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Commercial nuclear power plants in the United States are light water reactors (LWRs) that use water as a coolant, with temperatures ranging between 280°C and 320°C. Water purity is tightly controlled; nevertheless, the high temperatures expose components in the water circuit to degradation by corrosion by many different types of environmental attack. Uniform corrosion occurs across the entire surface of a material and is prevalent in engineering systems to some extent. Site-specific corrosion processes, such as crevice corrosion, intergranular attack, or galvanic corrosion, are common in complex engineering systems that consist of multiple materials joined by welds or other solid state joining processes. Even within a single component, if second phase strengthening is used or if surface defects are present, localized corrosion can occur.

Proton (H+) irradiation effects in polycrystalline UO2 have been studied. The irradiation was carried out using three ion energies and two different ion fluxes at 600 °C. Scanning electron microscopy (SEM) investigations showed that significant surface flaking took place. Focused ion beam (FIB) milling in SEM was successfully applied for extracting lamellas from uneven blistered surfaces for transmission electron microscopy (TEM) investigations allowing detailed investigations for the degradation mechanisms. High-resolution TEM for the flaked UO2 surfaces revealed that the implanted H+ formed sharp two-dimensional cavities at the peak ion-stopping region instead of diffusing to the matrix. The resulting lateral stress likely caused UO2 surface deterioration in good agreement with previous blistering and flaking studies on crystalline materials. Graphical abstract

A key strategy to increase the radiation resistance of materials has been to introduce a high density of interfaces that can act as sinks for radiation-induced defects. Twin boundaries are a type of interface that can be introduced through deformation but are usually considered to be ineffective sinks. Using heavy ion irradiation and transmission electron microscopy, this study investigates the influence of a high area per unit volume of twin boundaries on the radiation-induced swelling response of an austenitic stainless steel. The study shows that swelling can be suppressed in regions containing a high density of closely-spaced deformation twin boundaries.

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems will require compact heat exchangers (CHXs) for the next generation of nuclear reactors. The DOE is sponsoring research to support the development and deployment of CHXs for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) that includes university research institutes, national laboratories, manufacturers, and industry experts. The objective is to enable the use of CHX designs in advanced reactor service. A necessary step for achieving this objective is to ensure that the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, Division 5 has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a CHX in an advanced reactor. A strategy for ASME Boiler and Pressure Vessel Code, Section XI, Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will also be required. The purpose of this ASME Code Roadmap is to identify the research gaps impeding the development of suitable construction and ISI rules for CHXs for high temperature reactor service and to provide a framework to utilize the research project results consistent with the expectations and needs of the industry and future owners.

Transmission electron microscopy observation of Kr bubble evolution in polycrystalline UO 2 annealed at high temperature was conducted in order to understand the inert gas behavior in oxide nuclear fuel. The average diameter of intragranular bubbles increased gradually from 0.8 nm in as-irradiated sample at room temperature to 2.6 nm at 1600 °C and the bubble size distribution changed from a uniform distribution to a bimodal distribution above 1300 °C. The size of intergranular bubbles increased more rapidly than intragranular ones and bubble denuded zones near grain boundaries formed in all the annealed samples. It was found that high-angle grain boundaries held bigger bubbles than low-angle grain boundaries. Complementary atomistic modeling was conducted to interpret the effects of grain boundary character on the Kr segregation. The area density of strong segregation sites in the high-angle grain boundaries is much higher than that in the low angle grain boundaries.

A cluster dynamics model describing the formation of vacancy and interstitial clusters in irradiated oxides has been developed. The model, which tracks the composition of the oxide matrix and the defect clusters, was applied to the early stage formation of voids and dislocation loops in UO2, and the effects of irradiation temperature and dose rate on the evolution of their densities and composition was investigated. The results show that Frenkel defects dominate the nucleation process in irradiated UO2. The results also show that oxygen vacancies drive vacancy clustering while the migration energy of uranium vacancies is a rate-limiting factor for the nucleation and growth of voids. In a stoichiometric UO2 under irradiation, off-stoichiometric vacancy clusters exist with a higher concentration of hyper-stoichiometric clusters. Similarly, off-stoichiometric interstitial clusters form with a higher concentration of hyper-stoichiometric clusters. The UO2 matrix was found to be hyper-stoichiometric due to the accumulation of uranium vacancies.

Corrosion of structural components in nuclear power plants is a major issue impacting plant capacity factor and operational costs, and it will continue to grow in importance both in light-water reactors (LWRs) receiving life extensions and in next-generation systems. This chapter reviews the corrosion processes, mechanisms, and consequences of the coolants currently in use and expected in GenIV reactors. These include LWR coolants; primary and secondary water systems in pressurized water reactors, and normal and hydrogen water chemistries in boiling water reactors, as well as supercritical water in the supercritical water reactor. Corrosion, carburization, and decarburization due to impurities in high-temperature helium are covered as well as corrosion in molten salt and liquid metals including sodium and lead alloys. Together, these coolants cover the majority of current and next-generation environments in which structural materials must function.

The environmental degradation of intermediate heat exchanger (IHX) materials in impure helium has been identified as an area with major ramifications on the design of very high-temperature reactors (VHTR). It has been reported that in some helium environments, non-ductile failure is a significant failure mode for Alloy 617 with long-term elevated-temperature service. Non-ductile failure of intermediate exchangers can result in catastrophic consequences; unfortunately, the knowledge of creep crack initiation and creep crack growth (CCG) in candidate alloys is limited. Current codes and code cases for the candidate alloys do not provide specific guidelines for effects of impure helium on the high-temperature behavior. The work reported here explores creep crack growth characterization of Alloy 617 and Alloy 800H at elevated temperatures in air and in impure helium environments, providing information on the reliability of these alloys in VHTR for long-term service. Alloy 617 was found to exhibit superior CCG resistance compared to Alloy 800H. For Alloy 617 tested at 973 K (700 °C), a notable increase in the resistance to crack growth was measured in air compared to that measured in the helium environment; CCG results for Alloy 800H suggest that air and helium environments produce similar behavior. Testing of grain boundary-engineered (GBE) Alloy 617 samples revealed that, although the technique produces superior mechanical properties in many respects, the GBE samples exhibited inferior resistance to creep crack growth compared to the other Alloy 617 samples tested under similar conditions. Grain size is noted as a confounding factor in creep crack growth resistance.

Tristructural isotropic (TRISO) coated particle fuel is a promising fuel form for advanced reactor concepts such as high temperature gas-cooled reactors (HTGR) and is being developed domestically under the US Department of Energy's Nuclear Reactor Technologies Initiative in support of Advanced Reactor Technologies. The fuel development and qualification plan includes a series of fuel irradiations to demonstrate fuel performance from the laboratory to commercial scale. The first irradiation campaign, AGR-1, included four separate TRISO fuel variants composed of multiple, laboratory-scale coater batches. The second irradiation campaign, AGR-2, included TRISO fuel particles fabricated by BWX Technologies with a larger coater representative of an industrial-scale system. The SiC layers of as-fabricated particles from the AGR-1 and AGR-2 irradiation campaigns have been investigated by electron backscatter diffraction (EBSD) to provide key information about the microstructural features relevant to fuel performance. The results of a comprehensive study of multiple particles from all constituent batches are reported. The observations indicate that there were microstructural differences between variants and among constituent batches in a single variant. Insights on the influence of microstructure on the effective diffusivity of key fission products in the SiC layer are also discussed.

Recently, the diffusion coefficient of silver (Ag) in zirconium carbide (ZrC) at 1500 °C was determined to help elucidate the diffusive behavior of radioactive 110m Ag in the ZrC coating of the tri-structural isotropic (TRISO) nuclear fuel particle. This study was conducted using a recently developed diffusion experimental technique, known as the encapsulating source method. To verify the validity of this experimental method, modifications to the original experimental design have been made within the domain of this method. Primarily, the ability to retrieve the ZrC disk made it possible to create depth profiles using secondary ion mass spectrometry (SIMS) for more precise depth profiling of Ag concentration in ZrC. Applying this improved experimental design, a sample was fabricated and diffusion annealed at 1500 °C for 2 h. Finally, silver concentrations on the ZrC disk were measured using SIMS. Consequently, the diffusion coefficient of Ag in ZrC 0.95 at 1500 °C was experimentally determined to be about 1.7 × 10 −17 m 2 /s, remarkably consistent with the previously reported value of 2.8 × 10 −17 m 2 /s, also obtained by means of the encapsulating source method. This proves that the encapsulating source method is a reliable experimental technique to determine the Ag diffusivity in ZrC.

Zirconium carbide and zirconium nitride are candidate materials for new fuel applications due to several favorable physicochemical properties. ZrC and ZrN samples were irradiated at the Advanced Test Reactor National Scientific User Facility with neutrons at 800 °C to a dose of 1 dpa. Structural examinations have been made of the ZrC samples using high resolution transmission electron microscopy, and the findings compared with a previous study of ZrC irradiated with protons at 800 °C. The use of X-ray absorption fine structure spectroscopy (XAFS) to characterize the radiation damage was also explored including a model based on spherical crystallites that can be used to relate EXAFS measurements to microscopy observations. A loss of coordination at more distant coordination shells was observed for both ZrC and ZrN, and a model using small spherical crystallites suggested this technique can be used to study dislocation densities in future studies of irradiated materials.

The tri-structural isotropic (TRISO) coated particle fuel has been developed and used for high temperature gas-cooled reactors (HTGRs). It provides a unique robustness of the first barrier for the fission products. The TRISO fuel particle has typically consisted of a UO2 or UCO kernel, surrounded by successive layers of porous carbon, dense inner pyrocarbon, silicon carbide, and dense outer pyrocarbon. During operation, however, the SiC layer has been known to release radioactive silver 110mAg which makes maintenance more difficult and thus costly. Zirconium carbide has been considered as a promising alternative to the SiC fission product barrier. ZrC exhibits high temperature stability and possibly possesses superior Pd resistance, while the retention properties especially for silver have not been adequately studied. To help elucidate the diffusive behavior of silver in the ZrC coating of the TRISO-coated particle, a new diffusion experimental technique, called the encapsulating source method, has been developed by constructing a constant source diffusion couple between ZrC and Ag gas originated from Zr–Ag solid solution. Scanning electron microscopy (SEM), wavelength-dispersive X-ray spectroscopy (WDS), electron backscatter diffraction (EBSD) and optical methods were used to analyze the diffusion couple annealed at 1500 °C. The resultant diffusion coefficient of Ag in single-crystalline ZrC0.84 at 1500 °C was experimentally determined to be about 2.8 (±1.2) × 10−17 m2/s.

ASTM G67 is the standard test method for determining the susceptibility to intergranular corrosion (IGC) of 5xxx series aluminum alloys and requires measuring mass loss after exposure to nitric acid (the nitric acid mass loss test, NAMLT). This mass loss, and thus the IGC susceptibility, is dependent on orientation of the wrought product, but the quantitative degree of this dependence has not been well studied or documented. The orientation dependence of IGC susceptibility for AA5083 (UNS A95083) in H131 and H116 tempers is quantitatively presented herein. The underlying mechanism behind this dependence is also explored.

Accurate online condition monitoring and anomaly detection are crucial in nuclear applications to optimize economic performance and minimize safety risks. To achieve this goal, several major challenges exist which must be addressed. First, multi-sensor signals are often collected in the form of complex, multivariate time series. Second, relatively few anomaly records are available to train detection models. Lastly, the recorded data may contain uncertainties resulting from various sources, such as operator-induced variability and measurement error. In this paper, a recurrent neural network-based approach is proposed to tackle these issues by effectively utilizing historical data obtained during both normal and abnormal operations. Several advanced data preprocessing techniques are developed to improve the training process of the proposed neural network. The efficiency and sensitivity of the proposed method are evaluated on the multi-sensor signal measurements and operational reports obtained from a real case study. The results demonstrate much improved detection accuracy and practicality of the proposed method over conventional approaches.

While thin reactor structural components such as cladding and ducts do not experience significant gradients in dpa rate, gamma heating rate, temperature or stress, thick components can develop strong local variations in void swelling and irradiation creep in response to gradients in these variables. In this study we conducted microstructural investigations by transmission electron microscopy of two 52 mm thick 304-type stainless steel hex-blocks irradiated for 12 years in the EBR-II reactor with accumulated doses ranging from ∼0.4 to 33 dpa. Spatial variations in the populations of voids, precipitates, Frank loops and dislocation lines have been determined for 304 stainless steel sections exposed to different temperatures, different dpa levels and at different dpa rates, demonstrating the existence of spatial gradients in the resulting void swelling. The microstructural measurements compare very well with complementary density change measurements regarding void swelling gradients in the 304 stainless steel hex-block components. The TEM studies revealed that the original cold-worked-state microstructure of the unirradiated blocks was completely erased by irradiation, replaced by high densities of interstitial Frank loops, voids and carbide precipitates at both the lowest and highest doses. At large dose levels the amount of volumetric void swelling correlated directly with the gamma heating gradient-related temperature increase (e.g. for 28 dpa, ∼2% swelling at 418 °C and ∼2.9% swelling at 448 °C). Under approximately iso-thermal local conditions, volumetric void swelling was found to increase with dose level (e.g. ∼0.2% swelling at 0.4 dpa, ∼0.5% swelling at 4 dpa and ∼2% swelling at 28 dpa). Carbide precipitate formation levels were found to be relatively independent of both dpa level and temperature and induced a measurable densification. Void swelling was dominant at the higher dose levels and caused measurable decreases in density. Void swelling at the lowest doses was larger than might be expected based on the dpa level, an observation in agreement with earlier studies showing that the onset of void swelling is accelerated by decreasing dpa rates.

Oxide materials are used in numerous applications such as thermal barrier coatings, nuclear fuels, and electrical conductors and sensors, all applications where nanometer-scale stoichiometric changes can affect functional properties. Atom probe tomography can be used to characterize the precise chemical distribution of individual species and spatially quantify the oxygen to metal ratio at the nanometer scale. However, atom probe analysis of oxides can be accompanied by measurement artifacts caused by laser-material interactions. In this investigation, two technologically relevant oxide materials with the same crystal structure and an anion to cation ratio of 2.00, pure cerium oxide (CeO2) and uranium oxide (UO2) are studied. It was determined that electronic structure, optical properties, heat transfer properties, and oxide stability strongly affect their evaporation behavior, thus altering their measured stoichiometry, with thermal conductance and thermodynamic stability being strong factors.

The nature and magnitude of Ag diffusion in SiC has been a topic of interest in connection with the performance of tristructural isotropic (TRISO) coated particle fuel for high temperature gas-cooled nuclear reactors. Ion implantation diffusion couples have been revisited to continue developing a more complete understanding of Ag fission product diffusion in SiC. Ion implantation diffusion couples fabricated from single crystal 4H-SiC and polycrystalline 3C-SiC substrates and exposed to 1500–1625 °C, were investigated by transmission electron microscopy and secondary ion mass spectrometry (SIMS). The high dynamic range of SIMS allowed for multiple diffusion regimes to be investigated, including enhanced diffusion by implantation-induced defects and grain boundary (GB) diffusion in undamaged SiC. Estimated diffusion coefficients suggest GB diffusion in bulk SiC does not properly describe the release observed from TRISO fuel.

Designing materials for performance in high-radiation fields can be accelerated through a carefully chosen combination of advanced multiscale modeling paired with appropriate experimental validation. The studies reported in this work, the combined efforts of six universities working together as the Consortium on Cladding and Structural Materials, use that approach to focus on improving the scientific basis for the response of ferritic–martensitic steels to irradiation. A combination of modern modeling techniques with controlled experimentation has specifically focused on improving the understanding of radiation-induced segregation, precipitate formation and growth under radiation, the stability of oxide nanoclusters, and the development of dislocation networks under radiation. Experimental studies use both model and commercial alloys, irradiated with both ion beams and neutrons. Transmission electron microscopy and atom probe are combined with both first-principles and rate theory approaches to advance the understanding of ferritic–martensitic steels.

Swift heavy ion induced radiation damage is investigated for ceria (CeO2), which serves as a UO2 fuel surrogate. Microstructural changes resulting from an irradiation with 940 MeV gold ions of 42 keV/nm electronic energy loss are investigated by means of electron microscopy accompanied by electron energy loss spectroscopy showing that there exists a small density reduction in the ion track core. While chemical changes in the ion track are not precluded, evidence of them was not observed. Classical molecular dynamics simulations of thermal spikes in CeO2 with an energy deposition of 12 and 36 keV/nm show damage consisting of isolated point defects at 12 keV/nm, and defect clusters at 36 keV/nm, with no amorphization at either energy. Inferences are drawn from modeling about density changes in the ion track and the formation of interstitial loops that shed light on features observed by electron microscopy of swift heavy ion irradiated ceria.

A systematic x-ray diffraction (XRD) study was performed on room-temperature Xe-irradiated and postirradiation annealed CeO2. Large scale XRD did not show any additional irradiation-induced phases upon irradiation. Depth profiling the CeO2 (111) diffraction peak over the 150 nm deep Xe-irradiated layer (400 keV, 1 × 1020 Xe/m2) by grazing incidence XRD indicated a lattice expansion at the irradiated layer. Postirradiation annealing (1 h at 1000 °C) in an oxygen-containing environment removed the observed XRD features. Electron energy loss spectroscopy (EELS) was performed for cross-sectional samples before and after postirradiation annealing. EELS showed that the Ce charge state changed from +4 to +3 at the CeO2 surface indicating the presence of O vacancies in both as-irradiated and annealed samples. EELS also indicated that the amount of O vacancies was reduced at the irradiated region by annealing. The experimental results are discussed based on electronic properties of CeO2, annihilation of oxygen vacancies, and evolution of irradiation damage.

In situ and ex situ transmission electron microscopy observation of small Kr bubbles in both single-crystal and polycrystalline UO2 were conducted to understand the inert gas bubble behavior in oxide nuclear fuel. The bubble size and volume swelling are shown as weak functions of ion dose but strongly depend on the temperature. The Kr bubble formation at room temperature was observed for the first time. The depth profiles of implanted Kr determined by atom probe tomography are in good agreement with the calculated profiles by SRIM, but the measured concentration of Kr is about 1/4 of the calculated concentration. This difference is mainly due to low solubility of Kr in UO2 matrix and high release of Kr from sample surface under irradiation.

SiC is a promising material for nuclear applications and is a critical component in the construction of tristructural isotropic (TRISO) fuel. A primary issue with TRISO fuel operation is the observed release of 110mAg from intact fuel particles. The release of Ag has prompted research efforts to directly measure the transport mechanism of Ag in bulk SiC. Recent experimental efforts have focused primarily on Ag ion implantation designs. The effect of the thermal exposure system on the ion implantation surface has been investigated. Results indicate the utilization of a mated sample geometry and the establishment of a static thermal exposure environment is critical to maintaining an intact surface for diffusion analysis. The nature of the implantation surface and its potential role in Ag diffusion analysis are discussed.

14YWT oxide dispersion strengthened (ODS) ferritic steel was irradiated with of 5 MeV Ni2+ ions, at 300 °C, 450 °C, and 600 °C to a damage level of 100 dpa. The stability of Ti–Y–O nanoclusters was investigated by applying atom probe tomography (APT) in voltage mode, of the samples before and after irradiations. The average size and number density of the nanoclusters was determined using the maximum separation method. These techniques allowed for the imaging of nanoclusters to sizes well below the resolution limit of conventional transmission electron microscopy techniques. The most significant changes were observed for samples irradiated at 300 °C where the size (average Guinier radius) and number density of nanoclusters were observed to decrease from 1.1 nm to 0.8 nm and 12 × 1023 to 3.6 × 1023, respectively. In this study, the nanoclusters are more stable at higher temperature.

The microstructural changes and associated effects on thermal conductivity were examined in UO2 after irradiation using 3.9 MeV He2+ ions. Lattice expansion of UO2 was observed in x-ray diffraction after ion irradiation up to 5×1016 He2+/cm2 at low-temperature (< 200 °C). Transmission electron microscopy (TEM) showed homogenous irradiation damage across an 8 µm thick plateau region, which consisted of small dislocation loops accompanied by dislocation segments. Dome-shaped blisters were observed at the peak damage region (depth around 8.5 µm) in the sample subjected to 5×1016 He2+/cm2, the highest fluence reached, while similar features were not detected at 9×1015 He2+/cm2. Laser-based thermo-reflectance measurements showed that the thermal conductivity for the irradiated layer decreased about 55 % for the high fluence sample and 35% for the low fluence sample as compared to an un-irradiated reference sample. Detailed analysis for the thermal conductivity indicated that the conductivity reduction was caused by the irradiation induced point defects.

Molten FLiNaK salt [46.5%LiF-11.5%NaF-42%KF (mol%)] has been proposed for use as a secondary reactor coolant and medium for transfer of high-temperature process heat from nuclear reactors to chemic...

Fission products, such as krypton (Kr), are known to be insoluble within UO2, segregating toward grain boundaries and eventually leading to a lowering in thermal conductivity and fuel swelling. Recent computational studies have identified that differences in grain boundary structure have a significant effect on the segregation behavior of fission products. However, experimental work supporting these simulations is lacking. Atom probe tomography was used to measure the Kr distribution across grain boundaries in UO2. Polycrystalline depleted UO2 samples were irradiated with 0.7 MeV and 1.8 MeV Kr-ions and annealed to 1000°C, 1300°C, and 1600°C for 1 h to produce a Kr-bubble dominated microstructure. The results of this work indicate a strong dependence of Kr concentration as a function of grain boundary structure. Temperature also influences grain boundary chemistry with greater Kr concentration evident at higher temperatures, resulting in a reduced Kr concentration in the bulk. Although Kr segregation takes place at elevated temperatures, no change in grain size or texture was observed in the irradiated UO2 samples.

Zone-refined ultra high pure ZrC with five C/Zr ratios ranging from 0.84 to 1.17 was irradiated using a 2 MeV proton beam at 1125 °C. The stoichiometry effect on the irradiation response of ZrC microstructure was examined using transmission electron microscopy following the irradiation. The irradiated microstructures generally feature a high density of perfect dislocation loops particularly at away from the graphite precipitates, and the C/Zr ratio shows a notable effect on the size and density of dislocation loops. The dislocation loops are identified as interstitial type perfect loops, and it was indirectly proved that the dislocation loop core likely consists of carbon atoms. Graphite precipitates that form with excess carbon in the super-stoichiometric ZrC are detrimental, and the dramatic increases in the size of and density of dislocation loops in the vicinity of graphite precipitates in ZrC phase were observed. Irradiation-induced faceted voids were only observed in ZrC0.95, which is attributed to the pre-existing dislocation lines as biased sinks for vacancies.

In situ transmission electron microscopy observation of polycrystalline UO2 (with average grain size of about 5 µm) irradiated with Kr ions at 600°C and 800°C was conducted to understand the radiation-induced dislocation evolution under the influence of grain boundaries. The dislocation evolution in the grain interior of polycrystalline UO2 was similar under Kr irradiation at different ion energies and temperatures. As expected, it was characterized by the nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation lines and tangles at high doses. For the first time, a dislocation-denuded zone was observed near a grain boundary in the 1-MeV Kr-irradiated UO2 sample at 800°C. The denuded zone in the vicinity of grain boundary was not found when the irradiation temperature was at 600°C. The suppression of dislocation loop formation near the boundary is likely due to the enhanced interstitial diffusion toward grain boundary at the high temperature.

Zirconium carbide (ZrC) is being considered for utilization in deep burn TRISO fuel particles for high-temperature, gas-cooled reactors. Zirconium carbide has a cubic B1 type crystal structure along with a very high melting point (3420 °C), exceptional hardness and good thermal and electrical conductivities. Understanding the ZrC irradiation response is crucial for establishing ZrC as an alternative component in TRISO fuel. Until now, very few studies on irradiation effects on ZrC have been released and fundamental aspects of defect evolution and kinetics are not well understood although some atomistic simulations and phenomenological studies have been performed. This work was carried out to understand the damage evolution in float-zone refined ZrC with different stoichiometries. Proton irradiations at 800 °C up to doses of 3 dpa were performed on ZrC x (where x ranges from 0.9 to 1.2) to investigate the damage evolution. The irradiation-induced defects, such as density of dislocation loops, at different stoichiometries and doses which were characterized by transmission electron microscopy (TEM) is presented and discussed.

The stability of Y–Ti–O nanoclusters, dislocation structure, and grain boundary segregation in 9Cr-ODS steels has been investigated following proton irradiation at 400 °C with damage levels up to 3.7 dpa. A slight coarsening and a decrease in number density of nanoclusters were observed as a result of irradiation. The composition of nanoclusters was also observed to change with a slight increase of Y and Cr concentration in the nanoclusters following irradiation. Size, density, and composition of the nanoclusters were investigated as a function of nanocluster size, specifically classified to three groups. In addition to the changes in nanoclusters, dislocation loops were observed after irradiation. Finally, radiation-induced enrichment of Cr and depletion of W were observed at grain boundaries after irradiation.

In situ Transmission Electron Microscopy was conducted for single crystal UO2 to understand the microstructure evolution during 300 keV Xe irradiation at room temperature. The dislocation microstructure evolution was shown to occur as nucleation and growth of dislocation loops at low irradiation doses, followed by transformation to extended dislocation segments and tangles at higher doses. Xe bubbles with dimensions of 1–2 nm were observed after room-temperature irradiation. Electron Energy Loss Spectroscopy indicated that UO2 remained stoichiometric under room temperature Xe irradiation.

Detailed investigation of a recently failed static corrosion test involving molten eutectic LiF–NaF–KF salt at 850 C contained in 316L stainless steel crucibles shows that a single sealed crucible leaked molten salt into the electric furnace due to mechanical failure, causing a chemical reaction between the molten salt and the clay bonded silicon carbide insulation used in the furnace. As a result, corrosive vapors of Na2SiF6 ,K 2SiF6, SiF3, and F2, were formed. These vapors reacted with the stainless steel crucibles and lead to the formation of a porous corrosion crust and the eventual catastrophic structural failure of all nine sealed 316L stainless steel crucibles. 2014 Published by Elsevier Ltd.

Microstructure characterization of corrosion behavior of an alumina forming austenitic (AFA) steel exposed to supercritical carbon dioxide was conducted at 450–650 °C and 20 MPa. At low temperature and short exposure times, the oxidation kinetics were parabolic and the oxide scales were mainly composed of protective and continuous Al 2 O 3 and (Cr, Mn)-rich oxide layers. As the temperature and exposure time increased, the AFA steel gradually suffered breakaway oxidation and its oxide scales showed a multilayer structure mainly composed of Fe 3 O 4 , (Cr, Fe) 3 O 4 , NiFe/FeCr 2 O 4 /Cr 2 O 3 /Al 2 O 3 , FeCr 2 O 4 /Al 2 O 3 , and NiFe/Cr 2 O 3 /Al 2 O 3 , in sequence. The corrosion mechanism based on the microstructure evolution is discussed in detail.

We investigate the influence of proton irradiation on thermal conductivity in single crystal silicon. We apply a laser based modulated thermoreflectance technique to measure the change in conductivity of the thin layer damaged by proton irradiation. Unlike time domain thermoreflectance techniques that require application of a metal film, we perform our spatial domain measurement on uncoated samples. This provides greater sensitivity to the change in conductivity of the thin damaged layer. Using sample temperature as a parameter provides a means to deduce the primary defect structures that limit thermal transport. We find that under high temperature irradiation the degradation of thermal conductivity is caused primarily by extended defects.

Ferritic/Martensitic (F/M) steels with high Cr content posses the high temperature strength and low swelling rates required for advanced nuclear reactor designs. Radiation induced segregation (RIS) occurs in F/M steels due to solute atoms preferentially coupling to point defect fluxes which migrate to defect sinks, such as grain boundaries (GBs). The RIS response of F/M steels and austenitic steels has been shown to be dependent on the local structure of GBs where low energy structures have suppressed RIS responses. This relationship between local GB structure and RIS has been demonstrated primarily in ion-irradiated specimens. A 9 wt.% Cr model alloy steel was irradiated to 3 dpa using neutrons at the Advanced Test Reactor (ATR) to determine the effect of a neutron radiation environment on the RIS response at different GB structures. This investigation found the relationship between GB structure and RIS is also active for F/M steels irradiated using neutrons. The data generated from the neutron irradiation is also compared to RIS data generated using proton irradiations on the same heat of model alloy.

The effect of low temperature proton irradiation in depleted uranium dioxide was examined as a function of fluence. With 2.6 MeV protons, the fluence limit for preserving a good surface quality was found to be relatively low, about 1.4 and 7.0 × 10 17 protons/cm 2 for single and poly crystalline samples, respectively. Upon increasing the fluence above this threshold, severe surface flaking and disintegration of samples was observed. Based on scanning electron microscopy (SEM) and X-ray diffraction (XRD) observations the causes of surface failure were associated to high H atomic percent at the peak damage region due to low solubility of H in UO 2 . The resulting lattice stress is believed to exceed the fracture stress of the crystal at the observed fluencies. The oxygen point defects from the displacement damage may hinder the H diffusion and further increase the lattice stress, especially at the peak damage region.

In situ transmission electron microscopy (TEM) observation of UO2 single crystal irradiated with Kr ions at high temperatures was conducted to understand the dislocation evolution due to high-energy radiation. The dislocation evolution in UO2 single crystal is shown to occur as nucleation and growth of dislocation loops at low-irradiation doses, followed by transformation to extended dislocation segments and networks at high doses, as well as shrinkage and annihilation of some loops and dislocations due to high temperature annealing. Generally the trends of dislocation evolution in UO2 were similar under Kr irradiation at different ion energies and temperatures (150 keV at 600 °C and 1 MeV at 800 °C) used in this work. Interstitial-type dislocation loops with Burgers vector along 〈1 1 0〉 were observed in the Kr-irradiated UO2. The irradiated specimens were denuded of dislocation loops near the surface.

Ferritic–martensitic (FM) steels have been widely considered as structural materials for future fusion and fission systems due to their excellent swelling resistance and other properties under irradiation. NF616 is a newer generation FM steel that provides higher allowable operating temperatures, but there is limited information on the development of irradiated microstructures or associated mechanical properties in NF616. Studying the irradiation-induced defects as a function of dose could help to better understand the mechanisms of damage evolution. Proton irradiation at 773 K, up to 1, 2 and 3 dpa, was performed on NF616 to investigate damage evolution at different doses. Irradiation-induced defects, such as clusters, loops and dislocations, were investigated.

Development of TiO 2 diffusion barrier coating by electrophoretic deposition (EPD) has been studied to mitigate fuel-cladding chemical interactions (FCCI). Important EPD deposition parameters, including solvent, additives, particle size and crystal structure, current, and voltage were optimized for coating deposition on flat T91 ferritic steel substrates. Post-deposition sintering in the range of 850–1050 °C was investigated. Diffusion characteristics of the coatings were evaluated by diffusion couple experiments at 575 °C for 100 h using cerium as one of the fission products responsible for FCCI. Results showed that the coated steel exhibited up to 83% reduction in solid state inter-diffusion with cerium. Heat transfer calculations showed that the fuel center-line temperature would increase slightly due to the addition of the TiO 2 diffusion barrier coating; however, the maximum temperature still remains well below the melting point of uranium and is even lower than eutectic temperature between Fe 2 U and Fe 2 U 6 at cladding centerline and cladding/fuel interface, respectively.

Emissivity measurements for candidate alloys for very high temperature reactors were carried out in a custom-built experimental facility, capable of both efficient and reliable measurements of spectral emissivities of multiple samples at high temperatures. The alloys studied include 304 and 316 austenitic stainless steels, Alloy 617, and SA508 ferritic steel. The oxidation of alloys plays an important role in dictating emissivity values. The higher chromium content of 304 and 316 austenitic stainless steels, and Alloy 617 results in an oxide layer only of sub-micron thickness even at 700 °C and consequently the emissivity of these alloys remains low. In contrast, the low alloy SA508 ferritic steel which contains no chromium develops a thicker oxide layer, and consequently exhibits higher emissivity values.

Corrosion tests were performed in steam and supercritical water at 500 °C for two ferritic–martensitic alloys: HCM12A and NF616. The corrosion kinetics for the two alloys are similar in both environments showing near cubic behavior, but the corrosion rate was significantly higher in supercritical water than in steam. Examinations of the oxide layers using scanning electron microscopy and microbeam synchrotron diffraction and fluorescence show that both alloys form two-layer oxide structures in either environment. The outer layer contains only Fe 3 O 4 , while the inner layer contains a mixture of Fe 3 O 4 and FeCr 2 O 4 . Additionally, marker experiments using a novel photolithographic deposition process show that the original water–metal interface corresponds with the outer–inner layer interface as expected. The results are discussed in light of known corrosion mechanisms.

Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

Corrosion of PM2000 ODS steel in supercritical CO2 at 650 °C and 20 MPa for 3000 h was studied. The corrosion performance of the alloys was evaluated by weight change measurements, and the surface oxide scale was characterized by scanning electron microscopy, X-ray diffraction, and X-ray photoelectron spectroscopy. PM2000 ODS steel showed significantly better corrosion resistance compared to austenitic stainless steels and nickel based alloys tested in the same conditions. The high corrosion resistance is attributed to the formation of a dense and protective aluminum-rich oxide layer on the surface. The oxide layer thickness was only about 200 nm after 3000 h exposure.