Thermal-hydraulic Characteristics Research Articles

Development and validation of a one-dimensional solidification model — Part II: Model validation for molten lead-bismuth

The lead-bismuth eutectic (LBE) cooled fast reactor (LFR) has shown promising applications due to its excellent neutron economy and thermal-hydraulic characteristics. Nevertheless, the high-freezing point of LBE arises a risk of coolant solidification, which is identified as a design basis accident for LFRs. Accurate one-dimensional (1D) solidification models are critical for analyzing the coolant solidification behavior. However, the development and evaluation of solidification models are hindered by the lack of experimental data. In this study, a series of molten LBE horizontal penetration solidification experiments were designed and performed with different experimental conditions. The final penetration distances were measured, and the experimental conditions were used as boundary and initial conditions of 1D simulations based on the ASYST-SF code. The simulation results showed the relative error of the final penetration distance is within 12%. This study provides a series of LBE solidification experimental data and demonstrates the reliability of the ASYST-SF code.

Experimental and numerical investigations on thermal-hydraulic characteristics of supercritical CO2 flows in printed circuit heat exchangers

In the supercritical CO2 (SCO2) Brayton cycle, the pre-cooler is employed to cool SCO2 to be near the pseudo-critical point state to minimise compression work and achieve high cycle efficiency. The dramatic variation in thermophysical properties of SCO2 makes heat transfer and flow very complex in the pre-cooler, bringing a lot of difficulty to the design of the related components and system. This study attempts to address this issue. The experiment and simulation are carried out to analyse the influences of mass flux, inlet temperature, and pressure on the thermal-hydraulic characteristics of SCO2 flows in a straight-channel printed circuit heat exchanger (PCHE) employed as a pre-cooler, and the underlying heat transfer mechanism is analysed as well. The local heat transfer coefficient (h) is higher when the phenomenon that the local temperature of SCO2 (Tf) is equal to the pseudo-critical temperature (Tpc) appears in the laminar sublayer than when this phenomenon appears in the buffer layer. The new heat transfer and flow correlations are presented, and the genetic algorithm and back-prediction (GA-BP) neural network is employed to predict the local heat transfer and flow performance. The maximum relative error between the Nusselt number (Nu) predicted by the new heat transfer correlation and Nu obtained via experiment and simulation is within ±20%. The maximum relative error between the Nu predicted by the GA-BP neural network and the Nu obtained through experiment and simulation is within ±15%. The maximum relative error between the Fanning friction factor (f) predicted by the new friction factor correlation and f obtained via experiment and simulation is within ±5%. The maximum relative error between f predicted by the GA-BP neural network and f obtained through experiment and simulation is within ±4%. The present investigations analyse the thermal-hydraulic characteristics of SCO2 in the pre-cooler prototype and could provide useful guidance for the design of pre-coolers for the SCO2 Brayton cycle.

Thermal-Hydraulic Characteristics of Carbon Dioxide in Printed Circuit Heat Exchangers with Staggered Airfoil Fins

Airfoil fin printed circuit heat exchangers (PCHEs) have broad application prospects in the naval, aerospace, electric power, and petrochemical industries. The channel structure is a critical factor affecting their thermal-hydraulic characteristics. In this study, a novel PCHE channel structure with staggered NACA 0025 airfoil-shaped fins was proposed; accordingly, the thermal-hydraulic characteristics of the novel channel structure using carbon dioxide as the working fluid at different fin heights under different operating conditions (trans-, near-, and far-critical) were investigated. The results indicated that the thermal-hydraulic performance of the PCHE under the trans-critical operating condition was better than that under the near-critical and far-critical operating conditions. Compared with conventional airfoil fin channels, the novel airfoil fin channel attained comparable comprehensive performance while reducing the fin volume by 50%, thus achieving a more lightweight PCHE design. The comprehensive performance of the PCHE was the poorest when the fin height was slightly below the channel height, which should be avoided during the design of airfoil fin PCHEs. The results provide theoretical support for the design and optimization of airfoil fin PCHEs.

Evaluating two-phase fluid flow and heat transfer in pillow-plate heat exchangers with nanofluids for organic Rankine cycle in municipal solid waste power plant: A numerical simulation study

One of the key considerations in the design and operation of process industries nowadays is reducing energy waste. The purpose of this research is to assess fluid flow and heat transfer in pillow-plate heat exchangers (PPHEs) using nanofluids for the Organic Rankine Cycle (ORC) in Municipal Solid Waste (MSW) power plants. The thermal performance and hydraulic features of the PPHE were investigated using a two-phase numerical modeling approach. three nanoparticles of Cu, TiO2, and Al2O3 were examined. To precisely determine the temperature distribution within the heat exchanger, various boundary conditions were used. The results demonstrate a strong correlation between the velocity field and temperature, underlining the importance of boundary conditions in determining the temperature profile. Furthermore, the two-phase model illustrates that nanofluids are extremely reliant on inlet velocity and volume fraction. such that at a 2 m/s inlet velocity, increasing the volume percentage to 12.8% contributes to heat transfer coefficient improvements of 7.57%, 12.15%, and 2.71% for water_TiO2, water_Al2O3, and water_Cu, respectively. These findings give significant insight into the fluid flow and heat transfer properties of PPHEs, as well as prospective pathways for increasing thermal performance and decreasing pump power in ORC systems using nanofluids in MSW-generating plants.

Thermal-hydraulic CFD simulation of PWR 5 × 5 bowing fuel assembly

To investigate the effect of fuel assembly bowing on coolant flow and heat transfer, three-dimensional thermal–hydraulic characteristics of 5×5 bowing fuel assembly are investigated using CFD simulation. The effects of different bowing sizes (1 mm and 2 mm) on the coolant thermal–hydraulic characteristics under three patterns of bowing shapes (C-shape, S-shape, and W-shape) of fuel assembly are revealed. The results show that in case 6 bowing of the fuel assembly reduces the average coolant outlet temperature by 0.37 k. Compared with normal conditions, the pressure, velocity, and temperature of the corner channel and the side channel are more affected by bowing than the middle channel. The research has reference value for fuel assembly design.

Thermal Hydraulic Analysis of a Novel Passive Residual Heat Removal System with Loop Thermosyphons for the Pool-vessel Reactor

Advanced pool-vessel reactors require reliable, secure, and passive mechanisms to discharge residual heat from the reactor following shutdown. For this reason, this paper proposes a novel Passive Residual Heat Removal System (PRHRS) with loop thermosyphons. To evaluate the residual heat removal and heat transfer capabilities of the PRHRS, a numerical analysis tool is developed. To achieve this, A thermal hydraulic model is established by secondary development of ANSYS FLUENT, designed to couple the primary loop, reactor pool, and loop thermosyphons. A feasibility study is conducted utilizing the proposed model, and the results demonstrate that this system can efficiently and promptly transfer residual heat to the reactor pool and ultimately to the atmosphere, preventing the core from overheating as well as the pool water from saturated boiling. This model is highly effective in analyzing the thermal hydraulic and safety characteristics of pool-vessel reactors and providing technical support for the design and optimization of the PRHRS.

Heat transfer evaluation of liquid lead-bismuth eutectic cross flow tube bundle: Experimental part

Helical coil once-through tube steam generator (HCOTSG) is an excellent form of heat exchanger. HCOTSG is adopted in various industries, such as Lead-bismuth Fast Reactor (LFR). The flow and heat transfer in the shell side of HCOTSG are important for its design and optimization. It is of great significance to carry out relevant experiments to obtain empirical correlations. In this paper, experimental study on the lead-bismuth eutectic (LBE) cross flow around the tube bundle is carried out to investigate the thermal-hydraulic characteristics based on geometric approximation of local region. Three test sections with different inclination angles of 0°, 4°, 8° are adopted in the experiments with Peclect number (Pe) up to 350. The experimental sections are uniformly heated with heat flux ranging from 20 to 100 kW/m2. The heat transfer characteristic of LBE cross flow around inclined tube bundle is obtained, considering the effects of inclination angle of the test section. The heat transfer correlation of LBE cross flow around inclined tube bundle is proposed for the first time. By comparing the data of the experiments, the results are all within the error range of 10%. The correlation is appropriate for the program development of LBE HCOTSG. This study can provide essential experimental data for the design and code development of LBE HCOTSG.

Rapid analysis of packed pebble beds for thermal–hydraulic characteristics via reduced order models

In recent years, the pebble bed module has garnered increased attention due to its inherent safety and versatile applications, particularly in the form of High-Temperature Gas-Cooled reactors and fusion tritium breeding blankets. Understanding the thermal–hydraulic characteristics within the pebble bed is crucial for the safety evaluation and design of such reactors. To study pore flow in the pebble bed, discrete element coupling methods and computational fluid dynamics (DEM-CFD) are commonly employed. However, simulating thousands of pebbles can be computationally intensive and time-consuming. To address this, reduced order models (ROMs) have been developed to reduce the computational complexity by employing simplified systems instead of the original complex system, thereby significantly saving computational time while maintaining accuracy. In this work, ROMs have been established for body-centered cubic packing, face-centered cubic packing, and randomly packed beds of pebbles. Using the ROM results, subsequent analysis focused on two important thermal–hydraulic characteristics: pressure and effective thermal conductivity. A comparison was made between the results obtained from the full-order model (FOM) using CFD and the reconstructed ROM results. The case studies demonstrate the potential of the proposed approach in rapidly and accurately investigating significant flow and heat transfer features, such as pressure and temperature.

Research of geometrically distinct automobile radiators in terms of thermal-hydraulic characteristics

The fin-and-tube heat exchangers are traditionally used in the automobile cooling system to maintain a proper function and service life of an engine and its components. In the study, the effect of coolant and radiator mileage on thermal–hydraulic characteristics of radiators was investigated. Two types of automobile radiators with different heat exchange areas, mileage, number and size of fans, tube configurations, and coolant mileage were experimentally studied. The local and average Nusselt number of the tubes and fins, heat flux, pressure drop, and efficiency index JF were evaluated based on the numerical studies where the turbulent flow was solved by the k-ω SST model using the ANSYS Fluent software. The shortest cooling time was achieved at the new radiator II with the coolant mileage of 0 km (4 min), while the longest one was achieved at the old radiator II with the coolant mileage of 50,000 km (8.5 min). The radiator II with two fans achieved half the time of cooling when running both fans compared to the only one. The studied range of the coolant and radiator mileage do not have any significant effect on the fault-free operation of the engine cooling system. The efficiency index JF for the new radiator I with the coolant mileage of 0 km is 0.86 – 1.90% higher for the tubes and 0.42 – 1.22% higher for the fins compared with the other investigated radiator and coolant mileages.

Thermal-hydraulic performance of ammonia in manifold microchannel heat sink

With the increasing miniaturisation of electronic devices, the demand for cooling technologies with high efficiency has increased. Therefore, in this study, a manifold microchannel heat sink was designed and fabricated to enhance the dissipating capability of high-heat-flux hotspots. Owing to its excellent thermophysical properties, ammonia was selected as the working fluid, and its thermal–hydraulic characteristics were evaluated. In addition, the effects of flow rate, heat flux, saturation temperature, and inlet conditions were investigated through experiments. High-subcooled inlet experiments revealed that the thermal resistance decreased and pressure drop increased with increasing flow rate. Moreover, localised subcooled boiling was observed near the heated region. In near-saturated inlet experiments, jet impingement and microchannel cooling were identified as the dominant heat transfer mechanisms over different flow ranges. Notably, the higher heat flux, lower saturation temperature, and two-phase inlet condition reduced the thermal resistance under the operating conditions of the study. The proposed heat sink dissipated the heat flux (up to 1542 W/cm2) from the heat source (1 mm × 6 mm) with a pressure drop of 10.6 kPa, and the corresponding heating surface temperature and thermal resistance were 68.05 ℃ and 0.450 K/W, respectively.

Investigation on the DNB-type critical heat flux in a fuel assembly with 5 by 5 rods

In this paper, Eulerian two-phase model coupled with wall boiling model was employed to simulate the departure from nucleate boiling (DNB) in a 5 × 5 fuel assembly. The numerical simulation accurately predicted the critical heat flux (CHF) and also its location. Compared with the calculated results from W-3 formula and FLICA subchannel code, the CFD (Computational Fluid Dynamics) calculations showed good agreement against experimental measurements with the deviation of 5%, but presented more detailed flow distribution. Based on CFD results, the method of CHF prediction was discussed with an eye on wall superheat, void fraction and vapor convective heat flux in rod surfaces. The distribution of flow field and liquid–vapor fraction affected by mixing vane were further studied. Mixing vane can continually influence the downstream thermal hydraulic characteristics, forming the diagonal circulation and square circulation with the flow development. As a result, the bubble aggregation firstly occurred on the surface of corner hot rod instead of central hot rod, which induced the local heat transfer crisis. The CFD calculations presenting the evolution process of physical parameters can provide guidance for rod bundle CHF test and fuel assembly design.

Study on the deposition migration and heat transfer characteristics in the reactor core based on OpenFOAM

Corrosion deposition in heat exchangers may lead to a reduction in heat transfer efficiency, and even can affect equipment operational safety. Traditional CFD deposition heat transfer analysis methods are limited by model size, making it difficult to obtain global, long-term deposition heat transfer characteristics for complex heat exchange equipment. In this paper, a corrosion product deposition prediction and heat transfer model based on sub-channel analysis is proposed. By integrating it into the nuclear reactor core thermal–hydraulic characteristics analysis code CorTAF, which is developed on the OpenFOAM platform, the corrosion deposition and heat transfer calculation for the entire reactor is achieved. A numerical simulation of the deposition and heat transfer of the AP1000 reactor core during a 300-day operation is conducted. The calculation results clearly show the distribution of corrosion products in the reactor core and the temperature changes of the fuel cladding with the deposition days: the maximum deposition thickness on the core cladding surface is 38.3 μm, and the highest temperature increase is 11.0 K. The deposition and cladding temperature rise distribution is not linear with operation time but is closely related to the core power distribution. The modeling method successfully obtains the corrosion deposition and heat transfer characteristics of the entire reactor core, and has potential implications for the prediction of corrosion deposition distribution and thermal effect calculations in other large-scale heat exchange equipment.

Thermal–Hydraulic Calculation and Analysis on Water Wall System of a 700 MWe Ultra-Supercritical CFB Boiler

Ultra-supercritical circulating fluidized bed (CFB) boiler combustion technology has the advantages of environmental protection and good commercialization. As one of the key aspects in ultra-supercritical CFB combustion technology, the water wall system directly determines whether the CFB boiler can safely and effectively transform heat energy. This paper studies the thermal–hydraulic characteristics of the designed water wall system of a 700 MWe ultra-supercritical CFB boiler, the largest one that will be built next year. At four loads of the water wall system, hydrodynamic parameters are calculated based on the successfully verified mathematical models. The results show that the hydrodynamic characteristics of the system are good. The temperature distribution of the working fluid and metal of the water wall system is acceptable and safe at each load, which can support the technical development and improvement of the 700 MWe ultra-supercritical CFB boiler.

Experimental and numerical investigation of sulfuric acid decomposition for hydrogen production via iodine–sulfur cycle

In the context of global warming and the greenhouse effect, the iodine–sulfur cycle is an important method of producing hydrogen on a large scale with low carbon emissions using nuclear energy for future applications. The sulfuric acid decomposition reaction is a key step in this process. A sulfuric acid decomposer with good performance usually has the characteristics of high decomposition rate, small pressure drop loss and reasonable temperature distribution, which can ensure the efficient and stable process of sulfuric acid decomposition. Therefore, an accurate evaluation of the performance of the sulfuric acid decomposer unit and the reproducibility of the whole process is of utmost importance for designing and engineering an actual system for hydrogen generation. In this work, two sets of experimental systems were designed to measure the important kinetic parameters of sulfuric acid decomposition reaction and to determine the thermal hydraulics and chemical reaction performance indicators of the system. Then, a multi-field coupling model in which the phase change and chemical reactions were coupled by user-defined functions was developed based on the experimental kinetic parameters to account for the whole process. The thermal–hydraulic characteristics and the chemical reaction indexes predicted by the model are in good agreement with the experimental results. The maximum errors of the sulfuric acid decomposition fraction and pressure drop do not exceed 5% and 10%, respectively. Overall, this work provided a complete model for predicting the performance of the sulfuric acid decomposer.

Research Progress on Thermal Hydraulic Characteristics of Spent Fuel Pools: A Review

Nuclear power plants (NPPs) produce large amounts of spent fuel while generating electricity. After the spent fuel is taken out of the reactor core, it still has a high decay heat and needs to be cooled for years or even decades before it can be reprocessed or buried deeply. Due to the long storage period of spent fuel, storage safety evaluation is a concern. In this regard, cooling systems are critical for the safe storage of spent fuel. Here, the research progress of cooling methods for spent fuel pools (SFPs) is reviewed, and the structural characteristics, application limitations and heat transfer performance of active and passive cooling technologies under accident conditions are discussed in detail. Moreover, future developments of SFPs are discussed, and the results of this review confirm that there is a great deal of research scope to improve the cooling performance and safety of spent fuel. This paper aims to provide a reference guide for engineers and will be highly beneficial to researchers engaged in spent fuel storage.

Three-dimensional thermal-hydraulic characteristics analysis of plate-type fuel reactor core based on OpenFOAM

In this paper, a full scale three-dimensional thermal-hydraulic analysis method for plate-type fuel reactor core is proposed through integrating traditional sub-channel analysis concepts into CFD platform. The coupled heat transfer model between fuel plates and narrow rectangular coolant channels is established. The three-dimensional thermal-hydraulic analysis code CorTAF-PT is developed based on open-source CFD platform OpenFOAM. The CorTAF-PT is validated against with the Qiu-TAKAHASHI experiment and IAEA 10 MW MTR. The code simulation results agree well with the experiment and MTR operation data with relative error within 5%. The full-scale MTR simulation under steady state and LOFA is carried out through channel-level resolution. In partial blockage accident, maximum temperature of coolant, fuel pellet and cladding increase by 9.75 K, 60.65 K and 60.52 K respectively. Furthermore, the double temperature peaks of fuel plates and local coolant reflux with −0.121 m s−1 are revealed under partial blockage accident. This work provides reference value for correlational studies on thermal-hydraulic characteristics of plate-type fuel reactor core.

Investigation of nuclear reactor core thermal-hydraulic characteristics after partial loss of flow accident

In normal operation conditions of nuclear power plants, the distribution of primary coolant between fuel channels would be considered almost uniform. When different number of Reactor Circulation Pumps (RCPs) are switched off, known as an abnormal condition, this uniform distribution is disturbed and different conditions occur for each channel depending on its position in the core. In this research, the normal and abnormal condition (with one or two tripped RCPs) for a VVER-1000/446 is investigated. For evaluation of the core neutronic calculations and thermal power distribution, USNRC’s PARCS system code is employed. Then a thermal-hydraulics module was developed for performing the T/H calculation of the core zone. The input velocity of each channel in abnormal condition was calculated based on developed CFD model in downcomer and lower plenum of Reactor Pressure Vessel (RPV) by ANSYS-CFX. The results show that, in normal operation, the hot channel is related to the central fuel assembly of the reactor core with the highest relative power equal to 1.29 and total power of 23.74 MW. In this case, the minimum inlet velocity, the maximum coolant outlet velocity, and the maximum fuel temperature are 5.6 (m/s), 330.96 (°C), and 1345.8 (°C), respectively. In the cases of operation with one and two tripped RCPs, the hot channel is related to the fuel assemblies with the lowest inlet velocities. The lowest velocities are 0.32 m/s, 0.24 m/s, and 0.22 m/s respectively for the condition with one tripped pump, two tripped pumps placed oppositely, and two tripped pumps placed contiguously. The hot channel numbers in these cases are 158, 102, and 90, respectively. In these channel, the condition of outlet flow would be superheated, but the fuel temperature (1006.3, 1050.9, and 987.9) do not reach the maximum allowable margin. The study confirms the necessity of the coolant distribution consideration in OLCs as well as events that may disturb the symmetry of the coolant flow. It also showed that the lateral fuel assemblies are more at risk in this situation because of a significant reduction in coolant flow. Likewise, the investigations proved the safe continuation of the operation in PLOFA conditions with the preventive algorithm of the emergency protection system and without any need for immediate mitigation actions or operator intervention.

Thermal-hydraulic analysis of He–Xe gas mixture in 2×2 rod bundle wrapped with helical wires

Gas-cooled space reactor, which adopts He–Xe gas mixture as working fluid, is a better choice for megawatt power generation. In this paper, thermal-hydraulic characteristics of He–Xe gas mixture in 2 × 2 rod bundle wrapped with helical wires is numerically investigated. The velocity, pressure and temperature distribution of the coolant are obtained and analyzed. The results show that the existence of helical wires forms the vortexes and changes the velocity and temperature distribution. Hot spots are found at the contact corners between helical wires and fuel rods. The highest temperature of the hot spots reach 1600K, while the mainstream temperature is less than 400K. The helical wire structure increases the friction pressure drop by 20%–50%. The effect extent varies with the pitch and the number of helical wires. The helical wire structure leads to the reduction of Nusselt number. Comparing thermal-hydraulic performance ratios (THPR) of different structures, the THPR values are all less than 1. It means that gas-cooled space reactor adopting helical wires could not strengthen the core heat removal performance. This work provides the thermal-hydraulic design basis for He–Xe gas cooled space nuclear reactor.

Development and validation of a general two-fluid thermal-hydraulic analysis code for reactors with plate-type fuel assemblies

A general two-fluid thermal-hydraulic analysis code, TTHAC-PFA (Two-fluid Thermal-Hydraulic Analysis Code for reactors using Plate-type Fuel Assemblies), has been developed. The code was developed on basis of the fundamental governing equations. Reasonable mathematic and physical models for parallel rectangular channels and fuel plates have been built up to meet the command of the safety analysis for plate-type fuel reactors. The coolant distribution is conducted on basis of the pressure drop equation theory. Normal subchannel approach is adopted in the code. Fortran language is adopted. The RA-6 experiment is used to evaluate the heat transfer models. The max deviation of the simulations from the measured value is about 5.7 °C. Then, experiments on two-phase flow pressure drop are adopted to validate the code. Approximately 97% of the data points are within ±20% of the deviation range. In addition, CARR (China Advanced Research Reactor) is adopted as the benchmark for normal condition validation. Flow distribution validation and temperature profile validation are conducted. The flow validation presents negligible mean discrepancy and the temperature differences are within the scope of 6 °C. All validations are in good agreement, which indicates the developed code meet the requirements of analyzing thermal-hydraulic characteristics for parallel rectangular channels in reactors. Furthermore, TTHAC-PFA is applied to the investigate channel blockage accident occurred in the CARR. The mass flow is redistributed among channels due to the obstruction. In the obstructed channel and two nearby channels, the coolant temperatures increase a lot. The coolant temperature in the obstructed channel rises by 60 °C. In the rest of the channels, the coolant temperatures decrease with the increase of coolant flow. Besides, the two fuel plates, which form the blocked channel, are asymmetrically cooled due to flow maldistribution. The temperature difference between the cladding on both sides of the fuel plate can reach 40 °C. As a result of partial blockage, the cladding temperature on the side of the blocked channel rises by 63 °C. During the blockage accident, the plate temperature is under the melting temperature. The results indicate the significance of studying the effects of the adjacent channels near the obstructed channel for reactors with plate-type fuel elements.

Preliminary code development for single-phase liquid natural circulation flow simulation based on two-phase compressible model

It is of great significance to study the thermal-hydraulic characteristics of Lead-based Fast Reactor. In these years, the computational fluid dynamics method has become an attractive way in thermal-hydraulic and safety simulations of liquid metal cooled reactors. In order to investigate thermal-hydraulic characteristics of single-phase natural circulation flow in liquid metal cooled reactors, a self-developed CFD code was developed in this paper. The new code was proposed based on the two-phase compressible model, including liquid and vapor governing equations and a series of auxiliary models. The commercial CFD software ANSYS FLUENT was used to verify and validate the new code by comparing their simulation results under a typical heat driven cavity flow case. The comparison showed that the results simulated by the new code had good agreement with the ones calculated by ANSYS FLUENT. Furthermore, by coupling the above CFD model with the fuel pin heat transfer model, the application was conducted based on a full-scale reactor steady-state simulation of a natural circulation driven LBE-cooled fast reactor. From the results, it can be concluded that the new code can simulate and evaluate thermal-hydraulic behaviors of the LBE-cooled reactor well.