• Title/Summary/Keyword: high temperature gas cooled reactor

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Computer Simulation of Methanation Reactor with Monolith Catalyst (전산 모델링을 통한 모노리스 촉매형 메탄화 반응기의 성능 특성 연구)

  • Chi, Junhwa;Kim, Sungchul;Hong, Jinpyo
    • Journal of Hydrogen and New Energy
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    • v.25 no.4
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    • pp.425-435
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    • 2014

  • Simulation studies on catalytic methanation reaction in externally cooled tubular reactor filled with monolithic catalysts were carried out using a general purpose modelling tool $gPROMS^{(R)}$. We investigated the effects of operating parameters such as gas space velocity, temperature and pressure of feeding gas on temperature distribution inside the reactor, overall CO conversion, and chemical composition of product gas. In general, performance of methanation reaction is favored under low temperature and high pressure for a wide range of their values. However, methane production becomes negligible at temperatures below 573K when the reactor temperature is not high enough to ignite methanation reaction. Capacity enhancement of the reactor by increasing gas space velocity and/or gas inlet pressure resulted no significant reduction in reactor performance and heat transfer property of catalyst.

  • https://doi.org/10.7316/KHNES.2014.25.4.425 인용 PDF KSCI

Suggestion of Structural Sizing Methodology on a Coaxial Double-tube Type Hot Gas Duct for the VHTR (초고온가스로의 동심축 이중관형 고온가스덕트에 대한 구조정산 방법론 제안)

  • Song, Kee-Nam;Kim, Y.W.
    • 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.

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High-Temperature Tensile Strengths of Alloy 617 Diffusion Weldment (Alloy 617 확산용접재의 고온 인장강도)

  • Sah, Injin;Hwang, Jong-Bae;Kim, Eung-Seon
    • Transactions of the Korean Society of Pressure Vessels and Piping
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    • v.14 no.1
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    • pp.15-23
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    • 2018

  • A compact heat exchanger is one of critical components in a very high temperature gas-cooled reactor (VHTR). Alloy 617 (Ni-Cr-Co-Mo) is considered as one of leading candidates for this application due to its excellent thermal stability and strengths in anticipated operating conditions. On the basis of current ASME code requirements, sixty sheets of this alloy are prepared for diffusion welding, which is the key technology to have a reliable compact heat exchanger. Optical microscopic analysis show that there are no cracks, incomplete bond, and porosity at/near the interface of diffusion weldment, but Cr-rich carbides and Al-rich oxides are identified through high resolution electron microscopic analysis. In high-temperature tensile testing, superior yield strengths of the diffusion weldment compared to the code requirement are obtained up to 1223 K ($950^{\circ}C$). However, both tensile strength and ductility drop rapidly at higher temperature due to the insufficient grain boundary migration across the interface of diffusion weldment. Best fit curves for minimum yield strength and average tensile strength are drawn from the experimental tensile results of this study.

  • https://doi.org/10.20466/KPVP.2018.14.1.015 인용 PDF KSCI

THE BENCHMARK CALCULATIONS OF THE GAMMA+ CODE WITH THE HTR-10 SAFETY DEMONSTRATION EXPERIMENTS

  • Jun, Ji-Su;Lim, Hong-Sik;Lee, Won-Jae
    • Nuclear Engineering and Technology
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    • v.41 no.3
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    • pp.307-318
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    • 2009

  • KAERI (Korea Atomic Energy Research Institute) has developed the GAMMA+ code for a thermo-fluid and safety analysis of a VHTR (Very High Temperature Gas-Cooled Reactor). A key safety issue of the VHTR design is to demonstrate its inherent safety features for an automatic reactor power trip and power stabilization during an anticipated transient without scram (ATWS) accident such as a loss of forced cooling by a trip of the helium circulator (LOFC) or a reactivity insertion by a control rod withdrawal (CRW). This paper intends to show the ATWS assessment capability of the GAMMA+ code which can simulate the reactor power response by solving the point-kinetic equations with six-group delayed neutrons, by considering the reactivity changes due to the effects of a core temperature variation, xenon transients, and reactivity insertions. The present benchmark calculations are performed by using the safety demonstration experiments of the 10 MW high temperature gas cooled-test module (HTR-10) in China. The calculation results of the power response transients and the solid core temperature behavior are compared with the experimental data of a LOFC ATWS test and two CRW ATWS tests by using a 1mk-control rod and a 5mk-control rod, respectively. The GAMMA+ code predicts the power response transients very well for the LOFC and CRW ATWS tests in HTR-10.

  • https://doi.org/10.5516/NET.2009.41.3.307 인용 PDF KSCI