• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3775-3786
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• 2023

Mixing Vane Grid (MVG) is one of the most important structures in fuel assembly due to its high performance in mixing the coolant and ultimately increasing Critical Heat Flux (CHF), which avoids the temperature rising suddenly of fuel rods. To evaluate the mixing performance of the MVG, a Total Diffusion Coefficient (TDC) mixing coefficient is defined in the subchannel analysis code. Conventionally, the TDC of the spacer grid is obtained from the combination of experiments and subchannel analysis. However, the processing of obtaining and determine a reasonable TDC is much challenging, it is affected by boundary conditions and MVG geometries. In is difficult to perform all the large and costing rod bundle tests. In this paper, the CFD method was applied in TDC analysis. A typical 5 × 5 MVG was simulated and validated to estimate the mixing performance of the MVG. The subchannel code was used to calculate the TDC. Firstly, the CFD method was validated from the aspect of pressure drop and lateral temperature distribution in the subchannels. Then the effect of boundary conditions including the inlet temperature, inlet velocities, heat flux ratio between hot and cold rods and the arrangement of hot and cold rods on MVG mixing and TDC were studied. The geometric effects on mixing are also carried out in this paper. The effect of vane pattern on mixing was investigated to determine which one is the best to represent the grid's mixing performance.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.5
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• pp.1269-1279
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• 1995

An experimental study is carried out on turbulent diffusion flames stabilized by a circular cylinder in a divergent duct flow. A commercial grade gaseous propane is injected from two slits on the rod as fuel. Flame stability limits, as well as size and temperrature of recirculation zone, are measured by direct and schlieren photographs to clarify the characteristics and structure of diffusion flames and to assess the effect of various divergent angle of duct. The results of the present study are as follows. Temperature in the recirculation zone decreases with increasing divergent angle. The blow-off velocity in parallel duct is higher than that in divergent duct. Critical blow-off velocity is expected to be about 8-12 degree through blow-off velocity pattern. Regardless of divergent angles, the length of recirculation zone is nearly constant, and this length becomes longer with rod diameter. Pressure gradient has an effect on the eddy structure in shear layer behind the rod. With the increase of divergent angle, large scale eddies by dissipated energy in shear layer are split into small scale eddies, and the flame becomes a typical distributedreacting flame.

• Nuclear Engineering and Technology
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• v.30 no.4
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• pp.342-352
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• 1998

A nuclear design feasibility of soluble boron free(SBF core for the medium-sized(600MWe) PWR was investigated. The result conformed that soluble boron free operation could be performed by using current PWR proven technologies. Westinghouse advanced reactor, AP-600 was chosen as a design prototype. Design modification was applied for the assembly design with burnable poison and control rod absorber material. In order to control excess reactivity, large amount of gadolinia integral burnable poison rods were used and B4C was used as a control rod absorber material. For control of bottom shift axial power shape due to high temperature feedback in SBF core, axial zoning of burnable poison was applied to the fuel assemblies design. The combination of enrichment and rod number zoning for burnable poison could make an excess reactivity swing flat within around 1% and these also led effective control on axial power offset and peak pin power, The safety assessment of the designed core was peformed by the calculation of MTC, FTC and shutdown margin. MTC in designed SBF core was greater around 6 times than one of Ulchin unit 3&4. Utilization of enriched BIO(up to 50w1o) in B4C shutdown control rods provided enough shutdown margin as well as subcriticality at cold refueling condition.

• Transactions of the Korean Society of Mechanical Engineers
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• v.19 no.11
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• pp.2981-2994
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• 1995

In order to elucidate the effects of positive pressure gradient on flame properties, structure and stabilization, an experimental study is made on turbulent diffusion flame stabilized by a circular cylinder in a divergent duct flow. A commercial grade gaseous propane is injected from two slits on the rod as fuel. In this paper, stabilization, characteristics and flame structure are examined by varying the divergent angle of duct. Temperature, ion current and Schlieren photographs were measured. It is found that critical divergent angle is expected to be about 8 ~ 12 degree through blow-off velocity pattern to divergent angle and the positive pressure gradient influences the flame temperature, intensity of ion current and eddy structure behind the rod. With the increase of divergent angle, typical temperature of recirculation zone is low but intensity of ion current is high in shear layer behind rod. Energy distributions of fluctuating temperature and ion current signals turn up low frequency corresponding to large scale eddies but high frequency corresponding to small scale eddies as well as low with the increase of divergent angle. Therefore the flame structure becomes a typical distributed-reacting flame.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2709-2716
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• 2020

Compositions of large nuclear cores (e.g. boiling water reactors) are highly heterogeneous in terms of fuel composition, control rod insertions and flow regimes. For this reason, they usually lack high order of symmetry (e.g. 1/4, 1/8) making it difficult to estimate their neutronic parameters for large spaces of possible loading patterns. A detailed hyperparameter optimization technique (a combination of manual and Gaussian process search) is used to train and optimize deep neural networks for the prediction of three neutronic parameters for the Ringhals-1 BWR unit: power peaking factors (PPF), control rod bank level, and cycle length. Simulation data is generated based on half-symmetry using PARCS core simulator by shuffling a total of 196 assemblies. The results demonstrate a promising performance by the deep networks as acceptable mean absolute error values are found for the global maximum PPF (~0.2) and for the radially and axially averaged PPF (~0.05). The mean difference between targets and predictions for the control rod level is about 5% insertion depth. Lastly, cycle length labels are predicted with 82% accuracy. The results also demonstrate that 10,000 samples are adequate to capture about 80% of the high-dimensional space, with minor improvements found for larger number of samples. The promising findings of this work prove the ability of deep neural networks to resolve high dimensionality issues of large cores in the nuclear area.

• Nuclear Engineering and Technology
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• v.31 no.3
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• pp.314-327
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• 1999

A subchannel analysis code MATRA applicable to PWRs and ALWRs has been developed to be run on an IBM PC or HP WS based on the existing CDC CYBER mainframe version of COBRA-Rf-1. This MATRA code is a thermal-hydraulic analysis code based on the subchannel approach for calculating the enthalpy and How distribution in fuel assemblies and reactor cores for both steady-state and transient conditions. HATRA has been provided with an improved structure, various functions, and models to give more convenient user environment and to enhance the code accuracy. Among them, the pressure drop model has been improved to be applied to non-square-lattice rod arrays, and the models for the lateral transport between adjacent subchannels have been improved to enhance the accuracy in predicting two-phase flow phenomena. The predictions of MATRA were compared with the experimental data on the flow and enthalpy distribution in some sample rod-bundle cases to evaluate the performance of MATRA. All the results revealed that the predictions of MATRA were better than those of COBRA-IV-I.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.311-316
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• 1996

제어봉이탈사고시의 핵연료봉 거동을 연구로에서의 반응도사고 모사실험 결과와 기존의 핵연료 손상기준을 비교하여 분석하였다. 반응도사고시 고연소도 핵연료의 손상은 주로 PCMI 기구로 발생하는데, 고연소도에서의 피복관의 부식 및 수소화 그리고 방사선조사에 의한 연성감소와 산화층 박리로 인한 수소화합물의 국부적인 집중화로 인한 피복관의 현저한 연성감소가 주요 원인이었다. 기존의 핵연료 손상 기준에서 DNB가 일어날때 핵연료 손상이 발생한다는 가정은 낮은 핵연료엔탈피에서 핵연료 손상이 일어나는 것과 동일함을 확인하였으며, 현재까지 발표된 실험자료와 핵연료손상기구의 분석을 통해 연소도에 따른 반응도사고시의 핵연료손상기준을 예비적으로 유도하였다. 핵연료손상은 낮은 연소도에는 DNB로 발생하고 고연소도에서는 PCMI로 발생할수 있기 때문에, 과도상태에서의 고연소도 핵연료의 건전성 유지를 위해서는 피복관 산화층의 박리로 인한 수소화합물의 집중화로 피복관의 연성이 감소되는 것을 방지할 필요가 있다.

• Nuclear Engineering and Technology
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• v.52 no.3
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• pp.477-484
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• 2020

The control rod drive mechanism needs to be debugged after reactor fresh fuel loading. It is of great importance to monitor the subcriticality of this process accurately. A modified method was applied to the subcriticality monitoring process, in which only a single control rod cluster was fully withdrawn from the core. In order to correct the error in the results obtained by Neutron Source Multiplication Method, which is based on one point reactor model, Monte Carlo neutron transport code was employed to calculate the fission neutron distribution, the iterated fission probability and the neutron flux in the neutron detector. This article analyzed the effect of a coarse mesh and a fine mesh to tally fission neutron distributions, the iterated fission probability distributions and to calculate correction factors. The subcriticality before and after modification is compared with the subcriticality calculated by MCNP code. The modified results turn out to be closer to calculation. It's feasible to implement the modified NSM method in large local reactivity addition process using Monte Carlo code based on 3D model.

• Transactions of the Korean Society of Mechanical Engineers
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• v.18 no.2
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• pp.512-518
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• 1994

An experimental study is made on turbulent diffusion flames stabilized by a circular cylinder in a divergence flow. In this paper, stabilization characteristics and flame structure are examined by varying the divergence angle of duct and position of a circular cylinder. The fuel used is a commercial grade gaseous propane injected by two slit of rod. It is found that the positive pressure gradient greatly influences the eddy structure behind the rod. and that two different kinds of combustion patterns exist at the blowoff limit depending on the divergent angle of duct. They are distinguished by their wake structures: one associated with Karman vortex shedding, the other without it. Also, the blowoff velocity in the former is found to be higher than in the later.

• 전기의세계
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• v.15 no.5
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• pp.1-5
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• 1966

The stability and safety of operation of a reactor is determined mainly by the sign and magnitude of its reactivity responses to temperature changes. Reactors are subject to temperature fluctuations due to the changes in reactor power and ambient temperature. These temperature fluctuations cause reactivity disturbances through changes in the nuclear and physical properties of the core. Because of these important phenomena by the temperature effects, a large portion of study and testing on a reactor design has been conducted. In this experiment the overall temperature coefficient of the TRIGA MARK-II reactor is measured. The basic procedure is to change the tgemperature of the water moderator, and from the movements of a newly recalibrated control rod(this is necessary due to the effects of fuel burn-up and control rod depression) required to mintain criticality, the reactivity worth of the temperature change is determined. From this measurement, the overall temperature coefficient seems to be smoothly varying, almost a linear function of temperature, and a value of approximately -0.267${\c}$/$^{\circ}C$ can be obtained for an average temperature range from $17.6^{\circ}C$ to $32.5^{\circ}C$.