• Proceedings of the Korean Nuclear Society Conference
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• 2004.05a
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• pp.65.1-65.1
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• 2004

• Nuclear Engineering and Technology
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• v.51 no.4
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• pp.1024-1031
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• 2019

The Molten Salt Fast Reactor (MSFR) with its liquid circulating fuel and its fast neutron spectrum calls for a new safety approach including technological neutral methodologies and analysis tools adapted to early design phases. In the frame of the Horizon2020 program SAMOFAR (Safety Assessment of the Molten Salt Fast Reactor) a safety approach suitable for Molten Salt Reactors is being developed and applied to the MSFR. After a description of the MSFR reference design, this paper focuses on the identification of the Postulated Initiating Events (PIEs), which is a core part of the global assessment methodology. To fulfil this task, the Functional Failure Mode and Effect Analysis (FFMEA) and the Master Logic Diagram (MLD) are selected and employed separately in order to be as exhaustive as possible in the identification of the initiating events of the system. Finally, an extract of the list of PIEs, selected as the most representative events resulting from the implementation of both methods, is presented to illustrate the methodology and some of the outcomes of the methods are compared in order to highlight symbioses and differences between the MLD and the FFMEA.

• 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.

• Nuclear Engineering and Technology
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• v.54 no.5
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• pp.1560-1569
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• 2022

Liquid fueled molten salt reactors (MSRs) have seen renewed interest because of their inherent safety features, higher thermal efficiency and potential for efficient thorium utilisation for power generation. Thorium fluoride is one of the salts used in liquid fueled MSRs employing Th-U cycle. In the present study, ThF₄ was prepared by hydro-fluorination of ThO₂ using anhydrous HF gas. Process parameters viz. bed depth, hydrofluorination time and hydrofluorination temperature, were optimized for the preparation of ThF₄ in a static bed reactor setup. The products were characterized with X-Ray diffraction and experimental conditions for complete conversion to ThF₄ were established which also corroborated with the yield values. Hydrofluorination of ThO₂ at 450 ℃ for half an hour at a bed depth of 6 mm gave the best result, with a yield of about 99.36% ThF₄. No unconverted oxide or any other impurity was observed. Rietveld refinement was performed on the XRD data of this ThF₄, and Chi2 value of 3.54 indicated good agreement between observed and calculated profiles.

• Nuclear Engineering and Technology
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• v.54 no.12
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• pp.4431-4440
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• 2022

An advanced nuclear reactor based on molten salts including a molten salt reactor and pyroprocessing needs a sensitive monitoring system suitable for operation in harsh environments with limited access. Multi-element detection is challenging with the conventional technologies that are compatible with the in-situ operation; hence laser-induced breakdown spectroscopy (LIBS) has been investigated as a potential alternative. However, limited precision is a chronic problem with LIBS. We increased the precision of LIBS under high temperature by protecting optics using a gas protective layer and correcting for shotto-shot variance and lens-to-sample distance using a laser-induced acoustic signal. This study investigates cerium as a surrogate for uranium and corrosion products for simulating corrosive environments in LiCl-KCl. While the un-corrected limit of detection (LOD) range is 425-513 ppm, the acoustic-corrected LOD range is 360-397 ppm. The typical cerium concentrations in pyroprocessing are about two orders of magnitude higher than the LOD found in this study. A LIBS monitoring system that adopts these methods could have a significant impact on the ability to monitor and provide early detection of the transient behavior of salt composition in advanced molten salt-based nuclear reactors.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1005-1014
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• 2009

The Dual Unit Optimizer 2000 MWe (DUO2000) is put forward as a new design concept for large power nuclear plants to cope with economic and safety challenges facing the $21^{st}$ century green and sustainable energy industry. DUO2000 is home to two nuclear steam supply systems (NSSSs) of the Optimized Power Reactor 1000 MWe (OPR1000)-like pressurized water reactor (PWR) in single containment so as to double the capacity of the plant. The idea behind DUO may as well be extended to combining any number of NSSSs of PWRs or pressurized heavy water reactors (PHWRs), or even boiling water reactors (BWRs). Once proven in water reactors, the technology may even be expanded to gas cooled, liquid metal cooled, and molten salt cooled reactors. With its in-vessel retention external reactor vessel cooling (IVR-ERVC) as severe accident management strategy, DUO can not only put the single most querulous PWR safety issue to an end, but also pave the way to very promising large power capacity while dispensing with the huge redesigning cost for Generation III+ nuclear systems. Five prototypes are presented for the DUO2000, and their respective advantages and drawbacks are considered. The strengths include, but are not necessarily limited to, reducing the cost of construction by decreasing the number of containment buildings from two to one, minimizing the cost of NSSS and control systems by sharing between the dual units, and lessening the maintenance cost by uniting the NSSS, just to name the few. The latent threats are discussed as well.

• Nuclear Engineering and Technology
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• v.52 no.6
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• pp.1243-1251
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• 2020

Irradiation-induced damage of binderless nanoporous-isotropic graphite (NPIG) prepared by isostatic pressing of mesophase carbon microspheres for molten salt reactor was investigated by 3.0 MeV He⁺ irradiation at room temperature and high temperature of 600 ℃, and IG-110 was used as the comparation. SEM, TEM, X-ray diffraction and Raman spectrum are used to characterize the irradiation effect and the influence of temperature on graphite radiation damage. After irradiation at room temperature, the surface morphology is rougher, the increase of defect clusters makes atom flour bend, the layer spacing increases, and the catalytic graphitization phenomenon of NPIG is observed. However, the density of defects in high temperature environment decreases and other changes are not obvious. Mechanical properties also change due to changes in defects. In addition, SEM and Raman spectra of the cross section show that cracks appear in the depth range of the maximum irradiation dose, and the defect density increases with the increase of irradiation dose.

• 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.54 no.3
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• pp.817-825
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• 2022

A recent benchmarking effort, under the auspices of the Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA), has been made to evaluate the current state of modeling and simulation tools available to model fluoride salt-cooled high temperature reactors (FHRs). The FHR benchmarking effort considered in this work consists of several cases evaluating the neutronic parameters of a 2D prismatic FHR fuel assembly model using the participants' choice of simulation tools. Benchmark participants blindly submitted results for comparison with overall good agreement, except for some which significantly differed on cases utilizing a molybdenum-bearing control rod. Participants utilizing more recently updated explicit isotopic cross sections had consistent results, whereas those using elemental molybdenum cross sections observed reactivity differences on the order of thousands of pcm relative to their peers. Through a series of supporting tests, the authors attribute the differences as being nuclear data driven from using older legacy elemental molybdenum cross sections. Quantitative analysis is conducted on the control rod to identify spectral, reaction rate, and cross section phenomena responsible for the observed differences. Results confirm the observed differences are attributable to the use of elemental cross sections which overestimate the reaction rates in strong resonance channels.