• Economic and Environmental Geology
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• v.41 no.6
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• pp.731-746
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• 2008

Most multiphase flow simulators following fractional flow approach assume incompressibility of fluid and matrix or consider only two phase flow (water and air, water and NAPL). However, in this study, mathematical governing equations were developed for fully compressible three-phase flow using fractional flow based approach. Also, fully compressible multiphase flow simulator (CMPS) considering compressibilities of matrix and fluid was developed using the mathematical governing equations. In order to verify CMPS, the CMPS were compared with analytical solution and the existing multiphase flow simulator, MPS, which had been developed for simulating incompressible multiphase flow (Suk and Yeh 2007; Suk and Yeh 2008). According to the results, solutions of CMPS and MPS and analytical solutions are well matched each other. Thus, it is found that CMPS has the capability of simulating compressible three phase flow phenomena assuming compressibilities of fluids and matrix.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.59-68
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• 2020

ECCMIX component was introduced in RELAP5/MOD3 for calculating the interfacial condensation. Compared to other existing components in RELAP5, user experience of ECCMIX component is restricted to developmental assessment applications. To evaluate the capability of the ECCMIX component, ECCS experiment was conducted which included single-phase and two-phase thermal mixing. The experiment was carried out with test sections containing a main pipe (70 mm inner diameter) and a branch pipe (21 mm inner diameter) under the atmospheric pressure. The steam mass flow in the main pipe ranged from 0 to 0.0347 kg/s, and the subcooled water mass flow in the branch pipe ranged from 0.0278 to 0.1389 kg/s. The comparison of the experimental data with the calculation results illuminated that although the ECCMIX component was more difficult to converge than Branch component, it was a more appropriate manner to simulate interfacial condensation under two-phase thermal mixing circumstance, while the two components had no differences under single-phase circumstance.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.897-907
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• 2020

The flow characteristics of reactor fuel assembly always intrigue the designers and the experimentalists among the myriad phenomena that occur simultaneously in a nuclear core. In this work, the visual experimental method has been developed on the basis of refraction index matching (RIM) and particle image velocimetry (PIV) techniques to investigate the detailed flow characteristics in China fast reactor fuel subassembly. A 7-rod bundle of simulated fuel subassembly was fabricated for fine examination of flow characteristics in different subchannels. The experiments were performed at condition of Re=6500 (axial bulk velocity 1.6 m/s) and the fluid medium was maintained at 30℃ and 1.0 bar during operation. As for results, axial and lateral flow features were observed. It is shown that the spiral wire has an inhibitory effect on axial flow and significant intensity of lateral flow mixing effect is induced by the wire. The root mean square (RMS) of lateral velocity fluctuation was acquired after data processing, which indicates the strong turbulence characteristics in different flow subchannels.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.446-454
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• 2021

Critical heat flux (CHF) has traditionally been evaluated using look-up tables or empirical correlations for nuclear power plants. However, under complex moving condition, it is necessary to reconsider the CHF characteristics since the conventional CHF prediction methods would no longer be applicable. In this paper, the additional forces caused by motions have been added to the annular film dryout (AFD) mechanistic model to investigate the effect of moving condition on CHF. Moreover, a theoretical model of the natural circulation loop with additional forces is established to reflect the natural circulation characteristics of the loop system. By coupling the system loop with the AFD mechanistic model, a CHF prediction program called NACOM for natural circulation loop under moving condition is developed. The effects of three operating conditions, namely stationary, inclination and rolling, on the CHF of the loop are then analyzed. It can be clearly seen that the moving condition has an adverse effect on the CHF in the natural circulation system. For the calculation parameters in this paper, the CHF can be reduced by 25% compared with the static value, which indicates that it is important to consider the effects of moving condition to retain adequate safety margin in subsequent thermal-hydraulic designs.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.637-646
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• 2021

The free-piston Stirling engine (FPSE) has been widely used in aerospace owing to its advantages of high efficiency, high reliability, and self-starting ability. In this paper, a 20-kW FPSE is proposed by analyzing the requirements of space nuclear power reactor. A code was developed based on an improved simple analysis method to evaluate the performance of the proposed FPSE. The code is benchmarked with experimental data, and the maximum relative error of the output power is 17.1%. Numerical results show that the output power is 21 kW, which satisfies the design requirements. The results show that: a) reducing the pressure shell's thickness can improve the output power significantly; b) the system efficiency increases with the wire porosity, while the growth of system efficiency decreases when the porosity is higher than 80%, and system efficiency exhibits a linear relationship with the temperatures of the cold and hot sides; c) the system efficiency increases with the compression ratio; the compression ratio increases by 16.7% while the system efficiency increases by 42%. This study can provide valuable theoretical support for the design and analysis of FPSEs for space nuclear power reactors.

• Nuclear Engineering and Technology
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• v.53 no.12
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• pp.3902-3909
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• 2021

SACOS series of subchannel analysis codes have been developed by XJTU-NuTheL for many years and are being used for the thermal-hydraulic safety analysis of various reactor cores. To achieve fine whole core pin-level analysis, the input preprocessing and parallel capabilities of the code have been developed in this study. Preprocessing is suitable for modeling rectangular and hexagonal assemblies with less error-prone input; parallelization is established based on the domain decomposition method with the hybrid of MPI and OpenMP. For domain decomposition, a more flexible method has been proposed which can determine the appropriate task division of the core domain according to the number of processors of the server. By performing the calculation time evaluation for the several PWR assembly problems, the code parallelization has been successfully verified with different number of processors. Subsequent analysis results for rectangular- and hexagonal-assembly core imply that the code can be used to model and perform pin-level core safety analysis with acceptable computational efficiency.

• Journal of Ocean Engineering and Technology
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• v.35 no.1
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• pp.50-58
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• 2021

The demand for eco-friendly energy is expected to increase due to the recently strengthened environmental regulations. In particular, the flow inside the pipe used in a cargo handling system (CHS) or fuel gas supply system (FGSS) of hydrogen transport ships and hydrogen-powered ships exhibits a very complex pattern of multiphase-thermal flow, including the boiling phenomenon and high accuracy analysis is required concerning safety. In this study, a feasibility study applying the boiling model was conducted to analyze the multiphase-thermal flow in the pipe considering the phase change. Two types of boiling models were employed and compared to implement the subcooled boiling phenomenon in nucleate boiling numerically. One was the "Rohsenow boiling model", which is the most commonly used one among the VOF (Volume-of-Fluid) boiling models under the Eulerian-Eulerian framework. The other was the "wall boiling model", which is suitable for nucleate boiling among the Eulerian multiphase models. Moreover, a comparative study was conducted by combining the nucleate site density and bubble departure diameter model that could influence the accuracy of the wall boiling model. A comparison of the Rohsenow boiling and the wall boiling models showed that the wall boiling model relatively well represented the process of bubble formation and development, even though more computation time was consumed. Among the combination of models used in the wall boiling model, the simulation results were affected significantly by the bubble departure diameter model, which had a very close relationship with the grid size. The present results are expected to provide useful information for identifying the characteristics of various parameters of the boiling model used in CFD simulations of multiphase-thermalflow, including phase change and selecting the appropriate parameters.

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.579-588
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• 2022

In order to meet the needs of domestic reactor severe accident analysis program, a MIDAC (Module Invessel Degraded severe accident Analysis Code) is developed and maintained by Xi'an Jiaotong University. As the accuracy of the calculation results of the analysis program is of great significance for the formulation of severe accident mitigation measures, the article select three experiments to evaluate the updated severe accident models of MIDAC. Among them, QUENCH-06 is the international standard No.45, QUENCH-16 is a test for the analysis of air oxidation, and FROMA is an out-of-pile fuel rod melting experiment recently carried out by Xi'an Jiaotong University. The heating and melting model with lumped parameter method and the steam oxidation model with Cathcart-Pawel and Volchek-Zvonarev correlations combination in MIDAC could better meet the needs of severe accident analysis. Although the influence of nitrogen still need to be further improved, the air oxidation model with NUREG still has the ability to provide guiding significance for engineering practice.

• Nuclear Engineering and Technology
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• v.54 no.10
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• pp.3909-3918
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• 2022

In the event of a severe accident, the reactor core may melt due to insufficient cooling. the high-temperature core melt will have a strong interaction (FCI) with the coolant, which may lead to steam explosion. Steam explosion would pose a serious threat to the safety of the reactors. Therefore, the study of steam explosion is of great significance to the assessment of severe accidents in nuclear reactors. This research focuses on the development of a two-dimensional steam explosion integrated analysis code called SEINA. Based on the semi-implicit Euler scheme, the three-phase field was considered in this code. Besides, the influence of evaporation drag of melt and the influence of solidified shell during the process of melt droplet fragmentation were also considered. The code was simulated and validated by FARO L-14 and KROTOS KS-2 experiments. The calculation results of SEINA code are in good agreement with the experimental results, and the results show that if the effects of evaporation drag and melt solidification shell are considered, the FCI process can be described more accurately. Therefore, it is proved that SEINA has the potential to be a powerful and effective tool for the analysis of steam explosions in nuclear reactors.

• Nuclear Engineering and Technology
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• v.53 no.7
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• pp.2184-2194
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• 2021

The Free-Piston Stirling engine (FPSE) is of interest for many research in aerospace due to its advantages of long operating life, higher efficiency, and zero maintenance. In this study, a 1-kW FPSE was proposed by analyzing the requirements of Space Reactor Power Systems (SRPS), of which performance was evaluated by developing a code through the Simple Analysis Method. The results of SAM showed that the critical parameters of FPSE could satisfy the designed requirements. The heater of the FPSE was designed with the copper rectangular fins to enhance heat transfer, and the parametric study of the heater was performed with Computational Fluid Dynamics (CFD) software STAR-CCM⁺. The Performance Evaluation Criteria (PEC) was used to evaluate the heat transfer enhancement of the fins in the heater. The numerical results of the CFD program showed that pressure drop and Nusselt number ratio had a linear growth with the height of fins, and PEC number decreased as the height of fins increased, and the optimum height of the fin was set as 4 mm according to the minimum heat exchange surface area. This paper can provide theoretical supports for the design and numerical analysis of an FPSE for SRPSs.