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In this paper, the self-developed subchannel analysis program SACOS-PB was used to analyze the thermal hydraulic characteristics of 127 rods in horizontal lead-bismuth reactor under steady state and typical ocean transient conditions. The results of the steady-state analysis showed that the coolant flow and temperature of the vertical assembly were symmetrically distributed, the coolant flow of the horizontal assembly gradually increased along the gravity direction, the temperature rose first and then fell, and the peak coolant temperature of the cross-sectional channel was about 365 °C. Three typical ocean transient operating conditions, inclined, undulating and swing, were also carried out. The results showed that the maximum temperature of the coolant at the exit of the assembly increased with the increase of the transverse heeling angle and the decrease of the trim angle after the incline motion changes the gravity component. Under the undulating and swing conditions, the coolant hydrothermal parameters fluctuated periodically, and the parameter fluctuation period is 10s in line with the oceanic conditions. In the above transient conditions, the peak coolant temperature at the assembly outlet increases and exceeds 365 °C. Compared with the pitching motion on the Y-axis, the rolling and tumbling motion on the X-axis had a significant effect on the coolant flow rate and temperature distribution. The calculation results can provide support for the analysis study of the safety characteristics of lead-bismuth power units under marine conditions. [ABSTRACT FROM AUTHOR]

• Empirical correlations are fitted in the helical cruciform seven-rod bundle. • Mixed convection in the laminar flow region is analyzed. • Transverse mixing in inner, side, and corner subchannels is investigated. A novel type of nuclear fuel named helical cruciform fuel may realize self-supporting for adjacent elements and improve heat transfer, both of which contribute to increasing the power density of the reactor. The flow and heat transfer characteristics of a hexagonal helical cruciform seven-rod bundle were experimentally investigated in the present work. 200 sets of isothermal frictional resistance experiments and 199 sets of heat transfer experiments were finished. The laminar flow region (200< Re <1000), the transitional flow region (1000< Re <3000), and the turbulent flow region (3000< Re <16,000) were determined based on the experimental data. Several correlations are compared with the present experimental data. The helical cruciform single-rod correlation underestimates the Nusselt number in the seven-rod bundle for Re <7000, while it has an average relative error of 7.96 % with the resistance experimental data. The empirical correlations to calculate the frictional resistance coefficient and the Nusselt number were fitted. When the Richardson number is greater than 0.1 in the laminar flow region, the Reynolds number is less than 600, and flow is considered to be mixed convection. The temperature distributions of the inner, side, and corner subchannel are analyzed. The temperature of the inner subchannel is higher than the side and corner subchannel. The increase of Reynolds number is beneficial to the fluid transverse mixing between different subchannels. This work gives validation data to support the numerical model in the helical cruciform rod bundle. [ABSTRACT FROM AUTHOR]

In order to deal with the environmental pollution problem caused by excessive emission of greenhouse gases, hydrogen energy, a green low-carbon secondary energy, acts as an important role in the process of energy transformation and the realization of global carbon neutrality goals. With the development of nuclear reactors of Generation IV, there is a new way using the high-temperature residual heat of nuclear reactors to couple with advanced chemical reaction in producing hydrogen. This research based on a small Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR), gives a hydrogen production method derived by biomass. Exergy analysis of this method by Aspen HYSYS shows that the material exergy consumption (including physical exergy consumption and chemical consumption) is the main part affecting exergy losses. The energy utilization of a closed-loop heating system is better than that of an open system that directly heats gasification agents. Specifically, When the temperature reaches 700 °C, the mass flow rate of CO 2 and Biomass is 1:15 and the reaction finally reaches equilibrium, the total exergy consumption of the system is 1.47 × 109 kJ/h and the mole fraction of H 2 is 49.9 % in theory. In addition, the change of temperature mainly affects physical exergy consumption while the change of mass flow rate mainly affects chemical exergy consumption. Furthermore, when the system is coupled with Brayton cycle power generation system, in which a small amount of SCO 2 is drawn out as a high-temperature gasification agent for biomass hydrogen production, the exergy consumption is barely increased. • The concept of a biomass and nuclear hydrogen production system is presented. • Supercritical carbon dioxide acts as a gasification agent and heat-transfer medium. • Detailed exergy analysis on thermal-hydraulic parameters of FuSTAR. • Exergy consumption barely increases when coupled with the Brayton power cycle. • A reference for the optimal design of comprehensive utilization of nuclear energy. [ABSTRACT FROM AUTHOR]

Fluoride-salt-cooled high-temperature reactors (FHRs) are a promising reactor class that combines prominent attributes from the Gen-Ⅳ reactors, raising attractive interests of the international community. However, tritium control has been a major challenge in FHRs because of the neutron reactions of lithium. It's necessary to study the transport characteristics of tritium in the fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) and to provide theoretical support for appropriate tritium control measures and safe operation of the FHRs. In this paper, the inclusive transport characteristics of production, graphite adsorption, diffusion, dissolution, and permeation on tritium in the primary loop are analyzed. The mathematical and physical model of the tritium transport phenomenon is established, and an analysis model of tritium transport characteristics is also developed. Then, the tritium transport characteristics in the FuSTAR primary loop are analyzed. The results show that the tritium in the FuSTAR primary loop mainly exists in the form of Diatomic Tritium (T 2), and it's found that Tritium Fluoride (TF) and T 2 are adsorbed by graphite after being generated in the core, and this process reaches saturation in about 150 Effective Full Power Days (EFPDs). It also indicates that almost 48% of tritium penetrates the secondary loop through the heat exchanger, and only about 3.8% of tritium is adsorbed on the core graphite. TF makes the corrosion of the heat exchanger and downcomer more serious. Finally, the safety of tritium transport in FuSTAR is conservatively estimated, with a tritium emission dose of 0.1674 mSv/a. This study can provide a theoretical reference for the structural design, tritium recycling, and radiation protection of FHRs. • A mathematical and physical model of the tritium transport phenomenon in FHR is established. • Tritium mainly exists in the form of T 2 in the primary loop of FuSTAR. • Tritium penetration and corrosion should be focused on the downcomer and PCHE. • The tritium release dose to the environment of FuSTAR is assessed. [ABSTRACT FROM AUTHOR]

The isothermal steam oxidation experiments of Cr-coated Zr alloy cladding at 1350 ℃ were conducted, and the influence of Cr-Zr eutectic reaction on the oxidation behavior was investigated. The oxidation duration, ranging from 50 s to 3600 s, covers the cases from the initial formation of eutectic liquid to the complete oxidation of the cladding. The microstructures of the coating and the substrate were characterized. The eutectic-oxidation degradation mechanism of the coating is summarized. The migration of the eutectic liquid and the evolution of Cr-rich phases in the post-oxidation eutectic structure are studied and discussed, confirming the process of ZrCr 2 reacting with α-Zr(O) to form Zr 3 Cr 3 O. The degraded Cr coating and the Cr retained in ZrO 2 during the migration of the eutectic liquid can hardly affect the oxidation of cladding substrate. However, the Cr-rich phases, which eventually aggregated in the radial center of the cladding, can reduce the oxidation rate. [Display omitted] • Systematic isothermal steam oxidation experiments of Cr-coated Zr alloy cladding at 1350 ℃ were conducted. • Microstructural evolution mechanism of Cr coating and Zr substrate induced by eutectic and oxidation reactions is summarized. • The formation of Zr 3 Cr 3 O during the oxidation of Cr-Zr eutectic liquid is confirmed. • The oxidation kinetics are summarized and the influence of Cr-rich phases is analyzed. [ABSTRACT FROM AUTHOR]

The relocation mechanisms of molten core have an important effect on the core degradation process which is a complicated coupled process of oxidation, flow, and heat transfer with great uncertainty. However, most relocation experiments are complicated integral-effects bundle experiments, too many factors are involved, and it is difficult to obtain the flow resistance correlation for the molten pool flow in the bundle channel. So in this paper, a separate-effects visualization experiment study on the flow mechanism of the molten pool relocation was conducted by using glycerol-water solutions at room temperature with a 2 × 2 bundle channel. The flow front of glycerol-water solutions can be captured by the high-speed camera via a transparent plexiglass shroud. And a new flow resistance correlation of the bundle channel for the molten pool relocation is obtained. In addition, the influence of flowing gas on the relocation process of the molten pool is also investigated. This study is supposed to provide valuable references for the calculation model of the molten pool relocation in severe accident analysis codes. • A separate-effects visualization experiment study on the flow mechanism of the molten pool relocation was conducted in room temperature. • A new flow resistance correlation of the bundle channel for the molten pool relocation is obtained. • The influence of flowing gas on the relocation process of the molten pool is investigated. [ABSTRACT FROM AUTHOR]

• This study utilizes the increasing flow regime experimental data to extend the prediction range and improve the accuracy. • Integrated flow regime prediction models considering vertical and horizontal flow direction were developed and evaluated. • The MLP, RF, and SVM algorithms were utilized to develop the flow regime prediction model respectively. • The flow regime prediction model shows better performance than existing models in three directions respectively. The nuclear reactor core is the pivotal component in a nuclear power plant to generate and transfer heat, so accurate prediction of reactor core flow and heat transfer characteristics is a crucial problem for the reactor system design. Flow regime is closely related to the thermal–hydraulic characteristics of two-phase fluid. Still, flow regime models used in thermal–hydraulic calculation codes rely heavily on earlier experimental data, featuring a narrow application range and unsatisfactory accuracy. To extend the prediction range and improve the accuracy of flow regime prediction by making full use of the increasing experimental data, integrated flow regime prediction models considering vertical flow direction for light water reactor (LWR) and horizontal flow direction for Canadian Natural Deuterium Uranium (CANDU) were developed and evaluated based on three standard machine learning algorithms in this study. Firstly, the experimental data of horizontal and vertical flow collected from the literature was modified and preprocessed to be training data. Then, the multi-layer perceptron (MLP) algorithm, random forest (RF) algorithm, and support vector machine (SVM) algorithm were utilized to develop the flow regime prediction model respectively. The performance of the three models was evaluated and compared and the results showed that the flow regime prediction model based on RF was the optimal model with higher prediction and more efficiency than the other two models. Finally, the flow regime prediction model was compared with existing models in three directions respectively, which indicated that the range and accuracy of the model based on RF were superior to the existing models. This study provides an implantable and scalable approach for flow regime prediction, and the application range and accuracy of flow regime prediction can be continuously improved with updating training data. [ABSTRACT FROM AUTHOR]

• The transient response of FuSTAR in five design basis accidents are analyzed with the transient simulation code FUSY. • The sensitivity analysis is performed for some key parameters that may affect the transient behavior. • FuSTAR has a great ability to survive in URIA and ULOF. • The maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. FuSTAR is a small, modular fluoride-salt-cooled high-temperature advanced reactor, proposed by Xi'an Jiaotong University (XJTU), aiming at fulfilling the requirements of Gen IV nuclear system and meanwhile ensuring the inherent safety and economics. In order to evaluate the safety performance of FuSTAR, five typical design basis accidents are selected for safety analysis with different initial events, which are unprotected reactivity insertion accident (URIA), unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected overcooling (UOC), and the accident combing ULOF with ULOHS. The sensitivity analysis is also performed. FuSTAR is modeled by modified system analysis code and the simulations of the steady-state are verified by the results with CFD method. The results of transient accidents indicate that safety of FuSTAR has been guaranteed in URIA and ULOF. Meanwhile, the maximum limit of external parameters causing ULOHS and UOC are obtained by sensitivity analysis. In the ULOF and ULOHS, the peak of fuel temperature, cladding surface temperature and core outlet temperature of coolant are below their safety limits. Therefore, FuSTAR exhibits acceptable safety under design basis accidents. This study will provide supports in verifying the rationality of the design and the subsequent optimization of FuSTAR. [ABSTRACT FROM AUTHOR]

Spray cooling is an efficient cooling technology with high heat dissipation, and has been extensively utilized in various industry areas. The interaction mechanism between the spray droplet and environmental space with high pressure and temperature is extremely complex, especially with non-condensable gas. In this paper, a pseudopotential multicomponent multiphase lattice Boltzmann model for droplet migration with non-condensable gas in high temperature and pressure was proposed. The mechanism of transfer characteristics for droplet in steam were studied. It is found that the mass transfer of droplet is controlled by density and condensation, droplet growth is mainly impacted by density at initial stage, and in later period, the growth is greatly weakened by the condensation suppression of non-condensable gas. The calculated results show that when Fo > 0.4, mass transfer is significantly decreased. Heat transfer is also affected by non-condensable gas, and 2% gas declines the droplet heat transfer rate by about 10%. It is also found that the mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content inside and outside the droplet. The results can be used for development of spray cooling model, and have positive guiding significance for optimizing spray cooling system. • A heat and mass transfer model of droplet and steam with non-condensable is established. • The mechanism of transfer characteristics for droplet in steam were studied. • The mass transfer of droplet is controlled by density and condensation in different stage. • The mass transfer rate is approximately linearly associated with the ratio of non-condensable gas content. [ABSTRACT FROM AUTHOR]

Fuel pins assembled in Pressurized Water Reactor (PWR) and Liquid Metal-cooled Fast Reactor (LMFR) core are usually cooled by single-phase coolant such as water or liquid metal. Researches on the coolant flow and heat transfer phenomena through rod bundles is quite essential for the nuclear reactor core design. This paper briefly summaries the current developments of Computational Fluid Dynamics (CFD) studies on the thermal hydraulic analysis of single-phase flow in the rod bundles of nuclear reactor core. On account of the tremendous amount of publications related to each aspect above, this brief review would focus on the most representative researches in the past decades. The aim of this paper is to provide an outline and present a comprehensive overview of the CFD research and guidelines on single-phase flow through bare rod bundle rod with mixing vane spacer and wrapped wire spacer in reactor core. In addition, the analysis and validation methods for various CFD calculating methods are concluded. The applications of CFD approach with different type of rod bundles are summarized. Finally, some remaining challenges for future work are proposed and discussed. [ABSTRACT FROM AUTHOR]

In this study, we investigate the steam oxidation behaviors of molten zirconium alloy at exceptionally high temperatures of 1870 ℃ and 1970 ℃. A specialized high-temperature steam oxidation apparatus was established to conduct the steam oxidation experiment. The experiments yielded crucial data including oxidation weight gain results, macroscopic appearance of the oxide surface, and section morphology. Several new phenomena: α-Zr(O) precipitates and two-layer oxide structure, were observed in this study. Based on these observations, a complete mechanism of molten zirconium oxidation in steam was developed, providing an interpretive framework for the oxidation observed in this experiment. This newly proposed mechanism reevaluates the formation process of nonceramic precipitates in the molten bulk and its consequential impact on oxidation kinetics at temperatures below 2000 ℃. • High-temperature steam oxidation behaviors of molten zirconium alloy at 1870℃ and 1970℃ were studied. • The oxidation mechanism of molten zirconium was discussed. • The enhanced effect of the melt on the oxidation kinetics was analyzed. [ABSTRACT FROM AUTHOR]

• Two coupled parallel frameworks for shell and tube heat exchangers are proposed. • Fully three-dimensional parallel computation on both sides of the heat exchanger is realized. • One-to-one method achieves 85% parallel efficiency in 14 cores computation. • One-to-many method improves computational efficiency by about 20 times over serial in 28 cores computation. Shell and tube heat exchanger is widely used as a common type of heat exchanger. When the size and the number of heat exchanger tubes are enormous, it is difficult to use the direct modeling method for numerical simulation of heat exchangers. The simplified modeling method based on porous media utilizing user-defined function (UDF) has been widely applied. However, when the three-dimensional (3D) modeling is required on the tube side, the parallelization problem still restricts the computational efficiency of numerical simulation. In this paper, a one-to-one parallel method was first developed to solve the problem of coupled parameter transfer between the primary and secondary sides in the parallel framework, which was tested to achieve a parallel efficiency of 85% under 14 cores. However, the method acquires complicated data transfer between nodes, large memory occupation, and the modeling of the tube side is different from the actual structure under certain circumstances. To solve the problems, a one-to-many parallel method is proposed, and the computational efficiency is tested to be enhanced by about 20 times under 28 cores compared with single core, with a parallel efficiency of 70%. The two methods proposed in this paper are of great significance to the parallel transformation of the future 3D full-size code for shell and tube heat exchangers development and the calculation efficiency improvement. [ABSTRACT FROM AUTHOR]

• A Hi-fi space thermionic nuclear reactor system model is established, and its dynamic performance is obtained. • A model predictive controller is designed for the space reactor system using discrete state-space model and Kalman filter. • Several transient scenarios were designed to evaluate the control performance of the model predictive controller applied to space reactors from various perspectives. • Stable and rapid control of the space reactor has been achieved. Space nuclear reactors possess the inherent benefits of independence from solar energy and the capability to withstand intricate external circumstances. Distinguishing themselves from terrestrial reactors, space nuclear reactors function within the realm of outer space, operating autonomously throughout their entire operational cycles. Consequently, the establishment of a dependable automatic control system assumes utmost significance. In this paper, a high-fidelity nonlinear space thermionic nuclear reactor model is established, encompassing neutron physics, thermal–hydraulic, and thermoelectric conversion system models. An automatic control system with a model predictive controller (MPC) is designed based on the system. A simplified state-space model is established based on system identification as the internal model for the MPC controller, demonstrating good agreement with the dynamic response of the nonlinear system model. The deployment of the MPC controller for system control simulations exhibits excellent control performance under various operating conditions. In the case of a nuclear power step response, the MPC controller achieves an overshoot of 0.07 % and a settling time of 0.65 s. For an electric power step response, the MPC controller yields an overshoot of 0.28 % and a settling time of 36.6 s. These control performances are significantly better than those achieved by the PID controller. In power tracking control, the MPC controller demonstrates virtually error-free tracking performance. When external reactivity disturbances are introduced, the MPC controller outperforms the PID controller in swiftly compensating for the disturbances, maintaining the system's operating state at the desired level. [ABSTRACT FROM AUTHOR]

A deterministic neutron transport solver with SP 3 method based on OpenFOAM has been developed and validated in this paper. The physical models, numerical methods and schemes, boundary conditions, and solving strategy are discussed in detail. The approximate Marshak boundary condition of this solver is validated by the ISSA benchmark problem. Pin-by-pin lattice and simplified core neutron transport benchmark problems are introduced to make the validation of the solver. Numerical results show that SP 3 solver developed based on OpenFOAM has the promising prediction accuracy of k e f f , neutron scalar flux, fission power, and control rod worth. Parallel tests of SP 3 solver are also carried out under the million scale meshes, and the solver has an acceptable acceleration ratio and efficiency. No new computational methods or numerical schemes are proposed in this paper. Instead, the main new works in this paper are our detailed and quantified numerical validations of serval international neutron transport benchmark problems for SP 3 solver based on OpenFOAM. The capability of OpenFOAM in performing SP 3 calculations has been further tested and validated in this paper, which gives additional confidence that OpenFOAM can effectively be used for SP 3 analysis. • Detailed and quantified numerical validations of SP 3 solver based on OpenFOAM. • This study shows the accuracy of FV-based SP 3 method in specific benchmarks. • Additional confidence of SP 3 analysis based on OpenFOAM has been enhanced. [ABSTRACT FROM AUTHOR]

• The basic theory of improved quasi-static methodhas been studied in detail. • Provide the derivation of the first-order perturbation theory of neutron diffusion equation and kinetics parameters. • Implementing this method into the state-of-the-art FVM framework – OpenFOAM, and elaborating on the numerical details. To analyze the transient problems for neutronics and make the preparation for high-fidelity multi-physics simulation. In this study, a neutron space–time kinetics solver called "KaiyangIQS" with the improved quasi-static method based on OpenFOAM is developed. The basic principles of improved quasi-static method, numerical method, and solver strategy are derived and discussed in detail. Separate validations for the modules of solving amplitude function and adjoint scalar flux diffusion equation are carried out. The results show that OpenFOAM can effectively handle the stiff matrix of amplitude equations with high accuracy and robustness. The adjoint scalar flux diffusion equation solver is also validated by the IAEA 3D benchmark. For the improved quasi-static method solver, unidimensional rod ejection problem, TWIGL benchmark problem, and three-dimensional rod ejection problem are carried out for calculation and validation. Normalized power prediction, reactivity prediction, effect of adopting different discretization schemes of transient terms, rod cusping effect, and lag effect of delayed neutron precursors are analyzed in detail. The numerical results of the solver with improved quasi-static method that developed in this study are in good agreement with references, and this solver can be considered a promising tool to simulate the transient neutron kinetics problems. [ABSTRACT FROM AUTHOR]

Pressurizer is widely used in the industrial systems, which can maintain the pressure of the system in a certain range. Especially, as the system pressure rises, the spray droplets interacts with the saturated steam in pressurizer to reduce system pressure. In this paper, a heat and mass transfer model of droplet and steam in pressurizer is established through the lattice Boltzmann method (LBM), and the characteristics of moving droplet in steam space with different Reynolds numbers are analyzed. In this paper, these results show that the flow fields on both sides of the droplet have obvious symmetry at low Reynolds number. As the Reynolds number increasing, the wake region of flow field for moving droplets gradually loses symmetry. When the Reynolds number reaches 103 orders of magnitude, the droplet motion begins to appear horizontal deviation. The growth of droplet condensation radius is inhibited with the increase of Reynolds number. When Reynolds number reaches 2600 and Fo = 0.2, the growth of droplet radius stops and then decreases. Due to the influence of inertia force generated by internal circulation flow and droplet movement, as the Reynolds number increasing, the droplet heat transfer rate decreases at first, but begins increasing when Reynolds number raises to about 2000. The calculation results have reference value for the development of spray droplet model of pressurizer. • A heat and mass transfer model of droplet and steam in pressurizer is established. • The characteristics of moving droplet in steam under different Re are analyzed. • With the increase of Re , the growth of droplet condensation is inhibited. • Heat transfer rate decreases first and then increases with the increase of Re. [ABSTRACT FROM AUTHOR]

The conceptual design of China Fusion Engineering Test Reactor (CFETR) has been completed and is currently in the engineering design phase. In this paper, focusing on three categories of candidate blanket designs and divertor designs, the selection schemes for materials of plasma-facing components, solid structure, tritium breeding materials, neutron multiplier, coolant, and divertor were extracted. Thermo-physical property data was compiled to form a material thermo-physical property database, named Thermo-physical Property Database (ThePD), which is the first advanced thermo-physical database for materials of CFETR blankets. Besides, the callable code was generated, which can be directly applied in the system code for the blanket. Based on internationally available database from National Institute of Standards and Technology (NIST) and similar software AP1700, the correctness of the data and the transitional analysis of regions designed by International Association for the Properties of Water and Steam (IAPWS) were verified. Using the thermal-hydraulic system analysis code, integrated validation was performed for the validated thermo-physical property data and code, demonstrating the applicability and accuracy of the data. ThePD is conducive to addressing the issue of incomplete thermo-physical property data faced by blanket design and safety analysis software for CFETR, and it can fill this gap. [ABSTRACT FROM AUTHOR]

Based on the analysis of energy structure and the development trends of advanced nuclear energy technology, combining the complementary advantages of nuclear energy and other energy sources in China, the Fluoride-Salt-cooled high-Temperature Advanced Reactor (FuSTAR) system has been proposed. FuSTAR system is designed to address production capacity issues in areas with energy shortages or resource concentration and can be combined with high-temperature process heat to achieve high-value-added processes such as hydrogen production. In this paper, the Tri-structural isotropic fuel combined with helical cruciform elements has been studied in detail on the neutron physics and thermal-hydraulic properties, and the flow and heat transfer experiment is carried out for the first time. The multilayer nonlinear programming method has been applied to the design of thermal transport systems. Then the optimal parameters and configuration of the thermoelectric conversion system are obtained by the superstructure optimization method. The inherent safety of FuSTAR is demonstrated through deterministic safety analysis. In addition, the iodine-sulfur cycle technology and the characteristics of tritium production are also analyzed. This study highlights the significant potential of integrating nuclear energy with other energy sources in an optimized energy-production system. • Multi-purpose reactor energy system for power - hydrogen - tritium production. • Over 50 % maximum thermal efficiency. • Simultaneous analysis of design, optimization, and safety. [ABSTRACT FROM AUTHOR]

• A conceptual design of FuSTAR is developed, and validated by neutronics and thermal-hydraulics analysis. • Detailed core design and neutronics analysis with near-critical configuration and burnup calculation are carried out. • Detailed thermal-hydraulics analysis with turbulent flow and heat transfer is carried out. The fluoride-salt-cooled high-temperature advanced reactor (FuSTAR) is an advanced small modular reactor, which aims to the electric supplement and clean heat production to western China. In order to have good economy and safety, FuSTAR needs to have the characteristics of small size, light weight, deep burnup, and inherent safety. A conceptual design of FuSTAR is proposed in this study with neutronics and thermal-hydraulics assessments. In this study, its concept is established with two-region fuel to flatten core power, and the neutron spectrum of the core is in the thermal spectrum to improve the utilization of thermal neutrons and reduce fuel inventory. Neutronics and thermal-hydraulics analyses are also carried out, and established the core layout and material ratio under the near-critical condition. Burnup depth of the fuel, power distribution, neutron flux distribution, flow and heat transfer in rod-bundles channel, and 1/6 core porous media analysis are also investigated. The parameters of the designed core are evaluated quantitatively from various aspects. The results show that the core design in this study has sufficient shutdown depth, the equilibrium cycle is 3.24 years, and the discharge burnup depth is 106 GWd/MtU. The helical cruciform fuel (HCF) adopted by this reactor has a better flow and heat transfer capacity than cylindrical-type fuel at the inner channel. The analysis of 1/6 core porous media shows that the peak fuel temperature under normal operating condition is 1009.3 K, which is far below the limit of 1573 K. Relevant design and analysis results can provide the reference for the subsequent small modular FHR design and construction. [ABSTRACT FROM AUTHOR]

• Tie tube elements greatly improve temperature safety. • The 2:1 support ratio is still the best choice. • Deposition energy cannot be ignored in the SNTP system. • The issue of sudden temperature rise on the outer tie tube is worth noting. As a main structure, tie-tube elements are used both in the traditional and modern LEU nuclear thermal propulsion reactors. This complex structure and its corresponding complex convective heat transfer process have been studied, including CFD calculations. But these studies did not effectively evaluate the applicability of the models and typical problems. In this study, a general small nuclear thermal propulsion (SNTP) reactor model is established by using three-dimensional Monte Carlo code and commercial CFD program Fluent. The applicabilities of material properties, turbulence models and grids are clarified by analyzing various influencing factors. The calculation results show that due to the small enthalpy difference and heat transfer, the convection effect of inner flow based on hydrogen can be ignored. There is no factor that can significantly reduce the maximum fuel temperature, but the porosity of the insulator can significantly improve the heat transfer ratio. In terms of thermal safety and overall performance, 2:1 support ratio is still the best choice. Based on the simulation results, the phenomenon of steep temperature rise of outer tie-tubes is found, and the SST k-omega model with Y + <3 is recommended for simulation of similar problems. [ABSTRACT FROM AUTHOR]

The oxidation and degradation mechanisms of Cr coating on Zr alloy cladding under high temperature steam atmosphere are summarized. Six basic phenomena are considered for model development, including: oxidation of Cr, growth of ZrCr 2 , formation of Zr(Cr) solid solution, outward diffusion of Zr, reduction of Cr 2 O 3 , and volatilization of Cr 2 O 3. The mathematical analysis model is established and validated, demonstrating its good capacity to predict the increase and subsequent decrease in the thickness of Cr 2 O 3 scale, as well as the thickness changes in other individual layers. The developed model is applied in the analysis of the structure evolution of Cr coating under steady state, transient heating conditions and postulated accident scenarios. [ABSTRACT FROM AUTHOR]

• Sliding mesh method was used to simulate moving conditions. • Comparing the numerical results with the experimental results. • Two calculation methods of simulating moving conditions were compared. • It is concluded that the effect of rolling motion is the most sever. • Secondary flows and flow instabilities under moving conditions were analyzed. The investigation of flow and heat transfer of lead-bismuth eutectic (LBE) under moving conditions is vital to the design optimization of LBE loop system in offshore engineering. In this paper, we have developed a method for simulating moving conditions using sliding mesh. Moreover, we have conducted experiments to validate the numerical model, and the comparison results demonstrate a good agreement between them. The calculation results demonstrate that the period of the flow rate fluctuation is consistent with the motion period, and the amplitude of the fluctuation will increase with the increase of the maximum range of motion and the decrease of the motion period. In addition, the influence of rolling on system parameters is more severe than that of heaving. These results obtained in this study have important engineering application value for the design of LBE loop system under moving conditions. [ABSTRACT FROM AUTHOR]

• Subchannel code for the thermal hydraulic analysis of liquid metal cooled reactor. • Simulation of transition from forced to natural circulation by subchannel code. • Effect of non-uniform heating on the thermal field. • Validation of YAOGUANG code under steady and transient conditions for LBE coolant. Based on the experimental campaign performed on the NACIE-UP (NAtural Circulation Experiment-Upgraded) facility, the benchmark activity was proposed by ENEA to qualify and validate the numerical tools for the Heavy Liquid Metal (HLM) system. Xi'an Jiaotong University participated in the benchmark exercise of subchannel validation and the self-developed subchannel code YAOGUANG was applied to the thermal hydraulic analysis. The performed two tests with different power distributions on the Fuel Pin bundle Simulator (FPS) were simulated, characterized by transition from forced to natural circulation. The applicability of this code to the stable state and mass flow rate transient was evaluated by comparing the numerical outcomes with the experimental results. Then the effect of the non-uniform heating on the thermal field was analyzed. The results indicate that this subchannel code has good accuracy in simulating both steady state and transient conditions with uniform power distribution. For the non-uniform case, the subchannel calculations can capture the general trend of temperature distributions, with a larger simulation deviation observed. This discrepancy is mainly due to the limitations of existing models adopted because they didn't consider the impacts of non-uniform conditions on the flow and heat transfer in rod bundles. Finally, the influence of non-uniform heating is mainly manifested as the increase of axial and radial temperature gradients and it helps to flatten the radial thermal field to some degree. [ABSTRACT FROM AUTHOR]

Thermal stratification is an important phenomenon in the process of natural circulation decay heat removal in sodium-cooled fast reactors (SFR). It affects the establishment of the natural circulation and the flow rate of the natural circulation in a station black out (SBO) accident, and therefore affects the safety of the reactor. In safety evaluation, programs are often required to perform analysis for various conditions and long-term transients. However, it is difficult to meet this requirement by adopting a multi-dimensional method or one-dimensional and multi-dimensional coupling method. Therefore, this paper proposes applying a thermal stratification model to the system analysis program THACS to analyze the thermal stratification phenomenon. First, the thermal stratification model is validated with experimental data. The validated model is applied to analyze a station black out accident in the SFR. The lumped parameter model and thermal stratification model are used respectively for the hot pool. By comparing the calculation results of the two models, the effect of thermal stratification of the hot pool on key parameters such as the natural circulation flow rate of the primary circuit, natural circulation flow rate in the inter-wrapper region, average core outlet temperature, maximum core fluid temperature, and the maximum temperature of cladding was analyzed, and the effect of thermal stratification on natural circulation decay heat removal of the SFR was evaluated. This work can provide a reference for the safety analysis of the reactor and contribute to developing a more conservative safety analysis and evaluation strategy. [ABSTRACT FROM AUTHOR]

• GeN-Foam has been extended to predict the occurrence of DNB. • CHF look-up table and post-CHF flow and heat transfer models have been developed. • OECD/NRC PSBT benchmark (Phase II) is used to preliminarily validate the new solver. • Fluid temperatures in each sub-channel resulted within ±10 °C with respect to experimental values. • The predicted DNB power and time of occurrence fell within a ±10 % error range of the experimental values for all cases considered in this paper. The prediction of departure from nucleate boiling (DNB) has always been a crucial aspect of thermal-hydraulic codes for the analysis of Light Water Reactors. In this paper, GeN-Foam, a multi-physics code developed based on OpenFOAM, has been enhanced to incorporate DNB prediction using a CHF look-up table as well as post-CHF flow and heat transfer models. To assess GeN-Foam's accuracy in modeling DNB conditions in typical pressurized water reactors (PWRs), a preliminary validation was conducted utilizing Phase II of the PSBT benchmark, including the steady-state fluid temperature (Test Series 1), steady-state DNB (Test Series 4), and transient DNB (Test Series 11T) benchmarks. This paper presents the closure correlations implemented in GeN-Foam, provides a description of the PSBT benchmarks, and includes a comparison between GeN-Foam's predictions and experimental data, as well as code-to-code verification with other participants. The results demonstrate that GeN-Foam exhibits good performance in simulating two-phase flow boiling conditions. Furthermore, GeN-Foam accurately predicts the DNB power and time of occurrence within a ±10 % error range compared to measured values, preliminarily indicating its effectiveness in predicting DNB occurrence and two-phase flow boiling phenomena. [ABSTRACT FROM AUTHOR]

The threshold displacement energy of Fe and Fe-Cr-Al alloys and the influence factors were investigated using molecular dynamic simulations. The mechanism of defect recombination was discussed, as was the evaluation of the recombination radius for Fe and Fe-Cr-Al. The simulation results show that Fe and Fe-Cr-Al have big differences in threshold displacement energies and directional dependencies at low temperatures, but the differences become less pronounced at high temperatures. The cause might be that the replacement collision sequence can effectively separate an interstitial from a vacancy in Fe-Cr-Al and greatly reduce threshold displacement energy at low temperatures. In addition, the energy transfer efficiency of primary knock-on Al atoms is low due to the difference in atomic mass, and the recombination effect for Cr is hindered by the formation of Fe-Cr 〈110〉 dumbbell. However, the effect of solute atoms on threshold displacement energy is insignificant at high temperatures, due to the increase in the recombination radius and a reduction in the effect of the aforementioned factors. [ABSTRACT FROM AUTHOR]

Numerical simulation analyses of LIFUS5/MOD2 experiment facility for hyperbaric steam-water jet inside liquid lead bismuth eutectic (LBE) were conducted using computational fluid dynamics (CFD). Alchagirov surface tension correlation and Lee evaporation model were applied in Eulerian-Eulerian two-phase flow model. Benchmark study was performed by comparing the CFD results against experimental data. Four thermal-hydraulics phenomena of sensitivity analysis were established under three extended conditions. These include 1) argon zone compression risk, 2) pressure rising and initial pressure peak, 3) argon entrained by LBE fluctuation, and 4) LBE level rising and steam cavity diffusion path. Overall, CFD models have a great advantage simulating pressure change and pressure wave transfer with 0.37% maximum error. LBE level rising and the radial and even diffusion of steam cavity should happen simultaneously under steam generator tube rupture accident (SGTR). This study provides the engineering basis for Lead-cooled Fast Reactor (LFR) design to improve its safety margin. [ABSTRACT FROM AUTHOR]

MYRRHA is the Gen IV heavy liquid metal cooled fast reactor (LMFR), which is a pool-type reactor system developed and designed by the Belgian Nuclear Research Center SCK·CEN. The primary cooling system of the MYRRHA operates with molten lead-bismuth eutectic (LBE) as the working fluid, which offers several advantages. The equipment failures of fourth generation nuclear facilities will cause serious economic and environmental consequences, especially when flow blockage occurs, the threat to reactor safety is enormous. Rapidly detecting the occurrence of flow blockage and initiating shutdown procedures can effectively reduce the consequences of the accident. In this paper, a coupled sub-channel/CFD multi-scale platform is developed by implementing the sub-channel code SACOS-PB to the CFD code FLUENT based on the domine decomposition and explicit numerical scheme. The flow characteristics and temperature distribution in the upper plenum of a lead-cooled fast reactor(LFR) design based on MYRRHA reactor were investigated. The temperature at the core outlet increases 110K under the flow blockage condition, and the monitoring point for flow blockage is most suitable to be arranged in the center of the assembly and below the height of 0.6 m. The analysis results well demonstrated the 3D flow and local phenomena of the fluid in the upper plenum, which is significant to reactor design and reactor safety analysis of LFRs. [ABSTRACT FROM AUTHOR]