• Korean Chemical Engineering Research
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• v.52 no.4
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• pp.459-466
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• 2014

SI Cyclic process is one of the thermochemical hydrogen production processes using iodine and sulfur for producing hydrogen molecules from water. VHTR (Very High Temperature Reactor) can be used to supply heat to hydrogen production process, which is a high temperature nuclear reactor. IHX (Intermediate Heat Exchanger) is necessary to transfer heat to hydrogen production process safely without radioactivity. In this study, the strategy for the optimum design of IHX between SI hydrogen process and VHTR is proposed for various operating pressures of the reactor, and the different cooling fluids. Most economical efficiency of IHX is also proposed along with process conditions.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.33 no.9
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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.

• Journal of Energy Engineering
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• v.15 no.2 s.46
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• pp.127-137
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• 2006

To resolve the environmental and economics problems of fossil fuel energy, a hydrogen economy is promoted in many developed countries. Massive production of hydrogen using a nuclear power is a practical way to feed fuel required for the hydrogen economy. The author introduces a very high temperature reactor and its development status. He also reviews recent achievements and directions of research in hydrogen production process, such as sulfur-iodine thermochemical cycle, sulfur hybrid cycle, and high temperature electrolysis.

• The Transactions of The Korean Institute of Electrical Engineers
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• v.65 no.11
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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.

• Proceedings of the KSME Conference
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• 2008.11a
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• pp.717-724
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• 2008

Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source for nuclear hydrogen generation, which can produce hydrogen from water or natural gas. A primary hot gas duct (HGD) as a coaxial double-tube type cross vessel is a key component connecting the reactor pressure vessel and the intermediate heat exchanger for the VHTR. In this study, structural sizing methodology for the primary HGD with a coaxial double-tube of the VHTR that produces heat at temperatures in the order of $950^{\circ}C$ was suggested and a structural pre-sizing of it was carried out as an example.

• Nuclear Engineering and Technology
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• v.32 no.6
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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.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.5
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• pp.553-561
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• 2010

A medium-scale helium loop that can simulate a VHTR (very-high-temperature reactor) is now under construction at the Korea Atomic Energy Research Institute. The heaters of the test helium loop electrically heat helium fluid up to $950^{\circ}C$ at pressures of 1 to 9 MPa. To optimize the design specifications of the experimental helium loop, the conjugate heat transfer in the high-temperature helium heater was analyzed by performing a CFD simulation. The analysis results indicate that the maximum temperature does not exceed the allowable limit. It is confirmed that the thermal characteristics of the loop with the given geometry satisfy the design requirements.

• Journal of the Korean Society of Mechanical Technology
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• v.13 no.2
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• pp.145-153
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• 2011

The irradiation tests of materials in HANARO have been performed usually at temperatures below $300^{\circ}C$ at which the RPV(Reactor Pressure Vessel) materials of the commercial reactors such as the light water reactor and CANDU are operated. As VHTR(Very High Temperature Reactor) and SFR(Sodium-cooled Fast Reactor) projects are being carried as a part of the present Gen-IV program in Korea, the requirements for irradiation of materials at temperatures higher than $500^{\circ}C$ are recently being gradually increased. To overcome the restriction in the use at high temperature of the existing Al thermal media, a new capsule with double thermal media composed of two kinds of materials such as Al-Ti and Al-graphite was designed and fabricated more advanced than the single thermal media capsule. At the irradiation test of the capsule, the temperature of the specimens successfully reached $700^{\circ}C$ and the integrity of Al, Ti and graphite material was maintained.

• Transactions of the Korean hydrogen and new energy society
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• v.34 no.3
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• pp.267-273
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• 2023

As the impacts of global climate change become increasingly apparent, the reduction of carbon emissions has emerged as a critical subject of discussion. Nuclear power has garnered attention as a potential carbon-free energy source; however, the rapidity of load following in nuclear power generation poses challenges in comparison to fossil-fueled methods. Consequently, power-to-gas systems, which integrate nuclear power and hydrogen, have attracted growing interest. This study presents a preliminary design of a very high temperature reactor (VHTR) integrated blue hydrogen production process utilizing DWSIM, an open-source process simulator. The blue hydrogen production process is estimated to supply the necessary calorific value for carbon capture through tail gas combustion heat. Moreover, a thermodynamic assessment of the main recuperator is performed as a function of the helium flow rate from the VHTR system to the blue hydrogen production system.

• Journal of computational fluids engineering
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• v.16 no.3
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• pp.95-103
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• 2011

In the block type VHTR core, there are inevitable gaps among core blocks for the installation and refueling of the fuel blocks. These gaps are called bypass gap and the bypass flow is defined as a coolant flows through the bypass gap. Distribution of core bypass flow varies according to the reactor operation since the graphite core blocks are deformed by the fast neutron irradiation and thermal expansion. Furthermore, the cross-flow through an interfacial gap between the stacked blocks causes flow mixing between the coolant holes and bypass gap, so that complicated flow distribution occurs in the core. Since the bypass flow affects core thermal margin and reactor efficiency, accurate prediction and evaluation of the core bypass flow are very important. In this regard, experimental and computational studies were carried out to evaluate the core bypass flow distribution. A multi-block experimental apparatus was constructed to measure flow and pressure distribution. Multi-block effect such as cross flow phenomenon was investigated in the experiment. The experimental data were used to validate a CFD model foranalysis of bypass flow characteristics in detail.