• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(2)
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• pp.680-686
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• 1995

It has been known that water-side corrosion of fuel rods in nuclear reactor is accompanied with the loss of metallic wall thickness and pickup of hydrogen. This corrosion is one of the important limiting factors ill the operating life of fuel rods. In connection with the fuel cladding corrosion, a device to measure the water-side oxide layer thickness by means of the eddy-current method without destructing the fuel rod was developed by KAERI. The device was installed on the multi-function testing bench in the nondestructive test hot-cell and its calibration was carried out successfully for the standard rod attached with plastic thin films whose thicknesses are predetermined. It shows good precision within about 10% error. And a PWR fuel rod, one of the J-44 assembly discharged from Kori nuclear power plant Unit-2, has been selected for oxide layer thickness measurements. With the result of data analysis, it appeared that the oxide layer thicknesses of Zircaloy cladding vary with the length of the fuel rod, and their thicknesses were compared with those of the destructive test results to confirm the real thicknesses.

• Nuclear Engineering and Technology
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• 제42권5호
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• pp.590-599
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• 2010

Recently, efficient operation and practical management of power plants have become important issues in the nuclear industry. In particular, typical aging parameters such as stress and cumulative usage factor should be determined accurately for continued operation of a nuclear power plant beyond design life. However, most of the major components have been designed via conservative codes based on a 2-D concept, which do not take into account exact boundary conditions and asymmetric geometries. The present paper aims to suggest an effective fatigue evaluation methodology that uses a prototype of the integrated model and its transfer functions. The validity of the integrated 3-D Finite Element (FE) model was proven by comparing the analysis results of individual FE models. Also, mechanical and thermal transfer functions, known as Green's functions, were developed for the integrated model with the standard step input. Finally, the stresses estimated from the transfer functions were compared with those obtained from detailed 3-D FE analyses results at critical locations of the major components. The usefulness of the proposed fatigue evaluation methodology can be maximized by combining it with an on-line monitoring system, and this combination, will enhance the continued operations of old nuclear power plants.

• Nuclear Engineering and Technology
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• 제51권8호
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• pp.1939-1950
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• 2019

The MELCOR code useful for a plant-specific hydrogen risk analysis has inevitable limitations in prediction of a turbulent flow of a hydrogen mixture. To investigate the accuracy of the hydrogen risk analysis by the MELCOR code, results for the turbulent gas behavior at pipe rupture accident were compared with CFX results which were verified by the American National Standard Institute (ANSI) model. The postulated accident scenario was selected to be surge line failure induced by station blackout of an Optimized Power Reactor 1000 MWe (OPR1000). When the surge line failure occurred, the flow out of the surgeline was strongly turbulent, from which the MELCOR code predicted that a substantial amount of hydrogen could be released. Nevertheless, the results indicated nonflammable mixtures owing to the high steam concentration released before the failure. On the other hand, the CFX code solving the three-dimensional fluid dynamics by incorporating the turbulence closure model predicted that the flammable area continuously existed at the jet interface even in the rising hydrogen mixtures. In conclusion, this study confirmed that the MELCOR code, which has limitations in turbulence analysis, could underestimate the existence of local combustible gas at pipe rupture accident. This clear comparison between two codes can contribute to establishing a guideline for computational hydrogen risk analysis.

• Nuclear Engineering and Technology
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• 제52권11호
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• pp.2667-2677
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• 2020

A domain decomposition (DD) scheme for GPU-based Monte Carlo (MC) calculation which is essential for whole-core depletion is introduced within the framework of the modified history-based tracking algorithm. Since GPU-offloaded MC calculations suffer from limited memory capacity, employing DDMC is inevitable for the simulation of depleted cores which require large storage to save hundreds of newly generated isotopes. First, an automated domain decomposition algorithm named wheel clustering is devised such that each subdomain contains nearly the same number of fuel assemblies. Second, an innerouter iteration algorithm allowing overlapped computation and communication is introduced which enables boundary neutron transactions during the tracking of interior neutrons. Third, a bank update scheme which is to include the boundary sources in a way to be adequate to the peculiar data structures of the GPU-based neutron tracking algorithm is presented. The verification and demonstration of the DDMC method are done for 3D full-core problems: APR1400 fresh core and a mock-up depleted core. It is confirmed that the DDMC method performs comparably with the standard MC method, and that the domain decomposition scheme is essential to carry out full 3D MC depletion calculations with limited GPU memory capacities.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 추계학술발표회논문집(1)
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• pp.528-533
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• 1997

A CANFLEX(CANdu FLEXible fuelling) 43-element bundle has developed for a CANDU-6 reactor as an alternative of 37-element fuel bundle. The design has two diameter elements (11.5 and 13.5㎜) to reduce maximum element power rating and buttons to enhance the critical heat flux(CHF), compared with the standard 37-element bundle. The freon CHF experiments have performed for two series of CANFLEX bundles with and without buttons with a modelling fluid as refrigerant H-l34a and axial uniform heat flux condition. Evaluating the effects of buttons of CANFLEX bundle on CHF and Critical Channel Power(CCP) with the experimental results, it is shown that the buttons enhance CCP as well as CHF. All the CHF's for both the CANFLEX bundles are occurred at the end of fuel channel with the high dryout quality conditions. The CHF enhancement ratio are increased with increase of dryout quality for all flow conditions and also with increase of mass flux only lot high pressure conditions. It indicates that the button is a useful design lot CANDU operating condition because most CHF flow conditions for CANDU fuel bundle are ranged to high dryout quality conditions.

• 한국물환경학회지
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• 제22권1호
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• pp.30-36
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• 2006

This study was carried out to evaluate the performance of Modified Ludzsck Etinger (MLE) process with waste oyster shell media in aerobic tank. Influent flow was 36 L/d and the order of reactor was anoxic, aerobic and sedimentation tank and unit hydraulic retention time was 2 hr, 6 hr and 4 hr, respectively. Sludge recycling rate in sedimentation tank and internal recycling rate were 100%. Media fill rate in aerobic tank was 5%, 10% and 17% and fluid MLSS concentration in aerobic tank was 3000~4000 mg/L. Average TCOD removal rate was 91~93%, TBOD 92~96%, SS 95~96% and when media fill rate was 10% or more, in organic compound removal it could satisfy with wastewater discharge standard. Average total nitrogen removal rate was 70~76% and average total phosphorous removal rate was 58~65%. With media fill rate increasing, total phosphorous average removal rate also increased. For it was that released calcium ion from waste oyster shell reacted with soluble phosphorous. From these experiment results, the MLE process using waste oyster shell as media is a practical method for advanced sewage treatment in rural area.

• 전기학회논문지
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• 제58권2호
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• pp.232-238
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• 2009

Flicker, also known as voltage fluctuation, is a newest problem of power quality issues, because it is caused by nonlinear loads such as electrical arc furnace and large-scale induction motor, which are country-widely used as the heavy industries of a country develop. An international standard, International Electrotechnical Commission (IEC) 61000-4-15, was published in 1997 and revised in 2003. With increasing concerns about flicker, its mitigation methods have been also studied. General countermeasures for flicker are divided into three categories: a) enhancing the capacity of supplying system, b) Series elements including series reactor and series capacitor and c) power electronic devices including static VAR compensator (SVC) and static synchronous compensator (STATCOM). This paper introduces how to mitigate the voltage flicker at the point of common coupling (PCC) and presents how to simulate and compare the flicker alleviating effects by each mitigation method, using IEC flickermeter based on the Matlab/Simulink program.

• Korean Chemical Engineering Research
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• 제50권2호
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• pp.223-228
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• 2012

ABS와 폴리에틸렌(Polyethylene, PE) 및 ABS-PE 혼합물의 저온열분해를 회분식 반응기를 이용하여 상압 및 $450^{\circ}C$에서 실행하였다. 열분해 시간은 20~80분까지 하였고 열분해로 생성된 성분은 지식경제부에서 고시한 증류성상온도에 따라 가스, 가솔린, 등유, 경유, 중유로 분류하였다. ABS와 PE의 혼합 폐플라스틱의 열분해 전환율은 PE의 함량이 증가할수록 증가하는 것으로 나타났다. 열분해생성물의 수율은 PE의 함량이 높을수록 중유 > 가스 > 가솔린 > 경유 > 등유 순으로 회수되었다.

• 한국전기전자재료학회:학술대회논문집
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• 한국전기전자재료학회 1999년도 춘계학술대회 논문집
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• pp.434-437
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• 1999

An EM pump is used for the purpose of transporting the electrically conducting liquid sodium of the high temperature that is used as a coolant in the liquid metal reactor. In the present study, the pilot pump has been designed and manufactured for the high temperature of $600^{\circ}C$ by the equivalent circuit materials and the consideration of the materials and functions. The length and diameter of the pump are given as 84 cm and 10 cm each due to the fixed geometry of the circulation system to be installed. The characteristic of the developing pressure and efficiency is found out by using Laithewaite\`s standard design formula. It is shown that the developing pressure and efficiency are maximized at the frequency of 15 Hz from the curve. The annular channel gap of 3.95 mm is selected in the range of the reasonable hydraulic frictional loss. The components of the pump consist of the material for the high temperature. And then, the pump is manufactured to have the nominal flowrate of 40 1/min and developing Pressure of 1.3 bar.

• Nuclear Engineering and Technology
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• 제36권6호
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• pp.559-570
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• 2004

A computational fluid dynamics model for predicting moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the calandria tubes. The buoyancy effect induced by the internal heating is accounted for by the Boussinesq approximation. The standard $k-{\varepsilon}$ turbulence model with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the calandria tubes in the core region is simplified to a porous media in which the anisotropic hydraulic impedance is modeled using an empirical correlation of pressure loss. The governing equations are solved by DFX-4.4, a commercial CFD code developed by AEA technology. The resultant flow patterns of the constant-z slices containing the inlet nozzles and the outlet port are "mined-type", as observed in the former 2-dimensional experimental investigations. With 103% full power for conservatism, the maximum temperature of the moderator is $82.9^{\circ}C$ at the top of the core region. Considering the hydrostatic pressure change, the minimum subcooling is $24.8^{\circ}C$.