SEARCH · Search NASA
Results for “Pyroprocessing”
Search indexed NASA NTRS and DOE OSTI research on propulsion, heat transfer, battery materials and energy systems. Follow report and document links to the original sources.
Quote a phrase for an exact phrase match. Source license links do not imply unrestricted reuse.
Component effects on crystallization of RE-containing aluminoborosilicate glass
Lanthanide-aluminoborosilicate (LABS) glass is one option for immobilizing rare earth (RE) oxide fission products generated during reprocessing of pyroprocessed fuel. This glass system can accommodate a high loading of RE oxides and has excellent chemical durability. The present study describes efforts to model equilibrium crystallinity as a function of glass composition and temperature as well as liquidus temperature (TL) as a function of glass composition. The experimental method for determining TL was ASTM C1720-11. Typically, three crystalline phases were formed in each glass: Ce-borosilicate (Ce 3 BSi 2 O 10 ), mullite (Al 10 Si 2 O 19 ), and corundum (Al 2 O 3 ). Cerianite (CeO 2 ) was a common minor crystalline phase and Nd-silicate (Nd 2 Si 2 O 7 ) occurred in some of the glasses. In the composition region studied, TL decreased as SiO 2 and B 2 O 3 fractions increased and strongly increased with increasing fractions of RE oxides; Al 2 O 3 had a moderate effect on the TL but, as expected, it strongly affected the precipitation of Alcontaining crystals.
Calculated thermodynamic properties of GdCl 3 in LiCl-KCl eutectic molten salt
The thermodynamic behaviors of lanthanides in molten salt are significant for the pyroprocessing of spent nuclear fuel. In this study, the ternary phase diagram of LiCl-KCl-GdCl 3 system was assessed by the CALPHAD method using two-sublattice model. The liquidus surfaces were calculated and plotted from 720 K to 1000 K. The solubility of GdCl 3 from 720 K to 850 K was determined from the phase diagram developed. The activity coefficient and apparent potential of GdCl 3 with a concentration up to 3 mol% were also investigated. They both show a dependence on the concentration. In conclusion, From 0 to 3 mol%, the apparent potential increases by about 3% and the activity coefficient increases by about 50%.
Investigation of physical and chemical properties for upgraded SAP (SiO2Al2O3P2O5) waste form to immobilize radioactive waste salt
Not Available
Liquid cadmium cathode performance model based on the equilibrium behaviors of U and Pu in molten LiCl–KCl/Cd system at 500°C
The liquid cadmium cathode is a pyroprocessing technology used for the recovery of group actinide metals from the molten chloride salts that support uranium electrorefining. This article presents a model for the performance of a liquid cadmium cathode based on considerations of salt/alloy equilibrium thermodynamics, Cd-U-Pu phase diagram behaviors, and mass balance constraints. In conclusion, the model demonstrates how the salt/cadmium system responds as incremental additions of uranium are made to the liquid cadmium cathode.
Molten salt reactor waste and effluent management strategies: A review
Molten salt reactors (MSRs) are being considered as one of the potential nuclear options to meet future energy demands. While the MSR designs are drastically different from the more traditional light water reactor, many of the waste streams are similar between the concepts. The purpose is to outline strategies for the treatment and processing of MSR-type wastes from concepts of reconditioning and recycle of certain components to partitioning and direct immobilization of other waste components. To help bridge science and technology gaps, knowledge gained from similar efforts such as pyroprocessing of Experimental Breeder Reactor II salt wastes can be leveraged to develop concept-to-disposition pathways.
A multi-scale study of the adsorption of lanthanum on the (110) surface of tungsten
In this study, we utilize a multi-scale approach to studying lanthanum adsorption on the (110) plane of tungsten. The energy of the system is described from density functional theory calculations within the framework of the cluster expansion method. It is found that including two-body figures up to the sixth nearest neighbor yielded a reasonable agreement with density functional theory calculations as evidenced by the reported cross validation score. The results indicate that the interaction between the adsorbate atoms in the adlayer is important and cannot be ignored. The parameterized cluster expansion expression is used in a lattice gas Monte Carlo simulation in the grand canonical ensemble at 773 K and the adsorption isotherm is recorded. Implications of the obtained results for the pyroprocessing application are discussed.
Review—Fundamental Uranium Electrochemistry and Spectroscopy in Molten Salt Systems
Uranium is a key element used for nuclear energy production. Some advanced reactor designs, specifically molten salt reactors, will continue to use uranium as the fissile material for energy production. These new technologies require an intimate understanding of uranium chemistry during and after energy production. This review covers contemporary research on the coordination chemistry and behavior of uranium with the coolant and pyroprocessing salts as proposed for use in future reactor designs. Discussed topics include the nature of U redox reactions involving the reduction of U(III) to U metal and oxidation of U(III) to U(IV). These systems have been interrogated using cyclic voltammetry, chronopotentiometry, and optical and X-ray absorption spectroscopies. Insights obtained into the electrode potentials, the uranium species, and their diffusion coefficients in alkali halide melts from decades of research are summarized selectively. Further, perspectives are provided on the importance of unifying studies for comparison across multiple institutions. The application of synchrotron radiation research and multimodal approaches involving two (or more) probes, such as the widespread combination of UV–visible spectroscopy and electroanalysis known as spectroelectrochemistry, can provide new knowledge about the main process of uranium electrorefining—diffusion, as will be demonstrated in this review through the lack of comparable results.
Application of Cyclic Voltammetry with W Electrodes for Measurement of High CaO Concentration in Molten CaCl 2
Molten CaCl 2 is a useful electrolyte for certain pyroprocessing techniques, and the solubility of oxide ions in this salt is a critical characteristic of many of these techniques. CaO is a byproduct of direct oxide reduction (DOR), so the concentration of oxide ions can be used to determine the extent of these reactions. Using cyclic voltammetry (CV) with tungsten electrodes, the CaO concentration in the salt up to approximately 5 wt% was correlated to both oxidation peak height and power density values. This correlation was demonstrated in both equilibrium and transient conditions. CV is a promising method for the in situ real-time monitoring of CaO concentrations in DOR processes.
Electrochemistry of Europium(III) Chloride in 3 LiCl – NaCl, 3 LiCl – 2 KCl, LiCl – RbCl, and 3 LiCl – 2 CsCl Eutectics at Various Temperatures
Here we report the effect of changing the eutectic melt composition on the electrochemical properties of europium(III) chloride under pyroprocessing conditions. The number of electrons transferred, redox potentials and diffusion coefficients were determined using various electrochemical and spectroelectrochemical techniques in four different eutectic mixtures (3 LiCl - NaCl, 3 LiCl - 2 KCl, 3 LiCl - RbCl, and 3 LiCl - 2 CsCl) while varying the temperature of the melt. It was determined that Eu3+ undergoes a one electron reduction to Eu2+ in each melt at all temperatures evaluated. Within all the melts a positive shift in the redox potential as well as an increase in the diffusion coefficient for Eu3+ was observed as the temperature increased. Also observed was a positive shift in the redox potential and increase in the diffusion coefficient for Eu3+ as the weighted average of the cationic radii for the melt decreased.
Study of Potentiometry for Monitoring Activity of GdCl 3 in Molten LiCl-KCl Salt
In pyroprocessing spent nuclear fuels by electrorefining in molten LiCl-KCl salt, it is desired to monitor in real time the UCl 3 concentration in the salt for safeguards purposes. Current chemical analysis of the highly radioactive salt for electrorefining by an inductively coupled plasma technique is inconvenient and usually time-consuming in generating the salt composition results. In this paper, we evaluated whether a simple potentiometry approach can be used for real-time monitoring the concentration of GdCl 3 , which was used as a surrogate for UCl 3 , in LiCl-KCl-GdCl 3 salt by measuring the open circuit potential of a Gd metal electrode with respect to a Ag/AgCl reference electrode (RE) when GdCl 3 salt was incrementally added to the LiCl-KCl salt. Additions of LaCl 3 , CeCl 3 and NdCl 3 salts were used for evaluating the effects of other chloride salts on the selectivity of the Gd metal electrode vs Ag/AgCl RE. While using potentiometry to determine GdCl 3 concentrations, Gd metal was unexpectedly observed to be unstable and dissolved in LiCl-KCl salt when GdCl 3 is present.
Laser-Induced Breakdown Spectroscopy (LIBS) Measurement of Uranium in Molten Salt
In this current study, the molten salt aerosol–laser-induced breakdown spectroscopy (LIBS) system was used to measure the uranium (U) content in a ternary UCl 3 –LiCl–KCl salt to investigate and assess a near real-time analytical approach for material safeguards and accountability. Experiments were conducted using five different U concentrations to determine the analytical figures of merit for the system with respect to U. In the analysis, three U lines were used to develop univariate calibration curves at the 367.01 nm, 385.96 nm, and 387.10 nm lines. The 367.01 nm line had the lowest limit of detection (LOD) of 0.065 wt% U. The 385.96 nm line had the best root mean square error of cross-validation (RMSECV) of 0.20 wt% U. In addition to the univariate calibration approach, a multivariate partial least squares (PLS) model was developed to further analyze the data. Using partial least squares (PLS) modeling, an RMSECV of 0.085 wt% U was determined. The RMSECV from the multivariate approach was significantly better than the univariate case and the PLS model is recommended for future LIBS analysis. Overall, the aerosol-LIBS system performed well in monitoring the U concentration and it is expected that the system could be used to quantitatively determine the U compositions within the normal operational concentrations of U in pyroprocessing molten salts.
Sulfur Hexafluoride Treatment of Used Nuclear Fuel to Enhance Separations
Reactive Gas Recycling (RGR) technology development has been initiated at Savannah River National Laboratory (SRNL), with a stretch-goal to develop a fully dry recycling technology for Used Nuclear Fuel (UNF). This approach is attractive due to the potential of targeted gas-phase treatment steps to reduce footprint and secondary waste volumes associated with separations relying primarily on traditional technologies, so long as the fluorinators employed in the reaction are recycled for use in the reactors or are optimized for conversion of fluorinator reactant. The developed fluorination via SF 6 , similar to the case for other fluorinators such as NF 3 , can be used to address multiple fuel forms and downstream cycles including continued processing for LWR via fluorination or incorporation into a aqueous process (e.g. modified FLUOREX) or for subsequent pyro treatment to be used in advanced gas reactor designs such metal- or gas-cooled reactors. This report details the most recent experimental results on the reaction of SF 6 with various fission product surrogate materials in the form of oxides and metals, including uranium oxides using a high-temperature DTA apparatus capable of temperatures in excess of 1000°C. The experimental results indicate that the majority of the fission products form stable solid fluorides and sulfides, while a subset of the fission products form volatile fluorides such as molybdenum fluoride and niobium fluoride, as predicted thermodynamically. Additional kinetic analysis has been performed on additional fission products. A key result is the verification that SF 6 requires high temperatures for direct fluorination and subsequent volatilization of uranium oxides to UF 6 , and thus is well positioned as a head-end treatment for other separations technologies, such as the volatilization of uranium oxide by NF 3 as reported by colleagues at PNNL, advanced pyrochemical separations or traditional full recycle approaches. Based on current results of the research at SRNL on SF 6 fluoride volatility for UNF separations, SF 6 treatment renders all anticipated volatile fluorides studied to be volatile, and all non-volatile fluorides studied to be non-volatile, with the notable exception of uranium oxides. This offers an excellent opportunity to use this as a head-end separations treatment process because: 1. SF 6 can be used to remove volatile fluorides from a UNF matrix while leaving behind uranium oxides. Therefore an agent such as NF 3 should be able to very cleanly separate a pure UF 6 stream, leaving compounds in the bottoms such as PuF 4 , SrF 2 and CsF after the UNF matrix has been pre-treated with SF 6 . 2. Due to the fact that the uranium oxide is not separated in the volatilization step upon direct contact with SF 6 at moderately high temperatures (≤ 1000°C), this fluoride approach may be well-suited for head-end processing for Gen IV reactor designs where the LWR is treated as a fuel stock, and it is not desired to separate the uranium from plutonium, but it is desired to separate many of the volatile fission products. 3. It is likely that removal of the volatile fission products from the uranium oxide should simplify both traditional and next generation pyroprocessing techniques. 4. SF 6 treatment to remove volatile fission products, with or without treatment with additional fluorinators, could be used to simplify the separations of traditional aqueous processes in similar fashion to the FLUOREX process. Further research should be conducted to determine the separations efficiency of a combined SF 6 /NF 3 separations approach which could be used as a stand-alone separations technology or a head-end process.
Study of a Double Bubbler for Material Balance in Liquids (Training Report 2013)
The objective of this project was to determine the potential of a double bubbler to measure density and fluid level of the molten salt contained in an electrorefiner. Such in-situ real-time measurements can provide key information for material balances in the pyroprocessing of the nuclear spent fuel. This theoretical study showed this technique has a lot of promise. Four different experiments were designed and performed. The first three experiments studied the influence of a variety of factors such as depth difference between the two tubes, gas flow rate, the radius of the tubes and determining the best operating conditions. The last experiment purpose was to determine the precision and accuracy of the apparatus during specific conditions. The elected operating conditions for the characterization of the system were a difference of depth of 25 cm and a flow rate of 55 ml/min in each tube. The measured densities were between 1,000 g/l and 1,400g/l and the level between 34cm and 40 cm. The depth difference between the tubes is critical, the larger, the better. The experiments showed that the flow rate should be the same in each tube. The concordances with theoretical predictions were very good. The density precision was very satisfying (spread<0.1%) and the accuracy was about 1%. For the level determination, the precision was also very satisfying (spread<0.1%), but the accuracy was about 3%. However, those two biases could be corrected with calibration curves. In addition to the aqueous systems studied in the present work, future work will focus on examining the behavior of the double bubbler instrumentation in molten salt systems. The two main challenges which were identified in this work are the effect of the temperature and the variation of the superficial tension.
Fission Product Separation from Pyrochemical Electrolyte by Cold Finger Melt Crystallization
This work contributes to the development of pyroprocessing technology as an economically viable means of separating used nuclear fuel from fission products and cladding materials. Electrolytic oxide reduction is used as a head-end step before electrorefining to reduce oxide fuel to metallic form. The electrolytic medium used in this technique is molten LiCl-Li2O. Groups I and II fission products, such as cesium (Cs) and strontium (Sr), have been shown to partition from the fuel into the molten LiCl-Li2O. Various approaches of separating these fission products from the salt have been investigated by different research groups. One promising approach is based on a layer crystallization method studied at the Korea Atomic Energy Research Institute (KAERI). Despite successful demonstration of this basic approach, there are questions that remain, especially concerning the development of economical and scalable operating parameters based on a comprehensive understanding of heat and mass transfer. This research explores these parameters through a series of experiments in which LiCl is purified, by concentrating CsCl in a liquid phase as purified LiCl is crystallized and removed via an argon-cooled cold finger.
The Multi-Isotope Process (MIP) Monitor Project: FY13 Final Report
The Multi-Isotope Process (MIP) Monitor provides an efficient approach to monitoring the process conditions in reprocessing facilities in support of the goal of “… (minimization of) the risks of nuclear proliferation and terrorism.” The MIP Monitor measures the distribution of the radioactive isotopes in product and waste streams of a nuclear reprocessing facility. These isotopes are monitored online by gamma spectrometry and compared, in near-real-time, to spectral patterns representing “normal” process conditions using multivariate analysis and pattern recognition algorithms. The combination of multivariate analysis and gamma spectroscopy allows us to detect small changes in the gamma spectrum, which may indicate changes in process conditions. By targeting multiple gamma-emitting indicator isotopes, the MIP Monitor approach is compatible with the use of small, portable, relatively high-resolution gamma detectors that may be easily deployed throughout an existing facility. The automated multivariate analysis can provide a level of data obscurity, giving a built-in information barrier to protect sensitive or proprietary operational data. Proof-of-concept simulations and experiments have been performed in previous years to demonstrate the validity of this tool in a laboratory setting for systems representing aqueous reprocessing facilities. However, pyroprocessing is emerging as an alternative to aqueous reprocessing techniques.
Ion Selective Ceramics for Waste Separations. Input for Annual Accomplishments Report
This report discusses“Ion-Selective Ceramics for Waste Separations” which aims to develop an electrochemical approach to remove fission product waste (e.g., Cs+ ) from the LiCl-KCl molten salts used in the pyroprocessing of spent nuclear fuel.
Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP)
This document constitutes progress made on this project by the Georgia Tech led Integrated Approach to Fluoride High Temperature Reactor Technology and Licensing Challenges (FHR-IRP). This is the first draft of the project final report. Chapter 3 is focused on Tritium management. Tritium management in Fluoride-salt High-temperature Reactors (FHRs) presents a major operational concern and a significant challenge. Tritium is generated via multiple pathways including various interactions in fuel and structural materials, as well as coolant (primarily interaction with lithium and beryllium), leading to a complex evolution and spatial distribution of tritium generation rate. FHR generates orders of magnitude more tritium than in LWRs, and has potential for a significant release, via permeation through structural materials (due to high temperature), and through heat exchanger(s). Therefore, accurate estimate is necessary. For that purpose relevant tritium production pathways were identified. A methodology to obtain space-time dependent tritium generation rate was developed and implemented, employing SCALE and MCNP codes. Detailed models of the reference Advanced High Temperature Reactor (AHTR) design were developed. A multi-parametric metrics was introduced to capture relevant tritium production indicators, including the instantaneous tritium production rate, total tritium production (neglecting subsequent decay), and accumulated tritium production (accounting for the decay). The obtained results will inform tritium management, including the sizing of the components, and to gain insight on the impact of the core design parameters on tritium generation rates. Research effort on Liquid Salt Impurity Removal, Redox and Corrosion Control is discussed in Chapter 4. Section 4.1 is an introduction chapter. Based on the current available knowledge and operation experience on previous MSRs and other types of nuclear reactors, the report identifies the major impurities that are related to either material compatibility, salt properties or radioactive source terms, and the potential contamination sources of these impurities, their transport in the primary coolant loop and available cleanup methodologies for a particular group of impurity. In section 4.2, the available purification methods were discussed, including the liquid/liquid extraction using liquid Bi-Li which has been applied in MSBR and electrochemical purification which is developed based on Pyroprocessing. In this section, the new multi-stage and multi-salt-component was developed and discussed. Section 4.3 discussed the redox potential control for salt chemistry and corrosion mitigation, including the control theory and the soluble/soluble buffer control method. The section also developed method on how to analyze soluble/soluble redox reaction in a molten salt system and designed an electrochemical sensor for redox potential control. Corrosion models were also reported in this section. Section 4.4 focused on how to measure the fundamental properties in a molten salt. As an example, the section focused on La fluoride in the molten salt. The section developed methods on how analyze electrochemical measurement data to obtain the fundamental properties including the redox potential, diffusion coefficient and exchange current density. All these fundamental information is important for salt purification system development. Section 4.5 focused on measurements of corrosion products in the salts. Because of the different behaviors between Cr and La in the salt, new analysis method for Cr fluorides in the salt was developed. Some results for Fe fluoride in the salt and corrosion measurements were also reported. Section 4.6 demonstrated the separation of oxide from the salt. This section also measured the solubility of La 2 O 3 in the salt as function of temperature. Section 4.7 focused on the separation of La and Ce from the molten salt. These two elements were selected because La was found to be separated by using inert electrode, while for separation of Ce, reactive electrode must be selected. Section 4.8 demonstrated the separation iodine ions from the molten salt. Two kinds of working electrode materials (metal W and graphite) were investigated. Section 4.9 focused on the fundamental data of fission products and actinides in liquid Bi. The chapter reviewed the available data, then based a CALPHAD model, the available data were broadened to the concentration-dependent cases.