Duane Arnold Energy Center - Fifth Inservice Inspection ... · March 7, 2017 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Duane Arnold

March 7, 2017

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

Duane Arnold Energy Center Docket No. 50-331 Renewed Op. License No. DPR-49

Fifth lnservice Inspection Interval Program Plan

NEXTeraM ENERGY~

NG-17-0042 . 10 CFR 50.55a

The Fifth lnservice Inspection Interval Program Plan for the Duane Arnold Energy Center (DAEC) began on November 1, 2016 and ends on October 31, 2026. Pursuant to 10 CF'R 50.55a(g)(4)(ii), the ISi Program Plan for DAEG is based on the requirements of Section XI of the ASME Boiler and Pressure Vessel Code, 2007 Edition with Addenda through 2008.

NextEra Energy Duane Arnold, LLC (hereafter NextEra Energy Duane Arnold) is hereby submitting the Fifth lnservice Inspection Interval Program Plan for DAEC. The program is provided in the Enclosure. Four Relief Requests for the Fifth 10-Year lnservice Interval are included in Appendix A to the Enclosure. NextEra Energy Duane Arnold respectfully requests approval of the Relief Requests within one year. A fifth Relief Request for the Fifth 10-Year lnservice Interval will be submitted under different correspondence at a later date.

If you have any questions, please contact Michael Davis at (319) 851-7032.

Dean Curtland Site Director, Duane Arnold Energy Center . NextEra Energy Duane Arnold, LLC

Enclosure

cc: Administrator, Region Ill, USNRC Project Manager, DAEC, USNRC Senior Resident Inspector, DAEC, USNRC

NextEra Energy Duane Arnold, LLC, 3277 DAEC Road, Palo, IA 52324

Enclosure to NG-17-0042

Duane Arnold Energy Center Fifth lnservice Inspection. Interval Program Plan

· 82 pages to follow

NextEra Energy Nuclear Engineering Department

Fleet Programs Engineering Group Inspection Section

15430 Endeavor Drive Jupiter, FL 33478

Fifth lnservice Inspection Interval Program Plan

for

Duane Arnold Energy Center 3277 DAEC Road Palo, Iowa 52324

Commercial Service Date:

February 1, 1975

USNRC Docket Number:

Document Number: 5th lnterval-ISl-PDA-Program Plan Rev. 1

Prepared by: . Reference EC 288247 for Signature

Reviewed by: Reference EC 288247 for Signature

· Reviewed by: Reference EC 288247 for Signature Supervisor - Fleet Programs Engineering

Approved by: Reference EC 288247 for Signature Manager - Fleet Programs Engineering .

Duane Arnold Energy Center ISi Program Document No. 5th lnterval-ISl-PDA-Program Plan

Revision 1 January 27, 2017

Record of Revision

Rev Date Affected

Reason for Revision No. Pages

0 October 28, 2016 Entire Original Issue of the Fifth 10-Year Inspection Document Interval Program Plan

2, 3, 5, 6, 12, Prior to submitting to the NRC, editorial

1 January 27, 2017 15, 17, 24, 29, changes, all references to PWR only systems 31 , 32-36, 44 and submittal date was added for Relief Request and A2 1, 2, 3 and 4

2 of 50


List of Effective Pages

Section Pagels

2, 3, 5, 6, 12, 24, 15,

Program Plan Text 17,29,31, 32-36 and

. 44

Appendix A A2

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Date


January 27 2017

January 27, 2017


Table of Contents

Revision 1 January 27, ~017

Cover Page ................................................................................................................................. 1

Record of Revision ....................................................................................................................... 2

List of Effective Pages .......................................................... '. ........................................................ 3

Table of Contents .......................................................................................................................... 4

List of Tables ....................................................................... : ....................... : ................................ 7

Abbreviations ................................................................................................................................ 8

Abstract. ..... ~ .............................................................................................................................. 11

1.0 Introduction ...................................................................................................................... 12 1.1 ISi Program Plan Development.. ................................................ ~ .................. : ........ 12 1.2 Other ISi Programs ............... : ................................................................................... 12

1.2.1 lnservice Testing Program (IST) ........................................................... : ... 12 1.2.2 Snubber Program .......................... , ......... -................................................. 12 1.2.3 Pressure Test Program .................. : ............................ : ...... : .......... : ............ 12 1.2.4 Repair and Replacement Program ........................ -'· .................................. 13 1 :2.5 M~tal Containment lnservice Inspection Program (IWE) ........................... 13

1.3 Construction Permit ................................................................................. : .............. 13 1.4 Commercial Service Dates ........................................................... : ...... : ................. 13 1.5 Background ............................................................................................................. 13

1.5.1 Preservice Inspection ... : ..................... ~·-·············:······································ 13 1.5.2 First lnservice Inspection Interval. ............................................................. 14 1.5.3 Second lnservice Inspection Interval. ........................................................ 14 1.5.4 Third lnservice Inspection Interval ............................................................ 15 1.5.5 Fourth lnservice Inspection Interval ................................................. : .. : ...... 15

. 1.5.6 Fifth lnservice Inspection Interval .................................................... : ....... : 15 1.6 Applicable Editions and Addenda of ASME Section XI ......................... , ................ 16

1.6.1 10CFR50.55a Code of Federal Regulations Conditions ..... : ....................... 16 1. 7 System Classification .............................................................................................. 18

1. 7 .1 Optional Construction ........ , ...................................................................... 19 1. 7:2 Containment Penetrations·: ....................................................................... 19 1.7.3 Class MC Components ...................................................................... : ...... 19

1.8 Inspection Program Plan ........................................................................................ 19 1.9 Regulatory Guides ................................................ : ................................................ 20 1.10 ASME Section XI Code Cases ............................................................................... 20 1.11 Branch Technical Position MEB 3-4 ........................................................................ 22 1.12. Plant Life Extension ................................................................................................ 22

1.12.1 Licensing Renewal Commitment Documents ............................................ 22 1.13 Successive Examinations ...................................................................................... 24 1.14 NOE Examinations and Personnel Qualification/Certification ................................. 24

1.14.1 Alternative Examinations (IWA-4250(b)(2) and IWA-4521) ......................... 24 1.14.2 Certification and recertification (IWA-2314) ................................................. 24 · 1.14.3 Appendix VIII Requirements ........................................................................ 24

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Duane Arnold Energy Center ISi Prograrn Document No. 5th lnterval-ISl-PDA-Program Plan


2.0 Risk Informed (RI) Requirements ...................................................................................... 25

3.0 Development of Class 1 Examination Plan ......... :············································ .................. 25 3.1 Class 1 Code Exemptions ...................................................................................... 26 3.2 Component Examination Basis .............................................................................. 26

3.2.1 3.2.2 3.2.3 3.2.4 3.2.5 3.2.6 3.2.7 3.2.8 3.2.9 3.2.10 3.2.11 3.2.12 3.2.13 3.2.14 3.2.15

Category B-A ............................................................................................ 27 Category B-B ............................................................................................ 27 Category B-D ............................................................ · ................... -............. 28 Category B-F ............................................................................................ 28 Category B-G-1 .......................................................................................... 28 Category B-G-2: ....................... · ................................................................. 29 Category B-J ....................................... , ...................................................... 29 . c·ategor:y .. B-K ........................................................ ····················' ............. · .. _30 Category B-L-2 ...................................................................... : .................. 31 Category B-M~2 .... .-................................. _ ....................................... : .......... 31 Category B-N-1 ........................................................... , .................... · ......... 32 Category B-N-2 ........................................................ .-....................... : ........ 32 Category B:.N-3 ..... .-.................................................. , ................................ 33 Category B-0 ............................................................................................ 33 Category B-P .................. · .............. .-.............................. : .. : ......................... 33

4.0 Development of Class 2 Examination Plan ............................................................ : ........... 33 4. t Class 2 Code Exemptions ................................................................................. ~ ..... 33

4.1.1 IWC-12.21 ................................................ -................................................. 33 4.1.2 IWC-1222 ................................................................................................. 34 4~ 1.3 IWC-1223 .:· .................................................................. ~ .. -........ _. ..... · ... _..: ...... ·35

4.2 Component Examination Basis ................................... ~ ............... : .......................... 35 4.2.1 Category C-A .......... : ................................................................................. 36 4.2.2 Category C-B .......................................................... .-................................. 36 4.2.3 Category C-C ............................................................................ · ................ 36 4.2.4 Category C-0 ............................................................................................. 37 4.2.5 Category C-F-1 ........ .-· ................................... : ....... · ......................... ,: .......... 37 4.2.6 Category C-F-2 ....................................... , .................... , ..................... '. ... , .. 37 4.2. 7 Category C-H ............................................................................................ 37

5.0 Development of Class 3 Examination Plan ......... : ............ , ................................................. 38 5.1 Class 3 Code Exemptions: ................................... : ...................... ~ ........................ ~. 38

5.1.1 IWD-1220 ........ · .......................................................................... _ ............... 38 5.1.2 IWD-5222(c) ·····················································'······································· 39

5.2 Component Examination Basis .............................................................................. 39 5.2.1 Category D-A ......................................................... , .................................. 40 5.2.2 Category D-B .............. ~ ............ · ................................................................. 40

6.0 IWE Metal Containment Requirements ............. : ................................................................ 40

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7.0 Development of Component Supports Examination Plan .................................................. 40 7.1 Code Exemptions for Supports .............................................................................. 40 7.2 Support Examination Basis .................................................................................... 41

7.2.1 Category F-A ............................................................................................ 41 7 .2.2 Item Number Sufffixes .............................................................................. 41

7.3 Snubbers ........................................ ~ ........ , ............................................................. 42 8.0 Augmented and Other Programs ....................................................................................... 42

9.0 Evaluation/Acceptance Criteria .......................................................................................... 44 9.1 Supplemental Examinations .................................................................................. 44 9.2 Additional Examinations ......................................................................................... 44 9.3 Successive Inspections for Components ............................................................... 44

9.3.1 Class 1 Components [IWB-2420(b)] .............................................................. 44 9.3.2 Class 2 Components [IWC-2420(b)] .............................................................. 45 9.3.3 Class 3 Components [IWD-2420(b)] .............................................................. 45 9.3.4 Component Supports [IWF-2420(b)] ............................................................. 45

10.0 Repair/Replacement Activities ........................................................................ , .................. 45

11. 0 Relief Requests ................................................................ : ................................ ~ ... ·.· ............ 45

12.0 ISi Boundary Classifications ............................................................................................ :. 46

13.0 Addition of Welds, Components and Component Supports: .... : .................................... , .... 48

14.0 Records .................... , ............................................ · .......................................................... 48 14.1 General ................................................................................................................. 48 14.2 Nondestructive Examinations ........................................................... : ..................... 48 14.3 Reports ............................. : ................................................................ : .............. · .... 48 14.4 lnservice Inspection Summary Report ................................................................... 49

. 14.5 NIS-2 or NIS-2A Reports ................................................................................ : ...... 49

1s:o References .................................................................. : .................................................... 49

Appendix A - Relief Requests ....................................................................................................... A 1

Relief Request RR-01 Extension of Permanent Relief from Ultrasonic Examination

of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term .................................................... A3

Relief Request RR-02 Alternative Requirements for Buried Piping and Components ............... A11

Relief Request RR-03 Alternative Requirements for Nozzle Inner Radius and Nozzl~-to-Shell Welds .......................................................................... A16

Relief Request RR-04 Alternative for Seal Weld Proced.ure Qualification ................................... A27 · Relief Request RR-05 RI-ISi Program (to be supplied) ....................................... , .............. A32

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List of Tables

Table 1

Table 2

Table 3

Table 4

Table 5

Table 6

Table 7

Table 8

Table 9

Table 10.

Table 11

Table 12

Table 13

First lnservice Interval Dates .................................................................................. 14

Second lnservice Interval Dates ............................................................................ 14

Third lnservice Interval Dates ................................................................................ 15

Fourth lnservice Interval Dates ................ , ............................................................. 15

Fifth lnservice Interval Dates .................................................................................. 16

USNRC Regulatory Guides ............................................................................... , ... 20

Applicable Code Cases .......................................................................................... 20

Listing of Class 1 Exempt Systems (Portions) .......................... ~ ............................. 26

Listing of Class 1 Valves by Group .................................................. :~ ..................... 32

· Listing of Class 2 Exempt Systems ........................................................................ 35

Listing of Class 3 Exempt Systems ........................................................................ 39

ISi Boundary Classification Drawings .................................................................... 46

Fifth Interval Relief Requests ................. :·······························································A2

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---- ---- -- ·-·:--~--= ·:::----~· __ -.:,::::~---=:-=---~---~~~- =-=- ::--


ANll

ANSI

ASME

BC

B& PV

BWR

CFR.

CHR

CPS

.CRD

CRS

cs

CTMT

CV

cw

DAEC

DN

ECCS

ECT

ENG

FPL

FPS

Abbreviations

Authorized Nuclear lriservice Inspector

American Nuclear Standard Institute

American Society of Mechanical Engineers

Branch Connection

Boiler & Pressure Vessel

Boiling Water Reactor

Code of Federal Regulations

Containment Heat Removal

Code Programs Section

Control Rod Drive

Code Required Surface .

Core Spray System

Containment

Control Valve

Clockwise

Duane Arnold Energy Center

Diameter Nominal

Emergency Core Cooling Sys~er:n

Eddy Current Testing

Nuclear Engineering

Florida Power & Light Company

Fuel Pool System

8 of 50


I


Abbreviations

FW Feedwater System

HPCI High Pressure Coolant Injection

HS High Stress

HX Heat Exchanger

ID Identification

IE Inspection and Enforcement

ISi lnservice Inspection

IST lnservice Testing

JPN Juno Nuclear Engineering

LER License Event Report

LS Long Sear:n

MOV Motor Operated Valve

MRP Material Reliability Program

MSIV Main Steam Isolation Valve

MS Main Steam System

MT Magnetic Particle Testing

NIA Not Applicable

NOE Nondestructive Examination

NPS Nominal Pipe Size

PDA Plant Duane Arnold

PDS Program Plan Boundary Drawings

P&ID Piping and Instrumentation Diagram

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-----=- ·-- --...:_-::: ~= ":........_-:'-:-="-.....'.'.'....:::·~-'::.= ::._:-.".:--:--- ~-~--:---~

Duane Arnold Energy Center ISi Program Document No. 5th lnterval~ISl-PDA-Program Plan

POV

PSI

PT

QA

QC

QP

RC

RHR

RCIC

. RI-ISi

RPV

RWCU

SD

SRP

T

TE

USN RC

UFSAR

UT

VT

Abbreviations

Pneumatic Operated Valve

Preservice Inspection.

Liquid Penetrant Testing

Quality Assurance

Quality Control

Quality Procedure

Reactor Coolant

Residual Heat Removal

Reactor Core Isolation Cooling

Risk Informed lnservice Inspection

Reactor Pressure Vessel

Reactor Water Cleanup

Structural Discontinuity

Standard Review Plan

Thickness of Component, Pipe, etc.

Terminal End

United States NuClear Regulatory Commission

Updated Final Safety Analysis Report

Ultrasonic Testing

Visual Testing

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Abstract


This document describes the Class 1, 2, and 3 lnservice Inspection (ISi) Plan Fifth 10-Year lnservice Inspection Interval for the Duane Arnold Energy Center (Duane Arnold /PDA).

This Program Plan was developed and prepared to meet the requirements of the American Society of Mechanical Engineers, Boiler and Pressure Vessel Code, Section XI, 2007 Edition through 2008 Addenda, and 1 OCFR50.55a for Class 1, 2, and 3 systems. This Program Plan is subject to the conditions of 10CFR50.55a(b)(2), except design and access provisions and preservice examination requirements. This Plan identifies those components and/or systems and their supports that are subject to examination and testing.

Wh.ere applicable, ASME Code Cases are incorporated. The Code Cases used are either approved through publication in 10CFR50.55a, NRC Regulatory Guide 1.147, or are included in a Relief Request

Other alternatives to the Code requirements have been included as relief requests, or they reference specific NRC regulations. Areas where Code compliance is not possible are also . included as Relief Requests, along with proposed alternatives.

The ISi Programs for Containment, lnservice Pressure Tests, and Snubber Examinations are covered under separate plant documents. General requirements for these programs are included. · for completeness.

. . . . . . .

Additional requirements for augmented examinations are addressed. The ISi Program does not require these examinations; however, due to the nature of the augmented requirements, these programs have been included and administered at the request of the plant. ·

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1.0 · Introduction

Duane Arnold is a General Electric Mark 1 Boiling Water Reactor. NextEra Energy Resources, LLC (NextEra) is the Owner of Record. The commercial operation date for Duane Arnold was February 1, 1975.

1.1 ISi Program Plan Development

This document details the lnservice Inspection Plan of Class 1, 2, and 3 components for the Fifth 10-Year lnservice Inspection Interval for Duane Arnold.

The schedule of examinations is locat~d in the ISi component database. The IDDEAL management system is used to develop and track the outage schedule. The isometrics used for locations of welds during examinations are located at the Duane Arnold site. The schedule and isometrics are separate controlled documents.

lnservice examination of components and system pressure tests conducted during successive 120-month inspection intervals must comply with the requirements of the latest edition and· addenda of the Code incorporated by reference in 1 OCFR50.55a 12 months prior to the start of the 120-month inspection interval. These requirements are subject. to the limitations and modifications listed within 1 OCFR50.55a.

This Program Plan reflects the lnservice Inspection requirements of ASME Section XI, 2007 Edition through 2008 Addenda as modified by 1 OCFR50.55a.

1.2 Other ISi Programs

This document does not address every aspect of lnsenlice Inspection. The following text details the examinati'on and testing requirements of those components and parts covered by other documents.

1.2.1 · lnservice Testing Program (IST.)

The program for lnservice Testing of Class 1, Class 2, and Class 3 Pumps and Valves is covered by the Duane Arnold lnservice Testing (IST) Program, which is submitted and approved separately.

1.2.2 Snubber Program

The program for the examination and testing of safety-related snubbers is addressed by Duane Arnold plant procedures.

1.2.3 Pressure Test Program

The program for lnservice System Pressure Testing of ASME Code Class 1, Class 2, and Class 3 components and systems is addressed in a separate document.

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1.2.4 Repair and Replacement Program


The Repair and Replacement Program for ASME Code Class 1, 2, and 3 component and systems, Class MC and CC are addressed by Duane Arnold plant procedures.

1.2.5 Metal Containment lnservice Inspection Program (IWE)

The DAEC 2nd Interval Containment Inspection Plan controls the examination of Containment Building under Subsection IWE and is administered separately. ·

1.3 Construction Permit

The Construction permit for Duane Arnold was issued on June 1968.

1.4 Commercial Service Dates

The Operating License for Duane Arnold was issued in January 1974. The Commercial Service Date for Duane Arnold was February 1, 1975 ..

1.5 Background

The United States of America Standards (USAS) used for the original design ·and construction of DAEC were B31.1 (1967), Code for Power Piping, and B31.7 (1969 Edition with the 1970/1971 Addenda), Code for Nuclear Power Piping. The "General Design Criteria for Nuclear Power Plant Construction Permits" was published for comment in the Federal Register in July 1967. The final version of these design criteria was not incorporated into the Code of Federal Regulations· (1 OCFR50, Appendix A) until February 1971; approximately the same time that the DAEC FSAR was submitted to the NRC. The license for DAEC is based, in part, on design and construction of the plant to USAS B31.1, USAS B31.7, and the interpretation of t.he intent of the Draft General Design Criteria published in July 1967.

The initial Duane Arnold ISi program was based on the ASME Boiler ·and Pressure Vessel (B&PV) Code, Section XI, 1970 Edition. This program was submitted as part of the original FSAR (Appendix J), which was accepted by the NRC. However, the inspection rules and requirements of the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, 1970 Edition were minimal and have ch.anged significantly since then. Federal regulations require that ISi programs be updated, to the exterit practical, to comply with the inspection and testing requirements of the edition and addenda of the ASME Code incorporated by reference in 1 OCFR50.55a one year prior to the start of each ten-year inspection interval.

1.5.1 Preservice Inspection.

The Class 1 Preservice Inspection was conducted in accordance with the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, 1970 Edition. The Class 2 'preservice inspection was not performed because it was not required at that stage of Duane Arnold's construction when it would have been used. For these components, shop and in-plant examinations records of components and welds will . be used . for comparison with lnservice Inspection data. ·

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1.5.2 First lnservice Inspection Interval


The first 10-year lnservice Inspection Interval was conducted in accordance with the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, 1974 Edition through Summer 1975 Addenda.

·~:•" j~ t.~b,• ~ ... : ~~\;;.«' . .,_,_., ~--~;:.~~·~. ~·;,. .. ··~··~8t:il9":1lii1":~>''•H,·:. ~.:~··, .. ,,.~·,;, ___ ~:··~·:···~;:~~':J,~'-~•.~-~~·~ ... , :,<:"-. : .. "c~~ · ...... i:;,-ii:s(I r:rs~_r\/.[cerl rrspe~ctior:Unter.\lat.D.ates: · .. · ·., .· · "~ .. ~ ,

Interval 1 2/01/1975 - 10/31/1985

Period 1 2/01/1975 - 5/31/1978

Period 2 6/1/1978 - 6/30/1982

Period 3 7 /1 /1982 - 10/3111985

The end of th~ first interval was extended from February 1, 1985, to October 31, 1985, due to a recirculation inlet nozzle safe-end replacement outage that lasted from June 17, 1978 through March 10, 1979. The extended interval was consistent with the ASME B&PV Code Section XI, Paragraph IWA-2400(c) of the 1974 Edition through the Summer 1975 Addenda, and IES letters dated December 13, 1983 (NG-83-4036) and January 24, 1984 (NG-84-0213).

1.5.3 Second lnservice Inspection Interval

The second 10-Year lnservice Inspection: Interval was conducted in accordance with the ASME B&PV Code Section· XI, 1980 Edition through Winter 1981 Addenda. ·

Interval 2 11/01/1985-10/31/1996

Period 1 11/0V1985 - 212811989

Period 2 3101 /1989 - 6/30/1992

Period 3 7/01/1992 - 10/31/1996

The end of the second interval was originally scheduled for November 1, 1995. The second interval was also extended 1 year, as permitted by IWA-2430( d) of ASME B&PV Code Section XI. 1980 Edition through the Winter 1981 Addenda. The second interval was extended by one y~ar to incorporate the end of refueling outage (RF0-14) scheduled for October. 1996.

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Di.Jane Arnold Energy Center ISi Program Document No. 5th lnterval-ISl-PDA-Program Plan

1.5.4 Third lnservice inspection Interval


The third 10-Year lnservice Inspection Interval was conducted in accordance with the ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, 1989 Edition with no Addenda. The third interval started November 1, 1996 and ended October 31, 2006.

Interval 3 11 /01 /1996 - 10/31/2006

Period 1 11/01/1996 - 10/31/1999

Period 2 11/01/1999-10/31/2003

Period 3 11/01 /2003 - 10/31 /2006

1.5.5 Fourth lnservice inspection Interval

The fourth 10-Year lnservice Inspection lnter\tal was conducted in accordance with ASME Boiler and Pressure Vessel (B&PV) Code, Section XI, 20Q1 Edition through the 2003 Addenda.

Interval 4 11/01(2006 - 10/31/2016

Period 1 11101 /2006 - 10/31 /2009

·Period 2 11/01/2009-·10/31/2013

Period 3 . 11/01/2013 - 10/31/2016.

The fourth interval originally ended on February 21, .. 2014 to coincide with· the end of the Operating License which resulted in an interval of only 99 months. An extension of the License for 20 years was granted in December 2010, as such, the fourth 10-Year lnservice Inspection Interval was extended to a full 10 years. . · - ·

1.5.6 Fifth lnservice lnspedion Interval

The fifth 10-year lnservice Inspection Interval will be ·conducted in accordance with ASME Boiler and Pressure Vessel (B&PV) Code, Section · XI, 2007 Edition through the 2008 Addenda as modified by 10CFR50.55a.

The fifth 10-Year lnservice Inspection Interval is divided in.to three . successive Inspection periods as determined by calendar years of plant service within the interval. The dates of the fifth 10-year lnservice Inspection Interval and Periods are as follows: .

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·" -· ... · · : ~-.-.. ..0~~ ·-. _>;i /'.:~? ~~ ·:! ... 0 <~_-. ~:-'~ :. _, o/~~.1~~~.·;·.~,~~:~;~:·-~· · · -~'. · --,1 ~-,.L~,·tJ~_.. . , :<·- .~)!:~~ .;i,

;.· -.;~. O.·~ ... . l'.1:17iftt1; trfsS,t1V_LG8~~l_risp_eq:tiQti~JrJ.terv~I' IT)·~tcl$"." ~ .·.:: ;. · ~:· '. r:; ~~~~ ·_ij~ ;,;-,

5th Interval 11/1/2016 - 10/31/2026

1st Period 11/1/2016 - 10/31/2019

2nd Period 11/1/2019 - 10/31/2023

3rd Period 11 /1 /2023 - 10/31 /2026

1.6 Applicable Editions and Addenda of ASME Section XI

In accordance with 1 O CFR 50.55a(b)(2), the lnservice Inspection Requirements for the fifth lnservice Inspection Interval applicable to Class 1, 2, and 3 components at Duane Arnold are based on ASME Section XI, 2007 Edition through the 2008 Addenda.

Portions of the ISi Program Plan are based on other Editions and Addenda of ASME Section XI, Relief Requests, the Code of Federal Regulations, Regulatory Guides, and Plant Technical Specifications and. commitments. Where this has occurred, it is documented within this Program Plan.

1.6.1 1 OCFR50.55a Code of Federal Regulations Conditions

The following mandatory ·and optional Code of Federal Regulations conditions are included in 1 OCFR50.55a as of January 1, 2016. Only those 10CFR50.55a conditions applicable to the ASME Section XI, 2007 Edition through the 2008 Addenda nondestructive examination requirements for Class 1, 2, and 3 components and component supports are listed. The conditions for metal containment components, snubbers and repair/replacement are listed for completeness only and are not applicable to this Program Plan. These conditions were reviewed for inclusion in the ISi Program Plan and dispositioned as follows:

1.6.1.1 Duane Arnold will not utilize the option in 1 OCFR50.55a(b)(2)(ii), to examine Class 1 piping per ASME Section XI, 1974 Edition with the Summer 1975 Addenda. Duane Arnold will have a RiskInformed lnservice Inspection (RI-ISi) Program for Class 1 piping. This option is not applicable.

1.6.1.2 The Duane Arnold design does not include a concrete containment subject to ASME Section XI, Subsection IWL requirements. Therefore, the mandatory conditions in 10CFR50.55a(b)(2)(viii) do not apply.

1.6.1.3 The Duane Arnold design includes a metal shell subject to ASME Section ·XI, Subsection IWE requirements. Therefore, the mandatory conditions in 1 OCFR50.55a(b)(2)(ix) apply to Duane Arnold. In addition, DAEC will implement the optional conditions for the maximum direct examination distance for remote visual examinations presented in 1 OCFR50.55a(b)(2)(ix)(B). The examination of metal. containment components is addressed

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under DAEC 2nd Interval Containment Inspection Plan.

1.6.1.4 As required by 1 OCFR50.55a(b)(2)(x), Duane Arnold will apply the station 1 OCFRSO Appendix B Quality Assurance Program of NQA-1 to ASME Section XI activities.

1.6.1.5 The mandatory condition in 1 OCFR50.55a(b)(2)(xii), which limits the use of underwater welding per IWA-4660 to materials other than those that are irradiated, will be included in the Duane Arnold ASME Section XI Repair/Replacement Program.

1.6.1.6 As allowed by 1 OCFR50.55a(b)(2)(xiv), for Appendix VIII Qualified Personnel, Duane Arnold will use the annual practice requirements in Vll-4240 of ASME Section XI Appendix VII in place of th~ 8 hours of annual hands-on training (when deemed appropriate) as discussed in 1 OCFR50.55a(b)(2)(xiv). When utilizing this option, the annual practice requirements will be performed on material or welds that contain cracks, or by analyzing prerecorded data from material or welds that contain cracks. All training will be completed no earlier.than 6 months prior to performing ultrasonic examinations.

1.6.1.7 As required by 10CFR50.55a(b)(2)(xviii)(A), Level I and II nondestructive examination personnel at Duane Arnold will be recertified on a 3-year interval in lieu of the 5-year interval specified in IWA-2314(a) and IWA-2314(b) of the 2007 Edition through 2008 Addenda.

1.6.1.8 As required by 10CFR50.55a(b)(2)(xix), Duane Arnold will not apply the provisions in IWA-4520(b)(2) and IWA-4521 of the 2Q07 Edition through 2008 Addenda which allow the substitution of , ultrasonic examination for radiographic examination specified in· the Construction Code.

1.6.1.9 , As required by 1 OCFR50.55a(b)(2)(xx)(B), Duane Arnold will apply IWA-4540(a)(2) of the 2002 Addenda when performing system leakage tests after repair and replacement activities involving welding or brazing on a pressure retaining boundary.

1.6.1.10 Implementation of IWA-2220, "Surface Examination" that allows the use of an ultrasonic examination method to perform a surface examination, is prohibited by 1 OCFR50.55a(b)(2)(xxii).

1.6.1.11 The mandatory ·condition in 1 OCFR50.55a(b)(2)(xxiii); which prohibits use of the provisions for eliminating mechanical processing of thermally cut surfaces .in IWA-4461.4.2 of Section XI, 2001 Edition through the latest edition arid addenda incorporated by reference in paragraph (a)(1 )(ii), will be included in the Duane Arnold Section XI Repair/Replacement Program.

1.6.1.12 The mandatory condition in 1 OCFR50.55a(b)(2)(xxv), which prohibits use of the provisions in IWA-4340, "Mitigation of Defects by Modification;" Section XI, 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1 )(ii), will be included in the Duane Arnold Section XI

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Repair/Replacement Program.


1.6.1.13 The mandatory condition in 1 OCFR50.55a(b)(2)(xxvi), iri which the repair and replacement activity provisions in IWA-4540(c) of the ASME Section XI 1998 Edition for pressure testing Class 1, 2, and 3 mechanical joints must be applied when using the 2001 Edition through the latest edition and addenda incorporated by reference in paragraph (a)(1)(ii). This is included ·in the Duane Arnold Section XI Repair/Replacement Program .

. 1.6.1.14 As required by 1 OCFR50.55a(b)(2)(xxviii), Duane Arnold will use the stated conditions when implementing Equation 2 in ASME Section XI, Appendix A-4300(b)(1).

1.6.1.15 As required by 1 OCFR50.55a(b)(2)(xxix), Duane Arnold will submit a relief request to apply a Risk-Informed lnservice Inspection Programs· in accordance with ASME Section XI Nonmandatory Appendix R during the Fifth Interval.

1.6.1.16 As required by 10CFR50.55a(b)(3)(v), Duane Arnold will use Article IWF-5000, "lnservice Inspection Requirements for Snubbers," of ASME Section XI, when performing inservice inspection examinations and tests of snubbers. Duane· Arnold will also comply with second provision (b)(3)(v)(B), which states that Licensees must comply with the provisions for examining and testing of snubbers in Subsection ISTD of the ASME OM Code and m~ke appropriate changes to their technical specifications or license-controlled documents when using the 2006 Addenda and later editions and addenda of.ASME Section XI.

1. 7 System Classification ~ . .

During subsequent revisions of the ISi program, other safety-related systems were added to the ISi program and ASME Code Class designations were assigned to

·establish the examination boundaries and define the required inspections and tests for the associated components. Systems, or portions of systems, were considered .. safety-related if they were determined to mitigate the consequences of an accident based on the analyses contained in Section 15 of the UFSAR. Although the

· General Electric Design Classifications do not directly correlate to ASME Code Class 1, 2, and 3, and NRC Quality Groups A, B, C, and D of Regulatory Guide 1.26, they were used as the basis for establishing the ASME Section XI examination boundaries. For the purposes of ISi, the DAEC Safety Class (SC) I safety-related components were designated ASME Section XI Code Class 1, the SC II safetyrelated components were designated ASME Section XI Code Class 2, and the SC Ill safety-related components were designated ASME Section XI Code Class 3.

The DAEC D+QA systems, including both safety-related ·and nonsafety-related systems, (except the Main Steam lines outside MSIV to Stop valves, ·and portions of the Emergency Service Water piping) were generally designated Non-Code Class. Because Duane Arnold was designed and constructed prior to the issuance, of Regulatory Guide 1.26 (safety guide 26) and NUREG-0800, these documents were not used to establish the original ASME Section XI examination boundaries, however, in accordance with the requirements of ASME Section XI 1974 Edition

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through Summer 75 Addenda, IWA-1000 footnote 2, these guidance documents were used during the first 10-Year ISi Program update. Duane Arnold has formally committed to the use of either Regulatory Guide 1.26 or NUREG-0800, Section 3.2.2. The Duane Arnold ISi Program for the fifth 10-Year lnservice Inspection Interval will continue to employ Regulatory Guide 1.26, NUREG-0800 and other approved American Nuclear Society guidance documents to determine the applicability of component inspections and to determine examination boundaries.

The Duane Arnold UFSAR was used for guidance and provides the basis for establishing the app_licable system safety classifications contained in this document.

1.7.1 Optional Construction

Optional construction of a component within a system boundary to a classification higher than the minimum class established in the component design specification does not affect the overall system classification by which the applicable rules of ASME Section XI are determined.

1.7.2 Containment Penetrations

Portions of piping penetrating the containment vessel which are required to be constructed to Class 1 or 2 rules for piping, and which may differ from the classification of the balance of the piping system, may not affect the overall system classification that determ.ines the applicable rules of ASME Section XI.

1.7.3 Class MC Components

1 OCFR50.55a was amended, effective November 22, 1999, to address the requirements of ISi of metal containment buildings. The DAEC 2nd Ten Year . Containment Inspection Plan defines the classification of the areas to be examined.

1.8 Inspection Program Plan

Examinations for Class 1, 2, and 3 components and their supports are scheduled in accordance with the lnservice Inspection Program. This incorporates the criteria of IWB-2411, IWC-2411, IWD-2411, IWF-2410 and Tables IWB-2411-1, IWC-2411-1, IWD-2411-1 and IWF-2410-1 for the examination of Class 1, 2, and 3 components and their supports, respectively. Examinations are scheduled based upon previous 10-Year Intervals; to the extent practical and within the limits of IWB-2420(a), IWC-2420(a), IWD-2420(a) and IWF-2420(a).

In order to be consistent in determining compliance with the applicable Code required examination completion percentages, NeXtEra Energy chooses to multiply the number of examination areas by the minimum or maximum percentages of the Inspection Program and then round to the nearest whole number. The result is the number NextEra Energy uses for determining how many welds, components, or supports that will be examined each period to meet Code requirements

A listing of Duane Arnold Class 1, 2 and 3 components, component supports and their relevant ISi related details subject to examination, are maintained and updated on an ongoing basis in the Duane Arnold ISi database.

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The ISi component database management system (IDDEAL) is used to develop and track the outage schedule and satisfies the requirements of IWA-2420(b)(1) through (6), respectively. The 5th Interval Duane Arnold ISi database provides the data for the lnservice Inspection Schedule Tables for examination of the major components for Duane Arnold. This database includes tables with descriptions of each component subject to examination, the required ASME Code references, and any other pertinent information that is useful for determining examination requirements.

1.9 Regulatory Guides

The Regulatory Guides determined to be applicable to Duane Arnold for purposes of this ISi Program Plan are listed below:

1.26 Quality Group Classifications 1.147 Section XI Code Case Acceptability 1.193 ASME Code Cases Not Approved for Use

1.10 ASME Section XI Code Cases

ASME; Section XI Code Cases applicable to the ISi Plan are shown below. Each of · the Code Cases has been approved or conditionally approved and listed in USNRC

Regulatory Guide 1.147, or are the subject of a relief request. When Code Cases are approved for use through a relief request, and are later added to Regulatory Guide 1.147, NextEra Energy may continue to use them in accordance with the Regulatory Guide.

t:· ~~~~.·-/;:~~1~1:~ .:'.:::::~p:~~;:;r,~~f ~~:(;,:t~'.tiiefrzL~P.fHi1<~ pJ~~-J~d;?~~,e.a:~~s'.:'.~'.,:c:f:~' J.·:. ;~ .~:· ,·:~',/'.h:;~;;:: ~:,:··~·-~.'.~:, ;.-.N~_MB:~~( > . '.'..· : ·~ . <~ '.: :_~'.)~~ :~~~~~'. , ··: ~:.Y:~:. ·;:-~~';:.~~-d.s·Ff!t~t!·~ .. Q.:~~, ::3-: .'.. }.; ;~~ ~)~i ~: E;;' ~: ;'{ : f · .~ , :~,·~~?r.

Repair Welding Using Automatic or Machine Gas Tungsten-Arc N-432-1 Welding (GTAW) Temper Bead Technique, Section XI, Division 1

(Approved without conditions in Regulatory Guide 1.147, Rev. 17) Rotation of Snubbers and Pressure Retaining Items for the Purpose of

N-508-4 Testing or Preventive Maintenance, Section XI, Division 1 (Approved with conditions in Regulatory Guide 1.147, Rev. 17) Evaluation Criteria For Temporary Acceptance of Flaws in Moderate

N-513-3 Energy Class 2 and 3 Piping, Section XI Division 1, (Approved with conditions in Regulatory Guide 1.147, Rev. 17)

N-516-3 Underwater Welding (Approved with conditions in· Regulatory Guide 1.147; Rev. 17) Alternative Requirements For Successive Inspections of Class 1 and 2

N-526 Vessels, Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17) Alternative Requirements to Repair and Replacement Documentation Requirements and lnservice Summary Report Preparation and

N.:532-5 Submission as Required by IWA-4000 and IWA-6000, Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17)

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N-561-2

N-562-2

N-586-1

N-597-2

· N-600

N-606-1

N-613-1

N-629

N-639

N-641

N-648-1

N-651

N-660

N-661-2

N-702

Alternative Requirements for Wall Thickness Restoration of Class 2 and High Eriergy Class 3 Carbon Steel Piping, Section XI, Division 1 A roved with conditions in Re ulato Guide 1.14 7, Rev. 17

Alternative Requirements for Wall Thickness Restoration of Class 3 Moderate Energy Carbon Steel Piping,. Section XI, Division 1 A roved with conditions in Re ulator Guide 1.147, Rev. 17

Alternative Examination Requirements for Classes 1, 2, and 3 Piping Component and Supports, Section XI,· Division .1 (Approved without conditions in Re ulato Guide 1.147, Rev. 17 Requirements for Analytical Evaluation of Pipe_ Wall Thinning· (Ap roved with conditions in Re ulato Guide 1.147, Rev. 17 Transfer of Qualifications Between Owners, Section XI, Division 1 A roved without conditions in Re ulato Guide 1.147, Rev. 17

Similar and Dissimilar Metal Welding Using Ambient Temperature Machine GTAW Temper Bead Technique for BWR CRD Housing/Stub Tube Repairs, Section XI, Division 1 (Approved with conditions .. in Re ulato Guide 1.147, Rev. 17 ·ultrasonic Examination of Penetration Nozzles in Vessels, Examination Category B-D, Item Nos. B3.10 and B3.90, Reactor Noule-'-to-Vessel Welds, Figs. IWB-2500-7(a),. (b), and (c), Section XI, Division 1, A roved without conditions in Re ulato Guide 1.147, Rev. 17

Use of. Fracture Toughness Test Data to Establish .Reference Temperature for Pressure Retaining Materials, Section XI, Division 1 A roved without conditions in Re ulato Guide 1.147, Rev. 17 ·

Alternative· Calibration Block Material (Approved with· condi_tions in Re ulato Guide 1.147, Rev. 17 Alternative Pressure-Temperature Relationship and Low Temperature Overpressure Protection System Requirements, Section XI, Division 1-A roved without conditions in Re ulato Guide 1.147, Rev. 17

Alteq1ative Requirements for Inner Radius Examinati.on of Class 1 Reactor Vessel_ Nozzles, -Section XI, Division 1 (Approved with' condition in Re ulato Guide 1.147. Rev. 17 Ferritic and Dissimilar Metal Welding Using SMAW Temper Bead Technique Without Removing the Weld Bead Crown for the First Layer Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17 Risk-Informed Safety Classification for Use in Risk.,.lnformed Repair/Replacement Activities, Section XI,· Division 1 (Approved with conditions in Re ulato Guide.1.147, Rev. 17 Alternative Requirements for Wall Thickness Restoration of Classes 2 and 3 Carbon Steel Piping for Raw Water Service, Section XI, Division 1 A roved with conditions in Re ulato Guide 1.147, Rev. 17 Alternative. Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division 1 A roved with conditions in Re ulato Guide 1.147, Rev: 17

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- ..,_. .. .. · ~·· ._: • .:- .fl~ ?tao.1e:t_A-:pp.ii~~b·!~.'«iAa~~¢~~~~s: ' __ ,~;:_ .: >. :·_·;;, .; : .. :·<:. ~ .... ::'::~{ ,.

'· . . ..

.... 'l. " -1;" !."~ ~ . -

(JSi~ n]·~~I'~ .. ) ~>~ .: ,~;: . ~~ .. ::~ ~~-y·::· < n'·~-~~~!~s~_~.- ~·-;·~~:~-~--: -:~- -~~-:~rp~~:~r.r~t!~.~l .. ~~·. ~.~-- ;:.~-~v ~·~?~·:: ~( ~··:;,<~~~!~·· ~-· _ :._"'::.> · .. ~::·- '~:~-~-~~~ Evaluation Criteria for Temporary Acceptance of Degradation in

N-705 Moderate Energy Class 2 or 3 Vessels and Tanks, Section XI, Division 1 (Approved without conditions in ReQulatorv Guide 1.147, Rev. 17)

N-735 Successive Inspections of Class 1 and 2 Piping Welds (Approved without conditions in Regulatory Guide 1.147, Rev. 17)

N-747 Reactor Vessel Head-to-Flange Weld Examination, Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17)

N-751 Pressure Testing of Containment Penetration Piping (Approved with conditions in Regulatory Guide 1.147, Rev. 17) Temper Bead Procedure Qualification Requirements for

N-762 Repair/Replacement Activities Without Postweld Heat Treatment, Section XI, Division 1 (Approved without conditions in Regu.latory Guide 1.147, Rev. 17) Alternative to Inspection Interval Scheduling Requirements of IWA-

N-765 2430, Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17) Roll Expansion of Class 1 In-Core Housing Bottom Head Penetrations

N-769 in BWRs (Approved without conditions in Regulatory Guide t.147, Rev. 17) Alternative Qualification Criteria for Eddy Current Examinations of

N-773 Piping Inside Surfaces, Section XI, Division 1 (Approved without conditions in Regulatory Guide 1.147, Rev. 17)

1.11 Branch Technical Position MEB 3-4

Duane Arnold applies the Generic Letter 87-11, Relaxation in Arbitrary Intermediate Pipe Rupture Requirements.

1.12 Plant Life Extension

The purpose of the Renewed License Program (EN-AA-106) is to define the roles and responsibilities for the implementation of the License Renewal process and the actions that need to be completed for each unit after the Renewed Operating License is issued to include the post approval inspections and operation in the Period of Extended Operation (PEO).

·The ISi Program Plan implements the NextEra Energy commitment to manage the effects of aging for systems/structures/components within the scope of License Renewal. The ASME Section XI Plan is credited as an aging management program for License Renewal. As such, changes to the ASME Section XI· Program Plan shall consider License Renewal requirements described in the Duane· Arnold Design Basis Document, DBD-A61-012, "DAEC Topical DBD for License Renewal".

1.12.1 Licensing Renewal Commitment Documents

LRAP-M001, ASME Section XI; lnservice Inspection, Subsections IWB, IWC and IWD - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required

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by this document.


LRAP-M003, Reactor Head Closure Studs - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M004, BWR Vessel ID .Attachment Welds - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M005, BWR Feedwater Nozzle - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M006, BWR Control Rod Drive Return Line Nozzle - An existing engineering program has been credited as a License' Renewal commitment for performing the inspections as required by this document.

LRAP-M007, BWR Stress Corrosion Cracking - An existing engineering program has been credited as· a License Renewal commitment for performing the inspections as required by this document.

LRAP-M008, BWR Penetrations - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M009, Reactor Vessel Internals Inspection Program - An existing engineering program has been credited as a . License Renewal commitment for performing the in~pections as required by this document.

LRAP-M013, Thermal Aging and ·Irradiation Embrittlement of Cast Austenitic Stainless Steel ·(CASS) - A new engineering program was developed to meet this License Renewal commitment.

LRAP-M018, Bolting Integrity - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-:-M025, BWR Reactor Water Cleanup. - An existing ·engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M031, Reactor Vessel Surveillance - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

LRAP-M040, One Time Inspection of ASME Code Class Small Bore Piping - An existing engineering program has been credited as a License . Renewal commitment for performing the inspections as required by this document.

LRAP-S003, ASME, Section XI, Subsection IWF - An existing engineering program has been credited as a License Renewal commitment for performing the inspections as required by this document.

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1.13 Successive Examinations


The sequence of component examinations will be modified in accordance with the requirements of ASME Section XI, IWB-2420(a), IWC-2420(a), IWD-2420(a) and IWF-2420(a). This allows NextEra Energy to alter the sequence of examinations to allow the examination of several components in an area during one outage instead of over several outages. This will reduce costs and radiation exposure. The percentage requirements of IWB-2411, IWC-2411, IWD-2411, and IWF-2410 will be satisfied. The examination schedule is designed to optimize the performance of work within the plant to reduce radiation dose, eliminate interference with other work, and reduce costs.

1.14 NOE Examinations and Personnel Qualification/Certification

All NOE will be performed in accordance with the requirements of ASME Section XI, 2007 Edition through the 2008 Addenda with the following modifications as required by 10CFR50.55a.

1.14.1 Alternative Examinations (IWA-4520(b)(2) and IWA-4521)

Duane Arnold will not apply the provisions in IWA-4520(b)(2) and IWA-4521 of the 2007 Edition through the. 2008 Addenda which allow the substitution of ultrasonic examination for radiographic examination specified in the Construction Code. (Reference 1 OCFR50.55a(b)(2)(xix)).

1.14.2 Certification and Recertification (IWA-2314)

Level I and II nondestructive examination personnel and personnel qualified under the American SoCiety for Nondestructive Testing Central Certification Program and ANSl/ASNT CP-189 shall be recertified o_n a 3-year interval in lieu of the 5-year interval specified in IWA-2314(a) and IWA-2314(b) (Reference 10 CFR 50.55a(b)(2)(xviii)(A)). ·

1.14.3 Appendix VIII Requirements

1.14.3.1 1 OCFR50.55a(b)(2)(xi\I) requires that all personnel qualified for performing ultrasonic ·(UT) examinations in accordance with Appendix VIII shall ·receive additional annual hands-on training. This requirement consists of at least eight hours of hands on training on samples containing cracks no earlier than six months prior to performing examinations at a licensee's facility. Duane Arnold will comply with these additional training requirements for personnel performing ASME Section XI Appendix VIII, UT examinations.

1.14.3.2 In September 1999, 10CFR50.55a incorporated an expedited implementation schedule for ASME Section XI, Appendix VIII. Duane Arnold implemented the requirements in accordance with the expedited schedule within the third 10-year interval. For the fifth 10-Year lnservice Inspection Interval, NextEra Energy will implement the requirements of ASME Section XI, 2007 Edition through the 2008 Addenda as modified by 1 OCFR50.55a and relief request, wheri approved.

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2.0 Risk-Informed (RI) ISi Requirements


During the fourth 10-Year lnservice Inspection Interval, Duane Arnold implemented a RiskInformed selection criterion for Class 1 Examination Category B-F and B-J piping welds and Class 2 Examination Category C-F-2 piping welds using the methodology contained in EPRI Topical Report TR-112657 Revision B-A and described in ASME Code Case N-578-1. This alternative to the requirements of ASME Section XI was submitted to the NRC in Relief Request NOE-ROOS. This alternative was approved by the NRC on January 31, 2007 in ML070090357. This alternative to the requirements of ASME Section XI will be resubmitted to the NRC during the fifth 10-Year lnservice Inspection Interval as Relief Request RR-05.

3.0 Development of the Class 1 Examination Plan

Plant controlled isometrics, P&IDs, component drawings, and plant walkdowns were used to develop the ISi drawings and the scope of examinations. During examinations, drawings will be used to locate and identify each component. Other plant controlled drawings or documents will be used when additional information is required.

Refer to the Class 1, 2, and 3 ISi Schedule for a complete listing of components subject to examination and the proposed examination schedule.

DAEC requested General Electric (GE) perform an analysis to determine the applicability of IWB-1220(a) and identify those systems and piping line sizes that could be exempted. This analysis was performed by GE document 22A2750, and results documented in section 5.2.5.3.3 to the Updated Final Safety Analysis Report.

The calculation located in GE document 22A2750 identifies and provides that those portions of steam piping with an inside diameter of 2.24 inches and water piping with an inside diameter of 1.12 inches may be exempted from the surface and volumetric examination requirements of Table IWB-2500-1. The systems credited in this calculation, which provide normal makeup, are the Reactor Core Isolation Cooling (RCIC) and Control Rod Drive (CRD) systems. ·

3.1 Class 1 Code Exemptions

The following Class 1 exemption criteria are applicable to Duane Arnold components. Article·IWB-1220 of ASME Section XI, 2007 Edition through the 2008 Addenda lists those Duane Arnold components exempt from examination.

IWB-1220- Components Exempt from Examination

The following components 1 or parts of components are exempted from. the volu.metric and surface examination requirements of IWB-2500: ·

(a) components that are connected to the reactor coolant system and part of the reactor coolant pressure boundary, and that are of such a size and shape so that upon postulated rupture the resulting· flow of coolant from the reactor coolant system under normal plant. operating conditions is within the capacity of makeup systems which are operable from on-site emergency power. The emergency core cooling systems are excluded from the calculation of makeup capacity;

(b) (1) components and piping segments NPS 1 (ON 25) and smaller;

(2) components and piping segments which have one inlet and one outlet, both of which are NPS 1 (ON 25) and smaller;

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(3) components2 and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 1 (ON 25) pipe.

(c) reactor vessel head connections and associated piping , NPS 2 (ON 50) and smaller, made inaccessible by control rod drive penetrations.

(d) welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration , or encapsulated by guard pipe.

Note 1: The exemptions from examination in IWC-1220 may be applied to those components permitted to be Class 2 in lieu of Class 1 by the regulatory authority having jurisdiction at the plant site.

Note 2: For heat exchangers, the shell side and tube side may be considered separate components.

(e) The above exemptions affect portions of the following systems:

Table 8 Listing of Class 1 Exempt Systems (Portions)

System P&ID Nuclear Boiler M-114 Rx Vessel Instrumentation M-115 Rx Recirculation M-116 CRD Hydraulic M-117, M-118 Residual Heat Removal M-119, M-120 Core Spray M-121 HPCI M-122, M-123 RCIC M-124, M-125 Standby Liquid Control M-126 Reactor Water Cleanup M-127 MSIV Leakage Control M-184

3.2 Component Examination Basis

The following paragraphs discuss each Table IWB-2500-1 Examination Category and Item Number applicable to Duane Arnold . The Duane Arnold Class 1 components subject to examination and testing , the Code required extent of examinations as well as the limitations for these examinations and tests, are included . All other requirements are found in ASME Section XI , 2007 Edition through the 2008 Addenda.

The ISi component database management system (IDDEAL) is used to develop and track the outage schedule, and satisfies the requirements of IWA-2420(b )( 1) through (6) , respectively.

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3.2.1 Category B-A, Pressure Retaining Welds in Reactor Vessel

ASME Section XI, Appendix VIII requirements are implemented as required.

Item 81 .10 Shell Welds

81 .11 Circumferential

81 .12 Longitudinal

Examine essentially 100% of all longitudinal and circumferential shell welds (does not include shell-to..:flange weld).

Item 81 .20 Bottom Head Weld

81 .21 Circumferential

81 .22 Meridional

Examine essentially 100% of the accessible length of all circumferential and meridional welds.

Item 81 .20 Top Head Weld 81 .21 Circumferential

81 .22 Meridional

Examine essentially 100% of the accessible length of all circumferential and meridional welds. ·

Item 81 .30 Shell-to-Flange Weld

Examine essentially 100% of the shell-to-flange weld.

Code Table Note 3: The shell-to-flange weld examination may be performed during the first and third inspection periods, in which case 50% of the shellto-flange weld shall be examined by the end_ of the first period, and the remainder by the end of the third period. During 'the first· period, the examination need only be performed from the flange face, provided this same portio~ is examined from the shell during the third period.

Alternatively, NextEra Energy may defer the examination to the end of the interval per Code Table Note 5, provided the conditions identified in Code Table Note 5 are met.

Item 81 .40 Head-to-Flange Weld

Examine essentially 100% of the head to flange weld.

Items 81 .50 Repair Welds

81 .51 Beltline Region

There are no repair welds in the beltline region of the Duane Arnold Reactor Pressure Vessel.

3.2.2 Category 8-8, Pressure Retaining welds in Vessels other than Reactor Vessels

This Examination Category is not applicable to Duane Arnold.

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--- --~~~-"":-~~- - .. --·--:.----::- ~ --· ---- -::-__



3.2.3 Category B-D, Full Penetration Welds of Nozzle in Vessels

Reactor Vessel:

Item B3.90 Item B3.100

Nozzle-to-Vessel Welds Nozzle Inside Radius Section

Examine all nozzles during the interval. In place of the UT examination required by Table IWB-2500-1, NextEra Energy may implement the alternative requirements of Code Case N-648-1 as modified by NRC Reg. Guide 1.147. 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 in Table IWB-3512-1 with limiting assumptions on the flaw aspect ratio. The provisions of Table IWB-2500-1 for this examination category continue to apply except that, in place of the examination volumes, the surfaces to be examined are the external surfaces shown in the figures applicable to this table.

3.2.4 Category B-F, Pressure Retaining Dissimilar Metal Welds

Reactor Vessel:

Item BS.10 Item BS.20 Item BS.30

NPS 4 (ON 100) or Larger Nozzle-to,..Safe End Butt Welds Less Than NPS 4 (ON 100) Nozzle-to-Safe End Butt Welds Nozzle-to-Safe End Socket Welds

Examine 100% of the Examination Category_B-F welds.

The Rl-ISr application is based on EPRI methodology and includes the examination of Class 1 piping butt welds currently addressed by ASME Section XI, ·Examination Category B-F and the Risk Informed Examination Category R-A. The proposed alternative will be submitted via Relief Request RR-65 for approval by the NRC.

3.2.5 Category-B-G~1 - Pressure Retaining Bolting, Greater Than 2 in. (50 mm) in Diameter

For heat exchangers, piping, pumps, and valves, volumetric examination of bolting for heat exchangers, pumps, or valves may be conducted on one heat exchanger, one pump, or one valve among a group of heat exchangers, pumps, or valves that are similar in design, type, and function. In addition, when the component to be examined contains a group of bolted connections of similar design and size, such as flanged connections, the examination may be conducted on one bolted connection among the group

For heat_ exchangers, piping, pumps, and valves, visual· examinations are limited to components selected for examination under B-B (vessels other than RPV), B-J (piping), B-L-2 (pump casings), and B-M-2 (valve bodies exceeding NPS 4).

Reactor Vessel:

Item B6.1 o Item B6.20 Item B6.40 Item B6.50

Closure Head Nuts Closure Studs Threads in Flange Closure Washers, Bushings

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--- ---·--~-·- - --=-- ----. ---~ -~ -~-:---~~ -'"":'- -=---·~---~--=-----~--::::-~.:'.:---------::-· --


Examine 100% of the bolting each interval.

Pumps:

Item B6.180 Bolts and Studs


Item B6.190 Item B6.200

Flange Surface, when connection disassembled Nuts, Bushings, Washers

Examine 100% of the above items of one Reactor Recirculation Pump during the interval.

3.2.6 Category B-G-2, Pressure Retaining Bolting, 2 in. (50 mm) and Less in Diameter ·

For heat exchangers, piping, pumps, and valves, examinations are limited to components selected for examination under Category B-B (vessels other than RPV), Category B-J (piping), Category B-L-2 (pump casings), and Category B-M-2 (valve bodies exceeding NPS 4), if disassembled.

Reactor Vessel:

Item B7 .10 Bolts, Studs, and Nuts


Piping:

Item B7.50 Bolts, Studs, and Nuts

. Examine 100% of the bolting each interval.

Pumps:

Item B7.60 Bolts, Studs, and Nuts


Valves:

Item B7.70 BQlts, Studs, and Nuts

Examine 100% of the bolting on one of each group of valves each intervaL ·

3.2.7 Category B-J, Pressure Retaining Welds in Piping

Item B9.10 Piping NPS 4 or larger

Item B9.11 Circumferential welds

Item B9.30 Branch pipe connection welds

Item B9.31 Piping NPS 4 or larger

Item B9.32 Piping less than NPS 4

Item B9.40 Socket welds

Examinations shall include the following:

(a) All terminal ends in each pipe or branch run connected to vessels.

(b) All terminal ends and joints in each pipe or branch run connected to other components where the stress levels exceed either of the

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Revision 1 ·January 27, 2017

following limits under loads associated with specific seismic events and operational conditions:

(1) primary plus secondary stress intensity range of 2.4 Sm for ferritic steel and austenitic steel

(2) cumulative usage factor U of 0.4

(c) All dissimilar metal welds not covered under Examination Category 8-F.

(d) Additional piping welds so that the total number of circumferential butt welds (or branch connection or socket welds) selected for e~amination equals 25% of the circumferential butt welds (or branch connection or socket welds) in the reactor coolant piping system. This total does not include welds exempted by IW8-1220 or welds in Item No. 89.22.

The RI-ISi application is based on EPRI methodology and includes the examination of Class 1 piping butt welds currently addressed by ASME · Section. XI, Examination Categorj 8-:J and the Risk Informed Examination Category R-A. The proposed alternative will be submitted via Relief Request. RR-05 for approval by the NRC.

3.2:8 Category 8-K, Welded Attachments for Vessels, Piping, Pumps, and Valves

For piping, pumps, and valves, a sample of 10% of the welded attachments will be examined. Each welded attachment will receive a surface examination of' 100% of required areas of each welded attachment. Examination is also required whenever component support member ·. deformation (e.g., broken, bent, or pulled out parts) is identified during operation,· refueling, maintenance, examination, lnservice Inspection, or testing. Examinations performed as a result of support deformation cannot be credited under the requirements of the Inspection Program.

Pressure Vessels:

Item B 10.10 Welded Attachments

For multiple vessels of similar design, function and service, only one of the welded attachments of only one of the multiple components ·requires examination. For single· vessels, only one welded attachment shall be selected for examination.· The attachmerit selected for examination on one of the multiple vessels or the single vessel, as applicable, shall be an attachment under continuous load during normal system operation, or an attachment subject to a ·potential intermittent load (seismic, water hammer, etc.) during normal system operation if an attachment under continuous load . does not exist.

Piping:

Item 810.20 Welded Attachments .

Examine 10% of the _welded attachments associated with the component supports selected for examination under IWF-2510.

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Pumps:

Item B10.30 Welded Attachments


Examine 10% of the welded attachments associated with the component supports selected for examination under IWF-2510.

3.2.9 Category B-L-2, Pump Casing

Pumps:

Item B12.20 Pump Casing

Examine the internal surfaces of one of the two reactor recirculation pumps once per interval when disassembled for maintenance or repair.

3.2.10 Category B-M-2, Valve Bodies

Valves that are of the same size, constructional design, and manufacturing method, and that perform similar functions in the system are grouped together. This grouping for Duane Arnold is shown in Table 9.

Valves:

Item B12.50 Valve Body, Exceeding NPS 4 (ON 100)

Examine at least one valve of each group of valves once per interval when disassembled for maintenance or repair.

Table 9 Class 1 Valve Grouping

Group System Valve Number Type Size

Number cs M0-2117 Gate 8"

1 cs M0-2137 Gate 8" cs V21-0042 Gate 8" cs V21-0043 Check 8"

2 cs V21-0072 Check 8" cs V21-0073 Check 8"

3 FW M0-4441 Check 16" FW M0-4442 Check 16"

4 FW V14-0001 Check 16" FW V14-0003 Check 16"

5 FW V14-0002 Check 16" FW V14-0004 Check 16" MS CV-4412 Globe 20" MS CV-4413 Globe 20" MS CV-4415 Globe 20"

6 MS CV-4416 Globe 20" MS CV-4418 Globe 20" MS CV-4419 Globe 20" MS CV-4420 Globe 20" MS CV-4421 Globe 20"

7 MS PSV-4400 Relief 6"

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Table 9 Class 1 Valve Grouping

Group System Valve Number Type

Number MS PSV-4401 Relief MS PSV-4402 Relief MS PSV-4405 Relief MS PSV-4406 Rel ief MS PSV-4407 Relief

8 MS PSV-4403 Safety/Relief MS PSV-4404 Safety/Relief

9 HPSI M0-2238 Gate HPSI M0-2239 Gate

10 HPSI M0-2312 Gate

11 HPSI V23-0049 Check

RC M0-4601 Gate

12 RC M0-4602 Gate RC M0-4627 Gate RC M0-4628 Gate RH M0-1908 Gate

13 RH M0-1909 Gate RH V19-0148 Gate RH M0-1905 Gate

14 RH M0-2003 Gate RH V19-0147 Gate RH V20-0081 Gate

15 RH V20-0082 Check RH V19-0149 Check

3.2.11 Category 8-N-1 , Interior of Reactor Vessel

Reactor Vessel :

Item 813.10 Vessel Interior


Size

6" 6" 6" 6" 6" 6" 6" 10" 1 O"

12"

12"

22" 22" 22" 22" 18" 18" 18" 20" 20" 20" 20" 20" 20"

Examine accessible areas once each inspection period above and below the reactor core made accessible for examination by removal of components during normal refueling .

3.2.12 Category 8-N-2 , Welded Core Support Structures and Interior Attachments to Reactor Vessels

Reactor Vessel (8WR):

Item 813.20 Interior Attachments within 8eltline Region

Examine interior attachments within the beltl ine region once per interval. These examinations may be deferred until the end of the interval.

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Item 813.30 Interior Attachments Beyond Beltline Region

Examine interior attachments beyond the beltline region once per interval. These examinations may be deferred until the end of the interval.

3.2.13 Category B-N-3, Removable Core Support Structures

Item 813.40 Core Support Structure

Examine accessible surfaces of core support structures once per interval. The structure shall be removed from the vessel. These examinations may be deferred until the end of the interval.

3.2.14 Category B-0, Pressure Retaining Welds in Control Rod Housings

Reactor Vessel (BWR):

Item B 14.10 Welds in Control Rod Drive (CRD) Housings

Examine 10% of the peripheral housing welds.

3.2.15 Category B-P, All Pressure Retaining Components

The system leakage test shall be conducted prior to plant startup following each refueling outage.

Item 815.10 Pressure Retaining Components [(IWB-5222(a)]

Item 815.20 Pressure Retaining Components [(IWB-5222(b)]

Perform VT-2 visual examination in association with system leakage test each refueling outage for Item 815.10 and once per interval for Item 815.20.


Plant controlled isometrics, P&IDs, component drawings, and plant walkdowns were used to develop the ISi drawings and the scope of examinations. During examinations, drawings will be used to locate and identify each component. Other plant controlled drawings or documents will be used when additional information is required .



The following Class 2 exemption criteria are applicable. Article IWC-1220 of ASME Section XI , 2007 Edition through the 2008 Addenda lists the components exempt from examination.

IWC-1220 Components Exempt from Examination

The following components (or parts of components) are exempted from the volumetric and surface examination requirements of IWC-2500;

4.1.1 IWC-1221 , Components within RHR, ECC, and CHR Systems or Portions of Systems1

. Note that exemptions IWC-1221(b) and (c) are not listed because they are applicable only to pressurized water reactor plants (PWRs) .

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(a) For systems, except high pressure safety injection systems in pressurized water reactor plants:

(1) components and piping segments NPS 4 (ON 100) and smaller.

(2) components and piping segments which have one inlet and one outlet, both of which are NPS 4 (ON 100) and smaller.


(d) Piping and other components of any size beyond the last shutoff valve in open-ended portions of systems that do not contain water during normal plant operating conditions.

4.1.2 IWC-1222, Components within Systems or Portions of Systems Other Than RHR, ECC, and CHR Systems1

. Note that exemption IWC-1222(b) is not listed because it is only applicable to pressurized water reactor plants (PWRs) .

(a) For systems, except auxiliary feedwater systems in pressurized water reactor plants:

(1) components and piping segments NPS 4 (DN100) and smaller.

(2) components and piping segments which have one inlet and one outlet, both of which are NPS 4 (ON 100) and smaller.


(c) Vessels, piping , pumps, valves, other components, and component connections of any size in systems or portions of systems that operate (when the system function is required) at a pressure equal to or less than 275 psig (1900 kPa) and at a temperature equal to or less than 200° F (93° C).

(d) Piping and other components of any size beyond the last shutoff valve in open-ended portions of systems that do not contain water during normal plant operating conditions.

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(e) The above exemptions affect portions of the following systems:

Table 10 Listing of Class 2 Exempt Systems

System P&ID

Main Steam M-103 Turbine Seal M-104 Condensate Demineralizer M-109 Rx Building Cooling Water M-112 Residual Heat Removal M-113, M-119, M-120 Nuclear Boiler M-114 Rx Vessel Instrumentation M-115 CRD Hydraulic M-118 Core Spray M-121 HPCI M-122, M-123 RCIC M-124, M-125 Standby Liquid Control M-126 Compressed Air M-130 Fuel Pool Cooling & Cleanup M-134 Radwaste Sump M-137 Containment Atmosphere Control M-143 Turbine Bldg Sample M-147 Drywell Cooling M-157 Aux. Heating Boiler & Main Loop M-160 Containment Atmosphere Monitoring M-181 MSIV Leakage Control M-184 Radwaste Liquid Waste M-186 Post Accident Sampling M-187

4.1.3 IWC-1223, Inaccessible Welds

Welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration , or encapsulated by guard pipe.

Note 1: RHR, ECC, and CHR systems are the Residual Heat Removal , Emergency Core Cooling , and Containment Heat Removal Systems, respectively.



The following paragraphs discuss each Table IWC-2500-1 Examination Category and Item Number applicable to Duane Arnold . The Duane Arnold Class 2 components subject to examination and testing , the Code required extent of examinations as well as the limitations for these examinations and tests, are included. All other requirements are found in ASME Section XI , 2007 Edition through the 2008 Addenda.

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The ISi component database management system (IDDEAL) is used to develop and track the outage schedule, and satisfies the requirements of IWA-2420(b)(1) through (6) , respectively .

4.2.1 Category C-A, Pressure Retaining Welds in Pressure Vessels

Item C 1.10 Shell Circumferential Welds

Examine 100% of cylindrical-shell-to-conical-shell-junction welds and shell( or head) to-flange welds . The examinations may be limited to one vessel or distributed among a group of vessels .

Item C1 .20 Head Circumferential Welds

Examine 100% of head-to-shell welds. The examinations may be limited to one vessel or distributed among a group of vessels .

4.2.2 Category C-8, Pressure Retaining Nozzle Welds in Vessels

Item C2.20 Nozzles Without Reinforcing Plate in Vessels > "V2 Inch (13 mm) Nominal Thickness

C2.21 Nozzle-to-Shell (Nozzle-to-Head or Nozzle-to-Nozzle) Weld

C2.22 Nozzle Inside Radius Section

These requirements are applicable to Nozzle-to-Shell (Head) Welds for nozzles greater than NPS 4; reference General Note-Figures IWC-2500-4.

The nozzle inner radius examination requirement applies to nozzles greater than NPS 12; reference Figures IWC-2500-4(a) , (b) , (c) and (d) .

Examination applies to nozzles at terminal ends of piping runs. The examination may be limited to one vessel or distributed among the vessels .

4.2 .3 Category C-C, Welded Attachments for Vessels, Piping, Pumps, and Valves

For piping , pumps, and valves, a sample of 10% of the welded attachments will be examined. Each welded attachment will receive a surface examination of 100% of required areas of each welded attachment. Examination is also required whenever component support member deformation (e.g., broken, bent, or pulled out parts) is identified during operation, refueling , maintenance, examination, lnservice Inspection, or testing . Examinations performed as a result of support deformation cannot be credited under the requirements of the Inspection Program.

Pressure Vessels:

Item C3.10 Vessel Welded Attachments

For multiple vessels of similar design, function and service, only one of the welded attachments of only one of the multiple components requires examination. For single vessels, only one welded attachment shall be selected for examination . The attachment selected for examination on one of the multiple vessels or the single vessel , as applicable, shall be an attachment under continuous load during normal system operation, or an attachment subject to a potential intermittent load (seismic, water hammer,

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etc.) during normal system operation if an attachment under continuous load does not exist.

Examine one of the welded attachments on one of the Residual Heat Removal Heat Exchangers.

Piping :

Item C3.20 Piping Welded Attachments

Examine 10% of the welded attachments associated with the component supports selected for examination under IWF-2510.

4.2.4 Category C-D, Pressure Retaining Bolting> 2 in. (50 mm) in Diameter

This Examination Category is not applicable to Duane Arnold .

4.2.5 Category C-F-1 , Pressure Retaining Welds in Austenitic Stainless Steel or High Alloy Piping

This Examination Category is not applicable to Duane Arnold.

4.2.6 Category C-F-2, Pressure Retaining Welds in Carbon or Low Alloy Steel Piping

Item CS.SO Piping welds ~ 3/8 in. (10 mm) nominal wall thickness for piping > NPS 4 (ON 100)

CS.51 Circumferential Weld

Item CS.80 Pipe branch connections of branch piping ~ NPS 2 (ON 50)

CS.81 Circumferential Weld

Examine 7.5%, but not less than 28 welds, of all carbon and low alloy steel welds not exempted by IWC-1220. The welds to be examined shall be distributed among the systems in a manner such that a representative sample of each system and size is selected. The examinations shall be distributed as required by Note 2(a) ,(b), and (c) in Table IWC-2500-1 for this examination category.

Longitudinal welds are examined in accordance with Notes 6 and 7 in Table IWC-2500-1 for this examination category.

The RI-ISi application is based on EPRI methodology and includes the examination of Class 2 piping butt welds currently addressed by ASME Section XI, Examination Category C-F-2 and the Risk Informed Examination Category R-A. The proposed alternative will be submitted via Relief Request RR-05 for approval by the NRC.

4.2.7 Category C-H, All Pressure Retaining Components

Item C7.10 Pressure Retaining Components

The pressure retaining components within the Class 2 system boundaries are subjected to system leakage tests in accordance with IWC-5220 and visually examined per IWA-5240.

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Plant controlled isometrics , P&IDs, component drawings, and plant walkdowns were used to develop the ISi drawings and the scope of examinations. During examinations, drawings will be used to locate and identify each component. Other plant controlled drawings or documents will be used when additional information is requ ired .


In accordance with IWD-1210, the examination requirements of Subsection IWD apply to those pressure retaining components and their welded attachments on Class 3 systems in support of the following functions :

(a) reactor shutdown (b) emergency core cooling (c) containment heat removal (d) atmosphere cleanup (e) reactor residual heat removal (f) residual heat removal from spent fuel storage pool


The following Class 3 exemption criteria are applicable to Duane Arnold components. Article IWD-1220 of ASME Section XI , 2007 Edition through the 2008 Addenda lists those Duane Arnold components exempt from examination .

5.1.1 IWD-1220 - Components Exempt from Examination

The following components or portions of components are exempted from the VT-1 visual examination requirements of IWD-2500:

(a) components and piping segments NPS 4 (DN100) and smaller.

(b) components and piping segments which have one inlet and one outlet, both of which are NPS 4 (DN 100) and smaller.

(c) components 1 and piping segments which have multiple inlets or multiple outlets whose cumulative pipe cross-sectional area does not exceed the cross-sectional area defined by the OD of NPS 4 (ON 100) pipe.

(d) components that operate at a pressure of 275 psig (1900kPa) or less and at a temperature of 200° F (95° C) or less in systems (or portions of systems) whose function is not required in support of reactor residual heat removal, containment heat removal , and emergency core cooling .

(e) welds or portions of welds that are inaccessible due to being encased in concrete, buried underground, located inside a penetration, or encapsulated by guard pipe.


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(f) The above exemptions affect portions of the following systems:

Table 11 Listing of Class 3 Exempt Systems

System P&ID

RHR Service Water M-113 Emergency Service Water M-113 Nuclear Boiler M-114 Residual Heat Removal M-119, M-120 River Water M-129 Diesel Generator M-132 Fuel Pool CoolinQ M-134 Circulation Water M-142 Well Cooling Water M-144 Service Water Pumphouse M-146 Control Bldg Cooling M-169 Rx Bldg HVAC M-171 Air Flow Standby Filter Unit Control M-173

5.1.2 IWD-5222(c) System Leakage Test Exclusions

(c) The following portions of system are excluded from system leakage test examination requ irements:

(1) items outside the boundaries of IWD-5222(a)

(2) items outside the boundaries of IWD-5222(b)

(3) open-ended discharge piping that is not periodically pressurized to conditions described in IWD-5221

(4) portions of systems that are associated with a spray header or are normally submerged in its process fluid such that the external surfaces of the pressure-retaining boundary are normally wetted during its pressurized conditions


The following paragraphs discuss each Table IWD-2500-1 Examination Category and Item Number applicable to Duane Arnold. The Duane Arnold Class 3 components subject to examination and testing , the Code required extent of examinations as well as the limitations for these examinations and tests are included. All other requirements are found in ASME Section XI , 2007 Edition through the 2008 Addenda .

The ISi component database management system (IDDEAL) is used to develop and track the outage schedule, and satisfies the requi rements of IWA-2420(b)(1) through (6) , respectively.

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5.2.1 Category D-A, Welded Attachments for Vessels, Piping, Pumps, and Valves

Each welded attachment will receive a visual (VT-1) examination of 100% of required areas of each welded attachment. Examination is also required whenever component support member deformation (e.g., broken, bent, or pulled out parts) is identified during operation, refueling, maintenance, examination, lnservice Inspection, or testing. Examinations performed as a result of support deformation cannot be credited under the requirements of the lnspectio_h Program.

Piping:

Item D1 .20 Welded Attachments

The percentage sample shall be proportional to the total number of nonexempt welded attachments connected to the piping in each system subject to examination. Examine 10% of the welded attachments.

· 5.2.2 Category D-8, All Pressure Retaining Components

Item D2.10 · Pressure Retaining Components

A system leakage test (IWD-5220) shall be performed during each. inspection period.

6.0 IWE Metal Containment Requirements

The requirements for Code. Class MC (Metal Containment) are found in the Duane Arnold Containment Building Metal Containment ·1nservice Inspection Program, which is administered separately. The DAEC 2nd Ten Year Containment Inspection Plan establishes the administrative, managerial, and implementation control for the IWE Containment Inspection Program. -

7.0 Development of Component Support Examination Plan

Plant controlled isometrics, P&IDs, component drawings, and plant walkdowns were used to develop the ISi drawings and the scope of examinations. During examinations, drawings will be used to locate and identify each component. Other plant controlled drawings or documents will be used when additional information is required.

Refer to the Class 1, 2, and 3 ISi Schedule for a cor:nplete listing of components subject to exarniDRtion and the proposed examination schedule.

7.1 Code Exemptions for Component Supports

In accordance with IWF-1230, component supports exempt from the examination requirements of IWF-2000 are those connected to piping and other items exempted from volumetric, surface, or VT-1 or VT-3 visual examination by IWB-1220, IWC-1220, IWD-1220, and IWE-1220. In addition, portions of supports that are inaccessible by being encased in concrete, buried underground, or encapsulated by guard pipe are also exempt from the examination requirements of IWF-2000.

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7.2 Support Examination Basis

The following paragraphs discuss each Table IWF-2500-1 Examination Category and Item Number applicable to Duane Arnold. The Duane Arnold component supports subject to examination and testing, the Code required extent of examination as well as the limitations for these examinations and tests are included. All other requirements are found in ASME Section XI, 2007 Edition through the 2008 Addenda.

The ISi component database management system (IDDEAL) is used to develop and track the outage schedule, and satisfies the requirements of IWA-2420(b)(1) through (6), respectively.

7.2.1 Category F-A, Supports

Item F1.10 Class 1 Piping Supports

Examine 25% of Class 1 piping supports. The total percentage sample shall be comprised of supports from each system (e.g., Main Steam, Feedwater, or RHR), where the individual sample sizes are proportional to the total number of nonexempt supports of each type and function within each system.

Item F1.20 Class 2 Piping Supports

Examine 15% of Class 2 piping supports. The total percentage sample shall be comprised of supports from each system (e.g., Main Steam, Feedwater, or RHR), where the individual sample sizes are proportional to the total number of nonexempt supports of each type and function within each system_. ·

Item F1 .30 Class 3 Piping Supports

Examine 10% of Class 3 piping supports. The total percentage sample shall be comprised of supports from each system (e.g., Main Steam, Feedwater, or RHR), where the individual sample sizes are proportional to the total number of nonexempt supports of each type and function within each

·system.

Item F1 .40 Supports Other Than Piping Supports (Class 1, 2, 3, and MC)

Examine the supports . of only one of the multiple components within a system of similar design, function, and service.

7.2.2 Item Number Suffixes

__ -:" ___ -:;.-~---:-.: :..-:--.- - : - -

The following Item Numbers suffixes will be utilized to identify support types by component support function as required by Table IWF-2500-1, Note (1 ).

A - Single Acting Restraints or Hangers

8 - Double Acting Restraints C - Spring Hangers, Snubbers and Supports Other codes may be used as necessary. Several supports hold more than one classified line. These supports are counted only once and if scheduled for examination, will cover all of the

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applicable lines. These supports, if examined will be counted once for credit.

7.3 Snubbers

The Duane Arnold .snubber examination and functional testing program is implemented under a separate program. This program is implemented in accordance with the Fleet Snubber Program and Site Snubber Procedure as required by the OM Code, Subsection ISTD. Visual, VT-3, examination is performed on snubbers.

8.0 Augmented and Other Programs

This section identifies augmented inspection programs maintained within the ISi Program that are not required by ASME Section XI. However, due to the nature of the augmented requirements, these programs have been included within the ISi Program. These augmented programs satisfy NRC requirements, operating ·experience, engineering judgment, Licensing agreements, etc. Augmented program revisions or deviations shall be governed by the referenced documents. The following is the detailed description of the Duane Arnold's lnservice Inspection Program Plan Basis for Augmented Examination of additional components/systems.

8.1 Duane Arnold applies BWRVIP-75, "Technical Basis for Revision to Generic Letter 88-01 Inspection Schedules" to implement Generic Letter 88-01, NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping for identifying the basis and the successive inspection requirements for stainless steel piping welds. In GL 88-01, the NRC included staff positions on materials, inspections, mitigation options, repairs, and crack evaluation. The generic letter applied to "all BWR piping made of austenitic stainless steel that is four inches or larger in nominal diameter and contains reactor coolant at a temperature above 200°F during power operation regardless of code· classification." NUREG-0313, Rev.2, "Technical Report on Material Selection and Processing Guidelines for BWR Coolant Pressure Boundary Piping," [2] contained the_technical bases for the staff positions.

8.2 Duane Arnold has implemented BWRVIP-76, "BWR Core Shroud Inspection and Flaw Evaluation Guidelines," which provides reinspection requirements for Core Shrouds, after the initial inspection that was required by Generic Letter 94-03, "lntergranular Stress Corrosion Cracking of Core Shrouds In Boiling Water Reactors."

8.3 Baseline inspections were performed in accordance with the initial guidelines of NUREG 0619, "BWR Feedwater Nozzle and Control Rod Ddve Return Line Nozzle Cracking." Successive examinations of feedwater nozzles and the CRD nozzle inner radius will occur every 10 years with other feedwater sparger components examined once every two cycles. The CRD nozzle welds will be examined once every three cycles per commitment change C96-010.

8.4 BWRVIP 18 requires examinations ofthe Core Spray Spargers. Originally required by NRC IE Bulletin 80-13, "Cracking in Core Spray Spargers", these examinations have been upgraded to implement the recommended inspections contained in BWRVIP 18.

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8.5 Generic Letter 87-11, "Relaxation in Arbitrary Intermediate Pipe Rupture Requirements."

The adoption of GL 87-11 requires applying all of the requirements of BTP MEB 3-4 to a piping run. For portions of piping in containment penetration areas, this includes a requirement for 100% volumetric inservice examination of all pipe welds during each interval. Subsequent NRC implementation guidance clarifies that GL 87-11 may be applied to older plants licensed prior to 1975 and the issuance of MEB 3-1, Rev.O, may be applied to Class 2 or 3 piping systems without applying it to Class 1 piping systems, provided that the effected piping systems meet all requirements of BTP 3-4, may be applied to any piping run without changing the requirements for other piping runs, where a piping run is defined as an "analyzable segment of piping typically located between anchor points" and compliance with BTP 3-4 requires stress analyses to be based on the 1986 Edition of ASME Section II I instead of the original basis of B31.7-1969.

Based on the above criteria, the DAEC has applied GL 87-11 to the following piping runs: (a) HPCI steam supply line - Class 2 portion only (from M0-2239 to turbine inlet), (b) RCIC steam supply line - Class 2 portion only (from M0-2401 to turbine -inlet), (c) RWCU return line - Non-Class and Class 1 piping from regenerative heat exchanger 1 E-214A outlet to the tie in at the feedwater piping, and (d) RCIC injection line - Class 1 and Non-Class portions that tie in with RWCU return pipil")g.

8.6 Lice-nse Renewal commitments were made to examine 10% of the small bor~ ASME Class 1 butt welds and 10% of the small bore ASME Class 1 socket welds each ISi interval, since DAEC did fail one small bore weld during the initial 40 year operation.

8. 7 Baseline examinations for cast austenitic stainless steel components in Jet Pumps and Core Spray piping are scheduled to be examined in accordance with· License Renewal commitments contained in LRAP-M009 "BWR Vessel Internals."

8.8 Examinations of dry tubes to identify cracks and plunger engagement are performed in accordance with the latest SIL 409 and is dependent on the how long the dry tube has. been installed. Since DAEC has a small core, the LPRMs are replaced well before an inspection is required, so only IRMs and SRMs are tracked.

8.9 Examinations of jet pump sensing lines including the mitigation clamps, the standoff supports, and tubing are examined in accordance with SIL-420 when a jet pump that has a mitigation clamp or standoff is examined for BWRVIP-41.

8.1 O Owner elected inspection programs that are part of the augmented inspections are:

1. System pressure tests of the Main Steam Relief Valve pneumatic supply, Main Steam Isolation Valve pneumatic supply, Hard Pipe Vent Control Valve supply and the Diesel Air Start supply are pressure tested every period. A specific set of tank supports related to the above systems are. VT-3 examined each ISi interval.

2. A system pressure test of the Emergency Diesel Fuel system will be performed once per period and a specific set of day tank supports are VT-3 examined each ISi interval.

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3. VT-3 of the entire Moisture Steam Separator including the lifting eyes and a VT-1 examination of the tie bar-to-standpipe will be performed every 10 years due to damage sustained in the past wheri lifting the component.

4. Shroud head bolts will be VT-3 examined· every 10 years since other plants have seen wear after extended power uprate.

5. Guide rod brackets are inspected every two outages since a cracked tack weld was discovered in RFO 16.

9.0 Evaluation/Acceptance Criteria

NextEra Energy will pertorm non-destructive examinations using visual, surface (Penetr'ant and Magnetic Particle), and volumetric (Ultrasonic, Radiography, and Eddy Current) techniques. Other NOE techniques may be utilized when required._

During inservice inspections, NOE indications are evaluated against the . acceptance standards of ASME Section XI. Compof1ents with indications that do not exceed the acceptance criteria will be considered acceptable for continued service. Additional examinations are not required.

Examinations that reveal indications shall be evaluated in accordance with Articles IWA-3000, IWB-3000, IWC-3000, IWD-3000 ano IWF-3000, as applicable.

9.1 Supplemental Examinations

Examinations that detect flaws/conditions that require evaluation in accordance with the requirements of IWB-3100, IWC-3100 or IWF-3100, may be supplemented by other examination methods and techniques within the limits specified by IWB-3200, IWC-3200 or IWF-3200.

9.2 · Additional Examinations

Examinations that reveal flaws or relevant conditions that exceed the referenced acceptance standard, shall be extended to include additional examinations during ·the current outage. The adoitional examination requirements of IWB-2430, IWC-· 2430, IWD-2430, or IWF-2430, as applicable, shall be performed as determined by Nuclear Engineering.

9.3 Successive Inspections for Components

Where components are accepted for continued service by analytical evaluation, IWB-2420(b), IWC-2420(b), IWD-2420(b) and IWF-2420(b), the area containing the flaws or component support shall be subsequently reexamined· ih accordance with the following; ·

9.3.1 Class 1 Components [IWB-2420(b)]

If a component is accepted for continued service in accordance with IWB-3132.3 or IWB-3142.4, the areas containing flaws or relevant conditions shall be reexamined during the next three inspection periods listed in the schedule of the Inspection Program of IWB-2400.

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Provided the flaws or relevant conditions remain essentially unchanged for three successive inspection periods, the component examination schedule

· will revert to the original schedule of succ·essive inspections.

9.3.2 Class 2 Components [IWC-2420(b)]

If a component is accepted for continued service in accordance with IWC-3122.3 or IWC-3132.3, the areas containing flaws or relevant conditions shall be reexamined during the next inspection period of listed in -the schedule of the Inspection Program of IWC-2400. Pr<?vided the flaws or relevant conditions remain essentially unchanged for the next inspection period, the inspection schedule will revert to the original schedule· ·of successive inspections.

9.3.3 Class 3 Components [IWD-2420(b)]

If a component is accepted for continued service in accordance with IWD-3000, the areas containing flaws or relevant conditions shall be reexamined during the next inspection period of listed in the schedule of the Inspection Program of IWD-2400. Provided the flaws or relevant conditions remain essentially unchanged for the next inspection period, the inspection schedule will revert to the original schedule of successive inspections.

9-.3.4 Component Supports [IWF-2420(b)]

If a component support is accepted for continued service in accordance with IWF-3112.2 or IWF-3122.2, the component support shall be reexamined during the next inspection period listed in the schedule of the Inspection Program of IWF-2410. Provided the examinations do not require additional corrective measures during the next inspection period, the inspection schedule will revert to the original schedule of successive inspections. -

10.0 Repair/Replacement Activities

The requirements of ASME Section XI, 2007 Edition through.the 2008 Addenda, the latest 1 O CFR 50.55a requirements and the Repair and Replacement Program for Duane Arnold shall be met for Class 1, 2, and 3 piping, Class MC, and components and their supports, as applicable. Specific requirements for the repair, replacement, or modification of ISi components are detailed in Duane Arnold's ASME Section XI Administrative Document, "Repair, Replacement, and Modification".

11.0 Relief Requests

A Relief Request is required when there is situations where Code requirements cannot be met or where an alternative is desired. Relief Requests shall be prepared using the NEI guidance for the standard format for requests from commercial reactor licenses pursuant to 1 OCFR50.55a. Relief Requests will be reviewed for completeness, technical ad_equacy, and implementation. Reviewers may be the site ISi Coordinators, the ISi Specialist,· NOE personnel, and any other group the Relief Request may affect. A complete list of Duane Arnold's Fifth Interval Relief Requests are included in Appendix A. Typical examples where Relief Requests are submitted are as follows:

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-:-::::..:_:::.--.-.-:--::. =-.~:_: .=::':' - --_-::::: _--:- -==--~~ :.::-:·_--=-~·:.--::::-.~. - --- ... _. __

Duane Arnold Energy Center ISi Program Document No. 5 th lnterval-ISl-PDA-Program Plan

Revision 1 January 27 , 2017

11 .1 For Class 1 and 2 weld examinations, rel ief is required if 90% or less of the Code required coverage were achieved (if unable to meet Code examination requirements) .

11 .2 The request for use of an alternative, hardship or impractica lity to a requirement listed within ASME Section XI. An example is the use of a Code Case that has not been approved for use by the latest revis ion of NRC Reg . Guide 1.147.

12.0 ISi Boundary Classifications

The code required boundaries for all ISi Class 1, 2, 3, Non-class, and Augmented systems are color coded on Piping and Instrument Diagrams (P&IDs) . The following list includes those P&IDs applicable to the Duane Arnold ISi Program Plan .

Table 12 -ISi Boundary Classification Drawings

P&ID# DRAWING TITLE ASMECLASS M-100 LEGEND NIA M-1 01 LEGEND NIA M-102 LEGEND NIA M-103 MAIN STEAM TURBINE STOP & CONTROL VAL YES SH 1 2&AUG M-1 04 TURBINE STEAM SEAL SH 1 2 M-105 STEAM AIR EJECTOR AUG M-106 CONDENSATE FEEDWATER SHI NON CLASS M-1 07 CONDENSATE FEEDW ATER SH2 NON CLASS M-108 CONDENSATE DEMINERALIZER NON CLASS M-109 CONDENSATE DEMINERALIZER 2&AUG M-1 JO MAKE-UP DEMINERALIZER NON CLASS M-111 GENERAL SERVICE WATER NON CLASS M-112 REACTOR BUILDING COOLING WATER 2 M-113 RHR SERVICE WATER & EMERG. SERVICE WATER 2,3 M-114 NUCLEAR BOILER 1, 2 & 3 (AUG) M-11 5 REACTOR VESSEL INSTRUMENTATION 1&2 M-116 REACTOR RECIRCULATION 1 M-117 CRD HYDRAULIC, SH 1 1 M-118 CRD HYDRAULIC, SH 2 1, 2 M-119 RESIDUAL HEAT REMOVAL 1, 2 & 3 M-120 RESIDUAL HEAT REMOVAL 1,2 & 3 M-12 1 CORE SPRAY 1&2 M-122 HIGH PRESSURE COOLANT INJECTION SH 1 l ,2&AUG M- 123 HIGH PRESSURE COOLANT INJECTION SH 2 1, 2, AUG M-124 REACTOR CORE ISOLATION COOLING SH 1 1, 2&AUG M-1 25 REACTOR CORE ISOLATION COOLING SH 2 1&2 M-126 ST AND BY LIQUID CONTROL 1&2 M-127 REACTOR WATER CLEANUP l , AUG M-128 REACTOR WATER FILTER DEMINERALIZER NON CLASS M-129 RIVER WATER SUPPLY & INT AKE STRUCTURE 3 M-130 COMPRESSED AIR Sheets I - 8 & 10 NON CLASS M-130 COMPRESSED AIR Sheets 9 2

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Duane Arnold Energy Center ISi Program Document No. 5 th lnterval-ISl-PDA-Program Plan

Table 12 ISi Boundary Classification Drawings

P&ID# DRAWING TITLE M-131 TURBINE LUBE OIL M-132 DIESEL GENERA TOR SH's 1, 2, & 3 M-133 FIRE PROTECTION M-134 FUEL POOL COOLING & CLEANUP M-135 FUEL POOL DEMINERALIZER M-136 SERVICE CONDENSATE M-137 RADW ASTE SUMP SYSTEM Sheet 1 M-138 EQUIPMENT RADW ASTE M-139 FLOOR DRAIN RADW ASTE M-140 RADW ASTE SOLIDS HANDLING M-141 OFF GAS M-142 CIRCULATION WATER M-143 CONTAINMENT ATMOSPHERE CONTROL SH 1, 2, 3, and 4 M-144 WELL COOLING WATER SH 1 M-144 PRODUCTION WELL 1, 2, 3, AND 4 M-145 MISC. TURBINE GENERA TOR M-146 SERVICE WATERPUMPHOUSE M-147 TURBINE BUILDING SAMPLE M-148 AREA RADlA TION MONITORING M-149 OFF GAS RECOMBINER M-150 HY AC PLANT AIR FLOW M-151 CONTROL BUILDING & TSC AIR FLOW M-152 REACTOR BUILDING AIR FLOW M- 153 TURBINE BUILDING AIR FLOW M-154 HY AC RADW ASTE BUILDING AIR FLOW M-156 DRYWELL AIR FLOW M- 157 DRYWELL COOLING WATER SH 1 M-158 HVAC AIR FLOW AND STANDBY GAS TREATMENT M-159 VENTILATION TURBINE BUILDING M-160 AUX. HEATING BOILER & MAIN LOOP SH 1 M-161 AIR CONDITIONING CONTROL BUILDING M-162 AUX. HEATING REACTOR BUILDING M-163 AUX. HEATING TURBINE BUILDING M-164 VENTILATION RAD WASTE BUILDING M-165 MAIN PLANT AIRINTAKE&M.G. ROOM M-166 COOLING & HEATING PLANT AIR SUPPLY M-167 ADM. BUILDING HEATING AND COOLING M-168 ADM. BUILDING HEATING AND COOLING


ASMECLASS NONCLASS

3 NON CLASS

2&3 NON CLASS NON CLASS

2 NONCLASS NON CLASS NON CLASS NONCLASS

3 2 3

NONCLASS NON CLASS

3 2

NON CLASS NONCLASS NONCLASS NONCLASS NONCLASS NON CLASS NONCLASS NON CLASS

2 NON CLASS NON CLASS

2 NON CLASS NON CLASS NON CLASS NON CLASS NONCLASS NONCLASS NON CLASS NON CLASS

M-169 CONTROL BUILDING COOLING & PLANT CHILLED WTR SH 2, 3 3 M-170 HY AC MISC. CONTROL NON CLASS M-171 REACTOR BUILDING HY AC COOLING 3 M-1 72 AIR FLOW, HTG. CLG. MACH SHOP OFF GAS RETENTION BLDG. NON CLASS M-1 73 AIR FLOW ST AND BY FILTER UNIT CONTROL 3 M-174 DRYWELL HEATING & VENTILATION FAN NON CLASS M-175 AIR FLOW PUMPHOUSE NON CLASS M-176 VENTILATION & OFF GAS ST ACK REACTOR BUILDING NON CLASS M-177 INTAKE, TSC, & WELL HS. HTG. AND VENTILATION CONTROL NON CLASS

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Table 12 ISi Boundary Classification Drawings

P&ID# DRA W1NG TITLE ASMECLASS M-178 HY AC, MISC. CONTROL ROOM NON CLASS M-1 79 LEGEND (HY AC) NIA M-180 CHLORINATION & ACID FEED NON CLASS M-181 CONTAINMENT ATMOSPHERE MONITORING 2 M-182 RADWASTE EVAPORATION NON CLASS M-183 RADWASTE SAMPLE NON CLASS M-184 MSIV LEAKAGE CONTROL 1&2 M-185 FIRE PROTECTION CARBON MONOXIDE NONCLASS M-186 RADWASTE LIQUID WASTE STORAGE & HANDLING 2 M-187 POST ACCIDENT SAMPLING 2

13.0 Addition of Welds, Components, and Components Supports

The rules for selection and scheduling of examinations for new welds shall be in accordance with paragraphs IWB-2411 (b), IWC-2411 (b), IWD-2411 (b) and IWF-241 O(c) .

14.0 Records

14.1 General

Records of lnservice Inspection Program Plan, schedules, outage schedules, calibration standards, examination and test procedures , results of activities, final reports , certifications, and corrective actions will be developed and maintained in accordance with IWA-6000.

14.2 Nondestructive Examinations

Completed NOE examination data packages shall be submitted to the ISi Program Owner following completion of the inservice examination activity.

14.3 Reports

14.3.1 Reports/documents will be generated for the following activities:

(a) Nondestructive examination activities performed on Class 1, 2, and 3 systems, components and their supports

(b) Nondestructive examination activities performed on Class MC (c) System pressure tests (d) Repairs and replacements

14.3.2 The final reports shall contain, as a minimum, the information required to support submittal of the NIS-1 or OAR-1 (Code Case N-532-5 as modified by NRC Reg. Guide 1.147).

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14.4 lnservice Inspection Summary Report

NextEra Energy shall forward a summary report, NIS-1 or OAR-1 , of the ISi activity to the Nuclear Regulatory Commission in accordance with IWA-6230 or Code Case N-532-5 (as modified by NRC Reg. Guide 1.147), as applicable.

14.5 NIS-2 or NIS-2A Reports

NIS-2 forms, or if Code Case N-532-5 is followed , a NIS-2A form , will be completed for each repair or replacement.

15.0 References

The lnservice Inspection Program Plan for ISi Class 1, 2, and 3 (or Quality Groups A, B, and C respectively) systems and components and component supports, was developed after reviewing the following documents and procedures. Limitations of design, geometry, and materials of construction may have an impact on the implementation of some of these documents.

15.1

15.2

15.3

15.4

15.5

15.6

15.7

15.8

15.9

15.10

10 CFR 50.55(a) Code of Federal Regulations, Codes and standards.

American Society of Mechanical Engineers (ASME) Section XI Code, 2007 Edition through the 2008 Addenda .

Nondestructive Evaluation: Performance Demonstration Initiative (POI) Comparison to ASME Section XI , Appendix VIII 2007 Edition with 2008 Addendum and 10CFR50.55a, Year 2011 . EPRI , Palo Alto, CA 2012 1026510

American Society of Mechanical Engineers (ASME) Section XI Code, 1998 Edition.

USNRC Regulatory Guide 1.26 - Quality Group Classifications and standards for Water-, Steam-, and Radioactive-Waste-Containing Components of Nuclear Power Plants, Revision 3, dated February 1976.

USNRC Regulatory Guide 1.147 - lnservice Inspection Code Case Acceptability, ASME Section XI.

USNRC Regulatory Guide 1.178 - An Approach for Plant-Specific Risk Informed Decision Making lnservice Inspection of Piping.

USNRC Regulatory Guide 1.193 - ASME Code Cases Not Approved for Use.

Branch Technical Position MEB 3-1 , High Energy Fluid Systems, Protection Against Postulated Piping Failures in Fluid Systems Outside Containment.

Branch Technical Position APCSB 3.1. Paragraph B.2c (4) .

15.11 Duane Arnold Updated Final Safety Analysis Report.

15.12 Duane Arnold Technical Specifications, Docket number 50-331 .

15.13 Fleet lnservice Inspection Program - ER-AA-118, "ASME Section XI lnservice Inspection Program".

15.14 Duane Arnold Document - Program Engineering ASME Section XI Administrative Manual , "lnservice Inspection Administrative Document. "

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15.15 Duane Arnold Document - Program Engineering ASME Section XI Administrative Manual , "Repair, Replacement, and Modification Administrative Document

15.16 (Section XI) ."

15.17 Procedure - EN-AA-106, "Renewed License Program".

15.18 NUREG-1801 , Revision 1 dated September 2005, Generic Aging Lessons Learned (GALL) Report.

15.19 NUREG-1955, Safety Evaluation Report Related to the License Renewal of Duane Arnold Energy Center, November 2010.

15.19 DBD-A61 -012, DAEC Topical Design Basis Document for License Renewal.

15.20 Nuclear Energy Institute NEI 04-05, "Living Program Guidance to Maintain RiskInformed lnservice Inspection Programs for Nuclear Plant Piping Systems", April 2004.

15.21 Duane Arnold Document - DAEC 2nd Ten Year Containment Inspection Plan, "ASME Section XI , Subsection IWE, Containment Building Metal Containment lnservice Inspection Program for the Duane Arnold Energy Center."

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Appendix A

Relief Requests

Page A 1 of A32



Table 13 Fifth Interval Relief Requests

Relief Request Description Number

Extension of Permanent Relief from Ultrasonic

RR-01 Examination of Reactor Pressure Vessel Circumferential Shell Welds for the Renewed Operating License Term

RR-02 Alternative Requirements for Buried Piping and Components

RR-03 Alternative Requirements for Nozzle Inner Radius and Nozzle-to-Shell Welds

RR-04 Alternative for Seal Weld Procedure Qualification

RR-05 Risk Informed ISi Program

Page A2 of A32


Status

To Be Submitted: March 2017




To Be Submitted


Relief Request No. 1


"Extension Of Permanent Relief From Ultrasonic Examinatio.n Of Reactor Pressure Vessel Circumferential Shell Welds For The Renewed

Operating License Term"

March 201T

Page A3 of A32


DUANE ARNOLD ENERGY CENTER FIFTH INSERVICE INSPECTION INTERVAL

RELIEF REQUEST NUMBER RR-01, REVISION 0


EXTENSION OF PERMANENT RELIEF FROM VOLUMETRIC EXAMINATION OF REACTOR PRESSURE VESSEL CIRCUMFERENTIAL SHELL WELDS FOR THE

RENEWED OPERATING LICENSE TERM

Proposed Alternative In Accordance with 10 CFR 50.55a(z)(1)

--Alternative Provides Acceptable Level of Quality and Safety--

1. ASME Code Component(s) Affected

Code Class:

Reference:

Examination Category:

Item Number:

Component Numbers:

1

ASME Section XI, Table IWB-2500-1

B-A

81 .11

VCB-8001, VCB-A002, VCB-8003, and VCB~B004

2. Applicable Code Edition and Addenda

lnservice inspections (ISi) at Duane Arnold Energy Center (DAEC) are performed per the requirements of the ASME Boil~r and Pressure Vessel Code Section XI, 2007 Edition through the 2008 Addenda as required by 1 OCFR50.55a. ·

3. Applicable Code Requirement

The applicable Code requirement is contained in Subsection IWB, Table IWB-2500-1; "Examination Category B-A, Item Number 81 .11, "Circumferential Shell Welds", which requires a volumetric examination for all .Reactor Pressure . Vessel (RPV) circumferential shell welds each interval.

4. . Reason for Request

On January 6, 2005, Duane Arnold received U.S. Nuclear Regulatory Commission · (NRC) authorization for a technical alternative that eliminated performance of RPV. circumferential shell weld examinations for the duration of the full-term operating license that ended on February 21, 2014 (Reference 1 ). The primary basis was an analysis in accordance with Boiling Water Reactor Vessel and Internals Project (BWRVIP) report, BWRVIP-05, "BWR Reactor Pressure Vessel Shell Weld Inspection Recommendations," (Reference 2) and NRC guidance, which indicated that the limiting conditional failure probability for the circumferential shell welds would be satisfied through the expiration of the original full-term operating license.

Page A4 of A32



Anticipated changes in metallurgical conditions expected over the renewed license period required analysis and further evaluation to demonstrate the continued acceptability for not performing volumetric examinations of these RPV circumferential shell welds over the additional renewed operating license term of 20-years. The analysis was based on BWRVIP-05 and BWRVIP-74-A, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal" (Reference 3). Information on the projected acceptability for continuing permanent deferral--of volumetric examinations on RPV circumferential shell welds was provided in Section 4.2.4 of NUREG-1955, Safety Evaluation Report Related to the License Renewal of Duane Arnold Energy Center.

NUREQ-1955 (Reference -4), provides a ·summary of· the safety basis for the· acceptability of various aspects of the license renewal. NUREG-1955 Section 4.2.4,

. "Reactor Vessel Circumferential Weld Examination Relief," di_scusses specifics of the NRC evaluation for this area and the continued acceptability of this alternative for the renewed license period of operation. Subsection 4.2.4.4, "Conclusion," of the NUREG states:

On the basis of its review, as discussed above, the staff concludes that Duane Arnold has demonstrated, pursuant to 10 Ci=R 54.21(c)(1)(ii), that, for the RPV circumferential weld examination relief time-limited aging analysis (TLAA), -the-

. analyses have been projected to the ~nd of the period of extended operation .. ·However, Duane Arnold will· need to request an extension of the relief for circumferential weld examination for the period of extended operation. The staff also concludes that the UFSAR supplement contains an appropriate summary

_description. of. the activities for managing the effects of aging and the TLAA .. evaluation, as required by 10 CFR 54.21 (d).

This 10 CFR 50.55a request is provided to meet the license renewal commitment to resubmit a request for authorization of . permanent relief from the volumetric examination of RPV circumferential shell welds through the 20-year renewed operating license. As such, NextEra Energy· is proposing an alternative in accordance with 10 CFR 50.55a(z)(1) on the basis that this alternative provides ·an acceptable level of

. quality and safety. · ·

5. Proposed Alternative and Basis for Use

Proposed Alternative

The projected failure frequency of the subject welds at· Duane Arnold·· has been . determined to be sufficiently low for the duration of the renewed operating license term to justify eliminating the examinations required by 1 O CFR 50.55a(g) in accordance with ASME Code Section XI, Table IWB-2500-1, Examination Category B-:A, Item No. 81 .11, Circumferential Shell Welds. Pursuant to 10 CFR 50.55a, "Codes· and Standards," paragraph(z)(1 ), and consistent with guidance provided in NRC Generic Letter 98~05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request.· Relief from Augmented Examination Requirements on Reactor PressureVessel Circumferential Shell Welds" (Reference 5), and the final license renewal safety evaluation report for proprietary report, BWRVIP-

Page AS of A32



74, "BWRVessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal," (Reference 6), .NextEra Energy proposes the following altemate provisions for the subject weld examinations for the 20-year renewed operating license term.

• The examination requirements of ASME Code Section XI, Table IWB-2500:.1, Examination Category B-A, Item No 81 .12, for the RPV longitudinal shell welds,

· also known as vertical or axial welds, will be performed as required to the extent possible. · ·

• As a proposed alte.rnative to the requirements of ASME Code Item No. 81 .11 for the RPV circumferential shell welds, the longitudinal. weld examinations for ASME Code Item No. 81, 12 will include examination on the segment of RPV circumferential · welds that intersects with . the · longitudinal welds, or approximately 2 to 3 percent of the RPV shell circumferential welds.

• The proposed examination alternative for the RPV circumferential shell welds may be performed from either the internal inside diameter (ID) surface, or .from the external outside diameter (OD) surface of the RPV as determined by. NextEra Energy. ·

• Examination of the remaining portions of the RPV circumferential shell welds will be permanently deferred through the renewed operating license term, ending midnight of February 21, 2034.

• Examination will be ·completed in accordance with the ASME Section XI, Appendix VIII, "Performance · Demonstration- for Ultrasonic Examination Systems," for the intervals applicable ·code of Record edition and addenda as required· and modifiec;i by 10 CFR 50.55a, "Codes and standards."

Basis for Use

The BWRVIP.:.74 rE:lport provides generic guidelj~.e&\'l~r the appropriate in.spection and flaw evaluation recommendations to assure safefyi function integrity of reactor pressure vessel components continuing from the initial operating license term through the renewed operating license term. The NRC final license renewal safety evaluation for BWRVIP-74, concluded that Appendix E of the July 28, 1998 NRC safety evaluation for BWRVIP-05, conservatively evaluated BWR reactor pressure vessels to 64 Effective Full Power Years (EFPY), which is 10 EFPY greater than what is· realistically expected at the end of an additional 20-year license renewal period. - -

The NRC staff analysis provides a technical basis for an alternative. from the ASME Code Section XI requirements for the volumetric examination of RPV circumferential shell welds for the license renewal period. The associated safety evaluation stated that to obtain relief [similar to the conditions promulgated in Generic Letter 98-05 (Reference 5)] each licensee would have to demonstrate that:

(1) At the end of the license renewal period, the circumferential welds will satisfy the limiting conditional failure frequency for circumferential welds in Appendix E of the NRC staffs Final Safety Evaluation Report (FSER) for BWRVIP-05, and ·

Page A6 of A32



(2) That they have implemented operator training and established procedures that limit the frequency of cold over-pressure events to the amount specified in the NRC staffs FSER for BWRVIP-05.

The following discussion describes how each of these criteria will be met during the renewed operating license period.

Demonstrate that Circumferential Welds Will Satisfy the Limiting Conditional Failure Frequency at the End of the License Renewal Period.

The following discussion is taken from the staff evaluation under Section 4.2.4, "Reactor Vessel Circumferential Weld Examination Relief," within NUREG:-1955, the Duane Arnold license renewal safety evaluation report (SER), and summarizes the basis for use and the acceptability of the proposed alternative.

The staff reviewed LRA Section 4.2.4 to verify, in accordance with 10 CFR 54.21 (c)(1 )(ii), that the analyses have been projected to the end of the period of extended operation.

The technical basis for relief from the ASME Code Section XI, inservice inspection (ISi) requirements is discussed in the staffs final Safety Evaluation(SE) concerning the BWRVIP-05 report dated July 28, 1998. The staff's SE letter indicated that BWR applicants may request relief from the ISi requirements of 10 CFR 50.55a(z) for volumetric examination of circumferential RPV welds by demonstrating that: (1) at the expiration of the license, the circumferential welds satisfy the limiting conditional failure probability for circumferential welds in the s_taffs evaluation dated July 28, 1998, and (2) the applicant implemented operator training and established procedures that limit the frequency of cold overpressure events to the frequency specified in the staffs SE. The letter indicated that as part of any BWR LRA, the requirements for the inspection of RPV circumferential welds during an additional 20-year period of extended operation must be reassessed on a plant-specific basis .. In addition, the applicant must request relief from the ISi requirements for volumetric examination of circumferential welds for the period of extended operation in accordance with the requirements of 10 CFR 50.55a(z).

The staff reviewed Duane Arnold's probability of failure (PoF) results which were based .on the methodology presented in the BWRVIP-05 report. PoF results were calculated for 54 EFPY for the limiting RPV beltline circumfer_ential weld VCB-A2. The PoF evaluation included the ·consideration of the low temper~ture overpressure protection (L TOP) occurrence probability. Duane Arnold did not submit detailed calculations for obtaining PoF values for the limiting beltline circumferential weld and, therefore, in RAI 4.2.4-1 dated September 24, 2009, the staff requested that Duane Arnold submit the calculations for mec;in RT NoT·

In its response dated October 23, 2009, Duane Arnold provided a mean RT NOT

value for the limiting RPV circumferential weld. The staff confirmed the validity of the data for the copper and nickel contents and the initial RT NOT values for the limiting RPV beltline circumferential weld. The staff reviewed Duane Arnold's calculations for the 54 EFPY mean RT NOT value for the limiting RPV circumferential

Page A 7 of A32


Revision 1 January27, 2017

weld using the data presented in the submittal dated October 23, 2009, and found them to be acceptable. The 54 EFPY mean RT NDT value is bounded by the 64 EFPY mean RT NDT value used by the staff for determining the conditional failure probability of a circumferential weld. The 64 EFPY mean RT NDT .value from the staffs SE dated July 28, 1998, is representative of a Chicago Bridge and Iron Company (CB&I) weld because CB&I fabricated the Duane Arnold's Reactor Pressure Vessel (RPV). Since Duane Arnold 54 EFPY mean RT NDT value is less than the 64 EFPY value used in the staffs SE dated July 28, 1998, the staff concludes that the RPV conditional failure probability of the limiting RPV circumferential weld is bounded by the staff analysis, and the PoF value due to a L TOP event at 54 EFPY for the limiting RPV circumferential weld is less than the value used in the staffs SE dated July 28, 1998. Therefore, the staff accepted Duane Arnold's response and considers its concerns related to RAI 4.2.4-1 resolved.

In the July 28, 1998, staff's SE for the BWRVIP-05 report required that examination of the limiting RPV circumferential· shell welds must be performed if the corresponding volumetric examinations of the RPV axial shell welds revealed any presence of an age-related degradation mechanism. By letter dated September 24, 2009, the staff issued RAI 4.2.5-1 requesting that Duane Arnold confirm whether previous volumetric examinations of the RPV axial shell welds have shown any indication of cracking or other age-related degradation mechanisms in the welds. In its response dated October 23, 2009, Duane Arnold stated that no re9ordable indications of cracking were identified during the previous volumetric examinations of the RPV axial welds. The staff accepted this response and considers its concerns related to RAl4.2.5-1 resolved.

Implement Operator training and Establish Procedures that Limit the Frequency of Cold Over-Pressure Events to the Amount Specified in the NRC Staff Safety Evaluation for BWRVIP-05

Section 4.2.4 of NUREG-1955 also indicates that Duane Arnold stated it will use the same procedures and tra_ining used to limit RPV cold overpressure events as those approved by the staff for Duane Arnold to implement the BWRVIP-05 ·technical alternative for the term of the original operating license. · By letter dated October 23, 2009, Duane Arnold provided a response to RAI 4.2.4-2 stating it will use training procedures (to limit RPV cold overpressure events) described in Duane Arnold's relief request NDE-R047. This relief request was submitted on February 12, 2004, and was approved by the staff in a letter dated January 6, 2005. Since the training procedures were approved by the staff and they are consistent with the BWRVIP-05 report, the staff accepted Duane Arnold's response and considers its concerns related to RAI 4.2.4-2 resolved.

Therefore, use of the guidance provided in NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05 Report to Request Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds", and the final license renewal safety evaluation report for proprietary report, BWRVIP-74, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal" provides an acceptable level of quality and safety pursuant to 1 O CFR 50.55a(z)(1 ).

Page AB of A32


6. Duration for the Proposed Alternative


The proposed alternative will be applied for the 20-year term of the renewed operating license expiring at midnight February 21, 2034.

7. Precedents

The NRC has authorized similar requests to adopt an alternative to the ASME Section XI, Table IWB-2500-1, Examination Category B-A, Item. No. 81 .11 criteria for permanent relief from the volumetric examination of RPV circumferential shell welds for the period of extended operation.

Similar relief requests have been granted to the following plants:

1) NRC letter, "Pilgrim Nuclear Power Station - Issuance of Relief Request No. PRR-51 - Relief From Certain ASME Code, Table IWB-2500-1, Reactor Vessel. Circumferential Weld Examination Requirements (CAC Nos. MF6361 )," dated March 15, 2016. (ADAMS Accession No. ML 16042A291)

2) NRC letter, "Monticello Nuclear Generation Plant (MNGP) - Request for Relief No. 1.7 Regarding Examination· of Reactor Pressure Vessel Shell Circumferential

·Welds (TAC ·No. ME3526)" dc;lted February 8, 2011. (ADAMS Accession No. ML 110200700)

8; References

1. NRC letter, "Duane Arnold Energy Center - Re: Alternatives for Examination of Reactor Pressure Vessel Circumferential Shell Welds Relief Request NDE-R047 (TAC No. MC2181)" dated January 6, 2005. (ADAMS Accession No. ML043270051)

· 2. Electric Power Research Institute (EPRI) Report TR-105697, "BWR Reactor Pressure Vessel Shell Weld Inspection Recomnien.dations (BWRVIP-05)," dated September 1995.

3. BWRVIP-74-A Report, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines for License Renewal," dated June 2003.

4. NUREG-1955, "Safety Evaluation Report Related to the License Renewal of Duane Arnold Energy Center."

5. NRC Generic Letter 98-05, "Boiling Water Reactor Licensees Use of the BWRVIP-05. Report to Request. Relief from Augmented Examination Requirements on Reactor Pressure Vessel Circumferential Shell Welds," dated November 10, 1998.

Page A9 of A32



6. NRC letter, 'Acceptance for Referencing of EPRI Proprietary Report TR-113596, "BWR Vessel and Internals Project, BWR Reactor Pressure Vessel Inspection and Flaw Evaluation Guidelines (BWRVIP-74)," and Appendix A, "Demonstration of Compliance with the Technical Information Requirements of the License Renewal Rule (10 CFR 54.21 ),"' dated October 18, 2001.

Page A10 of A32




. "ALTERNATIVE FOR PRESSURE TESTING AND VISUAL' EXAMINATION OF BURIED PIPING AND COMPONENTS"

March 2017

Page A11 of A32




Revision 1 January 2·7, 2017

ALTERNATIVE REQUIREMENTS FOR BURIED PIPING AND COMPONENTS




Code Class:

Reference:


Item Number:

Component Numbers:

3

ASME Section XI, Table IWD-2500-1

D-B

82.10

Various


lnservice inspections (ISi) at Duane Arnold Energy Center (DAEC) are. performed 'per the. requirements of the ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda as required by 1 OCFR50.55a.


IWA-5244(b) states that for buried components where a VT-2 visual examination cannot be. performed, the examination requirement is satisfied by the following:

. .

1) The system pressure test for buried components that are isolable by means of valves shall consist of a test that determines the rate of pressure loss. Alternatively, the test may determine the change in flow between the ends of the buried components. The acceptable rate of pressure loss or flow shall be established by the Owner.

2) The system pressure test for non-isolable buried components shall consist of a test to confirm that flow during operation is not impaired. ·

3) Test personnel need not be qualified for VT-2 visual examination.

4. Reason for Request

IWA-5244(b)(1) requires either a pressure loss test or a test that determines the change in flow between the ends of the buried components for isolable sections of buried piping. The acceptable rate of pressure loss or flow shall be established by the Owner. Sections of River Water Supply, Emergency Service Water (ESW), and Residual Heat Removal

Page A12 of A32



Service Water (RHRSW) System buried piping were not designed with consideration for isolation valves adequate for performing a pressure loss type test or do not contain instrumentation adequate for measuring changes in flow between the ends of the buried piping.

The River Water Supply System contains large diameter buried piping (24 inch diameter) that runs from the River Intake Structure to the Pump House and is greater than 1500 feet in length. The ESW System and the RHRSW System contain large diameter buried piping (16 inch diameter for RHRSW and 8 inch and 6 inch diameter for ESW) that runs from the Pump House to the Turbine Building and is greater than 500 feet in length.

The subject piping design for these systems did not provide for isolation valves that are capable of supporting a pressure loss type test considering the volume of the piping and the available capacity of test pumps. The system isolation valves were only intended to provide isolation for maintenance activities with only static system pressure.

River Water Supply and ESW were designed with a single flow element per train located in the Pump House. ESW has some additiorial flow instrumentation on some downstream components, but not for every branch on a train. RHRSW was design~d with a single flow element per train located in the Reactor Building before the Residual Heat Removal System. Heat Exchanger. Therefore, the installed instrumentation is inadequate for measuring the flow difference at each end of the buried piping. The use of ultrasonic flow instrumentation was considered, but the piping configurations do not provide for the straight runs of piping required for accurate f'ow measurement. ·

Both the River Water Supply and RHRSW systems include four pumps each with two . pumps designated to each. of two independent trains. The River Water Supply pumps and · RHRSW pumps have installed excess capacity. Therefore, each of the independent trains of both the River Water and RHRSW systems.can accommodate a leak and still satisfy the accident analysis requirements. ESW has one pump per train. The ESW system supplies

· . various plant heat exchangers, which have flow margin due to heat transfer requirements.

5. Proposed Alternative

Pursuant to 10 CFR 50.55a, "Codes and Standards," paragraph(z)(1 ), NextEra Energy· . requests relief from performing the system leakage test for buried portions that are isolable by means of.valves by measuring rate of pressure loss or the change iri flow between the ends of buried components. In lieu of performing a system pressure test in accordance with the requirements specified in IWA-5244(b)(1 ), NextEra Energy shall use the provisions of IWA-5244(b)(2) to confirm that flow during operation is not impaired. IWA-5244(b)(2) states that the system pressure test for non-isolable buried components shall con~ist of a test to confirm that flow during operation is not impaired. The proposed alternative provides an acceptable level of quality and safety.

Basis for Use

The integrity of the buried piping for the River Water Supply and RHRSW, will be verified during quarterly pump testing. Using the downstream instruments, flow rate is set at the fixed reference value in accordance with ASME OM Code 2004 Edition through 2006 Addenda and documented in the test record. The pump discharge. pressure is then measured and used to determine the head produced by the pump. Head and flow rate are interdependent variables, which together define pump hydraulic

Page A13 of A32


Revision 1 · January 27, 2017

I

performance. As the pump degrades, the developed head will decrease at the reference flow rate. However, since the location of the flow rate instruments are downstream of the buried piping, a decrease in pump head, during testing may also indicate through-wall leakage in the buried portion of the RHRSW or River Water Supply piping.

Significant through-wall leakage would be evident because the total flow rate would · increase even though the downstream indicated flow rate is set at the reference value. Therefore, a satisfactory quarterly service water pump test also verifies the integrity of the buried system supply piping. Should the pump test results fall below the required action range of the code, then additional testing and evaluations will be performed to d~termine whether the unsatisfactory test results are due to degraded pump performance or through-wall leakage.

For the ESW System, the integrity of the. buried piping will be verified during the combination of the quarterly pump testing and the verification that adequate flow is supplied to cooling loads as provided by installed instrumentation. Using the upstream

. instruments, flow rate for the ESW pump is set at a fixed reference value in accordance with ASME OM Code 2004 Edition through 2006 Addenda and documented in the test record. Downstream critical heat loads supplied by ESW such as the Emergency Diesel Generator Coolers, Control Building Chillers, and the Residual Heat Removal and Core Spray Room Coolers, have installed flow instrumentation. The installed instrumentation represents approximately 90% of the total critical load flow required to be supplied by ESW. Significant through-wall leakage would be evident by a· marked decrease in supplied flow to the 90% instrumented downstream loads. Annual trending of the ESW .instrumented critical load flow rates compared fo the upstream. set reference value flow verifies the integrity of the buried piping.

In addition, NextEra Energy proposes to perform visual examination of the ground. surface area immediately above each buried section of ESW on a refuel cycle bases in lieu of performing the test required by IWA-5244(b)(1). The visual examinations will be performed only after the subject piping has been in operation at nominal operating

. conditions for a minimum of 24-hours. The ASME s·ection XI code only requirt?s a pressure test once each period (Every 3 to 4 years).

Pursuant to 10CFR50.55a(z)(1), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.


This relief is requested for the duration of the Duane Arnold Fifth lnservice Inspection Interval beginning on November 1, 2016 and scheduled to end on October 31, 2026.

7. Precedents

This relief request was approved for the Duane Arnold Generating Station in the fourth interval as NDE-R007. The NRC SER was issued on June 12, 2007 (TAC No. MD2523 and ADAMS Accession No. ML071380183). ·

Page A14 of A32




1. NRC Safety Evaluation dated May 20, 2015, "Point Beach Nuclear Plant, Units 1 and 2, Relief Request RR-8, Relief from The Requirements of The American Society of Mechanical Engineers Boiler and Pressure Vessel Code for Examination of Buried Components (TAC NOS. MF4140 and MF4141) (ADAMS Accession No. ML 15127A291).

2. NRC Safety Evaluation dated September 19, 2011, "Safety Evaluation of Relief Requests Regarding Pressure Testing Of Service Water System Buried Piping-Salem Nuclear Generating Station, Unit Nos. 1 and 2" (TAC Nos. ME4861 and ME4862, ADAMS Accession No. ML 15134A242). .

Page A15 of A32




"Alternative Requirements For Nozzle Inner Radius And · Nozzle-To-Shell Welds"

March 2017

Page A16 of A32


DUANE ARNOLD ENERGY CENTER FIFTH INSERVICE INSPECTION INTE~.V AL



ALTERNATIVE REQUIREMENTS FOR NOZZLE INNER RADIUS AND NOZZLE-TO-SHELL WELDS




Code Class:

Reference:


Item Number: Component Numbers:

1

ASME Section XI, Table IWB-2500-1

8-D

83.90 and 83.100 · Reactor Vessel Nozzles N1, N2, N3, N5, N6, N7, NS,

N11, N12, and N16 (see Attachment 1 for specific nozzle identifications)


lnservice inspections (ISi) at Duane Arnold Energy Cen'ter (DAEC) are performed per the requirements of the ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda as required by 1 OCFR50.55a.


Table IWB-2500-1 "Examination Category 8-D, Full Penetration Welded Nozzles in Vessels."

Class 1 nozzle-to-vessel weld and nozzle inner radii examination requirements are delineated in Item Number 83.90 "Nozzle-to-Vessel Welds," and 83.100 "Nozzle Inside Radius Section." The required method of examination is volumetric. All nozzles with full penetration welds to the vessel shell (or head) and integrally cast nozzles are examined each interval.

All of the nozzle assemblies identified in Attachment 1 are full penetration· welds.

Page A17 of A32


4. Reason for Request


lnservice inspection of the ASME Code Class 1, 2, and 3 components is performed in accordance with the ASME Section XI code and applicable addenda as a way to detect anomaly and degradation indications so that structural integrity of these components can be maintained. This is required by Title 10 of the Code of Federal Regulations (10 CFR) 50.55a(g), except where specific relief has been granted by the Commission pursuant to 1 OCFR50.55a(g)(6)(i). The provision of 10 CFR 50.55a(z) states that alternatives to the requirements of paragraph (g) may be used, when authorized by the NRC, if: ( 1) the proposed alternatives would provide an acceptable level of quality and safety, or (2) compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety.

For all RPV nozzle-to-vessel shell welds and nozzle inner radii, ASME Code, Section XI requires 100 percent inspection during each 10-year ISi interval. However, ASME Code Case N-702, "Alternative Requirements for Boiling Water Reactor (BWR) Nozzle Inner Radius and Nozzle-to-Shell Welds, Section XI, Division 1," is listed in RG 1.147, Table 2, "Conditionally Acceptable Section XI Code Cases" proposes an

_alternative which reduces the inspection of RPV nozzle-to-vessel shell welds and nozzle inner radius areas from 100 percent to 25 percent of the nozzles for each nozzle type during each 10-year interval. The Condition associated with Code Case N-702 is as follows:

The applicability of Code Case N-702 must be shown by demonstrating that _ the criteria in Section 5.0 of NRG Safety Evaluation regarding BWRVIP-1_08 dated December 18, 2007 (ML073600374) or Section 5.0 of NRG Safety Evaluation regarding BWRVIP-241 dated April 19, 2013 (ML13071A240) are · met. The evaluation demonstrating the applicability of the Code Case shall be reviewed and approved by the NRG prior to the applicability of the Code Case.

The identified nozzles (Attachment 1) are scheduled for examination during the fifth 10-year inspection interval for the Duane Arnold. The proposed alternative provides an acceptable level of quality and safety, and the reduction in scope could provide a dose savings of as much as 25 Rem over the current interval.

5. Proposed Alternative and Basis for Use

Proposed Alternative

Pursuant to 1 OCFR50.55a(z)(1 ), NextEra Energy is requesting relief from performing the required examinations on 100% of the identified ASME Code Table IWB-2500-1, Examination Category B-D, Class 1 nozzle assembly requirements as delineated in Item Number 83.90 "Nozzle-to-Vessel Welds," and 83.100 "Nozzle Inside Radius Section." (Attachment 1). Alternatively, in accordance with Code Case N-702 a minimum of 25% of the nozzle-to-vessel welds and inner radius sections, including at least one nozzle from each system and nominal pipe size, will be volumetrically examined. For the nozzle assemblies identified in Attachment 1, this will mean one

Page A18 of A32


from each of the groups identified below:

Group Total

Number

Recirculation Outlet (N 1 )* 2

Recirculation Inlet (N2) 8

Vessel Instrumentation 6

(N11, N12, N16)

Core Spray (NS) 2

Nozzles on Vessel Top 3 Head (N6, N7))

Jet Pump (N8) · 2

Main Steam (N3) 4

*This group does not meet Criteria 4 of the NRC SER1.

Revision 1 ·January 27, 2017

Number to be Examined

2

2

2

1

1

1

1

Footnote 1 - The RPV wall thickness was taken from the Form N-1 Data Sheet.

Code Case N-702 also stipulates that VT-1 examination may be used in lieu of the volumetric examination for the inner radii. NRC has not endorsed the use of Code Case N-702 for VT-1 examinations. Duane Arnold will utilize Code Case N-648-1,· "Alternative Requirements for Inner Radius Examination of Class 1 Reactor Vessel Nozzles, Section XI Division 1," with associated required conditions specified in RG · 1.147, if a VT-1 examination is performed in lieu of a volumetric examination .. Volumetric examination of the nozzle inner radii is the preferred method.

·Basis for Use

EPRI Technical Report 1003557, "BWRVIP-108: BWR Vessel and Internals Project Technical Basis for the Reduction of Inspection Requirements for the Boiling Water Reactor Nozile-to-Vessel Shell Welds and Nozzle Blend Radii," provides the basis for Code Case N-702. The evaluation 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 106 for 40 years) with or without inservice inspection. The report concludes that inspection of 25% of each nozzle type is technically justified.

This report received an NRC Safety Evaluation (SER) dated December 19, 2007. In· the SER, Section 5.0 "Plant Specific Applicability" requires each licensee who plans to request relief from the ASME Section XI requirements for RPV nozzle-to-vessel shell welds and nozzle inner radii sections, may reference the BWRVIP-108 report as the technical basis for use of ASME Code Case N-702 as an alternative. However, each licensee should demonstrate the plant specific applicability of the BWRVIP-108 report to t.heir units in the relief request, by showing that all the following general and nozzle-specific criteria are satisfied:

(1) The maximum RPV heatup/cooldown rate is limited to less than 115°F per

Page A19 of A32



hour. The Duane Arnold surveillance that monitors reactor vessel heatup/cooldown (STP 3.4.9-01) limits the rate to less than 100°F for Curve B and less than 20°F for Curve A.

(2) For the Recirculation Inlet Nozzles, the following criteria must be met:

a. (pr/t)/CRPV<1.15, the calculation for the Duane Arnold N2 Nozzle results in 0.97 48 which is less than 1.15.

b. [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE<1.15, the calculation for the Duane Arnold N2 Nozzle results in 1.0923 which is less than 1.15.

(3) For the Recirculation Outlet Nozzles, the following criteria must be met:

a. (pr/t)/CRPV<1.15, the calculation for the Duane Arnold N1· Nozzle results in 1, 17 which is higher than 1.15.

b. [p(ro2+ri2)/(ro2-ri2)]/CNOZZLE<1.15, the calculation for the Duane Arnold N1 No.zzle results in 0.87 which is less than 1.15.

So based on the above information, all RPV nozzle-to-shell welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles, meet the criteria and therefore Code Case N-702 is applicable. However, the Recirculation .outlet Nozzles do not meet all of the criteria and ·code Case N-702 would not be applied. See Attachment 2 for details. · ·

Therefore, use of Code Case N-702 provides an acceptable level. of quality and safety pursuant to 10 CFR 50.55a(z)(1) for all nozzle.-to-vessel shell welds and nozzle inner radii sections, with the exception of the Recirculation Outlet Nozzles.


This relief is requested for the duration of the Duane Arnold Fifth lnservice Inspection Interval beginning on November 1, 2016 and scheduled to end on . .

October 31 , 2026.

7. Precedents

This relief request was approved for Duane Arnold in the fourth interval as NDE-R013. The NRC SER was issued on August 29, 2008 (TAC No. ME7246 and ADAMS Accession No. ML082040046).


1. NRC Safety Evaluation dated Detember 6, 2016, "James A Fitzpatrick Nuclear Power Plant - Relief from the Requirements of the ASME Code Case N-702 and BWRVIP-241 For Plant Nozzle-To-Vessel Welds And Nozzle Inner Radii" (CAC NO. MF8301) (ADAMS Accession No. ML 16334A440).

Page A20 of A32



2. NRC Safety Evaluation dated October 5, 2016, "Columbia Generating Station - Relief Request For Alternative 41Sl-04 Applicable To The Fourth 10-Year lnservice Inspection Program Interval (CAC No. MF7331) (ADAMS Accession No. ML 16263A233).

3. NRC Safety Evaluation dated October 30, 2015, "LaSalle County Station, Units 1 and 2, Relief from the Requirements of the ASME Code Re: RR 13R14, Proposed Alternative to the Examination Requirements for Nozzle-to-Vessel Welds and Inner Radii Sections in Accordance with 10 CFR 50.55a(z)(1)" (TAC Nos. MF5654 and MF5655) (ADAMS Accession No. ML 15226A412).

4. NRC Safety Evaluation dated May 20, 2015, "Cooper Nuclear Station - Relief Request No. Rl-08, Revision 0 Applicable to Fourth 10-Year lnservice Inspection Interval" (TAC No. MF4429) (ADAMS Accession No. ML 15134A242).

5. NRC Safety Evaluation dated October 1, 201 o .. "Relief Request RR-A37 for Fermi 2 RE: Evaluation of Alternative to Reactor Pressure Vessel Nozzle-to-Vessel Welds and Inner Radius Examinations" (TAC No. ME3117) (ADAMS Accession No. ML 102590141).

8. Attachments

Attachment 1 : Applicable Nozzles Attachment 2: Responses to NRC Plant Specific Applicability

Page A21 of A32


Attachment 1 Applicable Nozzles

. cat Item No . Summary Component Nozzle

Dwg/ISO No. No. ID ID

8-D 83.90 9700 RRA-D001 N2A 1.2-22 SHT-01

8-D 83.100 9800 RRA-D001-

1.2-22 SHT-01 INNER RAD

8-D 83.90 9900 RR8-D001 N28 1.2-22 SHT-01

8-D 83.100 10000 RR8-D001-


8-D 83.90. 10100 RRC-D001 N2C 1.2-22 SHT-01

8-D 83.100 10200 RRC-D001-


8~D 83.90 10300 RRD-D001 N2D 1.2-22 SHT-01

8-D 83.100 10400 RRD-D001-


B-D 83.90 10500 RRE-D001 N2E 1.2-20 SHT-01

. 8-D 83.100 10600 RRE-D001-


8-D · 83.90 10700 RRF-D001 N2F 1.2-20 SHT-01

8-D 83.100· 10800 RRF-D001-


8-D 83.90 10900 RRG-D001 N2G 1.2-20 SHT-01

8-D 83.100 11000 RRG-D001-


8-D 83.90 11100 RRH-D001 N2H 1.2-20 SHT-01

8-D 83.100 11200 RRH-D001-


8-D 83.90 11300 VIA-D001 N11A 1.2-28 SHT-01

8-D 83.100 11400 VIA-D001-

1.2-28 SHT-01 INNER.RAD

8-D 83.90 11500 Vl8-D001 N118 1.2-29 SHT-01

8-D 83.100 11600 Vl8-D001-

1.2-29 SHT ~01 INNER RAD

8-D 83.90 11700 VIC-D001 N12A 1.2-30 SHT-01

8-D 83.100 11800 VIC-D001-


Page A22 of A32

System

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

RR

VI

VI

VI

VI

VI

VI


Nominal Comments Pipe Size

10"

10"

10"

10"

10"

10"

10"

10"

10".

10"

10"

10"

10"

10"

10"

10"

2.5"

2.5"

2.5"

2.5"

2.5"

2.5"



Cat Item No. Summary Component Nozzle


8-0 83.90 11900 VID-D001 N128 1.2-31 SHT-01

8-D 83.100 12000 VID-D001-


8-0 83.90 12100 VIE-D001 N16A 1.2-33 SHT-01

8-D 83.100 12200 VIE-0001-


8-D 83.90 12300 VIF-D001 N168 1.2-34 SHT-01

8-D 83.100 12400 VIF-0001-


1.2-12A SHT-8-0 83.90 4600 CRA-D001 N9

01

CRA-0001- 1.2-12A SHT-8~0 83.100 4700

INNER RAD 01

8-D 83.90 ·4800 CSA-0001. N5A 1.2-07 SHT-01

8-0 . 83.100 4900 CSA-0001-

1.2-07 .SHT-01 INNER RAD

8-D 83.90 5000 CS8-D001 N58 1.2-08 SHT-01

8-D 83.100 5100 CS8-D001-

1.2-08 SHT-01. INNER RAD

8-D 83.90 5300" FWA-D001 N4A 1.2-05 SHT-01

FWA-D001-8-D 83.100 5400 1.2-05 SHT-01

INNER RAD

8-D 83.90 5800 FW8-D001 N48. 1.2-05 SHT-01

FW8-D001-8-0 83.100 5900 1.2-05 SHT-01

INNER RAD

8-D 83.90 6300 FWC~Doo1. N4C 1.2-06 SHT"01

FWC-D001-8-D 83.100 6400 1.2-06 SHT-01

INNER RAD

8-0 . 83.90 6800 FWD-D001 N4D 1.2-06 SHT-01

Page A23 of A32

System

VI

VI

VI

VI

VI

VI

CR

CR

cs

cs

cs

cs

FW

FW

FW

FW

FW

FW

FW



2.5"

2.5'.'

2.5"

2.5"

2.5"

2.5" .. Cannot use for

2.5" Code Case N-702

Cannot use for

2.5" Code Case N-702

8"

8"

8"

8';

Cannot use for

10" Code Case N-702

Cannot use for

10" Code Case N-702

Cannot use for

10" Code Case N-702

Cannot use for

10" Code Case N-702

Cannot use for

10" Code Case N-702

Cannot use for 10" Code Case

N-702 Cannot use for

10" Code Case N-702

- ----". ~--=--=~--



C~t Item No. Summary Component Nozzle


8-D 83.100 6900 FWD-0001-


8-0 83.90 7200 HOA-D001 N15 1.2-32 SHT-01

8-D 83.100 7250 HDA-D001-

1.2-32. SHT-01 INNER RAD

8~D 83.90 7300 HS8-D001 N68 1.2-23 SHT-01

8-D 83.100 7400 HS8-0001-


8-D 83.90 7500 HVA-D001 N7 1.2-24 SHT-01

8-0 83.100 7600 HVA-D001-


8~D 83.90 9500 RHA-0001 N6A · 1.2-13 $HT-01

8-D 83.100 9600 RHA-0001-

1.2-13 SHT~01 INNER RAD

8-D 83.90 7700 JPA-0001 NBA 1.2-25 SHT-01

8-D 83.100 7800 JPA-D001-

1:2-25 SHT-01 INNER RAD

8-0 83.90 7900 JP8-0001 N88 1.2-26 SHT-01

8-D 83.100 8000 JP8-D001-

1.2-26 SHT-01 INNER RAD .

8-D 83.90 8100 LCA-0001 N10 1.2-27 SHT:.01

8-D 83.100 8200 LCA-0001-


8-D 83.90 8300 MSA-D00.1 N3A 1.2-01 SHT-01

8-D 83.100 8400 MSA-D001-


8-D 83.90 8500 MS8-D001 N38 1.2-02 SHT-01

8-0 83.100 8600. MS8-D001~


8-D 83.90 8700 MSC-D001 N3C 1.2-03 SHT-01

8-0 83.100 8800 MSC-D001-


8-D 83.90 8900 MSO-D001 N3D 1.2-04 SHT-01

Page A24 of A32

System

FW

HD

HD

HS

HS

HV

HV

RH

RH'.

JP

JP

JP

. JP

LC

LC

MS

MS

MS

MS

MS

MS

MS



Cannot use for

10" Code Case N-702

2" Exempt by

IW8-1220(c)

2" Exempt by

IW8-1220(c)

6".

6"

4"

4"

6"

6i'

4"

4"

4"

4" -.

2"

'2"

20"

20"

20"·

20"

20"

20"

20"



Cat Item No. Summary Component Nozzle


8~D 83.100 9000 MSD-D001-


1.2-19A SHT-8-D 83.90 9100 RCA-D001 N1A

01

RCA-D001- 1.2-19A SHT-8-D 83.100 9200

INNER RAD 01

1.2-21A SHT-8-D 83.90 9300 RC8-D001 N18

01

RC8-D001- 1.2-21A SHT-8~0 83.100 9400

INNER RAD 01

Page A25 of A32

System

MS

RC

RC

RC

RC

.Revision 1 January 27, 2017


20"

Does not meet 22" Criteria 4 of the

NRC SER


NRC SER

Does not meet 22"· Criteria 4 of the

NRC SER


NRC SER


Attachment 2

Revision 1 January 27 , 2017

Responses to NRC Plant Specific Applicability

1) The maximum Reactor Pressure Vessel (RPV) heat-up/cool-down rate is limited to less than 115 °F/hour.

Response: STP 3.4.9-01 limits the RPV heat-up/cool-down to::; 100 °F for Curve Band::; 20 °F for Curve A

Reci rculation Inlet Nozzles Recirculation Outlet Nozzles

2) (pr/t)/CRPv<1 .15 4) (pr/t)/CRPv<1 .15

p=RPVNormal Operating Pressure 1025 p=RPVNormal Operating Pressure 1025

r=RPVinner radius 92.5 r=RPVinner radius 92 .5

t=RPVwall thickness 5.031 t=RPVwall thickness 5.031

CRPV 19332 CRPV 16171

0.9748 <1 .15 1.17 <1.15

3) [p( ro2+ri2)/ ( ro2 -ri2) ]/CNOZZLE< 1.15 5) [p( ro2+ri2

)/ ( ro2 -ri2) ]/CNOZZLE< 1 .15

p=RPVNormal Operating Pressure 1025 p=RPVNormal Operating Pressure 1025

r0 = nozzle outer radius 10.56 r0= nozzle outer radius 19.628

ri= nozzle inner radius 5.5 ri= nozzle inner radius 9.875

CNOZZLE 1637 CNOZZLE 1977

1.0923 <1.15 0.87 <1 .15

Page A26 of A32

Duane Arnold Energy Center ISi Program Document No_. 5th lnterval-ISl-PDA-Program Plan



"Alternative For Seal Weld Procedure Qualification"

· March2017

Page A27 of A32 - - -- ·---~-------:---:_ ~----~--~"':- ·-· -----~ --- .---·-.-::·~-~-.:- -- ------·-~-·-.·-::--:--.-----~ -------- - _--

Duane Arnold Energy Center ISi Program Document No. 5th interval-ISl-PDA.,Program Plan




ALTERNATIVE FOR SEAL WELD PROCEDURE QUALIFICATION


. --Hardship or Unusual Difficulty without Compensating Increase in Level of Quality and Safety--

--:··-::- ~-- - - -~- -----:-- ......


Code Class:

Reference:

· Component Numbers:

1

ASME Section XI, IWA-4221

PSV-4400, PSV-4401, PSV-4402, PSV-4405, PSV-4406, PSV-4407, and six uninstalled spare Main Steam Relief Valves (MSRVs)


lnservice inspections (ISi) at Duane Arnold Energy Center (DAEC) are performed per the requirements of the ASME Boiler and Pressure Vessel Code Section XI, 2007 Edition with the 2008 Addenda as required by 1 OCFR50.55a.


IWA-4221, Construction Code and Owner's Requirements

IWA-4421, "Construction Code and Owner's Requirements," states that:

"(a) An item to be used for repairliep/acement activities shall meet the Owner's Requirements. Owner's Requirements may be revised, provided they are reconciled in

·accordance with /WA- 4222. Reconciliation documentation shall be prepared. ·

(b) An item to be used for repair/replacement activities shall meet· the Construction Code specified in accordance with (1 ), (2), or (3) below: ·

(1) When replacing an existing item, the new item shall meet the Construction · Code to which the original item was constructed.

(2) When adding a new component to an existing system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for construction of the system or of any originally installed component in that system.

(3) When adding a new system, the Owner shall specify a Construction Code that is no earlier than the earliest Construction Code used for other systems that perform a similar function.

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(c) As an alternative to (b) above, the item may meet all or portions of the requirements of different Editions and Addenda of the Construction Code, or Section Ill when the Construction Code was not Section Ill, provided the requirements of IWA-4222 through IWA-4226, as applicable, are met. Construction Code Cases may also be used. Reconciliations required by· this Article shall be documented. All or portions of later different Construction Codes may be used as listed below:

(1) Piping, piping subassemblies, and their supports: 831.1to831. 7 to Section Ill.

(2) Pumps, valves, and their supports: fiom 831.1 to Draft Code for Pumps and Valves for Nuclear Power to Section Ill.

(3) Vessels and their supports: Section VIII to Section Ill. (4) Atmospheric and 0-15 psig (0-103 kPa) storage tanks and their supports: .

Section VIII, AP/ 620, or AP/ 650 to Section Ill:"

Installed and spare MSRVs were originally manufact.ured to the 1968 Edition, Winter. 1968 Addenda of ASME, Section Ill, and General Electric Design Specification 21A9206 Rev. 6. All subsequent Repair/Replacement Activities were intended to meet .

. the original construction code and design specification. All Repair/Replacement Activity seal welding has been performed by the original Manufact!Jrer (Target R9ck) using the same Weld Prncedure Specifications (WPSs) as used during original fabrication.

4. Reason for Reg uest

The original weld procedure specifications (WPS) for seal weldirig were qualified in accordance with the Manufacturer's standard rather than ASME. requirements. The Manufacturer's standard included multiple surface Non-Destructive Examinations . (NOE) and macro examinations of sectioned specimens. The three seal welds affected · are the bellows-to-spacer plate seal weld, the pilot seat-to-body seal weld, and the second stage seat-to-body seal weld. (Reference Figure 1 for seal welds).

The 1968 Edition, Winter 1968, Addenda of ASME Code, Section Ill, did not inelude fabrication requirements for valves or provide any requirements for seal welding. The General Electric design specification required weld procedures to be qualified in accordance with ASME Section IX. However, the author of this design specification, and the original purchaser of the MSRVs (General Electric), believed that this requirement was never intended to be applicable to seal welding. It is reasonable to assume that the 1968 Edition of ASME Section IX should have been used to qualify seal welding procedures, since the 1968 Edition of ASME Section IX, paragraph Q-10 (b), requires all welding to be qualified using reduced section tension specimens and guided bend specimens.

Although the Target Rock procedures have since· been revised to reference the new procedure qualification records (PQRs) using the same seal weld parameters as the original seal welding procedure and meeting all tensile and bend tests requirements in accordance with ASME Section IX, a hardship still exists to meet the ASME Section XI Code requirements for having a qualified weld procedures prior to production welding and subsequent code stamping. As stated in the previous interval relief request, in order to meet this requirement, all seal welds would have to be removed and re-welded

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Revision 1 January 2,7, 2017

using the same welding procedure that is now pre-qualified. Also, removal of existing seal welds. would require th~ MSRV be completely disassembled, the seat rings replaced, ·and the reassembled valve tested. This unnecessary welding evolution could potentially degrade the carbon steel casting. Therefore, replacement of existing seal welds is considered a hardship, or unusual difficulty, without a compensating increase in the level of quality and. safety.

5. Proposed Alternative

Pursuant to 1 O CFR 50.55a, "Codes and Standards," paragraph(z)(2), NextEra Energy requests authorization to continue using the updated Target Rock WPSs for the seal welding of the installed and spare MSRVs that were post-production qualified in accordance with ASME Code, Section IX requirements. This proposed alternative is deemed to be acceptable because NextEra Energy considers implementation of the applicable Code requirements to be a hardship or unusual difficulty without a compensating increase in level of quality and safety.

Basis for Use .

In 2007, Target Rock completed three procedure qualification records (PQR's) using· the same seal welding parameters as in the original seal welding procedures and the weld coupons were tested in accordance with the 2004 Edition, 2006 Addenda, of ASME Section IX. All tensile and bend testing was found acceptable to the ASME . Code, Section IX requirements. All three seal weld WPSs have been revised to reference the new PQRs that were qualified via tensile and bend testing.

These post-qualifying PQRs verify that the seal welds made with the original seal welding WPSs meet all tensile and bend test requirements and justify continued use. The revised seal welding WPSs that now reference the new PQRs are planned to be used during future Repair/Replacement activities, if performed by Target Rock.

· 6. Duration for the Proposed Alternative

This relief is requested for the Duane Arnold Fifth lnservice Inspection Interval beginning on November 1, 2016 and scheduled to end on October 31, 2026.

7. Precedents

This relief request was approved for the Duane Arnold in the fourth interval as NDER012. The NRC SER was issued on July 2, .2008 (TAC No. MD6293 and ADAMS

·Accession No: ML081680709).

8. ·Attachments

Figure 1 - Seat Seal Welds

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Figure 1

Seat Seal Welds


Note: The 1968 Edition, Winter 1968 Addenda of ASME Section Ill, which was the design Code utilized for the Duane Arnold MS RVs, did not include fabrication requirements for valves or provide any requirements for seal welding.

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"Risk Informed ISi Program"

To be submitted at a later date

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