• Journal of Korean Society of Industrial and Systems Engineering
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• v.38 no.4
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.12 no.1
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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.

• 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.

• Proceedings of the KSME Conference
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• 2008.11b
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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.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.4 no.2
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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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• v.39 no.2
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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.

• Journal of computational fluids engineering
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• v.15 no.1
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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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.33 no.10
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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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• v.12 no.3
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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.

• Journal of Energy Engineering
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• v.25 no.4
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• pp.85-92
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• 2016

Since the power density of the VHTR(Very High Temperature Reactor) is lower, there is less possibility of core melt. VHTR has no risk of explosion caused by hydrogen generation when the loss of coolant accident occurs, which is another advantage. Along with safety benefit, it can be used as a process heat supplier near demand facilities because coolant temperature is very high enough to be used for industrial purpose. In this paper, we designed the primary system using VHTR and the secondary system providing electricity and process heat. Based on that 350 MW thermal reactor proposed by NGNP(Next Generation Nuclear Part), we developed conceptual model that the IHX(Intermediate Heat Exchanger) loop transports 300 MW thermal energy to the secondary system. In addition, we analyzed thermodynamic behavior and performed the efficiency analysis and optimization study depending on major parameters.