• 한국압력기기공학회 논문집
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• 제12권1호
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• pp.85-92
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• 2016

The operating temperature of VHTR components is much higher than that of conventional PWR due to high core outlet temperature of VHTR. Material requirements and technical issues of VHTR reactor components which are mainly dominated by high temperature service condition were discussed. The codification effort for high temperature material and design methodology are explained. The design class for VHTR components are classified as class A or B according to the recent ASME high temperature reactor design code. A separation of thermal boundary and pressure boundary is used for VHTR components as an elevated design solution. Key design characteristics for reactor pressure vessel, control rod, reactor internals, graphite reflector, circulator and intermediate heat exchanger were analysed. Thermo-mechanical analysis of the process heat exchanger, which was manufactured for test, is presented as an analysis example.

• 한국압력기기공학회 논문집
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• 제9권1호
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• pp.48-55
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• 2013

In this paper, W-DCAP-HTR(Web-based Design Compatibility Assessment Program for High Temperature Reactor) which will be used to check the design criteria for high temperature reactor is newly proposed. To do this, the assessment procedure of the ASME Sec.III Div.5 such as time-dependent primary stress limit, accumulated inelastic strain, and creep-fatigue damage evaluation were investigated. Furthermore, the trend of candidate materials for high temperature reactor was also reviewed. Then, all assessment procedures for high temperature reactor have been computerized to enhance the efficiency and to reduce the possibility of human error during calculating procedure by hand calculation. It can be directly conducted by adopting the actual thermal and structural analysis results. The validation of W-DCAP-HTR has been demonstrated by benchmark analysis.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2088-2093
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• 2008

In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gas-cooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method.

• 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.

• 대한기계학회논문집B
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• 제33권9호
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• pp.709-717
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• 2009

Heat transfer rate is a very important factor for the performance of a steam reformer because a steam reforming reaction is an endothermic reaction. Coaxial cylindrical reactor is the reactor design which can improve the heat transfer rate. Temperature, fuel conversion and heat flux in the coaxial cylindrical steam reformer are studied in this paper using numerical method under various operating conditions. Langmuir-Hinshelwood model and pseudo-homogeneous model are incorporated for the catalytic surface reaction. Dominant chemical reactions are assumed as a Steam Reforming (SR) reaction, a Water-Gas Shift (WGS) reaction, and a Direct Steam Reforming (DSR) reaction. Although coaxial cylindrical steam reformer uses 33% less amount of catalyst than cylindrical steam reformer, its fuel conversion is increased 10 % more and its temperature is also high as about 30 degree. There is no heat transfer limitation near the inlet area at coaxial-type reactor. However, pressure drop of the coaxial cylindrical reactor is 10 times higher than that of cylindrical reactor. Operating parameters of coaxial cylindrical steam reformer are the wall temperature, the inlet temperature, and the Gas Hourly Space Velocity (GHSV). When the wall temperature is high, the temperature and the fuel conversion are increased due to the high heat transfer rate. The fuel conversion rate is increased with the high inlet temperature. However, temperature drop clearly occurs near the inlet area since an endothermic reaction is active due to the high inlet temperature. When GHSV is increased, the fuel conversion is decreased because of the heat transfer limitation and short residence time.

• 전기학회논문지
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• 제65권11호
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• pp.1848-1853
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• 2016

Control Rod Position Indicator in nuclear reactor vessel has developed for small reactor in Korea. Because of severe environment in reactor vessel, target of this study is to develop the suitable position indicator. In this study, solenoid type position indicator made of Mineral Insulated Cable(MI Cable) was introduced to adapt in severe environment. And inductance of the solenoid was used to indicate the rod position for high precision. But problem of this concept is that a linear slope of inductance is changed by temperature effect. To resolve this problem, two sensing coils were introduced for temperature compensation. A role of the sensing coil is to make reference linear equation about certain temperature. To confirm this concept, also, inductance of solenoid and sensing coils were measured at room and high temperature (${\sim}300^{\circ}C$). The results of measurement show that the position error of sensing coil between room and high temperature was about 2%. But it was identified that this error was resulted from insufficient test environment (temperature error between solenoid and sensing coils was about 2% at high temperature condition). Therefore, solenoid type position indicator shows that it is very suitable in reactor vessel as a high precision rod position indicator.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• 한국연소학회:학술대회논문집
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• 한국연소학회 2015년도 제51회 KOSCO SYMPOSIUM 초록집
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• pp.157-159
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• 2015

In this study, $CF_4$ decomposition was experimentally investigated in a high temperature flow reactor. Effects of temperature, reactant composition and concentration, and residence time on its decomposition into other stable species were analyzed. Then the results were compared to numerical results obtained using Chemkin Plug Flow Reactor model with Princeton Chemistry. As a preliminary result higher decomposition rate is obtained for higher reactor temperature and long residence time when proper reactants are supplied.

• Nuclear Engineering and Technology
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• 제38권5호
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• pp.417-422
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• 2006

The present paper describes the 45 MW High Flux Reactor (HFR) which is located in Petten, The Netherlands. This paper focuses on selected technical aspects of this reactor and on nuclear fuel irradiation experiments. These fuel experiments are mainly experiments on Light Water Reactor (LWR) and Very/High Temperature Reactor (V/HTR) fuels, but also on Fast Reactor (FR) fuels, transmutation fuels and Material Test Reactor (MTR) fuels.

• Nuclear Engineering and Technology
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• 제32권6호
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• pp.566-594
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• 2000

In most LMFBR(Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER(Korea Advanced Liquid MEtal Reactor) reactor internal strictures is carried out for normal operating conditions which have the operating temperature 53$0^{\circ}C$ and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME Code Case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects.