• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3450-3463
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• 2023

In the two-step analysis of Pebble-Bed type High-Temperature Gas-Cooled Reactor (PB-HTGR), the lattice physics calculation for the generation of homogenized cross-sections is based on the fuel pebble. However, the randomly-dispersed fuel particles in the fuel pebble introduce double heterogeneity and randomness. Compared to the deterministic method, the Monte Carlo method which is flexible in geometry modeling provides a high-fidelity treatment. Therefore, the Monte Carlo code NECP-MCX is extended in this study to perform the lattice physics calculation of the PB-HTGR. Firstly, the capability for the simulation of randomly-dispersed media, using the explicit modeling approach, is developed in NECP-MCX. Secondly, the capability for the generation of the homogenized cross-section is also developed in NECP-MCX. Finally, simplified PB-HTGR problems are calculated by a two-step neutronics analysis tool based on Monte Carlo homogenization. For the pebble beds mixed by fuel pebble and graphite pebble, the bias is less than 100 pcm when compared to the high-fidelity model, and the bias is increased to 269 pcm for pebble bed mixed by depleted fuel pebble. Numerical results show that the Monte Carlo lattice physics calculation for the two-step analysis of PB-HTGR is feasible.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.103-110
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• 2007

The High Temperature Gas-cooled Reactor (HTGR) possesses inherent safety features and is recognized as a representative advanced nuclear system for the future. Based on the success of the HTR-10, the long-time operation test and safety demonstration tests were carried out. The long-time operation test verifies that the operation procedure and control method are appropriate for the HTR-10 and the safety demonstration test shows that the HTR-10 possesses inherent safety features with a great margin. Meanwhile, two new projects have been recently launched to further develop HTGR technology. One is a prototype modular plant, denoted as HTR-PM, to demonstrate the commercial capability of the HTGR power plant. The HTR-PM is designed as $2{\times}250$ MWt, pebble bed core with a steam turbine generator that serves as an energy conversion system. The other is a gas turbine generator system coupled with the HTR-10, denoted as HTR-10GT, built to demonstrate the feasibility of the HTGR gas turbine technology. The gas turbine generator system is designed in a single shaft configuration supported by active magnetic bearings (AMB). The HTR-10GT project is now in the stage of engineering design and component fabrication. R&D on the helium turbocompressor, a key component, and the key technology of AMB are in progress.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1310-1319
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• 2024

The reactor cavity cooling system (RCCS) is a passive reactor safety system commonly present in the designs of High-Temperature Gas-cooled Reactors (HTGR) that removes heat from the reactor pressure vessel by means of natural convection and radiation. It is one of the factors responsible for ensuring that the reactor does not melt down under any plausible accident scenario. For the simulation of accident scenarios, which are transient phenomena unfolding over a span of up to several days, intermediate fidelity methods and system codes must be employed to limit the models' execution time. These models can quantify radiation heat transfer well, but heat transfer caused by natural convection must be quantified with the use of correlations for the heat transfer coefficient. It is difficult to obtain reliable correlations for HTGR RCCS heat transfer coefficients experimentally due to such a system's size. They could, however, be obtained from high-fidelity steady-state simulations of RCCSs. The Rayleigh number in RCCSs is too high for using a Direct Numerical Simulation (DNS) technique; thus, a Reynolds-Averaged Navier-Stokes (RANS) approach must be employed. There are many RANS models, each performing best under different geometry and fluid flow conditions. To find the most suitable one for simulating an RCCS, the RANS models need to be validated. This work benchmarks various RANS models against three experiments performed on the HTTR RCCS Mockup by the Japanese Atomic Energy Agency (JAEA) in 1993. This facility is a 1/6 scale model of a vessel cooling system (VCS) for the High Temperature Engineering Test Reactor (HTTR), which is operated by JAEA. Multiple RANS models were evaluated on a simplified 2d-axisymmetric geometry. They were found to reproduce the experimental temperature profiles with errors of up to 22% for the lowest temperature benchmark and 15% for the higher temperature benchmarks. The results highlight that the pragmatic turbulence models need to be validated for high Rayleigh natural convection-driven flows and improved accordingly, more publicly available experimental data of RCCS resembling experiments is needed and indicate that a 2d-axisymmetric geometry approximation is likely insufficient to capture all the relevant phenomena in RCCS simulations.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1087-1100
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• 2009

A finite element method utilizing the Galerkin form of the weighted residuals procedure was developed to estimate the mechanical behavior for a coated fuel particle (CFP) of a high temperature gas-cooled reactor (HTGR). Through a weak formulation, finite element equations for multiple layers were set up to calculate the displacements and stresses in a CFP. The finite element method was applied to the stress analyses for three coating layers of a tri-isotropic coated fuel particle (TRISO) of a HTGR. The stresses calculated by the finite element method were in good agreement with those from a previously developed computer code and depicted the typical stress behavior of the coating layers very well. The newly developed finite element method performs a stress analysis for multiple bonded layers in a CFP by changing the material properties at any position in the layers during irradiation.

• Nuclear Engineering and Technology
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• v.36 no.5
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• pp.375-387
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• 2004

This paper presents a transient multicomponent mixture analysis tool developed to analyze the molecular diffusion, natural convection, and chemical reactions related to air ingress phenomena that occur during a primary-pipe rupture of a high temperature gas-cooled reactor (HIGR). The present analysis tool solves the one-dimensional basic equations for continuity, momentum, energy of the gas mixture, and the mass of each gas species. In order to obtain numerically stable and fast computations, the implicit continuous Eulerian scheme is adopted to solve the governing equations in a strongly coupled manner. Two types of benchmark calculations were performed with the data of prerious Japanese inverse U-tube experiments. The analysis program, based on the ICE technique, runs about 36 times faster than the FLUENT6 for the simulation of the two experiments. The calculation results are within a 10% deviation from the experimental data regarding the concentrations of the gas species and the onset times of natural convection.

• Nuclear Engineering and Technology
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• v.45 no.6
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• pp.745-752
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• 2013

For thermo-fluid and safety analyses of a High Temperature Gas-cooled Reactor (HTGR), intensive efforts are in progress in the developments of the GAMMA+ code of Korea Atomic Energy Research Institute (KAERI) and the AGREE code of the University of Michigan (U of M). One of the important requirements for GAMMA+ and AGREE is an accurate modeling capability of a bypass flow in a prismatic core. Recently, a series of air experiments were performed at Seoul National University (SNU) in order to understand bypass flow behavior and generate an experimental database for the validation of computer codes. The main objective of the present work is to validate the GAMMA+ and AGREE codes using the experimental data published by SNU. The numerical results of the two codes were compared with the measured data. A good agreement was found between the calculations and the measurement. It was concluded that GAMMA+ and AGREE can reliably simulate the bypass flow behavior in a prismatic core.

• Nuclear Engineering and Technology
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• v.41 no.4
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• pp.413-426
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• 2009

Nuclear energy plays an important role in world energy production by supplying 6% of the world's current total electricity production. However, 86% of the energy consumed worldwide to produce industrial process heat, to generate electricity and to power the transportation sector still originates in fossil fuels. To cope with dwindling fossil fuels and climate change, it is clear that a clean alternative energy that can replace fossil fuels in these sectors is urgently required. Clean hydrogen energy is one such alternative. Clean hydrogen can play an important role not only in synthetic fuel production but also through powering fuel cells in the anticipated hydrogen economy. With the introduction of the high temperature gas-cooled reactor (HTGR) that can produce nuclear heat up to $950^{\circ}C$ without greenhouse gas emissions, nuclear power is poised to broaden its mission beyond electricity generation to the provision of nuclear process heat and the massive production of hydrogen. In this paper, the features and potential of the HTGR as the energy source of the future are addressed. Perspectives on nuclear heat and hydrogen applications using the HTGR are discussed.

• Journal of Advanced Marine Engineering and Technology
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• v.21 no.4
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• pp.372-380
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• 1997

This paper describes experimental investigations of helium-air exchange flows through parti¬tioned opening and two-opening. Such exchange flows may occur following rupture accident of stand pipe in high temperature gas cooled reactor. A test vessel with the two types of small open¬ing on top of test cylinder is used for experiments. An estimation method of mass increment is developed and applied to measure the exchange flow rate. A technique of flow visualization by Mach-Zehnder interferometer is provided to recognize the exchange flows. In the case of exchange flow through the partitioned opening, flow passages of upward flow of the helium and downward flow of the air within the opening are separated by vertical partition, and the two flows interact out of entrance and exit of the opening. Therefore, an experiment of the exchange flow through two-opening is made to investigate effect of the fluids interaction of the partitioned opening sys¬tem. As a result of comparison of the exchange flow rates between the two types of the opening system, it is found that the exchange flow rate of the two-opening system is larger than that of the partitioned opening system due to absence of the effect of fluids interaction. Finally, the fluids interaction between the upward and downward flows through the partitioned opening is found to be an important factor on the helium-air exchange flow.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.14 no.1
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• pp.8-14
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• 2018

Printed circuit heat exchangers (PCHEs) are widely used with an increasing demand for industrial applications. PCHEs are capable of operating at high temperatures and pressure. We consider a PCHE as a candidate intermediate heat exchanger type for a high temperature gas-cooled reactor (HTGR). For conventional application using stainless steels, design and manufacturing of PCHEs are well established. For applications to HTGR, knowledge of longitudinal conduction and deformation of channel is required to estimate design margin. This paper analyzes the effects of longitudinal conduction and deformation of channel on thermal performance using a code internally developed for design and analysis of PCHEs. The code has a capability of two dimensional simulations. Longitudinal conduction is estimated using the code. In HTGR operating condition, about ten percent of design margin is required to compensate thermal performance. The cross-sectional images of PCHE channels are obtained using an optical microscope. The images are processed with computer image process technique. We quantify the deformation of channel with dimensional parameters. It is found that the deformation has negative effect on structural integrity. The deformation enhances thermal performance when the shape of channel is straight in laminar flow regime. It reduces thermal performance in cases of a zigzag channel and turbulent flow regime.

• Nuclear Engineering and Technology
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• v.56 no.5
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• pp.1679-1686
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• 2024

Graphite material plays an important role in nuclear reactors especially the high-temperature gas-cooled reactors (HTGRs) by its outstanding comprehensive nuclear properties. The structural integrity of graphite pebble fuel elements is the first barrier to core safety under any circumstances. The correct knowledge of the stiffness coefficient of the graphite pebble fuel element inside the reactor's core is significant to ensure the valid design and inherent safety. In this research, a vertical extrusion device was set up to measure the stiffness coefficient of the graphite pebble fuel element by the Institute of Nuclear and New Energy Technology (INET) of Tsinghua University in China. The stiffness coefficient equations of graphite pebble fuel elements at different temperatures are given (in a helium atmosphere). The result first provides the data on the high-temperature stiffness coefficient of pebbles in helium gas. The result will be helpful for the engineering safety analysis of pebble-bed nuclear reactors.