Ammonium Diuranate Research Articles

Fabrication of Thorium and Thorium Dioxide

Thorium based nuclear fuel is of immense interest to India by virtue of the abundance of Thorium and relative shortage of Uranium. Thorium metal tubes were being cold drawn using copper as cladding to prevent die seizure. After cold drawing, the copper was removed by dissolution in nitric acid. Thorium does not dissolve being passivated by nitric acid. Initially the copper cladding was carried out by inserting copper tubes inside and outside the thorium metal tube. In an innovative development, the mechanical cladding with copper was replaced by electroplated copper with a remarkable improvement in thorium tube acceptance rates. Oxalate derived thoria powder was found to require lower compaction pressures compared to ammonium diuranate derived urania powders to attain the same green compact density. However, the green pellets of thoria were fragile and chipped during handling. The strength improved after introducing a ball milling step before compaction and maintaining the green density above the specified value. Alternatively, binders were used later for greater handling strength. Magnesia was conventionally being used as dopant to enhance the sinterability of thoria. The normal sintering temperature for magnesia doped thoria was 1600℃ - 1700℃, which was achieved in electrically heated molybdenum element sintering furnaces with reducing atmosphere. 0.25 mole percent addition of niobia to the thoria was found to bring down the sintering temperature to 1150℃. Sintering could be done in ordinary furnaces in air atmosphere using silicon carbide or Kanthal heating elements. Electrical conductivity was measured for both magnesia and niobia doped sintered thoria and used in interpreting differences in sintering behavior.

Investigation of ammonium diuranate calcination with high-temperature X-ray diffraction

The thermal decomposition of ammonium diuranate (ADU) in air is investigated using in situ high-temperature X-ray diffraction (HT-XRD), thermogravimetry and differential thermal analysis. Data have been collected in the temperature range from 30 to 1000 °C, allowing the observation of phase transformation and the assessment of the energy changes involved in the calcination of ADU. The starting material 2UO3·NH3·3H2O undergoes a process involving several endothermic and exothermic reactions. In situ HT-XRD shows that amorphous UO3 is obtained after achieving complete dehydration at 300 °C, and denitration at about 450 °C. After cooling from heat treatment at 600 °C, a crystalline UO3 phase appears, as displayed by ex situ XRD. The self-reduction of UO3 into orthorhombic U3O8 takes place at about 600 °C, but a long heat treatment or higher temperature is required to stabilise the structure of U3O8 at room temperature. U3O8 remains stable in air up to 850 °C. Above this temperature, oxygen losses lead to the formation of U3O8−x, as demonstrated by subtle changes in the diffraction pattern and by a mass loss recorded by TGA.

Impurity characterization and thermal decomposition mechanism of ammonium diuranate during in-situ synthesis of U3O8 using simultaneous TG–DTA–FTIR and PXRD measurements

Current studies describe the application of simultaneous TG–DTA–FTIR technique in the in-situ synthesis of tri uranium octaoxide (U3O8) and impurity characterization of the ammonium diuranate (ADU) matrix. Formation of U3O8 could be due to the decomposition of ADU to uranium trioxide (UO3) and finally to U3O8 by the temperature of 500°C. Presence of carbon and carbonate could be seen as impurities in the matrix. Carbon content has been determined quantitatively using TG and FTIR techniques and the results are in good agreement. Ammonium nitrate was found as impurity in ammonium diuranate matrix. The PXRD studies confirm the formation of U3O8 as end product during the decomposition of ammonium diuranate. From our studies, it is suggested that ammonium diuranate decomposes to tri uranium octaoxide (U3O8) through evolution of nitrogen, hydrogen and oxygen, and not through ammonia evolution in the literature.

Concentration of ammonium diuranate effluent by reverse osmosis and forward osmosis membrane processes

In this study, both reverse osmosis (RO) and forward osmosis (FO) experiments are conducted to concentrate simulated inactive ammonium diuranate (ADU) filtered effluent solution (by mixing uranyl nitrate and ammonium nitrate) using indigenously developed cellulose acetate blend (CAB) and thin-film composite polyamide (TFCP) membranes. Both the membranes are prepared and characterized in terms of pure water permeability and solute rejection for 2000 ppm NaCl feed, water contact angle and surface average roughness. Subsequently, testing of volume reduction and concentration of simulated ADU effluent solution are carried out using custom made RO and FO testing systems. It is found that in RO process, the performance in terms of volume reduction factor and concentration factor with respect to uranium for both CAB and TFCP membranes are comparable but the CAB membranes show better performance than the TFCP membranes in FO modes. The concentration of ammonium nitrate is less in RO concentrate than in the FO concentrates. Similar experiments with the feed solution having different concentration of uranium (1–20 ppm) with same concentration of ammonium nitrate show that in FO, almost no leaching of uranium is found to the draw solution side but in RO, some of the uranium starts passing to the permeate side, particularly at lower concentration of uranium in the feed.

Effect of ammonium nitrate on precipitation of Ammonium Di-Uranate (ADU) and its characteristics

Scanning Electron Microscopic (SEM) images of ADU prepared with gaseous ammonia, (a) in presence of excess ammonium nitrate (80 g/L) as well as (b) in absence of excess ammonium nitrate at different magnification. Effect of ammonium nitrate on precipitation of Ammonium Di-Uranate (ADU) from nitric acid medium via gaseous ammonia route had been investigated. Studies on effect of ammonium nitrate on precipitation time, particle size, shape, surface morphology , flowability , oxygen/uranium (O/U) ratio and tap density of calcined ADU were carried out at various ammonium nitrate concentrations. It was observed that, the presence of excess ammonium nitrate influences the precipitation time, particle size distribution and surface morphology of the ADU. ADU and uranium oxide were characterized with Scanning Electron Microscopy (SEM) and X-ray diffraction (XRD). Presence of ammonium nitrate during precipitation leads to the formation of bigger, porous and uniform particles as compared to the ADU prepared without ammonium nitrate additions.

졸-겔 방법으로 제조된 Ammonium Diuranate 핵연료 분말의 공정장치 변수에 따른 특성

This paper describes the spherical ammonium diuranate gel particles which are the intermediated material of the <TEX>$UO_2$</TEX> microsphere for an VHTR(very high temperature reactor) nuclear fuel. The characteristics of the intermediate-ADU gel particles prepared by AWD(ageing, washing, and drying) and FB(fluidized-bed) apparatus were examined and compared in a sol-gel fabrication process. The electrical conductivity of washing filtrate from the FB treating and the surface area of dried-ADU gel particles were higher than those of AWD treating. Also, an internal pore volume in dried-ADU gel particles showed a more decrease in AWD treatment than FB treatment because of decomposition of PVA affected by the washing time. However, the internal microstructures of ADU gel particles were similar regardless of the process variation.

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.

Thermal Denitration of Ammonium Nitrate Solution in a Fluidized-Bed Reactor

Ammonium diuranate (ADU) filtrate, which contains mainly ammonium nitrate (80–100 g/L), is generated during hydrometallurgical processing of uranium. This filtrate stream poses a disposal problem because of its high nitrate content and residual radioactivity. Fluidized-bed thermal denitration is considered as a suitable chemical-free disposal option for the aqueous waste nitrate stream. Hence, investigations to explore the decomposition of ammonium nitrate in a fluidized bed have been carried out. To enable theoretical analysis and performance evaluation of the process, a mathematical model was developed. The model is based on two-phase theory of a bubbling fluidized bed. Model calculations were used to predict the axial concentration profile of ammonium nitrate in the emulsion and bubble phases and the axial temperature profiles of gas bubbles, emulsion gas, and emulsion particles. The mechanism of decomposition of ammonium nitrate in a fluidized bed was explored, and the conversion of ammonium nitrate w...

Feasibility of the recovery of uranium from alkaline waste by amidoximated grafted polypropylene polymer matrix

The amidoximated grafted polypropylene polymer matrix was prepared by post irradiation grafting of acrylonitrile (AN) onto thermally bonded non-woven matrix of poly(propylene) sheet using electron beams. This precursor polymer was reacted with hydroxylamine to convert AN to poly(acrylamidoxime) (AO) groups, and conditioned by treating them with 2.5 % KOH at 80 °C for 1 h. The polymer matrix was having the degree of AN grafting ~106 wt% and its subsequent conversion to AO groups ~70 %. The water uptake capacity of AO polymer matrix were found to be 100 ± 5 % (w/w). Quantitative recovery of uranium from alkaline waste (ammonium diuranate supernatant) solution was achieved by this polymer matrix. The other radionuclides present in the waste solution were not extracted by the polymer matrix. For all other radionuclides, the uptake was found to be <6 %.

Study of calcinations of ammonium diuranate at different temperatures

Effect of calcination temperature has been studied on tap density, surface area, porosity, O/U ratio, morphology and crystal phases of uranium oxides. The oxides were produced by calcination of ammonium diuranate (ADU). It has been observed that O/U ratio reduces with an increase in temperature. Surface area and porosity increases with temperature, passes through maxima and then reduces. These observations have been explained using high resolution SEM. The crystal phase analysis has shown that the heating of ADU results in to α-U3O8 via β-UO3.

Study of crystallization and morphology of ammonium diuranate and uranium oxide

Ammonium diuranate (ADU) and its heating product uranium oxides (UO3+U3O8) are important intermediates for nuclear fuel production. The microstructure and particle size distribution (PSD) of ADU powder produced at various stages of precipitation have been measured to study the particle growth of ADU. The change of morphology of ADU and UO3+U3O8 has also been studied when gaseous ammonia is added instead of aqueous ammonia solution and when a seeding agent is added to the uranyl nitrate solution before precipitate formation. It has been observed that the ADU particles form basically agglomerates of submicron platelets. The microstructure of ADU was retained in UO3+U3O8 even after calcination.

Use of microwave radiation for preparing uranium oxides from its compounds

The formation of uranium oxides by thermal decomposition of uranyl diaquadihydroxylaminate monohydrate, ammonium diuranate, ammonium tricarbonatouranylate, and uranium peroxide under the action of microwave (MW) radiation was studied. Uranium dioxide is formed by decomposition of these compounds in a reducing atmosphere at the MW radiation power of 600 W and treatment time of 5–10 min. In air, under the same conditions, U3O8 is formed. Under the action of MW radiation, substandard ceramic pellets of UO2 fuel can be readily converted in air to powdered U3O8. The use of MW radiation for thermal decomposition of uranium compounds allows the power and time consumption to be considerably reduced relative to the process with electrical resistance furnaces. A quick method for gravimetric testing of the composition of uranium oxides (UO2 or U3O8) using MW radiation was suggested.

Prediction model of microwave calcining of ammonium diuranate using incremental improved back-propagation neural network

Prediction model of microwave calcining of ammonium diuranate using incremental improved back-propagation neural network Yingwei LI , Bingguo LIU , Jinhui PENG 1,2)∗, Wei LI , Daifu HUANG 3) and Libo ZHANG 1,2) 1) Faculty of Metallurgical and Energy Engineering, Kunming University of Science and Technology, Kunming 650093, China 2) Key Laboratory of Unconventional Metallurgy, Ministry of Education, Kunming University of Science and Technology, Kunming 650093, China 3) No.272 Nuclear Industry Factory, China National Nuclear Corporation, Hengyang 421002, China Manuscript received 23 May 2010; in revised form 17 October 2010

Sorption of uranium on magnetite nanoparticles

Magnetite (Fe3O4) nanoparticle was synthesized using a solid state mechanochemical method and used for studying the sorption of uranium(VI) from aqueous solution onto the nanomaterial. The synthesized product is characterized using SEM, XRD and XPS. The particles were found to be largely agglomerated. XPS analysis showed that Fe(II)/Fe(III) ratio of the product is 0.58. Sorption of uranium on the synthesized nanomaterials was studied as a function of various operational parameters such as pH, initial metal ion concentration, ionic strength and contact time. pH studies showed that uranium sorption on magnetite is maximum in neutral solution. Uranium sorption onto magnetite showed two step kinetics, an initial fast sorption completing in 4–6 h followed by a slow uptake extending to several days. XPS analysis of the nanoparticle after sorption of uranium showed presence of the reduced species U(IV) on the nanoparticle surface. Fe(II)/Fe(III) ratio of the nanoparticle after uranium sorption was found to be 0.48, lower than the initial value indicating that some of the ferrous ion might be oxidized in the presence of uranium(VI). Uranium sorption studies were also conducted with effluent from ammonium diuranate precipitation process having a uranium concentration of about 4 ppm. 42% removal was observed during 6 h of equilibration.

Evaluation of N,N-dialkylamides as promising process extractants

Studies carried out at BARC, India on the development of new extractants for reprocessing of spent fuel suggested that while straight chain N,N-dihexyloctanamide (DHOA) is promising alternative to TBP for the reprocessing of irradiated uranium based fuels, branched chain N,N-di(2-ethylhexyl)isobutyramide (D2EHIBA) is suitable for the selective recovery of 233U from irradiated Th. In advanced fuel cycle scenarios, the coprocessing of U/Pu stream appears attractive particularly with respect to development of proliferation resistant technologies. DHOA extracted Pu(IV) more efficiently than TBP, both at trace-level concentration as well as under uranium/plutonium loading conditions. Uranium extraction behavior of DHOA was however, similar to that of TBP during the extraction cycle. Stripping behavior of U and Pu (without any reductant) was better for DHOA than that of TBP. It was observed during batch studies that whereas 99% Pu is stripped in four stages in case of DHOA, only 89% Pu is stripped in case of TBP under identical experimental conditions. DHOA offered better fission product decontamination than that of TBP. GANEX (Group ActiNide EXtraction) and ARTIST (Amide-based Radio-resources Treatment with Interim Storage of Transuranics) processes proposed for actinide partitioning use branched chain amides for the selective extraction of uranium from spent fuel feed solutions. The branched-alkyl monoamide (BAMA) proposed to be used in ARTIST process is N,N-di-(2-ethylhexyl)butyramide (D2EHBA). In this context, the extraction behavior of U(VI) and Pu(IV) were compared using D2EHIBA, TBP, and D2EHBA under similar concentration of nitric acid (0.5 — 6M) and of uranium (0-50g/L). These studies suggested that D2EHIBA is a promising extractant for selective extraction of uranium over plutonium in process streams. Similarly, D2EHIBA offered distinctly better decontamination of 233U over Th and fission products under THOREX feed conditions. The possibility of simultaneous stripping and precipitation of thorium (as oxalate) from loaded organic phase was explored using 0.05M oxalic acid. Ammonium diuranate (ADU) precipitation was performed on the oxalate supernatant for the recovery of uranium. Quantitative recovery (>99.9%) of Th as well as of U was achieved. Radiolytic studies suggested that irradiated DHOA and D2EHIBA behaved better with respect to fission product decontamination as compared to that of TBP.

Thermal decomposition of ammonium diuranate and uranium peroxide

Thermal decomposition of ammonium diuranate and uranium peroxide

Ammonium Polyuranate Precipitation from Fluoride Solutions: An Experimental Study of the UO3-HF-NH3-H2O System as Applied in UF6 Processing

In the ammonium diuranate (ADU) process, UF6 is reacted with water, and the acidic solution of uranyl fluoride is treated with aqueous ammonia to precipitate ammonium polyuranate for subsequent reduction to UO2 and production of fuel pellets for commercial nuclear reactors. Our experiments simulated adding aqueous ammonia to the reaction products of UF6 and water in typical ADU processes. Chemical and X-ray diffraction analysis of products from the experiments are consistent with postulated chemical equilibria in which solids with structures close to that of ammonium polyuranate are formed from co-precipitation of the NH4+(aq) cation with (previously unreported) anions of the form UO2F3-x(OH)x-(aq). More efficient separations of solid products were obtained at NH4OH:UF6 ratios of 19 or greater, with x closer to the value of 3 for the hypothetical formation of pure ammonium polyuranate. Supplementary experiments in the current study and a previous study in our laboratory indicated that nominal uranium concentrations of 90 mg/l in the filtrate resulting from such separations could be reduced to microgram per liter levels by batch mixing a 1-to-2.5 aqueous diluate of the filtrate with the Diphonix® ion exchange resin. Our study further demonstrated that reaction of the purified NH4OH-NH4F diluate with aqueous Ca(OH)2 at 80 to 90°C could produce essentially uranium-free CaF2 and an ammonia distillate, as useful waste-conversion end products from a modified ADU process.

Some results uranium dioxide powder structure investigation

Features of the macrostructure and microstructure of uranium dioxide powders are considered. Assumptions are made on the mechanisms of the behavior of powders of various natures during pelletizing. Experimental data that reflect the effect of these powders on the quality of fuel pellets, which is evaluated by modern procedures, are presented. To investigate the structure of the powders, modern methods of electron microscopy, helium pycnometry, etc., are used. The presented results indicate the disadvantages of wet methods for obtaining the starting UO2 powders by the ammonium diuranate (ADU) flow sheet because strong agglomerates and conglomerates, which complicate the process of pelletizing, are formed. The main directions of investigation that can lead to understanding the regularities of formation of the structure of starting UO2 powders, which will allow one to control the process of their fabrication and stabilize the properties of powders and pellets, are emphasized.

고온가스로용 핵연료 제조에서 열처리 조건이 우라늄산화물 입자 특성에 미치는 영향

The effects of thermal treatment conditions on ADU (ammonium diuranate) prepared by SOL-GEL method, so-called GSP (Gel supported precipitation) process, were investigated for <TEX>$UO_2$</TEX> kernel preparation. In this study, ADU compound particles were calcined to <TEX>$UO_3$</TEX> particles in air and Ar atmospheres, and these <TEX>$UO_3$</TEX> particles were reduced and sintered in 4%-<TEX>$H_2$</TEX>/Ar. During the thermal calcining treatment in air, ADU compound was slightly decomposed, and then converted to <TEX>$UO_3$</TEX> phases at <TEX>$500^{\circ}C$</TEX>. At <TEX>$600^{\circ}C$</TEX>, the <TEX>$U_3O_8$</TEX> phase appeared together with <TEX>$UO_3$</TEX>. After sintering of theses particles, the uranium oxide phases were reduced to a stoichiometric <TEX>$UO_2$</TEX>. As a result of the calcining treatment in Ar, more reduced-form of uranium oxide was observed than that treated in air atmosphere by XRD analysis. The final phases of these particles were estimated as a mixture of <TEX>$U_3O_7$</TEX> and <TEX>$U_4O_9$</TEX>.

Uranium/technetium separation for the UREX process – synthesis and characterization of solid reprocessing forms

AbstractIn the context of the demonstration of the UREX (uranium extraction) process, a separation of uranium and technetium using an anion exchange resin was performed on a simulant solution containing 98.95 g/L of238U and 130.2 mg/L of99Tc. After sorption on the resin, TcO4−was eluted with NH4OH, the eluting stream was treated, and the technetium converted to Tc metal (yield=52.5%). The purity of the compound was analyzed: it contains less than 23.8 μg of238U per gram of99Tc. Tc metal was characterized by X-ray diffraction (XRD) and X-ray absorption fine structure (XAFS) spectroscopy; EXAFS analysis clearly confirms the hexagonal structure of the metal obtained after treatment. Uranium was converted to ammonium diuranate and to U3O8in a yield of 88.2%, analysis indicates that the compound contains less than 0.16 μg of99Tc per gram of ammonium diuaranate.