Uranium Nitrate Solution Research Articles

Acute radiation sickness due to critical accidents at a nuclear production facility

Introduction. Nuclear accidents are rare, however, today they cannot be entirely excluded. Thereupon the wealth of global experience is of particular value for prediction of outcomes that potentially follow short-term exposures to ionizing radiation at high dose rates.The objective is to describe clinical simptoms and signs in workers exposed to acute gamma-neutron radiation.Methodology. This paper considers healths effects after radiation exposure due to a critical accident that happened on January, 2 1958 during operations with an experimental flask containing uranium nitrate solution at the Mayak PA. During accident four workers were exposed to short-term gamma-neutron radiation and subsequently suffered from acute radiation syndrome of different severity degree.Results and discussion. Symptoms of the initial response, development of disorders in various organs and systems and the course of the acute radiation syndrome in exposed individuals are described.Conclusion. Multiple organs or tissues are involved in severe acute radiation syndrome at an early period after acute exposure, at that cardiovascular collapse is the most critical physiological state. Multiple organs or tissues failure in the longtrem period after acute exposure is a secondary manifestation both of the primary nonspecific response to radiation and specific radiation-induced direct damage of highly sensitive cells

Treatment of Uranyl Nitrate Solution by Nanofiltration

The traditional evaporation process has obvious disadvantages when treating uranyl nitrate with a uranium concentration less than 10 g/L, such as more ancillary equipment, high energy consumption, and high cost. By contrast, nanofiltration equipment has low integration, and multivalent cations can be rejected effectively by nanofiltration membranes. In this work, a spiral-wound DK1812 nanofiltration membrane with an area of 0.325 m2 was used to treat a uranium nitrate solution with a uranium concentration of 10 g/L. The uranium concentration in the permeate is 16.91 mg/L, which means that the uranium rejection rate is 99.83% and the permeate flux of the solution is 71.1 L/(m2·h) under the conditions of a feed temperature of 30°C, a tangential velocity of 30 cm/s, and a transmembrane pressure of 1.5 MPa.

Studies on nucleation and crystal growth kinetics of uranium(IV) oxalate

Abstract An improved apparatus is used for nucleation measurements according to Nielsen’s method. A new method is proposed to calculate the dilution ratio N of the reaction solution during nucleation rate determination. With the rule, when the initial apparent supersaturation ratio S′ = f(N) in the dilution tank is controlled from 1.2 to 2.7, crystal nucleus dissolving and secondary nucleation can be avoided satisfactorily. Experiments are realized by varying the supersaturation ratio from 26.0 to 297.5 and temperature from 30 °C to 50 °C. Uranium(IV) oxalate is precipitated by mixing equal volumes of tetravalent uranium nitrate and oxalic acid solution. The experimental results show that the nucleation rate of uranium(IV) oxalate in the supersaturation range as show above is characterized by the primary homogeneous mechanism and can be expressed by the equation R N = A N exp ( − E a / R T ) exp [ − B / ( ln S ) 2 ] , ${R_N} = {A_N}{\rm{exp}}( - {E_a}/RT){\rm{exp}}[ - B/{({\rm{ln }}S)^2}],$ where AN = 1.9 × 1027 m−3s−1, Ea = 71.2 kJ mol−1, and B = 34.3. The crystal growth rate can be expressed by the equation G ( t ) = k g exp( − E ′ a / R T )( c − c eq ) g , $G(t) = {k_g}{\rm{exp(}} - {E^{\prime}_a}/RT{\rm{)(}}c - {c_{{\rm{eq}}}}{{\rm{)}}^g},$ where kg = 7.1 × 105 (mol/L)−0.98 (m/s), E ′ a = 33.1 kJ mol − 1 , ${E^{\prime}_a} = 33.1 \ {\rm{ kJ \ mo}}{{\rm{l}}^{ - 1}},$ and g = 0.98. The results indicate that the crystal growth of tetravalent uranium(IV) oxalate is controlled by the BCF model.

Uranium biosorption by Lemna sp. and Pistia stratiotes

Biosorption-based technologies have been proposed for the removal of radionuclides from radioactive liquid waste containing organic compounds. Nevertheless, pytoremediation potential of uranium (U) by nonliving aquatic macrophytes Lemna sp. and Pistia stratiotes has not been previously addressed. In this study, uranium biosorption capacity by Pistia stratiotes and Lemna sp. was evaluated by equilibrium and kinetics experiments. The biomasses were added to synthetic and real waste solutions. The assays were tested in polypropylene vials containing 10 mL of uranium nitrate solution and 0.20 g of biomass. Solutions ranging from 0.25 to 84.03 mmol l−1 were employed for the assessment of uranium concentration in each macrophyte. The equilibrium time was 1 h for both macrophytes. Lemna sp. achieved the highest sorption capacity with the use of the synthetic solution, which was 0.68 mmol g−1 for the macrophyte. Since Lemna sp. exhibit a much higher adsorption capacity, only this biomass was exposed to the actual waste solution, being able to adsorb 9.24 × 10−3 mmol g−1 U (total). The results show that these materials are potentially applicable to the treatment of liquid radioactive waste.

Preparation and certification of the uranium nitrate solution reference materials series IRMM-2019 to IRMM-2029 for the isotopic composition

The IRMM-2019-2029 series of uranium nitrate solutions was certified for the isotopic composition. The solutions were originally prepared for the certification of the IRMM-019-029 series of uranium hexafluoride (UF6) materials in 2014, for which certification measurements were performed by thermal ionization mass spectrometry after conversion of UF6 into uranyl nitrate solutions. Analyses were performed using the modified total evaporation method and for some materials with low 236U abundances, the n(235U)/n(238U) ratio was measured using a n(233U)/n(236U) double spike. The integrity of the certified values after dilution and ampouling of the original solutions was checked through process control measurements.

Validation of gravimetry for high-accuracy analysis of uranium

Gravimetry after calcination is an assay method for determining the quantity of a given element from the mass measurement. This method, adapted for high concentration and purity uranium solution, has the advantage of not being related to a chemical standard and of having an extremely high trueness and precision. Besides, the method is simple, robust and cost-effective. The results of the analytical method validation are shown. Expanded uncertainties, assessed thanks to a reference uranium nitrate solution, are lower than 0.2%. The technique can thus be used in addition to the well-known isotopic dilution thermal-ionisation mass-spectrometry.

Improved fluorimetric measurement of uranium uptake and distribution in spring wheat (Triticum aestivum L.)

Abstract Uranium uptake and (radio)toxicity was tested on spring wheat (Triticum aestivum L.) in a laboratory study using differently concentrated uranium nitrate solutions. Within these experiments, two analytical assays of uranium were comparatively tested: a fast and improved fluorimetric assay and the classical colorimetric (U(IV)–arsenazo(III) complexation) one. During the germination, the wheat seeds and plantlets supported well the uranium solutions of treatment within the entire concentration range (1E-4–5E-3 M). Uranium proved to be non (radio)toxic to wheat as compared with other natural and anthropogenic radiocations, probably because its uptake by spring wheat during the germination is low. Indeed, only a small fraction of uranium administered was located within the roots, whereas the uranium content of the stems was negligible. A high correlation between the results obtained by two analytical methods was found. However, the fluorimetric assay proved to be more reliable and fast, and accurate.

Reexamination of Criticality Accident in JCO

Nuclear chain reactions are, by now, commonly used in the nuclear reactors, and thus it seems that there is no basic problem in fission processes from the scientific point of view. However, the criticality accident that occurred in JCO in 1999 suggests that one should carefully examine this accident from the nuclear physics point of view. Indeed the chain nuclear reactions should have taken place in the small area of space with 45 cm diameter disk times 30 cm height tank. In fact, when people carry the uranium nitrate solution into sedimentation tank, then this solution with uranium should get into the critical state at the 45? of uranium nitrate solution. The root cause of the accident should not be very simple from the nuclear physics point, and it should be quite important to examine why the uranium nitrate solution with 45? could have become critical.

Bioaccumulation characterization of uranium by a novel Streptomyces sporoverrucosus dwc-3

The biosorption mechanisms of uranium on an aerobic bacterial strain Streptomyces sporoverrucosus dwc-3, isolated from a potential disposal site for (ultra-)low uraniferous radioactive waste in Southwest China, were evaluated by using transmission electron microscopy (TEM), energy dispersive X-ray (EDX) analysis, Fourier transform infrared spectroscopy (FT-IR), X-ray photoelectron spectroscopy (XPS), proton induced X-ray emission (PIXE) and enhanced proton backscattering spectrometry (EPBS). Approximately 60% of total uranium at an initial concentration of 10mg/L uranium nitrate solution could be absorbed on 100mg S. sporoverrucosus dwc-3 with an adsorption capacity of more than 3.0mg/g (wet weight) after 12hr at room temperature at pH3.0. The dynamic biosorption process of S. sporoverrucosus dwc-3 for uranyl ions was well described by a pseudo second-order model. S. sporoverrucosus dwc-3 could accumulate uranium on cell walls and within the cell, as revealed by SEM and TEM analysis as well as EDX spectra. XPS and FT-IR analysis further suggested that the absorbed uranium was bound to amino, phosphate and carboxyl groups of the cells. Additionally, PIXE and EPBS results confirmed that ion exchange also contributed to the adsorption process of uranium.

99Mo Separation from High-Concentration Irradiated Uranium Nitrate and Uranium Sulfate Solutions

We report separation data indicating that 99Mo can be separated from a vast excess of either uranyl sulfate or uranyl nitrate in irradiated dilute acid solutions. These results suggest that, if medical isotope 99Mo is produced during fission of high concentrations of low enriched uranium sulfate solution fuel, it is feasible to both recover >90% of the 99Mo for further purification and the uranium for recycle.

Adsorption and desorption of uranium (VI) in aerated zone soil

In this paper, the adsorption and desorption behavior of uranium (VI) in aerated zone soil (from Southwest China) was systematically investigated using a static experimental method in order to provide useful information for safety assessment of the disposal of (ultra-)low uraniferous radioactive waste, as well as a potential remediation method for uranium-contaminated soils. The adsorption behavior of uranium (VI) was firstly studied by batch experiments as functions of contact time, pH, liquid/solid ratio, temperature, colloids, minerals and coexistent ions. The results indicated that the adsorption of uranium (VI) by natural soil was efficient at an initial concentration of 10 mg/L uranium (VI) nitrate solution with 100 mg natural soil at room temperature when pH is about 7.0. The adsorption was strongly influenced by the solution pH, contact time, initial concentration and colloids. The adsorption equilibrium for uranium (VI) in soil was obtained within 24 h and the process could be described by the Langmuir adsorption equation. For uranium (VI) desorption, EDTA, citric acid and HNO3 were evaluated under different conditions of temperature, concentration and proportion of liquid to solid. The adsorbed uranium (VI) on natural soil could be easily extracted by all these agents, especially by HNO3, implying that the uranium-contaminated soils can be remedied by these reagents.

Preparation of ammonium diuranate particles by external gelation process of uranium in INET

In order to satisfy the mass spherical fuel elements (SFE) requirements for the High Temperature gas-cooled Reactor (HTR)–Pebble bed Module (HTR-PM), a modified and optimized fabrication of dense uranium dioxide (UO2) kernel via an acid-deficiency uranium nitrate (ADUN) solution was researched and developed in the institute of nuclear and new energy technology (INET). In this study, external gelation process of uranium (EGU) based on the total gelation of uranium (TGU) in the period of 10MW HTR (HTR-10) was used to prepare the ammonium diuranate (ADU) compound particles, an intermediate for UO2 kernel. The scale of batch production was improved from 500g U/batch at that time to current 3kg U/batch. Improvements in many aspects such as large scale preparation of ADUN, multiple-nozzle casting system and new aging, washing and drying (AWD) system without isopropanol were also presented. Under strict controls and measures, ADU particles with good sphericity and uniform diameter were acquired to ensure the subsequent preparation of dense UO2 kernels for HTR-PM.

Estimation of Generated Volume of Radiolytic Gas Voids during Prompt Criticality with 10% Enriched Uranium Nitrate Solution in TRACY

An oscillatory pressure response due to the generation of radiolytic gas voids was observed during a prompt criticality with 10% enriched uranium nitrate solution in the Transient Experiment Critical Facility (TRACY) of Japan Atomic Energy Agency (JAEA). In order to understand the mechanism of the nuclear thermal-hydraulic behavior in solution fuelcriticality accidents, the mechanism of the oscillatory pressure response was studied using both the measured data and visual observation of the solution behavior. A simplified vibration model for the estimation of the generated volume of radiolytic gas voids was proposed using the measured data of the pressure period and the solution mass. Rapid depressurization experiments with carbonic acid water were also carried out in order to verify the application of the simplified vibration model for radiolytic gas voids. Consequently, the generated volume of radiolytic gas voids for the oscillatory pressure case was correlated as a function of the energy for the generation of gas voids.

Installation Places of Criticality Accident Detectors in the Plutonium Conversion Development Facility

At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF.

微分パルスボルタンメトリーによる硝酸ウラン溶液中のU(IV)、U(VI)同時定量

微分パルスボルタンメトリーによる硝酸ウラン溶液中のU(IV)、U(VI)同時定量

Estimation of Generated Volume of Radiolytic Gas Voids during Prompt Criticality with 10% Enriched Uranium Nitrate Solution in TRACY

Estimation of Generated Volume of Radiolytic Gas Voids during Prompt Criticality with 10% Enriched Uranium Nitrate Solution in TRACY

Fabrication of Cermet Nuclear Fuels Designed for the Transmutation of Transuranic Isotopes

The Uranium Extraction (UREX) family of processes uses solvent extraction techniques designed to partition spent uranium and transuranic (TRU) isotopes from fission product waste. Once separated, the collective TRU elements (Np, Pu, Am, and Cm) can be recycled in advanced nuclear energy systems. A zirconium matrix cermet is proposed as a fuel form for this application. Processing methods have been designed to convert the TRU product and spent Zircaloy cladding into feed materials for the hot extrusion of the cermet fuel pins. The TRU conversion process is being developed using a surrogate mixture of uranium and cerium nitrate solutions to generate mixed oxide microspheres. The Zircaloy recovery process is a hydride-dehydride method that is being demonstrated at the bench scale. The powder products from these methods may be combined through hot extrusion into a cermet composite; demonstration experiments using zirconium powder and zirconia microspheres have been completed.

Validation of Simplified Evaluation Models for First Peak Power, Energy, and Total Fissions of a Criticality Accident in a Nuclear Fuel Processing Facility by TRACY Experiments

Simplified evaluation models are developed at the Japan Atomic Energy Research Institute (JAERI) to predict the first peak power, energy, and total fission numbers during a criticality accident for design and installation of a criticality alarm system and for quick response with measures to avoid excessive exposure of the general public. These models were first derived in previous papers only from theoretical considerations employing one-point reactor kinetic neutron behavior and thus are applicable to any geometrical shape of vessel containing fissile solution. Applicability concerning nuclide composition comes essentially from using empirical equations describing specific heat and density to give simplified forms of the models. The models developed originally for a stepwise reactivity insertion mode are shown in the current paper to approximately stand for the ramp reactivity insertion mode by giving their theoretical formation and are validated by applying experimental data from JAERI’s Transient Experiment Critical Facility (TRACY) on a low-235U-enriched uranium nitrate solution as well as CRAC experiments on high-235U-enriched uranium nitrate solution together with past accident data, including the most recent JCO accident.

Dielectric Properties of Uranium and Plutonium Nitrate Solution and the Oxide Compounds Formed in the De-nitration Process by the Microwave Heating Method

In the microwave heating method, a liquid mixture of uranium nitrate and plutonium nitrate solution (U/Pu nitrate solution) is co-converted to MOX powder through four steps. Dielectric properties of U/Pu nitrate solution and de- nitrated MOX powder were measured. The reflection method was used for U/Pu nitrate solution and U3O8 powder, while the perturbation method was used for 50%PuO2-UO3, β-UO3, γ-UO3 and U3O8 powders. The measured dielectric loss product epsilon′ tan δ of U/Pu nitrate solution was strongly affected by the concentration of nitrate, but there was little dependence on the dielectric oscillation of uranium and plutonium ions. This behavior was explained by the Maxwell-Wagner effect. When U/Pu nitrate solution was continuously heated by microwaves, dielectric loss product increased until the boiling point and then it decreased rapidly, but temperature increased steadily at the same time due to the sharp decrease in heat capacity of the media. The temperature histories in the solution and compounds formed in the de-nitration process by the microwave heating method could be explained from the measured dielectric properties.

Dielectric Properties of Uranium and Plutonium Nitrate Solution and the Oxide Compounds Formed in the De-nitration Process by the Microwave Heating Method

Dielectric Properties of Uranium and Plutonium Nitrate Solution and the Oxide Compounds Formed in the De-nitration Process by the Microwave Heating Method