• Nuclear Engineering and Technology
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• v.53 no.10
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• pp.3217-3228
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• 2021

Flow blockage of the fuel assembly in the lead-based fast reactor (LFR) may produce critical local spots, which will result in cladding failure and threaten reactor safety. In this study, the flow blockage characteristics were analyzed with the sub-channel analysis method, and the circumferentially-varied method was employed for considering the non-uniform distribution of circumferential temperature. The developed sub-channel analysis code SACOS-PB was validated by a heat transfer experiment in a blocked 19-rod bundle cooled by lead-bismuth eutectic. The deviations between the predicted coolant temperature and experimental values are within ±5%, including small and large flow blockage scenarios. And the temperature distributions of the fuel rod could be better simulated by the circumferentially-varied method for the small blockage scenario. Based on the validated code, the analysis of blockage characteristics was conducted. It could be seen from the temperature and flow distributions that a large blockage accident is more destructive compared with a small one. The sensitivity analysis shows that the closer the blockage location is to the exit, the more dangerous the accident is. Similarly, a larger blockage length will lead to a more serious case. And a higher exit temperature will be generated resulting from a higher peak coolant temperature of the blocked region. This work could provide a reference for the future design and development of the LFR.

• Nuclear Engineering and Technology
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• v.54 no.7
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• pp.2728-2735
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• 2022

The research on the flow and heat transfer characteristics of lead bismuth(LBE) is significant for the thermal-hydraulic calculation, safety analysis and practical application of lead-based fast reactors(LFR). In this paper, a new CFD model is proposed to solve the thermal-hydraulic analysis of LBE. The model includes two parts: turbulent model and turbulent Prandtl, which are the important factors for LBE. In order to find the best model, the experiment data and design of 19-pin hexagonal rod bundle with spacer grid, undertaken at the Karlsruhe Liquid Metal Laboratory (KALLA) are used for CFD calculation. Furthermore, the turbulent model includes SST k - 𝜔 and k - 𝜀; the turbulent Prandtl includes Cheng-Tak and constant (Pr_t =1.5,2.0,2.5,3.0). Among them, the combination between SST k - 𝜔 and Cheng-Tak is more suitable for the experiment. But in the low Pe region, the deviation between the experiment data and CFD result is too much. The reason may be the inlet-effect and when Pe is in a low level, the number of molecular thermal diffusion occupies an absolute advantage, and the buoyancy will enhance. In order to test and verify versatility of the model, the NCCL performed by the Nuclear Thermal-hydraulic Laboratory (Nuthel) of Xi'an Jiao tong University is used for CFD to calculate. This paper provides two verification examples for the new universal model.

• Nuclear Engineering and Technology
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• v.53 no.3
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• pp.763-775
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• 2021

The thermal hydraulic performances of Steam Generator (SG) under both steady and transient operation conditions are of great importance for the safety and economy in nuclear power plants. In this paper, based on our self-developed SG thermal hydraulic analysis code STAF (Steam-generator Thermalhydraulic Analysis code based on Fluent), an improved new version STAF-CT (fully Coupling and Transient) is developed and introduced. Compared with original STAF, the new version code STAF-CT has two main functional improvements including "Transient" and "Fully Three Dimensional Coupling" features. In STAF-CT, a three dimensional energy transferring module is established which can achieve energy exchange computing function at the corresponding position between two sides of SG. The STAF-CT is validated against the international benchmark experiment data and the results show great agreement. Then the U-shaped SG in AP1000 nuclear power plant is modeled and simulated using STAF-CT. The results show that three dimensional flow fields in the primary side make significant effect on the energy source distribution between two sides. The development of code STAF-CT in this paper can provide an effective method for further SG high fidelity research in the nuclear reactor system.

• 한국전산유체공학회:학술대회논문집
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• 2010.05a
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• pp.115-121
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• 2010

Cavitating flow is widely shown in many engineering systems, such as marine propellers, pump impellers, nozzles, injectors, torpedoes, etc. The present work focuses on the numerical analysis of the multiphase flow around the underwater vehicle which was launched from a submarine. The governing equation is the Navier-Stokes equation with a homogeneous mixture mode. The multiphase flow solver uses an implicit preconditioning scheme in curvilinear coordinate. For the code validation, the results from the present work are compared with the existing experimental and numerical results, and a reasonably good agrements are obtained. The multiphase flow around an underwater vehicle is simulated which includes submarine wake effects.

• Nuclear Engineering and Technology
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• v.49 no.6
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• pp.1318-1325
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• 2017

Direct Numerical Simulation (DNS) serves as an irreplaceable tool to probe the complexities of multiphase flow and identify turbulent mechanisms that elude conventional experimental measurement techniques. The insights unlocked via its careful analysis can be used to guide the formulation and development of turbulence models used in multiphase computational fluid dynamics simulations of nuclear reactor applications. Here, we perform statistical analyses of DNS bubbly flow data generated by Bolotnov ($Re_{\tau}=400$) and LueTryggvason ($Re_{\tau}=150$), examining single-point statistics of mean and turbulent liquid properties, turbulent kinetic energy budgets, and two-point correlations in space and time. Deformability of the bubble interface is shown to have a dramatic impact on the liquid turbulent stresses and energy budgets. A reduction in temporal and spatial correlations for the streamwise turbulent stress (uu) is also observed at wall-normal distances of $y^+=15$, $y/{\delta}=0.5$, and $y/{\delta}=1.0$. These observations motivate the need for adaptation of length and time scales for bubble-induced turbulence models and serve as guidelines for future analyses of DNS bubbly flow data.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.162-176
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• 2022

In Lead-based reactor (LBR) severe accident, the meltdown and migration inside the reactor core will lead to fuel fragment concentration, which may further cause re-criticality and even core disintegration. Accurately predicting the migration and solidification behavior of melt in LBR severe accidents is of prime importance for safety analysis of LBR. In this study, the Moving Particle Semi-implicit (MPS) method is validated and used to simulate the migration and solidification behavior. Two main surface tension models are validated and compared. Meanwhile, the MPS method is validated by the L-plate solidification test. Based on the improved MPS method, the migration and solidification behavior of melt in LBR severe accident was studied furthermore. In the Pb-Bi coolant, the melt flows upward due to density difference. The migration and solidification behavior are greatly affected by the surface tension and viscous resistance varying with enthalpy. The whole movement process can be divided into three stages depending on the change in velocity. The heat transfer of core melt is determined jointly by two heat transfer modes: flow heat transfer and solid conductivity. Generally, the research results indicate that the MPS method has unique advantage in studying the migration and solidification behavior in LBR severe accident.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.937-946
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• 2020

The aim in the present work is to simulate accident scenarios of AP1000 during the small-break loss-of-coolant accident (SBLOCA) and investigate the performance and behavior of automatic depressurization system (ADS) during accidents by using MIDAC (The Module In-vessel Degradation severe accident Analysis Code). Four types of accidents with different hypothetical conditions were analyzed in this study. The impact on the thermal-hydraulic of the reactor coolant system (RCS), the passive core cooling system and core degradation was researched by comparing these types. The results show that the RCS depressurization becomes faster, the core makeup tanks (CMT) and accumulators (ACC) are activated earlier and the effect of gravity water injection is more obvious along with more ADS valves open. The open of the only ADS1-3 can't stop the core degradation on the basis of the first type of the accident. The open of ADS1-3 has a great impact on the injection time of ACC and CMT. The core can remain intact for a long time and the core degradation can be prevent by the open of ADS-4. The all results are significant and meaningful to understand the performance and behavior of the ADS during the typical SBLOCA.

• 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.

• 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.