• Korean Journal of Fisheries and Aquatic Sciences
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• v.53 no.5
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• pp.790-793
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• 2020

Length-weight relationships (LWR) for 27 fish species inhabit Southern sea in Korea were investigated to describe several biological characters. Total 7,399 individuals were collected by R/V Tamgu-20 using bottom trawl between 2018 to 2019 and were identified as 19 families and 27 species. Parameter b ranged from 2.414 to 3.472. Thirteen species among 27 species showed isometric growth (b=3), six species showed negative allometry (b<3) and eight species showed positive allometry (b>3). The results of this study provide useful basic biological information about 27 fishes and are highly reliable due to use of data measured directly.

• Nuclear Engineering and Technology
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• v.52 no.11
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• pp.2415-2421
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• 2020

The axial heterogeneous void distribution in a fuel channel is a relevant and important issue during nuclear reactor analysis for LWR, especially for boiling water channel-type reactors. Variation of the coolant density in fuel channel has an effect on the neutron spectrum that will in turn have an impact on the values of absolute reactivity, the void reactivity coefficient, and the fuel isotopic compositions during irradiation. This effect is referring to as the history effect in light water reactor calculations. As the void reactivity effect is positive in RBMK type reactors, the underestimation of water density heterogeneity in 3D reactor core numerical calculations could cause an uncertainty during assessment of safe operation of nuclear reactor. Thus, this issue is analysed with different cross-section libraries which were generated with WIMS8 code at different reference water densities. The libraries were applied in single fuel model of the nodal code of QUABOX-CUBBOX/HYCA. The thermohydraulic part of HYCA allowed to simulate axial water distribution along fuel assembly model and to estimate water density history effect for RBMK type fuel.

• Proceedings of the KWS Conference
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• 1996.10a
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• pp.19-34
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• 1996

A review was made of fatigue failures of nuclear power plant components in Japan, which were experienced in service and during periodical inspection. No case has been recently reported of a service fatigue failure of a reactor pressure vessel itself, excluding nozzle corner cracks, that occurred many years ago. But, service fatigue failures have been occasionally experienced in piping systems, pumps, and valves, on which fatigue design seems to have been inadequately applied. The causes of fatigue failures can be divided into two categories: mechanical-vibration-induced fatigue and thermal-fluctuation-induced fatigue. Vibration-induced fatigue failure occurs more frequently than is generally thought. The lesson gleaned from the present survey is a recognition that a service fatigue failure may occur due to any one or a combination of the following factors: (1) lack of communication between designers and fabrication engineers, (2) lack of knowledge about a possibility of fatigue failure and poor consideration about the effects of residual stresses, (3) lack of consideration on possible vibration in the design and fabrication stages, and (4) lack of fusion or poor penetration in a welded joint.

• JSTS:Journal of Semiconductor Technology and Science
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• v.14 no.1
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• pp.61-69
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• 2014

With the extreme ultraviolet (EUV) lithography, the performance limit of chemically amplified resists has recently been extended to 16- and 11-nm nodes. However, the line edge roughness (LER) and the line width roughness (LWR) are not reduced automatically with this performance extension. In this paper, to investigate the impacts of the EUVL mask and the EUVL exposure process on LER, EUVL is modeled using multilayer-thin-film theory for the mask structure and the Monte Carlo (MC) method for the exposure process. Simulation results demonstrate how LERs of the mask transfer to the resist and the exposure process develops the resist LERs.

• Nuclear Engineering and Technology
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• v.30 no.6
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• pp.609-616
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• 1998

Cast duplex stainless steels are extensively used for primary pressure boundary components. These components are, however, embrittled due to the precipitation of $\alpha$' phase by spinodal decomposition and other processes when exposed to reactor operating temperature for a design lifetime or life extension conditions. This report presents a procedure for estimating the current condition and the residual life of safety-related stainless steel components by using ANL database and correlations. The database of Charpy impact energy suggests that CF-8M grade is the most susceptible to thermal aging and CF-3 grade is the least. Thus, the integrity of CF-8M alleys may be degraded seriously and the degree of deterioration may exceed acceptance limit after several years of service in the nuclear reactors.

• Proceedings of the Korean Information Science Society Conference
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• 1999.10a
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• pp.581-583
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• 1999

상속(Legacy) 소프트웨어 시스템은 오랜 기간 사용되었고 충분히 검증된 안정적인 서비스를 현재까지도 제공하는 유용한 시스템이다. 새로운 분산 객체 환경에서 기존의 시스템에서 제공하는 서비스를 사용하기 위한 방법으로, 기존 시스템을 대체할 새로운 시스템을 개발하는 방법과 기존 시스템의 코드를 수정하는 방법과 기존 시스템을 객체 포장기법으로 포장해서 사용하는 방법이 있다. 본 논문은 이 중에서 기존 시스템을 객체로 포장하여 분산 객체 기술인 CORBA 환경에서 이 시스템을 재사용하는 방법에 대하여 연구한다. 이 과정에서 다양한 형태의 인터페이스를 가진 기존 시스템을 효과적으로 포장할 수 있는 방법으로 LWR(Legacy Wrapping Rule)을 제안하고, 랩퍼(Wrapper)인 구현 객체 클래스를 만드는 랩퍼 생성기를 구현하였다. 이렇게 함으로써 상속 시스템을 보다 쉽고 강력하게 분산 환경으로 이주시킬 수 있다.

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.309-316
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• 2005

The contents of a new book currently in preparation are described. The dearth of books in the field of nuclear materials has left both students in nuclear materials classes and professionals in the same field without a resource for the broad fundamentals of this important sub-discipline of nuclear engineering. The new book is devoted entirely to materials problems in the core of light-water reactors, from the pressure vessel into the fuel. Key topics deal with the $UO_2$ fuel, Zircaloy cladding, stainless steel, and of course, water. The restriction to LWR materials does not mean a short monograph; the enormous quantity of experimental and theoretical work over the past 50 years on these materials presents a challenge of culling the most important features and explaining them in the simplest quantitative fashion. Moreover, LWRs will probably be the sole instrument of the return of nuclear energy in electric power production for the next decade or so. By that time, a new book will be needed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.131-136
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• 1996

An analytic nodal expansion method has been derived for the multigroup neutron diffusion equation in 2-D cylindrical(R-Z) coordinate. In this method we used the second order Legendre polynomials for source, and transverse leakage, and then the diffusion eqaution was solved analytically. This formalism has been applied to 2-D LWR model. $textsc{k}$$_{eff}$, power distribution, and computing time have been compared with those of ADEP code(finite difference method). The benchmark showed that the analytic nodal expansion method in R-Z coordinate has good accuracy and quite faster than the finite difference method. This is another merit of using R-Z coordinate in that the transverse integration over surfaces is better than the linear integration over length. This makes the discontinuity factor useless.s.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1602-1611
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• 2021

A varying degree of eccentricity always exists in the initial configuration of a nuclear fuel rod. Its impact on traditional LWR fuel is limited as the radial gap closes relatively early during irradiation. However, the effect of misalignment is expected to be more relevant in rods with highly conductive fuels, large initial gaps and low conductivity filling gases. In this paper, we study similar characteristics in the experimental setup of two fuel disc irradiation campaigns carried out in the OECD Halden Boiling Water Reactor. Using the multi-dimensional fuel performance code OFFBEAT, we combine 2-D axisymmetric and 3-D simulations to investigate the effect of eccentricity on the fuel temperature distribution. At the same time, we illustrate how the advent of modern tools with multi-dimensional capabilities might further improve the design and interpretation of in-pile separate-effect tests and we outline the potential of such an analysis for upcoming experiments.

• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1190-1197
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• 2018

Analysis of the GRSW-A model coupled into the FALCON code is extended by simulation of central void formation in fuel pellets due to high-temperature fuel restructuring. The extended calculation is verified against published, well-known experimental data. Good agreement with the data for a central void diameter in pellets of the rod irradiated in an Experimental Breeder Reactor is shown. The new calculation methodology is employed in comparative analysis of modern BWR fuel behavior under assumed high-power operation. The initial fuel porosity is shown to have a major effect on the predicted central void diameter during the operation in question. Discernible effects of a central void on peak fuel temperature and Pellet-Cladding Mechanical Interaction (PCMI) during a simulated power ramp are shown. A mitigating effect on PCMI is largely attributed to the additional free volume in the pellets into which the fuel can creep due to internal compressive stresses during a power ramp.