Operating temperature windows for fusion reactor structural materials

Effect of Carbon on Void Nucleation in Iron

The study reports the significance of carbon presence in affecting void nucleation in Fe. Without carbon, void nucleation rates decrease gradually at high temperatures but remain significantly high and almost saturated at low temperatures. With carbon present, even at 1 atomic parts per million, void nucleation rates show a low-temperature cutoff. With higher carbon levels, the nucleation temperature window becomes narrower, the maximum nucleation rate becomes lower, and the temperature of maximum void nucleation shifts to a higher temperature. Fundamentally, this is caused by the change in effective vacancy diffusivity due to the formation of carbon-vacancy complexes. The high sensitivity of void nucleation to carbon comes from the high sensitivity of void nucleation to the vacancy arrival rate in a void. The void nucleation is calculated by first obtaining the effective vacancy diffusivity considering the carbon effect, then calculating the defect concentration and defect flux change considering both carbon effects and pre-existing dislocations, and finally calculating the void nucleation rate based on the recently corrected homogeneous void nucleation theory. The study is important not only in the fundamental understanding of impurity effects in ion/neutron irradiation but also in alloy engineering for judiciously introducing impurities to increase swelling resistance, as well as in the development of simulation and modeling methodologies applicable to other metals.

Measurement of the [Formula: see text]Ta([Formula: see text]) cross sections up to stellar s-process temperatures at the CSNS Back-n

The neutron capture cross section of [Formula: see text]Ta is relevant to s-process of nuclear astrophysics, extraterrestrial samples analysis in planetary geology and new generation nuclear energy system design. The [Formula: see text]Ta([Formula: see text]) cross section had been measured between 1 eV and 800 keV at the back-streaming white neutron facility (Back-n) of China spallation neutron source(CSNS) using the time-of-flight (TOF) technique and [Formula: see text] liquid scintillator detectors. The experimental results are compared with the data of several evaluated libraries and previous experiments in the resolved and unresolved resonance region. Resonance parameters are extracted using the R-Matrix code SAMMY in the 1-700 eV region. The astrophysical Maxwell average cross section(MACS) from kT = 5 to 100 keV is calculated over a sufficiently wide range of neutron energies. For the characteristic thermal energy of an astrophysical site, at kT = 30keV the MACS value of [Formula: see text]Ta is 834 ± 75 mb, which shows an obvious discrepancy with the Karlsruhe Astrophysical Database of Nucleosynthesis in Stars (KADoNiS) recommended value 766 ± 15 mb. The new measurements strongly constrain the MACS of [Formula: see text]Ta([Formula: see text]) reaction in the stellar s-process temperatures.

Equation of state for He bubbles in W and model of He bubble growth and bursting near W{100} surfaces derived from molecular dynamics simulations

Molecular dynamics (MD) simulations are performed to derive an equation of state (EOS) for helium (He) bubbles in tungsten (W) and to study the growth of He bubbles under a W(100) surface until they burst. We study the growth as a function of the initial nucleation depth of the bubbles. During growth, successive loop-punching events are observed, accompanied by shifts in the depth of the bubble towards the surface. Subsequently, the MD data are used to derive models that describe the conditions that cause the loop punching and bursting events. Simulations have been performed at 500, 933, 1500, 2000, and 2500 K to fit the parameters in the models. To compute the pressure in the bubble at the loop punching and bursting events from the models, we derive an EOS for He bubbles in tungsten with an accompanying volume model to compute the bubble volume for a given number of vacancies ([Formula: see text]), He atoms ([Formula: see text]), and temperature (T). To derive the bubble EOS, we firstly derive the EOS for a free He gas. The derived free-gas EOS can accurately predict all MD data included in the analysis (which span up to 54 GPa at 2500 K). Subsequently, the bubble EOS is derived based on the free-gas EOS by correcting the gas density to account for the interaction between He and W atoms. The EOS for the bubbles is fitted to data from MD simulations of He bubbles in bulk W that span a wide range of gas density and sizes up to about 3 nm in diameter. The pressure of subsurface bubbles at the loop punching events as calculated using the bubble-EOS and the volume model agrees well with the pressure obtained directly from the MD simulations. In the loop punching model, for bubbles consisting of [Formula: see text] vacancies and [Formula: see text] helium atoms, the [Formula: see text] ratio that causes the event, the resulting increase in [Formula: see text], and the associated shift of the bubble depth are formulated as a function of [Formula: see text] and T. In the bursting model, a bubble must simultaneously reach a certain depth and [Formula: see text] ratio in order to burst. The burst depth and [Formula: see text] are also modeled as a function of [Formula: see text] and T. The majority of the loop punching events occur at bubble pressures between 20 and 60 GPa, depending on the bubble size and temperature. The larger the bubble and the higher the temperature, the lower the bubble pressure. Furthermore, our results indicate that at a higher temperature, a bubble can burst from a deeper region.

Microengineering Design for Advanced W-Based Bulk Materials with Improved Properties

In fusion reactors, such as ITER or DEMO, the plasma used to generate nuclear reactions will reach temperatures that are an order of magnitude higher than in the Sun's core. Although the plasma is not supposed to be in contact with the reactor walls, a large amount of heat generated by electromagnetic radiation, electrons and ions being expelled from the plasma will reach the plasma-facing surface of the reactor. Especially for the divertor part, high heat fluxes of up to 20 MW/m² are expected even in normal operating conditions. An improvement in the plasma-facing material (which is, in the case of ITER, pure Tungsten, W) is desired at least in terms of both a higher recrystallization temperature and a lower brittle-to-ductile transition temperature. In the present work, we discuss three microengineering routes based on inclusions of nanometric dispersions, which are proposed to improve the W properties, and present the microstructural and thermophysical properties of the resulting W-based composites with such dispersions. The materials' behavior after 6 MeV electron irradiation tests is also presented, and their further development is discussed.

Effect of Vacancy Behavior on Precipitate Formation in a Reduced-Activation V-Cr-Mn Medium-Entropy Alloy

In this work, we studied the evolution of vacancy-like defects and the formation of brittle precipitates in a reduced-activation V-Cr-Mn medium-entropy alloy. The evolution of local electronic circumstances around Cr and Mn enrichments, the vacancy defects, and the CrMn₃ precipitates were characterized by using scanning electron microscopy with energy-dispersive spectroscopy, X-ray diffraction, and positron annihilation spectroscopy. The microstructure measurements showed that the Mn and Cr enrichments in the as-cast sample significantly evolved with temperature, i.e., from 400 °C, the Cr/Mn-segregated regions gradually dissolved into the matrix and then disappeared, and from 900 °C to 1000 °C, they existed as CrMn₃ precipitates. The crystallite size of the phase corresponding to CrMn₃ precipitates was about 29.4 nm at 900 °C and 43.7 nm at 1000 °C. The positron annihilation lifetime results demonstrated that the vacancies mediated the migration of Cr and Mn, and Cr and Mn segregation finally led to the formation of CrMn₃ precipitates. The coincidence Doppler broadening results showed that the characteristic peak moved to the low-momentum direction, due to an increase in the size of the vacancy defects at the interface and the formation of CrMn₃ precipitates.

Distribution of Hydrogen and Defects in the Zr/Nb Nanoscale Multilayer Coatings after Proton Irradiation

Radiation damage is one of the significant factors limiting the operating time of many structural materials working under extreme conditions. One of the promising directions in the development of materials that are resistant to radiation damage and have improved physical and mechanical properties is the creation of nanoscale multilayer coatings (NMCs). The paper is devoted to the experimental comprehension of changes in the defect structure and mechanical properties of nanoscale multilayer coatings (NMCs) with alternating layers of Zr and Nb under irradiation. Series of Zr/Nb NMCs with different thicknesses of individual layers were fabricated by magnetron sputtering and subjected to H+ irradiation. The evolution of structure and phase states, as well as the defect state under proton irradiation, was studied using the methods of high-resolution transmission electron microscopy (HRTEM), X-ray diffraction analysis (XRD), glow discharge optical emission spectroscopy (GDOES), and positron annihilation spectroscopy (PAS). The layer-by-layer analysis of structural defects was carried out by Doppler broadening spectroscopy (DBS) using a variable-energy positron beam. To estimate the binding energy and the energy paths for the hydrogen diffusion in Zr/Nb NMCs, calculations from the first principles were used. When the thickness of individual layers is less than 25 nm, irradiation causes destruction of the interfaces, but there is no significant increase in the defect level, the S parameter (open volume defects amount) before and after irradiation is practically unchanged. After irradiation of NMC Zr/Nb with a thickness of layers 50 and 100 nm, the initial microstructure is retained, and the S parameter is significantly reduced. The GDOES data reveal the irregular H accumulation at the interface caused by significant differences in H diffusion barriers in the bulk of Zr and Nb multilayers as well as near the interface’s region.

High-Entropy Alloys for Advanced Nuclear Applications

The expanded compositional freedom afforded by high-entropy alloys (HEAs) represents a unique opportunity for the design of alloys for advanced nuclear applications, in particular for applications where current engineering alloys fall short. This review assesses the work done to date in the field of HEAs for nuclear applications, provides critical insight into the conclusions drawn, and highlights possibilities and challenges for future study. It is found that our understanding of the irradiation responses of HEAs remains in its infancy, and much work is needed in order for our knowledge of any single HEA system to match our understanding of conventional alloys such as austenitic steels. A number of studies have suggested that HEAs possess ‘special’ irradiation damage resistance, although some of the proposed mechanisms, such as those based on sluggish diffusion and lattice distortion, remain somewhat unconvincing (certainly in terms of being universally applicable to all HEAs). Nevertheless, there may be some mechanisms and effects that are uniquely different in HEAs when compared to more conventional alloys, such as the effect that their poor thermal conductivities have on the displacement cascade. Furthermore, the opportunity to tune the compositions of HEAs over a large range to optimise particular irradiation responses could be very powerful, even if the design process remains challenging.

Recent advances in characterising irradiation damage in tungsten for fusion power

Tungsten is the front-runner candidate for building the plasma-facing armour components for future fusion reactors. However, in-service irradiation by fusion-neutrons and helium will create lattice-defects in the material, compromising its properties and lifetime. Improving the component’s resilience to radiation damage and accurately predicting the lifetime of irradiated components is key for commercial feasibility of the reactor. For this purpose, understanding the creation and evolution of radiation damage is essential. This paper reviews recent advances in characterising radiation damage through experimental and modelling techniques. Tungsten-ion- and helium-ion-implantation are commonly used to mimic the damage created by neutron- and helium-irradiation respectively. Defects (> 1.5 nm) can be directly imaged using transmission electron microscopy while all defects (size-independent), may be indirectly probed by measuring lattice strains induced by them (using diffraction techniques; synchrotron X-rays or high-resolution electron-backscatter). Neutron-irradiation produces mainly ½〈111〉 prismatic loops. Loop-interaction and structural organisation evolves with changing implantation dose and temperature. Helium-irradiation, < 573 K, induces formation of small helium-vacancy clusters, which evolve into bubbles, blisters and “fuzz” structure with changing temperature and dose. Nano-indentation or micro-cantilever bending tests can be used to examine mechanical properties of ion-implanted layers. Both helium- and neutron-implantation defects induce increased hardening often followed by subsequent strain-softening and localised deformation. Such irradiation-induced alterations are detrimental to material ductility and long-term structural integrity of tungsten-based components. Development of physically-based material models that capture the physics of underlying irradiation-induced changes, inspire confidence of reliably using simulations to predict mechanical behaviour and in-service performance of irradiated engineering components in future.

Exploitation of thermal gradients for investigation of irradiation temperature effects with charged particles

The effects of radiation damage on materials are strongly dependant on temperature, making it arguably the most significant parameter of concern in nuclear engineering. Owing to the challenges and expense of irradiating and testing materials, material property data is often limited to few irradiation conditions and material variants. A new technique has been developed which enables the investigation of radiation damage of samples subject to a thermal gradient, whereby a wealth of data over a range of irradiation temperatures is produced from a single irradiation experiment. The results produced are practically inaccessible by use of multiple conventional isothermal irradiations. We present a precipitation-hardened copper alloy (CuCrZr) case-study irradiated with a linear temperature gradient between 125 and 440 °C. Subsequent micro-scale post irradiation characterisation (nanoindentation, transmission electron microscopy and atom probe tomography) highlight the capability to observe mechanical and microstructural changes over a wide range of irradiation temperatures. We observed irradiation-softening in CuCrZr that did not occur due to irradiation-enhanced aging of the Cr-precipitates. Excellent reproducibility of the new technique was demonstrated and replicated irradiation-hardening data from several isothermal neutron irradiation studies. Our new technique provides this data at a fraction of the time and cost required by conventional irradiation experiments.

Functional Multi-Nanolayer Coatings of Amorphous Carbon/Tungsten Carbide with Exceptional Mechanical Durability and Corrosion Resistance

A novel functional multilayer coating with periodically stacked nanolayers of amorphous carbon (a:C)/tungsten carbide (WC) and an adhesion layer of chromium (Cr) was deposited on 304 stainless steel using a dual magnetron sputtering technique. Through process optimization, highly densified coatings with high elasticity and shear modulus, excellent wear resistance, and minimal susceptibility to corrosive and caustic media could be acquired. The structural and mechanical properties of the optimized coatings were studied in detail using a variety of analytical techniques. Furthermore, finite element method simulations indicated that the stress generated due to contact against a steel ball was distributed well within the coating, which allowed the stresses to be lower than the yield threshold of the coating. Thus, an ultralow wear rate of ∼10^-12mm³/N mm could be achieved in dry sliding conditions under relatively high Hertzian contact pressures of ∼0.4-0.9 GPa. The amorphous and pinhole-free structure of the individual layers, sufficient number of pairs, and the relatively dense stacked layers resulted in significant polarization resistance (Z″ = 5.5 × 10⁶ Ω cm²) and increased the corrosion resistance of the coating by 10-fold compared to that of recently reported corrosion-resistant coatings.