• Transactions of the Korean Society of Mechanical Engineers A
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• v.39 no.3
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• pp.249-257
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• 2015

A fuel assembly consists of fuel rods composed of pellets (UO2) and a cladding tube (Zircaloy). The role of the fuel rods in the reactor is to generate heat by nuclear fission, as well as to retain fission products during operation. A simulation method using a computer program was used to evaluate the safety of the nuclear fuel rods. This computer program has been called the fuel performance code. In the analysis of a light water reactor fuel rod, the gap conductance, which depended on the distance between the pellets and cladding tube, mainly influenced the thermomechanical behavior of the fuel rod. In this work, a 3D gap element was proposed to simulate the thermo-mechanical behavior of the nuclear fuel rod, considering the gap conductance. To implement the proposed 3D gap element, a 3D thermo-mechanical module was also developed using FORTRAN90. The asymmetric characteristics of the nuclear fuel rod, such as the MPS (missing pellet surface) and eccentricity, were simulated to evaluate the proposed 3D gap element.

• The Journal of Engineering Geology
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• v.19 no.2
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• pp.139-144
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• 2009

The microseismic monitoring system with wide range of frequency has been operating in real time and it is remotely monitored at indoor and on-site for one year. This system was constructed and established in order to secure the safe and effective operation of the KAERI Underground Research Tunnel(KURT). For one year monitoring work, total 14 events were recorded in the vicinity of the KURT, and the majority of events are regarded as ultramicroseismic earthquake and artificial impacts around the tunnel. The major event is the magnitude 3.4 earthquake which was centered around Gongju city, Chungnam Province. It means that there is no significant evidence of high frequency microseismic event, which is associated with fracture initiation and/or propagation in the rock mass and shotcrete. Three components sensor was applied in order to analyze and define the direction of vibration as well as an epicenter of microseismic origin, and also properly designed and installed in a small borehole. This monitoring system is able to predict the location and timing of fracturing of rock mass and rock fall around an undreground openings as well as analysis on safety of various kinds of engineering structures such as nuclear facilities and other structures.

• Analytical Science and Technology
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• v.9 no.1
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• pp.35-42
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• 1996

A rapid analytical method based on capillary electrophoresis is described for the determination of trace anions in high-purity chemicals which is used to prevent corrosion demage in nuclear power plants. Separations are carried out at 20kV using trimethylsilane-coated fused-silica capillary ($70cm{\times}50$ or $75{\mu}m$ i.d.) with the electrolyte of 5mM Chromate(pH=8). Detection was achieved using on-column indirect photometry at 254nm. The simultaneous analysis of inorganic anions, chloride, nitrate, sulfate, azide and phosphate was performed using methods of hydrodynamic(>1ppm) and / or electrokinetic(<1ppm) injection. The results of studies on the coexisting anions on analyte ions shows that peak responses of analyte in hydrodynamic injection is constant without effect of coexisting anions, but those of analysis in electrokinetic injection is strongly dependant upon the kind of coexisting anions and its ionic mobility. The analyte enrichment rate, hence peak response, is positive relationship with the resistance of the sample solution. Thus, appropriate measures, such as standard addition or internal standard technique, must be used to account for differences in conductance of standard and sample solutions.

• Nuclear Engineering and Technology
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• v.22 no.2
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• pp.101-107
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• 1990

The PIXE (Proton Induced X-ray Emission) method is applied to the quantitative analysis of trace elements in tap water, red wine, urine and old black powder samples. Sample irradiations are performed with a 1.202 MeV proton beam from the SNU 1.5-MV Tandem Van de Graaff accelerator, and measurements of X-ray spectra are made by the Si(Li) spectrometer To increase the sensitivity of analysis tap water is preconcentrated by evaporation method. As an internal standard, Ni powder is mixed with black powder sample and yttrium solution is added to the other samples. The analyses of the PIXE spectra are carried out by using the AXIL (Analytical X-ray Analysis by Iterative Least-squares) computer code, in which the routine for least-squares method is based on the Marquardt algorithm. The elements such as Mg, Al, Si, Ti, Fe and Zn are analyzed at sub-ppm levels in the tap water sample. In the red wine sample prepared without preconcentration. the element Ti is detected in the amount of 3ppm. In conclusion, the PIXE method is proved to be appropriate for the analysis of liquid samples by relative measurements using the internal standard. and is expected to be improved by the use of evaluated X-ray production cross-sections and the development of sample preparation techniques.

• Nuclear Engineering and Technology
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• v.24 no.2
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• pp.154-162
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• 1992

Two major goals for chemical cleaning on the secondary side of nuclear steam generators are to remove sludge effectively and to minimize corrosion of base metals. In this work, electrochemical and sludge dissolution behaviors have been investigated in order to find out which parameters are critical and important during a copper removal process for chemical cleaning and to evaluate safety aspects and effectiveness of two major copper removal processes developed commercially in foreign countries. Hydrogen peroxide is vert effective for the process to use EDTA, NH$_4$OH and EDA at 38$^{\circ}C$ to control the potential of copper in a potential range sood for copper sludge removal. Corrosion rates for carbon steel SA 285 Gr.C and Alloy 600 are very small during this process if it is controlled properly. However, the corrosion rate of SA 285 Gr.C will be increased greatly if its corrosion potential is maintained below -450mV. The process to use EDA and ammonium carbonate is effective at 6$0^{\circ}C$ to dissolve copper sludge if the corrosion potential of copper can be controlled above -200mV. However, it is very difficult to raise the corrosion potential of copper to this range by air blowing and stirring.

• Journal of the Korea Concrete Institute
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• v.28 no.5
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• pp.611-620
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• 2016

In the construction of nuclear power plants, only 420 MPa reinforcing bars are allowed and, therefore, so many large-diameter bars are placed, which results in steel congestion. Consequently, re-bar works are difficult and the quality of RC structures may be deteriorated. To solve the steel congestion, 550 MPa bars are necessary. Among many items for verifying structural performance of reinforced concrete with 550 MPa bars, the 43 mm hooked bars are examined in this study. All specimens failed by side-face blowout and the side cover explosively spalled at maximum loads. The bar force was initially transferred to the concrete primarily by bond along a straight portion. At the one third of maximum load, the bond reached a peak capacity and began to decline, while the hook bearing component rose rapidly. At failure, most load was resisted by the hook bearing. For confined specimens with hoops, the average value of test-to-prediction ratios by KCI code is 1.45. The modification factor of confining reinforcement which was not allowed for larger than 35 mm bars can be applied to 43 mm hooked bars. For specimens with 70 MPa concrete, the average value of test-to-prediction ratios by KCI code is 1.0 which is less than the values of the other specimens. The effects of concrete compressive strength should be reduced. An equation to predict anchorage capacity of hooked bars was developed from regression analysis including the effects of compressive strength of concrete, embedment length, side cover thickness, and transverse reinforcement index.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.3 no.3
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• pp.159-165
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• 2005

The characterization of radiological contamination inside pipes generated during the decommission of a nuclear facility is necessary before pipes can be recycled or disposed. But, existing direct measurements of radioactive contamination level using the survey-meter can not estimate the characteristic of contamination on a local area such as the pipe inside. Moreover, the measurement of surface contamination level using the indirect methods has many problems of an application because of the difficulty of collecting sample and contamination possibility of a worker when collecting sample. In this work, plastic scintillator was simulated by using Monte Carlo simulation method for detection of beta radiation emitted from internal surfaces of small diameter pipe. Simulation results predicted the optimum thickness and geometry of plastic scintillator at which energy absorption for beta radiation was maximized. In addition, the problem of scintillator processing and transferring the detector into the pipe inside was considered when fabricating the plastic detector on the basis of simulation results. The characteristic of detector fabricated was also estimated. As a result, it was confirmed that detector capability was suitable for the measurement of contamination level. Also, the development of a detector for estimating the radiological characteristic of contamination on a local area such as the pipe inside was proven to be feasible.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.4 no.2
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• pp.179-186
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• 2006

Chemical wastes are generated from nuclear facilities and R&D laboratories, but the uranium concentration in the final dried cake is evaluated into 11.2 Bq/g, which exceeds the exemption level of 10 Bq/g for each U isotopes, so the cake is categorized into a radioactive waste. Acid dissolution was applied to extract uranium from the waste sludge, and uranium adsorption on the dissolved solution was experimented by using IRN-77 and Diphosil bead. A large amount of resin was required to get above 80% of uranium removal, which was found to be due to a large amount of metal ions simultaneously dissolved from the precipitates with uranium. As an alternative method, acid dissolution is applied to the dewatered wet cake of the sludge, and the natural evaporation method is adopted for the dissolved solution. The uranium concentration of the dissolved solution was estimated to be 6.97E-01 Bq/ml, and the specific activity of the final waste sheets is evaluated to be 4.3 Bq/g. These results lead to the suggestion that the application of acid dissolution to the wet cake and the natural evaporation for the dissolved solution is an effective treatment method for chemical wastes containing uranium.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.7 no.4
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• pp.229-236
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• 2009

There are two types of spent fuels generated from nuclear power plants, CANDU type and PWR type. PWR spent fuels which include a lot of reusable material can be considered to be recycled. CANDU spent fuels are considered to directly disposed in deep geological formation, since they have little reusable material. In this study, based on the Korean Reference spent fuel disposal System(KRS) which is to dispose both PWR and CANDU spent fuels, the more effective CANDU spent fuel disposal systems have been developed. To do this, the disposal canister has been modified to hold the storage basket which can load 60 spent fuel bundles. From these modified disposal canisters, the disposal systems to meet the thermal requirement for which the temperature of the buffer materials should not be over $100^{\circ}C$ have been proposed. These new disposals have made it possible to introduce the concept of long tenn storage and retrievabililty and that of the two-layered disposal canister emplacement in one disposal hole. These disposal concepts have been compared and analyzed with the KRS CANDU spent fuel disposal system in terms of disposal effectiveness. New CANDU spent fuel disposal concepts obtained in this study seem to improve thermal effectiveness, U-density, disposal area, excavation volume, and closure material volume up to 30 - 40 %.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.34 no.8
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• pp.1105-1111
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• 2010

The Korea Atomic Energy Research Institutes has been developing the SMART (System integrated Modular Advanced ReacTor), an environment-friendly nuclear reactor for the generation of electricity and to perform desalination. SMART reactors can be exposed to various external and internal loads caused by seismic and coolant flows. The CRDM(control rod drive mechanism), one of structures of the SMART, is a component which is adjusting inserting amount of a control rod, controlling output of reactor power and in an emergency situation, inserting a control rod to stop the reactor. The purpose of this research is performing the analysis of dynamic characteristic to ensure safety and integrity of structure of CRDM. This paper presents two FE-models, 3-D solid model and simplified Beam model of the CRDM in the coolant, and then compared the results of the dynamic characteristic about the two FE-models using a commercial Finite Element tool, ABAQUS CAE V6.8 and ANSYS V12. Beam 4 and beam 188 of simplified-model were also compared each other. And simplified model is updated for accuracy compare to 3-D solid.