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The deterministic two-step method for the fast reactor neutronics calculation, composed of cross-section homogenization and diffusion or transport core calculations, was widely applied in the fast reactor engineering design and analysis field. The homogenized cross-section calculation method based on Monte Carlo with continuous-energy and fine geometry can provide high-precision cross-sections for advanced fast reactors. The current development status and trends of coupling Monte Carlo-generated homogenized cross-sections with diffusion and transport core calculations were briefly reviewed in this paper. The methods discussed include the Monte Carlo flux-volume homogenization method, the superhomogenization equivalence technique (SPH), and the Monte Carlo flux-moment homogenization method (MHT). The MET-1000 metal fuel fast reactor is used as a benchmark. The SPH equivalent techniques are widely used to preserve the reaction rates of a reference heterogeneous model and a homogenous model. In this paper, the SPH was applied to the control rods' cross-section address to improve the diffusion core calculations. This equivalence technique reduces the overestimation of the control rod worth using the diffusion core solver from 13.5% to 0.35% and improves power distribution prediction accuracy. With SPH correction, the MC/diffusion in this work exhibits about

The steam generator tube rupture (SGTR) accident is one of the design basis accidents of lead-cooled fast reactor (LFR). In the event of such an accident, a subcooled water jet is generated in the lead pool of primary circuit, which forms a complicated distribution of the liquid metal-gas two-phase flow. The detection of the two-phase interface and migration of steam bubbles is an significant part of reactor safety analysis. Due to the opacity, corrosiveness, high temperature of liquid metal, and the current limitations of measurement methods involving liquid metal-gas two-phase flow, a diamond-like carbon (DLC) coated wire-mesh sensor (WMS) based on electrical principles was proposed in this paper. The sensor could achieve the image of the liquid metal-gas two-phase distribution inside the flow channel, while preventing the corrosion failure of the sensor. Based on the COMSOL software, a numerical model of the DLC coated WMS was established. Selecting the materials of two phases as liquid GaInSn alloy and air, the simulation of the electric field in the liquid metal environment containing a single bubble was performed, and an electric signal value matrix representing the liquid metal-gas two-phase was output. The signal matrix was further normalized through the linear relation to obtain the two-phase distribution image, which verified the application feasibility of the coated WMS in liquid metal. In addition, a prototype of the DLC coated WMS was manufactured, and the numerical simulation results were experimentally validated in order to demonstrate the rationality of the numerical model. Furthermore, the influence of different structural parameters on the measurement accuracy of coated WMS was studied in detail by changing the transverse interval, axial interval, electrode diameter, coat thickness, and the size of air bubble. The simulation results show that the electric field distribution in the liquid metal-gas two-phase environment differs significantly from that in conventional fluid medias such as water-gas due to the high relative permittivity of liquid metal. On the other hand, considering the resolution of the WMS and the phenomenon of electric signal crosstalk, the transverse interval of electrode should be selected between 2-3 mm, and the axial interval should be selected between 1.5-2 mm to achieve high measurement accuracy. The coat thickness and electrode diameter have little influence on the measurement accuracy, but to reduce the intrusion effect of the WMS, the electrode diameter should be selected between 0.4-0.5 mm and the coat thickness should not exceed half of the electrode radius. This study could guide the practical application of DLC coated WMS in the liquid metal-gas two-phase environment and provide a technical basis for the analysis of multi-phase distribution under SGTR accidents.

In the development of critical heat flux (CHF) correlation, the mixing factor β (i.e., the turbulent thermal diffusivity) is a characteristic parameter representing the cross-mixing effect between sub-channels. At present, the mixing factor obtained through the cross-mixing experiment is essentially the macroscopic equivalent quantity of averaging the cross-mixing effect (such as turbulent cross-mixing, flow sweeping, etc.). In fact, the mixing effect caused by the grid will decrease along the downstream, which will lead to a certain deviation in the local parameters calculated by the average mixing factor. The influence of the deviation on the DNBR limit value of the CHF correlation and the safety margin is not clear. Thus, the CHF correlation matching with the sub-channel analysis program FLICA-ⅢF was developed by using the minimum DNBR point method. Owen criterion was used to determine the DNBR limit. After achieving the development methods of CHF correlation, the prediction effect, DNBR limits and actual maximum power of correlations of different mixing factors were compared and analyzed. The results indicate that in the range of experimental conditions, the local pressure of minimum DNBR point does not change with the change of mixing factor, the local mass flow rate increases with mixing factor, and the local quality decreases with the increase of mixing factor. The deviation of mixing factor will increase the DNBR limit of its correlation, but there is no linear relationship between the two. Besides, in actual conditions, the maximum power calculated by the correlations of different mixing factors is different. While the maximum power value of the precise mixing factor is also the largest. In conclusion, the more precise mixing factor determined by experiment has higher accuracy and economy in fitting correlation. Therefore, it is valuable and meaningful to consider the fine subchannel of grid mixing effect.

The helically metallic fuel is promising to improve the power density and safety margin of reactors core by its advantages of high thermal conductivity, large heat transfer area-to-volume ratio and continual inter-channel mixing. However, helical geometry and metallic U-Zr fuel will introduce challenges in the analysis of neutron physics, thermohydraulics and mechanics performance. Thus, in the current work, the analysis methods were developed for the characteristics of helically metallic fuel in neutron physics, thermohydraulics, mechanics and multi-physics coupling. For neutron physics, a 3D continuous-energy Monte Carlo method was employed to address the complex geometries and the complex intermediate neutron energy spectrum and generate few-group cross-sections. In core calculations, the 3D method of characteristics (MOC) with high geometric flexibility, accuracy and convergence was utilized. This method has been verified in 3D calculations of helically metallic fuel core. The results demonstrate that both cross-section generation and core calculation have achieved high accuracy and effectively improved computational efficiency. For thermohydraulics, the experimental technique including visual measurement and wire mesh sensor were used to measure the mixing characteristics of single and two-phase flow in helically metallic fuel rod bundle. CFD and subchannel analysis codes were developed to predict the boiling and critical heat flux. The inter-channel mixing was dominated by the flow sweeping mixing, while the turbulent mixing was insignificant. The vapor phase crowded at the elbow of the rods, where the boiling crisis was triggered. For mechanics, the molecular dynamics method was employed to predict the fundamental thermal-mechanical properties. The correlations were proposed for the thermal conductivity and elasticity modulus of U-Zr alloy under irradiation conditions with pores. The finite element tool Abaqus was used to analysis the thermal-mechanical performance of helically metallic fuel rod bundle. The maximum stress was found at the blade tips where adjacent rods contact with each other. The maximum stress varied from 326.7 MPa under fresh condition to about 313.3 MPa at 14.1% FEMA because of creep, which was far lower than the failure strength of the rod. The structure integrity can be ensured during the full life cycle. Based on the individual analysis method, the multi-physics coupling framework combining the neutron physics, thermohydraulics and mechanics was developed for the design and performance evaluation of the fuel assembly and reactor core with helically metallic fuel rods. Based on this multi-physics analysis tool, a novel boron-free small modular pressurized water reactor NETH-HCF175M design using helical-cruciform metal fuel was proposed. The core can achieve a cycle length of about 1 360 EFPD as the reactivity of the core is completely controlled by burnable absorbers and control rods.

BackgroundNatural circulation systems are widely used in the nuclear industry for decay heat removal, post-accident containment cooling, and the cooling of radioactive waste storage facilities. Because a large number of spent fuel assemblies are densely arranged in the spent fuel pool, the spent fuel assemblies in the central area of the spent fuel pool can be approximated as having adiabatic boundary conditions. In the event of a coolant loss accident, the heat generated by the spent fuel assemblies can be exported only by the natural circulation of air.PurposeThis study aims to develop analytical models for fuel assembly natural circulation characteristics independent of the results of computational fluid dynamics (CFD) to investigate the natural circulation characteristics of spent fuel pools following a coolant loss accident.MethodsBased on the experimental results of a fuel assembly pressure drop of a full-scale pressurized water reactor (PWR), the Darcy–Forchheimer model was firstly revised to predict the pressure drop of humid air flowing through the fuel assembly. Models for the fuel assembly natural circulation flow rate and peak cladding temperature were then established, and the effects of the relative humidity of air on the models were considered. Finally, these models were applied to investigating effects of total heating power, ambient temperature, and relative humidity on the natural circulation flow rate and peak cladding temperature of the fuel assembly.ResultsThe results show that the models accurately predict the natural circulation flow rate and peak cladding temperature of the fuel assembly under different heating powers, and the errors are less than 25% and 20%, respectively, as compared with the experimental measurement values.ConclusionsThe developed models in this study can be used to study the natural circulation characteristics of full-scale PWR fuel assemblies.

Reactor fuel assemblies may be biased and bowed as a whole due to factors such as manufacturing and installation deviations or external forces. This consequence may have an impact on heat transfer and criticality, and even the possibility of edge-rod criticality. In order to evaluate the impact size, a set of high-quality dynamic mesh generation technology was developed through CFD. The dynamic mesh technology was used for the base condition to make the overall displacement or bowing of the fuel rods. Using the Eulerian two-fluid model and the improved RPI wall boiling model, the CHF of the 5×5 rod bundle channel with different offset distances and bowing degrees was predicted, and the CFD prediction results were verified with the CHF look-up table value. The results show that the CHF predicted by the non-deformation and small deformation conditions are in good agreement with the values of the sub-channel parameter CHF look-up table, while the large deformation condition triggers the side rods critical and CHF look-up table is not applicable. Because the CHF table is made based on an 8 mm round tube, the non-8 mm round tube or rod bundle structure (such as the central sub-channel) can be corrected by the hydraulic diameter correction factor, and the prediction accuracy is often high. However, at the edge and corner channel position of the rod bundle assembly, the heated boundary and wet circumference are not uniform, and the offset or bowing of the fuel rods squeezes the corner channel, resulting in a more unbalanced distribution of the heated boundary and wet circumference at this location. As a result, the CHF look-up table is not applicable. The critical occurrence position and numerical value under different deformation conditions were compared, the distribution of mass fluxes in different sub-channels at the critical height was also compared, and the sensitivity of various deformation types to CHF was summarized in this paper. For the calculation results of the undeformed fuel assembly, the flow rate of the central sub-channel is the highest, the side channel is the second, and the corner channel is the smallest. For the deformed fuel assembly, the offset and bowing change the flow distribution of the sub-channels, the flow of the corner channels decreases, and the flow of the central sub-channel increases. When the gap of the corner channel is so small that the coolant flow cannot effectively take away the bubbles on the surface of the fuel rods, causing local congestion of the bubbles, the side rods will be critical. The XY-type condition has the strongest influence on the corner channel flow, but the weakest effect on the overall flow distribution, and the C-type bowing condition has the strongest influence on the overall flow distribution. In the calculation condition, the axial power is uniformly biased and bowed by 0.5 mm without changing the critical axial position, while bowing by 0.8 mm and 1 mm affects the critical position. The larger the three deformation degrees, the smaller the predicted CHF value and the increase of the critical probability of the side rods. Among them, the XY-type deformation is the least sensitive to the CHF value, while the C-type bowing with a cosine axial heat flux has the greatest impact on the CHF value. Compared with the same degree of bowing, the critical heat flux density value of axial power cosine is smaller. When bowing 0.5 mm, the critical position of the axial power uniform condition occurs at the center, while the critical position of the axial power cosine condition occurs at the edge. That is, the cosine power may also be the cause of the criticality of the side rods.

BackgroundSteam generator tube rupture (SGTR) accidents in lead-bismuth cooled fast reactors result in the generation of numerous steam bubbles owing to the interaction of the high-temperature liquid lead-bismuth eutectic (LBE) from the primary circuit and high-pressure subcooled water of the secondary circuit. These steam bubbles carried by the LBE may enter the core, causing local heat transfer deterioration and a power transient, seriously affecting reactor safety.PurposeThis study aims to elucidate the movement and dynamic behaviors of steam bubbles in liquid LBE and develop a drag coefficient model applicable to bubble migration in LBE for assessing the safety of the core during SGTR accidents.MethodsBased on the coupled level-set and volume-of-fluid (CLSVOF) method, a three-dimensional numerical model for calculating the drag coefficient of steam bubbles in LBE was established to study their movement and dynamic behaviors. Firstly, the deformation, velocity, and trajectory characteristics of bubbles in LBE were analyzed, and their drag coefficients were estimated. Then, the simulated drag coefficient values were compared with those from existing models whilst the prediction performance of the model proposed by Tomiyama was superior to those of the others. Finally, Tomiyama's drag model was optimized by introducing the We number, and its applicability was analyzed.ResultsAnalysis results show that the errors of the existing drag coefficient models are large under the condition of bubble–LBE two-phase flow, and the calculation error of the optimized model for the bubble drag coefficient in LBE is within 15%.ConclusionsThe feasibility of drag calculation model optimized in this study is verified, demonstrating its suitability for calculating the drag coefficient of steam bubbles in LBE.

As a new type of reactor fuel assembly, helical-cruciform fuel (HCF) assembly has some advantages. The cruciform section shortens heat conduction path and the twist structure strengthens inter-channel mixing, which can upgrade core power density with enough safety margin. The periodic self-supporting structure fixes the assembly instead of spacer grid to reduce flow resistance. However, because of the small contact area at self-supporting structure the stress concentration is prone in contact position, which has a negative effect on the integrity of cladding. To solve this problem, the thermal-fluid-force coupling characteristics of typical HCF assembly were studied in this research. The flow and heat transfer of coolant were simulated by ANSYS Fluent to obtain velocity, temperature and vapor fraction distributions at single-phase and two-phase flow. Steady-State Thermal and Static Structural module in ANSYS were used to analyze the thermal and mechanical behavior of HCF assembly, respectively, and the differences of stress and strain distributions between single-phase flow and two-phase flow were found. The thermal-fluid coupling was two-way, and the data were exchanged in System Coupling module. Because the deformation of HCF assembly was too small to change flow and heat transfer of coolant, the thermal-force coupling was one-way. The results show that the cross flow intensity of coolant near the cladding is greater than that in the center of channel, and the direction of vortices at the two places is opposite. The cross flow affects the coolant temperature and vapor fraction distribution and makes the vapor fraction fluctuate with a period of 45° along flow direction at two-phase flow. The temperature at outer surface of cladding and pellet center also fluctuates with a period of 90° along flow direction except for the entrance part at single-phase flow and two-phase flow. Because a small amount of bubbles at the cladding surface can enhance the heat transfer, the heat transfer coefficient at two-phase flow is twice as large as that at single-phase flow. The maximum Mises stress of 300 MPa is in the lobe region of HCF assembly with 0.1 plastic strain. The thermal stress of HCF assembly in the lobe region is mainly related to the contact constraint, but in the valley region it is mainly related to temperature gradient inside HCF. At the cladding of HCF assembly, the Mises stress in the valley region and the plastic strain in the lobe region at single-phase flow and two-phase flow are the same, but in the lobe region the Mises stress at two-phase flow is 30 MPa smaller than that at single-phase flow.