Reprocessing Of Fuels Research Articles

The Chalkboard: Stretching Cyclic Voltammetry to its Potential Limit

Electrochemistry's reach is extensive, from fundamentals, material and commodity synthesis and reactor engineering to energy conversion devices. For most areas of interest, experimental conditions are well controlled. However, in some cases, experimental conditions are significantly more complex because applications are located in the field and exposed to the elements. Sensors and innovations designed to abate corrosion are in contact with the environment. These situations add complexity to design activities and data interpretation. Atmospheric air contains hundreds of species (mostly at trace levels), whereas oceans are effective solvents for ions, inorganic (geological features contact) and organic (biological activity) species. A growing interest in the circular economy will also result in similar and uncontrolled experimental conditions for envisaged electrochemical processes. Nuclear reprocessing of fuel rods to recover uranium involves separating noble, transuranic, lanthanide, and other elements. Battery electrodes recycling by hydrometallurgical processes will lead to multi-component solutions. Polymer recycling in bulk using electrochemical processes is desirable to minimize costs, which will create an assemblage of organic components (diverse polymers, leftover synthesis reactants, dyes, stabilizers, corrosion inhibitors). For process control of electrochemical recycling, effective methods are needed to assess the composition of multi-component solutions.

The Chalkboard: Stretching Cyclic Voltammetry to its Potential Limit

Electrochemistry’s reach is extensive, from fundamentals, material and commodity synthesis and reactor engineering to energy conversion devices. For most areas of interest, experimental conditions are well controlled. However, in some cases, experimental conditions are significantly more complex because applications are located in the field and exposed to the elements. Sensors and innovations designed to abate corrosion are in contact with the environment. These situations add complexity to design activities and data interpretation. Atmospheric air contains hundreds of species (mostly at trace levels), whereas oceans are effective solvents for ions, inorganic (geological features contact) and organic (biological activity) species. A growing interest in the circular economy will also result in similar and uncontrolled experimental conditions for envisaged electrochemical processes. Nuclear reprocessing of fuel rods to recover uranium involves separating noble, transuranic, lanthanide, and other elements. Battery electrodes recycling by hydrometallurgical processes will lead to multi-component solutions. Polymer recycling in bulk using electrochemical processes is desirable to minimize costs, which will create an assemblage of organic components (diverse polymers, leftover synthesis reactants, dyes, stabilizers, corrosion inhibitors). For process control of electrochemical recycling, effective methods are needed to assess the composition of multi-component solutions.

Development of computer code ADWITA and data library for the solution of transmutation chain equations and application to the analysis of nuclear fuel cycles

A fuel cycle analysis code, ADWITA (Activation, Decay, Waste Incineration and Transmutation Analysis) has been developed which can generate isotopic inventory, calculate radioactivity, and decay heat based on irradiation history. The code can be used for fuel cycle design and transmutation studies. The system of linearized transmutation chain equations is solved by application of Bateman series solution method and the matrix exponential method in which the matrix exponential is approximated by the exponential power series. The required data libraries for Pressurised Heavy Water Reactor (PHWR) and Advanced Heavy Water Reactor (AHWR) have also been developed. The code and the data have been validated using the analytical decay of 237Np, the PHWR fuel cycle benchmark data available in IAEA TECDOC and experimental measurement of isotopic composition and decay heat in AECL report. The code has been applied for the analysis of fuel cycle of AHWR which has been designed for the utilisation of thorium as nuclear fuel on large scale and envisages fuel reprocessing. The application of code ADWITA and its library to AHWR has helped to understand the unique features of AHWR fuel cycle. It has been found that in AHWR the activity and decay heat due to actinides and fission product (FP) in both (Th, 233U) MOX and (Th, Pu) MOX fuel is similar at the discharge but the activity and decay heat due to actinide takes over that due to FP in case of (Th, Pu) MOX earlier. The accumulation of uranium isotopes at the end of burnup in individual (Th, 233U) and (Th, Pu) MOX fuel pin has been estimated which is an important input to the reprocessing of these fuels. The rise in activity after reprocessing due to presence of 232U has also been estimated.

A simple spectrofluorimetric method to estimate uranium: application to nuclear reprocessing solution

A simple spectrofluorimetric method to estimate uranium: application to nuclear reprocessing solution

129I/127I and Δ14C records in a modern coral from Rowley Shoals off northwestern Australia reflect the 20th-century human nuclear activities and ocean/atmosphere circulations

Radionuclides produced by 20th-century human nuclear activities from 1945 (e.g., atmospheric nuclear explosions and nuclear-fuel reprocessing) made significant impacts on earth's surface environments. Long-lived shallow-water corals living in tropical/subtropical seas incorporate the anthropogenically-produced radionuclides, including 129I and 14C, into their skeletons, and provide time series records of the impacts of nuclear activities. Here, we present 129I/127I and Δ14C time series records of an annually-banded modern coral skeleton from Rowley Shoals, off the northwestern coast of Australia, in the far eastern Indian Ocean. The 129I/127I and Δ14C records, covering the period 1930s−1990s, exhibit distinct increases caused by the nuclear activities, and their increasing profiles are clearly different from each other. The first distinct 129I/127I increase occurs from 1955 to 1959, followed by a decrease in 1960–1963. The increase is probably due to US atmospheric nuclear explosions in Bikini and Eniwetok Atolls in 1954, 1956 and 1958. The 129I produced in those nuclear tests would be transported by the North Equatorial Current, a portion of which passes through the Indonesian Throughflow and then reaches Rowley Shoals. This initial increase from 1955 is, however, absent in the Δ14C record, which shows a distinct increase from 1959 and its peak around the mid-1970s, followed by a gradual decrease. This absence and the 4-year-delayed Δ14C increase are likely due to dilution of explosion-produced 14C with natural carbon (by seawater mixing and air-sea gas exchange) being much more intense than that of explosion-produced 129I with natural iodine (by the same processes), suggesting that the 129I/127I ratio is a more conservative anthropogenic tracer in surface ocean waters, as compared to Δ14C. The second 129I/127I increase is contemporaneous with a rapid Δ14C increase during 1964–1967, followed by a rapid 129I/127I decrease in 1968–1969; the increases can be ascribed to very large atmospheric nuclear explosions conducted in the former Soviet Union in 1961–1962. The third 129I/127I increase appears between 1969/1970 and 1992, which can be attributed to airborne 129I released from nuclear-fuel reprocessing facilities in Europe, the former Soviet Union and the US. The coral 129I/127I and Δ14C time series records, combined with previous studies, enhance our understanding of the behavior of anthropogenic 129I and 14C in the global ocean and atmosphere.

Spent Nuclear Fuel Reprocessing and Nuclear Materials Recycling in Two-Component Nuclear Energy

Spent Nuclear Fuel Reprocessing and Nuclear Materials Recycling in Two-Component Nuclear Energy

Nd(III) hypersensitive peak as an optical absorption probe for determining nitric acid in aqueous solution: An application to aqueous raffinate solutions in nuclear reprocessing

A new method using Nd(III) absorption peak as a probe is described for the measurement of nitric acid concentration in aqueous solution. The hypersensitive peak of Nd(III) at 575.1 nm shows a substantial enhancement in the absorbance in comparison to other absorption peaks with increasing nitric acid concentration. The integrated area and absorbance of this hypersensitive peak show a linear dependency over a large dynamic range of 0.5–15.5 M of nitric acid. A methodology for the correction of spectral interference to the probing absorption peak of Nd(III) is also reported. The method is applied for the measurement of nitric acid in synthetic high level liquid waste solution and shown to be comparable to that obtained by titrimetric method. The present method can be easily adopted for the measurement of nitric acid concentration in aqueous raffinate solutions of nuclear reprocessing streams.

Prospects of functionalized porous materials towards extraction of spent nuclear fuel radionuclides

Radioactive waste levels have continued to increase due to the growth and development of nuclear energy, industrial or medical radioactive use. In this regard, conventional radioactive waste generated by nuclear power plants cannot be ignored. Effective management of nuclear radioactive waste plays a vital role in alleviating negative impacts on the society and environment. Despite the progress that has been made concerning separation and recycling of spent nuclear fuel through the PUREX process, several gaps in knowledge still exist especially towards the development of a robust separation system based on solid-phase extraction using porous materials. Solid phase extraction is being viewed as one of the most convenient and effective approaches in the removal of cations in radionuclide solutions. This is due to its ability to increase the selectivity and sensitivity of the method as it permits discriminatory binding of analyte to a solid support where the analyte can be collected and thereafter eluted using small quantity of a different solvent. The review covers the current methods used in aqueous nuclear reprocessing, highlights their deficiencies and introduces the potential of applying solid-phase extraction in management of nuclear waste. This study gives the prospects of functionalized porous sorbent materials as solid support in solid-phase extraction of spent nuclear fuel elements.

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Application of Aluminium Nitrate for Extraction Reprocessing of Spent Nuclear Fuel

Experimental Modeling of Technetium(VII) Recovery from Raffinates after Extractive Reprocessing of Spent Nuclear Fuel

Experimental justification has been provided for the possibility of extractive isolation of technetium from raffinate of technical process of affinage of uranium–plutonium product resulting from hydrometallurgical reprocessing of spent nuclear fuel (SNF). Technetium (VII) from aqueous solution of Zr(IV) (10 g/L) and 4 mol/L HNO3 with small content of ZrOb+ and $${\text{Z}}{{{\text{r}}}_{{\text{2}}}}{\text{O}}_{{\text{3}}}^{{b + }}$$ oxocations is recovered by 80% in 30% TBP in hydrocarbon diluent over seven stages of counter-current continuous cascade at ratio O : A = 1 : 1. Variation in phase flow ratio for cascade part leads to virtually complete Tc(VII) extraction in organic phase over ten stages. Technetium(VII) is selectively back extracted from loaded organic phase with water virtually completely over eleven stages of counter-current continuous cascade. Technological extraction scheme for technetium isolation from SNF solutions has been proposed.

Demonstration of Aqueous Reprocessing of U-Zr and U-Pu-Zr Metallic Alloy Fuels Using an Ejector Mixer-settler with Tri-n-Butyl Phosphate (TBP) as the Extractant

ABSTRACT U-Zr and U-Pu-Zr metallic alloy fuels have been employed in fast reactors in the early days and these fuels have been considered for use in future fast reactors in India. It is proposed to develop an aqueous method based on the PUREX process for the reprocessing of spent metallic fuels. In the present study, an ejector mixer-settler was employed to demonstrate the feasibility of aqueous reprocessing of U-Zr and U-Pu-Zr metallic alloy fuels with tri-n-Butyl Phosphate (TBP) as the extractant. Continuous counter-current mixer settler runs were separately carried out with U-Zr and U-Pu-Zr feed solutions using 1.1 M TBP/n-DD solvent. Quantitative extraction of U and U + Pu was achieved within five stages leaving zirconium in the raffinate in the case of U-Zr and U-Pu-Zr feed solutions, respectively. Quantitative back extraction of uranium from loaded organic phase was achieved using 0.01 M HNO3 in the case of U-Zr mixer-settler runs. Dual strip solutions such as 2 M and 0.01 M HNO3 were used for the quantitative stripping of U and Pu from the loaded organic phase in the case of U-Pu-Zr mixer-settler runs. The proposed flow sheets demonstrate that the aqueous reprocessing of metallic fuels using a PUREX-based method is feasible.

Synroc technology: Perspectives and current status (Review)

AbstractDr Eric (Lou) Vance spent 32 years at the Australian Nuclear Science and Technology Organisation (ANSTO), where he was dedicated to the development of Synroc technology, a waste treatment solution for intractable nuclear wastes. The original form of Synroc, a multiphase ceramic wasteform based on stable and leach resistant titanate minerals, was invented by Australian scientists in the late 1970s. This formulation was directed toward the immobilization of PUREX wastes from the reprocessing of nuclear fuels. Synroc at ANSTO under the scientific leadership of Dr Vance since evolved beyond these original titanate ceramics into a waste treatment technology platform. This platform can be applied to produce glass, glass‐ceramic and ceramic wasteforms and offers distinct advantages in terms of waste loading and suppressing volatile losses. The platform therefore provides an opportunity to treat those waste streams that are problematic for glass matrices alone or existing vitrification process technologies. Such wastes include, for example, actinide‐bearing wastes, those that contain large proportions of refractory elements, those with significant fission product or corrosive volatile emissions and those wastes resulting from radiopharmaceutical production. The implementation of the latter will see the industrialization of Synroc technology via a first‐of‐a‐kind Synroc Waste Treatment Facility that is currently under construction at ANSTO. This paper will review Synroc technology, particularly noting the substantial and essential contributions from the late Dr Vance. The review will also provide some perspective on the development of the technology for nuclear waste immobilization and describe the significant recent advancements at ANSTO.

Simulation of Electrolyzer Processes for the Reprocessing of Spent Nuclear Fuel

The ANSYS software package is used to simulate a functioning electrolyzer for the reprocessing of spent nuclear fuel via electrolysis in LiCl molten salt with the addition of Li2O. The seven considered types of electrolyzer design differ in their mutual arrangement of electrodes and anode thickness. The temperature field, as well as distributions of the current density and electric potential, are calculated. The optimal electrolyzer design is determined in order to achieve the best temperature and electrical characteristics.

Uranium recovery via melt separation of rare earth elements and chemical oxidation

Pyroprocessing of spent fuels, especially the recovery of U, improves the overall efficiency of nuclear recycling. The recovery of U/Transuranic elements (TRU) during the removal of rare earth elements (REEs) via melting was evaluated. Uranium was used as a surrogate material instead of TRU, and the distributions of U, Nd, and Ce were evaluated after melting. Uranium and REEs were clearly separated after melting and the U component contained < 2 wt% of REEs as impurities. Melting recovery of U was applicable after the electrorefining process to obtain a high enough concentration of REEs above their solubility in U. The REE component contained ~3 wt% U, and U was also recovered through the chemical oxidation of REEs with UCl3. The U remaining in the product following chemical oxidation exceeded 87%. Based on these results, a new method of recovering U/Pu from U/Pu/REE-alloys following the electrorefining process is proposed. This approach can be used to improve the recycling of spent nuclear fuels, thereby enhancing the efficiency and sustainability of nuclear systems.

Spectrophotometric Analysis of Ternary Uranyl Systems to Replace Tri-N-butyl Phosphate (TBP) in Used Fuel Reprocessing

In this report, the interaction of monoamide/diamide and monoamide/diglycolamide mixtures with {text{UO}}_{2}^{2 + } are investigated in pH = 1 methanolic nitric acid media. These monoamides include N,N-dimethylacetamide (DMAA), N,N-diethylacetamide (DEAA), N,N-dibutylacetamide (DBAA) and N,N-dibutylbutanamide (DBBA). N,N,N′N′-tetraethylmalonamide (TEMA) and N,N,N′,N′-tetraethyldiglycolamide (TEDGA), which were chosen as model diamides and diglycolamides, respectively. Complex stability constants for each ligand were modelled using the Stability Quotients Using Absorbance Data program using UV–visible data. Complex stoichiometry of ligand mixtures was determined using Job plots and UV–Vis spectrometry. Monoamides were confirmed to produce only disolvate complexes with {text{UO}}_{2}^{2 + } in solution. The log10(K) values for monoamides were found to be independent of amine-side chain length, but were slightly dependent on the carbonyl-side chain length. TEDGA was found to produce multiple uranyl complexes in solution. Job plot data indicated that the uranyl cation strongly prefers to bond either only with the monoamide or diamide in ternary monoamide–diamide–UO2 systems. Monoamide–diglycolamide–UO2 systems were more complicated, with Job plot data indicating the potential for multiple ternary species being present is dependent on the monoamide structure.

Nucleophagy mediators and mechanisms.

Nuclear recycling is essential for cell and organismal homeostasis. Nuclear architecture perturbations, such as nuclear loss or nuclear enlargement, have been observed in several pathological conditions. Apart from proteasomal components which reside in the nucleus, specific autophagic proteins also shuttle between the nucleus and the cytoplasm. Until recently, only the microautophagic degradation of nuclear components had been described. Recent studies, dissecting nuclear material recycling in organisms ranging from yeast to mammals, provide insight relevant to other forms of nucleophagy and the mediators involved. Nucleophagy has also been implicated in pathology. Lamins are degraded in cancer through direct interaction with LC3 in the nucleus. Similarly, in neurodegeneration, Golgi-associated nucleophagy is exacerbated. The physiological role of nucleophagy and its contribution to other pathologies remain to be elucidated. Here we discus recent findings that shed light into the molecular mechanisms and pathways that mediate the autophagic recycling of nuclear material.

Radwaste Volume from VVER-1000 Spent Fuel Reprocessing and Fractionation Variants

The possibilities of minimizing the volume of waste are examined for the reprocessing of spent nuclear fuel from VVER-1000 and the matrix material for high-level wastes. The analysis takes into account the designbasis isotopic composition of VVER-1000 spent fuel and the current normative base in the sphere of waste management. It is shown that deep underground disposal of radioactive wastes cannot be avoided at the present stage of development of the reprocessing technologies for the spent fuel from thermal reactors and with the existing legislation associated with the conditions of use of matrices for solidification of wastes and the conditions of waste disposal. Variant methods for decreasing the volume of underground disposal of radwaste are proposed.

Mathematical Model of Cadmium Distillation at the Stage of Pyrochemical Reprocessing of Spent Mixed Uranium-Plutonium Nitride Nuclear Fuel

A mathematical model of the technological process of cadmium distillation at the stage of pyrochemical reprocessing of spent mixed uranium-plutonium nitride fuel for the purpose of studying the serviceability of and developing monitoring and process control systems is presented. The basic thermodynamic and kinetic characteristics and the principles of mathematical modeling of vacuum distillation of liquid mixtures, specifically, cadmium, are described. The requirements of the model and the assumptions are formulated. The information structure and mathematical description of the model were developed. Computational experiments were performed on the model and showed its serviceability and possibility of use for developing monitoring and process control systems.

Aqueous solubility of tri-iso-amyl phosphate (TiAP) - Effect of acidity, temperature and metal ions

In recent times, tri-iso-amyl phosphate (TiAP) has been emerged as an alternate solvent to the existing tri-butyl phosphate (TBP) mainly for the aqueous reprocessing of high Plutonium content fast reactor fuels adopting the conventional PUREX process. Little solubility of the solvent in aqueous phase during extraction can lead to complex issues in further operations especially during the evaporation of raffinate. In this context, the aqueous solubility of tri-iso-amyl phosphate (TiAP) in a wide range of nitric acid concentrations from 0 to 14 M has been determined at three different experimental temperatures 298, 313 and 328 K. Various process parameters such as aqueous phase acidity and presence of heavy metal ions which affect the solubility of solvent are also discussed.