• Applied Chemistry for Engineering
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• v.2 no.1
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• pp.30-37
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• 1991

The production of hydrogen from steam by reduced iron with additives such as CuO, $In_2O_3$, $MoO_3$ and $WO_3$ has been kinetically investigated. It was shown that all additives have a promoting effect on reaction activity in the order of $$MoO_3{\gg}In_2O_3{\sim_=}WO_3{\sim_=}CuO$$. The shrinking core model was applied to predict the complete conversion time and the results were quite comparable with experimental values. The reaction was carried out in a fixed flow reactor packed with reduced iron with 1 wt % of additives under the conditions, $600-750^{\circ}C$, Ar flow rate of 1 L/min and steam partial pressure of 0.085 atm. The apparent activation energies were 14.2, 20.9, 21.3, 22.4 and 27.9 kJ/mol with $MoO_3$, $In_2O_3$, $WO_3$, CuO and without additive, respectively.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1571-1578
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• 2020

As a critical material in very/high-temperature gas-cooled reactors, graphite material directly affects the safety of the reactor core structures. Owing to the complex structures of graphite material in reactors, the material typically undergoes complex stress states. It is, therefore, necessary to study its mechanical properties, failure modes, and strength criteria under complex stress states so as to provide guidance for the core structure design. In this study, compressive failure tests were performed for graphite material under the condition of different confining pressures, and the effects of confining pressure on the triaxial compressive strength and Young's modulus of graphite material were studied. More specifically, graphite material based on the fracture surfaces and fracture angles, the graphite specimens were found to exhibit four types of failure modes, i.e., tension failure, shear-tension failure, tension-shear failure and shear failure, with increasing confining pressure. In addition, the Mohr strength envelope of the graphite material was obtained, and different strength criteria were compared. It showed that the parabolic Mohr-Coulomb criterion is more suitable for the strength evaluation for the graphite material.

• Journal of the Korean Institute of Electrical and Electronic Material Engineers
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• v.22 no.8
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• pp.689-693
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• 2009

We investigate the comparison of fault current characteristics between the separates three-phase flux-lock type superconducting fault current limiter(SFCL) and integrated three-phase flux-lock type superconducting fault current limiter(SFCL). The single-phase flux-lock type SFCL consists of two coils. The primary coil is wound in parallel to the secondary coil on an iron core and superconducting elements are connected to secondary coil in series. Superconducting elements are used by the YBCO coated conductor. The separated three-phase flux-lock type SFCL consists of single-phase flux-phase type SFCL in each phase. But the integrated three-phase flux-lock type SFCL consists of three-phase flux-reactors wound on an iron core. Flux-reactor consists of the same turn's ratio between coil 1 and coil 2 for each single phase. To compare the current limiting characteristics of the separated three-phase flux-lock type SFCL and integrated three-phase flux-lock type SFCL, the short circuit experiments are carried out fault condition such as the single line-to-ground fault. The experimental result shows that fault current limiting characteristic of the separated three-phase flux-lock type SFCL was better than integrated three-phase flux-lock type SFCL. And the integrated three-phase flux-lock type SFCL has an effect on sound phase.

• Journal of the Computational Structural Engineering Institute of Korea
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• v.15 no.1
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• pp.59-68
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• 2002

The cylindrical Upper Guide Structure assembly of the reactor intervals wish the Core Support Barrel and the Inner Barrel Assembly is subjected to flow induced loads horizontally which include random pressure fluctuation due to turbulent flow and pump pulsation pressures. The purpose of this papers is to perform random vibration and harmonic response analyses fort flow induced loads. The dynamic response characteristics due to random turbulence and pump pulsation loads were evaluated using the lumped mass beam model. Especially the model considered the annulus effects due to water gaps existing between cylindrical structures such as the Upper Guide Structure Barrel, the Core Support Barrel, and the Inner Barrel Assembly. The effect of the Inner Barrel Assembly inside the Upper Guide Structure assembly was studied. The peak dynamic responses lot each loading condition due to the addition of IBA were affected by the natural frequencies of the structures. Therefore the peak dynamic responses of the structures should be conservatively obtained from evaluation of dynamic analysis for various loading conditions.

• Nuclear Engineering and Technology
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• v.19 no.1
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• pp.34-41
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• 1987

This work concerns with a comparison of the 1-dimensional (or 1-D) load follow analysis method with reference to the detailed 3-dimensional (or 3-D) computations. For this purpose a 1-D two-group finite difference code, HLOFO, and a 3-D coarse-mesh code based on the modified Borresen's method, CMSNAC, are developed. The CMSNAC code is used to obtain the 3-D power peaks and reactivity parameters in response to power swing from 100 to 50 and back to 100% in the 12-3-6-3 load cycle for the BOL of the KORI Unit 1 PWR core. The 3-D result is then compared with the 1-D HLOFO computations, the cross section and buckling inputs of which are obtained by combining the flux-volume weighting scheme with the approximate flux from the auxiliary 3-D computations. It is shown that the 1-D computation has a limited accuracy, yet it is confirmed that it can describe the core axial average behavior which is fairly consistent with the detailed 3-D computation.

• Journal of the Korean Society of Combustion
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• v.22 no.4
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• pp.19-26
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• 2017

This study investigates the gasification of coal char by $CO_2$ under high pressures in a drop tube furnace(DTF). The rate constants are derived for the shrinking core model using the conventional method based on the set reactor conditions. The computational fluid dynamic(CFD) simulations adopting the rate constants revealed that the carbon conversion was much slower than the experimental results, especially under high temperature and high partial pressure of reactants. Three reasons were identified for the discrepancy: i) shorter reaction time because of the entry region for heating, ii) lower particle temperature by the endothermic reaction, and iii) lower partial pressure of $CO_2$ by its consumption. Therefore, the rate constants were corrected based on the actual reaction conditions of the char. The CFD results updated using the corrected rate constants well matched with the measured values. Such correction of reaction conditions in a DTF is essential in deriving rate constants for any char conversion models by $H_2O$ and $O_2$ as well as $CO_2$.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.941-951
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• 2016

The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM) 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD) calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively) with experimental results.

• Nuclear Engineering and Technology
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• v.43 no.3
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• pp.257-270
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• 2011

A set of reflood tests has been performed using ATLAS, which is a thermal-hydraulic integral effect test facility for the pressurized water reactors of APR1400 and OPR1000. Several important phenomena were observed during the ATLAS LBLOCA reflood tests, including core quenching, down-comer boiling, ECC bypass, and steam binding. The present paper discusses those four topics based on the LB-CL-11 test, which is a best-estimate simulation of the LBLOCA reflood phase for APR1400 using ATLAS. Both homogeneous bottom quenching and inhomogeneous top quenching were observed for a uniform radial power profile during the LB-CL-11 test. From the observation of the down-comer boiling phenomena during the LB-CL-11 test, it was found that the measured void fraction in the lower down-comer region was relatively smaller than that estimated from the RELAP5 code, which predicted an unrealistically higher void generation and magnified the downcomer boiling effect for APR1400. The direct ECC bypass was the dominant ECC bypass mechanism throughout the test even though sweep-out occurred during the earlier period. The ECC bypass fractions were between 0.2 and 0.6 during the later test period. The steam binding phenomena was observed, and its effect on the collapsed water levels of the core and down-comer was discussed.

• Journal of Radiation Protection and Research
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• v.41 no.3
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• pp.237-244
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• 2016

Background: Korea Atomic Energy Research Institute is developing a fission product transport module for predicting the behavior of radioactive materials in the primary cooling system of a nuclear power plant as a separate module, which will be connected to a severe accident analysis code, Core Meltdown Progression Accident Simulation Software (COMPASS). Materials and Methods: This fission product transport (COMPASS-FP) module consists of a fission product release model, an aerosol generation model, and an aerosol transport model. In the fission product release model there are three submodels based on empirical correlations, and they are used to simulate the fission product gases release from the reactor core. In the aerosol generation model, the mass conservation law and Raoult's law are applied to the mixture of vapors and droplets of the fission products in a specified control volume to find the generation of the aerosol droplet. In the aerosol transport model, empirical correlations available from the open literature are used to simulate the aerosol removal processes owing to the gravitational settling, inertia impaction, diffusiophoresis, and thermophoresis. Results and Discussion: The COMPASS-FP module was validated against Aerosol Behavior Code Validation and Evaluation (ABCOVE-5) test performed by Hanford Engineering Development Laboratory for comparing the prediction and test data. The comparison results assuming a non-spherical aerosol shape for the suspended aerosol mass concentration showed a good agreement with an error range of about ${\pm}6%$. Conclusion: It was found that the COMPASS-FP module produced the reasonable results of the fission product gases release, the aerosol generation, and the gravitational settling in the aerosol removal processes for ABCOVE-5. However, more validation for other aerosol removal models needs to be performed.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.5
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• pp.836-842
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• 2016

When an accident occurs in the nuclear power plant, the faulted information might mislead to the high possibility of aggravating the accident. At the Fukushima accident, the operators misunderstood that there was no core exposure despite in the processing of core damage, because the instrument information of the reactor water level was provided to the operators optimistically other than the actual situation. Thus, this misunderstanding actually caused to much confusions on the rapid countermeasure on the accident, and then resulted in multiplying the accident propagation. It is necessary to be equipped with the function that informs operators the status of instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they are able to make a decision more safely. In this study, we have performed various tests for the fault detection sensitivity of an data-driven empirical model to review the usability of the model in the accident conditions. The test was performed by using simulation data from the compact nuclear simulator that is numerically simulated to PWR type nuclear power plant. As a result of the test, the proposed model has shown good performance for detecting the specified instrument faults during normal plant conditions. Although the instrument fault detection sensitivity during plant accident conditions is lower than that during normal condition, the data-drive empirical model can be detected an instrument fault during early stage of plant accidents.