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Lead-cooled fast reactors (LFRs) have a wide range of application scenarios, which require the thermal-hydraulic characteristics of LFRs to be reliable. In the present paper, the Lead-cooled fast reactor Thermal-Hydraulic Analysis Code LETHAC was developed, including the models of pipe, heat exchanger, and pool. To verify the correctness of LETHAC, two experimental facilities and three experimental cases were selected, including GFT and PLOFA tests for NACIE-UP and Test-1 for CIRCE. The calculated results show the same and consistent trend with the experimental data, but there are some discrepancies. It can be found that LETHAC is suitable and reliable in predicting the transient behavior of lead-cooled system.

Average and local convective heat transfer coefficients of nitrogen are measured experimentally in an electrically heated circular tube for a range of Reynolds number from 1.08 × 104 to 3.60 × 104, and wall-to-bulk temperature ratio from 1.01 to 1.77. The exit Mach number is up to 0.17, and the heat flux is up to 46 kW·m−2. The molybdenum test section has a 62 diameters heated section with an inside diameter of 5 mm and a 30 diameters entrance section to ensure the fully-developed flow. Uncertainty of Nusselt number is less than 1.6 % in this study. The results indicate that the average heat transfer correlations evaluated by both the bulk and the modified film Reynolds numbers agree well with the experimental data. The local heat transfer results based on bulk properties are compared with previous empirical correlations. New prediction correlations are recommended which are significantly affected by the property variation and heated length. The comparison between the proposed correlations and experimental points shows that 88 % of experimental data fall into an error of 10 %, and almost all data are within an error of 20 %.

Abstract In the extensive urban rail transit network, interruptions will lead to service delays on the current line and spread to other lines, forcing many passengers to wait, detour, or even give up their trips. This paper proposes an event-driven simulation method to evaluate the impact of interruptions on passenger flow distribution. With this method, passengers are regarded as individual agents who can obtain complete information about the current traffic situation, and the impact of the occurrence, duration, and recovery of interruption events on passengers’ travel decisions is analyzed in detail. Then, two modes are used to assign passenger paths: experience-based pre-trip mode and response-based entrap mode. In the simulation process, the train is regarded as an individual agent with a fixed capacity. With the advance of the simulation clock, the network loading is completed through the interaction of the three agents of passengers, platforms, and trains. Interruption events are considered triggers, affecting other agents by affecting network topology and train schedules. Finally, taking Chongqing Metro as an example, the accuracy and effectiveness of the model are analyzed and verified. And the impact of interruption on passenger flow distribution indicators such as inbound volume, outbound volume, and transfer volume is studied from both the individual and overall dimensions. The results show that this study provides an effective method for calculating the passenger flow distribution of an extensive urban rail transit network in the case of interruption.

The nanoscale liquid sodium film inside the microporous wick structure is of great importance to understanding the evaporation mechanism of the sodium heat pipe. The novel optimized wick structure is made of several layers of special screen. The surface of each screen exhibits a nanostructure type. Some non-ideal nanostructures may result from experimental faults or limits. And they will have an effect on the evaporation of film. In the present study, molecular dynamics is adopted to investigate this effect. The simulation system consists of the liquid sodium film and the solid surface. The flat surface is set as the reference. Based on the three non-ideal shapes of deposition, the sinusoidal nanostructures, conical nanostructures, and spherical nanostructures are built. The results indicate that the evaporation is suppressed by the above nanostructure surfaces. The weakening effect is through three forms: the potential gradient of the liquid film is intensified and the evaporation difficulty is increased; the heat transfer in the solid-liquid contact region is impeded; the collision heat transfer inside the liquid film is affected due to the delay of the aggregation variation between liquid atoms.

Based on the classical Gursone-Tvergaarde-Needleman (GTN) method, a theoretical model was developed and applied to predict the mechanical behavior and fracture characteristics of hydrogenated zirconium cladding. With full consideration of the influence of hydrogen on plastic stain and void volume in materials, the damage parameters in the classical GTN model were modified. Through benchmarking with experimental data, it was proved that the hydrogen embrittlement phenomenon in zirconium materials can be properly simulated by the modified GTN model proposed in this study. By integrating this modified model into ABAQUS software, numerical simulation of a cladding tensile test was further conducted at different temperature, hydrogen concentration and oxidation conditions. Influence of hydrogen embrittlement on the mechanical properties and fracture characteristics of zirconium cladding was analyzed. The results suggest that the effect of hydrogenation on the cladding mechanical performance varies in magnitude at different temperatures, because hydrogen exists in different forms and also varies significantly at different temperatures. At a relatively low temperature, a considerable amount of hydrogen is presented in the form of precipitates, which is more likely to cause hydrogen embrittlement failure. In addition, the numerical simulation of a cladding tube burst test was carried out under a loss of coolant accident. It was concluded that the influence of hydrogen embrittlement on the burst stress and burst strain was still considerable, despite of the rather high temperature at the burst position.

Recently, nanotechnology attracts more and more attention in heat transfer system. Many studies have shown that using nanofluid as the working fluid can improve the thermal performance of the heat pipe. In this paper, the preliminary effort of applying nano-surface engineering on the heat pipe wick to fabricate the nano-structured wick structure has been carried out. Experimental investigation has been performed to determine the effect of the heat source power, mesh aperture and the wick surface structure on the total thermal resistance of the wick structure. When the heat source power increased from 20 W to 30 W, then to 50 W, the evaporation resistance decreases 25.9% and 13.7% respectively, while the conduction resistance increases 24.8% at first and then decreases 9.5%. The evaporation resistance is less with than without the wick structure, and it increases with the increase of mesh aperture. When the mesh aperture increases from 75 μm to 150 μm, then to 460 μm, the evaporation resistance increases 1.9% and 1.7% respectively. Whereas, the wick structure, as well as the mesh aperture have slight influence on the conduction resistance. The experimental results suggest that the nano structure on the surface can efficiently decrease the evaporation resistance but increase the conduction resistance at the same time. Compared to the wick structure with smooth surface, the evaporation resistance of nano-surface deposited by 0.01% TiO2 and 0.02% TiO2 has decreased 4.3% and 15.5%, respectively. This study provides a reference for the preparation of the heat pipe with greater performance in the future.

Deeply understanding the phase change of thin liquid sodium film inside wick pore is very important for further studying high-temperature sodium heat pipe's heat transfer. For the first time, the evaporation and condensation of thin liquid sodium film are investigated by the Lennard-Jones potential of molecular dynamics. Based on the startup and normal operation of the sodium heat pipe, three different cases are simulated. First, the equilibrium is achieved and the Mass Accommodation Coefficients of the three cases are 0.3886, 0.2119, 0.2615 respectively. Secondly, the non-equilibrium is built. The change of liquid film thickness, the number of gas atoms, the net evaporation flux (Jnet), the heat transfer coefficient (h) at the liquid-gas interface are acquired. Results indicate that the magnitude of the Jnet and the h increase with the basic equilibrium temperature. In 520–600 K (the startup of the heat pipe), the h has approached 5–6 W m−2 K−1 while liquid film thickness is in 11–13 nm. The fact shows that during the initial startup of the sodium heat pipe, the thermal resistance at the liquid-gas interface can't be negligible. This work is the complement and extension for macroscopic investigation of heat transfer inside sodium heat pipe. It can provide a reference for further numerical simulation and optimal design of the sodium heat pipe in the future.

One-dimensional oxygen transport relation is indispensable to study the oxygen distribution in the LBE-cooled system with an oxygen control device. In this paper, a numerical research is carried out to study the oxygen transport characteristics in a gas-phase oxygen control device, including the static case and dynamic case. The model of static oxygen control is based on the two-phase VOF model and the results agree well with the theoretical expectation. The model of dynamic oxygen control is simplified and the gas-liquid interface is treated as a free surface boundary with a constant oxygen concentration. The influences of the inlet and interface oxygen concentration, mass flow rate, temperature, and the inlet pipe location on the mass transfer characteristics are discussed. Based on the results, an oxygen mass transport relation considering the temperature dependence and velocity dependence separately is obtained. The relation can be used in a one-dimensional system analysis code to predict the oxygen provided by the oxygen control device, which is an important part of the integral oxygen mass transfer models.

Study of single bubble behavior under rolling motions can prove useful for fundamental understanding of flow field inside the modern small modular nuclear reactors. The objective of the present study is to simulate the influence of rolling conditions on single rising bubble in a liquid using multiphase Moving Particle Semi-implicit (MPS) method. Rolling force term was added to 2D Navier-Stokes equations and a computer program was written using C language employing OpenACC to port the code to GPU. Computational results obtained were found to be in good agreement with the results available in literature. The impact of rolling parameters on trajectory and velocity of the rising bubble has been studied. It has been found that bubble rise velocity increases with rolling amplitude due to modification of flow field around the bubble. It has also been concluded that the oscillations of free surface, caused by rolling, influence the bubble trajectory. Furthermore, it has been discovered that smaller vessel width reduces the impact of rolling motions on the rising bubble. The effect of liquid viscosity on bubble rising under rolling was also investigated and it was found that effects of rolling became more pronounced with the increase of liquid viscosity.

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect its stress conditions, mechanical behavior and thermal-hydraulic performance. A reliable numerical method is of great importance to reveal the complex evolution of mechanical deformation, flow redistribution and temperature field for the plate-type fuel assembly under non-uniform irradiation. This paper is the first part of a two-part study developing the numerical methodology for the thermal-fluid-structure coupling behaviors of plate-type fuel assembly under irradiation. In this paper, the thermal-fluid-structure coupling methodology has been developed for plate-type fuel assembly under non-uniform irradiation condition by exchanging thermal-hydraulic and mechanical deformation parameters between Finite Element Model (FEM) software and Computational Fluid Dynamic (CFD) software with Mesh-based parallel Code Coupling Interface (MpCCI), which has been validated with experimental results. Based on the established methodology, the effects of non-uniform irradiation and fluid were discussed, which demonstrated that the maximum mechanical deformation with irradiation was dozens of times larger than that without irradiation and the hydraulic load on fuel plates due to differential pressure played a dominant role in the mechanical deformation.

The plate-type fuel assembly adopted in nuclear research reactor suffers from complicated effect induced by non-uniform irradiation, which might affect stress conditions, mechanical behaviors and thermal-hydraulic performance of the fuel assembly. This paper is the Part II work of a two-part study devoted to analyzing the complex unique mechanical deformation and thermal-hydraulic characteristics for the typical plate-type fuel assembly under irradiation effect, which is on the basis of developed and verified numerical thermal-fluid-structure coupling methodology under irradiation in Part I of this work. The mechanical deformation, thermal-hydraulic performance and Mises stress have been analyzed for the typical plate-type fuel assembly consisting of support plates under non-uniform irradiation. It was interesting to observe that: the plate-type fuel assembly including the fuel plates and support plates tended to bend towards the location with maximum fission rate; the hot spots in the fuel foil appeared at the location with maximum thickness increment; the maximum Mises stress of fuel foil was located at the adjacent location with the maximum plate thickness increment et al.

The boiling behavior on nano-structured surfaces is a frontier research direction in nuclear engineering. However, the mechanism of boiling heat transfer on nano-structured surfaces is still unclear. In this study, a depletable micro-layer model and the nano-structure model are proposed based on the Moving Particle Semi-implicit (MPS) method coupled with Meshless-Advection using the Flow-directional Local-grid (MAFL) scheme, also known as the MPS-MAFL method. The developed method in this paper establishes a bridge between the nano-scale surface structure heat transfer and the macroscopic bubble boiling. Only by knowing the nanoparticle size, porosity, and thickness of the nano-structure, the heat transfer of the nano-structure can be considered into the macroscopic boiling bubble growth process. The accuracy of the approach is validated by benchmark cases and experiments, respectively. The present method quantitatively simulates the bubble growth behaviors on nano-structured surfaces for the first time. The results indicate that the heat transfer contribution of the micro-layer to bubble growth was not neglectable, while the proportion of heat transfer rate of the micro-layer on the bared surface was 40.55% at ΔTw = 7.22 ℃ and 32.23% at ΔTw = 10.15 ℃. The heat transfer contributions of the micro-layer and the wicked fluid to the bubble growth in the nano-structured heater were about 42.13%, the ratio of them was 14:11. The current study provides a fundamental base for further investigations.

Due to the special structure, helical coil once-through steam generators (HCOTSGs) can withstand greater thermal expansion stress, have a larger heat exchange area, and cause the secondary flow phenomenon to improve the heat transfer capacity. Thus, it is inevitable to further study the thermal-hydraulic characteristics of the HCOTSG, a vital component of reactors. The optimal physical models were selected for different heat transfer areas by detailed investigation and comparison, to adopt in the TACS, a self-developed thermal-hydraulic analysis code that is accurate and widely-used in HCOTSG simulation. After verification with experiments of Santini, taking the IRIS reactor as the object, the TACS was employed to briefly study the thermal-hydraulic performances of the HCOTSG system at first, and then discuss the influences of thermal and geometric parameters on the HCOTSG. The analysis results showed that: the variation of the primary flow rate greatly affects the heat transfer coefficient of the primary side, while the maximum value of the heat transfer coefficient of the secondary side only changes drastically with the secondary flow rate; when the primary side flow drops to 70%, superheated steam is not generated on the secondary side; when the secondary mass flow rate decreases to a certain extent, the steam superheat degree increases by a small amplitude; the primary side inlet temperature has a little influence on the primary temperature difference, but significant influence on the two-phase zone length of the secondary side; the secondary inlet temperature has very little influence on the HCOTSG system; the inner diameter and the outer diameter of helical tubes have obvious influences, while the helical diameter has not.

Air is widely utilized as working fluid in heat exchangers with high heat flux and high temperature, and the empirical correlations for predicting local and total heat transfer coefficients are indispensable for thermal design and optimization of the heat exchange applications. Experimental investigation on heat transfer characteristics of high temperature air in round tube is conducted. The experimental system is designed and constructed. The data processing method for obtaining wall temperature and heat transfer coefficient is introduced. The measured local and total heat transfer coefficients are compared with the widely adopted correlations, including Taylor's correlation and Gnielinski's correlation. The suggested empirical correlations obtained from the experimental data are presented, which may contribute to the design of heat exchangers concerning with high temperature air.

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA. Keywords: MIDAC, ADS, SBLOCA, Thermal-hydraulic, Core degradation

Small nuclear reactor features higher power capacity, longer operation life than conventional power sources. It could be an ideal alternative of existing power source applied for special equipment for terrestrial or underwater missions. In this paper, a 25kWe heat pipe cooled reactor power source applied for multiple use is preliminary designed. Based on the design, a thermal-hydraulic analysis code for heat pipe cooled reactor is developed to analyze steady and transient performance of the designed nuclear reactor. For reactor design, UN fuel with 65% enrichment and potassium heat pipes are adopted in the reactor core. Tungsten and LiH are adopted as radiation shield on both sides of the reactor core. The reactor is controlled by 6 control drums with B4C neutron absorbers. Thermoelectric generator (TEG) converts fission heat into electricity. Cooling water removes waste heat out of the reactor. The thermal-hydraulic characteristics of heat pipes are simulated using thermal resistance network method. Thermal parameters of steady and transient conditions, such as the temperature distribution of every key components are obtained. Then the postulated reactor accidents for heat pipe cooled reactor, including power variation, single heat pipe failure and cooling channel blockage, are analyzed and evaluated. Results show that all the designed parameters satisfy the safety requirements. This work could provide reference to the design and application of the heat pipe cooled nuclear power source. Keywords: Small nuclear reactor system, Heat pipe cooled reactor, Thermal-hydraulic characteristics, Safety analysis

In a congested large-scale subway network, the distribution of passenger flow in space-time dimension is very complex. Accurate estimation of passenger path choice is very important to understand the passenger flow distribution and even improve the operation service level. The availability of automated fare collection (AFC) data, timetable, and network topology data opens up a new opportunity to study this topic based on multisource data. A probability model is proposed in this study to calculate the individual passenger’s path choice with multisource data, in which the impact of the network time-varying state (e.g., path travel time) on passenger path choice is fully considered. First, according to the number and characteristics of OD (origin-destination) candidate paths, the AFC data among special kinds of OD are selected to estimate the distribution of passengers’ walking time and waiting time of each platform. Then, based on the composition of path travel time, its real-time probability distribution is formulated with the distribution of walking time, waiting time, and in-vehicle time as parameters. Finally, a membership function is introduced to evaluate the dependence between passenger’s travel time and the real-time travel time distribution of each candidate path and take the path with the largest membership degree as passenger’s choice. Finally, a case study with Beijing Subway data is applied to verify the effectiveness of the model presented in this study. We have compared and analysed the path calculation results in which the time-varying characteristics of network state are considered or not. The results indicate that a passenger’s path choice behavior is affected by the network time-varying state, and our model can quantify the time-varying state and its impact on passenger path choice.

Three-layer back propagation network (BPN) and genetic neural network (GNN) were developed in this study to predict the flow boiling heat transfer coefficient (HTC) in conventional and small-diameter channels. The GNN has higher precision than BPN (with root mean square errors of 17.16% and 20.50%, respectively) and other correlations. The inputs include vapor quality x, mass flux G, heat flux q, diameter D and physical parameter φ, and the predicted flow boiling HTC is set as the outputs. Influences of input parameters on the flow boiling HTC are discussed based on the trained GNN: nucleate boiling promoted by a larger saturated pressure, a larger heat flux and a smaller diameter is dominant in small channels; convective boiling improved by a larger mass flux and a larger vapor quality is more significant in conventional channels. The HTC increases with pressure both in conventional and small channels. The HTC in conventional channels rises when mass flux increases but remains almost unaffected in small channels. A larger heat flux leads to the HTC growth in small channels and an increase of HTC was observed in conventional channels at a higher vapor quality. HTC increases inversely with diameter before dry out. Keywords: Back propagation network (BPN), Genetic neural network (GNN), Flow boiling heat transfer coefficient, Conventional channel, Small channel

Modern small modular nuclear reactors can be built on a barge in ocean, therefore, their flow characteristics depend upon the ocean motions. In the present research, effect of rolling motion on flow and friction characteristics of laminar flow through vertical and horizontal narrow channels has been studied. A computer code has been developed using MPS method for two-dimensional Navier-Stokes equations with rolling motion force incorporated. Numerical results have been validated with the literature and have been found in good agreement. It has been found that the impact of rolling motions on flow characteristics weakens with increase in flow rate and fluid viscosity. For vertical narrow channels, the time averaged friction coefficient for vertical channels differed from steady friction coefficient. Furthermore, increasing the horizontal distance from rolling pivot enhanced the flow fluctuations but these stayed relatively unaffected by change in vertical distance of channel from the rolling axis. For horizontal narrow channels, the flow fluctuations were found to be sinusoidal in nature and their magnitude was found to be dependent mainly upon gravity fluctuations caused by rolling. Keywords: Rolling motions, MPS method, Narrow channels, Flow fluctuations, Laminar flow, Friction coefficient

The ever-increasing travel demand has brought great challenges to the organization, operation, and management of the subway system. An accurate estimation of passenger flow distribution can help subway operators design corresponding operation plans and strategies scientifically. Although some literature has studied the problem of passenger flow distribution by analyzing the passengers’ path choice behaviors based on AFC (automated fare collection) data, few studies focus on the passenger flow distribution while considering the passenger–train matching probability, which is the key problem of passenger flow distribution. Specifically, the existing methods have not been applied to practical large-scale subway networks due to the computational complexity. To fill this research gap, this paper analyzes the relationship between passenger travel behavior and train operation in the space and time dimension and formulates the passenger–train matching probability by using multi-source data including AFC, train timetables, and network topology. Then, a reverse derivation method, which can reduce the scale of possible train combinations for passengers, is proposed to improve the computational efficiency. Simultaneously, an estimation method of passenger flow distribution is presented based on the passenger–train matching probability. Finally, two sets of experiments, including an accuracy verification experiment based on synthetic data and a comparison experiment based on real data from the Beijing subway, are conducted to verify the effectiveness of the proposed method. The calculation results show that the proposed method has a good accuracy and computational efficiency for a large-scale subway network.

Two-phase flow instability may occur in nuclear reactor systems, which is often accompanied by periodic fluctuation in fluid flow rate. In this study, bubble rising and coalescence characteristics under inlet flow pulsation condition are analyzed based on the MPS-MAFL method. To begin with, the single bubble rising behavior under flow pulsation condition was simulated. The simulation results show that the bubble shape and rising velocity fluctuate periodically as same as the inlet flow rate. Additionally, the bubble pairs’ coalescence behavior under flow pulsation condition was simulated and compared with static condition results. It is found that the coalescence time of bubble pairs slightly increased under the pulsation condition, and then the bubbles will continue to pulsate with almost the same period as the inlet flow rate after coalescence. In view of these facts, this study could offer theory support and method basis to a better understanding of the two-phase flow configuration under flow pulsation condition.

Underwater vehicle is designed to ensure the security of country sea boundary, providing harsh requirements for its power system design. Conventional power sources, such as battery and Stirling engine, are featured with low power and short lifetime. Micronuclear reactor power source featured with higher power density and longer lifetime would strongly meet the demands of unmanned underwater vehicle power system. In this paper, a 2.4 MWt lithium heat pipe cooled reactor core is designed for micronuclear power source, which can be applied for underwater vehicles. The core features with small volume, high power density, long lifetime, and low noise level. Uranium nitride fuel with 70% enrichment and lithium heat pipes are adopted in the core. The reactivity is controlled by six control drums with B4C neutron absorber. Monte Carlo code MCNP is used for calculating the power distribution, characteristics of reactivity feedback, and core criticality safety. A code MCORE coupling MCNP and ORIGEN is used to analyze the burnup characteristics of the designed core. The results show that the core life is 14 years, and the core parameters satisfy the safety requirements. This work provides reference to the design and application of the micronuclear power source.

Core degradation and material migration research is one of the key areas in severe accident research. A review of core degradation and melting materials immigration research in light-water reactor is important to encourage relevant research. In this paper, both relevant experiments and numerical analyses are reviewed. Due to their high cost, there have only been a few experiments on severe accidents performed. They focus on different aspects of core degradation and material migration, including early stage and late stage accidents in both PWRs and BWRs. All the current experimental data should be fully utilized due to the limited data available. On the other hand, the data available from numerical analyses for severe accidents is very extensive. There are already many systemic severe accident codes developed by different organizations. These codes provide severe accident sequences analysis, severe accident prediction, and are also important in security policy formulation. At the end of this paper, relevant work at Xi’an Jiaotong University is introduced.

A theoretical investigation on the thermal hydraulic characteristics of a new type of passive residual heat removal system (PRHRS), which is connected to the reactor coolant system via the secondary side of the steam generator, for an integral pressurized water reactor is presented in this paper. Three-interknited natural circulation loops are adopted by this PRHRS to remove the residual heat of the reactor core after a reactor trip. Based on the one-dimensional model and a simulation code (SCPRHRS), the transient behaviors of the PRHRS as well as the effects of the height difference between the steam generator and the heat exchanger and the heat transfer area of the heat exchanger are studied in detail. Through the calculation analysis, it is found that the calculated parameter variation trends are reasonable. The higher height difference between the steam generator and the residual heat exchanger and the larger heat transfer area of the residual heat exchanger are favorable to the passive residual heat removal system.