• 한국수소및신에너지학회논문집
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• 제19권3호
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• pp.239-247
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• 2008

Hydrogen production facility using very high temperature gas cooled reactor lies in situation of high temperature and corrosion which makes hydrogen release easily. In that case of hydrogen release, there lies a danger of explosion. However, from the point of thermal-hydraulics view, the long distance of them makes lower efficiency result. In this study, therefore, outlines of hydrogen production using nuclear energy are researched. Several methods for analyzing the effects of hydrogen explosion upon high temperature gas cooled reactor are reviewed. Reliability physics model which is appropriate for assessment is used. Using this model, leakage probability, rupture probability and structure failure probability of very high temperature gas cooled reactor are evaluated and classified by detonation volume and distance. Also based on standard safety criteria which is value of $1{\times}10^{-6}$, safety distance between the very high temperature gas cooled reactor and the hydrogen production facility is calculated.

• Nuclear Engineering and Technology
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• 제44권7호
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• pp.817-824
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• 2012

The most important component of a nuclear power plant is its nuclear reactor. Studies on the integrity of reactors have become an important part regarding the safety of a nuclear power plant. The US Nuclear Regulatory Commission Regulatory Guide (NRC RG) 1.20 presents a Comprehensive Vibration Assessment Program (CVAP) to be used to verify the structural integrity of the Reactor Vessel Internals (RVI) for flow-induced vibration prior to commercial operation. However, there are few published studies related to the RVI CVAP. We classified the Advanced Power Reactor 1400 (APR1400) RVI CVAP as a non-prototype category-2 reactor as part of an independent validation of its design. The aim of this paper is to present the results of structural response analyses of the Upper Guide Structure (UGS) assembly of the APR1400 reactor. These results show that the UGS and the Inner Barrel Assembly (IBA) meet the specified integrity levels of the design acceptance criteria. The vibration and stress analysis results in this paper will be used as basic information to select measurement locations of the vibration and stress for the APR1400 RVI CVAP.

• Nuclear Engineering and Technology
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• 제27권5호
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• pp.710-720
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• 1995

원자로에서 펌프에 의해 야기되는 맥동 압력은 원자로 내부 구조물에 진동과 손상을 줄 수 있기 때문에 관심이 증가되고 있다. 본 연구에서는 냉각관과 환형관(원자로 압력 용기와 노심 보호 지지대 사이)으로 구성된 기하 형태에서 펌프에 의해 야기되는 맥동 압력을 해석할 수 있는 수력학적 모델을 개발하였다. 수학적 지배 방정식은 압축성, 비점성 유체에 대해 선형화된 Navier-Stokes 방정식이다. 냉각관과 환형관을 따로 분리하여 해석하고 두영역의 커플링 영향을 고려하였다. 또한 본 기하 형태에서 펌프맥동 압력에 영향을 미치는 주요 기하 인자에 대한 평가를 수행하였다. 본 해석 결과와 실험차를 비교하여 만족할 만한 결과를 얻었다.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.377-383
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• 2019

Targeting at simulating the application of thorium-uranium (TU) fuel in the CANDU-6 reactor, this paper analyzes the process using the code DRAGON/DONJON where the discrete TU fuel pins are applied in the CANDU-6 reactor under the time-average equilibrium refueling. The results show that the coolant void reactivity of the assembly analyzed in this paper is lower than that of 37-element bundle cell with natural uranium and 37-element bundle cell with mixed TU fuel pins; that the max time-average channel/bundle power of the core meets the limits - less than 6700kW/860 kW; that the fuel conversion ratio is higher than that of the CANDU-6 reactor with natural uranium; and that the exit burnup increases to 13400 MWd/tU. Thus, the simulation in this paper with the fuel in the 37-element bundle cell using discrete TU fuel pins can be considered to be applied in CANDU-6 reactor with adequate modifications of the core structure and operating modes.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.284-292
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• 2019

For the continuous operation of a nuclear reactor, burnt fuel needs to be replaced with fresh fuel, where appropriate (ex-vessel) fuel handling is required. Particularly for the Sodium-cooled Fast Reactor (SFR) refueling, its process has unique challenges due to liquid sodium coolant. The ex-vessel spent fuel transportation should concern several design features such as the radiation shielding, decay-heat removal, and inert space separated from air. This paper proposes a new design optimization methodology of cask shielding to transport the spent fuel assembly in a prototype SFR for the first time. The Particle Swarm Optimization (PSO) algorithm had been applied to design trade-offs between shielding and cask weight. The cask is designed as a double-cylinder structure to block an inert sodium region from the air-cooling space. The PSO process yielded the optimum shielding thickness of 26 cm, considering the weight as well. To confirm the shielding performance, the radiation dose of spent fuel removed at its peak burnup and after 1-year cooling was calculated. Two different fuel positions located during transportation were also investigated to consider a functional disorder in a cask drive system. This study concludes the current cask design in normal operations is satisfactory in accordance with regulatory rules.

• Nuclear Engineering and Technology
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• 제54권1호
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• pp.357-373
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• 2022

The high-temperature gas-cooled reactor pebble-bed module project is the first commercial Generation-IV NPP(Nuclear Power Plant) in China. A new joint is used for the vertical support of RPV(Reactor Pressure Vessel). The steel corbel is integrally embedded into the reactor-cabin wall through eight asymmetrically arranged pre-stressed high-strength bolts, achieving the different path transmission of shear force and moment. The vertical monotonic loading test of two specimens is conducted. The results show that the failure mode of the joint is bolt fracture. There is no prominent yield stage in the whole loading process. The stress of bolts is linearly distributed along the height of corbel at initial loading. As the load increases, the height of neutral axis of bolts gradually decreases. The upper and lower edges of the wall opening contact the corbel plate to restrict the rotation of the corbel. During the loading, the pre-stress of some bolts decreases. The increase of the pre-stress strength ratio of bolts has no noticeable effect on the structure stiffness, but it reduces the ultimate bearing capacity of the joint. A simplified calculation model for the elastic stage of the joint is established, and the estimation results are in good agreement with the experimental results.

• Nuclear Engineering and Technology
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• 제55권8호
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• pp.2835-2843
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• 2023

After an emergency shutdown of a lead-bismuth fast reactor, thermal stratification occurs in the upper Plenum, which negatively impacts the integrity of the reactor structure and the residual heat removal capacity of natural circulation flow. The research on thermal stratification of reactors has mainly been conducted using an experimental method, a system program, and computational fluid dynamics (CFD). However, the equipment required for the experimental method is expensive, accuracy of the system program is unpredictable, and resources and time required for the CFD approach are extensive. To overcome the defects of thermal stratification analysis, a high-precision full-order thermal stratification model based on CFD technology is prepared in this study. Furthermore, a reduced-order model has been developed by combining proper orthogonal decomposition (POD) with Galerkin projection. A comparative analysis of thermal stratification with the proposed full-order model reveals that the reduced-order thermal stratification model can well simulate the temperature distribution in the upper plenum and rapidly elucidate the thermal stratification interface characteristics during the lead-bismuth fast reactor accident. Overall, this study provides an analytical tool for determining the thermal stratification mechanism and reducing thermal stratification.

• 대한기계학회논문집A
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• 제21권7호
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• pp.1042-1049
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• 1997

LMFBR(Liquid Metal Fast Breeder Reactor) vessel is operated under the high temperatures of 500-550.deg. C. Thus, transient thermal loads were severe enough to cause inelastic deformation due to creep-fatigue and plasticity. For reduction of such inelastic deformations, Y-piece structure in the form of a thermal sleeve is used in LMFBR vessel under repeated start-up, service and shut-down conditions. Therefore, a systematic method for inelastic analysis is needed for design of the Y-piece structure subjected to such loading conditions. In the present investigation, finite element analysis of heat transfer and inelastic thermal stress were carried out for the Y-piece structure in LMFBR vessel under service conditions. For such analysis, ABAQUS program was employed based on the elasto-plastic and Chaboche viscoplastic constitutive equations. Based on numerical data obtained from the analysis, creep-fatigue damage estimation according to ASME Code Case N-47 was made and compared to each other. Finally, it was found out that the numerical predictio of damage level due to creep based on Chaboche unified viscoplastic constitutive equation was relatively better compared to elasto-plastic constitutive formulation.

• 한국병원경영학회지
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• 제5권1호
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• pp.156-175
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• 2000

This study was designed to find out the relations between the major investment decision-making behaviors and profitability of the hospital. A total of 57 hospitals were analyzed on this study. The major findings were as follows; 1. Among the types of the investment decision-making, major factors affecting the profitability were where the top management belongs among the defender, analyzer, prospector, and reactor type. Other factors were whether or not hospital analyzes which is more economical between the purchase by cash and lease of the medical equipment and whether or not hospital changes the decision before the actual investment. 2, Among the types of the investment decision-making, major factors affecting the financial structure and efficient operation of the assets were ranking of the priority and whether or not hospitals can get enough revenue and cash flow when hospitals have to borrow a big amount of fund from outside. 3. Among the financial indices regarding the financial stability, major factor affecting the profitability was fixed assets to long-tenn capital. Other factors affecting the financial structure and efficient operation of the assets were value added to medical equipment, normal profit to medical equipment, liability to total assets, current ratio, value added to payroll expenses. 4. Investment decision-making behaviors are partially influencing on the financial structure and efficient operation of the assets. However it was proved that the profitability was the most influencial factor than other factors related with the operation of the hospital. 5. To improve the irrational investment decision-making behaviors strategic management system should be introduced, and the top mamagement's investment decision-making style should be changed from reactor and analyser styles to prospector and reactor ones.

• Nuclear Engineering and Technology
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• 제48권4호
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• pp.1002-1008
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• 2016

An understanding of the coarsening process of the large fission gas pores in the high burn-up structure (HBS) of irradiated $UO_2$ fuel is very necessary for analyzing the safety and reliability of fuel rods in a reactor. A numerical model for the description of pore coarsening in the HBS based on the Ostwald ripening mechanism, which has successfully explained the coarsening process of precipitates in solids is developed. In this model, the fission gas atoms are treated as the special precipitates in the irradiated $UO_2$ fuel matrix. The calculated results indicate that the significant pore coarsening and mean pore density decrease in the HBS occur upon surpassing a local burn-up of 100 GWd/tM. The capability of this model is successfully validated against irradiation experiments of $UO_2$ fuel, in which the average pore radius, pore density, and porosity are directly measured as functions of local burn-up. Comparisons with experimental data show that, when the local burn-up exceeds 100 GWd/tM, the calculated results agree well with the measured data.