Preparation of Uranium Carbonitride by Reaction of UC with Ammonia

With the view to establishing a method of directly converting uranium carbide into uranium carbonitride and hydrocarbons, an attempt has been made to induce reaction between UC and ammonia under various temperatures from 25° to 600°C and pressures from 1 to 1,500 kg/cm². The reaction aimed at was realized to the extent of practical significance under pressures exceeding 500 kg/cm² at 450°C and exceeding 250 kg/cm² at 500°C. The hydrocarbons produced thereby were found to be mainly methane, indicating that the formula of the predominant reaction was

• UC+NH₃→UN_2-x+ CH₄+H₂.

Upon heating the powdery fine black product to 1,800°C in vacuo, unexpectedly marked sintering was found to occur, resulting in dense uranium monocarbonitride without any compaction pretreatment.     

Preparation of Uranium Nitrides from Uranium Tetrafluoride

A method has been developed for preparing UN by two-step reaction using UF₄. As a result, the reaction of UF₄ with Si and N₂ has proved quite practical for obtaining U₂N₃. It was confirmed by chemical analysis and by X-ray diffraction analysis that single phase U₂N₃ was obtained at reaction temperatures above 1,200°C, and only 3% of the uranium was lost by evaporation even at 1,200°C. In the second reaction, the dissociation of U₂N₃ to UN in vacuum took place rapidly above 1,050°C, at which temperatures single phase UN was obtained, possessing a lattice constant of 4.888 Å.     

Preparation of Uranium Disulfide from Uranium Chloride

A method has been developed for the preparation of US₂ using UCl₄, as a first step in the preparation of uranium monosulfide. The reaction of UCl₄, A1 and H₂S has been proved to be quite practical. It was confirmed by chemical analysis and X-ray diffraction analysis that almost pure US₂ was obtained at temperatures above 400°C, and the conversion efficiency of uranium reached almost 100%.     

Preparation of Uranium Nitrides from Uranium Tetrachloride

A study was made of UN preparation by two-step reactions with UCl₄, Al and N₂ gas. In the first step, an intermediate uranium nitride with an N/U ratio of 1.68 resulted from the reaction of UCl₄ with A1 under constant flow of N₂ at reaction temperatures higher than 800°C. The X-ray pattern of the intermediate nitride did not correspond to any previously known uranium nitride. A maximum conversion efficiency of about 70% was obtained at temperatures between 800° and 1,000°G for the first reaction. In the second reaction, the intermediate nitride was heat treated under vacuum. To obtain single phase UN from the intermediate nitride, the heat treatment required a temperature of at least 1,100°C, at which the minimum holding time was 60 min.     

Preparation of High Density Uranium Nitride and Uranium Carbonitride Fuel Pellets

Uranium nitride and uranium carbonitride fuel pellets were prepared for irradiation in the Japan Material Testing Reactor. The pellets are 6.9 mm in diameter and 7 mm long, and are of natural and 5% enriched uranium. Uranium nitride powder was prepared from uranium metal via hydride and higher nitride. Uranium carbide powder was prepared from uranium metal by hydriding and then reacting with propane. The lowest possible reaction temperatures were selected to obtain fine and reactive powders. Uranium nitride and mixed powders of different ratios (UC: UN = 1: 3, 1:1 and 3: 1) were cold pressed without binder. Sintering was carried out in a tungsten crucible in vacuum (10~⁴ mmHg) for 2 hr at 1,900°–2,000°C. The density of the pellets obtained was in the range of 90~95% of the theoretical value with an oxygen content of 1,300~2,100 ppm. No second phase, such as metallic uranium, were observed in the specimens, either by metallography or X-ray diffraction. These pellets of unexpectedly high density without second phase must have been obtained thanks to the good powder characteristics combined with proper sintering conditions. The compositions of uranium carbonitride pellets were found to be slightly nitrogen deficient, compared with the reactants.     

Chlorination-Volatilization Processing of Irradiated Uranium Dioxide, (I)

A chlorination-distillation process for irradiated U0₂ fuel treatment is presented, which utilizes the difference in vapor pressure between the chlorides of uranium and those of F.P. Good recovery of uranium and decontamination of F.P. were not obtained with single distillation. Improvement was seen with the adoption of a process in which most of the F.P. were removed from the irradiated U0₂ fuel through a sorption-desorption process using a BaCl₂ bed. The present report describes the behavior of uranium chlorides in this process.

The vapor of uranium chlorides formed by chlorinating U0₂ powder with a mixed gas of Ar and CCl₄ vapor was passed through the bed whose temperature varied along its length from 500° to 100°C, under which conditions the chloride vapor was completely sorbed on the bed. The uranium chlorides thus sorbed did not desorb in a mixed gas stream of Ar and CCl₄ vapor at bed temperatures below 480°C, but when the bed was heated to 480°~560°C, and Cl₂ gas added to the gas stream, the chlorides were completely desorbed in the form of vapor of higher uranium chlorides.     

Quenched-in defects in UC and UCN

The quenched-in resistivity of uranium monocarbide (UC) and uranium carbonitride (UCN) was studied and the activation energies of formation and migration for defects that are mobile in two recovery stages at about 200 and 700° C were obtained. These values were compared with the published diffusion enthalpies and with the migration energies obtained from recovery experiments on neutron irradiated UC. It is commonly suggested that the two recovery stages are due to isolated carbon and uranium vacancies. This is not compatible with recent diffusion data. It is argued here that dissociation of impurity-vacancy complexes is a more probable alternative for the stage at 700° C. Higher defect energies and recovery temperatures are obtained for UCN (and UN) as compared to UC, in agreement with other kinetic data in the series UC-UCN-UN. The migration energy of uranium vacancies was determined to be 2.4 ± 0.4 eV for UC and to increase for UCN.     

Kinetic Studies of Fluorination of Uranium Carbides by Fluorine, (II)

The fluorination reaction of uranium dicarbide with fluorine to form UF₆ has been studied in the temperature range between 220° and 300°C using a thermobalance. The overall mechanism of reaction is similar to the case of uranium monocarbide, i.e. in the first step, UC₂ is rapidly converted to UF₄, polymeric fluorocarbon and gaseous fluorocarbon, while in the second step, UF₄ and polymeric fluorocarbon are converted rather slowly to UF6 and gaseous fluorocarbon. The amount of polymer is much larger than the case of uranium monocarbide, its weight ratio to UF₄ being 0.2 to 0.25 in the early stages of the reaction. The fluorination of ₄ to UF₆ always follows the linear law derived from the diminishing sphere model. The apparent activation energy was determined to be 19.5 kcal/mole. Differences in the fluorination between UC and UC₂ are discussed, with the effect of polymer taken into consideration.     

Preparation of Uranium Phosphide from Uranium Tetrafluoride

A study was made on UP preparation by two-step reactions starting with UF₄, Si and red phosphorus.

The first step was to produce an intermediate uranium phosphide U₃P₄ from the reaction of UF₄ with Si under phosphorus vapor at temperatures above 900°C.

In the second step, the intermediate phosphide was heat treated under vacuum. To obtain single phase UP from the intermediate phosphide, the treatment required a temperature above 1,050°C, at which temperature the minimum holding time was 60 min.     

News item

Stoichiometric UC can be quantitatively reacted with the stoichiometric amount of ZnS to yield a mixture of US and C by heating at 800–1000°C in vacuum or inert atmosphere. During the reaction the Zn is liberated as vapour. Intimately blending the US+C mixture with the requisite amount of stoichiometric UO₂ and heating at 1700°C under vacuum results in a reaction between the C and UO₂, $\text{2}\text{3}$ of the C being evolved as CO, the US under these conditions being inert. The UC formed dissolves in the US to give a solid solution of composition 75 mol % US-25 mol % UC. If desired, UC₂ can be used in place of UC, the initial reaction temperature being somewhat lower. UO₂ can be replaced by another uranium oxide such as U₃O₈. Solid solutions of reduced US content can be prepared by decreasing the amount of ZnS and uranium oxide employed, but attempts to increase the US mol % to over 90 by repeating the process on a 75 mol % US-25 mol % UC solid solution have not been entirely successful. The products obtained were probably solid solutions of US with nearly 10 mol % UC₂ and a small amount of a UOSUO₂ phase. The results of chemical analysis and examinationsby X-ray diffraction and metallurgical techniques are reported.     

Fluorination of Uranium Metal by Fluorine Gas

A study has been made on the fluorination of uranium metal chips to UF₆ with fluorine gas in the temperature range of 100° to 400°C. The formation of UF₆ was influenced by the concentration and the flow rate of fluorine gas and by the reaction temperature. The reaction occurred apparently above 200°C. Uranium metal was first converted to low fluorine content compound such as UF₃ or UF_4-x, then to UF₄ and finally to UF₆. The intermediate compounds were confirmed by X-ray analysis and by their color.     

Diffusion Coefficient of Fission Gas in Uranium Dioxide Powder Formed by Carbon Dioxide Oxidation of Uranium

In connection with a program to study the behavior of punctured fuel elements for the Tokai Atomic Power Reactor, the diffusion coefficient of fission gas in uranium oxide powder formed by CO₂ oxidation of U was determined by post-irradiation experiment, in which the fractional release of fission gas during isothermal heating of the powder was measured. The U was oxidized at 600° and 700°C, and in both cases the O/U ratio of the oxides, measured gravimetrically, was 2.0. The diffussion coefficients in the oxide powder formed by oxidation at 600°C were found to be 1.4× 10-²⁰, 1.3×10-¹⁹, 1.1×10¹⁸ and 1.0×10-¹⁷, cm²sec-¹, respectively at 450°, 550°, 650° and 750°C, and in the oxide powder formed at 700°C, 7.4×10-¹⁹ and 3.6×10-¹⁶cm²sec-¹ at 600° and 700°C, respectively. Activation energies calculated for the two oxide powders were comparatively low.     

Uranium nitride fuels in superheated steam

Uranium mononitride (UN) pellets of different densities were subjected to a superheated steam/argon mixture at atmospheric pressure to evaluate their resistance to hydrolysis. Complete degradation of pure UN pellets was obtained within 1 h in 0.50 bar steam at 500 °C. The identified reaction products were uranium dioxide, ammonia, and hydrogen gas, with no detectable amounts of nitrogen oxides formed. However, the reaction could not be carried to completion, and the presence of uranium sesquinitride and higher uranium oxides or uranium oxynitrides in the solid residue is indicated. Evolution of elemental nitrogen was seen in connection with very high reaction rates. The porosity of the pellets was identified as the most important factor determining reaction rates at 400–425 °C, and it is suggested that in dense pellets, cracking due to internal volume increase initiates a transition from slow surface corrosion to pellet disintegration. The implications for the use of nitride fuels in light water reactors (LWR) are discussed, with some observations concerning hydrolysis as a method for ¹⁵N recovery from isotopically enriched spent nitride fuel.     

Kinetic Studies of Fluorination of Uranium Carbides by Fluorine, (I)

The rate of the fluorination reaction of UC with fluorine to form UF₆ has been studied in the temperature range between 240° and 300°C by following with a thermobalance the change in weight of the solid phase. The overall reaction takes place in two steps: In the first step, UC is rapidly converted to UF₄, which, in the second step, fluorinates to UF₆. During the first step, polymeric fluorocarbon is also produced and is found dispersed throughout the intermediate UF₄ particles. At reaction temperatures below 270°C, the second step reaction roughly follows the linear law. At temperatures above 270°C, it comes to follow the parabolic law, presumably due to the formation of a compact layer of polymeric fluorocarbon on the surface of the particles. Analysis of the gaseous products using mass and infrared spectrophotometers revealed the formation of several kinds of lower fluorocarbons.     

Oxidation Behavior of Uranium Monosulfide

The oxidation behavior of uranium monosulfide powder in oxygen and air has been studied over a range from room temperature to 900°C, by means of DTA, TGA, X-ray diffraction and chemical analysis. Most of the reactions occurring during the oxidation process have been identified. Intermediate oxidation products observed are γ-US₂, UOS, UO₂SO₄, UO₂ and UO_2+x.     

Contribution a l'etablissement du diagramme de phases du systeme U-C-O a l'aide de la diffractometrie de rayons X a haute temperature sous pression controlee

During the progressive reduction of uranium dioxide by graphite under known pressures of carbon monoxide, we have measured the lattice parameters of the compounds “UO₂” (fcc), α-“UC₂” (tetragonal) and “UC” (fcc), and have evaluated the compositions of these non-stoichiometric phases near 1770°C. By this means, and for the first time thanks to these direct measurements at such high temperatures, it has been possible to establish with precision the range of the main univariant phase fields [“UO₂”, C, α-“UC₂”, CO] and [“UO₂”, α-“UC₂”, “UC”, CO]. The part of the U-C-O diagram thus mapped out completes the earlier isotherm for 1700°C determined by Henry et al.     

Uranium monocarbide

Uranium monocarbide is of interest as a possible nuclear fuel and material for nuclear thermoelectrical transformations. In order to accurately define the effect of the conditions for preparing uranium monocarbide (UC) on its composition, studies were made which established the optimum regime for preparing UC with a stoichiometric composition.Studies were made of the conditions for sintering and hot pressing of UC powder and also conditions for sintering UC + U alloys, giving specimens with a porosity of about 5%. The specific gravity of UC powder determined by a pycnometer is 12.97 ± 0.09 g/cm³, microhardness of the phase-923 ± 56 kg/mm².The thermal conductivity of UC in the range 100–700 °C varies from 0.028 to 0.04 cal/cm. sec.deg; the mean thermal coefficient of linear expansion in the range 20–1500 ° is 11.6 · 10^–6. Specimens of UC subjected to cyclic heat-treatment in the range 200–1000 ° withstood 500 cycles without failure. Specimens of UC + U withstood more than 1000 cycles.     

Experimental kinetic study of oxidation of uranium monocarbide powders under controlled oxygen partial pressures below 230 °C

Uranium monocarbide (UC) powders are known to be easily oxidised by gas mixtures containing oxygen. In this study, the oxidation of UC micron powders was followed by isothermal thermogravimetry at temperatures ranging from 100 °C to 230 °C in two different gas mixtures: synthetic air and 97%N₂ + 3%O₂. X-ray diffraction tests conducted on powders after their oxidation showed that small crystallites of UO₂ oxide are formed. Furthermore, an analysis of mass gain showed that the carbon initially linked with the uranium is not oxidised but retained in oxide layers. Additionally, this kinetic study revealed that the rate-limiting step mechanism governing the oxidation of UC powders is a diffusive process that follows the Arrhenius law regarding temperature. Finally, it was discovered that cracks occur in grains once a given fractional conversion has been reached, inducing a major increase in the volume of grains.     

In-Line Uranium Monitoring in Reprocessing Waste Solution by Time-Resolved Laser-Induced Fluorometry

In-line monitoring by fluorometry of uranium concentration in reprocessing waste solution has been realized. The reduction of U0₂ ²⁺ fluorescence by coexisting ions and solution temperature can be corrected by measuring fluorescence lifetime, absorbance of excitation beam wavelength and absorbance of fluorescence wavelength. The method applicability was examined by using a sample solution simulating the waste solution of the codecontamination process. When the coexisting ion concentrations were increased in the sample solution which included 50 mg//of uranium, the corrected value of the uranium concentration was constant in the range of 0~1.5 times the coexisting ion concentrations. When the temperature was changed in the range of 30~45°C, the corrected value was also constant. The precision of the correcting method was ±15%'. These results verified that the correcting method could be applied to in-line monitoring of uranium concentration even though species and amounts of the coexisting ions and solution temperature were changed in the waste solution.