VTechWorks Repository :: Browsing by Author "Zhang, Jinsuo"

Browsing by Author "Zhang, Jinsuo"

Now showing 1 - 19 of 19

Corrosion behavior of aluminum alloy in simulated nuclear accident environments regarding the chemical effects in GSI-191
Wang, Da; Leong, Amanda; Yang, Qiufeng; Zhang, Jinsuo (Korean Nuclear Society, 2022-11)
Long-term aluminum (Al) corrosion tests were designed to investigate the condition that would generate severe Al corrosion and precipitation. Buffer agents of sodium tetraborate (NaTB), trisodium phosphate (TSP) and sodium hydroxide (NaOH) were adopted. The insulation materials, fiberglass and calcium silicate (Ca-sil), were examined to explore their effects on Al corrosion. The results show that significant precipitates were formed in both NaTB/TSP-buffered solutions at high pH. The precipitates formed in NaTB solution raise more concerns on chemical effects in GSI-191. A passivation layer formed on the surfaces of coupon in solution with the presence of insulations could effectively mitigate Al corrosion. The Fe-enriched intermetallic particles (IPs) embedded in coupon appeared to serve as seeds to readily induce precipitation via providing extra area for heterogeneous Al hydroxide precipitation. X-ray spec-troscopy (EDS) and X-ray diffraction (XRD) analyses indicate that the precipitates are mainly boehmite (g-AlOOH) and no direct evidence confirms the presence of sodium aluminum silicate or calcium phosphate.
Corrosion of Silica-Based Optical Fibers in Various Environments
Leong, Amanda; Rountree, Steven Derek; Zhang, Jinsuo (MDPI, 2023-08-08)
This research article explores the potential of optical fibers as sensors, highlighting their ability to measure various parameters such as temperature, pressure, stress, and radiation dose. The study focuses on investigating the material compatibility of optical fibers in challenging sensing environments like Gen II/II+ and advance nuclear reactors, as well as concentrated solar power (CSP) plants. Material compatibility tests were conducted to determine the feasibility of using fluorine and germanium optical fiber sensors in these environments. The study found that raw fibers were corrosion-resistant to lead bismuth eutectic at 600 °C, regardless of the coating. In molten salt environments, raw fibers were incompatible with FLiNaK but showed corrosion resistance to MgCl₂-NaCl-KCl. However, the survivability of raw fiber optics improved with a gold coating in FLiNaK. Raw fiber optics were found to be incompatible in high-temperature steam at 1200 °C and in a pressurized water reactor (PWR) at 300 °C.
Corrosion Studies of Molten Chloride Salt: Electrochemical Measurements and Forced Flow Loop Tests
Zhang, Mingyang (Virginia Tech, 2023-08-23)
This study encompasses various aspects of corrosion in chloride molten salt environments, employing electrochemical techniques and a forced convection loop. It explores corrosion thermodynamic properties, electrochemical corrosion kinetics, and flow-induced dynamic corrosion. The study developed a novel electrochemical method for measuring thermodynamic properties of corrosion products and develops a new analysis theory for potentiodynamic polarization data obtained from cathodic diffusion-controlled reactions. Additionally, the design and operation experience of a forced convection chloride molten salt loop is shared. Particularly, the study presents novel findings on the turbulent flow-induced corrosion phenomenon and mechanism of Fe-based alloys in Mg-based chloride molten salt. These outcomes provide valuable insights into the corrosion mechanisms and flow-induced corrosion of Fe-based alloys in chloride molten salt. The results and experiences shared in this paper have implications for the successful implementation of molten salt as an advanced heat transfer fluid and thermal energy storage material in high-temperature applications, benefiting the nuclear and concentrating solar communities.
Development of Metallic Fuel Additives and Alloys for Sodium-cooled Fast Reactors
Zhuo, Weiqian (Virginia Tech, 2022-07-11)
The major goal of the work is to develop effective additives for U-10Zr (wt.%) metallic fuel to mitigate the fuel-cladding chemical interactions (FCCIs) due to fission product lanthanides and to optimize the fuel phase mainly by lowering the gamma-onset temperature. The additives Sb, Mo, Nb, and Ti have been investigated. Metallic fuels with one or two of the additives and with or without lanthanide fission products were fabricated. In this study, Ce was selected as the representative lanthanide fission product. A series of tests and characterizations were carried out on the additive-bearing fuels, including annealing, diffusion coupling, scanning electron microscopy (SEM), X-ray powder diffraction (XRD), and differential scanning calorimetry (DSC). Sb was investigated to mitigate FCCIs because available studies show its potential as a lanthanide immobilizer. This work extends the knowledge of Sb in U-10Zr, including its effect in the Zr-free region. Sb forms precipitates with fuel constituents, either U or Zr. However, it combines with the lanthanide fission product Ce when Ce is present. Those Sb-precipitates are found to be stable upon annealing, and are compatible with the cladding. The additive does not change the phase transition of U-10Zr. Mo, Nb, and Ti have been investigated for phase optimization based on the known characteristics shown in the binary phase diagrams. The quaternary alloys, i.e., two Mo-bearing alloys and two Nb-bearing alloys, were investigated. Compared to U-10Zr, a few weight percentages of Zr are replaced by those additives in the quarternary alloys. The solid-state phase transitions were determined (alpha and U2Ti transfer into gamma). The transition temperature varies depending on the compositions. The Mo-bearing alloys have lower -onset temperatures than the Nb-bearing alloys. All of them have lower gamma-onset temperatures than that of U-10Zr. Since low gamma-onset temperature is favorable, the results indicate that the fuel phase can be optimized by the replacement of a few weight percentages of Zr into those additives. All the experiments were out-of-pile tests. Therefore, in-pile experiments will be necessary to fully evaluate the performance of the additives in the future.
Diffusion behavior of lanthanide-additive compounds (Ce4Sb3, Ce2Sb, and CeTe) against HT9 and Fe
Xie, Yi; Zhang, Jinsuo; Benson, Michael T.; Mariani, Robert D. (2019-04)
Antimony and tellurium have been identified as promising additives in metallic fuel, which can immobilize free-lanthanide fission products into stable intermetallic compounds in order to mitigate the fuel-cladding chemical interaction. Ce4Sb3, Ce2Sb, and CeTe are the primary compounds formed by Sb or Te with the lanthanide Ce present in the fuel. If these compounds are present at the outer periphery of the fuel, they will come in contact with and react with the cladding after the fuel swells. The present study investigates the reactivity of these compounds with two cladding materials, HT9 and Fe. The diffusion couple tests between these compounds and HT9 or Fe were conducted at 853 K. Scanning electron microscopy and transmission electron microscopy were used to analyze the morphology, microstructure, and phase distribution of the diffusion region. It was observed that the diffusion region thickness formed by the three compounds was significantly reduced compared to free Ce. There was no observed diffusion or reaction between Ce4Sb3 or Ce2Sb with either HT9 or Fe. CeTe was found to diffuse and react with HT9, forming Cr3Te4 and TeFe at the diffusion region, as well as to penetrate into Fe, mostly by intergranular diffusion.
Direct Lithium-ion Battery Recycling to Yield Battery Grade Cathode Materials
Ge, Dayang (Virginia Tech, 2019-08-05)
The demand for Lithium-ion batteries (LIBs) has been growing exponentially in recent years due to the proliferation of electric vehicles (EV). A large amount of lithium-ion batteries are expected to reach their end-of-life (EOL) within five to seven years. The improper disposal of EOL lithium-ion batteries generates enormous amounts of flammable and explosive hazardous waste. Therefore, cost-effectively recycling LIBs becomes urgent needs. Lithium nickel cobalt manganese oxides (NCM) are one of the most essential cathode materials for EV applications due to their long cycle life, high capacity, and low cost. In 2008, 18.9% of Lithium-ion batteries used NCM cathode material worldwide while this number increased to 31% six years later. An environment–friendly and low-cost direct recycling process for NCM has been developed in this project. The goal of this project is to recycle the EOL NCM and yield battery-grade NCM with equivalent electrochemical performance compared to virgin materials. In order to achieve this goal, four different heat treatment conditions are investigated during the direct recycling process. From the experimental results, the charge and discharge capacities of the recycled material are stable (between 151-155 mAh/g) which is similar to that of the commercial MTI NCM when sintered at 850 °C for 12 hours in the air. In addition, the cycling performance of recycled NCM is better than the commercial MTI NCM up to 100 cycles.
Kinetic Property and SS 316/Alloy 617 Corrosion Study in Molten Chloride and Fluoride Salts
Yang, Qiufeng (Virginia Tech, 2022-10-04)
This study focused on the kinetic data measurements, such as diffusion coefficient D and exchange current density i_0 of the electrochemical reactions of corrosion products (Fe, Cr and Ni ions) and corrosive species (OH-), and corrosion studies of structural materials (SS 316H and Alloy 617), including static corrosion and galvanic corrosion, in molten MgCl2-NaCl-KCl and/or NaF-KF-UF4-UF3 salts in a temperature range of 600 to 800C. The study applied the semi-differential (SD) analysis method and innovative fitting method for the kinetic property data measurements in the multicomponent system of NaF-KF-UF4-UF3 salts. In molten MgCl2-NaCl-KCl salts, the measured D_(OH^- ) has the largest value followed by D_(〖Cr〗^(2+) ), D_(〖Fe〗^(2+) ), D_(〖Cr〗^(3+) ) and D_(〖Ni〗^(2+) ) at the studied temperatures, and none of the diffusion coefficients depends on the ion concentration in the studied concentration range and all of them followed the Arrhenius law. At the same temperature, the measured D_(Fe^(2+) ) and D_(〖Cr〗^(2+) ) values in molten NaF-KF-UF4-UF3 salts were slightly smaller than those obtained in molten MgCl2-NaCl-KCl salts. The non-linear curve fitting technique was applied to determine the exchange current density i_0, charge transfer coefficient α, limiting current density i_L and standard rate constant k^0 values. i_0 and k^0 followed the Arrhenius law. The obtained fundamental data can be applied to corrosion models which make the corrosion rate prediction possible in a static system from the experimental kinetic data. Corrosion studies of SS 316H and Alloy 617 in thermal purified molten NaF-KF-UF4-UF3 salts were performed for 120 hours. Based on the post-test analysis, the major metal species corrosion products were Cr, Fe and Mn in SS 316H tests, and Cr, Co, Ni in Alloy 617 tests. The measured UF4/UF3 ratio increased after corrosion tests because some of the U3+ was oxidized to U4+ by corrosive impurities and corrosion products during tests. Cr depletion and salt penetration were observed at grain boundaries (GBs) for both SS 316H and Alloy 617. For Alloy 617 specimens, the corroded area could be divided into two parts: the first part (near the surface) where Cr was completely depleted, and the second part (underneath the first part) where Cr was partially depleted. For SS 316H specimens, the average attack depth was larger than that of Alloy 617. Mo segregation was observed in the matrix of SS 316H specimens but was found to be enriched at GBs in the second part of Alloy 617 specimens. The corrosion study of Alloy 617 with time was also conducted for 72 hours and 32 hours, respectively. A thin layer composed of Fe, Co, Ni and Mo was found on the surface of the specimen, which was different from the previous 120-hour tests. In the salt, the concentration of Cr kept increasing with time, while for the other identified corroded elements, i.e., Fe, Co, Ni and Mo, their concentrations increased first, then decreased until becoming zero or stable. In the galvanic corrosion study of Alloy 617/graphite in molten NaF-KF-UF4-UF3 salts, the galvanic corrosion rate of Alloy 617 at 750C was about four times of that at 650C in the 2-hour tests, which indicated that temperature has a significant effect on the galvanic effect. In the 120-hour galvanic corrosion test, the galvanic corrosion rate became slightly larger with time in the studied system. Similar to the previous 120-hour Alloy 617 corrosion test, the corroded area of the post-test specimen was divided into two parts. The measured attack depth in both parts were much smaller compared with that in the 120-hour Alloy 617 test. This was because of the lower corrosive impurity concentrations in the salt used in the test. The salt in the galvanic corrosion test has been used in the previous corrosion test, during which the corrosive impurities were consumed, which made the salt less corrosive. Finally, it is necessary to point out that all the salts used in the present work were only thermally purified, which is effective in the removal of moisture but not in the removal of oxide impurities. Therefore, further studies are needed to understand the oxides' impacts on the corrosion behavior, especially on the salt penetration.
Material Corrosion by Nuclear Reactor Coolants
Leong, Amanda (Virginia Tech, 2022-09-19)
This work investigated material corrosion by nuclear reactor coolants, including pressurized water reactor (PWR) coolant, boiling water reactor (BWR) coolant, high-temperature steam, lead-bismuth eutectic (LBE), and molten salt. Novel cladding materials for accident tolerant fuel (ATF) and additive manufacture (AM) Ni-based alloy were studied in water coolants. Similarly, the ATF material and Ni-based alloys were also examined under high-temperature steam to understand the corrosion behavior in beyond design basis accident (BDBA) scenarios. In addition to isothermal corrosion, stress corrosion cracking (SCC) and oxide layer in situ measurements were also conducted. Unlike conventional studies in liquid LBE that focused on Fe-based alloys, the present studies also investigated Ni-based alloys to explore the Ni content effects on the corrosion by LBE at high temperatures under saturated oxygen conditions. In molten salt environments, the corrosion behaviors of both Ni-based and Fe-based alloys were investigated. This study developed a redox potential range for mitigating corrosion by using a redox couple of UF4 /UF3 and a novel approach of potential measurements against F2/ F- potential experimentally.
Material Degradation Studies in Molten Halide Salts
Dsouza, Brendan Harry (Virginia Tech, 2021-04-16)
This study focused on molten salt purification processes to effectively reduce or eliminate the corrosive contaminants without altering the salt's chemistry and properties. The impurity-driven corrosion behavior of HAYNES® 230® alloy in the molten KCl-MgCl2-NaCl salt was studied at 800 ºC for 100 hours with different salt purity conditions. The H230 alloy exhibited better corrosion resistance in the salt with lower concentration of impurities. Furthermore, it was also found that the contaminants along with salt's own vaporization at high temperatures severely corroded even the non-wetted surface of the alloy. The presence of Mg in its metal form in the salt resulted in an even higher mass-loss possibly due to Mg-Ni interaction. The study also investigated the corrosion characteristics of several nickel and ferrous-based alloys in the molten KCl-MgCl2-NaCl salt. The average mass-loss was in the increasing order of C276 < SS316L < 709-RBB* < IN718 < H230 < 709-RBB < 709-4B2. The corrosion process was driven by the outward diffusion of chromium. However, other factors such as the microstructure of the alloy i.e. its manufacturing, refining, and heat-treatment processes have also shown to influence the corrosion process. Lowering the Cr content and introducing W and Mo in the alloy increased its resistance to corrosion but their non-uniform distribution in the alloy restricted its usefulness. To slow-down the corrosion process, and enhance the material properties, selected alloys were boronized and tested for their compatibility in the molten KCl-MgCl2-NaCl salt. The borided alloys exhibited better resistance to molten salt attack, where the boride layer in the exposed alloy was still intact, non-porous, and strongly adhered to the substrate. The alloys also did not show any compensation in their properties (hardness). It was also found that the boride layer always composed of an outermost silicide composite layer, which is also the weakest and undesired layer as it easily cracks, breaks, or depletes under mechanical and thermal stresses. Various different grades of "virgin" nuclear graphites were also tested in the molten KF-UF4-NaF salt to assist in the selection of tolerable or impermeable graphites for the MSR operational purposes. It was found that molten salt wettability with graphite was poor but it still infiltrated at higher pressure. Additionally, the infiltration also depended on the pore-size and porosity of the graphite. The graphite also showed severe degradation or disintegration of its structure because of induced stresses.
Mean-Field Parameter Study of Radiation-Induced Segregation in a Binary Metal Alloy
Chan, Ryan James (Virginia Tech, 2020-01-29)
The purpose of this thesis is to broaden the tools and knowledge available for understanding the behavior of metals under irradiation to aid in the pursuit of advanced materials for deployment in Generation IV (Gen-IV) nuclear reactor designs. A mean-field study is conducted on a body-centered cubic (BCC) A-B binary metal alloy system. The performance of the simulated metal system is measured by assessing the degree of segregation that occurs at the grain boundary (GB) in the center of the one-dimensional simulation box. This mean-field method was developed using rate theory equations to observe the diffusion of defects and solute atoms in the binary BCC alloy modeled after a section of planes in the <100> direction of α-iron. The method in this thesis is adapted from a previous radiation-induced segregation (RIS) study that was similarly validated against thermal segregation isotherms. This adapted simulation code was used to study RIS by varying the initial values and conditions across ranges relevant to Generation IV reactor designs. The simulations run with this code were centered around segregation energy and the diffusion coefficient relationships between defects and solute atoms. The most influential conditions applied to both the segregation energy and diffusion coefficient relationship test suites were the temperature and dose rate. The interplay of the various segregation energies, manipulated diffusion coefficients, temperatures, and dose rates is explored in this thesis. The code used in this thesis is presented as a modular framework for further parameter study with a clear direction for more complex alloys.
Modeling of Effect of Alloying Elements on Radiation Damage in Metallic Alloys
Zhang, Yaxuan (Virginia Tech, 2020-05-26)
Metallic alloys are important structural and cladding materials for current and future reactors. Understanding radiation-induced damage on metallic alloys is important for maintaining the safety of nuclear reactors. This dissertation mainly focuses on radiation-induced primary damage in iron-based metallic. Systematic molecular dynamics simulations were conducted to study the alloying element effects on the primary damage in Fe-based alloys, including defect production and dislocation loop transformations, and their connections with defect thermodynamics. First, effects of alloying elements on the primary damage in three Fe-based ferritic alloy systems were studied, with a particular focus on the production behaviors of solute interstitials. The production behaviors of solute interstitials include over-production or under-production, compared with their solute concentration in the Fe matrix. The three alloy systems are: (1) a Fe-Cr alloy system; (2) a Fe-Cu alloy system; and (3) an ideal but artificial Fe-Cr alloy system, which is used as a reference system. It is found that the number ratio of solute interstitials to the total interstitials is distinct in these alloys. The solute interstitials are over-produced in the Fe-Cr systems but under-produced in the Fe-Cu system, compared with solute composition in the alloys. The defect formation energies in both dilute and concentrated alloys, interstitial-solute binding energies, liquid diffusivities of Fe and solute atoms, and heat of mixing have been calculated for both Fe-Cr and Fe-Cu alloys. Among these factors, our analysis shows that the relative thermodynamic stability between Fe self-interstitials and solute interstitials plays the most important role on the production behaviors of solute interstitials. Next, to obtain a correlation that can quantitatively estimate the solute interstitial fraction in the Fe-based alloys, molecular dynamics simulations were conducted to simulate the cascade damage in a series of "artificial" Fe-Cr alloys with tunable binding energies between a substitutional solute (Cr) atom and a Fe self-interstitial atom (SIA). To achieve this, the Fe-Cr cross pair interaction in the interatomic potential was modified by multiplying a scaling factor so that the solute-SIA binding energy varies linearly from positive to negative values. It is found that the solute interstitial fraction has a strong correlation with the solute-SIA binding energy, and the correlation can be approximately described by a Fermi-Dirac-Distribution-like equation. The independent defect production results reported in literature are found to align well with this correlation. The correlation may be used to estimate the solute interstitial fraction in a wide range of Fe-based alloys simply based on the solute-SIA binding energy, without conducting laborious cascade simulations. Furthermore, primary damage was further investigated in Fe-tungsten (W) alloys to investigate the atomic size effect. The large difference in atomic size between Fe and W can introduce both global volume expansion and local lattice distortion in the Fe matrix. In order to understand how oversized W influences the defect production behaviors in Fe-based alloys, molecular dynamics simulations were conducted to study the primary damage in three systems at 300 K: (a) unstrained pure Fe, (b) Fe-5at.%W alloy, and (c) strained pure Fe with the same volume expansion as the Fe-5%W. The investigation of defect production behaviors include the production of Frenkel pairs, and cluster formation preference. Based on the total number of Frenkel pairs, it indicates that the global volume expansion introduced by oversized W and external strain can lead to enhanced defect production. Meanwhile, the defect cluster analysis in all three systems indicates that the local lattice distortion induced by oversized W can significantly influence the morphologies and size distributions of defect structures. Defect formation energies were calculated to interpret the different defect production behaviors in these systems. Finally, radiation can produce not only point defects but also both <100> and ½<111> type dislocation loops in pure Fe and Fe-Cr alloys. However, contradictory experimental results have been reported on how the Cr concentration affects the ratio of <100> to ½<111> dislocation loops. In this section, molecular dynamics simulations were conducted to study how Cr concentration affects the formation probability of <100> dislocation loops from overlapping cascades on a pre-existing ½<111> dislocation loop in a series of Fe-Cr alloys with 0 – 15%Cr at 300 K. Our atomistic modeling directly demonstrates that the ratio of <100> to ½<111> dislocation loops decreases with the increasing Cr concentration, which is consistent with many experimental observations. Next, independent molecular statics calculations show that the formation energies of both <100> and ½<111> dislocation loops increase with the increasing of Cr content. However, the former has a much faster increase rate than the latter, indicating that the formation of <100> loops becomes energetically more and more unfavorable than ½<111> loops as the Cr content increases. The results provide a thermodynamics-based explanation for why Cr suppresses the formation of <100> dislocation loops in Fe-Cr alloys, which can be applied to all <100> loop formation mechanisms proposed in literature. The possible effects of other alloying elements on the formation probability of <100> loops in Fe-based alloys are also discussed.
Mullite Membrane Reference Electrode Evaluation and Application for Ni-Cr Corrosion Behavior in High Temperature Chloride Salts
Meilus, Emily Vanda (Virginia Tech, 2023-06-28)
Molten salt reactors (MSRs) using chloride-based salt-matrixes as coolants or fuels are a promising option for advanced nuclear reactors, but the extreme temperatures and corrosivity of molten salts pose a challenge for implementation. Molten MgCl2-NaCl-KCl is a viable candidate for MSRs that is considered in this work. Thermochemical properties are derived from electrochemical tests that aid in characterizing the properties of salts. To study these properties, some work has proposed using a three-electrode system with a reference electrode housed in a ceramic membrane. This research aims to develop a stable high-temperature reference electrode using a ceramic membrane that is then applied to develop an on-line monitoring system of Ni-Cr alloy corrosion in chloride salt. A mullite tube used as the membrane of a Ni(II)/Ni reference electrode in molten MgCl2-NaCl-KCl is studied. The performance of two different membrane thicknesses (1.325mm and 0.255mm) was studied in temperature ranges from 635oC to 835oC and data collected on the calculated formal potential of the Ni(II)/Ni system. Tests indicated that the results were stable and repeatable, and the formal potential for both systems differed from the previous experimental data by 0.12V at most, indicating that the system can be applied as an effective reference electrode. Using the reference electrode, on-line monitoring the corrosion of Ni-15wt.%Cr, Ni-20wt.%Cr, and Ni-30wt.%Cr was studied for 120 hours in MgCl2-NaCl-KCl. The on-line measurements showed the concentration changes of dissolved Cr and Ni by corrosion in the bulk molten salt. This work confirms that Ni(II)/Ni reference electrodes with a mullite tube membrane are stable and effective in molten chloride salt systems, particularly MgCl2-NaCl-KCl. The mullite membrane prepared by the manufacturer may be used directly for electrochemical applications without polishing, simplifying the reference electrode manufacturing process, and making it easier to replicate. The use of a Ni(II)/Ni reference electrode provides an avenue to study a different range of salt systems than previous reference electrodes allowed, particularly alloys in chloride salts at high temperatures. This work also confirms that the mullite tube may be used to perform on-line analysis of alloy corrosion in high temperature molten chloride salts. The study of Ni-Cr alloys in chloride salts better prepares the nuclear industry to select coolant salts and alloy containers with the best set of thermochemical and corrosion resistant characteristics for MSRs.
Reduction of Solid Uranium Dioxide in Calcium Salts
Karakaya, Nagihan (Virginia Tech, 2022-07-01)
Nuclear energy has gained crucial importance since it has a minor impact on climate change and greenhouse gas releases; additionally, the other energy sources are insufficient to reach the world's energy needs without nuclear energy. Another sign that the Generation IV International Forum (Kelly, Gen IV International Forum: A decade of progress through international cooperation, 2014) has pointed out is to utilize uranium resources to the maximum and recycle spent nuclear fuel through burn-up in the Generation IV reactor designs, one of which is the molten salt reactor (MSR). Therefore, the MSR can use the spent nuclear fuel as a fresh fuel when the actinides recycle. That reprocessing of spent fuel could be one of the opportunities to contribute to future nuclear energy goals. This study aims to develop a modified pyroprocessing method to prepare molten salt fuels for MSR from spent oxide nuclear fuel that was burned in light water reactors (LWRs). The process diagram illustrated as (1) spent fuel treatment, (2) chopping and voloxidation of spent oxide fuel, (3) oxide reduction of spent fuel, and then depending on the fuel structure and composition for the MSR, it continues by one or two of the following; – electrorefining, – chlorination, and – fluorination. The subject of this study focused on oxide reduction in two categories: chemical reduction and electrochemical reduction. The system designs have been optimized in calcium salts since they have high calcium metal and calcium oxide solubility. The significant results indicated that both methods would substantially reduce the solid uranium dioxide pellet. The chemical reduction will reduce the total solid pellet at 850oC in the composition of 55.73mol%CaCl2-12.37mol%CaF2-26.58mol%Ca-5.32mol%UO2 over 12 hours. The total reduction in the electrochemical test is seen at 850oC during 12 hours with a salt composition of 79mol%CaCl2-17mol%CaF2-4mol%CaO. These oxide reduction mechanisms are convenient ways to reprocess spent oxide fuel from LWRs to utilize in the MSR. Additionally, the reduced fuel is also applicable to using other next-generation reactors. The prospect of this research is the explicit comparison between chemical and electrochemical methods in calcium salts.
Rotating Disk Electrode Design for Concentration Measurements in Flowing Molten Chloride Salts
Sullivan, Kelly Marie (Virginia Tech, 2022-07-25)
Over the past several years as interest in cleaner energy sources has grown nuclear power has come to the forefront. However, as interest in nuclear power grows so does the concern over the amount of high-level radioactive waste produced. Currently, the most popular way to deal with spent nuclear fuel is interim storage until a viable treatment option becomes available. Simply waiting for spent fuel to become safe to handle will take thousands of years and is not a reasonable long-term solution. We will soon run out of space in our spent fuel pools and while more dry storage space can be found it is not an ideal solution. One answer to this problem is the reprocessing of spent nuclear fuel. This could be done with either the plutonium uranium reduction extraction (PUREX) method or the pyroprocessing method. Since PUREX does not have the same level of built-in proliferation resistance as pyroprocessing, pyroprocessing is starting to be seen as a good alternative method. Pyroprocessing would take the spent nuclear fuel from a light water reactor and make it into a metal-based fuel that could be used in certain advanced reactors. Molten salt reactors are of particular interest when it comes to reprocessing spent nuclear fuel because of their unique property of using a liquid fuel. Molten salt reactors and spent fuel reprocessors could be directly connected which would save both time and money as little storage and transportation would need to be considered. Regardless of how and where the used nuclear fuel is being recycled it is important to be able to keep track of the major actinides and fission products in the fuel as it moves through the process. Electrochemical concentration measurements are straightforward and well understood in static cases when there is only a single element to consider. When additional elements are added, or the system is flowing rather than static, things get slightly more complicated but are still decently well understood. However, in the case of spent fuel reprocessing the system is both be flowing and contains much more than a single element. This case is not well understood and is what this study attempts to understand. Two different rotating electrodes were designed to simulate flowing conditions in an electrochemical cell. The first was a tungsten rotating disk electrode (RDE) and the second was a graphite RDE. We were not able to fully insulate the tungsten RDE and were therefore unable to achieve reliable results. Because of this the tungsten design was put aside in favor of the graphite design, which did prove to be sufficiently insulated. The graphite RDE was tested in two different salt systems: LiCl-KCl-NiCl2-CrCl2 and LiCl-KCl-EuCl3-SmCl3. In the nickel-chromium system the graphite RDE produced the expected results. The calculated nickel concentration was found to be within 10% of the measured concentration. Calculations of the chromium concentration, however, were not possible due to the deposition of nickel on the graphite surface, which increased the surface area of the working electrode. When the graphite RDE was tested in the second system it was first tested in the ternary salt LiCl-KCl-EuCl3 and was able to produce decent results. The concentration of europium calculated from the scan was within 10% of the measured value. When the RDE was tested in the LiCl-KCl-EuCl3-SmCl3 salt the results did not come out as expected. Several rather noisy CV curves were obtained and no alterations to the cell seemed to affect them. At this point it was determined that the reason for the confused scans was a connection problem that could not be remedied within the time frame of this study. While this study does not accomplish the task it set out to do, it is a good step in the direction toward understanding flowing systems containing more than a single element of interest and has successfully designed a reliable graphite RDE.
Separation and Properties of La₂O₃ in Molten LiF-NaF-KF Salt
Yang, Qiufeng (Virginia Tech, 2018-12-21)
Studies on nuclear technology have been ongoing since nuclear power became uniquely important to meet climate change goals while phasing out fossil fuels. Research on the fluoride salt cooled high temperature reactor (FHR), which is funded by the United States Department of Energy (DOE), has developed smoothly with the ultimate goal of a 2030 deployment. One challenge presented by FHR is that the primary coolant salt can acquire contamination from fuel failure and moisture leaking into the system. If contamination happens, it will result in a low concentration of fission products, fuel, transuranic materials and oxide impurities in the coolant. These impurities will then affect the properties of the molten salt in the long term and need to be removed without introducing new impurities. Most of the research conducted recently has focused on impurity separation in chloride molten salts. More research urgently needs to be conducted to study the impurity separation method for the fluoride molten salts. In this study, the La₂O₃-LiF-NaF-KF (La₂O₃-FLiNaK) system is used to demonstrate impurity separation in molten fluoride salt. Since lanthanum oxide needs to be dissolved in the fluoride molten salt and studies in this field are still not complete, the solubility of lanthanum oxide in FLiNaK have been measured at different temperatures to obtain the temperature-dependent solubility and understand the corresponding dissolution mechanisms first. In the solubility related experiments, Inductively Coupled Plasma Mass Spectrometry (ICP-MS) is utilized to analyze the concentration of lanthanum ions in the molten FLiNaK salt, while X-ray powder diffraction (XRD) was applied to determine the phase patterns of molten salt. Second, electrochemical experiments with tungsten and graphite as working electrodes were conducted individually to demonstrate the separation of the dissolved oxide from the salt. When the tungsten working electrode was applied, the lanthanum ions were reduced to lanthanum metal at the tungsten cathode, while the fluorine ions reacted with the tungsten anode to form tungsten fluoride. In the experiments, the production of tungsten fluoride could lead to increasing current in the cell, even overload. Moreover, theoretically, tungsten fluoride WF4 is soluble in the fluoride salt thus introducing new impurities. All these issues make tungsten not the best choice when applied to the separation of oxygen ions. Therefore, another common working electrode graphite is used. It not only has all the advantages of tungsten, but also has good performance on separation of oxygen ions. When the graphite electrode was applied, the lanthanum ions were separated in the form of lanthanum carbide (LaC₂), while the oxygen ions can be removed in the form of carbon dioxide (CO₂) or carbon monoxide (CO). In addition, only graphite was consumed during the whole separation process, which is why the graphite anode electrode is called the “sacrificial electrode”. Third, First Principle Molecular Dynamics (FPMD) simulations with Vienne Ab initio Simulation Package (VASP) was conducted to study the properties of the fluoride molten salt. In this study, the structure information and enthalpy of formation were obtained. Generally, the simulation process can be divided into four steps: (1) the simulation systems are prepared by packing ions randomly via Packmol package in the simulation cell; (2) an equilibrium calculation is performed to pre-equilibrate the systems; (3) FPMD simulations in an NVT ensemble are implemented in VASP; (4) based on the FPMD simulations results, the first peak radius and the first-shell coordination number were evaluated with partial radial distribution function (PRDF) analysis to determine the statistics of molten salt structure information, while the transport properties, e.g., the self-diffusion coefficient was calculated according to the function of mean square displacement (MSD) of time generated by the Einstein-Smoluchowshi equation. The viscosity and ionic conductivity were obtained by combining the self-distribution coefficient with the Einstein-Stokes formula and Nernst-Einstein equation.
Species Chemistry and Electrochemical Separation in Molten Fluoride Salt
Wang, Yafei (Virginia Tech, 2019-09-11)
Fluoride salt-cooled high-temperature reactor (FHR) is a safer and potentially less expensive alternative to light water reactor due to the low pressure of primary system, passive decay heat cooling system, chemically inert coolant salt, and high-temperature power cycle. However, one challenge presented by this reactor is that fission products may leak into the primary system from its TRISO particle fuel during normal operation. Consequently, the circulating fission products within the primary coolant would be a potential radioactive source. On the other hand, the containment material of the molten salt such as nickel-based alloys may be corroded, and its species may stay in the salt. Thus, the installment of the primary coolant clean-up system and the study on the contaminant species' chemistry and separation are necessarily needed. Electrochemical separation technique has been proposed as the online coolant clean-up method for FHR for removing some impurities from the salt such as lanthanides and corrosion products. The present research focuses on the electrochemical separations of fission products and corrosion products in molten FLiNaK salt (46.5LiF-11.5NaF-42KF mol%) which is the surrogate of the primary coolant candidate FLiBe (67LiF-33BeF2, mol%) for FHR. The main objective is to investigate the electrochemical behaviors of fission products and corrosion products in molten FLiNaK salt to achieve its separations, and provide fundamental properties to instruct the conditions needed to be applied for a desired electrochemical separation. La and Ce are two main elements concerned in this study since they are major lanthanide fission products. Electrochemical behavior of LaF3 in molten FLiNaK salt was studied on both W and Mo inert working electrodes. Although the standard reduction potential of La (III) is more cathodic than that of the primary salt melt constituents K (I) and Na (I), the electrochemical separation of La from molten FLiNaK salt was achieved by merely using inert working electrode because of the formed LaF63- when KF or NaF exists as the salt constituents. Fundamental properties of La in molten FLiNaK salt were also studied at various situations by electroanalytical methods including cyclic voltammetry (CV), chronopotentiometry (CP), and potentiodynamic polarization scan (PS). Ce is another fission product to be separated out from molten FLiNaK salt. Both inert (W) and reactive working electrodes (Cu and Ni) were utilized to realize the extraction of Ce. The electrochemical behaviors of Ce observed on inert W electrode are similar to the ones obtained in FLiNaK-LaF3 system. Reactive electrodes Cu and Ni were used to precede the electrochemical deposition potential of Ce by forming intermetallic compounds. It turned out only Ni electrode was feasible for preceding the deposition potential and the intermetallic compound was identified as CeNi5. The dissolution of chromium metal in the form of chromium fluoride into molten FLiNaK salt is the main concern of alloy corrosion in FHR. To understand the alloy corrosion and removal of the corrosion products from the FHR salt coolant, the electrochemical behavior and fundamental properties of Cr in molten FLiNaK salt were investigated in the present study as well. A new analysis method for the Cr two-step electrochemical reaction in the salt was developed. The method can be applied to other two-step reactions as well. Liquid bismuth was proposed to be the extraction media for liquid/liquid multistage separation of fission products in molten salt reactor. It also can be used as the cathode to extract the fission product of which the electrodeposition potential is close to or more negative than that of the main constituents of molten salt. Activity and activity coefficient are essential factors for assessing the extraction behavior and viability of bismuth in separating fission products. Hence, in the present study, the activity and activity coefficient of fission products and alkali metals (Li and K) at different concentrations and temperatures were determined by experiment and simulation methods respectively. To conduct the parametric study for the electrochemical reaction process and predict fundamental properties, an electrochemical model including single-step reversible, irreversible, and quasi-reversible reactions, multiple-reaction, and two-step consecutive charge transfer reaction was developed based on MOOSE. Although the model was not applied to analyze the experimental data in the present study, this model provides an efficient and easy way to understand the effect of various parameters on electrochemical reaction process. The present study supplied a comprehensive study on the electrochemical separation of fission products and corrosion products in molten FLiNaK salt and will contribute greatly to the development of FHR.
Studies of Used Fuel Fluorination and U Extraction Based on Molten Salt Technology for Advanced Molten Salt Fuel Fabrication
Davis, Brenton Conrad (Virginia Tech, 2023-12-14)
This study focuses on techniques that can be used to fuel next generation reactors. The first two studies are new techniques for recycling used nuclear fuel (UNF) and the third is a method of separating uranium (U) from lithium fluoride (LiF) and thorium fluoride (ThF4) salt also known as FLiTh for a thorium (Th) fuel cycle. The first technique proposed for UNF recycling was to use the cladding as an anode to oxidize the zircaloy and dissolve it into a LiF, sodium fluoride (NaF), zirconium fluoride (ZrF4) salt. Zirconium (Zr) was also reduced and deposited on a tungsten (W) cathode at the same time transporting the Zr through the salt. As commercial zircaloy would be contaminated with UNF oxides, and the oxides will not oxidize as part of the electrochemical process, they would be left at the anode as the Zr is dissolved away. This means the deposited Zr, on the cathode, can be disposed of as low-level waste (LLW) or recycled back into the nuclear industry instead of being stored as high-level waste (HLW). The next technique was fluorination of UNF oxides using ZrF4. Using the same LiF-NaF-ZrF4 salt, uranium oxide (UO2), lanthanum oxide (La2O3), and yttrium oxide (Y2O3) were fluorinated into uranium fluoride (UF4), lanthanum fluoride (LaF3), and yttrium fluoride (YF3). By sampling and recording the change in concentration over time, the reaction rate of all three oxides was determined and a temperature dependent reaction rate was reported from 500°C to 650°C. A zirconium oxide (ZrO2) product layer developed on UO2, but it only slowed down the fluorination process but did not stop it. UO2 and Y2O3 fluorinated entirely but La2O3 did not. The solubility limit of LaF3 in the salt was determined to be the reason the reaction did not go to completion. The last technique was the electrochemical separation of U from FLiTh, to simulate irradiated Th that decays to protactinium (Pa). A constant, albeit small current, was used to deposit U on a W electrode without Th depositing with it. A liquid metal bismuth (Bi) electrode was used as well, and a constant current resulted in Th depositing with the U. To get just U to deposit, the current needed to be applied for a time and then no current applied for a time so the system could reach equilibrium. By cycling these two steps it was possible to get U to deposit in Bi without Th.
Studies on Molten Salt Fuels: Properties, Purification, and Materials Degradation
Park, Jaewoo (Virginia Tech, 2024-04-12)
The molten salt reactor (MSR) is one of the advanced nuclear reactors expected to be alternatives to the conventional water-cooled nuclear reactor systems. Despite many advantages of MSRs, properties of molten salts have not been sufficiently measured in previous studies. In addition, the corrosion of structural alloys by molten salt is the biggest challenge for the operation of MSRs. This study focuses on measurements of thermophysical and thermodynamic properties of fluoride salt fuels, salt purification, and the degradation of structural materials in static and flowing molten-salt fuels. For the measurements of properties, phase transition, specific heat capacity, vapor pressure, contact angle on nuclear-grade graphite, and density were measured. The methodologies for the property measurements used in this study were validated by measuring the properties of metals or salts that have been well studied. For the flow-induced corrosion tests, the salt flow with different velocities was simulated by rotating the stainless steel 316H (SS316H) specimens in molten NaF-KF-UF4 (FUNaK) contained in glassy carbon crucibles at 1073 K. Salt samples were intermittently collected to monitor concentration changes of corrosion products in the salt, and surfaces and cross-sections of post-test SS316H specimens were analyzed to study their corrosion behaviors. Different batches of FUNaK were synthesized using different methods of purification, such as thermal purification, U-metal purification, and hydrofluorination with electrochemical purification (chemical purification) to study impacts of salt purification on the corrosion of SS316H. The corrosion test of SS316H by thermally purified FUNaK showed that the Fe concentration increased at the beginning and then decreased while the Cr concentration continued increasing while the rate decreased. In addition, (Cr, Fe)7C3 layers, Cr-metal particles, and dendritic structures concentrated with Cr and Fe were observed on the glassy carbon crucible after the 2 m/s test. The U-metal purification and hydrofluorination with electrochemical purification reduced concentrations of oxygen and hydrogen in FUNaK and mitigated the corrosion of SS316H significantly. The infiltration of the fluoride fuel salts into graphite and the fluorination of graphite by the salts at different pressures and temperatures were also studied. The salt infiltration into graphite at pressures above its threshold pressure was observed, and the formation of carbon fluorides on the surface of post-test graphite specimens was identified.
Thermal Properties of Candidate Coolant Salts
Ridder, Cathleen Elise (Virginia Tech, 2024-07-23)
With the increasing research on advanced reactors, molten salt reactors have been recognized for their potential. As with any advanced reactor concept, each component and material must be thoroughly investigated before any reactors of that type are created. One of the most pressing issues in MSR research is that of the salts themselves. Though there are a multitude of salts to choose from when designing such a reactor, many of these salts lack the extensive research required to fully understand them. Across the decades there have been many studies that have investigated select molten salts, but there are a few problems with many of those studies. Those problems are the following: prior papers use obsolete and less reliable methods for their measurements, the papers don't investigate the salts across a wide enough range of temperatures nor at varying compositions, and finally many of the salts that are seen as candidates today were not given as much attention when molten salt reactors were first conceptualized which has resulted in a lack of research on them. Indeed, the research into these salts is lacking in many ways. This study seeks to investigate a collection of promising coolant salts in depth with acknowledgment to those past studies. LiF-NaF-KF (46.5-11.5-42.0 mol%) will be used as a calibration standard and for the purpose of verifying our methodology. Specifically, FLiNaK was used in the development of volume-height curves as calibration for density measurements. NaOH-KOH of four different compositions ( 0.5-0.5mol%, 0.55-0.45mol%, 0.6-0.4mol%, and 0.65-0.35 mol%) will be evaluated for their densities and heat capacities. And finally, BeF2-NaF(43-57mol%) will be evaluated within the question of if the properties are desirable enough that the dangers posed by beryllium are an acceptable risk. BeF2-NaF will have melting point, heat capacity, density, and vapor pressure measurements performed. Additionally, extensive impurity analysis and removal (via an HF gas system) was done to our BeF2-NaF samples. The melting point and heat capacity were evaluated using dynamic scanning calorimetry (DSC), the vapor pressure was evaluated using thermogravimetric analysis (TGA), and the density was measured using a system similar to the Arrhenius method that measures height.

Browsing by Author "Zhang, Jinsuo"

Results Per Page

Sort Options