• Nuclear Engineering and Technology
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• v.37 no.6
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• pp.537-556
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• 2005

A primary-pipe rupture accident is one of the design-basis accidents of a High-Temperature Gas-cooled Reactor (HTGR). When the primary-pipe rupture accident occurs, air is expected to enter the reactor core from the breach and oxidize in-core graphite structures. This paper describes an experiment and analysis of the air ingress phenomena and the method fur the prevention of air ingress into the reactor during the primary-pipe rupture accident. The numerical results are in good agreement with the experimental ones regarding the density of the gas mixture, the concentration of each gas species produced by the graphite oxidation reaction and the onset time of the natural circulation of air. A hydrogen production system connected to the High-Temperature Engineering Test Reactor (HTTR) Is being designed to be able to produce hydrogen by themo-chemical iodine-Sulfur process, using a nuclear heat of 10 MW supplied by the HTTR. The HTTR hydrogen production system is first connected to a nuclear reactor in the world; hence a permeation test of hydrogen isotopes through heat exchanger is carried out to obtain detailed data for safety review and development of analytical codes. This paper also describes an overview of the hydrogen permeation test and permeability of hydrogen and deuterium of Hastelloy XR.

• Proceedings of the Membrane Society of Korea Conference
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• 2003.11a
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• pp.25-33
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• 2003

It summarized about the application of the membrane technology in thermochemical water-splitting iodine-sulfur process that was hydrogen production using the nuclear heat from the High Temperature Gas-Cooled Reactor (HTGR). Thermochemical water-splitting hydrogen production method using the high temperater nuclear thermal energy could be realized and remained to be solved the investigation subject. And, it is possible for mass-production of hydrogen such as one of the clean energy in future.

• Proceedings of the KSME Conference
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• 2005.11a
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• pp.117.2-117.2
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• 2005

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.2088-2093
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• 2008

In this study, flow field measurement of the Pebble Bed Reactor(PBR) for the High Temperature Gas-cooled Reactor(HTGR) was performed. Large number of pebbles in the core of PBR provides complicated flow channel. Due to the complicated geometries, numerical analysis has been intensively made rather than experimental observation. However, the justification of computational simulation by the experimental study is crucial to develop solid analysis of design method. In the present study, a wind tunnel installed with pebbles stacked was constructed and equipped with the Particle Image Velocimetry(PIV). We designed the system scaled up to realize the room temperature condition according to the similarity. The PIV observation gave us stagnation points, low speed region so that the suspected high temperature region can be identified. With the further supplementary experimental works, the present system may produce valuable data to justify the Computational Fluid Dynamics(CFD) simulation method.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.1974-1982
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• 2020

Numerical prediction of fission product plateout and circulating coolant activities under normal operating conditions is crucial in the design of a high temperature gas-cooled reactor (HTGR). The results are used for the maintenance and repair of the components as well as the safety analysis regarding early source terms under loss of coolant accident scenarios. In this work, a new computer code named POSCA (Plate-Out Surface and Circulating Activities) was developed based on a one-dimensional model to evaluate fission product plateout and circulating coolant activities within the primary circuit of a HTGR. The verification and validation of study for the POSCA code was done using available analytical results and two in-pile experiments (i.e., OGL-1 and VAMPYR-1). The results of the POSCA calculations show that POSCA is able to simulate plateout and circulating coolant activities in a HTGR with fast computation and reasonable accuracy.

• Membrane Journal
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• v.14 no.3
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• pp.185-191
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• 2004

It summarized about the properties of thermochemical water-splitting iodine-sulfur process that was hydrogen production using the waste heat from the High Temperature Gas-Cooled Reactor (HTGR) recycling the heat of nuclear power. It was mainly explained about the application of membrane separation technique in IS process. Thermochemical water-splitting hydrogen production method using the high temperature nuclear thermal energy could be realized and remained to be solved the investigation subject. And, it is possible for mass-production of hydrogen such as one of the clean energy in future.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.1275-1276
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• 2005

• Corrosion Science and Technology
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• v.18 no.1
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• pp.8-15
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• 2019

A high-temperature gas-cooled reactor (HTGR) is recognized as the best candidate reactor for next generation nuclear reactors. Helium is used to be the coolant in the core of the HTGR with temperature expected to exceed $900^{\circ}C$ at the core outlet. Several iron- and nickel-based superalloys, including Alloy 800H, Hastelloy X, and Alloy 617, are potential structural materials for intermediate heat exchanger (IHX) in an HTGR. Oxidation behaviors of three selected alloys (Hastelloy X, Alloy 800H, and Alloy 617) were investigated at four different temperatures from $650^{\circ}C$ to $950^{\circ}C$ under helium environments with various concentrations of $O_2$ and $H_2O$. Preliminary results showed that chromium oxide as the primary protective layer was observed on surfaces of the three tested alloys. Based on results of mass gain and SEM analyses, Hastelloy X alloy exhibited the best corrosion resistance in all corrosion tests. Further details on the oxidation mechanism of these alloys are presented in this study.

• Nuclear Engineering and Technology
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• v.54 no.8
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• pp.3106-3116
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• 2022

The present work presents analysis results of the core design optimizations for an annular, prismatic High Temperature Gas-cooled Reactor (HTGR) with passive decay-heat removal features. Its thermal power is 100 MW_t and the operating temperature is 850 ℃ (1123 K). The neutronic calculations are done for the core with heterogeneous distribution of fuel and burnable poison particles (BPPs) to flatten the reactivity swing and power peaking factor (PPF) during the reactor operation as well as for control rod (CR) insertion into the core to restrain a small excess reactivity less than 1$. The next step of the study is done for evaluation of core reactivity coefficient of temperature.

• Nuclear Engineering and Technology
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• v.38 no.1
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• pp.45-60
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• 2006

Presently, the VHTGR (Very High Temperature Gas-cooled Reactor) is considered the most attractive candidate for a GEN-IV reactor to produce hydrogen, which will be a key resource for future energy production. A new concept for a reactor cavity cooling system (RCCS), a critical safety feature in the VHTGR, is proposed in the present study. The proposed RCCS consists of passive water pool and active air cooling systems. These are employed to overcome the poor cooling capability of the air-cooled RCCS and the complex cavity structures of the water-cooled RCCS. In order to estimate the licensibility of the proposed design, its performance and integrity were tested experimentally with a reduced-scale mock-up facility, as well as with a separate-effect test facility (SET) for the 1/4 water pool of the RCCS-SNU to examine the heat transfer and pressure drop and code capability. This paper presents the test results for SET and validation of MARS-GCR, a system code for the safety analysis of a HTGR. In addition, CFX5.7, a computational fluid dynamics code, was also used for the code-to-code benchmark of MARS-GCR. From the present experimental and numerical studies, the efficacy of MARS-GCR in application to determining the optimal design of complicated systems such as a RCCS and evaluation of their feasibility has been validated.