Uranium Fission Products Research Articles

Research Progress on Controllable Absorption Properties of Rare Earth Element Doped Electromagnetic Wave Absorbing Materials†

Comprehensive SummaryThe utilization of rare earth (RE) elements is pivotal in wave absorption. In particular cases, RE group elements manifesting unique 4f electron layer structures are doped as impurities into certain materials, which is a practical and reliable method for regulating these materials' magnetic and electrical properties. Moreover, ferrites and metal‐organic frameworks (MOFs), standing out among conventional and emerging wave‐absorbing materials, can achieve significantly enhanced wave absorption through RE doping or substitution. Numerous scholars have dedicated massive research over a substantially long time to explore the utilization of RE elements to reinforce the absorption of these two materials. Therefore, consolidating and summarizing such efforts are crucial and necessary. This review aims to clarify the underlined mechanism of electromagnetic wave (EMW) absorption and elaborate on the impact of RE doping by providing a comprehensive overview of recent progress in ferrites and MOFs dopped with RE elements. Finally, the limitations associated with RE doping in such materials are delineated, and the upcoming prospects for its application are highlighted. Key ScientistsIn 1947, J. A. Marinky et al. obtained promethium from uranium fission products. From the isolates of yttrium soil in 1794 to the discovery of promethium in 1947, it took more than 150 years. In 1975, Guangxian Xu found that the rare earth solvent extraction system has the basic "constant mixed extraction ratio" rule. In 1984, the theory of "one‐cavity multi‐mode" was proposed by Weigan Lin to improve the electromagnetic wave theory further and develop the theory of absorbing materials. In 1995, Omar M. Yaghi synthesized the first MOFs in history, and since then, they have been widely used in gas adsorption and separation and other fields because of their large specific surface area and adjustable pore structure. In recent years, various functional materials based on MOFS‐derived carbon have been studied endlessly. Masato Sagawa, known for inventing NdFeB magnets, won the "28th Japan Prize" in 2012. In recent years, Renchao Che has promoted the development of interface theory and magnetic theory in the field of wave absorption. In addition, Jiurong Liu is committed to developing various wave‐absorbing materials and the respective theoretical research and continues to promote the development of wave‐absorbing materials to "thin, light, wide and strong".

Control and surveillance of redox potential for 233Pa dissolution in 2LiF-BeF2 molten salt.

233Pa, the precursor nuclide of 233U in the thorium-uranium conversion is prone to reductive deposition in 2LiF-BeF2 (66 : 34 mol%, FLiBe) molten salt. We explored the adjustment and control of the redox potential of the FLiBe melt to avoid the 233Pa reduction deposition. The experimental data indicated that the deposited 233Pa can be completely dissolved and reentered into the molten salt with the addition of oxidant NiF2, and the distribution and behaviour of uranium, thorium, neptunium, and most fission products did not have any significant change in the NiF2-oxidised FLiBe molten salt, showing the feasibility of this manner to make 233Pa exist stably in the melt. The effects of NiF2-addition on the behaviour of the fission products 95Nb and 131I in the FLiBe molten salt were also investigated. It was found that 131I could be used as a redox indicator to monitor the redox potential of the oxidation-enhanced FLiBe molten salt. All the information drawn from this study could provide significant support for the control and surveillance of the redox potential of the FLiBe molten salt in the upcoming thorium molten salt reactor (TMSR).

Radiochemical and chemical characterization of fuel, salt, and deposit from the electrorefining of irradiated U-6 wt% Zr in hot cells

Abstract Metal fuels are considered as the promising candidates for future fast breeder reactors. Pyro-chemical reprocessing is the ideal method for reprocessing spent metallic fuels due to the inherent process advantages. Electrorefining run was demonstrated in a hot cell facility with irradiated U-6 wt% Zr alloy at 500 °C using LiCl–KCl eutectic melt. In order to understand the behavior of the actinides and various fission products during high-temperature electrolysis, various process streams, viz., irradiated metal alloy fuel, the eutectic salt, and the cathode deposit were analyzed for the uranium, plutonium, and other fission product contents. Various methods employed for characterizing the process streams and the behaviors of some of the fission products during the electrolysis process are highlighted. The major gamma emitting radionuclides present in the irradiated fuel were 106Ru, 125Sb, 134Cs, 137Cs, 144Ce, and 154Eu. During electrorefining, cesium, cerium and europium were oxidized and dissolved in the molten media, whereas ruthenium and antimony remained in the anode basket. A minor contamination of zirconium was found in the cathode deposit.

Reaction of Oxygen with Uranium (IV) Chloride in Fused Alkali Chlorides

Reaction of oxygen with solutions of uranium(IV) chloride in fused LiCl and three alkali chloride eutectic mixtures (LiCl–KCl, NaCl–KCl–CsCl, NaCl–CsCl) was investigated at 550–750 °C. Bubbling oxygen or oxygen-containing gas mixtures (O2–H2O, O2–Ar, O2–H2O–Ar) through LiCl–UCl4 melts resulted in significant precipitation of uranium (up to 87%) in the form of oxides and alkali uranates. Increasing mean radius of the solvent melt cations decreased the degree of uranium precipitation and uranyl chloride (soluble in the melt) became the main product of the reaction. High temperature spectroscopy measurements were employed to determine the kinetic parameters of the reaction in LiCl–KCl, NaCl–KCl–CsCl and NaCl–CsCl melts. Reaction rates, order, rate constants and activation energy values were estimated. Increasing temperature led to increased reaction rates but the effect of uranium chloride concentration depended on the cationic melt composition. Oxygen reacts with uranium(IV) containing melts much faster than with the melts containing rare earth chlorides and oxygen sparging can be implemented for separating uranium and rare earth fission products in pyrochemical reprocessing of spent nuclear fuels.

Formation, Diffusion, and Growth of Gas Bubbles in γ-Uranium with the Excess of Interstitial Atoms: Relation between Molecular Dynamics and Kinetics

The formation of gas nanobubbles through the merging of individual fission products of uranium is an important process for the evolution of nuclear fuels. The theoretical description of this process is very difficult because both the dynamics of individual atoms in the lattice and the kinetics of evolution of an ensemble of bubbles should be taken into account within a unified model. Such a model is constructed in this work on the basis of molecular dynamics simulations for xenon bubbles in bcc uranium in the case of the excess of interstitial atoms in the crystal matrix. The analysis is based on the molecular dynamics simulation of the nonequilibrium process of formation of xenon nanobubbles from individual Xe atoms dissolved in the crystal matrix. A relation between the size of bubbles and the number of gas atoms in them, as well as the dependence of the diffusion coefficient of bubbles on their radius and the number of interstitial atoms in the γ-U matrix, has been analyzed. A kinetic model of evolution of the ensemble of bubbles has been proposed to describe the molecular dynamics results and to extrapolate them to long times.

Model for the Conversion of Beta Spectra from Fission Products of Uranium and Plutonium Isotopes into Antineutrino Spectra

Model for the Conversion of Beta Spectra from Fission Products of Uranium and Plutonium Isotopes into Antineutrino Spectra

МЕДИКО-САНИТАРНЫЕ ПОСЛЕДСТВИЯ РАДИАЦИОННЫХ АВАРИЙ ПРИ ОБРАЩЕНИИ С ОТРАБОТАННЫМ ЯДЕРНЫМ ТОПЛИВОМ

Purpose: To consider the features of potential health effects for personnel and for the population in the event of radiation accidents at facilities for post-reactor spent nuclear fuel management (including spent nuclear fuel pools and general plant storage facilities, transportation of nuclear materials and their subsequent processing at radiochemical plants). Results: Based on the analysis of publications, the potential radiation consequences for personnel and for the population in the event of various types of radiation accidents at the stages of spent nuclear fuel management were systematized. Conclusion: Advanced technologies underlying the closed nuclear fuel cycle provide a high level of radiation safety for personnel and for the population. At the same time, a large amount of accumulated radioactive and nuclear materials makes it necessary to maintain the health care system in preparedness at all stages of spent nuclear fuel management. Health effects for personnel may be associated with exposure because of criticality accident and internal intake of fission products of uranium and actinides (by inhalation and through wound surfaces). In case of accidents at radiochemical production, combined radiation-thermal and radiation-chemical injuries are also possible. The main potential hazard for the population in the event of radiation accident at the spent nuclear fuel storage facility is the contamination of the environment and exposure to long-lived uranium fission products and actinides. This requires clarification and development of appropriate criteria and derived intervention levels for making decisions on protective measures. Hypothetically, the health effects in case of radiation accident when managing “fresh” spent nuclear fuel can be comparable to a large-scale reactor accident and require urgent protective measures, including evacuation and iodine prophylaxis. An important factor that should also be taken into account when planning health care measures is the need for decontamination of victims.

Introducing Zirconium Organic Gels for Efficient Radioiodine Gas Removal.

Iodine radioisotope, as one of the most important fission products of uranium, may cause severe damage to human health when it is accidentally discharged into the environment. Hence, efficient removal of radioactive iodine is one of the most critical issues for both used nuclear fuel (UNF) reprocessing and environmental remediation. In this work, three metal-organic gels (MOGs) were introduced for iodine removal. The presented zirconium-based MOGs, namely, CWNU, CWNU-NH2, and CWNU-2NH2, were prepared via moderate solvothermal reactions. These MOGs all exhibit excellent chemical stability and reusability, marked iodine sorption capability, and favorable machinability, which can even rival commercial ones. The sorption capacities are determined to be 3.36, 4.10, and 4.20 g/g, respectively. The increased amount of amino group is considered to be responsible for the elevated iodine sorption capacity and kinetics, as confirmed by combined sorption studies and XPS analysis. The presented work sheds light on the utilization of MOGs for radioiodine capture.

Sorption of Strontium to Uraninite and Uranium(IV)-Silicate Nanoparticles.

Spent nuclear fuel contains both uranium (U) and high yield fission products, including strontium-90 (90Sr), a key radioactive contaminant at nuclear facilities. Both U and 90Sr will be present where spent nuclear fuel has been processed, including in storage ponds and tanks. However, the interactions between Sr and U phases under ambient conditions are not well understood. Over a pH range of 4–14, we investigate Sr sorption behavior in contact with two nuclear fuel cycle relevant U(IV) phases: nano-uraninite (UO2) and U(IV)–silicate nanoparticles. Nano-UO2 is a product of the anaerobic corrosion of metallic uranium fuel, and UO2 is also the predominant form of U in ceramic fuels. U(IV)–silicates form stable colloids under the neutral to alkaline pH conditions highly relevant to nuclear fuel storage ponds and geodisposal scenarios. In sorption experiments, Sr had the highest affinity for UO2, although significant Sr sorption also occurred to U(IV)–silicate phases at pH ≥ 6. Extended X-ray absorption fine structure (EXAFS) spectroscopy, transmission electron microscopy, and desorption data for the UO2 system suggested that Sr interacted with UO2 via a near surface, highly coordinated complex at pH ≥ 10. EXAFS measurements for the U(IV)–silicate samples showed outer-sphere Sr sorption dominated at acidic and near-neutral pH with intrinsic Sr-silicates forming at pH ≥ 12. These complex interactions of Sr with important U(IV) phases highlight a largely unrecognized control on 90Sr mobility in environments of relevance to spent nuclear fuel management and storage.

Separation of 131I from Uranium Fission Products by Dry Distillation

Separation of 131I from Uranium Fission Products by Dry Distillation

Rapid Separation of Photofissioned Uranium Products via a Single-Pass Multiplexed Chromatographic Fission Product Separation System.

Current state-of-the-art fission product separations frequently involve multiple independent separation columns and sample manipulation processes; to couple these processes together, multiple evaporation and transposition steps are often required. The addition of these steps results in lengthy separation times, increased analysis costs, the potential for sample loss, and release of radioactive contamination. We report a new semiautomated method for the rapid separation of U, Zr, Mo, Ba, Sr, Te, and lanthanide fission products from irradiated uranium samples. Chemical yields for U, Zr, Ba, Sr, Te and the lanthanides from less than 3-day old uranium fission product samples are consistently greater than 90%, while those of Mo are greater than 70%. This method minimizes the use and addition of oxidation and reduction reagents that often cause issues with retention and separation. Uranium dissolution and fission product separations using this single-pass method are achievable in under 2 h, representing a significant improvement over traditional gravimetric uranium fission product separation procedures.

Форми знаходження радіонуклідів у донних відкладеннях приміщення 012/7 об’єкта «Укриття»

After installation of the Arch of New Safe Confinement (NSC) into design position, precipitations intake in the underroof space of the Shelter object ceased, and water evaporation began from the unorganized accumulations of radioactively contaminated water. As a result of physical and chemical processes of coagulation and deposition of sparingly soluble salts, bottom sediments, containing uranium and radionuclides, formed. The aim of the work is to determine speciation forms of uranium and radionuclides in the bottom sediments, which formed due to the drying out of unorganized accumulations of radioactively contaminated water in room 012/7 of the NSC – Shelter object. Results of the experimental determination of speciation forms of uranium and radionuclides 90Sr, 137Cs, 154Eu, 238Pu, 239+240Pu, 241Am and 244Cm in the bottom sediments of room 012/7 of the NSC – Shelter object are presented. An amount of water-soluble, exchangeable and acid-soluble uranium forms, fission products and transuranium elements in the bottom sediments of room 012/7 of the NSC – Shelter object was determined by the method of consecutive extractions. Content of uranium, 90Sr, 238Pu, 239+240Pu, 241Am and 244Cm in the leaching solutions and insoluble residue was determined by the ion-exchange method. 90Sr activity was determined by beta-radiometric measurements. Activity of 234,235,236,238U, 238Pu, 239+240Pu, 241Am and 244Cm was determined by alpha-spectrometric measurements of uranium, plutonium and americium sources. The concentration of uranium in the bottom sediments of room 012/7 was 1.02 ± 0.26 mg/g. Abundance content of uranium isotopes is 234U – 0.0159, 235U – 1.09, 236U – 0.192 and 238U – 98.71%. Isotope content of uranium in the bottom sediments, on the whole, corresponds to the isotope content of uranium in the irradiated fuel of the Chornobyl NPP 4th unit with an average burnup. The specific activity of 90Sr and 137Cs in the bottom sediments of room 012/7 is in the range of 6∙108−1∙109 Bq/kg. 239+240Pu and 241Am specific activity in the bottom sediments is in the range of 6∙105 – 8∙106 Bq/kg. 90Sr, 137Cs, 154Eu, 238Pu, 239+240Pu, 241Am and 244Сm radionuclides in the bottom sediments are in different physical and chemical forms, which determine their potential mobility. Mainly uranium and 137Cs in the bottom sediments of room 012/7 are in the exchangeable forms. An amount of water-soluble forms of uranium and 137Cs is about 1.5 – 3%. The main amount of 90Sr, more than 60% is in the carbonate forms, which are very soluble in weak acid medium, when pH is 4.8. More than 65% of 238Pu and 239+240Pu in the bottom sediments of room 012/7 are in the acid-soluble forms. Potential mobility of 241Am in the bottom sediments is notably higher than of plutonium in the soluble state, when pH 4.8, transfer more than 40% of 241Am. Relations of activities of radionuclides 137Cs/90Sr, 90Sr/239+240Pu, 241Am/239+240Pu and 44Cm/239+240Pu in the bottom sediments of room 012/7 differ significantly from the similar relations in the fuel containing materials of the Shelter object.

A history of Hanford tank waste, implications for waste treatment, and disposal

AbstractMore than 40 years of plutonium processing have left almost 56 million gallons of mixed radioactive waste sequestered in 177 underground tanks on the Hanford Site. Three different processing technologies were employed for plutonium purification in addition to uranium scavenging and fission product removal from the tank waste. All of these chemical processes have contributed to a complex waste stream that varies from tank to tank that presents downstream processing challenges to render the waste into a safe form for long‐term storage. The current disposition pathway for Hanford tank waste is vitrification. To maximize waste loading and minimize the number of high‐level waste canisters stored in a geologic repository, pretreatment of the waste is required. Both pretreatment and vitrification operations are impacted by the waste composition.

Simulation of ion flux of actinides and uranium fission products in the plasma separator with a potential well

This work is devoted to the development of a plasma mass separation method with a potential well for spent nuclear fuel reprocessing. The configuration of the separation chamber with an axial magnetic field up to 0.25 T and a radial electric field up to 3 kV/m is considered. Using numerical simulation, we study the ion flux motion with the same mass composition as the spent nuclear fuel injected along magnetic field lines. The effect of fields and initial injection parameters on the spatial separation of actinides from uranium fission products is investigated. The simulation of the ion flux motion is also performed taking into account elastic collisions of ions with background gas atoms. Elastic collision cross sections for U+, Pu+, Cs+, and Sr+ ions in helium and argon are obtained theoretically. We show that in argon, the separation is possible up to a pressure of the order of 1 mTorr, while in helium, it is possible to separate elements by mass groups in the collisional regime at pressures up to about 10 mTorr.

Influence of zirconium concentration on the extraction of technetium (VII) in the processing SNF

Technetium is part of the fission products of uranium formed during the operation of a nuclear reactor. The accumulation of technetium in the fuel is about 1090 g/t-U with a burn up of 40 GW day/t-U. The half-life of Tc-99 is 212 thousand years and its removal into radioactive waste together with relatively short-lived (T1/2 = 30–50 years), the handling of which implies near-surface storage, seems unacceptable. In view of this, it is advisable to remove Tc from the 1st cycle raffinate before concentrating and solidifying. All known variants of Tc isolation are based on its ability to complexate with TBP with subsequent extraction, as well as to co-extract in anionic form with U (VI) and tetravalent elements. To this end, we studied the effect of zirconium (VI) concentration on the extraction of technetium (VII) in the processing of spent nuclear fuel (SNF).

Identification of short-lived fission products in delayed gamma-ray spectra for nuclear material signature verifications

Short-lived Fission Products (FPs) of both uranium (U) and plutonium (Pu) isotopes were identified in neutron irradiated nuclear material (NM) standards in the high-energy gamma-rays range in the perspective of the utilisation of the Delayed Gamma-ray Spectrometry (DGS) technique for nuclear material signature verifications. DGS is a promising active Non-Destructive Assay (NDA) technique that found applications not only in nuclear safeguards but also in nuclear security.This paper presents results of measurement campaigns that include the development of experimental set-up, sample irradiation, gamma-ray spectra acquisitions and qualitative analysis of the obtained gamma-spectra. The experimental setup-up is based on an HPGe planar detector of high-energy resolution and an irradiation chamber made of high-density polyethylene (HDPE) that accommodates a sealed 252Cf source of 50 MBq activity. The delayed gamma-ray spectra of standard samples of U and Pu neutron interrogated samples are successfully acquired using the cyclic neutron irradiation method. 35 and 24 FPs with half-lives ranging from dozens seconds to minutes were identified in the irradiated U and Pu samples, respectively. In the energy range 2–4.9 MeV of the acquired DGS gamma-ray spectra; we have successfully identified 72 gamma-ray peaks of 21 FPs in U and Pu standard samples. This large number of identified FPs is auspicious to the full development of DGS as new active NDA technique for nuclear signature verifications.

Removal of cesium ions from aqueous solutions using various separation technologies

Cesium is the major fission product of uranium, which widely exists in radioactive wastewater. Radiocesium has potential adverse effects on human health and ecological environment. Different methods such as chemical precipitation, coagulation/co-precipitation, solvent extraction, membrane process, chemical reduction, and adsorption have been used to remove radioactive cesium from aqueous solution. However, the development of innovative technologies capable of selectively removing radioactive cesium is still imperative yet challenging. This review focused on cesium removal using various separation technologies, including chemical precipitation, solvent extraction, membrane separation, and adsorption. The key restraints for cesium removal, as well as the recent progress of these methods have also been discussed. Particular attention has been paid to the adsorption methods, which has been highlighted by introducing the latest advances in inorganic adsorbents (such as metal hexacyanoferrates, clay minerals, carbon-based-adsorbents, and ammonium molybdophosphate), organic adsorbents (such as ion exchange resin, metal–organic frameworks and supramolecular/indicator grafting adsorbents), and biosorbents (such as agroforestry wastes and microbial biomass). Adsorption-based methods are high efficient in separation of cesium ions from aqueous streams, and adsorption of cesium ions has been investigated intensively and even used in practical applications, there is still considerable scope for improvement in terms of adsorption capacity and selectivity.

Dispersion of Aluminum Hydroxide with Uranium Fission Products as a Possible Cause of Accelerated Release of Aluminum into the Primary Circuit of a Research Nuclear Reactor

The problem of an increase in the Al concentration in the primary coolant of a research nuclear reactor is considered. The dispersion of aluminum hydroxide by fission products of uranium sorbed on the reactor core surfaces is suggested as a possible cause of this phenomenon. The correlation between the Al concentration and the reactor core contamination with uranium was considered with the IVV-2M research nuclear reactor as an example. Contamination of the reactor core with uranium was accompanied by a considerable increase in the activity concentrations of 24Na, 27Mg, and 239Np washed out from oxide deposits on the reactor core surfaces, which confirms the hypothesis put forward.

The removal of radioactive strontium ions from aqueous solutions by isotopic exchange using strontium decavanadates and corresponding mixed oxides

Two types of strontium decavanadate samples (A and B) were synthesized conductometrically in different pH ranges of 6.7–5.6 and 4.2–3.5, respectively. The samples were used as isotope exchangers after heat treatment at 333, 378 and 573 K for the removal of 90Sr radionuclide from the aqueous solution which is one of the most important fission products of uranium. The content of crystalline water in the exchangers was determined by thermo-gravimetric analysis (TGA). They were characterized using the X-ray diffraction (XRD), diffuse reflectance infrared Fourier-transforms (DRIFT) and the energy dispersive spectral (EDS) analysis. The SEM micrographs showed that sample A dried at 333 K has an urchin-like structure but it decomposes as annealed at 573 K. Sample B dried at 333 K is composed of quadrangular-prism like micro-rods but nano-sized platelets also formed at 573 K. The time-dependent studies conducted at three different molar ratios of strontium ions in the exchanger to the solution showed that Sr2+ ions completely exchanged at the ratio of 3.5 and 2.0 for the samples A and B annealed at 573 K, respectively. The kinetic data were analyzed using McKay, Nernst Planck, and Paterson’s models and correlated to structural characteristics of the exchangers. The Sr-removal capacities of the exchangers evaluated by applying the Langmuir isotherm model increased with the annealing temperature. The results revealed that radioactive Sr2+ ions can be selectively removed by the exchangers annealed at 573 K from real wastewater because the exchange fractions are not significantly affected by the pH changes and the common coexisting cations.

Determination of concentrations of fission products by ICP–AES in solutions from spent nuclear fuel reprocessing

A procedure was developed for determining the gravimetric concentration of U and fission products (Ba, Mo, Zr, Sr, Rb, Pd, and Cd) in solutions from spent nuclear fuel (SNF) reprocessing, using inductively coupled plasma atomic emission spectrometry (ICP–AES). The optimum analysis parameters were found. The analytical lines that are not significantly affected by the matrix element were chosen for the elements being determined. The total margin of error in measurements of the metal concentrations (0.1–2500 mg L–1) by the suggested procedure is in the interval 1–20% (for the minimal concentration). The detection thresholds of uranium fission products in model solutions at optimum parameters vary in the interval from 0.020 to 0.1 mg L–1.