• 산업경영시스템학회지
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• 제38권4호
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• pp.39-44
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• 2015

The VHTR (Very High Temperature gas-cooled nuclear Reactor) has been considered as a major heat source and the most safe generation IV type reactor for mass hydrogen production to prepare for the hydrogen economy era. The VHTR satisfies goals for the GIF (Generation IV International Forum) policy such as sustainablility, economics, reliability and proliferation resistance and physical protection, and safety. As a part of a VHTR economic analysis, we have studied the VHTR construction cost and operation and maintenance cost. However, it is somewhat difficult to expect the ripple effect on the whole industry due to the lack of information about Inter-industries relationship. In many case, the ripple effect are based on experts' knowledge or uncertain qualitative assumptions. As a result, we propose quantitative analysis techniques for ripple effects such as the production inducement effect, added value inducement effect, and employment inducement effect for VHTR 600MWt${\times}$4 modules construction and operation ripple effect based on NOAK (Nth Of A Kind). Because inducement effect values have been published annually, we predict inducement effect's relation function and estimated values including production inducement effect value, added value inducement effect value, and employment inducement effect value using time series and estimated values are verified with published inducement effects' value. This paper presents a new method for the ripple effect and preliminary ripple effect consequence using a time series analysis and inter-industry table. This ripple effect analysis techniques can be applied to effect expectation analysis as well as other type reactor's ripple effect analysis including VHTR for process heat.

• 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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회A
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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.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2008년도 추계학술대회B
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• pp.2850-2855
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• 2008

Very High Temperature Reactor (VHTR) has been selected as a high energy heat source for a nuclear hydrogen generation. The VHTR heat is transferred to a thermo-chemical hydrogen production process through an intermediate loop. Both Process Heat Exchanger and sulfuric acid evaporator provide the coupled components between the VHTR intermediate loop and hydrogen production module. A small scaled Compact Nuclear Hydrogen Coupled Components test loop is developed to simulate the VHTR intermediate loop and hydrogen production module. Main objective of the loop is to screening the candidates of NHDD (Nuclear Hydrogen Development and Demonstration) coupled components. The operating condition of the gas loop is a temperature up to $950^{\circ}C$ and a pressure up to 6.0MPa. The thermal and fluid dynamic design of the loop is dependent on the structures that enclose the gas flow, especially primary side that has fast gas velocity. We designed and constructed a small scale sulfuric acid experimental system which can simulate a part of the hydrogen production module also.

• 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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• 제39권2호
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• pp.111-122
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• 2007

The current energy supply system is burdened by environmental and supply problems. The concept of a hydrogen economy has been actively discussed worldwide. KAERI has set up a plan to demonstrate massive production of hydrogen using a VHTR by the early 2020s. The technological gap to meet this goal was identified during the past few years. The hydrogen production process, a process heat exchanger, the efficiency of an I/S thermochemical cycle, the manufacturing of components, the analysis tools of VHTR, and a coated particle fuel are key areas that require urgent development. Candidate NHDD plant designs based on a 200 MWth VHTR core and I/S thermochemical process have been studied and some of analysis results are presented in this paper.

• 한국전산유체공학회지
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• 제15권1호
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• pp.1-8
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• 2010

RCCS is a passive safety-related system that removes the decay heat of VHTR when normal decay heat removal systems are in failure. Understanding thermo-hydraulics of RCCS is important to design a safer VHTR. RCCS consists of 292 cooling panels, which are placed in the reactor cavity. The layout of RCCS gives an idea that, for CFD simulations, cooling panels can be assumed to be one annulus tube. This assumption can reduce significantly the computational time, especially for the unsteady simulation. To simulate RCCS in an axisymmetric manner, three models were suggested and compared. Each model has (1) the same outer radius, (2) the same cross-sectional area (3) the same pressure drop, respectively, as the RCCS cooling panels. The steady-state simulation was conducted with these three models and the DO radiation model. It is found that over 90% of the heat from the outer wall of the reactor pressure vessel is transported to the RCCS by radiative heat transfer. The simulation with the third model, which has the same pressure drop as the design, estimates the closest wall temperature profiles to a thermo-hydraulic code, GAMMA+, result.

• 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.

• 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.

• 에너지공학
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• 제25권4호
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• pp.85-92
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• 2016

초고온가스로는 열출력 밀도가 낮아 노심용융의 가능성이 낮으며, 냉각재 상실사고 시 수소 발생 등으로 인한 폭발의 위험도 없다. 안전성 측면의 장점과 더불어 냉각재를 초고온으로 만들어 전력생산이외에 산업시설용 공정열로의 응용도 가능하다. 본 논문에서는 초고온가스로를 일차계통으로 하고, 전력 및 공정열 공급이 가능한 이차계통의 개념 설계를 담고 있다. 기존에 NGNP(Next Generation Nuclear Part)에서 제안한 350 MW 열출력 원자로 모델을 기반으로 수소생산 루프와는 별도로 전력생산을 위한 300 MW의 열에너지를 중간열교환기를 통해 이차계통으로 전달하는 참조모델을 개발하고, 이를 열역학적 측면에서 분석하였으며 이차계통 각 지점에서 주요 설계변수에 따른 효율분석과 최적화개념 연구를 수행하였다.