• 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.

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3071-3079
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• 2023

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1131-1136
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• 2020

Xenon behaves differently in molten salt reactors (MSRs) compared to solid fuel reactors. This behavior needs exploring due to the large reactivity effect of the ¹³⁵Xe isotope, given the current interest in MSR power plant development for commercial deployment. This paper focuses on select topics in xenon transport, reviews relevant past works, and proposes specific research questions to advance the state of the art in each of the focus areas. Specifically, the paper discusses the issue of xenon solubility in MSRs, the behavior of particulates circulating in MSR fuel salt and its influence on the xenon transport, the possibility of ionization of xenon atoms which changes its effective size and thus affects its mass transport, and finally the issue of circulating void fraction and how it is measured. This work presents specific recommendations for MSR designers to research the limits of Henry's law validity, circulating particulate scrubbers, validity of mass transport coefficients in high radiation fields, and the effects of pump speed on circulating void fraction.

• Nuclear Engineering and Technology
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• v.54 no.11
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• pp.4125-4133
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• 2022

KAERI (Korea Atomic Energy Research Institute) developed the conceptual design of PGSFR (Prototype Gen-IV Sodium Cooled Fast Reactor) and Burner Reactor. Since the reactor characteristics of the PGSFR and Burner Reactor are different, the shape, size and the arrangement of the main components in the reactors must be different. Therefore, the conceptual design for the fuel handling and transfer systems needs to be performed coinciding with the structure of the reactor. Especially, because a redan structure dividing hot and cold pool is installed in the reactor vessel, the conceptual design of the fuel handling and transfer system largely changes depending on the location of the redan structure. Various elements of the conceptual design and an integral arrangement for the fuel handling and transfer system were arranged according to the characteristics, sizes and shapes of the reactors. In this paper, the conceptual designs of the fuel handling and transfer system for PGSFR and Burner Reactor are described. Especially, an A-frame method is selected as the fuel handling and transfer system for the Burner Reactor, considering the layout of the internal structure. The tilt angle, diameter and length of A-frame is determined and the strength evaluation of the A-frame is performed.

• Nuclear Engineering and Technology
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• v.52 no.1
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• pp.51-58
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• 2020

Despite the recent growth of interest in molten salt reactor technology and the crucial role which heat transfer plays in the design of power reactors, specific studies on the design of heat exchangers for the Molten Salt Fast Reactor have not yet been performed. In this work we deliver a preliminary but quantitative analysis of the intermediate heat exchangers, based on reference design data from the SAMOFAR H2020-Euratom project. Two different promising reference technologies are selected for study thanks to their compactness features, the Printed Circuit and the Helical Coil heat exchangers. We present preliminary design results for each technology, based on simplified design tools. Results highlight the limiting effects of the compactness constraints imposed on the fuel salt inventory and the allowed size. Large pressure drops on both flow sides are to be expected, with negative consequences on pumping power and natural circulation capabilities. The small size required for the flow channels also represents possible fabrication issues and safety concerns regarding channel blockage.

• Journal of computational fluids engineering
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• v.21 no.1
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• Transactions of the KSME C: Technology and Education
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• v.3 no.4
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• pp.257-264
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• 2015

To maintain sustainability of nuclear energy as an important energy source, both safety problem and Spent Nuclear Fuels(SNFs) problem should be solved. In case of Gen-IV reactors such as fast reactor, SNFs can be used as fuels by using fast neutrons. It can be a suitable treatment method of high-level waste in near future. Liquid metals such as Sodium or Lead-Bismuth Eutectic (LBE) can be possibly used as a coolant to use fast neutrons. In this paper, it was described that natural circulation parameter studies, design analyses, material selections and a completion of facilities. To develop a natural circulation facility, thermal hydraulic analyses were performed. Installation technique of liquid metal natural circulation were secured.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.38 no.11
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• pp.1257-1264
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• 2014

Efforts are presently underway for developing an optimal design methodology for GEN-IV nuclear reactors based on target failure probabilities. A typical example is the system-based code, in which the results are represented in the form of partial safety factors (PSFs). Thus, a PSF is one of the crucial elements in either component design or integrity assessment based on target failure probabilities during the operation period. In the present study, a procedure for calculating the PSF of a circumferential through-wall cracked pipe based on the elastic-plastic crack initiation criterion is established, in which the importance of each input variable is assessed. Elastic-plastic J-integrals are calculated using the GE/EPRI and reference stress methods, and the PSF values are calculated using both first- and second-order reliability methods. Moreover, the effect of statistical distributions of assessment variables on the PSF is also evaluated.

• Nuclear Engineering and Technology
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• v.52 no.2
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• pp.337-343
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• 2020

Oxide dispersed strengthened (ODS) steel is an important candidate for Gen-IV reactors. Oxide embedded in Fe can help to trap irradiation defects and enhances the strength of steel. It was observed in this study that the size of oxide has a profound impact on the depinning mechanism. For smaller sizes, the oxide acts as a void; thus, letting the dislocation bypass without any shear. On the other hand, oxides larger than 2 nm generate new dislocation segments around themselves. The depinning is similar to that of Orowan mechanism and the strengthening effect is likely to be greater for larger oxides. It was found that higher shear deformation rates produce more fine-tuned stress-strain curve. Both molecular dynamics (MD) simulations and BKS (Bacon-Knocks-Scattergood) model display similar characteristics whereby establishing an inverse relation between the depinning stress and the obstacle distance. It was found that (110)_oxide || (111)_Fe (oriented oxide) also had similar characteristics as that of (100)_oxide || (111)_Fe but resulted in an increased depinning stress thereby providing greater resistance to dislocation bypass. Our simulation results concluded that critical depinning stress depends significantly on the size and orientation of the oxide.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.31-42
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• 2007

Energy supply is increasingly showing up as a major issue for electricity supply, transportation, settlement, and process heat industrial supply including hydrogen production. Nuclear power is part of the solution. For electricity supply, as exemplified in Finland and France, the EPR brings an immediate answer; HTR could bring another solution in some specific cases. For other supply, mostly heat, the HTR brings a solution inaccessible to conventional nuclear power plants for very high or even high temperature. As fossil fuels costs increase and efforts to avoid generation of Greenhouse gases are implemented, a market for nuclear generated process heat will be developed. Following active developments in the 80's, HTR have been put on the back burner up to 5 years ago. Light water reactors are widely dominating the nuclear production field today. However, interest in the HTR technology was renewed in the past few years. Several commercial projects are actively promoted, most of them aiming at electricity production. ANTARES is today AREVA's response to the cogeneration market. It distinguishes itself from other concepts with its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important of which is the design flexibility to adapt readily to combined heat and power applications. From the start, AREVA made the choice of such flexibility with the belief that the HTR market is not so much in competition with LWR in the sole electricity market but in the specific added value market of cogeneration and process heat. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source tree of Greenhouse gases emissions. The ANTARES module produces 600 MWth which can be split into the required process heat, the remaining power drives an adapted prorated electric plant. Depending on the process heat temperature and power needs, up to 80% of the nuclear heat is converted into useful power. An important feature of the design is the standardization of the heat source, as independent as possible of the process heat application. This should expedite licensing. The essential conditions for success include: ${\bullet}$ Timely adapted licensing process and regulations, codes and standards for such application and design ${\bullet}$ An industry oriented R&D program to meet the technological challenges making the best use of the international collaboration. Gen IV could be the vector ${\bullet}$ Identification of an end user(or a consortium of) willing to fund a FOAK