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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]

Background: The understanding of the factors causing decreased overall survival (OS) in older patients compared to younger patients in lung adenocarcinoma (LUAD) remains. Methods: Gene expression profiles of LUAD were obtained from publicly available databases by Kaplan‐Meier analysis was performed to determine whether age was associated with patient OS. The immune cell composition in the tumor microenvironment (TME) was evaluated using CIBERSORT. The fraction of stromal and immune cells in tumor samples were also using assessed using multiple tools including ESTIMATE, EPIC, and TIMER. Differentially expressed genes (DEGs) from the RNA‐Seq data that were associated with age and immune cell composition were identified using the R package DEGseq. A 22‐gene signature composed of DEGs associated with age and immune cell composition that predicted OS were constructed using Least Absolute Shrinkage and Selection Operator (LASSO). Results: In The Cancer Genome Atlas (TCGA)‐LUAD dataset, we found that younger patients (≤70) had a significant better OS compared to older patients (>70). In addition, older patients had significantly higher expression of immune checkpoint proteins including inhibitory T cell receptors and their ligands. Moreover, analyses using multiple bioinformatics tools showed increased immune infiltration, including CD4+ T cells, in older patients compared to younger patients. We identified a panel of genes differentially expressed between patients >70 years compared to those ≤70 years, as well as between patients with high or low immune scores and selected 84 common genes to construct a prognostic gene signature. A risk score calculated based on 22 genes selected by LASSO predicted 1, 3, and 5‐year OS, with an area under the curve (AUC) of 0.72, 0.72, 0.69, receptively, in TCGA‐LUAD dataset and an independent validation dataset available from the European Genome‐phenome Archive (EGA). Conclusion: Our results demonstrate that age contributes to OS of LUAD patients atleast in part through its association with immune infiltration in the TME. [ABSTRACT FROM AUTHOR]

Ethanol, also known as alcohol, is one of the most common drinks in the world. Chronic ethanol exposure has been reported to induce mental disorders. Ethanol also has a strong effect on the gut microbiota. The gut microbiota has been reported to affect the brain via multiple pathways, including changes in γ-aminobutyric acid (GABA) system, and cause a variety of mental disorders. The GABA system in the cortex is associated with anxiety. However, the role of gut microbiota played in ethanol exposure-induced changes in the GABA system and anxiety is still not clear. We established a 30-day ethanol exposure mouse model and investigated the effects of microbiota using the antibiotic minocycline. Minocycline alleviated ethanol-induced anxiety-like behaviour, dysbiosis of microbiota, intestinal barrier disruption, increased serum endotoxin and interleukin (IL)-6. Minocycline also attenuated ethanol-induced apoptosis and decreased expression of glutamate decarboxylases (GADs) and GABRA1 in the prefrontal cortex. Our results indicated that gut microbiota plays an important role in ethanol-induced anxiety-like behaviour by altering the function of GABA system. In addition, causal mediation analysis showed that endotoxin and IL-6 may mediate the connection between the gut microbiota and the expression of GABAA receptor in the prefrontal cortex. [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]

• 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]

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]

Train station delay prediction is always one of the core research issues in high-speed railway dispatching. Reliable prediction of station delay can help dispatchers to accurately estimate the train operation status and make reasonable dispatching decisions to improve the operation and service quality of rail transit. The delay of one station is affected by many factors, such as spatiotemporal factor, speed limitation or suspension caused by strong wind or bad weather, and high passenger flow caused by major holiday. But previous studies have not fully combined the spatiotemporal characteristics of station delay and the impact of external factors. This paper makes good use of the train operation data, proposes the multiattention mechanism to capture the spatiotemporal characteristics of train operation data and process the external factors, and establishes a Multiattention Train Station Delay Graph Convolution Network (MATGCN) model to predict the train delay at high-speed railway stations, so as to provide references for train dispatching and emergency plan. This paper uses real train operation data coming from China high-speed railway network to prove that our model is superior to ANN, SVR, LSTM, RF, and TSTGCN models in the prediction effect of MAE, RMSE, and MAPE. [ABSTRACT FROM AUTHOR]

• The influence of geometric parameters including twist pitch and cross section sizes on friction factor and heat transfer coefficient is studied in the helical cruciform single rod. • Geometric dimensionless numbers are proposed to fit the friction factor and heat transfer coefficient of helical cruciform. • The proposed empirical correlations of frction factor and heat transfer coefficient are validated respectively based on numerical simulation. Helical cruciform fuel is a novel type of nuclear reactor fuel with self-supporting and large surface-volume ratio features, potential to increase the reactor power density. However, the shape of the fuel is so complicated that influence of geometric parameters on thermal hydraulic characteristics was not clearly researched. In this paper, geometric dimensionless numbers are proposed and validated based on the numerical simulation to modify friction factor and heat transfer coefficient correlations proposed through experiments of helical cruciform fuels. Twist pitch of the fuel has a great impact on flow resistance and heat transfer, resulting in the modified coefficient ranging from 0.94 to 1.04 for friction factor and from 0.022 to 0.024 for heat transfer coefficient respectively. Change in cross-section parameters has little influence on the fuel performance, leading to the unit cross-section modified factor for both friction factor and heat transfer coefficient. After fitting the correlations with the dimensionless number, several cases are established to validate the empirical correlations, with a maximum error below 10.0 % for friction factor equation and 8.5 % for heat transfer coefficient correlation when Reynolds number is between 2.7 × 104 and 2.7 × 105. This paper provides a guide for study of heat transfer and fluid flow in helical structures, and a reference for establishment of reactor system model using helical cruciform fuels. [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]

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]

• Local condensation heat transfer in smooth channels and pin fin array is presented. • The effect of operating parameters on the heat transfer characteristics was analyzed. • Sensitivity analysis is performed of heat transfer to identify the critical parameters. • Condensation heat transfer increases with heat flux in the range of vapor quality above 0.4. In this paper, an experimental study of flow condensation was carried out of R134a in a multiport mini channel and a mini channel with pin fin array. The former comprised 20 parallel rectangular channels with an equivalent diameter of 0.64 mm. The latter is a narrow rectangular mini channel containing 10 rows of diamond pin fins with a staggered configuration, and height and longitudinal/transverse pitch of 0.5 mm and 2 mm respectively. To acquire the local value of heat flux and heat transfer coefficient, air jet impingement cooling devices were adopted to condense the vapor of refrigerants. The effects of vapor quality, mass flux, heat flux, and saturation pressure on flow condensation heat transfer coefficient were investigated with the operating conditions: vapor quality from 1 to 0, mass flux from 160 to 450 kg/(m2s), heat flux from 10.0 to 39.8 kW/m2, and saturation pressure from 600 to 1500 kPa. The experimental results indicated that the pin fin array significantly improves the condensation heat transfer coefficient up to nearly 186 % compared to the multiport mini channel in the high vapor quality region. For both channels, the local heat transfer coefficient increases with an increase in vapor quality, mass flux, and heat flux whereas decreases with increase in saturation pressure. The influence of heat flux and mass flux on heat transfer coefficient was more pronounced in the high vapor quality region than in the low vapor quality region. A sensitivity analysis was performed at different vapor quality levels. The most influential parameters in the smooth channel and the pin fin array are saturation pressure and mass flux, respectively. The relative contribution of heat flux is more obvious in the high vapor quality region. Some of the existing correlations have good prediction performance for the smooth channel experimental data in this paper, and the corresponding MAD is within 20 %. However, their deviations are much greater for the applications of pin fin array. [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]

• 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]

Summary: Closed type assembly leads to complicated core thermohydraulics which are significant for design and safety analysis in sodium‐cooled fast reactor (SFR). Most of the major phenomena can be dealt with by general parallel channel network and two‐dimensional model for inter‐wrapper flow (IWF) in system code transient thermal‐hydraulic code for analysis of SFR (THACS). A main deficiency left is the overestimation of inter‐assembly heat transfer effect due to lump temperature message that one‐dimensional assembly model can only provide. A methodology was thus developed to compute wall‐subchannel temperature for simulating energy transportation through duct wall. Intra‐assembly thermohydraulics were classified as two mechanisms and non‐dimensional parameter was introduced for characterizing each of them. Ratio between linear heat flux density through hexagonal wrapper surface and pin surface qw/p was employed to account for radial heat transfer. Flow redistribution effect arouse in mixed and natural convection was concretized by a importance symbol of buoyancy force, Richardson number Ri. Two criteria were also proposed for dividing flow state into three parts and specific model for each section was defined. Two validation cases against 3‐D computational fluid dynamics (CFD) simulation of China experimental fast reactor whole core and test data from plant dynamics test loop with direct heat exchanger experiment were performed with THACS. Noticeable improvement after incorporation of this model demonstrated its value in system analysis. Although the procedure for parameter computations is geometry related, this method can be reproduced to any type of assembly directly. [ABSTRACT FROM AUTHOR]

Summary: To evaluate residual heat removal capacity of next generation China sodium‐cooled fast reactor (SFR) and provide data for code validation, an integral natural circulation experiment was performed on a scaled water platform which had same structure with reactor primary system. A modular code THACS focusing on safety analysis of SFR was validated by this test. Some necessary development and verification for physical property, frictional loss, and heat transfer of water were carried out first. Whole simulation of an unprotected loss‐of‐flow (ULOF) case was divided into two phases and computation lasted until 5000 s of test time during second stage. Agreement in light of system temperature distribution in first stage provided a suitable initial condition for subsequent transient. Satisfied flow rate curve was reproduced with decaying pressure head and valve characteristic as input when pump was stopping. All natural circulation paths were predicted, including original flow in primary loop through main pipe and reversed flow in reactor vessel cooling system (RVCS). Transition to operative mode of direct heat exchanger (DHX) in direct reactor auxiliary cooling system (DRACS) and its cooling power were also forecasted successfully. This work means a significant advance for THACS in the field of engineering application of China SFR. [ABSTRACT FROM AUTHOR]

Summary: Accuracy and computational expenses are contradictory when performing simulation to hot plenum where multi‐dimensional phenomena voice loudly in sodium‐cooled fast reactor (SFR). A 2‐dimensional method was preferred in light of the cylindrical shape and drive force behind significant thermal stratification phenomena during transient case. One barrier related to buoyancy effect on turbulent flux transportation has been removed before. Asymmetric arrangement of components in circumferential direction makes another difficulty to reproduce actual flow field within hot plenum of pool‐type SFR. A series of special techniques for spatial discretization were implemented to restore geometry reality. Treatments about upper core structure, main heat exchanger and auxiliary heat exchanger were evaluated based on criterion of predicting well‐mixed inner plenum and stratification like outer plenum. Finally, optimal model came to an 8‐block partition scheme in which whole profiles of upper core structure and auxiliary heat exchanger were described. Validation was carried out through a normal state experiment on a water platform with reduced‐scale structure. Agreement of relative temperature distribution demonstrated ability of this model for tackling spatial layout in pool‐type SFR with the 2‐dimensional library. All real‐physics and efficient features indicate that this 2‐dimensional methodology is an excellent tool for engineering application when designing or evaluating a new SFR concept. [ABSTRACT FROM AUTHOR]

Summary: China Fusion Engineering Test Reactor (CFETR) has two operation phases: Phase I with Pf = 200 MW is to demonstrate the steady‐state operation; Phase II with Pf = 1 GW is to validate the DEMO technology. Helium Cooled Ceramic Breeder (HCCB) blanket is one blanket candidate for CFETR, and several research institutes have conducted extensive conceptual designs and comprehensive analysis works for Phase I HCCB blanket. However, because the operation phase of CFETR has transformed from Phase I to Phase II and the latest physical design parameters have been just determined, the design and analysis works for HCCB blanket should be initialized. In this work, first, the basic structure design of CFETR Phase II HCCB blanket is performed and a design scheme is proposed according to the design principles and requirements of blanket. Besides, the material selection of each blanket component has been determined. Second, based on the global neutronics model of CFETR Phase II, the neutron wall loadings (NWL) on different blanket modules are calculated. Afterwards, the radial structure layout optimizations of all poloidal blanket modules arranged in a 22.5° toroidal sector are conducted by NTCOC. As for Phase II blanket, which operates under the high fusion power condition, it should be noted that a dynamic feedback is incorporated into the original one‐way optimization method for solving the problems of inadequately optimized tritium breeding performance and insufficient tritium releasing ability. Finally, the radial structure layout parameters of all blanket modules in CFETR Phase II neutronics model are set as the obtained final optimal schemes. The calculation results show that the global TBR of CFETR Phase II is 1.182, which can well meet the tritium self‐sustaining requirement. This work can provide worthy guidance and reference for further conceptual design, comprehensive analyses and optimizations of the CFETR Phase II HCCB blanket. [ABSTRACT FROM AUTHOR]

• Validation of THACS and OpenMC for SFR has been enhanced by simulating FFTF LOFWST Test #13. • A new modeling method multi-parallel channel plus local subchannel model is proposed for reactor core. • Important local phenomenon like inter wrapper flow, thermal stratification is analyzed by the specific model. In order to support collaborative efforts within international partnerships on the development of sodium-cooled fast reactor (SFR), the International Atomic Energy Agency (IAEA) proposes a coordinated research project (CRP) about the unprotected loss-of-flow-without-scram test (LOSWOS) performed at Fast Flux Test Facility (FFTF) in 2018, which could validate simulation tool and numerical models for SFR. As one of the participants, the Nuclear THermal-hydraulic Laboratory at Xi'an Jiaotong University (NuTHeL) participates with the monte carol code OpenMC and system thermal hydraulic code THACS. This paperdeal with the standalone neutronic modeling by OpenMC and thermal hydraulic modelling by THACS. For neutronic part, the axial homogenous model was modeled to analyze the reactivity feedback coefficients and steady radial power distribution. For thermal hydraulic part, the system model, including 2D CFD outlet plenum model, subchannel model and inter-wrapper flow model, was modeled to analyze both system response and local phenomena during the LOSWOS. In general, both neutronic results and thermal hydraulic results agreed well with other participants or experiments, the reactivity feedback of GEM could provide major negative feedback, which could be used as a novel passive shutdown device. However, the remaining discrepancies were the underestimated peak temperature of PIOTA at the begin of transients, it may be caused by unknow heat transfer coefficients around hexagon tube, which has no correlation and should be further investigated by experiments. In addition, when considering IWF model, the maximum temperature of the outlet for the active fuel region was 10 K larger than the subassemblie' outlet in the PIOTA, this phenomenon should be carefully treated in loss of flow accidents of SFR. [ABSTRACT FROM AUTHOR]

The strong coupling heat transfer between subassemblies and inter-wrapper gaps is closely related to the safety of the sodium-cooled fast reactor (SFR). Therefore, it is very important to carry out a detailed thermal hydraulic analysis of the core. In this paper, the inter-wrapper channel is modeled using the subchannel analysis method. A core thermal-hydraulic analysis code SACOS-Na (Subchannel Analysis Code Of Safety for SFR) is developed by using SIMPLE algorithm. For the subassembly module, a total of three experiments, including ORNL 19-pin experiment, TOSHIBA 37-pin experiment and KNS 37-pin experiment, are adopted to demonstrate the feasibility of the mathematical and physical models and numerical algorithms of the SACOS-Na as well as the rationality of the selected auxiliary models. For the inter-wrapper gap module, WEISS 2-subassembly experiment and CCTL-CFR 3-subassembly experiment are used to verify not only the simulation capability for backflow and strong inter-subassembly heat transfer, but also the heat structure model and the subchannel modeling approach on inter-wrapper gaps. In addition, the prediction accuracy of SACOS-Na has reached the advanced international level by comparing with the worldwide well-known core thermal hydraulic analysis codes. • The subchannel modeling method is adopted to build the mesh for inter-wrapper channels. • The SIMPLE algorithm is used to solve the problems of the reverse flow and inter-subassembly heat transfer occurred in SFRs. • The universal core thermal-hydraulic analysis code SACOS-Na for SFRs is developed. • The accuracy and reliability of the SACOS-Na are discussed from all aspects. [ABSTRACT FROM AUTHOR]

Summary: China Fusion Engineering Test Reactor (CFETR) is a brand new Tokamak reactor which is currently being designed to fill the intervals between ITER and future DEMO fusion reactor. It has two operation phases: phase I with Pf = 200 MW is to demonstrate steady‐state operation; phase II with Pf = 1000 MW to validate DEMO technology. Helium‐cooled ceramic breeder (HCCB) blanket is one of the candidate blanket concepts for CFETR. Until now, there are many research institutes which have performed conceptual designs and comprehensive analyses works for phase I HCCB blanket in detail. However, lately, the operational stage of CFETR has transformed from phase I to phase II, and the latest core design parameters have been just determined (major radius equals to 7.2 m, minor radius equals to 2.2 m). Therefore, the design and analyses work for HCCB blanket should also be initialized. In this work, based on the original integrated optimization method, the radial structure layout optimization of the outboard equatorial phase II HCCB blanket module is conducted by NTCOC. However, the calculation results show that the original optimization method provides inadequately optimized tritium breeding performance and insufficient tritium releasing ability under the condition of high fusion power. To solve these problems, a dynamic feedback is incorporated into the original optimization method for the first time, and then the calculation results show that the problem has been solved well after revision. This work can supply worthy guidance and reference for the conceptual design and comprehensive analyses of CFETR phase II HCCB blanket. [ABSTRACT FROM AUTHOR]

• Discharge coefficient is not affected by Reynolds number. • Spray pattern is stable under rated operating condition. • With the increase of the inclined angle, spray coverage expands. • Distribution trends to uniform due to dispersion phenomenon caused by aerodynamic drag. • The dispersion creates the SMDs changing pattern at different measuring positions. Pressurized Water Reactor (PWR) is the most mature technology for nuclear power plants and nuclear power devices. Spray systems play an essential role in the safety of a PWR. The spray characteristics of the nozzle are of crucial importance for the performance of the spray system. In the present study, a test loop with four rotatable branches was established to obtain spray characteristics data, including flow characteristic, spray cone angle (SCA), spray flux distribution and droplet size spectrum. The effect of inclined angles was taken into consideration when designing and conducting the experiment. A specific type of pressure-swirl nozzle used in China advanced PWR was employed for the study. According to experimental data, variations of pressure drop and the spray angles versus volume flow rate were obtained, and typical curves of spray flux distribution were plotted, and the change in the flux distribution was evident with the change in inclined angle. The flux distribution results were used to build 3D spray flux distribution models together with the spray cone angles tested. The droplet size spectrum was measured by laser diffraction analysis. Typical droplet size distributions were obtained by the relevant software, and the experimental Probability Density Function (PDF) of droplet size was derived and compared with several empirical distributions. [ABSTRACT FROM AUTHOR]

To exploit the advantages of Fluoride-Salt-cooled high-Temperature Advanced Reactors (FuSTAR) such as high temperature, inherent safety, compact structure, and modularization, an energy conversion system based on a closed Brayton cycle was adopted, and the superstructure design was optimized to attain the best parameters and configuration. To enhance the convergence rate of a single computation, the simultaneous equation method was employed, while the particle swarm method was utilized to prevent local optimality. Irreversible analysis and sensitivity analysis were performed to evaluate the system's performance. The findings demonstrate that thermal efficiency can be significantly enhanced by incorporating reheating into the system. Furthermore, for air and Ar, the recompression intercooling configuration proved effective in reducing the heat transfer temperature difference and enhancing thermal efficiency. For CO2 and Xe, the incorporation of reheating, recompression, and intercooling significantly improved thermal efficiencies beyond 53%, indicating their immense potential for further development. Introducing reheating into the Xe cycle led to a substantial increase in both thermal efficiency and the area of the temperature-entropy graph. The irreversible analysis revealed that for CO2 and SF6, the predominant loss occurred in the cooler and recuperator. In contrast, for the Xe cycle, all kinds of irreversible losses were evenly distributed across each device. Sensitivity analysis indicated that the thermal efficiency of near-critical fluids was more sensitive to the lowest pressure than that of far-critical fluids. Both the near-critical split ratio and the far-critical minimum pressure had a significant impact on the mass flow rate. The design results of this method can guide the selection of the optimal power cycle system configuration for FuSTAR. [ABSTRACT FROM AUTHOR]

• Thermal-hydraulics of LBE is studied and an experiment is carried out. • A new empirical correlation of LBE is developed, and validated by XJTU 7-rod bundle, KIT 19-rod bundle, and ENEA 37-rod bundle. • Detailed CFD is carried out to analyze the characteristics of flow and heat transfer of LBE. • Turbulent heat transfer model of LBE is discussed and verified. Lead-bismuth eutectic (LBE) is a liquid with low Prandtl number and high thermal conductivity, generally used as the coolant in lead-cooled fast reactors of Generation-IV. The traditional Reynolds analogy is not appropriate for LBE, hence the theoretical analysis and experimental measurements are studied to interpret the flow and heat transfer of LBE in this article. In this way, a single-rod experiment on flow and heat transfer for LBE is performed and a new empirical correlation is fitted. To analyze the behavior of flow and heat transfer for low Prandtl number fluids, and also to validate the turbulence model and new correlation in the hexagonal rod bundles. The CFD simulations are performed for the XJTU 7-rod bundle, the KIT 19-rod bundle, and the ENEA 37-rod bundle based on ANSYS Fluent. The convective heat transfer coefficient of LBE, the temperature of heating wall at different sub-channels, and the temperature non-uniformity are investigated in this paper, to reveal the flow and heat transfer characteristics of LBE. The results show that the deviations of the section averaged Nusselt number are in 25%, and the deviations of the local temperature of LBE and heating wall are in 10℃. Grid spacers will result in a 30% rise in the section averaged Nusselt number, which will improve the turbulent intensity and heat transfer. Temperature non-uniformity is reduced as the Peclet number is increased. The validations demonstrate that the new empirical correlation provides accurate convective heat transfer coefficient predictions for LBE flow and heat transfer in hexagonal rod bundles with a relative derivation in 20%, which can be used as a reference for additional LBE simulations and experiments. [ABSTRACT FROM AUTHOR]

• A design analysis code for sodium heated OTSG is developed based on the power mesh method. • The code is verified with various design data both of integral type and modular type OTSGs. • The influence of tube diameter on heat transfer capacity has been analyzed by the code. Sodium heated once-through steam generator (OTSG) is a critical component of Sodium cooled Fast Reactor (SFR). It is a counter current heat exchanger. It transfers heat from hot sodium to water. Therefore, the steam generator is a necessity for the operation of the plant. However, due to the phase change exists in water side, it becomes challenging to design a steam generator for SFR. After the successful construction of CEFR (China Experiment Fast Reactor), a power-up demonstration SFR is being developed in China. For the design and analysis of the steam generator of the demonstration SFR, a thermal hydraulic Design Code of Once-through Steam generator for Sodium cooled fast reactor (DeCOSS) has been developed based on the movable mesh method. The DeCOSS code has been benchmarked based on the design data of OTSGs reported. The tube lengths and temperature distributions calculated by DeCOSS showed general agreements with published data. The influences of tube diameters on heat transfer capacity have been analyzed. The results indicated that the heat transfer capacity is determined by the thermal resistance of tube wall under the condition of fixed heat transfer coefficient of water side. In addition, under the same flow area of water and sodium, the heat transfer capacity could be enhanced by decreasing the tube diameters with constant tube wall thickness. Hence, the code can be utilized to do the sensitivity analyses and optimum thermal design of OTSGs. Moreover, it can support the design work of once-through steam generator of SFR. [ABSTRACT FROM AUTHOR]

Highlights • A one-dimensional code for thermal-hydraulic analysis of sodium heated OTSG is developed. • Both integral type and modular type sodium heated OTSG can be analyzed by the code. • The accuracy of the code is validated with various design data and the ETEC shutdown experiment. Abstract Sodium heated once-through steam generator (OTSG) is an essential component of Sodium cooled Fast Reactor (SFR). It transfers the heat from hot sodium to water and converts water to superheat steam. Steam can roll the turbine and generate electricity. Therefore, the safety of the steam generator is vital to the operation of the plant. China is developing the China Demonstration SFR (CDSFR) after successful construction of China Experiment Fast Reactor (CEFR). In order to make sure the safety of the steam generator under various transient conditions and support the design of the steam generator for CDSFR, a Transient analysis Code of Once-through Steam generator for Sodium cooled fast reactor (TCOSS) has been developed. To benchmark the developed physical model, the calculated results of the code have been compared with some design data of OTSGs and they are found in a good agreement. In addition, the transient results of the code have been compared with the experimental results of a shutdown experiment for the validation of the code. The predictions of the code agree well with the experimental data. Hence, the code can be utilized to predict the variations of thermal-hydraulic parameters in steam generator under different transient conditions. In addition, predictions can support the design of the steam generator for CDSFR. [ABSTRACT FROM AUTHOR]

Highlights • Heat transfer experiments for super-low finned-tube bundles in MSRs are performed. • Tests are carried out under the operating conditions of the nuclear power plant. • A more accurate heat transfer correlation under wider Re number range is proposed. Abstract Experimental investigation on the heat transfer characteristics of superheat vapor through a new designed moisture separator reheater with "super-low" finned-tube bundles for Generation III+ nuclear power plant, was carried out based on a 5*30 tube bundle test section. Impacts of the vapor mass flow rates are analyzed on the heat transfer characteristics of superheat vapor through the finned-tube bundle. A group of experiments are reviewed on single-phase fluid flowing across the triangular super-low finned-tube bundles whose fin heights are less than 2.75 mm and ratios of fin height to fin diameter are from 0.05 to 0.15. Combining the presented experimental data with the other open published data for single-phase turbulent gases flowing across triangular super-low finned-tube, a new heat transfer correlation is proposed for the MSRs super-low finned-tube bundles, which shows better accuracy and has wider Reynolds number range in application. [ABSTRACT FROM AUTHOR]

Abstract The main vessel plays an important role in containing the entire primary sodium for pool-type sodium-cooled fast reactors (SFRs). The Reactor Vessel Cooling System (RVCS) has great effect on cooling the main vessel. However, little attention has been given to the study on transient characteristics of RVCS in the previous SFR research. Thus, a home-made one-dimensional (1-D) code named Reactor Vessel Cooling system Analysis Code for Sodium-cooled fast reactor (VECAS) is proposed to evaluate the thermal-hydraulic characteristics for SFR. The detailed models of the developed VECAS are presented in this paper. Moreover, the developed models have been validated against an experimental study. Numerical data of the main vessel cooling circuit are compared with the measurements of the Demonstration Fast Breeder Reactor (DFBR). The simulation results are in good agreement with the experimental data. Furthermore, the validated VECAS is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). The transient characteristics of RVCS in China Experimental sodium-cooled Fast Reactor (CEFR) are simulated by the coupled code. Steady analysis shows that the main vessel is cooled effectively. The peak temperature appears at the top of the main vessel lower than the permissible upper temperature limit. During the transient analysis, VECAS has predicted a reverse flow in RVCS, which contributes to the core cooling. Furthermore, sensitivity analysis of the main parameter has also been performed. Therefore, it can be concluded that coupled VECAS has the ability to evaluate the thermal-hydraulic characteristics as well as the decay heat removal capacity of RVCS. The coupled code could provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor. Highlights • A one-dimensional (1-D) code named VECAS is proposed to evaluate the thermal-hydraulic characteristics for SFR. • The code has been validated against the main vessel cooling circuit of DFBR. • VECAS is coupled with THACS to calculate the transient characteristics of RVCS in CEFR. • VECAS has the ability to evaluate the thermal-hydraulic characteristics and the decay heat removal capacity of RVCS. [ABSTRACT FROM AUTHOR]

Highlights • Fluoride salt experimental system has been built using heat pipes. • Mutual interference between heat pipes impedes the heat transfer. • Heat pipe spacing has little difference on the natural convection heat transfer. Abstract In order to improve the passive safety performance of molten salt reactor, a new conceptual design of passive residual heat removal system (PRHRS) has been proposed using heat pipe technology. In this paper, an experimental system has been constructed to verify the PRHRS design of molten salt reactors. An eutectic salt mixture of 46.5 mol%LiF-11.5 mol%NaF-42 mol% KF compositions called FLiNaK is selected as working fluid. Tests on the natural convection between the fluoride salt and a vertical array of heat pipes in the drain tank have been conducted. The temperature field of fluoride salt in the tank with different heat pipe spacing is analyzed. Comparisons have been made on the heat transfer characteristics between the natural convection with a single heat pipe and the vertical array of heat pipes. In addition, the empirical correlation for the Nusselt number is obtained within the range of Rayleigh numbers from 5.7 × 10 6 to 1.31 × 10 7. A good agreement between the proposed correlation and the experimental data is obtained with the deviation of −2.9%–1.7%. This work could provide valuable experimental data and suggestions for the design and analysis of PRHRS of MSR. [ABSTRACT FROM AUTHOR]

Summary: Thermal stratification phenomena usually occur in the upper plenum after a scram of sodium‐cooled fast reactors, which should be closed concerned in the fields of structural integrity assessment and residual heat removal capacity. A 2‐dimensional analysis program under cylindrical coordinate was developed for predicting the in‐vessel thermal hydraulics. Non‐orthogonal block‐structured grids were generated to resolve the problems with complex structures. A second‐order discrete scheme based on midpoint rule was applied to the spatial discretization of convection and diffusion terms. Two sets of experiments characterized by distinct shapes of their apparatuses were used for the validation, mainly from the viewpoints of vertical temperature distribution and rising behaviors of the stratification interface. Results show that RANS‐type turbulent models make the significant impacts on different flow regimes. In the calculation of a scaled model with plenty of stagnant sodium in the upper region, the realizable k − ε model (RKE) considering the mean deformation rate gives better outcomes than the standard k − ε model (SKE). For the plant‐type upper plenum with considerable flow rate along the entire height, buoyancy modeling is the crucial issue to follow the upward movement of the interface and the relaxation process of the temperature gradient. In this case, employment of the turbulent Prandtl number reflecting the damping effect by incorporating the local Richardson number well reproduced the experimental results. [ABSTRACT FROM AUTHOR]

Highlights • A new pressure drop limit has been added into the optimization criteria. • The previous integrated optimization algorithm of NTCOC is improved. • The improved code has been applied on the HCSB blanket for CFETR Phase II. • The computational processes with two methods are compared. Abstract Blanket is the key component of Chinese Fusion Engineering Test Reactor (CFETR). In our previous work, considering that the radial arrangement of the internal functional zones has the greatest influence on the blanket performances and is also the most important basis of the 3D design, NTCOC, a Neutronics/Thermal-hydraulic Coupling Optimization Code which adopts 1D neutronics and 2D thermal-hydraulic simplified calculation models, has been independently developed for facilitating the radial build design and optimization of the CFETR helium cooled solid breeder (HCSB) blanket. Through setting three leading design requirements as the optimization criteria, this code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase I, which preliminarily verifies the reasonability of the code. However, recently, the design phase of CFETR has gradually changed from I to II, the fusion power of which will be quintuple to the previous I. This means that the internal nuclear power deposition in each blanket component under II operation condition will be much larger than the previous I. Therefore, much larger coolant inlet mass flowrate will be needed and more breeder units should to be inserted, both of which will make the total pressure drop of blanket for Phase II much larger than the previous I and even exceed the limit of CFETR HCSB blanket greatly. This can’t be sustained in CFETR, a power plant that should produce electricity. To solve this problem, in this paper, the previous integrated neutronics/thermal-hydraulic optimization algorithm is improved by adding the new pressure drop limit into the optimization criteria. Then the improved code has been applied on optimizing the radial build arrangement of a conceptual design of HCSB blanket for CFETR Phase II through constant and variable step length two methods. The calculation results verifies the feasibility and reasonability of the improved optimization algorithm preliminarily. [ABSTRACT FROM AUTHOR]

To investigate Counter-Current Flow (CCF) characteristics in the Pressurizer Surge Line (PZR SL) assembly of the large advanced passive nuclear reactors, steam-water CCF experiments are carried out on a stainless steel test section, which is a 1:4 scaled-down model of the AP1000 PZR SL assembly. The inner diameter of the test PZR SL pipe is 90 mm, which also belongs to the large-diameter pipes, as same as the prototype PZR SL pipe, in the field of Counter-Current Flow Limitation (CCFL) researches. The present steam-water CCFL experiments are conducted under normal pressure and saturated temperature, with the PZR simulator collapse water level ranging from 350 to 900 mm. CCFL becomes severer at higher steam flow rate, and the most limiting CCFL effect locates between the PZR simulator bottom head and the vertical part of PZR SL pipe. The onset of CCFL and zero liquid penetration (ZP) are two critical conditions in the CCF development process, dividing the process into three stages: Before-CCFL, Partial-CCFL, and CCFL-ZP. According to the local CCFL conditions, the development of the CCFL process is also divided into four-regions since onset of CCFL. The present steam-water CCF data are well normalized in terms of dimensionless Kutateladze (Ku) numbers, and a Ku-type empirical partial CCFL correlation is developed. The comparisons of the present CCF data and partial CCFL correlation with the CCF data and correlations of former experiment researches validate that the present empirical partial CCFL correlation is conservative to predict steam-water partial CCFL in the prototype PZR SL assembly of the large advanced passive nuclear reactors. [ABSTRACT FROM AUTHOR]

The evaluation of thermal-hydraulics of core under decay heat removal conditions is essential to the safety evaluation of a sodium-cooled fast reactor. The inter-wrapper flow (IWF) has an influence on the thermal-hydraulics of core. However, the study of the flow and heat transfer of IWF was limited. In this paper, a 2D layered IWF model was developed, and some tests were simulated to validate the IWF model, including heat removal tests SHRT-17 and SHRT-45R conducted in the experimental fast reactor EBR-II and a natural circulation test performed during the PHENIX end-of-life experiment. In order to simulate all the components of the primary system, the IWF model is coupled with the Transient Thermal-Hydraulic Analysis Code for Sodium-cooled fast reactors (THACS). In the simulation without IWF model, the predicted peak temperature of the instrumented subassembly XX10 in EBR-II is about 150 K lower than test data, and the predicted average outlet temperature of reactor core in PHENIX is about 20 K higher than test data. While the predictions of THACS with IWF model agree well with the test data. The results show that the IWF can be accurately simulated by the IWF model, and the IWF model improves the accuracy of the simulations of the reactor core. Further, some sensitivity analyses were conducted to provide better reference for the study of sodium-cooled fast reactors. [ABSTRACT FROM AUTHOR]

Fluorine High-temperature Reactor (FHR) is a novel reactor type without water-cooling, but with a high outlet temperature. Small Modular FHR is convenient for modular assembly and transportation. Therefore, it is suitable for remote and water shortage areas to provide efficient and clean nuclear energy. The neutronics and heat transfer of the fuel are the key issues when the small modular FHR has high thermal power output. This research proposes the fuel – TRISO (TRi-ISOtropic) particles dispersed in Helical Cruciform Fuel (HCF). HCF can be self-supported in radial direction, no additional grid-spacers are required, which can decrease the pumping power and improve the efficiency of the reactor system. For small modular FHR, it can be disassembled, transported, and assembled better without grid-spacers and simplifying the core structure. The heat transfer capacity of the coolant also could be further improved because of the special structure of HCF. Based on the Monte-Carlo particle transport code MCNP, the neutronic characteristic from the assembly to the whole core of FHR is studied. By obtaining the fission power distribution in near-critical condition, the thermal-hydraulics of the single most heated HCF and 7 bundle HCF in the heated assembly are calculated. In order to preliminary analyze the thermal-hydraulics of the whole core scale, the thermal-hydraulics analysis of porous media is carried out for 1/6 core. The isotropic resistance tensor and thermal equilibrium porous model have been introduced with reasonable approximation, and porous resistance relation has been carried out. The radial power peak factor is 1.185 and the axial power peak factor is 1.435 in near-critical condition, which proves the arrangement of the HCF and the control rods can flatten the power distribution effectively. The average temperature of the coolant outlet in the active region of the core is 974.44 K, and the average temperature of the upper plenum outlet is 972.85 K. High coolant outlet temperature can ensure effective supply of high-temperature process heat. And the highest temperature of the fuel is 1271.49 K, which is lower than the limitation 1573 K. The parameters of this research can supply the neutronics and thermal-hydraulics references for further FHR design. [ABSTRACT FROM AUTHOR]

• A two-fluid, six-equation thermal–hydraulic system code was validated. • The results show the code has the capability to simulate thermal–hydraulic phenomena for the pressurized water reactor. • Loss of feedwater accident experiment of LOFT Project were used to validate the code. CGN has developed a two-fluid thermal–hydraulic system analysis code, LOCUST 1.2. This paper carries out an integral validation of LOCUST 1.2 based on the LOFT L9-3 experiment. First, the facility has been modeled using "node and branch" approach. Second, steady-state calculation is performed to obtain the initial parameters, which shows that critical parameters are within the uncertainty of experimental measurements. Third, transient response is calculated. For primary coolant system, the core power, the temperatures are in good agreement with the experiment, and the pressure has an acceptable deviation after the main steam control valve shutdown. For secondary side of steam generator, the pressure trend is in good agreement with the experiment, and liquid level fits well to experiment before main steam control valve shutdown. As a result, LOCUST 1.2 has the capability to simulate thermal–hydraulic phenomena for reactor primary coolant system and secondary side of steam generator structures and equipment. [ABSTRACT FROM AUTHOR]

Summary: Molten salt reactor (MSR) as 1 candidate of the generation IV advanced nuclear power systems attracted more attention in China due to its top ranked in fuel cycle and thorium utilization. Two types of MSR concepts were studied and developed in parallel, namely the MSR with liquid fuel and that with solid fuel. Abundant fundamental research including the neutronics modeling, thermal‐hydraulics modeling, safety analysis, material investigation, molten salts technologies etc. were carried out. Some analysis software such as COUPLE and FANCY were developed. Several experimental facilities like high‐temperature fluoride salt experiment loop have been constructed. Some passive residual heat removal systems were designed, and 1 test facility is under construction. The key MSR techniques including the extraction and separation of molten salt and construction of N‐base alloy have been mastered. Based on these fundamental research, Chinese Academy of Sciences has completed the design of thorium‐based MSRs with solid fuel and liquid fuel and is promoting their construction in the near future. In China, future efforts should be paid to the material, online fuel processing, Th‐U fuel cycle, component design, and construction and thermal‐hydraulic experiments for MSR, which are rather challenging nowadays. [ABSTRACT FROM AUTHOR]

Summary: A new conceptual design of a passive residual heat removal system (PRHRS) has been proposed for molten salt reactor. High‐temperature heat pipes are used in this new design to improve the system inherent safety and make the PRHRS more compact. An experimental system using fluoride salt FLiNaK has been constructed to validate and support the future design of PRHRS of molten salt reactors. In this research, tests on the natural convection heat transfer of FLiNaK in the drain tank with an inclined heat pipe inserted at different heights were performed. The temperature distribution of fluoride salt in the tank was analyzed. The height of heat pipe and the bulk temperature of FLiNaK have little influence on the normalized salt temperature distribution. However, with the height of heat pipe increasing, the temperature difference of molten salt decreases and heat transfer coefficient of natural convection increases. In addition, the empirical correlations of natural convection heat transfer between liquid FLiNaK and inclined heat pipe are obtained within the range of Rayleigh numbers from 3.97 × 106 to 1.16 × 107. The comparisons show that a good agreement with less than 5% deviation is obtained between the proposed correlations and the test data. [ABSTRACT FROM AUTHOR]

Different from the square-arrayed fuel rod bundles in pressurized water reactors, the fuel subassembly for sodium-cooled fast reactors features with wire spacers helically wrapped on the fuel rod. This intricate geometry consequently imposes a significant challenge in numerical simulation. In the present study, a new marked and separated meshing method is developed to reduce the required mesh and modeling time. Instead of directly generating the complex geometry and mesh, the hexahedral wire wrapper mesh was dealt with locus equations. First, the numerical modeling method was validated by combining the experimental data of a 19-pin wire-wrapped rod bundle and the simplified Cheng and Todreas correlation. Furthermore, a comparative study of different meshing methods was conducted. Then, an investigation on the three-dimensional flow phenomenon and heat transfer characteristics of a 61-pin wire-wrapped fuel subassembly for the China Experimental Fast Reactor was carried out. Finally, complex axial and transverse flow behaviors and temperature field in different types of subchannels were quantitatively analyzed. Results indicate that the axial velocity exhibits periodic oscillation with a maximum normalized value of 1.54. The flow oscillation in the interior subchannels presents a smaller amplitude and higher frequency than that in the corner and edge subchannels. Local hot spots occur in the gaps beneath the spacer wires. Sweeping and cross flow patterns exist in the peripheral and interior subchannels, respectively. In addition, the secondary transverse flow induced by the spacer wires is stronger in the corner subchannel. The cross flow strength reaches a peak value equal to 30% of the average axial velocity, which promotes the thermal mixing. [ABSTRACT FROM AUTHOR]

Fusion-fission hybrid reactor is believed to be a feasible implementation of fusion energy in a short time as it requires less advanced technology in fusion physics and engineering than pure fusion reactor. In this paper, based on the mature technology of pressurized water reactor (PWR) and existing fusion reactor blanket concepts, a new conceptual design scheme of water-cooled pressure tube blanket was proposed for hybrid reactor and both the material selections and the structure design were performed. In the blanket, a trapezoidal U-shaped first wall (FW) was adopted to decrease the gaps between blanket modules. Due to high heat flux from plasma and intense irradiation by high-energy neutrons, it's essential to conduct thermal-mechanical coupling analyses and optimizations of the FW to ensure reasonable temperature and stress distributions on it. In this work, CFX-Workbench coupling method was adopted to perform the simulation. The influences of different FW geometrical configurations and boundary conditions on the temperature and stress distributions were researched in detail to optimize the FW structure. Besides, the thermal and mechanical characteristics of the helium tanks, the cooling channels and the breeding zones were researched in this paper. The calculation results verified the reasonability of the conceptual design scheme preliminarily and could provide a meaningful reference for the further conceptual design, analysis and optimization of the hybrid reactor blanket. [ABSTRACT FROM AUTHOR]

In this study, a three-step approach of ordinal 3D-1D&2D-3D is independently proposed to derive a coupling optimization design rule of the Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder (HCSB) blanket. Firstly, the three dimensional (3D) neutronics model of CFETR is built with all of the 15 blanket modules adopting the original radial build arrangement parameters. Based on this model, both the neutron wall loading (NWL), which will serve as the normalization factors for the following local neutronics calculations, and the tritium breeding ratio (TBR) distributions on each blanket module are obtained. Secondly, the radial build arrangements of all the 7 inboard and the 8 outboard blanket modules in an 11.25° toroidal CFETR sector are optimized by taking both the tritium breeding and thermal performances into account as the optimization objectives. NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which has been independently developed by Xi’an Jiaotong University for accelerating the radial build design and optimization of the fusion reactor HCSB blanket concept with adopting the simplified 1D neutronics and 2D thermal-hydraulic models, is used for these work. Thirdly, based on the final optimized schemes of all the blanket modules, a comprehensive research on the coupling optimization design rule of the CFETR HCSB blanket is performed in-depth and a coupling optimization design integration approach has been finally summarized. Finally, the 3D CFETR neutronics model is updated by replacing all of the original blanket radial build arrangement parameters with the final optimized schemes. The calculation results show that after the radial optimizations, the total TBR of all the 15 blanket modules has increased from the previous 1.092–1.274, which can well meet the self-sufficiency tritium breeding requirement of CFETR. [ABSTRACT FROM AUTHOR]

Nonwater fluids have been widely adopted in the designs of several advanced nuclear reactors, especially the fourth-generation nuclear reactors and International Thermonuclear Experimental Reactor (ITER) blankets. Relap5 Mod3, initially developed for the water-cooled nuclear reactors, can continue to be applied to conduct the transient analysis of nonwater-cooled nuclear reactor systems as recommended. Thermodynamic and transport properties of fluoride salts, sodium, and lithium-lead eutectic are summarized and embedded into the Relap5 Mod3 code. Flow and heat transfer empirical correlations specific for low-Prandtl-number liquid metal and high-Prandtl-number fluoride salts are also added into the corresponding modules of the code. Through the benchmark problems including the protected loss of forced cooling of Mark1 pebble bed fluoride-salts-cooled high-temperature reactor and the Shutdown Heat Removal Test (SHRT) 17 of Experimental Breeder Reactor II, the modified Relap5 Mod3 has been verified and validated for the transient simulation of the fluoride-salt-cooled high-temperature reactors and sodium-cooled fast reactors. An ITER demo lithium-lead eutectic blanket, China Dual-Functional Lithium-Lead Test Blanket Module (DFLL-TBM), also proves safe under the protected loss of forced cooling accident through the analysis of the modified Relap5 Mod3. [ABSTRACT FROM AUTHOR]

Reliable systematic uncertainty analysis is of great significance for reactor safety and licensing. In this paper, efforts have been made to quantify uncertainties of computational results. The uncertainty quantification and sensitivity analysis for a 20 MWth prismatic Transportable Fluoride-salt-cooled High temperature Reactor (TFHR) proposed by MIT is performed with the DAKOTA code coupled to the best-estimate nuclear system software RELAP5-3D. Eight input parameters are selected with specified variations and probability density functions (PDFs), a total number of 400 steady-state and Unprotected Loss of Flow (ULOF) transient calculations are then conducted in order to statistically obtain the 5% and 95% peaking fuel temperature tolerance intervals with confidence level of 95%. The Latin hypercube sampling (LHS) model in DAKOTA is verified against analytical results and the TFHR RELAP5 modeling is benchmarked based on CFD calculations. It is found that, for the steady-state, the uncertainties of the heat transfer coefficient and helium gap conductivity factor are the most remarkable contributors to the statistical results of peaking fuel temperature. 0.5% of calculated fuel temperatures exceeded the steady-state fuel temperature limit of 1573 K. Under ULOF accident condition, no temperature points exceeded the transient fuel failure limit 1873 K, but there is a possibility of 0.195 leading to the damage of structural material. Three primary contributors to the uncertainty in peak fuel and outlet coolant temperature are the reactor power, helium gap, and heat transfer coefficient. The findings indicate that there are needs for further reactor design optimization from thermal-hydraulics prospective. [ABSTRACT FROM AUTHOR]

As one of the most important components of fusion reactor, breeding blanket mainly accounts for tritium self-sufficiency, shielding and fusion energy collection. The performance of the blanket is mainly determined by the internal structural arrangement and material composition. However, if we directly adopt a three-dimensional calculation model for the blanket design and optimization, not only will it consume extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial arrangement of the functional zones, which has the greatest influence on the performance, is the basis of the detailed and accurate 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulics C oupling O ptimization C ode, which adopts simplified 1D neutronics and 2D thermal-hydraulic calculation models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the functions of both the automatic iterative updating the temperature-dependent cross-section libraries and the data transmission between the neutronics and thermal-hydraulic calculation results through the inner and external coupling between the MCNP4C and CFD software, respectively. Therefore, this can make the comprehensive optimization analyses progress much faster and more accurate. In addition, a variable step length method is adopted to further improve the coupling optimization speed of this code, and several leading design objectives are set as the optimization criteria instead of taking all the possible practical engineering constrains into account. Finally, this code has been applied to optimize the radially arranged zones of a conceptual design scheme of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts mixed pebble beds as both the tritium breeder and neutron multiplier, under both the normal and critical conditions. [ABSTRACT FROM AUTHOR]

Water hammer caused by system devices closing or opening are important factors that threat the safety of nuclear reactors. A coupling scheme is proposed to analyze water hammer in a designed parallel pump-valve system. The developed FLUENT/RELAP5 makes it possible to obtain three-dimensional (3-D) flow field of the key area with lower computer resource than traditional computational fluid dynamic (CFD) analysis. The coupled code can reflect many sophisticated phenomena which system code ignores. In the developed coupling scheme, RELAP5 is compiled as a Dynamic Link Library (DLL). The compiled library is invoked as a subroutine at every time step by FLUENT. A testing case has been proposed to verify the implementation of the coupling scheme. Testing results prove the implementation of the coupling scheme successful and coupled results reliable. During the period of valve close, the coupled code predicts considerably well over movement of the valve element by the moving grid method. During the period of pumps start, three shock waves occur, and the mechanical loading oscillates continuously after 12 s. Moreover, the coupled results are compared with the standalone RELAP5 results. Contrast analysis represents that the coupled FLUENT/RELAP5 has the potential to obtain credible transient numerical results. The predicted 3-D transient characteristics directly reflect the generated pressure surges and forces in the designed pump-valve system, which is a fundamental guide for mitigating the threat of water hammer. [ABSTRACT FROM AUTHOR]

As one of the key components of fusion reactor, breeding blanket is of vital importance for the functions of tritium self-sufficiency and fusion energy transformation. The performances of the blanket mainly depend on the internal structural arrangement and material composition. However, if we directly adopt the actual 3D calculation model for the blanket design and optimization, not only will it cost extensive human and computer resources, but also make the computational process extremely complicated and lengthy. Considering that the radial build arrangement of the functional zones is the most important determinant for the detailed 3D design of the blanket, NTCOC, a N eutronics/ T hermal-hydraulic C oupling O ptimization C ode, which adopts 1D neutronics and 2D thermal-hydraulic simplified models, is developed for accelerating the radial build design and optimization of the blanket in this work. This code realizes the iterative updating the temperature cross-section libraries through inner coupling, which can make the calculation more accurate. Besides, it realizes the automatic data transmission between the nuclear and thermal calculation results, which can facilitate the coupling optimization process greatly. In addition, a variable step length method is adopted for further improving both the coupling optimization efficiency and accuracy of this code, and several leading design requirements are set as the optimization criteria instead of taking all the practical engineering constrains into account. Finally, after the improvement of the algorithm of the integrated neutronics/thermal-hydraulic optimization for NTCOC, this code has been applied on optimizing the radial build arrangement of a conceptual design of Chinese Fusion Engineering Test Reactor (CFETR) helium cooled solid breeder blanket, which adopts separated Li 4 SiO 4 and Beryllium pebble beds as tritium breeder and neutron multiplier respectively, under normal, critical and DEMO three conditions. [ABSTRACT FROM AUTHOR]

Chinese Fusion Engineering Test Reactor (CFETR) is a new test tokamak device being designed in China aimed to bridge gaps between ITER and DEMO. As one of the candidate tritium breeding blankets, a conceptual design scheme of the helium cooled solid breeder blanket has been proposed and a series of preliminary analyses have been carried out to access the performances. However, both the required global tritium breeding ratio for CFETR not less than 1.2 and its poor working conditions under intense radiation need further thorough neutronics analyses and optimizations during the design phase. In this work, first, three-dimensional neutronics model of CFETR was built, and the neutron wall loading and global TBR were obtained. The nuclear and thermal calculation results were automatically coupled, which could make the neutronics calculation results more accurate and guaranteed they could always satisfy the corresponding thermal limits during the whole process. Then, the tritium breeding and shielding performances of both the outboard and inboard equatorial blanket modules were optimized for the comprehensive optimal schemes. The influences of Be/W armors on the shielding performance and TBR were also investigated. Finally, the nuclear heating rates and the neutron flux densities in different components were calculated based on the obtained comprehensive optimal scheme. In this paper, the neutronics analyses and optimizations verified that the optimized conceptual design could well meet the tritium self-sufficiency and neutron shielding requirements, and this could provide a valuable reference for the further thermal-hydraulic analysis and structural optimization of the CFETR helium cooled solid breeder blanket. [ABSTRACT FROM AUTHOR]

Electroencephalogram (EEG)-based brain–computer interfaces (BCIs) allow users to use brain signals to control external instruments, and movement intention detecting BCIs can aid in the rehabilitation of patients who have lost motor function. Existing studies in this area mostly rely on cue-based data collection that facilitates sample labeling but introduces noise from cue stimuli; moreover, it requires extensive user training, and cannot reflect real usage scenarios. In contrast, self-paced BCIs can overcome the limitations of the cue-based approach by supporting users to perform movements at their own initiative and pace, but they fall short in labeling. Therefore, in this study, we proposed an automated labeling approach that can cross-reference electromyography (EMG) signals for EEG labeling with zero human effort. Furthermore, considering that only a few studies have focused on evaluating BCI systems for online use and most of them do not report details of the online systems, we developed and present in detail a pseudo-online evaluation suite to facilitate online BCI research. We collected self-paced movement EEG data from 10 participants performing opening and closing hand movements for training and evaluation. The results show that the automated labeling method can contend well with noisy data compared with the baseline labeling method. We also explored popular machine learning models for online self-paced movement detection. The results demonstrate the capability of our online pipeline, and that a well-performing offline model does not necessarily translate to a well-performing online model owing to the specific settings of an online BCI system. Our proposed automated labeling method, online evaluation suite, and dataset take a concrete step towards real-world self-paced BCI systems. [ABSTRACT FROM AUTHOR]

• Gut microbiota regulate the behavior and emotion through microbiota-gut-brain axis. • Changes in gut microbiota are closely associated with depression disorder. • Gut microbiota connects with the brain through many ways in depression disorder. The microbiota-gut-brain axis is a bidirectional regulatory pathway between the brain and the gastrointestinal tract, which plays an important role in maintain homeostasis. Gut microbiota could influence the behavior, cognition, stress response and others via the axis. Depression is a complex psychiatric disease, giving rise to heavy social health and economic burden. In recent years, studies have shown that the gut microbiota are closely linked to the pathophysiological processes of depression. In this article, the interaction and its underlying mechanisms between depression and gut microbiota were summarized. [ABSTRACT FROM AUTHOR]