ML100920542

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American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section Xl, Inservice Inspection Program for the Fourth Ten-Year Inspection Interval
ML100920542
Person / Time
Site: Browns Ferry Tennessee Valley Authority icon.png
Issue date: 03/31/2010
From: Krich R
Tennessee Valley Authority
To:
Document Control Desk, Office of Nuclear Reactor Regulation
References
Download: ML100920542 (209)
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Tennessee Valley Authority 1101 Market Street, LP 3R Chattanooga, Tennessee 37402-2801 R. M. Krich Vice President Nuclear Licensing March 31, 2010 10 CFR 50.4 10 CFR 50.55a U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Unit 2 Facility Operating License No. DPR-52 NRC Docket No. 50-260

Subject:

American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section Xl, Inservice Inspection Program for the Fourth Ten-Year Inspection Interval In this letter, the Tennessee Valley Authority is submitting the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Xl, Inservice Inspection (ISI) Program for the Fourth Ten-Year Inspection Interval for Unit 2 of the Browns Ferry Nuclear Plant (BFN). The Code of Record for the Fourth Ten-Year Interval ISI Program is the 2004 Edition of the ASME Boiler and Pressure Vessel Code, Section Xl.

The Fourth Ten-Year Interval for BFN Unit 2 commences on May 25, 2011, and ends on May 24, 2021.

The applicable regulation, 10 CFR 50.55a(g), requires that ISI examinations for ASME Code Class 1, 2, and 3 components of a water-cooled nuclear facility meet the ISI requirements of the ASME Section Xl Code. Additionally, 10 CFR 50.55a(g)(4)(ii) requires that the ISI Program be updated every 120 months to the latest NRC approved Edition and Addenda of the ASME Section Xl Code, which is in effect 12 months prior to the start of the next 120-month Inspection Interval. The enclosed BFN Unit 2 ISI program update satisfies that requirement.

printed on recycled paper

U.S. Nuclear Regulatory Commission Page 2 March 31, 2010 The enclosure to this letter contains the updated BFN Unit 2 ASME Section Xl ISI Program for the Fourth Ten-Year Inspection Interval, which conforms to the 2004 Edition of the ASME Section Xl Code.

There are a total of six requests for relief submitted within the ISI program update. Four requests for relief (Attachments 7 through 10) submitted for staff review and approval are similar to, and consistent with, requests for relief that were reviewed and approved by the NRC for use in the BFN Unit 2 Third Ten-Year ASME Section Xl Inspection Interval. One request for relief (Attachment 11) has been submitted to the NRC for use in the BFN Unit 2 Third Ten-Year ISI Interval by TVA letter dated January 15, 2010, and is pending NRC approval.

One other request for relief (Attachment 6, Reactor Pressure Vessel Circumferential Weld Examinations) was approved by the NRC in a letter dated May 31, 2005, for the remaining term of operation under the existing license, which ends December 20, 2013, and is included in the Fourth Ten-Year Inspection Interval ISI Program for completeness. This request for relief will be resubmitted for NRC approval prior to the end of the original operating license period.

TVA anticipates that other ISI requests for relief may be necessary during the BFN Unit 2 Fourth Ten-Year ISI Interval. These requests for relief typically involve TVA's inability to obtain the specified examination coverage (greater than 90 percent) as a result of weld geometry or component interferences. These requests for relief will be submitted following the performance of the examinations when the specific component and examination coverage percentages have been determined.

There are no new regulatory commitments contained in this letter. If you have any questions, please contact Terry Cribbe at (423) 751-3850.

Respectfully, R. M. Krich

Enclosure:

American Society of Mechanical Engineers Boiler and Pressure Vessel Code, Section Xl, Inservice Inspection Program for the Fourth Ten-Year Inspection Interval cc: See Page 3

U.S. Nuclear Regulatory Commission Page 3 March 31, 2010 cc (Enclosure):

NRC Regional Administrator - Region II NRC Senior Resident Inspector - Browns Ferry Nuclear Plant

Enclosure Tennessee Valley Authority Browns Ferry Nuclear Plant Unit 2 American Society of Mechanical Engineers Boiler and Pressure Vessel Code,Section XI, Inservice Inspection Program for the Fourth Ten-Year Inspection Interval (See Attached)

Browns Ferry Nuclear Plant Unit 2 Surveillance Instruction 2-SI-4.6.G Inservice Inspection and Risk - Informed Inservice Inspection Program Unit 2 Revision 0040 Quality Related Level of Use: Information Use Effective Date: 05-25-2011 Responsible Organization: PGM, Engineering Program Group Prepared By: Frederick J. Nilsen Approved By: John Colvin

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 2 of 205 Current Revision Description Pages Affected: All.

Type of Change: Revision Tracking Number: 042 Updated for Fourth Ten Year Inspection Interval: May 25, 2011 to May 24, 2021.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 3 of 205 Table of Contents 1.0 INTRO DUCTIO N .................................................................................................. 7 1.1 P u rp o s e ........................................................................................................................ 7 1.2 S co p e ............................................................................................................................ 8 1.3 Frequency ..................................................................................................................... 9 1.3.1 Inspection Interval and Inspection Periods .................................................. 9 1.3.2 Extended Operating License ...................................................................... 9 1.4 Section XI Requirem ents ........................................................................................... 9 1.4.1 Section XI Code of Record and Risk Informed ISI Program ....................... 9 1.4.2 Code Cases ............................................................................................... 10 1.4.3 Preservice Inspection (PSI) History ........................................................... 12 1.4.4 First Inspection Interval History .................................................................. 12 1.4.5 Second Inspection Interval History ............................................................... 13 1.4.6 Third Inspection Interval History ................................................................ 14

2.0 REFERENCES

........................................................................................................... 15 2.1 Technical Requirem ents ......................................................................................... 15 2.2 Final Safety Analysis Report ................................................................................... 15 2.3 NRC Docum ents ...................................................................................................... L.15 2.4 Plant Procedures and Instructions .......................................................................... 16 2.5 Codes and Standards .............. ............................................................................... 17 2.6 Drawings ..................................................................................................................... 17 2.6.1 Unit 2 Section XI Code Class Boundary Drawings ................................... 17 2.6.2 Unit 2 ISI Component and Component Support Drawings ......................... 19 2.6.3 Unit 2 ISI Bolting, Nozzle, and W eld Drawings ......................................... 19 2.6.4 Unit 2 ISI Com ponent Support Drawings .................................................. 20 2.6.5 Unit 2 Risk-Informed ISI Segment Boundary Drawings ............................. 21 2.7 Vendor Manuals .......................................................................................................... 24 2.8 Reference Docum ents ............................................................................................. 24 2.9 Miscellaneous Docum ents ....................................................................................... 25 2.10 TVA Nuclear Standard Program s and Processes ................................................... 25 3.0 PRECA UTIO NS AND LIM ITATIO NS ...................................................................... 25 4.0 PREREQ UISITES ....................................................................................................... 26

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 4 of 205 Table of Contents (continued) 5.0 SPECIAL TOOLS AND EQUIPMENT .................................. 26 6.0 ACCEPTANCE STANDARDS ................................................................................ 26 7.0 INSTRUCTION STEPS/ELEMENTS ..................................................................... 26 7 .1 R e s p o n s ib ilitie s ........................................................................................................... 26 7.1.1 Materials Technology & Codes ................................................................. 26 7.1.2 Program Engineering Group ..................................................... 27 7.1.3 Site Engineering Design ............................................................................ 30 7.1.4 Site Licensing .......................................................................................... 30 7.1.5 Inspection Services Organization (ISO) .................................................... 30 7.1.6 Site Records Management (RM) ................................................................ 31 7.1.7 Authorized Nuclear Inservice Inspector (ANII) .......................................... 31 7.1.8 Nuclear Assurance ..................................................................................... 31 7 .2 Im p le m e ntatio n ........................................................................................................... 32 7.2.1 System for Maintaining Status of Examinations ....................................... 32 7.2.2 Notification of Indication (NOI) ................................................................. 33 7.2 .3 E xa m inatio ns ........................................................................................... . . 36 7.3 Components Subject to Examination ..................................................................... 37 7.3.1 ASME Class 1 Equivalent Components Subject to Examination (IW B ) ....................................................................................................... . . . 37 7.3.2 ASME Class 2 Equivalent Components Subject to Examination (IW C ) .................................................................................................... . . . 40 7.3.3 ASME Class 3 Equivalent Components Subject to Examination (IW D) and Non-Code Class Components ................................................ 41 7.3.4 Component Supports Subject to Examination (IW F) ................................. 41 7.3.5 Successive Examinations, Class 1, 2, 3, or Component S u p po rts ................................................................................................. . . 43 7.4 Calibration Standards ............................................................................................. 45 7.5 Records and Reports ............................................................................................. 45 7.5.1 ISI Summary Report ................................................................................ 45 7 .5 .2 S ite F ina l R e po rt ..................................................................................... .. 4 8 7 .5 .3 R ad iog ra phs ........................................................................................... . . 49 7.6 Requests for Relief (RFR) ....................................................................................... 49

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 5 of 205 Table of Contents (continued) 7.7 Repairs and Replacements ..................................................................................... 50 7.8 ASME Section Xl Programs Not Addressed By 2-SI-4.6.G ...................................... 50 7.8.1 System Pressure Tests ............................................................................ 50 7.8.2 Pump and Valve Inservice Testing ........................................................... 50 7.8.3 Snubber Inservice Testing ....................................................................... 50 7.8.4 Containment Inservice Inspection ............................................................. 50 7.9 ISI Data Base Update and Maintenance .................................................................. 50 7 .10 C o rre ctive A ctio n ........................................................................................................ 51 7.11 Augmented Examinations ....................................................................................... 51 7.11.1 Weld DSRHR-2-05A ................................................................................ 52 7.11.2 HPCI Pump Discharge Support Inspection Following Injection ................. 52 7.11.3 CRD Return Line Reroute .......................................................................... 53 7.11.4 Feedwater Nozzles .................................................................................. 53 7.11.5 Augmented Examination of Austenitic Stainless Steel and Dissimilar Metal Welds Susceptible to IGSCC (BWRVIP-75-A) ................ 54 7.11.6 Technical Surveillance Requirement (TSR) 3.4.3.2 .................................. 55 7.11.7 RPV Interior Examinations ........................................................................ 55 7.11.8 Level Instrumentation Nozzle Safe Ends BWRVIP-49 .............................. 55 7.11.9 Core Plate delta/P/Standby Liquid Control (SLC) Nozzle B W R V IP -27 ............................................................................................. . . 56 7.11.10 Core Spray and Recirc Inlet Safe Ends .................................................... 57 7.12 -N 12A Instrument Nozzle Safe End Weld Overlay .......................... 57 7.13 V oluntary Exam inations ........................................................................................... 57 7.14 Risk-Informed Inservice Inspection .......................................................................... 58 7 .14 .1 Introd uctio n ............................................................................................. . . 58 7 .14 .2 P u rpose ................................................................................................... . . 58 7 .1 4 .3 S c o p e ............................................................................................................ 58 7 .14 .4 F req ue ncy ................................................................................................ . . 59 7.14 .5 Living P rogram ......................................................................................... 59 7.14.6 Risk Informed Inservice Inspection Program Analysis .............................. 64 : Section 8.1 Examination Schedule ..................................................... 65 : Section 8.2 Listing of Welds for Generic Letter 88-01 ......................... 101

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 6 of 205 Table of Contents (continued) : Section 8.3 ASME Class 1 Equivalent Valve List ................................. 107 : Section 8.4 Class 1 Piping and Pump Flange Bolted Connections Group List Code Case N-652-1 ........................................ 111 : Section 8.4 Requests For Relief........................................................... 112 : TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION Xl, INSERVICE (ISI) AND AUGMENTED INSPECTION PROGRAM FOURTH TEN YEAR INSPECTION INTERVAL REQUEST FOR RELIEF 2- ISI-9, REVISION 1 ....................................... 113 : TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 ASME SECTION Xl INSERVICE INSPECTION PROGRAM (FOURTH TEN YEAR INSPECTION INTERVAL) REQUEST FOR RELIEF 2 -ISI-4 0 ..................................................................................................... 122 : TENNESSEE VALLEY AUTHORITY BROWNS FERRYNUCLEAR PLANT (BFN) UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

SECTION Xl, INSERVICE INSPECTION (ISI)PROGRAM (FOURTH TEN YEAR INSPECTION INTERVAL) REQUEST FO R RELIEF 2-ISI-41 ............................................................................... 159 : TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN) UNIT 2 ASME SECTION Xl INSERVICE INSPECTION PROGRAM (FOURTH TEN-YEAR INSPECTION INTERVAL) REQUEST FOR RELIEF 2-P D I-40 ................................................................................................... 185 0: Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN), Unit 2 American Society of Mechanical Engineers (ASME)Section XI, Inservice Inspection (ISI) Program Request for Relief 2-1S1-1, Updated Risk Informed Inservice Inspection Program ................................................................ 190 1: Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN) Unit 2 American Society of Mechanical Engineers,Section XI Inservice Inspection Program, Unit 2 Fourth Ten Year Inspection Interval Request for Relief 2-ISI-43 ..................... 196

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 7 of 205

1.0 INTRODUCTION

1.1 Purpose This Inservice Inspection (ISI) Program is an administrative Surveillance Instruction (SI) utilized to obtain data through nondestructive examinations (NDE) required by ASME Section XI. This procedure satisfies portions of the Technical Requirement 3.4.3 (TR 3.4.3.) and fulfills the requirements of SPP-9.1, related to NDE of Code Class 1, 2, and 3 (equivalent) components in accordance with applicable ASME Section XI requirements. NDE results are used to verify continued structural integrity of the subject components and their acceptability for continued service, and to determine if a flaw is an isolated case or of a generic nature.

This program shall serve as TVA's ISI/NDE plan and schedule for ASME Code Class 1, 2, and 3 (equivalent) components, in accordance with the requirements of ASME Section Xl, IWA-1400 for the fourth ten year ISI interval.

Owner's Statement data is as follows:

Owner: Tennessee Valley Authority Address of Corporate Office: Chattanooga Office Complex 1101 Market St.

Chattanooga, TN 37402-2801 Name & Address of Power Plant: Browns Ferry Nuclear Plant P. O. Box 2000 Decatur, AL 35609 Applicable Nuclear Power Unit: BFN, Unit 2 Construction Permit Date: Construction Permit was issued prior to January 1, 1971.

Commercial Operation Date: March 1, 1975 First 10 Year ISI Interval: March 1, 1975 to May 24, 1992 Second 10 Year ISI Interval May 24, 1992 to May 24, 2001 Third 10 Year ISI Interval May 25, 2001 to May 24, 2011 Fourth 10 Year ISI Interval May 25, 2011 to May 24, 2021

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page8 of 205 1.2 Scope This program is in effect for BFN Unit 2 during the fourth inspection interval which begins May 25, 2011. The Unit 1 and 3 ISI programs are contained in 1-SI-4.6.G and 3-SI.4.6.G, respectively.

The Inservice Inspection Program (ISI) is designed to comply with the Codes listed in Section 1.4 Requests for Relief are issued for regulatory review and approval when implementation of ASME Section XI requirements is determined to be impractical or when alternatives are adopted. These programs provide for implementation in accordance with the scheduling requirements of ASME Section XI, IWA-2400. Refer to Section 7.6.

Code Class (equivalent) boundaries are depicted on the color-coded drawings listed in Section 2.6. These drawings are prepared and maintained by Program Engineering Group and are issued and controlled through BFN Records Management (RM).

The ASME Section XI Code Class (equivalent) Boundary Drawings, ISI Drawings, and the RI-ISI Drawings, identify the components and systems to be examined. The Unit 2 ISI Component and Component Support Drawings are listed in Section 2.6.

Certain elements of ASME Section XI and OM Code (repairs and replacements, system pressure tests, pump and valve inservice testing, snubber examination and inservice testing, and containment inservice inspection) are implemented by other site procedures. Refer to Sections 7.7 and 7.8.

Specifics concerning performance of Nondestructive Examinations (NDE) are not part of this program, but are included in IEP-100, Nondestructive Examination Procedures.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 9 of 205 1.3 Frequency 1.3.1 Inspection Interval and Inspection Periods This inspection interval is from May 25, 2011 to May 24, 2021. This is the fourth inspection interval for BFN Unit 2 and is ten years long. The inspection interval is divided into three periods in accordance with ASME Section XI, IWA-2432, Inspection Program B.

The associated inspection period dates are listed below:

Inspection Period Minimum Exams Maximum Exams First (5/11-5/14) 16% 34%

Second (5/14-5/18) 50% 67%

Third (5/18-5/21) 100% 100%

The minimum and maximum examination percentages are applicable to those examination categories where deferral is NOT permissible.

The inspection interval may be extended in accordance with IWA-2430(e) if Unit 2 is out of service continuously for six months or more.

1.3.2 Extended Operating License The original operating license for BFN Unit 2 was to have expired on June 28, 2014.

The BFN Unit 2 license has been extended (Renewed License DPR-52) and will now expire at midnight on June 28, 2034. The fourth 10-Year ISI Interval will therefore overlap the original and extended license timeframes.

1.4 Section Xl Requirements 1.4.1 Section Xl Code of Record and Risk Informed ISI Program The ISI code of record for examination performance, including NDE method selection, examination volume/surface area, and evaluation, is the 2004 Edition of ASME Section XI.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Pagel of 205 1.4.1 Section XI Code of Record and Risk Informed ISI Program (continued)

The extent and frequency of examination of piping welds for the fourth inspection interval shall be in accordance with the Risk - Informed Inservice Inspection Program (RI-ISI). This program was implemented in the third inspection period of the second inspection interval as an alternative to the requirements of Subsections IWB and IWC for inservice inspection of Class 1 and 2 piping. The RI-ISI Program is prepared in accordance with 10CFR50.55a(a)(3)(i) and Code Case N-577. This code case provides risk-informed requirements for inservice inspection of Class 1, 2, 3, and Non-Code Class piping. These requirements are an alternative to the requirements of examination categories B-F, B-J, C-F-I, and C-F-2. Refer to Section 7.14.

ASME Section Xl Technical Interpretation #1N04-009, SC XI Technical Interpretation,Section XI, Table IWB-2500-1, Examination Category B-G-1, Item No. B6.50.

Certification of NDE personnel shall be in accordance with ANSI/ASNT CP-1 89, 1995, Standard for Qualification and Certification of Nondestructive Testing Personnel, as amended by the requirements of Division 1, ASME Section Xl, Division 1, including editions through the 2004 Edition.

The Nondestructive Examination (NDE) Program (NDE techniques, qualification of personnel, weld reference system, and standards for examination evaluation) will be in accordance with ASME Section XI, Division 1, including editions through the 2004 Edition.

NDE personnel performing ultrasonic examinations of bolting and piping, Reactor Pressure Vessel Welds, Weld "Overlay", and Reactor Pressure Vessel Nozzles and Dissimilar Metal welds are certified and qualified in accordance with the Performance Demonstration Initiative (PDI).

1.4.2 Code Cases The following Code Cases have been approved for use by the NRC in Regulatory Guide 1.147, Revision 15 and have been adopted by TVA for use at BFN Unit 2.

Code Case N-460, Alternative Examination Coverage for Class 1 and 2, Welds, Section Xl, Division 1.

Code Case N-504-3, Alternative Rules For Repair of Classes 1, 2, and 3 Austenitic Stainless Steel Piping,Section XI, Division 1 Nonmandatory Appendix Q, Weld Overlay Repair of Class 1, 2, and 3 Austenitic Stainless Steel Piping Weldments (2004 Edition with Addenda through 2005 of ASME Section Xl)

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 11 of 205 1.4.2 Code Cases (continued)

Code Case N-513-2, Evaluation Criteria for Temporary Acceptance of Flaws in Moderate Energy Class 2 or 3 Piping,Section XI, Division 1.

Code Case N-526, Alternative Requirements for Successive Inspections of Class 1, 2, and 3 VesselsSection XI, Division 1.

Code Case N-532-4, Repair/Replacement Activity Documentation Requirements and Inservice Summary Report Preparation and Submission,Section XI, Division 1.

Code Case N-552, Alternative Methods - Qualification For Nozzle Inside Radius Section From the Outside Surface section Xl, Division 1.

To achieve consistency with the 10 CFR 50.55a rule change published September 22, 1999 (64 FR 51370), incorporating Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems," to Section XI, add the following to the specimen requirements:

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

The number of false calls must not exceed three.

Code Case N-586-1, Alternative Additional Examination requirements for Class 1, 2, and 3 Piping, Components, and SupportsSection XI, Division 1.

Code Case N-613-1, Ultrasonic Examination of Full Penetration Nozzles in Vessels, Examination Category B-D, Item No.'s B3.10, and B3.90, Reactor Nozzle-To-Vessel Weld's Fig's. IWB-2500-7 (a), (b), and (c),Section XI, Division 1.

Code Case N-624, Successive Inspections,Section XI, Division 1.

Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1, subject to the following conditions:

In place of a UT examination, a visual examination with enhanced magnification that has a resolution sensitivity to detect a 1-mil width wire or crack, utilizing the allowable flaw length criteria of Table IWB-3512-1 with limiting assumptions on the flaw aspect ratio will be performed. The provisions of Table IWB-2500-1, Examination Category B-D, continue to apply except that, in place of examination volumes, the surfaces to be examined are the external surfaces shown in the figures applicable to this table (the external surface is from point M to point N in the figure).

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 2 of 205 1.4.2 Code Cases (continued)

Code Case N-652-1, Alternative Requirements to Categories B-G-1, B-G-2, and C-D Bolting Examination Methods and Selection Criteria Section XI, Division 1 Code Case N-658, Qualification Requirements for Ultrasonic Examination of Wrought Austenitic Pipe Welds,Section XI, Division 1.

Code Case N-663, Alternative Requirements for Classes 1 and 2 Surface Examinations,Section XI, Division 1.

Code Case N-664, Performance Demonstration Requirements for Examination of Unclad Reactor Pressure Vessel Welds, Excluding Flange Welds,Section XI, Division 1.

Code Case N-686, Alternative Requirements for Visual Examinations VT-1, VT-2, and VT-3 Section XI, Division 1.

Code Case N-695, Qualification Requirements for Dissimilar Metal Piping Welds,Section XI, Division 1.

Code Case N-700, Alternative Rules For selection of Class 1, 2, and 3 Vessel Welded Attachments for Examination, ASME Section XI, Division 1.

Code Case N-702, Alternative Requirements for Boiling Water Reactor (BWR)

Nozzle Inner Radius and Nozzle-to-Shell WeldsSection XI, Division 1.

(Reference 2-ISI-43)

Interpretation XI-1-04-18 Table IWB-2500-1, Examination Category B-G-1, Item No.

B6.50. Date Issued: March 8, 2005 IN04-009 1.4.3 Preservice Inspection (PSI) History For Unit 2, a preservice inspection (PSI) program was not required. TVA performed a self-imposed PSI program for Class 1 components to the 1971 Edition, Summer 1971 Addenda of ASME Section XI.

1.4.4 First Inspection Interval History The first period of the first interval, in effect from March 1,1975 through July 1, 1981, was to the 1971 Edition, Summer 1971 Addenda of ASME Section XI.

The long duration on this period was due to an extension for the fire outage and an additional one year extension in accordance with IWA-2400 to establish concurrent intervals for Units 1, 2, and 3 beginning with the second period.

See NRC letter dated June 20, 1986 (A02 860630 006) for approval of these adjustments.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Pagel3 of 205 1.4.4 First Inspection Interval History (continued)

The second period (July 1, 1981 through July 7, 1988) and the third period (February 26, 1986 through May 23, 1992) were to the 1974 Edition with Addenda through Summer 1975 of ASME Section XI.

Ultrasonic examination and evaluation of piping welds was upgraded to the 1977 Edition, Summer 1978 Addenda of ASME Section XI for these periods. This included examination per IWA-2232(b), IWA-2232(c), and Appendix III (to the extent specified in Request for Relief IS1-1 5) and evaluation per IWA-3000, IWB-3000, and IWC-3000 of the 1977 Edition, Summer 1978 Addenda.

The overlap of the second and third periods occurred because of the extended outage from September 15, 1984 to May 24, 1991 and TVA's decision to complete the second and third period examinations during this time to close out the first interval.

This decision was made since the first interval had been extended twice and it was prudent to end it and commence with a second inspection interval to a current edition of ASME Section XI.

Beginning January 1, 1992, the preservice inspection of pipe welds, including the extent of examination, (Examination Categories B-F, B-J, and C-F) were performed in accordance with the 1977 Edition, Summer 1978 Addenda of ASME Section Xl, IWA-2232, IWA-3000, IWB-2200(c), Table IWB-2500-1, and Table IWC-2500-1.

Code Cases N-234, 235, 307-1, 308, 341, 356, 416, 435-1, 460, and 461 thatwere approved by Regulatory Guide 1.147 were used at BFN during the first interval.

1.4.5 Second Inspection Interval History During the Second Interval (May 24, 1992 through May 24, 2001) the ISI Program was performed to Inspection Program B of the 1986 Edition of Section XI. The extent of examination for category B-J welds was in accordance with the 1974 Edition, Summer 1975 Addenda of ASME Section XI in accordance with 10CFR50.55a(b)(2)(ii).

Code Cases N-307-1, N-435-1, N-445, N- 457, N-460, N-461, N-491, and N-524 that were approved by Regulatory Guide 1.147 were used at BFN during the second interval.

The Risk - Informed Inservice Inspection Program (RI-ISI) was implemented in the third inspection period of the second inspection interval, as an alternative to the requirements of Subsections IWB and IWC for inservice inspection of Class 1 and 2 piping, the RI-ISI Program is prepared in accordance with 10CFR50.55a(a)(3)(i) and Code Case N-577.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 14 of 205 1.4.6 Third Inspection Interval History The code of record for the Unit 2 third inspection interval is the 1995 Edition through the 1996 Addenda of American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code,Section XI, Division I, in accordance with Title 10 Code of Federal Regulations (CFR) Part 50, 50.55a(g)(4). The effective code edition and addenda are determined in accordance with 10CFR50.55a(b)(2).

The extent and frequency of examination of piping welds for the Third Inspection Interval was in accordance with the Risk - Informed Inservice Inspection Program (RI-ISI). This program was implemented in the third inspection period of the second inspection interval as an alternative to the requirements of Subsections IWB and IWC for inservice inspection of Class 1 and 2 piping. The RI-ISI Program is prepared in accordance with 10CFR50.55a(a)(3)(i) and Code Case N-577. This code case provides risk-informed requirements for inservice inspection of Class 1, 2, 3, and Non-Code Class piping. These requirements are an alternative to the requirements of examination categories B-F, B-J, C-F-i, and C-F-2. Refer to Section 7.14.

Code Cases N-460, N-498-4, N-504-3, N-526, N-532-4, N-552, N-586-1, N-598, N-623, N-624, N-652, N-658, and N-686 that were approved by Regulatory Guide 1.147 were used at BFN during the third interval.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 15 of 205

2.0 REFERENCES

2.1 Technical Requirements BFN Unit 2 Technical Requirements Manual TR 3.4.3, Structural Integrity.

2.2 Final Safety Analysis Report Browns Ferry Nuclear Plant Updated Final Safety Analysis Report, Volume 2, Section 4.12.

2.3 NRC Documents 10CFR50.55a(g), Code of Federal Regulations.

10CFR50.2, Code of Federal Regulations.

Regulatory Guide 1.26, Quality Group Classifications and Standards for Water-,

Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants.

BWR (Boiling Water Reactor) Vessel and Internals Project, Technical Basis for Revisions to Generic Letter (GL) 88-01 Inspection Schedules (BWRVIP-75).

Reference Safety Evaluation Report (SER), RIMS # L44 020320 001 from NRC dated March 15, 2002. BWRVIP-75-A Final Report dated October 2005.

Regulatory Guide 1.147, Inservice Inspection Code Case Acceptability ASME Section XI Division I.

Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions of Plant-Specific Changes to the Licensing Basis.

Regulatory Guide 1.178, An Approach for Plant-Specific Risk-Informed Decision making: Inservice Inspection of Piping.

Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping.

NUREG-0313, Rev. 2, Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping, Final Report.

NUREG-0619, BWR Feedwater Nozzle and Control Rod Drive Return Line Nozzle Cracking.

IE Bulletin 80-13, Core Spray Spargers.

NRC Information Notice 98-42: Implementation of 10 CFR 50.55a (g) Inservice Inspection Requirements.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 16 of 205 2.4 Plant Procedures and Instructions QADP-1, Conduct of Quality Assessment and Inspection NEDP-3, Drawing Control 2-SI-4.6.H-1, Visual Examination of Hydraulic and Mechanical Snubbers 1-SI-4.6.G, Inservice Inspection Program for Unit 1 3-SI-4.6.G, Inservice Inspection Program for Unit 3 2-SI-4.6.G-A, Analysis For Risk - Informed Inservice Inspection Program Unit 2 0-TI-140, Monitoring Program For Flow Accelerated Corrosion 0-TI-364, ASME Section XI Pressure Tests 0-TI-365, Reactor Pressure Vessel Internals Inspection (RPVII) Units 1, 2, and 3 0-TI-376, ASME Section XI Containment Inservice Inspection Program Units 1, 2 and 3 0-TI-400, ASME Section XI Inservice Inspection Program Responsibilities And Interface Document.

MSI-0-001-INS001, Reactor Vessel Internals Visual and Ultrasonic Inspection MSI-0-001-VSLOO1, Reactor Vessel Disassembly and Reassembly MCI-0-068-PMPOO1, Maintenance of Reactor Water Recirculation Pumps MCI-0-001-VLVO01, Main Steam Isolation Valves Atwood Morrill Co. Disassembly, Inspection, Rework, and Reassembly.

MCI-0-001-VLVO02, Main Steam Relief Valves Target Rock Model 7567 Disassembly, Inspection, Rework, and Reassembly.

IEP-100, Administration of Nondestructive Examination (NDE) Procedures.

IEP-200, Qualification and Certification Requirements for TVA Nuclear Nondestructive Examination (NDE) Personnel.

IEP-300, Qualification and Certification of Ultrasonic TVA Nuclear Personnel for Preservice and Inservice ASME Section XI Examinations.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 17 of 205 2.5 Codes and Standards ASME Boiler and Pressure Vessel Code,Section XI, 1971 Edition with Addenda through Summer 1971.

ASME Boiler and Pressure Vessel Code,Section XI, 1974 Edition with Addenda through Summer 1975.

ASME Boiler and Pressure Vessel Code,Section XI, Appendix III, 1977 Edition with Addenda through Summer 1978.

ASME Boiler and Pressure Vessel Code,Section V, 1974 Edition with Addenda through Summer 1975.

ASME Boiler and Pressure Vessel Code,Section XI, 1986 Edition with No Addenda.

ASME Boiler and Pressure Vessel Code,Section XI, 1989 Edition with No Addenda.

ASME Boiler and Pressure Vessel Code,Section XI, 1995 Edition with 1996 Addenda.

ASME Boiler and Pressure Vessel Code,Section XI, 2001 Edition with 2003 Addenda.

ASME Boiler and Pressure Vessel Code,Section XI, 2004 Edition 2.6 Drawings 2.6.1 Unit 2 Section XI Code Class Boundary Drawings 2-47E2600-57A-ISI, RCS Instrumentation.

2-47E600-58-1SI, Mech. Instr. and Controls.

2-47E610-43-1-ISI, Mech. Control Diag. Sampling and Water Quality.

2-47E2600-301-1SI, CRD Hyd. System.

2-47E2600-302-1SI, CRD Hyd. System.

2-47E600-599-1SI, Mech. I&C.

0-117C2556-4-1SI, Rack 25-18.

0-117C2556-5-1SI, Rack 25-18.

2-164C5981-4-1SI, Rack 25-7.

2-164C5981-5-1SI, Rack 25-7.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 18 of 205 2.6.1 Unit 2 Section XI Code Class Boundary Drawings (continued) 2-164C5984-4-ISI, Rack 25-56A.

2-164C5984-5-1SI, Rack 25-56B.

2-164C5985-4-1SI, Rack 25-57.

2-47E801-1-ISI, Main Steam.

2-47E801-2-ISI, Main Steam.

2-47E803-1-ISI, Feedwater.

2-47E803-5-ISI, Feedwater.

2-47E805-3-ISI, Heater Drains, Vents & Miscellaneous Piping 2-47E807-2-ISI, Turbine Drains & Misc. Piping.

2-47E81 0-1-ISI, Reactor Water Cleanup.

2-47E81 1-1-1SI, Residual Heat Removal.

2-47E812-1-ISI, High Pressure Coolant Injection.

2-47E813-1-1SI, Reactor Core Isolation Cooling.

2-47E814-1-ISI, Core Spray.

2-47E815-4-ISI, Aux. Boiler Sys.

2-47E817-1-ISI, Nuclear Boiler.

2-47E820-2-ISI, Control Rod Drive Hydraulic.

2-47E2820-6-ISI, Control Rod Drive Hydraulic.

0-47E839-5-ISI, Hypochlorite System.

2-47E822-1-1SI, Reactor Bldg Closed Cooling Water.

2-47E844-2-ISI, Raw Cooling Water.

2-47E852-1-ISI, Floor and Dirty Radwaste Drainage.

2-47E852-2-ISI, Clean Radwaste & Decon Drainage.

2-47E854-1-ISI, Standby Liquid Control.

2-47E855-1-ISI, Fuel Pool Cooling.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 19 of 205 2.6.1 Unit 2 Section XI Code Class Boundary Drawings (continued) 2-47E856-2-1SI, Demineralized Water.

2-47E858-1-ISI, RHR Service Water.

2-47E859-1-ISI, Emergency Equipment Cooling Water.

2-47E867-3-1SI, Sampling and Water Quality.

2.6.2 Unit 2 ISI Component and Component Support Drawings ISI-0368-C, Sheets 1 - 15, EECW and RHRSW Pumping Station Class 3.

3-ISI-0390-C, Sheets 1 - 3, EECW Unit 3 Class 3.

1-ISI-0391-C, Raw Cooling Water Unit 1 Class 3.

2.6.3 Unit 2 ISI Bolting, Nozzle, and Weld Drawings 2-CHM-2046-C, Reactor Vessel Nozzle and Weld Locations Class 1.

ISI-0444-C, Reactor Vessel Bottom Head Assy. Class 1.

ISI-0316-A, Reactor Vessel Clad Patches.

ISI-0343-A, Core Differential Pressure and Liquid Control Nozzle Weld Locations.

ISI-0351-A, Instrumentation Nozzles Class 1.

2-1SI-0312-B, Main Steam Bolting Class 1.

ISI-0347-B, Recirculation Inlet Nozzles Class 1.

ISI-0031-C, Reactor Building Closed Cooling Water System Class 2 Welds.

ISI-0040-C, CRD Hydraulic Header Class 2 Welds.

2-1SI-0103-C, Core Spray System Class 2 Welds.

2-ISI-0128-C, HPCI System Class 2 Welds.

2-1SI-0129-C, RCIC System Class 2 Welds.

2-ISI-0221-C, RHR System Class 1 Welds.

2-1SI-0222-C, Main Steam System Class 1 Welds.

ISI-0266-C, Vessel Stud Locations Class 1.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 20 of 205 2.6.3 Unit 2 ISI Bolting, Nozzle, and Weld Drawings (continued) 2-ISI-0269-C, Feedwater System Class 1 Welds.

2-ISI-0270-C, Recirculation System Class 1 Welds.

2-ISI-0271-C, Core Spray System Class 1 Welds.

2-ISI-0272-C, RWCU, RCIC, and CRD Systems Class 1 Welds.

2-ISI-0273-C, HPCI System Class 1 Welds.

2-ISI-0292-C, Control Rod Drive Penetrations, Drain Nozzle, and Flux Monitor Nozzles Class 1.

CHM-1 090-A, RPV Control Rod Drive Penetration, BFN.

CHM-1 091-A, RPV Support Shirt Weld, BFN.

CHM-1094-A, RPV Nozzle to Vessel Welds, BFN.

CHM-1095-A, RPV Vessel and Head Welds, BFN.

2-ISI-0380-C, Standby Liquid Control System Class 1 Welds.

2-ISI-0383-C, Feedwater Instrumentation Class 1 Welds.

2-ISI-0406-C, RHR Heat Exchanger Welds and Supports Class 2.

2-ISI-0407-C, Recirculation Pump Bolting Class 1.

ISI-0408-C, Closure Head Assembly Class 1.

2-ISI-0410-C, Jet Pump Instrument Nozzle Class 1.

2-MSG-0018-C, RHR System Class 2 Welds.

2-MSG-0021-C, Main Steam Class 2 Welds.

2.6.4 Unit 2 ISI Component Support Drawings ISI-03110-B, RHR Pump Class 2.

ISI-0032-C, Reactor Building Closed Cooling Water System Class 2.

2-ISI-0041-C, CRD Header Class 2.

2-ISI-0079-C, Main Steam System Class 2.

2-ISI-0105-C, Core Spray System Class 2.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 21 of 205 2.6.4 Unit 2 ISI Component Support Drawings (continued) 2-1SI-0130-C, HPCI System Class 2.

2-1SI-0131-C, RCIC System Class 2.

2-1SI-0133-C, FPC System Class 3.

2-ISI-0145-C, RHR Service Water System Class 3.

2-ISI-0274-C, RWCU, RCIC, and CRD Systems Class 1.

ISI-0275-C, HPCI System Class 1.

2-ISI-0276-C, RHR System Class 1.

2-ISI-0277-C, Feedwater System Class 1.

2-ISI-0278-C, Recirculation System Class 1.

2-ISI-0279-C, Main Steam System Class 1.

ISI-0280-C, Core Spray System Class 1.

2-ISI-0324-C, RHR System Class 2.

2-1SI-0379-C, Standby Liquid Control System Class 1.

2-1SI-0412-C, Main Steam Relief Valve Blowdown Class 2.

2-ISI-0415-C, Reactor Vessel Class 1.

2.6.5 Unit 2 Risk-Informed ISI Segment Boundary Drawings 2-001-RIISI-01, Main Steam Risk Informed Segment Boundary 2-001-RIISI-02, Main Steam Risk Informed Segment Boundary 2-001-RIISI-03, Main Steam Risk Informed Segment Boundary 2-001-RIISI-04, Main Steam Risk Informed Segment Boundary 2-002-RIISI-01, Condensate Storage and Supply Risk Informed Segment Boundary 2-002-RIISI-02, Condensate Risk Informed Segment Boundary 2-002-RIISI-03, Condensate Risk Informed Segment Boundary 2-002-RIISI-04, Condensate Risk Informed Segment Boundary

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 22 of 205 2.6.5 Unit 2 Risk-Informed ISI Segment Boundary Drawings (continued) 2-003-RIISI-01, Reactor Feedwater Risk Informed Segment Boundary 2-003-RIISI-02, Reactor Feedwater Risk Informed Segment Boundary 2-003-RIISI-03, Reactor Feedwater Risk Informed Segment Boundary 0-023-RIISI-01, Unit 0 RHR Service Water Risk Informed Segment Boundary 1-023-RIISI-01, Unit 1 RHR Service Water Risk Informed Segment Boundary 2-023-RIISI-01, Unit 2 RHR Service Water Risk Informed Segment Boundary 3-023-RIISI-01, Unit 3 RHR Service Water Risk Informed Segment Boundary 0-024-RIISI-01, Units land 0 Raw Cooling Water Risk Informed Segment Boundary 1-024-RIISI-01, Unit 1 Raw Cooling Water Risk Informed Segment Boundary 2-024-RIISI-01, Unit 2 Raw Cooling Water Risk Informed Segment Boundary 3-024-RIISI-01, Unit 3 Raw Cooling Water Risk Informed Segment Boundary 3-024-RIISI-02, Unit 3 Raw Cooling Water Risk Informed Segment Boundary 3-024-RIISI-03, Unit 3 Raw Cooling Water Risk Informed Segment Boundary 2-027-RIISI-01, Unit 2 Condenser Circulating Water Risk Informed Segment Boundary 2-063-RIISI-01, Unit 2 Standby Liquid Control Risk Informed Segment Boundary 0-067-RIISI-01, Unit 0 Emergency Equipment Cooling Water Risk Informed Segment Boundary 0-067-RIISI-02, Unit 0 Emergency Equipment Cooling Water Risk Informed Segment Boundary 2-067-RIISI-01, Unit 2 Emergency Equipment Cooling Water Risk Informed Segment Boundary 2-067-RIISI-02, Unit 2 Emergency Equipment Cooling Water Risk Informed Segment Boundary 2-067-RIISI-03, Unit 2 Emergency Equipment Cooling Water Risk Informed Segment Boundary

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 23 of 205 2.6.5 Unit 2 Risk-Informed ISI Segment Boundary Drawings (continued) 2-068-RIISI-01, Unit 2 Reactor Water Recirculation, Drains, Vents, and Blowdown Risk Informed Segment Boundary 2-069-RIISI-01, Unit 2 reactor Water Cleanup Demineralizer Risk Informed Segment Boundary 2-069-RIISI-02, Unit 2 reactor Water Cleanup Risk Informed Segment Boundary 2-070-RIISI-01, Unit 2 Reactor Building Closed Cooling Water Risk Informed Segment Boundary 2-070-RIISI-02, Unit 2 Reactor Building Closed Cooling Water Risk Informed Segment Boundary 2-070-RIISI-03, Unit 2 Reactor Building Closed Cooling Water Risk Informed Segment Boundary 2-070-RIISI-04, Unit 2 Reactor Building Closed Cooling Water Risk Informed Segment Boundary 2-071-RIISI-01, Unit 2 Reactor Core Isolation Cooling Risk Informed Segment Boundary 2-071-RIISI-02, Unit 2 Reactor Core Isolation Cooling Risk Informed Segment Boundary 2-071-RIISI-03, Unit 2 Reactor Core Isolation Cooling Risk Informed Segment Boundary 2-073-RIISI-01, Unit 2 High Pressure Coolant Injection Risk Informed Segment Boundary 2-073-RIISI-02, Unit 2 High Pressure Coolant Injection Risk Informed Segment Boundary 2-074-RIISI-01, Unit 2 Residual Heat Removal Risk Informed Segment Boundary 2-074-RIISI-02, Unit 2 Residual Heat Removal Risk Informed Segment Boundary 2-074-RIISI-03, Unit 2 Residual Heat Removal Risk Informed Segment Boundary 2-074-RIISI-04, Unit 2 Residual Heat Removal Risk Informed Segment Boundary 2-074-RIISI-05, Unit 2 Residual Heat Removal Risk Informed Segment Boundary 2-075-RIISI-01, Unit 2 Core Spray System Risk Informed Segment Boundary

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 24 of 205 2.6.5 Unit 2 Risk-Informed ISI Segment Boundary Drawings (continued) 2-075-RIISI-02, Unit 2 Core Spray System Risk Informed Segment Boundary 2-078-RIISI-01, Unit 2 Fuel Pool Cooling Risk Informed Segment Boundary 2-085-RIISI-01, Unit 2 Control Rod Drive Risk Informed Segment Boundary 2-085-RIISI-02, Unit 2 Control Rod Drive Risk Informed Segment Boundary 2-085-RIISI-03, Unit 2 Control Rod Drive Risk Informed Segment Boundary 2-085-RIISI-04, Unit 2 Control Rod Drive Risk Informed Segment Boundary 2.7 Vendor Manuals A. BFN-VTM-B014-0010, B&W Reactor Pressure Vessel Manual, Contract 66C60-90744 B. BFN-VTM-B580-0010, B&J Recirculation Pump Manual, Contract 67C60-91750 C. BFN-VTM-B260-0030, Bingham Pump Co. RHR Pump Manual, Contract 66C60-90744 D. BFN-VTM-P160-0010, VTM-P160-0010, Vendor Technical Manual for Perfex Corp. Heat Exchangers, Types NEN, CEU, CES, and CEN 2.8 Reference Documents A. ASME Section Xl Code Cases as listed in Section 1.4.

B. GE SIL No. 571, Instrument Nozzle Safe End Inspection C. Boiling Water Reactor Owners Group (BWROG) Licensing Topical Report, "Alternate BWR Feedwater Nozzle Inspection Requirements",

GE-NE-523-A71-0594, Revision 1, August 1999. Reference TVA submittal dated October 23, 2000 (RIMS # R08 001023 713 D. Boiling Water Reactor Vessel and Internals Project (BWRVIP) BWRVIP-27, "BWR Vessel and Internals Project, BWR Standby Liquid Control System/Core Plate delta/P Inspection and Flaw Evaluation Guidelines."

E. Boiling Water Reactor Vessel and Internals Project (BWRVIP) BWRVIP-49, "BWR Vessel and Internals Project, Instrument Penetration Inspection and Flaw Evaluation Guidelines."

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 25 of 205 2.9 Miscellaneous Documents A. Incident Investigation No. 11-B-93-026 Unit #2 Steam Dryer Bracket Inspection.

B. DNE Calculation, Exclusion Criteria for ISI Scope. RIMS R14950829109, R14010222101, and R14020405108. (MD-Q0999-950033) This reference refers to Section Xl activities not covered under the Risk-Informed Program.

C. GE Letter Nos.: BFSE 93-143, BFSE 94-001, BFSE 94-002, BFSE 94-005, and BFSE 94-007.

D. Memorandum from K. L. Groom to F. W. Froscello, dated August 22, 1996, NRC IEB 88-01 IGSCC for Unit 2 and 3 Core Spray Safe-End Replacement Weld, RIMS R92960821851 2.10 TVA Nuclear Standard Programs and Processes A. SPP-2.2, Administration of Site Technical Procedures B. SPP-3.1, Corrective Action Program C. SPP-3.5, Regulatory Reporting Requirements D. SPP-9.1, ASME Section XI E. SPP-9.3, Plant Modifications and Engineering Change Control F. SPP-2.4, Records Management 3.0 PRECAUTIONS AND LIMITATIONS A. Radiation Protection (RADPRO) shall be contacted prior to any work in a radiologically controlled area (RCA). RADPRO shall determine the requirements for a radiological work permit (RWP) and any other radiological requirements.

B. Standard safety practices as outlined in the TVA Health and Safety Manual shall be followed.

C. Efforts should be made to ensure proper planning to reduce delays and radiation exposure during performance of examinations.

D. Any revisions to this instruction initiated by other groups shall be submitted to Program Engineering for concurrence prior to incorporation.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 26 of 205 4.0 PREREQUISITES Personnel responsible for performance of examinations should familiarize themselves with the requirements of this program prior to performing the examinations. Specifics concerning performance of NDE are not a part of this program, but are included in Inspection Services Organization (ISO) Program Manuals, Section 2, Engineering NDE Procedures.

5.0 SPECIAL TOOLS AND EQUIPMENT Equipment is specified in the applicable NDE procedure utilized for performance of the examination.

6.0 ACCEPTANCE STANDARDS Acceptance criteria are specified in the applicable NDE procedures of IEP-1 00, which are in compliance with ASME Section XI, Articles IWA-3000, IWB-3000, IWC-3000, IWD-3000, and IWF-3000 except where ASME Section III or other construction code examinations are employed to satisfy ASME Section Xl requirements.

Acceptance by analytical evaluations performed in accordance with IWX-3000 for Class 1, 2, and 3 components and component supports shall be submitted to the regulatory authority having jurisdiction at the plant site. This information may be submitted with the Inservice Inspection Summary Report including Form NIS-1 or the Owner's Activity Report (OAR-1) or, if deemed necessary, a separate report shall be submitted.

7.0 INSTRUCTION STEPS/ELEMENTS 7.1 Responsibilities 7.1.1 Materials Technology & Codes A. Providing ASME Section Xl interpretations as requested by various site organizations or as required in program development and implementation.

B. Providing assessment and oversight of ISI programs and activities, including review of ISI Program reports and submittals prior to issuance.

C. Review of Requests for Relief prior to issuance.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 27 of 205 7.1.2 Program Engineering Group A. Defining ASME Section Xi Code Class 1, 2, and 3 equivalent boundaries in accordance with applicable guidelines (e.g.: 10CFR50.2, 10CFR50.55a, ASME Section Xl, Regulatory Guide 1.26, and others).

B. Preparing/revising ASME Section Xl Code Class boundary drawings to identify the ASME Section XI Class 1, 2, and 3 equivalent boundaries within each plant system as defined in 7.1.2.A. Reference procedure 0-TI-400. See Section 2.6 for drawing list.

C. Preparing/revising ASME Section XI ISI and RI-ISI drawings that identify the Class 1, 2, and 3 equivalent components (including supports) that require NDE to comply with ASME Section XI requirements. See Section 2.6 for drawing list.

D. Preparing/revising this instruction (ISI Program) in accordance with SPP-2.2, and submitting it to:

1. Site Procedures for approval and issue as a controlled document.
2. Records Management (RM) for subsequent submittal to the ANII.
3. Site Licensing for subsequent submittal to the NRC, as required.

E. Ensuring this program includes the following information as a minimum:

1. The ASME Section XI Code of Record for ISI
2. Inspection interval number and begin/end dates
3. List of ASME Section XI Code Class Boundary drawings
4. List of ASME Section Xl ISI drawings
5. ASME Section Xl Examination Category and Item Number for components.
6. Examination schedule providing quantities for each applicable code item number distributed over each period of the inspection interval
7. NDE method required for each code item number
8. Applicable Requests for Relief
9. Name and address of Owner
10. Name and address of generating plant
11. Name or number designation of the unit
12. Commercial operation date of the unit

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 28 of 205 7.1.2 Program Engineering Group (continued)

13. Description of the system utilized for maintaining record of completed examinations
14. Description of scan plan contents and control
15. Applicable augmented examination requirements and their basis
16. Code Cases proposed for use and the extent of their application.
17. Ensuring the RI-ISI Program is maintained as a Living Program. Review the RI-ISI Program on a basis of periods that coincide with the inspection program requirements contained in Section Xl Inspection Program B.

F. Providing a list of components scheduled for examination during each refueling outage to Inspection Services Organization (ISO) for scan plan development.

This list shall include the component identifier, ASME Section XI examination category and item number, ISI drawing number and sheet number, and examination requirement source.

G. Program Engineering Group will provide prior to each refueling outage, a list of Flow Accelerated Corrosion (FAC) examinations to be taken credit for in the RI-ISI Program from the FAC Engineer to be included in the scan plan.

Agreement of the FAC examinations required for the RI-ISI Program will be documented by a signature from the FAC Engineer on the scan plan.

H. Approving scan plan and revisions and submitting copies of the approved scan plan to site management and the AN II.

1. Determining scope of additional samples and notification of site engineering when an indication(s) results from inservice inspection examinations.

J. Notifying site engineering of indications found during the final additional sample examination to allow evaluation for further actions to be taken.

K. Preparing a Request for Relief (RFR) as required when conformance with Code requirements is impractical (see Section 7.6). ISO responsibilities related to identification of limited examinations are listed in Section 7.1.5.

L. Submitting RFRs to Site Licensing in a timely manner to support ISI activities.

M. Performing NDE in accordance with this instruction.

N. Ensuring that ISI/PSI examinations are performed in accordance with approved TVA or contractor NDE procedures authorized by ISO.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 29 of 205 7.1.2 Program Engineering Group (continued)

0. Administering the Authorized Inspection Agency (AIA) contract and ensuring that services of AIA are utilized when performing Code required activities.

TVA's interface with the Authorized Inspector for ISI, repairs, and replacements is defined in SPP-9.1.

P. Providing AIA representative (ANII) with access to plant facilities and documentation in accordance with IWA-2130 of ASME Section XI.

Q. Notifying ANII prior to performing ASME Section XI examinations.

R. Preparing a Notification of Indication (NOI) to document rejectable indications detected during the performance of ASME Section XI examinations. The NOI process is defined in SPP-9.1.

S. Preparing examination reports and recording them (report number, date, and comments/NOI number) in the scan plan. When inservice examinations are implemented by instructions other than this program, copies of the examination performing organization. These data sheets shall be used as examination reports and incorporated into the scan plan.

T. Ensuring that scan plan examinations are complete prior to completion of an outage.

U. Preparing (or ensuring preparation of) the ISI Summary Report including Form NIS-I. Ensuring that Form NIS-1 is signed by the ANII. Submitting the ISI Summary Report or Owner's Activity Report to Site Licensing in accordance with site schedules, augmented examination summary reports, obtaining ANII signatures, coordinating summary report review with ISO, and submitting augmented examination summary reports to Site Licensing. The reports shall be submitted to Site Licensing at the end of the inservice inspection period.

V. Preparing and submitting the Site Final Report to RM as a QA record.

W. Ensuring records used as PSI records from manufacturers or construction organizations comply with SPP-9.1.

X. Ensuring the calculation of component support acceptance ranges, if required, are prepared in accordance with IEP-100, N-GP-7 and N-VT-I.

Y. Maintaining calibration blocks stored at the plant site.

Z. Initiating a pre-outage meeting to identify augmented examinations in accordance with Section 7.11.

AA. Ownership of the ISI Program, and assignment of an ISI Program Engineer with primary responsibility for ISI activities.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 30 of 205 7.1.3 Site Engineering Design A. Including provisions for inservice inspection access in designs in accordance with ASME Section XI, IWA-1400(b) and IWA-1 500.

B. Performing engineering evaluations in support of examination indications related to operability and corrective measures.

C. Performing evaluations of rejectable indications found during final additional sample examinations to determine if further action is required.

D. Determining those component supports that could be affected by observed failure modes and could affect nonexempt components.

E. Providing specific written details for augmented requirements (Refer to Section 7.11) and determining if a post examination meeting is required.

7.1.4 Site Licensing A. Filing this instruction, and/or including revisions, with the NRC in accordance with IWA-1400(c).

B. Submitting Requests for Relief (RFR) and the ISI Summary Report or Owner's Activity Report including Form NIS-1 and IW(X)-3600 analytical evaluations to the NRC.

7.1.5 Inspection Services Organization (ISO)

A. Developing and maintaining a computerized data base, at the direction of Program Engineering Group, to include components identified on the ISI weld and support drawings.

B. Preparing/revising scan plans for each refueling outage of the inspection interval, as directed by Program Engineering Group, utilizing the computerized data base. This includes providing additional information provided by NDE Level III personnel to complete the scan plan, such as NDE procedure references, calibration standard references, and UT scanning angles.

C. Providing NDE Level III approval of scan plan revisions that affect the additional information of Section 7.1.5B. and maintaining a scan plan revision history log.

D. Providing NDE Level III determination if a Request for Relief (RFR) is required because of areas that are inaccessible or partially inaccessible for examination or because it is determined that conformance with Code requirements is impractical and notifying Program Engineering Group. Reference Paragraph 7.2.3E.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 3 of 205 7.1.5 Inspection Services Organization (ISO) (continued)

E. Approving contractor NDE procedures (using IEP-100 as a guideline),

contractor written practices for qualification and certification of NDE personnel, certifications of contractor's NDE personnel and equipment performing ISI/PSI examinations.

F. Providing NDE Level III evaluation of successive examination results.

G. Packaging radiographs for storage and providing them with reader sheets as a life of plant record to RM.

H. Providing copies of IEP-100 NDE procedure revisions and evidence of personnel qualifications to RM as permanent records for the service lifetime of the plant in accordance with IWA-1400(k).

1. Maintaining as-built calibration standard drawings and the calibration standard material certifications.

7.1.6 Site Records Management (RM)

A. Issuing controlled copies of ASME Section XI Code Class Boundary Drawings and ISI drawings.

B. Issuing this instruction and providing controlled copies to Program Engineering Group, ANI/ANII, and other requesting organizations.

C. Maintaining the site final report as a life of plant QA document. Other records referenced in the final report (work plans, radiographs, etc.), NDE procedure revisions, and evidence of personnel qualifications shall be retained for the service lifetime of the plant.

7.1.7 Authorized Nuclear Inservice Inspector (ANII)

A. Performing the duties of IWA-21 10, including a detailed review of this instruction and subsequent revisions. He shall submit a report of the review to the Owner in accordance with IWA-21 10(a)(3).

b. Having the prerogative and authorization to require requalification of an operator or procedure when he has reason to believe code requirements are not being met.

7.1.8 Nuclear Assurance A. Ensuring the adequacy of contractor's QA programs in accordance with the TVA Standard Programs and Processes.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 32 of 205 7.2 Implementation 7.2.1 System for Maintaining Status of Examinations A. ISI Data Base A computerized data base shall be utilized for identification of the components requiring examination and for maintaining the status of completed examinations for ASME Section XI and/or augmented credit. Maintenance and updating of this data base is detailed in Section 7.9.

B. Scan Plan

1. The scan plan is developed from the ISI data base and details the examinations scheduled for performance during an outage. A scan plan may also be used for examinations performed pre-outage or between outages. It should contain as a minimum: components to be examined; Code examination category; Code item number; methods of examination; NDE procedure reference; calibration standard reference; ISI drawing and sheet number.
2. Prior to performing examinations, the scan plan shall be approved by Program Engineering Group.
3. When inservice examinations are performed as a result of instructions other than this program (e.g., maintenance instructions, work plans, etc.),

copies of the examination data sheets shall be submitted to Program Engineering Group by the performing organization for assignment of a report number and incorporation into the scan plan.

4. During implementation, it may become necessary to revise the scan plan.

Scan plan revisions may be initiated by Program Engineering Group, ISO, or by other personnel involved with implementation of the scan plan. All changes shall be coordinated with Program Engineering Group and, as needed, with the appropriate plant planning and scheduling personnel for facilitating the use of supporting craft personnel.

5. Revisions to the scan plan shall be controlled in the same manner as the original. ISO shall maintain a scan plan revision history log. Interim working copies may be handwritten to allow examinations to be performed before a formal revision is issued. These changes shall be approved by Program Engineering Group and a NDE Level Ill, as required by Section 7.1.2H or Section 7.1.5C. Approving individuals shall initial and date such changes.

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C. Configuration Changes

1. When major portions of existing pipe or supports are replaced or new systems are added, a system walkdown should be performed under the direction of Program Engineering Group to identify the pipe configuration, welds, components, and supports that are required to be in the inspection program.
2. If variations in configuration are discovered or modifications (including additions or deletions), repairs or replacements are made during the service lifetime of the unit, the changes shall be marked on field corrected copies of the appropriate drawing listed in Section 2.6 by a Program Engineering Group representative. The field corrected copies shall be used in the performance of examinations and as records until the drawing has been revised to reflect the change(s).
3. Program Engineering Group shall be responsible for reviewing the proposed change, revising the drawings as necessary, and ensuring the revised drawings are issued prior to the next refueling outage. The scan plan shall be revised to reflect any PSI examinations performed due to the variations in configuration. The ISI Engineer (or his designee) shall track the ISI field drawing revisions by utilization of a log book. This log book will utilize the assigned RIMS (Record of Information Management System) number on the NEDP-3 Form, "Request For Administrative Change To Drawings, " as the tracking number for the ISI field drawing revisions.

Guidelines for preparation and control of ISI examination drawings are delineated in SPP-9.1 and NEDP-3.

7.2.2 Notification of Indication (NOI)

A. NOI form, FORM SPP-9.1-2 of SPP-9.1, shall be used to document indication(s) exceeding the acceptance criteria of Article IWX-3000 of ASME Section XI. If engineering evaluation determines that the condition is unacceptable for continued service, corrective action shall be initiated Program Engineering Group shall provide/coordinate dispositions for NOI's in accordance with SPP-9.1 and SPP-3.1. Any Problem Evaluation Reports (PERs) or Work Orders (WOs) generated to support the NOI disposition should be referenced on the NOI.

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B. Additional Examinations

1. Additional examinations for Class 1 equivalent components (IWB) shall be in accordance with the requirements of IWB-2430.

The additional examination samples are defined as those items (welds, areas, or parts) in a particular examination category and item number.

Engineering judgment should be documented concerning expansion (or no expansion) into additional systems. The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.

a. Examinations of the initial sample that reveal indications exceeding the acceptance standards of Table IWB-3410-1 shall be extended to include additional examinations in the same outage as the initial examinations.

The first additional examination sample shall include items scheduled for this and the subsequent period. If examinations for that item are not scheduled in the subsequent period, the most immediate period containing scheduled examinations of that item shall be examined.

b. If the first additional examinations of (1)(a) detect indications exceeding the acceptance standards of Table IWB-341 0-1, further additional examinations shall be performed during the outage. The second additional examination sample shall include the remaining items of similar design, size, and function.
2. Additional examinations for Class 2 equivalent components (IWC) shall be selected in accordance with IWC-2430.

The additional examination samples are defined as those items (welds, areas, or parts) in a particular examination category and item number.

Engineering judgment should be documented concerning expansion (or no expansion) into additional systems. The initial sample is the sample scheduled for examination at a particular outage for ASME Section XI credit.

a. Examinations of the initial sample that reveal indications exceeding the acceptance standards of IWC 3000 shall be extended to include additional examinations in the same outage as the initial examinations. The first additional sample shall include approximately the same number of items examined in the initial sample. The items selected should be those available in the interval sample that have the longest service time from its previous inservice examination.

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b. Ifthe first additional examinations of (2)(a) detect indications exceeding the acceptance standards of Table IWC 3410 1, further additional examinations shall be performed during the outage. The second additional examination sample shall include the remaining items of similar design, size, and function.
3. Additional examinations for component supports (IWF) shall be in accordance with section IWF-2430.
a. If component supports in the initial sample must be subjected to corrective measures in accordance with IWF 3000, the component supports immediately adjacent to those for which corrective action is required shall be examined. Also, the examinations shall be extended to include a first additional sample that includes supports within the system, equal in number and of the same type and function as those scheduled for examination during the period.
b. When the additional examinations of (3)(a) require corrective measures in accordance with IWF 3000, a second additional sample of the remaining component supports within the system of the same type and function as in (3)(a) shall be examined.
c. When the additional examinations of (3)(b) require corrective measures in accordance with IWF 3000, examinations shall be extended to include a third additional sample of the remaining nonexempt supports potentially subject to the same failure modes that required corrective measures in (3)(a) and (3)(b). These additional examinations shall include nonexempt component supports in other systems when support failures requiring corrective measures indicate non system related failure modes. At the request of Program Engineering Group, Site Engineering shall make the determination of failure mode applicability and select the third additional sample.
d. When the additional examinations of (3)(c) require corrective measures in accordance with IWF-3000, examination shall be extended to those exempt component supports that could be affected by the same observed failure modes and could affect nonexempt components. At the request of Program Engineering Group, Site Engineering shall make the determination of failure mode applicability and select a fourth additional sample of exempt component supports that could affect nonexempt components.

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4. Ifthe final sample examinations in (3)(d) above detect indications exceeding the acceptance standards of Article IWX 3000 of ASME Section XI, Program Engineering Group shall notify Site Engineering to evaluate the indications and make recommendations(s) for further action, if needed.

These actions would be beyond those required by ASME Section XI.

7.2.3 Examinations A. NDE shall be performed in accordance with IWA-2200 of ASME Section XI utilizing the NDE procedures in Inspection Services Organization (ISO)

Program Manuals, Section 2, Engineering NDE Procedures, or approved contractor procedures, with the exception of NDE procedures for ultrasonic examination shall be qualified to the requirements of Appendix VIII of ASME Section XI as implemented by the Performance Demonstration Initiative Program (PDI).

B. Personnel performing NDE operations shall be qualified and certified in accordance with IWA-2300 of ASME Section Xl as specified in IEP-200 and qualified to the requirements of the 1995 Edition of ANSI/ASNT CP-189 with the exception of NDE personnel performing ultrasonic examinations shall be qualified to the requirements of Appendix VIII of ASME Section XI as implemented by the Performance Demonstration Initiative Program (PDI).

C. The inservice examinations may be performed by Program Engineering Group, ISO, or contractor personnel. Contract preparation, administration, and supervision shall be the responsibility of Program Engineering Group.

Inspection plans and/or quality assurance programs submitted by contractors shall be reviewed and approved by Nuclear Assurance prior to use. All contractor NDE procedures used during the inspection program shall be reviewed and approved by ISO using IEP-100 as a guideline.

D. A weld reference system shall be established for welds and areas subject to surface or volumetric examination in accordance with IWA-2600.

E. Every attempt shall be made to provide 100% code coverage (volume or area) when performing an exam. When 100% overage is not obtained/obtainable, a NDE Level III shall promptly notify Program Engineering Group.

If the coverage is limited due to an obstruction which is removable, an evaluation shall be performed by Program Engineering Group to either allow removal of the obstruction or justify why the obstruction cannot be removed.

When less than the required ASME Section XI code examination volume or area is examined, the percentage examined shall be documented on the examination data sheet.

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The cause of the limitation shall be clearly specified as a part of the data sheet documentation. An NDE Level III representative shall review the limitation or impractical examinations during the refueling outage and determine if a code examination was achieved. If greater than 90 % code coverage was not achieved, the NDE Level III representative shall notify Program Engineering Group immediately to determine if an alternative component can be selected. If an alternate component cannot be selected the examination volume or area is qualified for request for relief action in accordance with Section 7.6.

7.3 Components Subject to Examination 7.3.1 ASME Class I Equivalent Components Subject to Examination (IWB)

A. ASME Class 1 equivalent systems are listed below:

Control Rod Drive Hydraulic System (CRD)

Core Spray System (CS)

Feedwater System (FW)

High Pressure Coolant Injection System (HPCI)

Main Steam System (MS)

Reactor Core Isolation Cooling System (RCIC)

Recirculation System (RECIR)

Residual Heat Removal (RHR)

Reactor Pressure Vessel (RPV)

Standby Liquid Control System (SLC)

B. The specific components subject to examination are identified on ISI drawings listed in Section 2.6. Attachment 3 contains detailed information for selected Class 1 valves. The number of components within each system, the number selected for examination during the interval and the number selected for examination by period are provided in Attachment 1, Section 8.1 Examination Schedule - Class 1 Equivalent (IWB) Components.

C. Selection and scheduling of ASME Class 1 equivalent components is in accordance with IWB-2412, Inspection Program B, IWB-1200 exemptions, and applicable requirements of Table IWB-2500-1.

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D. The extent and examination for Category B-F and B-J welds are in accordance with the RI-ISI examination requirements outlined in Section 7.14 of this Program E. The examination of Class 1 equivalent component supports is in accordance with Section 7.3.4.

F. ASME Class 1 equivalent piping and components that meet the size exemption listed in reference 2.9B are exempt from the Class 1 requirements of 10CFR50.55a(c)(1). The exempted piping /components would then be considered ASME Class 2 equivalent since they are part of the reactor pressure coolant boundary, but excluded from classification as ASME Class 1. This reference refers to ASME Section XI activities not covered under the Risk - Informed ISI Program.

G. The extent of examination for Category B-D, Item No. B3.90, "Pressure Retaining Nozzle-To-Vessel Welds," shall be (next to the widest part of the weld) one-half (1/2) inch from each side of the weld crown in lieu of one-half (1/2) through-wall thickness from each side of the weld required by the 2004 Edition of ASME Section Xl, Code, Table IWB-2500-1, Figures 7 (a) and (b).

Reference Code Case N-613-1.

H. The 2004 Edition of ASME Section XI, Table IWB-2500-1, Examination Category B-D, Item B3.100, requires a volumetric examination of the reactor pressure vessel head nozzles inside radius section. In accordance with Code Case N-648-1, TVA will perform an enhanced visual (VT-1) examination capable of a 1-mil resolution of the reactor pressure vessel and reactor pressure vessel head nozzles inside radius sections, in accordance with ASME Section XI, VT-1 requirements, This does not apply to the six (N4) Feedwater nozzles. The six Feedwater nozzle inner radius sections will continue to be examined with ultrasonic techniques developed and qualified using GE-NE523-A71-0594-A, Revision 01 and ASME Section XI Code requirements.

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1. Reactor Vessel Interior ASME Code Category B-N-I, Item Number B13.10 The space above and below the vessel core that is made accessible by the removal of components during normal refueling outages shall be visually examined, VT-3, during the first refueling outage and at subsequent refueling outages at approximately three year intervals (a minimum of once an inspection period). Reference Attachment 1 Parts 7 and 84.

The areas that are normally accessible include main steam nozzles, feedwater nozzles and spargers, core spray nozzles, piping, and spargers, top guide assembly, instrumentation nozzles, CRD return nozzle, and the RPV annulus area.

ASME Code Category B-N-2, Item Number B13.20 The accessible RPV attachment welds within the beltline region shall be visually examined (VT-1). The attachment welds within the beltline region are defined as the lower surveillance specimen bracket welds on shell course 2 and the jet pump riser brace pad welds. Reference Attachment 1 Parts 7 and 8.

ASME Code Category B-N-2, Item Number B13.30 The accessible RPV attachment welds outside the beltline region shall be visually examined (VT-3). The attachment welds normally accessible outside the beltline region are defined as the guide rod brackets, steam dryer support brackets, feedwater sparger brackets, core spray piping brackets and pads, RPV shroud support to RPV bottom head, and the surveillance specimen bracket welds on shell course 3. The shroud support leg to bottom RPV head welds are located under the core plate and are normally inaccessible.

Reference Attachment 1 Parts 7 and 8.

ASME Code Category B-N-2, Item Number B13.40 The accessible surfaces of the core support structure shall be visually examined (VT-3). The core support structure is defined as the top guide, core plate, control rod guide tubes, control rod drive housings, shroud support ledges, and the fuel support castings. Areas that are accessible during normal refueling outages include the top surface of the top guide and the outer peripheral top surface of the core plate. Reference Attachment 1 Parts 7 and 8.

All augmented examination requirements and commitments for BFN vessel internal examinations during the ISI interval are stated in O-TI-365. Reference Attachment 1 Parts 7 and 8.

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1) When visual examinations of RPV internal components are being performed for compliance with 0-TI-365 requirements and maintenance or refueling activities, credit may be taken for ASME Section XI requirements provided the visual examination meets the minimum requirements of ASME Section Xl.
2) When specialized visual examinations are being performed, for compliance with 0-TI-365 requirements and maintenance or refueling activities, and access to areas are made available that are normally inaccessible, credit for ASME Section XI maybe taken provided the visual examination meets the minimum requirements of ASME Section XI. These examinations shall be considered supplemental examinations.
3) All visual examinations performed on RPV internal components that meet the minimum VT-3 criteria, as stipulated in ASME Section XI, shall be considered for ASME Section XI, Code Category B-N-1 credit.
4) It is permissible to defer the visual examinations for ASME Section XI, Code Category B-N-2 to the end of the inspection interval. However, these examinations maybe performed at any time during the interval.

7.3.2 ASME Class 2 Equivalent Components Subject to Examination (IWC)

A. ASME Class 2 equivalent systems are listed below:

Control Rod Drive Hydraulic System (CRD)

Core Spray System (CS)

High Pressure Coolant Injection System (HPCI)

Main Steam (MS)

Reactor Building Closed Cooling Water System (RBCCW)

Closed Cooling Water System (RBCCW)

Reactor Core Isolation Cooling System (RCIC)

Residual Heat Removal System (RHR)

B. The specific components subject to examination are identified on ISI drawings listed in Section 2.6. The number of components within each system, the number selected for examination during the interval and the number selected for examination by period are provided in Attachment 1, Section 8.1 Examination Schedule - Class 2 Equivalent (IWC) Components.

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C. Selection and scheduling of ASME Class 2 equivalent components is in accordance with IWC-2412, Inspection Program B, IWC-1200 exemptions, and applicable requirements of Table IWC-2500-1.

D. Class 2 Piping Welds (C-F-1 and C-F-2) shall be in accordance with the RI-ISI examination requirements outlined in Section 7.14 of this Program.

E. The examination of Class 2 equivalent component supports is in accordance with Section 7.3.4.

7.3.3 ASME Class 3 Equivalent Components Subject to Examination (IWD) and Non-Code Class Components A. ASME Class 3 equivalent systems are listed below:

Emergency Equipment Cooling Water System (EECW)

Fuel Pool Cooling System (FPC)

Residual Heat Removal Service Water System (RHRSW)

The specific components subject to examination are identified on ISI drawings listed in Section 2.6. The number of components within each system, the number selected for examination during the interval and the number selected for examination by period are provided in Attachment 1, Section 8.1 Examination Schedule - Class 3 Equivalent (IWD) Components.

B. Selection and scheduling of ASME Class 3 equivalent components is in accordance with IWD-2412, Inspection Program B, IWD-1200 exemptions, and applicable requirements of Table IWD-2500-1.

C. The examination of Class 3 equivalent component supports is in accordance with Section 7.3.4.

7.3.4 Component Supports Subject to Examination (IWF).

A. ASME Class 1, 2, 3, and MC equivalent component and piping supports shall be examined in accordance with IWF-1 000.

B. The specific components subject to examination are identified on ISI drawings listed in Section 2.6. The number of supports within each system, the number selected for examination during the interval and the number selected for examination by period are provided in Attachment 1, Section 8.1 Examination Schedule, Part 4 - Component Supports (IWF).

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C. Selection and scheduling of component supports is in accordance with Table IWF-2410-2, Inspection Program B, IWF-1200 exemptions, and applicable requirements of Table IWF-2500-1.

D. Supports depicted as snubbers on the ISI support drawings are subject to examination outside the boundaries of the snubber (including the pins) in accordance with IWF-5300(c). The examination of snubbers (excluding the pins) and any repair, replacement or adjustment to the snubber itself is addressed by the Snubber Inservice Testing Program (see Section 7.8) and RFR 2-ISI-40.

E. The acceptance range for constant force and variable springs shall be in accordance with the support drawing. If the setting range is NOT identified on the drawing, the applicable general notes contained in the 47B435 series of drawings shall be utilized in accordance with N-VT-1 and N-GP-7.

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F. Component supports that have been adjusted in accordance with IWF-3000, repaired, or replaced shall be examined prior to return to service per the applicable examinations listed in Table IWF -2500-1. Additionally, for systems that operate above 200 degrees F during normal operation, an additional preservice examination shall be performed on the affected component supports during or following the subsequent system heat up and cool down cycle unless determined unnecessary by evaluation. This examination shall be performed during operation or at the next refueling outage. Component supports requiring an additional preservice examination shall be scheduled for examination and added to the applicable scan plan.

7.3.5 Successive Examinations, Class 1, 2, 3, or Component Supports Any corrective actions required as a result of ISI examinations shall be handled in accordance with SPP-3.1.

Successive examinations shall be performed in accordance with the requirements of IWB-2420, IWC-2420, IWD-2420, and IWF-2420.

A. Successive Examinations - Class 1 Equivalent Components Areas containing flaw indications or relevant conditions evaluated in accordance with IWB-3132.3 or IWB-3142.4 that qualify for continued service shall be re-examined during the next three inspection periods as listed in the inspection schedules. Ifthese re-examinations reveal that the flaw indications remain essentially unchanged for three successive inspections, then the component examination frequency may revert to the original schedule.

Components requiring successive examinations shall be scheduled for examination and added to the applicable scan plan. If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of IWB-341 0-1, successive examinations shall be performed, if determined necessary, based on an evaluation.

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B. Successive Examinations - Class 2 Equivalent Components Areas containing flaw indications or relevant conditions evaluated in accordance with IWC-3122.4 or IWC-3132.3 that qualify for continued service shall be re-examined during the next inspection period as listed in the inspection schedule. If this re-examination reveals that the flaw indications remain essentially unchanged, then the component examination frequency may revert to the original schedule.

Components requiring successive examinations shall be scheduled for examination and added to the applicable scan plan. If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of IWB-3410-1, successive examinations shall be performed, if determined necessary, based on an evaluation.

C. Successive Examinations - Class 3 Equivalent Components Areas containing flaw indications or relevant conditions evaluated in accordance with IWD-3000 that qualify for continued service shall be re-examined during the next inspection period as listed in the inspection schedule. If this re-examination reveals that the flaw indications remain essentially unchanged, then the component examination frequency may revert to the original schedule.

Components requiring successive examinations shall be scheduled for examination and added to the applicable scan plan.

If welded attachments are examined as a result of identified component support deformation and the results of these examinations exceed the acceptance standards of IWD-3000, successive examinations shall be performed, if determined necessary, based on an evaluation.

D. Successive Examinations for Class 1, 2, and 3 Component Supports (IWF)

Successive examinations for component supports (IWF) shall be determined in accordance with IWF-2420. (See Section 7.3.1E for component supports requiring an additional preservice examination).

When a component support must be subjected to corrective measures in accordance with IWF-3000 that support shall be re-examined during the next inspection period listed in the inspection schedule. If this re-examination does not require additional corrective measures, then the examination frequency may revert to the original schedule. Components requiring successive examinations shall be scheduled for examination and added to the applicable scan plan.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 45 of 205 7.4 Calibration Standards Calibration standards are included in ASME Section Xl, Appendix I. This appendix includes references to ASME Section XI, Appendix III and ASME Section V for additional requirements. As-built calibration standard drawings and calibration standard material certifications are maintained by ISO. The calibration blocks are stored at the plant site and maintained by ISI personnel.

7.5 Records and Reports Records and reports shall be prepared in accordance with ASME Code Case N-532-4 or ASME Section XI, Subarticle IWA 1400, Article IWA 6000.

7.5.1 ISI Summary Report Prepare an Owner's Activity Report in accordance with 7.5.1 .A or a Summary Report in accordance with 7.5.1 .B A. An ISI Owner's Activity Report in accordance with ASME Code Case N-532-4 for Class 1 and 2 (equivalent) Components shall be prepared and submitted to Site Licensing and other review organizations on a schedule that permits submittal to the NRC within 90 days after turbine generator synchronization following a refueling outage.

Information related to the Containment Inservice Inspection Program inspection of Class MC (equivalent) components (IWE) shall also be included in the ISI Owner's Activity Report as applicable. This information is compiled in accordance with 0-TI-376 for inclusion in the ISI Owner's Activity Report (see Section 7.5.1 B.15).

B. An ISI summary report for Class 1 and 2 (equivalent) Components shall be prepared and submitted to Site Licensing and other review organizations on a schedule that permits submittal to the NRC within 90 days after turbine generator synchronization following a refueling outage. Examinations, tests, replacements, and repairs conducted since the preceding summary report shall be included.

Information related to the Containment Inservice Inspection Program inspection of Class MC (equivalent) components (IWE) shall also be included in the ISI summary report as applicable. This information is compiled in accordance with 0-TI-376 for inclusion in the ISI summary report (see Section 7.5.1B.15).

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.7.5.1 ISI Summary Report (continued)

1. Cover Sheet A cover sheet stating "ASME Section XI Inservice Inspection Summary Report for Browns Ferry Nuclear Plant, Unit 2, " and the Refueling Outage.

The cover sheet shall also provide:

a. Date of document completion.
b. Name and address of owner.
c. Name and address of generating plant.
d. Name or number assigned to the nuclear power unit by TVA.
e. Commercial operation date for the unit.
2. Table of Contents A table of contents for the report should follow the title page.
3. Form NIS-1 The Owner's Report for Inservice Inspections, Form NIS-1, as shown in Appendix II of ASME Section XI shall be completed and included.
4. Form NIS-2 The Owner's Report for Repair and Replacement Activity, Form NIS-2, as shown in Appendix IIof ASME Section XI shall be completed and included.
5. Introduction and Summary of the Inspection The introduction should include the following information: Plant, unit number, preservice or inservice examinations, RFO cycle, systems, components, and vessels examined, organizations examinations were performed by, dates examinations were performed, ASME Section Code of Record. The summary should include a brief description of the overall inspection. Included as part of the summary, ASME Class 1, 2, and 3 equivalent components and the integrally welded attachments whose examination results required evaluation analysis (IWB-3132.3 and 3142.4 for Class 1 and 3; and IWC-3122.3 and 3132.3 for Class 2) shall be submitted to the NRC as required by IWB-3134 and 3144 and IWC-3125 and 3134.

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6. Examination Summary The examination summary shall tabulate the ASME Section Xl examinations credited for the applicable period. Items should include the following information: category, total number of examinations required for the inspection interval, total number required for the applicable period, total number credited for the applicable period, and exclusions, exceptions, or deferrals.
7. Examination Plan The Examination Plan shall give a detailed description of all areas subject to examination during the inspection. It should contain the following information: examination area, Code Category and Item Number, reference drawing, examination method, examination procedure, examination report number, calibration block, date of examination, and examination results. This plan may be submitted as the computerized Outage Report.
8. Component Re-Examination Reports The component re-examination section shall give a detailed description of all components subject to re-examination due to rework, repair, or replacement resulting from a Notification of Indication (NOI). This section should contain the examination area, Code Category and Item Number, reference drawing, examination method, examination procedure, examination report number, calibration block, date of examination, and examination results.
9. Summary of Notifications of Indications (NOls)

The summary of NOls shall give a short summary of each NOI report along with the indication discrepancy. It should also contain the final disposition including a reference to the corrective action taken.

10. Additional Sample The additional sample section, if applicable, shall indicate additional sample examinations performed as a result of a failed component. The summary should include reference to the applicable system, the affected component, the number of components examined as a result of the failure, and a description of additional samples and results of the additional sample examinations.

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11. Successive Examinations The successive examination section, if applicable, shall indicate examinations performed as a result of ASME Section XI requirements.

This section should contain a reference to the applicable system, the affected components, and the results of the successive examinations.

12. Analytical Evaluation The analytical evaluation section for ASME Class 1, 2, and 3 equivalent components and the integrally welded attachments whose examination results require evaluation analysis, if applicable, shall include a short summary of each analytical evaluation, the indication discrepancy, and its location.

A copy of each analytical evaluation should be included, with a reference to the applicable NOI and the component identifier.

13. Augmented Examinations As applicable, a brief summary of the augmented examinations reportable to the NRC shall be included.
14. Requests for Relief The summary of requests for relief shall give a short summary of each Request for Relief resulting from the inspection. This section shall summarize any components that did NOT receive the required examination coverage. The results should indicate the applicable component, Code Class, Code Category, Code Item Number, examination method, and calculated examination coverage. In addition, a description should summarize the access limitations and applicable reason why examination coverage cannot be obtained.
15. Containment Inservice Inspection Program (IWE)

This section, if applicable, should contain evaluations performed in accordance with the requirements of 10CFR 50 55a(b)(2)(x)(A), evaluation of inaccessible areas, and 10CFR50.55a(b)(2)(x)(D), evaluation for additional examinations, as delineated in 0-TI-376.

7.5.2 Site Final Report A site final report shall be prepared following each refueling outage and submitted to Records Management for retention as a permanent record. The site final report should contain, but NOT be limited to, the following:

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 49 of 205 7.5.2 Site Final Report (continued)

  • An index to record file
  • The inservice and preservice NDE examination reports and calibration data sheets
  • The ISI Summary Report or Owner's Activity Report with appendices prepared per Section 7.5.1
  • Personnel certifications

" Reference to NDE procedures

  • Reference to NDE examination records including radiographs and review forms
  • Notification of Indication (NOI) Reports

" Scan plans and scan plan revision logs (if applicable)

  • Containment Inservice Inspection Report prepared in accordance with 0-TI-376 7.5.3 Radiographs Radiographs shall be packaged by ISO and transmitted to RM for storage as a life of plant record.

7.6 Requests for Relief (RFR)

Impractical code requirements or examinations shall be submitted to NRC as written Requests for Relief in accordance with 10CFR50.55a(g)(5). Proposed alternate examinations and information to support the basis and justification for relief shall be included. Requests for Relief are identified in Section 8.5 of this program and listed in Attachment 1, Section 8.1 next to the applicable examination category.

ISO is responsible for notifying Program Engineering Group of impractical examination requirements and limitations that are encountered during performance of examinations. Reference Paragraph 7.1.5D and 7.5.11B.14.

RFRs shall be prepared in accordance with SPP-9.1. Materials Technology and Codes will be provided an opportunity to review RFRs.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 50 of 205 7.7 Repairs and Replacements ASME Section Xl repair and replacement activities are performed in accordance with SPP-9.1 and 0-TI-363. Preservice examinations required for ASME Code Class 1, 2, and 3 (equivalent) repaired/replaced components are in accordance with the code of record specified in this surveillance instruction. The examination categories and NDE method for preservice examinations may be determined from those listed in Attachment 1, Section 8.1.

7.8 ASME Section Xl Programs Not Addressed By 2-SI-4.6.G 7.8.1 System Pressure Tests The system pressure test program is identified in SPP-9.1. Additional details are provided in 0-TI-364.

7.8.2 Pump and Valve Inservice Testing The pump and valve inservice testing program is identified in SPP-9.1 and 0-TI-362.

7.8.3 Snubber Inservice Testing Snubber inservice examination and testing is in accordance with 2-SI-4.6.H-1.

7.8.4 Containment Inservice Inspection The containment inservice inspection program is identified in SPP-9.1 and 0-TI-376.

7.9 ISI Data Base Update and Maintenance A. Program Engineering Group is responsible for maintaining the ISI Data Base.

ISO may perform update functions at the direction of Program Engineering Group.

B. Changes to the ISI Data Base may become necessary for a number of reasons, such as: maintenance activities requiring Code examinations; repair/replacement activities; design changes adding or deleting components; implementation of Code Cases or requests for relief; or changes in planned examination scope due to additional or supplemental examinations.

C. All changes or updates shall be authorized by the ISI Program Engineer prior to entry into the ISI Data Base.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 51 of 205 D. Upon completion of examinations for a given operating cycle, the ISI Data Base shall be updated to reflect the actual status of completed examinations. This should be done in a timely manner following the refueling outage (within 6 months as a guide) to ensure data base integrity. This update should be based on the completed NDE examination reports.

E. Scan plan revisions shall include a sign-off that the ISI Data Base has been updated as part of the revision approval cycle.

F. The ISI Program Engineer is responsible for ensuring that the ISI Data Base is updated in conjunction with ISI Program Plan revisions for items such as design changes, adopted Code Cases, and requests for relief.

7.10 Corrective Action Any corrective action required as a result of ISI examinations shall be documented in accordance with SPP-3.1, Corrective Action Program.

7.11 Augmented Examinations Augmented examinations are performed in addition to ASME Section Xl Code requirements. The augmented examinations may be required by the NRC or be self-imposed by TVA. Typical sources include generic letters, IE Bulletins, technical specifications, vendor recommendations, BWRVIP documents, and industry experience.

The responsible organization or owner shall have technical and administrative responsibility for each augmented examination identified in this section. This responsibility shall include scheduling any examinations through Program Engineering Group, tracking the status of examinations, and reporting completed examinations. Responsible organizations requesting inclusion of augmented examinations in this section shall submit a written request to the ISI Program Engineer. The written request shall include specific details such as requirement source, identification of components requiring examination, examination frequency, examination method, examination area/volume, acceptance criteria, types of flaws anticipated, areas of high suspect, probability of failure, and reporting requirements.

Copies of the written request shall be submitted to ISO and Program Engineering Group to facilitate nondestructive examination procedure preparation, establishment of training programs, and personnel familiarization.

Prior to each refueling outage, a meeting shall be initiated by the ISI Program Engineer. Meeting attendees shall include the responsible organizations, System Engineering, and ISO. The meeting agenda should include examination plans and schedules, updates on industry experience, and any additional pertinent information.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 52 of 205 Following the completion of the augmented examination, Program Engineering Group shall report to the responsible organization items such as examination results and changes in results from previous examinations. The responsible organization shall determine if a meeting with the Program Engineering Group and/or other appropriate organizations is necessary to discuss items such as additional examinations to be conducted during the current outage, trends, lessons learned, and identify any future actions such as changes in the frequency of examination.

SIL's and clarification letters listed in this Augmented Examination Section provide GE's recommendation for reactor internals inspection. The actual scope and criteria for reactor internals inspections will be reviewed and approved by TVA Site Engineering prior to each refueling outage. Any indications found during inspections will receive a review and will be dispositioned by TVA Site Engineering. The responsible organization shall report augmented examination results to the NRC as required by the document initiating the examination.

7.11.1 Weld DSRHR-2-05A Responsible organization: Site Engineering Weld DSRHR-2-05A has an indication that was determined to be lack of fusion between layers of welding. It shall receive augmented RT and UT examinations each inspection period to monitor the size of the indication. Evaluation of the indication shall be performed by an ISO NDE level III by comparison to previous examinations. Ifthere is any change, Site Engineering shall be formally requested to provide additional evaluation. A report of the examinations shall be forwarded to the NRC with the NUREG-0313 report of Section 7.11.7. [NRC/C] Reference NRC Inspection Report 86-03, Open Item 86-03-03 (RIMS L29 860925 984)].

ISI Data Base Exam Requirement Source: DOI-02.

The weld DSRHR-2-05A received an augmented UT and RT examination for three consecutive periods (cycles 6, 9, and 11 refueling outages) these examinations were reviewed and evaluated by an ISO NDE Level III and BFN Site Engineering and revealed no changes to the size of the aforementioned indication. This was in accordance with TVA Commitment Letter, dated September 30, 1986 (RIMS# L29 860925 984). This commitment has been fulfilled and is therefore no longer required. Reference letter from Engineering dated April 28, 2004 (RIMS# R06 040428 934).

7.11.2 HPCI Pump Discharge Support Inspection Following Injection NRC commitment NCO850144002 for the augmented examination of the supports on the HPCI discharge line following an injection was revised on March 7, 1995.

The revised commitment requires the examination of the HPCI discharge line supports in accordance with the normal ASME Section XI requirements. The revised commitment number is NCO950027001.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 53 of 205 7.11.3 CRD Return Line Reroute (continued) 7.11.3 CRD Return Line Reroute Responsible organization: Site Engineering The augmented examination requirements of the CRD return line reroute are contained in NUREG 0619. The welded connections joining the rerouted CRD return line to the reactor water cleanup system was ultrasonically examined during the Cycle 4, Cycle 5, Cycle 6, and Cycle 10 refueling outages.

The NRC commitment to NUREG-0619 to perform ultrasonic examination of the Thermal Mixing Tee base material, welds RCRD-2-44, RCRD-2-45, and RCRDS-2-03 has been fulfilled and is no longer required in accordance with procedure SPP-3.3. Reference Commitment Item Number NC0810101003 and NCO 810101004.

Reference memorandum, Commitment Evaluation Form from TVA to NRC dated October 24, 2000, RIMS # R08 001128 743. ISI Data Base Exam Requirement Source: B01-02 7.11.4 Feedwater Nozzles Responsible organization: Site Engineering The augmented examination requirements for the feedwater nozzles and spargers is contained in NUREG-0619 and BWR Owners Group (BWROG) Licensing Topical Report GE-NE-523-A71-0594, Revision 1, August 1999, Table 6-1. Beginning with cycle 13, an ultrasonic examination of all of the feedwater nozzle bores, and inside blend radii are required every fifth (cycle 18) refueling outage. Reference Section 6.3. The alternate examination requirements, contained in Table 6-1 of GE-NE-523-A71-0594, Revision 1, eliminate the need for liquid penetrant examinations. The feedwater spargers are internal to the reactor pressure vessel and shall be visually examined every fourth refueling outage in accordance with Table 6-1 of the above licensing topical report and 0-TI-365.

Reporting is required within 6 months after the outage when an inspection was performed. The report of these examinations shall be included with the ISI Summary Report unless a special report is deemed necessary by Program Engineering. REFER TO NUREG-0619, Section 4.4.3 for information to be included. Reference TVA submittal dated October 23, 2000 (RIMS # R08 001023 713)

ISI Data Base Exam Requirement Source: B01-02.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 54 of 205 7.11.5 Augmented Examination of Austenitic Stainless Steel and Dissimilar Metal Welds Susceptible to IGSCC (BWRVIP-75-A)

Responsible organization: Site Engineering. Austenitic stainless steel and dissimilar metal circumferential welds in piping four inches or larger in nominal pipe diameter which contain reactor coolant at a temperature above 200 degrees F during power operation shall be examined in accordance with the requirements of BWRVIP-75-A as modified by the referenced safety evaluation beginning with the Unit 2, cycle 12 outage. Sample expansion of IGSCC Category B, C, D, or E weldments shall be in accordance with BWRVIP-75-A. The welds requiring examination per this paragraph are listed in Section 8.2 - Part 1 by IGSCC category. The Stainless Steel and Dissimilar metal welds that are exempt from examination because they contain coolant of 200 degrees or less are listed in Section 8.2 - Part 2, In addition to the requirements for procedure and personnel qualification in Section 7.2.3B, the examination procedures and personnel used for IGSCC examinations per BWRVIP-75-A shall be qualified to the requirements of Appendix VIII of ASME Section XI as implemented by the Performance Demonstration Initiative Program (PDI).

The IGSCC category and corresponding examination frequency are listed below are in accordance with the BWRVIP-75-A and the referenced safety evaluation. This schedule is applicable when the successive examination requirements specified in the safety evaluation are satisfied.

IGSCC CATEGORY EXAMINATION EXTENT AND SCHEDULE A Sampled per the Risk - Informed ISI Program.

B None in Unit 2.

C 25% sample every 10 years.

D 100 percent every 6 years.

E 25% of the welds with corrosion resistant overlay material every 10 years. 50% of these welds within 6 years.

100% of the non-overlay welds with cracking every 6 years.

F 100 percent every refueling outage G 100 percent during current outage A. Risk-Informed Alternative Measures Generic Letter 88-01 provides the NRC positions on IGSCC in BWR austenitic stainless steel piping and requests the licensee to indicate if they intend to propose alternative measures.

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(continued)

NRC Regulatory Guides 1.174, 1.178 and ASME Code Case N-577 were utilized to develop alternatives to the inspection requirements of Generic Letter 88-01. These alternatives are identified in the RI-ISI section of this program.

For specific details on the RI-ISI Program see Section 7.14 of this program.

7.11.6 Technical Surveillance Requirement (TSR) 3.4.3.2 Responsible organization: Site Engineering. Additional examinations shall be performed each inspection interval on selected circumferential pipe welds to provide additional protection against pipe whip in accordance with TSR 3.4.3.2. This TSR identifies the need to meet as closely as possible the requirements of ASME Section XI. Piping welds are examined in accordance with the Risk Informed-ISI (RI-ISI) Program as alternative to ASME Section XI. Therefore, examination volumes, examination methods, and acceptance standards for piping welds examined in accordance with TSR 3.4.3.2 should be similar to the RI-ISI Program.

This examination criteria utilized for the RI-ISI Program is specified in Table 1, Examination Category R-A, of Code Case N-577.

The following welds will be examined each inspection interval for pipe whip protection using examination volumes, examination methods, and acceptance standards specified in Item No. R1.11 of Examination Category R-A of Code Case N-577:

GFW-2-15, GFW-2-32, KFW-2-38, KFW-2-39, GMS-2-24.

A report of these examinations shall be included with the augmented summary report portion of the ISI Summary Report.

Examination Requirement Source: B04-02 7.11.7 RPV Interior Examinations Augmented examinations of RPV interior components are performed in accordance with 0-TI-365, Reactor Pressure Vessel Internals Inspection (RPVII) Units 1, 2, and 3. Examination results shall be included in the augmented summary report (notification of unsatisfactory results may impose additional reporting requirements as denoted by the source requirement).

7.11.8 Level Instrumentation Nozzle Safe Ends BWRVIP-49 Responsible organization: Site Engineering

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Inspections prior to Unit 2 Cycle 12 refueling outage were required in accordance with GE SIL-571. According to BWRVIP-49, "Instrument Penetration Inspection and Flaw Evaluation Guidelines", it is the intent that the inspection and evaluation guidelines be followed in place of any prior GE SIL (i. e. GE SIL-571) related to essential safety functions of the instrument penetrations. The BWRVIP-49 document follows ASME Section XI Code examinations, with no additional augmented BWRVIP examinations. For commercial dependability, an ASME Section XI, IWB-2500, Code Category B-P, VT-2 examination for instrument penetrations shall be performed as an augmented examination. A VT-2 leakage inspection shall be performed of the safe end to nozzle weld during the drywell leakage test performed each outage. Insulation removal is NOT necessary to perform the leak check.

ISI Data Base Exam Requirement Source: B07-02.

7.11.9 Core Plate delta/P/Standby Liquid Control (SLC) Nozzle BWRVIP-27 Responsible organization: Site Engineering Inspections prior to Unit 2 Cycle 12 refueling outage were required in accordance with GE SIL-571. According to BWRVIP-27, "BWR Standby Liquid Control System/Core Plate deltaP Inspection and Flaw Evaluation Guidelines", it is the intent that the inspection and evaluation guidelines be followed in place of any prior GE SIL (i. e. GE SIL-571) related to essential safety functions of the instrument penetrations. The BWRVIP-27 document follows ASME Section XI Code examinations, with no additional augmented BWRVIP examinations. For commercial dependability, an ASME Section XI, IWB-2500, Code Category B-P, VT-2 examination for instrument penetrations shall be performed as an augmented examination. A VT-2 leakage inspection shall be performed of the safe end to nozzle weld and safe end during the drywell leakage test performed each outage.

Insulation removal is required to perform the leak check.

ISI Data Base Exam Requirement Source: B07-02.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 57 of 205 7.11.10 Core Spray and Recirc Inlet Safe Ends Responsible organization: Site Engineering Perform UT inspection of the Core Spray and Recirc Inlet Safe Ends per the recommendation of GE Letter No. BFSE 94-007. The Unit 2 Core Spray and Recirc Inlet Safe Ends were replaced during the Cycle 5 outage. The safe ends were replaced with IGSCC resistant material, and in the case of the Recirc-inlet safe ends an improved design was used which eliminated crevices. These changes in materials and design mitigate the possibility of future Inter Granular Stress Corrosion Cracking (IGSCC). Per guidance provided in NRC Generic Letter 88-01 (NUREG 0313 Rev 2) and the recommendation of GE Letter No. BFSE 94-007, these safe ends shall be inspected at the frequency established for Category "A"weldments.

The accessible areas of the safe end base material which has exposure to the annulus/crevice area created by the thermal sleeve shall be inspected with UT. This inspection should be conducted in conjunction with the augmented UT inspection of the safe end to nozzle weld. Techniques previously used to inspect for safe end IGSCC cracking should be utilized as practical in the inspection effort to detect Internal Diameter (ID) initiated IGSCC indications.

The implementation interval started with the Unit 2 Cycle 5 outage. The described safe end base material shall be inspected at the same interval as category "A" weldments. Ideally the inspection should be performed in conjunction with the safe end to nozzle welds which are also classified category "A".

7.12 N 12A Instrument Nozzle Safe End Weld Overlay During the Unit 2 Cycle 15 Refueling Outage an Ultrasonic examination of the N12A, Safe-End to Pipe weld N12A-1 revealed a flaw. The flaw was evauated to be 84%

through-wall and was repaired by weld overlay in accordance with Code Case N-504-3 and Nonmandatory Appendix Q. Inservice examination requirements of the weld overlay consist of ultrasonic examination during the first or second refueling outage following application. Inservice examination of the N12A weld overlay shall be in accordance with ASME Section Xl Nonmandatory Appendix Q. Inservice examination of the overlay in accordance with Appendix Q shall continue as long as the repair remains part of the pressure boundary. By letter to the NRC dated May 22, 2009, TVA submitted Request for Relief 2-ISI-21 Revision 01.

7.13 Voluntary Examinations Certain examinations are done on a voluntary basis to obtain additional information to support inservice inspections or to resolve a problem identified through the corrective action program. Key voluntary examinations are defined and documented in this Section.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 58 of 205 7.14 Risk-Informed Inservice Inspection 7.14.1 Introduction The objective of the Inservice Inspection Program, 2-SI-4.6.G, is to address all piping locations that are subject to service induced degradation, in accordance with the requirements as specified in 10CFR50.55a. In accordance with NRC Regulatory Guides 1.174, 1.178, and Code Case N-577, with NRC approval an alternative inspection program which meets the criteria of IOCFR50.55a(a)(3)(i) to provide an acceptable level of quality and safety can be utilized.

These Regulatory Guides provide guidance specific to incorporating risk insights to inservice inspection programs of piping.

By incorporating insights from probabilistic safety assessment(PSA), traditional analysis, and operating reactor data an alternative inspection program known as risk informed inservice inspection program (RI-ISI) may be submitted for NRC review and approval.

7.14.2 Purpose Procedure 2-SI-4.6.G-A of this program outlines an acceptable alternative approach to the existing Section XI requirements for the scope and frequency of inspection of the ISI Program by utilizing risk informed techniques.

10CFR50.55a(a)(3)(i) allows the use of alternatives when authorized by the Director of the Office of Nuclear Reactor Regulation, when the alternative provides an acceptable level of quality and safety.

This alternative approach provides an acceptable level of quality and safety per 10CFR50.55a(a)(3)(i) by incorporating insights from probabilistic safety assessment and traditional analysis calculations supplemented with reactor operating data. The RI-ISI Program therefore will be enforceable under 10CFR50.55a.

7.14.3 Scope Procedure 2-SI-4.6.G-A of this program outlines the details for risk-informed requirements for inservice inspection of ASME Class 1, and 2 equivalent piping.

The examination requirements of Section XA shall be used for piping evaluated by the risk-informed process. The RI-ISI Program shall be implemented for the fourth inspection interval for BFN Unit 2.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 59 of 205 7.14.4 Frequency The inspection periods and inspection interval are defined in the Section XI Edition and Addenda as committed to in 2-SI-4.6.G. The piping segments and inspection strategy (i.e. frequency, number of inspections, methods, or all three) are defined in Attachment 1 Part 6 - Risk Informed Inspections and 2-SI-4.6.G-A. Inspections may be increased or relaxed as experience dictates.

The number of inspections and the frequency of those inspections will be a product of the systematic application of the Risk-Informed Process.

7.14.5 Living Program A. Program Implementation Nuclear Energy Institute document NEI 04-05 "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems" was published in April 2004. The recommendations of this document will be implemented at Browns Ferry Unit 2. As a minimum, updates to 2-SI-4.6.G-A, RI-ISI Program, shall be performed at least on the basis that coincide with the inspection program requirements contained in Section Xl under Inspection Program B, 2-SI-4.6.G. Changes to the PSA, piping performance, plant procedures that affect system operating parameters, piping inspections, component and valve lineups, equipment operating modes, or the ability of plant personnel to perform actions associated with accident mitigation shall be reviewed for any RI-ISI Program updates. Leakage and flaws identified during scheduled inspections shall be evaluated for possible RI-ISI Program updates.Changes to the PSA, piping performance, plant procedures that affect system operating parameters, piping inspections, component and valve lineups, equipment operating modes, or the ability of plant personnel to perform actions associated with accident mitigation shall be reviewed for any RI-ISI Program updates. Leakage and flaws identified during scheduled inspections shall be evaluated for possible RI-ISI Program updates.

B. Performance Monitoring During each Period, the Program Owner will maintain an awareness of input changes. At the end of each Period, the effects of the changes will be evaluated to determine if a change to the Program is required.

The RI-ISI program will be updated, if required, before the next refueling outage. The Maintenance Rule Expert Panel will review proposed RI-ISI program changes and provide program oversight. The following provides an overview of the RI-ISI program inputs.

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1. Plant Design Feature Changes Design changes have the potential to change piping configuration and alter stress calculations which were used as input to the calculations performed in support of the RI-ISI program. New systems and branch piping will be evaluated for inclusion into the scope of the RI-ISI program. The existing design impact review process will be used to ensure the impact of design changes on RI-ISI has been appropriately considered prior to final approval. The calculations supporting the RI-ISI program will be entered into TVA's calculation tracking program to ensure appropriate predecessors and inputs are identified and considered during design change preparation and review.
2. Plant PSA Changes Since the PSA forms the basis for the RI-ISI program, any changes to the PSA or risk significance determination will be evaluated for impact on the RI-ISI program. PSA and design changes will be incorporated into the RI-ISI program as required.
3. Plant Procedure Changes Changes to plant procedures that affect ISI, such as system operating parameters, test intervals, or the ability of plant operations to perform actions associated with accident mitigation shall be considered in any RI-ISI program update. Additionally, changes in procedures that affect component inspection intervals, valve lineups, or operational modes of equipment shall also be assessed for their impact on changes in postulated failure mechanism initiation or Core Damage Frequency (CDF).
4. Equipment Performance Changes Equipment performance changes shall be reviewed to ensure that changes in performance parameters (e.g. valve leakage, increased pump testing, vibration problems) are considered in the RI-ISI program update. Specific attention shall be paid to these type conditions if not previously assessed in the qualitative inputs to the component selections of the RI-ISI program.

Adverse equipment performance will be evaluated for changes to the RI-ISI inspection scope.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 61 of 205 7.14.5 Living Program (continued)

5. Examination Results When scheduled RI-ISI program NDE examinations, pressure tests and corresponding VT-2 visual examinations for leakage have been completed, and unacceptable flaws, evidence of service related degradation, or indications of leakage have been identified, these conditions shall be evaluated in accordance with plant procedures as applicable to determine the adequacy of the scope of the inspection program and update the RI-ISI program as applicable.
6. Individual Plant and Industry Failure Information The Program Owner will consider applicable piping failures or degradations identified by the site's corrective action program. Industry awareness will be maintained through the sites Operating Experience program, NRC Generic Letters and Bulletins, site participation in Boiling Water Owners Group initiatives, and participation in the ASME Section Xl Code committee activities.
7. Program Review The Maintenance Rule Expert Panel will provide the oversight role for the RI-ISI program. The Expert Panel will review proposed changes to the program.

As with past reviews, personnel possessing expertise in RI-ISI evaluation and ISI inspection/evaluation will be present during presentation and review of the above items.

C. Corrective Action Program A corrective action program assures that conditions adverse to quality such as failures, malfunctions, deficiencies, deviations, defective material and equipment and non-conformances, are promptly identified and corrected.

When required by SPP-3.1, the measures must ensure that the cause of the condition is determined and corrective action taken to preclude repetition. The identification of the condition, the cause of the condition, and the corrective action are to be documented and reported to appropriate levels of BFN Site Management.

For Code Piping categorized as High Safety Significance (HSS) the corrective action shall be consistent with the provisions of ASME Section XI.

Any corrective action required as the result of RI-ISI examinations shall be handled in accordance with SPP-3.1.

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D. Acceptance Guidelines The acceptance guidelines for implementation, monitoring and corrective action programs for the RI-ISI program are as follows:

1. The implementation program will be evaluated based on the attributes stated in Sections 7.14.5A through 7.14.5C.
2. Assurance that a nonconforming component will be brought back into conformance in a timely manner. Corrective actions required by ASME Section Xl shall continue to be followed.
3. Evaluations within the corrective action program may also include:
a. Assuring the root cause of the condition is determined and the corrective actions taken preclude repetition. The identification of the condition, the cause of the condition, and the corrective action are to be documented and reported to appropriate levels of management.
b. Determining the impact of the failure or nonconformance on system/train operability since the previous inspection.
c. Assessing the applicability of the failure or nonconforming condition to other components in the RI-ISI program.
d. Correcting other susceptible RI-ISI components as necessary.
e. Incorporating the lessons in the plant data base and computer models, if appropriate.
f. Assessing the validity of failure rate and unavailability assumptions that can result from piping failure(s) used in the PSA or in support of the PSA.
g. Considering the effectiveness of the component's inspection strategy in detecting the failure or nonconforming condition.
h. Reducing the inspection interval and/or adjust inspection methods as appropriate, when the component (or group of components) experiences repeated failures or nonconforming conditions.
4. The corrective action evaluation shall be provided to the PSA and RI-ISI Groups for any model changes and regrouping as appropriate.
5. The RI-ISI program documents shall be revised to document any RI-ISI program changes resulting from corrective actions taken.

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6. A program is in place to monitor industry findings (i.e., SPP-3.9, Operating Experience Program).
7. Examination requirements include all piping evaluated by the risk-informed process and selected for examination.
8. Inspection Program The examinations shall be completed during each ten-year inspection interval with the following exceptions.
a. If, during the interval, a reevaluation using the RI-ISI process is conducted and scheduled items are no longer required to be examined, these items may be eliminated.
b. If, during the interval, a reevaluation using the RI-ISI process is conducted and items are required to be added to the examination program, those items shall be added.
9. Successive Inspections If piping structural elements are accepted for continued service by analytical evaluation in accordance with N-577-3200, the areas containing the flaws or relevant conditions shall be reexamined during the next three inspection periods referenced in the schedule of the inspection program of N-577-2400.

If the reexaminations required by N-577-2420 (b) reveal that the flaws or relevant conditions remain essentially unchanged for the three successive inspection periods, the piping examination schedule may revert to the original schedule of successive inspections.

10. Additional Inspections Examinations performed in accordance with N-577-2500 that reveal flaws or relevant conditions exceeding the acceptance standards of N-577-3000 shall be extended to include additional examinations. The additional examinations shall include piping structural elements described in Table 1 of Code Case N-577 with the same postulated failure mode and the same or higher failure potential.

The number of additional elements shall be the number of piping structural elements with the same postulated failure mode originally scheduled for that fuel cycle.

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The scope of the additional examinations may be limited to those more-safety-significant piping structural elements within the systems whose materials and service conditions are determined by an evaluation to have the same postulated failure mode as the piping structural element that contained the original flaw or relevant condition.

If the additional examinations required by N-577-2430 (a) reveal flaws or relevant conditions exceeding the acceptance standard of N-577-3000, the examinations shall be further extended to include additional examinations.

These examinations shall include all remaining piping elements within Table 1 of Code Case N-577 whose postulated failure modes are the same as the piping structural elements originally examined in N-577-2430 (a).

An evaluation shall be performed to establish when those examinations are to be conducted. The evaluation must consider failure mode and potential.

For the inspection period following the period in which the examinations of N-577-2430 (a) or (b) were completed, the examinations shall be performed as originally scheduled in accordance with N-577-2400.

11. Examination and Pressure Test Requirements
a. Pressure testing and VT-2 visual requirements are to be performed on ClassI, 2, and 3 (equivalent) piping systems in accordance with Section XI as specified in the BFN SPT Program (i.e., SPP-9.1 and SI-3.3 Series).
b. Examination qualification and methods and personnel qualification are to be in accordance with 2-SI-4.6.G, Section 7.2.3.
12. Acceptance standards for identified flaws and repair/replacement activities are to be in accordance with 2-SI-4.6.G, Sections 6.0 and 7.7.
13. Records and reports shall be prepared and maintained in accordance with 2-SI-4.6.G, Section 7.5.

7.14.6 Risk Informed Inservice Inspection Program Analysis The Analysis for Risk-Informed Inservice Inspection Program for BFN Unit 2 is outlined in 2-SI-4.6.G-A of this Program.

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-A B1.11 5 RPV 5 -* CHM-2046-C UT RPV Shell Circ Welds

  • See RFR 2-ISI-9 B-A B1.12 15 RPV 15 15 CHM-2046-C UT RPV Shell Long Weld B-A B1.21 3 RPV 1 1 - ISI-0408-C UT RPV Circ Top Hd WId ISI-0444-C (RCH-2-1C). *See Note 1 2 ISI-0408-C B-A B1.22 16 RPV 6 2 2 UT RPV Mer Top Hd WId ISI-0444-C (RCH-2-XV). *See Note 1 B-A B1.30 RPV 1 1 CHM-2046-C UT RPV FIg Weld (C-5-FLG)

B-A B1.40 RPV 1 1 ISI-0408-C UT RPV Hd-FIg Weld (RCH-2-2C)

B-A B1.40 RPV 1 1 ISI-0408-C MT RPV Hd-FIg Flex Area (RCH-2-2C-FLEX)

B-A B11.51 N/A RPV N/A RPV Repair Weld(s) None B-B Various N/A N/A None

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-D B3.90 31 RPV 31 *13 *10 *8 CHM-2046-C UT RPV Noz-Ves Weld (See Note 9) ISI-0408-C ISI-0380-C ISI-0444-C B-D B3.1 00 7 RPV 7 6 1 CHM-2046-C UT RPV Noz IR (N4-IR ISI-0269-C and N10-IR B-D B3.1 00 24 RPV 24 *7 *10 *7 CHM-2046-C EVT-1 RPV Noz IR ISI-0408-C ISI-0380-C ISI-0444-C B-F B5.10 17 RPV N/A ISI-0270-C N/A Noz-SE Weld > 4" (See Note 4) ISI-0271-C ISI-0272-C ISI-041 0-C B-F B5.20 N/A N/A N/A (See Note 4)

B-F B5.30 N/A N/A None B-G-1 B6.1 0 92 RPV 92 30 31 31 ISI-0266-C VT-1 Clos Hd Nuts

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Cormponents Examination Item Number of Syster n/ Interval Period Period Period Sho wn On ISI Ex am(s)

Category No Components Subtot.al Sample Sample Sample Sample Dr*awing # Re quired Remarks B-G-1 B6.20 92 RPV 92 92 ISI -0266-C UT(*2) Studs (In Place)

(See Note 2)

Code Case N-652-1 B-G-1 B6.40 92 RPV 92 92 ISI-0266-C UT Threads in Flange (When Head Removed)

B-G-1 B6.50 92 sets RPV 92 sets 30 sets 31 sets 31 sets ISI-0266 -C /T-1 Washer (Sets of 2)

B-G-1 B6.50 92 RPV 92 *4 - ISI-0266-C VT-1 *Bushings (when RPV Head disassembled.

Reference Code Inquiry # IN04-009).

B-G-1 B6.150 N/A N/A None B-G-1 B6.180 32 RECIR 16(*3) 16 - ISI-0407-C UT Recir Pump Bolting B-G-1 B6.190 2 RECIR 1 (*3) 1(*3) - ISI-0407-C VT-1 Flange Face (See Note 8)

B-G-1 B6.200 32 RECIR 16(*3) 16 - ISI-0407-C VT-1 Recir Pump Nuts B-G-1 B6.200 32 RECIR 16(*3) 16 - ISI-0407-C VT-1 Recir Pump Washers B-G-1 B6.210 N/A N/A None

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-G-2 B7.1O0 N/A N/A None B-G-2 B7.50 *30 *5 VT-1 Pipe Bolting < 2"

  • 1 **Bolting to RPV Head MS/12 ISI-0312-B
  • 1 Nozzles MS/13 ISI-0312-B
  • 1 *Code Case N-652-1 RECIR/2 ISI-0270-C RPV/3** *2 ISI-0408-C B-G-2 B7.60 *4 RECIR/4 *2 ISI-0407-C

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-G-2 B7.70 37 9(*6) VT-1 Valve Bolting _<2" CS/4 2(*6) ISI-0271-C See Section 8.3

  • t FW/4 1('6) ISI-0269-C FCV-1 -XX MS /8 1('6) ISI-0222-C PCV-1 -XXX ISI-0312-B *Code Case N-652-1 MS/13 1('6)

RECIR/6 3(*6) ISI-0270-C 1(*6) ISI-0221-C RHR/2

  • Code Case N-652-1 B-G-2 *See 185 *VT-1 CRD *See Note 16 ISI-0292-C Note When disassembled 12 12 and bolts reused.

B-J B9.11 403 N/A N/A Circumferential (See Note 4) Welds > 4"

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks CS/31 ISI-0271-C FW/77 ISI-0269-C RPV Noz-SE Alloy Steel Welds (NOT HPCI/21 ISI-0273-C Dism Metal)

MS/120 ISI-0222-C RCIC/6 ISI-0272-C RECIR/78 ISI-0270-C RHR/36 ISI-0221-C RPV/13 ISI-0222-C RWCU/21 ISI-0269-C ISI-0408-C ISI-0272-C B-J B9.21 47 N/A N/A Circumferential Welds (See Note 4) < 4"

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks FW/14 ISI-0222-C MS/2 ISI-0222-C RCIC/22 ISI-0272-C RECIR/3 ISI-0270-C RPV/1 ISI-0272-C RWCU/1 ISI-0272-C SLC/4 ISI1-0380-C B-J B9.31 43 N/A N/A Branch Connections (See Note 4) >4" MS/27 - ISI-0222-C RECIRI15 - ISI-0270-C RHR/1 - ISI-0221-C B-J B9.32 9 N/A N/A Branch Connections (See Note 4) <4"

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks RCIC/1 ISI-0272-C RHR/4 ISI-0221-C RECIR/2 ISI-0270-C RWCU/2 ISI-0272-C B-J B9.40 249 N/A N/A Socket Welds (See Note 4)

FW/62 ISI-0222-C MS/74 ISI-0222-C RCIC/16 ISI-0272-C RHR/12 ISI-0221 -C RECIR/12 ISI-0270-C SLC/40 ISI-0380-C RWCU/33 ISI-0272-C B-K B10.10 1 RPV/1 1 1 ISI-0415-C *MT *RPV SPRT Skirt ref.

Code Case N-700

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-K B10.20 101 13 4 6 3 PT or MT Piping Welded Attachments CS/6 1 1 - ISI-0280-C MT or PT FW/29 3 3 - ISI-0277-C MT or PT HPCI/2 1 1 - ISI-0275-C MT or PT MS/36 4 4 - ISI-0279-C MT or PT RECIRI2O 2 2 ISI-0278-C PT RHR/6 1 - ISI-0276-C PT RWCU/2 1 1 ISI-0274-C PT B-K B1I0.30 6 RECIR 1 1 ISI-0278-C PT Pump Welded Attachments B-K B1 0.40 2 FW 1 1 ISI-0277-C MT or PT Valve Welded Attachments B-L-1 B12.10 N/A N/A None B-L-2 B12.20 2 RECIR 1('7) 1('7) ISI-0407-C VT-3 Pump Casing Interior

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks B-M-1 B12.30 N/A N/A None B-M-1 B12.40 N/A N/A None B-M-2 B12.50 21(Groups) 21('5) VT-3 Valve Body > 4" See 54 Total Section 8.3 Valves CS/6 3(*5) ISI-0271-C

'k FW/6 2(*5) ISI-0269-C FCV-1-XX "k

HPCI/3 2(*5) ISI-0273-C PCV-1-XXX

'k MS/8 1('5) ISI-0222-C MS/13 1('5) ISI-0312-C RCIC/1 1('5) ISI-0272-C RECIR/6 ISI-0270-C 3(*5)

RHR/9 ISI-0221-C 6(*5)

RWCU/2 ISI-0272-C 2(*5)

B-N-1 B13.10 1 RPV 1 1 1 1 CHM-2046-C VT-3 RPV Interior B-N-2 B13.20 1 RPV 1 1 CHM-2046-C VT-1 RPV Interior Att in Beltline Region

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Section 8.1 Examination Schedule PART 1 - CLASS 1 EQUIVALENT (IWB) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required - Remarks B-N-2 B13.30 1 RPV 1 1 CHM-2046-C VT-3 RPV Interior Att Beyond Beltline Region B-N-2 B13.40 1 RPV 1 1 CHM-2046-C VT-3 Shroud Supp Surfaces B-O B 14.10 40 RPV ISI-0292-C PT CRD Housing welds

  • REFER TO Para.

7.3.1 .G B-P ALL ------------------------

See Section 7.8 and SPP-9.1 VT-2 Pressure Test Program N12A WELD OVERLAY AND ASSOCIATED INSTRUMENT NOZZLE SAFE END TO PROCESS PIPE WELD EXAMINATIONS (SECTION 7.12)

N/A N/A 1 FW 1 1 ISI-0383-C-01 1UT N12A Weld Overlay

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Section 8.1 Examination Schedule NOTES

1) The accessible length of the RPV circumferential and meridional head welds will be ultrasonically examined in accordance with the extent and frequency of examination in Table IWB-2500-1, Examination Category B-A, Item Numbers B1.21 and B1.22. The two bottom head circumferential welds (C-S-LH and C-S-BH) and ten bottom head meridional welds (V-BH-1 through V-BH-10) are inaccessible because of their location in the bottom head and proximity to the CRD and in-core instrumentation housings. These welds will NOT be scheduled for examination since they are inaccessible. The accessible portions of the seven top head circumferential and meridional welds will be ultrasonically examined.
2) Studs (Bolting) may be examined in place under tension (B6.20), when connection is disassembled, or when the studs are removed. The four studs (# 22, 23, 24, and 25) normally removed for refueling have been scheduled under Item Number B6.20. Reference Code Case N-652-1 and Table 1, Notes 1, through 7.
3) Examine bolting of only one pump in accordance with Table IWB-2500-1, B-G-1, NOTE: (3) in conjunction with B-L-2, NOTE: (1).
4) Reference SECTION 8.1 EXAMINATION SCHEDULE PART 6 - RISK - INFORMED INSPECTIONS for Examination Categories B-F Item No. B5.10 and B5.20 and B-J Item No. B9.11, B9.21, B9.31, B9.32, and B9.40.
5) Examine only one valve per Group in accordance with Table 2500-1, B-M-2, NOTE: (3). There are 21 Groups of Class 1 valves NPS 4 or larger.
6) Number of Groups of Class 1 valves exceeding NPS 4 contained within this system (Examination Category and/or Item Number). All of the bolts or studs and nuts in each connection of one valve within each group of valves shall be visually examined during the inspection interval in accordance with visual examination method VT-1. All of the bolting from one valve within each group of valves shall be examined during the inspection interval when the B-M-2 valve interior surface examination is performed.
7) If the B-M-2 valve interior surface examination is NOT performed during the interval, then all of the bolting from one valve in each group of valves shall be visually examined in place at the end of the interval.
8) Examine only one pump in accordance with Table IWB-2500-1, B-L-2, NOTE: (1), (2) and (3).
9) At least 25% but NOT more than 50% of the RPV nozzles shall be examined by the end of the first period and the remainder by the end of the inspection interval, in accordance with Table IWB-2500-1, Category B-D, Note: (2).
10) Examination includes 1 (one) inch annular surface of flange surrounding each stud, in accordance with TableIWB-2500-1, Category B-G-1, Note: (4).

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Section 8.1 Examination Schedule NOTES

11) TVA will take both ASME Section XI Code credit and Augmented credt in accordance with NUREG-0619/GE-NE-523-A71-0594-A, Revision 01 for all six (N4) Feedwater Nozzles.
12) Code Category B-G-2, Item No. B7.80, CRD Housing Bolts, Studs, and Nuts were deleted in the ASME Section XI, Division 1 Code 1995 Edition 1995 Addenda. 10CFR 50.55a published October 01, 2004, 69FR58804 includes a MANDATORY limitation in (b)(2)(xxi) (B), as follows: "9B) The provisions of Table IWB-2500-1, Examination Category B-G-2, Item No. B7.80, that are in the 1995 Edition are applicable only to reused bolting when using the 1997 Addenda through the latest edition and addenda incorporated by reference in paragraph (b)(2) of this section." If the RPV CRD Flange bolting is disassembled and the bolting is reused, they shall receive a VT-1i Visual examination during the second and third period.

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Section 8.1 Examination Schedule PART 2 - CLASS 2 EQUIVALENT (IWC) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks C-A C1.10 12 RHR 3('1) 1 1 1 ISI-0406-C UT RHRHX RHRG-2-07 Shell RHRG-2-08 Circ.

RHRG-2-09 Head C-A C1.20 4 RHR 1(*1) 1 ISI-0406-C UT RHRHX RHRG-2-10 Circ C-A C1.30 N/A N/A None C-B C2.10 N/A N/A None C-B C2.22 N/A N/A None C-B C2.31 16 RHR 4(*2) 2 2 ISI-0406-C MT RHRHX RHRG-2-05A, RHRG-2-05B, RHRG-2-06A, RHRG-2-06B Noz Reinforcing Plate Welds.

C-B C2.32 N/A N/A None C-B C2.33 8 RHR 2(*2)(*5) 2 2 2 ISI-0406-C VT-2 RHRHX RHRG-2-05 &

-06, Noz Reinforcing PIt relief "telltale" hole @

Press Test.

C-C C3.10 12 RHR 1('3) 1 ISI-0406-C MT Pressure Vessel Welded Attachments RHRHX

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Section 8.1 Examination Schedule PART 2 - CLASS 2 EQUIVALENT (IWC) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks C-C C3.20 145 16 6 4 6 MT or PT Piping Welded Attachments CRD/6 1 1 ISI-0041-C PT CS/16 2 2 ISI-0105-C MT or PT HPCI/18 2 1 1 ISI-0130-C MT or PT MS/9 1 ISI-0079-C MT or PT 1

RBCCW/1 1 ISI-0032-C MT or PT RCIC/9 1 ISI-0131-C MT or PT RHR/86 9 3 5 ISI-0324-C MT or PT C-C C3.30 4 RHR 1 1 ISI-031 0-B MT or PT Pump Welded Attachments C-C C3.40 N/A N/A None C-D N/A N/A N/A None C-F-1 C5.11 8 (*4) N/A N/A Dissim metal & SS circ welds > 4".

CS/6 ISI-0103-C RHRJ2 MSG-0018-C C-F-1 C5.20 N/A N/A BWR Plant

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Section 8.1 Examination Schedule PART 2 - CLASS 2 EQUIVALENT (IWC) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks C-F-1 C5.30 N/A N/A None C-F-1 C5.40 N/A N/A None C-F-2 C5.51 1104 N/A N/A CS Circ Welds > 4".

(See Note 4)

CRD/72 ISI-0040-C CS/1 64 ISI-0103-C Includes containment HPCI/168 heat removal.

ISI-0128-C MS/114 MSG-0021 -C RBCCW/16 ISI-0031-C RCIC/84 ISI-0129-C RHR1486 MSG-0018-C C-F-2 C5.60 N/A N/A BWR Plant C-F-2 C5.70 N/A N/A None

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Section 8.1 Examination Schedule PART 2 - CLASS 2 EQUIVALENT (IWC) COMPONENTS Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks C-F-2 C5.81 5 N/A N/A (See Note 4) Sweep-o-let RHR/3 MSG-0018-C branch connection CS/1 ISI-0103-C MS/1 MSG-0021 -C C-G C6.10 N/A N/A None C-G C6.20 N/A N/A None C-H ALL ALL --See VT-2 Pressure Section 7. Test Program.

8 and SPP-9.1.

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Section 8.1 Examination Schedule NOTES

1) Examinations limited to one vessel of similar design, size, function in accordance with Table IWC-2500-1, Examination Category C-A, NOTE: (3).
2) Examinations limited to one vessel of similar design, size, function in accordance with Table IWC-2500-1, Examination Category C-B, NOTE: (4).
3) Examination requirement: For multiple vessels of similar design, function, and service, only one welded attachment weld of only one of the multiple vessels shall be selected for examination in accordance with Table IWC-2500-1, Examination Category C-C, NOTE: (4).
4) Reference SECTION 8.1 EXAMINATION SCHEDULE PART 6 - RISK - INFORMED INSPECTIONS for Examination Categories C-F-1 Item No. C5.11, C-F-2 Item No.C5.51 and C5.81.
5) The telltale hole in reinforcing plate shall be examined for evidence of leakage while the vessel is undergoing the system leakage test (IWC-5220), as required by Examination Category C-H, Table IWC-2500-1, Note: (5).

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Section 8.1 Examination Schedule PART 3 - CLASS 3 EQUIVALENT (IWD) COMPONENTS Third Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks D-A D1.10 N/A N/A None D-A D1.20 65 9 3 2 4 VT-1 Welded Attachments.

EECW/34 5 2 1 2 ISI-0368-C FPC/1 1 1 ISI-0133-C RHRSW/30 3 1 1 1 ISI-0145-C D-A D1.30 N/A N/A D-A D1.40 N/A N/A D-B D2.10 N/A All See Section 7.8 and SPP-9.1 VT-2 Pressure Test Program

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks F-A F1.10 139 43 15 13 15 VT-3 Class 1 Pipe Supports.

Subtotals follow below.

F-A FI.10A 6 2 1 1 VT-3 ONE DIRECTIONAL RIGID SUPPORTS MS/5 1 I ISI-0279-C See Note 3 RECIR/1 1 ISI-0278-C F-A F1.10B 17 6 2 2 2 VT-3 MULTIDIRECTIONAL RIGID SUPPORTS.

FW/4 1 ISI-0277-C See Note 3 HPCI/1 1 ISI-0275-C MS/5 1 1 ISI-0279-C RHR/3 1 1 ISI-0276-C 1

RWCU/2 ISI-0274-C CS/2 1 ISI-0280-C F-A F1.10C 59 16 4 6 6 VT-3 Variable Supports.

(Constant Force, Springs, ETC.)

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks CS/6 1 1 ISI-0280-C See Note 3 FW/1 0 3 3 ISI-0277-C HPCI/2 1 1 ISI-0275-C MS/16 5 4 1 ISI-0279-C RECIRI1 3 2 2 ISI-0278-C RHR/10 3 2 ISt-0276-C RWCU/2 1 1 ISI-0274-C F-A F1.10D 57 19 8 4 7 VT-3 Variable Supports.

(Snubbers)

CS/4 2 2 ISI-0280-C See Note 3 FW/19 6 6 ISI-0277-C HPCI/2 1 1 ISI-0275-C MS/20 6 4 2 ISI-0279-C RECIRI9 3 3 ISI-0278-C RHR/3 1 1 ISI-0276-C

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks F-A F1.20 324 52 16 17 19 VT-3 CLASS 2 PIPE SUPPORTS.

SUBTOTALS FOLLOW BELOW F-A F1.20A 103 17 5 5 7 VT-3 ONE DIRECTIONAL RIGID SUPPORTS CRD/14 2 2 ISI-0041-C See Note 3 CS/16 2 2 ISI-0105-C HPCI/25 4 4 ISI-0130-C RBCCW/2 1 1 ISI-0032-C RCIC/10 2 2 ISI-0131-C RHR/36 6 5 1 ISI-0324-C F-A F1.20B 60 10 3 3 4 VT-3 MULTIDIRECTIONAL RIGID SUPPORTS

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks CRD/13 2 2 ISI-0041-C See Note 3 CS/8 1 ISI-01 05-C HPCI/8 ISI-01 30-C MS/4 ISI-0079-C RCIC/4 ISI-0131-C RHR/23 4 3 ISI-0324-C F-A F1.20C 122 21 6 8 7 VT-3 Variable Supports.

(Constant Force, Springs, ETC.)

CS/14 3 3 ISI-0105-C See Note 3 HPCI/15 3 3 ISI-01 30-C MS/32 4 4 ISI-0079-C RCIC/5 1 1 ISI-0131-C RHR/56 10 7 3 ISI-0324-C F-A F1.20D 39 4 2 1 1 VT-3 Variable Supports.

(Snubbers,)

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks HPCI/6 0 ISI-0130-C See Note 3 RCIC/3 0 ISI-0131-C RBCCW/1 1 1 ISI-0032-C RHR/29 3 1 1 1 ISI-0324-C F-A F1.30 213 27 4 16 VT-3 CLASS 3 PIPE SUPPORTS.

SUBTOTALS FOLLOW BELOW.

F-A F1.30A 73 9 3 5 1 VT-3 ONE DIRECTIONAL RIGID SUPPORTS.

EECW/42 5 5 ISI-0368-C See Note 3 FPC/1 1 1 ISI-0133-C RH RSW/29 2 .2 -- ISI-0145-C

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks F-A F1.30B 127 15 1 11 3 VT-3 MULTIDIRECTIONAL RIGID SUPPORTS.

EECW/99 11 9 2 ISI-0368-C See Note 3 FPC/4 1 ISI-01 33-C RHRSW/23 2 2 ISI-0145-C F-A F1.30C 11 2 0 0 2 VT-3 Variable Supports.

(Constant Force, Springs, ETC.)

EECW/1 0 ISI-0368-C See Note 3 RHRSW/10 2 2 ISI-0145-C F-A FI.30D 4 1 VT-3 Variable Supports.

1 (Snubbers)

EECW/4 1 ISI-0368-C See Note 3 F-A F1.40 74 33 11 11 11 ALL CLASSES OF COMPONENT SUPPORTS (NOT PIPE SUPPORTS).

SUBTOTALS FOLLOW BELOW.

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Section 8.1 Examination Schedule PART 4 - COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks F-A F1.40A RCIC/1 1 0 0 1 ISI-0131-C VT-3 See Note 3 F-A F1.40B 48 16('1) 6 6 4 VT-3 MULTIDIRECTIONAL RIGID SUPPORTS.

CRD/6 3('1) 3 ISI-0041-C TANK SUPPORTS.

CS/4 1(*1) 1 ISI-0105-C PUMP SUPPORTS EECW/16 2(*1) 2 ISI-0368-C PUMP & STRAINER SUPPORTS HPCI/3 3 3 ISI-0130-C TURBINE & PUMP SUPPORTS.

RCIC/2 2 2 ISI-0131-C TURBINE & PUMP SUPPORTS.

RHRJ12 3('1) 3 ISI-0406-C RHR HX SUPPORTS.

RHR/4 1('1) 1 ISI-310-B RHR PUMP SUPPORTS RPV/1 1 1 ISI-0415-A See Note 3

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Section 8.1 Examination Schedule PART 4 -COMPONENT SUPPORTS (IWF)

Fourth Inspection First Second Third Components Examination Item Number of System/ Interval Period Period Period Shown On ISI Exam(s)

Category No Components Subtotal Sample Sample Sample Sample Drawing # Required Remarks F-A F1.40C 13 10('1) 5 1 4 VT-3 Variable Supports.

(Snubbers, Constant Force, Springs, ETC.)

FW/2 2 2 ISI-0277-C See Note 3 RECIR/2 2 2 ISI-0278-C RECIR/6 3('1) 3 ISI-0278-C RH R/2 2 I 1 ISI-0324-C RPV/1 1 1 ISI-0415-C VT-3 Variable F-A FI.40D 12 Supports.(Snubbers,)

RECIRC/12 6 4 2 ISI-0278-C See Note 3

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Section 8.1 Examination Schedule NOTES

1) 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. See Table -2500-1, Examination Category F-A, Note: (3).
2) The Control Rod Drive Hydraulic System lateral restraint clamps and beams are NOT classified as ASME Code items since they are NOT part of the core support structure or part of the pressure boundary. Reference DCN # 7792 and General Electric Document # 23A6371.
3) F-A item number suffices (A, B, C, D) represent categorization in accordance with NOTE (1) to Table IWF-2500-1, i.e., A = one directional rod hangers, B = multi-directional restraints, C = supports that allow thermal movement such as springs, D = other, including snubbers.

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Section 8.1 Examination Schedule PART 5 - AUGMENTED EXAMINATIONS(*1)

Exam Augmented Regmt Refueling Refueling Refueling Refueling Refueling Exam Ref. Program Number of Source Cycle 17 Cycle 18 Cycle 19 Cycle 20 Cycle 21 Exams Category - Section Components Code Sample Sample Sample Sample Sample required Remarks A 7.11.5 48 B02-02 UT Section 8.2 See *note BWRVIP-75-A

  • Note: An alternative Risk-Informed Inservice Inspection Program in accordance with Appendix A is implemented in the Second Period of the Fourth Inspection Interval. Alternative examination requirements for Examination Categories B-F, B-J, C-F-i, and C-F-2 are provided in Examination Category R-A. The Risk-Informed Program includes the requirements for Augmented Examination Category A.

C 7.11.5 112 (25% in B02-02 9 6 4 5 6 UT Section 8.2 10 Years) BWRVIP-75-A D 7.11.5 11 B02-02 4 3 4 4 UT Section 8.2 BWRVIP-75-A E(SI) 7.11.5 7(100% B02-02 4 3 4 UT Section 8.2 every BWRVIP-75-A 6 Years)

E (OL) 7.11.5 9 (25% in B02-02 2 2 UT Section 8.2 10 Years) BWRVI P-75-A G 7.11.75 2 B02-02 2 2 2 2 2 VT-2 Section 8.2 BWRVIP-75-A

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Section 8.1 Examination Schedule PART 5 - AUGMENTED EXAMINATIONS(*1)

Exam Augmented Regmt Refueling Refueling Refueling Refueling Refueling Exam Ref. Program Number of Source Cycle 17 Cycle 18 Cycle 19 Cycle 20 Cycle 21 Exams Category Section Components Code Sample Sample Sample Sample Sample required Remarks NA(*2) 7.11.5 15 B02-02 N/A Stainless Welds, Temp Exclusion NUREG-03113/

BWRVIP-75-A B-D 7.11.6 6 B01-02 6 UT NUREG-0619 Feedwater, Nozzle Bore

&IR B-N-1 7.11.6 6 B01-02 6 6 6 VT-1 NUREG-0619 Reference Fw Nozzle 0-TI-365 Spargers B-J 7.11.6 5 B04-02 5 UT TSR 3.4.3.2 Pipe Whip N/A 7.11.8 7 B07-02 7 7 7 7 7 VT-2 BWRVIP-27 7.11.9 BWRVIP-49

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Section 8.1 Examination Schedule NOTES

1) Most of these components are considered within the Code examination numbers presented in Parts 1 through 4 of this Section.

Where one examination may serve as Code credit and as Augmented credit, it shall be so credited.

2) These stainless steel welds contain coolant at a temperature of 200 degrees or less during power operation and do NOT require examination under NUREG-0313. These welds are listed in Section 8.2, Part 2.

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Section 8.1 Examination Schedule PART 6 - RISK-INFORMED INSPECTIONS Interval 4 Interval 4 Interval 4 Examination Number of System/ Period 1 Period 2 Period 3 Exarm Category Item No. Components Subtotal Sample Sample Sample ISI Drawing Meth od Remarks R-A R1.11 15 FW/2 - - 2 2-ISI-0269-C UT HPCI/6 6 - - 2-ISI-0273-C UT 2-ISI-0128-C MS/2 - - 2 2-MSG-0021-C UT RCIC/1 - 1 2-ISI-0129-C UT RHRI4 2 2 2-MSG-0018-C UT R-A R1.16 Examination procedures shall satisfy requirements of GL 88-/BWRVIP-75A)

Cat A 6 RWCU/5 2 2 1 2-ISI-0272-C UT RECIRC/1 1 2-ISI-0270-C UT

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Section 8.1 Examination Schedule PART 6 - RISK-INFORMED INSPECTIONS Interval 4 Interval 4 Interval 4 Examination Number of System/ Period 1 Period 2 Period 3 Exam Category Item No. Components Subtotal Sample Sample Sample ISI Drawing Method Remarks Cat C 25 CS/8 3 3 2 2-ISI-0271-C UT RECIRC/17 3 7 7 2-ISI-0270-C UT RHR/4 1 2 2-ISI-0221-C UT Cat D 2 RHR/1 1 2-ISI-0221-C UT Examine every 6 years RWCU/1 1 1 2-ISI-0272-C UT Cat E (SI) 5 RECIRC/5 5 4 2-ISI-0270-C UT Examine every 6 years Cat E 4 RWCU/3 1 2 2-ISI-0272-C UT Examine every 10 years.

(OL)

RECIRC/1 1 - 2-ISI-0270-C UT Cat G 2 RHR/2 2 4 4 2-ISI-0221-C VT-2 Inaccessible for UT. Must be visually examined every outage R-A R1.18 16 FW/12 12 Only those locations required per Line Segments MS/4 4 SPP-9.7

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Section 8.1 Examination Schedule BFN Unit 2 RPV Interior Checklist Part 7 - ASME Section XI Required Examinations ASME Section XI, Code Category B-N-1 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.10 4 - MAIN STEAM NOZZLES (N-3'S) VT-3 B13.10 6 - FEEDWATER NOZZLES (N4'S) VT-3 B13.10 FEEDWATER SPARGERS VT-3 B13.10 CORE SPRAY PIPING VT-3 B13.10 2 - CORE SPRAY NOZZLES (N5'S) VT-3 B13.10 CORE SPRAY SPARGERS VT-3 B13.10 TOP GUIDE ASSEMBLY (TOP SIDE) VT-3 B13.10 4 -INSTRUMENTATION NOZZLES (2-N-1i'S, 2-N-12'S) VT-3 B13.10 CRD RETURN NOZZLE (N9) VT-3 ASME Section Xl, Code Category B-N-2 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.20 JP RISER BRACE PAD WELDS VT-1i 6 B1 3.20 SURVEILL. SPECMN. BRACKETS (SHELL COURSE 2) VT-1 ASME Section XI, Code Category B-N-2 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.30 GUIDE ROD BRACKETS VT-3 B13.30 STEAM DRYER SUPPORT BRACKETS VT-3 B13.30 STEAM DRYER SUPPORT BRACKETS ON TOP HEAD VT-3 B13.30 FEEDWATER SPARGER BRACKET VT-3 B13.30 CORE SPRAY PIPING BRACKETS AND PADS VT-3 B13.30 RPV SHROUD SUPPORT TO RPV BOTTOM HEAD (H-9) VT-3

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 99 of 205 Attachment I (Page 35 of 36)

Section 8.1 Examination Schedule BFN Unit 2 RPV Interior Checklist Part 7 - ASME Section XI Required Examinations ASME Section XI, Code Category B-N-2 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.30 SURVEILL. SPECMN. BRACKETS (SHELL COURSE 3) VT-3 B13.40 TOP GUIDE VT-3 B13.40 CORE PLATE VT-3

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Section 8.1 Examination Schedule BFN Unit 2 RPV Interior Checklist Part 8 - ASME Section XI Supplemental Examinations ASME Section Xl, Code Category B-N-1 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.10 ALL INTERNAL COMPONENTS VT-3 ASME Section Xl, Code Category B-N-2 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.30 RPV SHROUD SUPPORT LEGS TO RPV BOTTOM VT-3 HEAD (H-12) 1 ASME Section Xl, Code Category B-N-2 ITEM NO. DESCRIPTION REQUIREMENT CYCLE 17 CYCLE 18 CYCLE 19 CYCLE 20 CYCLE 21 B13.40 FUEL SUPPORT CASTINGS VT-3 B13.40 CONTROL ROD BLADE GUIDE TUBES VT-3 B13.40 CONTROL ROD DRIVE HOUSINGS VT-3 B13.40 CORE SHROUD SUPPORT ABOVE CORE PLATE VT-3 B13.40 CORE SHROUD SUPPORT BELOW CORE PLATE VT-3 B133.40 CONTROL ROD DRIVE STUB TUBES VT-3

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 1 - UNIT 2 WELDS REQUIRED TO BE EXAMINED PER GENERIC LETTER 88-01/BWRVIP-75-A Exam Pipe Size IGSCC Exam Weld Number System Method (Inches) Weld Confiq Category 2RA5 RECIR UT 12 P,P A 2RA6 RECIR UT 12 P,SE A 2RB5 RECIR UT 12 P,P A 2RB6 RECIR UT 12 P,SE A 2RC5 RECIR UT 12 P,P A 2RC6 RECIR UT 12 PSE A 2RD5 RECIR UT 12 P,P A 2RD6 RECIR UT 12 PSE A 2RE5 RECIR UT 12 P,P A 2RE6 RECIR UT 12 PSE A 2RF5 RECIR UT 12 P,P A 2RF6 RECIR UT 12 PSE A 2RG5 RECIR UT 12 P,P A 2RG6 RECIR UT 12 P,SE A 2RH5 RECIR UT 12 P,P A 2RH6 RECIR UT 12 P,SE A 2RJ5 RECIR UT 12 P,P A 2RJ6 RECIR UT 12 P,SE A 2RK5 RECIR UT 12 P,P A 2RK6 RECIR UT 12 P,SE A 2RA1 RPV UT 12 N,SE A 2RB1 RPV UT 12 N,SE A 2RC1 RPV UT 12 N,SE A 2RD1 RPV UT 12 N,SE A 2RE1 RPV UT 12 N,SE A 2RF1 RPV UT 12 N,SE A 2RG1 RPV UT 12 N,SE A 2RH1 RPV UT 12 N,SE A 2RJ1 RPV UT 12 N,SE A 2RK1 RPV UT 12 N,SE A JP-2-1A RPV UT 4 N,SE A JP-2-1 B RPV UT 4 N,SE A DRWC-2-07A RWCU UT 6 P,P A DRWC-2-07B RWCU UT 6 P,P A RWC-2-001-GO01 RWCU UT 4 V,V A RWC-2-001-G002 RWCU UT 4 E,V A RWCU-2-003-GO01 RWCU UT 6 E,P A RWCU-2-003-G002 RWCU UT 6 E,P A RWCU-2-003-G003 RWCU UT 6 PEN,P A RWCU-2-003-025 RWCU UT 6 P,V A RWCU-2-003-026 RWCU UT 6 V,P A RWCU-2-004-G082 RWCU UT 4 P,V A

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 1 - UNIT 2 WELDS REQUIRED TO BE EXAMINED PER GENERIC LETTER 88-01/BWRVIP-75-A Exam Pipe Size IGSCC Exam Weld Number System Method (Inches) Weld Confiq Cate-gory RWCU-2-004-G083 RWCU UT 4 P,V A RWCU-2-003-071 RWCU UT 6 P,P A TCS-2-401 RPV UT 10 N,SE A TCS-2-417 RPV UT 10 N,SE A TCS-2-403 CS UT 10 PSE A TSCS-2-418 CS UT 10 PSE A DCS-2-04 CS UT 12 P,P C DCS-2-05 CS UT 12 P,V C DCS-2-07 CS UT 12 P,P C DCS-2-13 CS UT 12 P,P C DCS-2-13A CS UT 12 P,P C DCS-2-14 CS UT 12 P,V C DSCS-2-01 CS UT 12 E,P C DSCS-2-02 CS UT 12 E,P C DSCS-2-09 CS UT 12 P,P C TCS-2-405 CS UT 12 E,V C TCS-2-406 CS UT 12 P,V C TCS-2-410 CS UT 12 E,V C TCS-2-422 CS UT 12 P,V C TCS-2-426 CS UT 12 E,P C GR-2-01 RECIR UT 28 P,PMP C GR-2-02 RECIR UT 28 P,V C GR-2-03 RECIR UT 28 E,V C GR-2-04 RECIR UT 4 C,P C GR-2-07 RECIR UT 4 C,P C GR-2-08 RECIR UT 28 T,X C GR-2-09 RECIR UT 12 P,P C GR-2-12 RECIR UT 12 P,P C GR-2-18 RECIR UT 22 H,X C GR-2-19 RECIR UT 12 P,P C GR-2-22 RECIR UT 12 P,P C GR-2-25 RECIR UT 22 H,V C GR-2-26 RECIR UT 22 H,V C GR-2-27 RECIR UT 28 P,PMP C GR-2-28 RECIR UT 28 P,V C GR-2-29 RECIR UT 28 E,V C GR-2-30 RECIR UT 4 C,P C GR-2-33 RECIR UT 4 C,P C GR-2-34 RECIR UT 28 P,X C GR-2-35 RECIR UT 12 P,P C GR-2-38 RECIR UT 12 P,P C GR-2-41 RECIR UT 12 P,R C

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 1 - UNIT 2 WELDS REQUIRED TO BE EXAMINED PER GENERIC LETTER 88-01/BWRVIP-75-A Exam Pipe Size IGSCC Exam Weld Number System Method (Inches) Weld Conficq Category GR-2-44 RECIR UT 22 H,X GR-2-48 RECIR UT 12 P,P GR-2-51 RECIR UT 22 H,V GR-2-52 RECIR UT 22 H,V GR-2-54 RECIR UT 28 E,P GR-2-55 RECIR UT 28 P,T GR-2-56 RECIR UT 28 E,V GR-2-57 RECIR UT 28 P,V GR-2-58 RECIR UT 28 E,PMP GR-2-60 RECIR UT 28 E,P GR-2-62 RECIR UT 28 E,V GR-2-63 RECIR UT 28 P,V GR-2-63A RECIR UT 4 F,F GR-2-63B RECIR UT 4 BC KR-2-01 RECIR UT 4 BC KR-2-02 RECIR UT 28 E,P KR-2-03 RECIR UT 28 P,T KR-2-04 RECIR UT 4 BC KR-2-11 RECIR UT 22 R,X KR-2-12 RECIR UT 22 H,X KR-2-13 RECIR UT 12 BC KR-2-15 RECIR UT 22 C,H KR-2-19 RECIR UT 12 BC KR-2-20 RECIR UT 12 BC KR-2-23 RECIR UT 4 BC KR-2-24 RECIR UT 28 E,P KR-2-25 RECIR UT 28 P,T KR-2-26 RECIR UT 4 BC KR-2-33 RECIR UT 22 R,X KR-2-34 RECIR UT 22 H,X KR-2-35 RECIR UT 12 BC KR-2-42 RECIR UT 12 BC KR-2-45 RECIR UT 28 E,P KR-2-46 RECIR UT 28 P,T KR-2-47 RECIR UT 28 E,P KR-2-48 RECIR UT 28 E,P KR-2-49 RECIR UT 4 BC KR-2-50 RECIR UT 28 E,P KR-2-51 RECIR UT 28 E,P KR-2-52 RECIR UT 28 E,P KR-2-53 RECIR UT 4 BC DRHR-2-02 RHR UT 24 P,V

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 1 - UNIT 2 WELDS REQUIRED TO BE EXAMINED PER GENERIC LETTER 88-01/BWRVIP-75-A Exam Pipe Size IGSCC Exam Weld Number System Method (Inches) Weld Confiq Cateqory DRHR-2-04 RHR UT 24 E,P C DRHR-2-05 RHR UT 24 P,V C DRHR-2-06 RHR UT 24 P,V C DRHR-2-07 RHR UT 24 P,V C DRHR-2-08 RHR UT 24 P,V C DRHR-2-13 RHR UT 24 E,P C DRHR-2-14 RHR UT 24 E,V C DRHR-2-15 RHR UT 24 P,V C DRHR-2-16 RHR UT 24 E,V C DRHR-2-17 RHR UT 24 P,V C DRHR-2-18 RHR UT 24 P,T C DRHR-2-19 RHR UT 20 P,T C DRHR-2-21 RHR UT 20 E,V C DRHR-2-23 RHR UT 20 P,V C DSRHR-2-01 RHR UT 24 E,P C DSRHR-2-02 RHR UT 24 P,P C DSRHR-2-03 RHR UT 24 P,P C DSRHR-2-04 RHR UT 24 E,P C DSRHR-2-04A RHR UT 24 E,P C DSRHR-2-05 RHR UT 24 E,P C DSRHR-2-05A RHR UT 24 E,P C DSRHR-2-06 RHR UT 24 P,P C DSRHR-2-07 RHR UT 24 E,P C DSRHR-2-08 RHR UT 6 BC C DSRHR-2-09 RHR UT 20 E,P C DSRHR-2-10 RHR UT 20 E,P C DSRHR-2-11 RHR UT 20 E,P C TRHR-2-191 RHR UT 20 E,V C N 1A-SE RPV UT 28 N,SE C N 1B-SE RPV UT 28 N,SE C RWCU-2-003-027 RWCU UT 6 E,V D RWCU-2-003-044 RWCU UT 6 P,E D DSRWC-2-01A RWCU UT 6 P,E C DSRWC-2-01 RWCU UT 6 E,P C DSRWC-2-02 RWCU UT 6 E,P C DSRWC-2-06 RWCU UT 6 E,P C RCRD-2-49 CRD UT 4 E,V D RCRD-2-50 CRD UT 4 E,V D CRD-2-005-003 CRD UT 4 P,V D DRHR-2-03 RHR UT 24 P,V D DRHR-2-11 RHR UT 24 P,V D DRHR-2-12 RHR UT 24 P,V D

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 1 - UNIT 2 WELDS REQUIRED TO BE EXAMINED PER GENERIC LETTER 88-01/BWRVIP-75-A Exam Pipe Size IGSCC Exam Weld Number System Method (Inches) Weld Confiq Category RCRD-2-33 RPV UT 4 C,N D RWCU-2-003-069 RWCU UT 6 P,P D RWCU-2-003-070 RWCU UT 6 P,P D TCS-2-421 (OL) CS UT 12 E,V E GR-2-15(OL) RECIR UT 12 P,R E GR-2-45(OL) RECIR UT 12 P,P E GR-2-53 RECIR UT 28 PSE E GR-2-59(OL) RECIR UT 28 PSE E GR-2-61 (OL) RECIR UT 28 P,P E GR-2-64(OL) RECIR UT 28 E,PMP E KR-2-14 RECIR UT 12 BC E KR-2-36 RECIR UT 12 BC E KR-2-37 RECIR UT 22 C,H E KR-2-41 RECIR UT 12 BC E DRHR-2-09 RHR UT 24 P,T E DRHR-2-22 RHR UT 20 P,V E DSRWC-2-03(OL) RWCU UT 6 E,P E DSRWC-2-04(OL) RWCU UT 6 E,P E DSRWC-2-05(OL) RWCU UT 6 E,P E DRHR-2-03B RHR VT-2 24 P,P G DRHR-2-13B RHR VT-2 24 P,P G

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Section 8.2 Listing of Welds for Generic Letter 88-01 PART 2 - UNIT 2 STAINLESS AND DISSIMILAR METAL WELDS NOT SUBJECT TO GENERIC LETTER 88-01 EXAMS WELD PIPE SIZE WELD IGSCC NUMBER SYSTEM (INCHES) CONFIG CATEGORY EXAM METHOD COMr' ,ENTS DCS-2-01 CS 12 P,V NA N/A TEMP ERATURE EXCL. JSION DCS-2-02 CS 12 E,V NA N/A TEMP ERATURE EXCL. JSION DCS-2-03 CS 12 P,V NA N/A TEMP ERATURE EXCL. JSION DCS-2-1 0 CS 12 E,V NA N/A TEMP ERATURE EXCL. JSION DCS-2-11 CS 12 P,V NA N/A TEMP ERATURE EXCLLJSION DCS-2-12 CS 12 P,V NA N/A TEMP ERATURE EXCLAJSION DSCS-2-14 CS 12 E,P NA N/A TEMP ERATURE EXCL. JSION DSCS-2-15 CS 12 E,P NA N/A TEMP ERATURE EXCLI JSION DSCS-2-16A CS 12 P,PEN NA N/A INACC.ESSIBLE IN PErNETRATION X-16A TEMP ERATURE EXCLLUSION.

DSCS-2-16B CS 12 P,PEN NA N/A INACC ESSIBLE IN PE!NETRATION X-1 6A TEMPIERATURE EXCLLUSlON.

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Section 8.3 ASME Class 1 Equivalent Valve List GROUP VALVE IS1 DWG VENDOR MATERIA L VALVE NUMBER NUMBER SIZE INCH SYSTEM NUMBER VENDOR DWG NO. SPEC. TYPE COMMENTS 1 3-554 24 FW ISI-0269-C ATWOOD & 20788-H A-216 WC ;B CHECK 3-558 MORRILL 3-568 3-572 2 HCV3-66 24 FW ISI-0269-C POWELL 035879-2 A-216 WCB GATE No B-G-2 HCV3-67 Bolting 3 FCV68-01 28 RECIR ISI-0270-C DARLING 94-12086 A351 CF8 GATE FCV68-77 4 FCV68-03 28 RECIR ISI-0270-C DARLING 94-12086 A351 CF8 GATE FCV68-79 5 FCV68-33 22 RECIR ISI-0270-C DARLING 94-12086 A351 CF8 GATE FCV68-35 6 FCV1-14 26 MS ISI-0222-C ATWOOD & 20851-H A216 WCB GLOBE FCV1 -15 MORRILL FCV1 -26 FCV1 -27 FCV1-37 FCV1-38 FCV1-51 FCV1 -52

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Section 8.3 ASME Class 1 Equivalent Valve List GROUP VALVE ISI DWG VENDOR MATERIAL VALVE NUMBER NUMBER SIZE INCH SYSTEM NUMBER VENDOR DWG NO. SPEC. TYPE COMMENTS 7 PCV1 -004 6 MS ISI-0312-B Target Rock PL-7657-1 00 A216 WCB PILOT SEE NOTE 1 PCV1 -005 & A105 OPERATED PCV1-018 RELIEF PCV1-019 PCV1 -022 PCV1 -023 PCV1 -030 PCV1-031 PCV1-034 PCV1-041 PCV1-042 PCVI -179 PCVI -180 8 HCV74-69 24 RHR ISI-0221-C POWELL 035880-3 A351 CF8M GATE No B-G-2 HCV74-55 Bolting 9 FCV74-54 24 RHR ISI-0221-C ATWOOD & 20800-H A351 CF8M CHECK FCV74-68 MORRILL No B-G-2 10 FCV74-53 24 RHR ISI-0221-C WALWORTH A-12334-Mi F A351 CF8M GATE FCV74-67 A-12334-Ml F Bolting.

11 HCV74-49 20 RHR ISI-0221-C POWELL 036207-2 A351 CF8M GATE No B-G-2 Bolting 12 FCV74-47 20 RHR ISI-0221-C WALWORTH A-12332-MiC A216 WCB GATE No B-G-2 Bolting 13 HCV75-27 12 CS ISI-0271-C POWELL 0360334-2 A351 CF8M GATE HCV75-55

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Section 8.3 ASME Class 1 Equivalent Valve List GROUP VALVE ISI DWG VENDOR MATERIAL VALVE NUMBER NUMBER SIZE INCH SYSTEM NUMBER VENDOR DWG NO. SPEC. TYPE COMMENTS 14 FCV75-26 12 CS ISI-0271 -C ROCKWELL 2-PD420868 A351 CF8M CHECK FCV75-54 15 FCV75-25 12 CS ISI-0271-C WALWORTH 0-IVP-1 1978 A351 CF8M GATE NO B-G-2 FCV75-53 Bolting.

16 FCV69-01 BW/IP INTNL. W 9825123 SA351 CF8M GATE No B-G-2 17 6 RWCU ISI-0272-C FCV69-02 BW/IP INTNL W 9825098 SA351 CF8M GATE Bolting.

18 FCV71-40 6 RCIC ISI-0272-C ROCKWELL PD-420688 A216 WCB CHECK No B-G-2 Bolting 19 FCV73-02 10 HPCI ISI-0273-C CRANE PB-1 39989 A216 WCB GATE No B-G-2 FCV73-03 FLOWSERVE W00025604 Bolting.

20 FCV73-45 14 HPCI ISI-0273-C ROCKWELL PD-420687 A216 WCB CHECK No B-G-2 Bolting 21 FCV74-48 20 RHR ISI-0221 -C WALWORTH A-12331-M1C A351 CF8M GATE No B-G-2 Bolting.

22 69-630 4 RWCU ISI-0272-C ANCHOR/ C23650 SA351 CF8M CHECK Exempt from DARLING B-M-2 exams.

Size < = 4"

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Section 8.3 ASME Class I Equivalent Valve List NOTE

1) MSRV's with serial numbers; 205, 206, 207, 208, 209, 1014, 1015, 1016, 1032, 1033, and 1034 are complete forgings (A105).

All other MSRV's have cast bodies (A216 WCB) with forged top works (A105). At least one cast and one forged valve body will be examined during the inspection interval.

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Section 8.4 Class I Piping and Pump Flange Bolted Connections Group List Code Case N-652-1 GROUP COMPONENT SIZE NUMBER NUMBER (INCHES) SYSTEM ISI DWG NUMBER COMMENTS 1 MSBC-2-01 6 Main Steam 2-ISI-0312-B-01 MSBC-2-02 6 Main Steam 2-ISI-0312-B-01 MSBC-2-03 6 Main Steam 2-ISI-0312-B-01 MSBC-2-04 6 Main Steam 2-ISI-0312-B-01 MSBC-2-05 6 Main Steam 2-ISI-0312-B-01 MSBC-2-06 6 Main Steam 2-ISI-0312-B-01 MSBC-2-07 6 Main Steam 2-ISI-0312-B-01 MSBC-2-08 6 Main Steam 2-ISI-0312-B-01 MSBC-2-09 6 Main Steam 2-ISI-0312-B-01 MSBC-2-10 6 Main Steam 2-ISI-0312-B-01 MSBC-2-11 6 Main Steam 2-ISI-0312-B-01 MSBC-2-12 6 Main Steam 2-ISI-0312-B-01 2 PCV1-2-004-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-005-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-018-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-019-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-022-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-023-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-030-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-031-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-034-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-041-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-042-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-179-PBC 6 Main Steam 2-ISI-0312-B-01 PCV1-2-180-PBC 6 Main Steam 2-ISI-0312-B-01 3 RBC-2-1 4 Recirc. 2-ISI-0270-C-01 RBC-2-2 4 Recirc. 2-ISI-0270-C-02 4 N6A-2-1-BC 6 RPV 2-ISI-0408-C-01 N6B-2-2-BC 6 RPV 2-ISI-0408-C-01 5 N7-2-3-BC 4 RPV 2-ISI-0408-C-01 6 RSF-A-1-BC 6 Recirc 2-ISI-0407-C-01 RSF-B-1-BC 6 Recirc 2-ISI-0407-C-01 7 RCH-A-2-BC 6 Recirc 2-ISI-0407-C-01 RCH-B-2-BC 6 Recirc 2-ISI-0407-C-01

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Section 8.4 Requests For Relief UNIT 2 REQUESTS FOR RELIEF

SUMMARY

LISTING RFR Description 2-ISI-9 Permanent Relief from inservice inspection requirements of 10 CFR Revision 50.55a(g) for the volumetric examination of the BFN Unit 2 RPV 001 circumferential welds (Code Category B-A, Item Nos. B1.11).

Approved by NRC on February 04, 2002.

Note: Request for Relief 2-ISI-9 has been approved for the remainder of the original license period. The original operating license period ends during the fourth interval; this Request for Relief will be resubmitted prior to the end of the original operating license period.

2-ISI-40 Relief to use alternate inspection and test plan for snubbers developed in accordance with Generic Letter (GL) 90-09, "Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions."

2-ISI-41 Relief from the requirements of ASME Section XI, Appendix VIII, Supplement 11, Examination of Piping Weld Overlays.

2-PDI-40 Relief to perform volumetric examination of the RPV Shell-To-Flange Weld and Head-To-Flange Weld, Code Category B-A, Item Nos. B1.30 and B1.40 to ASME Section XI, Appendix VIII, PDI requirements vs. ASME Section XI, Appendix I.

2-11-1 Updated Risk-Informed Inservice Inspection Program.

2-ISI-43 Relief to use ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-To-Shell Welds".

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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME)

SECTION XI, INSERVICE (ISI) AND AUGMENTED INSPECTION PROGRAM FOURTH TEN YEAR INSPECTION INTERVAL REQUEST FOR RELIEF 2- ISI-9, REVISION 1 EXECUTIVE

SUMMARY

TVA is requesting permanent relief from the inservice inspection requirements for volumetric examination of reactor pressure vessel (RPV) circumferential shell welds. This request applies to the remaining term of operation under the original license.

This request for relief will eliminate examination of the BFN Unit 2 RPV circumferential shell welds and is consistent with the guidance provided in NRC Generic Letter (GL) 98-05, "Boiling Water Reactor Licensees Use Of The BWRVIP-05 Report To Request Relief From Augmented Examination Requirements On Reactor Pressure Vessel Circumferential Shell Welds" dated November 10, 1998.

A final rule change to 10CFR50.55a was published in the Federal Register on August 6, 1992 (Federal Register Notice 57FR34666). The intent of this rule change was to require licensees to perform an expanded RPV shell weld examination as specified in the 1989 Edition of the ASME Section XI Code, on an "expedited" basis. Expedited in this context effectively means during the inspection interval that the rule was approved or the first period of the next inspection interval.

The examination schedule for the RPV axially oriented welds shall continue as required by the ASME Section XI Code.

TVA is scheduled to perform the RPV shell weld examinations required by the ASME Section XI Code in the third period of the Third Inservice Inspection Interval.

The BWRVIP-05 Report and the associated NRC SER supports exclusion of the examinations of the RPV circumferential shell welds provided certain limiting conditions regarding end of license vessel embrittlement and cold over-pressurization events are satisfied. TVA has satisfied the limiting conditions specified in GL 98-05 for BFN Unit 2.

This request for relief is consistent with ones submitted to NRC for BFN Unit 3 by TVA letters dated June 25, 1999, and October 22, 1999, and for BFN Unit 2 by TVA letter dated March 24, 2000. NRC letter to TVA dated November 18, 1999, approved the BFN Unit 3 request for relief. NRC letter to TVA dated August 14, 2000, approved the BFN Unit 2 request for relief.

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Therefore, in accordance with the guidance provided in GL 98-05 and pursuant to 10 CFR50.55a(a)(3)(i), TVA requests that relief be granted from performing the volumetric examinations of the BFN Unit 2 RPV circumferential shell welds.

Note: TVA is including this request for relief in the BFN Unit 2 ASME Section XI ISI Fourth Interval Program for information since the request was submitted in the Second ISI Interval for the remaining life of the plant under the original license.

Unit: Two (2)

System: Reactor Pressure Vessel (RPV)

Components: Table 1 lists the BFN Unit 2 RPV circumferential welds for which TVA is requesting permanent relief from volumetric examination.

The proposed relief is for the remaining term of operation under the original license.

TABLE 1 Weld Description CateQory and Table Exam Method IWB-2500-1 Item Number Vessel Shell to Shell Weld B-A, Volumetric B1.11 No.C-4-5 Vessel Shell to Shell Weld B-A, Volumetric B1.11 No. C-3-4 Vessel Shell to Shell Weld B-A, Volumetric B1.11 No. C-2-3 Vessel Shell to Shell Weld B-A, Volumetric B1.11 No. C-1-2 (Located in Belt-line Region)

Vessel Shell to Bottom B-A, Volumetric B1.11 Head Weld No. C-BH-1 ASME Code Class: ASME Code Class 1 Section XI Edition: 1986 Edition, no addenda Code Table: IWB-2500-1

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Examination Category: B-A (Pressure Retaining Welds in Reactor Vessel)

Examination Item Number: B 1.11 (Circumferential Shell Welds)

Code Requirement From Which Relief Is Requested: The inservice inspection requirements for the volumetric examination of RPV circumferential welds, ASME Section XI, Table IWB-2500-1, Examination Category B-A, Pressure Retaining Welds In Reactor Vessel, Examination Category B-A, Item B1.11, Circumferential Shell Welds, and the (expedited) augmented examination requirements of 10 CFR 50.55a(g)(6)(ii)(A) for vessel circumferential welds.

List Of Items Associated With The Relief Request: See Table 1 Basis for Relief: The basis for this request for relief is outlined in the NRC SER for the BWRVIP-05 Report and the guidance outlined in GL 98-05. These documents provide the basis for the elimination of examinations of the BWR RPV circumferential shell welds. The BWRVIP-05 Report SER concluded that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds. In addition, NRC conducted an independent risk-informed assessment of the analysis contained in the BWRVIP-05 Report SER. The NRC assessment and GL 98-05 concluded that the inspection of BWR RPV circumferential shell welds does not measurably affect the probability of failure. The industry examination results identified in the BWRVIP-05 topical report (Reference Electric Power Research Institute Report No. TR-1 05697) indicate that the necessity for performance of the circumferential shell weld volumetric examinations is not warranted based upon the low probability of failure of these welds.

TVA has addressed the two areas of concern outlined in the Permitted Action Section of Generic Letter 98-05: (1) the Unit 2 RPV level of embrittlement expected at the end of the period for which relief is requested in the most limiting RPV circumferential shell-weld areas, (2) the probability and expected frequency of the occurrence of a low temperature/high pressure transient on the Unit 2 RPV.

(1) Generic Letter 98-05 Permitted Action Item No. 1. Comparison Of The BFN Unit 2 RPV Brittle Fracture Information To The BWRVIP-05 And NRC Assessments Of The Probability Of Failure Of BWR RPV Circumferential Welds The BWRVIP-05 Report and the NRC Staffs independent risk-informed assessment of the initiative reports concluded that the probability of failure of the BWR RPV circumferential shell welds is orders of magnitude lower than that of the axial shell welds.

Additionally, the NRC assessment demonstrated that inspection of the RPV circumferential shell welds does not measurably affect the probability of failure.

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The independent NRC assessment included a Probabilistic Fracture Mechanics (PFM) analysis to estimate RPV failure probabilities. Three key assumptions in the PFM are: (1) the neutron fluence was that estimated to be the end-of-license mean fluence; (2) the chemistry values are mean values based on vessel types; and (3) the potential for beyond design basis events is considered. The BFN Unit 2 RPV was manufactured by Ishikawajima-Harima Heavy Industries Co. for Babcock & Wilcox. For plants with RPVs fabricated by Babcock and Wilcox (B & W), the mean end-of-license neutron fluence used in the NRC PFM analysis was 0.053 x 1019 n/cm 2 . The highest fluence anticipated at the end of the period of 32 EFPY for BFN Unit 2 (in the RPV belt line region, weld C-1-2) is 0.11 x 1019 n/cm 2 on the inside vessel surface. This fluence value was based on the BFN Unit 2 power uprate 32 EFPY operating curve information. The embrittlement for the BFN Unit 2 RPV due to fluence effects is less than the value obtained in the NRC limiting analysis for B & W RPV's shown in the SER (Table 2.6-4) for the BWRVIP-05 Report.

A comparison of the limiting BFN Unit 2 RPV circumferential shell weld analysis versus the NRC limiting analysis for B & W RPV's is provided in Table 2 below.

The BFN Unit 2 beltline region circumferential shell weld (C-1-2) was chosen for analysis to provide a basis for comparison to the NRC limiting analysis and as the Unit 2 RPV region where these calculated parameters would result in comparatively conservative values. The materials would also be representative of the Unit 2 RPV circumferential shell welds in general. The information in Table 2 represents the beltline region circumferential shell weld C-1-2, located between Unit 2 RPV shells course 1 and course 2. As shown in Table 2, the RTNDT for BFN Unit 2 is much lower than the NRC limiting case. Therefore, the conditional failure probability for BFN Unit 2 circumferential welds is bounded by the conditional failure probabilities in the NRC SER through the end of the original license period.

TABLE 2 PARAMETER BFN UNIT 2 Weld LIMITING B&W RPV C-1-2 Fluence (1019 n/cm2) 0.11 0.095 Initial RTNDT - 400F 200F Chemistry Factor 116.8 196.7 Cu (Wt %) 0.09% 0.31%

Ni (Wt %) 0.65% 0.59%

ARTNDT 50.90F 79.8%

Mean RTNDT [Initial 10.90F 99.80F RTNDT + ARTNDT]

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(2) Generic Letter 98-05 Permitted Action Item No. 2, Review Of BFN Unit 2 Procedural And Administrative Controls To Prevent RPV Low-Temperature / High-Pressure' Transient Events The NRC staff stated in GL 98-05 that beyond design-basis events occurring during plant shutdown could lead to cold over-pressure events that could challenge vessel integrity.

Although unlikely, the industry concluded that condensate and control rod drive pumps could cause conditions that could lead to cold over-pressure events that could challenge vessel integrity. For a BWR to experience such an event, the plant would require several operator errors. The NRC staffs assessment described several types of events that could be precursors to BWR RPV cold over-pressure transients.

These were identified as precursors because no cold over-pressure event has occurred at a U.S. BWR. The NRC assessment identified one actual cold over-pressure event that occurred during shutdown at a non-U.S. BWR. This event apparently included several operator errors that resulted in a maximum RPV pressure of 1150 psi with a temperature range of 79 0F to 88 0F.

The operating procedures for BFN Unit 2 are sufficient to prevent a cold over-pressure event from occurring during activities such as the system leak test performed at the conclusion of each refueling outage. Thus, the challenge to the BFN Unit 2 RPV from a non-design basis cold over-pressure transient is unlikely. The following discussion will provide further information to support TVA's conclusion.

BFN operating procedures and administrative control processes are in place to minimize the potential for occurrence of RPV cold over-pressurization events.

These processes include plant operating procedures, plant evolution planning and scheduling, administrative controls, and operator training.

Since cold over-pressurization events are most likely to occur during normal cold shutdown conditions, BFN operating procedures are written to require that RPV water level, pressure, and temperature are established and maintained in well controlled bands. BFN operators frequently monitor these parameters for abnormalities and indications of unwanted transients.

Also, any plant evolution which requires changes in these critical parameters is performed under the oversight of the Shift Manager who is also notified immediately of any abnormalities in the indications. Therefore, any deviation of these parameters from the established bands are promptly identified and corrected. In addition to these procedures, unit conditions for on-going activities which potentially can affect the maintenance of acceptable operating conditions and available contingency systems and plans are discussed by unit operations personnel at the time of shift turnover.

These administrative controls and procedures provide assurance that activities which could adversely effect RPV water level, temperature, and pressure are precluded.

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Nuclear Experience reviews and industry operating histories have shown that inadequate work-control processes and procedures could precipitate a cold over-pressurization event.

For BFN, outage work is controlled through planning and scheduling activities performed by the Outage management and Work Control Team. Unit and system work activities are carefully reviewed and coordinated to avoid conditions which could adversely affect the unit's RPV water level, temperature, and pressure.

Plant activities are routinely coordinated through the use of a plan-of-the-day (POD) which contains a list of activities to be performed and frequently contains cautionary notes on the activities.

These PODs are reviewed and discussed with station management and copies are maintained in appropriate locations. Changes to these PODs are approved through the Operations Department Management and the Shift Manager. In addition, during outages, work on unit systems and components is coordinated through work control centers which provide an additional level of unit operations oversight.

In the Main Control Room, the Shift Manager is required to maintain cognizance of any activity which could potentially affect reactivity, reactor water level, or decay heat removal.

BFN operators are required to provide positive control of reactor water level, temperature, and pressure within the specified bands, promptly report when operation outside the required bands occurs, and notify the Shift Manager of any restoration corrective measures being taken.

As part of the outage work control process, special procedures such as hydrostatic testing require pre-job briefings conducted with operations personnel for any activity which could potentially affect critical plant parameters. The pre-job briefing includes all cognizant individuals involved in the work activities. Expected plant system and component responses and contingency actions to mitigate unexpected conditions are also discussed.

When the plant is in cold shutdown, plant procedures require that the RPV head vent valves be opened after the reactor has been cooled to less than 212 0 F. Administrative and plant operations control procedures for this evolution and for controlling reactor water level, temperature, and pressure are an integral part of operator initial and re-qualification training.

Responses to abnormal water level and RPV conditions are also part of the operator's training.

In addition, unit-specific brittle-fracture operating pressure-temperature limit curves and procedures have been developed to provide the appropriate guidance for compliance with the operating limits and the associated Technical Specification requirements.

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REVIEW OF HIGH PRESSURE INJECTION SOURCES:

RPV water injection sources during cold shutdown conditions include three systems. During normal cold shutdown, RPV water level and pressure are controlled through the Control Rod Drive (CRD) and the Reactor Water Cleanup (RWCU) Systems. RPV conditions are controlled through a "feed and bleed" process using these two systems. The RPV and its piping system are not placed in solid water conditions and after the plant is cooled below 212 0 F, the head vent valves are opened. If either one of the RWCU or CRD Systems fail, the BFN operators would adjust the other system to maintain the proper water level and pressure. In addition, BFN also has water level instrumentation with set-points for high and low water levels that alarm at 39 inches high and 27 inches low to alert operators that a level transient is in progress and action is required. During these plant activities the CRD System typically injects water at a rate of less than 60 gallons per minute (gpm). Injection rates at this level allow the operator sufficient time to compensate for unanticipated level and pressure changes. Therefore, the probability of an occurrence of a high-pressure/low temperature event from these two systems, that places RPV conditions outside the pressure-temperature curve limits is low.

In addition to the RWCU and CRD Systems, the Standby Liquid Control System is another high-pressure source to the RPV. For BFN, SLC System operation occurs only if the system is manually initiated by operator action in accordance with emergency operating procedures.

Thus, SLC operation will not occur during cold shutdown operations except under stringently controlled test conditions. In the event of an inadvertent injection, the SLC injection rate (approximately 50 gpm) is sufficiently low to allow operators to intervene and control the reactor pressure.

During cold shutdown periods following refueling, the RPV is pressure tested in accordance with the applicable ASME Section Xl Code requirements. BFN hydrostatic tests of the RPV and the reactor coolant system are designated as complex and infrequently performed tests.

For these types of tests BFN requires a detailed pre-job briefing with all individuals participating in the test. Also, BFN has a dedicated operator for RPV water level and pressure control. RPV and reactor coolant system pressure testing is a carefully controlled plant evolution which receives special Operations management oversight and utilizes procedural controls to ensure that the test does not precipitate a transient outside the specified safety limits. These tests are also performed after the RPV and system are heated to the proper system inservice pressure test temperatures prior to increasing the system pressure. During these tests the RPV pressure, water level, and temperature are controlled through the CRD and RWCU Systems using the "feed and bleed" process. Increases (or decreases) in system pressure are limited to 50 pounds per square inch (psi) per minute. For example, if any RWCU valve fails, then the CRD pump is tripped and the RPV is depressurized. This practice minimizes the probability of exceeding the specified Technical Specification pressure-temperature limits during the system pressure test.

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During plant startup following a cold shutdown, the High Pressure Coolant Injection (HPCI) and the Reactor Core Isolation Cooling (RCIC) pumps provide a possible means to over-pressurize the RPV. However, for BFN, these systems have high pressure steam-driven pumps which have automatic isolation set-points of 100 psi and 50 psi respectively; and will not function when the plant is in cold shutdown.

Based upon the above evaluation the likelihood of a cold over-pressure transient event placing the Unit 2 RPV in non-design conditions is very low.

Therefore, the probability of an occurrence of a cold over-pressure transient is considered to be less than or equal to the probability used in the analysis described in the NRC independent evaluation performed in the assessment of the BWRVIP-05 Report.

ALTERNATIVE EXAMINATION:

As an alternative, TVA proposes to perform only the RPV longitudinal shell weld examinations during the third inspection period of the Fourth Ten-Year ISI Interval in conjunction with the scheduled ASME Section XI Code and augmented RPV Examinations.

JUSTIFICATION FOR THE GRANTING OF RELIEF:

Based upon the previous stated technical justifications, performance of the examination of the Unit 2 RPV circumferential shell welds in accordance with the ASME Code requirements, is not warranted. This position is supported by actual industry inspection experience, industry initiatives, and their supporting calculations. Further, the additional costs and personnel exposure that would be incurred without any apparent increase in safety does not warrant the performance of the examinations. These factors provide reasonable assurance of the continued structural integrity of the BFN Unit 2 RPV. Therefore, pursuant to 10 CFR 50.55a (a)(3)(i), TVA requests that permanent relief be granted from the inservice inspection and the augmented inspection requirements of 10 CFR 50.55a(g)(6)(ii)(A), for volumetric examination of reactor pressure vessel circumferential shell welds, ASME Section Xl, Table IWB-2500-1, Examination Category B-A, Item B1.11, Circumferential Shell Welds as permitted by GL 98-05.

Further, in accordance with the guidance specified in the NRC SER, Section 4.0 for the BWRVIP-05 Report, TVA intends to examine the RPV circumferential shell welds should axial weld examinations reveal an active mechanistic mode of degradation. The scope and schedule of these examinations would be submitted to NRC for approval. No relevant indications were identified by the examinations of the axial (longitudinal) welds conducted during the third period of the Second Ten-Year ISI Interval. RPV longitudinal shell weld examinations are currently scheduled during the third inspection period of the Third Ten-Year ISI Interval in conjunction with the scheduled ASME Section Xl Code and augmented RPV Examinations.

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This request for relief is consistent with ones submitted to NRC for BFN Unit 3 by TVA letters dated June 25, 1999, and October 22, 1999, and for BFN Unit 2 by TVA letter dated March 24, 2000. NRC letter to TVA dated November 18, 1999, approved the BFN Unit 3 request for relief. NRC letter to TVA dated August 14, 2000, approved the BFN Unit 2 request for relief.

IMPLEMENTATION SCHEDULE:

This Request for Relief will be implemented during the Second Ten Year ISI Inspection Interval for Browns Ferry Unit 2 and continue in effect for the remaining term of operation under the original license.

Reference NRC Safety Evaluation Report (SER) approval dated February 04, 2002 ATTACHMENT:

Brown Ferry Unit 2 RPV shell weld location schematic drawing These attachments are not contained in this relief request but are on file with the original in BFN Licensing.

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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM (FOURTH TEN YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-ISI-40 EXECUTIVE

SUMMARY

Pursuant to 10 CFR 50.55a(a)(3)(i), TVA is requesting relief from the identified ASME Section Xl Code requirements related to examination and testing of snubbers. TVA proposes to continue to use the examination and testing plans currently defined in the BFN Technical Requirements Manual (TR 3.7.4). The requirements for Snubbers were included in BFN Technical Specifications and were relocated to the TRM during the implementation of Improved Standard Technical Specifications at BFN. The current Technical Requirement Manual criteria have been promulgated and approved by the NRC, while ASME Section XI imposes overlapping requirements which do not enhance the quality or safety of the snubber examination and testing program.

Unit: Two (2)

System: Various Components: Component/Piping Snubbers ASME Code Class: 1, 2, 3 and MC Section Xl Edition: 2004 Edition Code Table: N/A Examination Category: N/A Examination Item Number: N/A REQUIREMENT FROM WHICH RELIEF IS REQUESTED:

The 2004 Edition of ASME Section XI, Article IWF-1 000 provides the requirements for inservice inspection (ISI) of Class 1, 2, 3, and MC component supports.

This includes the visual examination of snubbers. Article IWF-5000 contains the inservice test requirements (IST) for snubbers.

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BASIS FOR RELIEF:

ASME Section XI, Class 1, 2, 3 and MC equivalent snubbers are examined and tested in accordance with BFN Unit 2 Technical Requirements Manual (TRM), TR 3.7.4, "Snubbers" The TRM is prepared in accordance with the guidance provide by the NRC in Generic Letter (GL) 90-09, "Alternative Requirements for Snubber Visual Inspection Intervals and Corrective Actions". The scope for snubbers examined and tested in accordance with TR 3.7.4 is not limited by line size or other applicable code exemptions and includes a numerically greater population of snubbers than the ASME Section Xl program. Examination and testing of the snubbers in accordance with both ASME Section XI, and the BFN Unit 2 TRM would result in a duplication of effort utilizing different standards and require the preparation of a separate program and associated procedures. This would result in additional cost and unnecessary radiological exposure. In addition, the personnel performing snubber visual examination would also be required to be certified in accordance with the American Society of Nondestructive Examination (ASNT) SNT-TC-1A "Personnel Qualification and Certification in Nondestructive Testing" and ANSI/ASNT CP-1 89. This is an additional qualification and certification as compared to the task training qualification required to perform the TRM required examinations and testing of snubbers. The existing TRM program for examination and testing of snubbers was promulgated and approved by the NRC when the requirements were included in the BFN Unit 2 Technical Specifications. These requirements were relocated to the TRM during the conversion of the BFN Unit 2 Technical Specifications to the Improved Standard Technical Specifications.

Implementing the requirements for examination and testing snubbers of ASME Section XI, 2004 Edition would be a duplication of an existing program accepted by the NRC without a compensating increase in the level of quality and safety.

ALTERNATIVE EXAMINATION:

The BFN TRM, TR 3.7.4, requirements will be utilized for the examination and testing of snubbers for preservice, inservice, and repairs/replacement activities. The procedures utilized for these examinations are:

2-SI-4.6.H-1,"Visual Examination of Hydraulic and Mechanical Snubbers" 0-SI-4.6.H-2A, "Functional Testing of Mechanical Snubbers" 0-SI-4.6.H-2B, "Functional Testing of Bergen-Paterson, Anchor/Darling or Fronek Hydraulic Snubbers" 0-SI-4.6.H-2C, "Functional Testing of Bergen-Paterson Torus Dynamic Restraints" 0-SI-4.6.H-2E, "Functional Testing of Lisega Large Bore Torus Dynamic Restraint Snubbers" 0-SI-4.6.H-2F, "Functional Testing of Lisega Type 30 Hydraulic Snubbers"

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MPI-0-000-SNBOO2, "Hydraulic Shock and Sway Arrestor Bergen-Paterson Unit Disassembly and Reassembly" MPI-0-000-SNBOO4, "Instructions for Removing and Reinstalling Pacific Scientific Mechanical, Bergen-Paterson Grinnell Hydraulic, and Torus Dynamic Restraints" 0-TI-398, "Snubber Program Procedure" These examinations will include the pin-to-pin area inclusive of applicable snubbers. Testing of repaired and replaced snubbers will also be performed in accordance with TR 3.7.4.

Visual examination of repaired and replaced snubbers will be performed in accordance with MPI-0-OOOSNBOO4, "Instructions for Removing and Reinstalling Pacific Scientific Mechanical, Bergen-Paterson Grinnell Hydraulic, and Torus Dynamic Restraints."

Snubber examination and testing data will be maintained in accordance with the requirements of TR 3.7.4, the BFN corrective action program, SPP-3.1, "Corrective Action Program," and the implementing procedures (2-SI-4.6.H-1, O-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H 2E, 0-SI-4.6.H 2F, 0-TI-398, MPI-0-000-SNBOO2, and MPI-0-000-SNBO04).

The provisions of IWA-4000 will be met for the repair/replacement activities on snubbers by the current Repair and Replacement Program. TVA seeks relief from the examination and testing requirements of IWF-5200 required by IWF-5400 which requires snubbers installed, corrected, or modified by repair/replacement activities to be examined and tested in accordance with ASME/ANSI OM, Part 4. The examination and testing of snubbers [IWF-5200(a) and (b)] will be in accordance with the TRM Snubber Program.

In lieu of requirements of IWF-5200(a) and (b), the examination and testing requirements will be met by the Technical Requirements Manual (TRM) Snubber Program, prepared in accordance with the guidance contained in Generic Letter 90-09. As stated in TVA's request for relief, the areas inclusive of the pins back to the building structure and to the component/piping being supported will remain in the ASME Section XI examination boundary (ISI Program); therefore, IWF-5200(c) and IWF 5300(c) will be met by the Section XI Program.

JUSTIFICATION FOR THE GRANTING OF RELIEF:

The current program, as defined by TR 3.7.4, provides for a level of quality and safety equal to or greater than that provided by ASME/ANSI OM, part 4 ASME Section XI Code 1995 Edition, 1996 Addenda requirements.

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Examination, testing, repair and replacement of snubbers is currently performed in accordance with TR 3.7.4, which utilizes the guidance provided by NRC Generic Letter 90-09. ASME Section XI, 2004 Edition, has a different basis for the examination and testing plans. It is impractical to implement the requirements of both programs because of the resulting duplication of examination and testing efforts; e.g., different requirements for snubber quantities subject to examination or test; actually examined and/or tested and sample expansion requirements. This would result in additional cost and unnecessary radiological exposure. The existing TRM program for examination and testing of snubbers has been promulgated and approved by the NRC. The difference in the two programs could create confusion when selecting test samples, applying acceptance criteria, corrective actions, and examination schedules for failed snubbers.

This situation would increase the possibility of applying the wrong action due to conflicting requirements thus creating a nonconformance condition, an in-operability or even a violation of a TRM requirement.

To eliminate any misinterpretation or confusion in administering overlapping requirements for snubbers, and to remove the possibility of applying contradicting requirements to the same snubber(s), BFN proposes to examine and test BFN Unit 2 snubbers in accordance with BFN Unit 2 TR 3.7.4.

Subarticle IWF-5400 of the 2004 Edition of the code provides the requirements for repair and replacement of snubbers to be in accordance with IWA-4000. IWF-5200 provides that examinations shall be performed in accordance with ASME/ANSI OM, Part 4. This requirement is implemented in TR 3.7.4 (i.e., TSR 3.7.4.6). This program requires replacement snubbers and snubbers that have repairs which might affect the functional test results, to be tested to meet the functional test criteria prior to installation.

Maintenance procedure MPI-0-000-SNBOO4 provides visual examination criteria for installation of a snubber after repair or replacement. The ASME Section X1 repair/replacement program at BFN documents the verification of acceptability for repairs and replacements per IWA-4160.

ASME Section Xl VT-3 certification required by personnel performing snubber visual examinations is an additional certification as compared with the TRM program training qualifications. Personnel performing the TRM required visual examinations are "process qualified" to perform the examinations and testing as required by the TRM and implemented by the referenced procedures. This training currently includes visual acuity and specific training on the requirements and acceptance criteria associated with procedure 2-SI-4.6.H-1.

0-TI-398 states that personnel performing the visual inspections of 2-SI-4.6.H-1 must meet the visual acuity requirements of ASME Section XI, 2001 Edition, 2003 Addenda, Paragraph IWA-2321.

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The training and documentation of BFN personnel responsible for snubber visual examination and testing specified in the TRM implementing procedures provides an acceptable level of quality and safety.

Because relief is sought from the ASME Section XI snubber examination and test requirements there will be no ASME Section XI snubber examination and test activities to require Authorized Nuclear Inservice Inspector (ANII) involvement. The BFN TRM snubber program does not require the use of an ANII for examination and test requirements. The ANII' will not be involved in the TRM required visual examination or testing activities performed in lieu of the ASME Code requirements. A BFN snubber program engineer provides the oversight of the TRM snubber program implementation for both the visual examination and functional testing. This oversight includes both review and evaluation of visual examination and functional testing data to ensure TRM requirements are met. The BFN snubber program engineer provides the oversight that ensures an acceptable level of quality and safety exist without ANII involvement in these activities. ANII involvement will be maintained in inservice repair and replacement snubber activities, as required by IWA-21 10(g) and (h) and implemented by the BFN ASME Section XI repair and replacement program.

ASME Section XI, 2004 Edition Subarticle IWA-6230 provides requirements for ISl and IST documentation for snubbers in the framework of a summary report.

Under the alternate requirements for snubbers, there will be no ASME Section XI ISI and IST documentation to include in a summary report.

TR 3.7.4 is implemented by surveillance instructions 2-SI 4.6.H-1, 0-SI-4.6.H-2A, O-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H 2E, and O-SI-4.6.H 2F and maintenance instructions MPI-0-000-SNBOO2, and MPI-0-000-SNBOO4. These instructions are written and approved in accordance with the TVA Nuclear Quality Assurance Program, includes data sheets for documenting the visual examination and functional test data and results, and provides for documentation of non conforming results and evaluation of those results. The completed data sheets are Quality Assurance records and are controlled and maintained in accordance with the TVA Nuclear Quality Assurance Program and the BFN QA records program. These records are available onsite for review and inspection. The QA records documenting snubber visual examinations and functional tests provide an acceptable level of quality and safety when compared to the requirements of ASME Section XI and OM 1987, Part 4.

Completed data packages from O-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H-2E, and 0-SI-4.6.H-2F are permanent plant records prepared and stored in accordance with the TVA Nuclear Quality Assurance Program requirements.

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TR 3.7.4 is implemented by surveillance instructions 2-SI-4.6.H-1, 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H 2E, 0-SI-4.6.H 2F and maintenance instructions MPI-0-000-SNB002 and MPI-0-000-SNB0O4. These instructions are written and approved in accordance with the TVA Nuclear Quality Assurance Program (NQAP).

They include data sheets for documenting the visual examination and functional test data and results, provide for documentation of nonconforming results and evaluation of those results.

The completed data sheets are Quality Assurance (QA) records and are controlled and maintained in accordance with the TVA NQAP and BFN QA records program. These records are available onsite for review and inspection.

The alternate ISI and IST program, including the generated QA records documenting snubber ISI and IST provides an acceptable level of quality and safety when compared to the requirements of ASME Section XI, 2004 Edition.

Based on the justification provided, TVA's examination and testing of BFN snubbers, in accordance with TR 3.7.4 will provide an acceptable level of quality and safety. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), TVA requests that relief be granted from the ASME Section XI, 2004 Edition requirements related to ISI and IST of snubbers.

TVA's relief is consistent with request for relief 1-ISI-18 submitted by TVA letters dated May 27, 2008 and October 22, 2008, for the BFN Unit 1 Second Ten Year Inservice Inspection Interval. The NRC staff approved the request for relief by letter dated April 2, 2009.

Attachment K contains a comparison of the OM Part 4 requirements and the BFN TRM alternatives.

Based on the information provided above, TVA considers that the alternate BFN Unit 2 program, in accordance with TR 3.7.4, for the examination and testing of snubbers will provide an acceptable level of quality and safety.

Attachments:

Attachment A - BFN Unit 2 Technical Requirements Manual, TR 3.7.4, Snubbers Attachment B - BFN Technical Instruction, 0-TI-398, Snubber Program Procedure Attachment C - BFN Surveillance instruction, 2-SI-4.6.H-1, Visual Examination of Hydraulic and Mechanical Snubbers Attachment D - BFN Mechanical Preventative Instruction, MPI-0-000-SNBOO4, Removal and Reinstallation of Snubbers

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Attachment E - BFN Surveillance Instruction, 0-SI-4.6.H-2A, Functional Testing of Mechanical Snubbers Attachment F - BFN Surveillance Instruction, 0-SI-4.6.H-2B, Functional Testing of Bergen-Paterson, Anchor/Darling or Fronek Hydraulic Snubbers Attachment G - BFN Surveillance Instruction, 0-SI-4.6.H-2C, Functional Testing of Bergen-Paterson Torus Dynamic Restraints Attachment H - BFN Surveillance Instruction, 0-SI-4.6.H-2E, Functional Testing of Lisega Large Bore Torus Dynamic Restraint Snubbers Attachment I - BFN Surveillance Instruction, 0-SI-4.6.H-2F, Functional Testing of Lisega Type 30 Hydraulic Snubbers Attachment J - BFN Modification and Addition Instruction, MAI-4.10, Piping Clearance Instruction Attachment K - Comparison of TRM Program to ASME OM, Part 4 IMPLEMENTATION SCHEDULE:

This request for relief is applicable to the Fourth Ten Year Inservice Inspection Interval for BFN Unit 2 (May 25, 2011 through May 24, 2021)

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Attachment K Comparison of TRM Program to ASME OM, Part 4 OM Part 4, Section 1.5.6, Snubber Maintenance or Repair - Maintenance or repair activities which can alter the snubber's intended function shall be evaluated by considering the effects of maintenance or repair on the snubber's ability to meet the snubber examination criteria. Maintenance or repair activities that affect the ability of the snubber to satisfy its intended function shall be completed in accordance with written procedures. Snubbers which undergo maintenance or repair activities which could alter the snubber's ability to perform its intended function shall be examined and tested in accordance with the applicable requirements of paragraphs 2.3.1.2 and 3.2.1.1. The requirements selected shall ensure that the function(s) which may be affected are verified by the examination and tests to be acceptable.

For BFN Unit 2, the alternative requirements for Snubber Maintenance or repair is provided as follows.

TSR 3.7.4.6 states verify replacement snubbers and snubbers having repairs which might affect the functional test results meet the test criteria of TSR 3.7.4.2.

a. These snubbers shall have met the acceptance criteria subsequent to their most recent service; and
b. The functional test must have been performed within the 12 months prior to being installed in the unit.

The frequency of TSR 3.7.4.6 is once prior to installation in the unit for each replacement snubber and each snubber which has repairs which might affect functional test results.

0-TI-398, Section 7.18, Rebuilding of Hydraulic Snubbers states: Rebuilding of hydraulic snubbers shall be performed by task qualified and trained persons. Hydraulic snubbers shall be rebuilt in accordance with the MPI-0-000-SNBOO2, Hydraulic Shock and Sway Arrestor Bergen-Paterson, Anchor/Darling, Fronek Unit Disassembly and Reassembly, as appropriate.

MPI-0-000-SNBOO2, Section 6.2 states: Rebuilt snubber shall pass functional test criteria in O-SI-4.6.H-2B.

OM Part 4 Section 2.1.1a) There are no visible signs of damage or impaired operability as a result of storage, handling or installation.

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TSR 3.7.4.1 states Perform visual inspection of required snubber(s) based on the criteria of Table 3.7.4-1 for each population or category to verify:

a. No visible indications of damage or impaired OPERABILITY;
b. Attachments to the foundation or supporting structure are functional; and
c. Fasteners for the attachment of the snubber to the component or system and to the snubber anchorage are functional. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic.

The proposed alternative does not differentiate between damage or impaired operability storage, handling, of installation only that there is no visible indications of damage or impaired operability.

The alternate requirements in the BFN Surveillance Instruction (SI) 2-SI-4.6.H-1, , "Visual Examination Checklist for All Snubbers" checklist Item A reads: "The snubber has no visible indications of damage or impaired operability" and specifically includes observing the exposed parts of the snubber for broken parts, deformation, or other damage that could result in unacceptable performance.

OM Part 4 Section 2.1.1b) the snubber load rating, location, orientation, position setting, and configuration (attachments, extensions, etc.) are in accordance with design drawings and specifications. Installation records (based on physical inspections) verifying the snubbers were installed according to design drawings and specifications shall be acceptable in meeting this requirement.

TR 3.7.4. does not specifically address verification of the snubber load rating, location, orientation, position setting, and configuration (attachments, extensions, etc.) are in accordance with design drawings and specifications. However, these requirements are fulfilled by TR 3.7.4 program procedures.

The BFN alternate requirements in SI 2-SI-4.6.H-1 are as follows:

Section 7.2 G states "The snubbers are assigned a UNID number in an appropriate tracking program, which provides current and historical information for a specific snubber or support location." Section 7.2 H states "The snubbers are listed in Appendix A by exam number."

Appendix A provides additional information such as snubber drawing no., type/size, support number, and location. Snubber drawing number may be used to access the design drawing of a snubber, showing the location plan, material description, orientation, configuration (to include attachments, extensions and others), design requirements such as design travel/thermal movements and position settings. Section 7.2 I of the SI states that "A unique snubber/support number is given to each snubber location on a system."

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MPI-0-000-SNBOO4, Attachment 1, Note 2 states: "Document removal and reinstallation data for the snubber on Attachment 4." Snubber data listed in Attachment 4 relevant to these requirements includes: As-Found Position Indication Reading (for mechanical type),

As-Found Index (Plunger) Reading, As-Found Piston Reading and others (for hydraulic type).

Collection and recording data in Attachment 4 is done prior to removal of the snubber for functional testing and after reinstallation.

OM Part 4 Section 2.1.1c) adequate swing clearance is provided to allow snubber movement.

TSR 3.7.4.1 requires a visual inspection of required snubber(s) based on the criteria of Table 3.7.4-1 for each population or category to verify: a. No visible indications of damage or impaired OPERABILITY; b. Attachments to the foundation or supporting structure are functional; and c. Fasteners for the attachment of the snubber to the component or system and to the snubber anchorage are functional. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. This inspection is conducted in accordance with 2-SI-4.6.H-1.

The BFN alternate states that the item is to be observed in performance of visual SI 2-SI-4.6.H-1. Attachment 2 of the SI contains the following verifications: Centerline of the clamp assembly and structural attachment offset (i.e., a misalignment with the snubber axis exists) by no greater than plus or minus 6 degrees based on the clearances between the rod eyes, paddles, and the attachment clevis. Contact of these parts, which produces a side load on the snubber is unacceptable. Observe spacers are installed on each side of the snubber eye to reduce the misalignment and or binding. Space shall not exceed 1/16 inch on either side or 1/8 inch total. Observe for evidence of torsional binding (i.e. mechanical snubber twisted along its axis by the pipe clamp and structural attachments).

OM Part 4 Section 2.1.1d) if applicable, fluid is at the recommended level and fluid is not leaking from the snubber system.

TSR 3.7.4.1 requires a visual inspection of required snubber(s) based on the criteria of Table 3.7.4-1 for each population or category to verify: a. No visible indications of damage or impaired OPERABILITY; b. Attachments to the foundation or supporting structure are functional; and c. Fasteners for the attachment of the snubber to the component or system and to the snubber anchorage are functional. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. This inspection is conducted in accordance with 2-SI-4.6.H-1.

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The BFN Alternate fluid level acceptability check for hydraulic snubbers is included in the visual SI 2-SI-4.6.H-1 data sheets or Attachments 3, 5, 6, and 10. Immediate notification of the Snubber Engineer (SE)/SE Designee is required. A Work Order to remove the snubber and take to the snubber test facility to have functional test performed is directed by the applicable Attachment. Bergen-Paterson Torus Dynamic Restraint found with unacceptable fluid levels are considered inoperable and the Shift Manager/Unit Supervisor notified.

Examination of the snubber for location and cause of leaks is performed. Leakage locations are recorded in Remarks. The Snubber Engineer/SE Designee performs an evaluation and documents the unacceptable condition on Attachment 8 as applicable and initates appropriate corrective action.

MPI-0-000-SNBOO4, Attachment 1, Section 1.0[1] states: "perform visual inspection of the snubber for any visible damage or fluid leakage." Further, Section 1.0[8.2] states: "if the fluid level is unacceptable, but not empty, add GE SF 1154 silicon fluid using a hydraulic fluid gun until the fluid level reading is at or approximately the same as the piston rod extension given above. Notify the Snubber Engineer/Designee to take appropriate action."

OM Part 4 Section 2.1.1e) structural connections such as pins, bearings, studs, fasteners, and other connecting hardware such as locknuts, tabs, wire and cotter pins are installed correctly.

TSR 3.7.4.1 requires a visual inspection of required snubber(s) based on the criteria of Table 3.7.4-1 for each population or category to verify: a. No visible indications of damage or impaired OPERABILITY; b. Attachments to the foundation or supporting structure are functional; and c. Fasteners for the attachment of the snubber to the component or system and to the snubber anchorage are functional. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. This inspection is conducted in accordance with 2-SI-4.6.H-1.

The BFN Unit 2 alternate checklist items in SI 2-SI-4.6.H-1, Attachment 2, Item B states:

"Attachments to the foundation or supporting structure are functional" and verifies the following attributes: "Observe the exposed hanger structural steel, pipe clamps, base plates, lugs and other such plates of attachment for broken parts, deformation or other damage" and "Observe welds for visible damage at base plates, lugs, and other such points of attachment." , Item C states: "Fasteners for the attachment of snubber to the component and to the snubber anchorage are functional" and verifies the following attributes:

Observe to ensure the security of essential threaded fasteners such as anchorage bolts and pipe clamp bolts that are exposed.

0 Observe to ensure clevis bolts or pins are properly installed.

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Observe for the following attributes which may cause future problems but do not affect snubber operability or any acceptance criteria: pits, scratches, or rough places observed on the piston rod that would not contact the piston seal, cotter pins properly installed with legs bent sufficiently, but not completely, to prevent cotter pin from backing out, pivot pin retaining ring is properly installed, if required, and security locking devices (i.e., locking tabs or wire) on snubber attachment bolts properly installed, if required.

OM Part 4 Section 2.2, Preservice Thermal Movement Examination 2.2.1 Preservice Thermal Movement Examination Requirements. During initial system heatup and cooldown, snubber thermal movement for systems whose design operating temperature exceeds 250 degrees F (121degrees C) shall be verified as follows.

(a) During initial system heatup and cooldown at temperature plateaus specified by the Owner, record the thermal movement. Verify that the snubber movement during the thermal movement examination is within the design specified range. Any discrepancies or inconsistencies shall be evaluated to determine the movement acceptability prior to proceeding to the next specified plateau. The total thermal movement from cold to hot at full operating temperature shall be recorded. This value may be measured directly if maximum operating temperature was attained, or extrapolated from lower temperature readings. The cold or hot position setting shall be evaluated and adjusted if necessary to ensure adequate snubber clearance from fully extended or retracted positions.

(b) Verify that there is swing clearance at specified heatup and cooldown plateaus.

Preservice thermal movement examinations during inital system heatup and cooldown were conducted in accordance with BFN plant procedures during initial plant startup. Preservice thermal movement examination in accordance with this section are not required.

OM Part 4 Section 2.3.1, Inservice Examination Requirements; Visual Examination for impaired functional ability; Visual Examination Requirements to verify snubbers can carry load, snubbers do not restrict movement, and verification of special features required for actuation.

See discussion for OM Part 4 sections 2.1 .la through 2.1 .le.

OM Part 4, Section 2.3.1.2, Visual Examination Requirements states:

The snubber installation must meet the following requirements:

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b. Snubbers shall be installed in such condition that they do not restrict the thermal movement to the extent that unacceptable overstressing could develop in the pipe or other equipment that the installation is designed to protect or restrain. If no indication of binding, misalignment, or deformation is observed, the provisions of this requirement are considered to be satisfied.

TSR 3.7.4.1 requires a visual inspection of required snubber(s) based on the criteria of Table 3.7.4-1 for each population or category to verify: a. No visible indications of damage or impaired OPERABILITY; b. Attachments to the foundation or supporting structure are functional; and c. Fasteners for the attachment of the snubber to the component or system and to the snubber anchorage are functional. The discovery of loose or missing attachment fasteners will be evaluated to determine whether the cause may be localized or generic. This inspection is conducted in accordance with 2-SI-4.6.H-1.

The BFN alternative states that the item is to be observed in performance of visual SI 2-SI-4.6.H-1. Attachment 2 of the SI contains the following verifications: Centerline of the clamp assembly and structural attachment offset (i.e., a based on the. clearances between the rod eyes, paddles, and the attachment clevis. Contact of these parts, which produces a side load on the snubber is unacceptable. Observe spacers are installed on each side of the snubber eye to reduce the misalignment and or binding. Space shall not exceed 1/16 inch on either side or 1/8 inch total. Observe for evidence of torsional binding (i.e. mechanical snubber twisted along its axis by the pipe clamp and structural attachments).

c. Special features required for the actuation of the snubber shall be verified. For example, fluid supply or content for hydraulic snubbers shall be observed.

Observation that the fluid level is equal to or greater than the minimum amount which is sufficient for actuation at its operating extension is considered to satisfy the provisions of this requirement for hydraulic snubbers. If the fluid is less than the minimum amount, the installation is to be identified as unacceptable unless a test is performed establishing that the performance of the snubber is within specified limits. Tests shall be performed in accordance with paragraphs 3.2.1.1(b) and 3.2.1.1(c) and the initial test shall start with the piston at the as-found setting and be performed in the extension (tension) direction.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 135 of 205 Attachment 7 (Page 14 of 37) 0-TI-398, Appendix A, under Alternate Examinations states: The BFN Unit 2 TRM, TR 3.7.4, "Snubbers", requirements will be utilized for the examination and testing of snubbers for preservice, inservice, and repair/replacement activities. The procedures utilized for these examinations are 2-SI-4.6.H-1, "Visual Examination of Hydraulic and Mechanical Snubbers;"

O-SI-4.6.H-2A, "Functional Testing of Mechanical Snubbers;" 0-SI-4.6.H-2B, "Functional Testing of Bergen-Paterson, Anchor/Darling or Fronek Hydraulic Snubbers", 0-SI-4.6.H-2C, "Functional Testing of Bergen-Paterson Torus Dynamic Restraints", 0-SI-4.6.H-2E, "Functional Testing of Lisega Large Bore Torus Dynamic Restraints", and O-SI-4.6.H-2F, "Functional Testing of Lisega Type 30 Hydraulic Snubbers"; MPI-0-000-SNBOO2, "Hydraulic Shock and Sway Arrestor Bergen-Paterson Unit Disassembly and Reassembly;" and MPI-0-000-SNB004, "Instructions for Removing and Reinstalling Pacific Scientific Mechanical, Bergen-Paterson Hydraulic, Grinnell Hydraulic, and Bergen-Paterson or Lisega Torus Dynamic Restraints Snubbers." These examinations will include the pin-to-pin area inclusive of applicable snubbers.

When testing is performed by the manufacturer or a qualified vendor, the acceptance criteria contained in BFN surveillance procedures 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H)-2E, and 0-SI-4.6.H-2F are used to ensure testing performed meets program requirements.

Testing of repaired and replaced snubbers will also be performed in accordance with TR 3.7.4.

Visual examination of repaired and replaced snubbers will be performed in accordance with 2-SI-4.6.H-1, "Visual Examination of Hydraulic and Mechanical Snubbers".

Snubber examination and testing data will be maintained in accordance with the requirements of TR 3.7.4, the BFN corrective action program, SPP-3.1, and the implementing procedures (2-SI-4.6.H-1, 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H 2E, O-SI-4.6.H 2F, 0-TI-398, MPI-0-000-SNB002, and MPI-0-000-SNBO04).

The areas inclusive of the pins back to building structure and to the component/piping being supported will remain in the ASME Section Xl examination boundary.

OM Part 4 Section 2.3.2, Inservice Examination Frequency TSR 3.7.4.1 requires performance of a visual inspection of snubbers with the frequency based on TRM Table 3.7.4-1, "Snubber Visual Inspection Interval". Also, see discussion for OM Part 4 section 3.2.2.

OM Part 4 Section 2.3.3, Inservice Examination Sample Size The following criteria shall be utlized to establish inservice examination sample size.

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(a) The initial inservice examination of snubbers required by para. 2.3.2.1 shall include all (100%) of the snubbers of all groups as may have been established under the provisions of para. 1.6.

(b) Subsequent inservice examinations shall include all (100%) of the snubbers of all groups.

After two successive examination intervals, at the maximum time interval allowed in para. 2.3.2.2, the sample size for the next required examination of the group may be reduced if justified by the Owner and accepted by the regulatory authority having jurisdiction over the facility.

The inservice examination sample size for BFN Unit 2 is established in accordance with TRM Table 3.7.4 1. TSR 3.7.4.1 requires performance of a visual inspection of snubbers with the frequency based on TRM Table 3.7.4 1. Also, see discussion for OM Part 4 section 3.2.2.

OM Part 4, Section 2.3.4.1, Failure Evaluation. Snubbers which fail to meet the examination acceptance criteria shall be evaluated to determine the cause of the unacceptability.

TRM section TSR 3.7.4.1 states: Snubbers which appear inoperable as a result of visual inspection shall be classified unacceptable and may be reclassified acceptable for the purpose of establishing the next visual inspection interval, provided that (1) the cause of the rejection is clearly established and remedied for that particular snubber and for other snubbers irrespective of type that may be generically susceptible; and (2) the affected snubber is functionally tested in the as-found condition and determined OPERABLE per the criteria of TSR 3.7.4.2. A review and evaluation shall be performed and documented to justify continued operation with an unacceptable snubber. If continued operation cannot be justified, the system or train shall be declared inoperable.

Additionally, TRM section TSR 3.7.4.3 states: A failure analysis shall be made of each failure to meet the functional test acceptance criteria of TSR 3.7.4.2 to determine the cause of the failure.

OM Part 4, Section 2.3.4.2, Functional Test Evaluation - The snubber(s) that is found to be unacceptable as a result of inservice examination may be tested in accordance with requirements of paragraph 3.2, provided the testing can resolve the unacceptable condition. Results which satisfy the operability test criteria may be used to re-categorize the snubber(s) as acceptable.

See discussion for OM Part 4, Section 2.3.4.1 above.

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OM Part 4, Section 2.3.4.3, Examination Failure Mode Groups - Unacceptable snubbers shall be categorized into examination failure mode groups. An examination failure mode group shall include all unacceptable snubbers which have a given failure and all other snubbers subject to the same failure, except as permitted to be considered separately under paragraph 1.6. The examination failure mode groups shall be distinct for examination purposes from any testing failure mode groups. The following examination failure mode groups shall be used:

(a) Design/manufacturing (b) Application induced (c) Maintenance/repair/installation (d) Isolated (e) Unexplained The BFN snubber program does not categorize examination failures into failure mode groups, although as reported in the above response to OM Part 4, Section 2.3.4.1, the cause of all examination failures is established and remedied for that particular snubber and for other snubbers that may be generically susceptible.

In accordance with 0-TI-398, Snubber Program Procedure, snubber failures are classified as follows:

Location - A failure of a snubber(s) resulting from environmental conditions.

Examples include failure of a snubber due to excessive heat, failure due to excessive local system vibration or failure due to other localized anomalies.

Manufacturing - A failure resulting from a potential defect in manufacturing.

Examples include incorrectly assembled snubbers, inclusion of incorrect piece parts and failure of a snubber(s) due to improper or incorrect maintenance or repair practices performed by the vendor.

Design - A failure resulting from an error in the design. This classification would include incorrectly sized snubbers, snubbers provided with insufficient travel, snubbers with insufficient swing margins and design misapplications or errors.

Unknown - A failure that cannot be categorized as location, manufacturing, design or other. This includes all failures for which the cause of failure cannot be determined.

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Other - When the unknown cause of the failure cannot be categorized as location, manufacturing or design. This classification would include snubbers that fail due to end of service life. Additional examples include failure of snubbers due to plant transients, misapplication of rigging loads to the snubber and incidental interaction of tools or equipment on the snubber.

0-TI-398 sections 7.19.2A.3 and 7.19.2A.4 provide the following scope expansion and additional testing requirements due to the failure mode:

If the failure was found as part of the Technical Requirements Manual initial or any subsequent samples the following is the minimum required testing expansion:

a. Expand the sample by 10% of the remaining snubbers in the subgroup, and
b. Ifthe cause of the failure is determined to be a system transient, test all snubbers on that system, which could have been affected by the transient, and
c. Ifthe cause is a generic manufacturing defect, test all snubbers of that same type, and
d. Identify and test all other snubbers suspected of the same failure mode.

If the failure was found outside of the Technical Requirements Manual initial or any subsequent samples (i.e., other maintenance activity, testing directed/requested by the Snubber Engineer, Service Life Monitoring testing, testing of snubbers on non-Technical Specification or Technical Requirements Manual systems, etc.) the following is the minimum required testing expansion:

a. Ifthe cause of the failure is determined to be a system transient, test all snubbers on that system, which could have been affected by the transient, and
b. Ifthe cause is a generic manufacturing defect, test all snubbers of that same type, and
c. Identify and test all other snubbers suspected of the same failure mode.

OM Part 4, Section 2.3.4.4, Examination Failure Mode Group Boundaries. Once an examination failure mode group has been established, any snubber(s) in that examination failure mode group(s) will not be part of the examination group(s) from which the snubber originated except as noted in paragraph 2.3.4.5 below. The new examination failure mode group will remain as defined until:

(a) the examination failure mode group has reached the maximum time interval

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 139 of 205 Attachment 7 (Page 18 of 37) allowed (b) replacement/modification action in accordance with paragraphs 2.3.5.1(a) or (c), or paragraph 2.3.5.2 provides an examination failure group with all acceptable snubbers.

See discussion for OM Part 4, Section 2.3.4.3 above.

OM Part 4, Section 2.3.4.5, Snubbers in More Than One Failure Group - Any snubber(s) which is in more than one examination failure mode group, the examination schedule for that snubber will be determined by the examination failure mode group with the shortest examination schedule.

See discussion for OM Part 4, Section 2.3.4.3 above.

OM Part 4 Sections 2.3.4.3 through 2.3.5.4, which address Failure Mode Groups and Corrective Actions See discussion for OM Part 4 section 3.2.4.2.

OM Part 4 Sections 2.3.5.5, Supported Component(s)/System Evaluation.

See discussion for OM Part 4 section 3.2.4.1.

OM Part 4 Section 2.3.4.1, Failure Evaluation; Section 2.3.4.2, Functional Test Evaluation See discussion for OM Part 4 section 2.4 d.

OM Part 4 Section 2.4 a) checklists verifying preservice and inservice examination, fluids levels, and as-found conditions. Appendix A of this Part represents items normally included in a checklist (as follows):

  • Rotated reservoirs (hydraulic fluid could not reach valve blocks)

BFN Unit 2 does not have rotated reservoirs as a specific snubber checklist item. Reservoir inventory is verified in 2-SI-4.6.H-1 for the hydraulic snubber models installed on BFN Unit 2.

The models installed have pressurized hydraulic reservoirs that allows mounting in any spatial orientation.

0 Piston shaft painted, which could cause a frozen condition

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The BFN Unit 2 alternate requirements are provided in SI 2-SI-4.6.H-1, Attachment 2, Section A that contains the following checklist item: "The snubber has no visible indications of damage or impaired operability. This item includes the following verifications: "Observe for any signs of overstressing, Observe the exposed parts of the snubber for broken parts, deformation or other damage, such as weld arc strikes, paint, weld slag, adhesive, or other deposits on piston rod or support cylinder that could result in unacceptable snubber performance, Observe to see if there is any evidence that a snubber has experienced a potentially damaging transient since the last examination, and Observe the snubber and piston rod for excessive corrosion, solid deposits, which could impair operability of the snubber."

  • Units installed upside down The BFN Unit 2 Alternate requirements are provided in SI 2-SI-4.6.H-1, Section 6.0 A. 4 which states: "The snubber has the proper orientation, and adequate fluid level, if applicable." MPI-0-000-SNBOO4, Attachment 3, Section 2.0[11] verifies for Lisega Torus Dynamic Restraints, that the 2-way ball valve positioner is in the vertical position.

0 Sight glass broken The BFN Unit 2 Alternate requirements are provided in SI 2-SI-4.6.H-1, Attachment 2, Section A contains the following verification: "The snubber has no visible indications of damage or impaired operability. Observe the exposed parts of the snubber for broken parts, deformation or other damage, such as weld arc strikes, paint, weld slag, adhesive, or other deposits on piston rod or support cylinder that could result in unacceptable snubber performance."

Installed with preset shipment screws for shipping (screws must be removed before service )

The BFN Unit 2 functional testing of snubbers demonstrates free/unrestricted snubber movement in tension and compression directions. For a snubber to move freely, the preset shipment screws for shipping must be removed before the test. There is no preset shipment screw for shipping to remove on a Lisega Torus Dynamic Restraints.

Hydraulic fluid lines for snubber remote reservoir placed too close to hot pipe causing the lines to burst BFN currently has no snubbers with remote reservoirs. The BFN Unit 2 Alternate requirements are provided in MAI-4.2A Piping/Tubing Support Installation Data Sheet which documents clearances per MAI-4.10 applicable to SR, QR, or NQR supports in Category 1 Structures. MPI-0-000-SNBOO4, Attachment 1, section 2.1, Note (2) states that: "Any required relocation of the strut attachment to clear an interference should be brought to the attention of Site Engineering (Civil)."

\_

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  • Snubber placed in wrong location The BFN Unit 2 alternate requirements are provided in SI 2-SI-4.6.H-1, Section 7.2.G: "The snubbers are assigned a UNID number in an appropriate tracking program, which provides current and historical information for a specific snubber or support location." Section 7.2.H; "The snubbers are listed in Appendix A by exam number." Note: Appendix A also provides a specific location of a snubber with a unique identification number (UNID).
  • Clevis pins not attached to anchor The BFN Unit 2 alternate requirements are provided in SI 2-SI-4.6.H-1, Attachment 2, Section C which states: "Fasteners for the attachment of the snubber to the component and to the snubber anchorage are functional. Observe to ensure the security of essential threaded fasteners such as anchorage bolts and pipe clamp bolts that are exposed. Observe to ensure clevis bolts or pins are properly installed."

0 Snubber not installed at correct location The BFN Unit 2 Alternate requirements are provided in 2-SI-4.6.H-1, Section 7.2.G which states: "The snubbers are assigned a UNID number in an appropriate tracking program, which provides current and historical information for a specific snubber or support location.

Section 7.2.H states: "The snubbers are listed in Appendix A by exam number." Note:

2-SI-4.6.H-1, Appendix A also provides a specific location of a snubber with a unique identification number (UNID).

0 Bent or scored piston rod The BFN Unit 2 Alternate requirements are provided in 2-SI-4.6.H-1, Attachment 2, Section A which states: "The snubber has no visible indications of damage or impaired operability.

Observe the exposed parts of the snubber for broken parts, deformation or other damage, such as weld arc strikes, paint, weld slag, adhesive, or other deposits on piston rod or support cylinder that could result in unacceptable snubber performance."

0 Welding arc strikes The BFN Unit 2 alternate requirements are provided in 2-SI-4.6.H-1, Attachment 2, Section A which states: "The snubber has no visible indications of damage or impaired operability.

Observe the exposed parts of the snubber for broken parts, deformation or other damage, such as weld arc strikes, paint, weld slag, adhesive, or other deposits on piston rod or support cylinder that could result in unacceptable snubber performance."

0 Lubrication of pivot points

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The BFN Unit 2 alternate requirements are provided in MPI-0-000-SNBOO4, Attachment 2, Section 2.0[4] which states: "Apply anti-seize thread lubricant to the surface of the pivot pins and to all threaded fasteners that are being installed."

  • Abnormal spherical bearing position The BFN Unit 2 Alternate requirements are provided in MPI-0-000-SNBOO4 Section 7.2[3]

which states: "Ifa spherical bearing is found to be dislodged from the paddle housing, REINSERT by carefully pressing or tapping on the outer race. Use a Bergen-Paterson bearing installation tool or an appropriate sized pipe to assure proper alignment."

MPI-0-000-SNBOO4, Section 7.2[3.2] states: After the spherical bearing has been reinserted into the paddle if required, USE a center punch and MOVE approximately 1/32 inch away from the exterior of the spherical bearing race and with as little force as possible MAKE four punch indentations equally spaced around the race at approximately 90 degrees, on both sides of the paddle.

0 Protective coverings or plugs removed (after shipping or maintenance)

Snubbers received from the warehouse are free of protective coverings or plugs removed and ready for examination and testing as required.

  • Fluid level indicators and/or position indicators accessible for visual inspection The BFN Unit 2 Alternate requirements are provided in MPI-0-000-SNBOO4, Attachment 1 Section 2.21 10] which states: "At the end of each installation, CHECK each unit as a precaution for the following information."

" Piston rod extension dimension.

  • Fluid level indicator reading.
  • Whether or not fluid was added to bring unit to proper level.
  • Visible condition of the unit.
  • Condition of the strut assembly with particular attention to the clamp and the bolting tightness.

No visible corrosion or mechanical defects of working parts or surfaces

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The BFN Unit 2 alternate requirements are provided in 2-SI-4.6.H-1, Attachment 2, Section A which states: "The snubber has no visible indications of damage or impaired operability.

Observe the exposed parts of the snubber for broken parts, deformation or other damage, such as weld arc strikes, paint, weld slag, adhesive, or other deposits on piston rod or support cylinder that could result in unacceptable snubber performance. Observe the snubber and piston rod for excessive corrosion, solid deposits, which could impair operability."

OM Part 4 Section 2.4 b) examination records All examination records are documented in Attachments 2, 3, 4, and 6 of the 2-SI-4.6.H-1 and in Attachment 4 of MPI-0-000-SNBOO4. Visual examination records performed to facilitate testing are also documented in the appropriate attachments of applicable Sis 0-SI-4.6.H-2A, O-SI-4.6.H-2B, O-SI-4.6.H-2C, O-SI-4.6.H 2E, O-SI-4.6.H 2F.

OM Part 4 Section 2.4 c) thermal movement inspection records The stroke setting (As-Found/As-Left) which relates to thermal movement (shown in design drawings) for a given snubber are documented in Attachment 4 of MPI-0-000-SNB004, and Attachments 2, 3, 4, and 6 of 2-SI-4.6.H-1 and in the appropriate attachments of the applicable Sis O-Sl-4.6.H-2A, 0-SI-4.6.H-2B, O-Sl-4.6.H-2C, O-SI-4.6.H 2E, O-SI-4.6.H 2F.

OM Part 4 Section 2.4 d) nonconformance and corrective action required to be completed during the preservice and inspection interval The BFN Unit 2 alternate requirements are provided in 2-St-4.6.H-1, Section 7.2[F] which states: Evaluation sheets (Attachments 5, 6, and 7) shall be prepared and submitted with the data package, as appropriate, by the Snubber Engineer/Designee for each degraded or inoperable snubber identified by the performance of this instruction.

Section 7.3.1[1] states: "Ifany snubber is determined to be inoperable, Site Engineering Civil should initiate a Problem Evaluation Report (PER). Mechanical Maintenance Group (MMG)

Planning should write a minor maintenance Work Order (WO) to perform the necessary repairs required to return the inoperable snubber to operable status." The checklist in , Sections A, B, and C states: "Responses marked (UNAC) are unacceptable and require immediate notification of the Snubber Engineer/SE Designee at the time of discovery. Handling of deficiencies shall be completed in accordance with SPP-8.1 and SPP-3.1 ."

OM Part 4 Section 3.1 Preservice Operability Testing 3.1.1 Preservice Operability Testing Requirements.

Preservice operability testing shall be performed on all snubbers. Testing may be at the manufacturer's facility. The testing shall verify that:

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(a) the force that will initiate motion (breakaway force), the force that will maintain low velocity displacement (drag force), or both, as required by the documents of para.

1.5.2, are within specified limits, both in tension and compression; (b) activation is within the specified range of velocity or acceleration in both tension and compression; (c) release rate, where applicable, is within the specified range in tension and compression. For units specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be demonstrated.

For BFN Unit 2, the alternative requirements for Snubber Maintenance or repair is provided as follows.

TSR 3.7.4.6 states verify replacement snubbers and snubbers having repairs which might affect the functional test results meet the test criteria of TSR 3.7.4.2.

a. These snubbers shall have met the acceptance criteria subsequent to their most recent service; and
b. The functional test must have been performed within the 12 months prior to being installed in the unit.

When testing is performed by the manufacturer or a qualified vendor, the acceptance criteria contained in BFN surveillance procedures 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H)-2E, and 0-SI-4.6.H-2F are used to ensure testing performed meets program requirements.

OM Part 4 Section 3.2.1.1 a) the force that will initiate motion (breakaway force), the force that will maintain low velocity displacement (drag force), or both, as required by the documents of Paragraph 1.5.2 (Procedures and Instructions), is within specified limits, both in tension and compression Drag tests performed on snubbers are in accordance with 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, O-SI-4.6.H 2E, 0-SI-4.6.H 2F, generally using the STB 200 Test Bench. Drag test result/computer printout is a graph of Velocity (ipm), Position (inches) and Force (Ibs).

0-SI-4.6.H-2A, Section 6.OB states: The snubber functional test shall verify that:

1. "As-Found"
a. Activation (restraining action) occurs, in both tension and compression. Snubbers with activation values less than or equal to 0.04 g's are acceptable. The measured value shall be rounded to the second decimal place. Activation values greater than 0.04 g's shall be considered as a failure (inoperable snubber), require a failure analysis, and be evaluated for additional scope or expansion testing.

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b. The drag force shall not exceed the maximum acceptable drag force from Table 6.2-1 of the procedure.
2. "As-Left"
a. Activation (restraining action) occurs, in both tension and compression. Snubbers with activation values less than or equal to 0.02g's are acceptable.
b. The drag force shall not exceed 3% of the snubber rated load, except for PSA 1/4 the drag force shall not exceed 32.5 lbs.

Test data/computer printouts are documented in Attachment 2.

0-SI-4.6.H-2B, Section 6.OB states: The snubber functional test shall verify that:

1. "As-Found"
a. Corrected activation occurs in both directions of travel at a piston velocity greater than or equal 1 inch/minute and less than or equal 30 inches/minute.
b. Bleed takes place after activation of the snubber in both the tension and compression directions.
c. Drag forces do NOT exceed 2.0 percent of the snubber's rated load.
2. "As-Left"
a. Corrected activation occurs in both directions of travel at a piston velocity greater than or equal 5 inches/minute and less than or equal 20 inches/minute.
b. Corrected bleed takes place after activation of the snubber in both tension and compression and shall be within the ranges shown in Appendix C of the procedurefor each size of snubber.
c. Drag forces do NOT exceed 2.0 percent of the snubber's rated load.

Test data/computer printouts are documented in Attachment 2 of the procedure.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 146 of 205 Attachment 7 (Page 25 of 37) 0-SI-4.6.H-2C, Section 6.0 states: Responses which fail to meet the As Found Acceptance Criteria require immediate notification of the Snubber Engineer/SE designee at the time of failure.

1. "As-Found"
a. Activation shall occur in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed shall take place after activation of the snubber, in both tension and compression.
c. Drag forces shall be less than or equal to 15,000 pounds.
d. Activation, bleed, or drag force acceptance criteria may be other than that described in the steps above, if approved by Site Engineering on Attachment 5 of the procedure.
2. "As-Left"
a. Activation shall occur in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed shall take place after activation of the snubber, in both tension and compression and be greater than or equal 1 inch/minute and less than or equal 10 inches/minute.
c. Drag force shall be less than or equal to 15,000 pounds.
d. Activation, bleed, or drag force acceptance criteria may be other than that described in the steps above, if approved by Site Engineering on Attachment 5 of the procedure.
e. There shall be no loose or missing fasteners for attachment of the "As Found" dynamic restraint(s) to the component or the anchorage. Otherwise, complete Attachment 6 of the procedure.
f. The stroke setting shall be within the limits shown on the design drawing.

0-SI-4.6.H-2E, Section 6.0B states: The snubber functional test shall verify that:

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1. "As-Found"
a. Activation occurs in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed takes place after activation of the snubber, in both tension and compression directions.
c. Drag force does not exceed 15,000 pounds.
2. "As-Left"
a. Activation occurs in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed takes place after activation of the snubber, in both tension and compression, and is greater than 0.28 inch/minute and less than 1.18 inches/minute.
c. Drag force does not exceed 15,000 pounds.

Test data/computer printouts are documented in Attachment 2 of the procedure.

Breakaway force is a test parameter printed in the test data for drag test performed, normally less than the maximum drag force.

0-SI-4.6.H-2F, Section 6.08 states: The snubber functional test shall verify that:

1. "As-Found"
a. Activation shall occur in both directions of travel at lockup velocity greater than or equal 4.72 inches/minute and less than or equal 14.17 inches/minute.
b. Bleed rate shall take place after activation in tension and compression direction and shall be between 0.47 inch/minute and 4.72 inches/minute at full rated load plus/minus 5 percent.
2. "As-Left"
a. Activation shall occur in both directions of travel at lockup velocity greater than or equal 4.72 inches/minute and less than or equal 14.17 inches/minute.
b. Bleed rate shall take place after activation in tension and compression direction and shall be between 0.47 inch/minute and 4.72 inches/minute at full rated load plus/minus 5 percent.

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c. Drag force shall be less than or equal 2.0 percent of rated load of the snubber, or 20 lbs. (whichever is greater).

Above parameters are based on ambient temperature between 65 and 75 degrees F.

Test data/computer printouts are documented in Attachment 2 of the procedure.

OM Part 4 Section 3.2.1.1 b) activation is within the specified range of velocity or acceleration in both tension and compression Acceleration tests performed on mechanical snubbers are in accordance with 0-SI-4.6.H-2A. LOCKUP and BLEED TEST performed on hydraulic type snubbers are in accordance with 0-Sl-4.6.H-2B, 0 SI 4.6.H 2C, 0-SI-4.6.H-2E and 0-SI-4.6.H-2F. These tests are generally done on the STB 200 Test Bench. Acceleration Test, Lockup and Bleed Test data/computer printouts are graphs of Velocity (ipm), Time (seconds) and Force (Ibs).

0-SI-4.6.H-2A, Section 6.OB states: The snubber functional test shall verify that:

1. "As-Found"
a. Activation (restraining action) occurs, in both tension and compression. Snubbers with activation values less than or equal to 0.04 g's are acceptable. The measured value shall be rounded to the second decimal place. Activation values greater than 0.04 g's shall be considered as a failure (inoperable snubber), require a failure analysis, and be evaluated for additional scope or expansion testing.
2. "As-Left"
a. Activation (restraining action) occurs, in both tension and compression. Snubbers with activation values less than or equal to 0.02g's are acceptable.

Test data/computer printouts are documented in Attachment 2 of the procedure.

0-SI-4.6.H-2B, Section 6.0B states: The snubber functional test shall verify that:

1. "As-Found"
a. Corrected activation occurs in both directions of travel at a piston velocity greater than or equal 1 inch/minute and less than or equal 30 inches/minute.

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2. "As-Left"
a. Corrected activation occurs in both directions of travel at a piston velocity greater than or equal 5 inches/minute and less than or equal 20 inches/minute.

Test data/computer printouts are documented in Attachment 2 of the procedure.

O-SI-4.6.H-2BC, Section 6.0 states: Responses which fail to meet the As Found Acceptance Criteria require immediate notification of the Snubber Engineer/SE designee at the time of failure.

1. "As-Found"
a. Activation shall occur in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed shall take place after activation of the snubber, in both tension and compression.
d. Activation, bleed, or drag force acceptance criteria may be other than that described in the steps above, if approved by Site Engineering on Attachment 5 of the procedure.
2. "As-Left"
a. Activation shall occur in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
b. Bleed shall take place after activation of the snubber, in both tension and compression and be greater than or equal 1 inch/minute and less than or equal 10 inches/minute.
d. Activation, bleed, or drag force acceptance criteria may be other than that described in the steps above, if approved by Site Engineering on Attachment 5 of the procedure.

Test data/computer printouts are documented in Attachment 2 of the procedure.

0-SI-4.6.H-2E, Section 6.OB states: The snubber functional test shall verify that:

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1. "As-Found"
a. Activation occurs in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.
2. "As-Left"
a. Activation occurs in both directions of travel at a piston velocity greater than or equal 6 inches/minute and less than or equal 25 inches/minute.

Test data/computer printouts are documented in Attachment 2 of the procedure.

0-SI-4.6.H-2F, Section 6.0B states: The snubber functional test shall verify that:

1. "As-Found"
a. Activation shall occur in both directions of travel at lockup velocity greater than or equal 4.72 inches/minute and less than or equal 14.17 inches/minute.
b. Bleed rate shall take place after activation in tension and compression direction and shall be between 0.47 inch/minute and 4.72 inches/minute at full rated load plus/minus 5 percent.
2. "As-Left"
a. Activation shall occur in both directions of travel at lockup velocity greater than or equal 4.72 inches/minute and less than or equal 14.17 inches/minute.
b. Bleed rate shall take place after activation in tension and compression direction and shall be between 0.47 inch/minute and 4.72 inches/minute at full rated load plus/minus 5 percent.

Above parameters are based on ambient temperature between 65 and 75 degrees F.

Test data/computer printouts are documented in Attachment 2 of the procedure.

OM Part 4, Section 3.2.1.1 c) - release rate, where applicable, is within the specified range in tension and compression. For units specifically required not to displace under continuous load, the ability of the snubber to withstand load without displacement shall be demonstrated.

0-SI-4.6.H-2B, Section 6.OB states: The snubber functional test shall verify that:

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1. "As-Found"
b. Bleed takes place after activation of the snubber in both the tension and compression directions.
2. "As-Left"
b. Corrected bleed takes place after activation of the snubber in both tension and compression and shall be within the ranges shown in Appendix C of the procedure for each size of snubber.

Test data BLEED (RELEASE) Rate are recorded in Attachment 2 of the procedure.

0-SI-4.6.H-2E, Section 6.OB states: The snubber functional test shall verify that:

1. "As-Found"
b. Bleed takes place after activation of the snubber, in both tension and compression directions.
2. "As-Left"
b. Bleed takes place after activation of the snubber, in both tension and compression, and is greater than 0.24 inch/minute and less than 1.18 inches/minute.

Test data BLEED (RELEASE) Rate are recorded in Attachment 2 of the procedure.

Snubber specifically required not to displace under continuous load are addressed in TSR 3.7.4.2.

OM Part 4, Section 3.2.1.2, - Operability Test Loads - Snubbers shall be tested at a load sufficient to verify the operating parameters specified in paragraph 3.2.1.1. Testing at less than rated load must be correlated to operability parameters at rated load.

Test Load parameters are built into the Snubber Test Bench Machine Program, and are listed in the applicable Appendixes of 0-TI-398 as follows:

0-TI-398 Appendix B, Section 3.0 provides test parameters in the snubber test machine program for Bergen-Paterson, Anchor Darling or Fronek Hydraulic snubbers. Test Loads are in the Snubber Test Machine Program. These Test Loads are 80 percent of rated load for snubber sizes HSSA -3, -10, -20,- 30, ADH/FRONEK -20, -30, -50, -70, and -130.

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Lisega Torus Dynamic Restraints are functionally tested using the Lisega Surrogate snubber in accordance with 0-SI-4.6.H-2E. The test load in the Snubber Test Machine Program is 44.9 kip nominal value that is correlated to the 100 kip test load used in factory testing. Test data/results are recorded in Attachment 2 of the SI. Lisega Torus Dynamic Restraints may also be functionally tested by the manufacturer using a 100 kip test load prior to installation.

Test data/results are recorded in Attachment 2 of the SI.

0-TI-398 Appendix F, Section 3.0, provides test parameters in the snubber test machine program for Pacific Scientific Company (PSA) Mechanical Snubbers, Test Load parameters in the Snubber Test Machine Program are 60 percent of rated load for PSA sizes 1/4, 1/2, 3, and 10 and 50 percent of rated load for PSA sizes 1, 35, and 100.

OM Part 4, Section 3.2.1.3, Qualitative Testing. Qualitative testing may be used in lieu of quantitative measurements in meeting the requirements of paragraph 3.2.1.1, provided adequate justification can be presented and is acceptable to the regulatory authority having jurisdiction over the facility. In those cases, the Owner shall obtain sufficient data, based upon service history or life cycle testing, to justify the ability of the parameter in question to be within specifications over the life of the snubber (e.g.,

demonstrate that activation takes place without measurement of the activation level).

A test report shall be available for each snubber exempted from an inservice quantitative test requirement. The test report must verify that the parameter was within specifications to allow exemption of the snubber from quantitative testing of the parameter.

0-TI-398 has been revised to remove the "push-pull" functional testing of mechanical PSA snubbers.

0-TI-398, Section 7.17.2, Functional Testing of PSA Mechanical Snubbers states:

A. BFN Unit 2 Technical Requirements require drag force measurements and activation verifications.

B. Mechanical snubber functional testing is performed to verify two characteristics, activation and drag force. The limit for acceleration is not to exceed 0.02 g's. Pacific Scientific Company (PSC) performed qualification tests on new snubbers and performed tests on snubbers that have been in extended plant service to verify the activation levels at various loads. PSC concluded that there is no significant change in activation value at any level of rated load of the snubber.

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C. Performing a test for activation of PSA snubbers requires a specialized test machine.

BFN has a computer controlled API/Barker STB-200 snubber test bench. The test bench is capable of testing any size PSA snubber and most medium and small bore hydraulic snubbers. The software performs four basic functions:

1. Operator interface
2. Machine control
3. Data acquisition and conversion
4. Data analysis and presentation Test results are presented in the form of a graph with maximum and average test values.

OM Part 4, Section 3.2.2, Inservice Operability Testing Frequency. Testing shall take place at least every refueling outage using a sample of snubbers in the facility.

TSR 3.7.4.2, requires in-place or bench functional testing of a representative sample of 10%

of the total of each type of safety-related snubbers. The testing frequency specified is 24 months.

Surveillance Instructions 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, O-SI-4.6.H 2E, and 0-SI-4.6.H 2F Section 1.3A, Frequency/Conditions state:

This Surveillance Instruction shall be performed each refueling outage and portions of it may be performed to establish operability in accordance with 2-SI-4.6.H-1, Visual Examination of Hydraulic and Mechanical Snubbers.

Testing of the 10% sample lot of hydraulic and mechanical snubbers takes place every 24 months/ at each scheduled unit refueling outage in accordance with the TSR 3.7.4.2 and the applicable SI.

OM Part 4, Section 3.2.3, Inservice Operability Testing Sample Plans:

The inservice testing sample shall be selected using one of the three sample plans (a comparison of sampling plans is contained in Appendix C):

(a) 10% testing plan (b) 37 testing plan (c) 55 testing plan The snubbers of parallel and multiple installations shall be identified and counted individually.

All fractional sample sizes shall be rounded up to the next integer.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 154 of 205 Attachment 7 (Page 33 of 37) 0-TI-398, Section 7.17, 2nd paragraph of the Functional Test Program Guidelines states:

Functional tests are performed each operating cycle to meet Technical Requirements Manual TRM 3.7.4.2 verifying, by sampling 10% of each subgroup of snubber, that the safety-related or quality-related snubbers are operable. For each failure to meet the functional test acceptance criteria an additional 10% of the remaining snubbers in the subgroup shall be tested, until no additional failures occur. In addition to the required sampling, snubbers under service life monitoring and balance-of-plant programs should be addressed. Specific Technical Requirements are given in the appropriate functional testing SI.

2-SI-4.6.H-1, Appendix A is a snubber listing of each individual snubber shown on a pipe support drawing. Each snubber is given a unique identification number and are counted individually as listed in Appendix A.

OM Part 4, Section 3.2.4, Inservice Operability Testing Failure Evaluations TSR 3.7.4.3 states that: A failure analysis shall be made of each failure to meet the functional test acceptance criteria of TSR 3.7.4.2 to determine the cause of failure. The frequency is once for each discovery of snubber failure to meet functional acceptance criteria.

0-SI-4.6.H-2A, O-SI-4.6.H-2B, O-SI-4.6.H-2C, 0-SI-4.6.H 2E, and 0-SI-4.6.H 2F, Attachment 3 provides the requirements for performing failure evaluations of failed snubbers.

OM Part 4, Section 3.2.4.1, Failure Evaluation Requirements. Snubbers that do not meet the operability testing acceptance criteria in paragraph 3.2.1 shall be evaluated to determine the cause of the failure.

The BFN Unit 2 TSR 3.7.4.3 mandates that failure analysis be made for each snubber failure to meet the functional test acceptance criteria in TSR 3.7.4.2 or the applicable SIs:

0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H 2E, and 0-SI-4.6.H 2F state:

Evaluation of snubber operability, corrective actions, and selection of other suspect snubbers for verification shall be as specified on Attachment 3 or Attachment 6, as applicable.

For each failure to meet the functional test acceptance criteria an additional 10% of the remaining snubbers in the subgroup shall be tested, until no additional failures occur. In addition to the required sampling, snubbers under service life monitoring and balance-of-plant programs should be addressed. Specific Technical Requirements are given in the appropriate functional testing SI.

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An Engineering Failure Analysis for Inoperable Snubber must be performed or completed using the appropriate data sheets or attachments of the applicable SIs to determine the cause of failure. Results of this analysis shall be used to select snubbers to be tested in an effort to determine the operability of other snubbers with the same failure mode. Selection of snubbers for future testing may also be based on the failure analysis. This evaluation may also be used by Site Engineering when required to perform a supported system/component analysis.

For each failed snubber, a PER is initiated in accordance with SPP-8.1, "Conduct of Testing",

and SPP-3.1, a Work Order is initiated to replace the snubber (if necessary), perform in place or STB 200 Test Bench functional test. An additional 10% of the remaining snubbers in the subgroup shall be tested, until no additional failures occur. In addition to the required sampling, snubbers under service life monitoring and balance-of-plant programs should be addressed. Specific Technical Requirements are given in the appropriate functional testing SI.

0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, and 0-SI-4.6.H-2E state: For an inoperable snubber(s), within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, replace or restore inoperable snubbers to an operable status and perform an engineering evaluation on the supported component or system, if the snubber does NOT meet the functional test acceptance criteria of TSR 3.7.4.2. Otherwise, declare the system inoperable and follow the required actions specified in the TRM. The engineering evaluation is to determine if the component or system restrained by the snubber(s) was adversely affected by inoperability of the snubber(s) during the previous operating cycle and ensure that the restrained component or system remains capable of meeting its design function. The engineering evaluation(s) for supported component or system analysis are to be documented on Attachment 4 of the procedure.

0-SI-4.6.H 2E states: For an inoperable snubber(s), an engineering evaluation must be performed to determine if the component or system restrained by the snubber(s) was adversely affected by inoperability of the snubber(s) during the previous operating cycle and ensure that the restrained component or system remains capable of meeting its design function. The engineering evaluation(s) for supported component or system analysis are to be recorded on Attachment 4 of the procedure.

OM Part 4, Section 3.2.4.2, Test Failure Mode Groups. Unacceptable snubber(s) shall be categorized into test failure mode group(s). A test failure mode group(s) shall include all unacceptable snubbers that have a given failure mode, and all other snubbers subject to the same failure mode. The following failure modes shall be used:

(a) Design/manufacturing (b) Application induced (c) Maintenance/repair/installation (d) Isolated (e) Unexplained

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For BFN Unit 2, TSR 3.7.4.4, NOTE states: This testing is independent of the requirements of TSR 3.7.4.3. For any snubber which fails to lockup or fails to move (i.e., frozen in place),

evaluate the cause. If caused by manufacturer or design deficiency, perform in-place or bench functional test of all snubbers of the same design, subject to the same defect. The functional test acceptance criteria shall be as specified in TSR 3.7.4.2.

In addition, the applicable Surveillance Instructions require an engineering evaluation of the snubber failure, and classification of the snubber failure mode as isolated, location, manufacturing, design, or other. The engineering evaluation includes determination of subsequent testing required, based on the failure mode, which may involve testing of snubbers susceptible to the same failure mode. However, establishment of specific groupings based on failure modes is not performed.

OM Part 4, Section 3.2.4.3, Test Failure Mode Group Boundaries.

Once a test failure mode group has been established, any snubber(s) in that test failure mode group will not be part of the testing groups from which the snubbers originated except as noted in paragraph 3.2.4.4 below. The new test failure mode group will remain as defined until corrective action has been completed.

As stated in response to Section 3.2.4.2 above, establishment of test failure mode groups is not performed..

OM Part 4, Section 3.2.4.4, Snubbers in More Than One Test Failure Mode Group - In the event that a snubber(s) becomes included in more than one test failure mode group, it shall be counted in each failure mode group in which it is unacceptable and shall be subject to the corrective action of each test failure mode group.

As stated in response to Section 3.2.4.2 above, establishment of test failure mode groups is not performed. The corrective action for a snubber subject to multiple failure modes is as determined and documented in the engineering evaluation in accordance with the applicable Surveillance Instruction.

OM Part 4, Section 3.2.5, Inservice Operability Testing Corrective Action and Impact on Continued Testing - Snubbers which have been found unacceptable for the testing acceptance criteria of paragraph 3.2.1.1 shall be subjected to the following corrective actions(s) with its indicated impact on continued testing. Selection of the corrective action shall be governed by the sampling plan which is used.

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See discussion for OM Part 4 item 3.2.4.1 above. For each unacceptable or failed snubber, a PER is initiated in accordance with SPP-8.1 and SPP-3.1, failure analysis performed, Work Order initiated to replace the snubber (if necessary), or STB 200 Test Bench functional test performed on an additional lot equal to 10% of the remaining snubbers of that type. Testing shall continue until no additional inoperable snubbers are found within the subsequent lots or all snubbers of the original test type are tested or all suspect snubbers identified in the failure analysis have been tested, as applicable. The functional test criteria shall be as specified in TSR 3.7.4.2 or the applicable test Sis.

OM Part 4, Section 3.2.6, Inservice Operability Testing Methods.

The following test requirements shall apply:

(a) Testing to be performed on the snubbers in their as-found condition to the fullest extent practical regarding the features to be tested (b) Test methods employed must not alter the condition of the snubber to the extent the results are not representative of the parameters prior to test (c) inservice operability testing may be accomplished with the snubber installed in its permanent location by utilizing Owner approved test methods and equipment (d) The snubbers may be removed and bench tested in accordance Owner approved test procedures. After reinstallation, the snubber shall meet the requirements of paragraph 2.1.1(e)

(e) Where the physical size of the snubber, test equipment limitations, or inaccessibility of location prevent the use of methods in paragraphs (c) and (d) above, the snubber subcomponents shall be examined and tested in accordance with approved procedures. Reassembly of individual components must be in accordance with approved procedures.

(f) Testing methods may be used which measure parameters indirectly or parameters other than those specified if those results can be correlated to the specified parameters through established methods.

For BFN Unit 2, Operability testing methods for mechanical and hydraulic snubbers are quantitative (use of STB 200 Test Bench w/Windows program), in accordance with applicable Sis 0-SI-4.6.H-2A, 0-SI-4.6.H-2B, 0-SI-4.6.H-2C, 0-SI-4.6.H-2E, and 0-SI-4.6.H-2F and TSR 3.7.4.2 requirements and, MPI-0-000-SNBOO4 for Removing and Reinstalling Pacific Scientific Mechanical, Bergen-Paterson, Anchor/Darling, Fronek and Lisega Torus Dynamic Restraint Snubbers and, MPI-0-000-SNBOO2 for Hydraulic shock and Sway Arrestor Bergen-Paterson, Anchor/Darling, Fronek Unit Disassembly and Reassembly to meet the requirements (a) through (f) above.

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OM Part 4, Section 3.3, Testing Documentation Documents necessary to verify result of the preservice and inservice program shall include as a minimum:

(a) preservice operability test procedures and results (b) inservice operability test procedures and results (c) noncomformance results, noncomformance evaluations, and corrective action.

TR 3.7.4 documents the required action, completion time and conditions pertaining to the operability, inspection, testing, and acceptance criteria of snubbers at BFN. Surveillance instructions are utilized to implement the requirements of TR 3.7.4. Plant surveillance instructions O-SI-4.6.H-2A, 0-SI-4.6.H-2B, O-SI-4.6.H-2C, O-SI-4.6.H-2E, and O-SI-4.6.H-2F cover functional testing of all snubbers at BFN. Surveillance instruction 2-SI-4.6.H-1 covers the visual examination of Unit 2 hydraulic and mechanical snubbers. These surveillance instructions document pre-service operability test procedures and results. They also document in-service operability test procedures and results, nonconformance evaluations, and corrective actions. Plant procedure SPP 3.1 is utilized to document when snubbers are found to be outside of testing acceptance criteria or in a nonconforming condition.

Surveillance instructions are included in the work package for the snubber and transmitted to permanent record storage (EDMS) after closure of the work package/work order.

The proposed alternative for snubber visual examination training qualification and documentation is provided by 2-SI-4.6.H-1 Step 7.1.1 [A] through [F]. The visual acuity requirements of IWA-2320 are satisfied.

IWA-2300 addresses qualifications of NDE personnel. The BFN snubber program engineer and persons performing the snubber inspections meet the visual requirements described by IWA-2300.

For BFN Unit 2, O-SI-4.6.H-2A, O-SI-4.6.H-2B, O-SI-4.6.H-2C, O-SI-4.6.H-2E, and 0-SI-4.6.H-2F provide a means for the control and documentation of all snubber surveillance activities provided in this Surveillance Instruction. Snubber(s) operability tests results, including nonconformance results, nonconformance evaluations, and corrective actions are documented in the appropriate data sheets or attachments of the applicable Sis listed above.

These documents are included in the work package for the tested snubber and transmitted to permanent record storage (EDMS) after closure of the work package/Work Order for entry into the database.

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TENNESSEE VALLEY AUTHORITY BROWNS FERRYNUCLEAR PLANT (BFN)

UNIT 2 AMERICAN SOCIETY OF MECHANICAL ENGINEERS (ASME) SECTION XI, INSERVICE INSPECTION (ISI)PROGRAM (FOURTH TEN YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-ISI-41 Executive Summary: In accordance with 10 CFR 50.55a(a)(3)(i), TVA is requesting relief from inservice inspection requirements of the 2004 Edition of Section XI, Appendix VIII, Supplement 11, "Qualification Requirements For Full Structural Overlaid Wrought Austenitic Piping Welds," of the ASME Boiler and Pressure Vessel Code. The Performance Demonstration Initiative (PDI) Program for implementation of the Supplement 11 qualification program for overlay welds is not in strict compliance with the requirements of Supplement 11 of the 2004 Edition. TVA proposes to use the PDI Program for implementation of Appendix VIII, Supplement 11 as amended in the attachment to this request for relief. The amendments to Supplement 11 as shown in the attachment were coordinated with PDI, NRC, and Pacific Northwest National Laboratory (PNNL).

Unit: Two (2)

System(s): Various Components: Piping Welds with Structural Weld Overlay ASME Code Class: ASME Code Class 1 Section XI Edition: 2004 Edition as required by 10 CFR 50.55a(b)(2)(xxiv)

Code Table: N/A Examination Category: N/A Examination Item Number: N/A Code Requirement: The 2004 Edition of ASME Section XI as required by 10 CFR 50.55a(b)(2)(xxiv) requires a volumetric (UT) examination of the overlay including applicable portions of the pipe weld. The UT examination must be performed using personnel, procedures, and equipment qualified in accordance with Appendix VIII, Supplement 11.

Code Requirements From Which Relief Is Requested: Pursuant to 10 CFR 50.55a(a)(3)(i) relief is requested to use the enclosed Performance Demonstration Initiative (PDI) Program for implementation of Appendix VIII, Supplement 11 requirements.

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Relief is requested from the requirement to qualify personnel, procedures, and equipment in accordance with Appendix VIII, Supplement 11 as stated in the 2004 Edition of ASME Section X1, as required by 10 CFR 50.55a(b)(2)(xxiv). The Code requirements for which relief is requested are all contained within Appendix VIII, Supplement 11. For example, paragraph 1.1(d)(1), requires that all base metal flaws be cracks. Paragraph 1.1(e)(1 )

requires that at least 20% but less than 40% of the flaws shall be oriented within plus or minus 20 degrees of the pipe axial direction. Paragraph 1.1(e)(1) also requires that the rules of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws. Paragraph 1 .1(e)(2)(a)(1) requires that a base grading unit shall include at least 3 inches of the length of the overlaid weld. Paragraph 1.1(e)(2)(b)(1) requires that an overlay grading unit shall include the overlay material and the base metal-to-overlay interface of at least 6 square inches. The overlay grading unit shall be rectangular, with minimum dimensions of 2 inches. Paragraph 3.2(b) requires that all extensions of base metal cracking into the overlay material by at least 0.1 inch are reported as being intrusions into the overlay material.

List Of Items Associated With The Relief Request: Weld Overlay(s) that currently require examination in the Unit 2 ISI Program Basis For Relief: The requirements of ASME Section Xl, Appendix VIII, Supplement 11, as stated in the 2004 Edition as required by 10 CFR 50.55a(b)(2)(xxiv), are not practical to implement. The requirements were amended to improve the implementation process. The amended requirements are contained in the attachment to this relief request. The EPRI sponsored PDI amendments to Supplement 11, as shown in the attachment, were coordinated with PDI, NRC, and PNNL.

Paragraph 1.1(d)(1), requires that all base metal flaws be cracks. As illustrated below, implanting a crack requires excavation of the base material on at least one side of the flaw.

While this may be satisfactory for ferritic materials, it does not produce a useable axial flaw in austenitic materials because the sound beam, which normally passes only through base material, must now travel through weld material on at least one side, producing an unrealistic flaw response. To resolve this issue, the PDI program revised this paragraph to allow use of alternative flaw mechanisms under controlled conditions. For example, alternative flaws shall be limited to when implantation of cracks precludes obtaining an effective ultrasonic response, flaws shall be semielliptical with a tip width of less than or equal to 0.002 inches, and at least 70 percent of the flaws in the detection and sizing test shall be cracks and the remainder shall be alternative flaws.

Ec-- Mechanical fatigue crack are-in Base material

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Relief is requested to allow closer spacing of flaws provided they did not interfere with detection or discrimination. The existing specimens used to date for qualification to the Tri-party (NRC/Electric Power Research Institute (EPRI)/Boiling Water Reactor Owners Group (BWROG)) agreement have a flaw population density greater than allowed by the current Code requirements. These samples have been used successfully for all previous qualifications under the Tri-party agreement program. To facilitate their use and provide continuity from the Tri-party agreement program to Supplement 11, the PDI Program has merged the Tri-party test specimens into their weld overlay program. For example: the requirement for using IWA-3300 for proximity flaw evaluation in paragraph 1.1 (e)(1) was excluded, instead indications will be sized based on their individual merits; paragraph 1.1 (d)(1) includes the statement that intentional overlay fabrication flaws shall not interfere with ultrasonic detection or characterization of the base metal flaws; paragraph 1.1 (e)(2)(a)(1) was modified to require that a base metal grading unit include at least 1 inch of the length of the overlaid weld, rather than 3 inches; paragraph 1.1(e)(2)(a)(3) was modified to require sufficient unflawed overlaid weld and base metal to exist on all sides of the grading unit to preclude interfering reflections from adjacent flaws, rather than the 1- inch requirement of Supplement 11; paragraph 1.1 (e)(2)(b)(1) was modified to define an overlay fabrication grading unit as including the overlay material and the base metal-to-overlay interface for a length of at least 1 in, rather than the 6 square inches requirement of Supplement 11; and paragraph 1.1(e)(2)(b)(2) states that overlay fabrication grading units designed to be unflawed shall be separated by unflawed overlay material and unflawed base metal-to-overlay interface for at least 1 inch at both ends, rather than around its entire perimeter.

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

The PDI Program revised paragraph 2.0 allowing the overlay fabrication and base metal flaw tests to be performed separately. The requirement in paragraph 3.2(b) for reporting all extensions of cracking into the overlay is omitted from the PDI Program because it is redundant to the RMS calculations performed in paragraph 3.2(c) and its presence adds confusion and ambiguity to depth sizing as required by paragraph 3.2(c). This also makes the weld overlay program consistent with the Supplement 2 depth sizing criteria.

A comparison between the ASME Section Xl 2004 Edition of Supplement 11, Code Case N-653 (for information only), and the PDI Program is enclosed as an attachment.

To avoid confusion, several instances of the term "cracks" or "cracking" were changed to the term "flaws" because of the use of alternative flaw mechanisms. Additionally, to avoid confusion, the overlay thickness tolerance contained in paragraph 1.1 (b) last sentence, was reworded and the phrase "and the remainder shall be alternative flaws" was added to the next to last sentence in paragraph 1 .1 (d)(1).

The proposed amended requirements of Supplement 11 for the qualification of personnel, procedures, and equipment will provide an alternative with an acceptable level of quality and safety.

Alternate Requirement: In lieu of the requirements of ASME Section XI, 2004 Edition, of Appendix VIII, Supplement 11, the PDI Program shall be used. TVA proposes to utilize personnel, procedures, and equipment qualified in accordance with ASME section Xl, Appendix VIII, Supplement 11 as amended by the Attachment, which is the EPRI administered PDI Program.

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Justification For The Granting Of Relief: The U.S. nuclear utilities created the PDI to implement performance demonstration requirements contained in Appendix VIII of Section Xl of the Code. To this end, PDI has developed a program for qualifying equipment, procedures, equipment, and personnel in accordance with the ultrasonic testing criteria of Appendix VIII, Supplement 11. Prior to the Supplement 11 program, EPRI was maintaining a performance demonstration program for weld overlay qualification under the Tri-party Agreement "Coordination Plan for NRC/EPRI/BWROG Training and Qualification Activities of NDE Personnel", dated July 3, 1984. Instead of having two programs with similar objectives, the NRC staff recognized the PDI program for weld overlay qualifications as an acceptable alternative to the Tri-party Agreement in a letter to the PDI Chairman "Weld Overlay Performance Demonstration Administered by PDI as an Alternative for Generic Letter 88-01 Recommendations" (ADAMS accession number ML020160532). The NRC determined that the PDI program does not fully comport with the existing requirements of Supplement 11, but nevertheless, does satisfy the recommendation in GL-88-01 for weld overlay performance demonstrations and qualifications of procedures and personnel. The NRC concluded that the PDI program for weld overlays, which meets the spirit of Appendix VIII, Supplement 11, is an acceptable alternative to the requirements iin GL-88-01. The differences are discussed below.

Paragraph 1.1(b):

This paragraph states limitations to the maximum thickness for which a procedure may be qualified. The ASME Code states that "The specimen set must include at least one specimen with overlay thickness within minus 0.10 inch to plus 0.25 inch of the maximum nominal overlay thickness for which the procedure is applicable." The ASME Code requirement addresses the specimen thickness tolerance for a single specimen set, but is confusing when multiple specimen sets are used. The PDI proposed alternative states that "the specimen set shall include specimens with overlay not thicker than 0.10 inch more than the minimum thickness, nor thinner than 0.25 inch of the maximum nominal overlay thickness for which the examination procedure is applicable." The proposed alternative provides clarification on the application of the tolerance. The tolerance is unchanged for a single specimen set, however, it clarifies the tolerance for multiple specimen sets by providing tolerances for both the minimum and maximum thicknesses. The proposed wording eliminates confusion while maintaining the intent of the overlay thickness tolerance.

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Paragraph 1.11(d)(1):

The PDI program determined that certain Supplement 11 requirements pertaining to location and size of cracks would be extremely difficult to achieve. In an effort to satisfy the requirements, the PDI program developed a process for fabricating flaws that exhibited crack-like reflective characteristics. Instead of all flaws being cracks, as required by Paragraph 1.1(d)(1), the PDI program weld overlay performance demonstrations contain at least 70-percent cracks with the remainder being fabricated flaws exhibiting crack-like reflective characteristics. The application of alternative flaws are limited with a flaw tip dimension of 0.002 inches. Throughout the PDI proposal, the ASME Code term 'crack' is replaced with the term 'flaw' in order to substitute fabricated flaws for cracks.

Paragraph 1.1 (e)(1):

The ASME Code requires that at least 20-percent but not less than 40-percent of the flaws shall be oriented within plus or minus 20-degrees of the axial direction. In the proposed PDI program, the flaws satisfy the requirement and specifies that the flaws must be in the base metal. The PDI program is confining flaw placement to ensure that candidates do not have visual or physical access to the flaws. This is a tightening of the requirements.

The ASME Code also requires that the requirements of IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws. The proposed PDI program treats each flaw as an individual flaw and not as part of a system of closely spaced flaws. The proposed program controls the flaws going into a test specimen set such that the flaws are free of interfering reflections from adjacent flaws. In some cases, this would permit flaws to be closer together than what is allowed by IWA-3300, thus making the performance demonstration more challenging.

Paragraph 1.1(e)(2):

The ASME Code requires that specimens be divided into base metal and overlay grading units. The PDI program adds clarification with the addition of the word "fabrication" to denote the overlay process and ensures flaw identification by ensuring all flaws will not be masked by other flaws with the addition of, "Flaws shall not interfere with ultrasonic detection or characterization of other flaws." PDI's alternative provides clarification and assurance that the flaws are identified and associated with the appropriate material.

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Paragraph 1.1(e)(2)(a)(1):

The ASME Code requires that a base grading unit shall include at least 3 inches of the length of the overlaid weld, and the base grading unit includes the outer 25 percent of the overlaid weld and base metal on both sides. The PDI program reduced the criteria to 1 inch of the length of the overlaid weld and eliminated from the grading unit the need to include both sides of the weld. The test specimens from the existing weld overlay program have flaws on both sides of the welds, which prevents them from satisfying the base grading unit requirements. These test specimens have been used successfully for testing the proficiency of personnel for over 16 years. This is a more challenging test because the individual must locate the flaw on the correct side of the weld.

Paragraph 1.1(e)(2)(a)(2):

The ASME Code requires when base metal cracks penetrate into the weld overlay, the weld overlay within 1 inch of the crack becomes part of the base metal grading unit. The PDI program makes the base metal part of the weld overlay grading unit for cracks ending in the weld overlay. The object of the performance demonstration is to detect the crack tip that is in the weld overlay. The change redefines the grading unit to be representative of the intent of the performance demonstration.

Paragraph 1.1(e)(2)(a)(3):

The ASME Code requires that for unflawed base grading units, at least 1 inch of unflawed overlaid weld and base metal shall exist on either side of the base grading unit. This is to minimize the number of false identifications of extraneous reflectors. The PDI program stipulates that unflawed overlaid weld and base metal exist on all sides of the grading unit and be free of interfering reflections from adjacent flaws, which addresses the same concerns as the ASME Code.

Paragraph 1 .1(e)(2)(b)(1):

The ASME Code requires that an overlay grading unit shall include the overlay material and the base metal-to-overlay interface of at least 6 square inches. The overlay grading unit shall be rectangular, with minimum dimensions of 2 inches. The PDI program reduces the base metal-to-overlay interface to at least 1 inch (in lieu of a minimum of 2 inches) and eliminates the minimum rectangular dimension. This criterion is more challenging than the ASME Code because of the variability associated with the shape of the grading unit.

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Paragraph 11.1(e)(2)(b)(2):

ASME Code requires that overlay fabrication grading units designed to be unflawed shall be separated by unflawed overlay material and unflawed base metal-to-overlay interface for at least 1 inch around its entire perimeter. The PDI program changes the requirement of 1 inch around its perimeter to 1 inch at both ends with sufficient unflawed overlaid weld and base metal on both sides of the overlay fabrication grading unit to preclude interfering reflections from adjacent flaws. The PDI proposal accommodates test specimens that have been successfully used for previous weld overlay qualifications under the Tri-party (NRC/EPRI/BWROG) agreement program.

Paragraph 1.1(e)(2)(b)(3):

The ASME Code identifies the minimum number of flawed and unflawed grading units in the test set. The PDI proposal stipulates that for detection, the procedure test set must consist of at least three personnel test sets. The PDI proposal is more conservative than the ASME Code.

Paragraph 1.1(f)(1):

The ASME Code identifies the minimum number of flaws and flaw locations in the test set.

The PDI proposal stipulates that for sizing, the procedure test set must consist of at least three personnel test sets. The PDI proposal is more conservative than ASME Code.

Paragraphs 1.1(f)(3) and 1.1(f)(4):

The ASME Code requirements are clarified by the PDI program by replacing the term "cracking" with "flaws" because the use of alternative flaws would provide representative responses that are not achievable with implanted cracks.

Paragraph 1.1(f)(4):

The ASME Code stipulates base metal cracks extending in the overlay. The PDI proposal uses base metal flaws extending into the overlay. The flaws are fabricated alternatives to cracks that were determined to be acceptable in Paragraph 1.1 (d)(1).

Paragraph 2.0:

The ASME Code requirements are retained in the PDI program alternative. In addition, the PDI program provides clarification that the overlay fabrication flaw test and the base metal flaw test may be performed separately. The ASME Code is nonspecific on this issue. Since flaw specific criteria exist for the overlay fabrication process, separating the base metal test should produce the same results as the combined test.

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Paragraph 2.1 and 2.2(d):

The ASME Code requirements are clarified by the PDI program by the addition of the terms "metal" and "fabrication." The change provides consistency with the terms normally used for these applications in the manufacturing sector. The change is editorial in nature.

Paragraph 2.3:

The ASME Code requires 80-percent of the flaws shall be sized at a specific location on the surface of the specimen identified to the candidate. This requires detection and sizing performance demonstration to be performed separately. The PDI proposal permits detection and sizing performance demonstration to be performed together, when necessary, and permits identifying flawed regions instead of specific locations. The PDI proposal is more challenging for sizing because the candidate has no prior knowledge of specific flaw locations.

Paragraph 3.1:

The ASME Code requires examination procedures, equipment, and personnel (as a complete ultrasonic system) to be qualified for detection or sizing of flaws, as applicable, when certain criteria are met. The PDI program allows procedure qualification to be performed separately from personnel and equipment qualification. Historical data indicate that, if ultrasonic detection or sizing procedures are thoroughly tested, personnel and equipment using those procedures have a higher probability of successfully passing a qualification test. In an effort to increase this passing rate, PDI has elected to perform procedure demonstration separately in order to assess and modify essential variables that may affect overall system capabilities.

For a procedure to be qualified, the PDI program requires three times as many flaws to be detected (or sized) than required by Supplement 11. Personnel and equipment are still required to meet the Supplement 11 requirement; therefore, the PDI program criteria exceed the ASME Code requirements for personnel, procedures, and equipment qualification.

Paragraph 3.2(a):

The ASME Code requirements are clarified by the PDI program by replacing the term "cracking" with "flaws" because of the use of alternative flaw mechanisms. This clarification in the PDI program maintains the intent of the ASME Code requirement and is acceptable.

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Paragraph 3.2(b):

The ASME Code requires that all extensions of base metal cracking into the overlay material by at least 0.1 inch be reported as intrusions into the overlay material. The PDI program omits this criterion. The PDI program requires that cracks be sized to the tolerance specified in the Code, which is 0.125 inches. Since the Code tolerance is close to the 0.1 inch value of Paragraph 3.2(b), any crack extending beyond 0.1 inch into the overlay material would be identified from its dimensions. The reporting of an extension in the overlay material is redundant for performance demonstration testing.

Based on the above evaluation, the proposed alternative to use the EPRI-PDI program requirements in lieu of selected paragraphs of Supplement 11 to Appendix VIII of Section XI of the ASME Code will provide an acceptable level of quality and safety.

Implementation Schedule: This request for relief is applicable to the Unit 2 Fourth Ten-Year ISI inspection interval which expires on May 24,,2021.

Attachments:

Attachment A: Table - Comparison of ASME Section X1, Appendix VIII, Supplement 11, Code Case N-653, and PDI Alternative.

TVA to NRC Letter dated May 22, 2009.

NRC to TVA Letter (SER) dated August 12, 2009.

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REQUIREMENTS FOR FULL (Provided for Information Only) The Proposed Alternative to Supplement 11 STRUCTURAL OVERLAID WROUGHT Requirements AUSTENETIC PIPING WELDS 1.0 SPECIMEN REQUIREMENTS Qualification test specimens shall meet the No Change No Change requirements listed herein, unless a set of specimens is designed to accommodate.

specific limitations stated in the scope of the examination procedure (e.g., pipe size, weld joint configuration, access limitations).

The same specimens may be used to demonstrate both detection and sizing qualification.

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

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

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(b) The specimen set shall consist of at No Change (b) The specimen set shall consist of at least three specimens having different least three specimens having different nominal pipe diameters and overlay nominal pipe diameters and overlay thicknesses. They shall include the thicknesses. They shall include the minimum and maximum nominal pipe minimum and maximum nominal pipe diameters for which the examination diameters for which the examination procedure is applicable. Pipe diameters procedure is applicable. Pipe diameters within a range of 0.9 to 1.5 times a nominal within a range of 0.9 to 1.5 times a nominal diameter shall be considered equivalent. If diameter shall be considered equivalent. If the procedure is applicable to pipe the procedure is applicable to pipe diameters of 24 inches or larger, the diameters of 24 inches or larger, the specimen set must include at least one specimen set must include at least one specimen 24 inches or larger but need NOT specimen 24 inches or larger but need NOT include the maximum diameter. The include the maximum diameter.

specimen set must include at least one The specimen set shall include specimens specimen with overlay thickness within with overlays NOT thicker than 0.1 inches

-0.1 inches to +0.25 inches of the maximum more than the minimum thickness, nor nominal overlay thickness for which the thinner than 0.25 inches of the maximum procedure is applicable. nominal overlay thickness for which the examination procedure is applicable.

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

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(d) Flaw Conditions (1) Base metal flaws. All flaws must be (1) Base metal flaws. All flaws must be in (1) Base metal flaws. All flaws must be in cracks in or near the butt weld or near the butt weld heat-affected zone, or near the butt weld heat-affected zone, heat-affected zone, open to the inside open to the inside surface, and extending at open to the inside surface, and extending at surface, and extending at least 75 percent least 75 percent through the base metal least 75 percent through the base metal through the base metal wall. Flaws may wall. Intentional overlay fabrication flaws wall. Intentional overlay fabrication flaws extend 100 percent through the base metal shall NOT interfere with ultrasonic detection shall NOT interfere with ultrasonic detection and into the overlay material; in this case, or characterization of the cracking. or characterization of the base metal flaws.

intentional overlay fabrication flaws shall Specimens containing IGSCC shall be used Specimens containing IGSCC shall be used NOT interfere with ultrasonic detection or when available. At least 70 percent of the when available. At least 70 percent of the characterization of the cracking. flaws in the detection and sizing tests shall flaws in the detection and sizing tests shall Specimens containing IGSCC shall be used be cracks. Alternative flaw mechanisms, if be cracks and the remainder shall be when available. used, shall provide crack-like reflective alternative flaws. Alternative flaw characteristics and shall be limited by the mechanisms, if used, shall provide following: crack-like reflective characteristics and (1) Flaws shall be limited to when shall be limited by the following:

implantation of cracks precludes obtaining (a) The use of Alternative flaws shall be a realistic ultrasonic response. limited to when the implantation of cracks (2) Flaws shall be semi-elliptical with a tip produces spurious reflectors that are width of less than or equal to 0.002 inches. uncharacteristic of actual flaws.

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

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(2) Overlay fabrication flaws. At least No Change No Change 40 percent of the flaws shall be non-crack fabrication flaws (e.g., sidewall lack of fusion or laminar lack of bond) in the overlay or the pipe-to-overlay interface. At least 20 percent of the flaws shall be cracks. The balance of the flaws shall be of either type.

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

IWA-3300 shall be used to determine whether closely spaced flaws should be treated as single or multiple flaws.

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(2) Specimens shall be divided into base (2) Specimens shall be divided into base (2) Specimens shall be divided into base and over-lay grading units. Each specimen metal and overlay fabrication grading units. metal and overlay fabrication grading units.

shall contain one or both types of grading Each specimen shall contain one or both Each specimen shall contain one or both units. types of grading units. Flaws shall NOT types of grading units. Flaws shall NOT interfere with ultrasonic detection or interfere with ultrasonic detection or characterization of other flaws. characterization of other flaws.

(a)(1) A base grading unit shall include at (a)(1) A base metal grading unit shall (a)(1) The base metal grading unit includes least 3 inches of the length of the overlaid include at least 1 inch of the length of the the overlay material and the outer weld. The base grading unit includes the overlaid weld. The base metal grading unit 25 percent of the original overlaid weld.

outer 25 percent of the overlaid weld and includes the outer 25 percent of the The base metal grading unit shall extend base metal on both sides. The base overlaid weld and base metal on both circumferentially for at least 1 inch and shall grading unit shall NOT include the inner sides. The base metal grading unit shall start at the weld centerline and be wide 75 percent of the overlaid weld and base NOT include the inner 75 percent of the enough in the axial direction to encompass metal overlay material, or base overlaid weld and base metal overlay one half of the original weld crown and a metal-to-overlay interface, material, or base metal-to-overlay interface, minimum of 0.50" of the adjacent base material.

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

This portion of the overlay material shall unit.

NOT be used as part of any overlay grading unit.

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(a)(3) When a base grading unit is (a)(3) Sufficient unflawed overlaid weld and (a)(3) Sufficient unflawed overlaid weld and designed to be unflawed, at least 1 inch of base metal shall exist on all sides of the base metal shall exist on all sides of the unflawed overlaid weld and base metal grading unit to preclude interfering grading unit to preclude interfering shall exist on either side of the base reflections from adjacent flaws. reflections from adjacent flaws.

grading unit. The segment of weld length used in one base grading unit shall NOT be used in another base grading unit. Base grading units need NOT be uniformly spaced around the specimen.

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

shall be rectangular, with minimum dimensions of 2 inches.

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

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(b)(3) Detection sets shall be selected from (b)(3) Detection sets shall be selected from (b)(3) Detection sets shall be selected from Table VIII-S2-1. The minimum detection Table VIII-S2-1. The minimum detection Table VIII-S2-1. The minimum detection sample set is five flawed base grading sample set is five flawed base metal sample set is five flawed base metal units, ten unflawed base grading units, five grading units, ten unflawed base metal grading units, ten unflawed base metal flawed overlay grading units, and ten grading units, five flawed overlay fabrication grading units, five flawed overlay fabrication unflawed overlay grading units. For grading units, and ten unflawed overlay grading units, and ten unflawed overlay each type of grading unit, the set shall fabrication grading units. For each type of fabrication grading units. For each type of contain at least twice as many unflawed as grading unit, the set shall contain at least grading unit, the set shall contain at least flawed grading units. twice as many unflawed as flawed grading twice as many unflawed as flawed grading units. For initial procedure qualification, units. For initial procedure qualification, detection sets shall include the equivalent detection sets shall include the equivalent of three personnel qualification sets. To of three personnel qualification sets. To qualify new values of essential variables, at qualify new values of essential variables, at least one personnel qualification set is least one personnel qualification set is required. required.

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(f) Sizing Specimen (1) The minimum number of flaws shall be (1) The minimum number of flaws shall be (1) The minimum number of flaws shall be ten. At least 30 percent of the flaws shall ten. At least 30 percent of the flaws shall ten. At least 30 percent of the flaws shall be overlay fabrication flaws. At least be overlay fabrication flaws. At least be overlay fabrication flaws. At least 40 percent of the flaws shall be cracks 40 percent of the flaws shall be cracks 40 percent of the flaws shall be open to the open to the inside surface. open to the inside surface. For initial inside surface. Sizing sets shall contain a procedure qualification, sizing sets shall distribution of flaw dimensions to assess include the equivalent of three personnel sizing capabilities. For initial procedure qualification sets. To qualify new values of qualification, sizing sets shall include the essential variables, at least one personnel equivalent of three personnel qualification qualification set is required. sets. To qualify new values of essential variables, at least one personnel qualification set is required.

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

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

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

2.0 CONDUCT OF PERFORMANCE DEMONSTRATION The specimen inside surface and The specimen inside surface and The specimen inside surface and identification shall be concealed from the identification shall be concealed from the identification shall be concealed from the candidate. All examinations shall be candidate. All examinations shall be candidate. All examinations shall be completed prior to grading the results and completed prior to grading the results and completed prior to grading the results and presenting the results to the candidate. presenting the results to the candidate. presenting the results to the candidate.

Divulgence of particular specimen results or Divulgence of particular specimen results or Divulgence of particular specimen results or candidate viewing of unmasked specimens candidate-viewing of unmasked specimens candidate viewing of unmasked specimens after the performance demonstration is after the performance demonstration is after the performance demonstration is prohibited. prohibited. The overlay fabrication flaw test prohibited. The overlay fabrication flaw test and the base metal flaw test may be and the base metal flaw test may be performed separately. performed separately.

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

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

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

The candidate shall determine the length of the flaw in each region.

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

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

2.3 Depth Sizing Test.

For the depth sizing test, 80 percent of the The candidate shall determine the depth of (a) The depth sizing test may be conducted flaws shall be sized at a specific location on the flaw in each region. separately or in conjunction with the the surface of the specimen identified to the detection test.

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

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(b) When the depth sizing test is conducted in conjunction with the detection test and the detected flaws do NOT satisfy the requirements of 1.1 (f), additional specimens shall be provided to the candidate. The regions containing a flaw to be sized shall be identified to the candidate.

The candidate shall determine the maximum depth of the flaw in each region.

i (c) For a separate depth sizing test, the regions of each specimen containing a flaw to be sized shall be identified to the candidate. The candidate shall determine the maximum depth of the flaw in each region.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 182 of 205 Attachment 8 (Page 24 of 26) 3.0 ACCEPTANCE CRITERIA t I 3.1 Detection Acceptance Criteria.

t I Examination procedures, equipment, and Examination procedures are qualified for (a) Examination procedures are qualified personnel are qualified for detection when detection when all flaws within the scope of for detection when:

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

-t t

1) All flaws within the scope of the procedure are detected and the results of the performance demonstration satisfy the acceptance criteria of Table VIII-S2-1 for false calls.

.8. L

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(2) At least one successful personnel demonstration has been performed meeting the acceptance criteria defined in (b).

4 4 (b)Examination equipment and personnel are qualified for detection when the results of the performance demonstration satisfy the acceptance criteria of Table VIII-S2-1 for both detection and false calls.

(c) The criteria in (a), (b) shall be satisfied separately by the demonstration results for base metal grading units and for overlay fabrication grading units.

4 4 3.2 Sizing Acceptance Criteria.

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

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

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

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

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

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TENNESSEE VALLEY AUTHORITY BROWNS FERRY NUCLEAR PLANT (BFN)

UNIT 2 ASME SECTION XI INSERVICE INSPECTION PROGRAM (FOURTH TEN-YEAR INSPECTION INTERVAL)

REQUEST FOR RELIEF 2-PDI-40 Executive Summary:

In accordance with 10 CFR 50.55a(a)(3)(i), TVA is requesting relief from the specific requirements of performing the volumetric examination of the reactor pressure vessel (RPV) circumferential shell-to-flange weld and the RPV closure head-to-flange weld in accordance with the requirement of Appendix I of Section XI. TVA requests approval of an alternative to ASME Section XI, paragraph IWA-2232 of applicable editions for the 10-year Reactor Pressure Vessel (RPV) examinations. In lieu of the requirements of Appendix I and its associated sub-requirements of Article 4 of Section V, TVA will use the techniques, personnel, and equipment qualified to meet the requirements of ASME Section XI Appendix VIII, Supplements 4 and 6 of the 2001 Edition, in accordance with 10 CFR 50.55a(b)(2)(xxiv) and, as amended by Sections 10 CFR 50.55a(b)(2)(xv)(B) through 10 CFR 50.55a(b)(2)(xv)(G), and 10 CFR 50.55a(b)(2)(xvi)(A), by following the Electric Power Research Institute's (EPRI) Performance Demonstration Initiative (PDI) processes. TVA plans to use the proposed alternative during the next regularly-scheduled RPV examinations to be performed at or near the end of the current 10-year ISI Program interval, as required by ASME Code Section XI. No schedule change is requested. This proposed alternative represents the best available methodology in qualification of equipment and personnel performing ultrasonic examinations and uses an examination process that has provided, and will provide, the highest practical quality and greatest amount of coverage for the performance of the shell-to-flange and head-to-flange weld examinations. As such, the proposed alternative methodology provides an acceptable level of quality and safety. In addition, the approval of this relief results in savings in the cost of performing the examinations, with not having to incorporate the use of two different sets of examination equipment, and also results in lower personnel radiation exposure from not having to use a different methodology for the welds.

Note that this request for relief is similar to that requested by TVA for the BFN Units 2 and 3 reactor vessel-to-flange welds submitted initially in a letter to the NRC, dated February 23, 2005 and approved by the NRC in a letter dated August 2, 2005 (ADAMS Accession Number ML051730487). In addition, a similar request for relief has been approved by the Staff in a letter dated October 3, 2008 for the BFN Unit 1 reactor vessel shell-to-flange weld and the reactor vessel head-to-flange welds (ADAMS Accession Number ML082630051).

Unit: Two (2)

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System/Component:

Reactor Pressure Vessel (RPV) Upper Vessel Shell-to-Flange Weld. (Weld # C-5-FLG).

Reactor Pressure Vessel (RPV) Upper Head-to-Flange Weld. (Weld # RCH-2-2C).

ASME Code Class: ASME Code Class 1 Section Xl Edition: 2004 Edition, as amended by 10 CFR 50.55a, "Mandatory Limitations and Modifications".

Code Table: Table IWB-2500-1 Examination Category: Category B-A Examination Item Numbers: Item Number B13.30 (RPV Shell-to-Flange Weld)

Item Number B13.40 (RPV Head-to-Flange Weld)

Code Requirement From Which Relief Is Requested:

In accordance with ASME Section Xl, paragraph IWA-2232, "Ultrasonic examinations shall be conducted in accordance with Appendix I."

Further, in accordance with Appendix I, paragraph 1-21 10(b) "Ultrasonic examination of reactor vessel shell-to-flange welds, closure head-to-flange welds, and integral attachment welds shall be conducted in accordance with Article 4 of Section V, except that alternative examination beam angles may be used."

Relief Requested:

Pursuant to 10 CFR 50.55a(a)(3)(i), TVA requests relief from performing the designated vessel shell-to-flange weld and head-to-shell weld examinations in accordance with the requirements of ASME Section XI, paragraph IWA-2232, Appendix I, and the associated Article 4 of Section V methodology in accordance with paragraph 1-2110(b).

Basis For Relief:

In accordance with ASME Section XI, Subarticle IWA-2232, TVA is required to perform ultrasonic examinations (UT) of the RPV upper shell-to-flange and head-to-flange welds using Section XI, Appendix I, which in turn requires the use of the NDE methodologies and processes of ASME Section V, Article 4.

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The above listed welds are the only circumferential shell welds in the RPV that are not examined in accordance with the requirements of ASME Section Xl, Appendix VIII, as mandated in 10 CFR 50.55a with the issuance of the rule change dated September 22, 2009 (Federal Register Notice 64 FR 51370). This rule change mandated the use of ASME Section Xl, Appendix VIII, Supplements 4 and 6 for the conduct of RPV examinations.

Requests for relief are required to use the more technically-advanced Appendix VIII/PDI processes for the shell-to-flange weld exams and the closure head-to-flange weld exams, in lieu of the Section XI Appendix I and its associated Section V, Article 4 processes.

Proposed Alternative Examination:

TVA proposes the following alternative examination. In lieu of the requirements of Appendix I and its associated sub-requirements of Article 4,Section V, TVA will use the techniques, personnel, and equipment qualified to meet the requirements of ASME Section Xl Appendix VIII, Supplements 4 and 6 of the 2001 Edition, in accordance with 10 CFR 50.55a(b)(2)(xv) and, as amended by Sections 10 CFR 50.55a(b)(2)(xv)(B) through 10 CFR 50.55a(b)(2)(xv)(G), and 10 CFR 50.55a(b)(2)(xvi)(A), by following the Electric Power Research Institute's (EPRI) Performance Demonstration Initiative (PDI) processes.

Justification For Granting Relief:

ASME Section V, Article 4, describes the required techniques to be used for the Ultrasonic Test (UT) of welds in ferritic pressure vessels with wall thicknesses greater than 2 inches.

The techniques were first published in ASME Section V in the 1974 Edition, summer 1975 Addenda. The calibration techniques, recording criteria and flaw sizing methods are based upon the use of a distance-amplitude-correction (DAC) curve derived from machined reflectors in a basic calibration block. UT performed in accordance with Section V, Article 4, used recording thresholds of 50 percent DAC for the outer 80 percent of the required examination volume and 20 percent DAC from the clad/base metal interface to the inner 20 percent margin of the examination volume. Indications detected in the designated exam volume portions, with amplitudes below these thresholds, were therefore not required to be recorded. Use of the Appendix VIII/PDI processes would enhance the quality of the examination results reported because the detection sensitivity is more conservative and the procedure requires the examiner to evaluate all indications determined to be flaws regardless of their associated amplitude. The recording thresholds in Section V, Article 4, requirements and in the previously-applied guidelines of RG-1.150, Revision 1, are generic and somewhat arbitrary and do not take into consideration such factors as flaw orientation, which can influence the amplitude of UT responses.

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The EPRI Report NP-6273, "Accuracy of Ultrasonic Flaw Sizing Techniques for Reactor Pressure Vessels," dated March 1989, established that UT flaw sizing techniques based on tip diffraction are the most accurate. The qualified prescriptive-based UT procedures of ASME Section V, Article 4 have been applied in a controlled process with mockups of RPVs which contained real flaws and the results statistically analyzed according to the screening criteria in Appendix VIII of ASME Section XI. The results show that the procedures in Section V, Article 4, are less effective in detecting flaws than procedures qualified in accordance with Appendix VIII as administered by the PDI processes. Appendix VIII/PDI qualification procedures use the tip diffraction techniques for flaw sizing. The proposed alternative Appendix VIII/PDI UT methodology uses analysis tools based upon echo dynamic motion and tip diffraction criteria which has been validated, and is considered more accurate than the Section V, Article 4 processes.

UT performed in accordance with the Section V, Article 4 processes requires the use of beam angles of 00, 450, 600, and 700 with recording criteria that precipitates equipment changes.

Having to perform these process changes is time consuming and results in increased radiation exposure for the examination personnel.

Having to comply with the specific ASME Section Xl, Appendix I requirements for the RPV circumferential shell-to-flange weld and the head-to-flange weld, when the data is obtained using a less technically advanced process, results in an examination that does not provide a compensating increase in quality and safety for the higher costs and personnel exposures involved.

Past RPV shell-to-flange weld and head-to-flange weld examinations already performed at TVA plants used automated and manual UT systems operated by qualified vendors.

The examination coverage achieved during the Unit 2, 2001 exam (Cycle 11 outage, 04/03/2001) of the shell-to-flange weld (during the 2nd ISI program interval) resulted in a coverage of approximately 76.6 percent which is less than the required essentially 100 percent. Manual examination techniques were performed from the outside surfaces of the RPV during the Unit 2 examination in order to maximize the coverage. Examination coverage performed from the inside surfaces was limited due to the taper in the vessel wall at the edge of the weld area and the obstructions encountered with the guide rods and the steam nozzle plugs with the specific UT equipment used during the exam. The manual examination of the weld volume performed from the outside surfaces was limited by the flange configuration. This limited exam with a percentage of coverage of less than 90 percent was the subject of a BFN Unit 2 relief request number RR 2-IS1-14. This relief was reviewed by the NRC and found to be acceptable. A safety evaluation report (SER), on this relief, was issued by the NRC in a letter to J. A. Scalice, from A. G. Howe, dated April 3, 2003 (ADAMS Accession Number ML030970815). The examination performed on the Unit 3 RPV used a different set of newer designed UT equipment and thereby achieved a calculated coverage of 95 percent. Therefore, the Unit 3 examination results did not require the submittal and review of a relief request.

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For future Unit 2 RPV shell-to-flange weld examinations and closure head-to-flange weld examinations, TVA does not anticipate any less coverage than the required minimum of 90 percent of coverage. However, if any such limitations are encountered during the conduct of the examinations, a separate individual relief request will be submitted, as needed.

Procedures, equipment, and personnel qualified through the Appendix VIII, Supplements 4 and 6 PDI programs have shown to have a high probability of detection of flaws and are generally considered superior to the techniques employed earlier for RPV examinations. This results in increased reliability of RPV inspections and conditions where an acceptable level of quality and safety is provided with the proposed alternative methodologies. Accordingly, approval of this alternative evaluation process is requested pursuant to 10 CFR 50.55a(a)(3)(i).

BFN Unit-2 RPV shell-to-flange weld examination, utilizing PDI, is currently scheduled for the 3rd Interval, 3rd Period, Cycle 16 refueling outage, in the Spring of 2011.

Implementation Schedule And Duration:

This alternative is requested for the BFN Unit 2 Fourth Ten-Year Inspection interval as listed below:

Unit 2, Fourth Ten-Year Inspection interval (May 25, 2011 through May 24, 2021)

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Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN), Unit 2 American Society of Mechanical Engineers (ASME)Section XI, Inservice Inspection (ISI) Program Request for Relief 2-1S1-1, Updated Risk Informed Inservice Inspection Program EXECUTIVE

SUMMARY

This request for alternative proposes to adopt Risk-Informed Selection of Class 1 and Class 2 Piping Welds for Examination for the Fourth Ten-Year Inspection Interval for BFN Unit 2. The proposed process is similar to that originally submitted June 1, 2000. That request was authorized by NRC letter dated January 19, 2001, from R. P. Corella (NRC) to J. A. Scalice (TVA).

The BFN Unit 2 ISI program for the Fourth Ten-Year Inspection Interval will be based on the 2004 Edition of ASME Section Xl.

The risk-informed process provides an adequate level of quality and safety for selection of the Class 1 and Class 2 Piping Welds for examination. Therefore, pursuant to 10 CFR 50.55a(a)(3)(i) it is requested that authorization to utilize the alternative be granted.

SYSTEM/COMPONENTS(S) FOR WHICH ALTERNATIVE IS REQUESTED Class 1 and Class 2 Piping Welds CODE REQUIREMENTS Table IWB-2500-1, Examination Category B-F and Category B-J; Table IWC-2500-1, Examination Category C-F-1 and Category C-F-2 of the 2004 Edition of ASME Section XI.

ALTERNATIVE REQUESTED Authorization is requested to use a risk-informed process as an alternative for the selection of Class 1 and Class 2 Piping Welds for examination.

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BASIS FOR ALTERNATIVE The BFN Unit 2 ISI program for the Fourth Ten-Year Inspection Interval will be based on the 2004 Edition of ASME Section XI. The BFN Unit 2 ISI program for the examination of Class 1 and Class 2 piping welds is in accordance with a risk-informed process submitted June 1, 2000. That request was authorized by letter dated January 19, 2001, with enclosed Safety Evaluation. In the original submittal, the Tennessee Valley Authority (TVA) committed to review and adjust the risk ranking of piping segments as a minimum on an ASME period basis. This review and adjustment has been performed after each ASME period since approval, and the first periodic review was used as one of the bases for the development of the Nuclear Energy Institute (NEI) 04-05, "Living Program Guidance To Maintain Risk-Informed Inservice Inspection Programs For Nuclear Plant Piping Systems," dated April 2004.

ALTERNATIVE EXAMINATION In lieu of the requirements specified in the earlier "Code Requirements" section, the Risk-Informed process will be used for selection of Class 1 and Class 2 Piping Welds for examination.

JUSTIFICATION FOR GRANTING ALTERNATIVE A request for approval of an alternative Risk-Informed Inservice Inspection program for BFN Unit 3 was submitted April 23, 1999. That request was authorized by letter dated February 11, 2000, with enclosed Safety Evaluation. A request for approval of an alternative Risk-Informed Inservice Inspection (RI-ISI) program for BFN Unit 2 was submitted June 1, 2000. That request was authorized by NRC letter dated January 19, 2001. A request for approval of an updated alternative RI-ISI program for BFN Unit 3 was submitted October 19, 2005. That request was authorized by letter dated February 12, 2007, with enclosed Safety Evaluation. The BFN Unit 2 program update that is the subject of this request was developed in accordance with the process previously evaluated for BFN Units 2 and 3, including changes resulting from the review process and associated sensitivity studies.

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In the original BFN Unit 2 RI-ISI submittal, TVA committed to review and adjust the risk ranking of piping segments as a minimum on an ASME period basis. The first period of implementation of the RI-ISI program was the third period of the Second Ten-Year Inspection Interval, which ended May 25, 2001. The RI-ISI program was evaluated at the end of the Second Ten-Year Inspection Interval, with no update required, Several changes took place prior to the end of the first period of Interval 3, including revision of the Probabilistic Risk Assessment and initiation of hydrogen water chemistry/noble metals injection, so an interim review was performed in March 2002. This review and subsequent update was presented to and approved by the Expert Panel December 19, 2002. BFN Unit 2 Third Ten-Year Inspection Interval, First Period began May 25, 2001 and ended May 24, 2004. In accordance with BFN procedural requirements, an evaluation was performed following the end of that Period. Period 2 of the Third Ten-Year Inspection Interval began May 25, 2005 and ended May 24, 2008. In accordance with BFN procedural requirements, an evaluation was also performed following the end of that Period. Since the proposed program for the Fourth Ten-Year Inspection Interval is being submitted prior to the completion of the last period of the Third Ten-Year Inspection Interval, a verification of inputs to the calculation was performed as part of the update. The updated RI-ISI program resulting from these reviews is the subject of this request.

As part of the BFN Unit 1 Restart project, a RI-ISI program has been developed. Based on precedents in the rest of the industry and at the other TVA nuclear facilities, it was decided to limit the scope of the BFN Unit 1 program to Class 1 and Class 2 piping only. The updated alternative RI-ISI program for BFN Unit 3 submitted October 19, 2005 and authorized February 12, 2007 was also limited to this scope. For consistency, this revised BFN Unit 2 RI-ISI program is limited to this same scope.

In accordance with the guidance provided by NEI 04-05, a table is provided as Attachment 10 (Page 6 of 6) identifying the number of welds added to and deleted from the originally authorized RI-ISI program. The deletions from the previous program are attributable to lower failure rates due to the implementation of the hydrogen water chemistry/noble metal injection program, with the corresponding impact on Intergranular Stress Corrosion Cracking (IGSCC),

and a change in ASM Code Class boundary which resulted in the deletion of segment 2-085-031.

A new Delta Risk Evaluation was performed, and the revised RI-ISI program continues to represent a risk reduction when compared to the last deterministic ASME Section XI inspection program. The revised RI-ISI program represents a reduction of 1.869E-05 in regards to CDF and 1.596E-07 in regards to LERF.

PRA Quality The scope, level of detail, and quality of the BFN probabilistic risk assessment (PRA) is sufficient to support a technically defensible and realistic evaluation of the risk change for this proposed application.

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The BFN PRA model, dated December 7, 2009 was used to evaluate the consequences of pipe ruptures during operation in Mode 1. The base core damage frequency (CDF) and base large, early release frequency (LERF) from this version of the PRA model are 7.89E-06 /yr and 3.16E-06 /yr, respectively.

PRA model updates are scheduled for 48-month intervals. The administrative guidance for this activity is contained in Tennessee Valley Authority (TVA) administrative procedures.

The RI-ISI evaluation included a determination that the PRA model and supporting documentation accurately reflects the current plant configuration and operational practices consistent with its intended application.

After an extensive upgrade effort of the PRA for BFN, the BFN Units 1, 2 and 3 Internal Events PRA Peer Review was performed in May 2009 at the TVA offices in Chattanooga, TN, using the process described in NEI 05-04 ("Process for Performing Follow-on PRA Peer Reviews Using the ASME PRA Standard," dated January 2005), the ASME PRA Standard (ASME/ANS RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," dated 2009), and Regulatory Guide 1.200 ("An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, dated March 2009). A separate review was performed for the Internal Flooding portion of the BFN PRA in September 2009. The Internal Flooding Peer Review also used the NEI 05-04 process, the ASME PRA Standard, and Regulatory Guide 1.200, Revision 2. A team of independent PRA experts from nuclear utility groups and PRA consulting organizations carried out these Peer Review Certifications.

The purpose of these reviews was to provide a method for establishing the technical adequacy of the PRA for the spectrum of potential risk-informed plant licensing applications for which the PRA may be used. The 2009 BFN PRA Peer Reviews provided a full-scope review of the Technical Elements of the internal events, at-power PRA.

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These intensive peer reviews involved over two person-months of engineering effort by the review team and provided a comprehensive assessment of the strengths and limitations of each element of the PRA model. The Peer Review Certification of the BFN PRA model performed by Boiling Water Reactor Owners Group (BWROG) resulted in a total 125 findings for the three unit model for internal events and internal flooding. All findings from these assessments have been dispositioned. This resulted in a number of enhancements to the PRA model prior to its use to support these proposed changes. The certification team determined that with these proposed changes incorporated, the quality of all elements of the PRA model is sufficient to support "risk significant evaluations with deterministic input." As a result of the effort to incorporate the latest industry insights into the PRA model upgrades and certification peer reviews, TVA has concluded that the results of the risk evaluation are technically sound and consistent with the expectations for PRA quality set forth in Regulatory Guides 1.174 ("An Approach for Using Probabilistic Risk Assessment In Risk-Informed Decisions On Plant-Specific Changes to the Licensing Basis," Revision 1, November 2002) and 1.178 ("An Approach for Plant-Specific Risk-Informed Decisionmaking for Inservice Inspection of Piping," Revision 1, September 2003).

IMPLEMENTATION SCHEDULE The alternative will be used for BFN Unit 2 until the end of the'unit's fourth ten-year ISI program inspection interval, subject to the review and update guidance of NEI 04-05. The fourth inspection interval is currently scheduled to end on May 24, 2021.

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STRUCTURAL ELEMENT SELECTION RESULTS AND COMPARISON TO ORIGINAL PROGRAM AND PREVIOUS RI-ISI PROGRAM Original Program (1989 ASME and Generic Letter 88-01) Previous RI-ISI Program (a) (b) (c) Revised RI-ISI Program (a) (b) (c)

ASME Xl Elements (d) Auglmented Elements RI-ISI Examinations RI-ISI Examinations

  1. Dual FAC # R1.1 R1.1 R1.1 SystemSytm Segs B-J C-F-i B-_-_-F1CF2 C-F-2 A C D E G (XICredit

& Aug) (e)Sea R1.11 R1.16 R1.18 A C D E G Sags 11 6 6 88 A CC D EE G 001 MS 56 32 10 295 2 CI 2 4 30 24 2

003 FW 46 18 321 2 CI 1 10 32 11 12 063 SLC 5 5 1 1 15 068 RECIRC 16 21 10 58 10 21 ClI 1 17 C 58 0 16 1 A 1 65 0 1

7C 9 E 6 E 069 RWCU 19 6 1 8 2 3 5 CI 1 CI1 5 C 8 2 3 4 5A 5 4 4 3 1 D 2 D 3 E 070 RBCCW 17 2 071 RCIC 13 2 6 1 C12 11 2.

073 HPCI 11 5 5 14 4 CI 1 10 1 21 2C12 1 2 074 RHR 31 1 9 2 37 29 1 2 2 10 Cl1 9 C 29 1 22 28 42C 4 C 29 3 2 2 2 C12 1 D 1 D I I I 2 E 2 G 075 CS 15 8 5 6 12 14 1 14 CI 1 6 C 14 1 15 8 C 14 1 085 CRD 31 6 3 3 14 4 50 Cl 11 6CI Total 105 1 . 1 4 11 1 Examinations 9 98 13 85 11 109 6 16 2 51 14 6 6 2 9Cl 8 16 6 2 1 6 2 2 C2 9C12 2 Total Elements 9 392 173 940 Notes: (a) Svstem pressure test rearuirements and VT-2 visual examinatinns shall cnntinue to he nerformed in all ASME Code Class 1 2 and 3 svstems (b) Augmented programs including Flow Accelerated Corrosion (FAC) and Reactor Nozzle Thermal Fatigue Cracking (NUREG-0619) continue (c) Augmented program for IGSCC Categories C through G (VIP-075, Generic Letter 88-01, NUREG-0313) continues.

(d) The current ASME Section Xl ISI Program examines a minimum of 25% of the Class 1 and a minimum of 7.5% of the Class 2 elements (e) The FAC Augmented Program examines approximately 10% of the identified locations each refueling outage.

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Tennessee Valley Authority Browns Ferry Nuclear Plant (BFN)

Unit 2 American Society of Mechanical Engineers,Section XI Inservice Inspection Program, Unit 2 Fourth Ten Year Inspection Interval Request for Relief 2-ISI-43 Executive Summary: In accordance with 10 CFR 50.55a(a)(3)(i), the Tennessee Valley Authority (TVA) is requesting relief from inservice inspection requirements of the 2004 Edition of Section XI, Table IWB-2500-1, "Examination Category B-D, Full Penetration Welded Nozzles in Vessels-Inspection Program B," Item No.B3.90, "Reactor Vessel Nozzle to Vessel Welds," and Item No. B3.1 00, "Reactor Vessel Nozzle Inner Radius Section" of the ASME Boiler and Pressure Vessel Code for BFN Unit 2.

As an alternative, TVA proposes to use ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds," in lieu of the examination requirements of ASME Section XI, Table IWB-2500-1, "Examination Categories," to allow reduced percentage requirements for Nozzle-to-Vessel Weld and Inner Radius Section examinations.

Unit: Browns Ferry Nuclear Plant, Unit 2 ASME Code Components Affected: Reactor Pressure Vessel (RPV), Nozzle-to-Vessel Welds and RPV Nozzle Inner Radius Sections:

Reactor Recirculation Inlet Nozzles, N2A, N2B, N2C, N2D, N2E, N2F, N2G, N2H, N2J, and N2K Main Steam Nozzles, N3A, N3B, N3C, and N3D Core Spray Nozzles, N5A and N5B Reactor Pressure Vessel (RPV) Head Nozzles, N6A, N6B, and N7 Jet Pump Instrumentation Nozzles, N8A and N8B ASME Code Class: ASME Code Class 1 Section Xl Edition: 2004 Edition for BFN Unit 2. Additionally, for ultrasonic examinations,Section XI, Appendix VIII, "Performance Demonstration for Ultrasonic Examination Systems,"

of the 2001 Edition is implemented as required and modified by 10 CFR 50.55a(b)(2)(xv).

Code Table: Table IWB-2500-1, "Examination Categories" Code Examination Category: B-D, "Full Penetration Welded Nozzles in Vessels" -

Inspection Program B

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Code Examination Item Number: B3.90, "Nozzle-To-Vessel Welds" and B3.100, "Nozzle Inside Radius Section" Code Requirement: The 2004 Edition of ASME Section Xl, Table IWB-2500-1, Examination Category B-D, Item No. B3.90 and Item No. B3.100, require a volumetric examination of 100 percent each ten-year inspection interval of the reactor pressure vessel (RPV) nozzle-to-shell welds and nozzle inner radius section.

Reason for Request: The RPV Nozzle-to-Vessel Shell Welds and RPV Nozzle Inner Radius Sections listed in Attachment A are scheduled to be examined during the fourth inspection interval for BFN Unit 2. The proposed alternative provides an acceptable level of quality and safety and the reduction in examination scope could provide a total reduction in personnel radiation exposure of as much as 7.18 Person-REM (see Attachment A) over the fourth inspection interval for BFN Unit 2.

Proposed Alternative and Basis for Use: Pursuant to 10 CFR 50.55a(a)(3)(i), TVA is requesting relief from performing the required examinations on 100 percent of the identified nozzles. As an alternative, TVA proposes to examine 25 percent of the nozzle-to-vessel welds and nozzle inner radius sections, except for the Recirculation Outlet welds, including at least one nozzle from each system and nominal pipe size in accordance with ASME Code Case N-702. For the nozzles identified in Attachment A, the number of components to be examined from each group is provided in Table 1 below. This relief is not requested to be applied to the Recirculation Outlet Nozzle welds.

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Table 1 Number of Nozzles / Inner Radius to be Examined System and Nozzle Number of Nozzles Number to be Comments ID Examined Reactor 10 3 See Attachment C Recirculation Inlet for history of (N2A, N2B, N2C, previous N2D, N2E, N2F, examinations.

N2G, N2H, N2J, and N2K)

Main Steam 4 1 See Attachment C (N3A, N3B, N3C, for history of and N3D) previous examinations.

Core Spray 2 1 See Attachment C (N5A and N5B) for history of previous examinations.

Reactor Pressure 3 1 See Attachment C Vessel (RPV) Head for history of (N6A, N6B, and N7) previous examinations.

Jet Pump 2 1 See Attachment C Instrumentation (N8A for history of and N8B) previous examinations.

Code Case N-702 states that a VT-1 visual examination may be used in lieu of volumetric examination for the inner radii (Item B3.100). TVA is currently using Code Case N-648-1, Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles,Section XI Division 1, subject to the conditions provided in Regulatory Guide 1.147, Revision 15, dated October 2007.

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Basis For Relief: Electric Power Research Institute (EPRI) Technical Report 1003557, dated October 2002, "BWRVIP-1 08: Boiling Water Reactor Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozzle-to-Vessel Shell Welds and Nozzle Blend Radius," provides the basis for ASME Code Case N-702. EPRI Technical Report 1003557 found that failure probabilities at the nozzle blend radius region and nozzle-to-vessel shell weld due to a Low Temperature Overpressure event are very low (i.e., <1 x 10-6 for 40 years) with or without inservice inspection. The report concludes that inspection of 25 percent of each nozzle type is technically justified.'

The NRC documented their review of the EPRI report in an NRC Safety Evaluation Report (SER) dated December 19, 2007. In Section 5.0, "Plant Specific Applicability," of the SER, the NRC stated that each licensee who plans to request relief from the ASME Code, Section Xl requirements for RPV nozzle-to-vessel shell welds and nozzle inner radius sections may reference the BWRVIP-1 08 report as the technical basis for the use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant-specific applicability for the BWRVIP-1 08 report to each unit in the relief request by showing that all the general and nozzle-specific criteria addressed below are satisfied (See Attachment B):

(1) The maximum Reactor Pressure Vessel (RPV) heatup/cooldown rate is limited to less than 1150 F per hour. The BFN Unit 2 Technical Specification (TS) Surveillance Requirement (SR) 3.4.9.1 .b limits Reactor Coolant System (RCS) heatup and cooldown rates to < 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> for pressure and temperature limits specified in TS Figure 3.4.9-1, "Pressure/Temperature Limits for Mechanical Heatup, Cooldown following Shutdown, and Reactor Critical Operations." For the pressure and temperature limits specified in TS Figure 3.4.9-2, "Pressure/Temperature Limits for In-Service Leak and Hydrostatic Testing," Note 2 to BFN Unit 2, TS SR 3.4.9.1 limits RCS heatup and cooldown rates to < 150 F/hour.

The BFN Unit 2 surveillance procedures that require monitoring of reactor vessel heatup/cooldown (Surveillance Procedure 2-SR-3.4.9.1 (1) and 2-SR-3.4.9.1(2)) limit the heatup and cooldown rates to less than or equal to 100°F/hr for BFN Unit 2 Figure 3.4.9-1 pressure/temperature limits and less than or equal to 150 F/hr for TS Figure 3.4.9-2 pressure/temperature limits.

For the Recirculation Inlet Nozzles the following criteria must be met:

(2) (pr/t)/CRPv<1.15, the calculation for BFN Recirculation Inlet (N2) Nozzles results in 1.0986 which is less than 1.15 which satisfies Criterion 2.

(3) [p(r 0 2 +ri 2 )/(r0 2 -ri2 )]/CNozZLE<1 .15, the calculation for BFN N2 Nozzles results in 1.0012 which is less than 1.15 which satisfies Criterion 3.

For the Recirculation Outlet Nozzles the following criteria must be met:

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 200 of 205 Attachment 11 (Page 5 of 10)

(4) (pr/t)/CRPV< 1.15, the calculation for BFN Recirculation Outlet (N1) Nozzles results in 1.3134 which is higher than 1.15. Therefore, Criterion 4 is not satisfied for the BFN N1 nozzles. Therefore, these nozzles are not in the scope of this relief request.

(5) [p(r 0 2 +ri 2 )/(ro2-ri 2 )]/CNozZLE< 1.15, the calculation for the BFN N1 Nozzles results in 1.0751 which is less than 1.15 which satisfies Criterion 5.

Based upon the above information, all RPV nozzle-to-vessel shell welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles, meet the BWRVIP-108 Report criteria and therefore Code Case N-702 is applicable. The Recirculation Outlet Nozzles do not meet all of the BWRVIP-108 Report criteria. Therefore, Code Case N-702 is not applicable to these nozzles. As such, this relief is not requested to be applied to the Recirculation Outlet Nozzles. See Attachment B for details.

Therefore, pursuant to 10 CFR 50.55a(a)(3)(i), TVA considers that the use of Code Case N-702 provides an acceptable level of quality and safety for all RPV nozzle-to-vessel shell welds and nozzle inner radii sections for each BFN unit, with the exception of the Recirculation Outlet Nozzles (N1A and N1B).

Implementation Schedule:

This alternative is requested for the BFN Unit 2 Fourth Ten-Year Inspection interval as listed below:

Unit 2, Fourth Ten-Year Inspection interval (May 25, 2011 through May 24, 2021)

Precedent:

The NRC has approved a similar request for the Duane Arnold Energy Center. This approval is documented in an NRC safety evaluation dated August 28, 2008 (ML082040046).

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 201 of 205 Attachment 11 (Page 6 of 10)

Attachment A - Applicable Nozzles I Inner Radius for Unit 2 RPV Nozzle Component ID Category Item Number Nozzle ID (NOM) Comments/Dose Estimates Man-hr Dose rate Dose S

N2A N2A-NV B-D B3.90 11.56" 8 85 0.680 N2A-IR B-D B3.100 11.56" N2B N2B-NV B-D B3.90 11.56" 8 85 0.680 N2B-IR B-D B3.100 11.56" N2C N2C-NV B-D B3.90 11.56" 8 85 0.680 N2C-IR B-D B3.100 11.56" N2D N2D-NV B-D B3.90 11.56" 8 85 0.680 N2D-IR B-D B3.100 11.56" N2E N2E-NV B-D B3.90 11.56" 8 85 0.680 N2E-IR B-D B3.100 11.56" N2F N2F-NV B-D B3.90 11.56" 8 85 0.680 N2F-IR B-D B3.100 11.56" N2G N2G-NV B-D B3.90 11.56" 8 85 0.680 N2G-IR B-D B3.100 11.56" N2H N2H-NV B-D B3.90 11.56" 8 85 0.680 N2H-IR B-D B3.100 11.56" N2J N2J-NV B-D B3.90 11.56" 8 85 0.680 N2J-IR B-D B3.100 11.56" N2K N2K-NV B-D B3.90 11.56" 8 85 0.680 N2K-IR B-D B3.100 11.56" N3A N3A-NV B-D B3.90 23.75" 10 60 0.600 N3A-IR B-D B3.100 23.75" N3B N3B-NV B-D B3.90 23.75" 10 60 0.600 N3B-IR B-D B3.100 23.75" N3C N3C-NV B-D B3.90 23.75" 10 60 0.600 N3C-IR B-D B3.100 23.75"

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 202 of 205 Attachment 11 (Page 7 of 10)

RPV Nozzle Component ID Category Item Number Nozzle ID (NOM) Comments/Dose Estimates Man-hr Dose rate Dose S

N3D N3D-NV B-D B3.90 23.75" 10 60 0.600 N3D-IR B-D B3.100 23.75" N5A N5A-NV B-D B3.90 8.78" 6 30 0.180 N5A-IR B-D B3.100 8.78" N5B N5B-NV B-D B3.90 8.78" 6 30 0.180 N5B-IR B-D B3.100 8.78" N6A N6A-NV B-D B3.90 6-7/32" 6 15 0.090 N6A-IR B-D B3.100 6-7/32" N6B N6B-NV B-D B3.90 6-7/32" 6 15 0.090 N6B-IR B-D B3.100 6-7/32" N7 N7-NV B-D B3.90 4-1/4" 5 30 0.150 N7-IR B-D B3.100 4-1/4" N8A N8A-NV B-D B3.90 3-13/16" 5 40 0.200 N8A-IR B-D B3.100 3-13/16" N8B N8B-NV B-D B3.90 3-13/16" 5 40 0.200 N8B-IR B-D B3.100 3-13/16" Total Dose Savings for Remainder of the 2nd Interval (Note 1) 7.18 Rem Note 1: Dose savings of 7.18 Rem is based on performance of the reduced number of components to be examined as shown in Table 1.

BFN Inservice Inspection and Risk - 2-SI-4.6.G Unit 2 Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 203 of 205 Attachment 11 (Page 8 of 10)

Unit 2 (1) The maximum Reactor Pressure Vessel (RPV) heatup / cooldown rate is limited to less than 11 5°F/hour.

Technical Specification Surveillance Requirement SR 3.4.9.1 .b limits RCS heatup and cooldown rates to less than or equal to 100°F in any 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period.

Recirculation Inlet Nozzles Recirculation Outlet Nozzles (2) (pr/t)/CRPV<1.15 (4) (pr/t)/CRPV<1.15 p = RPV normal operating pressure 1035 p = RPV normal operating 1035 pressure r = RPV inner radius 125-11/16" r = RPV inner radius 125-11/16" t = RPV wall thickness 6.125" t = RPV wall thickness 6.125" CRPV 19332 CRPV 16171 1.0986 < 1.15 1.3134 > 1.15: Criterion Not Satisfied (3) [p(r 02 + ri2 )/(ro2 2

- ri )]/CNOZZLE < 1.15 (5) [p(ro2 2

+ ri2 )/(ro - ri )]/CNOZZLE <

2 1.15 p = RPV normal operating pressure 1035 p = RPV normal operating 1035 pressure

r. = nozzle outer radius 12.5" r, = nozzle outer radius 26.5" ri = nozzle inner radius 5.941" r = nozzle inner radius 15.566" CNOZZLE 1637 CNOZZLE 1977 1.0012 < 1.15 1.0751 < 1.15

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Unit 2 Applicable Nozzles(Note 1)

Component ID Category Item System Nominal CYCLE DATE Comments Number Number Pipe Size N2A-NV B-D B3.90 Recirc Inlet 12" 14 3/19/07 PDI Exam N2A-IR B-D B3.100 Recirc Inlet 12" 14 3/16/07 2

N2B-NV(Note ) B-D B3.90 Recirc Inlet 12" 12 14 3/3/03 3/19/07 PDI Exam N2B-IR B-D B3.100 Recirc Inlet 12" 12 3/7/03 N2C-NV B-D B3.90 Recirc Inlet 12" 14 3/19/07 PDI Exam N2C-IR B-D B3.100 Recirc Inlet 12" 14 3/16/07 N2D-NV(Note 2 ) B-D B3.90 Recirc Inlet 12" 14 3/22/07 PDI Exam N2D-IR B-D B3.100 Recirc Inlet 12" N2E-NV B-D B3.90 N2E-IR B-D B3.100 N2F-NV B-D B3.90 Recirc Inlet 12" 12 3/3/03 PDI Exam N2F-IR B-D B3.100 Recirc Inlet 12" 12 3/7/03 N2G-NV B-D B3.90 Recirc Inlet 12" 14 3/13/07 PDI Exam N2G-IR B-D B3.100 Recirc Inlet 12" 14 1 3/16/07 N2H-NV B-D B3.90 Recirc Inlet 12" 14 3/20/07 PDI Exam N2H-IR B-D B3.100 Recirc Inlet 12" 14 3/16/07 N2J-NV B-D B3.90 Recirc Inlet 12" 12 3/3/03 PDI Exam N2J-IR B-D B3.100 Recirc Inlet 12" 12 3/7/03 N2K-NV(Note 2 ) B-D B3.90 Recirc Inlet 12" 14 3/21/07 PDI Exam N2K-IR B-D B3.100 Recirc Inlet 12" 1 N3ANV(Note 2) B-D B3.90 Main 26" 14 3/22/07 PDI Exam Steam N3A-IR B-D B3.100 Main 26" Steam N3B-NV B-D B3.90 Main 26" 14 3/15/07 PDI Exam Steam N3B-IR B-D B3.100 Main 26" 14 3/16/07 Steam N3CNV(Note 2 ) B-D B3.90 Main 26" 14 3/22/07 PDI Exam Steam N3C-IR B-D B3.100 Main 26" Steam N3D-NV B-D B3.90 Main 26" 12 3/1/03 PDI Exam Steam N3D-IR B-D B3.100 Main 26" 12 3/3/03 Steam N5A-NV B-D B3.90 Core Spray 10" 14 3/14/07 PDI Exam N5A-IR B-D B3.100 Core Spray 10" 14 3/16/07 N5B-NV B-D B3.90 Core Spray 10" 14 3/23/07 PDI Exam N5B-IR B-D B3.100 Core Spray 10" 14 1 3/16/07

Inservice Inspection and Risk - 2-SI-4.6.G Informed Inservice Inspection Program Rev. 0040 Unit 2 Page 205 of 205 Attachment 11 (Page 10 of 10)

Unit 2 Applicable Nozzles(Note 1)

Component ID Category Item System Nominal CYCLE DATE Comments Number Number Pipe Size N6A-NV B-D B3.90 Head Vent 6" 12 2/28/03 PDI Exam N6A-IR B-D B3.100 Head Vent 6" 12 3/1/03 N6B-NV B-D B3.90 Head Vent 6" N6B-IR B-D B3.100 Head Vent 6" N7-NV B-D B3.90 Head Vent 4" 14 3/2/07 PDI Exam N7-IR B-D B3.100 Head Vent 4" 14 3/2/07 N8A-NV B-D B3.90 Jet Pump 4" 12 3/5/03 PDI Exam Inst N8A-IR B-D B3.100 Jet Pump 4" 12 3/7/03 Inst N8B-NV(NOte 2 ) B-D B3.90 Jet Pump 4" 14 3/22/07 PDI Exam Inst N8B-IR B-D B3.100 Jet Pump 4" I _Inst Note:1 Examinations listed for Unit 2 in this attachment are for the current Ten-Year ISI interval (i.e.,

Third Ten Year Inspection Interval). Examinations for the nozzle-to-vessel welds and inner radius sections in previous ISI intervals were completed in accordance with ASME Section XI Code requirements.

Note 2: RPV Nozzles; N2B-NV, N2D-NV, N2K-NV, N3A-NV, N3C-NV, and N8B-NV were expanded scope examinations for Cycle 14 due to the N9-NV failing UT examination. The N9-NV indication did not meet IWB-3500 criteria. It was determined to be a sub-surface flaw not serviced induced. The flaw was evaluated and accepted as is for continued operation.

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