Hydration thermodynamics of plutonium and transplutonium ions

Since thermodynamic properties are related to crystallographic radii, we first present sets of precise values of crystallographic radii (R) for ionic charge +2, +3 and +4 and different coordination numbers (CN). These data are based on the values reported by Shannon as well as variations of R with CN as observed for lanthanides, and linear correlations between isoelectronic $\text{M}{}^{\mathrm{m+}}$ and $\text{N}{}^{\mathrm{n+}}$ ions. Accurate measurements of the volume of the unit cell of oxides, fluorides and chlorides carried out recently give reliable radii values.Experimental entropies of actinide aquo ions are limited to Pu³⁺ and Th⁴⁺. Since this property depends on the structure of the hydrated ion, we decompose ionic entropies into three terms, related to electronic configuration, mass of the central ion, and structure of the aquo ion (S_h). We use our data on the structure of trivalent actinide ions to derive S_h and therefore to estimate values of trivalent ionic entropies. Entropies of divalent and tetravalent ions are also obtained similarly.Finally, taking into account the structure of the aqueous actinide ions, we calculate hydration enthalpies of +2, +3 and +4 actinide ions.     

The kinetics of actinide redistribution by vapor migration in mixed oxide fuels (II). By pores

The migration velocity of a closed lenticular pore in hyperstoichiometric mixed oxide fuel has been calculated, and the U/Pu ratio of the solid near the moving pore has been determined. Pore migration in mixed oxides differs from the analogous process in pure stoichiometric urania in that the composition as well as the temperature of the two faces of the pore are different. It was found that pore migration is not as effective a mechanism of actinide redistribution as vapor transport along cracks. The velocity of the pores in (U_0,8Pu_0,2) O_{2 + x} is $\text{～3}\text{1}\text{2}$ times faster than that in pure UO₂.     

On the interpretation of vapor pressure measurements on oxide fuel at very-high temperatures for fast reactor safety analysis

Safety analysis of fast reactors requires knowledge of the evaporation behavior and the total vapor pressure of oxide fuel materials in the temperature region from 3000 K upwards. Dynamic vapor pressure measurements on liquid oxide fuels performed in open-evaporation experiments with laser heating techniques imply strong alterations in the composition of the incongruently evaporating fuel surface, since, during evaporation, the depletion in the preferentially evaporating components cannot be resupplied by diffusion from the bulk material. After a short transient evaporation period stationary surface-evaporation is reached with a surface composition which differs greatly from the given fuel composition and depends on the actual evaporation temperature. When this stationary forced-congruent evaporation mode is reached, the gross vapor composition is well-defined and is identical to the bulk composition of the fuel but is quite different from the actual surface composition. In consequence, the total vapor pressure developing in open surface-evaporation of a liquid oxide fuel can substantially deviate from its thermodynamic equation-of-state, in the case of (U_0.80Pu_0.20) mixed oxide by a factor of 2 to 7 depending on the $\text{O}\text{M}$-ratio. Following these thermodynamic calculations direct measurement of the equation-of-state in open-evaporation experiments is practically impossible. Theoretically fitted expressions applicable in reactor safety analysis are presented for the equations-of-state and the vapor pressure equations for open surface-evaporation and also for the heats of evaporation of liquid (U_0.80Pu_0.20)O_1.95…2.00 mixed oxides.     

Oxygen potential and the chemical state of the fission products ruthenium,rhodium and palladium in irradiated oxide fuels

Thermodynamic properties of intermetallic compounds of the type M₃U and M₃Pu where $\text{M = Ru}$, Rh or Pd, are reviewed. The critical oxygen potential of the oxide fuel matrix necessary for their formation is suggested.     

Chemical thermodynamics of Cs and Te fission product interactions in irradiated LMFBR mixed-oxide fuel pins

A combination of fuel chemistry modelling and equilibrium thermodynamic calculations has been used to predict the atom ratios of Cs and Te fission products (Cs:Te) that find their way into the fuel-cladding interface region of irradiated stainless steel-clad mixed-oxide fast breeder reactor fuel pins. It has been $\text{concluded}$ that the ratio of condensed, chemically-associated Cs and Te in the interface region,?s:Te, which in turn determines the Te activity, is $\text{controlled}$ by an equilibrium reaction between Cs₂Te and the oxide fuel, and that the value of ?s:Te is, depending on fuel 0:M, either equal to or slightly less than 2:1. Since Cs and Te fission products are both implicated as causative agents in FCCI (fission product-assisted inner surface attack of stainless steel cladding) and in FPLME (fission product-assisted liquid metal embrittlement of AISI-Type 316), the observed out-of-pile Cs:Te thresholds for FCCI (4̃:1) and FPLME (2̃:1) have been rationalized in terms of Cs:Te thermochemistry and phase equilibria. Also described in the paper is an updated chemical evolution model for reactive/volatile fission product behavior in irradiated oxide pins.     

Composition and structure of fission product precipitates in irradiated oxide fuels: Correlation with phase studies in the Mo-Ru-Rh-Pd and BaO-UO2-ZrO2-MoO2 Systems

Composition and crystal structure of fission product precipitates in irradiated oxide fuels were studied by X-ray microanalysis and X-ray diffraction using instruments shielded for α-contamination and β-γ-radiation. Pin cross-sections, fuel micro-samples from 300 μm hollow drillings and residues from the dissolution of irradiated material in HNO₃ were investigated. The metallic phases found are hcp ?-Ru(Mo,Tc,Rh,Pd) solid solutions with broad variations in concentration of the components, bcc β-Mo(Tc,Ru) and fcc α-Pd(Ru,Rh). The dominating ceramic precipitate is composed of ($\text{Ba}{}_{\mathrm{1-x-y}}\text{Sr}{}_{\mathrm{x}}\text{Cs}{}_{\mathrm{y}}$)(U,Pu,RE,Zr,Mo)O₃ which crystallizes in the cubic perovskite type. The Mo fraction of these phases is related to the local oxygen potential of the fuels. The in-pile observed results agree well with phase studies in the quaternary Mo-Ru-Rh-Pd system where complete solid solubility exists between hcp Ru and the hexagonally stabilized Mo-Rh and Mo-Pd phases. Agreement is further attained with phase studies in the pseudoquaternary BaO-UO₂-ZrO₂-MoO₂ system which is characterized by a cubic perovskite phase Ba(U,Zr,Mo)O₃.     

In-pile creep study on mixed oxide (U,Pu)O2

The in-pile creep of a mixed oxide UO₂-PuO₂ under compression was studied up to fission rate of $\text{6 × 10}{}^{13}\text{f cm}{}^{\mathrm{?3}}\text{s}{}^{\mathrm{?1}}$, for stresses up to 26.5 MN m^?2, at temperatures ranging from 700 to 900°C. The results obtained agree with those of other authors. The creep rate is proportional to the applied stress and to the fission yield. However, it is athermal within the temperature range explored and is not affected by the burn-up, which has so far reached 30000 MWd t^?1 (3.6% FIMA). When the sample is under compression the fuel swells under the action of the fission products formed in the oxide during its irradiation. The swelling rate is about that commonly accepted for a clad fuel element. Finally it seems that the oxide swells more when free from stress than when subjected to a stress field, but this point has to be confirmed.     

Irradiation induced creep of ceramic nuclear fuels

This contribution gives a review of the experimental results and accompanying theoretical considerations. The mechanisms considered for irradiation creep are: relaxation of elastic stresses by fission spikes, promotion of dislocation slide by thermal spikes, preferential, stress-orientated nucleation of dislocation loops and preferential growth of dislocation loops. A survey over the irradiation creep rates attributed to steady-state creep shows $\text{ε}{}_{\mathrm{irr}}\text{~ σ · F}$ for oxide fuel in the stress and fission rate ranges of $\text{σ = 10–50}\text{MN/m}{}^2$ and $\text{F = 3 × 10}{}^{12}\text{–1 × 10}{}^{14}\text{f/cm}{}^3\text{· s}$ at burnups < 3%. There seems to be a continuous increase of the irradiation creep rate of oxide fuels with increasing temperature. However, that increase cannot be directly interpreted through a thermally activated process. It seems that the irradiation creep rate will also depend on fuel porosity, on plutonium distribution in mechanically blended UO₂-PuO₂, but not substantially on the plutonium content per se. Some results were already given for carbide and nitride fuels, which show the irradiation creep rate to be lower by about a factor of 10 than for oxide fuel under comparable conditions. Primary irradiation creep has been observed up to $\text{(3–5) × 10}{}^{19}\text{f/cm}{}^3$ and could prevail up to $\text{1 × 10}{}^{20}\text{f/cm}{}^3$.     

Equi-chemical-potential curves in ternary uranium oxides MyU1−yO2+x

Using Schuhmann's method the chemical potentials of three components in nonstoichiometric ternary uranium oxides $\text{M}{}_{\mathrm{y}}\text{U}{}_{\mathrm{1?y}}\text{O}{}_{\mathrm{2+x}}$(M = Mg and Pu) have been calculated. The obtained results are shown in the form of equi-chemical-potential curves (surface of partial molar free energy) in the composition triangle. In the hypostoichiometric region the contour lines of the surface are dense and run parallel with each other. Taking account of this fact the difficulty in attainment of homogeneity in the fabrication of mixed oxide fuel is discussed.     

Behavior of Fission Products Zirconium and Barium in Fast Reactor Fuel Irradiated to High Burnup

Barium and Zr generated in nuclear fuels can precipitate as multi-component oxide with some other fission products. In addition, the solubility of Ba in the fuel depends on the oxygen potential and the temperature and Zr can easily dissolve into the fuel matrix. Therefore, the behavior of the Ba-Zr oxide inclusions during irradiation is rather complex. In this work, the composition of multi-component oxides and the distributions of Ba and Zr as a function of relative radius were evaluated with X-ray microanalysis. As results, the oxide inclusions containing both Ba and Zr and containing only Ba were observed in the fuel irradiated to the burnup of 13.3 and 10.6 at%, respectively. These results were discussed in terms of the solubility of Ba and Zr in the fuel and in terms of the rO₂–UO₂ phase diagram, together with the radial distributions of Ba and Zr in fuel matrix.     

Equation of state of uranium oxide

The total and partial pressures over liquid UO₂ have been measured and calculated up to 5000K. A review of previous work is given. The equation of state of UO₂ as the main constituent of the fast breeder oxide fuel is required up to at least 5000K in order to estimate the energy release in a loss of flow (LOF) driven hypothetical core disruptive accident (HCDA) of the liquid metal fast breeder reactor (LMFBR).Two models, a macroscopic “mixture” model and a microscopic “defect” model have been developed to determine the oxygen potential of UO₂₀₀ up to 5000 K.A combination of mass spectrometric, Langmuir probe and high tension diode studies, applied for the first time to the laser vaporization process, revealed large quantities of ions emitted directly from the surface, and resolved previous discrepancies between measured and calculated vapour pressures by an enhanced rate of evaporation due to ion emission. As shown theoretically intrinsic ion emission can contribute to the net evaporation rate only if the resulting positive space charge can be neutralised. It is proposed that this can be accomplished by the presence of “hot” electrons in the plasma.The recommended equilibrium total pressure over liquid UO_2.00, valid between the melting point and 5000K, is $\text{log p (MPa) = ? 2.717 ? 20131/T + 1.925}\text{log T}$.     

Electrical conductivity measurement and thermogravimetric study of pure and niobium-doped uranium dioxide

The electrical conductivity and nonstoichiometric composition of UO_2+x and (U_1?yNb_y)O_2+x (y = 0.01, 0.05 and 0.10) were measured in the range $\text{1282 ≦ T ≦ 1373}$ K and Pa by tie four inserted wires method and thermogravimetry, respectively. The electrical conductivity of (U_1?yNb_y)O_2+x plotted against the oxygen partial pressure indicated a minimum corresponding to the transition between $\text{n}$- and $\text{p-}\text{type}$ cone uction. The band-gap energy of (U_1?yNb_y)O_2+x was calculated to be $\text{(248 ± 12)}$ kJmol.^?1, independent of niobium content, which is nearly the same as that of UO_2+x. From the oxygen partial pressure dependences of both the electrical conductivity and the deviation $\text{x}$ of UO_2+x and (U_1?yNb_y)O_2+x, the defect structures in these oxides were discussed with the complex defect model consisting of oxygen vacancies and two kinds of interstitial oxygens.     

Treatment of a waste salt delivered from an electrorefining process by an oxidative precipitation of the rare earth elements

For the reuse of a waste salt from an electrorefining process of a spent oxide fuel, a separation of rare earth elements by an oxidative precipitation in a LiCl-KCl molten salt was tested without using precipitate agents. From the results obtained from the thermochemical calculations by HSC Chemistry software, the most stable rare earth compounds in the oxygen-used rare earth chlorides system were oxychlorides (EuOCl, NdOCl, PrOCl) and oxides (CeO₂, PrO₂), which coincide well with results of the Gibbs free energy of the reaction. In this study, similar to the thermochemical results, regardless of the sparging time and molten salt temperature, oxychlorides and oxides were formed as a precipitant by a reaction with oxygen. The structure of the rare earth precipitates was divided into two shapes: small cubic (oxide) and large plate-like (tetragonal) structures. The conversion efficiencies of the rare earth elements to their molten salt-insoluble precipitates were increased with the sparging time and temperature, and Ce showed the best reactivity. In the conditions of 650 °C of the molten salt temperature and 420 min of the sparging time, the final conversion efficiencies were over 99.9% for all the investigated rare earth chlorides.     

Cesium-uranium-oxygen chemistry in uranium-plutonium oxide fast reactor fuel pins

The Cs-U-O system has been reinvestigated in light of recently reported thermodynamic data for Cs₂U₄O₁₂ and recent phase data showing the existence of a new Cs-U-O compound with an $\text{O}\text{(U + Cs)}$ atom ratio less than that of Cs₂UO₄. Our experiments have confirmed the existence of a new phase, allowed the formula Cs₂UO_3.56 to be assigned, and generated thermodynamic data for this new compound. This new phase exists only at oxygen potentials that are too negative to be encountered in uranium-plutonium oxide fast reactor fuel pins. The compound Cs₂UO₄ appears to be the most likely one to be formed, with the formation occurring at the fuel-blanket interface.     

A determination of the solubility of lithium oxide in liquid lithium by fast neutron activation

The solubility of lithium oxide in liquid lithium has been measured in the temperature range of 195 to 734°C. A direct sampling method was employed in which nickel sampling tubes fitted with 2 μm pore size filters were used to take samples of the oxide-saturated lithium. The samples, still contained in their sampling tubes, were analyzed for oxygen content by a fast neutron activation method. A value for the oxygen blank attributable to the sampling tubes was derived from a regression analysis of the individual solubility data points. The resulting solubility values are considerably lower than those reported by previous investigators and can be represented by the equations In S(mol% Li₂O) = 6.054–6669/T, $\text{log S}{}^1\text{(wppm oxygen) = 6.992–2896/T}$. The results indicate that removal of oxygen from lithium to a level of about 7 wppm (the solubility at 200°C) by cold trapping or by filtration should be feasible.     

Thermodynamic behavior of (U,Pu) mixed-oxide fuels

Transpiration experiments were performed at temperatures between 1000 and 1700°C to study the thermodynamics and defect structure of hypostoichiometric UO₂-20 wt % PuO₂ solid-solution systems. The oxygen partial pressures were established by using flowing $\text{H}{}_2\text{/H}{}_2\text{O}$ mixtures. After equilibration, the quenched products were analyzed by chemical, X-ray, neutron-diffraction and metallographic techniques. The versus temperature curves for the mixed oxide with different oxygen-to-metal ratios were plotted.Based on X-ray, neutron diffraction and metallographic data, it was concluded that the 20 wt % PuO₂ mixed oxide exists as a single phase under normal conditions, even at an oxygen-to-metal ratio as low as 1.92.The data from density measurements and neutron-diffraction analysis indicated that the predominant defects in the hypostoichiometric UO₂-20 wt % PuO₂ are anion vacancies.     

The solubility of hydrogen,deuterium and tritium in liquid lead-lithium alloys

The solubility of H₂, D₂ and T₂ in liquid Pb-Li alloys was calculated, as a function of composition and temperature, using the thermodynamic data of the binary systems. In the 17 at% Li alloy, at 600 K the D₂ solubility is log K = ?4.20 ± 0.076, K in appm. torr^{?$\text{1}\text{2}$}, that of H₂ is ?4.125 ± 0.076, and that of T₂ is ?4.244 ± 0.076. The calculated values give a much lower solubility than the experimental data. A ternary H-Li-Pb interaction term must be postulated to get agreement between experimental and calculated results.     

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.     

The mechanisms of fission gas release from (Th,U)O2

The releases of xenon from three (Th, U)O₂ specimens with different U contents were measured over a wide range of fission dose from $\text{2.9 × 10}{}^{19}$ to $\text{2.2 × 10}{}^{22}$ fissions m^?3 by using a post-irradiation technique. The releases were found to decrease with dose and to level off at higher doses. Measurements of the changes in lattice parameter and specific surface area of the same specimens enabled one to conclude that the decrease in release originates in the trapping of xenon by the vacancies and vacancy clusters induced by fission fragments. And the release mechanisms of fission gas were proposed based on the proper evaluation of the observation on radiation damage and recovery in oxide fuel.