• Korean Journal of Computational Design and Engineering
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• v.14 no.1
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• pp.1-9
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• 2009

ISO 15926 is an international standard for the representation of process plant lifecycle data. However, it is not easy to implement the part 2-data model and the part 4-initial reference data because of their complexity in terms of data structure and shortages of related development toolkits. To overcome this problem, ISO 15926-7(part 7) is under development. ISO 15926-7 specifies implementation methods for sharing and exchange of process plant lifecycle data, which is based on semantic web technologies such as OWL, Web Services, and SPARQL. For the application of ISO 15926-7, this paper discusses how to represent technical specifications of process plant equipment by defining user-defined reference data and object information model with an example of reactor coolant pumps located in the reactor coolant system of an APR 1400 nuclear power plant.

• Transactions of the Korean hydrogen and new energy society
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• v.29 no.5
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• pp.503-511
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• 2018

There have been many studies of combustion in the circulating fluidized bed. However, little study is available for combustion of wood pellet together fed with wood chip. The mixed ratio of two fuels is an useful information when thermal power company would receive the Renewable Energy Portfolio Standard (RPS) from government. In this study, the combustion behavior and kinetics of such biomass fuels are evaluated using fluidized bed reactor and thermogravimetric analyzers. The mixing ratio of wood chip relative to wood pellet was varied at different temperatures. The results show that a combustion reactivity changed significantly at the wood chip mixing ratio of 40%, particularly at low temperature condition.

• Journal of Korean Society of Water and Wastewater
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• v.11 no.3
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• pp.124-130
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• 1997

Wastewater containing phenol was treated using Pseudomonas sp. B3 in continuous reactor, reaction characteristics and kinetics according to variation of volumetric loading rate in continuous reactor were studied. The removal efficiencies of phenol were more than 99% at the whole range of experiment, and those of COD were 97% at the volumetric loading rate, $0.96kg/m^3{\cdot}d$ and 88% at $3.0kg/m^3{\cdot}d$, respectively. Kinetics constants of $q_m$, $K_s$, Y and $K_d$ were obtained 0.901 l/d, 0.620mg/l, 0.659 and 0.219 l/d, respectively. As compared with to constants of standard activated sludge process, these constants were remarkably different because of toxicity and inhibition of phenol to microbes. And also, kinetics constants of oxygen utilization, a, and b, were shown 0.384 kg $O_2/kg$ phenol and 0.029 l/d.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1024-1036
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• 2018

In this paper, McCARD code was verified using various models listed in the NUREG/CR-6361 benchmark guide, which provides specifications for single pin-cells, single assemblies, and the whole core classified depending on the nuclear properties and structural characteristics. McCARD code was verified by comparing its results with those of SCALE code for single pin-cell and single assembly benchmark problems. The difference in the multiplication factor obtained through the two codes did not exceed 90 pcm. The benchmark guide treats a total of 173 whole core experiments. The experiments are categorized as simple lattices, separator plates, reflecting walls, reflecting walls and separator plates, burnable absorber fuel rods, water holes, poison rods, and borated moderator. As a result of numerical simulation using McCARD, the mean value of the multiplication factors is 1.00223 and the standard deviation of the multiplication factors is 285 pcm. The difference between the multiplication factors and the experimental value is in the range of -665 pcm to + 1609 pcm. In addition, statistics of results for experiments categorized by reactor shape, additional structure, burnable poison, etc., are detailed in the main text.

• Journal of Environmental Health Sciences
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• v.22 no.2
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• pp.90-95
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• 1996

This study was performed to calculate the degration rate coefficient, operating parameters to meet the effluent standards, and the temperature adjustment coefficients to each parameter of pollution by seasonal variation of concentration and temperature of influent in livestock wastewater treatment by sequencing batch reactor process in field scale. The followings are the conclusions that were derived from this study. 1. In the field, temperature of livestock wastewater in reactor was 20.3$\circ$C in summer and 6.0$\circ$C in winter. The ratio of BOD:TKN: T-P in influent was 100:80:7. BOD loadings in winter and spring were 0.26 and 0.43 kg $BOD/m^3$ day, respectively. Those in summer and fall were 0.25 and 0.13 kg $BOD/m^3$ day, respectively. 2. The degradation rate coefficient for TKN was larger in summer and fall in which temperature was high than that in which temperature was high than that in winter and spring in which concentration was high. On the contrary, the phosphorus uptake rate was larger in winter and spring than that in summer and fall. 3. The hydraulic retention time in winter and spring was longer than that in summer and fall. Especially, in order to meet the standard for TKN of 120 mg/l in winter in which temperature of wastewater was 6.0$\circ$C, as the MLSS concentration was increased from 4, 000 to 7, 000 mg/l, the hydraulic retention time was increased from 212 to 121 hours. But, in order to shorten that less than 121 hours for the economical wastewater treatment, countermeasure to increase temperature of wastewater in the reactor should be considered. 4. the temperature adjustment coefficients for BOD, $COD_{Mn}$, TKN and T-P were 1.0241, 1.0225, 1.0541 and 1.0495, respectively. Namely, the treatment of TKN was most sensitively affected by temperature. For the purpose of the effective removal of nitrogen and phosphorus which are sensitive to temperature, it is necessary to keep the temperature of livestock wastewater more than 20$\circ$C which is the temperature of it in summer.

• Nuclear Engineering and Technology
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• v.44 no.7
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• pp.807-816
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• 2012

The Three Mile Island Unit 2 (TMI-2) accident has been studied extensively, as part of both post-accident technical assessment and follow-up computer code calculations. The models used in computer codes for severe accidents have improved significantly over the years due to better understanding. It was decided to reanalyze the severe accident scenario using current state of the art codes and methodologies. This reanalysis was adopted as a part of the joint standard problem exercise for the Atomic Energy Regulatory Board (AERB) - United States Regulatory Commission (USNRC) bilateral safety meet. The accident scenario was divided into four phases for analysis viz., Phase 1 covers from the accident initiation to the shutdown of the last Reactor Coolant Pumps (RCPs) (0 to 100 min), Phase 2 covers initial fuel heat up and core degradation (100 to 174 min), Phase 3 is the period of recovery of the core water level by operating the reactor coolant pump, and the core reheat that followed (174 to 200 min) and Phase 4 covers refilling of the core by high pressure injection (200 to 300 min). The base case analysis was carried out for all four phases. The majority of the predicted parameters are in good agreement with the observed data. However, some parameters have significant deviations compared to the observed data. These discrepancies have arisen from uncertainties in boundary conditions, such as makeup flow, flow during the RCP 2B transient (Phase 3), models used in the code, the adopted nodalisation schemes, etc. In view of this, uncertainty and sensitivity analyses are carried out using simulation based techniques. The paper deals with uncertainty and sensitivity analyses carried out for the first three phases of the accident scenario.

• Journal of the Korean Society for Nondestructive Testing
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• v.21 no.5
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• pp.561-564
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• 2001

The cladding area of HANARO Research Reactor fuel rods should be checked not to have any defects larger than the size required at QA documents by using eddy torrent testing method doting fabrication process. To apply eddy current testing inspection to the fuel rods, encircling differential probes and standard specimen were designed and fabricated. The impedance of the fabricated probes was measured with impedance analyzer in order to cheek that the probe has a suitable impedance for the inspection frequency, and with this probe and MIZ-40A eddy current equipment, the detectability of this probes was investigated. The developed probes could detect artificial notch with 2mm length 10% depth of cladding thickness in cladding area. In addition, the probe was successfully applied to detect the defects in cladding area doting fabrication of the research reactor rods.

• Journal of Radiation Protection and Research
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• v.14 no.1
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• pp.34-45
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• 1989

The habitability of a reactor control room in a French 1300 MWe P'4 type PWR has been evaluated through the exposure dose assessment for the reactor operator following a Loss of Coolant Accident. The main hypotheses adopted in this evaluation are based on the French Standard Safety Analysis Report. A simple computer program, named COREX(Control Room EXposure), was developed to calculate : the time-integrated radioactivities released from the reactor building, the volume factors for radionuclides concerned and the resulting time-integrated external whole body and internal thyroid doses to the reactor operators staying in the control room up to 30 days following the LOCA. The results obtained in this study, on the whole, well agreed with those proposed by the EDF(Electricite de France) except for the case of the whole body exposure, which was attributed to the differences in the volume factors for the radionuclides concerned.

• Korean Chemical Engineering Research
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• v.51 no.3
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• pp.297-302
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• 2013

Enhancement of photocatalytic activity of UV/photocatalysis was carried out to oxidize the gaseous $H_2S$ in a tubular reactor coated with photocatalyst of sol type $TiO_2$. EtOH was used as the standard material to select the photocatalyst, and it was confirmed that the reactor activity was dependent on the coated surface characteristics. The selected photocatalytic reactor, which coated with STS-01, showed about 80% conversion when the gas linear velocity was 0.01 m/s and relative humidity was 40%. However, the conversion level of the reaction decreased significantly with increasing gas linear velocity. Pt was loaded on the photocatalyst to enhance and maintain the performance of the reactor, which enhanced the conversion level of $H_2S$ more than 95% under the same experimental condition.

• Nuclear Engineering and Technology
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• v.54 no.1
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• pp.97-109
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• 2022

Multi-dimensional coupling analysis is a research hot spot in nuclear reactor thermal hydraulic study and both the full-scale system transient response and local key three-dimensional thermal hydraulic phenomenon could be obtained simultaneously, which can achieve the balance between efficiency and accuracy in the numerical simulation of nuclear reactor. A one-dimensional to three-dimensional (1D-3D) coupling platform for the nuclear reactor multi-dimensional analysis is developed by XJTU-NuTheL (Nuclear Thermal-hydraulic Laboratory at Xi'an Jiaotong University) based on the CFD code Fluent and system code RELAP5 through the Dynamic Link Library (DLL) technology and Fluent user-defined functions (UDF). In this paper, the International Standard Problem (ISP) No. 43 is selected as the benchmark and the rapid boron dilution transient in the nuclear reactor is studied with the coupling code. The code validation is conducted first and the numerical simulation results show good agreement with the experimental data. The three-dimensional flow and temperature fields in the downcomer are analyzed in detail during the transient scenarios. The strong reverse flow is observed beneath the inlet cold leg, causing the de-borated water slug to mainly diffuse in the circumferential direction. The deviations between the experimental data and the transients predicted by the coupling code are also discussed.