• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3969-3980
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• 2024

Over the years, numerous evaluations of material attractiveness have been performed for conventional light water reactors to better understand the nature of the spent fuel material and its desirability for misuse at different points in the nuclear fuel cycle. However, availability of such assessments for newer, Generation IV reactors such as Molten Salt Reactors is rather limited. In the present study we address the gap in knowledge of material attractiveness for molten salt reactor systems and describe the nature of irradiated fuel salts which the nuclear safeguards community might be faced with in the near future as more and more such reactors enter commission and operation. Within the scope of the paper, we use a large database of simulated irradiated fuel salt isotopics (and other derived quantities such as gamma activity, decay heat, and neutron emission rates) developed specifically for a molten salt reactor concept in order to shed some light on possible weapons usability of uranium and plutonium present in the irradiated fuel salts. This has been achieved by proposing a new attractiveness metric that is better suited for quantifying attractiveness of irradiated salts from a model molten salt concept. The said metric has been computed using a database that has been created by simulating the irradiation of molten fuel salt in a concept core over a wide range of operational parameters (burnup, initial enrichment, and cooling time) using the Monte-Carlo particle transport code, Serpent. With the help of this attractiveness metric, the findings from this study have shown that in relative terms, molten salt spent fuel is more attractive than spent fuel produced by a conventional light water reactor. The findings also underscore the need for strengthened safeguards measures for such spent fuel. These results are expected to be useful in the future for regulatory authorities as well as for nuclear safeguards inspectors for designing a functional safeguards verification routine for irradiated fuel of such unique nature.

• Tribology and Lubricants
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• v.40 no.3
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• pp.97-102
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• 2024

The next-generation Molten Salt Reactor is known for its high safety because it uses nuclear fuel dissolved in high-temperature molten salt, unlike traditional solid atomic fuel methods. However, the high-temperature molten salt causes severe corrosion in internal structural materials, threatening the reactor's safety. Therefore, it is crucial to investigate the high-temperature corrosion resistance and wear performance of materials used in reactors to ensure safety. In this study, the high-temperature corrosion resistances and wear performances of corrosion samples in a NaCl-MgCl2-KCl (20-40-40 [wt%]) molten salt are investigated to evaluate the applicability of economically viable stainless steels, 316SS and 304SS. Hastelloy C276 and a new alloy containing a small amount of Nb are used as reference samples for comparative analysis. The mass loss, mass loss rate per unit volume, and surface roughness of each sample are measured to understand the corrosion mechanisms. Scanning electron microscopy and energy-dispersive spectroscopy analyses are employed to analyze the corrosion mechanisms. Wear tests on the corroded samples are also conducted to assess the extent of corrosion. Based on the experimental results, we predict the lifespans of the materials and evaluate their suitability as candidate materials for molten salt reactors. The data obtained from the experiments provide a valuable database for structural materials that can enhance the stability of molten salt reactors and recommend high-temperature corrosion-resistant materials suitable for next-generation reactors.

• 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.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.20 no.1
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• pp.43-63
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• 2022

The aim of this review is to communicate some essential knowledge of the underlying mechanism of the corrosion of structural containment alloys during molten salt reactor operation in the context of prospective online monitoring in future MSR installations. The formation of metal halide species and the progression of their concentration in the molten salt do reflect containment corrosion, tracing the depletion of alloying metals at the alloy salt interface will assure safe conditions during reactor operation. Even though the progress of alloying metal halides concentrations in the molten salt do strongly understate actual corrosion rates, their prospective 1^st order kinetics followed by near-linearly increase is attributed to homogeneous matrix corrosion. The service life of the structural containment alloy is derived from homogeneous matrix corrosion and near-surface void formation but less so from intergranular cracking (IGC) and pitting corrosion. Online monitoring of corrosion species is of particular interest for molten chloride systems since besides the expected formation of chromium chloride species CrCl₂ and CrCl₃, other metal chloride species such as FeCl₂, FeCl₃, MoCl₂, MnCl₂ and NiCl₂ will form, depending on the selected structural alloy. The metal chloride concentrations should follow, after an incubation period of about 10,000 hours, a linear projection with a positive slope and a steady increase of < 1 ppm per day. During the incubation period, metal concentration show 1^st order kinetics and increasing linearly with time^1/2. Ideally, a linear increase reflects homogeneous matrix corrosion, while a sharp increase in the metal chloride concentration could set a warning flag for potential material failure within the projected service life, e.g. as result of intergranular cracking or pitting corrosion. Continuous monitoring of metal chloride concentrations can therefore provide direct information about the mechanism of the ongoing corrosion scenario and offer valuable information for a timely warning of prospective material failure.

• Nuclear Engineering and Technology
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• v.53 no.5
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• pp.1569-1579
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• 2021

This study aims to provide microstructural characterization for the matrix graphite which molten salt reactors (MSRs) use, and improve resistance to molten salt infiltration of the matrix graphite for fuel elements. Mesocarbon microbeads (MCMB) densified matrix graphite A3-3 (MDG) was prepared by a quasi-isostatic pressure process. After densification by MCMBs with average particle sizes of 2, 10, and 16 ㎛, the pore diameter of A3-3 decreased from 924 nm to 484 nm, 532 nm, and 778 nm, respectively. Through scanning electron microscopy, the cross-section energy spectrum and time-of-flight secondary ion mass spectrometry, resistance levels of the matrix graphite to molten salt infiltration were analyzed. The results demonstrate that adding a certain proportion of MCMB powders can improve the anti-infiltration ability of A3-3. Meanwhile, the closer the particle size of MCMB is to the pore diameter of A3-3, the smaller the average pore diameter of MDG and the greater the densification. As a matrix graphite of fuel elements in MSR was involved, the thermal and mechanical properties of matrix graphite MDG were also studied. When densified by the MCMB matrix graphite, MDGs can meet the molten salt anti-infiltration requirements for MSR operation.

• Journal of Energy Engineering
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• v.17 no.1
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• pp.23-30
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• 2008

The neutron kinetic analysis methods for the molten-salt reactors are quite different from those for conventional solid-fuel reactors, which do not take into account the flowing-fuel-induced neutronics effects. Therefore, for dynamics and safety analyses of the molten-salt reactor systems, the conventional kinetics codes would not be appropriate to accurately predict its transient behaviors. A point-kinetics with flowing- fuel model has been used to assess the fluid-fuel reactor system safety, but recognized as not to be sufficient to simulate spatial distributions of delayed-neutron precursors and neutron populations during transients for given detail reactor models. In order to meet this requirement, AMBIKIND, a 2-group, 2-dimensional neutron kinetics code suitable for the molten-salt reactor systems was developed. This paper explains the code's theoretical and numerical descriptions and, as a part of its verification, includes some simulation results of MSRE stability experiments. Even though the present reactor model does not include the recirculation effect of the fuel-salt through the reactor system, the AMBIKIN2D code should be able to predict the power and phase shift at various power levels and reactivity insertions with better accuracy.

• 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.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.73-78
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• 2010

The suppression of molten salt vaporization is one of the key technical issues in the electrolytic reduction process developed for recycling spent nuclear fuel from light-water reactors Since the Hertz-Langmuir relation previously applied to molten salt vaporization is valid only for vaporization into a vacuum, a diffusion model was derived to quantitatively assess the vaporization of LiCl, $Li_2O$ and Li from an electrolytic reducer operating under atmospheric pressure. Vaporization rates as a function of operation variables were calculated and shown to be in reasonable agreement with the experimental data obtained from thermogravimetry.

• Nuclear Engineering and Technology
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• v.56 no.4
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• pp.1135-1144
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• 2024

With the increasing ratio of renewables in the grid, a low-carbon and stable base load source that also is capable of load tracking is in demand. Sodium cooled fast reactors (SFRs) coupled to thermal heat storage system (THS) is a strong candidate for the need. This research focuses on the designing and performance validation of a two-tank THS based on molten salt to integrate with a 280 MW_th sodium cooled fast reactor. Designing of the THS includes the vital component, sodium-to-salt heat exchanger which is a technology gap that needs to be filled, and designing and parameter selection of the tanks and related pumps. Modeling of the designed THS is conducted followed by the description of operation strategies and control logics of the THS. Finally, the dynamic simulation of the designed THS is conducted based on Fortran. Results show, the proposed power system meets the need of the design requirements to store heat for 18 h during a day and provide 500 MW_th for peak demand for the rest of the day.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2315-2324
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• 2023

Space nuclear reactors are becoming popular in deep space exploration owing to their advantages of high-power density and stability. Following the fourth-generation nuclear reactor technology, a conceptual design of the dual drum-controlled space molten salt reactor (D²-SMSR) is proposed. The reactor concept uses molten salt as fuel and heat pipes for cooling. A new reactivity control strategy that combines control drums and safety drums was adopted. Critical physical characteristics such as neutron energy spectrum, neutron flux distribution, power distribution and burnup depth were calculated. Flow and heat transfer characteristics such as natural convection, velocity and temperature distribution of the D²-SMSR under low gravity conditions were analyzed. The reactivity control effect of the dual-drums strategy was evaluated. Results showed that the D²-SMSR with a fast spectrum could operate for 10 years at the full power of 40 kWth. The D²-SMSR has a high heat transfer coefficient between molten salt and heat pipe, which means that the core has a good heat-exchange performance. The new reactivity control strategy can achieve shutdown with one safety drum or three control drums, ensuring high-security standards. The present study can provide a theoretical reference for the design of space nuclear reactors.