This article discusses the main materials and water chemistry characteristics of the primary and secondary water circuits of a pressurized water reactor (PWR). It reviews the corrosion issues of PWR materials and the influence of corrosion and fouling on primary and secondary circuit radiation fields. The article explains the primary side intergranular stress corrosion cracking (IGSCC) in different materials, namely, nickel-base alloys, high-strength nickel-base alloys, low-strength austenitic stainless steels, and high-strength stainless steels. The secondary side corrosion in steam generator including denting, pitting, intergranular attack and IGSCC is also discussed. The article examines laboratory studies that have resulted in models and computer codes for evaluating and predicting intergranular corrosion, and considers the remedial actions for preventing or arresting intergranular corrosion. It concludes with information on the external bolting corrosion in nuclear power reactors.

Surface coatings are essential in all facilities that process nuclear materials or use nuclear fission for power generation. This article describes the coatings used in two basic types of Generation 3 nuclear reactor designs in the United States and their containment size. These reactors are the boiling water reactor (BWR) and pressurized water reactor (PWR). The article provides information on the loss-of-coolant accident (LOCA) identified as the design basis accident (DBA), which can rapidly de-water the core of an operating nuclear reactor. To avoid LOCA, both the BWR and the PWR include emergency core cooling systems. The article describes a DBA test and other coating performance parameters necessary for safety-related coating systems. It provides a detailed account of the selection criteria of coating types in a nuclear plant. The article concludes by highlighting protective coating strategies in Generation 3 Plants.

This article reviews a series of serious corrosion problems that have plagued the light water reactor (LWR) industry. It discusses the complex corrosion mechanisms involved, and the development of practical engineering solutions for their mitigation. The article contains tables that present the corrosion history of LWRs, and the ten most expensive operation and maintenance costs of corrosion for a particular reactor site.

The components used in light water reactors (LWR) often remain in contact with the primary coolant, whose typical temperatures and pressures are highly aggressive, therefore, initiating corrosion in most of the alloys. This article describes the corrosion behavior of zirconium alloys in water and heat flow conditions that causes irradiation on the zirconium alloy assemblies. It discusses the effect of irradiation on the microstructure and morphology of cladded linings. The article describes the impact of metallurgical parameters on the oxidation resistance of zirconium alloys. It concludes with a discussion on LWR coolant chemistry and corrosion of fuel rods in reactors.

This article focuses on the environmentally assisted cracking (EAC) of structural materials in boiling water reactors (BWRs), reactor pressure vessels, core internals, and ancillary piping. It discusses the effects of water chemistry on materials degradation, mitigation approaches, and their impact on aging management programs. The article reviews the effects of materials, environment, and stress factors on the cracking susceptibility of ferritic and austenitic structural alloys in BWRs. It describes the methods, such as data-based life-prediction approaches and mechanisms-informed life-prediction approaches, for predicting cracking kinetics in BWRs. The article provides information on several EAC mitigation techniques for BWR components, namely material solutions, stress solutions, and environmental solutions.

This article examines the understanding of persistent material changes produced in stainless alloys during light water reactor (LWR) irradiation based on the fundamentals of radiation damage and existing experimental measurements. It summarizes the overall trends and correlations for irradiation-assisted stress-corrosion cracking. The article addresses the effects of various radiation factors on corrosion. These include radiation-induced segregation at grain boundaries, radiation hardening, mode of deformation, radiation creep relaxation, and radiolysis. The article discusses a variety of approaches for mitigating stress-corrosion cracking in LWRs, in categories of water chemistry, operating guidelines, new alloys, design issues, and stress mitigation. It concludes with a discussion on the irradiation effects of irradiation on corrosion of zirconium alloys in LWR environments.

Damage to steels from neutron irradiation affects the properties of steels and is an important factor in the design of safe and economical components for fission and fusion reactors. This article discusses the effects of high-energy neutrons on steels. The effects of damage caused by neutron irradiation include swelling (volume increase), irradiation hardening, and irradiation embrittlement (the influence of irradiation hardening on fracture toughness). These effects are primarily associated with high-energy (greater than 0.1 MeV) neutrons. Consequently, irradiation damage from neutrons is of considerable importance in fast reactors, which produce a significant flux of high-energy neutrons during operation. Irradiation embrittlement must also be considered in the development of ferritic steels for fast reactors and fusion reactors. Although ferritic steels are more resistant to swelling than austenitic steels, irradiation may have a more critical effect on the mechanical properties of ferritic steels.

This article provides a summary of the concepts discussed in the Section “Corrosion in Specific Industries” in the ASM Handbook, Volume 13C:Corrosion: Environments and Industries. This Section applies the fundamental understanding of corrosion and knowledge of materials of construction to practical applications. The industries addressed are nuclear power, fossil and alternative fuel, land transportation, air transportation, microelectronics, chemical processing, pulp and paper, food and beverage, pharmaceutical and medical technology, petroleum and petrochemical, building, and mining and metal processing.

This article provides an overview of the safety aspects and integrity concept for pressure vessels, piping, and tubing. It focuses on the fracture mechanics approaches used to validate components with longitudinal cracks and circumferential cracks and to analyze crack growth behavior under cyclic loading. Full-scale testing facilities and the typical test results required for various applications are discussed. The article also presents information on the transferability of mechanical properties of materials.

Supercritical water oxidation (SCWO) is an effective process for the destruction of military and industrial wastes including wastewater sludge. This article discusses the unique properties of supercritical water and lists the main technological advantages of SCWO. For many waste streams, corrosion continues to be one of the central challenges to the full development of the SCWO technology. The article presents a summary of selected materials exposed to various environments as well as the observed form of corrosion in a table. It also illustrates the necessity to adopt a synergistic approach incorporating feed chemistry control, reactor design modifications, and intelligent materials selection, for mitigating degradation of SCWO systems.

Zirconium, hafnium, and titanium are produced from ore that generally is found in a heavy beach sand containing zircon, rutile, and ilmenite. This article discusses the processing methods of these metals, namely, liquid-liquid separation process, distillation separation process, refining, and melting. It also discusses the primary and secondary fabrication of zirconium and hafnium and its alloys. The article talks about the metallurgy of zirconium and its alloys with emphasis on allotropic transformation, cold work and recrystallization, anisotropy and preferred orientation, and the role of oxygen. It concludes by providing useful information on the applications of reactor and industrial grades of zirconium alloys.