• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.9-20
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• 2007

Design study on the Gas Turbine High Temperature Reactor 300-Cogeneration (GTHTR300C) aiming at producing both electricity by a gas turbine and hydrogen by a thermochemical water splitting method (IS process method) has been conducted. It is expected to be one of the most attractive systems to provide hydrogen for fuel cell vehicles after 2030. The GTHTR300C employs a block type Very High Temperature Reactor (VHTR) with thermal power of 600MW and outlet coolant temperature of $950^{\circ}C$. The intermediate heat exchanger (IHX) and the gas turbine are arranged in series in the primary circuit. The IHX transfers the heat of 170MW to the secondary system used for hydrogen production. The balance of the reactor thermal power is used for electricity generation. The GTHTR300C is designed based on the existing technologies of the High Temperature Engineering Test Reactor (HTTR) and helium turbine power conversion and on the technologies whose development have been well under way for IS hydrogen production process so as to minimize cost and risk of deployment. This paper describes the original design features focusing on the plant layout and plant cycle of the GTHTR300C together with present development status of the GTHTR300, IHX, etc. Also, the advantage of the GTHTR300C is presented.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• 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.

• Journal of Radiation Protection and Research
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• v.33 no.3
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• pp.113-120
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• 2008

In this study, The TRITGO code was introduced, which can predict the amount of tritium production, it's transport, removal, distribution and the level of contamination for the produced hydrogen by the tritium on the VHTR (very high temperature gas cooled reactor). The TRITGO code was improved so that the permeation to the IS Iodine Sulfide) loop for producing the hydrogen can be simulated. The contamination level of the produced hydrogen by the tritium was predicted by the improved code for the VHTR with 600MW thermal power. The contamination level for the produced hydrogen by tritium was predicted as 0.055 Bq/$H_2-g$. This level is three order of lower than the regulation value of 56 Bq/$H_2-g$ from Japan. From this study, the following results were obtained. it is important that the fuel coating (SiC layer) should be kept intact to prevent the tritium from releasing. Also it is necessary that the level of impurity such as 3He and Li in the helium coolant and the reflector consisting of the graphite should be kept as low as possible. It was found that the capacity of the purification system for filtering the impurities directly from the coolant will be the important design parameter.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.443-444
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• 2005

• Corrosion Science and Technology
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• v.13 no.5
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• pp.178-185
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• 2014

Alloy 617 is considered as a candidate Ni-based superalloy for the intermediate heat exchanger (IHX) of a very high-temperature gas reactor (VHTR) because of its good creep strength and corrosion resistance at high temperatures. Helium is used as a coolant in a VHTR owing to its high thermal conductivity, inertness, and low neutron absorption. However, helium inevitably includes impurities that create an imbalance in the surface reactivity at the interface of the coolant and the exposed materials. As the Alloy 617 has been exposed to high temperatures at $950^{\circ}C$ in the impure helium environment of a VHTR, the degradation of material is accelerated and mechanical properties decreased. The high-temperature strength, creep, and corrosion properties of the structural material for an IHX are highly important to maintain the integrity in a harsh environment for a 60 year period. Therefore, an alloy superior to alloy 617 should be developed. In this study, the mechanical and high-temperature corrosion properties for Ni-Cr alloys fabricated in the laboratory were evaluated as a function of the grain boundary strengthening and alloying elements. The ductility increased and decreased by increasing the amount of Mo and Cr, respectively. Surface oxide was detached during the corrosion test, when Al was not added to alloy. However the alloy with Al showed improved oxide adhesive property without significant degradation and mechanical property. Aluminum seems to act as an anti-corrosive role in the Ni-based alloy.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.35 no.10
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• pp.1249-1255
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• 2011

A PHE (Process Heat Exchanger) is a key component required to transfer heat energy of $950^{\circ}C$ generated in a VHTR (Very High Temperature Reactor) to a chemical reaction that yields a large quantity of hydrogen. A small-scale PHE prototype made of Hastelloy-X was scheduled for testing in a small-scale gas loop at the Korea Atomic Energy Research Institute. In this study, as a part of the evaluation of the high-temperature structural integrity of the PHE prototype, high-temperature structural analysis modeling, and macroscopic thermal and elastic-plastic structural analysis of the PHE prototype were carried out under the gas-loop test conditions as a preliminary qwer123$study before carrying out the performance test in the gas loop. The results obtained in this study will be used to design the performance test setup for the modified PHE prototype.

• Transactions of the Korean Society of Mechanical Engineers B
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• v.34 no.12
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• pp.1093-1099
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• 2010

For Very High Temperature Reactors (VHTR), the designs of the Intermediate Heat Transport Loop (IHTL) and the Intermediate Heat Exchanger (IHX) are particularly difficult because of the high-temperature operation (up to $950^{\circ}C$). In this study, Flinak molten salt, a eutectic mixture of LiF, NaF, and KF (46.5:11.5:42.0 mole %) is considered as the heat transporting fluid in the IHTL. To evaluate the flow and heat transfer performance of the Flinak molten salt in small channels with hydraulic diameters in the millimeter range, a double-pipe heat exchanger was constructed using small-diameter tubes for the heat exchange between the Flinak and the gas flow. The experimental data showed that, for laminar Flinak flow, the measured friction factors were close to the 64/Re curve and the Nusselt numbers were generally between 3.66 and 4.36.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.36 no.10
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• pp.1283-1288
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• 2012

A process heat exchanger (PHE) in a nuclear hydrogen system is a key component for transferring the considerable heat generated in a very high temperature reactor (VHTR) to a chemical reaction that yields a large quantity of hydrogen. A performance test on a medium-scale PHE prototype made of $Hastelloy^{(R)}$-X is scheduled in a small-scale gas loop at the Korea Atomic Energy Research Institute. In this study, as a preliminary study before carrying out the performance test in the gas loop, high-temperature structural analysis modeling and macroscopic thermal and structural analysis of the medium-scale PHE prototype by imposing the established displacement boundary constraints were carried out under the gas loop test condition. The results obtained in this study will be compared with the performance test results of the medium-scale PHE prototype in the gas loop.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.10 no.1
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• pp.64-69
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• 2014

Gr. 91 steel is used for the major structural components of Generation-IV reactor systems, such as a very high temperature reactor(VHTR) and sodium-cooled fast reactor(SFR). Since these structures are designed for up to 60 years at elevated temperatures, the prediction of long-term creep life is important for a design application of Gr. 91 steel. In this study, a number of creep rupture data were collected through world-wide literature surveys, and using these data, the long-term creep life was predicted in terms of three methods: the single-C method in Larson-Miller(L-M) parameter, multi-C constant method in the L-M parameter, and a modified method("sinh" equation) in the L-M parameter. The results of the creep-life prediction were compared using the standard deviation of error value, respectively. Modified method proposed by the "sinh" equation revealed better agreement in creep life prediction than the single-C L-M method.