In this article, Eulerian two-fluid model coupled with an extended RPI wall boiling model was applied to simulate the departure from nucleate boiling (DNB) in vertical heated tubes by commercial CFD software. A total of 48 different phase interaction models were simulated, of which 39 simulations’ results are compared with experimental data while the rest gives no results. Wall temperature, mean liquid temperature and mean profile of vapor fraction have been used for comparison of these models. It was found that nucleation site density (NSD) model has a greater effect on the prediction of wall temperature, while bubble departure diameter model shows the significant effect on both the prediction of wall temperature and the vapor fraction. The results show that the Kocamustafaogullari and Ishii model for bubble departure diameter show good accuracy in predicting the mean void fraction. In addition, two model combinations can well predict the steam distribution, the average liquid temperature and the wall temperature during high pressure boiling, whose simulation results show good agreement with the experiment results, which indicates that the CFD method can be used in simulating the bubble behaviour for subcooled water flow boiling.

The molten salt reactor (MSR) is one of the six advanced reactor concepts declared by the Generation IV International Forum (GIF), which can be characterized by attractive attributes such as inherent safety, sustainable development, nuclear non-proliferation, natural resource protection, and economic efficiency. The molten salt reactor is the only liquid fuel reactor in the fourth generation of nuclear power reactors. In order to study the influence of critical parameters on the natural circulation capacity of the molten salt reactor after shutdown. Based on the thermal-hydraulic analysis software RELAP-5, a numerical analysis model of the overall molten salt reactor first loop, including the core, heat exchanger, and main pump, was established to investigate the temperature change in the presence of decay heat in the loop piping by calculating the flow rate change in the loop during the conversion from forced to natural circulation. The results demonstrate that the decay heat of the molten salt during the conversion from forced to natural circulation will change the temperature field distribution and reduce the efficiency of the core in discharging the decay heat. Meanwhile, parameters such as the loop height, the channel area of loop pipeline and the transverse pipe length can affect the natural circulation process of liquid fuel molten salt reactor. This study confirms the rationality of using the natural circulation of molten salt fuel salt for decay heat discharge and provides a theoretical basis for the design of the decay heat discharge system.

The study of reactor fuel degradation behaviors under severe accidents have great significance for the safety of nuclear power plants and the formulation of management measures. Annular fuel is one kind of new-type fuel, which is composed of inner/outer cladding, inner/outer coolant channel and middle annual fuel pellet. Annular fuel has become one of the main development objects in many countries because of its economy and safety. At present, the research on the degradation behaviors of annular fuel under severe accident is still insufficient. Based on this reason, an out-of-pile visualization high temperature experimental system is built. The degradation experiment of annular fuel simulation rod is carried out based on the alternative cladding material-Al and the real cladding material-Zr. The temperature variation and visualization process data of the simulation rod are obtained. The degradation behaviors of annular fuel are further analyzed based on the experimental results.

High temperature gas cooled reactor (HTGR), which has the characteristics of inherent safety, modular design and high outlet temperature, is considered as the most promising fourth generation reactor type. The intermediate heat exchanger (IHX) is the key equipment to produce hydrogen from HTGR, and the uniformity of the flow field among tube bundles has significant effects on the heat transfer performance and safety of the IHX. In this study, an experimental system was built to explore the effect of inlet flow distributors on flow uniformity, and the internal mechanism was explored through numerical simulation. The experimental results show that the velocity is various at different positions in the intermediate heat exchanger, and the uniformity of flow field without flow distributor is poor. Selecting the appropriate flow distributor can significantly improve the flow uniformity. The numerical simulation results show that the existence of the flow distributor makes the inlet flow split into two streams of fluid. One directly enters the tube bundle area, while the other enters the tube bundle area after flowing along the distributor for half a circle. The proportion of two streams of fluid affects the flow uniformity, while the flow distributors with different lengths adjust the proportion of two streams of fluid. Finally, an optimal shape of the inlet flow distributor is determined by combining the results of experiment and numerical simulation.

A Coordinated Research Project on “Benchmark Analysis of FFTF Loss of Flow Without Scram Test” was launched by the International Atomic Energy Agency (IAEA) in 2018. A series of passive safety tests were conducted from 1980-1992 at the Fast Flux Test Facility (FFTF), 400 MW(th) liquid sodium cooled nuclear test reactor owned by U.S. Department of Energy (DOE) to demonstrate the potential of FFTF to survive severe accident initiators with no core damage. Amongst these tests was a series of Loss of Flow Without Scram (LOFWOS) tests from power levels up to 50%, also commonly referred to as Unprotected Loss of Flow (ULOF) tests, which were studied in the IAEA CRP. The data were provided by the Argonne National Laboratory (ANL) and Pacific Northwest National Laboratory (PNNL).

Another Research Coordinated Project on “Benchmark of Transition from Forced to Natural Circulation Experiment with Heavy Liquid Metal Loop” was launched by the IAEA in 2022. Three tests were conducted in 2017 to study the thermal-hydraulic behavior of a test fuel assembly cooled by lead-bismuth eutectic alloy during transition from forced to natural convection at the NACIE-UP facility at Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA), Italy. This project is the first IAEA CRP that is dedicated to the thermal hydraulics of lead and lead bismuth eutectic (LBE) technology.

The paper provides a general overview of the two CRPs within the framework of the IAEA activities on thermal hydraulics of fast reactors.

The fuel test rod is the basic element for carrying out the fuel irradiation test, and its temperature performance determines whether the irradiation test can be carried out smoothly. Many factors can affect the temperature performance of fuel test rods, including structure, materials, thermal and cooling conditions. This paper first establishes a one-dimensional radial heat transfer model of fuel test rod according to certain assumptions, and analyzes the main criteria for evaluating its temperature performance according to the basic criteria of irradiation test, including the maximum temperature of fuel, cladding and the outside of protective tube. Therefore, according to the different thermal resistance layer thickness, protective tube thickness and coolant flow rate established three conditions, the impact of these three factors on the maximum temperature of cladding was analyzed. Then, the temperature of fuel test rod is compared under theoretical calculation and numerical calculation, which proves that the one-dimensional model can accurately assess the influence of various factors. Finally, the effects of other factors such as the temperature of coolant, the linear heat rate (LHR) of fuel, the γ heat rate and the porosity of fuel are analyzed. The results show that the influence caused by coolant temperature change is relatively limited, the influence of LHR is very significant, the influence of γ heat rate cannot be ignored, and the influence of porosity of fuel is within acceptable limits.

Industrial processes involving gas-liquid two-phase flow are one of the most common phenomena in the field of nuclear energy. How to determine the flow pattern is a basic problem in two-phase flow analysis, and an effective two-phase flow pattern prediction model is very important for the research of multiphase flow. In many traditional studies, people still cannot get accurate enough prediction results. In this paper, machine learning tools and data-based algorithms will be used to determine the flow pattern, and the algorithm with the highest accuracy will be selected to establish a flow pattern prediction model that meets the requirements. By using the flow pattern prediction model in this paper, the accuracy of flow pattern prediction can reach more than 99%, which is higher than the traditional methods.

In this study, two-phase flow tests have been carried out in a 3×3 rod bundle channel to investigate the feasibility on flow regime classification using artificial neural network (ANN). The rod bundle channel was constructed with nine 11.5 mm-diameter rods in a rectangular array with a pitch of 15.4 mm and enclosed by a 52×52 mm rectangular channel, and the overall height was about 5 m. The test flow conditions covered the superficial gas and liquid velocities of <j_g>= 0.02-10 m/s and <j_f>= 0.02-2 m/s. The transient void fraction of each test was measured and converted into probability density function (PDF) and cumulative probability distribution function (CPDF) and rearranged as the input matrices for Kohonen neural network for training and classification with the training epoch of 1500 times. While assigning 3 groups, the bubbly and cap-bubbly flows can be clearly distinguished; whereas when assigning 5 groups, the bubbly and cap-bubbly flow can be further subdivided with an additional transition region, and the cap-turbulent and churn flows can be roughly separated. This study has preliminarily demonstrated the feasibility on classification of two-phase flow regimes in the 3×3 rod bundle channel using artificial neural network technique.

The critical heat flux (CHF) corresponding to the nucleate boiling crisis (DNB) is critical for the design, operation, and safety of nuclear facilities. The development of an accurate and robust CHF prediction model remains the primary goal of the thermal-hydraulic engineering community.

The goal of this study is to use a stacked ensemble learning model to achieve superior capabilities for CHF prediction. Our approach, in particular, is based on a stacking ensemble learning scheme, in which the predictions produced by nineteen base learning and linear regression are used as meta-learner (super learner) in the top-level method to produce final predictions.

We tested the proposed scheme on an experimental datasets that reported CHF in a variety of ranges and geometries. The results show that a CHF prediction approach based on ensemble learning combines predictions produced by weaker learning methods and then feeds them to a meta-learner to achieve superior results. More importantly, this case study demonstrated that using an improved stacking ensemble model can result in very accurate predictions with low errors, making it a suitable approach for addressing the CHF prediction problem.

For contaminated water management in decommissioning Fukushima Daiichi nuclear power stations (1F), reduction in water injection, intermittent injection water and air cooling are considered. However, since there are uncertainties of fuel debris in the primary containment vessels (PCVs), it is necessary to examine and evaluate in advance optimal cooling methods according to the distribution state of the fuel debris and the progress of the fuel debris retrieval work.

Japan Atomic Energy Agency (JAEA) has developed a numerical method by using JUPITER for estimating the thermal behavior in the air cooling, including the influence of the position, heat generation and the porosity of fuel debris. It is however difficult to perform the large-scale thermalhydraulics analysis with JUPITER by modeling the internal structure of the debris which may consist of a porous medium. In this study, we added a porous medium model to JUPITER to analyze the heat transfer of the porous medium. In this paper, we report the validation of JUPITER with the porous model and discuss the effect of various models of the effective thermal conductivity based on the array of the solid inside the porous medium to the natural convective heat transfer. To obtain validation data of JUPITER for the natural convective heat transfer analysis around the porous medium, we performed the heat transfer and the flow visualization experiments of the natural convection in the experimental system including the porous medium. In the comparison between the experiment and the simulations for various models of the effective thermal conductivity, the numerical result with the geometric mean model was the closest of the models to the experimental results.

DEMO is a nuclear fusion power station that has, among others, the objective of demonstrating the possibility to produce several hundred MWs of electrical power from fusion reaction by the middle of this century, increasing the production of carbonneutral electricity. In particular, the Balance of Plant of DEMO has the key role to demonstrate the feasibility of delivering the power produced within the Tokamak reactor to the grid.

The design approach for Water Cooled Lithium Lead (WCLL) Breeding Blanket (BB) Primary Heat Transfer Systems (PHTSs) considers the Nuclear Power Plant experience, adopting components commonly used in nuclear industry. However, their performances could be negatively affected by the discontinuous operation (pulse-dwell-pulse) of the DEMO machine, as well as by low-load operation in dwell time. This makes mandatory a full assessment of the functional feasibility of such components throughout an accurate design and validation.

ENEA Experimental Engineering Division at Brasimone R.C. supports the design and qualification of DEMO by realizing STEAM, a water operated facility conceived to investigate the water technologies applied to the DEMO BB and Balance of Plant systems and components. STEAM is mainly composed by a water primary system reproducing the DEMO WCLL BB PHTS thermodynamic conditions (15.5 MPa, 328-295°C) and a secondary two-phase (liquid/steam) water loop reproducing the DEMO power conversion system conditions (6.4 MPa, 238-300°C). The experimental validation aims at reproducing steady-state and transient operation in DEMO-relevant conditions, as well as to perform dedicated tests on a once through steam generator mock-up, representative of the one envisaged for DEMO, aiming at testing its performances during the power phases of the machine.

Thermal-hydraulic analyses of STEAM have been performed by RELAP5/Mod3.3 system code. Steady-state qualification results are presented in this paper, along with the characterization of the facility dynamic behavior, realized in order to optimize the system layout and to assess the performances of the main components during the prescribed operating conditions.

As an aftermath of the Fukushima nuclear accident and its implications for public acceptance of nuclear energy, the nuclear industry has begun to seriously rethink and improve the safety strategies of nuclear power plants. The beyond-design-basis external events (BDBEEs), such as the earthquake and tsunami in Fukushima nuclear accident, may lead to an extreme accident which may lead to the failure of the traditional strategies. In the recent decade, the concept of FLEX strategies has been accepted in the nuclear engineering community and attracted a large number of researchers to investigate it for nuclear accident mitigation. The development of FLEX strategies is divided into two steps. One is an uncertainty study of system response, especially the safety injection system. The other is training and testing of machine learning (ML)-based FLEX strategies and to verify the effectiveness of ML algorithms in the development of FLEX strategies for nuclear power plants. This paper focuses on first step uncertainty study of the nuclear accidents by using statistical algorithm to generate a database of incidents and system responses. The conclusion of this paper is useful for analyzing the characteristic of typical nuclear accidents and benefits to the FLEX strategy development, which will be focused on in the near future.

To study the depressurization process and parameters change near the breaking point during the rapid blowdown in the pipeline system, an experimental device is designed and built to do the transient experiment and measurement based on the loss of coolant accident in the water-cooled reactor. The apparatus of the experiment include a heating container tank to provide water with certain initiated pressure and temperature, an experimental section, an experimental control system, various kinds of transducers and a data collection system, as well as some auxiliary equipment. At the same time, the layout and size of the test section have taken the results of numerical calculation into account. In order to better simulate the transient process, a quick-opening solenoid valve is used to accomplish the pipe rupture. Besides that, a series of highprecision piezoelectric sensors with a maximum frequency response of 10kHz is used to meet the requirement of transient pressure measurement. In addition, a movable measuring rod is designed to acquire the spatial distribution and time response curve of back pressure when the discharged water is released into the outer area. The experimental scheme, including the specific procedures, with some calculations are presented in this study. The solution strategy for crucial points and challenging issues is also discussed. This separated experiment provides a solution for parameter measurement both in the pipeline system and the outside space during the strong transient flow process, and also provides a foundation for subsequent experiments under high parameter working conditions in the future.

Fukushima Daiichi Nuclear Power Plant accident resulted in a core meltdown, releasing a large amount of radioactive materials into the environment. This accident has reconfirmed the necessity and importance of the further in-depth development of core damage assessment technology (Level 2 PSA). In order to advance the core damage assessment technology, it is necessary to establish a source term migration assessment method through leak paths. We have started basic studies on the fission product (FP) migration behavior through leak paths, aiming to develop an evaluation method for aerosol transport based on transport mechanisms. In this paper, we will report basic decontamination factor (DF) data in narrow circular channels that simulate leak paths through containment vessel (CV) and reactor building. An experimental line is set up, and the experiments are performed under conditions simulate the environmental and flow conditions in the CV penetrations and failure locations at severe accident (SA). The tests are conducted to find the effects of flow path size and particle size on the DFs. DFs are derived from the experimental measurement of the aerosol concentrations at the inlet and outlet of the test sections. The obtained experimental DFs were compared with the existing models developed for aerosol deposition, considering the particle size distributions.

A series of numerical simulations have been carried out to investigate the ECC water injection in Hualong NO.1 with a Direct Vessel Injection (DVI) system during a Large Break Loss-of-Coolant Accident (LBLOCA) using RELAP5 code. Some parametric studies are performed to investigate the effect of the DVI nozzle elevation and the location of the broken cold leg. The results showed that the filling rate in the lower plenum is the highest at the elevation of 300 mm below the centerline of the cold leg. Comparing the result at the elevation of 200 mm below and above the centerline of the cold leg, it can be derived that the ECC water film is wider at a higher elevation resulting in a larger filling rate. And by considering the filling rate with different location of the broken cold leg at a given DVI nozzle elevation, the major findings show that a larger distance between the break and the DVI nozzle leads to a lower bypass fraction and a larger filling rate in the lower plenum. Also, the comparison of the ECC water injection mode and condensation effect are performed.

With the development large-scale space equipment becoming more and more obvious, its power consumption and waste heat generation increase, it is urgent to develop a higher efficient heat transfer system. Liquid droplet radiator (LDR) is a promising new space waste heat export system. LDR has many advantages compared with the existing technology, including small volume of droplet, large specific surface area, internal stability, and more compact storage in the space chamber. It is very important to study and improve the droplet generator that generates droplets, which generate micro-droplet. The rapid development of numerical simulation technology provides a more intuitive, effective and accurate help for predicting and analyzing multiphase flow in microfluidic channels, which is conducive to better research on the generation of droplets.

In this study, a 2D model of droplet generator is designed by using VOF model. The droplet uniformity is studied by changing the inlet velocity and two phase surface tension. Breakup behavior and the diameter of droplet show that surface tension takes part in a important role. At the same time, the surface tension also has an impact on the droplet diameter, and when the surface tension reaches a certain threshold, the droplet will produce serious mutual merge. Increasing the jet velocity helps to slow this behavior.

Nucleate boiling is an important part of boiling heat transfer, which is widely used in nuclear energy, thermal power, electronic equipment cooling and other fields. Scholars have done a lot of boiling heat transfer experiments and fitted the corresponding heat transfer correlations, but no correlations with high accuracy and wide applicability have been obtained. In this paper, the experimental data of nucleate boiling with different coolants are collected, and the existing data of nucleate boiling in the tube bundle channel are compiled into a database. The mechanism of nucleate boiling is discussed through data analysis. It is found that the heat flux q has a linear relationship with the temperature difference Tw-Ts during the nucleate boiling process. This phenomenon is demonstrated and analyzed, and by introducing the RPI model, the intercept and slope of the linear relationship are discussed and a new method for constructing the relationship form is proposed to predict the heat transfer coefficient under nucleate boiling. The effects of mass flow, inlet subcooling and pressure on the linear correlation were analyzed.

The mixing vane on the spacer grid can strengthen the turbulence anisotropy downstream of the rod bundle channel, break the wall boundary layer, and improve the critical heat flux of the fuel rod. Studying the flow field downstream of spacer grid can provide data support for nuclear reactor design and safety analysis, and can also provide a benchmark for numerical simulation. Both AFA 3G-type and AP1000-type spacer grids are mature commercial spacer grids, the comparison of the mixing effects of these spacer grids can provide data for the future development of spacer grid.

In the current study, Particle Image Velocimetry (PIV) technology is used to visualize and measure the flow fields downstream of the spacer grid. Steady state flow fields downstream of an AFA 3G-type spacer grid and an AP1000-type spacer grid in a 5×5 rod bundle channel were measured can compared for Reynolds numbers varying from 5000 to 30000. The results show that the AFA 3G-type spacer grid is more focused on the mixing of the channel center. The mixing coefficient of AP1000-type spacer grid hardly changes with Re. The unsteady force brought by the AP1000-type spacer grid is smaller than that of the AFA 3G-type. Compared with the AP1000-type spacer grid, the downstream vortex scale of the AFA 3G-type is larger.

The design and development of the third-generation nuclear power plants（NPP） puts forward higher requirements for safety and economy, and the third-generation NPP has a high level of safety. As the electricity market becomes more and more competitive, it is necessary to optimize the design, reduce the cost, and improve the economy of the unit under the premise of balancing the safety and economy of the power plant. In the current design of the third generation NPPs, The auxiliary feedwater system was a special safety facility, and adopts the configuration plan of 2×50% motor-driven-pumps + 2×50% turbine-driven-pumps. This scheme increases the diversity of auxiliary feed pumps, there are also the following problems: There are few suppliers of turbine-driven-pumps for nuclear power in the world. The cost of turbine-driven-pumps is much higher than that of motor-driven-pumps, and the operation and maintenance costs in the later stage are also higher. It is necessary to configure a complete steam supply system for the turbine-driven-pumps to ensure its operation, and the operation stability is poor.

Therefore, it is necessary to cancel the turbine-driven-pumps to improve the economy of the unit and simplify the configuration of the auxiliary feedwater system, eliminate the transient impact of the system caused by the operation and maintenance of the turbine-driven-pumps, and improve the flexibility of the unit.

This paper is based on the design scheme of canceling the turbine-driven-pumps and corresponding steam supply equipment and discharge pipeline, adding an automatic isolation signal to automatically isolate the broken steam generator injection loop in the event of an accident. Flowmaster software is used to model the layout of the auxiliary feedwater system and carry out the capacity calculation and analysis of the auxiliary water supply system. Auxiliary feedwater flow under different working conditions of typical large and small fractures has calculated, and then analyze whether the system can meet the requirements of accident analysis. From the current calculation and analysis results, it is theoretically feasible to cancel the auxiliary feedwater turbine-driven-pumps and set up isolation to meet the requirements of accident analysis.

For the two-phase flow study, the description of bubble swarm is often expanded from a single bubble. Its related mechanism model is extended and developed from the mechanism model of single bubble. For instance, researchers have developed the drag model of bubble swarm from the drag model of single bubble, and finally established the interphase resistance model of two-phase flow. In a typical two-phase flow, bubbles have different trajectories due to the difference of their dynamic characteristics, and they will migrate to the wall or center of the tube. Then, the bubbles gather locally and forming a wall-peak or core-peak distribution of local flow parameters such as void fraction or interfacial area concentration. While local aggregation of bubbles also means the increased bubble interaction at this position. Thus, single bubble’s dynamic characteristics affect the phase distribution characteristics of two-phase flow, and further influence the heat transfer characteristics under possible heating conditions. In this paper, the rising motion experiment is carried out for single bubble’s dynamic characteristics in hydrostatic water and linear shear flow. A curved screen was used to generate a stable low-turbulence linear shear flow field. PIV was used to measure the flow field. Motion information of rising bubble is obtained through image processing. While force coefficient was obtained by quasi-steady-state analysis. Single bubble’s dynamic characteristics in different flow field is analyzed, including bubble shape and bubble force.

Steam discharging through spargers and blowdown pipes into the Pressure Suppression Pool (PSP) is employed in Boiling Water Reactor (BWR) to prevent overpressure of the reactor vessel and containment. The capability of suppression can be reduced during the operation when the thermal stratification is developed. Direct modeling of steam injection into a water pool with long-term transient is computationally expensive due to the large-scale difference in space and time. To enable such prediction, Effective Heat source and Effective Momentum source (EHS/EMS) models are proposed.

In previous work, we demonstrated the implantation of EHS/EMS models in the Computational Fluid Dynamics (CFD) tool and its application to plant simulation. In this work, we use the developed model to further investigate the thermal stratification and mixing in the PSP of a Nordic BWR. The event to be analyzed is initiated by spurious activation of one valve in the safety injection system. The focus of the simulations is to investigate the possibility of stratification development and understand the effects of the activation of different systems on pool behavior. Pool transient is simulated by CFD code (ANSYS Fluent) with EHS/EMS models and the injection conditions of the steam are derived from the simulation results performed by the system-level codes (GOTHIC).

Compared with M310 and Hualong units, the boric acid makeup subsystem of AP1000 units is more concise, and some equipment such as volume control box, boric acid pump and demineralized water pump are removed. But it also improves the requirements for control logic and valve performance. In fact, during the actual operation and maintenance of the AP1000 unit, boron deviation alarms occur frequently, affecting the safe operation of the unit. In this study, the operation characteristics of the AP1000 boric acid makeup subsystem were simulated and analyzed. A unit operation database containing 37288 groups of data is first established and used to mine the key factors affecting the makeup subsystem performance. The simulation model of AP1000 boric acid makeup subsystem was immediately established and used to analyze the influence mechanism of various factors. The results show that: 1)the key to improve the operation performance of the system is to maintain the consistency between the inlet demineralized water pressure and the inlet boric acid pressure of the three-way valve. 2) Insufficient pressure stabilizing capacity of pressure reducing valve V140 may be the root cause of boric acid concentration deviation.

VVER are one of the most common reactor types in the world. Moreover, a significant percentage of the Gen-III/Gen III+ reactors that are currently being built or have recently come into operation are VVER. Therefore, there is a growing interest in studying their behavior under both anticipated and accidental transients.

A joint effort between the Universidad Politécnica de Madrid (UPM) and the Karlsruhe Institute of Technology (KIT) has been made, within the ISASMORE project, in order to develop an integral plant model of a VVER-1000/V-320 reactor for TRACEp5 code, [1].

The purpose of this work is to analyze the impact, in a Small Break Loss Of Coolant Accident (SBLOCA) with failure of the High Pressure Injection System (HPIS), of depressurizing the secondary side by the BRU-A relief valves, with the objective of cooling the Reactor Coolant System (RCS) so that the Hydroaccumulators (HA) and the Low Pressure Injection System medium (LPIS) can inject borated water into operation.

Small Innovative Helium-Xenon Cooled Mobile Nuclear Power Systems (SIMONS), which may function as a stand-alone power source or "electric island," is a nuclear-powered electric generator concept designed to meet the requirements of portability. Due to its low system-specific mass and excellent energy conversion efficiency, the SIMONS reactor employs a closed He-Xe Brayton cycle (CBC). A printed circuit heat exchanger (PCHE) would be used to recover the waste heat exhausted from the turbine in order to increase the efficiency of the Brayton cycle. This work investigated the effect of the helium-xenon mole fraction on the convective heat transfer coefficient of the PCHE. The thermal-hydraulic performance of a zigzag PCHE unit model is investigated using a mixture of helium and xenon. Several geometrical parameters are investigated, including relative pitch ratio and zigzag angle. The principle of least squares is used to conduct a nonlinear fitting of the Nusselt number criteria correlation of PCHE based on the results of the numerical simulation study.

The Full Scope Simulator (FSS) of CANDU6 is very important for operator training, but due to the lack of the severe accident models, the severe accident (SA) training cannot be carried out. Therefore, it is urgent to expand the severe accident model on the CANDU6-FSS in QinShan III NPP. By coupling MAAP5-CANDU model to develop CANDU6 SA simulator module, the FSS of QinShan III NPP can provide richer and more complete accident scenarios for operator training and accident drills.

This paper introduces the severe models (such as reactor core, primary system and containment) and the key technology of coupling MAAP5-CANDU software with FSS simulation platform. Based on the CANDU6 SA simulator and virtual reality technology, a severe accident 3D displays system (3D) is developed. By using the data of CANDU6 SA Simulator, the system can present the severe accident simulation results with 3D dynamic visualization to the audience. Furthermore, the digital severe accident management guidance (SAMG) system is developed, which successfully transforms manual processes into electronic and automated processes, such as paper document review, data search, logical judgment, and auxiliary calculations related to SAMG operations. In the meanwhile, it can reduce the work intensity of relevant personnel of QinShan III NPP and improve the efficiency of SAMG execution.

In summary, the development of CANDU6 SA simulator is not only an effective means for nuclear power plant severe accident scenario drills, but also can be used as an auxiliary tool for SAMG verification and drill scenarios optimization, which provide technical support for the safe and stable operation of nuclear power plant in the future.

Steam-water separation technology has a wide range of applications in energy, environment, chemical industry etc. In the nuclear energy field, the steam separator plays an important role in supplying steam to turbines and the study of the mechanism of steam-water separation devices is very significant for the safe and efficient use of nuclear energy. Experimental studies on steam-water separation are usually relatively macroscopic, making it difficult to obtain microscopic information about droplet motion, collisions, etc. In terms of numerical simulations, there are mainly two types of methods including the Euler-Euler methods and Euler-Lagrange methods, the Euler-Euler method cannot obtain detailed information about the droplet motion, while the computational effort of the traditional Euler-Lagrange method increases with the time step, and the droplet motion information of the full field is cannot be directly obtained. Thus, the Euler grid approximation model is established and the analysis of single droplets is performed based on the Euler grid approximation algorithm, but only the motion of droplets in a single direction is simulated, and the calculated results of droplets moving in different directions cannot be obtained. Therefore, based on the explanation of the physical mechanism of droplet motion in the water vapor separation device as well as the single droplet motion model of Euler grid approximation method, this paper gives the physical mechanism explanation and mathematical expressions of the Euler grid approximation model for different calculation directions of droplets, analyzes the relationship between droplet accumulation and the droplet velocity in different motion directions, studies and gives the velocity and density fields of multi-droplet motion. The accuracy of the droplet field calculation is verified by comparison with the Euler-Lagrange method and can be improved with the grid encryption, and the applicability of the Euler grid approximation method is extended to multi-droplet fields.

Direct contact condensation (DCC) is widely appeared in the nuclear power plants and will lead to serious temperature and pressure fluctuations. For ocean nuclear power plants, the DCC is inevitably affected by the sea conditions, such as rolling, heaving motion. In this study, the integration of volume of fluid model (VOF) and condensation model is adopted to simulate the DCC characteristics under heave conditions. The interfacial heat and mass transfer are calculated by user-defined functions (UDF). The additional inertial force method is used to describe the heave motion. First, the DCC numerical model under heave motion is validated by the experimental data, which coincides well with the experiment results. Then, the characteristics of DCC under different heave parameters are numerically studied. The results show that the average condensation rate increases with an increase in heave amplitude, decreases with an increase in heave period. This is mainly because the heave motion will cause the liquid phase to impact the upper pipe wall, and increasing the contact area of vapor and subcooled water. The pressure oscillations induced by DCC are more complicated under heaving conditions. With heave period continuing to change, the formation mechanisms of the pressure peak are different. When the heave period is relatively small, the pressure peak is mainly induced by the acceleration of water phase under high frequency heaving motion. With an increase in the heave period, the heaving motion accelerated the formation of isolated steam bubble, and caused the condensation induced water hammer (CIWH) events, which in turn induced the pressure peak. In addition, the pressure peak is proportional to the heave amplitude. In summary, the direct contact condensation characteristics are more sensitive to the heave period under heaving conditions.

These qualitative conclusions may serve as a reference for direct contact condensation simulations under heaving conditions.

Long-term material compatibility in heavy liquid metal (HLM) remains a challenge for the successful deployment of HLM-based technologies. Flow-accelerated corrosion and erosion (FACE) phenomena can lead to continual material deterioration, which needs to be considered throughout the reactor design stage. Nonetheless, known experimental data are inadequate to cover all the prototypical flow regimes during LFR’s operation. Modelling of the FAC/FACE phenomena remains mostly in lumped parameter/subchannel scales, where the FAC model is coupled to the bulk flow of the pipe or subchannel. These methodologies might produce a sufficient prediction for the core internals; however, this might not be suitable for the pump impeller due to comparatively greater relative velocity and the occurrence of transient flow patterns near the rotating impeller. To establish an understanding of the connection between turbulence and FACE, the liquid lead-based Separate Effect Flow Accelerated Corrosion and Erosion (SEFACE) facility is currently under design at KTH in the framework of the Sustainable Nuclear Energy Research In Sweden (SUNRISE) project. SEFACE attempts to investigate FACE phenomena in the liquid lead and produce quantifiable validation data for model development.

The paper divides itself into two parts. Part I refers to the study of operational conditions in SEFACE via Reynolds Averaged Navier Stokes (RANS) simulation, while Part II deals with the recent attempt on modelling time-dependent flow shear on rotating disks based on large eddy simulation (LES). The paper begins with a brief review of prior studies on flow-accelerated corrosion. Following that, the SEFACE facility's design concept is laid out considering several physical and operational constraints. A periodic wedge of the SEFACE test chamber is chosen to examine the facility's time-averaged behaviour. The k-ω shear stress transport (SST) model was employed for the simulations. The torque prediction on the rotating disk system is verified with the empirical frictional factor prediction. The latest hydrodynamic design enables SEFACE to be spun at 1200 revolutions per minute (corresponding to a maximum velocity of 21 m/s) without causing free surface deformation or excessive pressure. SEFACE permits the collecting of experimental data under the effect of various relative velocities in a single experiment round. The second part of the paper focuses on a recent attempt to determine the wall shear stress distribution on a rotating disk using wall-modelled large eddy simulation (WMLES S-Omega). The obtained amplitude and frequency of wall shear stress fluctuations will aid model development in future.

Thermal stratification in large pools or enclosures is a very important phenomenon that is critical to nuclear reactor safety, including the cold- and hot-pool mixing in pool-type reactors, reactor cavity cooling system behavior in high-temperature gas-cooled reactors, passive containment cooling in advanced light-water reactors, and thermal stratification in boiling-water reactor suppression pools. It is very important to accurately predict pool temperature and density distributions for both design optimization and safety analyses of these reactor systems. Currently, effective thermal stratification modeling is a multi-scale method, that couples 1-D system-level code and 3-D CFD code, and can provide detailed thermal stratification information while providing system-level information in other domains. However, coupling 2 codes operating at different spatial and temporal scales remains a challenging issue, and high-resolution CFD simulation is computationally intensive. This study developed a data-driven coarse-mesh turbulence model for thermal stratification to improve the efficiency and accuracy of reactor safety analysis. The 3-D fluid conservation equations are modeled with a data-driven turbulence model based on local flow features on a much coarser scale so that the corresponding mesh setup in the CFD code can be consistent with the 1-D system-level code. A machine learning framework has been introduced for the acceleration of RANS by training an deep neural network with a large amount of fine-mesh CFD-generated data to predict steady-state turbulent eddy viscosity. We choose OpenFOAM as the simulation framework, TensorFlow as machine learning framework, and use TensorFlow C API to realize the coupling of OpenFOAM/TensorFlow and the invocation of the data-driven turbulence model. The coupling is introduced by way of a surrogate model prediction task for the turbulence eddy viscosity for the SUPERCAVNA experimental facility problem. This demonstrates the feasibility of the developed data-driven turbulence model.

In severe accidents of nuclear power plants, large amounts of fission products existing as radioactive aerosols are generated. Pool scrubbing plays an important role in the removal of radioactive fission products. Since previous studies have shown that the washing out process within the swarm flow makes a very important contribution to aerosol removal efficiency. It is of great significance to investigate the hydrodynamic behavior in swarm flow. The main focus of the current work is on the flow velocity and the residence time of the gas phase as well as the turbulence of the flow. The Euler-Euler two fluid method and the Lagrangian Particle Tracking (LPT) method are applied, the former is used to get the flow field information, and the LPT method is used to track the bubble movement to obtain the bubble residence time. The feasibility of the baseline model under pool scrubbing conditions was examined. Finally, the preliminary results of bubble residence time distribution (RTD) are obtained. The results obtained with the baseline model show good agreements with the experimental data. Liquid recirculation generated near the water surface in large volume gas flow rate slightly increases the bubble residence time. Residence time shows a flat trend with radial injection position.

For the wire-wrapped fuel assemblies in liquid metal reactors, high-speed scouring of the liquid metal coolant, and the lateral flow caused by the presence of wires lead to the vibration of fuel rods. And the vibration may induce micro-motion wear and damage the integrity of the cladding. In this paper, the flow-induced vibration problem of a 7-pin bundle assembly with filament winding positioned under axial flow of lead-bismuth alloy was analyzed by FSI numerical simulation method. Firstly, the vibration response of the wire-wrapped rod bundle with different pitch was calculated by changing the pitches of the winding wire. Then the vibration response of the wire-wrapped rod bundle with different pitch to diameter ratio was calculated by changing the pitch value. Finally, the vibration response of the wire-wrapped rod bundle with different inlet velocities was calculated by changing the inlet velocity. The analysis results show that the change of pitch of wire will affect the amplitude, vibration balance position and vibration frequency of the wire-wrapped fuel bundle; the change of pitch to diameter ratio will affect the vibration balance position, but not the amplitude and vibration frequency; the change of inlet velocity will affect the vibration balance position and vibration frequency, but not the amplitude.

Transient behavior and heat transfer characteristics of the multi-layer corium pool is a significant problem that arises in the In-Vessel Retention analysis of the nuclear reactor severe accident. According to the results of the OECD MASCA project, a heavy metallic layer with internal heat source and high thermal conductivity may form on the bottom of the lower head of the reactor pressure vessel (RPV). Based on the existing empirical relationships, the CHF (critical heat flux) at the bottom of the outer wall of the RPV is smaller. Therefore, the heavy metallic layer may cause failure of the RPV. Some numerical simulations based on the newly disclosed heavy metallic layer heat transfer experiment were conducted to investigate the heat transfer characteristics and effects of changes in internal heat source.

Water, molten salt or other low melting point solutions are usually used as simulation materials in the heat transfer characteristics experiments because of the high melting point and radioactivity of the real heavy metallic layer materials. However, the physical properties, especially the Prandtl number, of these materials are quite different from real materials, which may affect heat transfer characteristics, such as Nu distribution. Therefore, the effects of the Pr on the heat transfer characteristics of the heavy metallic layer were also investigated in this study.

A phase-field model coupled with flow and heat transport is developed to study the oxidation of molten Zr alloys. This model uses the phase-field method to capture the convection and diffusion of oxygen in melt and the oxidation reaction between Zr and oxygen to form ZrO₂. The phase-field model is verified by comparing it with published one-dimensional oxidation experimental results. The model, which is fully coupled with flow and heat transport, can describe the complex interaction in the process of melt oxidation, including thermal properties variation and oxide crust formation. The results of 2D molten Zr oxidation in oxygen atmosphere simulations show that the melt flow trend strengthens the oxidation kinetics.

In the severe accident of PWR nuclear power plants, one of the mitigation measures for core melting accidents is Ex-Vessel Retention (EVR) that the core catcher will hold the entire mass of corium coming from the reactor and the corium can be cooled during the spreading. The study of ex-vessel corium spreading is of great significance for the application of EVR. The present paper aims to investigate on ex-vessel corium spreading process with an enhanced implicit viscosity Incompressible Smoothed Particle Hydrodynamics (ISPH) method that is especially suitable for simulating the free-surface flow with solid-liquid phase change problem. The enthalpy-based heat transfer model is added in the corrector step. Implicit viscosity algorithm is used to eliminate the time step limitation of highly viscosity in the solidification interval. The numerical simulation of the KATS15 simulant materials spreading test was carried out to check the validity of the developed model. The results show that the developed method can accurately predict the 3D corium spreading process, and is of great engineering application value and scientific significance for determining the success criteria of the corium spreading and optimizing the design of the core catcher.

This work uses the Modelica language to simulate a tandem accelerator with only one single particle based on parameters from the 2×1.7 MV tandem accelerator (under construction) at Harbin Engineering University, which is expected to provide heavy negative ions like Cu-, Fe-, Ni-, etc. for research in the nuclear material field. The work focuses on the beam optics system of the accelerator including accelerating tubes, magnets and lens components and establishes simplified models of these components in the open-source platform OpenModelica. The system is capable to simulate the trajectory of a single particle at low speeds moving in the tandem accelerator by assembling all the components into a complete system in a visualized layout. By manipulating parameters, different kinds of particles can be simulated. It proves that the calculation results of the simulator are quantitively comparable with the results from assessed code used for the accelerator evaluation and can be used to simulate the accelerator for different purposes.

The aqueous solution, as a form of fissile material, is particularly common in reprocessing, where the possibility of criticality exists. The simplified thermal-hydraulic model coupled with point kinetics typically results in the advantages that are relatively easier to solve and lower computational cost compared to more fundamentally based modeling. In consequence, this makes simplified thermal-hydraulic models coupled with point kinetics widely used to track critical accident transients which may have significant durations. This is usually considered lumped model or diffused model (also known as the heat conduction model) with a modified thermal diffusivity coefficient. However, the difference in temperature spatial distribution between the actual natural convection due to buoyancy and simplified thermal-hydraulic models may lead to different calculation results of reactivity feedback. Therefore, this paper simulates the R100 experiment on the TRACY facility to study the spatial distribution characteristics of solutions temperature to provide a reference for improving the accuracy of the thermal-hydraulic models as much as possible, based on the coupling of point kinetics and computational fluid dynamics (CFD). The results indicate that the thermal-hydraulic model should be considered as a natural convection model rather than a heat conduction model in the simulation of criticality excursion transient in solution, in the case of taking into account the spatial effect of temperature feedback.

A novel passive flow limiter has been designed for the integral nuclear heating reactor of NHR200-II to mitigate the consequences of LOCA (loss of coolant accident). To hold the spherical disc of the flow limiter under various external disturbances, restricting rods with labyrinth structures were designed to support the spherical disc. The circular-grooved square cavity labyrinth structure was chosen for the passive flow limiter, and the pressure drop characteristics of the labyrinth structure were studied numerically. The influences of three parameters, i.e., the clearance between the piston and the cylindrical side wall (δ), the number of cavities (N), and the distance between the piston and the front wall (L), were analyzed by numerical simulation of labyrinth structures with nine different configurations and two working conditions (water at 20℃,1MPa and water at 280℃,8MPa). The results showed that the pressure drops of the labyrinth structure increased rapidly when the clearance δ narrowed. The impact of cavity number N on the pressure drop was less than the clearance δ. The pressure drop of the labyrinth structure reduced slightly as the distance L between the piston and the front wall increased.

In the simulation of boiling, it is important to select the appropriate interface tracking model and phase change model. In this study, the heat and mass transfer at the phase interface during the phase change was simulated based on the VOF model and the interfacial heat flux model. The both bubble growth process in the two-dimensional uniformly superheated liquid and attachd to the superheated wall was investigated in this study. It can be concluded that the interfacial heat flux model can not predict the 2-D bubble growth well and some no physical phenomena will occur in the simulation, though it can accurately predict the position of the phase interface in the 1D Stefan problem. The simulation results can show a good agreement with the theoretical solution by modifying the phase change conditions and mass transfer intensity. In this paper, it is proposed that the refinement of the grid or the increase of the mass transfer intensity can make the simulation more in line with the real situation when the interfacial heat flux model is applied. However. it is suggested to make a modification of phase change intensity in the UDF from the perspective of mass source term discretization since the simulation convergence and computing resources should be comprehensively considered.

The main pump blades of lead cooled fast reactor have complex mechanical characteristics, and there are complex fluid-structure coupling problems in the structural analysis of the impeller. In this paper, CFX software is mainly used for numerical simulation to carry out one-way fluid-structure coupling research on the moving blade of the main pump. Generally, three kinds of materials as Ti3SiC2, T91 steel and 304 stainless steel are considered for the moving blade of main pump and the deformation distribution and equivalent stress distribution of the three models under the three moving blade materials are analyzed respectively. Finally, the modal analysis of the moving blade of the main pump is carried out. The results show that the resistance to deformation of moving blade made in Ti3SiC2 is better than those of T91 steel and 304 stainless steel; The equivalent stress on the pressure surface of Ti3SiC2 is slightly higher than those of T91 steel and 304 stainless steel, and the maximum equivalent stress on the pressure surface of moving blade occurs on the inner edge of the blade. This research can provide reference for the design of moving blades, and help to improve the service life of the main pump.

Loss Of Coolant Accident causes the propagation of a transient rarefaction wave within the primary circuit that generates a transient pressure load on the baffle surrounding the reactor core. This is the result of nonidentical travel times of the rarefaction wave between the two sides of the baffle: the reactor core on one side and the by-pass between the baffle and core barrel on the other. The two zones have different geometrical characteristics, in particular the perforated reinforcement plates in the by-pass significantly influence wave propagation. Representing these obstacles in numerical simulations of the primary circuit requires the use of simplified models. A study on the accuracy of these models is hereby proposed, through a detailed comparison between numerical simulations realized with the EUROPLEXUS software and experimental results obtained on the MADMAX facility.

The global fossil energy crisis has created an urgent need for diversity and reliability of energy systems, and energy storage provides effective assistance in renewable energy utilization and power regulation. With its high latent heat, appropriate temperature range and small volume change, solid-liquid phase change energy storage has occupied an important position in the field of thermal storage such as urban waste heat recovery and building heating and cooling. In this paper, the melting process is studied by numerical simulation for rectangular and square triangular arrangement of pipes. The results show that the melting process is divided into three periods, with the strongest interactions between pipes in the middle of melting. The phase interface of the rectangular arrangement is preferentially connected in the horizontal direction, while the phase interface of the triangular arrangement is preferentially connected in the vertical direction, so the phase transition power of the rectangular arrangement is greater in the middle melting period. The triangular arrangement is more uniform and has better thermal conductivity, so the phase change power of the triangular arrangement increases at the end of melting.

The low temperature pool heating reactor is a small reactor operating at low temperature and low pressure, and the maximum temperature of the coolant does not exceed the saturation temperature under the corresponding pressure. Because it is mainly used in people 's life heating, the system design has high inherent safety. The natural circulation valve is the key equipment of the passive residual heat export system of the ' Yanlong ' low temperature pool heating reactor of the China Nuclear Group. The performance of the natural circulation valve directly affects the normal export of the residual heat of the reactor under accident conditions. Therefore, the reliability of the natural circulation valve design and the stability of the work are related to the safety and stability of the reactor. The natural circulation valve is passively opened when the main pump fails, in order to explore the flow field changes of the reactor core and the pool during the opening process ; effect of different natural circulation valve diameters on coolant outlet in core. In this paper, the overlapping grid method is used to simulate and analyze the opening behavior of the natural circulation valve, and the velocity field of the core and the pool during the opening process of the natural circulation valve, the flow change of the natural circulation valve flow channel and the inlet and outlet of the primary circuit are obtained. The velocity field and the mass flow rate at the inlet and outlet of the core are compared, and the influence of the diameter of the natural circulation valve on the flow field of the reactor and the pool is analyzed. From the perspective of qualitative and quantitative analysis, the reliability of the natural circulation valve in the low temperature pool heating reactor is verified, and the safety of the natural circulation valve device is verified, which provides a reference for the optimal design of the natural circulation valve.

Removal of core decay heat by utilizing natural circulation is expected as a significant measure to enhance the safety of sodium-cooled fast reactors (SFRs). During natural circulation, flow and temperature fields in the fuel assembly (FA) are significantly influenced by heat removal due to the inter-wrapper flow (IWF), which appears in the gap between FAs. Therefore, accurate evaluation of the temperature distribution in the FA at the low Re regime in natural circulation operation is demanded. A detailed thermal-hydraulics analysis code named SPIRAL has been developed in Japan Atomic Energy Agency (JAEA) to clarify thermal-hydraulic phenomena in the FA at various operation conditions. In this study, as a part of the validation study, SPIRAL was applied to analyze a large-scale fuel assembly experiment of a 91-pin bundle for two cases at the mixed and the natural convection conditions respectively in low Re regime with heat transfer from outside of the FA formed by an external flow. The hybrid k-ε/k_θ-ε_θ turbulence model, which was established for SPIRAL to reproduce the transition characteristics between laminar and turbulent conditions, was applied. The applicability of the SPIRAL to the thermal-hydraulics evaluation of FA at mixed and natural convection conditions with heat transfer from outside of the FA was confirmed by the comparisons of temperatures predicted by SPIRAL with those measured in the experiment.

During a severe accident, containment integrity is threatened by a possible hydrogen explosion. Hydrogen flame can be accelerated or quenched depending on the mixture, turbulence, and geometry. At present, turbulence and steam effects on flame acceleration, deceleration, and quenching are not well reproduced by the combustion models usually implemented in the tools relevant to nuclear safety.

flameFoam is our custom open-source turbulent premixed combustion solver for OpenFOAM framework. It was developed to perform fast and adequately accurate technical simulations using RANS and TFC modeling approaches. However, in such a framework, there is no accepted flame quenching model. The most common model is based on the flame stretch effect and estimates the probability that the stretching will not quench the flame. This model is parametrized by the critical rate of strain.

This model has been recently implemented in flameFoam. This paper presents the initial validation of the implemented model based on ENACCEF facility experiments available from the SARNET2 project. Results obtained with and without quenching modeling are presented and compared. The sensitivity of the obtained results to the critical rate of strain is examined in the simulations with flame quenching. Initial validation shows that the implemented model shows high sensitivity to the critical rate of strain parameter, however, this parameter is not well defined and can be used to fit the simulation results. Implemented model tends to quench the flame in the high turbulence areas, mainly the acceleration tube of the ENACCEF facility, but underestimates the quenching intensity at the exit to the dome. This results in unsatisfactory simulations outcome, showing that quenching modeling under severe accident-relevant conditions in the given framework requires further development.

if the residual heat of the core cannot be exported in time after a severe accident, the temperature of the fuel element will continue to rise until melting occurs. The core melting is ejected into the lower head of the Reactor Pressure Vessel (RPV) and reacts violently with the residual coolant in the lower head. As a result, a debris bed is formed. The debris bed is a porous structure composed of particles of different sizes. Due to the continuous fission reaction in the debris bed, decay heat is released. Therefore, it needs to keep cooling, otherwise it may melt again. In this paper, the pool boiling phenomenon of porous debris bed is investigated, the two-phase flow and boiling heat transfer calculation model of porous medium with internal heat source is established. Inside, the conservation equations of flow and heat transfer are established for porous debris bed. For the flow resistance, the drag force between gas and liquid in continuous fluid is considered, and the flow resistance of solid to fluid in porous medium area is considered. The heat transfer model takes into account the heat transfer between solid liquid and solid gas, as well as the heat and mass transfer between gas and liquid. At last, the proposed models are validated by using the existing literature data, and the results show that the developed calculation model has a certain accuracy. Furthermore, the effect of different operation condition and structural parameters are discussed.

In the context of nuclear safety, a 4 m3 experimental ventilated enclosure called CARDAMOMETTE has been implemented at IRSN to study the risk of explosion in the event of a hydrogen leakage from a duct in a nuclear facility. Different configurations of hydrogen leakage have been studied allowing to identify those that could potentially lead to explosivity conditions. For safety reason, helium was considered to simulate the behavior of hydrogen. Thanks to high level instrumentation (PIV, He mass spectrometry) and a well-equipped facility allowing local measurements inside the enclosure, a lot of data has been acquired, ensuring a very accurate validation of the CFD code ANSYS CFX. The objective of this validation is to evaluate the capabilities of the CFD code to predict the potential risk of explosivity depending on gas leakage and ventilation configurations.

For this purpose, an experimental and numerical program has been launched to study the influence of ventilation (location of air inlets, renewal rates), gas leakage configurations (location and flowrate, impinging jet) and space clutter (cylindrical container, tubes network, suspended ceiling) on helium dispersion inside the experimental bench and to highlight those leading to hazardous situations.

First, code-experiment comparisons of airflows inside the enclosure were led to ensure the capability of the CFD code to reproduce experimental airflows for some configurations. PIV velocity fields and experimental air renewal curves have been compared to those obtained with CFD calculations, showing a satisfactory agreement. Thanks to this first step, optimal numerical parameters (turbulence model, mesh, boundary conditions) have been chosen.

Secondly, studies of helium dispersion were carried out according to the different configurations presented before. In this paper, only results for free helium jet and impinging helium jet on the wall are presented. Experimental and numerical results of local concentrations were compared, showing a very good agreement and hence the capability of the code to highlight the high concentration areas. Sensitivity studies about turbulent Schmidt number were also led, allowing to define the best numerical dataset depending on the helium injection configurations.

Other experimental and numerical comparisons are currently in progress, especially for the configuration of an impinging helium jet on a cylindrical container.

In order to evaluate the safety performance of microreactor air-cooling residual heat removal system and provide reference for the subsequent experimental design of microreactor air-cooling residual heat removal system. A two-dimensional simplified model of microreactor air-cooling residual heat removal system was established based on CFD software. The steady-state calculation and transient numerical simulation of reactor startup and shutdown and station black-out (SBO) accident were carried out. By analyzing the temperature field, velocity field distribution, residual heat removal capacity, structural temperature distribution and natural convection characteristics in the transient process in the system, the following conclusions are obtained: It takes about 50s for the system to establish the natural circulation initially when the reactor is started. When the SBO accident occurs, the system can change from forced ventilation condition to the natural circulation condition in a short time to ensure the long-term cooling of the microreactor. The air-cooling residual removal system can operate normally and remove the decay heat after the reactor shutdown.

Tornadoes are a common weather hazard to consider as they may damage nuclear power plant facilities. In such situations, damage to the power plant facilities due to windborne debris should be evaluated. Tornado-like vortex (TLV) generators have been built to determine the effect of wind load under various conditions. However, installing wind-borne debris inside these facilities was difficult to protect them from damage. Therefore, a numerically efficient model that could evaluate the wind load of an actual tornado is highly demanded. This study will focus on the VorTECH facility at Texas Tech University as a typical Ward-type TLV chamber and present a numerically efficient and reproducible TLV model based on its experimental information. A series of axisymmetric 2D RANS simulations targeting the VorTECH experiment was conducted. Based on results, the following results were derived. Firstly, the mesh resolution at the wall of the confluence region was found to be important for the development of the boundary layer. The blended mesh did not affect the convergence and the resulting velocity distribution. As for the domain effect, the existence of the confluence region, the wall boundary condition, and the existence of the porous model all affected the flow structure. It was concluded that the development of the boundary layer, the acceleration of the tangential flow inside the confluence region, the roughness of the touchdown region, and the flow resistance at the rectifier should be carefully discussed to develop a reproducible TLV model.

As one of the most promising concepts for GEN-IV reactors, Liquid Metal Fast Reactor (LMFR) has attracted more research attention. Helical Coil Once-through Tube Steam Generator is a proposed form of steam generator with its unique advantages. It is widely used in various reactor power systems, including LMFR. In this paper, a numerical simulation method of Lead-bismuth Helical Coil Once-through Tube Steam Generator (LHCOTSG) based on porous medium method is proposed. The correctness of the simulation method is validated by comparison with three-dimensional numerical simulation results. Based on this method, a LHCOTSG with a given working condition was designed and studied. Under the power target of 1.5MW, the arrangement of four-layer helical tubes has the optimal comprehensive performance. On this basis, the performance of LHCOTSG with different geometric parameters under different working conditions was evaluated by comprehensive performance evaluation index. The multivariate linear fitting method is adopted to obtain the best geometric model within the range of working conditions and geometric parameters. The highest performance was improved by 3.37%. This study provides a new method reference for numerical simulation and design optimization of LHCOTSG.

Modelling of the interphase momentum exchange mainly involves the bubble size and interphase forces in the subcooled boiling flow simulations utilizing the Euler-Euler two fluid framework. Several mechanisms comprising nucleation, coalescence, breakup and condensation should be primarily considered for bubble size modelling based on iMUSIG model. In present study, the closure models are developed in OpenFOAM and assessed with the DEBORA experiment data. Based on the previously developed five-component wall boiling model, different bubble coalescence and breakup models are employed to simulate the bubble size respectively. Guo model for coalescence, Luo model for breakup, Ranz-Marshall correlation for condensation were proved as the suitable closure models for bubble size modelling in boiling flow, by which the contributions of each mechanism are further analyzed. The liquid in the near-wall region approaches to the saturation state, the condensation effects may be neglected, while the condensation is of importance in the upstream due to the subcooled liquid. The breakup process may also be unimportant for small bubbles compared with the coalescence process in the near-wall or low-subcooling region.