• Journal of Energy Engineering
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• v.12 no.1
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• pp.49-57
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• 2003

Approx. 200.000 bpd vacuum residue oil is produced from oil refineries in Korea, and is supplied to use asphalt, high sulfur fuel oil and for upgrading at the residue hydro-desulfurization unit. Vacuum residue oil has high energy content, however its high sulfur content and high concentration of heavy metals represent improper low grade fuel. To meet growing demand for effective utilization of vacuum residue oil from refineries, recently some of the oil refinery industries in Korea, such as SK oil refinery and LG Caltex refinery, have already proceeded feasibility study to construct 435~500 MWe IGCC power plant and hydrogen production facilities. Recently, KIER (Korea Institute of Energy Research) are studying on the Vacuum Residue gasification process using an oxygen-blown entrained-flow gasifier. The experiment runs were evaluated under the reaction temperature: 1.100~l,25$0^{\circ}C$, reaction pressure: 1~6 kg/$\textrm{cm}^2$G, oxygen/V.R ratio: 0.8~0.9 and steam/V.R ratio: 0.4~0.5. Experimental results show the syngas composition (CO+H$_2$): 85~93%, syngas flow rate: 50~l10 Nm$^3$/hr, heating value: 2,300~3,000 k㎈/Nm$^3$, carbon conversion: 65~92, cold gas efficiency: 60~70%. Also equilibrium modeling was used to predict the vacuum residue gasification process and the predicted values were compared reasonably well with experimental data.

• Journal of Energy Engineering
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• v.24 no.3
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• pp.96-108
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• 2015

A passive containment cooling system has been designed to remove the heat inside a containment during accidents without external power supply. In this work, the PCCS was introduced in the APR1400 plant to replace the containment spray system and, then, the thermal-hydraulic performance of the PCCS was analyzed using the system thermal-hydraulic computer code, MARS. A double-ended cold-leg break accident, which is known to induce the maximum pressure in the containment, is simulated, where the thermal hydraulics of the PCCS, the reactor coolant system, and the containment are simultaneously simulated. The results of the calculations showed that the PCCS can replace the existing spray system and that the containment building and its internal structure also play a very important role for the heat removal during the accident. Some sensitivity calculations were carried out to evaluate the model uncertainty and the effects of design parameters. The limitations of the PCCS are also discussed.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.283-296
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• 2008

The numerical simulations for groundwater flow were carried out to support the input parameters for safety assessment in LILW repository site. As the input parameters for safety assessment, the groundwater flux into the underground facilities during construction, flow rate through the disposal silo after closure of disposal silo and flow pathway from the disposal silo to discharge area were analyzed using the 10 cases groundwater flow simulations. From the total 10 numerical simulation results, the statistics of estimated output were similar to among 10 cases. In some cases, the analyzed input parameters were strongly governed by locally existed high permeable fracture zone at radioactive waste disposed depth. Indeed, numerical simulation for well scenario as a human intrusion scenario was carried out using the hydraulically severe case model. Using the results of well scenario, the input parameters for safety assessment were also obtained through the numerical simulation.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.6 no.4
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• pp.265-282
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• 2008

Based on the site characterization works in a low and intermediate level waste(LILW) repository site, the numerical simulations for groundwater flow were carried out in order to understand the groundwater flow system of repository site. To accomplish the groundwater flow modeling in the repository site, the discrete fracture network(DFN) model was constructed using the characteristics of fracture zones and background fractures. At result, the total 10 different hydraulic conductivity(K) fields were obtained from DFN model stochastically and K distributions of constructed mesh were inputted into the 10 cases of groundwater flow simulations in FEFLOW. From the total 10 numerical simulation results, the simulated groundwater levels were strongly governed by topography and the groundwater fluxes were governed by locally existed high permeable fracture zones in repository depth. Especially, the groundwater table was predicted to have several tens meters below the groundwater table compared with the undisturbed condition around disposal silo after construction of underground facilities. After closure of disposal facilities, the groundwater level would be almost recovered within 1 year and have a tendency to keep a steady state of groundwater level in 2 year.

• Tunnel and Underground Space
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• v.31 no.5
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• pp.333-359
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• 2021

Gases such as hydrogen and radon can be generated around the canister in high-level radioactive waste disposal systems due to several reasons including the corrosion of metal materials. When the gas generation rate exceeds the gas diffusion rate in the low-permeability bentonite buffer, the gas phase will form and accumulate in the engineered barrier system. If the gas pressure exceeds the gas entry pressure, gas can migrate into the bentonite buffer, resulting in pathway dilation flow and advective flow. Because a sudden occurrence of dilation flow can cause radionuclide leakage out of the engineered barrier of the radioactive waste disposal system, it is necessary to understand the gas migration behavior in the bentonite buffer to quantitatively evaluate the long-term safety of the engineered barrier. Experimental research investigating the characteristics of gas migration in saturated bentonite and research developing numerical models capable of simulating such behaviors are being actively conducted worldwide. In this technical note, previous gas injection experiments and the numerical models proposed to verify such behaviors are introduced, and the future challenges necessary for the investigation of gas migration are summarized.

• Tunnel and Underground Space
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• v.30 no.4
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• pp.359-381
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• 2020

Within the framework of DECOVALEX-2019 Task D, full-scale engineered barriers experiment (FEBEX) at Grimsel Test Site was numerically simulated to investigate an applicability of implemented Barcelona basic model (BBM) into TOUGH2-MP/FLAC3D simulator, which was developed for the prediction of the coupled thermo-hydro-mechanical behavior of bentonite buffer. And the calculated heater power, temperature, relative humidity, total stress, saturation, water content and dry density were compared with in situ data monitored in the various sections. In general, the calculated heater power and temperature provided a fairly good agreement with experimental observations, however, the difference between power of heater #1 and that of heater #2 could not captured in the numerical analysis. It is necessary to consider lamprophyre with low thermal conductivity around heater #1 and non-simplified installation progresses of bentonite blocks in the tunnel for better modeling results. The evolutions and distributions of relative humidity were well reproduced, but hydraulic model needs to be modified because the re-saturation process was relatively fast near the heaters. In case of stress evolutions due to the thermal and hydraulic expansions, the computed stress was in good agreement with the data. But, the stress is slightly higher than the measured in situ data at the early stage of the operation, because gap between rock mass and bentonite blocks have not been considered in the numerical simulations. The calculated distribution of saturation, water content, and dry density along the radial distance showed good agreement with the observations after the first and final dismantling. The calculated dry density near the center of the FEBEX tunnel and heaters were overestimated compared with the observations. As a result, the saturation and water content were underestimated with the measurements. Therefore, numerical model of permeability is needed to modify for the production of better numerical results. It will be possible to produce the better analysis results and more realistically predict the coupled THM behavior in the bentonite blocks by performing the additional studies and modifying the numerical model based on the results of this study.

• Journal of Energy Engineering
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• v.23 no.1
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• pp.90-94
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• 2014

Environmental qualification (EQ) for safety-related equipment is required to ensure that those equipment will perform their required function even under the harsh environment conditions arising from design basis accident in the nuclear power plant. As a part of EQ program, the room temperature analysis in case of a loss of Heating, Ventilation, and Air Conditioning(HVAC) system was carried out to ensure the operability of the safety-related equipment of a nuclear power plant randomly chosen among the Korean nuclear power plants. In this paper, this analysis was performed in the conservative perspective using GOTHIC code. The room temperature analysis includes selecting the rooms in which the safety related equipment are located but not supported by safety related HVAC and determining the temperature of the selected rooms. Target rooms for the analysis consist of W229/W237 (Aux. feedwater pump room), W232 (Aux. feedwater tank room) and W230 (Equipment passageway). The results showed the temperature range from $43^{\circ}C$ to $83^{\circ}C$, in 72 hours after a loss of HVAC. Those values are far below of generic EQ temperature($171^{\circ}C$). Therefore, it is satisfied with EQ requirement of temperature limits on safety related equipment.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.1
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• pp.49-56
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• 2005

Many different kinds of radwastes are discharged from the nuclear power plants, and $^{129}$I is included in these radwastes. Recovery test of $^{129}$I was evaluated for different radwastes(dry active waste, sludge, spent resin and simulated evaporator bottom). Recovery of $^{129}$I for dry active waste by acid leaching with $1.8\%$ NaClO was $74.3\%$$(RSD,\;2.2\%)$ and l291 for spent rein by alkali fusion method with KOH as a flux agent was $87.7\%$$(RSD,\;0.9\%$), respectively. iodide in simulated evaporator bottom containing a high concentration of borate was adsorbed with anion exchange resin at pH 7 phosphate buffer solution. Recovery of $^{129}$I for anion exchange resin was $92.5\%$ and not affected up to 1,200 $\mu$g/mL $H_3BO)3$(as a Boron). Recovery of $^{129}$I for the spent resin from nuclear power plant was $87.2\%$ $(RSD,\;1.2\%)$.

• Transactions of the Korean Society of Automotive Engineers
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• v.17 no.2
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• pp.98-103
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• 2009

The environment and energy related problem has become one of the most important global issues in recent years. One of the most effective ways of improving the fuel efficiency of automobiles is the weight reduction. In order to obtain this goal the hydroforming technology has been adapting for the high strength steel and its application is being widened. In present study, the chassis components (mainly cross members of engine cradle) simulation and development by hydroforming technology to apply high strength steel having tensile strength of 440 MPa grade is studied. In the part design stage, it requires feasibility study and process design aided by CAE (Computer Aided Design) to confirm hydroformability in details. Overall possibility of hydroformable chassis parts could be examined by cross sectional analyses. Moreover, it is essential to ensure the formability of tube material on every forming step such as pre-bending, performing and hydroforming. In the die design stage, all the components of prototyping tool were designed and interference with press was investigated from the point of geometry and thinning.

• Nuclear Engineering and Technology
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• v.41 no.5
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• pp.691-708
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• 2009

In order to analyze thermal hydraulic phenomena during a DVI (Direct Vessel Injection) line break LOCA (Loss-of-Coolant Accident) in the APR1400 (Advanced Power Reactor 1400 MWe), we performed experimental studies with the SNUF (Seoul National University Facility), a reduced-height and reduce-pressure integral test loop with a scaled down APR1400. We performed experiments dealing with eight test cases under varied tests. As a result of the experiment, the primary system pressure, the coolant temperature, and the occurrence time of the downcomer seal clearing were affected significantly by the thermal power in the core and the SI flow rate. The break area played a dominant role in the vent of the steam. For our analytical investigation, we used the MARS code for simulation of the experiments to validate the calculation capability of the code. The results of the analysis showed good and sufficient agreement with the results of the experiment. However, the analysis revealed a weak capability in predicting the bypass flow of the SI water toward the broken DVI line, and it was insufficient to simulate the streamline contraction in the broken side. We, hence, need to improve the MARS code.