• Journal of Korean Society for Atmospheric Environment
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• v.26 no.1
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• pp.85-93
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• 2010

$CF_4$ removal characteristics were investigated using an elongated arc reactor. The advantage of the elongated arc reactor includes direct use of treated gas as plasma operating gas and the enhancement of the removal reaction by using a thermo-chemistry and a plasma induced chemistry at the same time. Geometrical configurations, such as the length of the reactor and the shape of a throat, were tested to get an optimized removal efficiency with low power consumption. As results, over 95% of $CF_4$ removal was obtained with 300 lpm of total flowrate for various $CF_4$ concentration (0.1~1%). Corresponding specific energy density (SED), which means required electrical energy to treat the unit volume of treated gas, is about 3.5 kJ/L, The present technique can be applied to real applications by satisfying three major concerns, those are the high flowrate of treated gas, high removal efficiency (> 95%), and low power consumption (< 10 kJ/L).

• Transactions of the Korean Society for Noise and Vibration Engineering
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• v.26 no.5
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• pp.552-558
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• 2016

A control rod drive mechanism(CRDM) is a reactor regulating system, which inserts, withdraws or maintains a control rod within a reactor core to control the reactivity of the core. The CRDM for Jordan Research and Training Reactor with 5MW power has been designed and fabricated based on the HANARO’s experience through KAERI and DAEWOO consortium. This paper describes the seismic test results to demonstrate the operability, the drop performance and the structural integrity of CRDM during or after seismic excitations. The seismic tests are carried out under 5 OBE and 1 SSE loads at three Test Rigs simulating the reactor structure and the pool top. From the tests, the CRDM is smoothly driven without a malfunction of stepping motor under OBE load. The pure drop time under OBE and SSE loads is measured as 1.169s and 1.855s to meet the design requirement. Also, it is found that the CRDM maintains the structural integrity without a change of the function and natural frequency before and after seismic loads.

• Proceedings of the KIPE Conference
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• 2002.07a
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• pp.711-714
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• 2002

Thermal infrared camera have poor image qualities compared to commercial CCD cameras, as in contrast, brightness, and. resolution. To compensate the poor Image quality problems associated with the thermal infrared camera, the technique of superimposing thermal infrared image into real ccd image is proposed. The mobile robot KAEROT/m2, loaded with sensor head system at the mast, is entered to monitor leakage of heavy water and thermal abnormality of the calandria reactor area in overhaul period. The sensor head system is composed of thermal infrared camera and cod camera In parallel. When thermal abnormality on observation points and areas of calandria reactor area is occurred, unusual hot image taken from thermal infrared camera is superimposed on real CCD image. In this inspection experiment, more accurate positions of thermal abnormalities on calandria reactor area can be estimated by using technique of mapping thermal infrared image into CCD image, which include characters arranged in MPOQ order.

• Structural Engineering and Mechanics
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• v.59 no.2
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• pp.291-312
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• 2016

Seismic performance evaluation of shear wall is essential as it is the major lateral load resisting member of a structure. The ultimate load and ultimate drift of the shear wall are the two most important parameters which need to be assessed experimentally and verified analytically. This paper comprises the results of monotonic tests, quasi-static cyclic tests and shake-table tests carried out on a midrise shear wall. The shear wall considered for the study is 1:5 scaled model of the shear wall of the internal structure of a reactor building. The analytical simulation of these tests is carried out using micro and macro modeling of the shear wall. This paper mainly consists of modification in the hysteretic macro model, developed for RC structural walls by Lestuzzi and Badoux in 2003. This modification is made by considering the stiffness degradation effect observed from the tests carried out and this modified model is then used for nonlinear dynamic analysis of the shear wall. The outcome of the paper gives the variation of the capacity, the failure patterns and the performance levels of the shear walls in all three types of tests. The change in the stiffness and the damping of the wall due to increased damage and cracking when subjected to seismic excitation is also highlighted in the paper.

• Carbon letters
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• v.4 no.1
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• pp.36-39
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• 2003

Carbon and carbon-based materials are used in nuclear reactors and there has recently been growing interest to develop graphite and carbon based materials for high temperature nuclear and fusion reactors. Efforts are underway to develop high density carbon materials as well as amorphous isotropic carbon for the application in thermal reactors. There has been research on coated nuclear fuel for high temperature reactor and research and development on coated fuels are now focused on fuel particles with high endurance during normal lifetime of the reactor. Since graphite as a moderator as well as structural material in high temperature reactors is one of the most favored choices, it is now felt to develop high density isotropic graphite with suitable coating for safe application of carbon based materials even in oxidizing or water vapor environment. Carboncarbon composite materials compared to conventional graphite materials are now being looked into as the promising materials for the fusion reactor due their ability to have high thermal conductivity and high thermal shock resistance. This paper deals with the application of carbon materials on various nuclear reactors related issues and addresses the current need for focused research on novel carbon materials for future new generation nuclear reactors.

• Nuclear Engineering and Technology
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• v.39 no.4
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• pp.299-312
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• 2007

The program on thermal-hydraulic evaluation by testing and analysis (THETA) for the development and licensing of the new design features in the APR1400 (Advanced Power Reactor-1400) is briefly introduced with a presentation on the research motivation and typical results of the separate effect tests and analyses of the major design features. The first part deals with multi-dimensional phenomena related to the safety analysis of the APR1400. One research area is related to the multidimensional behavior of the safety injection (SI) water in a reactor pressure vessel downcomer that uses a direct vessel injection type of SI system. The other area is associated with the condensation of steam jets and the resultant thermal mixing in a water pool; these phenomena are relevant to the depressurization of a reactor coolant system (RCS). The second part describes our efforts to develop new components for safety enhancements, such as a fluidic device as a passive SI flow controller and a sparger to depressurize the RCS. This work contributes to an understanding of the new thermal-hydraulic phenomena that are relevant to advanced reactor system designs; it also improves the prediction capabilities of analysis tools for multi-dimensional flow behavior, especially in complicated geometries.

• Journal of computational fluids engineering
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• v.17 no.4
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• pp.41-48
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• 2012

A numerical analysis of thermal stratification in the upper plenum of the MONJU fast breeder reactor was performed. Calculations were performed for a 1/6 simplified model of the MONJU reactor using the commercial code, CFX-13. To better resolve the geometrically complex upper core structure of the MONJU reactor, the porous media approach was adopted for the simulation. First, a steady state solution was obtained and the transient solutions were then obtained for the turbine trip test conducted in December 1995. The time dependent inlet conditions for the mass flow rate and temperature were provided by JAEA. Good agreement with the experimental data was observed for steady state solution. The numerical solution of the transient analysis shows the formation of thermal stratification within the upper plenum of the reactor vessel during the turbine trip test. The temporal variations of temperature were predicted accurately by the present method in the initial rapid coastdown period (~300 seconds). However, transient numerical solutions show a faster thermal mixing than that observed in the experiment after the initial coastdown period. A nearly homogenization of the temperature field in the upper plenum is predicted after about 900 seconds, which is a much shorter-term thermal stratification than the experimental data indicates. This discrepancy is due to the shortcoming of the turbulence models available in the CFX-13 code for a natural convection flow with thermal stratification.

• Microbiology and Biotechnology Letters
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• v.9 no.3
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• pp.165-171
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• 1981

The penicillin amidase from Escherichia coli (ATCC 9637) was immobilized by entrappment in gelatin and DEAE-cellulose mixture cross-linked with glutaraldehyde, and the kinetics in a differential column reactor was studied. The optimal operating condition of a differential reactor was reasonably met when the enzyme loading was 1g, and 30 mM substrate solution in 0.1 M phosphate buffer (pH 8.0) was fed at flow rate 4$m\ell$/min and 4$0^{\circ}C$. The optimal pH and temperature were found to be 8.0 and 55$^{\circ}C$, respectively. The Michaelis-Menten constant was 4.8 mM while the maximum velocity was 308 units/g of the immobilized enzyme under the condition of the differential reactor. The effect of substrate inhibition disappeared in the immobilized enzyme preparation. The differential reactor was proved to be good for studying the true kinetics since the pH drop and the external diffusional resistance could be eliminated.

• Nuclear Engineering and Technology
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• v.40 no.7
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• pp.615-624
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• 2008

In this paper, the contribution of task types and error types involved in the human-related unplanned reactor trip events that have occurred between 1986 and 2006 in Korean nuclear power plants are analysed in order to establish a strategy for reducing the human-related unplanned reactor trips. Classification systems for the task types, error modes, and cognitive functions are developed or adopted from the currently available taxonomies, and the relevant information is extracted from the event reports or judged on the basis of an event description. According to the analyses from this study, the contributions of the task types are as follows: corrective maintenance (25.7%), planned maintenance (22.8%), planned operation (19.8%), periodic preventive maintenance (14.9%), response to a transient (9.9%), and design/manufacturing/installation (6.9%). According to the analysis of the error modes, error modes such as control failure (22.2%), wrong object (18.5%), omission (14.8%), wrong action (11.1 %), and inadequate (8.3%) take up about 75% of the total unplanned trip events. The analysis of the cognitive functions involved in the events indicated that the planning function had the highest contribution (46.7%) to the human actions leading to unplanned reactor trips. This analysis concludes that in order to significantly reduce human-induced or human-related unplanned reactor trips, an aide system (in support of maintenance personnel) for evaluating possible (negative) impacts of planned actions or erroneous actions as well as an appropriate human error prediction technique, should be developed.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3732-3753
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• 2023

The Real-time Analysis for Particle-transport and In-situ Detection (RAPID) Code System, developed based on the Multi-stage Response-function Transport (MRT) methodology, enables real-time simulation of nuclear systems such as reactor cores, spent nuclear fuel pools and casks, and sub-critical facilities. This paper presents the application of a novel fission matrix-based burnup methodology to the well-characterized JSI TRIGA Mark II research reactor. This methodology allows for calculation of nuclear fuel depletion by combination and interpolation of RAPID's burnup dependent fission matrix (FM) coefficients to take into account core changes due to burnup. The methodology is compared to experimentally validated Serpent-2 Monte Carlo depletion calculations. The results show that the burnup methodology for RAPID (bRAPID) implemented into RAPID is capable of accurately calculating the k_eff burnup changes of the reactor core as the average discrepancies throughout the whole burnup interval are 37 pcm. Furthermore, capability of accurately describing 3D fission source distribution changes with burnup is demonstrated by having less than 1% relative discrepancies compared to Serpent-2. Good agreement is observed for axially and pin-wise dependent fuel burnup and nuclear fuel nuclide composition as a function of burnup. It is demonstrated that bRAPID accurately describes burnup in areas with high gradients of neutron flux (e.g. vicinity of control rods). Observed discrepancies for some isotopes are explained by analyzing the neutron spectrum. This paper presents a powerful depletion calculation tool that is capable of characterization of spent nuclear fuel on the fly while the reactor is in operation.