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Iron-chromium-aluminum (FeCrAl) alloys are being considered for the cladding of uranium dioxide fuel in light water reactors (LWRs). FeCrAl alloys have good mechanical properties at temperatures of 300 °C and higher, and have superlative resistance to attack by steam at temperatures of up to 1000 °C and higher. A concern has been raised that the use of FeCrAl for cladding would result in a higher content of tritium in the reactor coolant as compared with the current system where the cladding is a zirconium based alloy. This review shows that the flux of tritium from the fuel rod cavities to the coolant across the fuel cladding wall will be greatly reduced by the presence of oxides on the surface of the cladding. The review of current literature and permeation data show that (a) protective oxides are expected to be present on both sides of the FeCrAl cladding, and (b) depending on the characteristics of these oxide layers it is reasonable to expect about two–three orders of magnitude reduction in tritium permeation, relative to the permeation response in clean, unoxidized condition for FeCrAl steels of interest, around 277 °C–377 °C temperatures.

The effect of alloy composition and oxidation condition on specific mass gain of FeCrAl alloys was studied and analyzed using a combination of experimental and AI approaches. A Neural Network (NN) classification model was used on the experimental FeCrAl dataset produced at GE Research from steam oxidation studies at both high (~ 1000°C) and low temperature (~ 400°C). Furthermore, using the Shapley Additive exPlanations (SHAP) explainable Artificial Intelligence (XAI) tool, we explore how the NN can identify an alloy at specific oxidation condition to form a protective oxide or not. We found high Al and Cr concentration increases the chances of forming protective oxide layer, which is consistent with literature studies. Contrary to Al and Cr, the presence of Mo in FeCrAl creates thick unprotective oxide scale that results in high mass gain per unit area.

The US Department of Energy is working with fuel vendors to develop accident tolerant fuels (ATF) for the current fleet of light water reactors (LWRs). The ATF should be more resilient to loss of coolant accident scenarios and help extending the life of the operating LWRs. One of the proposed ATF concepts is to use iron-chromium-aluminum (FeCrAl) alloys for the cladding of the fuel. A concern in using ferritic FeCrAl is that this type of cladding may result in an increase in the concentration of tritium in the coolant. The objective of the current critical review is to collect and assess information from the literature regarding diffusion or permeation of hydrogen (H) and its isotopes deuterium (D) and Tritium (T) across industrial alloys (including FeCrAl) used or intended for the nuclear industry. Over a hundred years of data reviewed shows that the solubility of hydrogen in ferritic alloys is lower than in austenitic alloys but hydrogen permeates faster through a ferritic material than through austenitic materials. The tritium permeation rates in FeCrAl alloys are between those in austenitic stainless steels and in ferritic FeCr steels. The activation energy for hydrogen permeation is approximately 30 pct higher in the austenitic alloys compared with the ferritic (typically ∼ 50 kJ/mol in ferritic vs. ∼ 65 kJ/mol in the austenitic). None of the major elements in FeCrAl alloys react with hydrogen to form detrimental hydride phases. The effect of surface oxides on FeCrAl delaying hydrogen entrance into FeCrAl alloy is not part of this review.

The international materials community is engaged in finding safer alternatives to zirconium alloys for the cladding of fuel in light water reactors. One solution is to replace the zirconium cladding using ferritic iron-chromium-aluminum (FeCrAl) alloys, which offer extraordinary resistance to high-temperature reaction with air or steam due to the formation of a protective alumina layer on the external surface. It is important to characterize the behavior of FeCrAl not only during accident conditions but in the entire fuel cycle, which may include reprocessing of the used fuel after it is removed from the power reactors. The reprocessing may involve the dissolution of the fuel rods in mineral acids. Little or nothing is known on the dissolution of FeCrAl alloys in common mineral acids, therefore the objective of this research was to study the dissolution of typical cladding tubing having two compositions of FeCrAl (APMT and C26M) in three acids (H2SO4, HNO3, and HCl) as a function of the temperature using both standard ASTM immersion tests as well as electrochemical tests. The dissolution behavior of the FeCrAl alloys is compared to the dissolution capability of other traditional nuclear materials such as austenitic stainless steels (304SS and 316SS) and austenitic nickel alloys (Alloy 600 and Hastelloy C-276). Results show that both C26M and APMT have a higher dissolution capability in the studied mineral acids, which will be beneficial for reprocessing procedures.

Primary challenge of using FeCrAl in high temperature industrial setting is the formation of α’- precipitates that causes brittleness in the alloy, resulting in failure through fracture. The precipitation causes hardness change during thermal aging which is sensitive to both alloy composition and experimental condition (i.e., temperature and time of heat treatment). A Gaussian Process Regression (GPR) model is built on the hardness data collected at GE Research. Subsequently, for the first time, SHapley Additive exPlanations (SHAP) built upon the GPR is used as an Explainable Artificial Intelligence (XAI) tool to understand the effect of feature values in driving the hardness change. Several key insights are affirmed and obtained from a material science perspective.

Traditionally, FeCrAl alloys played an important role in high-temperature applications due to their ability to form a passive Al oxide film at temperatures above ~800°C. Recently, FeCrAl alloys became of interest for the application of accident tolerant nuclear fuel cladding. This study covers work done at GE Research for better understanding the role of Al, Cr, and Mo in oxidation kinetics and thermodynamics. Several models and commercial prototype alloys have been tested in hydrothermal corrosion autoclave loops, at low temperature steam exposure (~400°C), high temperature steam exposure (~1000°C or higher), and high temperature air exposures. The results provide insights on how chromium and aluminum play a significant role in both high temperature and low temperature oxidation of FeCrAl. Additionally, machine learning tools are used to gain further insights on both predicting future optimized chemistries for balancing the properties of hydrothermal corrosion, low and high temperature steam oxidation, and thermal aging (which is exacerbated due to radiation in a nuclear reactor environment). GE plans to use this framework to further optimize the FeCrAl alloy system for use in nuclear reactor environments.

Iron-chromium-aluminum (FeCrAl) alloys are being characterized to be used in light water reactors for the cladding of the fuel. This family of alloys were never used before in a reactor environment so their behavior should be evaluated. This manuscript describes the enhanced mechanical properties and thermal creep resistance of two FeCrAl alloys (PMC26M & FA-SMT or APMT-2) as compared with the current Zircaloy-2 based cladding. At 450°C and under a nominal hoop stress of 138 MPa (20ksi), Zircaloy-2 tubes suffer rapid creep while the two FeCrAl alloys do not show any evidence of creep at a higher nominal hoop stress of 324 MPa (47 ksi). FA-SMT had higher creep resistance than PMC26M at 600°C and 800°C mainly due to precipitate strengthening in FA-SMT.

Iron-chromium-aluminum (FeCrAl) alloys are being characterized to be used in light water reactors for the cladding of the fuel. The use of accident tolerant fuel (ATF) such as FeCrAl alloys would allow for reactors operating safer, more economically, and for an extending life beyond 60 years. It was shown before that the lower or 12%Cr alloy such as C26M may have higher corrosion rate than the higher 21%Cr APMT. This article shows that the addition of a small amount of zinc (Zn) into the water reduces by more than 60% the dissolution rate of the FeCrAl. Zn incorporates into the external layer of the oxide film making the alloy more resistant to dissolution.

Electrochemical studies and microstructure analysis of directionally solidified hypoeutectic and eutectic aluminum-copper alloys were performed. Optical and scanning electron microscopy studies of corroded specimens with columnar and equiaxed microstructures in 0.1 M, 0.5 M, and 1 M NaCl solutions were conducted. Low-rate potential scanning and alternating current (AC) electrode impedance measurements were conducted to study the corrosion resistance of four aluminum-copper alloys. The concentration of Cu in the alloys proved to be a key factor in the corrosion resistance of the Al-Cu alloys, which controlled the fraction of α and θ phases and the morphological distribution of these phases. The addition of Cu provides cathodic sites that increase adjacent anodic activity and higher corrosion susceptibility of the Al-Cu alloys, as compared with pure Al. Arise in the Cu amount that is linked to an increased concentration of the Al2Cu intermetallic or theta phase results in a higher susceptibility to corrosion for the studied alloys. A microstructural morphology related to a decreased area of contact between the α-phase and the Al2Cu intermetallic phase enhances the corrosion resistance of the Al-Cu alloys. For the Al-1wt.%Cu alloy increasing the content of Cl− produces a beneficial result related to a more resistive passive film. For the rest of the studied alloys with nobler corrosion potentials, the increase in Cl− results in a decrease in their corrosion resistance.

The sensitization of stainless steels may decrease their corrosion resistance in industrial applications. Traditional immersion tests exist to determine the degree of sensitization (DOS) of the stainless steels. However, electrochemical methods may be preferred because they are less expensive and faster to perform. The fast and robust double loop electrochemical potentiokinetic reactivation (DL-EPR) test has been introduced to the corrosion community some decades ago but an interlaboratory testing study was necessary to assess the repeatability and reproducibility of the DOS results. This work reports on a recent study where 11 laboratories returned results that show a high degree of confidence in the data obtained by DL-EPR.

The irradiation-induced microstructure and irradiation-assisted stress corrosion cracking (IASCC) behavior of additively manufactured (AM) 316L stainless steels produced by laser powder bed fusion were evaluated for the first time. Irradiation-induced dislocation loops, voids, and γ′ precipitates were observed in all processing conditions following 2.5 dpa at 360 °C. The cell structure and dense dislocation walls in the stress-relieved AM materials recovered and showed signs of recrystallization following irradiation. Anisotropy in both tensile property and IASCC susceptibility were observed in the stress-relieved AM 316L stainless steel due to the printing texture. The hot-isotropic pressed AM 316L had better radiation tolerance and lower IASCC susceptibility than the stress-relieved AM 316L and conventionally forged 316L. Therefore, post-printing hot-isotropic pressing (HIP) is recommended for enhancing radiation tolerance and IASCC performance in nuclear applications as it eliminates the anisotropic mechanical behavior and IASCC susceptibility associated with the printing texture.

The evolution of nano-scale oxide inclusions in 316L austenitic stainless steel (SS) manufactured by laser powder bed fusion (L-PBF) was investigated. The parts were heat treated at different temperature for same time, the size, morphology and chemistry of oxide inclusions were characterized. After heat treatment, oxide inclusions were found be maintain their chemistry. Statistical analysis was done by analyzing backscattered electron (BSE) images, it was found that the inclusion size increases with temperature and inclusion number decreases with temperature which indicates the Ostwald ripening coarsening mechanism behind.

Since 2011, the international nuclear materials community has been engaged in finding replacements for zirconium alloys fuel cladding for light water reactors. Iron-chromium-aluminum (FeCrAl) alloys are cladding candidates because they have high strength at high temperature and an extraordinary resistance to attack by superheated steam in the event of a loss of coolant accident. As FeCrAl alloys have never been used in nuclear reactors, it is important to characterize their behavior in the entire fuel cycle. Stress corrosion cracking (SCC) studies were conducted for two FeCrAl alloys (APMT and C26M) in typical simulated boiling water reactor conditions at 288°C containing either dissolved hydrogen or oxygen. Crack propagation studies showed that both ferritic FeCrAl alloys were resistant to SCC at stress intensities below 40 MPa√m. Current results for FeCrAl confirm previous findings for Fe-Cr alloys showing that ferritic stainless alloys are generally much more resistant to high-temperature water SCC than austenitic stainless steels.

The electricity in the interconnected commercial grid comes from many sources including fossil fuels, renewable energies (e.g., solar and wind), and nuclear energy. The energy harvested from nuclear fission sources is clean, it does not release greenhouse gases to the environment, and it does not contribute to climate change. Moreover, nuclear sourced electricity is the safest type for civilian use. In the United States and around the globe, the number of human deaths per kilowatt hour of generated electricity from nuclear sources is five orders of magnitude lower than for coal and other hydrocarbon sources. Nuclear electricity is even safer than solar or wind electricity. Nuclear power was first used for military purposes, especially for the propulsion of submarines. However, in the early 1950s, the United Nations program of Atoms for Peace, both the United States and the Soviet Union developed their first civilian nuclear power plants which were connected to their respective grids in the late 1950s. At around this time, the trusted fuel rod pair of zirconium alloy cladding containing pellets of urania fuel was born and used for many decades since. Currently, there are approximately 30 nations which use 451 civilian reactors to generate electricity. The largest expansion of nuclear reactors was in the decades of 1970s and 1980s. In 2019 there were 54 new reactors under construction, mainly in Asia.

The third and the last concept for ATF fuel cladding is the use of silicon carbide (SiC) composite materials. The implementation of SiC for cladding is further down the road than the other two most near-term concepts, namely coated zirconium alloys, and FeCrAl alloys (Chapter 4, Accident tolerant fuel cladding concept: coatings for zirconium alloys, and Chapter 5, FeCrAl—iron–chromium–aluminum monolithic alloys). The main reason SiC is considered for nuclear application is because of its high transparency to neutrons, its resistance to attack by air and steam at temperatures as high as 1700°C, and its high mechanical strength at the high temperatures. One of the handicaps for SiC is their high solubility in condensed water at approximately 300°C typical of light water reactor operation conditions. Another challenge is related to the fabrication of the rod to produce a reliable hermeticity to avoid the release of toxic compounds into the coolant. Other issues that may need to get resolved are the need for a consistent production route with controlled stoichiometry of the SiC compound and containing low impurities, which may harm not only their resistance to oxidation but also their resistance to neutron irradiation. Other unknown for nuclear applications is the SiC cladding response to pellet cladding interaction. Since hermeticity and fabrication issues need to be resolved before SiC is used for cladding application, it may be used earlier as channel material for BWR reactors which does not need the hermeticity of the fuel rods. The more near term application for channels will provide familiarity on the use of ceramics in a nuclear environment which could be a facilitation path for later use in fuel rods. For channel boxes, the issue of SiC high dissolution rate in water may be solved by coating it with a thin layer of chromium like for the zirconium cladding.

The implementation of accident tolerant fuel concepts will increase the safety of nuclear power plant operation. This increased safety will occur not only because of the actual performance of the material concepts under normal and accident conditions but also because of the full evaluation of plant operation. This assessment can be accomplished by considering the concept of source term, which is defined as the type and number of radioactive contaminants that can be released for a nuclear power station owing to an accident. The analyses of several areas of improvements by use of ATF concepts in the reactor showed a decrease in the source term when compared with the current fuel designs.

Light water reactors (LWRs) have been generating electricity for over five decades for the electrical grid of more than 20 countries. Most of these reactors are built using a series of alloys and materials that have changed very little over the many decades. The most common structural materials are based on the three elements, iron (Fe), chromium (Cr) and nickel (Ni) such as stainless steels and nickel-based alloys. Most of the Fe-Cr-Ni alloys contain enough Cr to render them passive with low general corrosion in typical LWR environments. Since the chemistry of the water in the reactor is highly controlled, without aggressive impurities such as chloride or sulfate ions, localized corrosion is not a problem for the common structural alloys. The most common failure mode of the Fe-Cr-Ni alloys is environmentally assisted cracking or stress corrosion cracking (SCC), which is currently well understood and successfully mitigated, mainly by dissolving hydrogen gas into the coolant. The other fundamental material in the reactor core is the zirconium-based alloy used for cladding of the fuel rods. The traditional fuel rod alloys were Zircaloy-2 and Zircaloy-4, which were used regularly in boiling water reactor and pressurized water reactor systems, respectively. Very early in the development of the alloys (early 1950s), it was understood that a small amount (

The use of monolithic FeCrAl alloys to replace zirconium alloys is the second nearest term for the accident-tolerant fuel claddings after coated zirconium alloys. Currently a single layer of less than 1-mm-thick zirconium alloy metal separates the fuel cavity and its toxic compounds from the coolant water. The simplest and most elegant improvement would be to replace the current alloy for another alloy, obviously as resistant to corrosion as zirconium under normal operation conditions but a thousand times more resistant to attack by steam under severe accident conditions. An alloy down selection was conducted in steam under severe accident scenarios (>1200°C) and the highest environmental accident resistance was always offered by FeCrAl alloys because of the formation of an adherent, thin, and shielding film of alumina, which protected the FeCrAl from steam attack until its melting point. FeCrAl are ferritic materials that have a low corrosion rate under normal operation conditions by the formation of a protective chromium oxide on the surface. Ferritic FeCrAl are also resistant to environmental cracking from the coolant side. FeCrAl alloys can be made into full-length fuel rod tubes with a wall thickness of less than 0.5 mm and they can be industrially welded to hermetically seal the fuel cavity by a solid-state method. The thermal neutron absorption cross section of FeCrAl is approximately ten times higher than that of the current zirconium alloys, but this fact can be partially mitigated by making the cladding wall thickness half of the current values for Zircaloy owing to the stronger mechanical properties at temperature of the FeCrAl alloys. The use of FeCrAl cladding may initially increase the tritium release into the coolant but is anticipated that the oxidation of the cladding from the fuel cavity side and the coolant side will be effective barriers for hydrogen diffusion across the tube wall. The use of FeCrAl cladding will eliminate the occurrence of debris fretting and shadow corrosion, two current Zircaloy cladding environmental degradation modes. FeCrAl tube cladding concepts are currently under irradiation both at the advanced test reactor in Idaho National Laboratory and in two BWR commercial nuclear power stations in the United States. Even though FeCrAl materials were never used in light water reactors before, their characterization in reactor anticipated environments is advancing rapidly through international development efforts.

Until a few years ago, very few considered the possibility of using fuel systems in light water reactor that were not based on a zirconium alloy and uranium dioxide (UO2). All standards and guidelines developed over the last 60 years in the commercial nuclear power business are based on Zr/UO2. Any improvement over the decades was to understand and solve issues related to the performance of Zr/UO2. Now, there is a possibility for newer materials in power reactors, and regulatory bodies and trade organizations may be accepting this change as an unavoidable reality. The newer accident tolerant fuels ATF can be classified broadly into (1) newer claddings and (2) newer fuels. Each one of the cladding and fuels main groups has subgroups. In the cladding, subgroups are (1) coated zirconium alloys, (2) monolithic FeCrAl alloys, and (3) ceramic composites or SiC/SiC made of SiC fibers embedded in an SiC matrix. The subgroups in the fuels are (1) modified or doped UO2 and (2) high-density fuels including uranium nitride and uranium silicide. The most near-term concepts for implementation may include coated zircaloy cladding and modified urania fuel. The second closest implementation could be FeCrAl cladding for urania fuel. The ceramic composite SiC/SiC for cladding and the higher density fuels UN and U3Si2 are concepts that may need longer term developments. In the United States (and in many other countries), the entire nuclear power industry is prepared for the changes. In the first few years of development of the ATF concepts, the regulatory agency and the commercial side of nuclear power were observant bystanders. However, in the last few years, they became fully engaged. Currently, utilities are working closely with the fuel producers to have one or more concepts installed in their commercial reactors within the next quinquennium. The supportive organizations for the utilities such as the Electric Power Research Institute and the Nuclear Energy Institute are dedicating technologists to evaluate the benefits offered by the newer ATF concepts. The regulatory body (Nuclear Regulatory Commission) is investing in personnel to work in parallel with the US Department of Energy and the fuel vendors to allow a full reload into a reactor as soon as technically possible. The funding at federal level in the United States is robust for universities, national laboratories, fuel vendors, commercial utilities, and trade organizations to be working collectively with the main goal of making nuclear power plants safer to operate and to continue having the option of using nuclear power to supplement other clean energy sources and to minimize the indiscriminate burning of fossil fuels in order to generate electricity. It is always rewarding to have variety of technological and engineering organizations working with the same goal of making nuclear power stations more robust and safer to operate. Different technologists with diverse backgrounds would look at and evaluate the same issue of increased safety in reactor operation from a different perspective. The Fukushima accident was a critical unfortunate event for nuclear power generation; however, it eventually galvanized the industry to offer solutions for the continuing use of clean nuclear energy.

This chapter presents an educated, yet subjective maturity evaluation of the five accident-tolerant fuel (ATF) concepts discussed in Chapters 4–7. The five evaluated fuel concepts include three cladding concepts (chromium-coated zirconium alloy, FeCrAl alloys, and silicon carbide composites) and two fuel concepts including modified urania fuel and high-density fuels. The rating of the maturity level of the concepts has been conducted using metrics from technical readiness level principles by a panel of international nuclear materials experts, mainly from the nuclear energy agency. Not surprisingly the highest maturity level corresponded to the two evolutionary concepts which may have a better chance to obtain near-term licensing from the regulatory entities. These two nearest term concepts include chromium-coated zirconium alloy for cladding and chromia-doped urania for fuel. The second nearest term cladding concept was the monolithic iron–chromium–aluminum FeCrAl alloy. The other two concepts, namely silicon carbide for cladding and higher uranium density for fuel, have a much longer term possibility of implementation in a commercial power reactor owing to the many technical issues that still need to be solved. A positive twist is that a ceramic silicon carbide design can be first implemented for channel boxes in BWRs as a near- to mid-term ATF concept. This will allow for data gathering in an actual reactor environment, which may accelerate the subsequent implementation of SiC/SiC CMC as cladding for the fuel.

The fueling of the power reactors happens at relatively long intervals of up to 2 years, which is a convenient way of reducing operation costs by cutting down on fueling shut down periods. To be able to accommodate long runs, the fuels must comply with strict characteristics and properties. Uranium dioxide (UO2) or urania has been used reliably in light water reactors for several decades; therefore newer proposed fuels must meet a high standard to be seriously considered. The nearest term accident tolerant fuel (ATF) is a slight modification of the current urania by doping it with chromia to increase its thermal conductivity and to make it more resistant to crumbling at high irradiation doses. Like the case of the chromium-coated zirconium alloys for cladding, the doped urania fuel may have the easiest and fastest path for regulatory licensing approval. The second family of ATFs considered is called the high uranium density fuel because due to their specific stoichiometry these compounds contain more atoms of uranium per unit mass of fuel than urania. The two most advanced fuels in the high-density family are uranium silicide and uranium nitride. The largest handicap of the two high-density fuels is their high reactivity with water and steam in the case of a cladding breach. Not much is known of their response to an actual nuclear reactor environment. The uranium silicide fuel is currently under irradiation in the advanced test reactor and in a commercial reactor as part of a fuel vendor evaluation.

Understanding the performance of cladding materials during a loss-of-coolant accident scenario is crucial for developing next-generation accident tolerant nuclear fuel rod claddings. We describe a multi-modal approach combining synchrotron-based diffraction, and Raman spectroscopy with nano-scale electron microscopy techniques for investigating oxidation of Inconel 600 (A600) in steam and air environments at 1200°C for 2 h. We report that A600 exposed to steam develops a Cr2O3-enriched surface layer while a mixed phase oxide layer containing NiFe2O4, Cr2O3, Fe3O4 and Fe(3–x)CrxO4 is formed in air. Mechanism of oxidation for A600 in air and steam environments is discussed.

Iron-chromium-aluminum (FeCrAl) alloys are being investigated as cladding material for urania nuclear fuel in light water power reactors. One extraordinary attribute of the FeCrAl alloys is their resistance to attack by air and steam up to their melting point. It was of interest to study the kinetics of oxidation of FeCrAl from 800°C to 1300°C in air and in steam to determine how the oxides evolve to provide protection to the alloys. The two investigated alloys were APMT (Fe-21Cr-5Al-3Mo) and C26M (Fe-12Cr-6Al-2Mo). Results show that both alloys had similar oxidation kinetics despite their different chemical compositions. For the testing times, the oxidation rate was higher in air than in steam at the higher temperatures (1100–1300°C) and higher in steam than in air at the lower temperatures (800–1100°C). In the lower temperature range, the surface oxide consisted of two layers, an internal layer rich in aluminum and an external layer containing Al, Cr, and Fe. In the higher temperature range, the oxide was a single layer of alumina (no Cr, no Fe, no Mo).

Intergranular and intragranular Si and Mn rich oxide inclusions are present in laser additive manufactured austenitic stainless steel. The uniform oxide dispersions in additive manufactured material promoted early initiation of microvoids and reduced its impact toughness relative to powder metallurgy (hot isostatic pressing) and wrought materials. For stress corrosion cracking in high temperature water, the silica inclusions along the grain boundaries preferentially dissolved and appeared to accelerate oxidation and caused extensive crack branching.

FeCrAl alloys are candidate materials for manufacturing accident-tolerant fuel (ATF) cladding intended for light water reactors to increase fuel reliability and safety during design-basis and beyond-design-basis accident scenarios. The evolution of the materials' surface characteristics, microstructure, and mechanical properties when exposed to the Critical Heat Flux (CHF) in flow boiling testing is crucial for safety analysis while providing insights into their thermal-hydraulic performance in nuclear reactors. After CHF, the surface chemistry of two FeCrAl alloys, APMT and C26M, was studied to understand their evolution at the early stage of high-temperature excursions in short time periods. The results indicated a thin layer composed of oxides and hydroxides of Al, Cr, and Fe with varying proportions at different depths in the layer, as indicated by X-ray photoelectron spectroscopy (XPS) and depth profiling. The cross-sections prepared by focused ion beam (FIB) revealed the growth of an oxide layer, in the range of 90-180 nm thick, on the alloys' surfaces. The evolution of the materials' surface chemistry also led to a noticeable post CHF excursion increase in their wettability, with a slight increase in roughness. The investigation of the materials' mechanical properties indicated a modest increase in hardness by 10-15% as well as an increase in their yield strength, as evidenced by the microindentation and ring compression tests conducted before and after CHF testing. Scanning electron Microscopy (SEM) and X-ray diffraction (XRD) were used to investigate microstructural features of the materials and their changes after CHF treatment.

The US has currently a fleet of 99 nuclear power light water reactors which generate approximately 20% of the electricity consumed in the country. Near 90% of the reactors are at least 30 years old. There are incentives to make the existing reactors safer by using accident tolerant fuels (ATF). Compared to the standard UO2–zirconium-based system, ATF need to tolerate loss of active cooling in the core for a considerably longer time while maintaining or improving the fuel performance during normal operation conditions. Ferritic iron-chromium-aluminum (FeCrAl) alloys have been identified as an alternative to replace current zirconium alloys. They contain Fe (base) + 10–22 Cr + 4–6 Al and may contain smaller amounts of other elements such as molybdenum and traces of others. FeCrAl alloys offer outstanding resistance to attack by superheated steam by developing an alumina oxide on the surface in case of a loss of coolant accident like at Fukushima. FeCrAl alloys also perform well under normal operation conditions both in boiling water reactors and pressurized water reactors because they are protected by a thin oxide rich in chromium. Under normal operation condition, the key element is Cr and under accident conditions it is Al.

Alloys of iron-chromium-nickel are being explored as alternative cladding materials to improve safety margins under severe accident conditions. Our research focuses on non-destructively investigating the oxidation behavior of the FeCrNi alloy “Alloy 33” using synchrotron-based methods. The evolution and structure of oxide layer formed in steam environments were characterized using X-ray diffraction, hard X-ray photoelectron spectroscopy, X-ray fluorescence methods and scanning electron microscopy. Our results demonstrate that a compact and continuous oxide scale was formed consisting of two layers, chromium oxide and spinel phase (FeCr2O4) oxides, wherein the concentration of the FeCr2O4 phase decreased from the surface to the bulk-oxide interface.

A novel idea to create new alloys for Additive manufacturing (AM) by mixing a small addition of nanoparticles with bulk material was put forward. Integrated Computational Material Engineering (ICME) may be used to guide the AM process, predict thermal behaviour, optimize process parameters, secure net-shape, and qualify parts. Modelling is needed to predict the effect of defects and the effect of inclusions on mechanical properties, build quality and in-service performance. ICME can reduce trial-and-error fabrication and establish an AM digital twin, thereby accelerating part qualification and AM adoption. AM creates complex thermal processes which impact material microstructure and result in changes to mechanical properties in terms of strength and plasticity. In this study, Grain-Boundary Engineering (GBE) and multi-scale modelling are performed to expedite qualification of new and existing AM polycrystalline alloys. AM parts exhibit cracks, low toughness, low plasticity, and high residual stresses. To mitigate these characteristics, a methodology for GBE and microstructural control was developed. Depending on part requirements a low angle, coincident site lattice, or high angle grain boundary (LAGB,CSL, HAGB) might be the preferred option. For example, HAGB occurs in equiaxed grain microstructure, prevents precipitation and crystallization formation, and inhibits cracks and corrosion. Some polycrystalline materials may be non-weldable and exhibit cracks. Microstructural control can be used to address this phenomenon. AM offers unique advantages over casting, including material composition change and unique thermal processing. By utilizing nano-inclusions, cracks may be mitigated and AM parts improved. ICME physics-based approach is used to implement GBE modelling to improve the microstructure of an AM polycrystalline material from Equiaxed/HAGB to LAGB and CSL considering nano-scale inclusions. The ICME tool integrates meltpool engineering; thermal transport; nano-micro-mechanics; material characterization; calibration and qualification; Voronoi-tessellation, finite element modelling, and design of experiment optimization. With this background, ICME predicts time-temperature transformation, thermal properties, process map, meltpool size, precipitation, porosity, and effects of nano inclusions. In this manner, ICME helps determine ideal AM machine parameters that secure qualified parts as a consequence of the AM process. In this study, mechanical properties and fracture toughness were validated against test. Predictions included creep, crack growth, and the effect of precipitates and voids under in-service crack growth behaviour. Stainless steel (SS316L) dog bones specimens were printed and stress relieved by air and water cooling supported by fractography and imaging. The objective of this effort was to develop technologies and processes in AM design and engineering that could predetermine the microstructure of AM parts. Future efforts will employ the tool to tailor grain boundaries to produce predictable mechanical properties - including mode of failure.

Corrosion fatigue crack growth of laser additively-manufactured 316L stainless steel was studied in oxygenated 288 °C water. Effects of heat treatment, crack orientation, and applied cold work were evaluated. Stress-relieved material showed higher crack growth rate than annealed material. Their difference was not significant when the loading frequency was high. When the loading frequency decreased, the effects of material conditions created by heat treatment, crack orientation and applied cold work slightly increased. Retained unrecrystallized grains in an annealed material did not affect significantly the cracking susceptibility. High-temperature recrystallization treatment of an AM 316L component may be necessary for nuclear applications.

To avoid accidents like that in Fukushima, the US Department of Energy is engaged with a nuclear fuel vendor to evaluate the performance of iron-chromium-aluminum (FeCrAl) alloys such as advanced powder metallurgy tubing (APMT) as accident-tolerant material for uranium dioxide fuel cladding in light water reactors (LWR). It was important to characterize the oxides formed on APMT under both boiling water reactor (BWR) and pressurized water reactor (PWR) environments for a better understanding of its environmental sustainability in LWRs. Coupons of APMT were exposed for 1 year to both hydrogenated and oxygenated high-purity water at 288°C (e.g. simulated BWR water chemistry without Pt injection) and hydrogenated high-purity water at 330°C (e.g. simulated primary PWR water chemistry without Li/B addition). Results show that after 1-year immersion, APMT always developed a chromium-rich protective oxide film on its surface. In oxygen-containing environments, the oxide consisted of a dual layer, an external thicker layer containing mostly iron oxides and a thinner internal layer rich in chromium oxide. In hydrogen environments, only a single oxide layer formed, consisting of chromium oxide. This is a similar finding as for type 304 and 316 stainless steels and for nickel-based alloy 600, which is extensively reported in the literature. General corrosion of APMT alloys under LWR operating conditions would not be a limiting factor for its performance as cladding material.

Nickel alloy 690 (UNS N06690) is one of the current choices for nuclear power plant steam generator tubing. The severity of certain stress corrosion cracking submodes in Ni-Cr-Fe alloys, such as Alloy 690, may be a function of the alloy sensitization condition. Sensitization results from chromium carbide precipitation at or near grain boundaries when the alloy is exposed to temperatures from 500°C to 800°C, which causes local chromium depletion in the adjacent matrix. The objective of this work was to study the parameters that control the double loop electrochemical potentiokinetic reactivation (DL-EPR) method and to develop an optimized test routine for Alloy 690. Alloy 690 specimens were tested under different thermal treatments, tailored to obtain a variety of concentrations of chromium in solid solution near the grain boundary (GB) region. The DL-EPR method was applied to Alloy 690 in solutions with different concentrations of sulfuric acid and potassium thiocyanate, at 30°C and 50°C. The optimal condition for detecting sensitization of Alloy 690 was 0.5 M H2SO4 + 0.001 M KSCN, at 30°C. Optical micrographs confirmed an intergranular type of attack under thermally aged (sensitized) conditions. The ratio of the reactivation to the activation peak intensities (Ir/Ia) obtained in the DL-EPR test agreed well with mass loss obtained in boiling nitric acid solution with Cr(VI) additions (modified Huey test). Chromium depletion profiles near the GB of thermally treated specimens under different conditions were measured with the energy dispersive x-ray spectroscopy technique, performed on a transmission electron microscope. A model available in the literature was fitted to those results. The Ir/Ia ratio for Alloy 690, obtained when using the optimal DL-EPR testing conditions, correlated well with the Cr depletion zone width and depth near the GB.

Nickel alloys are broadly used in upstream oil and gas applications as a result of an advantageous set of properties including high mechanical strength accompanied by good ductility and toughness, ...