Peter L. Andresen

Radiation-induced material changes and susceptibility to intergranular failure of light-water-reactor core internals

Abstract Current understanding of radiation-induced material changes that occur in light-water-reactor (LWR) core components is critically reviewed and linked to intergranular failure processes. Although the basic science of radiation damage processes in metals is reasonably well established, accurate prediction of microstructures, microchemistries and mechanical property changes in complex stainless alloys during irradiation at LWR temperatures is not possible at present. Mechanistic understanding of these radiation-induced changes in commercial alloys is considered to be of paramount importance for the mitigation of the intergranular environmental cracking that occurs in service. Fundamental research is needed to define defect–solute interactions and microstructural evolution at intermediate temperatures and dose rates pertinent to LWRs where transient effects often dominate behavior. In addition, it is essential that radiation effects on matrix microstructure and microchemistry and grain boundary microchemistry be understood. Finally, a stronger emphasis on accurately quantifying radiation effects on environmental cracking mechanisms and kinetics is needed.

Life prediction by mechanistic modeling and system monitoring of environmental cracking of iron and nickel alloys in aqueous systems

Abstract Environmentally assisted cracking of structural components is a life-limiting factor in the operation of, for example, light-water reactors, fossil plants, steam and gas turbines, nuclear waste containment and deaerators. In the context of an overall emphasis on light-water reactors, the shortcomings of existing design and life evaluation codes are highlighted and an improved approach is presented on the basis of the application of fundamental mechanistic understanding and a quantitative model of environmental crack advance. The integration of this model with real-time chemical sensors and reference crack growth monitors permits enhanced control and management of cracking in structural components. Validation of this approach was performed using laboratory and field data for specimens and prototypical components. For nucleaar steam supply, an on-line monitoring and predictive modeling framework has been developed for both in-core and out-of-core structural components fabricated from types 304 and 316L stainless steel, A533B and A508 low alloy steel, and Inconel 600 and Inconel 182 alloys. The conceptual and quantitative formulation of the film rupture-slip dissolution mechanism of crack advance is reviewed. The analyses and independent evaluation of the crack tip environment, chemical reaction rates and stress-strain fields, common to all fundamental models for environmentally assisted cracking are summarized. On the basis of these data, theoretical predictions of the subcritical crack propagation rate were made for different steady state and transient conditions of environment (e.g. degree of aeration of water purity), material (e.g. degree of grain boundary

Stress Corrosion Cracking of Stainless Steels and Nickel Alloys in High-Temperature Water

Abstract Stress corrosion cracking (SCC) studies in stainless steels and nickel alloys reveal that all grades and conditions are susceptible to SCC in high-temperature water, whether deaerated or aerated, high H2 or low, theoretical purity water or buffered/contaminated, lower temperature or higher. However, the kinetics of SCC growth vary enormously with stress intensity, yield strength, sensitization, water chemistry, irradiation, temperature, etc. The role of yield strength is especially important because it changes with surface cold work, bulk cold work, weld shrinkage strain, and irradiation hardening; the role of metallurgical strengthening mechanisms, e.g., nitrogen additions or precipitation hardening, may have a similar effect. SCC growth rate measurements were performed in high-temperature water on unsensitized stainless steels (and Alloy 600 [UNS N06600]) of various grades and compositions. Little effect of grade/heat of stainless steel, martensite content, or H2 fugacity/permeation rate was ob...

Effects of Hydrogen on Stress Corrosion Crack Growth Rate of Nickel Alloys in High-Temperature Water

Abstract Stress corrosion cracking (SCC) of Alloy 600 (UNS N06600) and Alloys 182 (UNS W86182), 132 (UNS W86132), and 82 (UNS N06082) weld metals in high-temperature water is important because thes...

Stress corrosion cracking behavior of alloys in aggressive nuclear reactor core environments

Abstract The effects of irradiation on stress corrosion cracking occur through changes in the water chemistry and in the alloy microstructure. Considerable reactor experience has shown that a high-temperature water environment and a radiation field combine to produce irradiation-assisted stress corrosion cracking (IASCC) in core components of light water reactors. The principal effect of irradiation on water chemistry is through radiolysis, which results in an increase in the corrosion potential through the formation of radiolytic species consisting of radicals and molecules that can be oxidizing or reducing. In addition, profound effects of irradiation on the microchemistry and alloy microstructure create numerous pathways for IGSCC to occur. Radiation-induced segregation, the formation of a dislocation loop microstructure, irradiation hardening, and irradiation creep all occur simultaneously in space and time. Unfolding these various effects to determine the primary factors governing the observed effect...

Emerging Issues and Fundamental Processes in Environmental Cracking in Hot Water

Abstract In the absence of definitive data, mechanistic and modeling approaches to describe environmental cracking in high-temperature water are forced to depend too strongly on speculative hypothe...

Irradiation-assisted stress corrosion cracking

Abstract Irradiation-assisted stress corrosion cracking (IASCC) refers to the process(es) by which irradiation enhances the inherent susceptibility of alloys to stress corrosion cracking (SCC). This article begins with an introduction to the IASCC problem experienced in light water reactors, followed by a summary of existing plant and laboratory data that reveal its dependence on irradiation, environment and alloy parameters. The specific effects of irradiation on IASCC are classified into two categories: the effects on water chemistry and the effects on microstructure. Water chemistry effects include radiolysis and its effects on corrosion potential, and effects of corrosion potential on IASCC. Microstructure effects include radiation- induced segregation (RIS), irradiated microstructure, swelling and creep, and hydrogen and helium generation. Leading mechanisms proposed to explain the role of radiation in the SCC process are: radiolysis and crack tip strain rate, grain boundary chromium depletion, irradiation hardening, localized deformation and RIS of minor elements. As nuclear power plants operate for longer, an increased incidence of SCC can be expected unless active mitigation steps are taken. Drawing on the accumulated understanding over the past four decades of the key processes believed to control IASCC, a set of attributes of an ideal, IASCC-resistant alloy are identified as a first step.

Fundamental modeling of environmental cracking for improved design and lifetime evaluation in BWRs

Abstract An approach for improved design and lifetime evaluation of environmental cracking is presented based on fundamental modeling of the underlying processes operative in crack advance. In outlining this approach and its application in energy industries, the requirements for a life prediction methodology will be highlighted and the shortcomings of the existing design and lifetime evaluation codes will be discussed. Examples will be given of its use in a variety of cracking systems, such as stainless steels and nickel alloys in hot water, and irradiation assisted stress corrosion cracking.

Irradiation-assisted stress-corrosion cracking in austenitic alloys

Irradiation-assisted stress-corrosion cracking (IASCC) is a complicated phenomenon that poses a difficult problem for designers and operators of nuclear plants. Because IASCC accelerates the deterioration of various reactor components, it is imperative that it be understood and modeled to maintain reactor safety. Unfortunately, the costs and dangers of gathering data on radiation effects are high, and the phenomenon itself is so complex that it is difficult to enumerate all of the causes. This article reviews current knowledge of IASCC and describes the goals of ongoing work.

Grain boundary segregation in austenitic stainless steels and its effect on intergranular corrosion and stress corrosion cracking

This paper reports a study of grain boundary segregation, intergranular corrosion, and intergranular stress corrosion cracking in austenitic stainless steels. The results show that phosphorus, nitrogen, and sulfur all segregate to grain boundaries in these materials and that they can affect one anothers segregation through site compctition. In particular, the results demonstrate that phosphorus segregation can be lowered by the presence of nitrogen and sulfur in the steel. Also, if manganese is present in the steel, sulfur segregation will be greatly decreased as a result of formation of manganese sulfides. Phosphorus, sulfur, and nitrogen will not initiate intergranular corrosion in the modified Strauss test, although if corrosion is initiated by chromium depletion, these elements might enhance the corrosion process. Phosphorus segregation does enhance corrosion in the Huey test, even in steels that have not undergone grain boundary chromium depletion, although there does not appear to be a precise correlation between the depth of corrosion penetration and phosphorus segregation. Intergranular stress corrosion cracking in 288 °C water at a pH of 2.5 and electrochemical potential of OVSHE can occur in these steels even in the absence of chromium depletion if sulfur is present on the grain boundaries. Phosphorus segregation appears to have very little effect.