Search Results (14)

Recent advancements in tritium separation technologies have significantly improved efficiency, particularly through the integration of vapor phase catalytic exchange (VPCE), liquid phase catalytic exchange (LPCE), and combined electrolysis catalytic exchange (CECE) methods. Combining these techniques overcomes individual limitations, enhancing separation efficiency and reducing energy consumption. The CECE process, which integrates electrolysis with catalytic exchange, offers high separation factors, making it effective for high-concentration tritiated water treatment. Solid polymer electrolyte (SPE) technology has also gained prominence for its higher efficiency, smaller equipment size, and longer lifespan compared to traditional alkaline electrolysis. While electrolysis offers high separation factors, its high energy demand limits its cost-effectiveness for large-scale operations. As a result, electrolysis is often combined with other methods like CECE to optimize both energy consumption and separation efficiency. Future research will focus on improving the energy efficiency of electrolysis for large-scale, low-cost tritiated water treatment. Full article

Enriched lithium isotopes (⁶Li and ⁷Li) are essential in the nuclear energy industry, where ⁶Li is bombarded with neutrons to produce tritium for fusion reactions, while ⁷Li is used as a core coolant and pH regulator. Separation of ⁶Li and ⁷Li by electromigration is a promising method for producing enriched lithium isotopes that fulfill industrial needs. In this work, based on a previously proposed biphasic system electromigration routine, a three-stage system of ‘LiCl aqueous solution (anolyte)|B12C4-[EMIm][NTf₂] organic solution|NH₄Cl aqueous solution (catholyte)’ was constructed and the rules of lithium isotope separation and lithium-ion migration investigated. It was shown that the isotope enrichment effect of the catholyte was greatly affected by the experimental conditions, while that of the organic solution was less affected. As the B12C4 concentration increased, enhancement of ⁷Li enrichment in the catholyte and ⁶Li enrichment in the organic solution was observed, and α(C/O) and α(O/A) reached 0.975 and 1.018 at B12C4 of 0.5 mol/L. With the increase in current, migration time, and LiCl concentration, the isotope that was enriched in the catholyte trended from ⁷Li to ⁶Li (about 6 mA, 12 h or LiCl of 5 mol/L). Taking lithium-ion transport efficiency and lithium isotope separation effect into consideration together, a current of at least 6 mA, duration of at least 12 h, LiCl concentration of at least 1 mol/L and B12C4 concentration of 0.2 mol/L are suggested for the electromigration process. The work provides an important reference for system construction and experimental design of a biphasic electromigration separation method, which is expected to be an industrial alternative because of its environmental protection and high efficiency. Full article

The Liquid Phase Catalytic Exchange (LPCE) process plays a pivotal role in the separation of hydrogen isotopes, particularly in applications such as tritium removal in heavy water reactors. Effective separation is crucial for maintaining reactor safety and efficiency. In this study, the optimal operating conditions for the LPCE process were determined through orthogonal experiments and validated in different hydrogen isotope systems. The experiments investigated key operational parameters, including the filling ratio of catalyst to packing (FR), operating temperature (T), superficial gas velocity (V), and gas-to-liquid flow rate ratio (λ), using a robust L₁₆ orthogonal experiment design. The results indicated that V and FR had the most significant effects on the height equivalent to a theoretical plate (HETP), while λ exhibited the greatest impact on dedeuterization efficiency (DE). The optimal conditions obtained were V = 0.1 m/s, FR = 1:2, T = 70 °C, and λ = 2.5. Furthermore, the reproducibility of the optimal conditions was verified in LPCE columns with varying diameters (1.5 cm, 2.5 cm, 4.5 cm). Additionally, the findings were applied to both H-D and D-T separation systems, demonstrating consistency in mass transfer efficiency and validating the applicability of the optimal conditions in different hydrogen isotope separations. This research provides critical insights for optimizing tritium removal systems in heavy water reactors, contributing to enhanced reactor safety and performance. Full article

Nuclear fusion is a phenomenon that is well known within the nuclear physics community as a viable option for alternative energy as many natural gases and fossil fuels are phased out of commercial use. Deuterium and tritium fusion reactions are currently the leading candidates for nuclear fusion, with a major limiting factor being a means to produce tritium on an industrial scale. Lithium-6 is a well-known isotope that can produce tritium and helium following a fission reaction with a neutron. Unfortunately, the lithium-6 enrichment methods are limited to the COLEX process, which leaves behind an alarming amount of mercury waste as a potential environmental contaminant. Deep eutectic solvents are believed to be a potential alternative to lithium isotope separations due to the ease of generation, in addition to the minimum environmental waste generated when these solvents are employed. Previous studies have suggested that deep eutectic solvents are capable of separating lithium isotopes by utilizing a 2-thenoyltrifluoroacetone and trioctylphosphine oxide system that can biphasically react with a buffered solution containing lithium chloride. This system displays a separation factor of 1.068, which when compared to the 1.054 separation within the COLEX process, makes it a potential candidate for lithium-6/7 separation. Within this study, we investigate this system in comparison to two newly synthesized deep eutectic solvents and find that within these acetylacetone-based systems, little isotopic separation is observed. We investigate these systems both experimentally and computationally, showing the different lithium cation affinities, in addition to proposing how the electron-donating or -withdrawing nature can influence these systems. Full article

It is a challenging but key task to reduce the tritium inventory in EU-DEMO to levels that are acceptable for a nuclear regulator. As solution to this issue, a smart fuel cycle architecture is proposed based on the concept of Direct Internal Recycling (DIR), in which the Metal Foil Pump (MFP) will play an important role to separate the unburnt hydrogen isotopes coming from the divertor by exploiting the superpermeation phenomenon. In this study, we will present the assessment of the performance of the lower port of EU-DEMO after the integration of the MFP. For the first time, a thorough comparison of three different MFP (parallel long tubes, sandwich and halo) designs is performed regarding conductance for helium molecules, the pumping speed and the separation factor for deuterium molecules under different physical and geometric parameters. All simulations were carried out in supercomputer Marconi-Fusion with our in-house Test Particle Monte Carlo (TPMC) simulation code ProVac3D because the code had been parallelized with high efficiency. These results are essential for the development of a suitable MFP design in the vacuum-pumping train of EU-DEMO. Full article

Hydrogen permeation sparked a renewed interest in the second half of the 20th century due to the favorable features of this element as an energy factor. Furthermore, niche applications such as nuclear fusion gained attention for the highest selectivity ensured by self-supported dense metallic membranes, especially those consisting of Pd-based alloys. In this framework, the ENEA Frascati laboratories have decades of experience in the manufacturing, integration, and operation of Pd-Ag permeators. Most of the experimental investigations were performed on single-tube membranes, proving their performance under relevant operational conditions. Nowadays, once the applicability of this technology has been demonstrated, the scalability of the single-tube experience over medium- and large-scale units must be verified. To do this, ENEA Frascati laboratories have designed and constructed a multi-tube permeator, namely the Medium-Scaled Membrane Reactor (MeSMeR), focused on scalability assessment. In this work, the results obtained with the MeSMeR facility have been compared with previous experimental campaigns conducted on single-tube units, and the scalability of the permeation results has been proven. Moreover, post-test simulations have been performed based on single-tube finite element modeling, proving the scalability of the numerical outcomes and the possibility of using this tool for scale-up design procedures. Full article

The treatment of tritiated nuclear wastewater is facing greater challenges with the continuous expansion of the nuclear industry. The key to solving the issue of detritium in low-abundance tritium water lies in developing highly efficient and cost-effective hydrogen isotope separation technology. Graphene oxide (GO) membrane separation method exhibits greater potential compared to other existing energy-intensive technologies for the challenging task of hydrogen isotope separation in nuclear wastewater. In recent years, researchers have explored few strategies to enhance the performance of graphene oxide (GO) membranes in hydrogen isotope water treatment, recognizing the current limitations in separation efficiency. In this study, the GO/g-C₃N₄ composite membrane has been successfully employed for the first time in the separation of hydrogen isotopes in water. A series of GO membranes were prepared and their performances were tested by a self-made experimental device. As a result, the separation performance of the GO membrane was enhanced by the modification with graphitic carbon nitride (g-C₃N₄). The permeation rate of the GO/g-C₃N₄ membrane was higher than that of the GO membrane, while maintaining a high separation factor. Our study also demonstrated that this phenomenon can be attributed to the changes in membrane structure at the microscopic scale. The H/D separation factor and the permeate flux of the composite membrane containing g-C₃N₄ of 6.7% by mass were 1.10 and 7.2 × 10⁻⁵ g·min⁻¹·cm⁻² are both higher than that of the GO membrane under the same experimental conditions, which is promising for the isotope treatment. Full article

The latest progress in the design of the water-cooled lithium–lead (WCLL) tritium extraction and removal (TER) system for the European DEMO tokamak reactor is presented. The implementation and optimization of the conceptual design of the TER system are performed in order to manage the tritium concentration in the LiPb and ancillary systems, to control the LiPb chemistry, to remove accumulated corrosion and activated products (in particular, the helium generated in the BB), to store the LiPb, to empty the BB segments, to shield the equipment due to LiPb activation, and to accommodate possible overpressure of the LiPb. The LiPb volumes in the inboard (IB) and outboard (OB) modules of the BB are separately managed due to the different pressure drops and required mass flow rates in the different plasma operational phases. Therefore, the tritium extraction is managed by 6 LiPb loops: 4 loops for the OB segments and 2 loops for the IB segments. Each one is a closed loop with forced circulation of the liquid metal through the TER and the other ancillary systems. The design presents the new CAD drawings and the integration of the TEU into the tokamak building, designed on the basis of an experimental characterization carried out for the permeator against vacuum (PAV) and gas–liquid contactor (GLC) technologies, the two most promising technologies for tritium extraction from liquid metal. Full article

The interest of the fusion community in Pd–Ag membranes has grown in the last decades due to the high value of hydrogen permeability and the possibility of continuous operation, making it a promising technology when a gaseous stream of hydrogen isotopes must be recovered and separated from other impurities. This is the case of the Tritium Conditioning System (TCS) of the European fusion power plant demonstrator, called DEMO. This paper presents an experimental and numerical activity aimed at (i) assessing the Pd–Ag permeator performance under TCS-relevant conditions, (ii) validating a numerical tool for scale-up purposes, and (iii) carrying out a preliminary design of a TCS based on Pd–Ag membranes. Experiments were performed by feeding the membrane with a He–H₂ gas mixture in a specific feed flow rate ranging from 85.4 to 427.2 mol h⁻¹ m⁻². A satisfactory agreement between experiments and simulations was obtained over a wide range of compositions, showing a root mean squared relative error of 2.3%. The experiments also recognized the Pd–Ag permeator as a promising technology for the DEMO TCS under the identified conditions. The scale-up procedure ended with a preliminary sizing of the system, relying on multi-tube permeators with an overall number ranging between 150 and 80 membranes in lengths of 500 and 1000 mm each. Full article

Presented are the results of ^99mTc and ¹⁰¹Tc production via neutron irradiation of natural isotopic molybdenum (Mo) with epithermal/resonance neutrons. Neutrons were produced using a deuterium-deuterium (D-D) neutron generator with an output of 2 × 10¹⁰ n/s. The separation of Tc from an irradiated source of bulk, low-specific activity (LSA) Mo on activated carbon (AC) was demonstrated. The yields of ^99mTc and ¹⁰¹Tc, together with their potential use in medical single-photon emission computed tomography (SPECT) procedures, have been evaluated from the perspective of commercial production, with a patient dose consisting of 740 MBq (20 mCi) of ^99mTc. The number of neutron generators to meet the annual 40,000,000 world-wide procedures is estimated for each imaging modality: ^99mTc versus ¹⁰¹Tc, D-D versus deuterium-tritium (D-T) neutron generator system outputs, and whether or not natural molybdenum or enriched targets are used for production. The financial implications for neutron generator production of these isotopes is also presented. The use of ¹⁰¹Tc as a diagnostic, therapeutic, and/or theranostic isotope for use in medical applications is proposed and compared to known commercial nuclear diagnostic and therapeutic isotopes. Full article

Like many biological compounds, proteins are found primarily in their homochiral form. However, homochirality is not guaranteed throughout life. Determining their chiral proteinogenic sequence is a complex analytical challenge. This is because certain d-amino acids contained in proteins play a role in human health and disease. This is the case, for example, with d-Asp in elastin, β-amyloid and α-crystallin which, respectively, have an action on arteriosclerosis, Alzheimer’s disease and cataracts. Sequence-dependent and sequence-independent are the two strategies for detecting the presence and position of d-amino acids in proteins. These methods rely on enzymatic digestion by a site-specific enzyme and acid hydrolysis in a deuterium or tritium environment to limit the natural racemization of amino acids. In this review, chromatographic and electrophoretic techniques, such as LC, SFC, GC and CE, will be recently developed (2018–2020) for the enantioseparation of amino acids and peptides. For future work, the discovery and development of new chiral stationary phases and derivatization reagents could increase the resolution of chiral separations. Full article

For the stable and self-sufficient functioning of the DEMO fusion reactor, one of the most important parameters that must be demonstrated is the Tritium Breeding Ratio (TBR). The reliable assessment of the TBR with safety margins is a matter of fusion reactor viability. The uncertainty of the TBR in the neutronic simulations includes many different aspects such as the uncertainty due to the simplification of the geometry models used, the uncertainty of the reactor layout and the uncertainty introduced due to neutronic calculations. The last one can be reduced by applying high fidelity Monte Carlo simulations for TBR estimations. Nevertheless, these calculations have inherent statistical errors controlled by the number of neutron histories, straightforward for a quantity such as that of TBR underlying errors due to nuclear data uncertainties. In fact, every evaluated nuclear data file involved in the MCNP calculations can be replaced with the set of the random data files representing the particular deviation of the nuclear model parameters, each of them being correct and valid for applications. To account for the uncertainty of the nuclear model parameters introduced in the evaluated data file, a total Monte Carlo (TMC) method can be used to analyze the uncertainty of TBR owing to the nuclear data used for calculations. To this end, two 3D fully heterogeneous geometry models of the helium cooled pebble bed (HCPB) and water cooled lithium lead (WCLL) European DEMOs were utilized for the calculations of the TBR. The TMC calculations were performed, making use of the TENDL-2017 nuclear data library random files with high enough statistics providing a well-resolved Gaussian distribution of the TBR value. The assessment was done for the estimation of the TBR uncertainty due to the nuclear data for entire material compositions and for separate materials: structural, breeder and neutron multipliers. The overall TBR uncertainty for the nuclear data was estimated to be 3~4% for the HCPB and WCLL DEMOs, respectively. Full article

This paper reviews the membrane processes for the nuclear fusion fuel cycle—namely, the treatment of the plasma exhaust gases and the extraction of tritium from the breeding blankets. With respect to the traditional processes, the application of membrane reactors to the fusion fuel cycle reduces the tritium inventory and processing time, thus increasing the safety and availability of the system. As an example, self-supported Pd-alloy membrane tubes have been studied for the separation of hydrogen and its isotopes from both gas- and liquid-tritiated streams through water-gas shift and isotopic swamping reactions. Furthermore, this paper describes an innovative membrane system (Membrane Gas–Liquid Contactor) for the extraction of hydrogen isotopes from liquid LiPb blankets. Porous membranes are exposed to the liquid metal that penetrates the pores without passing through them, then realizing a gas–liquid interface through which the mass transfer of hydrogen isotopes takes place. Compared to the conventional hydrogen isotope extraction processes from LiPb that use the “permeator against vacuum” concept, the proposed process significantly reduces mass-transfer resistance by improving the efficiency of the tritium recovery system. Full article

Dense self-supported Pd-alloy membranes are used to selectively separate hydrogen and hydrogen isotopes. In particular, deuterium (D) and tritium (T) are currently identified as the main elements for the sustainability of the nuclear fusion reaction aimed at carbon free power generation. In the fusion nuclear reactors, a breeding blanket produces the tritium that is extracted and purified before being sent to the plasma chamber in order to sustain the fusion reaction. In this work, the application of Pd-alloy membranes has been tested for recovering tritium from a solid breeding blanket through a helium purge stream. Several simulations have been performed in order to optimize the design of a Pd-Ag multi-tube module in terms of geometry, operating parameters, and membrane module configuration (series vs. parallel). The results demonstrate that a pre-concentration stage before the Pd-membrane unit is mandatory because of the very low tritium concentration in the He which leaves the breeding blanket of the fusion reactor. The most suitable operating conditions could be reached by: (i) increasing the hydrogen partial pressure in the lumen side and (ii) decreasing the shell pressure. The preliminary design of a membrane unit has been carried out for the case of the DEMO fusion reactor: the optimized membrane module consists of an array of 182 Pd-Ag tubes of 500 mm length, 10 mm diameter, and 0.100 mm wall thickness (total active area of 2.85 m²). Full article