Fission Gas Release from Nuclear Fuels, (II)

The release behavior of fission gases in U-metal, UO₂ and uranium carbides, irradiated at a relatively low temperature (below 100°C) to low dosage, was studied by out-of-pile experiments.

It was found that fission gas (¹³³Xe) released from a specimen by fission fragment recoil is mostly captured in the wall of the irradiating capsule or in the capsule support material.

The amount of fission gas released into the void space of the capsule is proportional to the surface area and to the fuel burn-up, and is controlled by a knock-out release mechanism. The number of U atoms considered to take part in the knock-out mechanism by evaporation or displacement due to the intrusion of a recoil fission fragment, is estimated to be 1.4×10⁵~2.7×10⁵ atoms for U-metal and 5×10⁴~10×10⁴ atoms for UO₂ and uranium carbides.     

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.     

Growth of the interaction layer around fuel particles in dispersion fuel

Corrosion of uranium particles in dispersion fuel by the aluminum matrix produces interaction layers (an intermetallic-compound corrosion product) around the shrinking fuel spheres. The rate of this process was modeled as series resistances due to Al diffusion through the interaction layer and reaction of aluminum with uranium in the fuel particle to produce UAl_x. The overall kinetics are governed by the relative rates of these two steps, the slowest of which is reaction at the interface between Al in the interaction layer and U in the fuel particle. The substantial volume change as uranium is transferred from the fuel to the interaction layer was accounted for. The model was compared to literature data on in-reactor growth of the interaction layer and the Al/U gradient in this layer, the latter measured in ex-reactor experiments. The rate constant of the Al-U interface reaction and the diffusivity of Al in the interaction layer were obtained from this fitting procedure. The second feature of the corrosion process is the transfer of fission products from the fuel particle to the interaction layer due to the reaction. It is commonly assumed that the observed swelling of irradiated fuel elements of this type is due to release of fission gas in the interaction layer to form large bubbles. This hypothesis was tested by using the model to compute the quantity of fission gas available from this source and comparing the pressure of the resulting gas with the observed swelling of fuel plates. It was determined that the gas pressure so generated is too small to account for the observed delamination of the fuel.     

Production and investigation of thorium-containing fuel compositions

A technology has been developed for obtaining fuel tablets with the compositions (U, Th)O₂, (U, Th, Ca)O₂, and (U, Th)O₂+MgO by combined precipitation of uranium, thorium, magnesium, or calcium components from inert solutions, followed by heat treatment of the powders, compression into pellets, and sintering of the pellets. Work on optimizing the technological processes for obtaining fuel pellets so as to obtain good pellet quality was performed. The effect of the properties of the precipitates and powders, fabricated using different technological regimes on the properties of the finished objects was studied. The work includes detailed investigations of powders (x-ray phase analysis, electron-microscopic investigation) and sintered fuel tablets (change in the geometric dimensions as a result of sintering, determination of the density, and study of the microstructure). The behavior of fuel compositions (U, Th)O₂ and (U, Th)O₂+MgO in contact, with coolants under conditions where the fuel elements become unsealed was studied: with water at 300°C and sodium at 700°C. 3 figures, 3 tables, 6 references. State Science Center of the Russian Federation-A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 88, No. 5, pp. 346–353, May, 2000.     

Dancoff Factor in Arrayed-Type Cladded Fuel Lattice

The initial-stage sintering mechanism of hyperstoichiometric urania prepared by sol-gel process was determined in relation to temperature during constant rate heating (CRH). The urania powder used in this experiment was prepared by crushing in Ar atmosphere the micro- spheric gel of UO₂ obtained by sol-gel process, and reducing the resulting powder by heating in H₂ for 1 hr at 500°C. The results obtained from densification measurements indicated that the initial-stage sintering proceeded in two phases governed by different shrinkage mechanisms, as follows.

1. The sintering up to 675°C would be due to a mechanism such as rearrangement of grains and/or plastic flow.

2. Sintering from 750° to 800°C was interpreted as being controlled by uranium volume diffusion.

The estimated diffusion coefficient D = 1.42×10^?6 exp(-52,500/RT) cm²/sec. This value agreed in order of magnitude with the uranium diffusion coefficients measured by other workers for hyperstoichiometric urania.     

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.     

Evaluation of UF6 Vapor Release in Postulated Accident

The rupture of UF₆ gas line connected to hot UF₆ cylinder, being one of various accidents in UF₆ vapor leak-out, is considered as a postulated accident for uranium enrichment plants. For this type of rupture, we will estimate the amount of UF₆ vapor release based on a simplified calculation model and then make an evaluation of UF₆ vapor release through a ventilation system of feed vaporization facility. Assuming an instantaneous steady state for the change of UF₆ states, an unsteady state thermodynamics process is solved. Numerical examples show that about 52% of the initial UF₆ quantity are vaporized at 80°C (the temperature of the liquid UF₆ in the cylinder). Furthermore, by using the amount of released UF₆ vapor and the collection capacity of HEPA filter for IiF gas, the amount of gaseous UO₂F₂, HF which may be dissipated to the environment are conservatively estimated.     

Preparation of Uranium-233

Thorium dioxide amounting to 1.3 kg was irradiated for 241 days at an average neutron flux of 6×10¹⁰n/cm²'sec, and allowed to stand for 1.2yr. This cooling lowered the γ-activity of the irradiated ThO₂ sufficiently to permit handling without special shielding. Nine milligrams of ²³³U were extracted with TBP in dodecane, and determined by α-ray counting. The mass ratio of the uranium isotopes obtained was determined by mass spectrometry. It was indicated that the ²³³U was contaminated with natural uranium.     

Fission Gas Release from Uranium Heated in Carbon Dioxide

An experiment to predict the amount of fission gas release from U heated in CO₂ was made in connection with the safely evaluation of the Tokai Atomic Power Reactor. Fission gas release from slightly irradiated U pieces was measured as a function of percent oxidation, and it was found that the percent release was proportional to the percent oxidation. The fission gas released before the U was completely oxidized was 11%, 23% and 48% at 500°, 600° and 700°C, respectively. The form of temperature dependence of the fission gas release was of the Arrhenius type, with an activation energy of 5.1kcal/mol.     

Study of the properties of ammonium polyuranate precipitates and UO<Subscript>2</Subscript> powders and pellets obtained with different technologies

The results of a differential-thermal analysis are used to compare the properties of ammonia polyuranate precipitates, UO₂ powders and pellets, obtained by different methods as well as metallic uranium. It is found that the phase NH₃·3UO₃·5H₂O forms in regular precipitation of ammonium polyuranate. When using nanotechnology, the phases NH₃·2UO₃·3H₂O and 4NH₃·6UO₃·8H₂O are also present in the precipitate. UO₂ powder prepared from such precipitate has high activity, since all phase transformations in it occur at a lower temperature. Modified fuel pellets of uranium dioxide, which are obtained by means of nanotechnology or mechanical addition of ammonia-containing reagents to powder, differ from the standard powders by a lower rate and more complex mechanism of oxidation, similar to metallic uranium. Modified UO₂ fuel pellets fabricated at the Physics and Power-Engineering Institute, are now undergoing tests in the BOR-60 reactor. After tests on the irradiated new modified fuel have been completed, it will be possible to judge its reliability.     

The migration of intragranular fission gas bubbles in irradiated uranium dioxide

The mobility of intragranular fission gas bubbles in uranium dioxide, irradiated at 1600–1800°C, has been studied following isothermal annealing at temperatures below 1600°C. The intragranular fission gas bubbles, average diameter approximately 2 nm, are virtually immobile at temperatures below 1500°C. The bubbles have clean surfaces with no solid fission product contamination and are faceted to the highest observed irradiation temperature of 1800°C. This bubble faceting is believed to be a major cause of bubble immobility. In fuel operating below 1500°C the predominant mechanism allowing the growth of intergranular bubbles and the subsequent gas release must be the diffusion of dissolved gas atoms rather than the movement of entire intragranular bubbles.     

MOX Co-deposition Tests at RIAR for SF Reprocessing Optimization

The Oxide Electrowinning method has been studied as one of the candidate dry reprocessing concepts of the future fuel cycle systems. On the MOX co-deposition process, the main process of that method, some fundamental experiments have been performed to confirm its feasibility. In the experiments, several parameters were set to study the suitable electrolysis condition to obtain desired granule of MOX. The concentrations of uranium, plutonium, fission products(FP) simulators, and corrosion products(CP) simulators were adopted as the parameters. The blowing gas composition (O₂, Cl₂, Ar) during the electrolysis was also set as the variable condition. Through these experiments, it was clarified that the partial pressure of chlorine gas during electrolysis was important to obtain MOX granule with high Pu concentration (about 30%) without generating bottom precipitation in melt. Finally, adequacy of the process control method for MOX co-electrolysis was confirmed through the test using spent fast reactor(FR) fuel.     

Investigations on swelling and fission gas behaviour in uranium dioxide

UO₂ irradiated at temperatures between 1000 and 2100 K was investigated with respect to fission gas behaviour and swelling. The amount of fission gas was measured in three steps as released fission gas, fission gas retained in bubbles and pores, and fission gas in the fuel matrix. The retained fission gas reaches concentrations up to $\text{1.6 × 10}{}^{\mathrm{?2}}$ gas atoms per uranium atom at temperatures below 1250 K and decreases with increasing temperature. The swelling was evaluated by measuring the volume changes and by immersion density measurements. The maximum fission gas swelling without extensive bubble migration is about 20% at 2000 K. It diminishes to about 5% at 1250 K.     

Irradiation behavior of experimental fuel particles containing chemically vapor deposited zirconium carbide coatings

Four experimental fuel particle designs, utilizing zirconium carbide coatings in combination with porous and dense pyrocarbon coatings, were tested under high-temperature irradiation. As a fission-product corrosion test for the zirconium carbide, two particle designs employed carbide coatings applied directly over either UC₂ or (8Th, 1U)O₂ fuel kernels. The other two designs utilized zirconium carbide outside of porous pyrocarbon coatings but without the conventional inner dense pyrocarbon coating on either UC₂ or (8Th, 1U)O₂ fuel kernels. The particles were irradiated at 1200°C to a fast-neutron fluence of $\text{5 × 10}{}^{21}\text{n/cm}{}^2\text{(E μ'0.18 MeV)}$ and fractional burnups of the initial metal atoms of 0.7 and 0.08 for the UC₂ and (8Th, 1U)O₂ kernels, respectively. Stereoscopic, metallographic and electron-beam microprobe examination of the irradiated particles showed that the zirconium carbide possesses exceptional resistance to chemical attack by fission products and good mechanical stability under irradiation.     

Some observations of density and porosity changes in UO2 fuel irradiated in water-cooled reactors

A technique has been developed for the hot-cell measurement of the apparent density of irradiated UO₂ fuel after extraction from a fuel pin. A single determination is accurate to $\text{± 3 \%}$ at the 95 % confidence limit. The method has been applied to fuel irradiated in thermal neutron fluxes in the Winfrith SGHWR and in the Halden BWR. Material has been examined at ratings of 1–51 W/g and in the burn-up range $\text{0.09–5.79 × 10}{}^{20}\text{fissions/cm}$. It is concluded that pellets with peak temperatures below 1100°C densify during irradiation, but at higher temperatures the pellets begin to swell. Fuel micrography has shown that the densification is principally due to the loss of micropores with a temperature dependency given by an activation energy of 5200 cal/mol. Above 1000°C the densification is masked by the formation and growth of intergranular fission gas bubbles, whose volume may exceed that of the manufactured pores which have sintered. In solid fuel pellets central swelling did not balance densification in the cooler rim until the fuel centre temperature exceeded 1700°C.     

Interaction between uranium sulphides and oxygen

Study of the oxidation of uranium monosulphide shows that certain phenomena occur at three temperatures: 350–380, 480, and 720°C. In the temperature range 350–380°C, an intensive incorporation of oxygen begins, accompanied by loss of SO₂ and S. Simultaneously UO_2+0.45 and UO₂SO₄ are formed. As the temperature increases, the amount of sulphur remains constant and only oxygen is incorporated. At 540°C the X-ray pattern of the product corresponded to that of U₃O₈, but the composition was UO_3.50 + UO₂SO₄. At higher temperatures the remaining sulphur was burnt and U₃O₈ was obtained. The reaction between uranium disulphide and oxygen proceeds in a similar way, except that at 345°C preferential oxidation of sulphur occurs. Investigation of the isothermal oxidation of US and US₂ at temperatures 250–305° C and under an oxygen pressure of 400 Torr showed that the rate law was initially exponential (lateral growth of oxide film), and that it later became parabolic (diffusion of oxygen through the oxide layer).     

A microstructure-dependent model for fission product gas release and swelling in UO2 fuel

A model for the release of fission gas from irradiated UO₂ fuel is presented. It incorporates the relevant physical processes: fission gas diffusion, bubble and grain boundary movement, intergranular bubble formation and interlinkage. In addition, the model allows estimates of the extent of structural change and fuel swelling. In the latter, contributions of thermal expansion, densification, solid fission products, and gas bubbles are considered. When included in the ELESIM fuel performance code, the model yields predictions which are in good agreement with data from UO₂ fuel elements irradiated over a range of water-cooled reactor conditions: linear power outputs between 40 and 120 kW m⁻¹, burnups between 10 and 300 MW h(kg U)⁻¹, and power histories including constant, high-to-low and low-to-high power periods.The predictions of the model are shown to be most sensitive to fuel power (temperature), the choice of diffusion coefficient for fission gas in UO₂, and burnup. The predictions are less sensitive to variables such as fuel restraint, initial grain size and the rate of grain growth.     

Controlled oxidation experiments of simulated irradiated UO2 fuel in relation to thermochemical modelling

A thermochemical model to describe the chemical state of irradiated nuclear fuel has been advanced and validated by comparison to results of experiments on naturally-enriched UO₂ with additions of selected simulated fission products. These experiments involved controlled oxidation in Ar/H₂O/H₂ gas mixtures conducted at the Atomic Energy of Canada Limited-Chalk River Laboratories. A coulombic titration technique provided measurements of moles of oxygen acquired by the samples in relation to oxygen potential. Emphasis was placed on the role of molybdenum in buffering the oxidation of fuel.This treatment is expected to be especially useful when integrated into fuel performance codes that make use of thermodynamics as boundary conditions in heat and mass transfer computations.     

Preliminary Study on Production of Xenon-133 from Neutron-Irradiated Uranium Metal and Oxides by Oxidation

A method was developed for recovering the fission product ¹³³Xe from several kinds of reactor-irradiated U targets, including Al-U alloy, metallic U, and uranium oxides.

In order to observe the release of ¹³³Xe from U compounds at high temperatures, irradiated targets were heated at 500°~1,000°C in vacuum or under gas flow. The liberated ¹³³Xe was trapped on charcoal beds, and the release rate of ¹³³Xe from the compounds was determined by measuring the activity accumulating in the traps.

More than 90% of the ¹³³Xe was liberated from the alloy upon melting and from metal and uranium oxide upon oxidation.

The isolated Xe was purified by a system embodying cold traps and cryogenic distillation.

The final products were sealed in ampoules. They proved to possess radiochemical purity exceeding 99.9%, and less than 1 μ/ampoule (1 ml) of non-radioactive gaseous contaminants.     

Preparation of Uranium Carbide from Uranium Chloride

A method has been developed for the preparation of uranium monocarbide using UCI₄ as starting material. Among several thermodynamically possible reactions, the reaction of UCL, Al and graphite powder mixture under Ar atmosphere was selected for reasons of availability of material and facility of post-reaction processing. It was confirmed by chemical and X-ray analyses that UC had been obtained at temperatures between 600° and 1,200°C. The purest UC was obtained at 900°C, and the conversion efficiency of uranium reached a maximum of about 80%.