Oxygen potential and thermal conductivity of (U, Pu) mixed oxides

(U, Pu) mixed oxides, (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 are being considered as fuels for the Prototype Fast Breeder Reactor (PFBR) in India. The use of urania-plutonia solid solutions in PFBR calls for accurate measurement of physicochemical properties of these materials. Hence, in the present study, oxygen potentials of (U_1−yPu_y)O_2−x, with y = 0.21 and 0.28 were measured over the temperature range 1073-1473 K covering an oxygen potential range of −550 to −300 kJ mol⁻¹ (O/M ratio from 1.96 to 2.000) by employing a H₂/H₂O gas equilibration technique followed by solid electrolyte EMFmeasurement. (U_1−yPu_y)O_2−x, with y = 0.40 is being used in the Fast Breeder Test Reactor (FBTR) in India to test the behaviour of fuels with high plutonium content. However, data on the oxygen potential as well as thermal conductivity of the mixed oxides with high plutonium content are scanty. Hence, the thermal diffusivity of (U_1−yPu_y)O₂, with y = 0.21, 0.28 and 0.40 was measured and the results of the measurements are reported.     

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.     

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.     

Plutonium redistribution in mixed oxide (U,Pu)O2 nuclear fuel elements

Relatively simple analytic expressions are found for actinide redistribution by lenticular pore motion and by solid-state thermal diffusion. Properties of the former, which is based on a previously determined expression for the pore velocity, $\text{v}$, are investigated, and calculations show that pore motion cannot significantly redistribute plutonium unless $\text{v}$ is greatly reduced by a mechanism such as fission gas pickup. The expression for redistribution from thermal diffusion is fitted to data for two fuel pins. Together with previous fits, the results for one pin are consistent with the use of a Pu-U diffusivity with an activation energy of $\text{105 ± 10 kcal/mole}$. The failure to fit the data on the other pin indicates the need for a reduction in uncertainties in specified temperatures. Redistribution by vapour transport in cracks is found to be unimportant for observed crack densities.     

Calculation of the oxygen potential in the mixed oxide (U,Pu)O2 − x

We discuss a number of models for the structure of the substoichiometric mixed oxide (U, Pu)O_{2 ? x}. These models have parameters that may be calculated directly and indeed most of the parameters are calculated.The oxygen potentials we obtain reproduce the Markin—McIver rule at low temperatures, but predict that it should break down at high temperatures. Areas where further experimental work is required are indicated.     

Fabrication of high-density UO2 and (U0.75Pu0.25)O2 by hot pressing

The hot pressing behavior of hyperstoichiometric UO₂ and (U, Pu)O₂ powders has been evaluated. Specimens with densities in excess of 99% $\text{ρ}{}_{\mathrm{th}}$ can be fabricated with this technique at temperatures of 1000°C or less.     

A study of the generation of 232U in UO2 and MOX fuels

To clarify the generation pathway of ²³²U, an important nuclide for dose evaluation at various stages in the reuse of uranium, concentrations of ²³²U generated through various pathways were evaluated for UO₂ and mixed oxide (MOX) fuels. Burnup calculation was conducted with ORIGEN2.2 code adopting ORLIBJ40 library, a set of cross-section libraries based on JENDL-4.0. It was found that differences in ²³²U concentrations in UO₂ and MOX fuels mainly arise from differences in the initial compositions of ²³⁴U, ²³⁵U, and ²³⁶U. It was also found that the contribution of plutonium and americium isotopes in MOX fuels is small compared with that of uranium isotopes. The results clarified that the capture cross sections of ²³⁰Th, ²³¹Pa, ²³⁵U, and ²³⁶U, as well as the (n,2n) cross sections of ²³⁷Np and ²³⁸U, have a large effect on the generation of ²³²U. Additional investigation showed that ²³²U concentration is strongly affected not only by time after irradiation but also by time before irradiation.     

Thermal conductivity of (U,Pu,Np)O2 solid solutions

The thermal conductivities of (U_,Pu_,Np)O₂ solid solutions were studied at temperatures from 900 to 1770 K. Thermal conductivities were obtained from the thermal diffusivity measured by the laser flash method. The thermal conductivities obtained below 1400 K were analyzed with the data of (U,Pu,Am)O₂ obtained previously, assuming that the B-value was constant, and could be expressed by a classical phonon transport model, λ = (A + BT)⁻¹, A(z₁, z₂) = 3.583 × 10⁻¹ × z₁ + 6.317 × 10⁻² × z₂ + 1.595 × 10⁻² (m K/W) and B = 2.493 × 10⁻⁴ (m/W), where z₁ and z₂ are the contents of Am- and Np-oxides. It was found that the A-values increased linearly with increasing Np- and Am-oxide contents slightly, and the effect of Np-oxide content on A-values was smaller than that of Am-oxide content. The results obtained from the theoretical calculation based on the classical phonon transport model showed good agreement with the experimental results.     

Thermal conductivity of near-stoichiometric (U, Ce)C and (U, Pu, Ce)C solid solutions

The thermal conductivities of near-stoichiometric (U, Ce)C and (U, Pu, Ce)C solid solutions containing CeC up to 10 mol% were determined in the temperature range from 740 to 1600 K by the laser flash method. The thermal conductivity decreased with the cerium content in the solid solutions. The electrical resistivities were also measured for the purpose of analyzing the heat conduction mechanism. It was found that the decrease of electronic heat conduction caused by the addition of cerium resulted in decreasing the thermal conductivities of (U, Ce)C and (U, Pu, Ce)C compared with UC and (U, Pu)C.     

The movement of lenticular pores in mixed oxide (U,Pu) O2 nuclear fuel elements

A thermodynamically consistent theory is developed for the motion of lenticular pores in mixed oxide fuel. The advance on a previous theory is that oxygen concentrations on the two pore faces are allowed to differ according to the local plutonium concentration, as it is shown that otherwise stability criteria for the pore are violated. Numerical results for pore velocities are insensitive to values of oxygen to metal (O/M) and plutonium to metal (Pu/M) ratios and are practically independent of the oxygen heat of transport.Pores are shown to transport plutonium towards the centre of the fuel above a critical temperature which depends strongly on the O/M and Pu/M ratios. A macroscopic current representing this transport is derived, but the need is pointed out for estimating gas pickup by pores before concluding whether this current contributes towards Pu redistribution in actual reactor fuels.     

Chemical reactions between sodium and (U,Pu)02 mixed oxides

The three phase field MO₂-Na₃MO₄-Na has been studied in detail. It was shown that Na₃MO₄ exists at least in two crystallographic forms and may be above 1200°C, a third one fcc, with a small cell $\text{a = 4.77}\text{A}\text{?}$. This latter form which is considered in the literature as the “normal” uranate cannot be stable in the given conditions of preparation. The low temperature form, simple cubic $\text{a = 9.54}\text{A}\text{?}$, is the one to which the wrong Na₁₁U₅O₁₆ formula has been attributed up to now. It transforms at ca 975°C into a face-centered cubic form with $\text{a = 9.56}\text{A}\text{?}$, stable at least up to 1200°C. Isomorphism between Na-uranate and Na-uranoplutonate has been established for Pu content up to 30%, but there is still some doubt about the nature of the plutonate which is formed by direct reaction between PuO₂ and Na.For the compositions studied (Pu ≤ 30%) the monovariant domain adjacent to the “MO₂”-Na₃MO₄-Na domain, contains the three phases Na₃MO₄, Na₆MO₆ and Na.Oxygen potentials corresponding to the Na₃MO₄-MO_2-x-Na phase field with M = U_0.8Pu_0,2 have been deduced from the measurement of x₀ in equilibrium conditions. It has been foun'd that X_₀ was dependent on T, varying from 0.007 at 550°C to 0.040 at 1200°C. Corresponding values of the oxygen potentials can be represented by the expression: $\text{}\text{?}\text{sm = ? 907, 394 + 224.9 T J/mol 0}{}_2$ A thermodynamic treatment shows that this expression is quasi-independent of Pu content (Pu≤30%) and must be very close to the one valid for the UO₂-Na₃UO₄-Na domain.In conclusion, we examine very schematically how these results can offer guidelines to appreciate the evolution of failed breeder pins in reactor conditions.     

Preliminary study of the tight lattice pressured heavy water reactor loaded with Pu／U and Th／U mixed fuels

To improve nuclear fuel utilization efficiency and prolong fuel cycle burn-up,a tight ptich lattice pressured heavy water reactor was investigated as an alternative of next generation of power reactors.It is shown that the high conversion ratio and negative coolant void reactivity coefficient are challenges in the reactor core physics designs.Various techniques were proposed to solve these problems.In this work.a tight pitch lattice and mixed fuel assemblies pressured heavy water reactor concept was investigated.BY utilizing numerical simulation technique,it is demonstrated that reactor core mixed with Pu/U and Th/U assemblies can achieve high conversion ratio(0.98) ,long burn-up(60GWD/t)and negative void reactivity coefficients.     

Thermal conductivity and thermal expansion of hot-pressed trisodium uranate (Na3UO4)

Thermal conductivity and thermal expansion of Na₃UO₄ prepared by two different reaction processes were determined over a temperature range of 20–1000°C. Compositional differences in the samples resulting from the different reaction processes have a pronounced effect on thermal expansion and on thermal conductivity below 500°C. Above 500°C, these compositional differences in the thermal conductivities decrease.     

Synthesis and thermal conductivity of Y6UO12

Y₆UO₁₂ was synthesized by solid-state reactions of Y₂O₃ and U₃O₈. The high-density pellet of Y₆UO₁₂ was prepared by the spark plasma sintering followed by heat treatment in air for oxygen supplementation. The thermal conductivity (κ) was evaluated using the laser flash method from room temperature to 1173 K. The κ of Y₆UO₁₂ decreased with increasing temperature in the whole temperature range, indicating that the phonon contribution was predominant. The room temperature κ value of Y₆UO₁₂ was 4.90 Wm^?1K^?1. The magnitude relationship of κ among Y₆UO₁₂, Y₆WO₁₂, and Yb₆WO₁₂, i.e. κ of Yb₆WO₁₂ < κ of Y₆UO₁₂ < κ of Y₆WO₁₂, was discussed based on the general lattice thermal conductivity theory.     

Thermal conductivities of hypostoichiometric (U, Pu, Am)O2−x oxide

The thermal conductivities of (U_0.68Pu_0.30Am_0.02)O_2.00−x solid solutions (x = 0.00-0.08) were studied at temperatures from 900 to 1773 K. The thermal conductivities were obtained from the thermal diffusivities measured by the laser flash method. The thermal conductivities obtained experimentally up to about 1400 K could be expressed by a classical phonon transport model, λ = (A + BT)⁻¹, A(x) = 3.31 × x + 9.92 × 10⁻³ (mK/W) and B(x) = (−6.68 × x + 2.46) × 10⁻⁴ (m/W). The experimental A values showed a good agreement with theoretical predictions, but the experimental B values showed not so good agreement with the theoretical ones in the low O/M ratio region. From the comparison of A and B values obtained in this study with the ones of (U,Pu)O_2−x obtained by Duriez et al. [C. Duriez, J.P. Alessandri, T. Gervais, Y. Philipponneau, J. Nucl. Mater. 277 (2000) 143], the addition of Am into (U, Pu)O_2−x gave no significant effect on the O/M dependency of A and B values.     

The use of NRA to study thermal diffusion of helium in (U, Pu)O2

A lot of work has been already done on helium atomic diffusion in UO₂ samples, but information is still lacking about the fate of helium in high level damaged UOX and MOX matrices and more precisely their intrinsic evolutions under alpha self irradiation in disposal/storage conditions.The present study deals with helium atomic diffusion in actinide doped samples versus damage level. The presently used samples allow a disposal simulation of about 100 years of a UOX spent fuel with a 60 MW d kg^?1 burnup or a storage simulation of a MOX spent fuel with a 47.5 MW d kg^?1 burnup.For the first time, nuclear reaction analysis of radioactive samples has been performed in order to obtain diffusion coefficients of helium in (U, Pu)O₂. Samples were implanted with ³He⁺ and then annealed at temperatures ranging from 1123 K to 1273 K. The evolution of the ³He depth profiles was studied by the mean of the non-resonant reaction: ³He(d, p)⁴He. Using the SIMNRA software and the second Fick’s law, thermal diffusion coefficients have been measured and compared to the ³He thermal diffusion coefficients in UO₂ found in the literature.