.147 Certain Code Cases that have been approved by the ASME Board on Nuclear Codes and Standards may not have been reviewed and approved by the NRC Staff for generic use and listed in Regulatory Guide 1.147. Use of such Code Cases may be requested in the form of a “Request for Alternative” in accordance with 10 CFR 50.55a(a)(3). Once approved, these Requests for Alternatives will be available for use until such time that the Code Cases are adopted into Regulatory Guide 1.147, at which time compliance with the conditions contained in the Regulatory Guide is required. Table A.1‐2 identifies those Code Cases that have been requested through Requests for Alternatives. For convenience to the user of this ISI Program, the appropriate internal correspondence number is provided to assist in their retrieval from Document Control. All other Requests for Alternatives and Relief Requests (those not associated with NRC approval of Code Cases) are addressed in Appendix B. Adoption of Code Cases Mandated by 10 CFR 50.55a Code Cases required by regulation in 10 CFR 50.55a are incorporated into the ISI Program and implemented at the specified schedule. Code Cases currently required by 10 CFR 50.55a are identified in Table A.1‐3. Use of Annulled Code Cases As permitted by Regulatory Guide 1.147, Code Cases that have been adopted for use in the current interval that are subsequently annulled by ASME, may be used for the remainder of the interval. Code Case Revisions Initial adoption of a Code Case requires use of the latest revision of that Code Case listed in Regulatory Guide 1.147. However, if an adopted Code Case is later revised and approved by the NRC, then either


CODE CASES


the earlier or later revision may be used. An exception to this provision would be the inclusion of a condition on the later revision necessary to enhance safety. In this situation, the condition imposed on the later revision must be incorporated into the program. Adoption of Code Cases Issued Subsequent to Filing the Inservice Inspection Plan Code Cases issued by ASME subsequent to filing the Inservice Inspection Plan with the NRC may be incorporated within the provisions of Regulatory Guide 1.147 by initiating a revision to this Appendix. Any subsequent Code Cases shall be incorporated into the program and identified in either Table A.1‐1 or A.1‐2, as applicable, prior to their use. Code Cases not approved for use by the NRC Certain Code Cases that have been approved by the ASME Board on Nuclear Codes and Standards have been reviewed and are not approved by the NRC Staff for generic use. These Code Cases are listed in Regulatory Guide 1.193, “ASME Code Cases Not Approved for Use.” However, the NRC may approve their use in specific cases. Code Cases listed in Regulatory Guide 1.193 will not be used at Monticello without an approved Request for Alternative in accordance with 10 CFR 50.55a(a)(3). Those will be identified in Appendix B. Regulatory Guide 1.147, Revision 16 Approved Code Cases

Table A.1‐1 ‐ Code Cases Adopted from Regulatory Guide 1.147 Code Case Number Title NRC Conditions

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

None

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

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

N‐526 Alternative Requirements for Successive Inspections of Class 1 and 2 Vessels

None

N‐586‐1 Alternative Additional Examination Requirements for Classes 1, 2, and 3 Piping, Components, and Supports

None


CODE CASES



N‐597‐2 Requirements for Analytical Evaluation of Pipe Wall Thinning

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

(2) Components affected by flow‐accelerated corrosion to which this Code Case are applied must be repaired or replaced in accordance with the construction code or record and Owner’s requirements or a later NRC approved edition of Section III, “Rules for Construction of Nuclear Power Plant Components,” of the ASME Code (Ref. 7) prior to the value of tp reaching the allowable minimum wall thickness, tmin, as specified in 3622.1(a)(1) of this Code Case. Alternatively, use of the Code Case is subject to NRC review and approval per 10 CFR 50.55a(a)(3).

(3) For Class 1 piping not meeting the criteria of ‐3221, the use of evaluation methods and criteria is subject to NRC review and approval per 10 CFR 50.55a(a)(3).

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

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

N‐600 Transfer of Welder, Welding Operator, Brazer, and Brazing Operator Qualifications Between Owners

None


CODE CASES



N‐606‐1

Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stub Tube Repairs.

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

N‐613‐1

Ultrasonic Examination of Penetration Nozzles in Vessels, Examination Category B‐D, Item Nos. B3.100 and B3.90, Reactor Nozzle‐to‐Vessel Welds, Figs. IWB‐2500‐7(a), (b), and (c).

None

N‐639 Alternative Calibration Block Material

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

N‐651

Ferritic and Dissimilar Metal Welding Using SMAW Temper Bead Technique Without Removing the Weld Bead Crown for the First Layer

None

N‐705 Evaluation Criteria for Temporary Acceptance of Degradation in Moderate Energy Class 2 or 3 Vessels and Tanks

None

N‐735 Successive Inspections of Class 1 and 2 Piping Welds None


CODE CASES


Code Cases Approved Through Request for Alternatives

The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 16 and require a request for alternative prior to implementation. See Appendix B of this plan for the applicable requests.

Table A.1‐2 – Code Cases Adopted Via NRC Approved Requests

Code Case Number Title

Request for Alternative No.

N‐532‐5 Repair/Replacement Activity Documentation Requirements and Inservice Inspection Summary Report Preparation and Submission

RR‐006

N‐661‐2 Alternative Requirements for Wall Thickness Restoration of Classes 2 and 3 Carbon Steel Piping for Raw Water Service

RR‐004

N‐702 Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle‐to‐Shell Welds

RR‐002

N‐716 Alternative Piping Classification and Examination Requirements RR‐003

N‐795 Alternative Requirements for BWR Class 1 System Leakage Test Pressure Following Repair/Replacement Activities

RR‐005


CODE CASES


Code Cases Required by 10 CFR 50.55a

The following ASME Code Cases are not contained in Regulatory Guide 1.147, Revision 16, but are mandated in 10 CFR 50.55a.

Table A.1‐3 – Code Cases Required by 10 CFR 50.55a

Code Case Number

Title Notes

NONE

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan APPENDIX B

RELIEF REQUESTS


RELIEF REQUESTS Throughout this program, the term “relief request” is used interchangeably referring to submittals to the NRC requesting permission to deviate from either an ASME Section XI requirement, a 10 CFR 50.55a regulation, to use provisions from later Editions or Addenda of Section XI approved by the NRC as referenced in 10 CFR 50.55a(b) or to use provisions from Editions or Addenda of Section XI not approved by the NRC as referenced in 10 CFR 50.55a(b). However, in communications with the NRC and within written requests to deviate, the terms as defined below must be used for clarity and to satisfy 10 CFR 50.55a. Submittals to the NRC must clearly identify which of the below rules are being used to request the deviation. Table B.1‐1 contains an index of Request for Alternatives and Relief Requests written in accordance with 10 CFR 50.55a(a)(3) and (g)(5). The applicable Xcel Energy Inc. submittal and NRC Safety Evaluation Report (SER) correspondence numbers are also included for each request. Request for Alternatives

When seeking an alternative to the rules contained in 10 CFR 50.55a(g), the request is submitted under the provision of 10 CFR 50.55a(a)(3). Once approved by the Director, Office of Nuclear Reactor Regulation, the alternative may be incorporated into the ISI program. These types of requests are typically used to request use of Code Cases, Code Edition, or Addenda not yet approved by the NRC. Request for Alternatives must be approved by the NRC prior to their implementation or use. Within the provisions of 10 CFR 50.55a(a)(3) there are two specific methods of submittal:

10 CFR 50.55a(a)(3)(i) allows alternatives when authorized by the NRC, if the proposed alternatives would provide an acceptable level of quality and safety. Requests submitted under these provisions are not required to demonstrate hardship or burden.

10 CFR 50.55a(a)(3)(ii) also allows alternatives when authorized by the NRC, if compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. When submitted under this provision, there must be evidence of unusual hardship or difficulty. Typically this hardship will be dose or excessive disassembly.

It should be noted that during the public comment period of the rule there was a comment provided that questions whether the requirement to have request for alternatives applied to both of the above paragraphs. It was pointed out that under (a)(3)(ii), a licensee may not know there is a hardship until after attempting an examination. This would result in the NRC being asked to verbally approve alternatives on an expedited basis. The NRC disagreed with this comment and indicated that approval was required prior to implementation of the alternative. Implementation is considered to occur at the time the licensee needs to rely on the alternative to satisfy ASME code requirements. During a public meeting discussing the implementation of Code Case N‐770‐1, a clarification question was asked “Does the NRC clarification contained in the rulemaking (36248) allow the alternative to be performed prior to


RELIEF REQUESTS


approval as long as it is not credited until the NRC has reviewed and approved the request?” The NRC responded “The NRC clarification contained in the rulemaking (36248) allows the alternative inspection to be performed prior to approval as long as the inspection is not implemented (credited to satisfy the inspection requirement) until the NRC has reviewed the alternative inspection and approved the relief request for the alternative inspection” (MO112240818).

Relief Request Required due to Impracticality or Limited Examinations

10 CFR 50.55a(g)(5)(iii) and (iv) allows relief to be requested in instances when a Code requirement is deemed impractical with (iv) being specific to examination requirements that are determined to be impractical. The provisions of these two paragraphs are typically used to address impracticalities like limited examination coverage. Under 10 CFR 50.55a(g)(5)(iii), relief requests must be provided to the NRC no later than 12 months after the expiration of the 120‐month inspection interval for which relief is sought. Under 10 CFR 50.55a(g)(5)(iv), relief requests for examination impracticalities must be provided to the NRC no later than 12 months after the end of the active 120‐month interval

In cases where the ASME Section XI requirements for inservice inspection are considered impractical, Xcel Energy Inc. will notify the NRC and submit information to support the determination, as required by 10 CFR 50.55a(g)(5). The submittal of this information will be referred to as a Request for Relief.

Requests to use Later Edition and Addenda of ASME Section XI

On July 28, 2004, the NRC published Regulatory Issue Summary (RIS) 2004‐12, “Clarification on Use of Later Editions and Addenda to ASME OM Code and Section XI” and RIS 2004‐16 “Use of Later Editions and Addenda to ASME Code Section XI for Repair/Replacement Activities”. These RIS clarify the NRC position on using Editions and Addenda of Section XI, in whole or in part, later than those specified in the ISI program or the R/R Program. If the desired Edition or Addenda are referenced in 10 CFR 50.55a(b)(2), the request is submitted following the guidance of the RIS. These types of request are not required to demonstrate hardship, difficulty, or provide evidence of quality and safety. They do need to ensure that all related requirements are also used. Requests to use edition and/or addenda of ASME Section XI that are referenced in 10 CFR 50.55a(b)(2) that are later than the initial Code of Record established for the ISI program shall be submitted under the provisions of 10 CFR 50.55a(g)(4)(iv).


RELIEF REQUESTS


Table B.1‐1 Monticello Nuclear Generating Plan

Fifth Interval Relief Requests Relief Request

Relief Request Description Xcel Energy Inc. Correspondence

NRC SER Correspondence

RR‐001 Reactor Vessel Circumferential Welds L‐MT‐10‐014 (dated 03/12/10)

(SER dated 02/08/11)

RR‐002 Request to Use Code Case N‐702 “Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle‐to‐Shell Welds, Section XI, Division 1”

L‐MT‐11‐056 (dated 09/28/11)

Awaiting NRC SER

RR‐003 Request to Use Code Case N‐716 “Alternative Piping Classification and Examination Requirements”

To be submitted at a later date

RR‐004 Request to Use Code Case N‐661‐2 “Alternative Requirements for Wall Thickness Restoration of Class 2 and 3 Carbon Steel Piping for Raw Water Service”

To be submitted with the ISI Plan

RR‐005 Request to Use Code Case N‐795 “Alternative Requirements for BWR Class 1 System Leakage Test Pressure Following Repair/Replacement Activities”


RR‐006 Request to Use Code Case N‐532‐5 “Repair/Replacement Activity Documentation Requirements and Inservice Inspection Summary Report Preparation and Submission”


RR‐007 Request in Accordance with 10 CFR50.55a(g)(4)(iv) for Repair/Replacement Activities (IWE Program)

L‐MT‐11‐056 (dated 09/28/11)

Awaiting NRC SER

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan APPENDIX C.1

QUALITY GROUP CLASSIFICATION DRAWINGS (ISI BOUNDARY DRAWINGS)


The following table contains a list of the ISI Boundary Drawings that are used in conjunction with the ISI Classification Basis Document to define the boundaries of the Monticello piping systems that are examined in accordance with ASME Section XI. The ISI Boundary Drawings are located in and maintained in accordance with procedure PEI‐01‐02‐03.

BOUNDARY DRAWING NO.

SYSTEM/DESCRIPTION

1.5‐1 ISI Index Key 1.5‐2 Main Steam 1.5‐3 Feedwater 1.5‐4 Reactor Circulation 1.5‐5 Core Spray 1.5‐6 Residual Heat Removal Loop A 1.5‐7 Residual Heat Removal Loop B 1.5‐8 High Pressure Coolant Injection (Steam Side) 1.5‐9 High Pressure Coolant Injection (Water Side) 1.5‐10 Reactor Core Isolation Cooling (Steam Side) 1.5‐11 Reactor Core Isolation Cooling (Water Side) 1.5‐12 Standby Liquid Control System 1.5‐13 Primary Containment Atmospheric Control 1.5‐14 Emergency Diesel Generator Emergency Service Water 1.5‐15 Emergency Service Water 1.5‐16 Residual Heat Removal Service Water 1.5‐17 Hydraulic Control Units 1.5‐18 Control Rod Drive (Scram Discharge Volume Piping) 1.5‐19 Compressed Air 1.5‐20 Demineralizer Water & Reactor Building Cooling Water 1.5‐21 Reactor Water Cleanup & Liquid Radwaste 1.5‐22 Traversing In‐core Probe 1.5‐23 Excess‐Flow Check Valves 1.5‐24 Not Used 1.5‐25 Not Used 1.5‐26 Primary Containment Sampling 1.5‐27 Reactor Vessel Instrumentation


ISI ISOMETRIC DRAWINGS


The following table contains a list of the ISI Isometric Drawings that are used for locating the components and/or welds that are selected to be examined during the current interval. The ISI Isometric Drawings are located in and maintained in accordance with PEI‐01‐02‐04.

ISO NO. SYSTEM/DESCRIPTION CLASS 1 & 2 DRAWINGS ISI Fig. 0 Reactor Vessel Interior ISI Fig. 1 Reactor Vessel Top Head ISI Fig. 2 CRD Location Reactor Vessel ISI Fig. 3 Reactor Vessel Bottom Head ISI Fig. 4 Circumferential & Longitudinal Reactor Vessel Welds ISI Fig. 5 Reactor Vessel Nozzles ISI Fig. 6 Reactor Vessel Bolting ISI‐13142‐17‐A Residual Heat Removal “A” Suction ISI‐13142‐17‐B High Pressure Coolant Injection (Water) ISI‐13142‐17‐C Residual Heat Removal “B” Suction ISI‐13142‐18‐A Residual Heat Removal “B” ISI‐13141‐18‐B Residual Heat Removal “B” Discharge ISI‐13142‐18‐C Residual Heat Removal “B” Discharge ISI‐13142‐19‐A High Pressure Coolant Injection (Steam Side) Discharge ISI‐13142‐19‐B Reactor Core Isolation Cooling (Steam Side) Discharge ISI‐13142‐20‐A Core Spray “A” Suction ISI‐13142‐20‐B Core Spray “B” Suction ISI‐13142‐26‐A Core Spray “B” Discharge ISI‐13142‐26‐B Core Spray “B” Discharge ISI‐13142‐26‐C Core Spray “B” Discharge ISI‐13142‐26‐D Core Spray “B” Discharge ISI‐13142‐29‐A Reactor Building Cooling Water ISI‐13142‐31‐A Core Spray “A” Discharge ISI‐13142‐31‐B Core Spray “A” Discharge ISI‐13142‐31‐C Core Spray “A” Discharge ISI‐13142‐31‐D Core Spray “A” Discharge ISI‐13142‐33‐A Main Steam “A” ISI‐13142‐34‐A Main Steam “B” ISI‐13142‐35‐A Main Steam “C” ISI‐13142‐36‐A Main Steam “D” ISI‐13142‐37‐A Residual Heat Removal “A” Discharge ISI‐13142‐37‐B Containment Spray (RHR “A”) ISI‐13142‐37‐C Residual Heat Removal “A” Discharge ISI‐13142‐37‐D Containment Spray (RHR “A”) ISI‐13142‐37‐E Containment Spray (RHR “A”) ISI‐13142‐40‐A High Pressure Coolant Injection (Water Side) Discharge ISI‐13142‐40‐B High Pressure Coolant Injection (Water Side) Discharge




ISO NO. SYSTEM/DESCRIPTION ISI‐13142‐41‐A Reactor Core Isolation Cooling (Water Side) Suction ISI‐13142‐42‐A High Pressure Coolant Injection (Steam Side) ISI‐13142‐43‐A Reactor Core Isolation Cooling (Steam Side) ISI‐13142‐48‐A Residual Heat Removal Service Water ISI‐13142‐48‐B Residual Heat Removal Service Water ISI‐13142‐49‐A Residual Heat Removal “A” Shutdown Cooling ISI‐13142‐51‐A Residual Heat Removal “A” ISI‐13142‐51‐B Residual Heat Removal “B” ISI‐13142‐51‐C Residual Heat Removal “B” ISI‐13142‐51‐D Residual Heat Removal “B” ISI‐13142‐52‐A Feedwater C & D ISI‐13142‐53‐A Feedwater A & B ISI‐13142‐62 Fuel Pool Emergency Cooling ISI‐13142‐67 Fuel Pool Emergency Cooling ISI‐16 Jet Pump Instrument Nozzle ISI‐19 Reactor Instrument Nozzles ISI‐47 Reactor Core Isolation Cooling Pump & Turbine ISI‐48 Residual Heat Removal Pumps ISI‐49 Core Spray Pump Supports ISI‐73880‐A Reactor Water Cleanup ISI‐74209‐1‐A Recirculation “A” Drain ISI‐74210‐1‐A Recirculation “B” Drain ISI‐74215‐A Standby Liquid Control ISI‐782‐A Reactor Head Vent ISI‐782‐A‐A Reactor Head Vent ISI‐786‐A Main Steam Condensate Leakoff ISI‐7905‐32‐A Residual Heat Removal Heat Exchanger “A” ISI‐7905‐32‐B Residual Heat Removal Heat Exchanger “B” ISI‐821‐A Reactor Bottom Head Drain ISI‐8292‐42‐A High Pressure Coolant Injection Pumps ISI‐8292‐48‐A High Pressure Coolant Injection Turbine ISI‐93268‐1‐A Control Rod Drive Scram Header “A” ISI‐93268‐3‐A Control Rod Drive Scram Header “B” ISI‐94699‐A Primary Containment & Atmospheric Control ISI‐94879‐A Spare Penetration X‐47 ISI‐94966‐A Primary Containment & Atmospheric Control ISI‐94966‐B Containment Air Purge ISI‐97003‐A Residual Heat Removal Return Loop “A” ISI‐97003‐B Residual Heat Removal Supply Loop “A” ISI‐97004‐A Residual Heat Removal Return Loop “A” ISI‐97005‐A Recirculation Loop “A” ISI‐97005‐B Recirculation Manifold “A”




ISO NO. SYSTEM/DESCRIPTION ISI‐97005‐C Recirculation Pump “A” Supports ISI‐97006‐A Recirculation Loop “B” ISI‐97006‐B Recirculation Manifold “B” ISI‐97006‐C Recirculation Pump “B” Supports ISI‐97007‐A Reactor Instrument Nozzle N‐11B ISI‐97008‐A Reactor Instrument Nozzle N‐11A ISI‐97027‐A Residual Heat Removal Equalizer ISI‐105531‐A Standby Gas Treatment & Reactor Plenum ISI‐158074‐A Torus Hard Pipe Vent CLASS 3 DRAWINGS ND‐ISI‐100 Residual Heat Removal Service Water ND‐ISI‐101 Residual Heat Removal Service Water ND‐ISI‐102 Residual Heat Removal Service Water ND‐ISI‐103 Residual Heat Removal Service Water ND‐ISI‐104 Residual Heat Removal Service Water ND‐ISI‐105 Residual Heat Removal Service Water ND‐ISI‐106 Residual Heat Removal Service Water ND‐ISI‐107 Residual Heat Removal Service Water ND‐ISI‐108 Residual Heat Removal Service Water ND‐ISI‐109 Residual Heat Removal Service Water ND‐ISI‐110 Residual Heat Removal Service Water ND‐ISI‐111 Residual Heat Removal Service Water CLASS MC DRAWINGS ISI‐8291‐76 Class MC Supports 1.5‐81 Downcomer Restraints 1.5‐82 Vent Line & Header Restraints NH‐95932‐A Ring Header Seismic Restraints (Bays 1‐4) NH‐95932‐B Ring Header Seismic Restraints (Bays 5‐8) NH‐95932‐C Ring Header Seismic Restraints (Bays 9‐12) NH‐95932‐D Ring Header Seismic Restraints (Bays 13‐16) NX‐8291‐34‐A Vent Line & Header NX‐8291‐34‐C Downcomer, Vent Line & Header Supports NON‐CODE AUGMENTED NC‐ISI‐37 Reactor Core Isolation Cooling Feedwater NC‐ISI‐51 Control Rod Drive to Reactor Water Cleanup

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan APPENDIX D

INSPECTION PLAN AND SCHEDULE TABLES


Code Category Summary

Category Item

Number Description

Exam Method

Number of Components in Item No.

Required to be

Examined During Interval

Examination Percentage Required

Number to be Examined in Interval

Number to be

Examined in First Period

Number to be

Examined in Second Period

Number to be

Examined in Third Period

B‐A Pressure Retaining Welds in Reactor Vessel

B‐A B1.11 Reactor Vessel Circumferential Shell Welds

Volumetric 4 1(1) 25% All Welds 1 0 0

B‐A B1.12 Reactor Vessel Longitudinal Shell Welds

Volumetric 8 8 100% All Welds 0 0 8

B‐A B1.21 Reactor Vessel Circumferential Head Welds

Volumetric 2 2 100% Accessible

Length of All Welds

0 0 2

B‐A B1.22 Reactor Vessel Meridional Head Welds

Volumetric 16 16 100% Accessible

Length of All Welds

0 0 16

B‐A B1.30 Reactor Vessel Shell‐to‐Flange Weld

Volumetric 1 1 100% Weld 0 0 1(2)

B‐A B1.40 Reactor Vessel Head‐to‐Flange Weld

Volumetric and Surface

1 1 100% Weld 0 0 1(2)





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


Category Total 32 29 1 0 28

Notes for Cat. B‐A

Note 1: Request for Alternative RR‐001 was submitted 03/12/10 (L‐MT‐10‐014) to use BWRVIP‐05. NRC Safety Evaluation Report was issued 02/08/11 approving this alternative. Approximately 2 to 3 percent of each circumferential weld shall be examined. Welds VCBA‐2, VCBB‐3, and VCBB‐4 will be examined with the longitudinal welds at the intersection point. VCBB‐1 will be examined through the Nozzle Window N1A.

Note 2: Deferral is permissible in successive intervals per Note 5 of Item Numbers B1.30 and B1.40 if (a) no welded repair/replacement activities have been performed either on shell‐to‐flange or head‐to‐flange; and (b) neither the shell‐to‐flange weld nor head‐to‐flange weld contains identified flaws or relevant conditions that require successive inspection in accordance with IWB‐2420(b).





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐D Full Penetration Welded Nozzles in Vessels

B‐D B3.90 Reactor Vessel Nozzle‐to‐Vessel Welds

Volumetric 30(1) 16 55%(2) Same as 1st

Interval 5 6 5

B‐D B3.100 Reactor Vessel Nozzle Inside Radius Section

Volumetric 30(1) 16 55%(2) Same as 1st

Interval 5 6 5


Notes for Cat. B‐D

Note 1: Bottom Head Drain Nozzle (N‐15) is inaccessible and exempted per IWB‐1220(c).

Note 2: Request for Alternative (RR‐002) uses Code Case N‐702 for selection of Nozzle‐to‐Vessel and Inner Radius Section on the following nozzles (N2, N3, N5, N6, and N8). A total of 7 nozzle‐to‐vessel and inner radius sections were selected for examination under RR‐002. The Code Case allows 25% to be selected for examination over the interval. Recirculation Outlet Nozzle (N1) did not pass the criteria and therefore was not included in the 25% population. BWR Feedwater and CRD Nozzles are excluded.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐G‐1 Pressure Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter

B‐G‐1 B6.10 Reactor Vessel Closure Head Nuts

Visual, VT‐1 1(5) 1 100% Same as for 1st interval

0 0 1

B‐G‐1 B6.20 Reactor Vessel Closure Studs

Volumetric(4) 1(5) 1 100% Same as for 1st interval

0 0 1

B‐G‐1 B6.40 Reactor Vessel Threads in Flange

Volumetric 1(5) 1 100% Same as for 1st interval

0 0 1

B‐G‐1 B6.50 Reactor Vessel Closure Washers, Bushings

Visual, VT‐1 2(5) (6) 2 100% Same as for 1st interval

0 0 2

B‐G‐1 B6.180 Pump Bolts and Studs Volumetric(4) 2 1 50%(2) Same as for 1st interval

1 0 0

B‐G‐1 B6.190 Pumps Flange Surface when connection disassembled

Visual, VT‐1 2 0 0%(1)(3) Same as for 1st interval

0 0 0





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐G‐1 B6.200 Pumps Nuts, Bushings, and Washers


0 0 0


Notes for Cat.B‐G‐1

Note 1: Not Required unless disassembled

Note 2: Volumetric examination of bolting of heat exchangers, pump, or valves may be conducted on one heat exchanger, one pump, or one valve among a group of heat exchangers, pumps, or valves that are similar in design, type, and function. (Ref. Table IWB‐2500‐1, Examination Category B‐G‐1, Note 3). There are two Reactor Recirculation Pumps with 16 studs each. Therefore, only one of the pumps will be examined.

Note 3: For heat exchangers, piping, pumps, and valves, visual examinations are limited to components selected for examination under Examination Categories B‐B, B‐J, B‐L‐2, and B‐M‐2. (Ref. Table IWB‐2500‐1, Examination Category B‐G‐1, Note 4)

Note 4: If bolts or studs are removed for examination, surface examination meeting the acceptance standards of IWB‐3515 may be substituted for volumetric examination.

Note 5: There are a total of 64 reactor pressure vessel studs, nuts, bushings, and washers. In addition there are 64 “threads in flange” ligaments to examine.

Note 6: For Item Number B6.50, there are two items, one for the washers and one for the bushings.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐G‐2 Pressure Retaining Bolting, 2in. (50 mm) and Less in Diameter

B‐G‐2 B7.10 Reactor Vessel Bolts, Studs, and Nuts

Visual, VT‐1 2 0 0%(1) Same as for 1st interval(2)

0 0 0

B‐G‐2 B7.50 Piping Bolts, Studs, and Nuts

Visual, VT‐1 31 0 0%(1)(4) Same as for

1st interval (2) 0 0 0

B‐G‐2 B7.70 Valves Bolts, Studs, and Nuts


(2)(3) 0 0 0






Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


Notes for Cat. B‐G‐2

Note 1: Not required unless disassembled

Note 2: Examination is only required once per interval

Note 3: For components other than piping, examinations are limited to components selected for examination under Examination Categories B‐B, B‐L‐2, and B‐M‐2. (Ref. Table IWB‐2500‐1, Examination Category B‐G‐2, Note 2)

Note 4: Examination is limited to at least one piping flange in each group of piping flanges that are of the similar design, size, function, and service in the system.

Note 5: Only one valve of each group of valves is required as outlined in B‐M‐2





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐K Welded Attachments for Vessels, Piping, Pumps, and Valves

B‐K B10.10 Pressure Vessels Welded Attachments

Surface 5 1 20%(1) Same as for 1st interval

0 0 1

B‐K B10.20 Piping Welded Attachments

Surface 44 2 4.5%(2) Same as for 1st interval

1 1 0

B‐K B10.30 Pump Welded Attachments

Surface 6 1 16.7%(2) Same as for 1st interval

1 0 0


Notes for Cat. B‐K

Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. The welded attachment selected for examination, shall be an attachment under continuous load during normal operation or an attachment subject to a potential intermittent load during normal operation.

Note 2: For piping and pumps, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF‐2510 shall be examined. This ten percent sample is equivalent to 4.5% of all piping attachments and 16.7% of all pump attachments.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐L‐2 Pump Casings

B‐L‐2 B12.20 Pumps Pump Casing

(B‐L‐2) Visual, VT‐3 2 0 0%(1)(2)

Same as for first interval

0 0 0


Notes for Cat. B‐L‐2

Note 1: Examination is limited to at least one pump in each group of pumps performing similar functions in the system.






Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐M‐2 Valve Bodies

B‐M‐2 B12.50 Valve Body, Exceeding NPS 4 (DN 100) (B‐M‐2)

Visual, VT‐3 43 0 0%(1)(2)(3)

Same as for first interval

0 0 0


Notes for Cat. B‐M‐2

Note 1: Examination is limited to at least one valve in each group of valves that are of the same size, constructural design, manufacturing method, and that perform similar functions in the system.


Note 3: Valves have been grouped into 14 categories using scheduling notes BB, EE, FF, GG, HH, II, JJ, KK, LL, NN, OO, PP, QQ, and RR. One from each group is required to be examined per interval, if disassembled, totaling 14 valves.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐N‐1 Interior of Reactor Vessel

B‐N‐1 B13.10 Reactor Vessel, Vessel Interior (B‐N‐1)

Visual, VT‐3 1 1 100% Each

inspection period(2)

1 1 1

Category Total 1 1(1) 1 1 1

Notes for Cat. B‐N‐1

Note 1: Examination of this item number is required each period. Because these examinations are performed every period, the number required during the interval is three times the number of components. In order to keep the percentages accurate, only the number of components is reflected in the category total and not the total number of examinations required to be completed over the interval.

Note 2: Examination is limited to the spaces above and below the core made accessible by removal of components during normal refueling operations looking for loose parts or foreign material.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐N‐2 Welded Core Support Structures and Interior Attachments to Reactor Vessels

B‐N‐2 B13.20

Reactor Vessel (BWR) Interior Attachments Within Beltline Region

(B‐N‐2)

Visual, VT‐1 15 15 100% Same as for 1st interval

0 0 15

B‐N‐2 B13.30

Reactor Vessel (BWR) Interior Attachments Beyond Beltline Region (B‐N‐2)

Visual, VT‐3 25 25 100% Same as for 1st interval

0 0 25

B‐N‐2 B13.40 Core Support Structure Visual, VT‐3 20 20 100% Same as for 1st Interval

0 0 20


Notes for Cat. B‐N‐2

None





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐O Pressure Retaining Welds in Control Rod Drive and Instrument Nozzle Housings

B‐O B14.10 Reactor Vessel (BWR) Welds in Control Rod Drive CRD Housing

Volumetric or surface

48 6 12.5%

10% peripheral

CRD housings(1)

0 0 6

Category Total 48(2) 6(2) 0 0 6

Notes for Cat. B‐O

Note 1: There are 24 Peripheral CRDs, 10% of 24 is 3.

Note 2: There are two welds per Peripheral CRD, therefore the total required to be examined is twice the number of CRDs selected for examination. This results in a total of six (6) welds being selected for examination or 12.5% of total population.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


B‐P All Pressure Retaining Components

B‐P B15.10 Pressure retaining components (IWB‐5222(a))

Visual, VT‐2 1 1(3) 100% Each

refueling outage(1)

2(3) 2(3) 0

B‐P B15.20 Pressure retaining components (IWB‐5222(b))

Visual, VT‐2 1 1 100% Once per interval(2) 0 0 1


Notes for Cat. B‐P

Note 1: The system leakage test (IWB‐5220) shall be conducted prior to plant startup following a reactor refueling outage.

Note 2: The system leakage test (IWB‐5220) of the boundary of IWB‐5222(b) shall be performed at or near the end of the interval.

Note 3: Because these examinations are performed every outage, the number required during the interval is four times the number of procedures scheduled. In order to keep the percentages accurate, only the number of procedures scheduled is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


C‐A Pressure Retaining Welds in Pressure Vessels

C‐A C1.10 Pressure Vessels Shell Circumferential Welds

Volumetric 4 2 50%(1)(2) Each

inspection interval

1 0 1

C‐A C1.20 Pressure Vessels Head Circumferential Welds

Volumetric 4 2 50%(1)(3) Each

inspection interval

0 1 1


Notes for Cat. C‐A

Note 1: The examination may be limited to one vessel among the group of vessels of similar design, size, and function. (Ref. Table IWC‐2500‐1, Examination Category C‐A, Note 3)

Note 2: There are 2 circumferential shell welds on each RHRHX, requiring only 1 RHRHX to be scheduled or 2 of 4 shell welds are required to be examined for 50%

Note 3: There are 2 circumferential head welds on each RHRHX, requiring only 1 RHRHX to be scheduled or 2 of 4 head circumferential welds are required to be examined for 50%.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


C‐B Pressure Retaining Nozzle Welds in Vessels

C‐B C2.31

Nozzles With Reinforcing Plate in Vessels > 1/2 in. (13mm) Nominal Thickness Reinforcing Plate Welds to Nozzle and Vessel

Surface 4 2 50%(1) Each

inspection interval

1 1 0

C‐B C2.33

Nozzles With Reinforcing Plate in Vessels > 1/2 in. (13mm) Nominal Thickness Nozzle‐to‐Shell (Nozzle to Head or Nozzle to Nozzle) Welds When Inside of Vessel Is Inaccessible

Visual, VT‐2 4 2(3) 50%(1)(2) Each

inspection period

2 2 2






Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


Notes for Cat. C‐B

Note 1: The examination may be limited to one vessel or distributed among the group of vessels of similar design, size, and function. (Ref. Table IWC‐2500‐1, Examination Category C‐B, Note 4)

Note 2: The telltale hole in the reinforcing plate shall be examined for evidence of leakage while vessel is undergoing the system leakage test (each period) as required by Examination Category C‐H.

Note 3: Because these examinations are performed every period, the number required during the interval is three times the number of components selected for examination. In order to keep the percentages accurate, only the number of components selected is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


C‐C Welded Attachments for Vessels, Piping, Pumps, and Valves

C‐C C3.10 Pressure Vessels Welded Attachments(1)

Surface 6 1 16.7%(2)

Each identified

occurrence and each

inspection interval

1 0 0

C‐C C3.20 Piping Welded Attachments

Surface 78(5) 2 2.6%(3)(4)

Each identified

occurrence and each

inspection interval

1 0 1





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


C‐C C3.30 Pump Welded Attachments

Surface 6 1 16.7%(3)(4)

Each identified

occurrence and each

inspection interval

0 1 0


Notes for Cat. C‐C

Note 1: For multiple vessels of similar design, function, and service, only one welded attachment of only one of the multiple vessels shall be selected for examination. For single vessels, only one welded attachment shall be selected for examination. (Ref. Table IWC‐2500‐1, Examination Category C‐C, Note 4)

Note 2: There are 3 welded attachments on each Heat Exchanger, therefore 1 of 6 welded attachments are required to be examined for 16.7%

Note 3: For piping and pump welded attachments, a sample of 10% of the welded attachments associated with the component supports selected for examination under IWF‐2510 shall be examined.

Note 4: Examination is required whenever component support member deformation is identified. (Ref. Table IWC‐2500‐1, Examination Category C‐C, Note 6)

Note 5: There are 80 components in C3.20, but 2 of them do not require examination (Ref. Table IWC‐2500‐1, Examination Category C‐C, Note 1(b)).





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


C‐H All Pressure Retaining Components

C‐H C7.10 System leakage test (IWC‐5220)

Visual,

VT‐2(1) 11(2) 11(2)

Each inspection

period 11(2) 11(2) 11(2)


Notes for Cat. C‐H

Note 1: Visual examination of IWA‐5240

Note 2: There are 11 surveillance procedures that address each system (loop of each system as applicable). The systems included are SBLC, Core Spray, RHR, HPCI, RCIC, Feedwater, CRDH, RWCU. Each procedure is completed each period. Because these examinations are performed every period, the number required during the interval is three times the number of procedures scheduled. In order to keep the percentages accurate, only the number of procedures scheduled is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


D‐A Welded Attachments for Vessels, Piping, Pumps, and Valves

D‐A D1.20 Piping Welded Attachments

Visual, VT‐1 14 2 10%(1)

Each identified

occurrence and each

inspection interval (2)

0 1 1


Notes for Cat. D‐A

Note 1: Selected samples of welded attachments shall be examined each inspection interval. All welded attachments selected for examination shall be those most subject to corrosion, such as the welded attachments of the Service Water or Emergency Service Water systems. For welded attachments of piping, a 10% sample shall be selected for examination. This percentage sample shall be proportional to the total number of nonexempt welded attachments connected to the piping in each system subject to these examinations. (Ref. Table IWD‐2500‐1, Examination Category D‐A, Note 3).

Note 2: Examination is required whenever component support member deformation is identified. (Ref. Table IWD‐2500‐1, Examination Category D‐A, Note 4)





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


D‐B Pressure Retaining Components

D‐B D2.10 System Leakage Test (IWD‐5220)

Visual, VT‐2 8(1) 8(1) Each

inspection period

8(1) 8(1) 8(1)


Note for Cat. D‐B

Note 1: There are 8 surveillance procedures that address each system (loop of each system as applicable). The systems included are RHR Service Water, Emergency Diesel‐Emergency Service Water, Emergency Service Water, HPCI and RCIC. Each procedure is scheduled each period. Because these examinations are performed every period, the number required during the interval is three times the number of procedures scheduled. In order to keep the percentages accurate, only the number of procedures scheduled is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


F‐A Supports

F‐A F1.10a Class 1 Piping Supports ‐ One Directional

Visual, VT‐3 49 12 (1)

Each inspection

interval 4 4 5

F‐A F1.10b Class 1 Piping Supports ‐ Multi‐directional

Visual, VT‐3 31 8 (1)

Each inspection

interval 3 3 2

F‐A F1.10c Class 1 Piping Supports ‐ Thermal Movement

Visual, VT‐3 64 16 (1)

Each inspection

interval 7 5 6

F‐A F1.10 Total Class 1 Piping Supports

Visual, VT‐3 144 36(2) 25% Each

inspection interval

14 12 13


Visual, VT‐3 123 19 (1)

Each inspection

interval 6 6 7





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be



Visual, VT‐3 78 12 (1)

Each inspection

interval 6 3 3


Visual, VT‐3 67 10 (1)

Each inspection

interval 4 3 4

F‐A F1.20 Total Class 2 Piping Supports

Visual, VT‐3 268 41(2) 15% Each

inspection interval

16 12 14


Visual, VT‐3 69 7 (1)

Each inspection

interval 2 2 3


Visual, VT‐3 29 3 (1)

Each inspection

interval 1 1 1





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be



Visual, VT‐3 1 1 (1)

Each inspection

interval 0 1 0

F‐A F1.30 Class 3 Piping Supports Visual, VT‐3 99 10(2) 10% Each

inspection interval

3 4 4

F‐A F1.40a Supports other than Piping Supports – One Directional

Visual, VT‐3 23 15 (3)

Each inspection

interval 3 8 4

F‐A F1.40b Supports other than Piping Supports ‐ Multi‐directional

Visual, VT‐3 6 3 (3)

Each inspection

interval 1 0 2

F‐A F1.40c Supports other than Piping Supports ‐ Thermal Movement

Visual, VT‐3 10 7 (3)

Each inspection

interval 1 2 4





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


F‐A F1.40 Supports other than Piping Supports (Class 1,2, and 3)

Visual, VT‐3 39 25 64%(4) Each

inspection interval(2)

5 10 10

Category Total 550 112 (1)(2) 38 39 41

Notes for Cat. F‐A

Note 1: The total percentage sample shall be comprised of supports from each system, where the individual sample sizes are proportional to the total number of non‐exempt supports of each type and function within each system. (Ref. Table IWF‐2500‐1, Examination Category F‐A, Note 2)

Note 2: Based on Engineering Judgment with the Proration by System and Item Number as required by Footnote 1 above, the total number of supports selected for examination exceeds the number of supports required. The number of supports selected are reflected in the First, Second, and Third Period columns whereas the total number of supports required is reflected in the “Required to be Examination During the Interval” column.

Note 3: For multiple components other than piping, within a system of similar design, function, and service, the supports of only one of the multiple components are required to be examined. (Ref. Table IWF‐2500‐1, Examination Category F‐A, Note 3)

Note 4: One RPV (5 supports), one of two Reactor Recirculation Pumps (7 supports each (14 total)), one of four RHR pumps (1 support each (4 total)), one of two RHR Heat Exchangers (3 supports each (6 total)), one of two CS Pumps (1 support each (2 total)), one HPCI Pump (8 supports). Therefore 25 of 39 component supports are required to be examined for 64%.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


R‐A Risk Informed Piping Welds

R‐A R1.11 N‐716 Elements Subject to Thermal Fatigue

Volumetric 101 19 18.8%(1)(3)(4) Each

Inspection Interval

10 2 7

R‐A R1.20s N‐716 Elements not Subject to a Damage Mechanism

Visual, VT‐2 257 20(2) 7.8%(2) Each

refueling outage

40 40 20

R‐A R1.20 N‐716 Elements not Subject to a Damage Mechanism

Volumetric 453 48 10.6%(1)(3) Each

Inspection Interval

14 22 12

Category Total 811 87 11% 64 64 39





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


Notes for Cat. R‐A

Note 1: Percentages were determined using SI Calc 1000515.302.

Note 2: These welds are Socket Welds <NPS 2 and require a Visual, VT‐2 Examination each refueling outage per Note 9 of Code Case N‐716. Because these examinations are performed every outage, the number required during the interval is five times the number of components selected for examination. In order to keep the percentages accurate, only the number of components selected is reflected in the category total and not the total number of examinations required to be completed over the interval.

Note 3: The examination shall include any longitudinal welds at the location selected for examination. The longitudinal weld examination requirements shall be met for both transverse and parallel flaws within the examination volume defined for the intersecting circumferential welds.

Note 4: The length of the examination volume shown in Fig. IWB‐2500‐8(c) shall be increased by enough distance to include each side of the base metal thickness transition or counterbore transition.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


NUREG‐0619 BWR Control Rod Drive Return Line Nozzle Cracking Post Modification

NC NCR95‐

068

Assure integrity of the re‐routed piping at the tee on Isometric NX‐13142‐49A (NC‐ISI‐51), Line REW6‐3”‐DC.

Volumetric 3 3(1) 100% Each

schedule outage

6 6 3


Note for Cat. NUREG‐0619

Note 1: Commitment is to perform a volumetric examination of Welds W‐11, W‐12, and W‐13 each scheduled outage. So the total number of items required is five times (5 scheduled outages in the 5th 10 year interval) the number of items. Because these examinations are performed every outage, the number required during the interval is five times the number of components selected for examination. In order to keep the percentages accurate, only the number of components selected for examination is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


Mod 79Z018 RWCU Return Line Modification

NC NCR95‐

068

RCIC welds under Modification 79Z018 as on ISI ISO NC‐ISI‐37.

Volumetric 6 6 100% Every

scheduled outage

12 12 6


Notes for Cat. Mod 79Z018

Note 1: Commitment is to perform a volumetric examination of Welds W‐1, W‐2, W‐3, W‐4, W‐12 and W‐12A (ISO NC‐ISI‐37) each scheduled outage. So the total number of items required is five times (5 scheduled outages in the 5th 10 year interval) the number of items. Because these examinations are performed every outage, the number required during the interval is five times the number of components selected for examination. In order to keep the percentages accurate, only the number of components selected for examination is reflected in the category total and not the total number of examinations required to be completed over the interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


HELB High Energy Line Break

HELB R‐A Generic Letter 87‐11 Volumetric 42 42(1) 100% Every

Inspection Interval

5 23 14


Note for Cat. HELB

Note 1: A minimum of 10% of the total population is required to be selected for examination under Code Case N‐716. Actual number selected was 17 or about 40%.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


OE23699 Standby Liquid Control Tank

NC NC‐SAC Verify Structural Integrity of the SBLC Tank

Visual(1) 1 1 100%

Every inspection

interval 0 1 0


Note for OE23699

Note 1: Requirement is to perform a visual examination of the tank internals every interval.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


License Renewal M05011A

Class MC Supports License Renewal Aging Management Plans and Commitments

F‐A F1.40a Supports other than Piping Supports – One Directional


inspection interval

0 3 3

F‐A F1.40b Supports other than Piping Supports ‐ Multi‐directional


inspection interval

80 46 40


Notes for Class MC Supports

None





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


License Renewal

USAR Appendix K (K2.1.1)

Small Bore Class 1 Piping

R‐A R1.20s Class 1 small bore SS welds >NPS2 and <NPS4

Volumetric(2) 10 2 10%(1)

Each inspection

interval 0 1 1

Category Total 10 2 10% Each

inspection interval

0 1 1

Notes for Appendix K

Note 1: Examine 10% each inspection interval (2 welds were selected under Code Case N‐716 (RR‐003)).

Note 2: Examine in accordance with Appendix VIII. Weld Volume required is specified by Category B‐F or B‐J as applicable.





Category Item

Number Description

Exam Method


Required to be




Number to be


Number to be


Number to be


License Renewal M05020A M05021A M05022A

Feedwater Nozzles

B‐D B3.100 Nozzle Inner Radius Sections

Volumetric(2) 4 4 100%(1) Each

inspection interval

1 2 1


Notes for FW Nozzles

Note 1: Successive interval examinations shall be approximately 10 years from previous examination.

Note 2: The required examination volume is based on EPRI Modeling Report IR‐2004‐63 and extends beyond the volume specified in Figure IWB‐2500‐7(b).

Monticello Nuclear Generating Plant 5th Interval Inservice Inspection Plan APPENDIX E

RI‐ISI LIVING PROGRAM UPDATES


This is Revision 0 of the Monticello Nuclear Generating Plant Inservice Inspection (ISI) Plan for the 5th Ten‐Year Inspection Interval

Therefore there is no RI‐ISI Living Program Update at this time.