1. Introduction

Molten salt reactors (MSRs) are a type of Generation IV reactor that, unlike traditional reactors that use solid fuel, utilize a liquid nuclear fuel created by dissolving actinide compounds in chloride- or fluoride molten salt at high temperatures. MSRs operate at high temperatures between 500 to 800°C without pressurization, providing high thermal efficiency and inherent safety against severe accidents such as meltdown due to loss of coolant. A significant advantage of MSRs is the greater thermal expansion coefficient of the liquid compared to solid fuels, allowing for immediate negative reactivity feedback and autonomous control of the core output temperature without the need for control rods as the temperature increases [1, 2].

Historically, MSRs were demonstrated at Oak Ridge National Laboratory (ORNL) through the Aircraft Reactor Experiment (ARE) and Molten Salt Reactor Experiment (MSRE) programs in the 1950s and 1960s, achieving outputs of 2.5 MWth and 7.4 MWth, respectively, using fluoride- based molten fuel salts [3, 4]. Since the 2010s, various research organizations and start-ups in various countries have been developing MSRs with different design concepts. Particularly, chloride-based MSRs, which are suitable for gas spectrum nuclear fuel cycles and currently under development in countries such as the USA and France, can operate for over 20 years due to high actinide solubility. These reactors can utilize transuranic elements (TRUs) recovered from spent PWR fuel through pyroprocessing as well as high-assay low-enriched uranium (HALEU), contributing to significant volume reductions of high-level wastes. Unlike fluoride-based thermal neutron spectrum MSRs that use ²³³U, ²³²Th, etc., fast spectrum MSRs have a relatively reduced need for aggressive online reprocessing, such as 233Pa removal, thereby simplifying reactor design.

In South Korea, development of a 100 MWth chloridebased molten salt reactor for marine propulsion applications began in 2023 under the Korea Advanced Research Program Accelerator (KARPA) program. The subprojects under the center for development of MSR fundamental technologies include the MSR core design technology development, MSR liquid fuel and materials technology development, MSR system design technology development, MSR safety technology development, and MSR marine application technology development. This paper will describe the results of experiments on liquid nuclear fuel fabrication and natural convection molten salt loop operation to evaluate the corrosion characteristics of structural materials, which are currently performing associated with the MSR liquid fuel and materials technology development project among the subprojects.

1.1 Uranium Chlorination and Liquid Fuel Fabrication for Molten Salt Reactors

Chloride-based nuclear fuels for molten salt reactors (MSRs) are synthesized by combining actinide chlorides (e.g., UCl₃, PuCl₃) with various appropriately refined carrier or coolant salts, such as NaCl, NaCl-KCl, or NaCl- MgCl₂ [5, 6]. The eutectic composition and melting temperature of these mixtures are highly dependent on the specific combinations of base salts and actinide chlorides, which in turn dictate the thermophysical properties of the MSR fuel.

The synthesis of actinide chlorides has traditionally been achieved via solid-liquid chlorination, wherein molten salts like LiCl-KCl are combined with chlorinating agents (e.g., PbCl₂, CdCl₂, CuCl₂, ZnCl₂, BiCl₃). A representative example of such a chlorination reaction is [7, 8]:

While effective, this process often produces undesired byproducts, such as metallic cadmium (Cd) or lead (Pb), as well as higher-valence actinide chlorides (e.g., UCl₄). Consequently, additional purification or separation steps are required to isolate trivalent actinide chlorides.

Chloride-based MSRs exhibit notable advantages, such as the high solubility of actinide chlorides—up to approximately 40mol% (~70wt%)—which is nearly an order of magnitude higher than that in fluoride-based MSRs. This high solubility supports long-term stable reactor operation without frequent reprocessing. However, to optimize core design for desired power output and operational longevity, careful control of the initial actinide chloride concentration and periodic replenishment during operation may be necessary. This underscores the importance of developing direct, efficient methods for synthesizing high-purity actinide chlorides like UCl₃ via solid-gas reactions using metallic or oxide actinides.

Efforts to improve the chlorination process have included the use of Cl₂ gas and carbon, as demonstrated by Shin et al. [9], who converted U₃O₈ to UCl₃. Similarly, researchers at Los Alamos National Laboratory [10] successfully chlorinated 500 g of PuO₂ using COCl₂ or Cl₂-CCl₄ gas mixtures. While effective, these methods often result in actinide chlorides of limited purity. For metallic uranium, direct chlorination with Cl₂ gas is possible without the need for catalysts; however, this approach frequently yields a mixture of UCl₃, UCl₄, UCl₅, and UCl₆, depending on the stoichiometry of Cl₂ supplied. Additional purification steps, such as condensation, reduction, or separation, are required to obtain pure UCl₃.

Recently, alternative approaches using ammonium chloride (NH₄Cl) have demonstrated significant promise for converting uranium and plutonium metals to chlorides [11-13]. The use of NH₄Cl has the distinct advantage of generating only gaseous byproducts, as shown in the reaction below, which facilitates the production of high-purity UCl₃. Although HCl gas directly participates in the chlorination of U, initiating the reaction using NH₄Cl, which is solid at room temperature, is advantageous for the precise control of the optimal reaction conditions.

At the Korea Atomic Energy Research Institute (KAERI), significant progress has been made in developing a closed-reactor system for uranium chlorination using NH₄Cl. Preliminary experiments successfully demonstrated the conversion of 10 g of uranium metal into high-purity UCl₃ [12].

Building upon this foundation, the present study introduces a hydrogenation pretreatment step to enhance the efficiency and scalability of the NH₄Cl-based chlorination process. By first reacting uranium metal with hydrogen gas to form UH₃ powder, the reactivity of the subsequent chlorination process is significantly increased. This approach enables higher chlorination efficiency and greater production capacity of UCl₃.

1.2 Evaluation of Corrosion Characteristics of MSR Materials Using a Natural Convection Molten Salt Loop

The natural convection loop or thermal convection loop has been developed as a method for evaluating corrosion characteristics in molten salt environments [14]. Designed to study the dynamic corrosion behavior of materials, the natural convection loop relies on buoyancy-driven circulation induced by temperature-dependent density differences in the molten salt. This density-driven flow is generated by positioning the hot leg below the cold leg. Unlike isothermal autoclave experiments, the natural convection loop enables mass transport, which cannot be achieved under isothermal conditions [15].

In isothermal autoclave experiments, corrosion occurs at a single operating temperature. In contrast, experiments utilizing a natural convection loop allow various ionic species in the molten salt to circulate between the hot and cold legs, driven by changes in equilibrium constants with temperature and moderated by redox reactions. The thermal gradientpumped loop, which exhibits a significantly faster flow rate than the natural convection loop described above, can more accurately simulate reactor operating conditions [16].

Corrosion in typical molten salt loops is most pronounced in the hot leg, where reaction rates are highest. Consequently, the most corrosion-prone elements in an alloy dissolve into the molten salt. The order of dissolution of alloying elements into molten salts is determined by their Gibbs free energy. For fluoride- and chloride-based molten salt systems, the corrosion propensity of common alloying elements increases in the order: Mo, Ni, Fe, and Cr [17, 18]. Chromium is particularly vulnerable due to its formation of stable corrosion products, such as chromium chlorides, which are more readily attacked than other elements in the alloy.

As molten salt flows through the natural convection loop, corrosion products form and dissolve into the salt [19]. When the salt enters the cold leg and cools, changes in equilibrium constants reverse the corrosion reactions, causing metal ion precipitation. These precipitates may deposit on the loop walls, diffuse into the matrix, or remain suspended in the molten salt [16, 20]. Once the salt recirculates to the hot leg, the corrosion reaction resumes as equilibrium constants shift toward corrosion of the alloy [17]. This dynamic, non-equilibrium process, driven by temperature gradients, is a significant departure from the static conditions of isothermal autoclave experiments. Such differences are crucial for simulating corrosion in MSRs, where molten salt flows from the high-temperature reactor core to the heat exchanger, mimicking real operational conditions.

To date, most studies on molten salt corrosion have focused on fluoride salts under isothermal autoclave or electrochemical testing conditions, without accounting for temperature-driven mass transport. Additionally, reported corrosion rates of various materials in chloride salt environments have shown significant variability, making it challenging to draw definitive conclusions. To address these limitations, we designed and fabricated a natural convection loop for corrosion experiments in chloride-based molten salts. Using this system, we evaluated the corrosion characteristics of candidate structural materials for MSRs under realistic operating conditions.

2. Experimental

2.1 Uranium Chlorination and Liquid Fuel Fabrication for Molten Salt Reactors

Experiments for the chlorination of uranium metal were conducted in an argon-purged glovebox with oxygen and moisture levels controlled below 10 ppm. To hydride uranium metal, a gas mixture of Ar-4wt% H₂ was employed, while ammonium chloride (NH₄Cl, Sigma Aldrich, 99.99%) was used as the chlorinating agent. Prior to use, NH₄Cl was dried at 150°C for 5 hours to remove residual moisture. The uranium metal used in the experiments consisted of depleted uranium pellets with a cylindrical shape, approximately 1 cm in diameter and height.

The schematic design of the reactor used for powdering uranium metal through reaction with hydrogen gas is shown in Fig. 1(a). The reactor, approximately 10 cm in diameter and height, was equipped with two gas lines passing through the reactor cover. The inlet gas line, located at the center of the cover, extended to the bottom of the reactor, enabling the supply of hydrogen gas from the base (H₂-in). The outlet gas line was installed to discharge gases produced during the reaction (H₂-out). A mesh container, attached to the H₂ inlet gas tube approximately 5 cm above the base, was used to load uranium metal. An Al₂O₃ collection crucible was placed at the bottom of the reactor to recover the UH₃ powder formed during hydrogenation.

Fig. 1

Schematic designs of (a) hydriding reactor and (b) chlorination reactor.

The chlorination reactor shares the same structure as the hydrogenation reactor, with modifications made only to the flange (refer to Fig. 1(b)). The chlorination reactor cover included a single exhaust line to facilitate the removal of residual gases and was designed to connect to a vacuum pump for forced evacuation when necessary. Inside the stainless steel reactor, an additional Al₂O₃ reactor and cover (Al₂O₃ 99.8%) were installed. This setup ensured that the decomposed HCl gas remained within the Al₂O₃ reactor for an extended period to maximize its participation in the chlorination reaction, while also preventing contamination from chlorinated byproducts (e.g., FeCl₃) introduced by HCl gas.

For this study, the hydride product (UH₃) and the chlorinated product (UCl₃) were analyzed qualitatively and quantitatively using X-ray diffraction (XRD) and scanning electron microscopy-energy dispersive spectroscopy (SEM/EDS). Quantitative analysis of uranium chlorination efficiency was conducted using inductively coupled plasma optical emission spectroscopy (ICP-OES).

2.2 Evaluation of Corrosion Characteristics of MSR Materials Using a Natural Convection Molten Salt Loop

A NaCl–MgCl₂ (57–43mol%) molten salt mixture was prepared using commercial-grade NaCl (99.5%, Junsei) and anhydrous MgCl₂ (99%, Alfa Aesar). A total of 8.5 kg of salt (3.8 kg NaCl and 4.7 kg MgCl₂) was dried sequentially at 117, 185, 242, 304, 350, and 554°C to remove residual moisture [19, 21]. These steps facilitated the dehydration of MgCl₂, which forms hydrates under specific conditions, resulting in HCl generation during heating. HCl removal was critical to prevent corrosion in experimental equipment and molten salt circulation loops. A vacuum pump and a scrubber system were incorporated to maintain a vacuum pressure of 0.1 MPa, effectively removing HCl and moisture.

Corrosion studies employed three alloys—SS304, SS316, and Hastelloy-N—with 12 samples of each (10 mm × 15 mm × 1 mm) mounted on pure Ni wires. The samples were placed in the hot and cold legs of a natural convection loop designed to replicate Molten Salt Reactor (MSR) conditions (Fig. 2 [22]). The loop included a pre-treatment tank, salt reservoir, and drain tank connected by a 1-inch SS316 tube (Fig. 3 [22]). Six thermocouples, positioned in the hot and cold legs, monitored salt temperatures during operation.

Fig. 2

Two Ni wires with 36 samples of SS304, SS316 and Hastelloy-N. Samples with capital letter “C” are for the cold leg, without are for the hot leg of natural convection loop [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

Fig. 3

A schematic of a high-temperature molten salt natural convection loop [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

Fig. 4 [22] illustrates the loop system, equipped with heating elements for each tank and totaling 6 kW of power. The pre-treatment tank was directly linked to the loop, and two sample chains containing corrosion coupons were introduced into each leg. Salt was melted by activating all heaters, with T1–T6 temperatures gradually increasing over 8 hours. Once T6 reached the eutectic melting point, temperatures equalized, and natural salt circulation commenced. A stabilization phase of 4 hours followed, during which only the hot leg heater, maintained at 950°C, remained active as shown in Fig. 5 [22].

Fig. 4

An image of natural convection loop [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

Fig. 5

The heater temperature during start-up process of the loop [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

During 500 hours of normal operation, the highest T6 temperature in the hot leg peaked at ~580°C, while the lowest T1 temperature in the cold leg remained around 500°C as illustrated in Fig. 6 [22]. Current sensors attached to each heater calculated the salt flow rate, confirming steady natural circulation. After the experiment, the salt was drained into a tank for ICP-OES analysis to quantify metal dissolution.

Fig. 6

The heater temperature during start-up process of the loop. The image on the right is an enlarged view between stabilization and normal operation [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

The corrosion coupons from the cold and hot legs were retrieved, cleaned, and divided into 36 samples for analysis. Weight loss measurements were performed with a high-precision balance, and 18 samples were mounted and polished for SEM and EDS characterization. These analyses provided insights into material corrosion behavior under the tested conditions.

3. Results and Discussion

3.1 Uranium Chlorination and Liquid Fuel Fabrication for Molten Salt Reactors

3.1.1 Equilibrium calculations for U hydriding and U chlorination

A previous study [11] successfully synthesized uranium trichloride (UCl₃) by directly reacting metallic uranium (U) with ammonium chloride (NH₄Cl), as described by the following reaction:

However, for large-scale chlorination of uranium metal, pulverization is essential to increase the reaction surface area. In this study, a hydriding process was introduced to facilitate the pulverization of metallic uranium.

Fig. 7 presents the equilibrium calculations of the reaction between 1 mol of uranium metal and hydrogen gas (H₂) at various temperatures, performed using HSC Chemistry [23]. Below approximately 200°C, uranium reacts with hydrogen gas to form uranium trihydride (UH₃), whereas dehydrogenation occurs above 250°C.

Fig. 7

Equilibrium calculation of U hydriding reaction with H₂ gas against temperature.

In this study, UH₃ was synthesized via hydriding and then directly reacted with NH₄Cl to produce UCl₃, avoiding dehydrogenation, as represented by the following reactions:

Equilibrium calculations for 1 mol of UH₃ and 3 mol of NH₄Cl were performed, as shown in Fig. 8(a), confirming complete conversion to UCl₃ within the temperature range of 350°C to 450°C. Fig. 8(b) illustrates the equilibrium calculations at 400°C with a fixed 1 mol of UH₃ while varying the NH₄Cl input. When the NH₄Cl amount is below the stoichiometric ratio of 3, unreacted uranium interacts with N2 gas, forming byproducts such as UN0.965 and UNCl. Conversely, an excessive NH₄Cl input results in the overproduction of HCl, leading to an increased yield of UCl₄.

Fig. 8

Equilibrium calculation of UH₃ (1 mol) chlorination reaction with NH₄Cl (a) with respect to temperature and (b) amount of NH₄Cl addition.

3.1.2 Results of uranium metal hydriding

To achieve uranium metal pulverization, a dedicated hydriding reactor was designed, as shown in Fig. 9. Metallic uranium was loaded into a mesh-type container attached to the reactor cover, and the system was connected to a furnace, an H₂ gas inlet, and an H₂ outlet. For safety, an Ar-3.9 wt% H₂ gas mixture was used, and the gas flow rate was controlled via a mass flow controller. A pressure gauge was installed in the exhaust line to monitor the pressure of the exhaust gas throughout the experiment.

Fig. 9

Photographs of U hydriding reactor.

A total of 13 hydriding experiments were conducted, with experimental conditions and results summarized in Table 1. Experiments UH₃_#1 to UH₃_#7 explored different conditions (H₂ gas flow rate, reaction time, and temperature) to determine the optimal parameters for efficient uranium hydriding. The theoretical amount is the amount of UH₃ assuming that all of the input hydrogen gas reacts with U metal. Fig. 10 depicts the progressive pulverization of uranium pellets during hydriding. Experiments UH₃_#1 to UH₃_#3, which involved a single uranium pellet, exhibited low contact probability between uranium and hydrogen gas, resulting in a hydriding efficiency below 1%. However, increasing the initial uranium loading (UH₃_#3 to UH₃_#7) improved the hydriding efficiency up to 6.5%, indicating that the hydrogen gas flow rate significantly influenced the reaction. A flow rate exceeding 2 L∙min⁻¹, however, was predicted to induce fluidization issues with the produced UH₃ powder. Temperature effects were minimal below 200°C.

Fig. 10

Photograph of (a) depleted uranium pellet (left) and (b) partial UH₃ powder after hydriding with H₂ gas in UH₃_#1 experiment.

In UH₃_#8, a cubic uranium metal sample was used without physically removing the surface oxide layer prior to hydriding. Fig. 11(a) shows the cross-sectional image of uranium metal after 320 minutes of reaction, while Fig. 11(b) presents its SEM surface image. The uranium surface underwent partial hydriding, forming crater-like structures. The reaction predominantly progressed at initiation sites rather than uniformly across the surface. Compared to experiments UH₃_#9 and UH₃_#10, where the oxide layer was removed beforehand, the hydriding efficiency of UH₃_#8 was approximately one-third lower, indicating that prior removal of the oxide layer is crucial for efficient hydriding.

Fig. 11

(a) Photograph U pellet surface after hydriding reaction (UH₃_#8) and (b) microscopic morphology of the surface.

Experiments UH₃_#11 to UH₃_#13 assessed the scalability of the hydriding process using a 388 g uranium plate, as shown in Fig. 12(a). During hydriding, a substantial amount of UH₃ powder accumulated not only in the crucible at the bottom of the reactor but also on the residual uranium plate (Fig. 12(b)). To enhance powder collection in future large-scale setups, a vibrator system could be implemented. The uranium plate exhibited uniform hydriding, with abundant UH₃ powder observed in the collection crucible. The XRD analysis of the product (Fig. 12(c)) confirmed the formation of UH₃. The hydriding efficiency reached up to 34.1%, and further improvements could be achieved by increasing the hydrogen concentration.

Fig. 12

(a) U plate preparation for large scale U hydriding, (b) U plate and UH₃ product after hydriding experiment, and (c) X-ray diffraction pattern for the product.

3.1.3 Chlorination of UH₃ using NH₄Cl

A dedicated chlorination reactor (Fig. 13) was constructed to convert UH₃ powder into UCl₃. The hydrided UH₃ powder was mixed with NH₄Cl and loaded into an Al₂O₃ crucible, which was then sealed with an Al₂O₃ lid to minimize external contamination. The stainless steel reactor cover was equipped with a 1/4-inch pipe to allow gas discharge.

Fig. 13

Photographs of the chlorination reactor.

Five chlorination experiments were conducted, and the experimental conditions and results are summarized in Table 2. NH₄Cl was used in a stoichiometric excess of 3−5%. To maximize chlorination efficiency, the heating rate was controlled between 0.5–0.2°C∙min⁻¹, and the reaction was maintained at 450°C for five hours. UCl₃_#1 and UCl₃_#2 were conducted on a 10 g scale, UCl₃_#3 and UCl₃_#4 on a 100 g scale, and UCl₃_#5 on a 250 g scale to evaluate the effect of increasing UH₃ input.

Fig. 14 shows the reactants, post-chlorination UCl₃ products, and their characterization results for UCl₃_#1. The final UCl₃ product appeared as sponge-like agglomerates that disintegrated easily under slight pressure. XRD analysis confirmed the formation of UCl₃, while SEM images revealed elongated, needle-like structures smaller than 50 μm. In UCl₃_#5, a slower heating rate (0.2°C∙min⁻¹) was applied to accommodate the increased UH₃ input, as shown in Fig. 15. The reactants initially occupied over 70% of the Al₂O₃ crucible volume, with no observed loss or ejection of material post-reaction. The final UCl₃ product was denser and more compact compared to those from UCl₃_#1 and UCl₃_#2, likely due to the slower reaction kinetics. The conversion rate remained above 98% despite the increased scale, and ICP-OES analysis confirmed a purity exceeding 99%, with an RSD below 1% across three repeated measurements.

Fig. 14

Photographs of the reactant and UCl₃ product in UCl₃_#1 experiment.

Fig. 15

Photographs of the UCl₃ product in UCl₃_#5 experiment.

The UCl₃ synthesis method developed in this study has the advantage of producing pure UCl₃ without any residual byproducts compared to conventional chlorinating agents (e.g., PbCl₂, CdCl₂, CuCl₂, ZnCl₂, BiCl₃, etc.). Additionally, compared to the process of first producing U metal powder and then chlorinating it with NH₄Cl [13], this method eliminates the need for a separate UH₃ dehydrogenation step, simplifying the overall process. Furthermore, it achieves the highest batch-scale UCl₃ production capacity reported in the literature to date.

Fig. 16 illustrates the potential application of the synthesized UCl₃ as a precursor for various molten salt reactor (MSR) liquid fuels by mixing it with carrier salts such as NaCl, KCl, and MgCl₂. As shown in Fig. 16, the concentration of UCl₃ in the liquid fuel can be independently controlled. It was confirmed that, depending on the eutectic composition of the liquid fuel, up to 11mol% of UCl₃ can be incorporated in NaCl-MgCl₂-UCl₃ and up to 54mol% in KCl-UCl₃.

Fig. 16

Photographs of various MSR salt compositions using pure UCl₃ synthesized by the reaction with NH₄Cl ((left) after mixing (right) after melting).

3.2 Evaluation of Corrosion Characteristics of MSR Materials Using a Natural Convection Molten Salt Loop

3.2.1 Salt analysis

Table 3 displays the results of the ICP-OES analysis conducted on the salt before and after the corrosion experiment. Prior to the experiment, the concentrations of Fe and Cr stood at 300.5 and 33.8 ppm, respectively. However, post-experiment, these concentrations rose to 460.7 and 218 ppm, correspondingly. Furthermore, trace amounts of Mn were initially detected, which escalated to 18.3 ppm thereafter. Fe, Cr, and Mn are constituents of SS316, SS304, and Hastelloy-N (refer to Table 4), suggesting their leaching into the salt during the corrosion experiment.

The increase in these elemental concentrations within the salt highlights the material degradation occurring in the molten salt loop. The dissolution of Fe and Cr suggests active corrosion processes that may be driven by thermal and chemical factors inherent to the system. The presence of Mn also indicates the potential breakdown of stabilizing phases within the stainless steel alloys, which could affect the mechanical and chemical integrity of these materials over prolonged exposure.

Understanding the salt chemistry before and after corrosion exposure is crucial in assessing the degradation mechanisms at play. The observed elemental changes suggest that a dissolution-precipitation mechanism may be occurring, where metallic elements from the exposed materials dissolve into the salt and subsequently re-deposit under varying temperature conditions. The implications of this behavior include localized material loss, altered mechanical properties, and potential clogging of flow pathways within the molten salt loop.

3.2.2 Microstructural analysis

Fig. 17 illustrates the weight fluctuations of the samples pre- and post-corrosion experiments [22]. The red and blue dots denote weight changes in samples from the hot and cold legs, respectively. Different shapes such as squares, circles, and triangles represent weight changes in the Hastelloy- N, SS304, and SS316 samples, respectively. Notably, SS304 exhibited the highest weight loss under both hot- and cold-leg conditions, suggesting its least satisfactory corrosion behavior in the molten salt. The corrosion progression followed this sequence: SS304, SS316, and Hastelloy-N. Particularly noteworthy is that when temperatures in the cold and hot legs were similar (positions T2–T4, T3–T5), the weight loss in the hot leg slightly surpassed that in the cold leg, indicating mass transport within the natural circulation loop. Additionally, only the Hastelloy-N samples installed in the cold leg showed weight gain, suggesting potential deposit accumulation on the Hastelloy-N surface in the cold leg.

Fig. 17

The weight change graph of the samples before and after the corrosion experiment in natural convection loop [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

To further investigate these weight fluctuations, the microstructural characteristics of the corroded surfaces were analyzed. Eighteen samples were subjected to microstructure analysis, comprising three samples each of SS304, SS316, and Hastelloy-N from both the cold and hot legs. Each sample underwent cleaning in DI water at 30°C for over 30 minutes to eliminate residual salt, followed by mounting and polishing. SEM and EDS analyses were then conducted to examine surface degradation, elemental distribution, and potential deposit formations.

Fig. 18 illustrates the SEM analysis findings for the hot-leg samples [22]. The initial row of data showcases the analysis outcomes of the Hastelloy-N, SS304, and SS316 at the T6 position from left to right; subsequent rows represent data at the T5 and T4 positions, respectively. The SEM results indicated corrosion damage to the SS304 material, which was evident in surface roughness. Notably, at the highest temperature data in the first row, SS304 displayed corrosion exceeding 40 μm. However, corrosion was scarcely observed in the Ni alloy and SS316 at the lowest temperature data in the third row in the hot leg.

Fig. 18

The SEM analysis results of hot leg samples [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

The microstructural analysis of the hot-leg samples further revealed the presence of intergranular attack and localized corrosion pits in the SS304 samples. These features suggest the breakdown of the protective oxide layer and the penetration of molten salt into the material structure. SS316 exhibited moderate corrosion resistance, while the Hastelloy- N alloy demonstrated superior stability with minimal surface degradation.

Fig. 19 displays the SEM analysis outcomes of the cold-leg samples [22]. The data in the initial row from left to right represent the analysis findings of the Hastelloy-N, SS304, and SS316 at the T3 position, while subsequent rows represent data at the T2 and T1 positions, respectively. The damage from corrosion of the SS304 matrix was most pronounced in the SEM analysis of the cold-leg samples. However, unlike the hot leg samples, at the highest temperature data in the cold leg, SS304 exhibited corrosion of approximately 10 μm. Additionally, a distinctive phenomenon was observed in the microstructure of the Hastelloy-N, wherein weight gain was detected. A deposit was observed between the wavy surfaces, specifically in CH₃ and CH5 of the Hastelloy-N.

Fig. 19

The SEM analysis results of cold leg samples [22]. Reprinted from [22], Int. J. Energy Res., under the terms of the Creative Commos CC BY license.

The deposition of corrosion products on the Hastelloy- N is indicative of a mass transport mechanism involving dissolved metal species from the SS304 and SS316 samples. These metal species migrate through the molten salt loop and precipitate out onto the Hastelloy-N in regions of lower temperature. This observed deposition behavior corroborates with the weight gain measurements, providing further evidence of mass transport occurring within the system. The implication of these deposits is significant from an operational performance standpoint. Specifically, the presence of such deposits can lead to localized alterations in the surface chemical composition, potentially causing material embrittlement or the formation of stress concentration points. These changes can degrade the structural integrity and functional reliability of the Hastelloy-N over prolonged operational periods.

Moreover, it is crucial to monitor and understand these changes, as they might necessitate the implementation of corrective measures to mitigate potential detrimental effects. For instance, periodic inspections and surface treatments could be employed to address any surface composition changes or to remove harmful deposits. Additionally, optimizing the system’s operational conditions, such as temperature management and the composition of the molten salt, might also help in reducing the extent of such detrimental processes. By adopting a comprehensive approach that encompasses both preventative and corrective actions, the impact of these corrosive phenomena on longterm operational performance can be effectively managed. Long-term implications of these deposits include potential increases in maintenance costs, reduced system longevity, and elevated risks of equipment failure, which can lead to significant economic losses and operational downtime.

Overall, the results from the SEM and EDS analyses provide critical insights into the corrosion resistance of the tested materials. The data strongly support the conclusion that SS304 is the least resistant to molten salt corrosion, while SS316 offers moderate protection. The Hastelloy-N emerges as the most promising material, demonstrating minimal weight loss and enhanced stability, even under varying temperature conditions. These findings contribute to the understanding of material selection for molten salt reactor applications and highlight the need for further investigations into mitigating corrosion-induced material degradation.

4. Conclusions

This study introduced a hydriding process to enhance the pulverization of uranium metal into UH₃, thereby improving the efficiency of uranium chlorination. A customdesigned hydriding reactor was fabricated, and experimental results confirmed that hydriding efficiency was influenced by uranium metal loading and hydrogen gas flow rate. The presence of an oxide layer hindered uniform hydriding, highlighting the necessity of pre-treatment. Under large-scale conditions, the maximum hydriding efficiency achieved was 34.1%.

For liquid fuel production, a chlorination reactor was designed to convert UH₃ into UCl₃ using NH₄Cl as a chlorination initiator. Optimized heating conditions (0.5–0.2°C∙min⁻¹) and a slight NH₄Cl excess (3–5%) enabled the efficient production of high-purity UCl₃ (>99%) with a conversion rate exceeding 98%. The synthesis scale was systematically expanded from 10 g to 250 g, confirming the feasibility of large-scale production. Furthermore, the synthesized UCl₃ was successfully mixed with NaCl, KCl, and MgCl₂, demonstrating its applicability for molten salt reactor (MSR) fuels.

To assess the compatibility of structural materials with chloride-based molten salts, corrosion tests were conducted using a natural convection molten salt loop system. Candidate materials were exposed to NaCl–MgCl₂ eutectic salt at 580°C for 500 hours under a significant thermal gradient. SS304 exhibited the most severe corrosion, characterized by extensive mass loss and Cr depletion. SS316 showed improved corrosion resistance, while Hastelloy-N, a high- Ni alloy, exhibited superior stability with minimal degradation, highlighting its effectiveness in mitigating corrosion under high-temperature molten salt conditions. The observed corrosion behavior underscores the importance of structure/coating material selection for MSR applications.

Based on the findings of this study, we plan to conduct a natural convection loop test using liquid nuclear fuel produced through our fuel fabrication process. This test will allow for a more realistic evaluation of structural material corrosion in an actual nuclear fuel environment, providing crucial data for assessing material stability in molten salt reactor conditions.