Materials Handbook for Nuclear Plant Pressure Boundary Applications.pdf

Electric Power Research Institute 3420 Hillview Avenue, Palo Alto, California 94304-1338 PO Box 10412, Palo Alto, California 94303-0813 USA

800.313.3774 650.855.2121 [email protected] www.epri.com

NOTICE: This report contains proprietary information that is the intellectual property of EPRI. Accordingly, it is available only under license from EPRI and may not be reproduced or disclosed, wholly or in part, by any licensee to any other person or organization.

LICE

NS E D

M A TE

RIAL

LICE

NS E D

M A TE

RIAL

2013 TECHNICAL REPORT

Materials Handbook for Nuclear Plant Pressure Boundary Applications (2013)

EPRI Project Manager G. Ilevbare

ELECTRIC POWER RESEARCH INSTITUTE 3420 Hillview Avenue, Palo Alto, California 94304-1338 PO Box 10412, Palo Alto, California 94303-0813 USA

800.313.3774 650.855.2121 [email protected] www.epri.com

This document does NOT meet the requirements of 10CFR50 Appendix B, 10CFR Part 21,

ANSI N45.2-1977 and/or the intent of ISO-9001 (1994)

Materials Handbook for Nuclear Plant Pressure Boundary Applications (2013)

3002000122

Final Report, March 2013

DISCLAIMER OF WARRANTIES AND LIMITATION OF LIABILITIES THIS DOCUMENT WAS PREPARED BY THE ORGANIZATION(S) NAMED BELOW AS AN ACCOUNT OF WORK SPONSORED OR COSPONSORED BY THE ELECTRIC POWER RESEARCH INSTITUTE, INC. (EPRI). NEITHER EPRI, ANY MEMBER OF EPRI, ANY COSPONSOR, THE ORGANIZATION(S) BELOW, NOR ANY PERSON ACTING ON BEHALF OF ANY OF THEM: (A) MAKES ANY WARRANTY OR REPRESENTATION WHATSOEVER, EXPRESS OR IMPLIED, (I) WITH RESPECT TO THE USE OF ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT, INCLUDING MERCHANTABILITY AND FITNESS FOR A PARTICULAR PURPOSE, OR (II) THAT SUCH USE DOES NOT INFRINGE ON OR INTERFERE WITH PRIVATELY OWNED RIGHTS, INCLUDING ANY PARTY'S INTELLECTUAL PROPERTY, OR (III) THAT THIS DOCUMENT IS SUITABLE TO ANY PARTICULAR USER'S CIRCUMSTANCE; OR

(B) ASSUMES RESPONSIBILITY FOR ANY DAMAGES OR OTHER LIABILITY WHATSOEVER (INCLUDING ANY CONSEQUENTIAL DAMAGES, EVEN IF EPRI OR ANY EPRI REPRESENTATIVE HAS BEEN ADVISED OF THE POSSIBILITY OF SUCH DAMAGES) RESULTING FROM YOUR SELECTION OR USE OF THIS DOCUMENT OR ANY INFORMATION, APPARATUS, METHOD, PROCESS, OR SIMILAR ITEM DISCLOSED IN THIS DOCUMENT.

REFERENCE HEREIN TO ANY SPECIFIC COMMERCIAL PRODUCT, PROCESS, OR SERVICE BY ITS TRADE NAME, TRADEMARK, MANUFACTURER, OR OTHERWISE, DOES NOT NECESSARILY CONSTITUTE OR IMPLY ITS ENDORSEMENT, RECOMMENDATION, OR FAVORING BY EPRI.

THE FOLLOWING ORGANIZATION, UNDER CONTRACT TO EPRI, PREPARED THIS REPORT: Dominion Engineering, Inc.

THE TECHNICAL CONTENTS OF THIS DOCUMENT WERE NOT

PREPARED IN ACCORDANCE WITH THE EPRI NUCLEAR QUALITY ASSURANCE PROGRAM MANUAL THAT FULFILLS THE REQUIREMENTS OF 10 CFR 50, APPENDIX B AND 10 CFR PART 21, ANSI N45.2-1977 AND/OR THE INTENT OF ISO-9001 (1994). USE OF THE CONTENTS OF THIS DOCUMENT IN NUCLEAR SAFETY OR NUCLEAR QUALITY APPLICATIONS REQUIRES ADDITIONAL ACTIONS BY USER PURSUANT TO THEIR INTERNAL PROCEDURES.

NOTE For further information about EPRI, call the EPRI Customer Assistance Center at 800.313.3774 or e-mail [email protected].

Electric Power Research Institute, EPRI, and TOGETHERSHAPING THE FUTURE OF ELECTRICITY are registered service marks of the Electric Power Research Institute, Inc.

Copyright 2013 Electric Power Research Institute, Inc. All rights reserved.

This publication is a corporate document that should be cited in the literature in the following manner:

Materials Handbook for Nuclear Plant Pressure Boundary Applications (2013). EPRI, Palo Alto, CA: 2013. 3002000122.

iii

ACKNOWLEDGMENTS

The following organization, under contract to the Electric Power Research Institute (EPRI), prepared this report: Dominion Engineering, Inc. 12100 Sunrise Valley Drive, Suite 220 Reston, VA 20191 Principal Investigators J. Gorman P. Krull C. Marks P. Loo

This report describes research sponsored by EPRI.

v

PRODUCT DESCRIPTION

Utility engineers require accurate structural materials properties and performance data to make decisions regarding the adequacy of materials for nuclear power applications. To meet this need, the Electric Power Research Institute (EPRI) publishes the Materials Handbook for Nuclear Plant Pressure Boundary Applications, which has been updated and revised since its initial publication in 1998. The primary focus of the handbook is on pressure boundary materials such as those used for piping and components, pressure boundary bolting, and heat exchanger tubing. Additional information is included on materials used in related applications such as pump shafts, springs, and non-pressure boundary bolting.

Background As nuclear power plants age, structural components are being repaired or replaced due to operational or environmental deterioration. Utility engineers need accurate mechanical and physical properties data and updated service performance information on structural materials to assist them in making informed failure analysis, repair, and replacement decisions.

Objectives To provide accurate mechanical and physical properties data on structural materials used in

nuclear pressure boundary applications.

To provide a concise source of information on materials performance in light water reactor service.

Approach In developing the Materials Handbook, the authors focused on gathering and reviewing relevant reports of the service experience of structural materials from EPRI reports, technical literature, Institute of Nuclear Power Operations (INPO) databases, and U.S. Nuclear Regulatory Commission (NRC) document databases. The data collected not only summarize materials performance in service, but also include materials property data from sources such as vendor brochures, industry handbooks, and the American Society of Mechanical Engineers (ASME) Code.

Results Following is a summary of each section in the Materials Handbook for Nuclear Plant Pressure Boundary Applications:

Section I, Base Materials for Piping and Pressure Vessel Pressure BoundariesThis section includes chapters on carbon and low alloy steel piping; carbon and low alloy steel for pressure vessels; stainless steel for piping, components, and pressure vessels; and nickel base alloy for piping and pressure vessels. In this 2013 edition of the handbook, Chapters 1 and 2 of Section Icovering carbon and low alloy steels for pressure vessels, components, and pipinghave been revised and updated.

vi

Section II, High Strength Materials for Bolting, Valve Stems, Springs, etc.This section provides information on precipitation hardened alloys, with emphasis on Alloys 17-4PH, X-750, 718, and A-286. Also addressed are wrought high-strength austenitic stainless steels, including cold-worked Types 304 and 316 and XM-19 in the cold-worked and annealed conditions. Among the martensitic stainless steels represented are Types 403, 410, 414, CA15, and CA6NM. Non-stainless fastener steels discussed include low alloy quenched and tempered steels and maraging steels. Silicon bronze bolting alloys are covered in this section as well. In this 2013 edition of the handbook, Chapter 3 of Section IIcovering Alloy 718, a precipitation-hardened nickel-base alloyand Chapter 5 of Section IIcovering high strength wrought austenitic stainless steels (non-precipitation hardened)have been revised and updated.

Section III, Tubing AlloysThis section discusses copper tubing, titanium tubing, stainless steel tubing, carbon and low alloy steel tubing, and nickel tubing alloys (not including steam generator tubing). In this 2013 edition of the handbook, Chapter 5 of Section IIIcovering carbon and low alloy steel tubinghas been revised and updated.

Section IV, Pump and Valve Trim MaterialsThis section contains information regarding materials used in nuclear power plant pump and valve trim applications. The entire section was updated in the 2010 edition of the handbook.

Section V, Nonmetallic MaterialsThis section includes information on impurity limits for nonmetallic materials that can come into contact with pressure boundary materials. In this 2013 edition of the handbook, Section V has been updated.

Applications, Value, and Use The intent of the Materials Handbook for Nuclear Plant Pressure Boundary Applications is to provide utility engineers with information that is potentially vital to performing root-cause evaluations and making run-replace decisions on important structural components. Although the focus of this handbook is on pressure boundary components, materials used in other critical applications are included. This handbook serves as an important resource for nuclear power plant engineers as they evaluate structural components in their plants.

Keywords Bolting alloys Mechanical properties Nonmetallic materials Physical properties Pressure boundary materials Tubing

EPRI Proprietary Licensed Material

vii

ABSTRACT

The intent of the Materials Handbook for Nuclear Plant Pressure Boundary Applications is to provide utility engineers with a concise source for accessing materials information that is needed when performing failure analyses or selecting replacement materials. The main focus of the Materials Handbook is on pressure boundary materials such as the ones used for piping and components, pressure boundary bolting, and heat exchanger tubing. However, information regarding materials used in related applications such as pump shafts, springs, and non-pressure boundary bolting is also included for materials that are important to nuclear plants. The following types of information are included for each material covered in the Materials Handbook:

General description and typical applications Typical product forms and specifications Main limitations

Material properties, including mechanical and physical properties Weldability Extent to which the material is allowed for use in American Society of Mechanical Engineers

(ASME) Code applications and applicable material property data from the ASME Code Ordering information and practices Service experience Main results of research directed at nuclear applications of the material References with more detailed information


ix

EXECUTIVE SUMMARY

Introduction The intent of the Materials Handbook is to provide utility engineers with a concise source for the materials information that is often needed in connection with performing failure analyses or when selecting replacement materials. The main focus of the Materials Handbook is on pressure boundary materials such as those used for piping and components, pressure boundary bolting, and heat exchanger tubing. However, information regarding materials used in related applications such as pump shafts, springs, and non pressure boundary bolting is also included for materials that experience indicates are important to nuclear plants. Development of each chapter of the Materials Handbook involves a significant effort. The typical work performed for each chapter involves:

Gathering and reviewing relevant reports of service experience from Electric Power Research Institute (EPRI) reports, technical articles, INPO databases, and information filed by licensees with the Nuclear Regulatory Commission (NRC).

Gathering and reviewing relevant research information from the published technical literature.

Gathering material property data from sources such as vendor brochures, industry handbooks and the American Society of Mechanical Engineers (ASME) Code.

Drafting the chapter using the information gathered in the above reviews. Checking questionable issues with relevant industry experts, including experts at utilities,

EPRI, nuclear steam supply systems (NSSSs), and other engineering organizations. Reviewing each chapter by EPRI and utility experts. Resolving comments and issuing the chapter in final form.

Base Materials for Piping and Pressure Vessel Boundaries This section of the Materials Handbook covers the main base materials used for light water reactor (LWR) plant pressure boundary applications, that is, carbon and low-alloy steels, stainless steels, and nickel-base alloys.

Carbon and Low Alloy Steels for Pressure Vessels Carbon and low alloy steels have been widely used for the main vessels of nuclear power plants, such as reactor vessels, pressurizers, and steam generators. The reasons for use of carbon and low alloy steels for these pressure vessels are their combination of relatively low cost, good mechanical properties in thick sections, good weldability, and high resistance to stress corrosion cracking (SCC). With regard to reactor vessels, the grades of low alloy steels that are used also


Executive Summary

x

have acceptably low rates of embrittlement when subjected to neutron flux for long periods of time. In many reactor coolant applications, the pressure vessel steels have been clad on the inside wetted surface with corrosion resistant materials such as austenitic stainless steels or nickel alloys of the Alloy 600 type. Experience with carbon and low alloy steels in nuclear power plant pressure vessel service has generally been good, with relatively few service induced problems. In summary, experience has been as follows:

Reactor vessel core belt line region: Steels in the core beltline region are subject to embrittlement due to neutron irradiation. Embrittlement of the base materials of western design LWRs has generally not been a serious problem. However, some welds in early generation pressurized water reactors (PWRs) have been found to be especially sensitive to embrittlement and have required significant programs to address the resulting embrittlement concerns. A small lead plant, Yankee Rowe, was shut down because of issues related to reactor vessel embrittlement, but this is judged unlikely to occur at any of the currently operating plants.

A significant number of flaw indications have been detected in reactor vessels by ultrasonic testing (UT) performed for baseline or inservice inspections. Most of these flaws have been associated with welding or cladding, although a few have been due to laminations or inclusions in the steel plates or forgings. The base material flaws have rarely if ever, required repair. However, some of the weld and clad related flaws have led to repairs being made, especially when the flaws were detected before operation. There appear to be no reported cases of service induced growth of flaws present in the base plates or forgings, or of service induced growth of weld flaws that were present since initial construction.

There have been significant service induced flaws at vessel nozzles associated with mixing of lower temperature water with hot water. For example, thermal fatigue cracks developed in boiling water reactor (BWR) reactor vessel feedwater nozzles and control rod drive return line nozzles. Significant inspections and repairs were required in the late 1970s and early 1980s to address this problem. The design and procedure changes made at that time seem to have been effective since there have been no further reported occurrences.

There have been a few cases of crack initiation and growth in PWR steam generator shells at transition cone girth welds. These cracks appear to have been initiated as a result of weld damage, thermal stress cycles, and the occasional presence of oxidizing conditions. No new cases of this type of cracking have been reported since about 1991, and it appears that current water chemistry controls minimize the likelihood of serious cracking of this type in the future.

Significant numbers of cracks have developed in the cladding of BWR reactor vessel heads. In some cases, the cracks have penetrated short distances into the alloy steel base material. This cracking has required significant inspection and analysis to demonstrate the continued safe condition of the affected parts. In a few cases, it has been concluded that the cladding cracks may have penetrated into the base material as the result of service, but it appears more likely that such penetration occurred during fabrication.

A large boric acid corrosion induced cavity developed in the reactor vessel head at the Davis-Besse plant in 2002. The cavity was the result of leakage of primary coolant at a primary water stress corrosion cracking (PWSCC) induced crack in an Alloy 600 control rod drive


Executive Summary

xi

mechanism nozzle. This incident resulted in many industry and regulatory actions to ensure that appropriate inspections for boric acid corrosion are regularly performed and, more generally, that material degradation issues are given appropriate management attention and followup.

Carbon and Low Alloy Steel Piping Carbon steel piping has been widely used in nuclear power plants, especially in the following applications: reactor coolant system piping in PWRs, with internal cladding; reactor feeder headers in pressurized heavy water reactors (PHWRs); nuclear auxiliary systems such as residual heat removal and core spray piping in BWRs; conventional steam plant or balance of plant (BOP) piping, including condensate, feedwater, steam and steam drain systems; closed cooling water systems; service water systems; and condenser circulating water systems. The main reasons for using carbon steel piping are its combination of low cost, good strength, acceptable general corrosion resistance in many environments, and its relative freedom from SCC. Low alloy steel materials have also been used for some types of nuclear power plant piping, for example, steam exhaust, extraction, and drain lines. The main application is in portions of BOP piping where flow-accelerated corrosion (FAC) (also known as erosion-corrosion and hereafter called FAC/erosion-corrosion) has resulted in the need for replacement of carbon steel piping with more resistant material. Experience with carbon steel and low alloy steel piping in nuclear power plants has been as follows:

Reactor coolant system piping: Reactor coolant piping in Combustion Engineering and Babcock & Wilcox PWRs is made of carbon steel with internal stainless steel cladding. There have been no significant problems with this type of piping.

Reactor feeder piping: This piping in PHWRs has generally performed well, but some thinning of outlet piping has been experienced that has required chemistry adjustments to be made, and there have been a few cases of inside diameter (ID) and outside diameter (OD) cracking, attributed to oxygen driven intergranular stress corrosion cracking (IGSCC) and to low temperature creep cracking (LTCC).

Steam piping: Essentially all steam piping is carbon steel, and it has performed well, with no systematic serious problems.

Steam exhaust, extraction, and drain line piping: While many carbon steel steam exhaust, extraction, and drain lines have performed acceptably, many other lines have experienced wall thinning due to FAC/erosion-corrosion. This is generally attributed to the presence of two phase flow with high velocities and water droplet impingement and sometimes to the presence of abrasive magnetite particles. As a consequence of wall thinning problems, these lines have been found to require frequent monitoring and, in selected areas, replacement with more FAC/erosion-corrosion resistant materials, for example, low alloy steels or austenitic stainless steels.

Condensate and feedwater piping: Most carbon steel piping in these systems, which have single phase flow, has performed well. However, some serious cases of wall thinning due to FAC/erosion-corrosion have occurred. As a result, periodic inspections have been found to be necessary, and repair or replacement of excessively thinned areas has been required. Replacement has generally been with low alloy or austenitic stainless steels.


Executive Summary

xii

Closed cooling water system piping: The water in these systems is generally of controlled purity and is treated to inhibit corrosion. As a result, most of this piping has performed satisfactorily. However, some cases of SCC have occurred (possibly due to microbiologically influenced corrosion [MIC]), resulting in the need for more frequent inspections and increased attention to water chemistry control.

Service water system piping: The water in this piping is often untreated raw water. Numerous problems of corrosion and clogging have occurred with this type of piping, especially where the piping ID does not have a protective coating. A variety of remedial actions have been taken, including more frequent inspections, monitoring and cleaning, coating the ID with corrosion resistant materials, replacement with more corrosion resistant materials, mechanical and/or chemical cleanings followed by increased use of biocides, and conversion to recirculating closed systems. Buried portions of this piping were generally coated on the OD before burial. However, some OD initiated corrosion has occurred, leading to increased use of cathodic protection.

Condenser circulating water system piping: The water in this piping is often untreated raw water. In cases where biological growths could be a problem, treatment with biocides is sometimes used. Much of this piping is of large diameter, and it is often coated on both the ID and OD surfaces. Much of it is buried. In general, this piping has performed well, but some cases of OD corrosion damage have occurred, leading to case-specific remedial actions, such as local repairs and application of cathodic protection. Some cases of ID coating degradation have also occurred, but this has generally been detected and corrected before significant ID initiated corrosion has occurred.

Most of the carbon and low alloy steel piping material used in power plants is in the wrought or forged form. Both seamless and seam welded piping materials are covered by applicable ASTM International and ASME specifications and the ASME Code, and both types are widely used. Cast materials are also widely used, especially for fittings such as elbows and flanges.

Stainless Steel for Piping, Components, and Pressure Vessels This chapter covers stainless steels that are used for piping, components, and pressure vessels in nuclear power plants. It does not address stainless steels used for BWR or PWR reactor vessel internals, which are being addressed by the Boiling Water Reactor Vessel Internals Project for BWRs and the PWR Materials Reliability Program for PWRs. The stainless steels used for piping, components, and pressure vessels are of two main types. The first is wrought austenitic stainless steel used for parts such as piping, forged valve bodies, and pressure vessel shells. The second is austenitic cast stainless steel used for applications such as large diameter primary coolant piping, cast pump casings, and cast valve bodies; the material is, in fact, a duplex austenitic-ferritic material, usually with 10%25% ferrite. Welds in wrought material and in cast material normally also have a duplex austenitic-ferritic structure, with a few percent ferrite. Experience with stainless steels used for piping, components, and pressure vessels in nuclear power plants can be summarized as follows. Stainless steel has been widely used in BWR reactor coolant systems and safety and auxiliary

systems. In these systems, it has been used for piping, valves, pumps, and other special fittings and parts. In high temperature BWR applications such as the reactor coolant system, these materials have been subject to IGSCC in areas where the material was sensitized during


Executive Summary

xiii

fabrication, such as furnace sensitized safe ends and at weld joint heat affected zones (HAZs). The term sensitized refers to a process of chromium carbide precipitation at grain boundaries that can occur when the material is held in the temperature range of 800F1600F (427C871C). The precipitation of chromium carbide reduces the chromium concentration at the grain boundary and makes them susceptible to corrosive attack in BWR reactor coolant environments with normal water chemistry (NWC). In this regard, BWR reactor coolant with NWC has about 200 ppb oxygen, which raises the electrochemical potential (ECP) and makes it aggressive toward sensitized material. Remedial actions have been taken for all BWRs in the United States, and incidents of IGSCC in piping are now rare. Nuclear grades of piping have been developed that are resistant to this type of attack and are available for use in replacements and new applications.

Stainless steel has also been widely used in PWR reactor coolant systems and safety and auxiliary systems. Stainless steels have provided relatively troublefree service in these PWR applications. The absence of systematic IGSCC problems of the type that have affected BWRs is attributed to the low oxygen content of PWR reactor coolant, which keeps the ECP well below the zone in which IGSCC occurs in pure water environments. The relatively limited number of problems that have occurred in stainless steel parts in PWRs are generally due to either mechanical or thermal fatigue or, in a few cases, to the development in stagnant areas of aggressive environments with chlorides, concentrated boric acid, and trapped oxygen.

Cast austenitic stainless steels (CASSs) are subject to embrittlement as they age as a result of metallurgical changes that occur in the ferrite phase. This can become an issue for reactor coolant applications as plant service life increases. The factors that influence the degree of embrittlement include: service time and temperature, ferrite content, molybdenum concentration, and fabrication method. While this issue has not resulted in any reported physical problems, it has been identified as an aging management issue that needs to be addressed in license renewal. It appears that most, if not all, potential embrittlement problems can be adequately addressed by susceptibility assessments, increased inspections, and fracture mechanics analyses.

Stainless steel is sometimes used for service water systems and has generally performed satisfactorily. However, austenitic stainless steels have been found to be susceptible to pitting or crevice corrosion in freshwater service, especially under deposits. This type of under deposit attack occurs when deposits build up due to accumulation of materials such as silt or clay or biological growths. Prevention of such failures involves continuous attention to keeping the stainless steel surfaces clean and free of biological growths, for example, by preventing the entry of materials that can deposit and by the use of biocides. Stainless steels have also experienced attack under layup conditions when the water was allowed to dry, resulting in concentration of the impurities in the remaining water to aggressive levels.

The conventional grades of austenitic stainless steel, such as Types 304 and 316, were found many years ago not to work well in seawater and brackish water, mainly because of pitting and crevice corrosion. However, higher alloy stainless steels with 6% molybdenum, such as AL-6XN, have worked acceptably in seawater and brackish water environments. These grades also have increased resistance to the types of pitting attack in freshwater service described above.


Executive Summary

xiv

Both seam welded and seamless piping have been used, with seamless piping having been used more extensively than seam welded piping in reactor system applications. No reports of problems associated with use of seam welds were found during preparation of this chapter.

Nickel-Base Alloys for Pressure Vessels, Components, and Piping This chapter covers nickel-base alloys that are used for pressure vessels, components, and piping in nuclear power plants, exclusive of steam generator tubes. The base material alloys covered are Alloys 600 and 690. These alloys are widely used in BWR applications such as reactor vessel nozzle safe ends, core support structures, and shroud bolts. These alloys are also widely used in PWRs for applications such as penetrations and nozzles in PWR reactor coolant system (RCS) components, control rod drive mechanism (CRDM) and control element drive mechanism (CEDM) nozzles in reactor vessel heads, and instrument nozzles in pressurizers and RCS piping. Steam generator tubes are not covered in this chapter since they are adequately covered in other specialized books, reports and guidelines (for example, the original and revised Steam Generator Reference Book; EPRI guidelines for tube materials, inspections, and water chemistry; and reports of numerous workshops covering primary side corrosion, secondary side corrosion, and remedial measures). The performance of Alloy 600 and 690 type weld metals is covered in this chapter, in addition to Alloy 600 and 690 base materials. In this regard, Alloys 82 and 182 are the Alloy 600 type weld metals for gas tungsten arc welding (GTAW) (also known as tungsten inert gas or TIG welding) and shielded metal arc welding (SMAW), respectively, and Alloys 52 and 152 are the corresponding Alloy 690 type weld metals. These materials are used to weld Alloys 600 and 690 to themselves, to austenitic stainless steels, and to carbon and low alloy steel parts. In addition, these alloys are used as buttering and welds on carbon and low alloy parts, especially where these materials are joined to stainless steel piping, such as at nozzles. Experience with nickel-base alloys in pressure boundary service in nuclear power plants has been good in some applications. However, in other applications, some nickel-base alloys have experienced significant degradation. In summary, experience has been as follows:

Alloy 600 in some BWR applications has experienced intergranular stress corrosion cracking (IGSCC), such as at crevice locations in reactor vessel nozzle safe ends and core support structures. This IGSCC has generally occurred at welds and has been aggravated by the sensitization and residual stresses associated with welds. Alloy 600 type weld metals in BWRs have experienced cracking in locations such as nozzles and core support structures. This cracking has sometimes, but not always, been associated with crevices.

Steam generator tubes made of Alloy 600 in the mill annealed (MA) condition have experienced widespread corrosion problems, including primary water stress corrosion cracking (PWSCC), secondary side intergranular attack and stress corrosion cracking (IGA/SCC), and pitting. As discussed earlier, this experience is not addressed in this chapter because it has been adequately covered in other EPRI documents.

Alloy 600 in PWR penetration and nozzle applications has exhibited an increasing amount of PWSCC as PWRs have aged. This type of cracking was first experienced in pressurizer nozzles, with the early cracking attributed to the high temperatures in pressurizers (about 650F [343C]). The problem later occurred in other lower temperature penetrations and nozzles such as control rod drive mechanism (CRDM), and control element drive mechanism (CEDM) penetrations, and reactor coolant loop instrument nozzles.


Executive Summary

xv

Starting about 2000, a number of cases of PWSCC have been observed in Alloy 600 type weld metals in PWRs, for example, in reactor vessel outlet nozzles and at CRDM nozzle to reactor vessel head welds. PWSCC has also been observed at welds in steam generator divider plates in non-domestic PWRs.

High Strength Materials for Bolting, Valve Stems, Springs, Etc. This section of the Materials Handbook includes chapters covering precipitation hardening Alloys 17-4PH, X-750, 718, and A-286, high strength wrought austenitic stainless steels, martensitic stainless steels, high strength non-stainless fastener steels, and silicon bronze bolting alloys. With the exception of the non-stainless fastener steels and silicon bronze bolting alloys, these alloys combine good corrosion resistance in high temperature environments with high strength and are commonly used in nuclear power plants for applications such as bolting, springs, valve stems, and pump shafts where such a combination is required. The most important points regarding the use of these high strength corrosion resistant materials in these applications are summarized below: General Despite their generally high corrosion resistance, none of the alloys can be considered to be

completely resistant to SCC in reactor coolant environments. Accordingly, it is important to observe the stress, temperature, and processing limitations outlined in the chapters of this Materials Handbook when using any of these materials.

All of the high strength corrosion resistant alloys are resistant to general corrosion due to boric acid when used in external bolting applications of borated systems. In addition, except for martensitic stainless steels in poorly controlled heat treatment conditions, there have been no service failures of these materials in external bolting applications of borated systems due to SCC. However, tests indicate that at moderately high temperature (482F [250C]) and high boric acid concentrations (40%), all of these alloys except hot worked XM-19 are susceptible to SCC.I, II

Alloy 17-4 PH

While service experience indicates that any of these materials are suitable for use in external bolting applications on borated systems, except for Alloy 17-4 PH at temperatures over 482F (250C), tests indicate that there is some susceptibility to SCC in hot concentrated borated solutions of these alloys, except for hot worked XM-19, which the researchers involved indicate is a promising alloy but not yet fully qualified. Of the fully qualified alloys, Alloy A-286 performed the best in the tests. Considering this situation, the best choice for external bolting applications in high temperature borated water systems is considered to be Alloy A-286.

Alloy 17-4 PH, when it has been age hardened at 1075F (579C) or higher, has satisfactory corrosion resistance for use in external pressure boundary bolting and also in applications involving exposure to reactor coolant. However, even with these high hardening temperatures, it is still susceptible to SCC if exposed to adverse environments, for example, concentrated chlorides.

I J. M. Gras, et al., Corrosion par l acide borique concentr de materiaux pour boulangerie de composants de circuit primaire, Proceedings of the International Symposium Fontevraud I, Contribution of Materials Investigation to the Resolution of Problems Encountered in PWR Plants. pp. 178187, SFEN Sept. 1985.

II B. Prieux and J. M. Gras, Corrosion of Bolting Stainless Steels in High Temperature Concentrated Boric Acid, International Symposium Fontevraud II. Vol. 2, pp. 558567, SFEN Sep. 10-14, 1990.


Executive Summary

xvi

Alloy 17-4 PH is susceptible to SCC when it has been age hardened at temperatures below about 1075F (579C), and is very susceptible in the 900F (482C) heat treatment condition. For example, material in the 900F (482C) heat treatment condition can experience SCC when exposed to aggressive chemistries at ambient temperatures, if highly stressed.

Tests and service experience have shown that 17-4 PH, even in the more resistant 1100F (593C) condition, is susceptible to embrittlement at service temperatures of about 500F (260C) and higher. This embrittlement increases its susceptibility to both brittle fracture under impact loads and to SCC.

In summary, Alloy 17-4 PH can be used safely at service temperatures below 500F (260C) if it was aged at 1075F (579C)or higher. For service temperatures above 500F (260C), it is suggested that the allowable operating time be based on the guidance in the Alloy 17-4 PH chapter.

Alloy X-750 Alloy X-750, when in the age hardened conditions commonly used during the 1970s (for

example, the two step aged condition known as the AH condition), has relatively high susceptibility to SCC in reactor coolant environments.

Extensive testing indicates that Alloy X-750, when given a higher temperature solution anneal followed by a single step aging treatment, has increased resistance to SCC. However, it is not immune, and limitations on peak in-service stresses need to be observed. Guidance for proper use of this material in PWRs is contained in an EPRI specification, which covers a condition suitable for PWR reactor coolant service (Condition CIB or HTH).III For BWR applications, the guidance of the latest issue of BWRVIP 84 is applicable.IV

Susceptibility of Alloy X-750 to SCC is higher in BWR reactor coolant environments than in PWR environments. Accordingly, stress levels in BWR environments need to be lower than in PWR environments, and there is increased incentive to use a lower strength more SCC resistant material condition.

The susceptibility of Alloy X-750 to SCC has been found to be strongly dependent on its surface condition. Surface damage from precipitation heat treatments can increase its susceptibility, and final machining after heat treatment has been found to improve its resistance to SCC. Guidance regarding proper treatment of Alloy X-750 during manufacture is contained in an EPRI report.IV

Alloy 718

Alloy 718 has had generally good service experience, but it has experienced a limited number of cases of service induced SCC in nuclear plant applications, and some fatigue initiated failures have been reported. In addition, testing indicates that Alloy 718 exhibits relatively high SCC crack growth rates when in commercially available heat treatment conditions. Thus, the good service experience probably indicates a relatively high resistance to crack initiation.

III Material Specification for Alloy X-750 for Use in LWR Internal Components (Revision 1). EPRI, Palo Alto, CA: 1990. NP-7032.

IV BWRVIP-84, Revision 2: BWR Vessel and Internals Project, Guidelines for Selection and Use of Materials for Repairs to BWR Internal Components. EPRI, Palo Alto, CA: 2012. 1026603.

IV Design and Manufacturing Guidelines for High-Strength Components in LWRs Alloy X-750. EPRI, Palo Alto, CA: 1991. NP-7338-L.


Executive Summary

xvii

Tests supported by EPRI have shown that Alloy 718 in alternative heat treatment conditions exhibits lower SCC crack growth rates. These specific alternative heat treatments are not covered by the ASME Code. However, Code Case N-60-6 has recently been issued by the ASME that covers use of a modified grade of Alloy 718 with a revised heat treatment schedule that results in somewhat lower strengths and, based on tests, improved SCC properties.

Tests indicate that Alloy 718 has better IASCC resistance than most other alloys.

Alloy A-286 Alloy A-286 has poor resistance to SCC in oxygenated BWR environments and is generally

not appropriate for use in applications involving exposure to BWR reactor coolant.

Tests and service experience indicate that Alloy A-286 in commercially available heat treatments is susceptible to SCC in PWR reactor coolant environments if peak stresses are at the yield level or higher. Applications with peak stress levels below the yield stress, taking into account stress concentration factors and residual stresses, have not resulted in service induced SCC in PWR reactor coolant environments.

Improved heat treatment conditions that increase Alloy A-286s resistance to SCC in reactor coolant environments have not been systematically explored or identified.

Of the fully qualified corrosion resistant alloys that can be considered for use as external bolting for borated reactor coolant systems, Alloy A-286 is indicated by tests in hot concentrated boric acid as having the best resistance to SCC.

High Strength Wrought Austenitic Stainless Steels Two types of high strength wrought austenitic stainless steels are used in nuclear

applications: cold worked materials, such as cold worked Type 304 and 316 stainless steels, and nitrogen strengthened wrought Alloy XM-19, which may also be further strengthened by cold or hot work.

Cold worked austenitic stainless steels have been found to be susceptible to SCC in BWR environments and, as a result, are not widely used in BWRs.

Austenitic stainless steels that have been purposely cold worked to raise the yield strength have not experienced SCC in normal flowing PWR reactor coolant environments except in some cases where exposed to high neutron fluence. In this regard, cold worked Type 316L bolting has experienced irradiation assisted SCC (IASCC) in French plants. However, no plants in the United States have detected this condition, and cold worked normal grade Type 316 stainless steel bolting has been widely used in the United States in reactor internals applications with no reported failures. Cold worked Type 316Ti has been used for a long time in Siemens PWR reactor internals applications with some limited SCC detected.

While austenitic stainless steels that have been purposely cold worked to raise the yield strength have generally not experienced PWSCC in normal flowing PWR coolant environments (except for the 316L and 316Ti bolting discussed above), some cases of SCC have been found where superficial cold work has occurred. This has mainly affected pressure boundary base materials and not the bolting applications covered in this chapter; nevertheless this indicates that care needs to be taken to avoid accidental cold work of high strength austenitic stainless steel parts.


Executive Summary

xviii

Despite the generally satisfactory performance of cold worked austenitic stainless steel in PWRs, there are few published tests or qualification data for PWR environments (except for service experience with Type 316Ti, which is documented). Thus, the specific types of materials and ranges of stress and environment for which the materials are qualified are not well documented.

Alloy XM-19 is being used in increasing amounts in both BWR and PWR environments. Alloy XM-19 has mainly been used in the annealed condition, but has recently also been used in BWRs in a higher strength hot rolled condition. It is not known whether Alloy XM-19 has been used in the hot rolled or cold worked conditions in PWRs.

Martensitic Stainless Steels Martensitic stainless steels are widely used in LWRs in applications such as turbine blades,

bolting, valve stems and hardware, and pump shafts and hardware. Martensitic stainless steels have performed satisfactorily in many of these applications. However, there have been some failures, often related to the use of material with improper heat treatment.

The most common problem experienced with martensitic stainless steels has been the use of material with too low a tempering temperature, with a resultant high hardness and relatively high susceptibility to SCC.

Temper embrittlement has also been a problem with martensitic stainless steels. This occurs due to tempering in, or slow cooling through, the temperature range of about 750F1020F (399C549C). In addition, other sequences of improper thermo-mechanical processing or heat treatment can apparently result in material meeting specified mechanical properties but nevertheless having poor corrosion resistance and/or low ductility.

The main protection against use of martensitic stainless steels with poor corrosion resistance and/or low ductility is to assure via quality control that the material receives the specified austenitizing, quenching, and tempering treatments. Performance of impact tests, micro hardness tests, and metallographic examinations can also help assure that such conditions are not present.

Some grades of martensitic stainless steel are subject to embrittlement caused by aging at service temperatures. Tests indicate that 13% chromium steels such as Types 403, 410, 414, and CA15 with about 1% nickel or less will not embrittle significantly due to time at temperatures of about 575F (302C) for 300,000 hours, but that alloys with significant amounts of nickel, such as Type 431 (16% chromium, 2% nickel) and CA6NM (13% chromium, 4% nickel, 0.7% molybdenum) may be subject to embrittlement and increased susceptibility to SCC for aging times of several hundred thousand hours or more,

depending on their composition and service temperature (the higher the temperature, the more the embrittlement).V, VI

V M. Tsubota, et al., Characterization of Long Term Aged Martensitic Stainless Steels, Proceedings of the Fifth International Symposium on Environmental Degradation of Materials in Nuclear Power Systems Water Reactors. ANS, pp. 305310, 1992.

VI B. Yrieix and M. Guttmann, Aging Between 300 and 450C of Wrought Martensitic 13-17% Cr Stainless Steels, Materials Science and Technology. V9, N2, pp. 125134, Feb. 1993.


Executive Summary

xix

While significant problems have been experienced with martensitic stainless steels, it is considered that material with heat treatments, quenching practices, and hardness values of the types required by ASME/ASTM specifications and suggested in this Materials Handbook have reasonable resistance to SCC and can be used in applications where they have demonstrated satisfactory performance.

High Strength Non Stainless Fastener Steels These steels do not contain enough chromium to be corrosion resistant, or stainless; that is,

they have less than about 10.5% chromium. There are two main types of high strength non-stainless fastener steels that have been used in nuclear plants: (1) low alloy steels that are quenched and tempered to achieve a balance of strength and toughness and (2) maraging steels with very low carbon (0.03% or less) and high nickel (18%) that achieve their strength through precipitation hardening (for example, 18 Ni maraging steels).

With certain exceptions, high strength non-stainless fastener steels have provided satisfactory performance in U.S. LWRs. The main exceptions are as follows: Degradation of reactor coolant pressure boundary bolting due to general borated water

corrosion (wastage or erosion/corrosion). Low alloy steels experience rapid general corrosion when in contact with concentrated liquid boric acid solutions. Therefore, when these steels are used for bolted closures in reactor coolant pressure boundaries, leaks can give rise to rapid loss of material due to boric acid attack.

Degradation of pressure boundary bolting due to SCC. Some bolts in flanged joints have failed due to SCC; generally these types of failures can be tied to leaking gaskets and aggressive impurities from certain lubricants (particularly MoS2) or sealants. Fastener materials with specified yield strengths as low as 105 ksi (724 MPa) have failed due to SCC in reactor coolant pressure boundary applications. Out-of-specification material has not been implicated as a cause for SCC of pressure boundary bolting failures.

Degradation of supports and embedment bolting due to SCC. These failures are attributed to a combination of high stress, susceptible material condition, and a wet environment. Materials that have failed generally fall into one of two categories: (1) those that are specified with greater than 150 ksi (1034 MPa) (minimum yield strength, and (2) those that are specified with less than 150 ksi (1034 MPa) minimum yield strength but that have been supplied with a hardness higher than the specified allowable.

SCC failures of non-stainless fastener steels are currently not occurring in sufficient number to be considered a general problem. The reduction in SCC failures has been achieved by three methods: (1) elimination of MoS

2 based lubricants and use of alternatives such as

nickel-base anti-seize and graphite-alcohol lubricants, (2) reduction in bolt preload, and (3) use of lower strength material, particularly materials with yield strengths below 150 ksi (1034 MPa).

Degradation of low alloy steel fasteners due to boric acid corrosion continues to be a general concern because of the difficulty associated with eliminating leaks at reactor coolant pressure boundary joints. The main approaches used by the industry to address this concern have been to (1) minimize the occurrence of leaks through improved bolting procedures, for example, use of higher, more precisely controlled, preloads, and (2) switch to corrosion resistant alloys for applications where susceptibility to boric acid attack remains a concern.


Executive Summary

xx

Silicon Bronze Bolting Alloys Silicon bronze bolting alloys are mainly used in electrical equipment such as switchgear and

motor control centers. They are included in the Materials Handbook (even though they are rarely if ever used for pressure boundary applications) because they have been subject to stress corrosion failures in nuclear plants, and it is considered desirable to compile information about this problem.

Silicon bronzes are copper-base alloys with small amounts of zinc and silicon levels ranging from about 1% to 3%. Alloys with approximately 1% silicon are known as low silicon bronze and alloys with about 3% silicon are known as high silicon bronze. The only application of these alloys in nuclear plants covered in the NRCs Public Document Room holdings and in individual utility experience reports is as bolting in electrical equipment. In addition, some informal reports of service experience show that these alloys are also used in electrical equipment for services such as leaf spring/terminal connectors.

Silicon bronze alloys are covered in the ASME Code for use in Class 3 bolting applications and for low temperature (<

Significant SCC problems have been experienced with silicon bronze bolting in nuclear plant electrical equipment, and these problems have led to replacement of the silicon bronze with alternative alloys in several cases. However, it is possible that silicon bronze would perform satisfactorily if manufacturing practices were controlled so as to limit residual stresses and if controls were placed on installation practices to limit applied stresses.

300F [< 149C]) Class 2 and 3 pressure boundary applications and may have been used for these applications, although no reports of such use were located.

The main limitation with regard to use of silicon bronze bolting alloys is that they have a relatively high susceptibility to SCC. Susceptibility to SCC is especially acute in an ammoniated environment, which can result from use of ammoniated cleaning solutions, but it also is present in industrial and marine atmospheres. To protect against SCC, it is important to prohibit use of ammoniated cleaning solutions on or near silicon bronze bolting, to control torques and lubrication to ensure that applied stresses are not excessive, and to ensure that the manufacturing process used does not result in high residual stresses.

There are no reports of SCC of bolts known to have been machined from bar stock; whereas SCC has occurred fairly often with bolts formed by cold heading, especially carriage bolts. For this reason, it appears that it would be prudent to approach the use of cold headed bolts with caution, that is, use them only in cases where service experience has shown that SCC is not a concern for the specific manufacturing method and application involved. In addition, it is suggested that application of a thermal stress relief following cold heading be considered, for example, for 1 hour at 700F (371C).VII

Tubing Alloys

This section of the Materials Handbook includes chapters covering copper alloy tubing, titanium tubing, stainless steel tubing, carbon steel tubing, and nickel-base alloy tubing. The most important points regarding the use of these tubing alloys are summarized below:

VII Metals Handbook, Ninth Edition, Volume 13, Corrosion. ASM International, p. 615, 1987.


Executive Summary

xxi

Copper Alloy Tubing Copper alloy tubing has been widely used for many years in nuclear power plants, primarily

because it has good general corrosion resistance, high thermal conductivity, and good resistance to biological fouling. Typical applications have been in condensers, feedwater heaters, moisture separator reheaters, containment air coolers, and service water heat exchangers. While copper alloy heat exchanger tubing is still providing acceptable service in some plants, copper alloy tubing has been replaced in many of these applications at many other plants, and is generally not being selected for most new secondary system component applications in nuclear power plants. This is because of three main problems: Copper alloys are susceptible to several types of corrosion such as erosion-corrosion and

sulfide induced attack, and this corrosion has led to corrosion induced leaks. Copper carried by the secondary system to steam generators in PWRs has been found to

contribute to corrosion problems in the steam generators. Similarly, copper carried by the condensate-feedwater system in BWRs has been implicated in fuel clad corrosion problems.

The corrosion response of copper alloys to water chemistry variables such as pH and oxygen concentration is quite different than that of ferrous alloys. For this reason, the presence of copper alloys in a PWR secondary system results in use of chemistry parameters that lead to higher iron corrosion product transport rates to the steam generators than can be achieved in copper free systems.

Copper tubing is supplied in both seamless and seam welded forms, and both forms have been used in the power industry. However, the large bulk of the material used in nuclear plants has been seamless.

Titanium and Titanium Alloys Titanium and titanium alloy tubing has been used in nuclear power plants for over 30 years.

This is primarily because titanium has excellent corrosion resistance. Titanium is used in applications where copper alloys and stainless steels are not sufficiently resistant to corrosion. There have been no reports of corrosion induced failures of titanium tubing in condenser and service water applications. However, some hydriding failures have occurred as a result of inappropriate voltages applied to cathodic protection systems in condensers.

Titaniums relatively low thermal conductivity and relatively high cost has limited its use in conventional heat exchangers. However, using thinner wall tubes usually offsets the conductivity issue. Titaniums low modulus of elasticity is also a consideration, especially when used as a replacement tubing material in heat exchangers that have been designed for other alloys. Lack of consideration of this property has occasionally resulted in serious vibration problems and even tube failure.

Titanium tubing is supplied in both seamless and seam welded forms. However, the vast majority of material used in power plants has been seam welded. In theory, the use of seam welded tubes could raise concerns about the possible influence of the seam weld on defects and corrosion behavior. However, to date, there have been no reports of any significant problems associated with the use of seam welded tubing.


Executive Summary

xxii

Stainless Steel Tubing Austenitic stainless steel tubing has been widely used in freshwater cooling systems and in

condensate-feedwater systems and has generally performed well. However, the austenitic stainless steels have been found to be susceptible to pitting or crevice corrosion in freshwater service, especially under deposits. This type of under deposit attack occurs when deposits build up on tube surfaces due to deposition of materials such as silt or clay, precipitation of hardness salts, biological growths, etc. Prevention of such failures involves continuous attention to keeping the stainless steel surfaces clean and free of biological growths, for example, by using sponge ball cleaning and chlorination. The austenitic grades of stainless steel have also experienced attack under layup conditions when the water in the tubes was allowed to dry, resulting in concentration to aggressive levels of the impurities in the remaining water.

The conventional grades of austenitic stainless steel such as Types 304 and 316 were found many years ago not to work well in seawater and brackish water, mainly because of pitting and crevice corrosion. However, higher alloy stainless steels of either the austenitic type, such as AL-6XN and 904L, or the ferritic type, such as Sea-Cure, E-Brite and AL29-4C, have worked acceptably in seawater environments, although a few corrosion induced failures have been reported. These grades also have increased resistance to the types of pitting attack in freshwater service described above.

Ferritic stainless steels, especially Type 439, have been widely used in moisture separator reheater applications and have worked well, with very few failures reported. They have also been used to a limited extent in feedwater heaters.

Both seam welded and seamless tubing have been used, with seam welded tubing being used more extensively than seamless tubing. For the most part, the seam welds have not caused problems. However, there have been cases where seam welds in non-solution annealed austenitic stainless steel tubing have led to increased susceptibility to a pitting attack in seawater service. Thus, some caution is appropriate when considering the use of seam welded tubing, and it is prudent to use only seam welded tubing if the tubing is heat treated after welding.

Carbon and Low Alloy Steel Tubing Carbon and low alloy steel tubing has been widely used in two main applications in nuclear

power plants: feedwater heaters, especially high pressure feedwater heaters, and moisture separator reheaters (MSRs). Carbon steel has also had limited use in service water applications where the service water is fresh water.

A main reason for an increased use of carbon steel for feedwater heaters in the 1960s and 1970s was reluctance to use copper alloys because of concerns about the effects of copper on boilers and turbines, especially in fossil plants. This fossil plant experience was carried over to nuclear plants, especially PWRs. Other reasons for using carbon and low alloy steel for feedwater heaters and MSRs, as opposed to more corrosion resistant alloys, are its low cost, its good thermal conductivity, resistance to SCC, and the fact that its thermal coefficient of expansion matches that of pressure vessel structural parts, thus simplifying design.


Executive Summary

xxiii

The main drawbacks of carbon and low alloy steels are their susceptibility to (1) general corrosion and pitting, (2) erosion-corrosion or wall thinning in single phase flow regions (also called flow accelerated corrosion (FAC)), and (3) steam impingement erosion-corrosion or cutting due to effects of high velocity water droplet impingement in regions of two phase flow.

Experience with carbon steel and low alloy steel tubing in nuclear power plants can be summarized as follows: Carbon steel tubes have performed well in many high pressure feedwater applications and

carbon steel continues to be widely used for this application, especially in Europe. However, some plants have had sufficient problems due to corrosion and erosion-corrosion that carbon steel tube bundles have been replaced with stainless steel bundles.

Finned carbon steel tubes are continuing to perform satisfactorily in MSRs in some plants, but they have been replaced in a substantial number of MSRs with ferritic stainless steel tubes as a result of erosion-corrosion problems, deposit buildup, and corrosion caused bridging of the fins (which reduces heat transfer performance). Finned low alloy steel MSR tubes have performed well in most plants. The low alloy steel used is similar in composition to Cor-Ten, which is a copper bearing weathering steel with improved atmospheric corrosion resistance.

In some service water applications, especially those with treated fresh water, carbon steel tubes have performed satisfactorily. However, in other service water applications, carbon steel tubes have been found to be susceptible to pitting and under deposit corrosion and have had to be replaced.

Both seam welded and seamless carbon and low alloy steel tubing are covered by applicable ASTM and ASME specifications and the ASME Code, and both types have been used.

Nickel Alloy Tubing This chapter covers use of nickel alloy tubing for use in balance-of-plant (BOP) and other

non-steam generator heat exchanger applications in nuclear power plants (steam generator tubing is covered in other EPRI documents such as the Steam Generator Reference Book, Revision 1 (TR-103824), Guidelines for Procurement of Alloy 690 Steam Generator Tubing (TR-016743-V2R1), the annual Steam Generator Progress Report in EPRIs Steam Generator Degradation Database, and the PWR Secondary Water Chemistry Guidelines (1016555).

Nickel alloy tubing has had limited use in BOP and other non-steam generator applications in nuclear plants, with these applications mainly being the use of Alloy 400 in a few early BWR feedwater heaters and in moderator and shutdown heat exchangers in PHWRs. Nickel alloy tubing (for example, Alloy 600) has also been used occasionally for services such as sample coolers.

The Alloy 400 tubing used in early BWR feedwater heaters was replaced in the 1960s as part of programs to remove copper alloys from the secondary system because of copper deposit induced fuel failures. Alloy 400 used in auxiliary heat exchangers in PWHRs, for example, for moderator and shutdown heat exchangers, has performed satisfactorily at some plants, but has suffered rapid under deposit corrosion and has been replaced by other alloys at other plants.


Executive Summary

xxiv

Despite its rather limited use in non steam generator applications in nuclear power plants, nickel alloy tubing has some attractive features, such as good corrosion resistance and mechanical properties, and can be considered as a replacement alloy for some heat exchanger applications.

Based on the published literature, only seamless nickel alloy tubing has been used in nuclear plants. However, some applicable specifications allow use of seam welded tubing, and seam welded tubing of stainless steel alloys has performed well in similar applications. Thus, it seems reasonable to consider use of seam welded tubing for nuclear power plant heat exchanger applications.

Pump and Valve Trim Materials This chapter contains information regarding materials used in nuclear power plant pump and

valve trim applications. These trim materials generally are not part of the pressure boundary and, therefore, are not within the original scope of the Materials Handbook for Nuclear Pressure Boundary Applications, which, as its name implies, was limited to pressure boundary materials. However, feedback from utility users of the Materials Handbook indicated that information to aid in the selection and specification of materials for pump and valve trim would be useful. Accordingly, this chapter has been included in the handbook.

The scope of the chapter covers materials for the following types of pump and valve trim applications: Pumps: shafts, impellers, diffusers, wear rings, mechanical seals, internal bolting and

hardware, etc. Valves: shafts and stems, discs, cages, plugs, seats, internal bolting and hardware, disc

hinge pins, pressure seals, internal springs, etc.

In addition to identifying the materials used in different trim applications, the chapter covers some general technical topics that affect trim materials, such as galling and wear.

The discussion of trim materials in this chapter covers: Lists of materials that are known to have been used in each application in order to

provide assistance to engineers attempting to identify alternative materials. Comments describing problems that have occurred with specific trim materials. Discussion of the state of knowledge regarding several general topics important to trim,

such as wear and galling resistance.

Non-Metallic Materials

This chapter describes the results of an evaluation of requirements regarding impurities in non-metallic materials for nuclear plant application. The reason for including a discussion of non-metallic materials in the Materials Handbook is that impurities from non-metallic materials can affect the integrity of pressure boundary materials. In this regard, it is necessary to control the composition and quality of non-metallic materials to ensure that they do not lead to corrosion problems in pressure boundary materials.


Executive Summary

xxv

The main conclusion reached by this evaluation of impurities in non-metallic materials is that the limits on impurities imposed on non-metallic materials have no well defined technical bases, but rather are largely historical in nature and appear to be mainly based on an as low as reasonably achievable (ALARA) principle. It was further concluded that, since the impurity limits are based on ALARA rather than defined technical considerations, they can be modified using ALARA principles, that is, relaxed where meeting them is not reasonably achievable, but retained where reasonably achievable.

It was found that high temperature gaskets and packing typically involve graphite materials, and that such materials meeting nuclear grade impurity limits are readily available. However, gaskets for lower temperature applications can involve non graphite materials such as rubbers, and meeting impurity limits for such materials can be difficult. It appears likely that higher impurity limits could be justified for such applications, but developing alternate impurity limits would require significant effort on a case-by-case basis.

References for Executive Summary References used in the Executive Summary are listed as footnotes. References for each chapter are located at the end of each chapter.


xxvii

ACRONYMS

ABB-Atom ASEA Brown Boveri (a European NSSS) ABB/CE ASEA Brown Boveri/Combustion Engineering (an NSSS) AEC Atomic Energy Commission AECL Atomic Energy of Canada Limited ALARA As low as reasonably achievable AMS Aerospace Material Specification ANL Argonne National Laboratory ANSI American National Standards Institute ASM American Society for Metals ASME American Society of Mechanical Engineers ASTM American Society for Testing and Materials AVT All volatile treatment B&W Babcock & Wilcox (a NSSS) BOP Balance of plant BWR Boiling water reactor BWROG BWR Owners Group BWRVIP BWR Vessels and Internals Project CANDU Canadian deuterium uranium (a type of reactor) CEA Commisariat lEnergie Atomique (French AEC) CASS Cast austenitic stainless steel CEDM Control element drive mechanism CERT Constant extension rate CRD Control rod drive CRDM Control rod drive mechanism CRDRL Control rod drive return line CT Compact tensile CVCS Chemical and volume control system


Acronyms

xxviii

EAC Environmentally assisted cracking ECP Electrochemical potential ECT Eddy current test EDF Electricit de France EDM Electrical discharge machining EDS Energy dispersive spectroscopy EFPY Effective full power year EPR Electrochemical potentiokinetic reactivation EPRI Electric Power Research Institute FAC Flow-accelerated corrosion FWH Feedwater heater GALL Generic aging lessons learned GE General Electric GTAW Gas tungsten arc welding HAZ Heat affected zone HSW Heat sink welding HV Hardness, Vickers HWC Hydrogen water chemistry IAEA International Atomic Energy Agency IARDATA Irradiation-Anneal-Reirradiation Database IASCC Irradiation accelerated stress corrosion cracking ID Inner diameter IGA Intergranular attack IGSCC Intergranular stress corrosion cracking IHSI Induction heating stress improvement INPO Institute of Nuclear Power Operations KWU Kraftwerk Union, a division of Siemens (a European NSSS) LAS Low alloy steel LOCA Loss of coolant accident LPHSW Last pass heat sink welding LTCC Low temperature creep cracking LTS Low temperature sensitization LTSCC Low temperature stress corrosion cracking


Acronyms

xxix

MSIP Mechanical stress improvement process MIC Microbiologically influenced corrosion MRP Materials Reliability Program MSR Moisture separator reheater NDE Nondestructive examination NDT Nil ductility temperature NG Nuclear Grade NMCA Noble metal chemical addition NPS Nominal Pipe Size NRC Nuclear Reactor Commission NSSS Nuclear Steam System Supplier NUMARC Nuclear Management and Resources Council NWC Normal water chemistry OD Outer diameter PHWR Pressurized heavy water reactor PORV Pilot operated relief valve PTS Pressurized thermal shock PWHT Post weld heat treatment PWR Pressurized water reactor PWSCC Primary water stress corrosion cracking R Ratio of minimum to maximum, stress or stress intensity RCS Reactor coolant system RG Regulatory Guide RT Reference temperature or Radiographic test SANS Small angle neutron scattering SCC Stress corrosion cracking SGOG Steam Generator Owners Group SHE Standard hydrogen electrode SICC Strain induced corrosion cracking SMAW Shielded metal arc welding SSRT Slow strain rate test STP Standard temperature and pressure SWAP Service water assistance project (an EPRI project)


Acronyms

xxx

TEM Transmission electron microscopy or microscope TIG Tungsten inert gas UNS Unified Numbering System for materials USE Upper shelf energy UT Ultrasonic test VCD Vacuum carbon deoxidation

Chapter Preparation and Latest Revision Dates

Number Name First Issued Latest

Revision (I)1 CARBON AND LOW ALLOY STEELS FOR PRESSURE VESSELS 2002 2011 (I)2 CARBON AND LOW ALLOY STEEL PIPING 2002 2011 (I)3 STAINLESS STEEL FOR PIPING, COMPONENTS AND PRESSURE

VESSELS 2002 2009

(I)4 NICKEL-BASE ALLOYS FOR PRESSURE VESSELS, COMPONENTS AND PIPING

2002 2009

(II)1 17-4 PH PRECIPITATION HARDENING STAINLESS STEEL 1998 2006 (II)2 ALLOY X-750 PRECIPITATION HARDENING NICKEL-BASE ALLOY 1998 2008 (II)3 ALLOY 718 PRECIPITATION HARDENING NICKEL-BASE ALLOY 1998 2012 (II)4 ALLOY A-286 PRECIPITATION HARDENING AUSTENITIC

IRON-BASE ALLOY 1998 2010

(II)5 HIGH STRENGTH WROUGHT AUSTENITIC STAINLESS STEELS (NONPRECIPITATION HARDENED)

1998 2012

(II)6 MARTENSITIC STAINLESS STEELS (NON-PRECIPITATION HARDENED)

1998 2006

(II)7 HIGH STRENGTH NON-STAINLESS FASTENER STEELS (2008 UPDATE)

1998 2008

(II)8 SILICON BRONZE BOLTING ALLOYS 1998 2008 (III)1 COPPER ALLOY TUBING 1999 2010 (III)2 TITANIUM AND TITANIUM ALLOYS 1999 2008 (III)3 STAINLESS STEEL TUBING 1999 2008 (III)4 NICKEL ALLOY TUBING 2002 2008 (III)5 CARBON AND LOW ALLOW STEEL TUBING 2002 2011 (IV)1 PUMP AND VALVE TRIM MATERIALS 2006 2010 (V)1 NON-METALLIC MATERIALS 2002 2012


xxxi

CONTENTS

SECTION I BASE MATERIALS FOR PIPING AND PRESSURE VESSEL PRESSURE BOUNDARIES .........................................................................................................

1 CARBON AND LOW ALLOY STEELS FOR PRESSURE VESSELS.............................. (I) 1-1 1 General Description ..................................................................................................... (I) 1-1 2 Applications ................................................................................................................. (I) 1-2 3 Typical Product Forms and Specifications ................................................................... (I) 1-5 4 Main Limitations .......................................................................................................... (I) 1-5 5 Material Properties ...................................................................................................... (I) 1-6

5.1 Room Temperature Mechanical Properties .......................................................... (I) 1-6 5.2 Elevated Temperature Mechanical Properties ...................................................... (I) 1-6 5.3 Physical Properties .............................................................................................. (I) 1-7

6 Welding and Heat Treatment ....................................................................................... (I) 1-8 7 Application Specific Comments ................................................................................... (I) 1-9

7.1 ASME Boiler and Pressure Vessel Code Applications ......................................... (I) 1-9 7.2 Parts Immersed in Reactor Coolant ................................................................... (I) 1-14 7.3 Fracture Toughness Requirements .................................................................... (I) 1-14

7.3.1 Introduction to Fracture Toughness and Radiation Embrittlement .............. (I) 1-14 7.3.2 ASME Code Requirements ........................................................................ (I) 1-18 7.3.3 ASME Code Cases That Deal with Radiation Embrittlement Issues ........... (I) 1-23 7.3.4 NRC Requirements .................................................................................... (I) 1-23

8 Ordering Information and Practices ........................................................................... (I) 1-24 9 Service Experience.................................................................................................... (I) 1-24

9.1 Summary ........................................................................................................... (I) 1-25 9.2 Radiation Embrittlement ..................................................................................... (I) 1-26 9.3 Cracking of BWR Feedwater Nozzles ................................................................ (I) 1-28 9.4 Cracking of BWR Control Rod Drive Return Line (CRDRL) Nozzles .................. (I) 1-29 9.5 PWR Steam Generator Girth Weld Cracking ...................................................... (I) 1-30


xxxii

9.6 BWR Secondary Steam Generator Shell Cracks ............................................... (I) 1-34 9.7 Cracks in Sensitized Cladding in BWR Reactor Vessels with Possible Small Penetration into the Base Material ........................................................................... (I) 1-36 9.8 Fabrication-Induced Base Material and Weld Joint Flaws Detected After Installation or in Service ........................................................................................... (I) 1-36

9.9 Boric Acid Corrosion (BAC) of Pressure Vessel Shells ........................................... (I) 1-38 9.10 Flow Accelerated Corrosion (FAC) of Feedwater Heater and Moisture Separator Reheater Shells/Pressure Vessels ............................................................................... (I) 1-38 10 Laboratory Investigations......................................................................................... (I) 1-39

10.1 Radiation Embrittlement ................................................................................... (I) 1-39 10.2 Stress Corrosion Cracking in BWR Environments ............................................ (I) 1-47 10.3 Stress Corrosion Cracking in PWR Environments ............................................ (I) 1-56 10.4 Corrosion Fatigue in LWR Environments ......................................................... (I) 1-60 10.5 Crack Growth Rate Model and Crack Tip Chemistry ........................................ (I) 1-80 10.6 Improved Initial Properties ............................................................................... (I) 1-80 10.7 Warm Prestressing .......................................................................................... (I) 1-81 10.8 Fracture Properties of Clad Material................................................................. (I) 1-81 10.9 Boric Acid Corrosion ........................................................................................ (I) 1-81

11 Alternative Materials ................................................................................................ (I) 1-81 12 References .............................................................................................................. (I) 1-82

2 CARBON AND LOW ALLOY STEEL PIPING ................................................................. (I) 2-1 1 General Description ..................................................................................................... (I) 2-1 2 Applications ................................................................................................................. (I) 2-7 3 Typical Product Forms and Specifications ................................................................... (I) 2-8 4 Main Limitations .......................................................................................................... (I) 2-9 5 Material Properties .................................................................................................... (I) 2-10

5.1 Room Temperature Mechanical Properties ........................................................ (I) 2-10 5.2 Elevated Temperature Mechanical Properties .................................................... (I) 2-10 5.3 Physical Properties ............................................................................................ (I) 2-10

6 Welding and Heat Treatment ..................................................................................... (I) 2-11 7 Application Specific Comments ................................................................................. (I) 2-11

7.1 ASME Boiler and Pressure Vessel Code Applications ....................................... (I) 2-11 7.2 Parts Immersed in Reactor Coolant ................................................................... (I) 2-16



xxxiii

9.1 Summary ........................................................................................................... (I) 2-16 9.2 Flow-Accelerated Corrosion (FAC/Erosion-Corrosion) ....................................... (I) 2-17 9.3 Boric Acid Corrosion of Carbon and Low Alloys Steels ...................................... (I) 2-24 9.4 Thermal Stratification Induced Fatigue Failures of Feedwater Piping ................. (I) 2-24 9.5 Fatigue Failures ................................................................................................. (I) 2-26 9.6 Open System Service Water Piping ................................................................... (I) 2-27 9.7 SCC of Closed Cooling Water Piping ................................................................. (I) 2-29 9.8 OD Corrosion of Buried Piping ........................................................................... (I) 2-30 9.9 Strain-Induced Corrosion Cracking (SICC) ........................................................ (I) 2-30 9.10 Cracking of PHWR Outlet Feeder Pipes .......................................................... (I) 2-31

10 Laboratory Investigations......................................................................................... (I) 2-31 10.1 General Corrosion and Corrosion Product Release ......................................... (I) 2-31 10.2 Flow-Accelerated Corrosion and Erosion-Corrosion ......................................... (I) 2-44 10.3 Pitting, Under Deposit Corrosion, and Crevice Corrosion ................................. (I) 2-50 10.4 Stress Corrosion Cracking ............................................................................... (I) 2-51 10.5 Crack Initiation Due to Fatigue/Corrosion Fatigue ............................................ (I) 2-60 10.6 Crack Propagation Due to Fatigue/Corrosion Fatigue ...................................... (I) 2-61 10.7 Hydrogen Water Chemistry Effects on Carbon and Low Alloy Steel Piping ...... (I) 2-65 10.8 Boric Acid Corrosion ........................................................................................ (I) 2-66

11 Alternative Materials ................................................................................................ (I) 2-67 12 References .............................................................................................................. (I) 2-67

3 STAINLESS STEEL FOR PIPING, COMPONENTS, AND PRESSURE VESSELS ......... (I) 3-1 1 General Description ..................................................................................................... (I) 3-1 2 Applications ................................................................................................................. (I) 3-4 3 Typical Product Forms and Specifications ................................................................... (I) 3-5 4 Main Limitations .......................................................................................................... (I) 3-6 5 Material Properties ...................................................................................................... (I) 3-7

5.1 Room Temperature Mechanical Properties .......................................................... (I) 3-7 5.2 Elevated Temperature Mechanical Properties ...................................................... (I) 3-7 5.3 Physical Properties .............................................................................................. (I) 3-8

6 Welding and Heat Treatment ....................................................................................... (I) 3-9 6.1 Welding and Post Weld Heat Treatment .............................................................. (I) 3-9 6.2 Stress Relief ........................................................................................................ (I) 3-9

7 Application Specific Comments ................................................................................... (I) 3-9


xxxiv

7.1 ASME Boiler and Pressure Vessel Code Applications ......................................... (I) 3-9 7.2 Parts Immersed in Reactor Coolant ................................................................... (I) 3-19


9.1 Summary ........................................................................................................... (I) 3-21 9.2 BWR Piping ID Initiated IGSCC of Sensitized Standard Austenitic Stainless Steels ...................................................................................................................... (I) 3-22 9.3 SCC of PWR Stainless Steel at Surfaces Wetted by Reactor Coolant ............... (I) 3-27 9.4 Fatigue Failures of Small (


xxxv

4 NICKEL-BASE ALLOYS FOR PRESSURE VESSELS, COMPONENTS AND PIPING .. (I) 4-1 1 General Description ..................................................................................................... (I) 4-1 2 Applications ................................................................................................................. (I) 4-2 3 Typical Product Forms and Specifications ................................................................... (I) 4-7 4 Main Limitations .......................................................................................................... (I) 4-7 5 Material Properties ...................................................................................................... (I) 4-8

5.1 Room Temperature Mechanical Properties .......................................................... (I) 4-8 5.2 Elevated Temperature Mechanical Properties ...................................................... (I) 4-9 5.3 Physical Properties ............................................................................................ (I) 4-10

6 Welding and Heat Treatment ..................................................................................... (I) 4-11 7 Application Specific Comments ................................................................................. (I) 4-12

7.1 ASME Boiler and Pressure Vessel Code Applications ....................................... (I) 4-12 7.2 Parts Immersed in Reactor Coolant ................................................................... (I) 4-16


9.1 Summary ........................................................................................................... (I) 4-17 9.2 IGSCC of Alloy 600 Base Material in BWR Environments .................................. (I) 4-17 9.3 IGSCC of Alloy 182 in BWR Environments ........................................................ (I) 4-21 9.4 PWSCC of Alloys 600, 182 and 82 in PWR Environments ................................. (I) 4-24

9.4.1 Background ................................................................................................ (I) 4-25 9.4.2 Examples of Significant Incidents ............................................................... (I) 4-28

10 Laboratory Investigations......................................................................................... (I) 4-31 10.1 IGSCC Initiation in BWR Reactor Coolant Environments ................................. (I) 4-32 10.2 Crack Growth Rates in Nickel-Base Alloys and Weld Metals in BWR Reactor Coolant Environments.............................................................................................. (I) 4-54 10.3 PWSCC Initiation in PWR Reactor Coolant Environments ............................... (I) 4-77

10.3.1 Selection of Alloy 600 and Early History ................................................... (I) 4-79 10.3.2 Situation as of 2009 .................................................................................. (I) 4-81

10.3.2.1 General ............................................................................................. (I) 4-81 10.3.2.2 Applied Stress and Plastic Strain ...................................................... (I) 4-81 10.3.2.3 Alloy Type and Composition ............................................................. (I) 4-82 10.3.2.4 Processing History (Including Cold Work) ......................................... (I) 4-84 10.3.2.5 Microstructure ................................................................................... (I) 4-85 10.3.2.6 Strength ............................................................................................ (I) 4-88 10.3.2.7 Temperature ..................................................................

Documents

Materials Handbook for Nuclear Plant Pressure Boundary Applications.pdf