• Nuclear Engineering and Technology
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• 제36권6호
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• 제53권1호
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• pp.44-60
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• 2021

In this paper, we validate the decay heat calculation capability via a two-step method to analyze spent nuclear fuel (SNF) discharged from pressurized water reactors (PWRs). The calculation method is implemented with a lattice code STREAM and a nodal diffusion code RAST-K. One of the features of this method is the direct consideration of three-dimensional (3D) core simulation conditions with the advantage of a short simulation time. Other features include the prediction of the isotope inventory by Lagrange non-linear interpolation and the use of power history correction factors. The validation is performed with 58 decay heat measurements of 48 fuel assemblies (FAs) discharged from five PWRs operated in Sweden and the United States. These realistic benchmarks cover the discharge burnup range up to 51 GWd/MTU, 23.2 years of cooling time, and spanning an initial uranium enrichment range of 2.100-4.005 wt percent. The SNF analysis capability of STREAM is also employed in the code-to-code comparison. Compared to the measurements, the validation results of the FA calculation with RAST-K are within ±4%, and the pin-wise results are within ±4.3%. This paper successfully demonstrates that the developed decay heat calculation method can perform SNF back-end cycle analyses.

• Nuclear Engineering and Technology
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• 제55권6호
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• pp.2112-2124
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• 2023

With the rise of economic and safety standards for nuclear reactors, new concepts of Gen-IV reactors and modular reactors showed more complex designs that challenge current tools for reactor physics analysis. A Monte Carlo (MC) two-step method was proposed in this work. This calculation scheme uses the continuous-energy MC method to generate multi-group cross-sections from heterogeneous models. The multi-group MC method, which can adapt locally-heterogeneous models, is used in the core calculation step. This calculation scheme is verified using a Gen-IV modular lead-based fast reactor (LFR) benchmark case. The influence of homogenized patterns, scatter approximations, flux separable approximation, and local heterogeneity in core calculation on simulation results are investigated. Results showed that the cross-sections generated using the 3D assembly model with a locally heterogeneous representation of control rods lead to an accurate estimation with less than 270 pcm bias in core reactivity, 0.5% bias in control rod worth, and 1.5% bias on power distribution. The study verified the applicability of multi-group cross-sections generated with the MC method for LFR analysis. The study also proved the feasibility of multi-group MC in core calculation with local heterogeneity, which saves 85% time compared to the continuous-energy MC.

• 한국응용과학기술학회지
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• 제39권4호
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• pp.535-541
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• 2022

The radical reaction position was calculated by varying the calculation level for ACEL and ANT, which are detected with the highest frequency and concentration in PAHs pollution sites. The results of each calculation level were compared and evaluated with the existing literature. HF, B3LYP, B3LYP-D, and MP2 were used as the method for each level used for calculation. Except for HF, the MK charge by B3LYP, B3LYP-D, and MP2 was consistent with the experimental results. It was found that the dispersion effect was negligible in the calculation of ACEL and ANT because the calculation results by the B3LYP and B3LYP-D methods were the same. In particular, it was found that the MK charge calculation result by MP2 agrees well with the product/PAH ratio obtained as a result of the experiment. Considering the calculation cost, it would be preferable to use B3LYP to predict the radical reaction site of ACEL and ANT. However, considering the product/PAH ratio, it takes more time to calculate, but it is judged that it is better to use the MP2.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.19-29
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• 2022

The 2D/1D method has become the mainstream of the direct transport calculation considering the balance of accuracy and efficiency. However, the 2D/1D method still suffers from stability issues. Recently, a quasi-3D method has been proposed with axial Legendre expansion. Analysis and comparison of the 2D/1D and quasi-3D method is conducted in theory from the equation derivation. Besides, the C5G7 benchmark, the KUCA benchmark and the macro BEAVRS benchmark are calculated to verify the theory comparisons of these two methods with the direct transport code SHARK. All results show that the quasi-3D method has better stability and accuracy than the 2D/1D method with worse efficiency and memory cost. It provides a new option for direct transport calculation with the quasi-3D method.

• 한국정보통신학회논문지
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• 제13권10호
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• pp.2052-2058
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• 2009

디지털 시스템의 동작주파수가 증가하고 전압스윙폭이 감소함에 따라 PCB보드의 정확하고 빠른 해석이 중요하게 되었다. 단위 기둥 행열의 다중곱을 이용하는 전송선로 행열을 이용한 방법은 PCB보드 해석에 있어서 가장 빠른 방법이다. 본 논문에서 PCB보드 임피던스를 계산하는 새로운 방법이 제안되었다. 우선, 이 방법에서 PCB의 단위기둥에 대한 전송선로행열의 고유치와 고유벡터가 계산되고, 단위기둥에 대한 전송선로 행열은 행열요소의 곱셈횟수를 줄이기 위해 행열유사변환을 통해 변환된다. 이러한 유사변환을 방법은 기존방법에 비해 계산시간을 대폭 줄여 줄 수 있다. 제안된 방법은 가로 1.3인치 세로 1.9인치의 PCB기판에 적용되었고, 10배 정도의 계산시간저감 효과를 보였다. 제안된 방법은 보드임피던스의 반복적인 계산을 필요로 하는 PCB설계에 응용될 수 있다.

• 한국환경과학회지
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• 제18권2호
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• pp.197-203
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• 2009

The aim of this study is to calculate process emission of GHGs(greenhouse gases) in nonferrous-metal industry, such as Zn, Pb, Cu and Ni. In addition, variation and emission of GHGs generated from these company were defined. And then, GHGs algorithm and calculation formular which were considered as production process in each part of nonferrous-metal industry were developed to develop calculation program of GHGs emission. These algorithm and calculation formular would present fundamental direction about other nonferrous-metal industry in the future.

• Journal of Electrical Engineering and Technology
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• 제12권5호
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• pp.1883-1890
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• 2017

Transformers are one of the most precious elements of the electric power system. Stability and reliability of the electric power network mainly depend on the working of the transformer. Leakage reactance of the transformer is one of the important factors and accurate calculation of the leakage reactance is necessary for the transformer designers and electric distributors. Leakage reactance of the transformer depends on the geometry of the transformer. There are many different methods for the calculations of the leakage reactance however mostly are usable when the axial heights of the high voltage and low voltage windings are equal. When the axial heights of high voltage and low voltage windings are asymmetric most of the analytical methods are not reliable. In this study, a new analytical method is introduced for the calculation of the leakage reactance. Fourteen different transformers are investigated in this study and four of them are presented in this paper. The results of the new analytical method are compared with the experimental results. Other analytical and numerical methods are also compared with this new method. Results show that this method is more reliable and accurate as compared to the other analytical methods. The maximum relative error between short-circuit test and proposed method for these fourteen transformers was less than 2.8%.

• Journal of the Optical Society of Korea
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• 제18권2호
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• pp.134-145
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• 2014

The number of lens groups in modern zoom camera systems is increased above that of conventional systems in order to improve the speed of the auto focus with the high quality image. As a result, it is difficult to calculate zoom loci using the conventional analytic method, and even the recent one-step advanced numerical calculation method is not optimal because of the time-consuming problem generated by the iteration method. In this paper, in order to solve this problem, we suggest a new unified analytic method for zoom lens loci with finite object distance including infinite object distance. This method is induced by systematically analyzing various distances between the object and other groups including the first lens group, for various situations corresponding to zooming equations of the finite lens systems after using a spline interpolation for each lens group. And we confirm the justification of the new method by using various zoom lens examples. By using this method, we can easily and quickly obtain the zoom lens loci not only without any calculation process of iteration but also without any limit on the group number and the object distance in every zoom lens system.

• 대한전기학회논문지
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• 제41권7호
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• pp.712-722
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• 1992

Several voltage instability/collapse problems that have occurred in the electric utility industry worldwide have gained the attention of engineers and researchers of electric power systems. This paper proposes an effective calculation scheme of the extreme loading point and nose curves(P-V curves) using modified Newton-Raphson(N-R) load flow method and the Continuation Method. This method provides detail and visual information of the power system voltage profile and operating margin ro operators and planners. In this paper, a modified load flow claculation method for ill-conditioned power systems is introduced for the purpose of seeking more precise load flow solutions and nose curves, and the Continuation Method is also used as a part of the solution algorithm for the calculation of extreme loading point and nose curves. The conventional polar coordinate based N-R load flow program is modified to avoid numerical difficulties caused by the singularity of the Jacobian matrix occuring in the vicinity of extreme loading point of heavily loaded systems. Application results of the proposed method to Klos-Kerner 11-bus system and modified IEE-30-bus system are presented to assure the usefulness of the approach.