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Recent Advances in Materials for Molten Salt Nuclear Reactor Technology
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Recent Advances in Materials for Molten Salt Nuclear Reactor Technology

A special issue of Materials (ISSN 1996-1944). This special issue belongs to the section "Energy Materials".

Deadline for manuscript submissions: closed (20 December 2024) | Viewed by 12897

Special Issue Editors

Dr. Anna Smith

E-Mail Website
Guest Editor
Radiation Science and Technology Department, Faculty of Applied Sciences, Delft University of Technology, Mekelweg 15, 2629 JB Delft, The Netherlands
Interests: nuclear materials chemistry; experimental and computational thermodynamics; molten salts; thermochemical and thermophysical properties; structure-property relationships
Dr. Aimen Gheribi

E-Mail Website
Guest Editor
Department of Chemical Engineering, Centre for Research in Computational Thermochemistry, Ecole Polytechnique Montreal, C.P. Succursale “Downtown”, Montreal, Quebec H3C 3A7, Canada
Interests: multiscale thermodynamics applied to material science; atomistic simulations of liquid; solid and amorphous solutions; physico-chemical properties of melts at high temperature; molten salts/electrolytes

Special Issue Information

Dear Colleagues,

Molten salts are receiving increasing attention worldwide as key materials for sustainable and low-carbon energy technologies in relation with their appealing thermochemical and thermophysical properties. Areas of technological application include, for instance, thermal energy storage and concentrated solar power systems, nuclear fission and fusion reactors, the electrochemical recycling of spent nuclear fuel, electrolytes for fuel cells and molten metal batteries, etc. In particular, the interest in nuclear molten salt reactor (MSR) technology, where molten salts are used both for the nuclear fuel and coolant materials, is growing very rapidly.

The development of the MSR technology towards commercialization requires a thorough safety analysis of all components of the reactor during operation and accident conditions, especially of the molten salt fuel, coolant, and structural materials that are subject to extreme conditions (e.g., high temperatures, irradiation, and corrosive environment) during reactor operation, as well as their interaction.

We invite investigators to contribute original research or review articles reporting recent advances in the materials developed for MSR designs, including fuel and coolant salt synthesis, thermochemical and thermophysical properties, phase equilibria, thermodynamic modelling assessments, modelling and simulation of transport properties (e.g., density, viscosity, thermal conductivity), performance of structural materials with respect to corrosion at high temperature and radiation damage, unraveling of the degradation processes etc. This Special Issue aims especially at highlighting the relationships between structure and properties in the aforementioned research areas. Papers including modelling, simulation and experimental studies are all welcome.

It is our pleasure to invite you to submit a manuscript for this Special Issue. Full papers, communications, and reviews are all welcome.

Dr. Anna Smith
Dr. Aimen Gheribi
Guest Editors

Manuscript Submission Information

Manuscripts should be submitted online at www.mdpi.com by registering and logging in to this website. Once you are registered, click here to go to the submission form. Manuscripts can be submitted until the deadline. All submissions that pass pre-check are peer-reviewed. Accepted papers will be published continuously in the journal (as soon as accepted) and will be listed together on the special issue website. Research articles, review articles as well as short communications are invited. For planned papers, a title and short abstract (about 100 words) can be sent to the Editorial Office for announcement on this website.

Submitted manuscripts should not have been published previously, nor be under consideration for publication elsewhere (except conference proceedings papers). All manuscripts are thoroughly refereed through a single-blind peer-review process. A guide for authors and other relevant information for submission of manuscripts is available on the Instructions for Authors page. Materials is an international peer-reviewed open access semimonthly journal published by MDPI.

Please visit the Instructions for Authors page before submitting a manuscript. The Article Processing Charge (APC) for publication in this open access journal is 2600 CHF (Swiss Francs). Submitted papers should be well formatted and use good English. Authors may use MDPI's English editing service prior to publication or during author revisions.

Keywords

  • molten salt reactor
  • molten salts
  • structural materials
  • corrosion
  • radiation damage
  • CALPHAD
  • molecular dynamics
  • DFT

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Published Papers (7 papers)

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Research

20 pages, 5343 KiB  
Article
Synthesis, Purification, and Characterization of Molten Salt Fuel for the SALIENT-03 Irradiation Experiment
by Pavel Souček, Ondřej Beneš, Pieter Ralph Hania, Konstantin Georg Kottrup, Elio D’Agata, Alcide Rodrigues, Helena Johanna Uitslag-Doolaard and Rudy J. M. Konings
Materials 2024, 17(24), 6215; https://doi.org/10.3390/ma17246215 - 19 Dec 2024
Viewed by 879
Abstract
This work presents the synthesis, purification, and characterization of a molten salt fuel for the irradiation experiment SALIENT-03 (SALt Irradiation ExperimeNT), a collaborative effort between the Nuclear Research and Consultancy Group and the Joint Research Centre, European Commission. The primary objective of the [...] Read more.
This work presents the synthesis, purification, and characterization of a molten salt fuel for the irradiation experiment SALIENT-03 (SALt Irradiation ExperimeNT), a collaborative effort between the Nuclear Research and Consultancy Group and the Joint Research Centre, European Commission. The primary objective of the project is to investigate the corrosion behavior of selected Ni-alloy based structural materials which are being considered for the construction of fluoride molten salt reactors. During the test, these materials will be exposed to selected liquid molten fuel salts under irradiation in the High Flux Reactor in Petten, the Netherlands. In addition, the properties and distribution of the fission products formed in the fuel salt during burn-up will be studied by various post irradiation examinations. In the SALIENT-03 fuel, U and Pu fluorides, as fissile material, are dissolved in a carrier melt based on a 787LiF-22ThF4 eutectic mixture to form fuel salts with four different compositions, containing PuF3, UF4, UF3, and CrF3. This article comprehensively describes all the steps of this fuel synthesis process: the synthesis of the required pure fluoride powders (7LiF, ThF4, UF4, UF3, and PuF3); the mixing, melting, and purification of the different fuel salt compositions; and the fabrication of solid ingots to be loaded into the irradiation capsules. The characterization of the intermediate and final products is also carried out, following a rigorous quality assurance protocol. The quality assurance is achieved using an analytical scheme consisting of mass balance-based conversion efficiency evaluation, X-ray diffraction, and differential scanning calorimetry analyses. All experimental goals were successfully achieved, highlighting promising prospects for advancing future research and development regarding fuel production methods for fluoride-based molten salt reactors. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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8 pages, 2455 KiB  
Communication
An MeV Proton Irradiation Facility: DICE
by Sören Möller, Daniel Höschen, Wim Arnoldbik and Beata Tyburska-Pueschel
Materials 2024, 17(15), 3646; https://doi.org/10.3390/ma17153646 - 24 Jul 2024
Viewed by 982
Abstract
Materials applied in nuclear environments such as fission or fusion power-plants face severe conditions. The irradiation by neutrons induces thermal loads and irradiation damage. Furthermore, coolants in contact with the materials induce corrosion, which is particularly challenging for liquid salts intended for the [...] Read more.
Materials applied in nuclear environments such as fission or fusion power-plants face severe conditions. The irradiation by neutrons induces thermal loads and irradiation damage. Furthermore, coolants in contact with the materials induce corrosion, which is particularly challenging for liquid salts intended for the next generation of fission reactors. A new device (DICE) is installed at the 3.5 MV accelerator at DIFFER for the accelerated testing of such materials under combined irradiation and corrosion conditions. The DICE enables irradiation of samples at temperatures of up to 1050 K and in contact with liquid salts. An integrated shielding and a low power temperature control concept based on radiation cooling enables high-duty cycle application in a standard accelerator laboratory. Ion currents of up to 30 µA are possible with continuous irradiation. This work outlines the technical concept of the device and presents the first data. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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24 pages, 8724 KiB  
Article
Corrosion of Two Iron-Based Aluminaforming Alloys in NaCl-MgCl2 Molten Salts at 600 °C
by Louis Pellicot, Nathalie Gruet, Jérôme Serp, Romain Malacarne, Sophie Bosonnet and Laure Martinelli
Materials 2024, 17(13), 3224; https://doi.org/10.3390/ma17133224 - 1 Jul 2024
Viewed by 1326
Abstract
Molten salts have been used as heat transfer fluids since the middle of the 20th century. More recently, molten chloride salts have been studied for use in concentrated solar power plants or molten salt reactors. However, none of the materials studied to date [...] Read more.
Molten salts have been used as heat transfer fluids since the middle of the 20th century. More recently, molten chloride salts have been studied for use in concentrated solar power plants or molten salt reactors. However, none of the materials studied to date has been able to withstand this highly corrosive environment without controlling the salt’s redox potential. The alumina-forming alloy was a promising option, as it has not yet been widely studied. To investigate this possibility, two iron-based alumina-forming alloys were corroded in NaCl-MgCl2 eutectic at 600 °C for 500 h after being pre-oxidised to grow a protective layer of α-alumina on each alloy. A salt purification protocol based on salt electrolysis was implemented to ensure comparable and reproducible results. During immersion, alumina was transformed into MgAl2O4, as shown by FIB-SEM observation. Inter and intragranular corrosion were observed, with the formation of MgAl2O4 in the corroded zones. The nature of the oxides was explained by the predominance diagram. Intragranular corrosion was 2 µm deep, and intergranular corrosion 10 µm deep. Alumina formed at the bottom of the intergranular corrosion zones. The depth of intergranular corrosion is consistent with O diffusion control at the grain boundary. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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21 pages, 4302 KiB  
Article
Thermophysical Properties of FUNaK (NaF-KF-UF4) Eutectics
by Maxime Fache, Laura Voigt, Jean-Yves Colle, John Hald and Ondřej Beneš
Materials 2024, 17(11), 2776; https://doi.org/10.3390/ma17112776 - 6 Jun 2024
Cited by 1 | Viewed by 1563
Abstract
General interest in the deployment of molten salt reactors (MSRs) is growing, while the available data on uranium-containing fuel salt candidates remains scarce. Thermophysical data are one of the key parameters for reactor design and understanding reactor operability. Hence, filling in the gap [...] Read more.
General interest in the deployment of molten salt reactors (MSRs) is growing, while the available data on uranium-containing fuel salt candidates remains scarce. Thermophysical data are one of the key parameters for reactor design and understanding reactor operability. Hence, filling in the gap of the missing data is crucial to allow for the advancement of MSRs. This study provides novel data for two eutectic compositions within the NaF-KF-UF4 ternary system which serve as potential fuel candidates for MSRs. Experimental measurements include their melting point, density, fusion enthalpy, and vapor pressure. Additionally, their boiling point was extrapolated from the vapor pressure data, which were, at the same time, used to determine the enthalpy of vaporization. The obtained thermodynamic values were compared with available data from the literature but also with results from thermochemical equilibrium calculations using the JRCMSD database, finding a good correlation, which thus contributed to database validation. Preliminary thoughts on fluoride salt reactor operability based on the obtained results are discussed in this study. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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18 pages, 6105 KiB  
Article
Corrosion Mitigation in Molten Salt Environments
by Sylvie Delpech, Charly Carrière, Alexandre Chmakoff, Laure Martinelli, Davide Rodrigues and Céline Cannes
Materials 2024, 17(3), 581; https://doi.org/10.3390/ma17030581 - 25 Jan 2024
Cited by 10 | Viewed by 3092
Abstract
The aim of this paper is to present methods for corrosion mitigation in molten salt environments. The corrosion of structural materials depends directly on the redox potential of the salt. When the redox potential of the salt is higher than the standard potentials [...] Read more.
The aim of this paper is to present methods for corrosion mitigation in molten salt environments. The corrosion of structural materials depends directly on the redox potential of the salt. When the redox potential of the salt is higher than the standard potentials of the elements constituting the structural materials, corrosion occurs. If the reverse is true, no corrosion is observed. Herein, a methodology for calculating the theoretical potential of a molten salt is provided and compared with experimental measurements. Three ways to mitigate corrosion by modifying the salt redox potential are proposed: (i) using a soluble/soluble redox system; (ii) using a potentiostatic method; and (iii) using an amphoteric compound such as UCl3, TiCl2, or TiCl3. Immersion tests were conducted under the above conditions to validate the methodology. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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10 pages, 1344 KiB  
Article
Applications of Thermochemical Modeling in Molten Salt Reactors
by Theodore M. Besmann, Juliano Schorne-Pinto, Mina Aziziha, Amir M. Mofrad, Ronald E. Booth, Jacob A. Yingling, Jorge Paz Soldan Palma, Clara M. Dixon, Jack A. Wilson and Donny Hartanto
Materials 2024, 17(2), 495; https://doi.org/10.3390/ma17020495 - 20 Jan 2024
Cited by 1 | Viewed by 2125
Abstract
The extensively evaluated and consistent thermodynamic database, the Molten Salt Thermal Properties Database—Thermochemical (MSTDB-TC), was used along with additional thermodynamic values from other sources as examples of ways to examine molten salt reactor (MSR) fuel behavior. Relative stability with respect to halide potential [...] Read more.
The extensively evaluated and consistent thermodynamic database, the Molten Salt Thermal Properties Database—Thermochemical (MSTDB-TC), was used along with additional thermodynamic values from other sources as examples of ways to examine molten salt reactor (MSR) fuel behavior. Relative stability with respect to halide potential and temperature for likely fuel and fission product components were mapped in Ellingham diagrams for the chloride and fluoride systems. The Ellingham diagrams provide a rich, visual means for identifying halide-forming components in proposed fuel/solvent salt systems. Thermochemical models and values from MSTDB-TC and ancillary sources were used in global equilibrium calculations to provide compositions for a close analysis of the behavior of a possible Molten Chloride Salt Fast Reactor and a Molten Salt Reactor Experiment-type system at high burnup (100 GWd/t). The results illustrated the oxidative nature of burnup in MSRs and provided information about redox behavior and possible control. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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14 pages, 17131 KiB  
Article
Corrosion Behavior of Incoloy®800H, Hastelloy®G35® and 316L Stainless Steel in the Molten Eutectic Fluoride Mixture FLiNaK and Its Vapors
by Ambati Ramu, Viliam Pavlik, Veronika Sillikova and Miroslav Boca
Materials 2023, 16(7), 2679; https://doi.org/10.3390/ma16072679 - 28 Mar 2023
Cited by 2 | Viewed by 1698
Abstract
This paper discusses the findings of a corrosion experiment which investigates a transition area which is between the lower area and upper area of the sample. In this experiment, the lower half of the sample surface is exposed to molten salt and the [...] Read more.
This paper discusses the findings of a corrosion experiment which investigates a transition area which is between the lower area and upper area of the sample. In this experiment, the lower half of the sample surface is exposed to molten salt and the upper half of the sample surface is exposed to the vapors generated by molten FLiNak salt. Incoloy®800H and Hastelloy®G35® alloys and 316L stainless steel were selected for the corrosion experiment on the basis of their Cr content. The experiment was conducted at 600 °C for a period of 100 h. The results of the experiment show that, in the transition area, no abrupt change in corrosion mechanism takes place; the experiments also give us information on how the degree of degradation varies across the three areas of the samples. The experiment also showed two distinct corrosion mechanisms operating in the test samples: intergranular corrosion in the SS316L stainless steel test sample and continuous corrosion in Hastelloy®G35®. The results also show a progressive reduction in the concentrations of F, K and Na in the upper areas compared to the lower areas for Hastelloy©G35®. Cr is shown to have a critical role in the corrosion process, even when the sample surface is not in direct contact with the molten salt and is only in contact with the vapors generated by the molten salt. Full article
(This article belongs to the Special Issue Recent Advances in Materials for Molten Salt Nuclear Reactor Technology)
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