• 대한기계학회:학술대회논문집
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• 대한기계학회 2005년도 창립60주년 기념 추계 학술대회
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• pp.117.2-117.2
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• 2005

• Nuclear Engineering and Technology
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• 제53권8호
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• pp.2753-2760
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• 2021

This paper proposes using sodium-cooled fast reactor technologies for use in hydrogen vapor methane (SMR) modification. Using three independent energy rings in the Russian BN-600 fast reactor, steam is generated in one of the steam-generating cycles with a pressure of 13.1 MPa and a temperature of 505 ℃. The reactor's second energy cycles can increase the gas-steam mixture's temperature to the required amount for efficient correction. The 620 ton/hr 540 ℃ steam generated in this cycle is sufficient to supply a high-temperature synthesis current source (700 ℃), which raises the steam-gas mixture's temperature in the reactor. The proposed technology provides a high rate of hydrogen production (approximately 144.5 ton/hr of standard H₂), also up to 25% of the original natural gas, in line with existing SMR technology for preparing and heating steam and gas mixtures will be saved. Also, exergy analysis results show that the plant's efficiency reaches 78.5% using HTR heat for combined hydrogen and power generation.

• 한국압력기기공학회 논문집
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• 제4권2호
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• pp.1-6
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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. The VHTR can produce hydrogen from heat and water by using a thermo-chemical process or from heat, water, and natural gas by steam reformer technology. A co-axial double-tube primary hot gas duct (HGD) is a key component connecting the reactor pressure vessel and the intermediate heat exchanger (IHX) for the VHTR. In this study, a preliminary design analysis for the primary HGD of the nuclear hydrogen system was carried out. These preliminary design activities include a determination of the size, a strength evaluation and an appropriate material selection. The determination of the size was undertaken based on various engineering concepts, such as a constant flow velocity model, a constant flow rate model, a constant hydraulic head model, and finally a heat balanced model.

• Nuclear Engineering and Technology
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• 제55권11호
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• pp.4295-4306
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• 2023

The vibration of fuel rods in axial flow is a universally recognized issue within both engineering and academic communities due to its significant importance in ensuring structural safety. This paper aims to thoroughly investigate the stability and nonlinear vibration of a fuel rod subjected to axial flow in a newly designed high temperature gas cooled reactor. Considering the possible presence of thermal expansion and large deformation in practical scenarios, the thermal effect and geometric nonlinearity are modeled using the von Karman equation. By applying Hamilton's principle, we derive the comprehensive governing equation for this fluid-structure interaction system, which incorporates the quadratic nonlinear stiffness. To establish a connection between the fluid and structure aspects, we utilize the Galerkin method to solve the perturbation potential function, while employing mode expansion techniques associated with the structural analysis. Following convergence and validation analyses, we examine the stability of the structure under various conditions in detail, and also investigate the bifurcation behavior concerning the buckling amplitude and flow velocity. The findings from this research enhance the understanding of the underlying physics governing fuel rod behavior in axial flow under severe yet practical conditions, while providing valuable guidance for reactor design.

• 멤브레인
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• 제14권3호
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• pp.185-191
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• 2004

원자력 발전의 고온 가스로(high temperature gas-cooled reactor, HTGR)의 냉각제로 사용되는 He가스의 열에너지를 이용하여 물을 분해해서 수소를 생산하는 "열화학적 수소제조 IS프로세스"에 대해 설명하였다. 특히, 분리막 기술의 이용에 관한 연구를 중점으로 정리하였다. 고온 원자력 열에너지를 이용한 열화학적 수소 제조법은 실현 가능한 단계까지 왔다고 생각되며, 아직 연구 개발 과제가 많이 남아 있지만, 미래의 청정에너지 중의 하나인 수소를 대량 생산할 수 있는 가능성을 갖고 있다.

• Nuclear Engineering and Technology
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• 제53권11호
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• pp.3685-3693
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• 2021

This paper presents a disturbance observer-based robust backstepping load-following control (DO-RBLFC) scheme for modular high-temperature gas-cooled reactors (MHTGRs) in the presence of actuator saturation and disturbances. Based on reactor kinetics and temperature reactivity feedback, the mathematical model of the MHTGR is first established. After that, a DO is constructed to estimate the unknown compound disturbances including model uncertainties, external disturbances, and unmeasured states. Besides, the actuator saturation is compensated by employing an auxiliary function in this paper. With the help of the DO, a robust load-following controller is developed via the backstepping technique to improve the load-following performance of the MHTGR subject to disturbances. At last, simulation and comparison results verify that the proposed DO-RBLFC scheme offers higher load-following accuracy, better disturbances rejection capability, and lower control rod speed than a PID controller, a conventional backstepping controller, and a disturbance observer-based adaptive sliding mode controller.

• Corrosion Science and Technology
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• 제12권3호
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• pp.132-141
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• 2013

This work is concerning a methodology of Ni-base superalloy development for a very high temperature gas-cooled reactor(VHTR) using design of experiments(DOE) and thermodynamic calculations. Total 32 sets of the Ni-base superalloys with various chemical compositions were formulated based on a fractional factorial design of DOE, and the thermodynamic stability of topologically close-packed(TCP) phases of those alloys was calculated by using the THERMO-CALC software. From the statistical evaluation of the effect of the chemical composition on the formation of TCP phase up to a temperature of 950 oC, which should be suppressed for prolonged service life when it used as the structural components of VHTR, 16 sets were selected for further calculation of the mechanical properties. Considering the yield and ultimate tensile strengths of the selected alloys estimated by using the JMATPRO software, the optimized chemical composition of the alloys for VHTR application, especially intermediate heat exchanger, was proposed for a succeeding experimental study.

• 대한기계학회논문집A
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• 제33권10호
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• pp.1108-1118
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• 2009

A Very High Temperature Gas Cooled Reactor (VHTR) has been selected as a high energy heat source of the order of $950^{\circ}C$ 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 a reactor pressure vessel and an intermediate heat exchanger in the VHTR. In this study, a structural sizing methodology for the primary HGD of the VHTR is suggested in order to modulate a flow-induced vibration (FIV). And as an example, a structural sizing of the horizontal HGD with a coaxial double-tube structure was carried out using the suggested method. These activities include a decision of the geometric dimensions, a selection of the material, and an evaluation of the strength of the coaxial double-tube type cross vessel components. Also in order to compare the FIV characteristics of the proposed design cases, a fluid-structure interaction (FSI) analysis was carried out using the ADINA code.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2004년도 추계학술발표회 발표논문집
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• pp.171-172
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• 2004

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 2005년도 춘계학술발표회
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• pp.443-444
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• 2005