Advanced Nuclear Energy Systems Research Articles

Preparation and characterization of medium entropy alloy CrCoNi toughened SiC ceramics

SiC is a promising structural material for future advanced nuclear energy systems. However, the inherent brittleness and difficulty of densification of SiC limits its practical application. In this study, medium entropy alloy (CrCoNi) reinforced silicon carbide composites (CCN/SiC) were sintered at 1950–2200°C by spark plasma sintering (SPS). The influence of CCN addition on the CCN/SiC composites was investigated in terms of microstructure, mechanical and thermal properties. The results show that SiC sintered at 2200℃ has the highest density, and the addition of 7.5 wt% CCN can significantly improve the fracture toughness of SiC. Transgranular fracture, crack deflection, and premature crack termination due to interfacial passivation are considered as potential toughening mechanisms. Although the addition of CCN can degrade the thermal conductivity of SiC ceramics, it has little effect on the structural stability of the SiC matrix at high temperatures.

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.

Comparative study on responses of three candidate materials for ADS beam/target windows to proton irradiation: T91, SIMP and Ti-6Al-4V

Accelerator-Driven System (ADS) represents an advanced nuclear energy system designed for the purpose of transmuting and eliminating nuclear waste. The irradiation tolerance of a material stands out as a critical factor in determining its availability for ADS beam/target windows. This study delves into the response of three promising candidates (T91, SIMP, and Ti-6Al-4V) for ADS beam/target windows to proton irradiation, as investigated by a 1.4 MeV H2+ irradiation experiment to dose of 1.5 peak dpa at 460 °C. Their cavity swelling and hardening behavior were estimated using Transmission Electron Microscope (TEM) and Nanoindentation, respectively. It was demonstrated that SIMP steel shows lower cavity swelling but slightly higher irradiation hardening compared to T91. Notably, no cavity swelling is observed in Ti-6Al-4V, but it experiences a higher level of irradiation hardening. The nucleation and evolution mechanisms of irradiation-induced defects (including cavities, dislocation loops, and precipitates) and their contributions to the proton irradiation response are discussed. Furthermore, a summary of the advantages and disadvantages of the three materials as candidates for proton beam/target windows is provided.

Tritium storage and release mechanism in nuclear graphite using first-principles combined thermal desorption theory

Nuclear graphite (NG) is widely applied in nuclear facilities, and its strong interaction with tritium is an essential factor for tritium transport behavior in various reactors and the treatment of radioactive NG waste. Given that NG is a polycrystalline and porous material, interactions between tritium and NG exist in various microstructures, including graphite edges, graphene basal planes, and amorphous structures. In this study, density functional theory which is integrated in Vienna ab initio simulation package (VASP) was combined with thermal desorption theory to elucidate the underlying mechanism of hydrogen isotope storage corresponding to the experimentally observed desorption peaks in NG. The four desorption temperature peaks, ranging from low to high, were mainly attributed to hydrogen isotope desorption from the graphite basal plane surface, amorphous carbon surface, unsaturated armchair and zigzag edges, respectively. The interaction between hydrogen isotopes and carbon nanostructures provides deeper insights into the understanding of tritium enrichment on the NG surface, which is primarily attributed to the adsorption of tritium on graphite basal plane-like surfaces at relatively low temperatures. This study provides a practical method to bridge the gap between the atomic binding structure and desorption temperature peaks from a kinetic perspective, providing insights into temperature-dependent hydrogen desorption in typical carbon structures and deepening our understanding of tritium transport behaviors in advanced nuclear energy systems.

Editorial: Complex flow and heat transfer in advanced nuclear energy systems

Editorial: Complex flow and heat transfer in advanced nuclear energy systems

A reduced order sulfuric acid decomposition model for a nuclear-powered hybrid sulfur cycle

A reduced order model of sulfuric acid decomposition within a bayonet chemical reactor was developed to support the U. S. Department of Energy Integrated Energy System program, and address the lack of knowledge in scaling and integration for joint chemical and nuclear processes. Sulfuric acid decomposition within a bayonet reactor was modeled to provide chemical and thermodynamic data relevant to advanced nuclear reactor-driven integrated energy systems based on desired operational scale and operational conditions. The temperature range required for high-temperature advanced nuclear reactor integrated energy systems, 750-850 °C, was shown to produce reasonable agreement (within a few percent relative error) with past models and experimental data, and yielded good efficiency results for bayonet reactor operations. The results of the reduced order model agreed with previous work from Savannah River National Labs within a maximum of 3.4% error on the decomposition of sulfur trioxide, and on previous Hybrid Sulfur flowsheets from Gorensek and Summers that showed operational temperature, pressure, and composition ranges for efficiency which made the Hybrid Sulfur cycle competitive with water electrolysis. The agreement with previous high-fidelity models provided a framework for future Integrated Energy System grid evaluations with an advanced nuclear reactor and large-scale hydrogen production using a mathematical model to represent chemical operations.

The development of nuclear reactor three-dimensional neutronic thermal–hydraulic coupling code: CorTAF-2.0

Numerical reactor technology is the development direction of advanced nuclear energy system simulation, which has the ability of high-fidelity multi-physical field coupling calculation. The thermal–hydraulic and neutron physics fields in reactor core exhibit strong feedback relationships. Establishing the neutronic thermal–hydraulic coupling calculation model of the whole core “pin-by-pin” is the key technology for the development of numerical reactors. In this study, the neutron diffusion analysis model and cross-section database have been developed based on the open source CFD platform OpenFOAM. The three-dimensional neutron diffusion analysis model based on the finite volume method has been validated by benchmark problems. By integrating the neutron diffusion analysis model and cross-section database into nuclear reactor core three-dimensional thermal–hydraulic characteristics analysis code CorTAF, the strongly coupled neutronic thermal–hydraulic coupling analysis code CorTAF-2.0 has been developed. The neutronic thermal–hydraulic coupling phenomena in the full core of AP1000 have been simulated using CorTAF-2.0, yielding the sub-channel level parameters distribution of the whole core. This work offers valuable references for reactor full-core neutronic thermal–hydraulic coupling analysis.

Effects of Alloying Elements and Mechanical Alloying on Characteristics of WVTaTiCr Refractory High-Entropy Alloys.

Refractory high-entropy alloys (RHEAs) are among the promising candidates for the design of structural materials in advanced nuclear energy systems. The effects of Cr, V, Ta, and Ti elements and ball milling on the microstructural evolution and mechanical properties of model RHEAs were investigated. The results show that W-rich BCC1 and Ta-rich BCC2 solid solution phases were generated after a long milling duration. After high-temperature sintering, the (Cr, Ta)-rich phase associated with the Laves phase was observed in the Cr-containing model RHEAs. In addition, a high level of Ti, Ta, and V contents promoted the in situ formation of oxide particles in the alloys. Complex TiTa2O7 and Ta2VO6 oxide phases were identified by TEM, which suggests a solid-state reaction of Ti-O, Ta-O, and V-O subjected to high-energy ball milling. The oxide particles are uniformly dispersed in the BCC matrix, which can result in dispersion strengthening and the enhancement of mechanical properties.

Atomistic simulation of brittle-to-ductile transition in silicon carbide embedded with nano-sized helium bubbles

The tensile response of cubic silicon carbide (SiC) bulk containing cavities (voids and He bubbles) has been investigated using molecular dynamic simulations. The formation of cavities in SiC leads to a significant degradation in the mechanical properties of SiC with more influence on material fracture than initial elastic deformation. The brittle-to-ductile transition occurs in cavity-embedded SiC as the pressure in He bubbles increases. This is associated with the deformation mechanism that bond breaking at a low He bubble pressure transfers to extensive dislocation activities at a higher He bubble pressure. The cavities can effectively concentrate stress around them in the direction perpendicular to the tension, which leads to preferred cracking in the region with a higher tensile stress. The failure mechanism as revealed by this study improves understanding of property degradation in SiC that may be useful for applications of SiC in advanced nuclear energy systems.

Effects of cold rolling and heat treatment on microstructure and mechanical properties of 15Cr-15Ni ODS austenitic steel

Oxide dispersion strengthened (ODS) austenitic steels are considered as the promising candidate cladding materials application for advanced nuclear energy systems. However, the microstructure of ODS austenitic steels fabricated by mechanical alloying is usually not very homogeneous. Thermal mechanical treatment is an effective method for controlling microstructure of alloys. In this work, the effect of cold deformation and subsequent heat treatments on the improvement of microstructure homogeneous and mechanical property of a 15Cr-15Ni ODS austenitic steel was systemically investigated by means of electron backscatter diffraction, transmission electron microscopy and tensile test. Results show that the average grain size decreased and the distribution of grains was more homogeneous after optimized thermal mechanical treatment. The deformation of cold rolling and heat treatment temperatures shows obvious influence on the average size of dispersion particle, the proportion of high angle grain boundary, Schmidt factor and tensile property of 15Cr-15Ni ODS austenitic steel.

He-ion Irradiation Effects on the Microstructures and Mechanical Properties of the Ti-Zr-Hf-V-Ta Low-Activation High-Entropy Alloys.

High-entropy alloys (HEAs) have shown promising potential applications in advanced reactors due to the outstanding mechanical properties and irradiation tolerance at elevated temperatures. In this work, the novel low-activation Ti2ZrHfxV0.5Ta0.2 HEAs were designed and prepared to explore high-performance HEAs under irradiation. The microstructures and mechanical properties of the Ti2ZrHfxV0.5Ta0.2 HEAs before and after irradiation were investigated. The results showed that the unirradiated Ti2ZrHfxV0.5Ta0.2 HEAs displayed a single-phase BCC structure. The yield strength of the Ti2ZrHfxV0.5Ta0.2 HEAs increased gradually with the increase of Hf content without decreasing the plasticity at room and elevated temperatures. After irradiation, no obvious radiation-induced segregations or precipitations were found in the transmission electron microscope results of the representative Ti2ZrHfV0.5Ta0.2 HEA. The size and number density of the He bubbles in the Ti2ZrHfV0.5Ta0.2 HEA increased with the improvement of fluence at 1023 K. At the fluences of 1 × 1016 and 3 × 1016 ions/cm2, the irradiation hardening fractions of the Ti2ZrHfV0.5Ta0.2 HEA were 17.7% and 34.1%, respectively, which were lower than those of most reported conventional low-activation materials at similar He ion irradiation fluences. The Ti2ZrHfV0.5Ta0.2 HEA showed good comprehensive mechanical properties, structural stability, and irradiation hardening resistance at elevated temperatures, making it a promising structural material candidate for advanced nuclear energy systems.

Solving the issue of reliability data for FOAK equipment in an innovative nuclear energy system

There is a demand for the improved safety and economics of advanced nuclear energy systems. The world's first high-temperature gas-cooled reactor pebble-bed module nuclear power plant (HTR-PM Demo Project) is about to be completed in China. This power plant achieved initial full power at the end of 2022. The success of the HTR-PM Demo Project shows the importance of risk-informed initiatives throughout the project with solid support from comprehensive risk assessment. This paper focuses on a technical issue that needs to be solved when conducting risk-informed initiatives for HTR-PM, namely the issue of reliability data for first-of-a-kind (FOAK) equipment. In short, no data for the FOAK equipment are available in existing reliability databases. On the basis of the HTR-PM pilot experience, this paper proposes a process of estimating the required data for the reliability parameters of FOAK equipment. The process has five steps: (1) identification of key parts and assemblies, (2) preliminary determination of the reliability parameters and the associated reference values for the key parts and assemblies, (3) analysis of maintenance management tactics, (4) estimation of reliability parameters for the FOAK equipment using a Bayesian network, and (5) Bayesian updating with operational data. The helium circulator of HTR-PM is taken as an example to demonstrate the feasibility of the proposed process. A lack of reliability data is a common issue for innovative nuclear energy systems. The proposed process is supposed to be universal to all types of innovative design.

In-situ observation of damage evolution in Mo-SiCf/SiC heterogeneous composite

Innovative metal-SiCf/SiC heterogeneous structure composites are promising accident tolerance fuel cladding materials for the Gen IV advanced reactors. The mechanical damage of these composites is a critical issue for their safe application. In this work, the damage process and mechanism of the Mo-SiCf/SiC cladding are revealed by combining In-situ C-ring test and Finite Element Analysis. The results show that the mirco-cracks originate from the pores in the outermost side of the SiCf/SiC layer, and propagate along the fiber bundles boundaries. At the same time, the “bamboo joint”-like damage occurs in the SiCf/SiC layer during C-ring test due to the periodic pore distribution. The fact that the formation of the main crack in SiCf/SiC layer prior to the fracture of Mo layer indicate the Mo-SiCf/SiC cladding can maintain the hermeticity and structure integrity until the sample fails completely. Moreover, the C-ring strength of the Mo-SiCf/SiC cladding is 360 MPa, which is much higher than 201 MPa of SiCf/SiC cladding. Overall, the Mo-SiCf/SiC cladding structure exhibits better crack growth resistance and higher strength compared to SiCf/SiC cladding. Meanwhile this work will provide technical support and theoretical basis for the structural design of cladding materials for advanced nuclear energy systems.

Simulation of steel corrosion and iron yield in LFR with lead coolant

Liquid lead is the main coolant of the fourth generation of advanced nuclear energy system — lead cooled fast reactor (LFR), due to its good neutron economy, high heat transfer performance, stable chemical properties, constant low melting point, high boiling point, etc. Although there are many advantages, the corrosion of metal materials in liquid lead is one of the decisive factors restricting the development of lead cooled fast reactor. In this study, an engineering model for simulating the oxide film growth in liquid lead coolant is established, and the time-dependence of steel flux is analyzed based on the experimental data in the literature. The effects of circuit temperature, hydraulic diameter of section simulated, coolant velocity and steel types on the steel corrosion were investigated. The results showed that the oxide film formed on the steels is of micron grade and the diffusion yield of iron produced in pure lead coolant is of kilogram grade after ten years’ corrosion. The temperature and the coolant velocity have significant effects on the steel corrosion while the effect of hydraulic diameter is mild. Also, the steel types affect the growth of oxide film. The findings provide basic data for the evaluation of radioactive corrosion products of LFR.

Studies on high-temperature stability and strengthening mechanisms of high/medium-entropy alloys for potential nuclear applications: The case of FeCrV-based alloys

High-temperature stability and strengthening mechanisms of high/medium-entropy alloys (HEAs/MEAs) have been studied for Ti-added FeCrV-based MEAs. Experimental results show that FeCrVTix MEAs can maintain good high-temperature stability in chemical element, phase structure and mechanical property. A mixed-strengthening-mechanism model dominated by precipitation and fine-grain strengthening has been proposed by comparing experimental results and model calculations of the yield strength increments for the FeCrV-based MEAs owing to Ti addition. The calculated strength increments with this mixed-strengthening-mechanism model closely match the series experimental results. The excellent compressive fracture strain in the FeCrVTix MEAs mainly comes from crack tip plasticity, and Ti addition not only enhances compressive strain by grain refinement, but also improves compressive strain by crack deflection and bridging. Based on this strengthening model, the content, size and distribution of precipitates can be designed and controlled to regulate and optimize the mechanical properties of HEAs/MEAs. Therefore, this study provides new approaches and ideas for the design and development of HEAs/MEAs with high strength and good ductility for advanced nuclear energy systems.

Point defect properties in high entropy MAX phases from first-principles calculations

MAX phase materials exhibit both metal and ceramic properties due to their unique laminated atomic structures, making them promising candidates in advanced nuclear energy systems. Recently, high entropy MAX (HE-MAX) phases have been developed and attracted much attention due to their unique properties. However, the role of chemical disorder in HE-MAX phases in their point defects properties is still not clear. In this work, we investigated the point defect properties in (TiVNb)2SnC, (TiZrHf)2SnC, (TiVNbZrHf)2SnC, and five corresponding single-component M2SnC phases (M=Ti, V, Nb, Zr, Hf) using first-principles calculations. The average vacancy (VM, VSn, VC) formation energies in the HE-MAX phases are (TiZrHf)2SnC > (TiVNbZrHf)2SnC > (TiVNb)2SnC. With the addition of Zr and Hf atoms, the charge transfer between atoms in the HE-MAX phase increases, hindering the formation of these vacancies. Meanwhile, the obtained migration energies show that the migration barrier of VM through Ti is lower than that of V, Nb, Zr, or Hf, while Zr and Hf atoms increase the VC migration barrier due to their large atomic sizes. Additionally, the formation energies of antisite defects in all three HE-MAX phases are lower than the single-component M2SnC phases, indicating that the HE-MAX phases are more resistant to radiation-induced amorphization. This work provides a fundamental understanding of the effect of chemical disorder on point defect properties in MAX phases and proposes a new strategy for designing novel HE-MAX phases with better performance in nuclear applications.

Editorial for Special Issue on “Intelligent Technologies and Processes for Advanced Nuclear Power and Energy Engineering”

This Special Issue, entitled “Intelligent Technologies and Processes for Advanced Nuclear Power and Energy Engineering”, was organized by the journal Processes as a way to collect original research articles on the latest developments in intelligent technologies and processes for advanced nuclear power and energy systems [...]

Effect of irradiation temperature on radiation hardening of CLF-1 steel

A kind of reduced-activation ferritic-martensitic steel, CLF-1, was developed by the Southwestern Institute of Physics, China, as a candidate structural material for fusion and fourth-generation fission reactors. This work reports the results from the nanoindentation and transmission electron microscopy study performed on this steel after irradiation with 2.75 MeV Fe11+ ions up to 1 and 10 dpa at temperatures of 300 - 500 °C. An increase in hardness was observed in the damaged layer of samples irradiated at 400 °C, while the samples irradiated at 500 °C exhibited a softening of the damaged layer. The hardening can be primarily, although not exclusively, attributed to the increase of <100>-type dislocation loops observed at the increasing irradiation temperature. Nano-sized cavities around the dislocation loops and along dislocation lines were also observed, explaining the lattice swelling after irradiation at 500°C. The research results provide a valuable reference for the development of irradiation-hardening-resistant Fe-Cr alloys for advanced nuclear energy systems.

Effect of dislocation on helium irradiation defects in low-activation martensitic steel

Reduced-activation martensitic steel is a main candidate structural material for key components of advanced nuclear energy systems because of its good mechanical properties at room temperature, low neutron activation characteristics and satisfactory radiation resistance. In this work, we study the interaction between dislocations and helium-irradiation-induced defects in the steel and the effect of dislocations on the behavior of helium migration and desorption. The well-annealed samples are pre-deformed to 10% and 20% reductions in thickness by using cold rolling mill. After pre-deformation, samples are heat-treated to remove vacancy defects and retain dislocation defects. Then, the samples with reserved dislocations are irradiated with helium at room temperature (50 keV, 1×10&lt;sup&gt;17&lt;/sup&gt; He/cm&lt;sup&gt;2&lt;/sup&gt;). After irradiation, the samples are characterized by synchrotron radiation grazing incident X-Ray diffraction, positron annihilation Doppler broadening spectroscopy, and thermal desorption spectroscopy. The results show that dislocations hinder the diffusion of helium and helium-vacancy complexes, and reduce the accumulation of radiation damage. Such an effect becomes more significant with the increase of dislocation density. The BCC → FCC phase transition of low activation martensitic steel occurs at 1179 K. The increase of dislocation density will lead to the forward shift of helium desorption peak induced by phase transition. The retention of helium in the undeformed sample, 10% deformed sample and 20% deformed sample is 10.3%, 15.7% and 17.9%, respectively, indicating that high density dislocations promote the retention of helium.

Effects of He-ion irradiation on the microstructures and mechanical properties of the novel Co-free VxCrFeMnNiy high-entropy alloys

High-entropy alloys (HEAs), as a promising candidate structural material for advanced nuclear energy system, have been widely studied recently due to the combination of the outstanding irradiation tolerance and enhanced mechanical properties at elevated temperatures. In this study, an FCC-structured CrFeMnNi2 HEA and a BCC-structured VCrFeMn HEA without Co element were designed and prepared. The effects of He-ion irradiation on the microstructures and mechanical properties were investigated, so as to preliminarily evaluate the irradiation tolerance of the novel Co-free HEAs. At fluences of 3 × 1016 and 6 × 1016 ions/cm2, the CrFeMnNi2 and VCrFeMn HEAs showed good structural stability at 1023 K; the transmission electron microscope results revealed no amorphization/precipitations. The average sizes and number densities of He bubbles increased as the fluence improved for the two alloys. At each fluence, the average size of the He bubbles in the VCrFeMn HEA was larger than that in the CrFeMnNi2 HEA, while the number density and volume fraction were lower. The CrFeMnNi2 HEA possessed a higher hardening fraction (77%), and an irradiation-induced hardening saturation was observed in the VCrFeMn HEA with a lower hardening fraction of 33%. The experimental results indicated that the novel low-activation VCrFeMn HEA might be a competitive candidate structural material for advanced reactors. The mechanisms of the irradiation tolerance of Co-free HEAs were also discussed.