• 한국추진공학회지
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• 제22권5호
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• pp.81-87
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• 2018

극초음속 비행체에서는 공기와의 마찰열과 엔진열의 증가로 기체 내부의 열적 부하가 발생한다. 이는 비행체 내부 구조물의 변형을 일으키고 오작동을 발생시킬 수 있다. 흡열연료는 액체 탄화수소 연료로써 흡열반응을 통해 열을 흡수할 수 있는 연료이다. 본 연구에서는 실제 반응조건과 비슷한 고정층 흐름형 반응기에서 Exo-tetrahydrodicyclopentadiene(exo-THDCP)를 연료로 사용하여 흡열 촉매 종류에 따른 흡열 반응 시 생성물, 코크 생성량과 촉매 특성 변화 간 관계에 대한 연구를 수행하였다.

• Nuclear Engineering and Technology
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• 제51권6호
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• pp.1681-1688
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• 2019

In accordance with ASME Boiler and Pressure Vessel (B&PV) Code Sec.XI Appendix. G, a postulated crack is located at the beltline of a reactor pressure vessel because the neutron flux at the beltline is higher than elsewhere. This means that the distance between the core and the semi-spherical bottom head is longer than the distance between the core and the cylindrical beltline. However, several Small and Medium sized Reactors have bottom heads with diverse shapes, including dished or semi-elliptical shapes, to satisfy the requirement and performance. So, the aim of this paper is to evaluate the effect of crack location on Pressure-Temperature limit curve. To do this, two types of postulated crack location, such as beltline and semi-elliptical bottom head, were adopted to derive the Pressure-Temperature limit curve. Also, parametric studies for neutron flux, crack shape and so on were performed. As a result, core critical temperature of semi-elliptical bottom head is found to higher than that of beltline even when they have same values of thickness and neutron flux. This result will be useful to enhance the understanding of Pressure-Temperature limit curve.

• Nuclear Engineering and Technology
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• 제53권7호
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• pp.2407-2417
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• 2021

A structural integrity evaluation program (STEP) was developed for the high temperature reactor design evaluation according to the ASME Boiler and Pressure Vessel Code (ASME B&PV), Section III, Rules for Construction of Nuclear Facility Components, Division 5, High Temperature Reactors, Subsection HB. The program computerized HBB-3200 (the design by analysis procedures for primary stress intensities in high temperature services) and Appendix T (HBB-T) (the evaluation procedures for strain, creep and fatigue in high temperature services). For evaluation, the material properties and isochronous curves presented in Section II, Part D and HBB-T were computerized for the candidate materials for high temperature reactors. The program computerized the evaluation procedures and the constants for the weldment. The program can generate stress/temperature time histories of various loads and superimpose them for creep damage evaluation. The program increases the efficiency of high temperature reactor design and eliminates human errors due to hand calculations. Comparisons that verified the evaluation results that used the STEP and the direct calculations that used the Excel confirmed that the STEP can perform complex evaluations in an efficient and reliable way. In particular, fatigue and creep damage assessment results are provided to validate the operating conditions with multiple types of cycles.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.532-548
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• 2021

Nuclear power plants contain several monitoring systems that can identify the in-vessel phenomena of a severe accident (SA). Though a lot of analysis and research is carried out on SA, right from the development of the nuclear industry, not all the possible circumstances are taken into consideration. Therefore, to improve the efficacy of the safety of nuclear power plants, additional analytical studies are needed that can directly monitor severe accident phenomena. This paper presents an interacting multiple model (IMM) based fault detection and diagnosis (FDD) approach for the identification of in-vessel phenomena to provide the accident propagation information using reactor vessel (RV) out-wall temperature distribution during severe accidents in a nuclear power plant. The estimation of wall temperature is treated as a state estimation problem where the time-varying wall temperature is estimated using IMM employing three multiple models for temperature evolution. From the estimated RV out-wall temperature and rate of temperature, the in-vessel phenomena are identified such as core meltdown, corium relocation, reactor vessel damage, reflooding, etc. We tested the proposed method with five different types of SA scenarios and the results show that the proposed method has estimated the outer wall temperature with good accuracy.

• Korean Chemical Engineering Research
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• 제60권3호
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• pp.452-458
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• 2022

3가지 종류의 기-액 하이브리드 수평형, 수직형 그리고 needle-to-cylinder형 플라즈마 반응기가 제작되었다. 이들 반응기를 통하여 대기압 플라즈마 방전에서 발생하는 반응 활성종 생성과 전극 내의 전위차를 통한 세정성분의 기-액 활성화 반응을 일으키는 고효율 친환경 기반의 세정 개념을 제시하였다. 세정성능에 대한 효율성을 비교한 결과, needle-to-cylinder형 반응기가 가장 우수한 특성을 가졌다. 본 연구를 통해 기-액 하이브리드 대기압 플라즈마 반응기가 반도체 공정 등 초정밀 세정공정에 응용 가능성이 있음을 확인하였다.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1244-1249
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• 2023

Recently, the Kingdom of Saudi Arabia (KSA) announced the development of first-of-a-kind(FOAK) and most advanced futuristic vertical city and named as 'The LINE'. The project will have zero carbon dioxide emissions and will be powered by clean energy sources. Therefore, a study was designed to understand which clean energy sources might be a better choice. Because of its nearly carbon-free footprint, nuclear energy may be a good choice. Nowadays, the development of very small modular reactors (vSMRs) is gaining attention due to many salient features such as cost efficiency and zero carbon emissions. These reactors are one step down to actual small modular reactors (SMRs) in terms of power and size. SMRs typically have a power range of 20 MWe to 300 MWe, while vSMRs have a power range of 1-20 MWe. Therefore, a study was conducted to discuss different vSMRs in terms of design, technology types, safety features, capabilities, potential, and economics. After conducting the comparative test and analysis, the fuel cycle modeling of optimal and suitable reactor was calculated. Furthermore, the levelized unit cost of electricity for each reactor was compared to determine the most suitable vSMR, which is then compared other generation SMRs to evaluate the cost variations per MWe in terms of size and operation. The main objective of the research was to identify the most cost effective and simple vSMR that can be easily installed and deployed.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.807-812
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• 2001

This paper represents the wall thinning analysis results for secondary side piping of two types of domestic nuclear power plants based on the DB establishment and F AC analysis study for NPP secondary system piping. CHECWORKS code utilized in this study has been applied world widely to wall thinning analyses for secondary side piping and its reliability has also been proved. The predicted wear rates for several piping systems of a pressurized water reactor NPP are compared with those of a pressurized heavy water reactor NPP and with the measured wear rates. On the basis of comparison results of the predicted and measured wear rates, the analysis results can be effectively applied to the development of a standard thinned pipe management program targeted all domestic nuclear power plants.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.495-501
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• 1997

A preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only neutronic aspects of the accident were considered, independent of the accident scenario, and efforts were made to estimate the quantity of molten fuel which must be ejected out of the core to assure a sub-critical state after the accident. Two types of parameters were examined : characteristic parameters of molten core such as geometry, molten pool type (homogenized or stratified), fuel temperature, environment, and relative parameters to normal core such as core size(small or large), and fuel type (oxide, nitride, metal). The first type of parameters was found to intervene more directly in the recriticality risk than the second type of parameters.

• 대한전기학회논문지:전기물성ㆍ응용부문C
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• 제54권5호
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• pp.218-223
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• 2005

Impedance matching characteristics of planar type inductively coupled plasma sources are investigated utilizing the previously reported two-dimensional theory[1] of the anomalous skin effect. Two types of matching networks are considered, and the values of the circuit elements are expressed as functions of various reactor parameter. Also, two cases of perfect and imperfect matching conditions are considered and the functional dependence of the values of matching capacitance and reflection coefficient on the various reactor parameters are investigated using the present circuit model.

• Asian-Australasian Journal of Animal Sciences
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• 제5권2호
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• pp.199-215
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• 1992

Sources of organic waste materials for aerobic and/or anaerobic degradation, or for composting of solid wastes in Germany were estimated. The basic microbiology and the energetics of these processes were compared with special emphasis on anaerobic degradation, for which a general degradation scheme of carbohydrates is presented. Advantages of anaerobic over aerobic treatment processes are pointed out and conditions for maintaining a highly stable anaerobic process as well as producing a sanitized, hygienic product are discussed. Reactor systems suitable for efficient treatment of wastes with a high or low proportion of suspended solids are principally compared and results of laboratory studies on the degradation of several wastes and animal manures summarized. Finally, a piggery slurry treatment factory for an ultimate slurry processing to obtain a dry fertilizer and a harmless, disposable liquid, as it is in operation in Helmond/Holland, is presented and preliminary process data are presented.