Managing PWSCC in 

Butt Welds by Mitigation 

and Inspection 

Office of Nuclear Regulatory Research

NUREG/CR-7187

PNNL-23659
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Managing PWSCC in 

Butt Welds by Mitigation 

and Inspection 

Manuscript Completed: September 2014

Date Published: November 2014

Prepared by

E. J. Sullivan and M. T. Anderson

Pacific Northwest National Laboratory

P.O. Box 999

Richland, WA 99352

W. E. Norris, NRC Project Manager

NRC Job Codes V6085/V6411

Office of Nuclear Regulatory Research

NUREG/CR-7187

PNNL-23659

ABSTRACT

Operating experience has shown that Alloy 82/182/600 materials exposed to primary coolant

water (or steam) under the normal operating conditions of pressurized water reactors (PWRs) 

are susceptible to primary water stress corrosion cracking (PWSCC). These materials were 

widely used in the original construction of domestic PWR reactor coolant systems (RCS). 

Cracking typically initiates at the inside surface of these materials when exposed to reactor 

coolant and in the presence of high tensile residual stresses introduced by welding. The 

locations in which PWSCC has been identified include partial penetration nozzles, nozzle welds,

and piping butt welds. PWSCC increases the probability of a loss-of-coolant accident. This 

NUREG/CR pertains to managing PWSCC in Alloy 82/182 butt welds.

In response to this degradation mechanism, the nuclear industry and the U.S. Nuclear 

Regulatory Commission (NRC) undertook a variety of actions. These actions included 

conducting crack growth rate studies for Alloy 82/182/600 materials, developing augmented 

inspection programs for varied component configurations, and establishing an improved 

regulatory environment for addressing PWSCC. In Alloy 82/182 butt welds, the nuclear industry 

began implementing mitigation strategies for the most susceptible of these welds.

This report provides a summary of the operating experience with PWSCC in Alloy 82/182 butt 

welds and an overview of industry and regulatory activities to address the safety concern. This 

report describes the strategies being used to manage potential PWSCC in butt welds, the basic 

design and regulatory requirements of each strategy/mitigation technique, and an assessment 

of the issues that may impact the effectiveness of each management strategy. This report also 

summarizes the NRC regulatory requirements, ASME Code Cases, and the regulatory status of 

ASME Code Cases related to management of PWSCC in Alloy 82/182 butt welds. A historical 

summary and timeline of the major events, activities, and decision processes is provided in an 

appendix.

The effectiveness of the mitigation and inspection strategies being used is evaluated and issues 

that could affect the reliability of inspections that are required to be performed are also 

discussed.

PAPERWORK REDUCTION ACT STATEMENT

This NUREG does not contain information collection requirements and, therefore, is not subject 

to the requirements of the Paperwork Reduction Act of 1995 (44 U.S.C. 3501 et seq.).

PUBLIC PROTECTION NOTIFICATION

The NRC may not conduct or sponsor, and a person is not required to respond to, a request for 

information or an information collection requirement unless the requesting document displays a 

currently valid OMB control number.

iii

FOREWORD

The U.S. Nuclear Regulatory Commission (NRC) started approving leak-before-break (LBB) 

analyses in 1984 by granting exemptions from General Design Criterion (GDC)-4, 

“Environmental and dynamics effects design bases.” In 1987, GDC-4 was revised to allow 

dynamic effects associated with postulated pipe ruptures to be excluded from the design basis 

when analyses reviewed and approved by the Commission demonstrate that the probability of 

fluid system piping rupture is extremely low (i.e., less than 10 -6 /RY). The statement of 

considerations for the proposed revision to GDC-4 in 1986 said that “the leak-before-break 

approach should not be considered applicable to fluid system piping that operating experience 

has indicated is particularly susceptible to failure from the effects of corrosion.” Standard 

Review Plan (SRP) 3.6.3, “Leak-Before-Break Evaluation Procedures,” states that, “... 

evaluations must demonstrate that these [degradation] mechanisms are not potential sources of 

pipe rupture.” In practice, review criteria were implemented by excluding systems with potential 

corrosion degradation mechanisms. Satisfying SRP review criteria was considered a 

demonstration that the probability of fluid system piping rupture is extremely low. 

SRP 3.6.3 also contains guidance on the application of LBB to boiling water reactor (BWR) 

piping which is susceptible to intergranular stress corrosion cracking (IGSCC). The SRP 

indicates that LBB could be considered for this piping provided at least two mitigation methods 

(e.g., resistant materials, stress improvement, enhanced water chemistry) were applied to the 

piping. In the regulatory actions taken to provide acceptable inspection intervals for managing 

IGSCC in BWR piping, credit has been given for the number of mitigation techniques employed. 

At the time these criteria were developed, they were based on engineering judgment. However, 

it has been observed through operating experience that two mitigation methods in BWR piping 

provide improved resistance to IGSCC as compared to one method, and that the use of two

mitigation methods renders the piping highly resistant to cracking. Nevertheless, owners of 

BWRs have not requested NRC approval to apply LBB to this piping.

Beginning in 2001, operating experience in U.S. pressurized water reactors (PWR) occurred 

indicating that Alloy 82/182/600 materials in butt-welded configurations in the reactor coolant 

pressure boundary (RCPB) were susceptible to primary water stress corrosion cracking 

(PWSCC). The NRC determined that there was a low probability of near-term failure of welds, 

and that PWRs could continue to operate safely while the industry performed additional 

analyses and inspections (Sheron 2001). The Materials Reliability Program (MRP), an industry 

group formed to develop, in part, strategies for managing aging effects in nuclear plant systems 

and components, developed mandatory inspection and evaluation guidelines for dissimilar metal 

butt welds in the reactor coolant system of PWRs. These guidelines are contained in MRP-139, 

“Primary System Piping Butt Weld Inspection and Evaluation Guidelines,” and were 

implemented under Nuclear Energy Institute (NEI) 03-08, “Guidelines for the Management of 

Materials Initiatives.” The objective of MRP-139 was to manage PWSCC through a combination 

of inspection and mitigation, and the guidelines do not discriminate between welds approved by 

the NRC for LBB and other dissimilar metal butt welds.

The development of mitigation and inspection strategies, and the assessment of their

effectiveness, to ensure that the probability of rupture remains extremely low has been 

v
complicated. Several projects and multiple experts have been required. The purpose of this 

report is to provide a historical summary and timeline of the major events, an overview of 

industry and regulatory activities to address the safety concern, and the strategies being used to 

manage potential PWSCC in butt welds. The basic design and regulatory requirements of each 

strategy/mitigation technique, and an assessment of the issues that may impact the 

effectiveness of each management strategy, are also provided. This report will be a useful 

resource in understanding the history of LBB and recent efforts to address a new degradation 

mechanism.

vi
CONTENTS

ABSTRACT...............................................................................................................................iii

FOREWORD ..............................................................................................................................v

FIGURES...................................................................................................................................ix

TABLES ....................................................................................................................................ix

EXECUTIVE SUMMARY ...........................................................................................................xi

ACKNOWLEDGMENTS...........................................................................................................xv

ACRONYMS AND ABBREVIATIONS....................................................................................xvii

1 INTRODUCTION..............................................................................................................1-1

2 BACKGROUND ...............................................................................................................2-1

2.1 Major PWSCC Events ...............................................................................................2-1

2.2 Industry, ASME Code, and NRC Actions...................................................................2-3

2.3 Mitigation Techniques ...............................................................................................2-5

2.3.1 Weld Overlays ..............................................................................................2-5

2.3.2 Mechanical Stress Improvement Process.....................................................2-7

2.3.3 Inlays/Onlays................................................................................................2-8

2.3.4 Emergent Mitigation Techniques ................................................................2-10

3 MANAGEMENT OF PWSCC IN UNMITIGATED WELDS BY INSPECTION ...................3-1

3.1 NDE Reliability Issues in Unmitigated Welds.............................................................3-1

3.1.1 Introduction...................................................................................................3-2

3.1.2 Discussion....................................................................................................3-2

3.2 Inspection Limitations................................................................................................3-4

3.3 Uncertainty for Appendix VIII Piping Examinations Conducted from the Inside 

Surface......................................................................................................................3-7

3.4 Assessment of Code Case N-770-1 Inspection Requirements ..................................3-8

3.4.1 Overview of Code Case N-770-1..................................................................3-8

3.4.2 Code Case N-770-1, Table 1 Notes............................................................3-10

3.4.3 NRC Conditions on Code Case N-770-1.....................................................3-11

3.4.4 Assessment of Unmitigated Weld Inspection Intervals................................3-13

3.5 Conclusions ............................................................................................................3-15

4 MANAGEMENT OF PWSCC BY MSIP............................................................................4-1

4.1 Operating Experience with BWR Welds Mitigated by MSIP.......................................4-1

4.2 MSIP Design Requirements ......................................................................................4-2

4.3 NDE Reliability ..........................................................................................................4-3

4.3.1 Ignalina Nuclear Power Plant........................................................................4-3

4.3.2 PNNL Dissimilar Metal Weld Specimen ........................................................4-5

4.3.3 Comparison of Ignalina and PNNL 9C003 Specimen Data ...........................4-6

4.3.4 Inspection Limitations ...................................................................................4-7

4.4 Assessment of the Effectiveness of MSIP as a PWSCC Mitigation Strategy .............4-8

4.5 Assessment of Code Case N-770-1 Inspection Requirements ..................................4-9

vii
4.6 Conclusions ............................................................................................................4-12

5 MANAGEMENT OF PWSCC BY WELD OVERLAY ........................................................5-1

5.1 Operating Experience with Welds Mitigated by Weld Overlays..................................5-1

5.2 WOL Design Requirements.......................................................................................5-4

5.3 NDE Reliability ..........................................................................................................5-7

5.3.1 Inspection of Dissimilar Metal Weld Overlay Specimens...............................5-8

5.3.2 Inspection Limitations .................................................................................5-10

5.4 Assessment of the Effectiveness of Weld Overlays as a PWSCC Mitigation 

Strategy...................................................................................................................5-12

5.5 Status of ASME Code Cases N-740-X, “Full Structural Dissimilar Metal Weld 

Overlay for Repair or Mitigation of Class 1, 2, and 3 Items, Section XI, Division 

1,” and N-754, “Optimized Structural Dissimilar Metal Weld Overlay for 

Mitigation of PWR Class 1, Section XI, Division 1 ...................................................5-14

5.6 Assessment of ASME Code Case N-770-1 Inspection Requirements .....................5-14

5.7 Conclusions ............................................................................................................5-18

6 MANAGEMENT OF PWSCC BY WELD INLAYS/ONLAYS.............................................6-1

6.1 Operating Experience with Weld Inlays/Onlays .........................................................6-1

6.2 Code Case N-766 Design and Examination Requirements for Inlays and 

Onlays.......................................................................................................................6-6

6.3 NDE Reliability ..........................................................................................................6-8

6.3.1 Qualification and Acceptance Standards for Surface Examination................6-8

6.3.2 Ultrasonic Equivalency Testing of Weld Inlaid Components .........................6-9

6.4 Assessment of the Effectiveness of Weld Inlays/Onlays as a PWSCC 

Mitigation Strategy ..................................................................................................6-11

6.5 Status of Code Case N-766, “Nickel Alloy Reactor Coolant Inlay and Onlay for 

Mitigation of PWR Full Penetration Circumferential Nickel Alloy Dissimilar Metal 

Welds of Class 1 Items” ..........................................................................................6-15

6.6 Assessment of Code Case N-770-1 Inspection Requirements ................................6-17

6.7 Conclusions ............................................................................................................6-20

7 OVERALL CONCLUSIONS .............................................................................................7-1

8 REQUIREMENTS SUMMARY .........................................................................................8-1

9 REFERENCES.................................................................................................................9-1

APPENDIX A – MAJOR EVENTS AND ACTIVITIES.............................................................A-1

APPENDIX B – OPERATING EXPERIENCE WITH BUTT WELD PWSCC...........................B-1

APPENDIX C – DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL WESTINGHOUSE PLANT............................................................ C.1

APPENDIX D – DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL COMBUSTION ENGINEERING PLANT....................................... D.1

APPENDIX E – ASSESSMENT OF WELD RESIDUAL STRESS AND CRACK 

GROWTH ANALYSIS IN UNMITIGATED WELDS.......................................E-1

viii
FIGURES

6.1 V.C. Summer, Unit 1 Replacement Steam Generator Primary Inlet/Outlet 

Nozzle to Safe End Weld Design ....................................................................................6-3

6.2 Weld Inlay Configuration Proposed by Industry (Tsao 2008c).........................................6-4

6.3 Weld Onlay Configuration Proposed by Industry (Tsao 2008b).......................................6-4

TABLES

3.1 RCS Alloy 82/182 Butt Weld Code Required Volume Coverage Obtained for a 

Typical Westinghouse Plant............................................................................................3-5

3.2 RCS Alloy 82/182 Butt Weld Code Required Volume Coverage Obtained for a 

Typical Combustion Engineering Plant ...........................................................................3-5

3.3 Examinations for Inspection Items A-1, A-2, and B .........................................................3-9

ix

EXECUTIVE SUMMARY

Operating experience has shown that Alloy 82/182/600 materials exposed to reactor coolant

water (or steam) under the normal operating conditions of pressurized water reactors (PWRs) 

are susceptible to primary water stress corrosion cracking (PWSCC). PWSCC was observed in 

Alloy 600 mill annealed steam generator tubing by the early 1980s (Cullen and Mintz 2004). 

PWSCC has affected a number of other nickel alloy reactor coolant systems, such as upper and 

lower reactor vessel head penetrations and various vessel shell/nozzle and piping butt welds. 

This report provides an assessment of PWSCC in reactor pressure vessel (RPV) nozzle and 

piping butt welds.

Operating experiences beginning with the V.C. Summer through-wall axial cracking event in 

2000 prompted the nuclear industry to issue inspection guidelines, as the safety consequences 

of inadequate inspections could have been significant. In 2005 the industry Materials Reliability 

Program (MRP) issued MRP-139, Primary System Piping Butt Weld Inspection and Evaluation 

Guidelines (MRP 2005). MRP-139 provided industry guidance for volumetric and visual 

inspections of unmitigated and mitigated butt welds in PWR primary coolant systems. 

Appendix B of this report provides a summary of operating experience with PWSCC in butt 

welds from the first incidence at Palisades in 1993 (Rogers 1993; NRC 2008a) through the 

present. 

In 2005 the American Society of Mechanical Engineers (ASME) approved the development of a

Code Case on appropriate inspection requirements to address PWSCC in Class 1 butt welds 

containing Alloy 82/182. This case was later numbered Code Case N-770 (ASME 2009b). 

Code Case N-770 was revised in 2009 to address NRC concerns and Code Case N-770-1 was 

issued later that year (ASME 2009a). Code Case N-770-1 contains requirements for inspection 

of unmitigated as well as Alloy 82/182 reactor coolant system (RCS) butt welds mitigated by 

certain techniques. The NRC incorporated ASME Code Case N-770-1 by reference into 

10 CFR 50.55a(g)(6)(ii)(F) (76 FR 36232) in June 2011. This rule includes 10 conditions the 

NRC placed on the use of this Code Case to address safety concerns the NRC had with the 

implementation of Code Case N-770-1.

The predominant methods used by the nuclear industry to mitigate and repair PWSCC 

susceptible welds include weld overlays and the mechanical stress improvement process

(MSIP). With weld overlays, Alloy 52/152 is applied by welding on the outside pipe wall over the 

susceptible weld and on some distance on either side of the weld. Weld overlays may have a 

thickness that exceeds one-third the thickness of the original pipe weld. Weld overlays may 

generate compressive residual stresses in the inner region of the pipe weld to minimize the 

likelihood of crack initiation or propagation of PWSCC through the weld. However, the main 

function of weld overlays is to provide structural reinforcement and an improved platform for 

performing inservice inspection. MSIP modifies the existing ID residual tensile stresses in the 

weld metal and heat-affected zone of butt welds in piping. A load is applied to the outside 

surface of the pipe with a large two-piece mechanical hydraulic clamping device connected by 

two pairs of tangentially positioned studs. MSIP results in compressive residual stresses in the 

interior region of the pipe weld to minimize the likelihood of initiation or propagation of PWSCC 

xi
into the weld. These two methods had been used successfully in boiling water reactors (BWRs) 

to repair and mitigate welds susceptible to intergranular stress corrosion cracking. 

Another mitigation method that has been used by the nuclear industry, to a much less extent,

involves the application of Alloy 52/152 material over the inside of a susceptible weld to provide 

a barrier between the susceptible material and reactor coolant. Alloy 52/152 material is more 

resistant to PWSCC than the underlying Alloy 82/182 weld material. This barrier is called either 

an inlay or onlay, depending upon the preparation of the weld to be mitigated.

This report provides an assessment of the effectiveness of nondestructive examination (NDE) 

performed on mitigated and unmitigated Alloy 82/182 welds, a summary of ASME Code Case 

N-770-1 inspection requirements and the NRC conditions in 50.55a, an assessment of weld 

inspection intervals, and conclusions on the effectiveness of Code Case N-770-1 requirements 

for mitigated and unmitigated welds.

Inspection limitations can alter the effectiveness of NDE. Geometry, surface conditions, or 

access may limit performing complete weld examinations. For monitoring of potential PWSCC, 

and to enable mitigated and unmitigated welds to perform as expected, the NDE applied has to 

be effective and reliable. Information on geometric limitations to performing inspections of 

typical Westinghouse and Combustion Engineering plant RCS Alloy 82/182 welds is contained 

in Appendices C and D. This information is applicable to unmitigated welds as well as pre- and 

post-MSIP, pre-weld overlay (WOL) examinations, and pre-inlay/onlay examinations.

The examination requirements of ASME Code Case N-770-1 appear to be complete and 

parallel existing Section IWB Class 1 requirements. For unmitigated welds the requirements 

reflect a philosophy of more frequent monitoring for an active degradation mechanism. An 

assessment for unmitigated welds showed times for crack growth of initiated flaws 10% to 25% 

through-wall to be shorter than Code Case N-770-1 inspection intervals. Baseline examinations 

have been completed and, although axial cracks have been found, some involving through-wall 

leakage, no instances of circumferential or axial PWSCC have been observed that exceeded 

ASME Code structural factors. It will be important to assess the results of second-round 

inspections against Code Case N-770-1 inspection intervals requirements.

An evaluation showed that WOLs provide effective mitigation against the initiation of PWSCC 

and against the growth of existing PWSCC that has been detected and allowed to remain in 

service. Based on a review of Code Case N-770-1, as implemented with NRC conditions, it is 

concluded that the inspection requirements for WOLs provide an acceptable approach for 

monitoring potential PWSCC.

Assuming that the performance criteria of Code Case N-770-1, Appendix I, are satisfied, an 

evaluation also showed MSIP provides effective mitigation against the initiation of PWSCC and 

against the growth of existing PWSCC that has been detected and allowed to remain in service. 

Based on a review of Code Case N-770-1, as implemented with NRC conditions, it is further

concluded that the requirements for MSIP provide an acceptable approach for monitoring 

potential PWSCC.

xii
There are a number of issues that need to be addressed for the application of inlays and onlays 

as described in Code Case N-766. Code Case N-766, which has not been accepted by the 

NRC, contains ASME Code rules for the design, installation, and fabrication inspections of 

inlays and onlays. NRC staff’s issues with Code Case N-766 and the revision of this Code 

Case, N-766-1, are discussed in Section 6.5 of this report. Once the issues are resolved, Code 

Case N-766 and Code Case N-770-1 would be expected to provide an acceptable approach for 

designing, installing, and examining welds mitigated by inlays/onlays and for monitoring 

potential PWSCC in inlays and onlays. Resolution of the issues with Code Case N-766 and 

Code Case N-766-1 may necessitate some changes to Code Case N-770-1.

This report does not contain an assessment of chemical mitigation. As of the date of this report, 

the industry is studying this mitigation technique and has not sought regulatory credit for 

reduced examination frequencies of Alloy 82/182 butt welds based on chemical mitigation. 

Accordingly, NRC has limited its resources on chemical mitigation of Alloy 82/182 to keeping 

abreast of industry activities and progress on this topic. An Electric Power Research Institute 

(EPRI) report on the status of industry activities in this area is contained in MRP-263 NP (EPRI 

2012b), which is available for download at www.epri.com. 

Likewise, this report does not contain an assessment of mitigation by peening, although a brief 

description of mitigation by peening is provided in Section 2.3.4.1. The industry refers to this 

mitigation method by the term surface stress improvement. PWSCC is caused by a 

combination of three factors—tensile surface stress, susceptible material, and the elevated 

temperature environment of PWR coolant. The mitigation methods implemented by industry 

have the objective of removing one or more of those three factors. The industry has focused on 

the surface stress improvement techniques of laser peening and water jet peening. These 

methods operate by imparting a pressure shock wave to the surface of an Alloy 82/182 weld, 

which introduces a compressive residual stress on the treated surface. Industry prepared 

MRP-335 (EPRI 2012d) to document its basis for the effectiveness of peening as a mitigation 

technique for PWSCC and to support a potential reduction in examination frequencies as 

currently specified in ASME Code Case N-770-1. MRP discussed its request for NRC staff 

review of this report in a letter from EPRI dated May 1, 2013 (Crooker 2013). MRP-335 was 

under NRC staff review at the time of the issuance of this report. 

xiii

ACKNOWLEDGMENTS

The authors acknowledge Pacific Northwest National Laboratory (PNNL) staff members Susan 

Crawford, Tracy Moran, and Tony Cinson (currently with the NRC); and Jeff Devers and Todd 

Blechinger from LMT, Inc., for their contributions to the NDE workshop held in May 2012. The 

authors acknowledge the contributions made by John Tsao of the NRC for his assistance and 

insights regarding operating experience with PWSCC and John Lareau of PNNL for his 

assistance and insights regarding NDE issues. The authors thank Ms. Kay Hass for her 

assistance in preparing this manuscript.

xv

ACRONYMS AND ABBREVIATIONS

AFEA advanced finite element analyses

ASME Code ASME Boiler and Pressure Vessel Code

ASME American Society of Mechanical Engineers

B&W Babcock & Wilcox

BCL Battelle Columbus Laboratory

BWR boiling water reactor

BWROG BWR Owners’ Group

BWRVIP Boiling Water Reactors Vessels and Internals Project

CAL Confirmatory Action Letter

CASS cast austenitic stainless steel

CE Combustion Engineering

CFR Code of Federal Regulations

CGR crack growth rate

CRV Code-required volume

DHR decay heat removal

DM dissimilar metal

DMW dissimilar metal weld

EPRI Electric Power Research Institute

ET eddy current testing

FSWOL full structural weld overlay

GDC General Design Criteria

HAZ heat-affected zone

ID inner/inside diameter

IGSCC intergranular stress corrosion cracking

IN Information Notice

INPP Ignalina Nuclear Power Plant

ISI inservice inspection

LBB leak before break

MRP Materials Reliability Program

MSIP mechanical stress improvement process

MWe megawatt electric

NAPS-1 North Anna Power Station, Unit 1

NDE nondestructive examination

NEI Nuclear Energy Institute

NIFG NDE Improvement Focus Group

NPS nominal pipe size

xvii
NRC U.S. Nuclear Regulatory Commission

OD outer/outside diameter

OWOL optimized weld overlay

PA phased array

PAUT phased array ultrasonic testing

PDI Performance Demonstration Initiative

PEWR partial excavation and weld repair

PNNL Pacific Northwest National Laboratory

POD probability of detection

PORV power operated relief valve

PT liquid penetrant testing

PWR pressurized water reactor

PWROG PWR Owners’ Group

PWSCC primary water stress corrosion cracking

PZR pressurizer

RCP reactor coolant pump

RCPB reactor coolant pressure boundary

RCS reactor coolant system

RES NRC Office of Nuclear Regulatory Research

RIL Research Information Letter

RMSE root-mean-square error

RPV reactor pressure vessel

RT radiographic testing

SCC stress corrosion cracking

SG steam generator

SNR signal-to-noise ratio

SRP Standard Review Plan

SS stainless steel

SSI surface stress improvement

TFC thermal fatigue crack

TI Temporary Instruction

TMI-1 Three Mile Island, Unit 1

TRL transmit-receive longitudinal 

TRS transmit-receive dual shear-wave

UT ultrasonic testing

VE visual examination

WOL weld overlay

WRS weld residual stress

xviii
1 INTRODUCTION

Operating experience has shown that Alloy 82/182/600 materials exposed to primary coolant

water (or steam) under the normal operating conditions of pressurized water reactors (PWRs) 

are susceptible to primary water stress corrosion cracking (PWSCC). These materials were 

widely used in the original construction of domestic PWR reactor coolant systems (RCS). 

Cracking typically initiates at the inside surface of these materials when exposed to reactor 

coolant and in the presence of high tensile residual stresses introduced by welding. The first 

incidence of PWSCC that was observed in a butt weld occurred at Palisades in 1993 in the 

heat-affected zone of a power operated relief valve (PORV) Alloy 600 safe end pipe segment. 

PWSCC had been identified earlier than 1993 in components such as Alloy 600 steam 

generator tubes and Alloy 82/182 pressurizer control rod drive partial penetration welds. 

PWSCC in butt welds was also observed in 2000 at V.C. Summer in a field-welded Alloy 82/182 

reactor vessel nozzle-to-RCS piping weld. Since the V.C. Summer event, there have been well 

over a dozen indications recorded in butt welds in domestic PWRs that have been attributed to 

PWSCC (Rogers 1993; NRC 2008a).

Operating experiences beginning with the V.C. Summer through-wall axial cracking event in 

2000 prompted the nuclear industry to issue inspection guidelines, as the safety consequences 

of inadequate inspections could have been significant. The absence of an effective inspection 

regime could, over time, result in unacceptable circumferential or axial cracking or the

degradation of RCS components by corrosion from leaks through these welds. These 

degradation mechanisms increase the probability of a loss-of-coolant accident. The American 

Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code)-required

inspections of Alloy 82/182 butt welds were not sufficiently frequent to ensure that ASME Codeallowable safety margins would continue to be met in the event that PWSCC initiates. The 

growth rate of PWSCC in these welds is rapid; thus, PWSCC could lead to leakage or rupture

before the degradation would be detected by the inspections required by the ASME Code in 

Section XI, which is incorporated into the U.S. Nuclear Regulatory Commission (NRC)

regulations in Title 10, Section 50.55a, Codes and Standards, of the Code of Federal 

Regulations (10 CFR 50.55a).

Prior to regulatory actions being initiated by the NRC, the commercial nuclear power industry 

began to implement strategies to manage potential or existing PWSCC at Alloy 82/182 butt 

welds in PWRs. One general strategy consists of management by a combination of mitigation 

plus inspection, where three or four mitigation techniques have been used by industry. The 

other general strategy was to manage potential PWSCC by inspection alone. Materials 

Reliability Program (MRP)-139 provided industry guidance for the volumetric and visual 

inspections of unmitigated and mitigated butt welds in PWR primary systems (EPRI 2005, 

2008a). MRP-139 augmented the inspections of these locations that were already required by 

Section XI of the ASME Code.

In December 2005, the NRC sent a letter to ASME (Dyer 2005) requesting that the Section XI 

standards body address inspection requirements for Class 1 PWR piping butt welds fabricated 

with Alloy 82/182 weld materials. ASME approved the development of an ASME Code Case on 

appropriate inspection requirements to address PWSCC in Class 1 butt welds containing 

1-1
Alloy 82/182. This case was later numbered Code Case N-770 (ASME 2009b). Code Case 

N-770 was revised in 2009 to address NRC concerns and Code Case N-770-1 was issued later 

that year (ASME 2009a). The NRC incorporated ASME Code Case N-770-1 by reference with 

conditions into 10 CFR 50.55a (76 FR 36232) in June 2011. Code Case N-770-1 contains 

requirements for inspection of unmitigated as well as mitigated Alloy 82/182 RCS butt welds.

The purpose of this NUREG is to provide (a) a description of the strategies being used to 

manage potential PWSCC in butt welds, the basic design and regulatory requirements of each 

strategy/mitigation technique, and an assessment of the issues that may impact the 

effectiveness of each management strategy; (b) a summary of operating experience with 

PWSCC in dissimilar metal welds (DMWs); (c) a summary of regulatory requirements, ASME 

Code Cases, and the regulatory status of ASME Code Cases related to management of 

PWSCC in DMWs; and (d) a historical summary and timeline of the major events, activities, and 

decision processes.

Section 2 provides a high-level discussion of PWSCC events affecting Alloy 82/182 butt welds; 

industry, ASME Code, and NRC actions taken to maintain the structural integrity; each of the 

mitigation techniques that have been used; and the status of emergent mitigation techniques.

Section 3 provides an assessment of the effectiveness of nondestructive examination (NDE)

performed on unmitigated Alloy 82/182 welds, a summary of ASME Code Case N-770-1 

inspection requirements for unmitigated welds and NRC conditions in 50.55a on unmitigated 

welds, an assessment of unmitigated weld inspection intervals, and conclusions on the 

effectiveness of Code Case N-770-1 requirements for unmitigated welds.

For welds mitigated by the mechanical stress improvement process, weld overlays, and inlays 

and onlays, Sections 4–6, respectively, provide discussions of operating experience, design 

requirements, NDE reliability, ASME Code Case N-770-1 inspection requirements and related 

NRC conditions in 50.55a(g)(6)(ii)(F). These sections also provide an assessment of weld 

inspection frequencies and conclusions on the effectiveness of Code Case N-770-1 

requirements for mitigated welds. Section 7 contains overall conclusions based on information 

provided in Sections 2 through 6. Section 8 contains a summary of NRC requirements that 

apply to Alloy 82/182 butt welds.

Appendix A contains a summary of major events, industry and regulatory activities, and decision 

processes related to management of PWSCC in Alloy 82/182 butt welds. Appendix B provides 

a summary of U.S. and foreign operating experience with PWSCC in Alloy 82/182 butt welds. 

Appendices C and D provide information on Alloy 82/182 butt weld ultrasonic testing inspection 

limitations for a typical Westinghouse and a typical Combustion Engineering plant. These 

appendices provide drawings of scan plots showing assessments of axial and circumferential 

weld inspection coverage and photographs of the welds. Lastly, Appendix E contains a

summary of the results of four crack growth studies that were assessed for insights they provide 

regarding the inspection intervals for unmitigated Alloy 82/182 welds specified in ASME Code 

Case N-770-1. 

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2 BACKGROUND

2.1 Major PWSCC Events

In 2000, a large accumulation of boric acid deposits observed during a refueling outage at 

V.C. Summer led to the discovery of cracking in the ‘A’ reactor pressure vessel (RPV)-to-hot leg

nozzle Alloy 82/182 weld (Casto 2001; Cotton 2001; Byrne 2002; NRC 2008a). The weld had a 

through-wall axial flaw with a small circumferential component and other small part-through-wall 

axial flaws. The axial crack growth of the flaw stopped or arrested when the crack reached the 

low-alloy (ferritic) steel nozzle on the one side of the weld and the stainless steel (SS) pipe on 

the other side of the weld.

The circumferential flaw grew a short distance through a portion of the nickel-based alloy butter 

on the inside of an undercut portion of the low-alloy steel nozzle. The outward radial growth of 

the circumferential flaw was arrested when it ran into the low-alloy steel nozzle. Because the 

sizes of the axial and circumferential cracks were bounded, it would not have been possible for 

these cracks to lead to a piping rupture. The safety significance of this occurrence was limited 

to leakage of primary coolant. If a circumferential flaw had initiated closer to the center of the 

weld, it could have caused a more safety-significant situation.

Ultrasonic testing (UT) examination of the remaining nozzle-to-pipe welds in all the loops found 

no flaws. Eddy current testing (ET) identified a number of small indications on the inside 

diameter (ID) of the ‘A’ hot leg nozzle weld, which were not identified by UT. ET also identified 

similar small indications on the ID of four of the other five nozzle-to-pipe welds. The ‘A’ hot leg

weld was cut out and replaced with a nickel-alloy weld more resistant to primary water stress 

corrosion cracking (PWSCC). Metallurgical analysis of the weld confirmed the through-wall leak 

was from an axial crack 64-mm (2.5-in.) in length at the ID surface with a 5-mm (3/16-in.)

opening length on the outside diameter (OD) surface of the nozzle-to-pipe weld. The analysis 

also showed a number of the small ET indications to be shallow cracks. The cracking was 

determined to be PWSCC. An integrity evaluation was performed to justify continued plant 

operations without removal of the indications in the other four welds. The destructively 

determined flaw sizes of the ET indications in the loop ‘A’ hot leg nozzle-to-pipe weld were used 

to support the crack growth calculations of the ET indications found in the other nozzle-to-pipe 

welds. On February 20, 2001, the U.S. Nuclear Regulatory Commission (NRC) issued a safety 

evaluation documenting its review of the integrity evaluation. The safety evaluation concluded 

that V.C. Summer nuclear plant could be safely operated for one fuel cycle (1.5 years) with the 

cracks discovered via ET indications in the other nozzle-to-pipe welds allowed to remain in 

service (Cotton 2001). During the subsequent outage, a mechanical stress improvement 

process (MSIP) was applied to the ‘B’ and ‘C’ hot leg nozzle welds.

The first incidence of PWSCC in butt welds that was observed in a foreign plant occurred at 

Ringhals Unit 3 in 1999 in a reactor vessel hot leg nozzle-to-safe–end weld. Two axial flaws 

were found in 1999 and were left in service to be reexamined during the next outage. These 

flaws were measured by UT to be 9 mm (0.35 in.) deep in 2000 and 13 mm (0.51 in.) and 

16 mm (0.63 in.) deep in 2001. A new flaw was found in 2001 and measured at 8 mm (0.31 in.)

deep. All three flaws were removed by excavation (boat samples were obtained for 

2-1
metallurgical analysis) and repaired by Alloy 52 filler material in 2001. At Ringhals Unit 4 in 

2000, four axial flaws were found by UT. All four flaws were also removed by excavation and 

repaired by Alloy 52M. All four flaws were confirmed to be surface-breaking and the two 

deepest flaws were determined to be 22 mm (0.87 in.) deep by destructive examination 

(Skanberg 2001; EPRI 2004a; Gérard and Daoust 2004b; Miteva and Taylor 2006; NRC 

2008a).

Between 2001 and 2006, axial and circumferential indications attributed to PWSCC were found 

in reactor coolant system (RCS) welds at six other plants. None of these indications resulted in 

through-wall leakage. Where sizing information was available, none of the indications were 

believed to be deeper than 50% through-wall. On October 13, 2006, the Wolf Creek Nuclear 

Operating Corporation performed pre-weld overlay inspections using UT techniques on the 

pressurizer surge, spray, relief, and safety nozzle-to-safe end Alloy 82/182 and safe end-to-pipe 

SS butt welds (Garrett 2006; NRC 2008a). The inspection identified three circumferential 

indications in the surge nozzle-to-safe–end Alloy 82/182 butt weld, one in the relief nozzle-tosafe–end Alloy 82/182 butt welds, and one in the safety nozzle-to-safe–end Alloy 82/182 butt 

weld. The licensee attributed the indications to PWSCC. The UT sizing techniques used were 

not qualified; however, based on the sizes estimated by the examiners, some of the flaws were 

significantly longer circumferentially than flaws previously seen in the industry. All five flaws 

were estimated to be less than approximately one-third of the wall thickness in depth. The 

licensee applied full structural weld overlays (FSWOLs) to these welds.

Initial flaw evaluation studies based on the estimated flaw sizes at Wolf Creek indicated that 

flaws of similar size in uninspected pressurizer nozzle welds could result in leakage or rupture. 

The NRC concluded that licensees needed to complete inspections or mitigations of the 

pressurizer nozzle Alloy 82/182 welds by the end of 2007 and implement interim enhanced 

leakage monitoring (Evans 2007c). Subsequently, industry and the NRC staff independently 

performed advanced finite element analyses (AFEA) that indicated the plants with uninspected 

pressurizer nozzle welds could continue to operate until their next scheduled inspections (EPRI 

2007a; Rudland et al. 2007b). Innovative analytical techniques were developed to perform the 

AFEA and large resources were required to develop these techniques and perform the 

analyses. The Wolf Creek inspection results had major implications for the operation of about 

ten pressurized water reactors (PWRs) that had yet to inspect the Alloy 82/182 pressurizer 

nozzle welds by the end of 2007. If the AFEA techniques had not been developed or if the 

results of the AFEA had not been successful, the PWRs with uninspected pressurizer nozzle 

welds would have faced mid-cycle inspections to maintain safety. Additional events related to 

PWSCC in nickel-alloy butt welds are discussed in the paragraphs that follow. 

On January 4, 2008, while Davis-Besse was in cold shutdown with the reactor head in place 

and fuel in the reactor vessel, the licensee began to install a weld overlay on the decay heat 

removal drop line-to-RCS nozzle weld (NRC 2008a). This weld is located inside containment

and the RCS was in a partially drained condition. During the first weld pass on the drop line 

weld, the welding operator noticed water seeping from the weld. Visual inspection confirmed a 

small leak, yet surface examinations before welding showed no abnormal conditions or leakage. 

The licensee subsequently determined that the leak was from an axial through-wall flaw and

attributed the flaw to PWSCC. The licensee was able to complete a FSWOL to this weld.

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The licensee for Salem Unit 1 performed examinations on all eight reactor vessel nozzle hot and 

cold leg Alloy 82/182 welds during a fall 2008 refueling outage (Braun 2009). During the 

examinations, a circumferential inside surface-connected flaw was reported on one of the hot 

leg Alloy 82/182 welds. The flaw was measured to be 24% through-wall (16.1 mm [0.634 in.]) 

and 52.3-mm (2.06-in.) long on the ID surface. All eight reactor vessel nozzle welds were 

mitigated by MSIP during this outage.

During the October 2009 refueling outage inservice inspection (ISI) examinations at Seabrook 

Station, an axial 21% through-wall flaw was reported in a reactor vessel hot leg nozzle-to-safe–

end weld (O'Keefe 2010; Freeman 2011). The flaw was attributed to PWSCC and the licensee 

performed MSIP on this weld.

On March 24, 2012, axial indications at North Anna Power Station, Unit 1, were found as a 

result of leakage in the ‘B’ reactor coolant loop hot leg-to-steam generator nozzle weld. The 

leakage was found during the performance of work activities to support weld overlay work. After 

preparatory OD surface machining, workers noted a small amount of water seeping from two

flaws in the nozzle weld area. Approximately 2.54 cm (1 in.) of weld material thickness 

360 degrees around the circumference had been removed prior to identification of the seepage. 

A UT examination of this weld had been conducted prior to the excavation, and none of the five 

flaws was detected. It has been determined that two of the cracks were greater than 80% 

through-wall and three were greater than 40% through-wall (Anderson et al. 2012). A FSWOL 

was applied to this weld. Questions were raised regarding why the flaws were missed during 

the UT examination performed prior to the pre-weld overlay machining. This topic is further 

discussed in Appendix A on major events and activities associated with PWSCC.

Alloy 82/182 welds are found in PWR Class 1 systems. Graphics showing locations of 

Alloy 82/182 welds in these systems may be found in Chapter 2 of MRP-139, Revision 1 (EPRI 

2008a).

2.2 Industry, ASME Code, and NRC Actions

Because of the incidence of cracking and leakage in Alloy 82/182 butt welds in pressure 

boundary components, the Electric Power Research Institute (EPRI) Materials Reliability 

Program (MRP) prepared both proprietary and non-proprietary versions of the report, "PWR 

Materials Reliability Project Interim Alloy 600 Safety Assessment for U.S. PWR Plants 

(MRP-44), Part 1: Alloy 82/182 Pipe Butt Welds." These reports were submitted to the NRC 

staff for information in April 2001 (EPRI 2001). The NRC staff subsequently reviewed these

reports to evaluate MRP’s assessment of the generic implications of cracking in Alloy 82/182 

pipe butt welds. The NRC concluded that results of this interim safety assessment were 

sufficient to justify the continued operation of PWRs while the industry completes the 

development of the final version of the safety assessment. The NRC stated that, “the presently 

required in-service inspection (ISI) examinations need to be augmented in the near term” and 

“should the industry not be timely in resolving inspection capabilities to identify PWSCC in Alloy 

600 welds, regulatory action may result” (Sheron 2001). 

Soon after the V.C. Summer event, the U.S. commercial nuclear power industry began

implementing strategies to manage potential or existing PWSCC at Alloy 82/182 dissimilar metal 

2-3
welds (DMWs) in PWRs. One general strategy consisted of management by a combination of 

mitigation plus inspection, with industry applying different mitigation techniques to Alloy 82/182 

piping butt welds. The other general strategy was to manage potential PWSCC by more 

frequent inspections alone. In July 2005, the industry issued inspection guidelines, MRP-139, 

for Alloy 82/182 piping welds. This report and its revision in 2008 provided industry guidance for 

the volumetric and visual inspections of unmitigated and mitigated butt welds in PWR primary 

systems (EPRI 2005, 2008a). MRP-139 augmented the inspections of welded locations that 

were already required by Section XI of the ASME Code, which is incorporated into the NRC 

regulations by Title 10, Part 50, Section 50.55a, Codes and Standards, of the Code of Federal 

Regulations (10 CFR 50.55a 2011).

After the events at V.C. Summer and Ringhals in 2000, the NRC began to evaluate the 

adequacy of existing inspection requirements for Alloy 82/182 butt welds in the ASME Code, 

Section XI. In December 2005, the NRC sent a letter to ASME (Dyer 2005) requesting the 

standards body to address inadequate inspection requirements for Class 1 PWR piping butt 

welds fabricated with Alloy 82/182 weld materials. ASME approved the development of an 

ASME Code Case on appropriate inspection requirements to address PWSCC in Class 1 butt 

welds containing Alloy 82/182. This case was later numbered Code Case N-770 (ASME 

2009b). Code Case N-770 was revised in 2009 to address NRC concerns and Code Case 

N-770-1 was issued later that year (ASME 2009a). The NRC incorporated ASME Code Case 

N-770-1 by reference into 10 CFR 50.55a(g)(6)(ii)(F) (76 FR 36232) in June 2011. Code Case 

N-770-1 contains requirements for inspection of unmitigated as well as Alloy 82/182 RCS butt 

welds mitigated by certain techniques. Paragraph 10 CFR 50.55a(g)(6)(ii)(F) required PWR 

licensees to implement the requirements of ASME Code Case N-770-1, subject to the 

conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10), by the first refueling 

outage after August 22, 2011. Paragraph 10 CFR 50.55a(g)(6)(ii)(F) also required PWR 

licensees to complete baseline examinations for unmitigated welds by the end of the next 

refueling outage after January 20, 2012. ASME has continued to develop revisions to Code 

Case N-770-1. Code Case N-770-2, approved by ASME on June 9, 2011, was developed to 

provide separate inspection items (categories) for welds mitigated by stress improvement. 

Specifically, the revision of the Code Case contains inspection items for welds mitigated by 

optimized weld overlays and welds mitigated by stress improvement without welding. Code 

Case N-770-3, approved by ASME on April 7, 2013, was developed to change the inspection 

interval for cold leg temperature welds from every second period not to exceed 7 years to once 

per interval with the time between examination not to exceed 13 years. At the time this report 

was issued, Code Case N-770-4 was being developed to include inspection intervals for welds 

mitigated by peening and 10 CFR 50.55a had incorporated only ASME Code Case N-770-1.

ASME has been working on Code Cases to provide requirements for applying mitigation 

techniques to welds susceptible to PWSCC. Code Case N-740-2 provides requirements for 

design, installation, and examination of full structural weld overlays for Class 1, 2, and 3 items in 

PWRs and boiling water reactors (BWRs). Code Case N-740-2 provides requirements for both 

repair and pre-emptive mitigation. Code Case N-754 provides requirements for design, 

installation, and examination of optimized weld overlays for Class 1 items in PWRs. The Case 

is limited to mitigation repair of as-found flaws that measure not more than 50% in depth from 

the inside surface. Code Case N-766 provides alternative rules for the design, installation, and 

2-4
examination of a weld inlay or onlay onto full penetration, nickel alloy, DMWs associated with 

Class 1 component nozzles and piping in PWRs. Inlays and onlays are installed to isolate the 

Alloy 82/182 weld material from the reactor coolant. Code Case N-766 contains requirements 

for both repair and pre-emptive mitigation. At the time this report was issued, the NRC had not 

approved Code Cases N-740-2, N-754, or N-766 in U.S. NRC Regulatory Guide 1.147 or in 

Draft Regulatory Guide DG-1231 (NRC 2013).

2.3 Mitigation Techniques

Decisions to mitigate welds susceptible to PWSCC have been driven by several factors. 

Because the majority of PWR butt weld mitigations implemented to date have been pressurizer 

(PZR) nozzle welds, and because those welds have the highest operating temperature and tend 

to present considerable inspection challenges because of tapers, contours, and materials 

issues, weld overlays have been applied extensively. Weld overlays create a new surface from 

which to inspect and result in a new examination volume that can, typically, be fully inspected. 

Mitigation by MSIP was generally not a viable option for these welds because of their design 

geometries and accessibility. These issues caused volumetric inspection limitations for the 

majority of PZR nozzle-to-head welds; thus, the required pre-MSIP examinations could not be 

performed, although MSIP was used to mitigate the PZR spray, safety/relief and surge nozzle 

welds at Calvert Cliffs 1 and 2, and the pressurizer surge outlet nozzle Alloy 600 safe end-tostainless steel surge line upper elbow at Palisades (EPRI 2006). 

Weld inlays and onlays are designed to be applied to the inside surface of DMWs to isolate the 

susceptible Alloy 82/182 material from the reactor coolant environment. Weld inlays and onlays 

are designed to be about 3-mm (1/8-in.) thick and made of Alloy 52, a material believed to be 

much less susceptible to PWSCC.

As noted previously, the industry began performing baseline examinations under MRP-139 and 

10 CFR 50.55a(g)(6)(ii)(F), with these examinations of unmitigated welds to be completed by 

the end of the next refueling outage after January 20, 2012. The baseline examination results 

for primary system hot leg and cold leg temperature weld locations have been generally 

favorable, which has decreased the urgency to mitigate these welds in the short term.

2.3.1 Weld Overlays

Weld overlay (WOL) repairs were first applied on austenitic SS piping welds in BWRs in 1982 

(EPRI 1991). WOLs were initially applied as interim measures to welds affected by 

intergranular stress corrosion cracking (IGSCC) to allow the plants to return to power while longterm repairs or piping replacements were planned. NUREG/CR-4877 (Scott 1987) provides an 

assessment of the load-carrying capacity design basis for weld-overlay repairs which concludes, 

in part, that a WOL based on a design basis flaw of 100% through the original thickness of the 

pipe/weld wall overlay would be suitable for long-term plant operation. WOLs were recognized 

by the NRC in NUREG-0313 (Hazelton and Koo 1988) as a mitigation and repair technique, 

which could be credited for reducing the frequency of inspection for welds susceptible to 

IGSCC. The Electric Power Research Institute (EPRI 1991) provides a compilation of the 

results of experimental, analytical, and operational WOL programs conducted by nuclear 

utilities, the NRC, and EPRI.

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ASME Code Case N-504 was developed to address design and preservice examination aspects 

of full structural weld overlays for BWRs. Regulatory Guide 1.147, Revision 16, issued in 

October 2010 (NRC 2010a), conditionally accepted Code Case N-504-4 (ASME 2006). 

FSWOLs and optimized weld overlays (OWOLs) were evaluated by Fredette and Scott in 2010 

as a mitigation or repair strategy for PWSCC in Alloy 82/182 welds in PWRs (Fredette and Scott 

2010). The results of this study showed that for the geometries considered, weld overlays are 

an effective method to reduce weld residual stresses at the pipe weld inside surface which can 

lead to PWSCC in DMWs in PWR piping systems.

There are two types of weld overlays for PWR Alloy 82/182 welds susceptible to PWSCC—

FSWOLs, which have a minimum thickness of one-third the pipe wall thickness, and OWOLs, 

which have much less weld metal, in some cases having as few as three layers of applied 

weldment. Weld overlays extend in both axial directions some distance beyond the weld. Weld 

overlays extend 360 degrees circumferentially around the pipe. For FSWOLs, no structural 

credit is taken for the original pipe thickness. For OWOLs, structural credit is taken for the outer 

25% wall thickness of the DMW. These weld overlays are constructed of Alloy 52/152-type 

weld materials, which during testing at Pacific Northwest National Laboratory (PNNL) did not 

show any significant stress corrosion cracking (SCC) in PWR primary water (Toloczko and 

Bruemmer 2009).

The owner for Davis-Besse elected to apply Alloy 52 FSWOLs to the core flood nozzle-to-safe 

end welds, Alloy 52 OWOLs to the reactor coolant pump (RCP) discharge nozzle welds, and 

FSWOLs to the RCP inlet nozzle welds (White 2010).

In February 2011, the owner of Calvert Cliffs, Units 1 and 2, received NRC approval to apply 

Alloy 52 FSWOLs on a contingency basis in case any unacceptable indications of cracking are 

found during the fourth ISI interval in the reactor coolant pump inlet and outlet nozzles, PZR 

surge hot leg nozzles, safety injection cold leg nozzles, PZR spray cold leg nozzles, PZR relief 

nozzles, cold leg letdown drain nozzles, cold leg charging inlet nozzles, and cold leg loop drain 

nozzles (Salgado 2011). Since 2006, no indications of unacceptable cracking attributed to 

PWSCC have been found at Calvert Cliffs 1 or 2. Accordingly, no FSWOLs have been applied 

in these locations at these units under the licensee’s alternative approved by the NRC in 2011.

The owner of North Anna Power Station, Unit 1 (NAPS-1) requested NRC authorization in 

March 2011 to mitigate PWSCC susceptibility of steam generator (SG) hot leg nozzle DMWs by 

installing an Alloy 52M FSWOL on each of these welds during the NAPS-1 spring 2012 refueling 

outage (Price 2011). The NRC authorization for this licensee proposal was issued on 

January 27, 2012 (Salgado 2012). Axial flaws in a NAPS-1 hot leg SG nozzle were not 

detected during a manual pre-WOL inspection performed in March 2012. However, leaks 

through two flaws were found when the nozzle weld OD region was machined in preparation to 

install weld overlays (Anderson et al. 2012).

These mitigation approaches provide an alternative to relying on inspection alone as a means to 

manage potential PWSCC at these locations. In PWRs, the weld overlay process consists of 

applying an annulus of Alloy 52-type weld material on the outside of a pipe over the susceptible 

Alloy 82/182 DMW and the materials on either side of the DMW. Weld overlays are expected to 

2-6
continue to be relied upon in the mitigation of Alloy 82/182 welds in PWRs. Details on design 

and evaluation of WOLs are provided in Chapter 5 of this report.

2.3.2 Mechanical Stress Improvement Process

MSIP was first used in 1986. The NRC Office of Nuclear Regulatory Research issued Research 

Information Letter (RIL) Number 149, “Evaluation of the Mechanical Stress Improvement 

Process,” on February 12, 1987 (NRC 1987). RIL No. 149 was based on a review by Argonne 

National Laboratory (ANL) of information on MSIP submitted to the staff by O’Donnell &

Associates, Inc. and Westinghouse Electric Corporation, the developers of the process. Also, 

ANL performed analyses and tests to determine the changes in weld residual stress using two 

large-diameter pipe sections supplied by Vermont Yankee that were modified by MSIP. The RIL 

states that, “Based on the results of our research work and the data and analyses provided by 

O’Donnell & Associates, MSIP is judged to be an effective means of improving the residual 

stress state of piping system weldments. It is as effective for large diameter piping as small 

diameter piping.” MSIP was approved by the NRC for BWR applications in 1988 with the 

issuance of USNRC Generic Letter 88-01 (NRC 1988) and the guidelines contained in 

NUREG-0313, Revision 2 (Hazelton and Koo 1988).

Mitigation provides an alternative to relying on inspection alone as a means to manage potential 

PWSCC at these locations. MSIP modifies the existing ID residual tensile stresses in the weld 

metal and heat-affected zone of butt welds in piping. A load is applied to the outside surface of 

the pipe with a large two-piece mechanical hydraulic clamping device connected by two pairs of 

tangentially positioned studs. The clamping device is placed a short distance away from the 

weld being mitigated. Stainless steel, “waffled” contact plates are placed between the tool and 

the pipe outer surface to aid in uniformly distributing the load. The studs are tightened with 

hydraulic tensioners. Displacement is controlled to pre-assigned limits by using shims,

providing safety from over-straining the pipe. Loading plastically strains the pipe causing the 

pipe diameter to decrease in the region under the clamps. In the nearby weld and counterbore 

region, the plastic strain caused by the clamps generates compressive residual stress around 

the weld ID and counterbore region. Weld residual stresses (WRS) after application of MSIP 

are most compressive in the weld on the pipe inside surface and the WRSs increase and 

become tensile toward the outside surface. Likewise, WRSs at the inside surface increase at 

increasing distances away from the weld fusion line (Smith et al. 1987).

The generation of residual compressive stresses has been verified and confirmed by 

independent tests and analyses. These tests include residual stress measurements on BWR 

mockups and welds taken from discontinued plants. Weld residual stress analyses and crack 

growth analyses have been performed as well as sensitivity studies on process parameters, 

such as the location and axial length of the clamping device and the amount of plastic strain 

imparted. The following references discuss the results of these studies: Smith et al. (1987),

Phillips et al. (1993), Findlan et al. (2004), and Fredette and Scott (2009).

The current holders on the patent for MSIP, NuVision Engineering, have indicated that MSIP 

has over 25 years of successful operating experience in the United States with more than 

1000 welds treated (NuVision 2011). 

2-7
The largest diameter Alloy 82/182 welds in the primary loop of the RCS are the reactor vessel 

nozzle-to-pipe welds in Westinghouse plants and the RCP nozzle-to-pipe welds in Combustion 

Engineering and Babcock and Wilcox plants. MSIP has been implemented on reactor vessel 

nozzle DMWs at V.C. Summer, Salem Unit 1, and Seabrook. MSIP was used at these plants to 

stabilize the reactor vessel nozzle welds with indications of cracking. Except at Seabrook, MSIP 

was used to preemptively mitigate other reactor vessel nozzle welds that did not have 

indications of cracking (Dodson 2008; Burritt 2009; O'Keefe 2010). 

In 2010, the owners for D.C. Cook, Unit 1, elected to perform UT and ET of the RPV hot leg

nozzle welds pre-MSIP, and the cold leg nozzle welds post-MSIP (Cameron 2010). For the hot 

leg nozzle welds, the licensee performed pre- and post-MSIP Section XI, Appendix VIII, 

qualified UT examinations, as modified by an NRC-approved ASME Code relief request (Kobetz 

2006). For the cold leg nozzle welds, the licensee performed only a post-MSIP Section XI, 

Appendix VIII, qualified UT, as modified by an NRC-approved ASME Code relief request, and 

an ET examination from the inside surface of these piping welds. The licensee noted that 

ASME Code Case N-770 has provisions to accept post-MSIP UT examinations for these 

particular nozzles, if supplemented by ET. No indications were found as a result of the 

inspections of these welds.

MSIP has been applied at Watts Bar, Unit 2, to the six pressurizer nozzles, the four reactor 

vessel hot leg nozzles, and the four reactor vessel cold leg nozzles. This plant is undergoing 

licensing review (Watts Bar, Unit 2, Final Safety Evaluation Report, Section 5.5.3.3.1).

MSIP has been reported to stabilize existing cracks, or even close them sufficiently, such that 

they are no longer detected with ultrasonic NDE techniques (Findlan et al. 2004; Crawford et al. 

2011). Details on design and evaluation of MSIP are provided in Section 4 of this report.

2.3.3 Inlays/Onlays

Weld inlays and weld onlays are methods to mitigate potential PWSCC by applying Alloy 52/152

material to the inside pipe wall over an existing Alloy 82/182 weld. Alloy 52/152 material is 

considered more resistant to PWSCC than Alloy 82/182. Weld inlays and onlays are applied to 

isolate the existing Alloy 82/182 material from the reactor coolant. 

In larger bore piping (e.g., RPV hot leg outlet nozzle), mitigation by a method such as FSWOL is

very costly because of the amount of weld metal that must be deposited and outage time 

required to perform this welding. Additionally, OD access space may limit the ability to use 

MSIP in certain plants. To provide an alternative mitigation approach, industry developed inlay 

and onlay methods. The inlay process consists of excavating a small portion of the susceptible 

material on the ID surface of the pipe weld and applying Alloy 52/152 material in its place to 

form a more corrosion-resistant barrier between the Alloy 82/182 weld and butter materials, and 

reactor coolant. With the onlay process, no excavation of material is performed prior to applying 

the barrier of Alloy 52/152 material. During these methods, only a small layer of 2 to 3 weld 

beads would be deposited on the inside surface of the pipe. Industry proposed a weld inlay and 

onlay thickness of approximately 3 mm (1/8 in.). The ASME Code developed Code Case 

N-766, “Nickel Alloy Reactor Coolant Inlay and Onlay for Repair or Mitigation of PWR Full 

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Penetration Circumferential Nickel Alloy Welds in Class 1 Items,” to document the process for 

applying weld inlays and onlays. 

Based on industry presentations to NRC staff (Tsao 2008c, d), the first layer of weld metal 

would be a type 308L or 309L SS buffer on the existing SS safe end on one side of the 

susceptible weld and on the SS cladding on the nozzle adjacent to the butter. These buffer 

layers are to prevent hot cracking, which is possible if Alloy 52/152 material is deposited directly 

onto the existing SS. These SS buffer layers would stop short of application to the existing 

Alloy 82/182. A buffer layer of Alloy 82 may be applied directly to the existing Alloy 82/182 in 

between the SS buffer layers. The Alloy 82 and SS buffer layers would be blended to form a 

smooth first layer. Two weld metal layers of Alloy 52/152 material would be applied 

continuously over the SS and Alloy 82 first layer. Subsequent layer(s) of 52/152 material may 

be applied to provide material for machining and adequate thickness after surfacing, if needed. 

After welding, the surface of the weld would be machined to a relatively smooth condition. The 

ends of the welds would be blended into the surrounding areas.

Industry also indicated that the addition of Alloy 52/152 weld material to the inside surface of 

each nozzle/safe end junction would have no adverse effect on the flow through these nozzles. 

The inside diameter of the completed onlay would be larger than the supply pipe inside 

diameter; therefore, the onlay should not be a limiting feature of the flow path. Industry 

indicated that the structural qualification of the joint would not be adversely affected. No credit 

is taken for the strength of the onlay, which is considered as added cladding (Tsao 2008c).

The design of inlays is expected to be similar to onlays except that the existing weld and 

portions of the safe-end and nozzle cladding are excavated to accommodate the thickness of 

the weld inlay (Tsao 2008c). As part of a confirmatory analysis, the NRC staff and its contractor, 

Engineering Mechanics Corporation of Columbus, conducted both welding residual stress and 

flaw evaluation analyses to determine the effectiveness of inlay welds as a mitigative technique

(Rudland et al. 2010).

In 2010, the ASME Code, Section XI, approved Code Case N-766, “Nickel Alloy Reactor 

Coolant Inlay and Onlay for Mitigation of PWR Full Penetration Circumferential Nickel Alloy 

Dissimilar Metal Welds of Class 1 Items” (ASME 2010). The ASME prepared Code Case 

N-766-1 in response to NRC staff comments on Code Case N-766. ASME approved Code 

Case N-766-1 on April 7, 2013. At the time this report was issued, the NRC had not approved 

Code Case N-766 or Code Case N-766-1 in U.S. NRC Regulatory Guide 1.147 or in DG-1231 

(NRC 2013). NRC staff issues with these code cases are discussed in Section 6.5.

Weld onlays were applied to the core flood nozzles to reactor vessel DMWs at Arkansas 

Nuclear One Unit 1 during the fall 2008 outage (Clark 2009) and the ‘A’ core flood nozzle weld 

at Three Mile Island Nuclear Station, Unit 1 (TMI-1) in 2009 (Bellamy 2010). V.C. Summer has

SG inlet and outlet DMWs with Alloy 152/152 inlays from its SG replacement (Gatlin 2012a, b, 

c). A discussion on the V.C. Summer SG inlays is contained in Section 6.1 of this report.

Details on design and evaluation of inlays and onlays are provided in Section 6 of this report.

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2.3.4 Emergent Mitigation Techniques

2.3.4.1 Peening

In the United States, the techniques used to mitigate butt welds against the initiation or 

propagation of PWSCC have been weld overlays, MSIP and, to a lesser extent, inlays and 

onlays. Due to obstructions and limited space, weld overlays and MSIP may not be practical for 

some locations. Surface stress improvement (SSI) (i.e., peening) is another mitigation method 

that has been extensively used in Japanese PWRs and BWRs and is being considered as an 

additional or alternative PWSCC mitigation technique in the United States for Alloy 82/182 butt 

welds and other Alloy 82/182 locations. The SSI treatment is applied to the entire area of the 

wetted surface that is susceptible to PWSCC. This technique is reported to mitigate PWSCC by 

reversing the tensile stress at the surface exposed to reactor coolant to compressive residual 

stress.

Peening is considered by EPRI to be cost-effective and lower in dose than mitigations involving 

welding. EPRI has been working in recent years to establish the technical basis for applying 

SSI treatments as an alternative PWSCC mitigation method to protect key PWR plant assets. 

The program underway is intended to establish the effectiveness of SSI treatments and to

identify Alloy 600/82/182 locations for which relaxation of inspection requirements is appropriate 

after applying SSI mitigation treatments. EPRI is investigating laser peening and water jet 

peening (WJP). Japanese vendors are working with EPRI on this effort, including Toshiba on 

laser peening and Mitsubishi and Hitachi-GE on water jet peening. EPRI reported that in 

mockups of reactor vessel bottom-mounted nozzles (BMN) constructed of Alloy 600/82/182 

where the BMN ID, OD, and J-groove weld were peened with laser peening or WJP, the 

residual stress was compressive to a depth of at least 1.0–1.5 mm in most cases and for WJP 

of the inner surface of BMNs, the residual stress was compressive to a depth of about 0.5 mm.

Representatives of Toshiba, Mitsubishi, and Hitachi-GE made presentations to the ASME Task 

Group on High Strength Nickel Alloy Issues at several meetings in 2011 and 2012. These 

presentations can be found by ASME volunteers on the ASME Codes and Standards website 

but are not publicly available. Similar but less detailed information on peening is available in 

Crooker and Sims (2011) and White et al. (2012e). The ASME Code, Section XI, is developing 

proposed inspection requirements for peened welds, which would be a relaxation compared to

current inspection requirements for unmitigated welds in Code Case N-770-1. EPRI prepared

two reports on peening of DMW butt welds: Materials Reliability Program: Technical Basis for 

Primary Water Stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-267, 

Revision 1) (EPRI 2012c) and Materials Reliability Program: Topical Report for Primary Water 

stress Corrosion Cracking Mitigation by Surface Stress Improvement (MRP-335) (EPRI 2012d). 

EPRI made these reports available to download from its website at www.epri.com. These

reports were developed, in part, to support regulatory review of peening techniques and to 

support development of ASME Code rules on peening. MRP discussed its request for NRC 

staff review of this report in a letter from EPRI dated May 1, 2013 (Crooker 2013). MRP-335 

was under NRC staff review at the time of the issuance of this report.

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2.3.4.2 Partial Excavation and Weld Repair

An alternative mitigation and emergent repair strategy is being developed called partial 

excavation and weld repair (PEWR). The industry has discussed this strategy with the NRC 

staff, including during a meeting held on April 29, 2010 (EPRI 2010). The PEWR repair 

technique involves removal of a significant portion (approximately 50%) of the wall thickness of 

a DMW from the OD, and replacing the removed material with Alloy 52, which is less 

susceptible to PWSCC than Alloy 82/182. PEWR has been proposed as a technique that can 

be performed preemptively as a PWSCC mitigation technique or as a repair technique upon 

discovery of a PWSCC flaw. In addition, it has been proposed that PEWR can be applied to a 

partial arc of the weld or to the full weld circumference.

The PEWR concept was proposed as being similar to FSWOLs in that all design loads would be 

carried by the newly applied Alloy 52M weld material for full circumferential 360-degree repair,

but without the compressive residual stress benefit at the weld ID.

The three options that have been proposed are preemptive mitigation with a full 360-degree

PEWR, full 360-degree PEWR repair in the presence of PWSCC, and partial arc repair only in 

the area of an axial or limited circumferential flaw. The partial arc repair has been characterized 

as possibly being used as a temporary repair.

ASME Code Section XI is working to establish rules for partial excavation and weld repair in the 

form of an ASME Code Case. This work is being coordinated by the ASME Section XI, Task 

Group on Partial Excavation and Weld Repair of Dissimilar Welds. The Code Case will address 

fatigue and PWSCC crack growth from flaws detected in the weld, or flaws assumed for 

analysis purposes if no flaws are actually detected prior to installation of PEWR. The Code 

Case will also address welding considerations. The examination requirements and acceptance 

criteria for PEWRs are planned to be incorporated in a future revision of Code Case N-770-1. 

There are potential regulatory issues with PEWR because NRC approval would be necessary 

and approval for repairs requested on short notice, that is, during a refueling outage, may 

involve certain risks for the licensee.

2-11

3 MANAGEMENT OF PWSCC IN UNMITIGATED WELDS 

BY INSPECTION

3.1 NDE Reliability Issues in Unmitigated Welds

Nondestructive examination (NDE) reliability can be expressed in terms of the degree to which 

an examination system consistently achieves its purpose of detecting and characterizing 

targeted flaws. For this section on NDE reliability, the volumetric examination method being 

assessed is ultrasonic testing (UT), notwithstanding that there are a number of different qualified 

UT techniques, such as:

• Conventional UT – utilizes probes that insonify the material at a single angle to examine the 

intended examination volume for each direction scanned.

• Phased Array UT – utilizes search units that insonify the intended examination volume with 

multiple examination angles.

• Both conventional UT and Phased Array UT 

– May have data acquired and reviewed by the analyst in real time, or

– May have data acquired and encoded for subsequent review.

○ Encoded data may be gathered with automated or manually manipulated systems.

In assessing reliability, Pacific Northwest National Laboratory (PNNL) considered various 

factors that may influence UT performance. This section provides a discussion on UT issues 

related to detection and characterization of stress corrosion cracking (SCC) in unmitigated 

welds. Section 3.2 summarizes information on limitations to performing inspections to ASME 

Code, Section XI requirements. Section 3.3 provides an assessment of uncertainty for 

Appendix VIII piping examinations conducted from the inside surface. Section 3.4 provides an 

assessment of NDE required by ASME Code Case N-770-1. The conclusions reached on NDE 

reliability are presented in Section 3.5. Probability of detection (POD) is an accepted analytical 

method for quantifying NDE reliability and is often represented as a relationship between 

detection probability as a function of flaw size, for a given flaw type and examination technique. 

POD curves may also provide information about the threshold of detection of flaws and can

often be used to assess false call probabilities. However, this NUREG/CR does not include an 

assessment on POD of flaws in nickel alloy butt welds. In 2009, the Materials Reliability 

Program (MRP) issued MRP-262, Development of Probability of Detection Curves for Ultrasonic 

Examination of Dissimilar Metal Welds. The MRP report discusses the development of POD 

curves based on testing of UT examination candidates under the Performance Demonstration 

Initiative (PDI) managed by the Electric Power Research Institute (EPRI) and presents results

based on these demonstration tests. Based on a review of the MRP report, PNNL identified a 

number of issues with the resulting POD curves. These concerns were provided to EPRI;

however, as of the date of this report EPRI has not responded to these concerns. Additional 

work will need to be performed on POD by the nuclear industry, and possibly also by the U.S. 

Nuclear Regulatory Commission (NRC), to be able to include POD in an assessment of NDE 

reliability for nickel alloy welds.

3-1
3.1.1 Introduction

On May 8 and 9, 2012, PNNL held a workshop in Richland, Washington, with Lambert, MacGill, 

Thomas, Inc. (LMT), an NDE vendor for the commercial nuclear power industry. LMT uses UT 

examination methods qualified under ASME Code, Section XI, Appendix VIII. Both phased 

array (PA) and conventional UT techniques in encoded and non-encoded delivery modes are 

employed, as specified by the utility. LMT performs examinations of piping welds from the 

outer-diameter (OD) surface and has been involved with the inspection of a number of welds in 

pressurized water reactors (PWRs) and boiling water reactors (BWRs) in which indications were 

attributed to SCC.

The topics that were discussed during this workshop included:

• Characterization of flaws attributed to SCC,

• Encoded versus non-encoded examinations, and

• Characterization of embedded flaws in field welds.

3.1.2 Discussion

Encoded phased-array UT (PAUT) techniques were used to detect and characterize SCC found 

in a BWR control rod drive (N9) nozzle dissimilar metal weld (DMW) during a spring 2008 

refueling outage and in a PWR pressurizer (PZR) surge line DMW during a spring 2007 

refueling outage. In both cases, the stress corrosion cracking was subsequently mitigated with 

NRC-approved full structural weld overlays (FSWOLs). Additionally, encoded UT examinations 

were performed in 2008 on safety and relief nozzle welds on a PZR removed from St. Lucie, 

Unit 1, to assess results obtained previously from manual non-encoded PA examinations.

In the case of the BWR N9 weld, the full ASME Code-required inspection volume was not 

achieved with encoded PAUT techniques because of an interfering weld crown, geometry 

limitations from the nozzle contour, and the presence of an end cap. Only axial scanning for 

circumferentially-oriented flaws was conducted using PA, as the component was being prepared 

for a FSWOL procedure and full ASME Code UT volume coverage was not needed or required.

With minor geometric reflections in this weld, a circumferential flaw was readily detected and 

was localized at the weld-to-butter interface, determined to be inner-diameter (ID)-connected, 

and was attributed to intergranular stress corrosion cracking (IGSCC). The overall thickness of 

the weld was approximately 19 mm (0.75 in.) and the crack was depth-sized at approximately 

60% through-wall. A general observation made by LMT analysts during the workshop was that 

this circumferential flaw was relatively easy to detect and characterize on this somewhat thinwalled component. They noted that inspection of heavier-walled components can present 

additional challenges.

Regarding the surge line flaw, full volumetric inspection of the flawed region was precluded by 

the specimen’s taper and by the presence of lugs near the nozzle weld. During the initial 

conventional manual (non-encoded) examination, an axial flaw was detected. A decision was 

made to remove the lugs and use encoded PAUT to fully characterize the axial flaw as well as 

to conduct a more complete inspection of the Code-required inspection volume. Using encoded 

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PAUT, an ID-connected circumferential flaw was observed that was not initially detected by the 

manual conventional UT examination. The weld thickness is approximately 38 mm (1.50 in.) at 

the flaw location, and the flaw was depth-sized at 34% through-wall. It is unclear whether 

interference by the lugs prevented detection of the circumferential flaw during the initial manual 

examination. However, general observations made by the LMT analysts were that use of the 

encoded PAUT technique facilitated detection and characterization of this flaw, which would not 

have been straightforward in this weld using manual non-encoded techniques, even with the 

lugs removed. This observation is based on the presence of spurious reflections caused by a 

short safe end, numerous geometrical reflections, and various mode-conversion signals. The 

use of encoded techniques provides a more comprehensive assessment of the examination 

volume and enables analysts to use imaging techniques to sufficiently distinguish 

circumferential flaws from other spurious reflections and characterize them properly.

In the case of the St. Lucie, Unit 1, PZR safety and relief nozzle welds, the conclusions of the 

initial manual (non-encoded) PA examinations were that these welds contained deep 

non-uniform 360-degree ID-connected circumferential flaws indicative of SCC. This result 

raised questions about the applicability of the advanced finite element analyses (AFEA) safety 

assessment performed after circumferential flaws were detected in 2006 in Wolf Creek PZR

nozzles (EPRI 2007a). The Wolf Creek flaws were attributed to primary waster stress corrosion 

cracking (PWSCC). A number of operating PWRs had deferred the PZR nozzle weld 

inspections to their next refueling outage based on the AFEA assessment. Because of the 

significance of the initial St. Lucie, Unit 1, inspection result to these operating PWRs, initial 

manual PA examinations were reevaluated and the welds were reexamined shortly thereafter 

with encoded techniques. A PNNL NDE expert closely observed the reexamination activity. 

Further scanning and data analysis indicated that the welds were not cracked, but rather 

contained embedded fabrication flaws, attributed to slag, porosity, and/or lack of fusion, and 

machining geometry. The 360-degree indications at the ID were signals from fabrication-related

flaws. There was also an indication from a machining flaw due to a fabrication boring operation 

performed from both the vessel ID and the safe end. The boring operations were not aligned 

properly and resulted in an eccentric step. The indication from the eccentric step could be 

confused with a root signal, but it was about an inch away from the actual weld root. This 

eccentric step was confirmed by eddy current testing (ET) during the reexamination process. 

The presence of this geometrical discontinuity was demonstrated to the PNNL NDE expert who 

observed the reexamination activity.

Despite the abundance of fabrication-related flaws, a reevaluation of the fabrication flaws in 

accordance with ASME Code Section III criteria confirmed that the welds were acceptable. 

During the reevaluation, numerous problems were found with the initial examinations. These 

problems included use of probes that were inappropriately focused for the welds being 

examined, improper execution of PDI-qualified procedures, and the inability of the examiners to 

distinguish the signals from the fabrication flaws from potential PWSCC flaws with the manual 

techniques applied.

The workshop participants noted the disadvantage of using manual PA techniques in this 

situation because of the inability to review the UT data and carefully determine the origin of the 

signals present; this problem is inherent in all non-encoded UT, including both conventional and 

PA techniques. It was also noted that opportunities in the PDI qualification process for 

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candidates to be required to interpret reflections from false call signals such as those weld 

roots, counterbores, and welding fabrication flaws, may not be sufficiently realistic. Thus, 

examiners may pass the PDI qualification process with manual (both conventional and PA) nonencoded techniques without adequately demonstrating an ability to distinguish ID-connected 

flaws from other signals that are expected to be present in a weld inspection volume, especially 

in DMWs.

Workshop participants indicated that there are numerous factors that affect the reliability of UT

to detect and characterize flaws. These factors include the ability of the techniques used to 

obtain full ASME Code-required coverage of the inspection volume, access limitations posed by 

component geometry or surface conditions, the presence of interfering signals from geometric 

or metallurgical reflectors, the use of manual versus encoded techniques, the quality of the UT 

data, the reflectivity of the flaw, the impedance(1) of the materials being examined, the signal 

attenuation and dispersion(2) through those materials, and the experience and skill of the 

examiner. Other related factors may include the use of proper modeling methods to design the 

examination and the representativeness of the mockups used for performance demonstration to 

qualify UT examiners. The St. Lucie, Unit 1 experience and the recent North Anna Power 

Station, Unit 1 (NAPS-1) experience in 2012 (discussed in Section 2.1 above), reinforce the 

importance of carefully considering these factors before conducting an examination.

In this NUREG/CR, statements are made regarding the significance of effective and reliable

ultrasonic examinations, which in the authors’ opinions, involve techniques having optimized 

sound field intensities, with a sufficient number of beam angles to produce adequate volumetric 

coverage within the areas of interest (especially in coarse-grained austenitic materials) and 

applying proper contouring of wedges for full coupling across the entire transmitting surfaces of 

the probes. In addition, the applied sound fields should cover the optimum range of 

impingement angles for flaw detection using appropriate interrogation frequencies for the 

materials within the examination volume. Finally, for a general population of nondestructive 

examiners, the most effective and reliable UT methods are those using independent review of 

encoded data, thus limiting the use of manual non-encoded techniques to low risk examinations.

3.2 Inspection Limitations

Information on limitations to performing inspections of a typical Westinghouse and a typical 

Combustion Engineering (CE) plant reactor coolant system (RCS) Alloy 82/182 welds was also 

obtained from LMT during the workshop PNNL held in May 2012. Appendices C and D contain 

weld coverage assessment diagrams for these plants showing the probe angles used and 

percent coverage obtained with Section XI, Appendix VIII, qualified techniques. These 

appendices also indicate the materials of construction for the welds depicted. Qualified UT 

(1) Impedance of a material is inversely related to the amount of acoustic energy that is 

transmitted through it and reflected at the boundary with another material having a different 

acoustic impedance value.

(2) Dispersion is the phenomenon of a sound wave separating into its component frequencies

as it passes through a material. In the presence of dispersion, sound velocity is no longer 

uniquely defined, resulting in a decrease in signal amplitude. 

3-4

 
examinations are required prior to mitigation by the mechanical stress improvement process 

(MSIP) and optimized weld overlay (OWOL), as noted in Sections 4 and 5, respectively. 

For the plant-specific designs used to prepare Appendices C and D, Tables 3.1 and 3.2 provide 

the Code-required volume coverage achievable for Alloy 82/182 RCS butt welds in the typical 

Westinghouse and CE plants used for this assessment. The Westinghouse reactor pressure 

vessel (RPV) nozzle welds in this study were examined from the OD, although the RPV nozzle 

welds in many Westinghouse plants are examined from the inside surface, where inspection 

limitations differ and greater Code-required inspection coverage is generally understood to be 

obtained for axial scans at hot leg locations.

Table 3.1 RCS Alloy 82/182 Butt Weld Code Required Volume Coverage Obtained for a 

Typical Westinghouse Plant

Description Axial Scan, % Circumferential Scan, %

PZR surge nozzle 80 100

PZR surge nozzle 40 34

PZR safety and relief nozzle 0 0

RPV hot leg nozzle 28.8 100

SG hot leg nozzle 95.4 100

SG cold leg nozzle 95.4 100

RPV cold leg nozzle 100 100

Table 3.2 RCS Alloy 82/182 Butt Weld Code Required Volume Coverage Obtained for a 

Typical Combustion Engineering Plant

Description Axial Scan, % Circumferential Scan, %

RPC suction to RC pipe 100 44.3

RCP discharge to RC pipe 77 34

RC(a) loop surge nozzle 82.4 31.3

Letdown and drain nozzle 100 100

Hot leg drain nozzle 100 100

Charging inlet nozzle 100 100

Safety injection nozzle 94.3 26.6

Shutdown cooling nozzle 98 100

Spray nozzle 100 100

PZR surge nozzle 100 100

PZR spray nozzle 100 68.9

PZR safety nozzle 100 100

(a) RC – reactor coolant

In CE plants, the safe ends of medium- and large-bore piping are typically made of cast 

austenitic stainless steel (CASS). The Code-required volume (CRV) coverage presented in 

Table 3.2 takes credit for the scan coverage in the CASS safe ends using the Supplement 10 

techniques used for the carbon steel and Alloy 82/182 volumes. This approach for reporting 

CRV coverage was used because Code Case N-770-1 requires the examination volume to be 

examined by Appendix VIII procedures to the maximum extent practical, including 100% of the 

susceptible material volume.

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Inspection limitations exist for welds adjacent to CASS products. Currently, there are no 

performance demonstration qualification requirements for inspection of cast piping welds, as 

Appendix VIII, Supplement 9 has yet to be developed. Some NRC regulatory guidance on 

performing examinations of welds adjacent to CASS products is contained in a summary of the 

July 12, 2011, public meeting with industry (Collins 2011), held by the NRC to discuss 

implementation of Code Case N-770-1. Additional regulatory guidance may be found in a

summary of the NRC public meeting with the PWR Owners’ Group on January 10, 2012 

(Rowley 2012). This meeting was held to discuss an industry template for a generic relief 

request from the requirements of Code Case N-770-1, as conditioned by the NRC. The NRC 

staff noted in Collins (2011) that under 50.55a(g)(6)(ii)(F)(3) a licensee may use an Appendix III 

procedure under the requirements of Appendix VIII to meet the ASME Code-required 

examination volume of essentially 100% to credit previous examinations as baseline 

inspections. However, all other required inspections under 10 CFR 50.55a(g)(6)(ii)(F) for 

inspection of DMWs joined to cast stainless steel items require the licensee to meet the 

inspection coverage requirements of -2500(b) of ASME Code Case N-770-1. Subparagraph 

−2500(b) states that, “For cast stainless steel items for which no supplement is available in 

Appendix VIII, the required examination volume shall be examined by Appendix VIII procedures 

to the maximum extent practical including 100% of the susceptible material volume (nonstainless steel volume).” This wording requires that Appendix VIII procedures, not Appendix III 

procedures, be used for the examination of the required examination weld volume, including 

100% of the susceptible material volume. 

There appears to be some uncertainty regarding the requirement to examine “100% of the 

susceptible material volume.” Figure 1 in CC-770-1 shows the required examination volume for 

an unmitigated weld. This figure shows a typical weld with no “wrap around.” The term “wrap 

around” refers to Alloy 82/182 cladding that may extend from the end of the stainless steel (SS)

cladding on the inside wall of a ferritic nozzle to the nozzle weld butter. Because Figure 1 in 

CC-770-1 shows a typical weld with no wrap around, the inspection volume may have been 

interpreted to exclude this area per the figure, but the words in the Code Case say 100% of the 

susceptible material has to be examined. 

ASME, Section XI, prepared Code Case N-824, entitled, “Alternate Requirements for Ultrasonic 

Examination of Cast Austenitic Piping Welds from the Outer Diameter Surface.” The basis for 

Code Case N-824 is research performed by PNNL for the NRC Office of Nuclear Regulatory 

Research (RES). The requirements in this Code Case are deterministic rather than 

performance-based. However, this Code Case is considered to be a first step in the 

development of Appendix VIII, Supplement 9. NDE reliability would be expected to be 

substantially improved by applying this Code Case, with certain conditions based on NRC 

research, in situations where 100% of the susceptible material of the DMW cannot be examined 

from the ferritic side of a weld. The NRC staff supported the development of ASME Code Case 

N-824, however, at the time this report was issued, the NRC had not yet taken regulatory action 

to approve or impose ASME Code Case N-824.

Research conducted by RES on inspection of CASS demonstrated that current ASME 

Section XI, Appendix III, UT techniques are inadequate when applied to cast materials. This

research has shown that through-wall cracks of 10 to 50 percent depth in PZR surge line CASS 

components 5.1 cm (2.0 in.) or less in wall thickness could be readily detected when using 

3-6
phased array techniques at examination frequencies up to 1.0 MHz. Much lower frequency

ranges are needed for primary system CASS piping [greater than 5.1 cm (2.0 in.) in thickness]. 

In thick-walled CASS specimens, that is, specimens between 6.6 and 8.1 cm (2.6 and 3.2 in.), 

using phased array techniques at an inspection frequency of 500 kHz, examinations have been 

shown to reliably detect flaws in an approximately 30% and greater through-wall depth range. 

Additional information is contained in NUREG/CR-6933, Assessment of Crack Detection in 

Heavy-Walled Cast Stainless Steel Piping Welds Using Low-Frequency Ultrasonic Methods

(Anderson et al. 2007). NUREG/CR-7122, An Evaluation of Ultrasonic Phased Array Testing for 

Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds (Diaz et al. 2011), provides 

information on crack detection with UT in thinner-walled CASS. The information in these 

NUREGs should be helpful to industry in improving the reliability of inspections and to the NRC 

Office of Nuclear Regulatory Research in evaluating relief requests and inspection-related 

issues involving CASS.

3.3 Uncertainty for Appendix VIII Piping Examinations Conducted 

from the Inside Surface

The PDI specimens for large-bore Appendix VIII testing have representative geometries with 

challenging ID surfaces. These surfaces can provide poor contact for ultrasonic probes and the 

ID geometry can prevent a probe from impinging directly on certain crack features. Procedures 

for examination of DMWs from the ID surface have been demonstrated to satisfy the detection 

requirements of Appendix VIII, Supplement 10. However, currently no procedure for ID 

examination of these welds has been successfully demonstrated to satisfy ASME 

Supplement 10 through-wall depth sizing root mean square error (RMSE) requirements. 

Dozens of NRC-approved relief requests, beginning in 2003, have allowed licensees to add the 

difference between a procedure-demonstrated RMSE and the ASME acceptance criterion of 

3.18 mm (0.125 in.) to the size of any flaw found in the field. The 3.18-mm (0.125-in.) value is 

the Appendix VIII, Supplement 10, allowed RMSE for flaw depth measurements as compared to 

flaw true depth during performance demonstration. This allowance was expected in 2003 to be 

a short-term remedy; however, no alternative approach has been found and agreed upon

between industry and the NRC to supersede this allowance.

This short-term allowance has been used very rarely because of low numbers of flaw detections

in DMWs and has not yet been used to demonstrate that PWSCC flaws are acceptable for 

continued service without mitigation or alternate examination. The issue of flaw sizing 

uncertainty is germane for unmitigated welds as well as welds intended for mitigation by MSIP 

or OWOL. Indications of flaws that are detected in DMWs (and other butt welds in primary 

system piping) must be dispositioned by repair, replacement, mitigation, or acceptance/

evaluation for continued service. The replacement, and often the repair, option remove the 

indication, but the mitigation and acceptance/evaluation options require the licensee to consider 

the depth of the flaw indication determined by NDE.

As discussed in Section 2.3.2, MSIP mitigates SCC by introducing a permanent compressive 

residual stress field on the inside surface of the DMW by way of mechanical surface 

deformation. A standard provision for applying MSIP to a weld is that an examination must be 

performed to show that there are no crack indications on an ID surface deeper than 30% of the 

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wall or having a total extent greater than 10% of circumference. This provision ensures that 

such flaws are effectively mitigated by the process, considering factors such as the uncertainty 

in the flaw depth, the uncertainty in the depth of the compressive residual stress zone, and the 

effect of operating load stresses.

For installation of OWOLs, the flaw depth limit per ASME Code Case N-754 is 50% of preOWOL wall thickness. For installation of FSWOLs, there is no pre-FSWOL flaw depth limit.

Industry met with the NRC staff on March 16, 2012, to discuss a permanent approach for 

addressing the depth-sizing uncertainty issue (Cumblidge 2012). The approach proposed by 

industry was not accepted by the NRC staff. However, another approach was developed that is 

contained in ASME Code Case N-696-1, “Qualification Requirements for Appendix VIII Piping 

Examinations Conducted from the Inside Surface, Section XI, Division 1.” This Code Case 

would replace the 3.18-mm (0.125-in.) RMSE acceptance criterion for depth-sizing of dissimilar 

metal and austenitic piping welds 54 mm (2.1 in.) and greater in wall thickness with a maximum 

RMSE of 6.35 mm (0.250 in.). Code Case N-696-1 was approved by the ASME Code, Section 

XI Committee on Nuclear Inservice Inspection, on May 7, 2014. The NRC staff member of the 

Section XI Committee on Nuclear Inservice Inspection voted affirmatively on this Code Case. 

This Code Case approach appears to be promising but is yet to be accepted by NRC.

3.4 Assessment of Code Case N-770-1 Inspection Requirements

3.4.1 Overview of Code Case N-770-1

ASME Code Case N-770-1 specifies examination methods, volumes or areas, and frequencies 

for Alloy 82/182 butt welds that are unmitigated as well as for Alloy 82/182 butt welds that have 

been mitigated against PWSCC by one of several specified methods. The requirements of this 

Code Case pertain to inspections for potential PWSCC. The ASME Code, Section XI, 

IWB-2500, requirements continue to apply to Alloy 82/182 welds to monitor for other 

degradation mechanisms, such as fatigue. Code Case N-770-1 contains baseline and inservice 

inspection (ISI) requirements for unmitigated Alloy 82/182 butt welds and preservice and ISI 

requirements for mitigated Alloy 82/182 butt welds. 

The structure of Code Case N-770-1 is patterned after the requirements of Section XI, 

Subsection IWB. The Code Case has requirements in Sections -1000, -2000, and -3000 that 

parallel the requirements in Subsection IWB-1000, -2000, and -3000. Section XI addresses 

repair/replacement activities that may be needed so there are no -4000 requirements in this 

Code Case. Similarly, pressure test requirements are unaffected so there are no -5000 

requirements in the Code Case. 

The examination requirements are contained in Table 1 of the Code Case in the -2000 section. 

Inspection Items in Table 1 are assigned for each “type and condition” of component. 

Inspection Items in the Code Case are similar to Examination Categories and Item Numbers 

used in the Section XI, Table IWB-2500-1. The “type and condition” of component refers to: 

(1) unmitigated welds at three different service temperatures addressed by Inspection Item A-1, 

A-2, and B; and (2) cracked and uncracked welds, mitigated by one of several specified 

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methods addressed by Inspection Items C through K. The Code Case includes various figures 

needed to implement the examination requirements. 

The examination methods specified in the Code Case are visual, volumetric, and surface 

examinations. The visual examination (VE) requirements are the same as the VE requirements 

of Code Case N-722-1. The requirements of Code Case N-722-1, with conditions, were 

imposed by 10 CFR 50.55a(g)(6)(ii)(E). Volumetric examinations are required to be performed 

by UT methods that meet the applicable requirements of Appendix VIII. Surface examinations 

by eddy current are required to satisfy IWA-2223 and by liquid penetrant testing are required to 

satisfy IWA-2222.

Welds in Inspection Items A-1, A-2, and B are required to be examined by VE and UT. 

Inspection Item A-1 welds, hot leg welds with an operating temperature greater than 625°F 

(325°C) (i.e., pressurizer nozzle DMWs), are required to be examined visually each refueling 

outage and volumetrically every second refueling outage. Inspection Item A-2 welds, hot leg

welds with an operating temperature less than or equal to 625°F (325°C), are required to be 

examined visually each refueling outage and volumetrically every 5 years. Inspection Item B 

welds, cold leg welds with an operating temperature greater than or equal to 525°F (274°C) and 

less than or equal to 580°F (304°C), are required to be examined visually once per interval and 

volumetrically every second inspection period not to exceed 7 years. This information is 

presented in Table 3.3.

Table 3.3 Examinations for Inspection Items A-1, A-2, and B

Inspection 

Item Parts Examined

Examination 

Method

Extent and Frequency of 

Examination

A-1 Unmitigated butt weld at Hot leg 

operating temperature > 625°F (329°C)

Visual Each refueling outage

A-1 Volumetric Every second refueling 

outage

A-2 Unmitigated butt weld at Hot Leg 

operating temperature ≤ 625°F (329°C)

Visual Each refueling outage

A-2 Volumetric Every 5 years.

B Unmitigated butt weld at Cold Leg 

operating temperature ≥ 525°F (274°C) 

and < 580°F (304°C)

Visual Once per interval

B Volumetric Every second inspection 

period not to exceed 7 years

Table 1 of Code Case N-770-1 contains the examination requirements and includes the parts 

examined, the examination volumes and surfaces shown in figures, examination methods, 

references to acceptance standards, extent and frequency of examination, and rules regarding 

deferral of the examination to the end of the interval. In Code Case N-770-1, Table 1 includes 

18 explanatory notes that contain additional requirements. These notes are necessary because 

of the range of weld types and conditions covered by the case.

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The examination requirements of Table 1 and the notes were developed to provide a set of 

comprehensive inspection requirements that parallel existing Section IWB Class 1 

requirements.(3)

 

3.4.2 Code Case N-770-1, Table 1 Notes

The notes to Table 1, which are applicable to welds in Inspection Items A-1, A-2, and B, are (1), 

(2), (3), (4), and (5).

Note (1) states that, “Volumetric examination requirements, methods, acceptance standards 

and frequencies are applicable to Class 1 PWR piping and vessel nozzle butt welds nominal 

pipe size (NPS) 2 (DN 50) or greater.” Paragraph-1100(a) indicates that Code Case N-770-1 

contains visual examination requirements of greater than NPS 1 (DN 25) pressure-retaining 

Class 1 PWR piping and vessel nozzle butt welds fabricated with Alloy 82/182 materials. 

Volumetric examination requirements do not apply to butt welds less than NPS 2, while visual 

examination requirements apply to unmitigated butt welds larger than NPS 1.

Note (2) states that, “A visual examination (VE) shall consist of the following:

“(a) A direct examination of the bare metal surface of the entire outer surface of the weld 

with the insulation removed or lifted to allow access for the VE.

“(b) The direct VE shall be performed at a distance not greater than 4 ft (1.2 m) from the

weld and with a demonstrated illumination level sufficient to allow resolution of lower case

characters having a height of not greater than 0.105 in. (2.7 mm).

“(c) Alternatively, the VE may be performed with insulation in place or removed using 

remote visual equipment that provides resolution of the weld metal surface equivalent to a 

bare metal direct VE as defined in (a) and (b), above.

“(d) Personnel performing the VE shall be qualified as a VT-2 visual examiner and shall 

have completed at least four hours of additional training in detection of borated water 

leakage from Alloy 600/82/182 components and the resulting boric acid corrosion of 

adjacent ferritic steel components.

“(e) Examination may be performed with the system depressurized.”

The requirements of Note (2) are essentially the same as the VE requirements of Code Case 

N-722-1.

Note (3) states that, “A VE may be performed during an outage when a volumetric examination 

is performed from the weld outer surface. An ultrasonic examination performed from the 

component inside or outside surface in accordance with the requirements of Table 1 and 

Appendix VIII (1995 Edition with the 1996 Addenda or later) shall be acceptable in lieu of the VE 

requirement of this table.” 

(3) This overview is general in nature and is not reiterated in Sections 4, 5, or 6.

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This note is essentially the same as a note in Code Case N-722 that applied to visual 

examination of full penetration welds (i.e., butt welds). Requirements for visual examination of 

full penetration welds were removed in the development of Code Case N-722-1 because these 

requirements were relocated to Code Case N-770. 

Note (4) states that, “Ultrasonic volumetric examination shall be used and shall meet the 

applicable requirements of Appendix VIII.” The requirements of this note are consistent with 

ASME Code Section XI and are implemented by NRC requirements in 10 CFR 50.55a.

Note (5) states that with regard to subsequent inservice inspection of unmitigated welds with 

inside surface-connected planar flaws, “(a) If planar surface flaws are detected in the butt 

weld/base metal inside surface, this weld shall be reexamined at the shorter frequency of every 

refueling outage or the frequency determined by the crack growth analysis of Code Case 

N-770-1 paragraph -3132.3; and (b) This weld shall be subsequently examined at the frequency

required by (a) unless mitigated.” The requirements of this note are consistent with ASME 

Code, Section XI requirements.

3.4.3 NRC Conditions on Code Case N-770-1

The NRC conditions for unmitigated welds in the 10 CFR 50.55a(g)(6)(ii)(F), which imposed 

Code Case N-770-1, are discussed in detail below. The NRC imposed these conditions, which 

are in addition to the Code Case requirements, to ensure that, in combination with the Code 

Case, adequate protection is provided for monitoring Alloy 82/182 welds.

On May 4, 2010, the NRC issued a notice of proposed rulemaking to amend its regulations to 

incorporate by reference the 2005 Addenda through 2008 Addenda of Section XI, Division 1 of 

the ASME Code, as well as other recent Code updates (75 FR 24324). The NRC proposed to 

add a new Section 50.55a(g)(6)(ii)(F) [§50.55a(g)(6)(ii)(F)] to require licensees to implement 

ASME Code Case N-770, with 15 conditions. The NRC stated that the application of ASME 

Code Case N-770 is necessary because the inspections currently required by the ASME Code, 

Section XI, were not written to address degradation of Alloy 82/182 butt welds by PWSCC, and 

the safety consequences of inadequate inspections can be significant. The NRC stated that it 

was concerned that the absence of an effective inspection regime could, over time, result in 

unacceptable circumferential cracking or the degradation of reactor coolant system components 

by corrosion from leaks in these welds. If not addressed, degradation by PWSCC would 

increase the probability of a loss-of-coolant accident.

About half of the NRC proposed conditions were related to specific issues that the NRC raised 

after Code Case N-770 was finalized and approved by ASME. These issues were addressed 

by the ASME, Section XI, Task Group on Alloy 600/182/82 Issues in Code Case N-770-1, which 

was approved by the ASME in December 2009. Although Code Case N-770-1 was approved 

by ASME before the proposed rule was issued, the NRC decided not to revise the proposed rule 

package to include Code Case N-770-1, because it would have delayed the entire rulemaking 

process. The proposed NRC conditions that were addressed by Code Case N-770-1 were not 

considered controversial and were accepted by ASME. As a result of public comments, the final 

rule imposed Code Case N-770-1 instead of Code Case N-770. The rule was issued on 

June 21, 2011 (76 FR 36232).

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The remainder of this section discusses the conditions in the final rule imposing Code Case 

N-770-1 (76 FR 36232) that pertain to unmitigated welds; specifically, conditions 

§50.55a(g)(6)(ii)(F)(1), (2), (3), and (4).

Condition §50.55a(g)(6)(ii)(F)(1) of the rule requires licensees of existing operating PWRs as of 

July 21, 2011, to implement the requirements of ASME Code Case N-770-1, subject to the 

conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) by the first refueling 

outage after August 22, 2011. This is the basic implementing requirement imposing Code Case 

N-770-1 with conditions.

Condition §50.55a(g)(6)(ii)(F)(2) states that, “Full structural weld overlays authorized by the 

NRC staff may be categorized as Inspection Items C or F, as appropriate; welds that have been

mitigated by stress improvement without welding may be categorized as Inspection Items D or 

E, as appropriate, provided the criteria in Appendix I of the code case have been met; for ISI

frequencies, all other butt welds that rely on Alloy 82/182 for structural integrity shall be 

categorized as Inspection Items A–1, A–2, or B (Inspection Items for unmitigated welds) until 

the NRC staff has reviewed the mitigation and authorized an alternative code case Inspection 

Item for the mitigated weld, or until an alternative code case Inspection Item is used based on

conformance with an ASME mitigation code case endorsed in Regulatory Guide 1.147 with 

conditions, if applicable, and incorporated in this section.” This condition applies to Alloy 82/182 

butt welds with inlays or onlays or optimized weld overlays which have not been authorized by 

NRC, either on a weld-specific basis or through Regulatory Guide 1.147, to be categorized as 

Inspection Items D, E, or G–K. This condition is consistent with NRC philosophy to review and 

approve mitigation techniques prior to implementation and prior to a licensee taking credit for 

the reduced inspection frequencies for mitigated welds. The NRC evaluates mitigation methods 

to verify continued compliance with applicable General Design Criteria (GDC). An exception to 

the previous statement applies to MSIP welds, and this exception is discussed in some detail in 

Sullivan and Anderson (2013).

Condition §50.55a(g)(6)(ii)(F)(3) of the rule requires that, “Baseline examinations for welds in 

Table 1, Inspection Items A–1, A–2, and B, shall be completed by the end of the next refueling 

outage after January 20, 2012. Previous examinations of these welds can be credited for 

baseline examinations if they were performed within the re-inspection period for the weld item in 

Table 1 using Section XI, Appendix VIII requirements and met the Code-required examination 

volume of essentially 100 percent. Other previous examinations that do not meet these 

requirements can be used to meet the baseline examination requirement, provided NRC 

approval of alternative inspection requirements in accordance with paragraphs (a)(3)(i) or 

(a)(3)(ii) of this section is granted prior to the end of the next refueling outage after January 20, 

2012.”

This condition was imposed by the NRC, because the NRC did not agree with -2200 of Code 

Case N-770-1, which would allow two refueling outages from adoption of the Code Case to 

complete baseline examinations. Welds in Inspection Items A-1, A-2, and B are the welds most 

likely to experience PWSCC and some of these welds may not have received a baseline 

examination or received a complete examination, even under the industry initiative, MRP-139

(EPRI 2005; NRC 2010a). The NRC clarified in a public meeting (Collins 2011) on the rule that 

previously approved alternatives to ASME Code requirements regarding inspection frequency or 

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coverage remain applicable during the duration authorized by the NRC. Further, licensees may, 

if the examination was previously performed within the inspection frequency of Table 1 of ASME 

Code Case N-770-1, count these inspections as meeting the baseline inspection requirement of 

10 CFR 50.55a(g)(6)(ii)(F)(3). If a licensee’s previous inspection did not meet ASME Code,

Section XI, Appendix VIII requirements and the Code-required examination volume of 

essentially 100%, then relief in accordance with 10 CFR 50.55a(a)(3) remains an option to 

count these examinations as meeting the baseline examination requirement.

Code Case N-770-1, paragraph -2500(c) states that, “For axial and circumferential flaws, 

examination shall be performed to the maximum extent practical using qualified personnel and 

procedures. If essentially 100% coverage for circumferential flaws (100% of the susceptible 

material volume) can be achieved, the examination for axial flaws shall be completed to achieve 

the maximum coverage practical.” The NRC disagreed with this provision. In the proposed 

rule, the NRC noted that the -2500(c) requirement on inspection limitations is inconsistent with 

comparable inspection requirements of the ASME Code, Section XI. Axial flaws can lead to 

through-wall cracks and leakage of reactor coolant, which is a safety concern. Condition 

§50.55a(g)(6)(ii)(F)(4) requires that the axial and circumferential (flaw) examination coverage

requirements of -2500(c) may not be considered to be satisfied unless essentially 100%

coverage is achieved. The industry guidelines of MRP-139 (EPRI 2005) allow less than 

essentially 100% coverage in some cases; therefore, a number of previously conducted 

baseline examinations may not satisfy the rule. This condition was added for the NRC to 

ensure that, through NRC review of an authorization of alternative inspection coverage, 

appropriate actions are being taken to address potential inspection limitations for axial flaws.

3.4.4 Assessment of Unmitigated Weld Inspection Intervals

The results of four deterministic flaw evaluations are presented in Appendix E. In these 

evaluations, an initial flaw size or sizes are assumed and flaw growth analyses were performed 

to determine the time for a crack to progress to either a Code maximum allowed flaw depth or to 

progress through-wall. In some of the studies, a number of parameters, such as loads or initial 

flaw sizes, were varied to conduct a sensitivity study. The results of these evaluations were 

used to assess the Code Case N-770-1 inspection interval requirements. Three of the 

evaluations pertain to circumferential cracking in pressurizer, hot leg temperature, and cold leg 

temperature welds. The fourth evaluation was performed to assess axial cracking. A 

deterministic assessment of inspection intervals has limitations, compared to a probabilistic 

assessment, in that it cannot account for a full range of variables and associated uncertainties. 

Nevertheless, the results from these evaluations are considered to provide useful insights.

In the first study, which is discussed in Appendix E.1, results from mitigation studies performed 

by Battelle Columbus Laboratory were provided for an RPV hot leg nozzle weld without a safe 

end weld, a surge nozzle weld, and a pressurizer safety nozzle weld. The welds were analyzed 

for normal operating pressure and temperature without bending-moment loading. In all three 

cases, it was concluded that either a postulated circumferential flaw would not propagate or 

would arrest before reaching one-third through the wall. This result is similar to Stress Case “C” 

analyzed with a safe end closure weld in the NRC staff’s Hot Leg and Cold Leg Flaw Evaluation 

and Sensitivity Analyses summarized in Appendix E.3.

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In the second study, which is discussed in Appendix E.2, results from pressurizer nozzle 

advanced finite element analyses of circumferential flaws were provided for a sensitivity study of 

a surge nozzle weld under various loading conditions and with various postulated initial flaws as 

shallow as 10% through-wall. A sensitivity case with high loads and long shallow flaws

simulating multiple crack initiation sites and omitting the beneficial effect of a secondary SS

weld showed little if any time between first leakage and rupture. For this bounding case, 

monitoring pressurizer nozzle Alloy 82/182 welds with ISI would not appear to be an adequate 

strategy. However, from information on the NRC website along with Stall (2007), Wadley (2008)

and King (2006), all pressurizer surge nozzle DMW welds have been mitigated except at 

Palisades. As of mid-2014, there were unmitigated pressurizer nozzle butt welds at three 

plants. These welds have all been inspected at least once and no flaws were detected. The 

plants have committed to re-inspect these unmitigated welds at least once every 4 years, 

consistent with the requirements of Code Case N-770-1.

In the third study in Appendix E.3, results were provided for circumferential flaw evaluation 

sensitivity studies of hot leg and cold leg reactor vessel nozzle welds under various loadings for 

an initial flaw size of 24% through-wall. This initial flaw depth was selected based on the depth 

of a flaw found in a PWR hot leg reactor vessel nozzle weld. The flaw was attributed to PWSCC

and the study was performed to assess the significance of the flaw. The results for the hot leg 

analyses showed that there is sufficient time between leakage and rupture to shut down the 

plant should through-wall leakage occur. Analyses with safe end welds showed times to initial

leakage that bound the ASME Code Case N-770-1 inspection intervals for Inspection Item A-2 

(hot leg temperature) welds. Addressing uncertainties by including analysis cases of welds 

without attached safe ends, the times to initial leakage for Load Cases A, B, and D do not bound 

the inspection intervals in Code Case N-770-1 for Inspection Item A-2 welds.

In the cold leg cases the times to initial leakage, with one exception, bound the maximum 

inspection interval of 7 years in Code Case N-770-1 for Inspection Item B, unmitigated butt 

welds at cold leg operating temperature ≥ 525°F (274°C) and < 580°F (304°C). The exception 

is a high load case with a partial arc repair and no safe end closure weld. For this case, the 

time to leakage was 6.71 years, just short of the maximum interval of 7 years. The times 

between leakage and rupture for the cold leg cases are equal to or longer than 7.4 years. The 

results of the cold leg analysis appear to support an inspection interval of at least 7 years. 

In the fourth study in Appendix E.4, results were provided for an axial flaw evaluation of a steam 

generator nozzle to hot leg pipe DMW. The analysis showed for this example that an assumed 

10% through-wall flaw would grow to 75% through-wall in approximately 3-1/3 years or 

40 months. This interval of 3-1/3 years is considerably shorter than the required UT inspection 

interval of once every 5 years for Code Case N-770-1 Inspection Item A-2 welds.

Except for the results of the Battelle study, the analyses showed the time from an initial flaw size 

to through-wall leakage or to 75% through-wall cracking (in Appendix E.4) to be shorter than the 

inspection intervals for the Code Case N-770-1 Inspection Item A-1 and A-2 welds analyzed. 

Initial flaw sizes were assumed to be either 10% or approximately 25% through-wall depending 

on the study. Various flaw lengths were also assumed, again depending on the study, to model 

different sensitivity cases. Ten percent through-wall is frequently assumed to be a threshold of 

detection for UT. Initial circumferential flaw depths of approximately 25% were assumed to 

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model flaws detected during actual inspections of DMWs. Flaws approximately 25% throughwall were found at a number of plants. In the fourth study involving an axial flaw, an analysis 

was performed to estimate the time it could have taken for the deepest flaw to grow to the depth 

measured when discovered at NAPS-1. By contrast, in the cold leg cases the times to initial 

leakage, with one exception, bound the maximum inspection interval of 7 years in Code Case

N-770-1 for Inspection Item B, unmitigated butt welds at cold leg operating temperature.

Ideally, results from these studies for time to reach the ASME Section XI maximum depth of 

75% through-wall or the Code stability limit would bound the ASME Code Case inspection 

intervals, because this is the regulatory basis for degradation as opposed to showing that there 

is sufficient time to shut down a plant after leakage is discovered. Nevertheless, there have not 

been any occurrences of circumferential cracking that have exceeded ASME Code limits. There 

has been one case of axial through-wall leakage during plant operation at V.C. Summer in 2000 

and two cases of axial through-wall leakage that occurred during plant shutdown conditions at 

Davis Besse in 2008 and North Anna in 2012. With only a few exceptions, pressurizer (hightemperature) DMWs have been mitigated. Some hot leg temperature DMWs and a smaller 

number of cold leg temperature DMWs have been mitigated by a variety of techniques. 

Baseline visual and volumetric inspections of pressurizer, hot leg, and cold leg temperature 

Alloy 82/182 butt welds have been performed.

These deterministic analyses generally indicate that Code Case N-770-1 inspection intervals 

may be too long for use as a monitoring strategy to prevent PWSCC from exceeding the ASME 

Code limit of 75% maximum through-wall depth. The analyses also showed that there is 

sufficient time between leakage and rupture to shut down the plant should through-wall leakage 

occur. The Code Case N-770-1 inspection intervals were developed by industry and received 

NRC staff agreement during the Code Case development process and by virtue of its adoption 

in 10 CFR 50.55a. The welds to which this Code Case apply have received baseline 

examinations but few if any second-round examinations. Although axial cracking and axial 

through-wall leakage were found, no circumferential or axial PWSCC has exceeded ASME 

Code structural factors. While deterministic analyses indicate that Code Case N-770-1 

inspection intervals may be too long, additional bases would be needed to support a 

modification to Code Case N-770-1 inspection intervals. Such support would have to be the 

result of unfavorable results of realistic probabilistic analyses and/or operating experience, 

including results of second and later sets of examinations. 

3.5 Conclusions

In assessing the overall reliability of UT examinations to detect and accurately characterize 

service-induced SCC, there are a number of factors that must be considered. These include: 

• ability of the applied NDE to obtain full coverage of the susceptible inspection volume, 

• presence of interfering signals such as from geometric or welding fabrication reflectors, 

• the use of non-encoded versus encoded techniques, 

• general reflectivity of the service-induced flaw, 

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• acoustic impedance of the materials being examined and the signal attenuation and 

dispersion through those materials, 

• relative experience and skill of the examiner (and/or data analyst),

• use of modeling to design and optimize examination parameters, and

• representativeness of mockups used for performance demonstration qualifications.

The use of encoded UT techniques offers a distinct and significant advantage over non-encoded

techniques in detection and characterization of flaws, because of the ability to review the UT 

data off-line and carefully determine the origin of signals (both relevant and non-relevant) 

present in the images produced.

Current mockups and protocols used by industry during the PDI qualification process may not 

provide sufficient opportunities such that candidates may demonstrate adequate skills for 

interpreting realistic false call signals from non-crack reflection sources that may be 

encountered in the field; for example, fabrication flaws, weld root response variations, or other 

geometrical/metallurgical responses. Thus, examiners may pass the PDI qualification process 

without sufficiently demonstrating the ability to distinguish ID-connected flaws from other signals 

that may be expected to be present in a weld inspection volume, especially in DMWs.

Inspection limitations can affect the effectiveness of NDE requirements. Appendices C and D to 

this report contain weld coverage assessment diagrams for typical plants that show probe 

angles used and percent coverage obtained with Section XI, Appendix VIII, qualified techniques. 

ASME Code-required volumetric coverage appears achievable for many, although not all, 

Westinghouse and CE Alloy 82/182 RCS butt welds.

ASME Section XI prepared Code Case N-824, entitled “Alternate Requirements for Ultrasonic 

Examination of Cast Austenitic Piping Welds from the Outer Diameter Surface.” NDE reliability 

would be expected to be substantially improved by applying this Code Case, with certain 

conditions based on NRC research, in situations where 100% of the susceptible material cannot 

be examined from the ferritic side of the weld. The examination requirements of ASME Code 

Case N-770-1 appear to be complete and parallel existing Section IWB Class 1 requirements. 

For unmitigated welds, the requirements reflect a philosophy of more frequent monitoring for an 

active degradation mechanism. The NRC imposed a number of additional conditions in 

combination with Code Case N-770-1 for monitoring Alloy 82/182 welds.

The results of the cold leg temperature deterministic flaw analyses support an inspection 

interval of at least 7 years. However, in general, the bounding cases from the hot leg 

temperature deterministic crack flaw growth sensitivity studies do not bound the inspection 

intervals in Code Case N-770-1 for unmitigated welds. All pressurizer welds, which would be

the most likely to experience PWSCC, have been inspected and most have been mitigated or 

replaced with Alloy 52/152 welds. Baseline examinations of the welds addressed by MRP-139 

and Code Case N-770-1 have been completed, and although axial cracking and axial throughwall leakage have been found, no instances of circumferential or axial PWSCC have been 

observed that exceeded ASME Code structural factors. While deterministic analyses indicate 

that Code Case N-770-1 hot leg temperature inspection intervals may be too long, additional 

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bases would be needed to support a modification to Code Case N-770-1 inspection intervals. 

Such support could arise from unfavorable results of realistic probabilistic analyses and 

operating experience, including results of second and later rounds of examinations of these 

DMWs. It will be important to assess the results of the second and subsequent sets of Code 

Case N-770-1 DMW examinations against the Code Case N-770-1 inspection interval

requirements.

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4 MANAGEMENT OF PWSCC BY MSIP

4.1 Operating Experience with BWR Welds Mitigated by MSIP

Based on an information search performed by the authors as well as discussions with a number 

of industry experts familiar with decades of operating experience with boiling water reactors 

(BWRs), no instances were identified of leaks in welds mitigated by stress improvement 

(Findlan et al. 2004). However, there have been instances in which cracks, not identified when 

mitigating by the mechanical stress improvement process (MSIP), have been found during 

subsequent inspections.

During meetings held between U.S. Nuclear Regulatory Commission (NRC) staff and industry 

on May 25–26, 2010, representatives of the Boiling Water Reactors Vessels and Internals 

Project (BWRVIP) provided a summary of BWR operating experience in 2007 and 2008 for 

Category C and D welds containing Alloy 82/182 (Wirtz 2010). It was noted that NUREG-0313, 

Category C welds are those susceptible to intergranular stress corrosion cracking (IGSCC) that 

were treated by stress improvement after 2 years of operation while Category D contains welds 

susceptible to IGSCC with no stress improvement. Ten cases of these welds were found to 

have planar flaws; these were distributed among several BWRs. The presentation did not 

separate the cases by weld category. In 2008, after reviewing these ten cases of welds with 

planar flaws, the BWRVIP developed an accelerated inspection program for Category C and D 

welds containing Alloy 82/182 exposed to BWR environments that had not received an

Appendix VIII, Supplement 10 examination.

Surface contour of the outside-diameter (OD) weld region appeared to be the major factor 

affecting accurate detection and sizing. The presence of weld crowns and radial shrinkage near 

the weld caused incomplete probe contact resulting in a loss of ultrasonic testing (UT) signal. 

As part of the accelerated inspection program, the BWRVIP provided guidelines to owners on 

surface conditioning to improve UT effectiveness.

During the meetings on May 25–26, 2010, the BWRVIP reported that five of the ten BWR 

Category C and D welds identified with flaws requiring IWB-3600 evaluations were 

subsequently overlaid. Some of the welds were determined to contain fabrication flaws and 

were found to be acceptable for continued service. During a meeting on July 20, 2012 (Wirtz 

2012), it was reported that two flawed Category C welds were found by the accelerated 

inspection program since 2009. These welds were subsequently overlaid. It was also reported 

that four Category C welds remain to be examined and were scheduled to be inspected by June 

2014. The owners of these BWRs believe that none of these flaws were the result of new or 

continuing IGSCC. Rather, they believe that the cracks were present at the time stress 

improvement was applied. It is reasonable to conclude that these cracks were not the result of 

new IGSCC. However, one has to consider that OD weld surface conditions may have had a 

negative effect on earlier examination results; therefore, the conclusion that none of the flaws 

were the result of new or continuing IGSCC cannot be proven.

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This operating experience highlights the importance of weld OD surface conditioning, and in 

some cases, includes removing weld crowns, and the benefit of performing effective and reliable 

examinations by applying advanced technology such as Appendix VIII-qualified, encoded 

phased-array UT techniques, before and after application of MSIP.

4.2 MSIP Design Requirements

The ASME Code does not address the mechanical stress improvement process in Section XI 

requirements for repair and replacement as MSIP does not involve welding or replacement of 

components. Additionally, MSIP is essentially a proprietary process and, as such, the details of 

the design requirements are not publicly available.

During discussions between Pacific Northwest National Laboratory (PNNL) and the Chief 

Engineer of NuVision Engineering, the company that holds the patent for MSIP, it was indicated 

that the design process meets the criteria contained in ASME Code Case N-770-1, Appendix I, 

Performance Criteria and Measurement or Quantification Criteria for Mitigation by Stress 

Improvement. At a high level, this appendix requires that WRS calculations be performed to 

bench-mark the “squeezing” process. The applied radial load permanently reduces the OD and, 

to a lesser extent, the inside diameter (ID) in the region where the radial load is applied. A

typical change in OD caused by MSIP is 1% of the as-built size, although the vendor may apply 

somewhat more or less than 1% to achieve the desired change in residual stresses in the pipe 

weld. This “squeeze” creates a beneficial compressive stress on the ID of the welded region of 

interest. Candidate welds have to be reviewed during the design process to ensure that the 

weld can be inspected before and after MSIP by a qualified technique. With one exception 

discussed later in this report, a pre-MSIP UT inspection is required to be performed. MSIP may 

be applied to uncracked welds and, depending on the severity of the crack, on cracked welds. 

Any flaws that are detected during the pre-MSIP UT have to be evaluated prior to mitigation 

using the rules of Section XI, IWB-3600, Analytical Evaluation of Planar Flaws. Westinghouse 

Electric Company works with NuVision Engineering to implement MSIP. Westinghouse 

indicates that the process is only applied to locations that have circumferential cracks where the 

lengths add to no longer than 10% of the circumference and the maximum through-wall depth of 

any crack does not exceed 30% of the piping wall thickness (Elder 2008). This follows the 

guidelines on application of MSIP that are provided in NUREG-0313, Rev. 2 (Hazelton and Koo 

1988). A discussion of UT uncertainty for Appendix VIII piping examinations conducted from the 

inside surface is contained in Section 3.3.

MSIP causes a slight elongation of the region of the pipe being compressed. This elongation 

has to be analyzed to ensure that it does not adversely affect the piping system or cause ASME 

Code-allowable stresses to be exceeded. A more in-depth discussion of the Code Case 

N-770-1, Appendix I, criteria for MSIP is included in Section 5 of the report by Sullivan and 

Anderson (2013).

Fredette and Scott (2009), from Battelle Columbus Laboratory (BCL), performed finite element 

analyses on applying MSIP to pipe welds in multiple piping systems. They assessed initial 

WRS, including welds with and without an attached safe end weld. Weld residual stresses in 

both the dissimilar metal weld (DMW) and adjacent stainless steel (SS) weld (if present), after 

application of MSIP, were determined. Flaw growth calculations were also performed to 

4-2
understand the effect of applying MSIP to welds containing flaws. These analyses were 

reviewed by Sullivan and Anderson (2011) and confirm that compressive residual stresses at or 

near the inside surface are produced by MSIP. These analyses confirm the guidance in 

NUREG-0313, Rev. 2 (Hazelton and Koo 1988) on acceptable sizes of flaws that may be left in 

welds repaired by MSIP. The analyses by BCL also addressed the possibility of a deep flaw 

being missed or undersized during the pre-MSIP UT examination. The analyses showed that if

MSIP is applied to a weld with a deep pre-existing flaw (> 60% through-wall), MSIP increases 

the stress intensity factor (K), such that the time required for the crack to grow through the 

remaining wall thickness of the treated weld (post-MSIP) is very short. From the perspective of 

applying MSIP to welds with flaws, Sullivan and Anderson (2011) also indicates that the Codeallowable flaw sizes under design basis seismic loading in some cases are small. These results 

highlight the importance of effective and reliable nondestructive examination (NDE) prior to the 

application of MSIP.

BCL considered a series of WRS sensitivity analyses and included the following effects: the 

effect of modeling fixed and free displacement boundary conditions at the ends of the modeled 

sections; the effect on WRS of having or not having a nearby secondary SS weld; and the 

effects of changes in MSIP clamp location, the level of plastic deformation introduced, and tool 

size. BCL also performed sensitivity studies on 3D versus 2D analyses. These sensitivity 

studies assessed factors that might affect the WRS state prior to and after application of MSIP, 

and showed that large changes in the analyzed variables could impact the effectiveness of 

MSIP, but that ranges of variations normally expected to be encountered would not have a 

significant impact on MSIP effectiveness.

4.3 NDE Reliability

To evaluate the reliability of ultrasonic inspections, PNNL considered flaws evaluated both 

before and after MSIP application. The data used to make UT comparisons came from multiple 

sources: UT data on unmitigated welds presented during a workshop held by PNNL in May 

2012, a nuclear power plant in Lithuania, and a specimen built by PNNL. Section 3.1

summarizes the findings of the PNNL UT workshop on unmitigated welds, which is relevant to 

the reliability of pre-MSIP inspections. Section 4.3.1 summarizes the results of an investigation 

based on pre- and post-MSIP data acquired from cracked areas in 325-mm (12.8-in.) diameter 

piping at the Ignalina Nuclear Power Plant (INPP) in Lithuania. Section 4.3.2 summarizes the 

results of a follow-on exercise in which PNNL acquired and evaluated UT data from a PNNL 

DMW specimen containing implanted thermal fatigue cracks (TFCs). Section 4.3.3 contains a 

comparison of the UT data from Ignalina and the PNNL specimen (Crawford et al. 2011). 

Section 4.3.4 provides a discussion on potential limitations to conducting pre- and post-MSIP 

inspections.

4.3.1 Ignalina Nuclear Power Plant

4.3.1.1 Test Materials

Data was acquired from piping in the INPP (Crawford et al. 2011). The piping base material is 

08X18H10T, 18% Cr, 10% Ni, C wrought SS with a nominal outside diameter of 325 mm 

4-3
(12.8 in.) and a wall thickness of approximately 16 mm (0.63 in.). Two field welds and five 

shop-fabricated welds were included in this study.

Service-induced IGSCC was present in the heat-affected zone of many of the INPP piping 

welds because of high residual stresses imparted by autogenous root-welding procedures and 

the welds having been exposed to oxygen-rich water chemistry. Eight circumferentially-oriented

flaws from the Ignalina plant were included in this study.

4.3.1.2 MSIP Application

During the MSIP process, plastic deformation of the pipe wall adjacent to the weld (under the 

mechanical clamp) results in contouring of the inner and outer surfaces of the weld. These

surface conditions can potentially misalign subsequent ultrasonic transducer placement,

causing examination volumes to be limited except at extremely high insonification angles, and 

may produce flaw reflections to be redirected at a higher angle than would occur on a flat 

surface. When scanning over this contour, this surface condition introduces a gap between

transducers and the OD surface of the pipe, which decreases ultrasonic coupling, and causes

lack of sufficient sound penetration and/or irrelevant signals in the data. Because the effects of 

possible ultrasonic beam redirection and loss of coupling in the ultrasonic data from the MSIPapplied side is a concern, only UT data from the non-MSIP side of the INPP welds is discussed.

4.3.1.3 Ultrasonic Methods

The pipes at INPP were scanned using a phased-array (PA) transmit-receive dual shear-wave 

(TRS) probe, which allowed examiners to scan the piping welds with good coverage in a short 

amount of time. A ZETEC Z-Scan phased-array system with a 32/128PR channel configuration 

was used to record rectified A-scan data in line scans. The shear-wave phased-array probe 

had 2 × 32 elements and was designed to operate at 4.0 MHz with refracted angles from 50° to 

80°.

A 4.0-MHz shear-wave probe would likely be ineffective for penetrating austenitic welds in 

commercial U.S. reactors, as the dendritic grain structure of the weld metal absorbs and 

redirects sound beams, especially in this frequency regime. For this reason, most through-weld 

applications in austenitic piping at U.S. plants use refracted longitudinal waves in the 1.0- to 

2.25-MHz frequency range. However, the titanium-stabilized SS used for INPP piping and 

welds produced a fine, equiaxed, and randomly oriented grain structure only slightly coarser 

than would be found in carbon steel. This grain structure allows higher frequency shear waves 

to penetrate the welds much more effectively, thus facilitating ultrasonic examination and 

producing higher-resolution images.

Line scans adjacent to the welds were acquired with a calibrated probe orientated perpendicular 

to the weld for detection of circumferential flaws; that is, flaws parallel to the weld. Data was 

acquired at 1-mm (0.04-in.) increments circumferentially while the sound beam was swept from 

50° to 80° through the weld in 1° increments. The resolution of the mechanical scanning 

movement along the weld was 1.0 mm (0.04 in.).

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4.3.1.4 Data Evaluation

Staff from INPP identified areas of the subject welds containing IGSCC found during previous 

inspections. Phased-array data were collected from the cracked pipe welds and on the crackfree pipe weld, typically for 360° around the pipe. Each of the areas containing indications was 

analyzed to determine the signal-to-noise ratio (SNR) and the depth and length of the indication. 

The SNR was calculated from the peak flaw response and the average noise level in a flawless 

zone at the same part path or depth. Flaw depth was estimated using tip-diffracted signals 

where present, and flaw length was measured to the loss-of-signal level.

One IGSCC indication in weld P27z1 was detected prior to MSIP but not during the post-MSIP 

inspection. The amplitude response for all but one of the indications decreased after MSIP, with 

the one indication exhibiting a very small increase in amplitude response. The measured length 

of all the indications decreased after MSIP.

4.3.2 PNNL Dissimilar Metal Weld Specimen

4.3.2.1 Test Materials

Data were acquired from a DMW specimen, 9C003, fabricated by PNNL to simulate a 

pressurizer surge nozzle (Crawford et al. 2011). The weld of interest is an A106B/A105 carbon 

steel nozzle welded to a 316 SS safe end. The safe end is welded to a 316 SS pipe, and a 

large carbon steel flange was added to the nozzle to provide rigidity during the MSIP. The 

nozzle butter and nozzle-to-safe end weld material are NiCrFe3 (Alloy 82/182) with the safe 

end-to-pipe weld material being 308 SS. The nozzle-to-safe end and safe end-to-pipe weld 

crowns were ground smooth and flush with the OD surface.

Six circumferential and one axial thermal fatigue cracks were implanted in the butter-to-weld

region nearer the carbon steel nozzle side. The circumferential cracks were tilted between 8°

and 15° and varied in depth from 16% to 90% through-wall. Flaws were implanted within the 

butter material to minimize the potential for disturbing the parent material, thus avoiding the 

introduction of implantation anomalies that could result in the reflection of coherent sound energy.

4.3.2.2 MSIP Application

MSIP on the 9C003 specimen was applied over the safe end-to-pipe weld adjacent to the

targeted nozzle DMW containing the implanted flaws. A 0.94% reduction was achieved on the 

specimen as measured by NuVision during the MSIP application. Circumferential change 

measurements were taken at an axial mark on the safe end before and after MSIP. 

4.3.2.3 UT Methods

The PNNL 9C003 specimen was examined with two phased-array probes with designed center 

frequencies of 1.5 and 2.0 MHz. Both probes were operated in a transmit-receive longitudinal 

mode.

Phased-array data was acquired with a ZETEC DYNARAY system in conjunction with a 

manual-encoded scanner mounted directly on the specimen. Two encoders provided positional 

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information in the circumferential direction for line scans and in both circumferential and axial 

directions for raster scans. The probes were water-coupled and data were acquired over 

inspection angles of 30° to 70° in 1° increments. Line scan and raster scan data were acquired 

from both sides of the weld with the 1.5-MHz probe for the six circumferential flaws and the axial 

flaw. Line and raster data were also acquired from both sides of the axial flaw at 1.5 MHz. At 

2.0 MHz, line scan data were acquired from both sides of the weld on the six circumferential 

flaws. Raster data was limited to three of the circumferential flaws from both sides of the weld 

and no data were acquired on the axial flaw.

4.3.2.4 Data Evaluation

All six of the implanted TFCs were detected with both probes before and after the MSIP 

application. Because the true dimensions of the implanted flaws are known, the error in 

measurements could be determined. The measurement error data show that the pre- and postMSIP values overlap and do not indicate a reduction in flaw length after MSIP application. The 

smaller flaws are over length-sized in both the pre- and post-MSIP data. Root-mean-square 

errors (RMSEs) were within the ASME guidelines of 19.05 mm (0.75 in.) for acceptable length 

sizing.

The pre- and post-MSIP flaw depth values overlap, showing little difference between the two 

sets of measurements. The depth sizing was very good for all but the two deepest flaws. 

Neither PA probe focused well at the higher angles needed for detection of deep flaw tips. 

RMSEs were within the ASME guidelines of 3.18 mm (0.125 in.) in depth sizing for only half of 

the scans.

4.3.3 Comparison of Ignalina and PNNL 9C003 Specimen Data

A comparison of the changes in flaw characteristics between pre- and post-MSIP data for both 

types of flaws was made. One of the eight INPP flaws was no longer detected after MSIP; 

however, all six of the implanted flaws in the 9C003 specimen remained detectable after MSIP. 

On average, the Ignalina data show a length change of −18.4 mm (−0.72 in.) if all flaws are 

included. The change in length for the implanted flaws in the PNNL specimen does not show a 

negative or a positive trend. An increase in flaw length is unreasonable; therefore, any such 

observation of small growth is attributed to measurement error rather than an actual growth in 

the flaw length. This lack of a noticeable decrease in flaw length with the MSIP application in 

the PNNL specimen was not expected.

The Ignalina data show a loss in flaw response as measured by SNR at −6.4 dB if all eight flaws 

are included. The implanted flaw data on 9C003 for both probes, and line and raster images, 

trend toward a reduction in amplitude, but on average a small change of only −1.9 dB is 

4-6
observed.(1)

 On average, the six implanted flaws in the 9C003 specimen showed minimal 

change, if any, after MSIP. Because the flaw implantation process may have introduced a slight 

misalignment of the two crack faces on each flaw, it is hypothesized that during MSIP, this 

misalignment would have prevented the crack from closing, leading to similar ultrasonic 

responses for the pre- and post-MSIP data.

4.3.4 Inspection Limitations

Condition §50.55a(g)(6)(ii)(F)(2) of the rule that imposes Code Case N-770-1 allows welds 

mitigated by mechanical stress improvement to be categorized as Inspection Items D or E, as 

appropriate, provided the criteria of Appendix I of the Code Case have been met. Code Case 

N-770-1, Appendix I, Criterion 3 requires that mockup testing and NDE qualified to 

Appendix VIII performance demonstration requirements be performed to demonstrate that 

examination of the relevant volume of the mitigated component can be accomplished 

subsequent to mitigation including changes to the component geometry caused by MSIP. Code 

Case N-770-1, Appendix I, Criterion 5, requires that, “the mitigated weld shall be inspectable by 

a qualified process.” Criterion 5 further requires that, “An evaluation shall be performed to 

confirm that the required examination volume of the mitigated configuration is within the scope 

of an Appendix VIII supplement or supplements and that the examination procedures to be used

have been qualified in accordance with Appendix VIII. The evaluation shall confirm that the 

geometric limitations (e.g., weld crown, nozzle contour) of an Appendix VIII qualification are not 

exceeded for the mitigated weld.” If examination of the required examination volume cannot be 

satisfied, a licensee would be expected to obtain NRC approval of an alternative examination 

prior to performing the mitigation.

Information on geometric limitations to performing inspections of typical Westinghouse and 

Combustion Engineering (CE) plant reactor coolant system (RCS) Alloy 82/182 welds was 

obtained from LMT, Inc., an NDE vendor for the commercial nuclear power industry, during a 

workshop PNNL held in May 2012. The information on inspection limitations acquired during 

this workshop is discussed in Section 3.2 of this report and in Appendices C and D. These 

appendices contain weld coverage assessment diagrams for a typical Westinghouse and a 

typical CE plant. 

Another limitation to inspection of candidate welds for MSIP pertains to inspection of cast 

austenitic stainless steel (CASS). Research conducted by NRC Office of Nuclear Regulatory 

Research on inspection of CASS demonstrated that the Appendix III UT techniques are 

inadequate when applied to cast materials. This work is discussed in Section 3.2.

(1) An earlier experiment that resulted in a loss of signal reflectivity was conducted at PNNL in 

the 1980 time frame. The experiment was performed by generating a compressive loading 

across a flaw face and measuring changes in UT flaw response. This work showed that a 

12 ksi compressive load caused a −10 dB change in flaw response (Becker et al. 1981).

4-7

 
One additional potential examination limitation in the context of MSIP involves Westinghouse 

reactor vessel nozzle welds that are inspected from the nozzle ID surface. This limitation is 

discussed in Section 4.5 in connection with Code Case N-770-1, Note (12)(e). 

4.4 Assessment of the Effectiveness of MSIP as a PWSCC Mitigation 

Strategy

Sullivan and Anderson (2013) performed an assessment of the effectiveness of MSIP as a 

PWSCC mitigation strategy based on the criteria in Appendix I to ASME Code Case N-770-1. 

Code Case N-770-1 requires that these criteria be satisfied and documented prior to being able 

to take credit for the reduced examination requirements for stress improved welds in Inspection 

Items D and E as compared the requirements for unmitigated welds in Inspection Items A-1, 

A-2, or B. These criteria are paraphrased as follows.

• The mitigation technique has to minimize the likelihood of crack initiation.

• The effect on the susceptible weld material produced by the mitigation process has to be 

permanent.

• The capability to perform UT of the required inspection volume of the component cannot be 

adversely affected by the mitigation.

• The mitigation process cannot have degraded the component or adversely affected other 

components in the system.

• The mitigated weld has to be inspectable by a qualified process.

• Existing flaws, if any, have to be addressed as part of the mitigation.

• The effect of the mitigation on any existing flaws has to be analyzed.

These performance criteria were developed for Code Case N-770-1 by ASME Section XI. They 

provide rules for the implementation of MSIP, because the rules of IWA-4000, Repair/

Replacement Activities, do not apply to MSIP. These criteria have been approved by ASME 

and adopted by the NRC as the set of factors needed to ensure effective mitigation by stress 

improvement. Therefore, they were used by the authors of this report to assess the 

effectiveness of MSIP.

The following conclusions were reached by Sullivan and Anderson (2013) in performing this 

assessment. 

MSIP is effective in producing compressive WRS in the inner region of welds (Fredette and 

Scott 2009), thereby minimizing the likelihood of initiation and growth of PWSCC in Alloy 82/182 

DMWs. Based on creep properties of the materials involved, the operating temperature of MSIP 

welds, and analyses to show that residual stresses do not relieve during subsequent loading 

cycles, the MSIP mitigation method is expected to be permanent. Analyses are required to be 

performed on the effect of MSIPs on other components in the system and ASME Code criteria 

are required to be satisfied, which makes it reasonable to conclude that the designs for 

mitigation by MSIP will satisfactorily preclude potential adverse effects it may have on other 

components in the system.

4-8
Mockup testing and NDE qualified to Section XI, Appendix VIII, performance demonstration 

requirements would be expected to show that the capability to perform UT of the weld has not

been adversely affected by the mitigation, including possible geometric changes that may result 

from MSIP. Welds mitigated by MSIP are required to be inspectable by techniques and 

procedures qualified in accordance with Appendix VIII. An evaluation has to be performed to 

confirm whether the weld to be mitigated is within the scope of an Appendix VIII supplement and 

that the procedures to be used have been qualified. For these examinations to perform as 

intended, the NDE used has to be effective and reliable. Limitations involving incomplete 

coverage have to be reviewed by the NRC to ensure that alternative inspections satisfy 

regulatory criteria; for example, to ensure that an adequate level of quality and safety is 

provided.

Existing flaws must be shown using IWB-3600 to not become unacceptable over the life of the 

weld or before the next scheduled examination. 

Assuming that these criteria are satisfied as discussed above, it is reasonable to conclude that 

MSIP provides effective mitigation against the initiation of PWSCC and against the growth of 

any existing PWSCC that has been detected and allowed to remain in service.

4.5 Assessment of Code Case N-770-1 Inspection Requirements

An overview of Code Case N-770-1 inspection requirements is contained in Section 3.4.1. An 

assessment of each note in Table 1 of the Code Case that applies to MSIP welds is contained 

in Appendix B of Sullivan and Anderson (2013).

Inspection Item D welds are uncracked butt welds mitigated with stress improvement. 

Inspection Item E welds are cracked butt welds mitigated with stress improvement. 

Pre-MSIP examinations are required to be performed in the same outage as the MSIP is applied 

with one exception—for reactor vessel nozzle welds at cold leg temperatures that require the 

core internals to be removed to perform an examination, the volumetric examinations are not 

required prior to application of the MSIP. If the pre-MSIP volumetric examination is not 

performed, a post-MSIP preservice ID surface examination, in conjunction with the required 

post-MSIP volumetric examination, must be performed after removal of the core internals.

Examination of Alloy 82/182 reactor pressure vessel (RPV) cold leg nozzle welds in 

Westinghouse plants requires core internals to be removed to access these welds when 

examined from the ID. The exception to performing pre-MSIP examinations of cold leg

temperature RPV nozzle welds was developed because pre-MSIP examination that can only be 

performed from the ID (thus, requiring core internals removal) would negatively impact outage 

durations and cold leg weld locations are expected to be less susceptible to PWSCC than welds 

at hot leg temperatures. Per Code Case N-770-1, this exception can only be used for cold leg

nozzles that require core internals to be removed to perform a full volumetric examination from 

the ID, which means that the cold leg exception can only be used when the examination 

requirements cannot be satisfied by accessing and examining these welds from the OD surface.

4-9
However, in order to perform MSIP on these welds, access to the OD surfaces must be made 

available. If these welds have enough access for MSIP, one would logically conclude that there 

should also be enough space to conduct an examination from the OD, especially if a technology 

such as encoded phased-array UT is applied. It is the authors’ understanding that through 

proper OD surface conditioning (including weld crown smoothing or removal), volumetric 

examinations of many Westinghouse RPV cold leg nozzle welds could allow full axial and 

circumferential scan coverage. Scan coverage diagrams of the RPV cold leg nozzle welds for 

an example plant are shown in Appendix C, Figures C.20 and C.21. See also discussion in 

Section 3.2. On this basis, the RPV cold leg nozzle exception should not be used.

Post-MSIP examinations are required and are considered the preservice baseline examination. 

Per 10 CFR 50.55a(g)(6)(ii)(F)(9), Inspection Item D welds (uncracked butt welds mitigated by 

stress improvement) must all be examined no sooner than the third outage and no later than 

10 years following MSIP. Examination volumes that show no indication of cracking shall be 

placed into a population to be examined once per interval on a 25% sample basis. Inspection 

Item E welds (cracked butt welds mitigated by stress improvement) must be examined once 

during the first or second refueling outage following MSIP. Weld examination volumes that 

show no indication of crack growth or new cracking shall be placed into a population to be 

examined once per interval on a 25% sample basis.

As noted above, Code Case N-770-1 requires both uncracked and cracked MSIP welds to be 

placed into 25% inspection samples. The once-per-interval inspection frequency reflects an 

implicit conclusion by the ASME and, in imposing the Code Case, by the NRC that the MSIP 

technique results in effective mitigation and that inspection of MSIP welds serves a defense-indepth monitoring function rather than a degradation management function.

The rule imposing Code Case N-770-1 (76 FR 36232) imposed conditions (F)(1), (2), (4), (6), 

and (9) on welds mitigated by mechanical stress improvement. These conditions are in addition 

to the Code Case requirements and were imposed by the NRC to ensure that, in combination 

with the Code Case, adequate protection is provided for monitoring Alloy 82/182 welds.

Condition §50.55a(g)(6)(ii)(F)(1) of the rule requires licensees of existing operating pressurized 

water reactors (PWRs) as of July 21, 2011, to implement the requirements of ASME Code Case 

N-770-1, subject to the conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) 

by the first refueling outage after August 22, 2011. This is the basic implementing requirement 

imposing Code Case N-770-1 with conditions.

Condition §50.55a(g)(6)(ii)(F)(2) of the rule allows welds mitigated by mechanical stress 

improvement to be categorized as Inspection Items D or E, as appropriate, provided the criteria 

of Appendix I of the Code Case have been met. As noted previously, the Code Case provides 

performance criteria in Appendix I for ensuring that the basic objectives of Section XI for 

maintaining structural integrity are met. Prior to this Code Case, MSIPs were performed 

following industry-qualified procedures and guidance in NUREG-0313 (Hazelton and Koo 1988). 

MSIP had not been subject to NRC review. Accordingly, under condition (F)(2) licensees are 

not required to obtain NRC approval to categorize MSIP welds as Inspection Items D or E but, 

per Code Case N-770-1, the criteria of Appendix I must be satisfied and an evaluation 

documented in accordance with Code Case N-770-1, Note (12)(d).

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Code Case N-770-1, paragraph -2500(c) states that, “For axial and circumferential flaws, 

examination shall be performed to the maximum extent practical using qualified personnel and 

procedures. If essentially 100% coverage for circumferential flaws (100% of the susceptible 

material volume) can be achieved, the examination for axial flaws shall be completed to achieve 

the maximum coverage practical.” The NRC disagreed with this provision. Axial flaws can lead 

to through-wall cracks and leakage of reactor coolant, which is a safety concern. Condition 

§50.55a(g)(6)(ii)(F)(4) requires that the axial (flaw) examination coverage requirements 

of -2500(c) may not be considered to be satisfied unless essentially 100% coverage is

achieved. The industry guidelines of MRP-139 (EPRI 2005) allow less than essentially 100%

coverage in some cases; therefore, a number of previously conducted baseline examinations 

may not satisfy the rule. This condition was added by the NRC to ensure that, through NRC 

review of an authorization of alternative inspection coverage, appropriate actions are being 

taken to address potential inspection limitations for axial flaws. This condition applies to preand post-MSIP examinations.

Condition §50.55a(g)(6)(ii)(F)(6) is an ISI reporting requirement for mitigated welds if growth of 

existing flaws is found that exceeds the previous IWB-3600 flaw evaluations or if new flaws are 

detected. In such cases, licensees are required to provide a report to the NRC prior to entering

reactor re-start Mode 4 that summarizes the licensee’s flaw evaluation with inputs, 

methodologies, assumptions, and the cause of new flaws or flaw growth. If volumetric 

examination detects new flaws or growth of existing flaws in the required examination volume, 

the mitigation will not be performing as designed and the NRC will need to evaluate the 

licensee’s actions to address the problem. Therefore, this condition was added to verify the 

acceptability of the weld prior to being placed back in service.

Condition §50.55a(g)(6)(ii)(F)(9) pertains to scheduling and deferral of initial examinations. 

Condition (F)(9) clarifies that the first examination for Inspection Items D and E welds shall be 

performed as specified in Table 1 and not deferred.

These conditions were developed by the NRC to ensure that the level of quality and safety 

provided by the requirements for MSIP welds is consistent with that provided by existing ASME 

Code and NRC requirements for butt welds that are not susceptible to PWSCC.

Specific comments regarding ASME Code Case N-770-1, as implemented, are provided below. 

These comments are based on the assessment in Appendix B to Sullivan and Anderson (2013).

As noted above, it is the authors’ understanding that if OD weld surfaces can be adequately 

conditioned to allow full volumetric examination from the OD to be performed, the reactor vessel 

cold leg nozzle weld exception should not be used. However, if using the cold leg nozzle weld 

exception can be justified, Note (12)(e) permits cold leg RPV nozzles mitigated by MSIP to 

forego a pre-MSIP examination provided that post-MSIP preservice UT and ID surface 

examinations are performed. The surface examination shall be performed on the butt weld 

inside surface and shall consist of an eddy current examination (ET) in accordance with Section 

XI, IWA-2223. IWA-2223 requires that ET be conducted in accordance with Appendix IV. 

Appendix IV contains demonstration requirements only and allows the demonstrations to be 

performed on notches. Because the MSIP may decrease or eliminate the UT response of 

cracks, the ability of the surface examination to detect cracks takes on greater significance. 

4-11
Qualification of ET procedures and personnel by performance demonstration would increase the 

reliability of ET examination. Code Case N-773 contains ET performance demonstration 

requirements that may be used in lieu of Appendix IV, Supplement 2 when ET is used to 

complement UT performed on the inside surfaces of austenitic, DMW, and clad piping welds. 

The NRC proposed to unconditionally approve Code Case N-773 in DG-1231 and no public 

comments were received to the contrary, so it is expected that the final Regulatory Guide 1.147, 

Revision 17, scheduled to be published in fall 2014 will approve Code Case N-773. The ASME 

Code, Section XI, has initiated an action to incorporate the requirements of Code Case N-773 

into Appendix IV. This proposed Code change incorporates ASME XI Code Case N-773 and 

provides rules for the qualification of procedures and personnel for automated ET examination 

of stainless steel and stainless clad piping and is intended as a complement to the ultrasonic 

examinations performed from the inside surface of piping. At the time this report was issued, 

this proposed Code change had not yet been approved by ASME. 

Also, if using the cold leg nozzle weld exception can be justified, Code Case N-770-1 specifies 

that if no cracks are found during the post-MSIP preservice examinations of cold leg RPV 

nozzle welds, the welds will be considered uncracked and subject to the examination 

requirements of Inspection Item D. However, these provisions do not explicitly say what the 

user is required to do if, during the surface examination, the weld is determined to be cracked. 

There are currently no Appendix VIII, Supplement 10 procedures to examine welds from the ID

surface qualified for depth sizing flaws in unmitigated welds. The compressive stresses in the 

mitigated weld may cause crack faces to close or partially close and change the UT response of 

the flaws. This phenomenon may increase the difficulty of qualifying procedures for detection 

as well as depth and length sizing. If a weld were determined by the surface examination to be 

cracked, the issue may have to be resolved outside of the rules of the Code Case, prior to 

restart, and involve the NRC staff review of a relief request.

Based on a review of the requirements of Code Case N-770-1, as implemented with NRC 

conditions, it is concluded that these requirements for MSIP, including the inspection intervals,

provide an acceptable approach for monitoring potential PWSCC. It is appropriate to reiterate 

that for this monitoring to perform as expected, the NDE used has to be effective and reliable.

4.6 Conclusions

Based on information searches performed by the authors as well as discussions with a number 

of industry experts familiar with decades of operating experience with BWRs, no instances of 

leaks were identified in welds mitigated by stress improvement. However, there have been 

instances in which cracks, not identified when mitigating by MSIP, have been found during 

subsequent inspections.

Operating experience with MSIP in BWRs highlights the importance of removing weld crowns

and the benefit of performing effective and reliable examinations including, for example, the use 

of Appendix VIII-qualified, encoded phased-array UT techniques, prior to and subsequent to 

application of MSIP.

A comparison was made of the changes in flaw characteristics between pre- and post-MSIP 

data for service-induced flaws and flaws implanted in a mockup. One of the eight service4-12
induced flaws was no longer detected after MSIP; however, all six of the implanted flaws 

remained detectable after MSIP. On average, the service-induced flaw data show a decrease in 

length and SNR flaw response. The length of the implanted flaws remains basically unchanged 

and the SNR flaw response decreased only slightly. This difference in the observed responses 

was not expected.

Inspection limitations can affect the effectiveness of NDE requirements. Geometry or access 

may limit performing a complete pre-MSIP weld examination. Information on geometric 

limitations to performing inspections of typical Westinghouse and CE plant RCS Alloy 82/182 

welds is contained in Appendices C and D.

It is the authors’ understanding that if OD weld surfaces can be adequately conditioned to allow 

full volumetric examination to be performed from the OD, the exception in Code Case N-770-1 

to performing a pre-MSIP reactor vessel cold leg nozzle weld examination should not be used.

Inspection limitations exist for welds adjacent to CASS products and qualification requirements 

do not yet exist in Appendix VIII for inspection of CASS piping welds. This limitation is 

discussed in Section 3.2 for unmitigated welds and is equally applicable to welds considered for 

mitigation by MSIP. 

During the MSIP process, plastic deformation of the pipe wall adjacent to the weld (under the 

mechanical clamp) results in contouring of the inner and outer surfaces of the weld. These 

surface conditions can potentially misalign subsequent ultrasonic transducer placement causing 

examination volumes to be limited.

Assuming that the performance criteria of Code Case N-770-1, Appendix I, are satisfied and the 

NDE used is effective and reliable, it is reasonable to conclude that MSIP provides effective 

mitigation against the initiation of PWSCC and against the growth of any existing PWSCC that 

has been detected and allowed to remain in service.

Based on a review of the requirements of Code Case N-770-1, as implemented with NRC 

conditions, it is concluded that these requirements for MSIP provide an acceptable approach for 

monitoring potential PWSCC. For monitoring of potential PWSCC by MSIP to perform as 

expected, the NDE used has to be effective and reliable.

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5 MANAGEMENT OF PWSCC BY WELD OVERLAY

5.1 Operating Experience with Welds Mitigated by Weld Overlays

A literature survey was performed to obtain information on operating experience with weld 

overlays (WOLs). The Electric Power Research Institute (EPRI) report (1999) contains 

summaries of the results of boiling water reactor (BWR) WOL inspections performed through 

about 1996. The inspections summarized in EPRI (1999) after about 1985 were performed 

using procedures qualified in accordance with the NDE Coordination Plan agreed upon by the 

U.S. Nuclear Regulatory Commission (NRC), EPRI, and the BWR Owners’ Group (BWROG), 

generally referred to as the Tri-Party Agreement (DeYoung 1984). The conclusions of this 

report state that, “Volumetric examination of each weld overlay volume and a portion of the 

underlying piping material has been performed using ultrasonic techniques that have been 

qualified for detection of intergranular stress corrosion cracking (IGSCC). Each weld overlay 

has been examined multiple times, on an inspection frequency consistent with the requirements 

of NUREG-0313, Revision 2 for Category E welds. None of the examinations have revealed 

any indication of flaw propagation into the weld overlay repair or within the outer 25% of the 

component base metal. Some of the repairs [weld overlays] have been in service in excess of 

fifteen years, with no indication of weld overlay degradation.” No other similar reports were 

located summarizing collective operating experience. 

In an effort to acquire a different perspective on operating experience with weld overlays, the 

authors contacted industry representatives and laboratory staff who have spent careers working 

on piping weld degradation mechanisms, inservice inspection, and structural integrity issues

since the early 1980s. The authors were informed that there have been no instances of leaks in 

WOL welds or flaws propagating into overlays. In summary, no service-related operating 

experience involving WOLs indicating crack growth or any other issues of concern was found.

Two experiences on difficulties with the installation of weld overlays in pressurized water 

reactors (PWRs( are discussed below.

(a)

By letter to the NRC dated April 28, 2006 (Hoots 2006a), as supplemented by letters dated 

August 18, 2006 (Hoots 2006b) and September 14, 2006 (Jury 2006), Exelon submitted a 

request for relief from the ASME Code, Section XI, for Byron, Unit 1, for the third 10-year 

interval, which began June 30, 2006. The licensee requested relief from the repair/replacement 

requirements for structural weld overlays (FSWOLs) on pressurizer (PZR) spray, relief, safety, 

and surge nozzle safe ends. This relief request was motivated, in part, by the industry MRP139 guidelines for PWR licensees to complete the inspections of PZR temperature Alloy 82/182 

butt welds by December 31, 2007. This relief request was one of the early requests for NRC 

approval to apply FSWOLs to PWR butt welds. During a teleconference call between the NRC 

staff and the licensee on September 14, 2006, verbal relief was authorized for Byron, Unit No. 1,

to install the WOLs. NRC’s written authorization for the installation of these weld overlays was 

(a) The experiences described are based, in part, upon the knowledge of E. Sullivan while at 

the NRC.

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provided to the licensee in a letter dated January 29, 2007 (Marshall 2007). During the 

installation of many of the FSWOLs to the PZR butt welds at Byron, Unit 1, the licensee’s 

examinations detected fabrication flaws in the weld overlays. These flaws exceeded the 

acceptance criteria proposed by the licensee in its relief request. These fabrication flaws were 

removed and the WOLs were subsequently successfully installed. This experience resulted in 

the NRC requesting that licensees seeking relief from ASME Code requirements to install WOLs 

include a commitment to provide the results of the ultrasonic examination of the structural weld 

overlays within 14 days after the completion of the last weld overlay preservice ultrasonic 

examination.

In 2009, problems with welding fabrication flaws occurred during the installation of FSWOLs on 

each of the reactor vessel hot leg nozzle welds at Catawba, Unit 2. Several attempts were 

made to resolve the welding issues, but prior to restart the WOLs were essentially removed by 

grinding (Bartley 2009). As of the writing of this report, the licensee had not attempted to 

reinstall WOLs on the Catawba, Unit 2, reactor vessel hot leg nozzles.

An experience with laminar flaws detected in WOLs at Diablo Canyon, Unit 2 is worth noting. 

During refueling outage 2R17 in early 2013 required inservice inspection (ISI) examinations 

detected laminar indications in the dissimilar metal weld (DMW) weld overlays on the 

pressurizer A, B, and C safety nozzle and the spray nozzle to piping DMWs. These laminar 

fabrication flaws were detected in the WOLs using a qualified phased-array ultrasonic testing 

(UT) technique. The licensee did not detect these indications after WOL installation in 2008 nor 

during the follow-on ISI examinations in October 2009 when using a qualified conventional UT

technique. For the 2013 inspection results, the licensee reported that the indications in the 

WOL installed on safety nozzles A and B and spray nozzle exceeded the fabrication acceptance 

standards for examinations described in the original relief request for installing the weld overlays

(Polickoski 2013).

In lieu of removing the laminar indications, the licensee performed flaw evaluations and stress

analyses to demonstrate that the overlaid DMWs are acceptable for continued operation. The 

licensee performed the flaw evaluations of the laminar indications found in pressurizer safety 

nozzles A, B, and C, and the spray nozzle, using the requirements of ASME Code, Section XI, 

IWB-3600. The licensee also performed a stress analysis demonstrating that the interface 

length between the WOL and the nozzle/pipe base metal is sufficient in light of laminar 

indications using the rules of the ASME Code, Section III, NB-3227.2. The licensee committed 

to perform successive examinations over the next three ISI periods of safety nozzles A, B, and 

C and the spray nozzle to monitor the growth, if any, of the laminar indications in accordance 

with the ASME Code, Section XI, IWB-2420(b). The three successive examinations should 

verify the validity of the above crack growth and stress analyses and monitor the laminar 

indications. On March 8, 2013, the NRC provided verbal authorization to Diablo Canyon, Unit 2 

for the fuel cycle following the 2013 refueling outage, to use their proposed alternative in lieu of 

removing the laminar indications (Polickoski 2013). Following the phased-array UT examination 

that detected the laminar flaws, the utility formed a team to investigate the causes that resulted 

in the flaws not being identified during previous examinations. Subsequently, the industry NDE 

Integration Committee (IC) formed a focus group to evaluate the utility’s root cause report and 

determine appropriate industry actions.

5-2
The NDE IC issued NDE Alert 2014-02 dated February 10, 2014, to inform utilities and vendors 

of the inspection issue identified at Diablo Canyon, to convey industry actions, and to provide 

utilities with the NDE IC’s recommendations. This NDE Alert is not publicly available. 

Nevertheless, during the annual NRC staff/Industry meeting held to discuss materials related 

issues, industry presented the following NDE IC’s actions and recommendations during the

NDE Program Update presentation (Hacker 2014). 

The industry representative indicated that the following actions were taken based on the work of 

the NDE IC focus group.

• PDI-UT-8 was revised to provide enhancements for the areas identified in the Diablo 

Canyon root cause report. PDI-UT-8 is the generic Performance Demonstration Initiative 

(PDI) procedure for examining WOLs.

• New WOL demonstration samples were examined with both the non-encoded conventional 

and phased-array procedures to assure that no additional procedure changes are required 

to address these configurations. 

• NDE IC approved a multi-year project to develop a training module that can be used to 

prepare examiners prior to qualification and examinations at the plant. This training will 

address the operating experience and contain specific guidance and recommendations 

highlighted by the utility root cause. 

• PDI Program fabricated practice mockups representative of non-standard weld overlays. 

The NDE IC focus group recommended the use of these mockups prior to performing 

ultrasonic examinations of non-standard WOLs on site to allow the examiner to become 

familiar with the intricacies of performing examinations of these configurations. 

• The NDE IC approved a project to modify the non-standard WOL practice mockups to 

implant fabrication flaws similar to those missed at Diablo Canyon for the purpose of 

providing indications representative of known field conditions to assist in preparing the 

examiners for the examinations.

• The NDE IC approved projects to perform additional research in 2014 to evaluate 

approaches to improve the surface contact for the zero-degree examinations of small 

diameter (< 8-in.-diameter) weld overlays.

• The NDE IC focus group evaluated the Diablo Canyon root cause evaluation to determine if 

this event was the direct cause of inadequate procedure, equipment, personnel 

qualifications, or the qualification process that requires an extent of condition evaluation at 

other sites. The focus group concluded that this event is related to implementation of the 

procedure guidance and does not require an industry extent of condition evaluation.

NDE IC made the following recommendations on this event to the Industry.

• Perform a detailed review of the Diablo Canyon root cause and corrective actions taken, 

specifically the enhancements made to the examination procedures, to address nonstandard overlays, which include scan speed guidance and sensitivity adjustments.

• When applying non-encoded conventional UT examinations of WOL, implement the latest 

revision of the qualified examination procedure.

5-3
• Prior to performing examinations of non-standard overlays, utilize the practice mockups to 

prepare examiners.

This NDE event at Diablo Canyon is not discussed in Appendix B, Operating Experience with 

Butt Weld primary water stress corrosion cracking (PWSCC), because it did not involve 

PWSCC.

5.2 WOL Design Requirements

Depending on the weld geometry, fabrication practices, and the presence of a nearby safe endto-pipe weld, Alloy 82/182 piping butt welds may have axial and hoop residual tensile stresses 

within a zone near the inside surface of the weld. This tensile zone contributes to the 

susceptibility of Alloy 82/182 to PWSCC. Weld overlays convert tensile residual stresses at and 

near the inside surface of piping in the weld to a zone of compressive residual stresses or 

substantially reduced tensile residual stresses. Weld overlays are beneficial from the standpoint 

of providing additional structural reinforcement. In addition, there is additional potential benefit 

that the WOLs may provide geometries more favorable for inspection.

There are two types of weld overlays—full structural weld overlays, which have a minimum 

thickness of one-third the pipe wall thickness, and optimized weld overlays (OWOLs), which 

have less weld metal applied—in some cases having as few as three layers of applied 

weldment. Weld overlays extend in both axial directions some distance beyond the weld. Weld 

overlays extend 360° circumferentially around the pipe. For FSWOLs, no structural credit is 

taken for the original pipe thickness.

Fredette and Scott (2010) performed finite element analyses on applying WOLs to pipe welds in 

multiple systems. Initial weld residual stresses (WRSs) were assessed, including welds with 

and without an attached safe end. Weld residual stresses after application of WOLs were 

determined and flaw stress intensity calculations were performed to understand the effect of 

applying WOLs to welds containing flaws. These analyses were reviewed by Sullivan and 

Anderson (2011). The analyses confirm that compressive residual stresses at or near the inside 

surface are produced by WOLs and, in contrast to the finding for the mechanical stress 

improvement process (MSIP), confirmed that FSWOLs should not increase the stress intensity 

factor of deep preexisting flaws potentially present in the weld. 

The analyses by Fredette and Scott (2010) included sensitivity studies that were conducted to 

evaluate the effect of WOL weld sequencing on the resulting WRS in the field. Changing the 

welding sequence from left-to-right versus right-to-left weld deposition did not have a substantial 

effect on the resultant residual stresses in the weld. However, for the case where two weld 

heads start at different locations and move in the same direction, the resultant WRSs were 

larger (less favorable) than when a single head was used. This method might be used to 

complete the FSWOL in reduced time by doubling the amount of welding done in a given time. 

The analyses showed that some changes in weld sequencing can negate any claimed WRS 

benefit predicted for the WOL in the design process. In applying a FSWOL or an OWOL for a 

given geometry, it is crucial that the weld overlay fabricated in the field accurately reflects the 

design that was evaluated and approved.

5-4
Sensitivity studies on the effect of additional weld layers on the resulting WRS did not produce a 

consistent trend. The results for these sensitivity studies were presented for stresses at room 

temperature. Results for other analyses performed by Fredette and Scott (2009) show that for 

operating temperature the stresses change somewhat but follow similar trends to stresses at 

room temperature. For the surge nozzle geometry considered, the number of weld overlay 

passes had a minimal effect on the through-thickness axial stresses near the weld centerline. 

There was some effect of more layers on the inner diameter (ID) axial stresses at the 

butter/ferritic steel interface, but that effect tended to saturate after about three layers of overlay. 

However, the ID axial tension in the DMW area of the surge nozzle weld remains in tension in 

all cases and is only reduced to near zero with the application of the FSWOL thickness. For the 

hoop stress case, the effect was more pronounced with each successive layer further reducing 

the stresses. The safety nozzle and the reactor vessel cold leg nozzle results showed 

oscillation in the through-thickness axial stress with additional material. For the reactor coolant 

pump outlet nozzle, the axial stresses were also shown to oscillate. Application of layers one 

through three reduce the maximum stress, and layers five through eight raise the stress from 

the minimum achieved with layer three. OWOL actually produced better results than the 

FSWOL in the axial stresses, and the results were virtually the same for both OWOL and 

FSWOL in the hoop direction. Results for the geometries considered by Battelle Columbus 

Laboratory (BCL) indicate that the OWOL design can be an effective mitigation strategy for 

dealing with PWSCC in PWRs. However, these studies highlight the need for weld-specific 

analyses as part of the design process for defining the effect of weld sequencing and weld 

overlay thickness on the resulting WRS state.

The WOL thickness is designed based on a postulated flaw (the design-basis flaw) in the 

original pipe, DMW, or nozzle. The ASME Code, Section XI, establishes an allowable flaw size 

for the end-of-evaluation period based on the maximum flaw size that can be sustained in the 

component without exceeding the structural (safety) factors specified in the ASME Code. The 

end-of-evaluation period is the time period during which the flaw is calculated to be acceptable. 

The maximum allowable flaw size in accordance with the ASME Code, Section XI, IWB-3640, 

“Evaluation Procedures and Acceptance Criteria for Flaws in Austenitic and Ferritic Piping,” is 

75% of the component nominal wall thickness, which if applicable includes the thickness of the 

WOL. The weld overlay sizing requirements are described in MRP-169, Revision 1 (EPRI 

2008b) and are also contained in ASME Code Cases N-740-2 for FSWOLs (ASME 2008) and

Code Case N-754 for OWOLs (ASME 2011).

For the purpose of this discussion, pipe wall thickness and DMW wall thickness are the same. 

In addition, “pipe wall thickness” refers to the original pipe/weld thickness, not the thickness of 

the pipe wall plus the weld overlay thickness.

The FSWOL thickness is designed based on ASME Code, Section XI, IWB-3640 and 

Appendix C, “Evaluation of Flaws in Piping.” The design-basis flaw is a circumferential flaw 

100% through-wall and 360° around the weld circumference that is postulated to be present in 

the DMW. Once the FSWOL is applied, the design-basis flaw becomes the allowable flaw size 

for the end-of-evaluation period. The weld is not required to be inspected prior to application of 

a FSWOL. If a weld is not inspected prior to installation of a FSWOL, a post-FSWOL inspection 

must be performed to verify that the outer 25% of the original weld is free of planar flaws. If a 

pre-FSWOL inspection is not performed, then an initial flaw that is 75% through the original weld 

5-5
thickness is assumed in both the axial and circumferential directions in the design analysis for 

the service life of the weld. Any actual observed or postulated flaws in the DMW must be 

demonstrated, by a crack growth calculation, not to grow beyond the design basis flaw size 

before the next scheduled ISI.

If a pre-FSWOL inspection is conducted, an initial flaw size for the crack growth calculation is 

specified in ASME Code Case N-740-2. For example, if a qualified UT examination is 

performed prior to application of a FSWOL, and no inside surface-connected planar flaws are 

detected, initial flaws originated from the inside surface of the weldment equal to 10% of the 

original wall thickness are to be assumed in both the axial and circumferential directions in the 

design analysis for the service life of the weld.

The required ISI volume for a FSWOL includes the weld overlay thickness and the outer 25% of 

the original pipe wall as shown in Figure 2(a) in Code Case N-770-1 (ASME 2009a). Weld 

overlay sizing for FSWOLs is governed, in many cases, by the general requirements in the 

ASME Code, Section XI, IWB-3643, that no flaws of depth greater than 75% through-wall are 

acceptable. This maximum flaw depth dictates that the minimum FSWOL thickness, regardless 

of the applied loading, is one-third the thickness of the original pipe thickness, regardless of the 

applied loads. Thicknesses greater than this may be required depending upon the magnitude of 

the applied loadings and other factors, such as the resulting axial length of the overlay.

Code Cases N-754 (ASME 2011) and N-740-2 (ASME 2008) provide guidance for weld overlay 

length sizing, and these are the same for both FSWOLs and OWOLs. The underlying 

requirement is that sufficient weld overlay length be provided on either side of the observed 

crack to allow for adequate transfer of axial loads between the pipe and the weld overlay. For 

axisymmetric loading of a cylinder, local loading effects can be shown to attenuate to a small 

fraction of their peak value at an axial distance of 0.75√Rt from the point of loading (where R is 

the outer radius and t is the nominal wall thickness of the cylinder). Thus, if the weld overlay 

length is set equal to 0.75√Rt on either side of the crack, resulting in a total weld overlay length 

of 1.5√Rt, the overlay will extend beyond any locally elevated stresses due to the crack. In 

application of weld overlays pre-emptively, however, no crack will have been detected, so the 

above criterion is conservatively applied such that the minimum weld overlay length must be 

0.75√Rt beyond either side of the susceptible material. This will result in a total weld overlay 

length equal to 1.5√Rt plus the length of susceptible material (Alloy 82 or 182 weld metal and 

buttering) on the outer-diameter surface of the original DMW. The 0.75√Rt recommendation is 

only a rule of thumb, and shorter lengths may be used, if justified by stress analysis of the 

specific pre-emptive WOL configuration to demonstrate that adequate load transfer and stress 

attenuation are achieved. The length of the weld overlay can also be affected by the 

requirements to obtain full coverage of the required inspection volume.

The NRC notes in Blount (2010) that the structural integrity of the FSWOL does not take credit 

for the underlying DMW because the design-basis flaw is 100% through the original pipe wall 

and 360° around the circumference. Design requirements for FSWOLs applied over cast 

material, for which ASME Section XI, Appendix VIII, qualification requirements have not been 

written, are provided in Code Case N-740-2.

5-6
An OWOL may be used pre-emptively or for repair of DMWs with flaws of limited depth. The 

design-basis flaw for an OWOL is a circumferential flaw 75% through the original pipe wall and 

360° around the DMW circumference. The OWOL design takes credit for the structural support 

of the outer 25% wall thickness of the DMW. The design-basis flaw depth assumption of 75% of 

original wall thickness already meets the general Section XI, IWB-3643 maximum through-wall 

flaw depth requirement of 75% without an overlay. Thus, the minimum OWOL thickness is not 

controlled by this limit, as with FSWOLs, but instead is based on the actual internal pressure 

and pipe loads at the location of the DMW being overlaid and the ASME Code, Section XI 

IWB-3641-allowable flaw size requirements. In addition, Code Case N-754 requires that the 

minimum thickness of the OWOL be sufficient to reduce residual stresses to less than 69 MPa 

(10 ksi) tensile at operating temperature and pressure on the internal wetted surface of all stress 

corrosion cracking-susceptible materials.

The OWOL design requires that the pipe weld be inspected immediately prior to the overlay 

application, using an inspection technique qualified in accordance with ASME Section XI, 

Appendix VIII and found to exhibit no evidence of circumferential or axial cracking greater than 

50% of the wall thickness in the original weld. If a detected flaw has a depth greater than 50% 

through-wall, the OWOL design will not be applicable for the repair. A discussion of uncertainty 

for Appendix VIII piping examinations conducted from the inside surface is contained in 

Section 3.3.

The required inspection volume for OWOLs includes the OWOL thickness and the outer 50% of 

the original pipe wall as shown in Figure 5(a) in Code Case N-770-1 (ASME 2009a). Any actual 

observed flaws in the DMW must be demonstrated, by a crack growth calculation, not to grow 

beyond the allowable size before the next scheduled ISI. A qualified UT examination is 

performed prior to application of an OWOL. If no inside surface-connected planar flaws are 

detected, initial flaws originated from the inside surface of the weldment equal to 10% of the

original wall thickness are to be assumed in both the axial and circumferential directions in the 

design analysis for the service life of the OWOL. In cases where no inside-connected planar 

flaws are detected, the OWOL is considered preemptive. Specific design and inspection 

requirements for OWOLs applied over cast material, for which ASME Section XI, Appendix VIII, 

qualification requirements have not been written, are provided in Code Case N-754 (ASME 

2011). Post-overlay baseline UT and ISI examinations are required to verify the integrity of the 

applied OWOLs. Because OWOLs take credit for the underlying 25% wall thickness of the 

DMW material, the design may be required to account for the potentially lower toughness of the 

DMW material (particularly at the fusion line with the low-alloy or carbon steel nozzle) (EPRI 

2008b).

Additional details on design and evaluation of WOLs are provided in MRP-169, Rev. 1 (EPRI 

2008b) and the NRC’s evaluation of MRP-169, Rev. 1 (Blount 2010).

5.3 NDE Reliability

To evaluate the reliability of ultrasonic inspections, Pacific Northwest National Laboratory 

(PNNL) considered flaws evaluated both before and after WOL application. The data used to 

make pre- and post-WOL UT comparisons came from multiple sources: UT data on unmitigated 

welds presented during a workshop held by PNNL in May 2012 and two specimens with

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implanted thermal fatigue cracks (TFCs). Section 3.1 summarizes the findings of the PNNL UT 

workshop on unmitigated welds, which is relevant to the reliability of pre-WOL inspections, 

required for OWOL but optional for FSWOL. Section 5.3.1 summarizes the results of an 

investigation based on pre- and post-WOL examination data acquired from two specimens 

containing TFCs. Section 5.3.2 provides a discussion on potential limitations to conducting 

pre- and post-WOL inspections.

5.3.1 Inspection of Dissimilar Metal Weld Overlay Specimens

Industry has been installing weld overlays with Alloy 52/152 as a means of corrective or preemptive action for PWSCC. The reliability of NDE of welds mitigated by overlays is important 

for determining whether any new flaws have initiated or existing flaws have propagated, and for 

assessing the effect of any potential flaws on the service life of the weld. 

The reliability of NDE is relevant to welds mitigated by weld overlays for several reasons. Pipe 

welds to be mitigated by optimized weld overlays are required to be inspected immediately prior 

to the overlay application, using an inspection technique qualified in accordance with ASME 

Section XI, Appendix VIII, and found to exhibit no evidence of cracking greater than 50% of the 

wall thickness in the original weld. Undetected or mischaracterized flaws could compromise the 

structural factors required for OWOLs. 

The reliability of NDE of post-WOL welds is also important for determining whether any new 

flaws have initiated or existing flaws have propagated, and for assessing the effect of any 

potential flaws on the service life of the weld. PNNL performed an ultrasonic NDE study to 

investigate the inspectability of weld overlay specimens. The Technical Evaluation Report that 

summarizes this work is Ultrasonic Evaluation of Two Dissimilar Metal Weld Overlay 

Specimens, PNNL-21502 (Crawford et al. 2012). To conduct this study, thermal fatigue cracks 

were implanted in two DMW specimens. The specimens were composed of a cast austenitic 

stainless steel (CASS) pipe welded to a carbon steel nozzle with Alloy 82/182 weld material. 

Each specimen was evaluated with phased-array ultrasonic testing to determine flaw detection 

and characterization capabilities. An Alloy 82/182 WOL was added to the specimens, and they 

were again evaluated ultrasonically for flaw detection and characterization. A WOL made of 

Alloy 52/152 was cost-prohibitive, so Alloy 82/182 was selected instead; this choice does not 

impact ultrasonic propagation as it is essentially equivalent in both alloys. Flaw characterization 

included length and depth measurements as well as a signal-to-noise determination. Objectives 

of this work included determining ultrasonic detection capabilities in DMW specimens with a 

WOL and assessing ultrasonic examinations through CASS material. 

Specimen 10C-011 is a 516 Grade 70 nozzle and Alloy 82/182 butter section welded to a CASS 

pipe. Five circumferentially-oriented ID-connected TFCs were implanted in the center of the 

weld during the welding process to prevent implantation artifacts that could show in the 

ultrasonic data. The flaw depths range from 49.8% to 89.9% (49.8%, 59.7%, 69.5%, 79.7%, 

and 89.9%) through the original weld thickness, are tilted by 2 or 3 degrees, and are all 

circumferentially oriented. The flaw lengths vary from 3.8 to 6.4 cm (1.5 to 2.5 in.). This DMW 

mockup specimen is approximately 32.4 cm (12.8 in.) in diameter at the flaw position, and 

68.4 cm (26.9 in.) in length, with a nominal 3.3 cm (1.3 in.) wall-thickness at the flaw location

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within the DMW. The weld overlay is approximately 2.0 cm (0.8 in.) thick at the weld location. 

The flaws were easily detected with ultrasonic techniques similar to those qualified to ASME 

Section XI, Appendix VIII.

PNNL next evaluated specimen 9C-034 which has shallower TFCs implanted in the weld-tobutter region. Specimen 9C-034 contains a DMW between a 516 Grade 70 carbon steel nozzle 

and a CASS pipe. The nozzle side butter and weld consisted of Alloy 82/182 material. Four 

TFCs were implanted in the weld-to-butter region with flaw depths ranging from 13% to 31% 

(13%, 19%, 26%, and 31%) deep through the original weld thickness. The flaws are ID surfacebreaking with lengths at the ID in the range of 5.1 to 6.4 cm (2 to 2.5 in.). The flaws are vertical, 

have no tilt, and are circumferentially oriented. The individual section of CASS pipe was sent to 

FlawTech where it was welded to a carbon steel nozzle segment and the four TFCs were 

implanted into the butter region on the nozzle side of the welded specimen, reducing potential 

excavation and implantation artifacts. This DMW mockup specimen is approximately 32.5 cm 

(12.8 in.) in diameter at the flaw position and 58.4 cm (23 in.) in length, with a nominal 3.3 cm 

(1.3 in.) wall-thickness at the flaw location within the DMW. The four TFCs were inserted via a 

coupon-implant technique and were all circumferentially-oriented flaws. After the specimen was 

examined by UT, it was overlaid with Alloy 82/182. The weld overlay is approximately 2.0 cm 

(0.8 in.) thick at the weld location. 

The thickness requirements for weld overlays are discussed in MRP-169 (EPRI 2008b). The 

minimum WOL thickness requirement for a FSWOL is one-third of the weld thickness. Based 

on the nominal weld thickness for Specimens 9C-034 and 10C-011, it is clear that the 2.0-cm

(0.8-in.)-thick overlay represents a type of FSWOL. However, PNNL evaluated the length and 

depth sizing measurement errors of the flaws in Specimen 10C-011 as though it were an OWOL

specimen. This approach is conservative because the sound path to the three flaws between 

approximately 50% (49.8%) and 75% through the original wall is longer than would be the case 

in an actual OWOL specimen.

A weld root signal was often ultrasonically detected and may have precluded a strong corner 

response from the flaw itself. The presence of a strong weld root signal that could not be 

separated from the flaw signal complicated flaw length sizing.

The specimens were examined using four phased-array (PA) probes: a 2.0-MHz transmitreceive longitudinal (TRL) probe arrangement, a 1.5-MHz TRL probe, a 1.0-MHz TRL probe, 

and a 0.8-MHz TRL probe. The 1.5-MHz transducer was chosen based on frequencies 

commonly used in reactor inspections of stainless steel (SS). The additional 0.8-, 1.0-, and 

2.0-MHz probes were employed to more fully evaluate the entire frequency spectrum and 

provide data to better assess lower and higher frequency limits for inspecting the various DMW 

components. PA data were acquired on the nine flaws in the two specimens. In general, line 

scan and raster data were acquired from both sides of the flaw.

Supplement 11, Qualification Requirements for Full Structural Overlaid Wrought Austenitic 

Piping Welds, states that, “For flaws in base metal grading units (as opposed to overlay 

fabrication flaws), the candidate shall estimate the length of that part of the flaw that is in the 

outer 25% of the base metal thickness.” In Code Case N-653-1 (ASME 2012a) flaw length 

measurement is based on the part of the flaw that is in the examination volume. Extending the 

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same approach to OWOLs, the examination volume includes the OWOL thickness plus the 

outer 50% of the base metal. At the time this report was issued, Code Case N-653-1 was not 

approved in U.S. NRC Regulatory Guide 1.147. 

The four flaws in post-WOL Specimen 9C-034 with flaw depths ranging from 13% to 31% deep 

through the original weld thickness would not be included in the examination volume of either an 

OWOL or a FSWOL. Length-sizing measurements on the flaws in Specimens 9C-034 and 

10C-011 were made for the ID flaw length rather than for the part of the flaw that is in the outer 

25% or 50% of the base metal thickness. The reason for this approach is that the flaw profiles 

were not known, making it impossible to calculate root-mean-square error (RMSE) for length 

measurements at these depths. It is not clear why the Code requires length measurements of 

the part of the flaw that is in the outer 25% or 50% of the base metal thickness and comparison 

of these RMSE values with an acceptance standard because these length values appear to 

have no application to flaw evaluation rules. 

Summarizing the results of the study, two DMW specimens, one with four shallow 

circumferential flaws (13% to 31% through-wall) and one with five deeper circumferential flaws 

(50% to 90% through-wall) were ultrasonically examined. The DMW specimens were examined 

from the nozzle and the CASS pipe side with both raster and line scans using four PA probes in 

the 0.8- to 2.0-MHz frequency range, although not every combination of scan, scan direction, 

and frequency was used to examine each flaw. An Alloy 82/182 WOL was added to each 

specimen and they were re-evaluated ultrasonically for flaw detection and characterization. All 

nine flaws were detected in the pre- and post-WOL condition.

For purposes of this report, sizing errors were only considered for the five flaws that were 

greater than or equal to about 50% (49.8%) of the original material, because these flaws are 

contained within the required inspection volume for OWOLs. Using data from a second analyst 

who emphasized time-of-flight tip-diffraction sizing when these tip signals could be observed, 

pre- and post-WOL depth sizing RMSE for both the raster and line scan data were found to be 

within the ASME depth-sizing allowable of 3.2 mm (0.125 in.). Length-sizing error for these 

same flaws was calculated based on their ID surface-breaking length. The pre- and post-WOL

length-sizing RMSE values were found to be within the ASME qualification acceptance criterion 

of 19.1 mm (0.75 in.).

5.3.2 Inspection Limitations

Inspection limitations can affect the effectiveness of NDE requirements. A number of apparent 

inspection limitations are discussed in the following paragraphs. 

Code Case N-770-1, Note (12) requires that stress improvement techniques meet the 

performance criteria in Appendix I of this Code Case, and requires that satisfaction of these 

criteria be demonstrated in an evaluation. In Code Case N-770-1, stress improvement 

techniques include OWOLs.

The criteria of Code Case N-770-1, Appendix I, paragraph 5 require that, “An evaluation ... be 

performed to confirm that the required examination volume of the mitigated configuration is 

within the scope of an Appendix VIII supplement, or supplements, and that the examination 

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procedures to be used have been qualified in accordance with the ASME Code, Section XI, 

Appendix VIII.” Code Case N-770-1, Appendix I, paragraph 6, requires that, “An examination 

qualified to Section XI, Appendix VIII performance demonstration requirements shall have been 

performed ... before the application of the mitigation process to identify and size any existing 

flaws.” Any departure from these requirements of Code Case N-770-1, Appendix I, because of 

pre- or post-WOL inspection limitations beyond those permitted by Code Case N-770-1, as 

imposed by §50.55a(g)(6)(ii)(F), have to be evaluated and an alternative examination authorized 

by the NRC staff.

Geometry or access may limit performing a complete pre-WOL weld examination. 

Appendices C and D contain weld coverage assessment diagrams for typical Westinghouse and 

Combustion Engineering (CE) plant reactor coolant system (RCS) Alloy 82/182 welds that show 

the probe angles used and percent coverage obtained with Section XI, Appendix VIII-qualified 

techniques. These appendices also indicate the materials of construction for the welds 

depicted. As noted above, qualified UT examinations are required prior to installing optimized 

weld overlays.

For the plant-specific designs used to prepare Appendices C and D, Tables 3.1 and 3.2 provide 

the Code-required volume (CRV) coverage achievable for Alloy 82/182 RCS butt welds in the 

typical Westinghouse and Combustion Engineering plants used for this assessment. These 

tables indicate the welds for which Code-required inspection coverage would not be achieved 

based on the plant-specific designs. The Westinghouse reactor pressure vessel (RPV) nozzle 

welds in this study were examined from the OD, although the RPV nozzle welds in many 

Westinghouse plants are examined from the inside surface, where the inspection limitations 

differ and the CRV coverage obtained is generally understood to be higher. 

In CE plants, the safe ends of medium- and large-bore piping are typically made of CASS. The 

CRV coverage presented in Table 3.2 takes credit for the scan coverage in the CASS safe ends 

using the Appendix VIII procedures applied for the carbon steel and Alloy 82/182 volumes. This 

approach for reporting CRV coverage was used by this vendor because Code Case N-770-1,

Paragraph-2500(b) requires that Appendix VIII procedures be used to examine the Coderequired volume, including 100% of the susceptible material volume, to the maximum extent

practical. Inspection limitations exist for welds adjacent to CASS products. Because Appendix 

VIII, Supplement 9 requirements for examination of CASS piping welds have yet to be 

developed, there are no performance demonstration qualifications being performed for 

inspection of cast materials. Subparagraph-2500(b) states that, “For cast stainless steel items 

for which no supplement is available in Appendix VIII, the required examination volume shall be 

examined by Appendix VIII procedures to the maximum extent practical including 100% of the 

susceptible material volume (non-stainless steel volume).” Inspection limitations, such as an 

inability to examine 100% of the susceptible material volume, would be reviewed as part of the 

NRC process for authorizing the installation of WOLs.

ASME Section XI prepared Code Case N-824, entitled, “Alternate Requirements for Ultrasonic 

Examination of Cast Austenitic Piping Welds from the Outer Diameter Surface” (ASME 2012b). 

The requirements in this Code Case are deterministic rather than performance-based. 

However, this Code Case is considered to be a first step in the development of Supplement 9. 

Substantial improvement in NDE reliability would be expected by applying this Code Case, with 

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certain conditions based on NRC research, in situations where 100% of the susceptible material 

cannot be examined from the ferritic side of the weld. 

Research conducted by the NRC Office of Nuclear Regulatory Research on inspection of CASS

demonstrated that current Appendix III UT techniques are inadequate when applied to cast 

materials. However, some of the techniques used during Appendix VIII qualifications may 

perform well on CASS, especially for thinner-walled CASS. NRC research has shown that, for 

piping less than approximately 4 to 5.1 cm (1.6 to 2.0 in.) in thickness, refracted longitudinal 

waves (transmit-receive L-waves) operating at frequencies between 0.8 and 2.0 MHz, work 

fairly well. Much lower frequency ranges, on the order of 0.4 to 1.0 MHz, are needed for 

primary system CASS piping (greater than 5.1 cm [2.0 in.] in thickness). To date using these 

frequencies, inspections can reliably detect flaws in the 25%–30% through-wall range. 

Additional information is contained in NUREG/CR-6933, Assessment of Crack Detection in 

Heavy-Walled Cast Stainless Steel Piping Welds Using Low-Frequency Ultrasonic Methods

(Anderson et al. 2007). NUREG/CR-7122, An Evaluation of Ultrasonic Phased Array Testing for 

Cast Austenitic Stainless Steel Pressurizer Surge Line Piping Welds (Diaz et al. 2011), provides 

information on crack detection with UT in thinner-walled CASS. The information in these 

NUREGs should be helpful to industry in improving the reliability of inspections and to the NRC 

Office of Nuclear Regulatory Research in evaluating relief requests and inspection-related 

issues involving CASS.

5.4 Assessment of the Effectiveness of Weld Overlays as a PWSCC

Mitigation Strategy

Sullivan and Anderson (2012) performed an assessment of the effectiveness of WOLs as a 

PWSCC mitigation strategy based on the criteria in Appendix I to ASME Code Case N-770-1. 

These performance criteria were developed for Code Case N-770-1 by ASME Section XI. They 

provide rules for the implementation of stress improvement. These criteria were originally 

written for MSIP, because the rules of IWA-4000, Repair/Replacement Activities, do not apply to 

MSIP. However, Note (12)(d) of Code Case N-770-1 requires that these criteria be satisfied 

and documented for stress improved welds in Inspection Items D and E, which includes welds 

mitigated by OWOLs. These criteria have been approved by ASME and adopted by the NRC 

as the set of factors needed to ensure effective mitigation by stress improvement. Therefore, 

they were used by the authors of this report to assess the effectiveness of OWOLs and 

FSWOLs. 

These criteria are paraphrased as follows.

• The mitigation technique has to minimize the likelihood of crack initiation.

• The effect on the susceptible weld material produced by the mitigation process has to be 

permanent.

• The capability to perform UT of the required inspection volume of the component cannot be 

adversely affected by the mitigation.

• The mitigation process cannot have degraded the component or adversely affected other 

components in the system.

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• The mitigated weld has to be inspectable by a qualified process.

• Existing flaws, if any, have to be addressed as part of the mitigation.

• The effect of the mitigation on any existing flaws has to be analyzed.

The following conclusions were reached by Sullivan and Anderson (2012) in performing this 

assessment. 

It is clear that WOLs can produce significant compression in the inner regions of the pipe wall 

and result in weld residual stresses at the ID that have compressive or substantially reduced 

tensile stresses. However, as discussed in Section 5.2, analyses performed by BCL showed 

that some changes in weld sequencing can negate any claimed weld residual stress benefit 

predicted for the weld overlay in the design process. In applying a full structural weld overlay or 

an optimized weld overlay for a given geometry, it is crucial that the WOL fabricated in the field 

accurately reflects the design that was evaluated and approved.

Compressive or substantially reduced tensile stresses minimize the possibility that flaws would 

initiate in an overlaid weld. Based on creep properties of the materials involved, the operating 

temperature of WOLs, and analyses BCL performed showing that re-yielding does not occur on 

subsequent loading, the WOL mitigation method is expected to be permanent. Code Cases 

N-740 (FSWOLs) and N-754 (OWOLs) require analyses to be performed on the effect of WOLs 

on other components in the system and ASME Code criteria are required to be satisfied. 

Although these Cases are not approved by Regulatory Guide 1.147, Revision 16 (NRC 2010a), 

many of the provisions of these Cases, such as the one just discussed, are relied upon in 

requests for NRC authorization to install WOLs. Weld overlays are required by Code Case 

N-770-1 to be examined by qualified Appendix VIII procedures and WOLs tend to provide an 

improved platform for performing examinations. Any flaw detected must satisfy the structural 

factors required by the ASME Code, Section XI, under all postulated loading conditions. 

Limitations that prevent achieving the inspection coverage required by Code Case N-770-1, as 

incorporated by reference in 10 CFR 50.55a, have to be reviewed by the NRC to ensure that 

alternative inspections satisfy regulatory criteria. Nevertheless, even if flaws were not detected 

and propagate completely through the original weld thickness, WOLs provide reinforcement 

sufficient to maintain structural integrity under design-basis loading. Under these conditions,

OWOLs satisfy safety factors required by Standard Review Plan 3.6.3 for leak before break, 

although not necessarily the factors required by the ASME Code. On this basis, it is reasonable 

to conclude that WOLs provide effective mitigation against the initiation of PWSCC and against 

the growth of any existing PWSCC that has been detected and allowed to remain in service.

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5.5 Status of ASME Code Cases N-740-X, “Full Structural Dissimilar 

Metal Weld Overlay for Repair or Mitigation of Class 1, 2, and 3 

Items, Section XI, Division 1,” and N-754, “Optimized Structural 

Dissimilar Metal Weld Overlay for Mitigation of PWR Class 1, 

Section XI, Division 1

Regulatory Guide 1.193, Rev. 3, ASME Code Cases Not Approved for Use, was issued in 

October 2010 (NRC 2010b). This guide includes ASME Code Case N-740-1, “Dissimilar Metal 

Weld Overlay for Repair of Class 1, 2, and 3 Items, Section XI, Division 1.” The guide states 

that Code Case N-740-1 is unacceptable because of the total number of issues and the nature 

of some of the issues (e.g., lack of certain fundamental design details). The guide further states 

that the “staff believes that it would be inappropriate to attempt to conditionally approve either 

version 0 or 1 in Regulatory Guide 1.147. The staff will consider Revision 2 for approval when it 

is published by the ASME.”

Code Case N-740-2 was approved in November 2008 and issued in Supplement 7 to the 2007 

Edition of the ASME Boiler and Pressure Vessel Code. ASME stated that this revision was 

made to address previous NRC negatives as well as work done on MRP-169. This revision was 

also made to add requirements for fatigue crack growth and take out all preservice and 

inservice NDE for PWRs, because these requirements are contained in Code Case N-770-1. 

NRC listed Code Case N-740-2 in DG-1233 (Draft Regulatory Guide 1.193, Rev. 4), ASME 

Code Cases Not Approved for Use, which was issued in June 2013. The DG states that “(w)hile 

Revision 2 addresses some of the NRC staff concerns, significant issues remain. For example, 

the definition of nominal weld and base material appear to be inconsistent with the provisions of 

Section III. Also, additional detail is required on how to perform the flaw growth or design 

analysis. Finally, additional detail is required on how the overlays are designed.” 

Code Case N-754, “Optimized Structural Dissimilar Metal Weld Overlay for Mitigation of PWR 

Class I Items, Section XI, Division 1,” was approved in June 2011 and issued in Supplement 6 

to the 2010 Edition. Because of when the Code Case Regulatory Guides were published, at the 

time this report was issued, Code Case N-754 had not been addressed in either Regulatory 

Guide 1.147 or the draft Regulatory Guide on inservice inspection code case acceptability.

5.6 Assessment of ASME Code Case N-770-1 Inspection 

Requirements

The NRC incorporated ASME Code Case N-770-1 by reference into 10 CFR 50.55a(g)(6)(ii)(F) 

(76 FR 36232) in June 2011. Code Case N-770-1 contains requirements for inspection of 

unmitigated as well as Alloy 82/182 RCS butt welds mitigated by certain techniques. 

An overview of Code Case N-770-1 inspection requirements is contained in Section 3.4.1. An 

assessment of each note in Table 1 of the Code Case that applies to WOL welds is contained in 

Appendix B of Sullivan and Anderson (2012). 

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Code Case N-770-1, Inspection Item D welds are uncracked butt welds mitigated with stress 

improvement, which includes stress improvement by welding of Alloy 52 OWOLs. Inspection 

Item E welds includes cracked butt welds mitigated by OWOL.

Pre-OWOL examinations are required to be performed in the same outage as the OWOL is 

applied with one exception—for reactor vessel nozzle welds at cold leg temperatures that 

require the core internals to be removed to perform an examination, the volumetric 

examinations are not required prior to application of the OWOL. If the pre-OWOL volumetric 

examination is not performed, a post-OWOL preservice ID surface examination, in conjunction 

with the required post-OWOL volumetric examination, must be performed after removal of the 

core internals. If these examinations do not detect cracks, the OWOL will be considered 

uncracked and be subject to the examination requirements of Inspection Item D. If these 

examinations detect cracks, the OWOL will be considered cracked and subject to the 

examination requirements of Inspection Item E. Additional flaw evaluation requirements apply if 

cracking is detected.

Examination of Alloy 82/182 RPV cold leg nozzle welds in Westinghouse plants requires core 

internals to be removed to access these welds when examined from the ID. The exception to 

performing pre-OWOL examinations of cold leg temperature RPV nozzle welds was developed 

because pre-OWOL examination that can only be performed from the ID (thus, requiring core 

internals removal) would negatively impact outage durations and cold leg weld locations are 

expected to be less susceptible to PWSCC than welds at hot leg temperatures. Per Code Case 

N-770-1, this exception can only be used for cold leg nozzles that require core internals to be 

removed to perform a full volumetric examination from the ID, which means that the cold leg

exception can only be used when the examination requirements cannot be satisfied by 

accessing and examining these welds from the OD surface. 

However, in order to perform OWOL on these welds, access to the OD surfaces must be made 

available. If these welds have enough access for OWOL, one would logically conclude that 

there should also be enough space to conduct an examination from the OD, especially if a 

technology such as encoded phased-array UT is applied. It is the authors’ understanding that 

through proper OD surface conditioning (including weld crown smoothing or removal),

volumetric examinations of many Westinghouse RPV cold leg nozzle welds could allow full axial 

and circumferential scan coverage. Scan coverage diagrams of the RPV cold leg nozzle welds 

for an example plant are shown in Appendix C, Figures C.20 and C.21. See also discussion in 

Section 3.2. On this basis, the RPV cold leg nozzle exception should not be used.

Post-OWOL examinations are required and are considered the preservice baseline 

examination. Per 10 CFR 50.55a(g)(6)(ii)(F)(9) Inspection Item D welds must all be examined 

no sooner than the third outage and no later than 10 years following application of the OWOL. 

Inspection Item E welds must be examined once during the first or second refueling outage 

following OWOL. 

Per 10 CFR 50.55a(g)(6)(ii)(F)(8) OWOL welds are not permitted to be placed into a population 

to be examined on a sample basis and must be examined once each inspection interval.

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Inspection Item C welds are uncracked welds reinforced by a FSWOL of Alloy 52/152 material. 

Welds inspected prior to application of a FSWOL and found to be uncracked may be placed in 

Inspection Item C. Inspection Item F welds are cracked butt welds reinforced by FSWOL of 

Alloy 52/152 material. If volumetric examination is not performed on the weld prior to FSWOL, 

the weld shall be assumed cracked and shall be classified Inspection Item F. If cracking is not 

observed during post-FSWOL preservice volumetric examination performed from the outside 

surface of the overlay, axial and circumferential cracks at least 75% through the original wall 

thickness are required to be assumed.

For reactor vessel nozzle welds at cold leg temperatures requiring the core internals to be 

removed to perform the examination, the volumetric examinations are not required prior to 

application of the FSWOL. If the pre-FSWOL volumetric examination is not performed, a postweld overlay preservice examination consisting of a surface examination and a volumetric 

examination shall be performed after removal of the core internals. If these examinations do not 

detect cracks, the weld shall be considered uncracked and shall be subject to the examination 

requirements of Inspection Item C. Post-overlay examinations are required and are considered 

the preservice examination. If these examinations do detect cracks, the FSWOL will be 

considered cracked and subject to the examination requirements of Inspection Item F. 

Additional requirements related to evaluating the crack apply if cracking is detected.

Inspection Item C and F welds are required to be placed into a population to be examined on a 

25% sample basis.

As noted, Code Case N-770-1 permits both uncracked and cracked FSWOLs to be placed into 

25% inspection samples. The once-per-interval inspection frequency reflects an implicit 

conclusion by the ASME and, in imposing the Code Case, by the NRC that the FSWOL 

technique results in effective mitigation and that inspection of FSWOLs serves a defense-indepth monitoring function rather than a degradation management function.

The rule imposing Code Case N-770-1 (76 FR 36232) imposed conditions (F)(1), (2), (4), (6), 

(8), (9), and (10) on welds mitigated by WOL. The NRC imposed these conditions, which are in 

addition to the Code Case requirements, to ensure that, in combination with the Code Case, 

adequate protection is provided for monitoring Alloy 82/182 welds.

Condition §50.55a(g)(6)(ii)(F)(1) of the rule requires licensees of existing operating PWRs as of 

July 21, 2011, to implement the requirements of ASME Code Case N-770-1, subject to the 

conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) by the first refueling 

outage after August 22, 2011. This is the basic implementing requirement imposing Code Case 

N-770-1 with conditions.

Condition §50.55a(g)(6)(ii)(F)(2) of the rule allows existing FSWOLs authorized by the NRC staff

to be categorized as Inspection Items C or F, as appropriate. At the time the rule was issued, 

only one plant had authorization to install OWOLs. The OWOLs at this plant could not be 

categorized as Code Case N-770-1 Inspection Item D or E without specific authorization by the 

NRC. Categorization of future FSWOLs as Code Case Inspection Items C or F and future 

OWOLs as Code Case N-770-1 Inspection Items D or E has to be specifically authorized by the

NRC, generally, as part of NRC’s authorization to install the WOLs. Plant-specific authorization 

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from the NRC to install FSWOLs or OWOLs and categorize the WOLs as Inspection Items C, D, 

E, and F, as appropriate, will not be needed if ASME Code Cases N-740 and N-754, or a 

suitable revision of these Code Cases, are endorsed in a future revision of Regulatory Guide 

1.147 with conditions, if applicable, and incorporated in §50.55a. This condition is intended to 

ensure that WOL mitigations are designed, installed, and examined in a manner that will ensure 

an acceptable level of quality and safety before credit can be taken for the inspection 

frequencies of Inspection Items C, D, E, and F, as appropriate.

Code Case N-770-1, paragraph -2500(c) states that, “For axial and circumferential flaws, 

examination shall be performed to the maximum extent practical using qualified personnel and 

procedures. If essentially 100% coverage for circumferential flaws (100% of the susceptible 

material volume) can be achieved, the examination for axial flaws shall be completed to achieve 

the maximum coverage practical.” The NRC disagreed with this provision. Axial flaws can lead 

to through-wall cracks and leakage of reactor coolant, which is a safety concern. Condition 

§50.55a(g)(6)(ii)(F)(4) requires that the axial (flaw) examination coverage requirements 

of -2500(c) may not be considered to be satisfied unless essentially 100% coverage is

achieved. This condition was added for the NRC to ensure that, through NRC review of an 

authorization of alternative inspection coverage, appropriate actions are being taken to address 

potential inspection limitations for axial flaws. The industry guidelines of MRP-139 (EPRI 2005)

allow less than essentially 100% coverage in some cases. This condition of the rule applies to 

the examination of all inspection items, including C and F (for FSWOLs) and D and E (for 

OWOLs).

Condition §50.55a(g)(6)(ii)(F)(6) is an inservice inspection reporting requirement for mitigated 

welds if growth of existing flaws is found that exceeds the previous IWB-3600 flaw evaluations

or if new flaws are detected. In such cases, licensees are required to provide a report to the 

NRC, prior to entering Mode 4, which summarizes the licensee’s flaw evaluation with inputs, 

methodologies, assumptions, and cause of the new flaw or flaw growth. This condition applies 

to FSWOLs and OWOLs as well as welds mitigated by other techniques. If volumetric 

examination detects new flaws or growth of existing flaws in the required examination volume, 

the mitigation will not be performing as designed and the NRC will need to evaluate the 

licensee’s actions to address the problem. Therefore, this condition was added to verify the 

acceptability of the weld prior to being placed back in service.

Condition §50.55a(g)(6)(ii)(F)(8) states that welds mitigated by OWOLs in Inspection Items D 

and E are not permitted to be placed into a population to be examined on a sample basis, as 

permitted by the Code Case, and must be examined once each inspection interval. It was the 

staff’s view that if significant cracking were to occur in the Alloy 82/182 weld material, the more 

crack-resistant Alloy 52/152 optimized weld overlay material would substantially diminish or 

inhibit crack growth when the crack encounters the overlay. The postulation of a long flaw that 

is completely through the original weld is beyond the OWOL design basis, but it has been 

shown that OWOLs maintain structural integrity for such a flaw under design-basis loading,

although the structural factors may be somewhat less than required by the ASME Code (Blount 

2010). The NRC staff evaluated structural factors for this postulated flaw during its review of 

MRP-169. The NRC staff was concerned that an OWOL with a weld cracked to this degree 

could prevent the weld from leaking and could potentially rupture without prior evidence of

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leakage under design-basis conditions. The NRC considered Condition (F)(8) to be necessary 

to ensure that all optimized weld overlays are periodically inspected for potential degradation.

Condition §50.55a(g)(6)(ii)(F)(9) pertains to scheduling and deferral of initial examinations. The 

scheduling of initial examinations is discussed for Inspection Item D and E welds above. With 

respect to deferral of the initial examination, Condition (F)(9) clarifies that the first examination 

for Inspection Items D, E, and F welds shall be performed as specified. 

Condition §50.55a(g)(6)(ii)(F)(10) pertains to the alternative examination volume for OWOLs 

discussed in Note (b) of Figure 5(a) of the Code Case. This condition states that this alternative 

examination volume may not be applied unless NRC approval is authorized under 

§50.55a(a)(3)(i) or (ii). This condition was applied because the alternative examination volume 

of Note (b) of Figure 5(a) was not provided for public comment in the proposed rule (75 FR 

24324).

These conditions were developed by the NRC to ensure that the level of quality and safety 

provided by the requirements for WOLs is consistent with that provided by existing ASME Code 

and NRC requirements for butt welds that are not susceptible to PWSCC.

Based on a review of the requirements of Code Case N-770-1, as implemented with NRC 

conditions, it is concluded that these requirements for WOLs, including the inspection intervals,

provide an acceptable approach for monitoring potential PWSCC. It is appropriate to reiterate 

that for monitoring to perform as expected, the NDE used has to be effective and reliable.

5.7 Conclusions

Based on a survey of operating experience with WOLs, the authors did not find any instances of 

leaks in WOL welds or flaws propagating into overlays. No service-related operating experience 

involving WOLs indicating crack growth was found.

The reliability of NDE is relevant to welds mitigated by weld overlays for several reasons. Pipe 

welds to be mitigated by optimized weld overlays are required to be inspected immediately prior 

to the overlay application, using an inspection technique qualified in accordance with ASME 

Section XI, Appendix VIII, and found to exhibit no evidence of cracking greater than 50% of the 

wall thickness in the original weld. Undetected or mischaracterized flaws could compromise the 

structural factors required for OWOLs. Reliable NDE is also of interest for pipe welds to be 

modified by FSWOLs. If a pre-FSWOL inspection is performed, the initial flaw size for the crack 

growth calculation is based on the results of the pre-FSWOL inspection. The calculated life of 

the FSWOL depends, in part, upon the reliability of the NDE performed. 

The reliability of NDE of post-WOL welds is also important for determining whether any new 

flaws have initiated or existing flaws have propagated, and for assessing the effect of any 

potential flaws on the service life of the weld. 

5-18
In WOL mockup testing performed by PNNL, all targeted flaws were detected in the pre- and 

post-WOL condition. Using carefully selected NDE parameters and encoded NDE techniques, 

pre- and post-WOL length and depth-sizing RMSE for flaws greater than 50% through the 

original wall thickness were found to be within the ASME depth-sizing qualification acceptance 

criterion.

Inspection limitations can affect the effectiveness of NDE requirements. Geometry or access 

may limit performing a complete pre-WOL weld examination. Information on geometric 

limitations to performing inspections of typical Westinghouse and CE plant RCS Alloy 82/182 

welds is contained in Appendices C and D. Inspection limitations exist for welds adjacent to 

CASS products.

It is the authors’ understanding that if OD weld surfaces can be adequately conditioned to allow 

full volumetric examination to be performed from the OD, the exception in Code Case N-770-1 

to performing a pre-OWOL reactor vessel cold leg nozzle weld examination should not be used.

Based on an evaluation of WOLs using the criteria of Code Case N-770-1, Appendix I, it is 

reasonable to conclude that WOLs provide effective mitigation against the initiation of PWSCC 

and against the growth of any existing PWSCC that has been detected and allowed to remain in 

service. This statement is based, at least in part, on the proviso that the design that is 

evaluated and approved is the design that is actually created in the field.

Based on a review of the requirements of Code Case N-770-1, as implemented with NRC 

conditions, it is concluded that these requirements for WOLs provide an acceptable approach 

for monitoring potential PWSCC. For monitoring of potential PWSCC to perform as expected,

the NDE used has to be effective and reliable.

5-19

6 MANAGEMENT OF PWSCC BY WELD INLAYS/ONLAYS

6.1 Operating Experience with Weld Inlays/Onlays

Weld inlays and onlays are methods to apply weld layers of Alloy 52/152 material on the inside 

pipe wall surface to mitigate potential primary waster stress corrosion cracking (PWSCC) in 

Alloy 82/182 dissimilar metal butt welds. Alloy 52 material has been shown to be more resistant 

to PWSCC than Alloy 82/182. Weld inlays and onlays provide a barrier to separate the Alloy 

82/182 weld and butter from reactor coolant. 

In pressurized water reactors (PWRs) in the United States, inlays and onlays have been used in 

two locations—Westinghouse plant replacement steam generator (SG) hot leg nozzle-to-safe 

end welds and Babcock & Wilcox (B&W) plant reactor vessel core flood nozzle to safe end 

welds. 

In the 1990s, a number of plants replaced the SGs with Westinghouse-built SGs with Alloy 690 

thermally treated tubing. Some of these SGs were installed with Alloy 52 or 152 inlays or 

cladding on the inside of the Alloy 82 hot leg nozzle-to-safe end welds (Karwoski 2003). These 

welds were made at the factory. The closure welds when the SG safe end welds were attached 

to the piping were made of stainless steel. Plants with these inlays included North Anna Power 

Station, Unit 2 (McCoy 2009), V.C. Summer (Gatlin 2012c), Point Beach, Unit 2 (Kunowski 

2008b), and Kewaunee (Kunowski 2008a). 

A description of the V.C. Summer SG nozzle-to-safe end weld inlays is provided in Gatlin (Gatlin 

2012a, b, c). Although these letters were submitted to the U.S. Nuclear Regulatory Commission 

(NRC) in 2012, the V.C. Summer SG nozzle weld inlays have been in service since 1994. The

letters were submitted, in part, to request NRC authorization to recategorize these welds as 

Code Case N-770-1, Inspection Item G, uncracked butt weld mitigated with an inlay, per 

10 CFR 50.55a(g)(6)(ii)(F)(2). Gatlin (2012c) contains a summary description of the fabrication 

sequence as follows:

• The carbon steel channel head integral nozzle end is buttered with weld-deposited Alloy 82 

and preliminarily machined per drawing.

• The channel head nozzle inside diameter (ID) is clad with weld-deposited stainless steel 

(SS) to within approximately 2 inches of the top of the buttered nozzle end per drawing. The

first layer of cladding is 309L SS followed by 308L SS for the remaining layers.

• The nozzle ID region from the SS cladding to the end of the Alloy 82 buttered nozzle end is 

clad with weld-deposited Alloy 152.

• The nozzle end is machined with the safe end weld preparation.

• Post-weld heat treatment of the channel head nozzle end for 1.75 to 2.25 hours at 1125 

±25 degrees Fahrenheit is performed.

• The safe end forging is welded to the channel head nozzle with a double V-groove Alloy 82 

weld.

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• The final 0.27 inch of the nozzle inside diameter V-groove Alloy 152 weld (or inlay) is

completed.

• The outside diameter (OD) and the ID of the nozzle with safe end are final machined.

• A shop hydrotest is performed and shipping covers are welded to the end of the nozzle safe 

end.

• The safe end is machined at site with the piping side weld preparation.

Gatlin (2012c) indicates that, “There is no issue regarding identification of the edges of the 

dissimilar metal welds, to ensure the coverage is complete, for the replacement steam 

generators. As shown in Attachment 1 (Figure 6.1, below), the inlay merges with the stainless 

cladding at a location over the carbon steel nozzles, so even if there was incomplete coverage, 

the susceptible materials would not be exposed. At the upper end of the configuration, the 

Alloy 52/152 extends to the edge of the weld preparation groove, thus ensuring that the 

coverage will be complete. Further verification that the coverage is complete comes from the 

NDE performed on the weld region before the replacement steam generators left the shop.”

The configuration of the inlay at V.C. Summer, and possibly at the other replacement SG 

nozzles with Alloy 52 inlays from the 1990s time period, differs somewhat from the inlay designs 

proposed more recently. The newer inlay designs would extend the Alloy 52 inlay material into 

the safe end well past the safe end-to-Alloy 82/182 weld interface. 

On September 18, 2008, the NRC staff met with Nuclear Energy Institute and other industry 

representatives in Rockville, Maryland, to discuss the regulatory approach to be used when 

applying weld inlays and onlays to mitigate potential PWSCC in Alloy 82/182 dissimilar metal 

butt welds (Tsao 2008a). The weld inlay and weld onlay configurations proposed during this 

meeting are shown in Figures 6.2 and 6.3.

At the September 18, 2008 meeting, industry presented technical papers that addressed ASME 

Code requirements relative to materials, applicable Code Cases, and Sections III and XI 

requirements. Per the requirements in draft Code Case N-766, the technical papers discuss 

chromium content, the implementation process, and nondestructive examination (NDE). Inlays 

and onlays are minimally 3-mm (1/8-in.) thick and installed using a minimum of two layers of 

weldment. These technical papers included the industry’s bases for its conclusions that the use 

of inlays would require relief from ASME Code requirements, while onlays may not require relief 

depending upon welding issues, such as the ability to comply with ASME Code Case N-638-1 

on the ambient temperature machine gas tungsten arc temper bead welding technique. At this 

meeting, industry representatives indicated that Entergy had plans to install weld onlays at 

Arkansas Nuclear One, Unit No. 1 in fall 2008. Code Case N-638-4 was conditionally accepted 

by the NRC in Regulatory Guide 1.147, Revision 16.

6-2
Figure 6.1 V.C. Summer, Unit 1 Replacement Steam Generator Primary Inlet/Outlet 

Nozzle to Safe End Weld Design

6-3
Figure 6.2 Weld Inlay Configuration Proposed by Industry (Tsao 2008c)

Figure 6.3 Weld Onlay Configuration Proposed by Industry (Tsao 2008b)

During the Fall 2009 refueling outage, Three Mile Island installed an Alloy 52 onlay over the 

core flood ‘A’ nozzle weld (Bellamy 2010). Attempts to mitigate the core flood ‘B’ nozzle-to-safe 

end Alloy 82/182 dissimilar metal weld (DMW) by onlay were unsuccessful.

6-4
The core flood ‘B’ nozzle-to-safe end mitigation process was abandoned due to an inability to 

reliably apply Alloy 52M weld material.

(1) The weld configuration that was returned to service 

includes a new approximately 3-mm (1/8-in.) thick layer of Alloy 82 weld material over the DMW 

along with newly applied 309L SS weld build-up adjacent to the DMW to maintain a 

Performance Demonstration Initiative (PDI) ultrasonic testing (UT) inspectable configuration.

The dissimilar metal weld and required adjacent base material were UT, liquid penetrant testing 

(PT), and magnetic particle testing examined prior to performing any welding processes. There 

was no PWSCC identified during the examinations. Numerous small fabrication discontinuities 

that were acceptable to ASME Section III and Section XI were identified. These discontinuities 

and other unidentified original weld material impurities are believed to be the main cause of 

welding difficulties, because the most prevalent of these difficulties were encountered in the 

regions of NDE indications.

Sulfur mitigation layers of Alloy 82 filler metal were applied over the original DMW, which was 

the region requiring application of Alloy 52M weld material. In order to maintain an onlay 

configuration that was inspectable with a Supplement 10 procedure and had at least 3-mm 

(1/8-in.) thickness of PWSCC-resistant Alloy 52M material, the sulfur mitigation layer was limited 

to a thickness of 0.5 mm to 1 m (0.020 in. to 0.040 in.) in the dissimilar metal weld region. This 

thin sulfur mitigation layer allowed the Alloy 52M weld material to penetrate through the Alloy 82 

sulfur mitigation layer and pick up original weld material contaminants that interfered with the 

normal weld puddle wetting process. The Alloy 52M weld beads were described as "ropey" and 

did not have good wetting and fluidity, resulting in unacceptable conditions. Attempts to dilute 

the original weld material contaminants resulted in similar unacceptable Alloy 52M welding 

results.

The root cause was concluded to be the incomplete consideration of the aggregate effect of 

material, geometry, and design on the ‘B’ core flood weld when developing the appropriate weld 

parameters. It was also determined that a contributing factor was inadequate pre-outage 

mockup fidelity. The welding techniques were not fully demonstrated on materials that 

possessed similar high sulfur conditions with simulated acceptable fabrication flaws using the 

designed thin Alloy 82 barrier layer.

In addition to the domestic use of inlays and onlays discussed above, weld inlays were applied

on the reactor vessel-to-safe end welds at Ringhals 3 and 4 in Sweden. Inlays were applied at 

Ringhals 4 in 2002 and inspected with UT and eddy current testing (ET) in 2005. No indications 

were found. These welds were inspected again with UT and ET in 2010. No indications were 

found and these welds have now been placed on a 10-year re-inspection frequency (Hardies 

2011).

(2)

(1) Information on Three Mile Island core flood ‘B’ nozzle onlay obtained by email on 

October 16, 2012, from Exelon Nuclear.

(2) Information on Ringhals is in slides presented at NRC/industry meeting summarized in 

Hardies (2011).

6-5

 
Inlays were applied at Ringhals 3 in 2003 and inspected with UT and ET in 2006. No 

indications were found. These welds were inspected with UT and ET in 2010. No indications 

were found and these welds have now been placed on a 10-year re-inspection frequency

(Hardies 2011).

A summary of operating experience with PWSCC in butt welds in nuclear power plants that was 

compiled by the authors is provided in Appendix B. Based on this summary, the authors did not 

find any instances of leaks in welds with inlays or onlays. No service-related operating 

experience involving cracking in inlays or onlays was found. It should be noted, however, that 

the number of welds mitigated by inlays and onlays is small compared to the number of weld 

overlays and MSIP applied to butt welds susceptible to PWSCC, and information on inspections 

performed would not be available unless requested from each utility.

6.2 Code Case N-766 Design and Examination Requirements for 

Inlays and Onlays

Code Case N-766, “Nickel Alloy Reactor Coolant Inlay and Onlay for Mitigation of PWR Full 

Penetration Circumferential Nickel Alloy Dissimilar Metal Welds in Class 1 Items, Section XI, 

Division 1,” was used as a source for the design requirements summarized in this section

(ASME 2010). NRC concerns with two of the design requirements of this Case are noted in this 

section and discussed in Section 6.5.

For the final configuration, the applicable stress limits of the Construction Code have to be met. 

An inlay or onlay that is 10% or less of the design wall thickness may be neglected in the 

evaluation of the primary plus secondary stress intensity (NB-3222.2) and analysis for cyclic 

operation (NB-3222.4).

The Code Case requires that a fatigue crack growth evaluation be performed and specifies the 

assumptions to be used for initial flaw sizes for both circumferential and axial flaws. The 

acceptance standard for the fatigue crack growth evaluation is that a surface-connected flaw 

may not grow through the full thickness of the inlay or onlay, considering all applicable loads 

subject to Level A and B Service Limits.

The rules for performing fatigue crack growth evaluation are specified in Section XI, IWB-3640. 

The inlay and onlay designs are both required to be evaluated for fatigue crack growth. Two 

postulated cases are required for evaluation—one case for embedded planar flaws remaining 

within the DMW and one case for surface-connected planar flaws within the inlay/onlay.

Axial and circumferential flaw depths are assumed to be 10% of the original DMW thickness if 

as-left embedded flaw depths do not exceed 10% of the original DMW thickness, or if there are 

no embedded flaws. If as-left embedded flaw depths exceed 10% of the original DMW 

thickness, the actual as-left embedded flaw depths are to be used. The minimum initial flaw 

depth assumption of 10% through the DMW initiating at the inlay/onlay to DMW interface is 

based on the minimum detection capability inferred from Table IWB-3514-1 and is consistent 

with the crack growth and design requirements of Case N-740-2. For the circumferentiallyoriented flaw, the postulated flaw length is the entire circumference of the DMW. For the axially6-6
oriented flaw, the postulated flaw depth is the same, but the length is equal to the full DMW 

width plus any additional susceptible material.

In draft Code Case N-766-1, even if a flaw greater than 10% of the DMW thickness is detected 

and analyzed as an embedded flaw, as discussed below, the postulated 10% embedded flaw 

case is still required to be evaluated in order to cover undetected flaws that may be left between 

the inlay, or onlay, and the DMW at a location away from the detected flaw.

The weld inlay, or onlay, must also be evaluated for fatigue crack growth considering a 

postulated planar surface-connected flaw. The postulated initial flaw depth is specified to be 

1.5 mm (1/16 in.). This postulated flaw size is specified to conservatively account for 

undetected flaws within the inlay or onlay based on current detection methods. The depth is 

equivalent to approximately one layer of weld material. The flaw lengths are specified to be 

360° for circumferential flaws and the width of the DMW plus any additional susceptible material 

for axial flaws. For both the postulated embedded and surface-connected flaw cases, the flaw 

must not grow through the inlay, or onlay, for the remaining life of the plant, thereby maintaining 

its function as a barrier between the Alloy 82/182 material and reactor coolant (Marlette et al. 

2012). Although much more resistant than Alloy 82, the NRC has concluded in its negative 

ballot on Code Case N-766 that, in addition to performing a design analysis for fatigue, Alloy 52 

inlays and onlays should also be analyzed for PWSCC.

The effects of weld shrinkage on the component are generally not of great concern for 

inlay/onlay applications because the inlay/onlay thickness is so small compared to the overall 

DMW cross-sectional thickness. However, the Case requires that an inlay or onlay with a 

thickness greater than 12.5% of the original DMW thickness be evaluated for shrinkage effects. 

The thickness requirement is based on engineering judgment and the assumption that the inlay 

or onlay will be applied to a nozzle with a wall thickness of 50.8 mm (2 in.) or greater (Marlette 

et al. 2012).

Prior to installation of an inlay or onlay, the DMW and area to be welded has to be examined by 

volumetric and surface methods. Volumetric examinations have to be performed in accordance 

with the requirements of Code Case N-770-1 or later revisions. A surface examination has to 

be performed using liquid penetrant or ET on the area to be welded. Indications detected by 

volumetric examination that exceed IWB-3514 standards have to be corrected. Alternatively, 

the Code Case allows indications that do not meet the acceptance standards of IWB-3514 to be 

accepted by analytical evaluation in accordance with IWB-3600. The NRC disagreed with this 

provision and in its negative ballot on Code Case N-766 expressed a concern with returning to 

service welds with deep embedded flaws.

If a pre-existing flaw cannot be qualified to allow leaving it in place as an embedded flaw, the 

flaw has to be removed or reduced to an acceptable size by grinding out part or all of the 

material containing the flaw, and performing a localized welded repair prior to installing the inlay 

or onlay in accordance with IWA-4000 and IWB-3600. If weld repair of indications is required, 

the area where the weld inlay or onlay is to be deposited, including any local weld repairs, has 

to be examined using PT or ET. The area shall be free of surface indications with major 

dimensions greater than 1.5 mm (1/16 in.) prior to installation of the inlay or onlay. After 

installation, the inlay or onlay surface, including at least 13 mm (1⁄2 in.) of adjacent material has 

6-7
to receive a surface examination that may be performed by liquid penetrant or eddy current. 

Acceptance criteria for the surface examination of inlays or onlays has to satisfy Section III,

NB-5352, except that indications with major dimensions greater than 1.5 mm (1/16 in.) have to 

be removed or reduced in size or repaired by welding and reexamined. The inlay or onlay 

volume, including the fusion zone, and ferritic steel heat-affected zone, when temper bead 

welding is used, has to be examined by UT. The volumetric examination performed after 

installation is for the purpose of detecting lack of bond and clad flaw indications only and is 

performed in accordance with Section V requirements. If temper bead welding is used, the 

examination shall be conducted no sooner than 48 hours after the completion of the third 

temper bead layer over the ferritic steel base material. Preservice and inservice examinations 

are required to be performed in accordance with Code Case N-770-1 or later.

To reduce the potential for hot cracking, the Code Case permits users to apply austenitic filler 

material over the austenitic stainless steel or ferritic steel materials. Surface examination of the 

seal layer(s) must also be performed using the same acceptance criteria as noted in the 

preceding paragraph. The location of the DMW fusion zones has to be identified and the 

accuracy of the locating technique has to be demonstrated on representative mockups and 

documented. 

The Case specifies nickel alloy filler metal with a minimum 28% chromium content be used and 

an as-deposited weld metal with a minimum of 24% chromium content be achieved. Inlays and 

onlays have to be deposited using a Welding Procedure Specification qualified for groove 

welding in accordance with the Construction Code and Owner’s Requirements identified in the 

Repair/Replacement Plan. Likewise, welders and welding operators have to be qualified in 

accordance with the Construction Code and Owner’s Requirements identified in the 

Repair/Replacement Plan. The preheat and post-weld heat treatment requirements of the 

Construction Code and Owner’s Requirements identified in the Repair/Replacement Plan have 

to be met if welding on ferritic base materials or if 3 mm (1/8 in.) or less of nonferritic weld 

deposit exists above the fusion line of the ferritic steel base material. Alternatively, the Code 

Case prescribes that ambient temperature temper bead welding may be performed in 

accordance with Appendix I of the Code Case.

Inlays and onlays are required to consist of at least two layers after final surface preparation. 

The minimum required thickness is 3 mm (1/8 in.) and all inlay and onlay layers credited toward 

the minimum thickness have to contain at least 24% chromium, as determined by chemical 

analysis of a representative mockup. Weld inlays and onlays are required to extend beyond the 

DMW and butter fusion zones by at least twice the demonstrated accuracy of the technique 

used for locating the fusion zones or 6 mm (1⁄4 in.), whichever is greater.

6.3 NDE Reliability

6.3.1 Qualification and Acceptance Standards for Surface Examination

Code Cases N-770-1 (inspection of Alloy 82/182 butt welds) and N-766 (inlay/onlay mitigation 

technique) both use ET examinations to ensure the quality of welds. However, there are two 

issues with the application of ET as a surface examination method in these Code Cases. 

6-8
The first issue pertains to ET acceptance standards prescribed in these Code Cases. Code 

Case N-770-1 states that the acceptance standards of NB-5352 apply for the surface 

examination of weld inlays or onlays, except that rounded indications with dimensions greater 

than the smaller of 1.5 mm (1/16 in.) or 50% of the thickness of the inlay or onlay are 

unacceptable. Code Case N-766 also invokes NB-5352 as the acceptance standard for surface 

examinations. Section III, NB-5352 contains acceptance standards for liquid penetrant. These 

standards are not applicable to ET. For example, the 1.5-mm (1/16-in.) dimension allowed for 

liquid penetrant would be indicative of a smaller flaw than 1.5-mm (1/16-in.) because of 

penetrant bleed-out.

The second issue pertains to ET qualification. Code Case N-770-1 states that the ET 

examination is required to be performed in accordance with IWA-2223. Code Case N-766 

specifies that, “nondestructive examination methods shall meet the requirements of IWA-2200.” 

Section XI, IWA-2223 requires that eddy current examination for detection of surface flaws be 

conducted in accordance with Appendix IV. Appendix IV, Supplement 2, contains requirements 

for qualification of ET techniques by demonstration for surface examination of piping and 

vessels. Supplement 2 allows notches with a maximum width of 0.25 mm (0.010 in.) to be used 

for the demonstration. The Appendix IV qualification is applicable only to materials whose 

acceptance standard is 3 mm (1/8 in.) or more in length. The width of a PWSCC flaw would be 

much smaller than a 0.25-mm (0.010-in.) wide notch allowed to be used in an Appendix IV, 

Supplement 2, demonstration and ET used for inlays and onlays has to be capable of reliably 

detecting rounded indications as small as 1.5 mm (1/16 in.). 

At the ASME Code, Section XI meetings in Phoenix from November 5–8, 2012, the NRC staff 

(Collins et al. 2012) indicated that it is considering the adequacy of the ET qualification 

requirements of Appendix IV, Supplement 2, which are based on technique demonstration as 

opposed to performance demonstration techniques. ET performance demonstration techniques 

are provided in the recently approved Code Case N-773, “Alternative Requirements for 

Automated Eddy Current Examination of Class 1 Stainless Steel or Stainless Steel Clad Carbon 

Steel Piping from the Inside Surface.” This Case provides an optional ET qualification program

to that in Appendix IV, Supplement 2, and may serve as a guideline for a new Code Case for ET 

of inlays and onlays. At these meetings, the NRC staff also indicated an interest in performance 

qualification techniques that would address the possibility that very shallow subsurface defects 

will become exposed during operation. This has occurred at San Onofre, Unit 3, for upper head 

seal/cladding welds using Alloy 52 materials (Scherer 2008).

6.3.2 Ultrasonic Equivalency Testing of Weld Inlaid Components

The PWR Owners’ Group (PWROG) sponsored testing to evaluate the effectiveness of current 

automated ultrasonic procedures applied to inner surface inspection of inlaid welds without 

modification of essential variables (Latiolais 2008).

The primary focus of this project was to evaluate the detection and sizing capabilities of 

Appendix VIII, Supplement 10, qualified procedures used for examination of DMWs from the 

inside surface on an inlaid component without changing the essential variables of the qualified 

procedure. ASME Section XI, Appendix VIII, Supplement 10, is not applicable to components 

with supplemental corrosion-resistant cladding applied to mitigate stress corrosion cracking

6-9
(SCC). Therefore, no PDI-qualified technique currently exists that can be used to inspect inlaid 

DMWs. The PWROG project was intended to show the effect of inlay material on the ability to 

ultrasonically detect and size flaws that might develop later in the service life of an inlaid 

component. It was the PWROG’s plan for this testing that if equivalency was demonstrated, the 

need for additional qualification activities would not be necessary because currently available 

inspection procedures could be used.

An inlay test mockup was created that was identical to an existing PDI-601 series weld

configuration. The PDI-601 series weld configuration is for the Westinghouse PWR vessel 

outlet shop weld, which has a SA-508 nozzle, an Inconel butter and weld to a SA182-316 safe 

end, which is welded to a wrought 308/316 stainless steel pipe (Latiolais 2002). This mockup is 

a full-scale representation. Both the 601 and the inlay mockup were fabricated from actual 

dropouts taken from a cancelled PWR to ensure that materials and welding processes are 

typical of those found in an operating plant. The inlay material was applied using welding 

techniques and hardware developed by AREVA NP. In order to identify welding imperfections 

that might exist in the DMW, AREVA NP ultrasonically inspected the inlay mockup by using a 

qualified Appendix VIII procedure before inlay material was applied.

The inlay test mockup has four 90° quadrants with varying inlay thicknesses that were blended

smooth on the inner diameter surface. The first quadrant of the mockup was used for a 

baseline for noise comparison and had no weld inlay or flaws. The second and third quadrants 

have 0.508-cm and 0.178-cm (0.20-in. and 0.07-in.) thick inlays, respectively. In addition to the 

inlay, quadrants two, three, and four have the same set of four flaws identified as Flaws 1, 2, 3, 

and 12, which are identical to flaws contained in the 601 practice mockup. Flaws 1, 2, and 12 

were circumferentially oriented with depths of 8.6 mm (0.339 in.), 8.9 mm (0.350 in.), and 

9.5 mm (0.374 in.), respectively. Flaw 3 was axially oriented with a flaw depth of 20.7 mm

(0.815 in.). In the fourth quadrant, an embedded flaw represents the partial removal and 

subsequent 25.4-mm (1-in.) deep inlay repair of a flaw. The flaws were implanted in the 

mockup using the same techniques used for the fabrication of the PDI-601 mockup.

Two vendors were contracted to inspect the inlay mockup. Previously collected data on similar

flaws in the 601 mockup were used as a comparative reference. All data were collected 

following each vendor's applicable Appendix VIII-qualified automated ultrasonic procedure.

Both vendors performed an open procedure demonstration and reported similar findings with 

respect to flaw detection, sizing, and relative noise levels within the areas of the inlay. Both 

vendors observed an apparent increase in the noise level in the areas where the inlays were 

applied. For a given flaw, the signal-to-noise ratio was less in the Electric Power Research 

Institute (EPRI) block 601 when comparing the corresponding flaws in the inlayed areas. This 

reference noise level increased as the thickness of the inlay material increased. In the axiallyoriented flaws, the signal-to-noise ratio was more noticeably affected, but it was not at a level 

that would impact the results of the examination. This increase in noise generally made flawdepth sizing more difficult, but it had a lesser effect on detection capabilities. The increase in 

the relative noise within the areas of the inlays for circumferentially-oriented flaws depended 

upon the scanning direction. Based on discussions with multiple qualified UT inspectors, the 

increase in noise was attributable to the microstructure of the inlay and appeared to be affected 

by the orientation of the dendrites based on the direction of welding. 

6-10
Each of the axially orientated flaws was sized using the 60° probe instead of the 45° probe, 

which is not the optimum angle for through-wall sizing but is allowed by procedure. Both 

vendors were able to detect, characterize, length size, and depth size all the implanted axial and 

circumferential flaws without deviating from the procedurally qualified techniques and the results 

were within the tolerances imposed by EPRI’s PDI-qualified inspection procedures; that is, 

within the acceptance criteria in Section XI, Appendix VIII.

The reliability of UT is of interest to welds mitigated by weld inlays and onlays for several 

reasons. Pipe welds to be mitigated by weld inlays and onlays are required to be inspected 

immediately prior to the application, using an inspection technique qualified in accordance with 

ASME Section XI, Appendix VIII, and any flaws detected are to be evaluated in accordance with 

IWB-3600. Undetected or mischaracterized flaws could compromise the structural integrity of 

the weld. The calculated life of the inlay or onlay depends, in part, upon the reliability of the 

NDE performed. The reliability of NDE of post-mitigated welds is also important for determining 

whether any new flaws have initiated or existing flaws have propagated, and for assessing the 

effect of any potential flaws on the service life of the weld.

PNNL held a workshop in May 2012 with LMT, Inc., a commercial nuclear NDE vendor, to 

review UT data from service-induced flaws and to discuss factors that affect the reliability of 

various NDE techniques qualified in accordance with Section XI. The conclusions of this 

workshop, which relate to inspection of both pre- and post-inlay/onlay piping welds, are provided 

in Section 3.1.2.

6.4 Assessment of the Effectiveness of Weld Inlays/Onlays as a 

PWSCC Mitigation Strategy

In a study performed by Rudland et al. in 2010, confirmatory welding residual stress and flaw 

evaluation analyses were conducted to evaluate the inlay as a mitigation technique. Two largebore, reactor coolant nozzle geometries were considered. The first was an 872-mm (34.3-in.)

diameter, 68.1-mm (2.68-in.) thick reactor coolant outlet nozzle taken from an AREVA inlay 

submittal. The second was a 923-mm (36.3-in.) diameter, 83.8-mm (3.3-in.) thick reactor 

coolant nozzle used to develop welding residual stress results for probabilistic fracture 

mechanics calculations. For each of these geometries, detailed finite element simulations were 

conducted to predict the welding residual stress through the application of the inlay. The effects 

of temper bead welding and weld repairs were considered. The following repair sizes were 

considered:

• 50% deep ID preservice repair (before stainless steel safe end)

• 50% deep ID PWSCC repair (after stainless steel safe end but before inlay)

• 75% deep ID preservice repair

• 75% deep ID PWSCC repair

• 50% deep ID preservice repair and a 75% deep ID PWSCC repair

• 12% deep ID preservice repair

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• 12% deep ID PWSCC repair

• 12% deep ID preservice repair and a 12% deep ID PWSCC repair

For each case, a final inlay thickness of 3 mm (0.125 in.) was considered because this 

corresponded to the minimum thickness required by Code Case N-766. In addition, an inlay 

with a thickness of 6 mm (0.25 in.) was considered for the 50% deep ID preservice and 50% 

deep ID PWSCC repair cases.

The results from the welding residual stress analyses suggest that regardless of the repair

history, the ID stress state is driven by the inlay repair. In most cases analyzed, the ID axial 

stress due to the inlay was about 58 ksi (400 MPa), which is just above the yield strength for the 

DMW material. Even the beneficial effect of the SS safe end weld was eliminated by the thin 

layer of inlay material on the ID surface of the pipe. For all of the inlay cases with preservice

repairs, the high axial stress on the ID dropped to zero at about 15% percent of the wall

thickness for all repair depths. For the cases with PWSCC repairs, the axial stress dropped to

zero between 40% and 50% of the wall thickness for repairs greater than 50% deep, and it 

dropped to zero at about 20% of the wall thickness for the 12% deep repair.

The results from the welding residual stress analyses were used in PWSCC growth analyses

using both idealized-shaped and natural-shaped cracks assuming simulated crack growth rates.

The idealized flaw analyses followed the techniques documented in the ASME Section XI flaw

evaluation guidelines. Sensitivity studies were conducted on crack growth rate, bending stress,

operating temperature, initial flaw size and orientation, and inlay depth.

For the idealized flaw analyses, the ASME Section XI flaw evaluation guidelines were used in

conducting PWSCC growth calculations. The effects of crack initiation were ignored and an

initial surface-breaking defect was assumed in all analyses. This assumption is consistent with

the Code Case N-766 requirement that a postulated surface-breaking planar flaw that is 1.5-mm

(1/16-in.) deep will not grow through the full thickness of the inlay due to fatigue for the life of the 

inlay. Crack growth times through the inlay, to through-wall penetration, and to rupture were 

calculated. The following significant conclusions were obtained:

• The time for the initial defect to grow through-wall was dominated by the time for the initial 

defect to grow through the inlay; that is, a large portion of the crack growth time is spent in 

the first 3 mm of wall thickness. The only exception to this case was for the small weld 

repair (12%). In this case, the flaw arrested after passing through the inlay.

• For the 3-mm deep inlay, the time through the inlay was typically about 10 years. After that 

point, the time to leakage varied with repair size, ranging from an additional 5 years for a 

50% deep repair to arrest for a 12% deep weld repair.

• For the 6-mm deep inlay, the time through the inlay was about 25 years, with the additional 

time to leakage the same as for the 3-mm cases.

• The crack growth times through the inlay are directly proportional to the crack growth rate 

used. More experimental data on Alloy 52 crack growth rates is needed to quantify the 

uncertainty in the crack growth rate.

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• Bending stress did not have a large effect on the time for the crack to pass through the inlay, 

but had a large effect on time to leakage; that is, the lower bending stresses caused the

crack to slow considerably near the middle of the wall thickness.

• Temperature had a large effect on the crack growth behavior. The time through the inlay,

time to leakage, and time to rupture were increased by about a factor of 6 by reducing the

temperatures from hot leg (327°C) to cold leg (288°C) conditions. Even for the fastest crack 

growth considered, the time to leakage was over 50 years at the cold leg temperatures.

• Initial crack length did not have a large impact on the time through the inlay or to leakage.

• Axial cracks grew faster than circumferential cracks due to the larger hoop stresses in the

inlayed welds.

For the natural flaw growth analyses, the PipeFracCAE code, which was used in the Wolf Creek

analyses, was modified to handle the unusual crack growth that may occur in an inlayed DMW. 

In a natural flaw growth analysis, the flaws are driven by PWSCC and tensile stresses caused 

by the weld residual stress and the service loading. A finite element analysis is performed at 

each increment of flaw growth to calculate stress intensity factors along the crack front. 

Because of the differences in the crack growth rates between the inlay material (Alloy 52) and

the DMW (Alloy 82/182), a discontinuous, balloon-shaped crack forms. After modifications to 

the PipeFracCAE code were complete, several sample cases were run for comparisons to the 

idealized flaw analyses. The intent of these natural flaw analyses was to verify that the unusual 

flaw shapes did not highly impact the time through the inlay and leakage calculated using the 

idealized flaw assumption. The results from these runs illustrated that in some cases, the 

natural crack analyses gave slightly longer times to leakage, while in other cases the times to 

leakage were slightly shorter. The differences stemmed from the effect of the natural crack 

shape on the crack driving force and the more accurate representation of the welding residual 

stress field in the natural flaw analyses.

Overall, the natural crack shape analyses demonstrated that reasonable approximations of 

crack growth time through the inlay and to leakage can be made with idealized flaw analyses.

Several general conclusions can be made regarding the effectiveness of inlays as a mitigation 

technique. First, the sensitivity studies conducted assuming a postulated initial defect indicate 

that for the hot leg locations, the time to leakage values are less than the inlay design life even 

with Alloy 52 PWSCC crack growth rate improvement factors of 30–100. For the case of a 

3-mm (0.125-in.) inlay, the times to leakage from the postulated defect ranged from about 

12 years to 30 years. This conclusion assumes that the effect of crack initiation in Alloy 52 inlay 

materials is neglected and can be considered an additional conservatism in the results.

Second, the results presented in this report are highly influenced by the Alloy 52 crack growth

rates assumed. Because of the limited Alloy 52 crack growth data, additional crack growth data 

is needed for these materials to fully understand and quantify the uncertainty in the results.

Third, the times to leakage calculated in this effort do not support sample inspections for inlaymitigated hot leg temperature welds. However, the effect of temperature on the growth rates is

dramatic and therefore the results from this study support sample inspection for inlay-mitigated

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cold leg temperature welds. As noted in Section 6.6 below, the NRC incorporated Code Case 

N-770-1 into 10 CFR 50.55a and imposed a condition prohibiting sample inspection of hot leg

welds mitigated by inlays or onlays.

Finally, the results from this study suggest that a large portion of the time to leakage is spent

growing the postulated defect through the inlay. Because the inlay may only be about three to 

five percent of the total wall thickness, the study performed by Rudland et al. (2010)

recommended that both volumetric and ID surface examination be performed to locate a defect 

that may be present in the inlay material. Code Cases N-766 and N-770-1 require volumetric 

and surface examinations.

Regarding studies performed on cracking in inlay/onlay materials, Toloczko et al. (2010)

provides the results of initial examinations to assess defects in prototypic industry-produced 

Alloy 52, 52M, and 152 mockup welds. Weld metals were characterized in three different 

mockup configurations including an Alloy 52M inlay mockup made by Ringhals AB. Most of the 

weld cracks were found near the final Alloy 52M weld pass in this thick inlay. Only isolated 

cracks were identified near the Alloy 52M-to-Alloy 82 interface and nearly all of these were in 

the Alloy 82 material. The weld cracks in the Alloy 52M were typically less than 100 μm in 

length, but there was one possible instance of an approximately 500-μm crack in the final weld 

pass.

Toloczko et al. (2010) reported that the preliminary weld crack examinations indicate that small 

intergranular weld cracks will be present in most Alloy 52, 52M, and 152 weldments. Brief 

comments can be made on implications to the use of thin, high-Cr inlays to provide corrosion 

resistance. The observed cracks do not appear to provide a continuous path for water to reach 

the more SCC-susceptible, lower-Cr Alloy 182 or 82 weld metal. While the typical crack size in 

all the welds was 100 μm or less, in three of the weld mockups, several 500-μm-long cracks 

were observed. This length is insufficient for a single crack to span even a thin overlay, but a 

clustering of interconnected cracks may potentially provide a path. This seems unlikely based 

on the current limited results; however, more detailed studies of crack size and threedimensional distributions would need to be performed to better assess the probability that such 

interconnected cracks could exist.

Toloczko et al. (2010) reported further that the more important question is whether these preexisting flaws can act as sites for accelerated crack growth during light-water reactor service. 

The location and length of a weld crack in the inlay or overlay would obviously have a strong 

effect on the local stress intensities. A long weld crack that intersects the surface of an inlay or 

overlay would be subject to higher fatigue-driven stresses, and clusters of cracks could more 

easily become interconnected. Similar issues can be envisioned for a weld crack on the surface 

acting as a nucleation site for the growth of an SCC crack. Measured SCC propagation rates of 

these materials are extremely slow, with perhaps a maximum rate of approximately 0.1 mm/year 

at relatively high stress intensity values. Relying on a simple calculation, it was estimated that it 

would take roughly 20 years for an SCC crack to span a 2-mm-thick inlay repair.

An update on Pacific Northwest National Laboratory SCC crack growth testing on Alloy 152, 52, 

52M, and 52MSS welds is provided in Toloczko et al. (2011). These results support the crack 

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growth rate results previously reported and discussed above and are consistent with or more 

conservative than the crack growth rates assumed in Rudland et al. (2010).

Based on the results of the 2010 assessment by Rudland et al., NRC developed condition 

10 CFR 50.55a(g)(6)(ii)(F)(5) on the inspection requirements for inlays in Code Case N-770-1 

that prohibits sample inspection of hot leg temperature inlays/onlays. Assuming that reliable 

NDE techniques are used to inspect inlays/onlays, inlays/onlays of appropriate thickness would 

be expected to provide effective mitigation against the growth of PWSCC into Alloy 82/182 

welds to which they are attached.

6.5 Status of Code Case N-766, “Nickel Alloy Reactor Coolant Inlay 

and Onlay for Mitigation of PWR Full Penetration Circumferential 

Nickel Alloy Dissimilar Metal Welds of Class 1 Items”

ASME approved Code Case N-766, “Nickel Alloy Reactor Coolant Inlay and Cladding for Repair

or Mitigation of PWR Full Penetration Circumferential Nickel Alloy Welds in Class 1 Items” on

December 20, 2010, and issued it in Supplement 4 of the 2010 Edition of Section XI (ASME 

2010). The NRC staff, in accordance with ASME voting procedures, submitted a negative ballot 

on this Code Case.

The staff presented its issues with the Code Case during discussions at the Code committee 

meetings and in writing with the negative ballot. The staff’s major concerns were as follows.

1. The Code Case does not require a design analysis for flaw growth by PWSCC. The staff 

believes that an evaluation of PWSCC should be required for mitigative Alloy 52 inlays and 

onlays (Hardies 2011) to assure adequate thickness of the inlay/onlay for the Code Case 

inspection intervals requirements. In the staff’s PWSCC flaw growth analysis for 

inlays/onlays (Rudland et al. 2010), the staff postulated an initial flaw size of 1.6 mm

(1/16 in.) or 50% through the thickness of the inlay, as used in the Code Case for fatigue 

evaluation. The purpose of the staff’s analysis was to evaluate the inspection intervals for 

inlays and onlays in Code Case N-770-1. As discussed in Section 6.4, the NRC staff 

concluded that from this analysis that the sample inspection provisions of Code Case 

N-770-1 inspection requirements were not acceptable for hot leg welds mitigated by inlays 

or onlays.

2. Another concern expressed by the NRC staff prior to approval of Code Case N-766 by the 

ASME pertains to the allowable size of embedded flaws that could be left in service after 

application of an inlay or onlay. The Code Case allows a flaw in the original Alloy 82/182 

dissimilar metal weld to remain in service as long as IWB-3600 is satisfied. The NRC staff 

indicated that it does not find this to be acceptable for Class 1 piping. The NRC staff 

indicated its belief that Code Case N-766 must provide a limitation on the size of a flaw that 

is allowed to remain in service after an inlay is installed. The NRC staff indicated that it does 

not object to allowing flaws to remain in service after inlay or onlay installation. However, in 

casting its ballot on Code Case N-766, the NRC staff stated that it finds it to be 

unacceptable if a “large” flaw remains in service regardless of inlay repair and the 

requirements of the ASME Code, Section XI.

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Regarding the first of the NRC staff’s major concerns, industry developed a technical basis to 

justify a small initial flaw size for a PWSCC evaluation (Marlette et al. 2012). Marlette et al. 

report that the initial flaw size for the crack growth analysis was conservatively based on known 

eddy current detection capabilities. The ASME Boiler and Pressure Vessel Code contains 

qualification requirements for eddy current examination with specified flaw depths of 0.5 mm

(0.02 in.) and less. For example, Supplement 2, Appendix IV of Section XI specifies a 

maximum depth of 0.5 mm (0.02 in.) for machined notches within a specimen to be used in 

eddy current qualification testing. They also report that eddy current testing documented in 

Kobayashi (2010) has demonstrated the ability to detect surface flaws as shallow as 0.3-mm

(0.012-in.) deep within a weld inlay surface. Results are documented for several test specimens 

within the paper. Underwater laser beam welding was used to fill an inlay groove in one 

specimen. Surface slits with depths of 0.3 mm (0.012 in.), 0.5 mm (0.020 in.) and 1 mm

(0.04 in.) were placed in the inlay surface. The length and width of all the surface flaws were 

1.5 mm (0.06 in.) and 0.2 mm (0.008 in.), respectively. The test results demonstrated that eddy 

current is capable of detecting flaws as shallow as 0.3 mm (0.012 in.) deep within an inlay 

surface. 

Accordingly, Marlette et al. (2012) assumed an initial surface flaw with a depth of 0.5 mm

(0.02 in.) and a length-to-depth aspect ratio of 6, as used by Rudland et al. (2010). The 

analysis also assumed the same crack growth rate for the Alloy 52 weld inlay material that was 

used in Rudland et al. (2010); namely, a PWSCC crack growth rate for Alloy 52 weld material of 

one hundredth of the growth rate for Alloy 182. Marlette et al. (2012) report that the new 

analysis used the same two physical models, worst-case repair cases, residual stress models, 

bending and axial stresses, stress intensity factor calculation methodology, and crack paths as 

used by Rudland et al. (2010). Benchmarking runs were performed in order to provide 

validation that the generic analysis would produce similar results to the methods used by 

Rudland et al. (2010). All of the key parameters used in Rudland et al. (2010) including the 

1.5-mm (1/16-in.) initial flaw depth were used for the benchmarking. Once the benchmark 

results were verified to be in good agreement with the results reported in Rudland et al. (2010), 

the new analysis was run using the 0.5-mm (0.02-in.) initial flaw depth.

Marlette et al. (2012) report that the results of the new analysis demonstrate that a postulated 

flaw will not grow through the remaining inlay thickness within an inspection interval; that is, a 

10-year period. The analysis shows that the worst-case result is for axial flaw growth, which still 

provides over 15 years of operation before the postulated flaw would grow though the minimum 

inlay thickness. The authors indicate that the results from this analysis demonstrate that even in 

the unlikely case that PWSCC would become an active mechanism for crack growth within the 

resistant inlay material, it would not grow through the minimum required thickness within a 

10-year inspection interval.

ASME, Section XI, developed a revision to Code Case N-766 that was intended to address the 

NRC staff’s concerns and other issues that need to be clarified. Nevertheless, Code Case 

N-766-1 does not address either of the staff’s concerns. For example, regarding a design 

analysis for flaw growth, the revised Code Case does not require users to perform a casespecific analysis or verify that the industry’s generic analysis bounds weld-specific conditions. 

Final resolution of the staff positions on Code Case N-766 and Code Case N-766-1 will be 

developed as part of a future revision to Regulatory Guide 1.147 (Hardies 2011).

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6.6 Assessment of Code Case N-770-1 Inspection Requirements

An overview of Code Case N-770-1 inspection requirements is contained in Section 3.4.1. The 

paragraphs that follow in this section describe the Code Case N-770-1 requirements for inlays 

and onlays and the conditions that NRC imposed on the examination of inlays and onlays when 

the NRC incorporated ASME Code Case N-770-1 by reference into 10 CFR 50.55a (76 FR 

36232) in June 2011.

Inspection Item G welds are uncracked butt welds mitigated with an inlay and Inspection Item J 

welds are cracked butt welds mitigated with an inlay. Figure 3 of the Case shows the 

examination volume in weld inlays and indicates that inlays shall be constructed of an Alloy 52 

material. Inspection Item H welds are uncracked butt welds mitigated with an onlay and 

Inspection Item K welds are cracked butt welds mitigated with an onlay. Figure 4 of the Case 

shows the examination volume in weld onlays and indicates that onlays shall be constructed of 

an Alloy 52 material.

Table 1 of the Case provides requirements for the examination methods, acceptance standards, 

and extent and frequency of examination of unmitigated and mitigated welds. For Inspection 

Items G and H, Table 1 requires inservice volumetric and surface examinations to be performed 

no sooner than the shorter of 10 years following the application of the inlay or onlay and the 

design life of the inlay or onlay. Examination volumes that show no indication of cracking are to 

be placed into a population to be examined on a sample basis. For Inspection Items J and K, 

Table 1 requires inservice volumetric and surface examinations to be performed once during the 

first or second refueling outage following application of the inlay or onlay. Examination volumes 

that show no indication of cracking are to be placed into a population to be examined on a 

sample basis. Based on the analysis performed by Rudland et al. (2010), the NRC disagreed 

with sample inspection provisions of Inspection Items G, H, J, and K and imposed condition 

10 CFR 50.55a(g)(6)(ii)(F)(5). This condition is discussed below. The Code’s once-per-interval 

inspection frequency reflects an implicit conclusion by the ASME and, in imposing the Code 

Case, by the NRC that inlays and onlays are expected to result in effective mitigation and that 

inspection serves a defense-in-depth monitoring function rather than a degradation 

management function.

Table 1 indicates that Notes (15), (16), and (17) apply to weld inlays and onlays. A discussion 

of these Notes follows.

Note (15)(a) requires a volumetric examination to be performed immediately before application 

of inlays or onlays and after application as a preservice baseline inspection. This requirement is 

consistent with Section XI requirements in IWA-4000 for activities involving welding on pressure 

boundary components.

Notes (15)(b) and (c) indicate that if an inlay or onlay weld does not permit acceptable weld 

inspection coverage as defined in the paragraph -2500(c) of the Case or if the capabilities of the 

volumetric examination for detection, length or depth sizing are adversely affected by the inlay 

or onlay, the weld shall be examined as an unmitigated weld in accordance with the 

requirements of Inspection Items A or B. These requirements are needed to be able to monitor

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weld inlays and onlays and ensure they remain capable of isolating the susceptible Alloy 82/182 

weld from the reactor coolant and preventing potential PWSCC from occurring in the 

Alloy 82/182 weld.

Notes (15)(d) requires preservice surface examination to be performed and defines the surface 

examination requirements and the acceptance criteria. Note (15)(e) requires preservice 

volumetric examination to be performed and defines the volumetric examination requirements 

and acceptance criteria. 

The examination requirements and acceptance criteria of Notes (15)(d) and (e) draw on existing 

ASME Code, Section XI, rules except that rounded indications found by surface examination to 

be larger than 1.5 mm (1/16 in.) or 50% of the thickness of the inlay or onlay are unacceptable. 

The acceptance standards of NB-5352 would allow rounded indications as large as 3 mm

(1/8 in.).

Note (15)(f) allows welds, with cracks that have been detected and completely removed and 

repaired in accordance with IWA-4000 before application of the inlay or onlay, to be reclassified 

as Inspection Item G or H, respectively. This note clarifies the status of welds with cracks that 

have been completely removed and repaired and subsequently mitigated by inlays or onlays.

Note (16) applies to inservice volumetric examination of weld inlays and onlays that reveal crack 

growth or new cracking. Note (16)(a) requires subsequent volumetric examinations if the 

cracking meets the acceptance standards of -3132.3, “Acceptance by Evaluation,” of the Case. 

Note (16)(b) requires surface examinations in addition to the subsequent volumetric 

examinations required by Note (16)(a). These surface examinations are required to be 

performed in accordance with Note (17). Note (16)(c) states that if examinations required by 

Note (16)(a) reveal that the flaws remain essentially unchanged for three successive 

examinations, the weld examination schedule may revert to the sample and schedule of 

examinations identified in Table 1. These rules are consistent with ASME Code, Section XI,

requirements for subsequent examinations except that they also require a surface examination 

if crack growth or new cracking is revealed by a volumetric examination. Note (16)(c) also 

requires that if cracking penetrates beyond the thickness of the inlay or onlay, the weld shall be 

reclassified as Inspection Item A-1, A-2, or B, as appropriate, until corrected. This requirement 

is needed because loss of integrity of the inlay or onlay would expose the underlying 

Alloy 82/182 material to reactor coolant and the potential for PWSCC to occur.

Note (17) pertains to inservice examination requirements for surface examinations of weld 

inlays and onlays and provides qualification requirements and acceptance standards for surface 

examinations by liquid penetrant and ET. As in Note (15)(d), these examination requirements 

and acceptance criteria draw on existing ASME Code, Section XI, rules except that rounded 

indications found by surface examination to be larger than 1.5 mm (1/16 in.) or 50% of the 

thickness of the inlay or onlay are unacceptable. The acceptance standards of NB-5352 would 

allow rounded indications as large as 3 mm (1/8 in.).

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The rule imposing Code Case N-770-1 (76 FR 36232) imposed conditions (F)(1), (2), (4), (5), 

(6), and (7), which apply or may apply to welds modified by inlays or onlays. The NRC imposed 

these conditions, which are in addition to the Code Case requirements, to ensure that, in 

combination with the Code Case, adequate protection is provided for monitoring Alloy 82/182 

welds.

Condition §50.55a(g)(6)(ii)(F)(1) of the rule requires licensees of existing operating PWRs as of 

July 21, 2011, to implement the requirements of ASME Code Case N-770-1, subject to the 

conditions specified in paragraphs (g)(6)(ii)(F)(2) through (g)(6)(ii)(F)(10) by the first refueling 

outage after August 22, 2011. This is the basic implementing requirement imposing Code Case 

N-770-1 with conditions.

Condition §50.55a(g)(6)(ii)(F)(2) of the rule requires butt welds that rely on Alloy 82/182 for 

structural integrity that have been modified by inlays or onlays to be categorized as Inspection 

Items A-1, A-2, or B until the NRC staff has reviewed the mitigation and authorized an 

alternative Code Case for the mitigated weld or until an alternative Code Case Inspection Item 

is used based on conformance with an ASME mitigation Code Case endorsed in Regulatory 

Guide 1.147 with conditions, if applicable, and incorporated into 10 CFR 50.55a. This condition 

is intended to ensure that inlay and onlay mitigations are designed, installed, and examined in a 

manner that will ensure an acceptable level of quality and safety before credit can be taken for 

the inspection frequencies of Inspection Items G, H, J, and K, as appropriate.

Condition §50.55a(g)(6)(ii)(F)(3) of the rule requires that welds in Inspection Items A-1, A-2, and 

B, that have not received a baseline examination using Section XI, Appendix VIII requirements, 

shall be examined at the next refueling outage after January 20, 2012. The rule also provides 

provisions for when previous examinations can be credited for baseline examinations. Among 

other welds, this condition applies to butt welds modified by inlays or onlays that are required by

§50.55a(g)(6)(ii)(F)(2) to be categorized as Inspection items A-1, A-2, or B.

Code Case N-770-1, paragraph -2500(c) states that, “For axial and circumferential flaws, 

examination shall be performed to the maximum extent practical using qualified personnel and 

procedures. If essentially 100% coverage for circumferential flaws (100% of the susceptible 

material volume) can be achieved, the examination for axial flaws shall be completed to achieve 

the maximum coverage practical.” The NRC disagreed with this provision. Axial flaws can lead 

to through-wall cracks and leakage of reactor coolant, which is a safety concern. Condition 

§50.55a(g)(6)(ii)(F)(4) requires that the axial (flaw) examination coverage requirements 

of -2500(c) may not be considered to be satisfied unless essentially 100% coverage is

achieved. This condition was added for the NRC to ensure that, through NRC review of an 

authorization of alternative inspection coverage, appropriate actions are being taken to address 

potential inspection limitations for axial flaws. The industry guidelines of MRP-139 (EPRI 2005)

allow less than essentially 100% coverage in some cases. This condition of the rule applies to 

the examination of all inspection items, including G, H, J, and K.

Condition §50.55a(g)(6)(ii)(F)(5) of the rule requires all hot leg operating temperature welds in 

Inspection Items G, H, J, or K to be inspected each interval and supersedes the provisions of 

Table 1 that allow a sample inspection plan to be used. Condition §50.55a(g)(6)(ii)(F)(5) also 

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requires that a 25% sample of cold leg operating temperature welds must be inspected 

whenever the core barrel is removed (unless it has already been inspected within the past 

10 years) or has reached 20 years, whichever is less.

Condition §50.55a(g)(6)(ii)(F)(6) is an inservice inspection reporting requirement for mitigated 

welds if growth of existing flaws is found that exceeds the previous IWB-3600 flaw evaluations

or if new flaws are detected. In such cases, licensees are required to provide a report to the 

NRC, prior to entering Mode 4, that summarizes the licensee’s flaw evaluation with inputs, 

methodologies, assumptions, and cause of the new flaw or flaw growth. This condition applies 

to inlays and onlays as well as welds mitigated by other techniques. If volumetric examination 

detects new flaws or growth of existing flaws in the required examination volume, the mitigation 

will not be performing as designed and the NRC will need to evaluate the licensee’s actions to 

address the problem. Therefore, this condition was added to verify the acceptability of the weld 

prior to being placed back in service.

Condition §50.55a(g)(6)(ii)(F)(7) of the rule applies for Inspection Items G, H, J, and K when 

applying the acceptance standards of the ASME Code, Section XI, IWB-3514, Standards for 

Examination Category B-F, Pressure Retaining Dissimilar Metal Welds in Vessel Nozzles, and 

Examination Category B-J, Pressure Retaining Welds in Piping. For planar flaws contained

within the inlay or onlay, the thickness “t” in IWB-3514 is the thickness of the inlay or onlay. For 

planar flaws in the balance of the DMW examination volume, the thickness “t” in IWB-3514 is 

the combined thickness of the inlay or onlay and the DMW. NRC added this condition to ensure 

that IWB-3514 was used correctly, because other interpretations may be non-conservative.

These conditions were developed by the NRC to ensure that the level of quality and safety 

provided by the requirements for inlays and onlays is consistent with that provided by existing 

ASME Code and NRC requirements for butt welds that are not susceptible to PWSCC.

The NRC staff became aware of the concerns discussed in Section 6.3.1 on ET qualification 

and acceptance criteria subsequent to incorporating Code Case N-770-1 in 10 CFR 50.55a. 

Once these NDE reliability issues are resolved, Code Case N-770-1, with the NRC conditions 

on inlays/onlays, would be expected to provide an acceptable approach for monitoring potential 

PWSCC in inlays and onlays.

6.7 Conclusions

Based on a survey of operating experience with Alloy 82/182 butt welds, the authors did not find 

any instances of leaks in welds with inlays or onlays or flaws. No service-related operating 

experience involving inlays or onlays indicating crack growth was found, although the number of 

welds mitigated with inlays and onlays is small compared to the number of welds mitigated by 

weld overlays and MSIP and information on inspections performed did not appear to be 

available unless requested from each utility.

The acceptance standards in Code Cases N-770-1 and N-766 for indications detected by ET 

were written for PT and are not applicable to ET.

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The qualification requirements for ET in Code Cases N-770-1 and N-766 were not developed for 

flaws as small as may occur in inlays or onlays and, because they are based on technique 

demonstration rather than performance demonstration, are not considered by either NRC or 

PNNL to be as rigorous as needed to ensure effective surface inspections are performed.

The PWROG sponsored testing to evaluate the effectiveness of current automated UT

procedures applied to inner surface inspection of inlaid welds without modification of essential 

variables. EPRI concluded from this evaluation that the currently available UT procedures for 

automated inspection of welds from the ID could be used for inlaid welds and that additional 

qualification activities would not be needed. The results of this evaluation should also be 

applicable to onlay welds of similar thickness to the inlay welds tested.

ASME, Section XI, developed a revision to Code Case N-766 that was intended to address the 

NRC staff’s concerns and other issues that need to be clarified. Nevertheless, the revision does 

not address either of the staff’s concerns. With respect to the Code Case N-770-1 requirements 

on inlays/onlays, there are significant regulatory concerns on ET qualification and acceptance 

criteria.

ASME approved Code Case N-766, “Nickel Alloy Reactor Coolant Inlay and Cladding for Repair

or Mitigation of PWR Full Penetration Circumferential Nickel Alloy Welds in Class 1 Items” on

December 20, 2010, and issued it in Supplement 4 of the 2010 Edition of Section XI. The NRC 

staff, in accordance with ASME voting procedures, submitted a negative ballot on this Code 

Case. The major concerns of the staff were that the Code Case does not require a design 

analysis for flaw growth by PWSCC to verify adequate thickness of the inlay/onlay and the Code 

Case allows welds mitigated by inlays and onlays with large embedded flaws to be returned to 

service. Industry performed a generic PWSCC evaluation to address the first concern. ASME, 

Section XI is working on a revision to Code Case N-766 to address the NRC staff’s concerns 

and other issues that may need to be clarified.

Based on a review of the requirements of Code Case N-770-1, as implemented with NRC 

conditions, it is concluded that the inspection requirements for inlays and onlays, except for the 

outstanding issues noted above on ET qualification and acceptance criteria and on inlay/onlay 

thickness, provide an acceptable approach for monitoring potential PWSCC. For monitoring of 

potential PWSCC to perform as expected before and after mitigation by inlays and onlays, the 

NDE being applied has to be effective and reliable.

As noted, outstanding regulatory issues associated with the implementation of inlays and onlays

as a mitigation method pertain principally to analyses to verify the adequacy of the thickness of 

inlays and onlays and the adequacy of the NDE to be used pre- and post-mitigation. These 

issues will have to be resolved for the industry to obtain regulatory credit for the reduced 

examination frequencies in Code Case N-770-1. Assuming that effective and reliable NDE 

techniques are used to inspect inlays/onlays, inlays/onlays of appropriate thickness would be 

expected to provide effective mitigation against the growth of PWSCC into Alloy 82/182 welds to 

which they are attached.

6-21

7 OVERALL CONCLUSIONS

In the United States, primary water stress corrosion cracking (PWSCC) in butt welds has been 

found in the range of piping systems and sizes from drain lines to reactor coolant loop piping 

and at all Class 1 temperature conditions, including a drain line at the reactor coolant loop cold 

leg temperature.

Leakage to date through piping dissimilar metal butt welds that has been attributed to PWSCC has 

only been observed from axial cracks in unmitigated pressurizer and hot leg temperature welds.

In assessing the overall reliability of the ultrasonic testing (UT) examinations to detect and 

accurately characterize service-induced stress corrosion cracking (SCC), there are a number of 

factors that must be considered, including inspection volume coverage, the presence of 

interfering signals, the use of non-encoded versus encoded techniques, relative experience and 

skill of the examiner, use of modeling to design examination parameters, and representativeness of mockups used for performance demonstration qualifications. The use of encoded UT 

techniques offers a distinct and significant advantage over non-encoded techniques in detection 

and characterization of flaws, because of the ability to review the UT data off-line and carefully 

determine the origin of the signals present in the images produced.

The examination requirements of ASME Code Case N-770-1 appear to be complete and parallel 

existing Section IWB Class 1 requirements. For unmitigated welds the requirements reflect a 

philosophy of more frequent monitoring for an active degradation mechanism. Deterministic flaw 

growth analyses for unmitigated welds generally showed times for crack growth of initiated flaws 

10% to 25% through-wall to be shorter than Code Case N-770-1 inspection intervals. Baseline 

examinations have been completed and, although axial cracking and axial through-wall leakage 

have been found, no instances of circumferential or axial PWSCC have been observed that 

exceeded ASME Code structural factors. While deterministic analyses indicate that Code Case 

N-770-1 inspection intervals may be too long, additional bases would be needed to support a 

modification to Code Case N-770-1 inspection intervals. Such support could arise from 

unfavorable results of realistic probabilistic analyses and operating experience, including results 

of second and later rounds of examinations. Inspection limitations can alter the effectiveness of 

nondestructive examination (NDE). Geometry, surface conditions, or access may limit performing 

complete weld examinations. Information on geometric limitations to performing inspections of 

typical Westinghouse and Combustion Engineering plant reactor coolant system Alloy 82/182 

welds is contained in Appendices C and D. This information is applicable to unmitigated welds as 

well as pre- and post-mechanical stress improvement process (MSIP), pre-weld overlay (WOL)

examinations, and pre-inlay/onlay examinations.

Based on information searches, no instances of leaks were identified in welds mitigated by 

stress improvement. However, there have been instances in which cracks have been identified 

in boiling water reactor (BWR) welds previously mitigated by MSIP. Clearly these cracks were 

present when MSIP was applied. It is unclear whether these cracks continued to grow during 

subsequent plant operation. Operating experience with MSIP in BWRs highlights the 

importance of removing weld crowns and the benefit of performing effective and reliable 

7-1
examinations including, for example, the use of Appendix VIII-qualified, encoded phased-array 

UT techniques, prior to and subsequent to the application of MSIP.

Limited pre- and post-MSIP UT data from field welds with SCC indicates that MSIP may 

produce a decrease in flaw length measured by UT and a loss in UT flaw response as 

measured by signal-to-noise ratio.

Assuming that the performance criteria of Code Case N-770-1, Appendix I, are satisfied, it is 

reasonable to conclude that MSIP provides effective mitigation against the initiation of PWSCC 

and against the growth of any existing PWSCC that has been detected and allowed to remain in 

service. Therefore, based on a review of the requirements of Code Case N-770-1, as 

implemented with NRC conditions and the evaluation of MSIP as a mitigation strategy for 

PWSCC by Fredette and Scott (2009), it is concluded that these requirements for MSIP provide 

an acceptable approach for monitoring potential PWSCC.

Based on information searches, no instances of leaks in WOL welds or flaws propagating into 

overlays were identified. No service-related operating experience involving WOLs indicating 

additional crack growth was found.

In WOL mockup testing performed by Pacific Northwest National Laboratory all targeted flaws 

were detected in the pre- and post-WOL condition. Pre- and post-WOL length and depth-sizing 

root-mean-square error for flaws greater than 50% through the original wall thickness were 

found to be within the ASME depth-sizing qualification acceptance criterion.

An evaluation of WOLs using the criteria of Code Case N-770-1, Appendix I, has shown it to be

reasonable to conclude that WOLs provide effective mitigation against the initiation of PWSCC 

and against the growth of any existing PWSCC that has been detected and allowed to remain in 

service. Therefore, based on a review of the requirements of Code Case N-770-1, as 

implemented with NRC conditions and the evaluation of full structural weld overlays and 

optimized weld overlays as mitigation strategies for PWSCC by Fredette and Scott (2010), it is 

concluded that these requirements for WOLs provide an acceptable approach for monitoring 

potential PWSCC. 

During a survey of operating experience with Alloy 82/182 butt welds, the authors did not find 

any instances of leaks in welds with inlays or onlays, or unexpected flaws. No service-related 

operating experience involving inlays or onlays indicating crack growth was found, although the 

number of welds mitigated with inlays and onlays is very small compared to the number of 

welds mitigated by weld overlays and MSIP, and information on inspections performed would 

not be available unless requested from each plant owner.

There are a number of issues that need to be addressed for inlays and onlays in Code Cases 

N-766 and 770-1 concerning embedded flaws, the adequacy of inlay/onlay thickness for the 

prescribed inspection intervals, and qualification requirements and acceptance criteria for 

examinations by eddy current testing. Once these issues are resolved, the Code Cases would 

be expected to provide an acceptable approach for designing, installing, and examining welds 

mitigated by inlays/onlays and for monitoring potential PWSCC in inlays and onlays. 

For monitoring of potential PWSCC in mitigated and unmitigated welds to perform as expected, 

the NDE applied has to be effective and reliable.

7-2
8 REQUIREMENTS SUMMARY

This section contains a tabular summary of applicable NRC regulatory requirements, ASME 

Code Cases, and the regulatory status of applicable ASME Code Cases applicable to 

Alloy 82/182 butt welds. 

Applicable Requirement Description and Status, if Applicable

10 CFR 50, Appendix A, 

GDC 14

Reactor coolant pressure boundary (RCPB) is to be designed, fabricated, 

erected, and tested so as to have an extremely low probability of 

abnormal leakage, rapidly propagating failure, and gross rupture.

10 CFR 50, Appendix A, 

GDC 31

Fracture prevention of RCPB requires that the probability of rapidly 

propagating fracture of the RCPB is to be minimized.

10 CFR 50, Appendix A, 

GDC 32

Inspection of RCPB requires that components that are part of the RCPB 

have the capability of being periodically inspected to assess their 

structural and leak-tight integrity.

10 CFR 50, Appendix B, 

Quality Assurance 

Requirements, Criterion V

Instructions, Procedures, and Drawings requires that activities affecting 

quality shall be prescribed by documented instructions, procedures, or 

drawings of a type appropriate to the circumstances and shall be 

accomplished in accordance with these instructions, procedures, or 

drawings.

10 CFR 50, Appendix B, 

Quality Assurance 

Requirements, Criterion IX

Control of Special Processes requires that special processes, including 

nondestructive testing, shall be controlled and accomplished by qualified 

personnel using qualified procedures in accordance with applicable 

Codes, standards, specifications, criteria, and other special requirements.

10 CFR 50, Appendix B, 

Quality Assurance 

Requirements, Criterion 

XVI

Corrective Action requires that measures be established to assure that 

conditions adverse to quality are promptly identified and corrected.

Plant Technical 

Specifications

Reactor Coolant System Operational Leakage shall be limited to no 

pressure boundary leakage.

10 CFR 50.55a Codes and standards rule requires that ASME Code Class 1 components

meet the requirements of Section XI of the ASME Code. Section XI 

contains requirements for inspection, evaluation of inspection results, 

acceptance criteria for inspection results, repair and replacement, 

pressure testing, and qualification of NDE methods. Section XI 

inspection requirements, beyond inspection requirements of Code Case 

N-770-1, continue to apply for the detection of degradation mechanisms,

such as fatigue.

Rules for evaluation of flaws are contained in Section XI, IWB-3600 and 

Appendix C.

10 CFR 50.55a(b)(5) In-service Inspection Code Cases incorporates by reference Regulatory 

Guide 1.147, Revision 16, with conditions.

Code Case N-504-4 Code Case N-504-4 provides alternative rules for repair of Class 1, 2, and 

3 austenitic stainless steel piping by weld overlay. This Code Case was 

conditionally accepted by the NRC in Regulatory Guide 1.147, 

Revision 16. NRC imposed several conditions on the use of this Code 

Case.

8-1
Applicable Requirement Description and Status, if Applicable

Code Case N-638-4 Code Case N-638-4 provides rules of similar and dissimilar metal welding 

using the ambient temperature machine gas tungsten arc welding temper 

bead technique was conditionally accepted by the NRC in Regulatory 

Guide 1.147, Revision 16. NRC imposed two conditions on the use of 

this Code Case.

Code Case N-653-1 Code Case N-653-1 provides qualification requirements for detection and 

length and depth sizing for both service-induced and fabrication-induced 

flaws. It is applicable for wrought austenitic, ferritic, or dissimilar metal 

welds (DMWs), overlaid with austenitic weld material. Code Case 

N-653-1 has not been addressed in either Regulatory Guide 1.147 or 

1.193.

Code Case N-696-1 Code Case N-696-1 modifies the ultrasonic depth-sizing qualification 

acceptance criterion for piping 2.1 in. (54 mm) and greater in wall 

thickness, for examinations performed from the inside surface. The RMS 

error acceptance criterion is changed from 0.125 in. (3.18 mm) to 

0.250 in. (6.35 mm).

Code Case N-740-1 Code Case N-740-1 provides rules for DMW overlay for repair of Class 1, 

2, and 3 items. The NRC found Code Case N-740-1 unacceptable for 

use in Regulatory Guide 1.193, Revision 3. NRC listed Code Case 

N-740-2 in DG-1233 (Draft Regulatory Guide 1.193, Rev. 4), ASME Code 

Cases Not Approved for Use, which was issued in June 2013. The DG 

states that “(w)hile Revision 2 addresses some of the NRC staff 

concerns, significant issues remain.”

Code Case N-754 Code Case N-754 provides rules for optimized structural DMW overlays 

for mitigation of PWR Class 1 items. Code Case N-754 has not been 

addressed in either Regulatory Guide 1.147 or the draft Regulatory Guide 

on inservice inspection Code Case acceptability.

Code Case N-766-1 Code Case N-766-1 provides alternative rules for the design, installation, 

and examination of a weld inlay or onlay onto full penetration, nickel alloy, 

DMWs associated with Class 1 component nozzles and piping in PWRs. 

Code Case N-766-1 has not been addressed in either Regulatory Guide 

1.147 or 1.193.

Code Case N-770-1 Code Case N-770-1 contains requirements for visual and ultrasonic 

testing (UT) inspection of mitigated and unmitigated Class 1 Alloy 82/182 

butt welds and requirements for the evaluation of the inspection results. 

The NRC implementing requirements and conditions on the use of Code 

Case N-770-1 are contained in 10 CFR 50.55a(g)(6)(ii)(F)(1-10).

Code Case N-773 Code Case N-773 contains eddy current testing (ET) performance 

demonstration requirements that may be used in lieu of Appendix IV, 

Supplement 2 when ET is used to complement UT performed on the 

inside surfaces of austenitic, DMW, and clad piping welds.

Code Case N-824 Code Case N-824 provides rules that may be used in lieu of the 

requirements of the ASME Code, Appendix III, for ultrasonically 

examining cast austenitic piping welds from the outside surface.

8-2
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Overlays for Mitigating Primary Water Stress Corrosion Cracking at Nickel Alloy Butt Welds in 

Piping Systems Approved for Leak-Before-Break. PNNL-21660, Pacific Northwest National 

Laboratory, Richland, Washington. ADAMS Accession No. ML13330A803.

Sullivan EJ and MT Anderson. 2013. Assessment of the Mechanical Stress Improvement 

Process for Mitigating Primary Water Stress Corrosion Cracking in Nickel Alloy Butt Welds in 

Piping Systems Approved for Leak-Before-Break. PNNL-22070, Pacific Northwest National 

Laboratory, Richland, Washington. ADAMS Accession No. ML13330A972.

Thayer JK. 2007. Letter to JE Dyer. "Refined Crack Growth Calculations Supporting Industry 

Response to Wolf Creek Pressurizer Dissimilar Metal Weld Indications." February 14, 2007, 

Nuclear Energy Institute, Washington, D.C. ADAMS Accession No. ML070600674 and 

ML070600679.

Toloczko MB and SM Bruemmer. 2009. "Crack Growth Responses of Alloy 152 and 52 Weld 

Metals in Simulated PWR Primary Water -- White Paper." Pacific Northwest National 

Laboratory, Richland, Washington

Toloczko MB, MJ Olszta and SM Bruemmer. 2011. "Update on PNNL SCC Crack Growth 

Testing on Alloy 152, 52, 52M and 52MSS Welds." Presented at NRC - Industry 2011 Meeting 

on Alloy 690 Research, July 6-7, 2011, Rockville, Maryland. Available at 

http://pbadupws.nrc.gov/docs/ML1116/ML111661954.pdf.

Toloczko MB, MJ Olszta, RJ Seffens and SM Bruemmer. 2010. Preliminary Examination of 

Cracks in Alloy 152, 52, and 52M Weld Metals. PNNL-19119, Pacific Northwest National 

Laboratory, Richland, Washington. ADAMS Accession No. ML100471184.

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Institute and Industry Regarding Inlay and Onlay Applications on Dissimilar Metal Welds. U.S. 

Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession No. ML082770164.

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Institute and Industry Regarding Inlay and Onlay Applications on Dissimilar Metal Welds; 

Enclosure 2 - Slide 2, Weld Onlay Mitigation & Repair. U.S. Nuclear Regulatory Commission, 

Washington, D.C. ADAMS Accession No. ML082770353.

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Enclosure 3, Technical Papers 1 -- Technical Paper: Use of Weld Inlays as a PWSCC 

Mitigation for Reactor Vessel Safe-End-to-Primary Nozzle Welds, American Electric Power Lead 

Application. U.S. Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession No. 

ML082800043.

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Institute and Industry Regarding Inlay and Onlay Applications on Dissimilar Metal Welds; 

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Commission, Washington, D.C. ADAMS Accession No. ML082800061.

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October 21, 2008, Prairie Island Nuclear Generating Plant, Welch, Minnesota. ADAMS 

Accession No. ML083080007.

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June 9, 2014, U.S. Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession No. 

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ML082480225.

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EPRI-NRC Technical Exchange Meeting, May 25-26, 2010, Rockville, Maryland.

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2003, Gaithersburg, Maryland. U.S. Nuclear Regulatory Commission, Washington, D.C.

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APPENDIX A

MAJOR EVENTS AND ACTIVITIES

APPENDIX A

MAJOR EVENTS AND ACTIVITIES

A.1 Summary of Major Events and Activities

A.1.1 Introduction

Reactor coolant systems (RCSs) in pressurized water reactors (PWRs) have a number of butt 

welds containing Alloy 82/182 materials that are used in joining components. Welds containing 

Alloy 82/182 may be used in the butt welds that join ferritic vessel nozzles to stainless steel 

RCS loop piping. Some PWR designs may use Alloy 82/182 butt welds to connect stainless 

steel clad ferritic reactor coolant loop piping to stainless steel branch piping, such as emergency 

core cooling system piping. Operating experience has demonstrated that Alloy 82/182 materials 

exposed to primary coolant water (or steam) under the normal operating conditions of PWR 

power plants are susceptible to primary water stress corrosion cracking (PWSCC).

PWSCC in dissimilar metal butt welds was first observed in 1993 at Palisades because of a 

leaking circumferentially-oriented crack in an Alloy 82/182 weld that joined a stainless steel clad 

ferritic power-operated relief valve nozzle to an Alloy 600 safe end (NRC 1993). The Palisades 

event in 1993 is the only plant to date with a leaking circumferential crack determined to be 

PWSCC. A limit load analyses based on the circumferential crack size measured at the inner 

diameter (ID) resulted in a margin-to-failure of about 5 for faulted loads (Rogers 1993).

The next occurrence of PWSCC was found in 2000 because of a leaking axially-oriented crack 

at the V.C. Summer plant (Casto 2001; Cotton 2001).

Leaks due to axial PWSCC in butt welds have also occurred at Tsuruga, Unit 2 in Japan (NRC 

2004), at Davis-Besse in 2008 (NRC 2008a), and North Anna Power Station, Unit 1 (NAPS-1) in 

2012 (Anderson et al. 2012).

Indications of cracking both circumferential and axial have been found in numerous plants since 

2000. Domestic and foreign operating experience with PWSCC in butt welds is contained in 

Appendix B to this report. 

Prior to 2005, inspection of dissimilar metal butt welds was performed under the ASME Code 

Section XI requirements. In September 2005, the industry was provided with the final guidelines 

on the implementation of an industry “mandatory” initiative for inspection of Alloy 82/182 butt 

welds in PWRs to be performed on a more frequent basis than the inspections required by the 

ASME Code (Hartz 2005). This initiative was developed by the industry Materials Reliability 

Program (MRP) and is documented in MRP-139, “Primary System Piping Butt Weld Inspection 

and Evaluation Guideline,” Electric Power Research Institute (EPRI) Report 1010087 (EPRI 

2005). MRP-139 was issued under the implementation protocol of Nuclear Energy Institute

(NEI) 03-08 (discussed below) as an industry mandatory initiative. MRP-139 provided industry

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guidance for the volumetric and visual inspections of butt welds in PWR primary systems, which 

augmented the inspections of these locations that were already required by Section XI of the 

ASME Code.

In about 2005 industry began to mitigate welds by techniques such as weld overlays and the 

mechanical stress improvement process (MSIP). MRP-139 specifies inspection intervals for 

mitigated welds that are longer than the intervals specified for unmitigated welds and provides 

for sample inspections of mitigated welds. MRP-139 also specified schedules for plants to 

complete baseline inspections of all Alloy 82/182 welds within the scope of MRP-139 based on 

system temperature and pipe diameter. 

The U.S. Nuclear Regulatory Commission (NRC) staff reviewed the MRP-139 inspection 

program and provided comments to industry on its safety concerns in Mayfield (2005). One of 

the most important concerns the NRC staff raised dealt with MRP-139 not “requiring” volumetric 

inspections for piping less than 4-inches nominal pipe size (NPS). By issuing MRP-139: Interim 

Guidance, MRP 2007-038, on November 1, 2007, the MRP modified the MRP-139 guidelines to 

specify that Alloy 82/182 butt welds greater than or equal to 2-inches NPS but less than 

4-inches NPS and either exposed to temperatures equivalent to the hot leg or serve an 

Emergency Core Cooling System function (e.g., Babcock & Wilcox [B&W] High Pressure 

Injection nozzles) will be volumetrically inspected (King 2007). A complete listing and brief 

explanation of the MRP responses to the NRC staff’s concerns is contained in slides presented 

during a meeting between the EPRI MRP and the NRC staff held on February 21, 2008 

(Kammerdeiner 2008).

EPRI issued MRP-139, Revision 1, in 2008, EPRI Report 1015009 (EPRI 2008a) to address 

concerns raised by industry in the implementation of MRP-139 and to address some of the 

concerns the NRC staff raised. The MRP-139 ultrasonic testing (UT) baseline inspection 

schedules were revised by MRP-139, Revision 1, to specify that Alloy 82/182 welds were to be 

completed per the following schedule:

• By December 31, 2007, all Alloy 82/182 butt welds greater than or equal to 2-inches NPS 

associated with the pressurizer and exposed to pressurizer-like temperatures will be 

volumetrically inspected per this guideline (includes B&W pressurizer safety relief valve 

nozzle welds). Note that this applies to surge line nozzle welds near the pressurizer 

because of the potential for fatigue synergy.

• By December 31, 2008, Alloy 82/182 butt welds that are greater than or equal to 4-inches

NPS and less than or equal to 14-inches NPS and exposed to temperatures equivalent to 

the hot leg will be volumetrically inspected per this guideline. This implementation schedule 

also applies to the surge line nozzle weld at the hot leg and to the B&W makeup/high 

pressure injection nozzle weld.

• By December 31, 2009, Alloy 82/182 butt welds that are greater than 14-inches NPS and 

exposed to temperatures equivalent to the hot leg will be volumetrically inspected per this 

guideline.

• By December 31, 2010, Alloy 82/182 butt welds greater than or equal to 4-inches NPS that 

are exposed to temperatures equivalent to the cold leg will be volumetrically inspected per 

this guideline.

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The NRC staff monitored industry’s implementation of MRP-139 through NRC regional 

inspections conducted under Temporary Instruction (TI) 2515/172, “Reactor Coolant System 

Dissimilar Metal Butt Welds,” which the NRC issued in 2008 (NRC 2008b). This TI was revised 

in May 2010 (NRC 2010c) and expired at the end of June 2011. The regional inspection reports 

for TI 2515/172 were issued in integrated inspection reports for each plant. 

The industry actions to develop and implement the MRP-139 inspections allowed the NRC staff 

to pursue a process to codify requirements involving all stakeholders. The NRC concluded that 

the approach of working with ASME to revise inspection requirements and subsequently revise 

10 CFR 50.55a was both necessary and in the best interest of protecting public health and 

safety (NRC 2008a).

The ASME Code inspection requirements, incorporated by reference in 10 CFR 50.55a, provide 

a regulatory foundation for ensuring the integrity of pressure-retaining components. The NRC 

concluded that the current ASME Code requirements for inspection of Alloy 82/182 butt welds 

are not frequent enough to ensure that ASME Code-allowable limits will continue to be met in 

the event that PWSCC initiates. This conclusion was based on MRP-55 and MRP-115 crack 

growth rates (discussed below) developed by the EPRI/MRP and on the ASME Code maximum 

flaw depth allowable limit of 75% through-wall. This issue was addressed in the short term by 

the MRP-139 examinations. However, the NRC concluded that the ASME Code requirements 

needed to be revised to provide a regulatory framework for ensuring that ASME Code-allowable 

limits would not be exceeded, leakage would not occur, and potential flaws would be detected 

before they challenged the structural or leakage integrity of piping welds. The NRC concluded

that utility commitments to the industry initiative, MRP-139, would not provide a sufficient 

regulatory foundation for the long term (NRC 2008a).

In December 2005, the NRC sent a letter to ASME (Dyer 2005) requesting that the Section XI 

standards body address the inspection requirements for Class 1 PWR piping butt welds 

fabricated with Alloy 82/182 weld materials. In 2006 ASME approved the development of an 

ASME Code Case (Code Case N-770) on appropriate inspection requirements to address 

PWSCC in Class 1 butt welds containing Alloy 82/182 (Balkey 2006). ASME Code Case N-770 

was approved by ASME on January 30, 2009, and was published in Supplement 8 of the 2007 

Edition of the ASME Boiler and Pressure Vessel Code Nuclear Code Cases book. The title of 

ASME Code Case N-770 is “Alternative Examination Requirements and Acceptance Standards 

for Class 1 PWR Piping and Vessel Nozzle Butt Welds Fabricated with UNS N06082 or UNS 

W86182 Weld Filler Material With or Without the Application of Listed Mitigation Activities, 

Section XI, Division 1” (ASME 2009b).

Code Case N-770 was revised in response to a number of issues raised by the NRC. Code 

Case N-770-1 (ASME 2009a) was approved by the ASME in December 2009. On June 21, 

2011, the NRC issued a final rule (76 FR 36232, p. 36278) incorporating ASME Code Case 

N-770-1 into the regulations by reference. The new rule is contained in 10 CFR 

50.55a(g)(6)(ii)(F) and imposes Code Case N-770-1 with ten conditions the NRC concluded 

were necessary to ensure that these welds are appropriately monitored and assessed over their 

design life. This Code Case and the conditions imposed by the NRC in 10 CFR 

50.55a(g)(6)(ii)(F) are discussed in detail in Chapters 3–6 of this report.

A-3
The sections that follow discuss each of the major events involving PWSCC in primary system 

Alloy 82/182 butt welds and other significant industry and regulatory activities related to PWSCC 

in butt welds.

A.1.2 V.C. Summer

On October 7, 2000, during a normally scheduled containment inspection after entering a 

refueling outage, the licensee for V.C. Summer Nuclear Power Station identified a 

circumferential indication of PWSCC in the first weld between the reactor vessel nozzle and the 

‘A’ loop RCS hot leg piping. Specifically, the licensee found more than 200 pounds of boric acid 

crystals on the containment floor and protruding from the air boot around the ‘A’ loop RCS hot 

leg piping. Further examinations revealed a short through-wall crack in the hot leg nozzle safe 

end weld, approximately 0.9 m (3 ft) from the reactor vessel. Eddy current testing (ET) testing 

also identified similar small indications on the ID of four of the other five nozzle-to-pipe welds. 

The NRC issued Information Notice 2000-17 on October 18, 2000 (NRC 2000b) and 

supplemented the Information Notice on November 16, 2000 (NRC 2000a). The weld throughwall cracking was attributed to high residual stresses generated by extensive repairs made to 

this weld during fabrication. The reactor vessel nozzles at V.C. Summer are welded directly to 

reactor coolant piping, without safe ends. The hot leg and cold leg reactor vessel to piping 

welds were fabricated in the field (Casto 2001).

The licensee added a new section of stainless steel pipe with Alloy 52/152 welds to replace the 

30.5-cm (12-in.) section of the hot leg pipe containing the leaking weld. The licensee performed 

an analysis of the weld indications found by inspection in the Loops ‘B’ and ‘C’ RPV hot leg

nozzle welds and concluded from the analysis that the plant could be safely operated for a 

period of at least three years (Bamford et al. 2000). Hence, the licensee proposed to operate 

for two refueling cycles. The NRC staff performed an independent analysis of the flaws found 

by the licensee and determined that the licensee’s request for two fuel cycles of operation was 

not justified (Cotton 2001). The NRC concluded that the unit could be operated with ET 

indications in its Loops ‘B’ and ‘C’ RPV hot leg nozzle welds for one fuel cycle, at which time the 

weld indications would be re-examined. MSIP was subsequently applied to the Loops ‘B’ and 

‘C’ RPV hot leg nozzle welds during Refueling Outage 13 (Stuart 2002).

In response to this event, the NRC formed a Special Inspection Team to determine the 

adequacy of the licensee's previous inspection, confirm that the licensee had completed an 

analysis and examination to determine the root cause, and review the overall corrective action 

plan and the extent of the condition. Through that Special Inspection Team, the NRC identified 

potentially generic issues involving limitations of required nondestructive examinations to detect 

certain small inside-diameter stress-corrosion cracks, and the potential for multiple weld repairs 

to result in high residual stresses, which can contribute to stress-corrosion cracking (Casto 

2001).

In a letter dated December 14, 2000, the Nuclear Energy Institute informed the NRC that the 

Materials Reliability Project intended to lead the industry's actions to address the generic 

implications of the cracking experienced at V.C. Summer (Modeen 2000).

A-4
A.1.3 Wolf Creek

On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed pre-weld 

overlay UT inspections on the surge, spray, relief, and safety nozzle-to-safe end dissimilar metal 

and safe end-to-pipe stainless steel butt welds. The inspection identified five circumferential 

indications in the surge, relief, and safety nozzle-to-safe end dissimilar metal (DM) butt welds 

that the licensee attributed to PWSCC (Garrett 2006) and that were significantly larger and more 

extensive than previously seen in the industry (NRC 2007).

During a public meeting with the industry on December 20, 2006 (Mensah 2007), the NRC staff 

presented the results of a fracture mechanics-based scoping study that assessed the safety 

significance of the UT indications found at Wolf Creek. As a result of these analyses (Rudland 

et al. 2007a), the NRC staff concluded that there may be little or no time margin between the 

onset of leakage and rupture in pressurizer nozzle DM butt welds containing flaws similar to 

those found at Wolf Creek.

Over the course of many NRC staff discussions, the staff considered a number of options to 

assess what regulatory action to take to address this issue. The staff concluded that licensees 

needed to complete inspections or mitigations of the pressurizer nozzle Alloy 82/182 welds by 

the end of 2007 and implement interim enhanced leakage monitoring. This decision was based 

on the judgment that completing the actions by the end of 2007 would provide an appropriate 

balance of restoring safety margins within a time frame that would avoid compromising weld and 

inspection quality without placing undue reliance on the compensatory measure of enhanced 

leakage monitoring. A discussion of the risk-informed process used to recommend this decision 

is documented in Evans (2007c).

In March 2007, the NRC issued Confirmatory Action Letters (CALs) to 40 licensees with PWRs, 

confirming commitments from those licensees to resolve concerns regarding potential flaws in 

specific RCS Alloy 82/182 butt welds by performing inspections by the end of 2007. One 

example of the CALs issued was the CAL sent to Seabrook Station dated March 12, 2007 (Dyer 

2007a). The remaining 29 PWR plants had either completed the requisite actions or do not 

have welds susceptible to these flaws (Evans 2008b).

Nine of the plants receiving CALs did not have outages scheduled in 2007. These plants 

committed to accelerate outages into 2007 if the industry was not able to demonstrate an 

adequate level of safety to the NRC. The nine plants were Braidwood 2, Comanche Peak 2, 

Diablo Canyon 2, Palo Verde 2, Seabrook, South Texas Project 1, V.C. Summer, Vogtle 1, and 

Waterford 3) (Evans 2008b).

By letter dated February 14, 2007, the Nuclear Energy Institute informed the NRC staff that the 

EPRI MRP would be undertaking a task to refine the crack growth analyses pertaining to the 

Wolf Creek pressurizer dissimilar metal weld (DMW) ultrasonic indications (Thayer 2007). 

These studies would reduce unnecessary conservatisms and address some of the uncertainties 

in previous analyses (Thayer 2007). In a response to the letter from Thayer (2007), the NRC 

staff indicated that results from the proposed improved modeling will be useful for regulatory 

purposes provided several areas of uncertainty and potential non-conservatism are also 

addressed (Dyer 2007b). A discussion of the NRC staff’s concerns in the areas of uncertainty 

A-5
and potential non-conservatism is contained in the enclosure to Dyer (2007b). These refined 

crack growth analyses were referred to as advanced finite element analyses (AFEA) and were 

performed to address the NRC staff’s concerns regarding the potential for rupture without prior 

evidence of leakage from circumferentially-oriented PWSCC in pressurizer nozzle welds. The 

goal of the AFEA was to demonstrate that PWSCC in pressurizer DM butt welds would progress 

through-wall and exhibit detectable leakage prior to causing a possible rupture event. 

Industry took a two-phase approach to the AFE analyses. The Phase-I effort was conducted to 

establish the feasibility of using these uniquely developed crack growth codes by re-analyzing 

the critical case from the NRC staff’s scoping analysis (Rudland et al. 2007a). The NRC staff 

transmitted concerns with the Phase-I effort in a letter to NEI from Evans (2007a). The Phase-II 

effort was conducted to assess the pressurizer nozzle integrity of the nine PWRs through over a 

hundred sensitivity cases by varying nozzle-specific loads, dimensions, geometries, weld 

residual stresses, PWSCC crack growth rates, initial flaw assumptions, and other parameters.

Industry completed these analyses and documented the results in MRP-216, Revision 1, 

“Advanced FEA Evaluation of Growth of Postulated Circumferential PWSCC Flaws in 

Pressurizer Nozzle Dissimilar Metal Welds: Evaluations Specific to Nine Subject Plants” (EPRI 

2007a). MRP-216, Revision 1, concluded that PWSCC in pressurizer DM butt welds of the nine 

plants analyzed would progress through-wall and exhibit detectable leakage prior to causing a 

possible rupture event. These results were provided to the NRC staff by letter dated August 13, 

2007 (Marion 2007).

The NRC’s evaluation of MRP-216, Revision 1, concluded that there is reasonable assurance 

that the nine plants addressed by this evaluation would be able to operate safely until their next 

scheduled refueling outages in the spring of 2008. The NRC’s safety assessment (NRC 2007)

was transmitted to C. Haney, Director of the NRC Office of Nuclear Regulatory Research

Division of Licensing, by a memorandum from Evans (2007b). This safety assessment provides 

the basis for notification letters sent to the licensees of the nine plants planning to inspect the 

pressurizer nozzle dissimilar metal butt welds in the spring of 2008. This safety assessment is 

supported by a report that contains the NRC staff’s independent confirmatory analyses 

performed by the NRC Office of Nuclear Regulatory Research (Rudland et al. 2007b) . The 

NRC staff’s independent analyses enabled the staff to perform an in-depth review and critique 

of the industry’s analyses and to extend the industry’s analyses in some key respects.

A.1.4 Davis-Besse

On January 4, 2008, while Davis-Besse was in cold shutdown with the reactor head in place 

and fuel in the reactor vessel, the licensee began to install a weld overlay on the decay heat 

removal drop line-to-RCS nozzle weld (NRC 2008a). This weld is located inside containment. 

The RCS was in a drained condition. During the first weld pass on the drop line weld, the 

welding operator noticed water seeping from the weld. Visual inspection revealed a small leak. 

Surface examinations before welding showed no abnormal conditions or leakage. The licensee 

subsequently determined that the leak was from an axial through-wall flaw that the licensee 

attributed to PWSCC. The licensee applied a full structural weld overlay (FSWOL) to this weld.

A-6
A.1.5 Inspection of a Retired Pressurizer

In mid-February 2008, the NRC staff received the results of initial inspections of the nozzles of a 

retired pressurizer. These initial inspections were provided in MRP 2008-012 (King 2008c) . 

This pressurizer was removed from service to eliminate the possibility of extended plant outages 

resulting from cracking associated with the heater sleeves and replaced with a new pressurizer 

fabricated with materials with increased resistance to PWSCC. The pressurizer was donated to 

NRC for research purposes and the inspections were performed by EPRI to assist in 

determining the research value of the nozzle welds. These inspections found indications by dye 

penetrant (PT) and manual-phased array ultrasonic examination (UT). Circumferential and axial 

indications were found in five of six nozzles. The nozzle welds of most interest were the three 

safety nozzles. The inspection concluded that these nozzles had 360°, circumferentiallyoriented indications with non-uniform depths around the circumference. The deepest 

indications found were sized at 89% through-wall on the ‘A’ safety nozzle. The deepest 

indication found in the ‘B’ and ‘C’ safety nozzles were 75% and 69% through-wall, respectively.

Based on this information, NRC staff determined that the inspection results needed to be 

evaluated against the AFEA work the staff completed in September 2007 because the AFEA 

formed the basis for the continued operation of plants with pressurizer welds that had not yet 

been inspected, as mandated by industry guidelines. To help perform such an evaluation, the 

NRC staff requested that EPRI estimate the flaw profile for safety nozzle ‘A’ and provide some 

of the raw UT signals recorded during the inspection. EPRI provided this information to the 

NRC staff by letter MRP 2008-014, dated March 4, 2008 (King 2008a). EPRI estimated that the 

‘A’ safety nozzle weld contained a continuous deep indication 360° around the circumference. 

This reported flaw profile was more severe than any of the predicted flaws in the abovereferenced advanced finite element analyses that led to leakage that would be detected with 

sufficient time for plant shutdown prior to rupture. The flaw profile caused NRC staff to question 

whether the AFEAs would still support the spring 2008 pressurizer inspection schedules.

In making a regulatory decision to address the retired pressurizer nozzle weld inspection results 

in MRP 2008-012, the NRC staff considered three options. Option 1, the base case, would 

result in no change to existing regulatory and industry programs; that is, the affected plants 

would operate until their scheduled spring 2008 outage and inspect or mitigate the affected 

welds at that time. Option 2 would allow continued operation of the plants for a short time 

period while NRC staff gathered additional information. Option 3 would require all affected 

plants to shut down immediately and not restart until the basis for operation was reestablished 

or until inspection or mitigation activities were completed. The NRC staff based its regulatory 

decision on an assessment of the pros and cons of these options using the principles of riskinformed decision making.

On March 7, 2008, the NRC staff concluded that Option 2 was the appropriate decision; the staff 

judged that it had an appropriate basis to take a short period of time (e.g., within a week) to 

gather information to make a more informed decision. The initial inspection results were 

somewhat uncertain given that they were manual inspections with no recorded UT data. More 

refined inspection was judged to be prudent to reduce uncertainty regarding the results of the 

initial inspections. The NRC staff determined that the questions raised as a result of its review 

of the March 4, 2008, EPRI letter (King 2008a) were safety-significant questions and the NRC 

A-7
staff had verbally informed industry executives that it was considering regulatory action. A 

discussion of the risk-informed decision making process used to recommend Option 2 is 

documented in Evans (2008a).

Industry representatives took actions to rapidly put in place a comprehensive inspection effort 

that consisted of more advanced UT examinations, specifically ASME Code, Section XI, 

Appendix VIII-qualified encoded phased-array and radiographic testing (RT) examinations, 

followed a few days later by eddy current testing (ET) examinations. These inspections 

commenced on March 8, 2008, at the Studsvik-RACE facility in Memphis, the location of the 

nozzles from the retired pressurizer (Evans 2008b).

On March 9, 2008, an NRC inspector from Region I, and an expert in nondestructive 

examination (NDE) from PNNL, arrived at the Studsvik-RACE facility in Memphis. These NRC 

representatives immediately began auditing the weld inspection activities by the industry. Their 

audit included review of the UT inspection procedure, adherence to the UT procedure by the 

industry inspectors, the validity of the qualification of the inspection procedure by the 

Performance Demonstration Initiative (PDI) to ASME requirements, and evaluation of the UT 

inspection data. Their audit report also addressed the ET examinations performed on the inside 

surface of the safety nozzles. The inspectors concluded that the UT procedure was a qualified 

procedure with one exception that EPRI subsequently addressed. The NRC representatives 

verified that the industry inspectors followed the requirements of the UT procedure (Evans 

2008b).

As recommended in the March 4, 2008, letter from EPRI (King 2008a), EPRI concluded that a 

PDI-qualified automated phased-array UT (PAUT) technique would provide a more accurate 

profile of any potential degradation in the welds than a PDI-qualified manual PAUT technique. 

The NRC staff agreed with this conclusion. Phased-array ultrasonic inspection is a good 

technique for detecting critical flaws especially in welds with limited access and difficult 

microstructures. As with conventional manual UT, with manual phased-array ultrasonics the 

inspector is physically scanning the weld while looking at the equipment screen and doing data 

evaluation in real time. Although screen shots of areas of interest can be recorded, all data 

analysis is done simultaneously with the manual UT scanning. In comparison with encoded 

PAUT examination, the weld is scanned and a full set of position encoded ultrasonic data is 

recorded. This means that the equipment is recording everything that the inspector performing 

the manual scan was seeing, but each ultrasonic waveform is recorded along with the position 

information. This is a critical difference between the two methods in that the inspector can go 

back and carefully process and review the data and create a set of images that enable clearer 

interpretations of the data. These images include "B," "C," and "D" scans where the "B" scan 

shows a projected side view of the weld, the "C" scan shows a projected top view of the weld, 

and the "D" scan shows a projected end view of the weld. In all three views, the software can 

project a more three-dimensional-like profile of the weld on top of the ultrasonic data so that 

using these views, the inspector can easily visualize and analyze the data (locate and size flaws 

in the material). Thus, the ability of the encoded PAUT inspection to provide accurate 

representations of any flaws in the inspection volume and to characterize indications is superior 

to that of the manual PAUT inspection (Evans 2008b).

A-8
The automated PAUT examinations began on the evening of Saturday, March 8, 2008, and 

continued into the following week. Each of the St. Lucie pressurizer safety nozzle dissimilar 

metal welds was inspected.

On March 13, 2008, EPRI provided to the NRC the draft automated PAUT examination results 

summary. This document stated, in part, that the retired pressurizer safety nozzle welds ‘A,’ ‘B,’ 

and ‘C’ had multiple embedded fabrication flaws. This document concluded that these 

fabrication flaws were attributed to slag, porosity, and/or lack of fusion. Also, these indications 

were found to be clustered as well as individual fabrication flaws. The document discussed that 

the ET examination identified several small circumferentially-oriented linear indications in safety 

nozzle weld ‘A’ located at intermittent circumferential weld positions which appear to originate in 

the weld near the weld-to-austenitic base material interface (on the opposite side of the weld to 

the buttering). There was also an indication from a machining flaw due to a fabrication boring 

operation done from both the vessel ID and the safe end. The boring operations were not 

aligned properly and resulted in an eccentric step. The indication from the eccentric step could 

be confused with a root signal, but it was about an inch away from the actual weld root. This 

eccentric step was confirmed by ET during the reexamination process and demonstrated to the 

PNNL NDE expert who observed the reexamination activity. 

Based on the UT results, the EPRI document concluded that the inspection identified no flaws 

within the welds that were connected with the surface (Evans 2008b). The final report on the 

nondestructive examination summary of retired pressurizer safety nozzles was provided to the 

NRC staff on March 19, 2008 (King 2008b). This document concluded, in part, based on the 

results of the volumetric and surface examination methods employed that the pressurizer nozzle 

welds contain benign fabrication defects, but no specific indication of stress corrosion cracking. 

The final version of the summary reports was reviewed and determined to be essentially 

identical to the draft reports reviewed previously. In support of this document on March 13, 

2008, EPRI provided a series of UT scans of the inspection data for independent NRC staff and 

contractor interpretation on site in Memphis (Evans 2008b).

NRC staff completed its review of the encoded PAUT results. The NRC staff found that there 

was sufficient data available to provide reasonable assurance that there were no structurally 

significant service-induced flaws within the retired pressurizer safety nozzle welds ‘A,’ ‘B,’ and 

‘C.’ The NRC found that (1) the assumption made in the development of the manual flaw profile 

that the stacked indications identified in EPRI Letter 2008-012 were connected to the surface 

and (2) the determination that the flaw profile provided in EPRI letter MRP 2008-014 was

potentially due to service-induced cracking, while conservatively based on the data available, 

were not supported by the evaluation of the encoded phased-array UT data (Evans 2008b). 

The NRC concluded that the AFEAs supported the continued operation of the nine plants 

beyond December 31, 2007, to their respective spring outages. Letters were sent to each 

operating PWR that had not yet completed inspections of the pressurizer safety valve Alloy 

82/182 welds informing them of the NRC’s decision. One example of the letters issued was the 

letter sent to Braidwood Station, Unit 2, dated March 14, 2008 (Haney 2008).

Upon completion of the nondestructive examinations, the portion of the ‘Safety A’ nozzle 

containing the DMW was removed and sent to a laboratory for destructive examination to verify 

the NDE results. The destructive evaluation confirmed the indications and flaws found within 

A-9
the retired pressurizer safety ‘A’ nozzle are fabrication defects with no evidence of PWSCC. 

The destructive evaluation confirmed the intermittent circumferentially-oriented linear indications 

identified by ET and showed these to be fabrication hot tears, including one which was depthsized at approximately 17% through-wall. The flaws were confirmed to be non-safety significant 

and did not challenge the structural integrity of the component (Weakland 2008).

A.1.6 North Anna Power Station, Unit 1

During a spring 2012 inservice inspection of an Alloy 82/182 weld in an inlet (hot leg) steam 

generator nozzle at NAPS-1, several axially-oriented flaws were not detected by the licensee’s 

manual UT technique. The flaws were subsequently detected as a result of outer-diameter 

(OD) surface machining in preparation for a full structural weld overlay. The machining 

operation uncovered the existence of two through-wall flaws, based on the observance of 

primary water leaking from the DMW. Further ultrasonic tests were then performed, and a total 

of five axially-oriented flaws, classified as PWSCC, were detected in varied locations around the 

weld circumference (Anderson et al. 2012).

The North Anna hot leg safe end-to-nozzle weld configuration has an approximate 11° OD taper 

from the thinner austenitic piping side up to the thicker carbon steel nozzle, and is typical of a 

DMW created during steam generator replacement at Westinghouse-designed plants. 

However, the level of OD taper exhibited by this particular design is not included as a blind 

performance demonstration mockup used by the industry’s PDI, which is administered by EPRI. 

For this reason, the licensee engaged EPRI to assist in the development of a technical 

justification to support the basis for a site-specific qualification (Anderson et al. 2012).

The NRC Office of Research requested PNNL to assess the manual UT technique that was 

applied at North Anna, and evaluate potential causes for the failure of the examination to detect 

these significant flaws. The purpose of the PNNL assessment was to provide insights as to the 

nature of the event and provide a technical basis for regulatory consideration (Anderson et al. 

2012).

To accomplish these objectives, PNNL was asked to perform the following activities: 

• Model the acoustic performance of the manual UT probe arrangement used at North Anna 

to analyze its potential to detect inside diameter (ID) surface-connected, axially-oriented 

flaws, given predicted sound field characteristics; 

• Visit the North Anna site to evaluate the actual probe and UT responses produced from sitespecific DMW flaw mockups; and 

• Assess the licensee’s technical justification, developed by EPRI, for qualification acceptance 

of the manual technique that was applied (Anderson et al. 2012).

PNNL determined from ultrasonic modeling of the specific probe parameters for the manual 

examinations employed on the subject DMW configuration at North Anna that insufficient 

acoustic energy would be available near the inner surface of the weld to assure detection of ID 

surface-connected axially-oriented flaws. These results further suggested inadequate 

performance for manual, real-time examinations to properly discriminate and characterize 

surface-connected flaws from other welding or metallurgical features in the inner one-third of the 

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weld volume on these DMWs. This finding was corroborated by site observations that showed

low signal-to-noise ratio (SNR) for the targeted mockup flaws in the presence of spurious 

indications that interfere with detection and classification of simulated ID-connected cracks. 

PNNL found the technical justification developed to extend PDI qualifications to the North Anna 

steam generator Alloy 82/182 welds to be inadequate to make a reasonable case for using the 

manual, non-encoded procedure that was employed (Anderson et al. 2012).

As a follow-on effort, Anderson et al. (2014) addressed remaining technical questions and 

issues that were raised during the first evaluation process. Anderson et al. (2014) assesses the

failure of the licensee’s applied UT technique to detect multiple flaws, describes subsequent 

studies conducted, and provides final results and conclusions. Conclusions from this follow-on 

effort include the following:

• The data acquired on the mockups show, in some cases, that weld fabrication and flaw 

implantation signal responses are higher in amplitude than nearby implanted target flaws. 

These conditions render assessments (detections) of flaw signal responses in the presence 

of such ambient noise highly improbable when using a manual non-encoded UT technique.

• Based on site observations and laboratory measurements of UT responses from the flaws 

and other artifacts during examinations conducted by PNNL on the mockups, one would not 

expect the manual, non-encoded technique employed to consistently detect axial flaws (with 

low numbers of false calls) without examiners having prior knowledge of where these flaws 

are located.

• The PNNL assessment of the EPRI NDE Center-developed Technical Justification

supporting the manual ultrasonic examination demonstration at NAPS-1 shows that 

essential variables were changed, which contradict the requirements of Appendix VIII.

• The work conducted in assessing non-encoded and encoded examination techniques on the 

mockups indicates that encoded scans not only provide enhanced detection rates, but also 

allow for post-analysis and diagnosis of issues as well as appropriate oversight of 

examination records.

PNNL findings suggest that employment of a site-specific approach for an examination must 

have a solid technical basis to support the methodology and this basis would need to be 

rigorously and systematically developed in order to meet the intent of Appendix VIII. As a result 

of the PNNL findings, industry committed in a meeting on July 20, 2012, to take a broad look at 

the NAPS-1 and other operating experience, and work to understand its impact and determine 

the appropriate industry corrective actions. An initial action by industry was the formation of the 

NDE Improvement Focus Group (NIFG). The NIFG efforts focused on DMWs where sitespecific mockups have been used and non-encoded examinations conducted. 

In 2013, the Electric Power Research Institute published the following reports to provide 

guidance for nuclear power plant owners with respect to improving the reliability of NDE. These 

reports are available from the EPRI website.

A-11
• Nondestructive Evaluation: Guideline for Conducting Ultrasonic Examinations of Dissimilar 

Metal Welds, Revision 1, EPRI Report 3002000091 (EPRI 2013a), and

• Nondestructive Evaluation: Performance Demonstration Initiative (PDI) Guidance for 

Improved Reliability in Ultrasonic Examinations – Guideline for Hands-on Practice PDI-GL001 Revision B, Site Specific Mockup Requirements for Dissimilar Metal Welds, EPRI 

Report 3002000204 (EPRI 2013b).

These reports include:

• Performance Demonstration Initiative (PDI), Site Specific Configuration Mockup 

Requirements for Dissimilar Metal Welds, Revision C; and

• Guidelines for the Application of Team Scanning for Ultrasonic Examination of Dissimilar 

Metal Welds (DMW).

NIFG requested that NRC staff review these guidance documents. The NRC staff provided 

comments to NIFG on these guidance documents, and these comments were discussed with 

NIFG prior to NIFG’s issuance of these two reports. It appeared that NIFG chose not to address 

the NRC staff comments by revising the guidance documents. This matter was discussed 

during a public meeting held on January 8–9, 2014 (Hardies 2014). As an action item from the 

January 2014 meeting, NIFG requested the NRC staff to formally transmit their comments by 

letter to NIFG. The NRC staff’s comments were transmitted to NIFG by letter to the industry 

MRP NDE Integration Committee dated June 9, 2014 (Lupold 2014). The NRC staff noted in 

this letter that it believes that these comments are substantive and necessary to improve the 

performance of NDE when using either site-specific mockups or team scanning. The NRC staff 

also requested that the NDE Integration Committee provide a response to these comments. At 

the time of issuance of this report, the NDE Integration Committee had not provided a response 

to the NRC staff’s comments.

A.1.7 Other Significant Industry and Regulatory Activities

Industry and regulatory activities such as the development of MRP-139 and ASME Code Case 

N-770-1 and its codification in 10 CFR 50.55a were discussed in the Introduction to this 

appendix. The following paragraphs discuss other significant activities that have taken place 

related to management of PWSCC in butt welds.

A.1.7.1 NEI-03-08, Guideline for the Management of Materials Issues

In May 2003, the industry adopted an initiative on materials management and issued Nuclear 

Energy Institute (NEI) 03-08, “Guideline for the Management of Materials Issues” (NEI 2003). 

The industry initiative and the associated NEI 03-08 guidance document established policy, 

direction, oversight, and support for industry programs involving the management of materials 

issues. The initiative committed each nuclear utility to adopt the responsibilities and processes 

described in NEI 03-08.

Through the activities described in NEI 03-08, the industry stated that it would ensure that its 

management of materials degradation and aging would be forward-looking. In addition, the 

industry stated that it would continue to rapidly identify and effectively respond to emerging 

A-12
issues and would emphasize safety and operational risk significance. Guidelines developed for 

power plant licensees under NEI 03-08 address assessment, inspection, repair, mitigation, 

replacement, and regulatory interface.

A.1.7.2 MRP-55 and 115, Crack Growth Rate Studies

The incidence of PWSCC of Alloy 600 components in the PWR RCS highlighted the need for 

qualified equations for crack growth rates (CGRs) to evaluate flaws found by inservice 

inspection of thick-walled parts, including welds. In 2002, the MRP developed CGRs for 

PWSCC of thick-wall components fabricated from Alloy 600 base material, such as reactor 

vessel head nozzles. The Alloy 600 CGR were issued in MRP-55 (EPRI 2002).

In 2005, EPRI issued MRP-115NP (EPRI 2004b) to extend the MRP-55 work to cover crack 

growth rates in Alloy 82/182/132 welds.

The CGRs of MRP-55 and MRP-115 were incorporated into the 2010 Edition of Section XI of 

the ASME Code in Non-mandatory Appendix C, Section C-8500, “Stress Corrosion Cracking 

Growth Rate.” The current Codes and Standards rule, 10 CFR 50.55a, incorporates by 

reference the ASME Code through the 2008 Addenda of Section XI. The 2010 Edition of the 

ASME Code Section XI, with possible conditions on the use of Section C-8500 SCC CGRs for 

PWRs, will be address in a future NRC rule.

A.1.7.3 MRP-169, Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt 

Welds in PWRs

By letter dated September 7, 2005, the NEI submitted Topical Report (TR) Materials Reliability 

Program: Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt Welds in 

Pressurized Water Reactors (MRP-169), Revision 0, to the NRC staff for review (Marion 2005). 

Information on meetings between industry and NRC staff on the review of MRP-169, 

correspondence related to the NRC staff’s requests for additional information, and responses to 

those requests, including a revision to MRP-169 is provided in Blount (2010).

TR MRP-169, Revision 1, provides the technical basis for the design, analyses, and inspections

of the FSWOL and optimized weld overlays (OWOL) in PWRs (EPRI 2008b). The weld 

overlays are used to repair degraded nickel-based Alloy 82/182 DMWs or piping components 

made of Alloy 600 material by depositing weld metal on the outside surface of the welds and 

piping components. In addition, the overlays may be used as a preemptive measure to mitigate 

PWSCC.

Various aspects of TR MRP-169 are discussed in Chapter 5 of this report, in particular 

Sections 5.2, WOL Design Requirements; 5.3.1, Inspection of Dissimilar Metal Weld Overlay 

Specimens; 5.5, Status of ASME Code Cases N-740-X, “Full Structural Dissimilar Metal Weld 

Overlay for Repair or Mitigation of Class 1, 2, and 3 Items, Section XI, Division 1,” and N-754, 

“Optimized Structural Dissimilar Metal Weld Overlay for Mitigation of PWR Class 1, Section XI, 

Division 1; and 5.6, Assessment of ASME Code Case N-770-1 Inspection Requirements. The 

details of the NRC’s safety evaluation of MRP-169 are contained in Blount (2010).

A-13
The NRC’s safety evaluation of MRP-169 concluded, in part, that:

“MRP-169, Revision 1, as revised by letter dated February 3, 2010 (Riley 2010),

adequately describes the methods for the weld overlay design, the supporting analyses 

of the design, the experiments that verified the analyses, and the inspection 

requirements of the overlaid DMWs. The design of the FSWOL and OWOL follows the 

intent of the acceptance standards of the ASME Code, Sections III and XI. The effect of 

the weld overlay on the residual stress benefits has been demonstrated by the analysis 

and verified by the laboratory experiments with mockups.

“The ASME Code, Section XI has not yet included requirements for the weld overlay 

application, and the NRC staff has not yet approved ASME weld overlay Code Cases

N-740-2 and N-754 in 10 CFR 50.55a via Regulatory Guide 1.147. Therefore, licensees 

planning to install an FSWOL or an OWOL on DMWs need to request relief from the 

ASME Code requirements before overlay installation. Licensees may reference TR 

MRP-169 in their weld overlay relief request as the technical basis for the weld overlay 

design.... The NRC staff concludes that MRP-169 has provided an adequate technical 

basis to demonstrate that the DMW overlaid with either the FSWOL or the OWOL will 

provide reasonable assurance that the structural integrity of the piping system will be 

maintained to perform its intended function. Therefore, the NRC staff approves the use 

of MRP-169.”

A.1.7.4 MRP-317, Welding Residual Stress Dissimilar Metal Butt-Weld Finite Element 

Modeling Handbook

In October 2009, the 10-year inservice inspection at Seabrook identified an axial indication in 

the DMW at one of the reactor outlet nozzles at 158°. The axial indication did not meet the 

acceptance standards of IWB-3514. An analytical evaluation was performed by the licensee per 

IWB-3640 to demonstrate how long the as-found flaw could be left in service before repair was 

required. The analysis concluded that the cracked weld would have an allowable service life of 

just under 36 months to remain in compliance with the requirements of Section XI (O'Keefe 

2010), which would have provided a full operating cycle for the utility to plan for and mitigate the 

weld. However, the NRC staff raised questions about the weld residual stress analysis used in 

the licensee’s flaw evaluation. 

As a result of the NRC staff’s questions, the industry undertook an effort to develop guidelines 

for a consistent approach for performing weld residual stress analyses. This effort resulted in 

the development of MRP-317, “Welding Residual Stress Dissimilar Metal Butt-Weld Finite 

Element Modeling Handbook” (EPRI 2011). This handbook addressed subjects such as model 

geometry issues (e.g., design versus fabrication, weld bead geometry and sequencing, finite 

element mesh considerations), material property considerations, thermal models and welding 

parameters, structural modeling considerations, and model validation and verification. The NRC 

staff did not provide any formal comments or conclusions on this report.

This Seabrook reactor vessel outlet nozzle weld was mitigated by MSIP during the Fall 2009 

refueling outage on an emergent basis (NuVision 2009).

A-14
A.2 Timeline of Major Events and Activities

Date Event or Activity

September 1993 Circumferential through-wall crack at Palisades in the heat-affected zone of the 

power-operated relief valve Alloy 600 safe end.

1999 Two part-through-wall axial flaws found by UT in Ringhals, Unit 3, in a reactor 

vessel hot leg nozzle-to-safe end weld. 

2000 Four part-through-wall axial flaws found by UT in Ringhals, Unit 4, in a reactor 

vessel hot leg nozzle-to-safe end weld. 

October 2000 Axial through-wall crack in a reactor vessel nozzle-to-loop piping weld occurs at 

V.C. Summer.

May 2003 NEI issues NEI 03-08, Guidelines for the Management of Materials Issues.

2003 Boric acid deposits from an axial through-wall flaw found on the pressurizer relief 

nozzle of Tsuruga, Unit 2, in Japan. Subsequent UT from the OD detected two axial 

linear indications in the relief nozzle-to-safe–end weld.

September 2005 Industry implemented an initiative for more frequent inspections of Alloy 82/182 butt 

welds in PWRs than inspections required by the ASME Code. Initiative documented 

in MRP-139, “Primary System Piping Butt Weld Inspection and Evaluation 

Guideline.” 

December 2005 NRC staff sent a letter to ASME Section XI requesting Code development of rules 

for inspection of Alloy 82/182 butt welds.

October 2006 Wolf Creek UT inspections of pressurizer Alloy 82/182 butt welds identified five 

circumferential indications in surge, relief, and safety nozzle-to-safe–end butt welds 

that licensee attributed to PWSCC.

February 2007 EPRI MRP initiated a task to refine PWSC crack growth analyses of Wolf Creek 

pressurizer weld UT indications to address the NRC staff’s concerns regarding the 

potential for rupture without prior evidence of leakage. Refined analyses referred to 

as advanced finite element analyses (AFEA).

March 2007 NRC issued Confirmatory Action Letters to 40 licensees with PWRs, confirming 

commitments from those licensees to resolve concerns regarding potential flaws in 

specific RCS Alloy 82/182 butt welds by performing inspections by the end of 2007.

September 2007 NRC documented its safety assessment on the industry’s AFEAs. NRC staff 

performed independent AFEA and concluded that there was reasonable assurance 

that the nine plants addressed by the evaluation could operate safely until their next 

scheduled refueling outages in the spring of 2008.

January 2008 Leakage observed from the Davis-Besse decay heat removal drop line-to-RCS 

nozzle weld after application of the first weld pass for a weld overlay. Leaking flaw 

determined by UT to be axial.

Mid-February 

2008

NRC staff received the results of initial inspections of the nozzle welds of a retired 

pressurizer that needed to be evaluated against the AFEA results completed in 

September 2007.

March 8, 2008 EPRI implemented Section XI, Appendix VIII-qualified encoded phased-array, 

radiographic testing, and eddy current examinations of the retired pressurizer nozzle 

welds.

March 14, 2008 NRC informed affected licensees of its findings that there was sufficient NDE data 

available to provide reasonable assurance that there were no structurally significant 

service-induced flaws within the retired pressurizer safety nozzle welds.

June 21, 2011 NRC issues final 10 CFR 50.55a rule that adopts the ASME Code through the 2008 

Addenda and incorporates Code Case N-770-1 with conditions.

April 2012 As a result of OD surface machining in preparation for installation of a FSWOL, two 

axial flaws were found to be leaking in NAPS-1 steam generator hot leg nozzle-tosafe–end weld. Further ultrasonic tests were performed, and a total of five axiallyoriented flaws, classified as PWSCC, were detected in varied locations around the 

weld circumference.

A-15
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Standards for Class 1 PWR [pressurized water reactor] Piping and Vessel Nozzle Butt Welds 

Fabricated with UNS N06082 or UNS W86182 Weld Filler Material With or Without the 

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A-18
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Service at Port St Lucie Unit 1." March 19, 2008, Electric Power Research Institute, Palo Alto, 

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Report. Electric Power Research Institute, Palo Alto, California. ADAMS Accession 

No. ML080701017.

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Examination Improvements Focus Group (NIFG) Guidance Documents." June 9, 2014, U.S. 

Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession No. ML14099A291.

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2005, Nuclear Energy Institute, Washington, D.C. ADAMS Accession No. ML052520325.

Marion A. 2007. Letter to J Dyer. "Transmittal of EPRI Report, “Advanced FEA Evaluation of 

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Welds (MRP-216, Revision 1): Evaluations Specific to Nine Subject Plants”." August 13, 2007, 

Nuclear Energy Institute, Washington, D.C. ADAMS Accession Nos. ML072410236, 

ML072410240, and ML072410241.

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2005, NRC Office of Nuclear Regulatory Research, Washington, D.C. ADAMS Accession 

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Mensah TM. 2007. Memorandum to SL Rosenberg. "Summary of the December 20, 2006, 

Category 2 Public Meeting Re: Meeting with the Nuclear Energy Institute (NEI) to Discuss 

Implications of Wolf Creek Dissimilar Metal Weld Inspections." March 13, 2007, U.S. Nuclear 

Regulatory Commission, Washington, D.C. ADAMS Accession No. ML070330381.

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December 14, 2000, Nuclear Energy Institute, Washington, D.C. ADAMS Accession 

No. ML003779569.

NEI. 2003. Guideline for the Management of Materials Issues. NEI 03-08, Nuclear Energy 

Institute (NEI), Washington, D.C. ADAMS Accession No. ML032190048.

A-19
NRC. 1993. Palisades Pressurizer Penetrations Safe End Crack Results in Primary Coolant 

Leakage. Licensee Event Report 255-1993-009, U.S. Nuclear Regulatory Commission (NRC), 

Washington, D.C. November 8, 1993.

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Information Notice 2000-17, Supplement 1, U.S. Nuclear Regulatory Commission (NRC), 

Washington, D.C. November 16, 2000. ADAMS Accession No. ML003769172.

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October 18, 2000. ADAMS Accession No. ML003760993.

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Information Notice 2004-11, U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. 

May 6, 2004. ADAMS Accession No. ML041260136.

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Postulated Primary Water Stress Corrosion Cracking Flaws in Pressurizer Nozzle Dissimilar 

Metal Butt Welds. U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. ADAMS 

Accession No. ML072400199.

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Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping. Regulatory 

Information Summary 2008-25, U.S. Nuclear Regulatory Commission, Washington, D.C. 

October 22, 2008. ADAMS Accession No. ML081890403.

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Welds. TI 2515/172, U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. 

February 21, 2008. ADAMS Accession No. ML073330037.

NRC. 2010c. Temporary Instruction 2515/172, Revision 1, Reactor Coolant System Dissimilar 

Metal Butt Welds. U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. May 27, 

2010. ADAMS Accession No. ML100890642.

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Conditions on a Reactor Vessel Nozzle at Seabrook. NuVision Engineering. Pittsburgh, 

Pennsylvania. Accessed May 7, 2013. Available at 

http://nuvisioneng.com/News/November102009.aspx (last updated November 10, 2009).

O'Keefe M. 2010. Letter to NRC Document Control Desk. "Seabrook Station, Inservice 

Inspection Examination Report." February 9, 2010, NextEra Energy Seabrook, LLC, Seabrook, 

New Hampshire. ADAMS Accession No. ML100490381.

Riley JH. 2010. Letter to T Mensah. "Changes to Materials Reliability Program (MRP-169 

Rev. 1), Technical Basis for Preemptive Weld Overlays for Alloy 82/182 Butt Welds in PWRs." 

February 3, 2010, Nuclear Energy Institute, Washington, D.C. ADAMS Accession 

No. ML100360470 and ML100360482.

A-20
Rogers DW. 1993. Letter to NRC Document Control Desk. "Palisades Plant – Reply to NRC

Request for Information Regarding the Pressurizer Safe End Crack – Critical Flaw Size and 

Margin to Failure Analysis." November 30, 1993, Palisades Nuclear Plant, Covert, Michigan. 

U.S. Nuclear Regulatory Commission NUDOCS Accession No. 9312080126.

Rudland DL, DJ Shim, H Xu and G Wilkowski. 2007a. Summary Report on Evaluation of 

Circumferential Indications in Pressurizer Nozzle Dissimilar Metal Welds at the Wolf Creek 

Power Plant. Engineering Mechanics Corporation of Columbus, Columbus, Ohio. ADAMS 

Accession No. ML071560398.

Rudland DL, DJ Shim, T Zhang and G Wilkowski. 2007b. Final Report on Implication of Wolf 

Creek Indications – Final Report to Nuclear Regulatory Commission, Washington DC. 

Engineering Mechanics Corporation of Columbus, Columbus, Ohio. August 31, 2007. ADAMS 

Accession No. ML072470394.

Stuart W. 2002. "V. C. Summer Hot Leg Repair Update." Presented at Post-Refueling Outage 

Meeting between NRC Region II and V.C. Summer Nuclear Station, August 27, 2002, 

U.S. Nuclear Regulatory Commission, Region II Office, Atlanta, Georgia. ADAMS Accession 

No. ML022460180. Slides are at ML022460095, pp. 11-21.

Thayer JK. 2007. Letter to JE Dyer. "Refined Crack Growth Calculations Supporting Industry 

Response to Wolf Creek Pressurizer Dissimilar Metal Weld Indications." February 14, 2007, 

Nuclear Energy Institute, Washington, D.C. ADAMS Accession No. ML070600674 and 

ML070600679.

Weakland D. 2008. Letter to MRP Technical Advisory Group. "MRP 2008-053, St. Lucie 

Unit 1, Retired Pressurizer Safety “A” Nozzle DM Weld Destructive Examination Report." 

August 12, 2008, Electric Power Research Institute, Palo Alto, California. ADAMS Accession 

No. ML082480225.

A-21

APPENDIX B

OPERATING EXPERIENCE WITH BUTT WELD PWSCC

APPENDIX B

OPERATING EXPERIENCE WITH BUTT WELD PWSCC

B.1 U.S. Operating Experience

Operating experience has demonstrated that Alloy 82/182 materials exposed to primary coolant

water (or steam) under the normal operating conditions of pressurized water reactor (PWR)

power plants are susceptible to primary water stress corrosion cracking (PWSCC). The first 

incidence of PWSCC that was observed in a butt weld occurred at Palisades in 1993 in the 

heat-affected zone (HAZ) of the power-operated relief valve (PORV) Alloy 600 safe end. 

PWSCC had been identified earlier than 1993 in components such as Alloy 600 steam 

generator tubes and Alloy 82/182 pressurizer control rod drive partial penetration welds. 

PWSCC in butt welds was observed in 2000 at V.C. Summer in a field-welded Alloy 82/182 

reactor vessel nozzle-to-reactor coolant system (RCS) piping weld. Since the V.C. Summer 

event, there have been well over a dozen indications recorded in butt welds in domestic PWRs 

that were attributed to PWSCC. Summaries of each event are provided below followed by an 

overview of U.S. operating experience event in Table B.1.

Palisades. On September 16, 1993, plant personnel identified a leak in the PORV line near the 

nozzle connection to the pressurizer (NRC 1993). The pressurizer had a stainless steel clad 

ferritic PORV nozzle joined by an Alloy 82/182 weld to an Alloy 600 safe end. The crack 

initiated in the HAZ of the Alloy 600 safe end. Nondestructive examination (NDE) and visual 

inspection found a circumferential crack approximately 76 mm (3 in.) in length, about 30% of the 

circumference.

Consumers Power indicated the crack initiated due to PWSCC in the HAZ of the PORV line-topressurizer nozzle safe end weld. The cracking mode was intergranular from the inside 

diameter pipe surface with the final 5% to 10% of crack growth being transgranular.

V.C. Summer. In 2000, a large accumulation of boric acid deposits observed during a refueling 

outage at V.C. Summer led to the discovery of cracking in the ‘A’ hot leg pipe-to-RPV nozzle 

Alloy 82/182 weld (Casto 2001; Cotton 2001; Byrne 2002; NRC 2008a). The weld had a 

through-wall axial flaw with a small circumferential component and other small part-through-wall 

axial flaws. Based on destructive examinations of the piping and the weld material that was 

removed, the licensee determined that PWSCC caused the flaws. The low-alloy (ferritic) steel 

and the stainless steel at the ends of the weld arrested the axial crack growth of the flaw.

The circumferential flaw grew a short distance through a portion of the nickel-based alloy butter 

on the inside of an undercut portion of the low-alloy steel nozzle. The circumferential flaw was 

arrested when it ran into the low-alloy steel nozzle. Because the sizes of the axial and 

circumferential cracks were bounded, it would not have been possible for these cracks to lead to 

a piping rupture. The safety significance of this occurrence was limited to leakage of primary 

coolant. If a circumferential flaw had initiated closer to the center of the weld, it could have 

B-1
caused a more safety-significant situation. An ultrasonic testing (UT) examination of the 

remaining nozzle-to-pipe welds in all the loops found no flaws.

The licensee performed eddy-current examinations of the vessel nozzle-to-pipe welds from the 

inside surface of the pipe. This technique is sensitive enough to find surface flaws but is not 

capable of determining their depth in thick-walled piping. Small axial and circumferential flaws 

were identified in the ‘B’ hot leg pipe-to-reactor pressure vessel (RPV) nozzle nickel-based alloy 

weld; a small circumferential flaw was identified in the ‘C’ hot leg pipe-to-RPV nozzle nickelbased alloy weld; and a small circumferential flaw was found in both the ‘A’ and ‘C’ cold leg

pipe-to-RPV nozzle nickel-based alloy welds.

The staff of the U.S. Nuclear Regulatory Commission (NRC) reviewed the evaluation submitted 

by the licensee to justify the continued operation of V.C. Summer without repairing existing eddy 

current testing (ET) indications in the ‘B’ loop and ‘C’ loop welds. The Westinghouse evaluation 

concluded that V.C. Summer could be operated for at least two fuel cycles without repairing 

existing ET indications. The staff performed an independent evaluation using a bounding 

PWSCC growth rate and concluded that V.C. Summer could be operated with ET indications in 

the ‘B’ loop and ‘C’ loop hot leg welds for one fuel cycle.

The flawed region of the RPV ‘A’ nozzle weld was removed and a new spool piece welded in 

place with an Alloy 52 weld. The mechanical stress improvement process (MSIP) was applied 

to the RPV ‘B’ and ‘C’ nozzle welds. The UT and ET examinations performed by the licensee in 

subsequent refueling outages found no evidence of crack growth.

Three Mile Island, Unit 1. During a refueling outage in October 2003, an indication was 

detected in a surge line nozzle-to-safe–end Alloy 82/182 weld at Three Mile Island, Unit 1 

(TMI-1) (NRC 2002). The nozzle is a 25.4-mm (10-in.) diameter, Schedule 140, American 

Society for Testing and Materials (ASTM) A-105, Grade 2, carbon steel product with an 

Alloy 82/182 filler metal butter and welded with Alloy 82/182 filler metal to an ASTM A-336 

Class F8M forged stainless steel safe end. The surge line nozzle is connected to the steam 

generator ‘A’ hot leg of the primary coolant loop and normally is operating at 317°C (602°F).

TMI-1 was performing planned manual UT of the surge line nozzle-to-safe–end weld and found 

an axial indication in the weld material. The licensee characterized the indication as spanning 

the width of the weld on the inside surface and extending 12 mm (0.48 in.) into the weld.

The indication was confined in the Alloy 82/182 weld material and stopped at the base metal 

interface on either side of the weld. The indication was in a region that was repaired during 

original fabrication. Based on the location, acoustic response, and operating temperature, 

TMI-1 concluded that the indication was due to PWSCC.

TMI-1 installed an Alloy 52 full structural weld overlay repair to maintain weld integrity.

B-2
Calvert Cliffs, Unit 2. As a result of the UT inspections at Calvert Cliffs, Unit 2, in the spring of 

2005, two reactor coolant system nozzles were identified with indications of flaws requiring 

disposition (Holm 2005). No through-wall leakage was detected during the inspections. Both 

nozzles were determined to have a flaw or flaws that could not be found acceptable under 

ASME Section XI, IWB-3600.

The RCS nozzles requiring repair were the 50-mm (2-in.) diameter 21 Hot Leg Drain Line and 

22A Cold Leg Letdown Line. These are carbon steel nozzles buttered using Alloy 82/182, and a

stainless steel safe end was welded to the buttered nozzle using Alloy 82/82. The 21 Hot Leg

Drain Line Nozzle UT identified one circumferential indication and two axial indications in the

dissimilar weld regions that required further disposition. An engineering evaluation of this 

nozzle determined that the circumferential indication was acceptable under ASME Section XI 

standards, while the axial indications were not. The results of this evaluation indicated repair 

would be required prior to returning Unit 2 to service. The 22A Cold Leg Letdown Nozzle UT 

identified one axial indication that required disposition. Evaluation of this nozzle determined that 

this indication could not be accurately sized in accordance with ASME Section XI requirements 

and repair was also required prior to returning Unit 2 to service. The welds were overlaid.

The licensee reported that the root cause of the 21 Hot Leg Drain Nozzle (circumferential 

indication) was an original fabrication defect, without evidence of having grown in service. This 

indication was determined to be acceptable under ASME Section XI standards. The two axial 

indications on the 21 Hot Leg Drain Nozzle were caused by primary water stress corrosion 

cracking.

The 22A Cold Leg Letdown Nozzle axial indication was not found to be a degraded condition, 

but was instead considered by the licensee to be a local ultrasonic anomaly. Repair was 

performed because of the inability to size the anomaly within ASME Section XI Code 

requirements.

D.C. Cook, Unit 1. Examinations performed at D.C. Cook, Unit 1, in spring 2005, detected 

unacceptable cracking in a pressurizer safety nozzle-to-safe–end Alloy 82/182 weld (1-PZR-23) 

(Fadel 2005a, b). The crack was oriented in the axial direction. The crack was contained within 

the Alloy 82/182 weld, and the characteristics were concluded to be consistent with stress 

corrosion cracking. The weld was overlaid.

Millstone, Unit 3. In October 2005, the licensee was performing UT examinations at Millstone,

Unit 3, in accordance with ASME Code, Section XI, requirements (Price 2006; EPRI 2007b). 

The licensee identified an axial flaw in the pressurizer spray nozzle-to-safe–end weld 03-X5641-E-T. Flaw sizing information was not available. The licensee applied a full structural weld 

overlay (FSWOL).

Calvert Cliffs, Unit 1. In 2006, during the spring outage season at Calvert Cliffs, Unit 1, a 

25% through-wall circumferential indication was found in a hot leg surge nozzle-to-safe–end 

weld, a 20% through-wall circumferential indication was located in a hot leg drain line weld, and 

an 8% through-wall axial flaw was found in a pressurizer relief valve nozzle-to-safe–end weld

(Spina 2006; NRC 2008a). The licensee addressed the potential for further growth of these 

flaws by PWSCC by application of MSIP to these welds..

B-3
Davis-Besse. During the 2006 spring outage season at Davis-Besse, an axial indication of an 

indeterminate depth was found in a cold leg drain nozzle weld (NRC 2008a). Geometric 

interference prevented the UT from measuring the depth of the flaw. The licensee assumed the 

flaw was the result of PWSCC. The licensee completed a FSWOL repair. Because of the 

nature of the phenomenon, PWSCC at cold leg temperatures has a lower probability of initiating 

than at hot leg temperatures.

Wolf Creek. On October 13, 2006, the Wolf Creek Nuclear Operating Corporation performed 

pre-weld overlay inspections using UT techniques on the pressurizer surge, spray, relief, and 

safety nozzle-to-safe end Alloy 82/182 and safe end-to-pipe stainless steel butt welds (Garrett 

2006; NRC 2008a). The inspection identified three circumferential indications in the surge

nozzle-to-safe end Alloy 82/182 butt weld, one in the relief nozzle-to-safe end Alloy 82/182 butt 

welds, and one in the safety nozzle-to-safe end Alloy 82/182 butt weld. The licensee attributed 

the indications to PWSCC. The sizing techniques used were not qualified; however, based on 

the sizes estimated by the examiners, some of the flaws were significantly longer 

circumferentially than flaws previously seen in the industry. All five flaws were estimated to be 

less than approximately one-third of the wall thickness in depth. The licensee applied FSWOLs

to these welds.

Farley, Unit 2. The licensee performed manual UT of the pressurizer Alloy 82/182 butt welds 

during a Farley, Unit 2, refueling outage in April 2007. Examinations scanning the pressurizer 

surge nozzle-to-safe–end weld for axially-oriented flaws detected an unacceptable axial 

indication (George 2007; NRC 2008a). The licensee performed examinations for 

circumferentially-oriented indications but did not complete the manual data analysis, deciding 

rather to perform phased-array UT, which is better suited to the analysis of complex geometries. 

Subsequently, the licensee performed encoded PAUT examinations on the pressurizer surge 

nozzle weld. The scan for axially-oriented flaws revealed an axial indication in the same area 

as the manual call, characterized with a depth of 7.9 mm (0.31 in.), which is approximately 20% 

through-wall. The scan for circumferentially-oriented flaws detected an unacceptable 

circumferential indication, approximately 150 mm (6 in.) from the axial indication. It identified 

the circumferential indication as approximately 75-mm (3-in.) long (outside diameter dimension), 

with a maximum depth of 12.7 mm (0.5 in.), which is approximately 33% through-wall, and 

located the indication at or near the butter-to-Alloy 82/182 weld interface at or near the inside 

diameter surface. Because the indications were in PWSCC-susceptible material, the licensee 

based its corrective action on the assumption that the indications were the result of PWSCC. 

The licensee applied a FSWOL to this weld.

Davis-Besse. On January 4, 2008, while Davis-Besse was in cold shutdown with the reactor 

head in place and fuel in the reactor vessel, the licensee began to install a weld overlay on the 

decay heat removal (DHR) drop line-to-RCS nozzle weld (NRC 2008a). This weld is located 

inside containment. The RCS was in a drained condition. During the first weld pass on the drop 

line weld, the welding operator noticed water seeping from the weld. Visual inspection revealed 

a small leak. Surface examinations before welding showed no abnormal conditions or leakage. 

The licensee subsequently determined that the leak was from an axial through-wall flaw that the 

licensee attributed to PWSCC. The licensee applied a FSWOL to this weld.

B-4
Crystal River, Unit 3. During a March 2008 outage to replace a degrading reactor coolant 

pump seal at Crystal River, Unit 3, the licensee identified two circumferential indications in a 

weld that joins the DHR system drop line to an RCS hot leg (NRC 2008a). Further evaluation of 

the DHR hot leg nozzle connection determined that the two circumferential indications were 

actually one indication about 380-mm (15-in.) long. The UT measured the maximum throughwall depth of the flaw at 65% in one localized area. The licensee’s fracture mechanics 

evaluation of the circumferential indication on the DHR drop line determined that the flaw did not 

exceed the structural requirements for pipe in ASME Code, Section Xl. The licensee applied a 

FSWOL to this weld.

Salem, Unit 1. The licensee for Salem, Unit 1, performed examinations on all eight reactor 

vessel nozzle hot and cold leg Alloy 82/182 welds during a fall 2008 refueling outage. All 

examinations were performed from the outside surface using a combination of automated and 

manual phased-array techniques (Braun 2009; Burritt 2009). During the examinations, a 

circumferential inside surface-connected flaw was reported on one of the hot leg Alloy 82/182 

welds. The flaw was measured to be 24% through-wall (16.1 mm [0.634 in.]) and 52.3-mm

(2.06-in.) long on the inside surface.

The licensee analyzed the flaw in accordance with IWB-3600 and found it to be acceptable for 

up to 36 months of operation before repair or re-inspection. Further evaluation of the flaw size 

determined that it was within the acceptable limits of NUREG-0313 (Hazelton and Koo 1988) for 

the application of MSIP; that is, below 30% through-wall (24.2% actual) and less than 10% of 

the circumference (2.3% actual). All eight reactor vessel nozzle welds were mitigated by MSIP.

Seabrook. During the October 2009 refueling outage inservice inspection (ISI) examinations at 

Seabrook Station, an axial 21% through-wall flaw was reported in a reactor vessel hot leg

nozzle-to-safe–end weld (O'Keefe 2010; Freeman 2011). The flaw was detected in the weld 

and did not meet the acceptance standards of IWB-3514. An analytical evaluation was 

performed per IWB-3640 to determine how long the as-found flaw could be left in service before 

repair was required. The analysis resulted in an allowable service life of less than 36 months to 

remain in compliance with the requirements of Section XI. The flaw was attributed to PWSCC. 

The licensee for Seabrook performed MSIP on this weld.

North Anna Power Station, Unit 1 (NAPS-1). On March 24, 2012, axial indications were 

found as a result of leakage in the ‘B’ reactor coolant loop hot–leg-to-steam generator nozzle 

weld (Anderson et al. 2012; Anderson et al. 2014). The leakage was found during the 

performance of work activities to support weld overlay work on the ‘B’ loop hot leg steam 

generator nozzle weld. The workers noted a small amount of water seeping from the indications 

in the nozzle weld area. The indications were in the area of excavation that was being

performed prior to installing the weld overlay. Approximately 1 inch of weld material thickness 

360° around the circumference had been removed prior to identification of the seepage.

An examination of this weld had been conducted prior to the weld excavation, and none of the 

five indications were detected. It has been determined that two of the cracks were greater than 

80% through-wall and three were greater than 40% through-wall. A full structural weld overlay 

was applied to this weld.

B-5
B-6

Table B.1 U.S. Operating Experience with Butt Weld PWSCC

Plant Year System/Location Flaw Orientation Estimated Flaw Size 

Palisades 1993 HAZ of the PORV Alloy 600 

safe end

Circumferential NDE and visual inspection found a 

circumferential crack ~76 mm (3 in

length, about 30% circumference

V.C. Summer 2000 RPV nozzle-to-pipe weld TW axial flaw spanning the width 

of the weld with a small 

circumferential component and 

other small part-TW axial flaws

TW axial flaw 

Three Mile 

Island 1

2003 Surge line safe end-to-RCS hot 

leg nozzle weld

Axial Spanned the width of the weld on t

surface, extending about 50% TW 

deepest location 

Calvert Cliffs 2 2005 RCS hot leg drain nozzle-tosafe end weldOne circumferential and two axial No sizing information available 

D.C. Cook 1 2005 PZR safety valve nozzle-tosafe–end weldAxial Sizing information not found 

Millstone 3 2005 PZR spray valve nozzle-tosafe–end weldAxial No sizing information available 

Calvert Cliffs 1 2006 Surge line safe end-to-RCS hot 

leg nozzle weld 

Circumferential 25% TW, 2.4-in. long with a 123⁄4 in

Calvert Cliffs 1 2006 RCS hot leg drain nozzle-tosafe end–weldCircumferential 20% TW, 0.45-in. long with a 2-7/8 

Calvert Cliffs 1 2006 PZR relief valve nozzle-tosafe–end weldAxial >10% TW, 0.6-in. long with a 6-1/6 

Davis-Besse 2006 Cold leg drain nozzle weld Axial No sizing information available 

Wolf Creek 2006 PZR surge, safety and relief 

valve nozzle-to-safe–end welds

Circumferential Estimated sizes of the five flaws in 

(2008b); sizing technique was not 

Farley 2 2007 PZR surge nozzle-to-safe–end 

weld

Axial and circumferential Axial 20% TW; circumferential 33% 

3.5-in. long 

Davis-Besse 2008 Decay heat removal to RCS 

nozzle weld 

Axial 100% TW 

Crystal River 3 2008 Decay heat removal to RCS 

nozzle weld 

Circumferential 65% TW deep in one localized area

the flaw depth under 1/3 wall; 15-in

Salem 1 2008 Reactor vessel hot leg nozzleto-safe–end weldCircumferential 24% TW and 2-in. long on the insid

diameter (ID)

Seabrook 2009 Reactor vessel hot leg nozzleto-safe–end weldAxial Sizing information not found 

North Anna 2012 Reactor vessel hot leg nozzleto-safe–end weldAxial 5 flaws with the deepest estimated 

TW prior to machining. Two flaws 

after machining to prep for a FSWO
B.2 Foreign Operating Experience

As noted in Section B.1, operating experience has demonstrated that Alloy 82/182 materials 

exposed to primary coolant water (or steam) under the normal operating conditions of PWR 

power plants are susceptible to PWSCC. The first incidence of PWSCC in butt welds that was 

observed in a foreign plant occurred at Ringhals, Unit 3, in 1999 in a reactor vessel hot leg

nozzle-to-safe–end weld. PWSCC had been identified earlier in foreign PWRs in components 

such as steam generator tube and pressurizer control rod drive penetrations made of Alloy 600 

and in Alloy 82/182 partial penetration welds. Since the Ringhals, Unit 3, event, there have 

been a number of indications recorded in butt welds that were attributed to PWSCC. 

Summaries of each event are provided below followed by an overview of foreign operating 

experience event in Table B.2. 

A summary of cracking in Japanese steam generator inlet nozzle-to-safe–end welds is

contained in Table 4-3 of MRP-349 (EPRI 2012a). This table indicates that, in addition to the 

Japanese plants discussed below, cracking was found in the steam generator inlet nozzle-tosafe–end welds at Genkai, Unit 1, in 2008 and at Tomari, Unit 2, in 2008. No additional 

information on the cracking at these two plants beyond that contained in MRP-349 was located. 

The number and size of the cracks found in the steam generator inlet nozzle welds at these two 

plants is bounded by cracking found at other Japanese plants.

Mihama, Unit 2. Mihama, Unit 2, is a Westinghouse-type, two-loop 500-megawatt electric 

(MWe) PWR. In November 2007, it was reported that the licensee (Kansai Electric Power Co.) 

was preparing to shot peen the ID surface of steam generator (SG) nozzle dissimilar metal 

welds (DMWs) in September 2007 for preemptive mitigation against the effects of PWSCC. The 

licensee performed visual and ET of the weld ID surface in September 2007 and identified a 

number of axial indications—1 by visual examination and 13 by ET at the SG ‘A’ inlet nozzle 

(Nomura et al. 2007). Follow-up liquid penetrant testing (PT) confirmed the presence of the ETdetected indications. The maximum length of the largest indication (No. 4) was determined to 

be approximately 17 mm (0.75 in.). Follow-up UT estimated the depth to be approximately 

13 mm (0.5 in.) and the crack was depth-sized at approximately 11.5 mm (0.45 in.) based on 

observation of the fracture surface. The 17-mm indication was comprised of intermittently 

connected multiple small axial cracks approximately 3–5 mm (0.12–0.2 in.) in length (Nomura 

2008a). The axial indications were detected at various locations around the circumference, all 

of which with the exception of one indication (No. 13), exhibited PWSCC-like characteristics and 

were contained within the Alloy 600 butter and weld. Indication No. 13 was identified in the 

stainless steel safe end near the Alloy 600 weld and exhibited branching along the grain 

boundary. UT of this indication from the ID surface estimated the depth to be less than 5 mm 

(0.2 in.). Based on observation of the fracture surface, the crack depth was estimated to be 

approximately 0.9 mm (0.04 in.).

The SG inlet nozzle is low-alloy carbon steel with an Alloy 600 buttering and is stainless steel 

clad on the ID. The main loop piping is stainless steel, approximately 940-mm (37-in.) outer 

diameter (OD), 778-mm (30.6-in.) ID with an 81-mm (3.19-in.) wall thickness, and connects to 

the nozzle via a stainless steel safe end approximately 160 mm (6.3 in.) in length. The DMW 

joint is of a double-V configuration made with Alloy 600/82 filler material.

B-7
A circumferential ring containing the safe end, weld, and Alloy 600 butter, which contained the 

indications, was removed from the system for physical and chemical analysis. Detailed 

metallographic analyses were performed on indications Nos. 4 and 13. The licensee concluded 

that machining of the ID surface of the DMW and stainless steel safe end resulted in high 

residual tensile stresses on the ID surface, which caused PWSCC in the DMW and cracking to 

form near the safe end weld–to–safe end interface. A spool piece was used to replace the 

removed circumferential ring and was attached to the nozzle with Alloy 690 weld material 

(Nomura 2008b).

Ohi, Unit 3. Ohi, Unit 3, is a Westinghouse-type four-loop 1127-MWe PWR (ANS 2011). In 

November 2008, it was reported that ET of the RPV outlet nozzle weld was performed in March 

2008 by the licensee (Kansai Electric Power Co.) prior to the water jet peening operation of 

DMWs. Japanese PWR utilities had been applying water jet peening to reactor vessel nozzle 

DMWs since 2004 to mitigate PWSCC at those locations (Nomura 2008a). During this 

examination, a small axial flaw was detected. Visual examination from the ID surface estimated 

the flaw to be approximately 3 mm (0.12 in.) in length. Follow-up UT sized the flaw to be 

approximately 10 mm (0.4 in.) in length and less than 5 mm (0.2 in.) in depth. A detailed visual 

examination revealed evidence of machining on the ID surface. The crack exhibited branching 

along the dendritic boundaries similar to those observed on SG nozzle inlet welds.

The Ohi, Unit 3, RPV outlet nozzle is made of low-alloy steel with Alloy 82 butter, clad with 

stainless steel. The adjoining safe end is stainless steel 316F, with dimensions of 882-mm 

(34.7-in.) OD, 736-mm (29-in.) ID, and approximately 74-mm (2.9-in.) wall thickness. The DMW 

is of a double-V configuration and is made with Alloy 600/82 weld filler material.

The crack was presumed by the licensee to be relatively shallow, so repair by defect removal 

was implemented. The repair involved excavation of material at 0.5-mm (0.2-in.) to 1.0-mm 

(0.04-in.) depth increments, followed by an overall depth measurement, and visual examination 

and ET of the excavation to verify the absence or presence of the crack. The location and 

length of the crack was mapped for each increment. The flaw was completely removed at a 

depth of 20.3 mm (0.8 in.) (Mensah 2007). The maximum length of the crack was 13.5 mm 

(0.53 in.) at a depth of 5.5 mm (0.22 in.). Throughout the entire depth of the excavation, the 

crack exhibited turns and branching similar to that associated with PWSCC and was contained 

within the Alloy 82 butter and weld.

Water jet peening was performed on the surface of the divot created by the excavation, and the 

plant was placed in service for one operating cycle without weld repair. At a later time, the divot 

was planned to be filled with Alloy 82 weld material; a groove would be machined around the 

circumference of the joint spanning the stainless steel clad nozzle, Alloy 600 butter, DMW and 

safe end; and an inlay of Alloy 52 would be applied within the machined groove.

Ringhals, Units 3 and 4. Ringhals, Units 3 and 4, are Westinghouse three-loop PWRs, 

1051 MWe and 935 MWe, respectively (ANS 2011). There were no documented repairs of safe 

ends in Ringhals 3, but there were two extensive and one more confined safe end repairs 

documented on Ringhals 4 (Skanberg 2001; EPRI 2004a; Miteva and Taylor 2006; NRC 

2008a).

B-8
In 1999 during regularly scheduled ISI of the Ringhals, Unit 3, RPV outlet nozzle-to-safe–end 

DMWs, two shallow axial indications were detected in one of the DMWs. These indications 

were flaws that were found to be acceptable by evaluation and were left in service. Follow-up 

examinations in 2000 and 2001 revealed crack growth as well as the presence of a new crack. 

The flaw at 265°/299° in 2000 measured 13 mm (0.5 in.) in length and 9 mm (0.35 in.) in depth; 

in 2001, the same flaw measured 18 mm (0.7 in.) in length and 13 mm (0.5 in.) in depth. The 

flaw at 265°/323° in 2000 measured 16 mm (0.63 in.) in length and 9 mm (0.35 in.) in depth; in 

2001 the same flaw measured 20 mm (0.8 in.) in length and 16 mm (0.63) in depth. The new 

flaw detected in 2001 at the RPV outlet nozzle 25°/323° measured 8 mm (0.3 in.) in length and 

8 mm (0.3 in.) in depth. These cracks were subsequently removed through a boat sample.

During regularly scheduled ISI in 2000 of Ringhals, Unit 4, RPV outlet nozzle-to-safe–end 

DMWs, four axial flaws were detected in one of the outlet nozzle-to-safe–end welds. Three 

were evaluated as subsurface planar flaws and one as a surface-breaking planer flaw. All four 

were removed by boat samples. Metallographic examination showed that all four cracks were 

surface-breaking. The two deepest flaws were determined by metallographic examination to be 

22 ± 3-mm (0.87 ± 0.1-in.) deep. The depths of these flaws according to ultrasonic 

measurements were 13 ± 3 mm (0.5 ± 0.1 in.) and 16 ± 3 mm (0.63 ± 0.1 in.). Metallographic 

examination of the flaws determined that stress corrosion cracking (SCC) occurred in a weldrepaired Alloy 182 region, branching and crack-tip characteristics were similar to those 

associated with SCC, hot cracking and minor lack of fusion was evident, and propagation of the 

SCC crack into the carbon steel nozzle or stainless steel safe end materials did not occur. 

Alloy 52M was used at both plants to replace/repair the material removed through boat samples.

The RPV outlet nozzle DMW joint is of a double-V configuration and consists of a carbon steel 

SA508 Class 2 nozzle clad with stainless steel, with Alloy 182 buttering, connected to a 

stainless steel SA182 F316 safe end with Alloy 182 weld filler material.

Takahama, Units 2 and 3. Takahama, Units 2 and 3, are Westinghouse-type three-loop 

PWRs, 780 MWe and 830 MWe, respectively (ANS 2011). In August 2008, it was reported by 

the licensee (Kansai Electric Power Co.) that indications of PWSCC were detected in the SG 

inlet nozzles of Takahama, Units 2 and 3 (Nomura 2008a). For Takahama, Unit 2, SG ‘A’ inlet 

nozzle had 3 indications detected by ET that could not be confirmed with UT; SG ‘B’ inlet nozzle 

had 2 indications detected by ET for which UT sized to be up to 7 mm (0.28 in.) in length and 

6 mm (0.24 in.) in depth; and SG ‘C’ inlet nozzle had 4 indications for which UT sized to be up to 

14 mm (0.55 in.) in length and 8 mm (0.3 in.) in depth. For Takahama, Unit 3, SG ‘A’ inlet 

nozzle had 7 indications detected by ET for which UT sized to be up to 28 mm (1.1 in.) in length 

and 9 mm (0.35 in.) in depth; SG ‘B’ inlet nozzle had 16 indications detected by ET for which UT 

sized to be up to 38 mm (1.34 in.) in length and 15 mm (0.59 in.) in depth; and SG ‘C’ inlet 

nozzle had 9 indications detected by ET for which UT sized to be up to 14 mm (0.55 in.) in 

length and 9 mm (0.35 in.) in depth. For each of the Takahama steam generator inlet nozzles, 

the indications were distributed at various locations throughout the entire circumference of the 

DMW.

No weld joint details or component material specifications were provided for the original SG inlet 

nozzle weld, but the original DMW filler material used at Takahama, Unit 2, was Alloy 600/82, 

and at Takahama, Unit 3, was Alloy 600/132.

B-9
The repair methods reported to be used were dependent upon the significance of the indication 

detected. For relatively small and shallow cracks, the defects would be removed by fine 

grinding or electro-discharge machining, followed by ID surface stress mitigation using shot 

peening. For relatively large cracks, the elbow would be removed to facilitate weld repair of the 

defects, then reinstallation of the elbow using either Alloy 600 or Alloy 690 weld filler material 

followed by surface prepping and installation of a circumferential weld inlay of Alloy 690 material 

that spanned from the nozzle to safe end.

Takahama, Unit 4. Takahama, Unit 4, is a Westinghouse-type four-loop 830-MWe PWR (ANS 

2011). In November 2008, it was reported by the licensee (Kansai Electric Power Co.) that 

indications of PWSCC were detected in the SG inlet nozzles of Takahama, Unit 4, during the 

performance of visual and ET examinations in preparation for shot peening of the DMW ID 

surface for preemptive mitigation against PWSCC (Nomura 2008b). In all, 36 indications were 

detected—7 in the inlet nozzle of SG ‘A’, 8 in the inlet nozzle SG ‘B’, and 21 in the inlet nozzle 

of SG ‘C’. UT sized the indications in SG ‘A’ to be up to 14 mm (0.55 in.) in length and 12 mm 

(0.47 in.) in depth; in SG ‘B’ to be up to 30 mm (1.2 in.) in length and 13 mm (0.5 in.) in depth; 

and in SG ‘C’ to be up to 33 mm (1.3 in.) in length and 16 mm (0.63 in.) in depth. For each of 

the SG inlet nozzles, the indications appeared to be contained within the Alloy 600 butter and 

Alloy 600/132 weld material, and were distributed over the circumference of the DMW joint.

No weld joint details or component material specifications were provided for the original SG inlet 

nozzle weld, but the original DMW filler material used at Takahama, Unit 4, was Alloy 600/132.

At the time this information was presented, it was the utility’s intent to remove the defects, fill the 

excavations as necessary, and apply a circumferential weld inlay of Alloy 690 material that 

would span the DMW joint from the stainless steel cladding on the SG inlet nozzle to the 

stainless steel safe end.

Tihange, Unit 2. Tihange, Unit 2, is a three-loop 1008-MWe PWR (ANS 2011). In October 

2002, indications were detected in five areas on the ID surface of the pressurizer surge line 

nozzle-to-safe–end weld (Roussel 2005). Four of the indications detected were circumferential 

and were found to be within the acceptable size limits of Section XI. The fifth indication, an axial 

indication, could not be distinguished easily, but was ultimately sized to be approximately 

26 mm (1 in.) in length and 4 mm (0.16 in.) in depth. The flaw was located close to a repair 

made during fabrication, and it was thought that the indication could be a fabrication flaw that 

went undetected at that time, but the possibility of PWSCC could not be excluded (Gérard and 

Daoust 2004a, b). The component was placed back in service without repair or mitigation and 

was reexamined, initially at six-month intervals, in May 2003 and October 2003, and then 

inspected again in May 2005. The results of those examinations did not differ from the October 

2002 results.

At the time of the October 2002 examination, the 1992 Edition without Addenda of Section XI of 

the ASME Code was the applicable inservice inspection code for all Belgian plants. UT 

examination of the subject weld was required to be performed to the requirements of Section XI, 

Appendix VIII with the UT examination volume defined to be the inner one-third of the thickness 

of the weld over an axial length equal to the width of the weld projected radially from the weld 

crown plus 6 mm (1⁄4 inch) from each side of the weld crown. The pressurizer surge line nozzle 

B-10
joint is composed of a SA508 Class 3 nozzle, which is connected to a stainless steel 316 safe 

end by an Alloy 600/182 weld (Miteva and Taylor 2006).

In May 2003, the RPV outlet nozzle-to-safe–end welds were inspected, and consisted of ID UT 

and ID ET examinations. The ID UT examination did not identify any significant indications, but 

the ET examination detected a 10-mm (0.4-in.) crack-like axial indication on the hot leg No. 2 

nozzle weld. Follow-up UT examination of the indication using UT probes specifically focused 

to detect cracks within 1 mm (0.04 in.) of the surface could not detect the indication, so the 

indication was presumed to be 10 mm × 1 mm (0.4 in. × 0.04 in.). The weld was reexamined in 

May 2005 and the results did not differ from those of the May 2003 examination.

The ultrasonic inspection of the dissimilar welds at the RPV nozzles was performed from the ID 

of the vessel. Qualification of the procedure was based on the requirements of Appendix VIII to 

Section XI of the ASME Code, Edition 1992 without addenda, but the European (European 

Network for Inspection Qualification) methodology was also used as a guideline. The 

qualification program took place in 1999.

ID detection of circumferential indications in the first (inner) third of the weld thickness is

achievable from 5-mm (0.2-in.) depth and characterization from 8-mm (0.3-in.) depth knowing 

that the dead zone can extend from 4 to 6 mm (0.16 to 0.24 in.). ID detection of axial 

indications in the first third of the weld is achievable from 5-mm (0.2-in.) depth and 

characterization from 5-mm (0.2-in.) depth.

The UT examination of DMWs of pressurizer nozzle-to-safe–end welds is performed from the 

OD using automated techniques, which are qualified to detect axial cracks from 2 mm (0.08 in.) 

in depth and to size them from 4 mm (0.16 in.) in depth.

Tsuruga, Unit 2. Tsuruga, Unit 2, is a Westinghouse-type four-loop 1160-MWe PWR. In 2003 

during a regularly scheduled ISI, boric acid deposits were found on the pressurizer relief nozzle 

(NRC 2004; Yonazawa 2004). Subsequent UT from the OD of all of the pressurizer relief and 

safety nozzles detected two axial linear indications in the relief nozzle-to-safe–end weld, the 

longest of which was 35 mm (1.38 in.), and one axial linear indication in the ‘A’ safety nozzle-tosafe–end weld approximately 50 mm (1.97 in.) in length. These indications were confirmed with 

radiographic testing from the OD and PT from the ID. All of the indications were contained 

within the weld material, including a portion of the nickel-based alloy butter on the inside of an 

undercut portion of the low-alloy steel nozzle.

The configuration of the pressurizer safety and relief nozzles are all the same; a carbon steel 

nozzle clad with stainless steel, buttered with Alloy 600, and attached to a stainless steel 

Type 316 safe end with Alloy 600 weld filler material. The OD of the nozzle is approximately 

190 mm (7.48 in.); ID is approximately 130 mm (5.12 in.) with a wall thickness approximately 

30 mm (1.18 in.).

The samples removed for destructive examinations contained the entire weld and a portion of 

the base metal on each side of the weld.

B-11
Radiography was performed on the severed pieces, confirming the linear flaws. Metallurgical 

failure analysis was performed on these samples. The results showed that the cracks initiated 

from the inside diameter surface, were axially oriented, and were intergranular or interdendritic 

in nature. A through-wall crack was confirmed at the 90° location in the weld on the relief 

nozzle. The conclusion of the metallurgical analysis was that the nozzle failures were caused 

by PWSCC in the nozzle weld.

Plant personnel stated that no indications were detected during the previous inspections.

Tsuruga, Unit 2. In November 2007, it was reported that the licensee (Japan Atomic Power 

Co.) was preparing to shot peen the ID surface of SG nozzle DMWs for preemptive mitigation 

against the effects of PWSCC. In September 2007, the licensee performed ET of the weld ID 

surface and identified 29 indications by ET on the SGs ‘A’, ‘B’, and ‘C’ inlet nozzle DMWs—1 on 

SG ‘A’, 5 on SG ‘B’, and 23 on SG ‘C’ (Nomura et al. 2007). Follow-up UT was performed on 

the nozzles; UT did not detect the indication in SG ‘A’. The largest indication in the SG ‘B’ inlet 

nozzle was sized to be approximately 21 mm (0.83 in.) in length and 12 mm (0.47 in.) in depth. 

The largest indication in the SG ‘C’ inlet nozzle was sized to be approximately 14 mm (0.55 in.) 

in length and 13 mm (0.5 in.) in depth. The axial indications in SGs ‘B’ and ‘C’ were detected at 

various locations around the circumference. The licensee indicated that the indication on SG ‘A’ 

and the largest indication on SG ‘B’ would be further investigated using Suzuki’s Universal 

Micro-Printing method; however, no results were identified by our review.

The SG inlet nozzles are low-alloy carbon steel with an Alloy 600 buttering, and are stainless 

steel clad on the ID. The main loop piping is stainless steel, has an approximately 943-mm 

(37.1-in.) OD and 788-mm (31-in.) ID with a 77.5-mm (3.05-in.) wall thickness, and connects to 

the nozzle via a stainless steel safe end. The DMW joint is of a double-V configuration made 

with Alloy 600/132 filler material (Nomura 2008a).

The affected SG nozzles were not removed. Local weld repairs were performed on the 

indications using either Alloy 600 or Alloy 690 material, followed by the installation of a 

circumferential weld inlay of Alloy 690 over the repaired areas.

B-12
B-13

Table B.2 Foreign Operating Experience with Butt Weld PWSCC

Plant Year System/Location Flaw Orientation Estimated Flaw Size 

Mihama 

Unit 2

2007 SG-A inlet nozzle-tosafe–end weldNumerous axial indications 

in weld. Shallow flaw in SS 

safe end near nickel-based 

weld.

Largest flaw measured to be 17-mm long 

and 12-mm deep. Wall thickness is 81 

mm.

2 samples 

metallurgic

PWSCC in 

cracking ne

to-safe–en

Ohi Unit 3 2008 RPV outlet nozzle-tosafe–end weld1 axial flaw identified. Flaw removed by grinding. Maximum flaw 

length was 13.5-mm long at 5.5-mm depth 

and the final flaw depth was 20 mm. Wall 

thickness is 74 mm.

Returned t

peening. L

an inlay.

Ringhals 

Unit 3

1999 RPV outlet nozzle-tosafe–end weld2 axial flaws in 1999. 

Additional axial flaw found 

in 2001.

Initial 2 flaws measured 9-mm deep in 

2000 and 13- and 16-mm deep in 2001. 

New flaw found in 2001 measured 8-mm 

deep.

Left in serv

removed b

repaired by 

Ringhals 

Unit 4

2000 RPV outlet nozzle-tosafe–end weld4 axial flaws found by UT. 

Confirmed to be surfacebreaking by metallurgical 

examination.

2 deepest flaws determined to be 22-mm 

deep by destructive examination.

All 4 flaws 

sample an

Takahama 

Units 2 & 

3

2008 SG inlet nozzle-to-safe 

end–welds

Numerous indications in 3 

nozzle welds. Orientation 

not specified.

Largest flaw in Unit 2 was 14-mm long 

and 8-mm deep and largest flaw in Unit 3 

was 38-mm long and 15-mm deep. Wall 

thickness not provided.

Repair tech

the flaw de

Takahama 

Unit 4

2008 SG inlet nozzle-tosafe–end weldsNumerous indications in 3 

nozzle welds. Orientation 

not specified.

Largest flaw was 33-mm long and 16-mm 

deep. Wall thickness not provided.

At time of r

remove an

Alloy 690 i

Tihange 

Unit 2

2002 Pressurizer surge line 

nozzle-to-safe–end 

weld

Circumferential and axial Circumferential < IWB-3514.

Axial 26-mm long, 4-mm deep.

Left in serv

subsequen

Tihange 

Unit 2

2003 RPV outlet nozzle-tosafe–end weldAxial by ET; not identified 

by UT

Estimated at 10-mm long and 1-mm deep. Left in serv

subsequen

Tsuruga 

Unit 2

2003 Pressurizer relief and 

safety nozzle-to-safe–

end welds

2 axial flaws in relief nozzle 

weld. 1 axial flaw in ‘A’

safety nozzle weld.

1 through-wall flaw in relief nozzle weld. 

Other depths not available.

Samples re

analysis co

Tsuruga

Unit 2

2007 SG inlet nozzle-tosafe–end weldsNumerous axial indications 

in 2 of 4 nozzle welds.

Largest flaw in ‘B’ nozzle weld was about 

21-mm long and 12-mm deep. Largest 

flaw in ‘C’ nozzle weld was 14-mm long 

and 13-mm deep. Wall thickness is 

77.5 mm.

Local repa

Alloy 690 i
B.3 Conclusions

Based on a review of the available operating experience with PWSCC in butt welds, the 

following conclusions are provided.

• PNNL identified over two dozen occurrences of circumferential and axial PWSCC in 

Alloy 82/132/182 butt welds in U.S. and foreign plants since 1993. Some of these 

occurrences involved multiple cracks. Axial PWSCC has been observed more frequently 

than circumferential cracking.

• In the United States, PWSCC in butt welds has been found in the range of piping systems 

and sizes from drain lines to reactor coolant loop piping and at all Class 1 temperature 

conditions, including drain line piping at a reactor coolant loop cold leg temperature.

• The first incidence of PWSCC that was observed in a butt weld was a leaking through-wall

circumferential crack at Palisades in 1993 in the HAZ of the PORV Alloy 600 safe end.

• Leakage to date from PWSCC has only been observed from axial cracks in pressurizer and 

hot leg temperature welds.

• From the data available, PWSCC in Japan has been observed in pressurizer valve nozzle 

welds, RPV outlet nozzle welds, and steam generator inlet nozzle welds. All of the 

occurrences that were identified in Japan have been axial cracking. Information on weld 

configuration was provided on some of these welds that indicated a double V-groove 

configuration.

• Comparison of UT-measured depth and the depth measured by metallurgical analysis in 

Ringhals 4 (2000) and Ohi 3 (2008) showed flaw depths by UT that were substantially 

undersized.

• Based on the information obtained on the events described above no circumferential or axial 

PWSCC has exceeded ASME Code structural factors.

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Cliffs Nuclear Power Plant, LLC, Lusby, Maryland. ADAMS Accession No. ML051180015.

Mensah TM. 2007. Memorandum to SL Rosenberg. "Summary of the December 20, 2006, 

Category 2 Public Meeting Re: Meeting with the Nuclear Energy Institute (NEI) to Discuss 

Implications of Wolf Creek Dissimilar Metal Weld Inspections." March 13, 2007, U.S. Nuclear 

Regulatory Commission, Washington, D.C. ADAMS Accession No. ML070330381.

Miteva R and NG Taylor. 2006. General Review of Dissimilar Metal Welds in Piping Systems 

of Pressurised Water Reactors, Including WWER Designs. EUR 22469 EN, European 

Commission, Joint Research Centre, Luxembourg.

Nomura T, M Iwasaki and K Dozaki. 2007. "Presentation: Inspection for Steam Generator 

Nozzle Welds (Dissimilar Metal Welds) of Japanese PWR Plant." Presented at ASME Boiler 

and Pressure Vessel Code Meeting, November 2007, Dallas, Texas.

Nomura T. 2008a. "Presentation: Investigation Result of SG Nozzles Cracking in Mihama-2 

and RV Nozzle Cracking in Ohi-3." Presented at ASME Boiler and Pressure Vessel Code 

Meeting, August 2008, San Francisco, California.

Nomura T. 2008b. "Presentation: SG Nozzles Cracking in Takahama-4 and RV Nozzle 

Cracking in Ohi-3." Presented at ASME Boiler and Pressure Vessel Code Meeting, November 

2008, Los Angeles, California.

B.16
NRC. 1993. Palisades Pressurizer Penetrations Safe End Crack Results in Primary Coolant 

Leakage. Licensee Event Report 255-1993-009, U.S. Nuclear Regulatory Commission (NRC), 

Washington, D.C. November 8, 1993.

NRC. 2002. Recent Experience with Degradation of Reactor Pressure Vessel Head. 

Information Notice 2002-11, U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. 

ADAMS Accession No. ML020700556.

NRC. 2004. Cracking in Pressurizer Safety and Relief Nozzles and in Surge Line Nozzle. 

Information Notice 2004-11, U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. 

May 6, 2004. ADAMS Accession No. ML041260136.

NRC. 2008a. Regulatory Approach for Primary Water Stress Corrosion Cracking of Dissimilar 

Metal Butt Welds in Pressurized Water Reactor Primary Coolant System Piping. Regulatory 

Information Summary 2008-25, U.S. Nuclear Regulatory Commission, Washington, D.C. 

October 22, 2008. ADAMS Accession No. ML081890403.

NRC. 2008b. Temporary Instruction 2515/172, Reactor Coolant System Dissimilar Metal Butt 

Welds. TI 2515/172, U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. February 

21, 2008. ADAMS Accession No. ML073330037.

O'Keefe M. 2010. Letter to NRC Document Control Desk. "Seabrook Station, Inservice 

Inspection Examination Report." February 9, 2010, NextEra Energy Seabrook, LLC, Seabrook, 

New Hampshire. ADAMS Accession No. ML100490381.

Price JA. 2006. Letter to NRC Document Control Desk. "Dominion Nuclear Connecticut, Inc., 

Millstone Power Station Unit 3, Inservice Inspection Program - Owner's Activity Report." 

January 24, 2006, Dominion Nuclear Connecticut, Inc., Millstone Power Station, Waterford, 

Connecticut. ADAMS Accession No. ML060380567.

Roussel G. 2005. "Management of the Nickel-Base Alloy Cracking in Butt Welds at the Belgian 

Nuclear Power Plants." In Proceedings of NuPEER 2005 (Nuclear Pressure Equipment 

Expertise and Regulation) Dijon Symposium, Ageing Issues in Nuclear Power Plants. 

June 22-24, 2005, Dijon, France. Autorité de Sûreté Nucléaire, Paris, France. Available at 

http://www.asn.fr/fichiers/nupeer/2/2_2C_ROU.PDF.

Skanberg L. 2001. "PWSCC Observations in Ringhals 3 and 4 Safe Ends." In 16th 

International Conference on Structural Mechanics in Reactor Technology (SMiRT16). 

August 12-17, 2001, Washington, D.C. Available at 

http://www.iasmirt.org/transactions/16/WO_06.pdf.

Spina JA. 2006. Letter to NRC Document Control Desk. "Calvert Cliffs Nuclear Power Plant 

Unit No. 1; Docket No. 50-317, ASME Code Section XI Flaw Evaluation of Dissimilar Metal 

Weld Flaws Identified by Ultrasonic Testing." May 31, 2006, Constellation Energy Generation 

Group, Calvert Cliffs Nuclear Power Plant, LLC, Lusby, Maryland. ADAMS Accession No. 

ML061580443.

B.17
Yonazawa T. 2004. "Presentation: Tsuruga Unit-2 Pressurizer; Leakage from the Dissimilar 

Weld (Alloy 600) of the Relief Nozzle." In Proceedings of the Conference on Vessel Penetration 

Inspection, Crack Growth and Repair, Vol. 1, pp. 1013-1020. September 29, 2003–October 2, 

2003, Gaithersburg, Maryland. U.S. Nuclear Regulatory Commission, Washington, D.C. 

ADAMS Accession No. ML050410188.

B.18
APPENDIX C

DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL WESTINGHOUSE PLANT

APPENDIX C

DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL WESTINGHOUSE PLANT

C.1 Introduction

The information contained in this appendix was obtained during a workshop held with LMT, Inc., 

an inspection vendor for the commercial nuclear power industry. This vendor uses ultrasonic 

testing (UT) inspection methods qualified under ASME Code, Section XI, Appendix VIII, 

including phased-array (PA) techniques and conventional UT techniques in encoded and nonencoded delivery modes, as specified by the utility. This workshop was held to discuss the 

characteristics of UT data of indications attributed to stress corrosion cracking (SCC). This data 

was from encoded examinations of piping welds acquired from scans of the outside diameter 

(OD) surface. This workshop was also held to discuss limitations to achieving Code-required 

volume (CRV) coverage on Alloy 82/182 butt welds in a typical Westinghouse plant and a 

typical Combustion Engineering plant. This appendix provides information on limitations to 

obtaining the CRV coverage that may be encountered during examinations of Alloy 82/182 butt 

welds in the reactor coolant system in a Westinghouse plant.

Table C.1 provides a listing of the Alloy 82/182 butt welds that are typically found in a 

Westinghouse plant. The location of these welds is depicted in Figure C.1. The CRV coverage 

for the Alloy 82/182 in butt welds in the piping systems in Table C.1 is dependent on sitespecific configurations and allowable scan access thereof. Site-specific designs vary in regards 

to material and component configuration(s). Therefore, the coverage depicted is applicable only 

to the site-specific design used to prepare this document. Standard 45° and 60° angles were 

utilized for CRV coverage assessments unless specified otherwise in the illustrations. 

Table C.1 Alloy 82/182 Butt Welds and Dimensions in Typical Westinghouse

Description Quantity

Circumference

(Weld Centerline)

Thickness

(Weld Centerline)

PZR Surge Nozzle 1 47.5 in. (120.6 cm) 1.60 in. (4.06 cm)

PZR Spray Nozzle 1 18.7 in. (47.5 cm) 0.87 in. (2.21 cm)

PZR Safety and Relief Nozzle 4 25 in. (63.5 cm) 1.37 in. (3.48 cm)

RPV Hot Leg Nozzle 3/4 113 in. (287.0 cm) 2.50 in. (6.35 cm)

SG Hot Leg Nozzle 3/4 132.3 in. (336.0 cm) 4.90 in. (12.45 cm)

SG Cold Leg Nozzle 3/4 132.3 in. (336.0 cm) 4.90 in. (12.45 cm)

RPV Cold Leg Nozzle 3/4 113 in. (287.0 cm) 2.50 in. (6.35 cm)

PZR – pressurizer; RPV – reactor pressure vessel; SG – steam generator; 3/4 – 3 or 4 loops

C-1
Figure C.1 Typical Locations of Alloy 82/182 Butt Welds in Westinghouse Design Plants

C-2
C.2 Pressurizer Surge Nozzle

Figure C.2 Photograph of Pressurizer Surge Nozzle, Nozzle Weld, and Safe End

The examination limitations for the pressurizer surge nozzle Alloy 82/182 butt weld were 

minimal scanning surface and the OD examination surface contour and condition.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.3. The axial scan direction CRV coverage achievable was 80%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure C.4. The circumferential scan direction CRV coverage achievable was 

100%. 

C-3
Figure C.3 Pressurizer Surge Nozzle Weld Axial Scan CRV Coverage Assessment

Figure C.4 Pressurizer Surge Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

C-4
C.3 Pressurizer Spray Nozzle

Figure C.5 Photograph of Pressurizer Spray Nozzle, Nozzle Weld, and Safe End

The examination limitations for the pressurizer spray nozzle Alloy 82/182 butt weld were 

minimal scanning surface and the OD examination surface contour and condition.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.6. The axial scan direction CRV coverage achievable was 40%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure C.7. The circumferential scan direction CRV coverage achievable was 34%.

C-5
Figure C.6 Pressurizer Spray Nozzle Weld Axial Scan CRV Coverage Assessment

Figure C.7 Pressurizer Spray Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

C-6
C.4 Pressurizer Safety and Relief Nozzle

Figure C.8 Photograph of Pressurizer Safety and Relief Nozzle, Nozzle Weld, and Safe 

End

The examination limitations for the pressurizer safety and relief nozzle Alloy 82/182 butt weld 

were minimal scanning surface and the OD examination surface contour and condition. 

The cross-sectional examination coverage in both the axial scan direction for circumferential 

flaws and in the circumferential direction for axial flaws was 0%. This configuration does not 

allow a meaningful examination for either circumferential or axial flaws. The cross-sectional 

examination coverage for both scan directions is depicted in the assessment diagram in 

Figure C.9. 

C-7
Figure C.9 Pressurizer Safety and Relief Nozzle Weld Axial Scan CRV Coverage 

Assessment

C.5 Reactor Pressure Vessel Hot Leg Nozzle

Figure C.10 RPV Hot Leg Nozzle, Nozzle Weld, and Safe End

C-8
The examination limitations for the RPV hot leg nozzle Alloy 82/182 butt weld were minimal 

scanning surface and OD examination surface contour and condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.11. The axial scan direction CRV coverage achievable was 28.8%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure C.12. The circumferential scan direction CRV coverage achievable was 

100%. 

Figure C.11 RPV Hot Leg Nozzle Weld Axial Scan CRV Coverage Assessment

Figure C.12 RPV Hot Leg Nozzle Weld Circumferential Scan CRV Coverage Assessment

C-9
C.6 Steam Generator Hot Leg Nozzle

Figure C.13 Photograph of Steam Generator Hot Leg Nozzle, Nozzle Weld, and Safe End

The examination limitations for the steam generator hot leg nozzle Alloy 82/182 butt weld were

minimal scanning surface and OD examination surface contour and condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.14. The axial scan direction CRV coverage achievable was 95.4%. This 

examination consisted, in part, of a phased-array ultrasonic testing (PAUT) technical justification 

of performance demonstration per site-specific mockup criteria. The cross-sectional 

examination coverage in the circumferential scan direction for axial flaws is depicted in 

Figure C.15. The axial scan direction CRV coverage achievable was 100%.

C-10
Figure C.14 Steam Generator Hot Leg Nozzle Weld Axial Scan CRV Coverage Assessment

Figure C.15 Steam Generator Hot Leg Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

C-11
C.7 Steam Generator Cold Leg Nozzle

Figure C.16 Photograph of Steam Generator Cold Leg Nozzle, Nozzle Weld, and Safe End

The examination limitations for the steam generator cold leg nozzle Alloy 82/182 butt weld were

minimal scanning surface and OD examination surface contour and condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.17. The axial scan direction CRV coverage achievable was 95.4%. This 

examination consisted, in part, of a PAUT technical justification of performance demonstration 

per site-specific mockup criteria. The cross-sectional examination coverage in the

circumferential scan direction for axial flaws is depicted in Figure C.18. The axial scan direction 

CRV coverage achievable was 100%.

C-12
Figure C.17 Steam Generator Cold Leg Nozzle Weld Axial Scan CRV Coverage Assessment

Figure C.18 Steam Generator Cold Leg Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

C-13
C.8 RPV Cold Leg Nozzle

Figure C.19 Photograph of RPV Cold Leg Nozzle

The examination limitations for the RPV cold leg nozzle Alloy 82/182 butt weld were minimal 

scanning surface and OD examination surface contour and condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure C.20. The axial scan direction CRV coverage achievable was 100%. This 

examination consisted, in part, of a PAUT technical justification of performance demonstration 

per site-specific mockup criteria. The cross-sectional examination coverage in the

circumferential scan direction for axial flaws is depicted in Figure C.21. The axial scan direction 

CRV coverage achievable was 100%.

C-14
Figure C.20 RPV Cold Leg Nozzle Weld Axial Scan CRV Coverage

Figure C.21 RPV Cold Leg Nozzle Weld Circumferential Scan CRV Coverage

C-15

APPENDIX D

DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL COMBUSTION ENGINEERING PLANT

APPENDIX D

DISSIMILAR METAL WELD INSPECTION LIMITATIONS OF A 

TYPICAL COMBUSTION ENGINEERING PLANT

D.1 Introduction

The information contained in this appendix was obtained during a workshop held with LMT, Inc., 

an inspection vendor for the commercial nuclear power industry. This vendor uses ultrasonic 

testing (UT) inspection methods qualified under ASME Code, Section XI, Appendix VIII, 

including phased-array (PA) techniques and conventional UT techniques in encoded and nonencoded delivery modes, as specified by the utility. This workshop was held to discuss the 

characteristics of UT data of indications attributed to stress corrosion cracking (SCC). This data 

was from encoded examinations of piping welds acquired from scans of the outside diameter 

(OD) surface. This workshop was also held to discuss limitations to achieving Code-required 

volume (CRV) coverage on Alloy 82/182 butt welds in a typical Westinghouse plant and a 

typical Combustion Engineering (CE) plant. This appendix provides information on limitations to 

obtaining the CRV coverage that may be encountered during examinations of Alloy 82/182 butt 

welds in the reactor coolant system in a CE plant.

Table D.1 provides a listing of the Alloy 82/182 butt welds that are typically found in a CE plant. 

The location of these welds is depicted in Figure D.1. The CRV coverage for the Alloy 82/182 

butt welds in the piping systems in Table D.1 is dependent on site-specific configurations and 

allowable scan access thereof. Site-specific designs vary in regards to material and component 

configuration(s). Therefore, the coverage depicted is applicable only to the site-specific design 

used to prepare this document. Standard 45° and 60° angles were utilized for CRV coverage 

assessments unless specified otherwise in the illustrations. 

Table D.1 Alloy 82/182 Butt Welds and Dimensions in Typical CE Plant

Description Quantity

Circumference

(Weld Centerline)

Thickness

(Weld Centerline)

RCP Suction to RC Pipe 4 36.60 in. (92.96 cm) 3.45 in. (8.76 cm )

RCP Discharge to RC Pipe 4 36.60 in. (92.96 cm) 3.35 in. (8.51 cm)

RC Loop Surge Nozzle 1 42.00 in. (106.68 cm) 1.57 in. (3.99 cm)

Letdown and Drain Nozzle 4 3.90 in. (9.91 cm) 0.90 in. (2.29 cm)

Hot Leg Drain Nozzle 1 3.90 in. (9.91 cm) 0.90 in. (2.29 cm)

Charging Inlet Nozzle 2 18.00 in. (45.72 cm) 1.20 in. (3.05 cm)

Safety Injection Nozzle 4 42.00 in. (106.68 cm) 1.55 in. (3.94 cm)

Shutdown Cooling Nozzle 1 50.25 in. (127.64 cm) 1.57 in. (3.99 cm)

Spray Nozzle 2 16.00 in. (40.64 cm) 0.93 in. (2.36 cm)

PZR Surge Nozzle 1 41.00 in. (104.14 cm) 1.59 in. (4.04 cm)

PZR Spray Nozzle 1 5.20 in. (13.21 cm) 0.83 in. (2.11 cm)

PZR Safety Nozzle 3 7.95 in. (20.19 cm) 1.40 in. (3.56 cm)

RCP – reactor coolant pump; RC – reactor coolant; PZR – pressurizer

D-1
Figure D.1 Typical Locations of Alloy 82/182 Butt Welds in CE Design Plants

D-2
D.2 Reactor Coolant Pump (RCP) Suction

Figure D.2 Photograph of RCP Suction Nozzle, Nozzle Weld, and Safe End

The examination limitations for the RCP suction nozzle Alloy 82/182 butt weld were the cast 

austenitic stainless steel (CASS) pump nozzle material on the downstream side of the 

Alloy 82/182 weld and the examination surface (OD) contour /condition.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.3. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.4. The circumferential scan direction CRV coverage achievable was 

44.3%.

D-3
Figure D.3 RCP Suction Nozzle Weld Axial Scan Coverage Assessment

Figure D.4 RCP Suction Nozzle Weld Circumferential Scan Coverage Assessment

D-4
D.3 Reactor Coolant Pump Discharge

Figure D.5 Photograph of RCP Discharge Nozzle, Nozzle Weld, and Safe End

The examination limitations for the RCP discharge nozzle Alloy 82/182 butt weld were the 

CASS pump nozzle material on the upstream side of the Alloy 82/182 weld; the examination 

surface (OD) contour/condition; and instrumentation, safety injection, charging and spray nozzle 

connections. No coverage was obtainable in the locations where nozzle connections interfere 

with probe travel.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.6. The axial scan direction CRV coverage achievable was 77%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.7. The circumferential scan direction CRV coverage achievable was 34%.

D-5
Figure D.6 RCP Discharge Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.7 RCP Discharge Nozzle Weld Circumferential Scan CRV Coverage Assessment

D-6
D.4 Reactor Coolant (RC) Loop Surge

Figure D.8 Photograph of Surge Nozzle, Nozzle Weld, and Safe End

The examination limitations for the surge nozzle Alloy 82/182 butt weld were the CASS safe end 

material on downstream side of the Alloy 82/182 weld and the examination surface (OD) 

contour/ condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.9. The axial scan direction CRV coverage achievable was 82.4%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.10. The circumferential scan direction CRV coverage achievable was 

31.3%.

D-7
Figure D.9 Surge Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.10 Surge Nozzle Weld Circumferential Axial Scan CRV Coverage Assessment

D-8
D.5 Letdown and Drain

Figure D.11 Photograph of Letdown and Drain Nozzle, Nozzle Weld, and Safe End

There were no examination limitations for the letdown and drain nozzle Alloy 82/182 butt weld.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.12. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.13. The circumferential scan direction CRV coverage achievable was 

100%. 

D-9
Figure D.12 Letdown and Drain Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.13 Letdown and Drain Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

D-10
D.6 Hot Leg Drain

Figure D.14 Photograph of Hot Leg Drain Nozzle, Nozzle Weld, and Safe End

There were no examination limitations for the hot leg drain nozzle Alloy 82/182 butt weld. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.15. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.16. The circumferential scan direction CRV coverage achievable was 

100%. 

D-11
Figure D.15 Hot Leg Drain Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.16 Hot Leg Drain Nozzle Weld Circumferential Scan CRV Coverage Assessment

D-12
D.7 Charging Inlet

Figure D.17 Photograph of Charging Inlet Nozzle, Nozzle Weld, and Safe End

There were no examination limitations for the charging inlet nozzle Alloy 82/182 butt weld.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.18. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.19. The circumferential scan direction CRV coverage achievable was 

100%.

D-13
Figure D.18 Charging Inlet Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.19 Charging Inlet Nozzle Weld Circumferential Scan CRV Coverage Assessment

D-14
D.8 Safety Injection

Figure D.20 Photograph of Safety Injection Nozzle, Nozzle Weld, and Safe End

The examination limitations for the safety injection nozzle Alloy 82/182 butt weld were the CASS 

safe end material on upstream side of the Alloy 82/182 weld and the examination surface (OD) 

contour/ condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.21. The axial scan direction CRV coverage achievable was 94.3%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.22. The circumferential scan direction CRV coverage achievable was 

26.6%.

D-15
Figure D.21 Safety Injection Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.22 Safety Injection Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

D-16
D.9 Shutdown Cooling

Figure D.23 Photograph of Shutdown Cooling Nozzle, Nozzle Weld, and Safe End

The examination limitations for the shutdown cooling nozzle Alloy 82/182 butt weld were the 

CASS safe end material on the downstream side of the Alloy 82/182 weld and the examination 

surface (OD) contour /condition. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.24. The axial scan direction CRV coverage achievable was 98%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.25. The circumferential scan direction CRV coverage achievable was 

100%. 

D-17
Figure D.24 Shutdown Cooling Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.25 Shutdown Cooling Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

D-18
D.10 Spray

Figure D.26 Photograph of Spray Nozzle, Nozzle Weld, and Safe End

There were no examination limitations for the spray nozzle Alloy 82/182 butt.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.27. The axial scan direction CRV coverage achievable was 100%. The

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.28. The circumferential scan direction CRV coverage achievable was 

100%.

D-19
Figure D.27 Spray Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.28 Spray Nozzle Weld Circumferential Scan CRV Coverage Assessment

D-20
D.11 Pressurizer Surge

Figure D.29 Photograph of Pressurizer Surge Nozzle, Nozzle Weld, and Safe End

There were no examination limitations for the pressurizer surge nozzle Alloy 82/182 butt weld. 

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.30. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.31. The circumferential scan direction CRV coverage achievable was 

100%. 

D-21
Figure D.30 Pressurizer Surge Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.31 Pressurizer Surge Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

D-22
D.12 Pressurizer Spray

Figure D.32 Photograph of Pressurizer Spray Nozzle, Nozzle Weld, and Safe End

The examination limitation for the pressurizer spray nozzle Alloy 82/182 butt weld was the 

examination surface (OD) contour /condition.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.33. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.34. The circumferential scan direction CRV coverage achievable was 

68.9%.

D-23
Figure D.33 Pressurizer Spray Nozzle Weld Axial Scan CRV Coverage Assessment

Figure D.34 Pressurizer Spray Nozzle Weld Circumferential Scan CRV Coverage 

Assessment

D-24
D.13 Pressurizer Safety & Relief

Figure D.35 Photograph of Pressurizer Safety & Relief Nozzle, Nozzle Weld, and Safe 

End

There were no examination limitations for the pressurizer safety and relief nozzle Alloy 82/182 

butt weld.

The cross-sectional examination coverage in the axial scan direction for circumferential flaws is 

depicted in Figure D.36. The axial scan direction CRV coverage achievable was 100%. The 

cross-sectional examination coverage in the circumferential scan direction for axial flaws is 

depicted in Figure D.37. The circumferential scan direction CRV coverage achievable was 

100%.

D-25
Figure D.36 Pressurizer Safety and Relief Nozzle Weld Axial Scan CRV Coverage 

Assessment

Figure D.37 Pressurizer Safety and Relief Nozzle Weld Circumferential Scan CRV 

Coverage Assessment

D-26
APPENDIX E

ASSESSMENT OF WELD RESIDUAL STRESS AND CRACK 

GROWTH ANALYSIS IN UNMITIGATED WELDS

APPENDIX E

ASSESSMENT OF WELD RESIDUAL STRESS AND CRACK 

GROWTH ANALYSES IN UNMITIGATED WELDS

This appendix contains a summary of the results of four crack growth studies that were 

assessed for insights they provide regarding the inspection intervals for unmitigated 

Alloy 82/182 welds specified in ASME Code Case N-770-1. The conclusions reached regarding 

the Code Case inspection intervals are reported in Section 3.4.4.

E.1 Results from BCL Mitigation Studies

U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES)

initiated a program entitled, “PWSCC in Leak-Before-Break Systems,” to assess strategies for 

managing primary water stress corrosion cracking (PWSCC) in susceptible welds in leakbefore-break systems. Under this RES program, Battelle Columbus Laboratory (BCL)

performed two weld residual stress (WRS) and crack growth studies (Fredette and Scott 2009, 

2010). The stress and crack growth results shown in Table E.1 are summarized from the

studies performed by BCL under this program. Pacific Northwest National Laboratory (PNNL)

assessed these results in an effort to gain insights on Code Case N-770-1 inspection intervals 

for unmitigated welds. Three dissimilar metal weld (DMW) pipe/nozzle geometries were 

considered in a BCL study on mitigation against PWSCC by the mechanical stress improvement 

process (MSIP) (Fredette and Scott 2009): (1) a large-diameter hot leg (HL) reactor pressure 

vessel (RPV) nozzle weld, (2) a medium-diameter pressurizer surge nozzle weld, and (3) a 

small-diameter pressurizer safety nozzle weld. In the second study, BCL analyzed welds being 

mitigated against PWSCC by weld overlays (Fredette and Scott 2010). However, crack growth 

results were only reported for the surge nozzle and the safety nozzle and, for unmitigated welds, 

were the same as provided in BCL’s study on MSIP.

In modeling these pressurized water reactor (PWR) dissimilar metal welds, the application of the 

butter to the ferritic nozzle was modeled along with the post-weld heat treatment of the butter 

material. In addition, the weld root passes were ground out and re-welded as part of these 

analyses to simulate this common practice in the field. This full-circumferential grind-out/rewelding resulted in a residual stress field similar to that of a 360° inside surface repair weld. 

The results in Table E.1 are based on axisymmetric models. The stress intensities were 

calculated for normal operating pressure and temperature conditions and did not consider 

external bending moments.

E-1
Table E.1 Crack Growth Results from BCL Studies

System

Safe 

End

Crack 

Location ID Axial Stress K-Value (ksi-√in)(a)

Crack Growth 

Comments

Hot Leg RV 

Nozzle

No Butter/Weld 40 ksi at weld 

centerline

17 ksi-√in at ID; 30 at 

15%; 0 from 33% to 

70% TW

Arrest by 33% 

TW

Hot Leg RV 

Nozzle

No Ferritic/Butter 40 ksi at weld 

centerline

17 ksi-√in at ID; 25 at 

11% TW; 0 from 33% 

on

Arrest by 33% 

TW

Surge 

Nozzle

Yes Butter/Weld –30 ksi at weld 

centerline

0 ksi-√in to 14% TW; 5 

at 31% TW; 0 from 

40% to 70% TW

Unlikely to 

initiate 

Surge 

Nozzle

Yes Ferritic/Weld –30 ksi at weld 

centerline

0 ksi-√in to 14% TW; 8 

at 22% TW; 0 from 

36% to 68% TW

Unlikely to 

initiate 

Safety 

Nozzle

Yes Butter/Weld –50 ksi at weld 

centerline

0 ksi-√in to 10% TW; 

2.5 at 25% TW; 0 from 

33% to 60% TW

Unlikely to 

initiate 

Safety 

Nozzle 

Yes Ferritic/Butter –50 ksi at weld 

centerline

0 ksi-√in to 64% TW Unlikely to 

initiate 

RV = reactor vessel; TW = through-wall

(a) 1 ksi-√in = 1.099 MPa-√m

The following concluding points can be made about this study.

• The analyses for the three nozzle welds shown in Table E.1 showed that if cracking were to 

initiate, it would arrest before cracking reached about one-third through-wall. This same 

result is shown even for the HL nozzle weld that does not have the beneficial compressive 

WRS effect of the secondary stainless steel safe end weld. 

• These analyses were not conventional crack growth analyses and did not include external 

bending loads. Sensitivity analyses discussed below resulted in many fewer cases of 

arrested flaws.

E.2 Results from Pressurizer Nozzle Advanced Finite Element 

Analyses

Results were assessed of finite element analyses published in August 2007 by industry (EPRI 

2007a) and by NRC staff (Rudland et al. 2007b). These results were used to assess Code 

Case N-770-1 ultrasonic testing (UT) inspection intervals for Inspection Item A-1, unmitigated 

butt welds at operating temperature greater than 625°F. This inspection item applies to welds 

attached to the pressurizer. Some of the finite element analyses were based on an initial flaw 

size derived from an inspection performed in October 2006 at the Wolf Creek plant on 

pressurizer nozzle Alloy 82/182 butt welds. Cracking in these welds at Wolf Creek was

attributed to PWSCC.

The emphasis of the August 2007 analyses was to use advanced finite element analysis (AFEA) 

methods to remove the semi-elliptical flaw assumption that is typical in ASME Section XI flaw 

evaluations. Using this AFEA method, detailed sensitivity cases were analyzed to demonstrate 

E-2
that sufficient margins were available to allow pressurizer nozzle weld inspections of nine plants 

to be performed at their next refueling outages, rather than during forced mid-cycle outage 

shutdowns.

This assessment was based on the AFE sensitivity analyses performed on the surge nozzle 

welds. Table E.2 summarizes a number of the cases performed by both sets of analysts, 

industry, and the NRC. The first nine entries are from industry’s analyses (EPRI 2007a) and the 

last seven entries are from the NRC staff’s analyses (Rudland et al. 2007b).

Table E.2 Advanced Finite Element Analysis of Circumferential Flaw Growth in Surge 

Nozzle Welds

Case 

No. Loads Initial Flaw(s) Safe End

Time to 

Leakage

Time from 1 gpm 

to Stability 

Margin = 1.2

18a Low 10% TW, 360° With safe end Arrest Not Applicable

19b High 26% TW, 21/1 AR With safe end Arrest Not Applicable

20b Low 10% TW, 360° With safe end Arrest Not Applicable

26a Low, ID repair 10% TW, 360° With safe end 2.2 years >>40 days

17b High 26% TW, 21/1 AR Without safe end 1.2 years 35 days

18b Low 10% TW, 360° Without safe end 11.5 yrs 43 days

25b High, ID repair 26% TW, 21/1 AR Without safe end 0.5 year 68 days

S2b High 10% TW, 360° Without safe end 3.4 years 4 days(a)

S3b High 10% TW, 5.6/1 AR Without safe end 2.2 years 74 days

17-4 High, DEI WRS 10% TW, 360° With safe end 5.8 years 1.07 month

17-6 High, Emc2 WRS 10% TW, 360° With safe end Arrest Not applicable

17-8 High, DEI WRS 26% TW, 21/1 AR With safe end 6.15 years Not calculated(b)

17-9 High, DEI WRS 10% TW, 360° Without safe end 1.39 years Not available(c)

17-10 Medium, DEI WRS 10% TW, 360° Without safe end 2.84 years 0.64 month

17-11 High, DEI WRS, 

ID repair

10% TW, 360° Without safe end 0.86 year 0.31 month

17 High, DEI WRS 10% TW, 360° Without safe end 1.3 years Not available(c)

AR = aspect ratio

DEI = industry analyst

Emc2 = Engineering Mechanics Corporation of Columbus

(a) Initial leakage was calculated to be 4.9 gpm

(b) Relatively large margin so not calculated for this case

(c) Crack near critical at first leakage

All the finite element cases in Table E.2, except 19b and 20b, applied to a WRS model that

included inner-diameter (ID) back-gouging and re-welding. Back gouging and re-welding 

increases the tensile stresses in the region of the ID. With the exception of Case 26a, cases 

with a safe end-to-pipe weld either arrest or have a time to through-wall (TW) leakage of about 

6 years, which is longer than the Inspection Item A-1 inspection interval. Depending on the 

length of the safe end, installing a safe end-to-pipe weld introduces compressive residual 

stresses near the ID of the adjacent nozzle-to-safe end weld. Case 26a has low loads and a 

safe end-to-pipe weld and has a partial arc ID repair. The time to TW leakage for this case is 

2.2 years. The AFEA showed that Case 25b with high loads and a limited-extent ID repair had 

a shorter time to TW leakage than the same case without repair (Case 17b). However, Case 

E-3
25b with limited-extent ID repair had a longer time from initial through-wall leakage to failure 

than Case 17b without the repair.

Cases S2b, 17-10, and 18b with high, medium, and low loading, respectively, and without a safe 

end-to-pipe weld had times to through-wall leakage of 3.4 years, 2.84 years, and 11.5 years, 

respectively. These three cases had between 4 and 43 days for the plant to shut down before 

the stability margin established for this analysis was no longer satisfied. While the pressurizer 

surge nozzle configurations do have safe ends, the sensitivity cases analyzed without a safe 

end were included to account for uncertainties in the calculation of WRS and crack growth.

Case S2b, with 4 days calculated for the time between first leakage and crack instability, 

approximated a case with multiple shallow initial flaws around the ID. Because the external 

operating loading was medium, the long initial crack progressed through-wall around the 

circumference with a profile that left relatively little remaining wall ligament at first leakage.

The remaining six surge line cases included in Table E.2 were analyzed without safe end-topipe welds, with high loads, and with three different initial crack sizes. The times to leakage for 

the subset of cases without ID repairs (17b, S3b, 17-9, and 17) ranged from 1.2 years to 

2.2 years. Two cases (17 and 17-9) had cracks near the critical size at first leakage. The two 

cases with ID repairs and high loads (25b and 17-11) had times to through-wall leakage of 

0.5 year and 0.86 year but had adequate time between leakage and loss of stability margin to 

shut down the plant.

Code Case N-770-1 requires that an unmitigated pressurizer surge nozzle-to-safe–end weld be 

categorized as Inspection Item A-1 and requires that a volumetric examination of these welds 

be performed every second refueling outage. This interval would typically be about once every 

3 years but could range from 2 to 4 years.

The following concluding points can be made about this study.

• Overall, the times-to-failure results of the AFEA shown in Table E.2 do not bound the 

inspection intervals in Code Case N-770-1 for Inspection Item A-1. The high load cases 

without a safe end may be considered bounding cases. These bounding cases would 

experience through-wall leakage before the typical 3-year interval between UT examinations 

of these welds.

• For the cases with a safe end, the times to through-wall leakage were longer than the Code 

Case N-770-1–required inspection interval, except for Case 26a, which has an ID repair. 

One would expect that ID repairs were fairly common and EPRI (2007a) indicates that five of 

the nine surge nozzle welds in the plants studied had documented ID repairs. Therefore, 

these analyses do not tend to support the Code Case N-770-1 inspection interval for 

Inspection Item A-1 welds.

• From a review of information on the NRC website pertaining to reactor coolant system weld 

issues (http://www.nrc.gov/reactors/operating/ops-experience/pressure-boundaryintegrity/weld-issues/plant-specific-info.html) and Stall (2007), Wadley (2008), and King 

(2006), all the Alloy 82/182 pressurizer surge nozzle-to-safe–end welds have been mitigated 

by either full structural weld overlays or MSIP. With one exception, the NRC staff’s 

E-4
sensitivity cases for pressurizer safety, relief, and spray valve nozzle butt welds showed 

times to first leakage of longer than 4 years. One case resulted in time to first leakage of 

3.3 years. All the cases showed considerable time between first leakage and crack 

instability. A number of sensitivity cases showed crack arrest.

• As of early 2013, there were unmitigated pressurizer safety, relief, and/or spray valve nozzle 

butt welds at three plants. These welds have all been inspected at least once and the plants 

have committed to re-inspect these unmitigated welds at least once every 4 years, 

consistent with the requirements of Code Case N-770-1. The authors are not aware of any 

operating experience that would challenge the Inspection Item A-1 re-inspection interval for 

these welds.

E.3 Hot Leg and Cold Leg Flaw Evaluation and Sensitivity Analyses

The NRC staff conducted a scoping evaluation to determine the leak and rupture characteristics 

of DMWs in primary piping containing a hypothetical pre-existing circumferential flaw at hot leg 

locations and at cold leg reactor coolant pump (RCP) nozzle locations (Wallace 2014). In

conducting these analyses, it was assumed that the pre-existing flaws grew due to PWSCC and 

the analyses were based on ASME Boiler and Pressure Vessel Code, Section XI, IWB-3640, 

requirements. This analysis did not consider fatigue crack growth.

The NRC staff chose four representative hot leg cases that span the range of typical leakbefore-break approved plants and three RCP suction and discharge nozzle bounding cases 

from Ganta and Bamford (2009). Table E.3 provides details of the geometry and conditions that 

were modeled. The NRC staff evaluated the time to leak and time to rupture for stresses 

resulting from Normal Operating (NO) conditions. For conservatism, as the bounding loading 

case, the NRC staff also evaluated the times to leak and rupture under Safe Shutdown 

Earthquake (SSE) conditions.

Table E.3 Summary of Geometries and Conditions

Hot Leg RCP Nozzle

Outside diameter (OD) 86.25 cm (33.96 in.) 91.44 cm (36.00 in.)

Wall thickness (t) 6.07 cm (2.39 in.) 7.62 cm (3.00 in.)

Temperature (T) 325°C (617°F) 286.11°C (547°F)

Pressure (P) 15.46 MPa (2243 psi) 15.49 MPa (2247 psi)

Initial flaw depth (a) 1.61 cm (0.634 in.) 1.61 cm (0.634 in.)

Initial flaw length (2c) 5.23 cm (2.06 in.) 5.23 cm (2.06 in.)

The NRC staff performed the evaluations for each of four different WRS distribution cases. Two 

of the cases are from values that NRC staff used for the hot leg geometry in the original Wolf 

Creek scoping analyses (Rudland et al. 2007a). The first case is for an as-welded 

Westinghouse-type outlet nozzle with no stainless steel safe end closure weld and no repair. 

Case 2 is the same geometry and conditions, but with a 15% inside diameter backchip and last 

pass weld. For Case 3, NRC staff developed a WRS profile from Dominion Engineering, Inc. 

(DEI) results presented in EPRI report MRP-216 (EPRI 2007a). In that report DEI took a 90-

E-5
degree partial arc deep weld repair and presented the results in the center of the repair, at the 

end of the repair, and far removed from the repair. NRC staff developed Case 4 from Case 2 as 

part of the inlay program (Rudland et al. 2010), but included the effect of a stainless steel safe 

end weld closure weld.

The NRC staff calculated the PWSCC growth of the assumed initial flaw using the 75th

percentile crack growth rates from EPRI MRP-115-NP (EPRI 2004b) for the different 

geometries, stress cases, and WRS cases. NRC staff also calculated the time to leakage and 

used the net-section collapse criteria and the Alloy 182 Z-factor to calculate time to rupture and 

the critical crack size for rupture.

In addition to those assumptions, the staff assumed idealized surface and through-wall flaw 

shapes. The work in the AFEA project showed this assumption to be typically conservative for 

the time between leakage and rupture, but reasonably accurate for time to leakage.

The results of these evaluations are shown in Tables E.4 and E.5.

Table E.4 Results of Hot Leg Evaluation for Time to Leak and Time to Rupture

Stress 

Case WRS Case

Normal Operation

Normal Operation + 

Safe Shutdown 

Earthquake

Time to 

Leak 

(year)

Time to 

Rupture 

(year)

Time between 

Leak and 

Rupture (year)

Time to 

Leak 

(year)

Time to 

Rupture 

(year)

A

As-Welded 2.33 4.33 2.00 2.00 3.75

15% Backchip 1.83 3.50 1.67 1.58 3.00

Partial Arc Repair 0.83 3.08 2.25 0.75 2.67

With Safe End 7.92 9.42 1.50 5.67 7.08

B

As-Welded 2.08 3.92 1.83 1.50 2.67

15% Backchip 1.67 3.17 1.50 1.25 2.17

Partial Arc Repair 0.75 2.83 2.08 0.67 1.83

With Safe End 6.25 7.67 1.42 3.08 4.08

C

As-Welded 9.33 17.5 8.17 2.83 5.42

15% Backchip 5.83 11.5 5.67 2.17 4.25

Partial Arc Repair 0.33 16.2 14.8 0.92 3.83

With Safe End Arrest - - 15.3 17.0

D

As-Welded 2.67 5.08 2.42 1.25 2.17

15% Backchip 2.08 4.00 1.92 1.08 1.75

Partial Arc Repair 0.92 3.58 2.67 0.58 1.50

With Safe End 12.5 14.2 1.67 2.25 3.08

E-6
Table E.5 Results of RCP Nozzle Evaluation for Time to Leak and Time to Rupture

Stress 

Case WRS Case

Normal Operation

Normal Operation + 

Safe Shutdown 

Earthquake

Time to 

Leak 

(year)

Time to 

Rupture 

(year)

Time between 

Leak and 

Rupture (year)

Time to 

Leak 

(year)

Time to 

Rupture 

(year)

E

As-Welded 63.1 106. 42.9 15.2 25.1

15% Backchip 38.6 63.2 24.6 11.3 19.4

Partial Arc Repair 9.10 100. 90.9 5.19 16.8

With Safe End Arrest - - 47.8 54.9

F

As-Welded 83.8 136. 52.2 16.9 28.2

15% Backchip 47.8 75.6 27.8 12.4 21.5

Partial Arc Repair 9.80 165. 155. 5.48 18.9

With Safe End Arrest - - 62.6 70.2

G

As-Welded 26.5 45.1 18.6 14.2 23.5

15% Backchip 18.2 32.8 14.6 10.7 18.2

Partial Arc Repair 6.71 31.5 24.8 5.04 15.7

With Safe End 400.0 407.0 7.40 41.3 48.0

Review of the results provided the following observations.

• The hot leg cases generally had shorter times to leak and rupture than those for the cold leg 

cases. This is the result of generally higher bending stresses as well as higher hot leg 

temperatures.

• For all cases where growth of the assumed flaw by PWSCC resulted in leakage in NO 

conditions, the time between leakage and rupture was well over 1 year, providing significant 

margin to identify the leak and safely shut down the reactor.

• Backchipping the weld root and rewelding decreased the time to leakage and the time 

between rupture and leakage as compared to the as-welded condition.

• Assuming that the NO+SSE loading could occur for extended time periods, the shortest time 

to rupture under NO+SSE loading conditions was 1.5 years.

• Addition of the residual stress because of the stainless steel safe end closure weld resulted 

in a significant increase in the time to leakage, sometimes even resulting in crack arrest, but 

decreased the time between leakage and rupture.

• The weld residual stresses that resulted from the partial arc repair resulted in the shortest 

times to leakage and rupture but did not generally result in the shortest times between 

leakage and rupture.

The following concluding points can be made about this study.

• The “time to rupture” calculations are probably conservative, but the level of conservatism is 

not quantified.

E-7
• In the hot leg weld case with the shortest time to leakage (Load Case B with a partial arc 

repair on a weld with high loads and no safe end closure weld), initial leakage would occur 

in 0.75 years. With an additional 2.08 years between leakage and rupture, even if such a 

weld condition were to exist and started to leak, sufficient time would exist to repair it before 

it becomes critical.

• In the hot leg weld with the lowest loading (Load Case C), the time to leakage for the aswelded case and the case with a 15% backchip without a safe end weld exceeded the 

5-year inspection interval in Code Case N-770-1 for Inspection Item A-2, unmitigated butt 

welds at hot leg operating temperature less than or equal to 329°C (625°F). Load Case C 

with a partial arc repair, which did not have a safe end closure weld, does not bound the 

inspection interval for Inspection Item A-2. Load Case C with a 15% backchip and a safe

end closure weld showed crack arrest. 

• Overall, the time-to-initial leakage results for the hot leg temperature Load Cases A, B, and 

D analyzed without a safe end closure weld do not bound the 5-year inspection interval in 

Code Case N-770-1 for Inspection Item A-2. The times-to-initial leakage for all four hot leg 

temperature Load Cases analyzed with a safe end closure weld bound the inspection 

intervals in Code Case N-770-1 for Inspection Item A-2. The time between leakage and 

failure in all the cases analyzed would be sufficient to shut down a plant, assuming the 

leakage was detected. These sensitivity studies show the strong influence of weld residual 

stresses and external loading on the time-to-leakage, time-to-failure, and time between 

leakage and failure for the nozzle analyzed.

• In the cold leg case with the shortest time to leakage (Load Case G – partial arc repair on a 

weld with high loads and no safe end weld), initial leakage would occur in 6.71 years. This 

time to leakage is just short of the maximum inspection interval of 7 years in Code Case 

N-770-1 for Inspection Item B, unmitigated butt welds at cold leg operating temperature 

greater than or equal to 274°C (525°F) and less than 304°C (580°F). The time to leakage 

for the remainder of the cold leg cases were 9.1 years or longer. The times between 

leakage and rupture for the cold leg cases are equal to or longer than 7.4 years. The results 

of the cold leg analysis appear to support an inspection interval of at least 7 years. 

• Ideally, results for time to reach the ASME Section XI maximum depth of 75% through-wall 

would bound the ASME Code Case inspection intervals, because this is the regulatory basis 

for degradation as opposed to showing that there is sufficient time to shut down a plant after 

leakage is discovered.

E.4 Results from the North Anna Power Station, Unit 1, Flaw Growth 

Analysis

The NRC staff and its contractor, Engineering Mechanics Corporation of Columbus, completed 

a flaw growth analysis (NRC 2012) of axially-orientated flaws found in the steam generator hot 

leg piping-to-steam generator nozzle DMW. Five axial flaws of varying depths were found in 

March 2012 (see Appendix B.1). A detailed weld residual stress analysis was performed using 

typical computational weld modeling processes that have been validated by mockup testing in 

the nuclear industry. The North Anna Power Station, Unit 1 (NAPS-1) hot leg to steam 

generator nozzle weld is a double-V groove weld with two repairs in the region where 

E-8
through-wall flaws were observed following pre-overlay machining. These repairs were a large 

inner-diameter repair and small outer-diameter repair. The weld material is Alloy 82 for the 

butter and Alloy 182 for the main weld and repair welds. The nozzle and safe end materials are 

assumed to be A508 and A316 stainless steel, respectively.

An axi-symmetric finite element model was used for the nozzle. The weld procedure simulated 

consisted of application of the Alloy 82 butter layer, post-weld heat treat, machining of the 

butter, application of the Alloy 182 weld, machining of the repair grooves followed by application 

of the repair welds, and then finally application of the closure stainless steel weld. The flaws 

found in the steam generator DMW were all axially orientated and varying depths. Because

these are axially-orientated flaws, the loads that will affect the PWSCC growth will be the weld 

residual stress and hoop stress due to pressure.

The analysis considered idealized flaw growth and natural flaw growth models. The idealized 

flaw growth analysis was used to address questions such as, “What is the likelihood that a flaw 

was undetectable in the spring 2009 refueling outage and grew to the size observed in the early 

2012 outage?” The staff concluded from this analysis that it is highly likely that a detectable 

flaw was present in the 2009 outage and grew to the size observed in 2012.

Natural flaw growth analyses were performed on the NAPS-1 steam generator nozzle DMW to 

obtain a deterministic estimate of the progress of an axial flaw through the weld from 2009 until 

the flaws were discovered in 2012. In a natural flaw growth analysis, the flaws are driven by 

PWSCC and tensile stresses caused by the WRS and the service loading. For axial flaw growth 

in the DMW, the loading consists of internal nozzle pressure and flaw face pressure of 2248 psi 

(15.5 Mpa) combined with the WRS. In the natural flaw growth analysis method, a finite 

element analysis is performed at each increment of flaw growth to calculate stress intensity 

factors along the crack front. Using the stress intensity factor results and the assumed flaw 

growth law and the 75-percentile flaw growth rate from EPRI MRP-115 (EPRI 2004b), an 

increment of crack advance is calculated and the procedure is repeated. For this study, 

PWSCC of DMW was the only subcritical cracking mechanism that was considered.

Typically, the flaw growth increments were kept less than 10 mm (0.4 in.), with smaller 

increments as the crack depth approached mid-thickness. For every crack growth step, the 

WRS were mapped to the flaw mesh and the WRS were compared to ensure accuracy and as a 

check of the proper boundary conditions and surface loading.

The evolution of the flaw and the corresponding complex flaw shapes can be seen in Figure 7 of 

(NRC 2012). The initial flaw shape (at time = 0 years) was assumed to be 6-mm (0.24-in.) deep 

and a total of 12 mm (0.48 in.) in width. The flaw reaches the right side weld butter line at about 

0.46 years and then begins to grow slower in the butter. The flaw begins to touch the safe end 

at about 0.64 years and then the flaw follows the fusion boundary between the weld and safe 

end as it grows deeper. At 1.17 years, the flaw nearly reaches the butter-nozzle line and then 

can only grow vertically from that point on. The flaw depth growth slows down just past the 

neck area after about 1.76 years to 4 years. This is the region of the weld shape neck down 

where most of the material is represented by Alloy 82 butter. In addition, this is the region 

where the weld residual stresses become slightly negative, also contributing to a slower growth 

rate.

E-9
Figure E.1 shows the flaw depth at the deepest point versus time. The flaw growth rate falls 

near the narrow region between the double-Vs where the weld residual stress is lower, and a 

larger portion of the flaw is in the weld butter (with corresponding lower flaw growth). Later, the 

flaw growth rate increases as the deepest point in the main DMW with higher flaw growth rate

experiences higher weld residual stress.

Figure E.1 Flaw Growth Estimate from 2009 Outage to 2012 (NRC 2012)

Figure E.1 allows an estimate of the time that the flaw grew during the 2.75 years from the 2009 

inspection to 80% this spring. Starting with the observed 80% through-wall flaw in 2012, and 

moving back in time 33 months (corresponding to the preceding operating cycle), Figure E.1

shows that the estimated flaw depth was 47% in the 2009 outage. Similarly, starting with the 

80% flaw in 2012, it is seen that the 75% ASME Code acceptability limit would have been 

reached about 4 months prior. The NRC staff concluded that this “more accurate deterministic 

natural flaw growth analysis indicates with greater likelihood (than from the idealized flaw growth 

analysis) that a flaw was present and detectable in 2009 and grew to the size observed in 

2012.”

E-10
The following observations can be made from the results of this study.

• If the threshold of detection is conservatively (for this case) assumed to be 10% throughwall, Figure E.1 shows that the time to grow from 10% to 75% through-wall in this example 

is approximately 3-1/3 years or 40 months.

• This interval of 3-1/3 years is considerably shorter than the required UT inspection interval of 

once every 5 years for Code Case N-770-1 Inspection Item A-2 welds.

E.5 References

EPRI. 2004. Materials Reliability Program: Crack Growth Rates for Evaluating Primary Water 

Stress Corrosion Cracking (PWSCC) of Alloy 82, 182, and 132 Welds (MRP-115NP). 

TR-1006696, Electric Power Research Institute (EPRI), Palo Alto, California. NRC ADAMS 

Accession No. ML051450555.

EPRI. 2007a. Materials Reliability Program: Advanced FEA Evaluation of Growth of 

Postulated Circumferential PWSCC Flaws in Pressurizer Nozzle Dissimilar Metal Welds: 

Evaluations Specific to Nine Subject Plants. MRP-216, Rev. 1, EPRI Report No. 1015499, 

Electric Power Research Institute (EPRI), Palo Alto, California. ADAMS Accession Nos. 

ML072410240 and ML072410241.

Fredette L and PM Scott. 2009. Evaluation of the Mechanical Stress Improvement Process 

(MSIP) as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized 

Water Reactors. Battelle Columbus, Columbus, Ohio. ADAMS Accession No. ML092990646.

Fredette L and PM Scott. 2010. Evaluation of Full Structural and Optimized Weld Overlays as 

Mitigation Strategies for Primary Water Stress Corrosion Cracking in Pressurized Water 

Reactors. Battelle Columbus, Columbus, Ohio. ADAMS Accession No. ML101260540.

Ganta BR and WH Bamford. 2009. Flaw Evaluation of CE Design RCP Suction and Discharge, 

and Safety Injection Nozzle Dissimilar-Metal Welds, WCAP-16925-NP, Revision 1. 

Westinghouse Electric Company, LLC, Pittsburgh, Pennsylvania. ADAMS Accession No. 

ML092740086.

King C. 2006. Letter to T Mensah. "PWR Fleet Survey - MRP-139 Implementation Plans for 

Pressurizers, Revision 1." December 18, 2006, U.S. Nuclear Regulatory Commission, 

Washington, D.C. ADAMS Accession No. ML063190476.

NRC. 2012. NRC Staff Technical Position for North Anna Hot Leg to Steam Generator Nozzle 

Flaws, Executive Summary. U.S. Nuclear Regulatory Commission (NRC), Washington, D.C. 

July 20, 2012. ADAMS Accession No. ML12202B154.

Rudland DL, DJ Shim, H Xu and G Wilkowski. 2007a. Summary Report on Evaluation of 

Circumferential Indications in Pressurizer Nozzle Dissimilar Metal Welds at the Wolf Creek 

Power Plant. Engineering Mechanics Corporation of Columbus, Columbus, Ohio. ADAMS 

Accession No. ML071560398.

E-11
Rudland DL, DJ Shim, T Zhang and G Wilkowski. 2007b. Final Report on Implication of Wolf 

Creek Indications – Final Report to Nuclear Regulatory Commission, Washington DC. 

Engineering Mechanics Corporation of Columbus, Columbus, Ohio. August 31, 2007. ADAMS 

Accession No. ML072470394.

Rudland DL, F Brust, T Zhang, D-J Shim and G Wilkowski. 2010. Evaluation of the Inlay 

Process as a Mitigation Strategy for Primary Water Stress Corrosion Cracking in Pressurized 

Water Reactors. U.S. Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession 

No. ML101260554.

Stall JA. 2007. Letter to NRC Document Control Desk. "Inspection and Mitigation of Alloy 

82/182 Pressurizer Welds - Seabrook Station." January 31, 2007, U.S. Nuclear Regulatory 

Commission, Washington, D.C. ADAMS Accession No. ML070330591.

Wadley MD. 2008. Letter to NRC Document Control Desk. "Praire Island, Unit 2, Completion 

of Full Structural Weld Overlay in Response to Confirmatory Action Letter (NRR-07-024)." 

October 21, 2008, Prairie Island Nuclear Generating Plant, Welch, Minnesota. ADAMS 

Accession No. ML083080007.

Wallace J. 2014. Memorandum to D Rudland. "Coolant Loop Piping Flaw Growth Analyses." 

June 9, 2014, U.S. Nuclear Regulatory Commission, Washington, D.C. ADAMS Accession 

No. ML14141A302.

E-12



UNITED STATES

NUCLEAR REGULATORY COMMISSION

WASHINGTON, DC 20555-0001

------------------

OFFICIAL BUSINESS
NUREG/CR-7187 Managing PWSCC in Butt Welds by Mitigation and Inspection