A detailed safety assessment of innovative Generation IV reactor designs with heavy-liquid metal coolants, such as lead and lead-bismuth eutectic (LBE), requires an evaluation of the maximum core temperature in several postulated scenarios. Considering the complex geometry of fuel assemblies (FAs), and the low Prandtl number of the coolants, this flow scenario is challenging for the models used in numerical simulations, e.g. for relating the turbulent transport of momentum and heat. Thus, reliable experimental data are needed for validation. In recent years, a series of comprehensive heat transfer tests in fuel pin bundle simulators was performed at ENEA (Italy) and KIT (Germany) in the framework of the European collaborative projects THINS and SEARCH. Both grid and wire spacer geometries are considered, in a wide range of operating conditions (temperature, flow velocity and power density) representative of the ALFRED and MYRRHA fuel assemblies, in natural and forced-convective flow regimes. Although different experimental approaches were followed by each group (e.g. thermocouple position and average data treatment), there is a relatively good agreement on results in the overlapping regions. These experimental studies indicate that the mean Nusselt number is in well agreement with the predictions of empirical correlations developed for sodium systems. In particular, for wire-spaced FA, heat transfer results show values close to the Kazimi-Carelli correlation both for low and high flow rates at ENEA and KIT respectively. For grid-spaced FA, results are more in agreement with Ushakov correlation. Furthermore, large temperature differences are measured by thermo-couples installed at selected rods and sub-channels. A discussion on the influences of the spacer design and bundle size is included. This wide comparison allows an overview of the research on the HLM cooled fuel assembly in Europe.

Heat removal systems are of major importance for both present and future nuclear power plants as they belong to the set of systems devoted to ensure the integrity of the reactor core and to avoid core damage. The past experience and lessons learned on this topic suggest to adopt passive safety systems which can perform the safety function independently from operators’ actions and external energy sources, ensuring long term reactor cooling. Application of these systems to innovative reactor concepts such as (heavy) liquid metal reactors poses a problem related to the characteristic properties of the coolant: as the final heat sink of passive safety systems is often the external environment, the liquid metal will eventually undergo a phase change and solidify at the end of a complex dynamic process. The solidification of the coolant may have important effects on the transient behavior if it happens at an early stage of an accident, as the main flow path of the heat exchanger can be blocked by the coolant freezing while the decay heat in the core is still sufficiently high and need to be efficiently removed. An innovative passive safety system has been proposed for the decay heat removal system of ALFRED reactor (DEMO LFR, Gen.IV) where the issue of early coolant freezing is prevented. The innovation has been object of a patent and the system is potentially able to avoid solidification by reducing the amount of heat removed from the primary system by means of non-condensable gases passively injected into the water/steam mixture, which induce heat transfer degradation. Several numerical studies have been performed during the past years, but a complete validation of the operating principle requires an experimental assessment and characterization. To this aim the SIRIO experimental facility, scaled on the DHR of ALFRED, has been conceived. Several design activities have been performed so far for the development of the facility, such as scaling analysis on the basis of ALFRED DHR to determine the facility size, numerical simulations by means of RELAP5-3D to determine whether the facility is able to reproduce the expected physical phenomena and numerical simulations by means of Ansys CFX to investigate the performance of a heating system simulating the primary system of ALFRED based on a molten salt annulus. The present paper describes the design activities performed and provides insights on the methodologies adopted, as well as the current status of the design of the SIRIO facility.

The Brayton cycle with supercritical carbon (S-CO 2 ) as working medium is one of the most promising new nuclear power systems. Turbine is the key device during the working process in the Brayton power cycle. The turbine structural presents small size and extremely high rotational speed for the special physical properties of S-CO 2 , which increase the difficulty for the structural design and strength safety significantly. According to the aerodynamic design and optimization results of 200 kW S-CO 2 radial inflow turbine, this paper proposes a detail structural design and analysis method for turbine impeller. Based on the three-dimensional blade profile data and meridional planes data, key structural design parameters are chosen and the parametric geometry model is established by CAD tools. On this basis, numerical simulation models of turbine are established to analyze the structural strength in detail. Then the influence of parameters on the turbine impeller strength is studied by a series of finite element numerical procedures. The influence mechanisms of key structural design parameters on impeller strength are discussed. Moreover, the final model of turbine impeller is obtained by parameter comparison and selection. The results show that for the initial model, the maximum von-Mises equivalent stress is 400.10 MPa, the maximum radial deformation is 0.0333 mm and the maximum axial deformation is 0.0770 mm. For the final model, the maximum von-Mises equivalent stress is 294.26 MPa, the maximum radial deformation is 0.0279 mm and the maximum axial deformation is 0.0769 mm. The maximum von-Mises equivalent stress and maximum radial deformation of structural decreases 26.45 % and 16.22 % respectively compared with the initial model. As a result, the impeller structural strength safety margin is obviously improved by the parameter analysis.

Nuclear energy is an important way to solve energy shortage and pollution problems today. Therefore, China is vigorously developing nuclear energy and is facing huge demands for talents. However, nuclear engineering education has been severely hampered, particularly in its experiment aspect because of the Fukushima accident. The original sub-critical nuclear reactor at Tsinghua University (THU) was stopped, forcing students to use computers to conduct relevant nuclear simulation experiments. The emerging of supercomputers and commercial numerical simulation softwares has provided enough hardware and software support for complex calculations required in numerical simulation of nuclear reactors. Thus numerical simulation could be integrated into nuclear engineering education. With its ease of use, quick acquisition and direct visualization of results, numerical simulation can save much time and money. Besides, it is convenient to change simulation conditions which is helpful for academic research and talent development. This paper starts with THU’s nuclear engineering talent training mode, and taking the course “Advanced Nuclear Reactor Thermal Analysis” as an example, discusses the applications of numerical simulation softwares FLUENT and COBRA-TF in this course. Finally, the analysis shows that numerical simulation performs well in conceptual understanding and experimental design. It plays a significant role in nuclear engineering education, which provides an important reference for the new mode of nuclear engineering education.

In this study, computational fluid dynamics simulation of gas-droplet flow in a spray injection in a THAI vessel is performed and the heat and mass transfer between the gas and the droplets are investigated. The purpose of the numerical simulation was validation of a CFD model to use in a hydrogen risk assessment in a containment building under severe accident conditions. The numerical simulation has to cover the physical phenomena that occur when spray water is injected downward into stationary gas in a large closed vessel. In order to model the two phase flow of the gas and droplets, two different modelling approaches are applied, one based on the Lagrangian approach and the other based on the Eulerian approach. In both approaches, the gas-droplet interactions are modelled using the two-way Lagrangian particle model and dispersed multiphase model to simulate gas-droplet interaction, two way coupling of momentum, and the heat and mass transfer between gas phase and droplet phase. For droplet heat and mass transfer, convective heat transfer and diffusion limited mass transfer are assumed. The calculated change in the temperature of the gas shows qualitative agreement with the experimental results but the gas temperature decrease was over predicted with both approaches. The calculated pressure shows good agreement with experimental results in both approaches. Both approaches also show similar prediction of temperature and pressure; therefore, they can both be applied to the containment analysis. The heat transfer coefficient in the gas-droplet heat transfer should be modified to be suitable for the spray flow. When considering that the spray system in the containment building is to be operated de-pressurized, the CFD model for the spray flow should be suitable for the thermal hydraulic analysis of the containment building.

Molten salt reactors are a very promising option for the development of highly innovative solutions for the nuclear energy production of the future. The techniques used to model thermal hydraulics of a molten salt fast reactor when frozen salt wall technology is applied to the core vessel wall are presented here for 2D numerical models of a hyperboloid reactor core region with a heat exchanger was applied in Code_Saturne. A 3D simulation of the fluid flow and heat transfer with 16 recirculation loops containing the heat exchangers is also presented. It was found that there is strong cooling in separated flow regions in the external heat exchanger, which freezes where the porous model is applied.

This article reports a CFD-benchmark with the purpose of validating different turbulence modelling approaches for the transient heat transfer due to mixing of hot and cold flow in a T-junction including the wall. This validation exercise has been carried out within the MOTHER project. In the framework of the project, new experiments were performed with a novel measurement sensor allowing the measurements of the fluctuating wall temperature inside the solid pipe wall. The tests were performed for two different Reynolds numbers (Re) 40000 and 60000 and for two different T-junction geometries; a sharp corner and a round corner. The present article reports the synthesis of the CFD validation for a sharp corner T-junction for Re = 40 000. The CFD validation study has been performed using four different CFD softwares, namely STAR-CCM+, Code_Saturne, LESOCC2 and Fluent. In addition, five different turbulence models i.e. wall-function Large Eddy Simulation (LES), Deatched Eddy Simulation (DES), Partially Resolved Numerical Simulation (PRNS), Unsteady Reynolds Averaged Navier-Stokes URANS and RANS were used to perform the CFD computations. The validation exercise has shown that LES gives the best agreement with the experimental data followed by hybrid (LES/RANS), URANS and RANS models, respectively. The velocity and the thermal fields in the fluid region are correctly predicted by the proper use of the LES modelling, whereas, the accurate prediction of the thermal field in the solid requires very long sampling time in order to achieve a statistically converged solution, which of course requires an enormous computational power. Therefore, the statistical convergence of the thermal field in the solid has been found to be a bottleneck in order to accurately predict the temperature fluctuations in the wall. However, measuring the small amplitude temperature fluctuations is also associated with an uncertainty so the disagreement between CFD and measurements (of the order of 10 %) can also be attributed, in part, to uncertainties in the measurements.

In the pipeline system of nuclear industry, shock wave pressure in a pipe will be caused by the fast closing check valve after the pumping stops. This phenomenon is known as water hammer, which brings hidden danger to the security and reliability of the pipeline system. Specially, water hammer may cause serious damage on the pipeline system by the valve misoperation, by the valve malfunction, or by other unexpected events. A vortex diode is used as a highly reliable check-valve in nuclear applications, where it mainly benefits from the intrinsic properties of no moving parts and no leakage. In this paper, we proposed a novel method based on a vortex diode to protect water hammer. In the traditional analysis, a simple one-dimensional (1D) model is often used to simulate the water hammer. However, it is difficult to get the transient flow characteristics in a vortex diode using a 1D model. Thus, a three-dimensional (3D) model using computational fluid dynamics (CFD) is proposed to analyze water hammer in a pipeline system with a vortex diode. The 3D model was firstly verified by comparing the numerical results of CFD with experimental results of a water hammer test. Based on the 3D model, the water hammer was simulated at different inlet conditions in a pipeline system with a vertex diode. In order to investigate the vortex diode used as a leaky check-valve, the inlet pressure was decreased by the corresponding value of pump head to simulate the pump stop after the quasi-steady state was achieved in the vortex diode. It is found that the pressure fluctuation of water hammer is comparable to the pump value, which is not varying with initial velocity in the pipeline system. Thus, we have proved that a vortex diode in the pipeline system acts significantly in suppressing pressure fluctuation of water hammer. This study presents a CFD-based numerical method for water hammer and could be useful in protecting water hammer in nuclear industry.

In order to study the dynamic characteristics and fluid structure interactions of tubular structures under the action of fluid in reactor, such as fuel rod bundles and heat transfer tube bundle of steam generator, the dynamic equations and the acoustic wave equations of structures are discretized by finite element method. The acoustic wave equations are simplified from continuity equation and momentum equation of fluid field. Based on the fluid structure coupling method, the dynamic characteristics of the tube under the internal flow, external flow and combined action of internal and external flow are calculated respectively. The influence of flow field domain, element type and grid number on the dynamic characteristics of the tube is also analyzed. Secondly, based on the computational fluid dynamics and computational structural dynamics, the interaction between the two physical fields of fluid and structure is considered simultaneously. The finite volume method is used to discretize the fluid control equations and the turbulent flow is investigated using the large eddy simulation method (LES). The Newmark algorithm is used to solve the structural dynamic equations. Combined with the dynamic mesh control technique, a numerical model for flow induced vibration of three-dimensional flexible tube is established. Finally, the flow induced vibration of a three dimensional flexible single tube and a square arrangement tube bundle is calculated using the numerical model. By comparing with the existing research results, it is found that the numerical simulation results are in good agreement with the experimental results. Thus, the correctness of the model is verified. It is also shown that the numerical model established in this paper can be used to simulate the dynamic characteristics and flow induced vibration of tubular structures.

In the accident at the Fukushima Daiichi Nuclear Power Plant, fuel rods and reactor internals were overheated due to the radioactive decay heat and the oxidization of fuel cladding, and relocated to the lower region of the pedestal. To retrieve the fuel debris, location, distribution and characteristics of the fuel debris are quite important. However, this information still has not been clarified. To estimate the distribution of fuel debris, since experiments cannot provide sufficient information, numerical simulation will be a useful tool. There are some numerical codes for calculating conditions inside a reactor at severe accidents, however, those codes include assumptions and/or related to melt relocation, and cannot simulate relocation process directly. In order to simulate the relocation and corium spreading behavior in PCVs without assumptions and scenario, a numerical simulation code that can phenomenologically evaluate the melting phenomena is required. Then, a phenomenologically-based numerical simulation code for predicting the melting core behavior, including solidification and relocation based on the computational fluid dynamics, JUPITER, has been developed in JAEA. In this paper, by using JUPITER, we simulated the corium spreading process inside a pedestal. As a result, we confirmed that JUPITER has potential to estimate the debris distribution inside the pedestal region.

Detailed knowledge of a coolant flow in a fuel assembly of a reactor core has always been a major factor in the design of new nuclear systems. In this regard, traditionally adopted subchannel analysis codes cannot take into account local phenomena, which are quite essential. On the other hand, Computational Fluid Dynamic (CFD) is being recognized as a valuable research tool for thermal-hydraulics phenomenon in the fuel assembly geometries. Because of the high Reynolds number and geometric complexities, the practical CFD calculations are mostly limited to pragmatic Reynolds Averaged Navier-Stokes (RANS) type modelling approaches. A good prediction of the flow and heat transport inside the fuel rod bundle is a challenge for such RANS turbulence models and these models need to be validated. Although the measurement techniques are constantly getting improved, however, the CFD-grade experiments of flow mixing and heat transfer in the subchannel scale are often impossible or quite costly to be performed. In addition, lack of experimental databases makes it impossible to validate and/or calibrate the available RANS turbulence models for certain flow situations. In that context, Direct Numerical Simulation (DNS) can serve as a reference for model development and validation. The aim of this work is to design a numerical experiment in order to generate a high quality DNS database for a tight lattice bare rod bundle, which will serve as a reference for the validation purpose. The considered geometric design is based on the well-known Hooper experiment, which contains a bare rod bundle with pitch-to-diameter ratio of P/D = 1.107. Performing a DNS computation corresponding to the Hooper experiment requires a huge computational power. Hence, a wide range of unsteady RANS (URANS) study has been performed to scale-down the Reynolds number such that it is feasible for a DNS computation and at the same time it still preserves the main flow characteristics. In addition to the flow field, a parametric study for three different passive scalars is performed to take into account the heat transfer analysis. These passive scalars correspond to the Prandtl numbers of air, water and liquid metal fluids. The heat transfer of these three fluids has been studied in combination with two different boundary conditions at the walls, i.e. a constant temperature and a constant heat flux. Finally, the obtained URANS results are used to compute the Kolmogorov and Batchelor length scales in order to estimate the overall meshing requirements for the targeted DNS.

Steam explosion is one of the consequences of fuel-coolant interactions in a severe accident. Melt jet fragmentation, which is the key phenomenon during steam explosion, has not been clarified sufficiently which prevents the precise prediction of steam explosion. The focus of this paper is on the numerical simulation of the melt jet behavior falling into a coolant pool in order to get a qualitative and quantitative understanding of initial premixing stage of fuel-coolant interaction. The objective of our first phase is the simulation of the fragmentation process and the estimation of the jet breakup length. A commercial CFD code COMSOL is used for the 2D numerical analysis employing the phase field method. The simulation condition is similar to our steam explosion test supported by the ALISA (Access to Large Infrastructure for Severe Accidents) project between European Union and China, and carried out in the KROTOS test facility at CEA, France. The simulation result is in relatively good agreement with the experimental data. Then the effect of the initial jet velocity, the jet diameter and the instability theory are presented. The preliminary data of melt jet fragmentation is helpful to understand the premixing stage of the fuel-coolant interaction.

Jet breakup and fragmentation are important phenomena to be well understood during a core-disruptive accident of sodium-cooled fast reactors. The three-dimensional two-phase lattice Boltzmann model developed previously by the authors is improved in numerical stability used to simulate the hydrodynamic process of melt jet breakup. Nonorthogonal central moments is successfully introduced into the model. Numerical simulations of FARO-TERMOS experiments demonstrate the enhancements in stability of the present model. The simulations with two types of grid resolutions show the effect of spatial resolution on the results.

The highly turbulent flow inside a pressurized water reactor makes unpractical the use of scale resolving simulations, due to the large number of space and time turbulent structures. The high computational cost associated with typical large eddies simulations or direct numerical simulations techniques is unsuitable due to the large spatiotemporal resolution required. Partially averaged Navier-Stokes turbulence model is presented as bridging model between Reynolds averaged Navier-Stokes equations and direct numerical simulations. As filtered representation of the Navier-Stokes equations, the model is able to continuously shift its energy-based filter, inside the turbulence spectrum, being able to resolve the turbulent scales of interest. The choice of energy based cut-off filters gives the chance to directly impose the degree of needed resolution, where the most important large scales unsteadiness are resolved at minimal computational expenses. The partially averaged Navier-Stokes modelling approach has been tested for a Reynolds number of 14,000, inside a 5 × 5 fuel bundle, with a single spacer grid and split-type mixing vanes. Four different filters have been tested, whose resolution ranged from Reynolds averaged Navier-Stokes and large eddy simulation. A comparison with large eddy simulation will be presented. The results show that the partially averaged Navier-Stokes modeling produces results comparable to those of large eddy simulation when the appropriate cut-off energy filter is chosen. The turbulence models results will be compared with the available particle image velocimetry experimental data.

In this study we used Nek5000, an open-source, high-order spectral element CFD code developed at Argonne National Laboratory (ANL), to model the coolant flow in spacer grids. Two fuel assembly configurations were studied: 2 × 2 and 5 × 5 fuel rod arrangements. The simulations for the 2 × 2 case were based on previous studies, simulating one span of the 2 × 2 fuel rod configuration including a surrogate spacer grid and mixing vane design with typical features of spacers for energy production. Dual periodic boundary conditions were applied in the spanwise direction to take the crossflow into consideration. The study of the 5 × 5 fuel assembly was performed as part of the ANL–Framatome collaboration for advancing computational fluid dynamics (CFD) tools. An advanced numerical model was developed to simulate the experimental setup provided by Framatome. For the 5 × 5 fuel assembly study, two cases of flow geometry were simulated with Nek5000: balanced and unbalanced configurations. In the balanced flow the coolant was entering the fuel rod assembly through 121 uniformly spaced inlet holes arranged in an 11 × 11 matrix. The unbalanced case, on the other hand, featured 14 larger holes placed on only one side of the horizontal plane. Nek5000 accepts only hexahedral meshes, which bring a great challenge to the meshing process for a spacer grid fuel assembly. A tet-to-hex meshing strategy was applied to handle the complex geometric features. A tetrahedral mesh was created first, and then each tetrahedral element was converted into four hexahedral elements. Boundary layers were extruded to fit to the exact geometry. In order to account for transient flow characteristics, the large eddy simulation approach was applied in this study. The employed subgrid-scale model relies on explicit filtering, which has been proven valid for many engineering-scale simulations. We present here the simulation results obtained for both the 2 × 2 and 5 × 5 fuel assemblies.

In fast reactor, the coolant flows into the inside of the fuel assembly through the inlet hole, so according to the design requirement, it is necessary to determine a suitable inlet hole model to make the pressure drop and velocity reasonably. In this paper, three hole models are researched and the distance of adjacent hole of every model is different. Based on the hydraulic test of fuel assembly in fast reactor, a simplification model of distribution header and assembly pin has been built. The computational fluid dynamic code—CFX was used to simulate the pressure field and velocity field of the assembly pin with k-e turbulence model. The results show when the distance of adjacent hole increases, the change of pressure and velocity in the hole area is gently, which makes little vibration. The research will be the reference for the following simulation work and assembly hydraulic test.

In the reactor core of a pebble bed HTGR, fuel pebbles are in random arrangement. The coolant flow passages are complicated. Under the air and water ingress accident conditions, the partially enriched oxidizing gas corrodes the fuel pebbles by oxidation reaction. How the oxidizing gases (oxygen and water vapor) diffuse in the pebble bed will greatly influence the oxidation process. Analysis of the mass mixing is important for analyzing the fuel element graphite corrosion and safety of HTGR. In this paper, a three-dimensional simple cubic structure packed cells of pebble bed core model with surface contact method was established. The flow, and mass mixing processes were numerically and theoretically studied using numerical simulation and porous media methods. The effects of molecular diffusion, turbulent diffusion and mass dispersion on mass mixing effect were discussed respectively.

Radiation-induced damage and degradation to shielding and structure materials in nuclear reactors is one of the key limiting factors that affect the safety considerations and the lifetime. Neutron radiation damages materials mainly by exciting a number of Primary Knock-on Atom (PKAs). PKAs induce displacement cascades, causing microstructure changes and mechanical degradations in materials. Computer simulations are used to model this complex process. Knowing the PKA spectrum accurately is important because PKA spectrum and the input of computer simulation are coupled. In this work, we aim to obtain the PKA spectrum in Zr-based alloys with relatively low Zr concentration by using the Geant4 software. Some new functions were added by reprogramming in Geant4. We found that the energy spectra of PKA in Zr 2 Cu and Zr 2 Ni are mainly caused by Zr atom, and the shape of the average energy spectra are similar with pure Zr. The number of PKA distributions and the energy deposition in these two Zr-based metals are similar with pure Zr but different than those in pure Ni. These finding indicate that the metallic elements of Cu and Ni have small impact on PKA spectrum in Zr-based alloys, which illustrate that simplified simulation models are feasible when using computer for simulating. Moreover, it has great significant for the calculation of irradiation damage and the computer simulation for the process of collision cascade after one PKA is formed.

Control Rod Hydraulic Drive System (CRHDS) is a new type of built-in control rod drive technology, and the Integrated Valve (IV) is the key control component of it. The pulse water flowing into the control rod hydraulic mechanism (CRHDM) is controlled by the IV to drive the hydraulic cylinders to move in a predefined sequence to make the control rod perform step-up, step-down and scram functions. Flow resistance of the IV flow channels is the key design parameter of IV which influences the step motion of the hydraulic cylinder and thus affects the performance of the CRHDS. Experiments on the flow resistance of IV flow channels at different working temperatures were conducted to obtain differential pressures of IV under various temperature and flow rate operating conditions. Based on the experimental conditions and results, three dimensional flow field analysis of the IV flow channels was carried out to get the flow field distribution and hydraulic parameters of the IV flow channels. Flow resistance of the IV flow channels at different working temperatures were obtained using the calculation results and agree well with the experimental results. It verified the correctness of the CFD model. On the basis of the numerical simulation results, the velocity and pressure distribution schemes in the IV flow channels under different working temperature conditions were compared and analyzed. The research results show that the flow resistance of the IV in-rod flow channels remains largely unchanged at different working temperatures, the peak flow velocity appears at the entrance of the valve core section which is also the main flow resistance loss area. The theoretical model was then applied to analyze the influence of the design parameters which include the valve core size, the angle between flow channels, etc., on the total flow resistance of IV at high temperatures. And the analysis results show that, the angle between flow channels has little influence on the flow resistance coefficient. The increase of valve core radius can significantly reduce the total flow resistance of IV flow channels. Numerical simulation on one out-rod flow channel is also carried out, which shows that the flow resistance in out-rod flow channel is much lower than the corresponding in-rod flow channels. The research results can give guidance for the design and optimization of the IV.

When a severe accident of a nuclear reactor occurs, decommissioning work becomes important task. In the decommissioning work of a boiling water reactor after the severe accident, estimation of the sedimentation place of the molten debris is important. However, the technique to estimate exactly the sedimentation place has not been enough developed. Therefore, the detailed and phenomenological numerical simulation code named “JUPITER” is under development for predicting the molten core behavior in JAEA (Japan Atomic Energy Agency). The comparison between experimental and numerical results is necessary to clarify the validity of the numerical analysis code. The study provides the experimental data to examine the numerical simulation code. As a basic study to examine the numerical simulation code, a liquid film flowing in a modeling flow channel was studied by using water. The flow was visualized, and the flow data were obtained by image processing.