Advanced Nuclear Materials Research Articles

Investigation of nano-clusters evolved with Ti and Zr in ferritic ODS steel and their effect on microstructure and mechanical properties

In the present study, four nominal compositions Fe-15Cr-2W-0.3Y2O3-x (x = 0, 0.3Ti, 0.3Zr, 0.3Ti-0.3Zr) of ODS steel have been consolidated via spark-plasma-sintering of mechanically alloyed powders. All the samples were characterized via SEM, XRD, EBSD and TEM analysis and tested for hardness and compressive strength at room as well as elevated temperatures. Further, strength contributions of temperature dependent strengthening mechanisms were determined and correlated with the experimental values of compressive yield strength. Results suggest that contrary to Ti, addition of Zr was found with the precipitation of dense and much finer nano-clusters (∼3 nm) and observed to be more effective in enhancing the hardness and compressive strength at room as well as high temperatures. Composition containing both Ti and Zr solutes resulted in ultrafine grains of nearly uniform sizes and exhibited considerably enhanced hardness as well as compressive strength at all temperatures. However, it favoured the formation of Y-Zr-O nano-clusters instead of Y-Ti-O due to exhibiting lower formation enthalpy as determined from their reaction energies. As per the theoretically evaluated strength contributions, Orowan strengthening was found prevailed at high temperatures and it had a major contribution in compressive yield strength at room temperature as well. The study suggests the prevailing effect of formation of Y-Zr-O nano-clusters over Y-Ti-O nano-clusters in terms of obstacles to grain boundaries and mechanical properties which may provide an insight for the development of advanced nuclear materials.

High-temperature dimensional stability and radiation behavior of high-entropy rare earth tungstate ceramics

Materials with high-temperature dimensional stability and resistance to radiation damage are receiving increasing attention in advanced nuclear energy systems. In this research, (Sm0.2Eu0.2Gd0.2Dy0.2A0.2)2W3O12-type (A = Y, Tm,Ho) high-entropy ceramics is presented. The coefficients of thermal expansion of (Sm0.2Eu0.2Gd0.2Dy0.2Y0.2)2W3O12 (HE-RE2W3O12) is 7.21 × 10−6 K−1 at 300–1100K, which is the lowest among the three synthesized high-entropy ceramics and indicates good dimensional stability at high temperatures. After irradiation with Kr+ fluence of 4.58 × 1016 ions/cm2 at 400 °C, both HE-RE2W3O12 and Gd2W3O12 maintained crystalline, but the lattice volume expansion rate of HE-RE2W3O12 (0.71 %) is much lower than that of Gd2W3O12 (1.49 %). Additionally, in the post-irradiated HE-RE2W3O12, the Raman spectra vibration peak remained unchanged and no radiation-induced segregation was observed. Nanoindentation tests displayed that the degradation of hardness and elastic modulus in HE-RE2W3O12 is rarely degrade. Overall, these results provide new insights for research on advanced nuclear materials in the future.

A multiclass classification model for predicting the thermal conductivity of uranium compounds

ABSTRACT Advanced nuclear fuels are designed to offer improved performance and accident tolerance, with an emphasis on achieving higher thermal conductivity. While promising fuel candidates like uranium nitrides, carbides, and silicides have been widely studied, the majority of uranium compounds remain unexplored. To search for potential candidates among these unexplored uranium compounds, we incorporated machine learning to accelerate the material discovery process. In this study, we trained a multiclass classification model to predict a compound’s thermal conductivity based on 133 input features derived from element properties and temperature. The initial training data consist of over 160,000 processed thermal conductivity records from the Starrydata2 database, but a skewed data class distribution led the trained model to underestimate compound’s thermal conductivity. Consequently, we addressed the issue of class imbalance by applying Synthetic Minority Oversampling TEchnique and Random UnderSampling, improving the recall for materials with thermal conductivity higher than 15 W/mK from 0.64 to 0.71. Finally, our best model is used to identify 119 potential advanced fuel candidates with high thermal conductivity among 774 stable uranium compounds. Our results underscore the potential of machine learning in the field of nuclear science, accelerating the discovery of advanced nuclear materials.

Crack-healing mechanisms in high-entropy alloys under ion irradiation

High-entropy alloys (HEAs) are potential candidates for advanced nuclear structural materials due to their impressive mechanical properties under extreme conditions. However, micro-cracks, the most common material damage, are introduced upon of the material synthesis and in service. In this work, an atomistic investigation of the crack-healing mechanisms of a FeCoCrNiAl0.5 HEA under ion irradiation is provided by molecular dynamics simulations. Quantitative analysis of the generation and recombination of point defects during the process of overlapping collision cascades is implemented to assess the crack-healing mechanisms in the HEA. The interstitial defects generated in the core of the cascade during the first collision event diffuse to the crack surface, resulting in crack-healing during subsequent recrystallization. In addition, the corresponding vacancies accumulate and forms large-size vacancy clusters that generate stacking faults and complex dislocation networks distributed around the location of the healed crack. With increasing the number of overlapping cascades the defects recombination rate increases and the phase stability is further improved. Crack-healing engineering in HEAs subjected to ion irradiation could pave the way towards designing advanced nuclear materials.

Cladding Failure Modelling for Lead-Based Fast Reactors: A Review and Prospects

Lead-cooled fast reactors (LFRs) are considered one of the most promising technologies to meet the requirements introduced for advanced nuclear systems. LFRs have higher neutron doses, higher temperatures, higher burnup and an extremely corrosive environment. The failure studies of claddings play a vital role in improving the safety criteria of nuclear reactors and promoting research on advanced nuclear materials. This paper presented a comprehensive review of the extreme environment in LFRs based on the fuel performance analyses and transient analyses of reference LFRs. It provided a clear image of cladding failure, focusing on the underlying mechanisms, such as creep, rupture, fatigue, swelling, corrosion, etc., which are resulted from the motions of defects, the development of microcracks and accumulation of fission products to some extent. Some fundamental parameters and behavior models of Ferritic/Martensitic (F/M) steels and Austenitic stainless (AuS) steels were summarized in this paper. A guideline for cladding failure modelling was also provided to bridge the gap between fundamental material research and realistic demands for the application of LFRs.

High-entropy (Sm0.2Eu0.2Gd0.2Dy0.2Er0.2)2Hf2O7 ceramic with superb resistance to radiation-induced amorphization

Nuclear engineering materials are required to possess outstanding extreme environmental tolerance and irradiation resistance. A promising novel pyrochlore-type of (Sm0.2Eu0.2Gd0.2Dy0.2Er0.2)2Hf2O7 high-entropy ceramic (HE-RE2Hf2O7) for control rod was prepared by solid-state reaction method. The ion irradiation of HE-RE2Hf2O7 with 400 keV Kr+ at 400 °C was investigated using a 400 kV ion implanter and compared with single-component pyrochlore Gd2Hf2O7 to evaluate the irradiation resistance. For HE-RE2Hf2O7, the phase transition from pyrochlore to defective fluorite is revealed after irradiation at 60 dpa. After irradiation at 120 dpa, it maintained crystalline, which is comparable to Gd2Hf2O7 but superior to the titanate pyrochlores previously studied. Moreover, the lattice expansion of HE-RE2Hf2O7 (0.22%) is much lower than that of Gd2Hf2O7 (0.62%), indicating excellent irradiation damage resistance. Nanoindentation tests displayed an irradiation-induced increase in hardness and a decrease in elastic modulus by about 2.6%. Irradiation-induced segregation of elements is observed on the surface of irradiated samples. In addition, HE-RE2Hf2O7 demonstrates a more sluggish grain growth rate than Gd2Hf2O7 at 1200 °C, suggesting better high-temperature stability. The linear thermal expansion coefficient of HE-RE2Hf2O7 is 10.7 × 10-6 K-1 at 298–1273 K. In general, it provides a new strategy for the design of the next advanced nuclear engineering materials.

Effective self-healing behavior of nanocrystalline-amorphous laminated alloy under irradiation

An extensive investigation on the microstructural evolution of nanocrystalline–amorphous laminated alloys (NALAs) by molecular dynamics simulations and mechanistic analysis have been conducted to apprehend the interplay of complex phenomena governing structural changes in this alloy under neutron irradiation. It was discovered from the evolution profiles of free volumes, atomic unfilled spaces, and irradiation-induced vacancies that the profound structural response of the NALA was orchestrated by the rapid and spontaneous recovery of free volumes that indicate a self-healing ability in the amorphous zone, while the phenomenon of geometric atomic reconstitution in local structures governs the effective self-healing capacity for annihilated nanocrystal regions. Furthermore, a distinctive, self-migration/diffusion capture dynamics for the annihilation of defects by phase boundaries was discovered as an effective self-healing mechanism in NALAs. These findings will potentially facilitate the development of advanced nuclear materials with high irradiation resistance.

Fundamental Issues Identified for Thermodynamic Description of Molten Salt Systems

Thermodynamic modeling of the molten salt systems is critical in the design of advanced nuclear power plants, materials recycling, and many other important clean energy applications. As one of the most capable computational thermodynamic approaches, the CALPHAD (Calculation of Phase Diagrams) method has been widely used for multicomponent thermodynamic database development and further support more physics-based modeling. Although considerable CALPHAD modeling efforts for molten salt systems have been made, we found that no evaluation has been performed regarding the unary thermodynamic models for the molten salt systems despite the fact that more than one pure substance database is available for the CALPHAD modeling. Therefore, the thermodynamic descriptions of the pure salts should be critically evaluated to ensure a high-fidelity molten salt thermodynamic database. In this work, we comprehensively analyzed some selected molten salts using different available thermodynamic descriptions from two databases as an example. One is the SSUB database released by the SGTE (Scientific Group Thermodata Europe), and the other is the FactPS database from the FactSage. The significant discrepancies observed in the thermodynamic description between the existing CALPHAD pure substance databases and experiments call for an urgent effort to improve.

Preface to the special issue on Advanced Nuclear Materials

Preface to the special issue on Advanced Nuclear Materials

Nuclear data uncertainty propagation applied to the versatile test reactor conceptual design

The Versatile Test Reactor (VTR) currently under development is a 300 MWth sodium-cooled fast reactor (SFR) fueled with ternary metal alloy fuel, which aims to accelerate the testing of advanced nuclear fuels, materials, instrumentation, and sensors in high flux environments that are necessary to license the next generation of advanced reactor concepts. To support the VTR design process, uncertainties associated with the nuclear data has been propagated through the reactor core neutronics calculation to global parameters of interest, such as the core multiplication factor, kinetic parameters, and various reactivity feedback coefficients, following the sensitivity based uncertainty propagation approach. By folding the sensitivity coefficients, separately computed by the generalized perturbation theory code PERSENT and Monte Carlo code Serpent 2, with the variance–covariance matrices from COMMARA-2.0, we obtain the reaction-wise, isotope-wise, and overall uncertainties for each response of interest due to nuclear data uncertainty. With Serpent 2, the statistical error of the uncertainty is obtained by propagating the statistical error of the sensitivity coefficients through the same process using a newly developed uncertainty propagation method. From both codes, the overall top uncertainty contributors are found to be the cross section of Fe-56 elastic scattering, Na-23 elastic scattering, and U-238 inelastic scattering. The large contributions of the Fe-56 elastic scattering cross sections to global parameters are due to its relatively large relative uncertainty of 5–10% in nuclear data and the large volume of Fe-containing reflector assemblies in the fairly compact VTR core design. Both codes agreed well for the overall uncertainty estimates of all responses of interest, except the delayed neutron fraction, prompt neutron generation time, and the coolant density feedback coefficient, where Serpent 2 yielded a much larger value than PERSENT due to the large statistical error of sensitivity coefficients. The calculated uncertainties are also compared to those associated with other SFR cores. Another outcome of this study is a variance–covariance matrix of reactivity coefficients, which can be used in the subsequent uncertainty propagation to the system level to investigate the impact of identified uncertainties on system responses in the safety analysis.

A multiphysics model of the versatile test reactor based on the MOOSE framework

The traditional modeling approach for sodium fast reactor cores relies on separate physics models, where the fuel performance, thermal–hydraulics, and neutronics calculations required to predict the core physics characteristics for nominal conditions are decoupled by relying on user-imposed boundary conditions. This paper aims at evaluating the impact of multiphysics simulations for predicting the core characteristics of the Versatile Test Reactor, which is being designed as a 300-MWt sodium-cooled fast reactor. The purpose of the Versatile Test Reactor is to accelerate the testing of advanced nuclear materials in the United States. The proposed multiphysics model relies on the Griffin reactor physics code, the SAM thermal–hydraulic system code, the BISON fuel performance code, as well as generic Multiphysics Object-Oriented Simulation Environment capabilities implemented in the open-source tensor mechanics module. For keff calculations, the introduction of a tight coupling between the neutronics, thermo-mechanical and thermal–hydraulics models induces a change of around 543 pcm in the eigenvalue, compared to the traditional standalone neutronics calculation where approximate temperature profiles are used. The multiphysics model is then employed for quantifying the impact of the thermal conductivity uncertainties on some of the key figures of merit, such as the fuel centerline temperature, assembly powers, and keff for nominal core conditions. As anticipated, uncertainties on fuel thermal conductivity mostly impact the fuel centerline temperature, and to a lesser extend the keff.

Development of Advanced Nuclear Structural Materials for Sustainable Energy Development

Development of Advanced Nuclear Structural Materials for Sustainable Energy Development

Characterization of advanced nuclear materials under extreme environments

Characterization of advanced nuclear materials under extreme environments

Intermediate energy proton irradiation: Rapid, high-fidelity materials testing for fusion and fission energy systems

Fusion and advanced fission power plants require advanced nuclear materials to function under new, extreme environments. Understanding the evolution of mechanical and functional properties during radiation damage is essential to the design and commercial deployment of these systems. The shortcomings of existing methods could be addressed by a new technique - intermediate energy proton irradiation (IEPI) - using beams of 10–30 MeV protons to rapidly and uniformly damage bulk material specimens before direct testing of engineering properties. IEPI is shown to achieve high fidelity to fusion and fission environments in both primary damage production and transmutation, often superior to nuclear reactor or typical (low-range) ion irradiation. Modeling demonstrates that high dose rates (0.1–1 DPA/per day) can be achieved in bulk material specimens (100–300 μm) with low temperature gradients and induced radioactivity. The capabilities of IEPI are demonstrated through a 12 MeV proton irradiation and tensile test of 250 μm thick tensile specimens of a nickel alloy (Alloy 718), reproducing neutron-induced data. These results demonstrate that IEPI enables high throughput assessment of materials under reactor-relevant conditions, positioning IEPI to accelerate the pace of engineering-scale radiation damage testing and allow for quicker and more effective design of nuclear energy systems.

Re-evaluating ion and neutron irradiation data with calculated applicable arc-dpa

ABSTRACT To develop advanced nuclear materials, lots of irradiation experiments have been done to various materials in different conditions. Accurate quantification of the defects induced by irradiation is important for researches in nuclear materials, especially when evaluating the damage caused by different particles. An advanced model – athermal recombination corrected dpa (arc-dpa) is recently introduced. In order to apply arc-dpa on experimental data, overall arc-dpa in ion and neutron irradiated iron, copper and tungsten are calculated based on the arc-dpa function for single cascade and the calculated PKA energy spectra. Data of irradiation experiments are re-evaluated with our applicable arc-dpa. The use of arc-dpa on experimental data showed good fit, giving new insights into the evolution of irradiation damage done by different particles.

Atomic-scale mechanisms of He/V ratio effect on helium bubble hardening in iron for neutron irradiated F/M steels

The helium-to-vacancy (He/V) ratio is a key factor related to the helium bubble hardening mechanisms induced by neutron irradiation in ferritic/martensitic (F/M) steels as advanced nuclear structural materials. According to the helium bubble hardening results in F/M steel irradiated in the spallation neutron source, the interactions between a 1/2<111>{110} edge dislocation and helium bubbles with various sizes of 0.7–3.0 nm in bcc-Fe under irradiation temperatures and room temperature was investigated by using molecular dynamics simulation. The effect of various He/V ratios in the ranges of zero to the highest He/V ratios, 1.85–4.0 of stable helium bubbles with no dislocation loop emission, on the interaction mechanisms was analyzed. The results indicated that the critical resolved shear stress (CRSS) drops rapidly to a very low stress level, and immediately bottoms out and rebounds to a very high stress level, when the He/V ratio increases from 1.0 to the highest He/V ratio. The significant change of CRSS values is not only related to the climbing degree of dislocation after releasing, but also related to the moment when the climb appears. A new repulsion mechanism of interaction between dislocation and bubble was proposed to elucidate the abnormal phenomena of the bottoming rebound of CRSS with a great climbing degree, according to the repelling interaction between helium bubble and superjog of dislocation with both strong compressive stress field.

Application of a Graded Approach to Support the National Research Universal Reactor U-2 Experimental Loop Return to Service

Abstract The National Research Universal (NRU) reactor at Canadian Nuclear Laboratories (CNL) operated safely for over 60 years and supported a wide range of applications including, testing of fuels and materials under typical power reactor conditions in two experimental loops (U-1 and U-2). Both experimental loops had been taken out of service to address seismic deficiencies. CNL applied a graded approach to successfully return one of these loops, the U-2 Loop to service. The graded approach, without compromising safety, applied a risk informed methodology commensurate to the potential risk posed by the operation of the U-2 Loop. The work enabled the U-2 Loop to resume operation until the NRU reactor was permanently shut down in Mar. 31, 2018, generating valuable data that will be used in the development of advanced nuclear fuels and materials. This paper describes the graded approach employed by CNL that supported U-2 loop return to service (RTS). The use of graded approach is articulated to support development of safety and licensing cases for small modular reactor projects.

A physics-based mesoscale phase-field model for predicting the uptake kinetics of radionuclides in hierarchical nuclear wasteform materials

Multiscale, hierarchical, porous materials are promising nuclear wasteform materials with potential for efficiently adsorbing and sequestering radionuclides. However, fundamental understanding of such complex micro- and meso-structures on radionuclide diffusion and uptake kinetics is lacking, which is essential for designing advanced nuclear wasteform materials. In this work we develop a microstructural-dependent diffusion model that accounts for three-dimensional (3D) microstructural features, including heterogeneous thermodynamic and kinetic properties, to predict radionuclide uptake kinetics in complex porous structures. Na+ and Sr2+ ionic exchange in LTA zeolite multiscale materials in batch tests is taken as a model system to demonstrate the model and its application to hierarchical materials with multiscale porosity. It is found that the reduction of ionic mobility associated with chemistry, structure, and/or phase changes has more profound impact on the uptake kinetics in a large 3D porous particle than that in a small one. Uptake kinetics in large particles are limited by two diffusion processes: bulk diffusion and surface layer transport. Decreasing particle size and increasing mesoscale pore volume fraction changes the uptake kinetics from two diffusion processes to a single process and dramatically increases the uptake kinetics. Predicted uptake kinetics and the effect of microstructures on the effective capacity of radionuclides and uptake efficiency are consistent with results observed in Na+/Sr2+ exchange experiments. The results demonstrate that the developed model should be adaptable to transport problems in other hierarchical material systems.

Selective laser melting additive manufacturing of advanced nuclear materials V-6Cr-6Ti

Selective laser melting (SLM) additive manufacturing technology has been a prevailing method to fabricate components with complex physical geometry or novel structural design. However, original nuclear powder material such as vanadium-based alloy appropriate for SLM processing has yet not been obtained commercially, which significantly restricts the development of nuclear component manufacturing. In this study, near-sphere, uniform and fine V-6Cr-6Ti pre-alloy powder which met the performance demands for SLM processing was successfully obtained by high-energy ball milling. Subsequently, the V-6Cr-6Ti part was fabricated by SLM consolidation of as-prepared pre-alloy powder with double-region orthometric scanning strategy, forming strong texture feature within molten pool. The compression test was performed, showing that the maximum compression stress reached 1078MPa and the accumulated strain was about 0.32.

RETRACTED: First-Principles Study of Advanced Nuclear Materials: Defect Behavior and Fission Products in U-Si System

Uranium silicides are envisaged as potential nuclear materials for the next generation. U3Si is featured by the high actinide density and the better thermal conductivity relative to UO2. To properly and safely utilize nuclear materials, it is crucial to understand their chemical and physical properties. First-principles in theory is mostly used to analyze the point defect structures for uranium silicides nuclear fuels. The lattice parameters of U3Si and USi2 are calculated and the stability of different types of point defects are predicted by their formation energies. The results show that silicon vacancies are more prone to be produced than uranium vacancies in β-USi2 matrix. The most favorable sites of fission products are determined in this work as well. According to the current data, rare earth elements cerium and neodymium are found to be more stable than alkaline earth metals strontium and barium in a given nuclear matrix. It is also determined that in USi2 crystal lattice fission products tend to be stabilized in uranium substitution sites, while they are likely to form precipitates from the U3Si matrix. It is expected that this work may provide new insight into the mechanism for structural evolutions of silicide nuclear materials in a reactor as well as to provide valuable clues for fuel designers.