Search Results (31)

The lead-cooled fast reactor (LFR), a Generation IV nuclear system candidate, presents unique neutronic characteristics distinct from pressurized water reactors. Its neutron spectrum spans wider energy ranges with fast neutron dominance, exhibiting resonance phenomena across energy regions. These features require a fine energy group structure for fuel lattice calculations, significantly increasing computational demands. To balance local heterogeneity modeling with computational efficiency, researchers across the world adopt fuel assembly equivalence methods using 1D cylindrical models through volume equivalence principles. This approach enables detailed energy group calculations in simplified geometries, followed by lattice homogenization for few-group parameter generation, effectively reducing whole-core computational loads. However, limitations emerge when handling strongly heterogeneous components like structural/control rods. This study investigates the 1D equivalence method’s accuracy in lead–bismuth fast reactors under various fuel assembly configurations. Through comprehensive analysis of material distributions and their equivalence impacts, the applicability of the one-dimensional equivalence approach to fuel assemblies of different geometries and material types is analyzed in this paper. The research further proposes corrective solutions for low-accuracy scenarios, enhancing computational method reliability. This paper is significant in its optimization of the physical calculation and analysis process of a new type of fast reactor component and has important engineering application value. Full article

Motivated by the growth of global energy demand and the goal of carbon neutrality, lead-cooled fast reactors, which are core reactor types of fourth-generation nuclear energy systems, have become a global research hotspot due to their advantages of high safety, nuclear fuel breeding capability, and economic efficiency. However, its engineering implementation faces key challenges, such as material compatibility, closed fuel cycles, and irradiation performance of structures. This paper comprehensively reviews the latest progress in the core design of lead-cooled fast reactors in terms of the innovation of nuclear fuel, optimization of coolant, material adaptability, and design of assemblies and core structures. The research findings indicate remarkable innovation trends in the field of lead-cooled fast reactor core design, including optimizing the utilization efficiency of nuclear fuel based on the nitride fuel system and the traveling wave burnup theory, effectively suppressing the corrosion effect of liquid metal through surface modification technology and the development of ceramic matrix composites; replacing the lead-bismuth eutectic system with pure lead coolant to enhance economic efficiency and safety; and significantly enhancing the neutron economy and system integration degree by combining the collaborative design strategy of the open-type assembly structure and control drums. In the future, efforts should be made to overcome the radiation resistance of materials and liquid metal corrosion technology, develop closed fuel cycle systems, and accelerate the commercialization process through international standardization cooperation to provide sustainable clean energy solutions for basic load power supply, high-temperature hydrogen production, ship propulsion, and other fields. Full article

A negative value of the void reactivity coefficient (α_V) is one of the most important passive safety properties for the operation of nuclear reactor. Herein, are presented calculated values of the void reactivity coefficient for different geometries of reactors cooled by liquid lead (LFR) and sodium (SFR) with U-238-Pu-239 and Th-232-U-233 fuels. The calculations were carried out for the reactors filled with either one or two types of fuel assemblies. The most interesting results are obtained for reactor filled with two different types of fuel assemblies (hybrid reactor). Hybrid reactors consist of central and peripheral types of fuel assemblies using low enrichment fuel and high enrichment fuel, respectively. Both hybrid reactors based on the uranium cycle (U-cycle) and the thorium cycle (Th-cycle) can maintain a negative void reactivity coefficient value for wide range of reactor parameters. The calculation results of the hybrid reactor matched those from FBR-IME reactor. Full article

Lead-bismuth eutectic (LBE) is a eutectic alloy of lead (44.5 at%) and bismuth (55.5 at%) that can be used as the coolant for the fast nuclear reactors. In the event of specific conditions or even accidents of the reactors, the temperature of liquid LBE decreases, and it may undergo solidification and volume expansion during the aging process after solidification, which can easily cause damage to the reactor’s internal structure as well as the reactor vessels. In this study, the microstructure and mechanical properties of solidified LBE obtained at different cooling rates are systematically investigated after different aging times. It was found that the internal structure of LBE after aging remained a eutectic microstructure, consisting of the γ-phase (Bi-rich phase) and β-phase (Pb₇Bi₃). After a long period of static aging, the white γ-phase precipitated into the black β-phase, which further confirms the phase transition mechanism. Meanwhile, the acceleration of the cooling rate can aggravate volume expansion. As the aging time increases, there is no significant difference in the compressive yield strength σ of the LBE samples with the same cooling rate and only a certain degree of fluctuation. The elastic modulus E also shows similar results, indicating that aging time has a minor effect on the compressive yield strength σ and elastic modulus E of the LBE. With the increase in cooling rate, the compressive yield strength σ shows an upward trend, while the elastic modulus E is not significantly affected, with a small amplitude of fluctuation. Meanwhile, the hardness of LBE samples after long-term aging treatment is enhanced. After long-term aging, the overall density of the LBE samples shows a decreasing trend, the density fluctuation range of the fast cooling rates (5 K/min and 10 K/min) are significantly larger than that of the slow cooling rates. The decrease in density leads to volume expansion of the LBE during the aging process after solidification. Full article

Alumina-forming austenitic steels (AFA steels) exhibit excellent creep resistance and oxidation capabilities, making them a strong candidate for cladding materials in lead-cooled fast reactors. This study investigates the corrosion resistance of Mn-containing AFA steels in lead–bismuth eutectic (LBE) at 550 °C with a controlled oxygen concentration of 10⁻⁶ wt.%. The results demonstrate that under these experimental conditions, the addition of Al enhances the material’s resistance to lead–bismuth corrosion. Moreover, Mn incorporation significantly improves corrosion resistance, with the optimal composition being an AFA alloy containing 16 wt.% Ni, 12 wt.% Cr, 3 wt.% Al, and 4 wt.% Mn. Mn addition alters the type of oxide product formed on the alloy surface, shifting from Fe₃O₄ or (Fe, Cr)_xO_y to (Cr, Mn)_xO_y. Full article

Radiolysis of water and aqueous solutions refers to the decomposition of water and its solutions under exposure to ionizing radiation, such as γ-rays, X-rays, accelerated particles, or fast neutrons. This exposure leads to the formation of highly reactive species, including free radicals like hydroxyl radicals (^●OH), hydrated electrons (e⁻_aq), and hydrogen atoms (H^●), as well as molecular products like molecular hydrogen (H₂) and hydrogen peroxide (H₂O₂). These species may further react with each other or with solutes in the solution. The yield and behavior of these radiolytic products depend on various factors, including pH, radiation type and energy, dose rate, and the presence of dissolved solutes such as oxygen or ferrous ions, as in the case of the ferrous sulfate (Fricke) dosimeter. Aqueous radiation chemistry has been pivotal for over a century, driving advancements in diverse fields, including nuclear science and technology—particularly in water-cooled reactors—radiobiology, bioradical chemistry, radiotherapy, food preservation, wastewater treatment, and the long-term management of nuclear waste. This field is also vital for understanding radiation effects in space. Full article

With the advancement of lead–bismuth fast reactors, there has been increasing attention directed towards the design of and manufacturing technology for electromagnetic pumps employed to drive liquid lead–bismuth eutectic (LBE). These electromagnetic pumps are characterized by a simple structure, effective sealing, and ease of flow control. They exploit the excellent electrical conductivity of liquid metals, allowing the liquid metal to be propelled by Lorentz forces generated by the traveling magnetic field within the pump. To better understand the performance characteristics of electromagnetic pumps and master the techniques for integrated manufacturing and performance optimization, this study conducted fundamental research, development of key components, and the assembly of the complete pump. Consequently, an annular linear induction pump (ALIP) suitable for liquid lead–bismuth eutectic was developed. Additionally, within the lead–bismuth thermal experimental loop, startup and preheating experiments, performance tests, and flow-head experiments were conducted on this electromagnetic pump. The experimental results demonstrated that the output flow of the electromagnetic pump increased linearly with the input current. When the input current reached 99 A, the loop achieved a maximum flow rate of 8 m³/h. The efficiency of the electromagnetic pump also increased with the input current, with a maximum efficiency of 5.96% during the experiments. Finally, by analyzing the relationship between the flow rate and the pressure difference of the electromagnetic pump, a flow-head model specifically applicable to lead–bismuth electromagnetic pumps was established. Full article

With the increasing global demand for sustainable energy, the importance of advanced nuclear technologies, such as fourth-generation reactors, has become increasingly prominent. Fourth-generation reactors often use non-water coolants, among which liquid lead and lead–bismuth eutectics (LBEs) are highly promising, making lead-cooled fast reactors (LFRs) a popular area of research internationally. On the basis of extensive analysis and comparison conducted previously, this article summarizes and analyzes the advantages, problems, and differences in lead and LBE as LFR coolants. Overall, both lead and LBE have excellent neutronic characteristics, good heat transfer performance, and chemical inertness, and make LFRs highly efficient in nuclear fuel utilization, inherently safe, and relatively simplified in design. However, both of them corrode materials severely and produce highly toxic ²¹⁰Po, which are the problems that need to be considered for further engineering development. Moreover, LBE has a lower melting point, which allows a wider temperature range and lower insulation requirements in its design, making it easier to achieve engineering and miniaturization under the same conditions. Lead has a lower cost, is less corrosive to materials, and produces less ²¹⁰Po, which makes it a more ideal coolant for future development. Full article

Since F/M steel is one of the leading candidate materials for the lead-cooled fast reactor (LFR), its compatibility with the liquid LBE environment is an essential issue before application. One major way to improve LBE corrosion resistance is to control the oxygen concertation in liquid LBE for the growth of a stable, protective oxide layer on the surface of the structure material. However, the influence of the surface state on corrosion behavior is a more realistic issue when it comes to practical applications. In this study, the corrosion behavior of Si-reinforced 9Cr and 11Cr F/M steels with different surface states was investigated by a static liquid LBE corrosion test under solid-phase oxygen-controlled conditions. The result showed that at 550 °C, the coarse surface state caused dissolution behavior at the initial stage of corrosion, while the fine surface state formed the oxide layer. Moreover, at 610 °C, Si-reinforced 11Cr F/M steel shows better liquid LBE corrosion resistance due to its thinner oxide layer formation. Full article

Lead-cooled fast reactors exhibit strong inherent safety performance and good economic features, while material degradation due to corrosion and irradiation is still challenging. Oxide dispersion-strengthened steels are one of the promising candidates for fuel cladding materials. The effects of both irradiation and corrosion on ODS steel need to be further studied. In this work, MX-ODS steel was irradiated by Fe ions at 500 °C up to 46 dpa. Later, the as-received specimen and the irradiated specimen were used to conduct corrosion tests in oxygen-saturated Pb at 550 °C for 1 h. In the as-received specimen, discontinuous oxides penetrated by Pb and Pb in contact with steel matrix were observed, demonstrating unsatisfactory corrosion resistance of the material. However, in the irradiated specimen after corrosion experiment, a protective oxide layer formed and prevented Pb attack. The oxidation behavior differences between the two specimens can be attributed to the defects produced by irradiation and the structural discrepancy in oxides caused by the formation process. A possible mechanism of irradiation on the corrosion is discussed. In the as-received specimen, Fe atoms loss led to voids in the oxides, and lead penetrated the oxides through these voids. In the irradiated specimen, defects left by previous irradiation helped to form a more uniform oxide layer. The adhesive outer magnetite oxide and the Fe ions generated from where grain boundary oxidation developed retarded the presence of voids and made the oxide layer protective. Full article

The accurate calculation of reactor core heating is vital for the design and safety analysis of reactor physics. However, negative KERMA factors may be produced when processing and evaluating libraries of the nuclear data files ENDF/B-VII.1 and ENDF/B-VIII.0 with the NJOY2016 code, and the continuous-energy neutron cross-section library ENDF71x with MCNP also has the same problem. Negative KERMA factors may lead to an unreasonable reactor heating rate. Therefore, it is important to investigate the influence of negative KERMA factors on the calculation of the heating rate. It was also found that negative KERMA factors can be avoided with the CENDL-3.2 library for some nuclides. Many negative KERMA nuclides are found for structural materials; there are many non-fuel regions in fast reactors, and these negative KERMA factors may have a more important impact on the power distribution in non-fuel regions. In this study, the impact of negative KERMA factors on power calculation was analyzed by using the RBEC-M benchmark and replacing the neutron cross-section library containing negative KERMA factors with one containing normal KERMA factors that were generated based on CENDL-3.2. For the RBEC-M benchmark, the deviation in the maximum neutron heating rate between the negative KERMA library and the normal library was 6.46%, and this appeared in the reflector region. In the core region, negative KERMA factors had little influence on the heating rate, and the deviations in the heating rate in most assemblies were within 1% because the heating was mainly caused by fission. However, in the reflector zone, where gamma heating was dominant, the total heating rate varied on account of the gamma heating rate. Therefore, negative KERMA factors for neutrons have little influence on the calculation of fast reactor heating according to the RBEC-M benchmark. Full article

The hydrodynamic and thermal interaction of water with the high-temperature melt of a heavy metal was studied via the Volume-of-Fluid (VOF) method formulated for three immiscible phases (liquid melt, water, and water vapor), with account for phase changes. The VOF method relies on a first-principle description of phase interactions, including drag, heat transfer, and water evaporation, in contrast to multifluid models relying on empirical correlations. The verification of the VOF model implemented in OpenFOAM software was performed by solving one- and two-dimensional reference problems. Water jet penetration into a melt pool was first calculated in two-dimensional problem formulation, and the results were compared with analytical models and empirical correlations available, with emphasis on the effects of jet velocity and diameter. Three-dimensional simulations were performed in geometry, corresponding to known experiments performed in a narrow planar vessel with a semi-circular bottom. The VOF results obtained for water jet impact on molten heavy metal (lead–bismuth eutectic alloy at the temperature 820 K) are here presented for a water temperature of 298 K, jet diameter 6 mm, and jet velocity 6.2 m/s. Development of a cavity filled with a three-phase melt–water–vapor mixture is revealed, including its propagation down to the vessel bottom, with lateral displacement of melt, and subsequent detachment from the bottom due to gravitational settling of melt. The best agreement of predicted cavity depth, velocity, and aspect ratio with experiments (within 10%) was achieved at the stage of downward cavity propagation; at the later stages, the differences increased to about 30%. Adequacy of the numerical mesh containing about 5.6 million cells was demonstrated by comparing the penetration dynamics obtained on a sequence of meshes with the cell size ranging from 180 to 350 µm. Full article

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. Full article

For a sodium-cooled fast reactor, the capability for stable cooling and avoiding re-criticality on the debris bed is essential for achieving in-vessel retention when severe accidents occur. However, an unexploited uncertainty still existed regarding the compound effect of the heterogeneous configuration and dynamic particle redistribution for the debris bed’s criticality and cooling safety assessment. Therefore, this research aims to develop a numerical tool for investigating the effects of the different transformations of the heterogeneous configurations on the debris bed’s criticality/cooling assessment. Based on the newly proposed methodology in this research, via integrating the Discrete Element Method (DEM) with Computational Fluid Dynamics (CFD) and Monte-Carlo-based Neutronics (MCN), the coupled CFD–DEM–MCN solver was constructed with the originally created interface to integrate two existing codes. The effects of the different bed configurations’ transformations on the bed safety assessments were also quantitively confirmed, indicating that the effect of the particle-centralized fissile material had the dominant negative effect on the safety margin of avoiding re-criticality and particle re-melting accidents and had a more evident impact than the net bed-centralized effect. This coupled solver can serve to further assess the debris bed’s safety via a multi-physics simulation approach, leading to safer SFR design concepts. Full article

This study presents the design and computational fluid dynamics (CFD) analysis of a mini demonstrator for a dual fluid reactor (DFR). The DFR is a novel concept currently under investigation. The DFR is characterized by the implementation of two distinct liquid loops dedicated to fuel and coolant. It integrates the principles of molten salt reactors and liquid metal cooled reactors; thus, it operates in a high temperature and fast neutron spectrum, presenting a distinct approach in the field of advanced nuclear reactor design. The mini demonstrator serves as a scaled-down version of the actual reactor, primarily aimed at gaining insights into the CFD analysis intricacies of the reactor while minimizing computational costs. The CFD modeling of the MD intends to add valuable data for the purpose of modeling validation against experiments to be conducted on the MD. These experiments can be used for DFR licensing and design optimization. The coolant and fuel utilized in the mini demonstrator are of low Prandtl number (Pr = 0.01) liquid lead, operating at two distinct inlet temperatures, namely 873 K and 1473 K. The study showed a rapid increase in turbulence due to intense mixing and abrupt changes in flow areas and directions, despite the relatively low inlet velocities. Hot spots characterized by elevated temperatures were identified, analyzed, and justified based on their spatial distribution and flow conditions. Flow swirling within pipes was identified and a remedy approach was suggested. Inconsistent mass flow rates were observed among the fuel pipes, with higher rates observed in the lateral pipes. Although lower fuel temperatures were observed in the lateral pipes, they consistently exhibited higher heat exchange characteristics. The study concludes by giving physical insights into the heat transfer and flow behavior, and proposing design considerations for the dual fluid reactor to enhance structural safety and durability, based on the preliminary analysis conducted. Full article