Properties of uranium vacancies in uranium monocarbide

From the excess resistivity obtained by quenching from temperatures between 1300 and 1600 °C a formation energy of 1.7 eV is determined. The quenchedin resistivity recovers with an activation energy of 2.2 eV in the temperature region between 400 and 600 °C. These two values are attributed to the formation and migration activation energies of uranium vacancies, respectively. It is further shown that uranium monocarbide remains single-phase, if the deviation from stoichiometry does not exceed more than a few tenths of one percent and if gases (mainly oxygen) are in solid solution.     

Neutron damage and fission product gas release in the initial burst region of uranium monocarbides

Thermal neutron damage and fission product gas (¹³³ Xe) release in a burst region of uranium monocarbides were studied. After neutron irradiation, the electrical resistivity was measured from room temperature to 800° C. Three recovery stages were revealed in the resistivity of UC irradiated to $\text{4.0 × 10}{}^{16}$ nvt. The lattice parameter of UC with the same irradiation also showed three stages of recovery up to 1050°C. The initial burst of Xe from UC was studied in a dose range between $\text{1.6 × 10}{}^{15}$ and $\text{2.9 × 10}{}^{18}$ nvt. The burst occurred in three steps for lightly irradiated specimens, while there were two steps of the burst in heavily irradiated specimens. The activation energies for each burst step were calculated. From the results obtained here, we concluded that the burst was correlated with the recovery of damage in the neutron-irradiated UC.     

Atomic diffusion mechanism of Xe in UO2

We have investigated vacancy-assisted diffusion of Xe in uranium dioxide (UO₂) calculating incorporation, binding, and migration energies. All the energy values have been obtained using the density functional theory (DFT) within the generalized gradient approximation (GGA) and the projector-augmented-wave (PAW) method. Considering spin-polarization effect, we find that the computed migration energy is reduced by and agrees well with experimental data compared to those obtained from non-magnetic calculations. We also find that an oxygen vacancy lowers the migration energy of a uranium vacancy by about 1 eV, enhancing an effective movement of vacancy clusters consisting of both uranium and oxygen vacancies. Furthermore, the strain energy of Xe is large enough to contribute to the clustering of vacancies making it the driving force for the vacancy-assisted diffusion of Xe in UO₂. In summary all the calculated results suggest that the trivacancy is a major diffusion pathway of Xe in UO₂.     

The effect of Fe,Ni and W impurities on uranium diffusion in uranium monocarbide

The self-diffusion of uranium with U-233 as tracer was measured in stoichiometric UC which was doped with Fe, Ni or W impurities. The impurities were added by flash evaporation on the UC and subsequent diffusion into the UC. Uranium diffusion was increased in the doped samples as compared to undoped UC at all temperatures studied (1380 to 2200°C). The increase was most pronounced (more than a factor of 100) at low temperatures. Simultaneously, pronounced grain-boundary penetration was observed at low temperatures, possibly due to eutectic formation at the grain boundaries. The ratio of grain boundary to lattice diffusion coefficients at 1380°C was of the order of 3 × 10⁴. The present data serve to explain some of the scatter of the literature data and correlate well with the known increased rate of sintering of Ni-doped UC.     

Molten Salt Extraction for Removing Iodine from Liquid Sodium

The initial burst and the diffusion of fission produced Xe were investigated on UC₂ irradiated to thermal neutron doses between 2.1 x l0¹⁴ and 1.3 x 10¹⁶ fission/cm⁸. In heating-runs with 5 and 10°C/min and below 1,000°C, the burst revealed three steps at around 200, 400 and 600°C (in 5°C/min heating run), to which activation energies were obtained. It was found that the temperatures and the activation energies for each burst were close to those appearing in recovery processes of fission induced defects. In diffusion process above 1,000°C, on the other hand, the activation energy of Xe in UC₂ increased with increasing fission dose. In the specimens once heated up to 1,500°C, however, the diffusion coefficient could be expressed as, D?10^-1 exp (— 344±50 kJ/RT) (cm²/s), which was independent of the level of the fission dose. An enhancement and a suppression of the diffusion were suggested also in the dicarbide as was done previously for the monocarbide of uranium.     

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%.     

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.     

Positron annihilation studies on alpha-irradiated and deformed niobium

Positron lifetime and annihilation lineshape measurements have been made to study the defect kinetics and He-vacancy interactions in alpha-irradiated Nb. A comparative study on plastically deformed Nb has also been performed. The isochronal annealing results are discussed with a brief review of previous investigations. Dislocation/vacancy loops annealing occurs above 700°C. Positron trapping rate at He-vacancy clusters is seen to increase with the addition of He atoms to the clusters. Results favour the idea of He-bubble growth by an addition of He atoms or vacancies at intermediate temperatures (350–750°C) and by bubble migration and coalescence at high temperatures (800–900 ° C). Annealing out of He-bubbles/He-vacancy complexes is seen above 900°C.     

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.     

Theory of the production and depth distribution of helium defect complexes by ion implantation

The complex defect formations and migrations occurring under helium ion bombardment of Cu have been modeled by a system of coupled equations, including diffusion. Atomistic binding and migration energies were determined from two-body calculations. A new mathematical scheme was developed in order to take diffusion into account in a self-consistent fashion. The calculations were applied to the low temperature implantation and annealing experiments of Bauer and also to the (proton backscattering) profiling experiments of Blewer. The calculations indicate that the ~0° C release peak of Bauer may be due to helium interstitial migration. When applied to the Blewer experiments, the calculations indicate that the dominant defect after room temperature implantation is six helium atoms in a vacancy (He₆V). The shape and position of the total helium distribution is not altered by isochronal annealing (because of the trapping of vacancies by the helium), but the helium is released directly from the damage which traps it.     

An atomistic approach to self-diffusion in uranium dioxide

The formation and mobility of point defects in UO₂ have been studied within the framework of the Density Functional Theory. The ab initio Projector Augmented Wave method is used to determine the formation and migration energies of defects. The results relative to intrinsic point defect formation energies using the Generalized Gradient Approximation (GGA) and GGA+U approximations for the exchange-correlation interactions are reported and compared to experimental data. The GGA and GGA+U approximations yield different formation energies for both Frenkel pairs and Schottky trios, showing that the 5f electron correlations have a strong influence on the defect formation energies. Using GGA, various migration mechanisms were investigated for oxygen and uranium defects. For oxygen defects, the calculations show that both a vacancy and an indirect interstitial mechanism have the lowest associated migration energies, 1.2 and 1.1 eV respectively. As regards uranium defects, a vacancy mechanism appears energetically more favourable with a migration energy of 4.4 eV, confirming that oxygen atoms are much more mobile in UO₂ than uranium atoms. Those results are discussed in the light of experimentally determined activation energies for diffusion.     

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.     

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.     

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.     

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.     

Positron annihilation and nano-indentation analysis of irradiation effects on the microstructure and hardening of A508-3 steels used in Chinese HTGR

The irradiation and annealing behavior of Chinese A508-3 reactor pressure vessel (RPV) steel (0.04 wt% Cu) after 3 MeV Fe-ion irradiation ranging from 0.1 to 20 dpa at room temperature (called RTRPV) and high temperature (250?°C, called HTRPV) was studied by positron annihilation Doppler broadening (PADB) spectroscopy and nano-indentation hardness. PADB showed that the density of vacancy-type defects was higher for low-temperature irradiations. The higher hardness was found after high-temperature irradiation because of the formation of solute clusters during irradiation. Positron annihilation measurements revealed the interaction and clustering of vacancies with solute clusters which were introduced by Fe-ion irradiation. For both RTRPVs and HTRPVs, the positron defect parameter and positron diffusion length showed the recovery of the irradiation-induced defects. Total recovery was observed after annealing at 450 °C.     

Kinetics and mechanisms of carbon self-diffusion in uranium carbides

Depending upon the temperature, uranium carbide, $\text{UC}{}_{\mathrm{x}}$, has the B1 (NaCl type) crystal structure and is characterized by a wide range of the carbon-to-uranium ratio ($\text{0.95 < x < 2.0}$). In this report, the kinetics of carbon diffusion in uranium carbides are examined and mechanisms are proposed which explain the diffusivities determined by numerous investigators for compositions in the range $\text{0.95 < x < 2.0}$. When the composition is hypostoichiometric, the defect structure in the carbon sub-lattice consists of constitutional vacancies; the concentration of these vacancies is fixed by the carbon-to-uranium ratio and equals ($\text{1?x}$). Carbon diffusion occurs by the random migration of carbon atoms from one octahedral site to an adjacent vacant octahedral site. The defect structure in the carbon sub-lattice of stoichiometric and hyperstoichiometric uranium carbides consists of C₂ groups and single carbon atoms in the octahedral sites; the concentration of unoccupied octahedral sites (vacancies) is negligibly small. Diffusion occurs by the random migration of single carbon atoms from doubly occupied sites to adjacent singly occupied octahedral sites. Quantitative expressions of these mechanisms are developed which accurately describe the diffusion kinetics. The preponderant discordance in the experimental results of other investigators are reviewed and examined; it is concluded that the wide variability in these results was probably caused by microstructural effects,     

A Model of Turbulent Vortex Generation in Boundary Layer

Reactivity and compatibility studies were carried out on some vanadium base alloys with ceramic fuel materials. The vanadium base alloys studied were V-20%Ti, V-15%Ti-7.5%Cr and V-5%Cr. The fuel materials with which these alloys were put in contact were UC, UN, (U, Ce)N and (U,Ce)(C,N). Experiments were carried out at 700° and 900°C for both 500 and 1,000 hr. Within the conditions studied, V-5%Cr showed good compatibility with all of these fuel materials. At 700°C, no detectable interaction was observed between any of the alloy-fuel combinations.     

Production of 40 TBq Tritium Using Neutron-Irradiated 6Li-Al Alloy

A facility was reconstructed for producing tritium in 40 TBq per batch. Gaseous tritium was extracted from neutron-irradiated ⁶Li-Al alloy targets by heating at 700°C under vacuum and collected in uranium. The recovery yield of tritium was about 100% and the isotopic purity of the product was about 95%. Through the production run, no leakage of tritium from the facility was observed.     

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.