Journal of Nuclear Materials最新文献

Small-scale mechanical testing (SSMT) is of great interest in the nuclear materials community as it permits the use of surrogate irradiation techniques, such as ion beam irradiation when investigating the impact of irradiation damage on mechanical properties. The design of a micro-electro-mechanical-system (MEMS) micro-tensile stage is demonstrated to enable micron-sized specimen testing up to failure under ambient and reactor-relevant temperatures. Copper and Zircaloy-2 microscale specimens, having gauge lengths of 200 μm, gauge widths of 48 μm, and specimen-dependent constant gauge thicknesses of 10 μm to 26 μm, are fabricated using micro-wire electrical discharge machining and focused ion beam milling. From each gauge section, microstructure data is captured prior to testing using electron backscatter diffraction analysis, and an optimized digital image correlation speckle pattern is deposited on the specimen surface. The speckle pattern is tracked during deformation and post-processed to obtain strain-field evolution maps throughout deformation. Strain maps provide a means to account for and explain microstructure-dependent features observed in the captured stress-strain curves. The combination of using micro-wire electrical discharge machining, focused ion beam cleaning, and ion beam induced platinum deposition for micron-sized specimen preparation as well as utilizing microfabrication techniques to fabricate an integrated heating microtensile load frame reported herein serve as a demonstration of SSMT approaches that can be adopted to tackle challenging problems in nuclear materials research, including but not limited to the effects of ion beam irradiation on mechanical performance and providing experimental data to support small-scale modeling efforts of embrittling precipitates or radiation-induced segregation.

Zircon is a promising candidate for use as a matrix in the immobilization of radioactive waste. However, high temperatures (1400–1600 °C) and calcination durations are generally required to produce it. In this letter, we report the synthesis of cerium-containing solid solution based on zircon Zr_1-xCe_xSiO₄ by means of a straightforward and environmentally sustainable milling-assisted solid-phase approach conducted at a reduced temperature. Our study demonstrated that the utilization of mechanical activation of the oxide mixture in a mill not only diminished the temperature (to 1200 °C) and the duration of calcination but also increased the Ce content in zircon up to 6.4 at.%. Additionally, we propose a novel method for determining the phase composition of calcined samples using Vegard's law and mass balance.

Structural materials for laser-based inertial fusion energy (IFE) reactor concepts are expected to operate under pulsed irradiation conditions, with cycles consisting of microsecond-long neutron bursts followed by inter-pulse periods of up to one second in duration. During each laser shot, irradiation damage is introduced at dose rates that are up to six orders of magnitude higher than those in their magnetic fusion energy (MFE) counterparts. Under certain conditions, the inter-pulse periods may last an amount of time sufficient to anneal much of the damage introduced during each shot. This phenomenon is highly temperature dependent, with pulsed heating directly linked to the pulsed damage, large surface temperature spikes may also occur. As such, this intermittent mode of operation has the potential to lead to fundamental differences in how irradiation damage accumulates in structural reactor materials. However, damage to structural materials under IFE conditions has received comparatively much less attention than in MFE, as pulsed conditions add yet an extra dimension to the already extremely challenging problem of microstructural evolution under fusion neutron irradiation in structural materials. In this work we use the stochastic cluster dynamics (SCD) method to simulate the evolution with time of defect cluster concentrations under IFE conditions. We consider the Laser Inertial Fusion Energy (LIFE) reactor concept as the representative IFE design for our study, for which detailed spectral information is available, including gas transmutant production. We simulate several pulse frequencies and three different temperatures, and compare the results with continuous irradiation cases under identical average dose rates. The simulations are run in Fe-9Cr system as a model alloy for reduced-activation ferritic/martensitic (RAFM) steels, which are the leading structural material candidates for first-wall structures in MFE and IFE devices. We find that, in practically all scenarios, pulsed irradiation restricts the formation of helium-vacancy clusters relative to the levels seen under equivalent steady irradiation conditions. As well, although self-interstitial atom clusters do accumulate under pulsed operation, their number densities remain up to an order of magnitude lower than in continuous irradiation conditions. Based on the SCD results, we provide a temperature-pulse rate map to identify regions where pulsed irradiation may lead to larger defect accumulation than under continuous irradiation.

The oxidation-induced microstructure evolution of nuclear graphite (IG-110 and NBG-17) is studied. Graphite samples were oxidized in air at 500 °C. Complicated oxidation paths were observed on the ion-milled surface of oxidized graphite samples. The weight loss of graphite samples during oxidation is mainly attributed to the oxidation paths which are extended and broadened constantly. In contrast, the original gas-escape pores do not show obvious change in sizes during oxidation. The oxidation paths are characterized to be much thinner and more tortuous than the gas-escape pores, which should be taken into account in estimating the oxygen diffusivity in oxidized graphite. The edges of graphite sheets exposed in the oxidation paths could scatter the phonon and electrons strongly. Therefore, the thermal conductivity declines dramatically with oxidation time. At weight loss around 30 %, the thermal conductivity of the oxidized graphite decreased to ∼16.8 % of the corresponding original value.

The application of Liquid Lead-Bismuth Eutectic (LBE) as a coolant in Lead-cooled Fast Reactors (LFRs) presents significant challenges due to its corrosive nature, especially at elevated temperatures. This study investigates the corrosion kinetics and mechanisms of 15–15Ti stainless steel in flowing LBE at 500 °C with oxygen concentrations ranging from 1 to 3 × 10⁻⁶ wt% and a flow rate of 1 m s^-1. The research highlights the formation and growth of the Fe-Cr spinel oxide layer, which follows a parabolic rate law indicative of a diffusion-controlled process, crucial for long-term material stability predictions. Despite the initial effectiveness of the Fe-Cr spinel layer in mitigating corrosion, its integrity is compromised over time due to spallation and dissolution, allowing penetration of LBE elements and selective dissolution of Fe, Ni, and Cr. The study reveals a complex interplay between direct dissolution of steel components and spallation of the oxide layer, providing critical insights into material loss mechanisms and the degradation of structural materials in LBE-cooled reactors.

We report results from atom probe tomography (APT) experiments capturing Zr(Fe,Cr)₂ second phase particles (SPPs) in Zircaloy-2-type fuel cladding after reactor operation. In light of recent reports of H trapping around SPPs, we assess the feasibility of H analysis in modern commercial atom probe instruments on this system. To this end we employed voltage and laser pulsing APT on specimens prepared by focused ion beam (FIB) at room and cryogenic temperature. Room temperature FIB caused transformation of the α-Zr matrix into δ-hydride, but left SPPs mostly unaffected. This indicates that α-Zr has a higher affinity for H than SPPs. However, even under optimized conditions, we were not able to find evidence for H trapping near SPPs located within the α-Zr matrix in cryogenically FIB sharpened specimens, where no hydride transformation occurs.

Cantilever beam samples with V-notches were employed to measure the radial delayed hydride cracking rate (DHCR) in Zr-2.5Nb alloy pressure tubes within the temperature range from 120 °C to 250 °C in this study. Both initiation and propagation of DHC were monitored using the direct current potential drop (DCPD) technique. The results show that radial DHCR and temperature fit the Arrhenius relationship within the temperature as mentioned above range, yielding a crack propagation activation energy of 38.66 kJ/mol. It was observed that the incubation periods for radial DHC initiation typically shortened with rising temperature from 180 °C to 250 °C, and the anisotropy of DHCR is also discussed in this study. Besides, the presence of undissolved circumferential hydrides appears to affect radial DHC behavior.

High-entropy alloys (HEAs) are promising materials for nuclear applications. Understanding the influence of local chemical order (LCO) on defect diffusion and evolution in these alloys is crucial for enhancing their resistance to radiation damage. In this study, we used molecular dynamics simulation to investigate the effect of LCO on monovacancy diffusion in CoNiCrFe HEA. Alongside the Warren-Cowley parameter, which quantifies the degree of local elemental ordering, we propose a new parameter to characterize the spatial scale of LCOs. Based on their degree, LCO structures have been classified as either chemical medium-range order (CMRO) or chemical short-range order (CSRO). Our study reveals a non-monotonic variation in vacancy diffusion coefficients, transitioning from random solid solution to CSRO and then to CMRO structures. Moreover, we observe a preference for vacancy diffusion in low-energy Fe, Co-rich regions, with their spatial distribution and spatial connectivity significantly influencing the vacancy diffusion. Due to the spatial scale of the inhomogeneity introduced by LCO, both global average diffusion parameters and single migration barriers cannot fully reflect the actual diffusion dynamics. Therefore, our study emphasizes the importance of understanding the preferred diffusion pathways and the associated energy landscapes to fully assess the defect diffusion dynamics in HEAs. This deeper investigation into localized diffusion behaviors influenced by LCO is crucial for evaluating and enhancing the radiation damage tolerance of HEAs.

Ion-irradiated quartz, albite, and microcline were analyzed to understand the mechanism of neutron irradiation degradation of concrete structures in nuclear power plants for long-term operation. While mineral amorphization has previously been directly associated with the expansion of concrete aggregates, our study showed that volume expansion is more pronounced after crystalline structures become amorphous. The temperature dependence of volume expansion correlated well with the recovery of unpaired electrons and viscous flow of the amorphized quartz.

In nuclear power plants, CuCrZr alloy and pure copper (T2) are widely used as the structural materials and service at high temperatures, and the comprehensive understanding of the mechanical properties is essential. In this paper, the small punch test (SPT) is used to understand the variations of mechanical properties and fracture mechanism with temperature for them. The soft temperature of CuCrZr alloy is identified as 400 °C, with the sharp decrease in SPT strength parameters and fracture energy, which is caused by the oxidation. The SPT fracture mode changes from the central line crack to “O” shape circumferential crack, and the mixed oxidation and plastic deformation fracture mechanism is observed at 500 °C. The soft temperature of T2 is 300 °C, with the sharp decrease in the SPT fracture displacement. The “O” shape circumferential crack and the necking phenomenon are unchanged for the fracture mode of T2, but the fracture mechanism changes from transgranular fracture to intergranular fracture, with increase in temperature. This paper describes the deformation behaviour of CuCrZr alloy and T2 using SPT as a function of temperature, investigates the fracture behaviour and attempts to bring out the effect oxidation during high temperature tests.