Calorimetric measurements on (U,Th)O2 solid solutions

Enthalpy increments of urania - thoria solid solutions, (U_0.10Th_0.90)O₂, (U_0.50Th_0.50)O₂ and (U_0.90Th_0.10)O₂ were measured by drop calorimetry in the temperature range 479 - 1805 K. Heat capacity, entropy and Gibbs energy function were computed. The heat capacity measurements were carried out also with differential scanning calorimetry in the temperature range 298 - 800 K. The heat capacity values of (U_0.10Th_0.90)O₂, (U_0.50Th_0.50)O₂ and (U_0.90Th_0.10)O₂ at 298 K are 59.62, 61.02, 63.56 J K⁻¹ mol⁻¹, respectively. The results were compared with the data available in the literature. From the study, the heat capacity of (U,Th)O₂ solid solutions was shown to obey the Neumann - Kopp’s rule.     

Separation of uranium from (U, Th)O2 and (U, Pu)O2 by solid state reactions route

Solid state reactions of UO₂, ThO₂, PuO₂ and their mixed oxides (U, Th)O₂ and (U, Pu)O₂ were carried out with sodium nitrate upto 900 °C, to study the formation of various phases at different temperatures, which are amenable for easy dissolution and separation of the actinide elements in dilute acid. Products formed by reacting unsintered as well as sintered UO₂ with NaNO₃ above 500 °C were readily soluble in 2 M HNO₃, whereas ThO₂ and PuO₂ did not react with NaNO₃ to form any soluble products. Thus reactions of mixed oxides (U, Th)O₂ and (U, Pu)O₂ with NaNO₃ were carried out to study the quantitative separation of U from (U, Th)O₂ and (U, Pu)O₂. X-ray diffraction, X-ray fluorescence, thermal analysis and chemical analysis techniques were used for the characterization of the products formed during the reactions.     

Properties of lenticular pores in UO2, (U,Pu) O2 and PuO2

Lenticular pore migration rates in oxide muclear fuels were experimentally measured in out-of-pile heating experiments. It is deduced that those pores which are in part responsible for the formation of columnar grains, are only produced in the absence of relevant amounts of filling gas. Specimens containing important concentrations of He, produced by Pu alpha decay, show columnar grain restructuring by grain boundary migration. Some consequences are drawn concerning the possible role played by lenticular pores in the mechanisms of fission gas release from nuclear fuels.     

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.     

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.     

Lattice parameters of (U, Pu, Am, Np)O2−x

The solid solutions of (U_{1−z−y’−y”}Pu_zAm_y’Np_y”)O_2−x (z = 0-1, y’ = 0-0.12, y” = 0-0.07) were investigated by X-ray diffraction measurements, and a database for the lattice parameters was updated. A model to calculate the lattice parameters was derived from the database. The radii of the ions present in the fluorite structure of (U, Pu, Am, Np)O_2−x were estimated from the lattice parameters measured in this work. The model represented the experimental data within a standard deviation of σ = ±0.025%.     

Thermal and mechanical properties of (U,Er)O2

Erbium is considered as a slow burnable poison suitable for use in light water reactors (LWRs). Addition of a small amount of Er₂O₃ to all UO₂ pellets will make it possible to develop super high burnup fuels in Japanese nuclear facilities which are now under the restriction of the upper limit of ²³⁵U enrichment. When utilizing the (U,Er)O₂ fuels, it is very important to understand the thermal and mechanical properties. Here we show the characterization results of (U_1−xEr_x)O₂ (0 ? x ? 0.1). We measured their thermal and mechanical properties and investigated the effect of Er addition on these properties of (U,Er)O₂. All Er completely dissolved in UO₂, and the lattice parameter decreased linearly with the Er content. Both the thermal conductivity and Young’s modulus of (U,Er)O₂ decreased with the Er content. These results would be useful for us in evaluating the performance of the (U,Er)O₂ fuels in LWRs.     

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.     

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.     

Crack formation,crack healing and porosity redistribution during irradiation of UO2 and (U,Pu) O2

The irradiation-induced void volume redistribution in the fuel was analysed. The radial crack volume and porosity distributions, the central radii and the radial gap width were measured after irradiation and compared with the calculated values. Short-time (He-loop experiments in the FR2 reactor), medium-time (bundle irradiation in the BR2 reactor) and long-time (trefoil-irradiation in the DFR reactor) irradiated fuel pins were examined. The model of pore migration, used in the computer code SATURN-la, is based on the evaporation-condensation mechanism. Measured swelling rates were extrapolated to higher temperatures and used. The crack volume distribution was calculated on the basis of a multifractured fuel model. One can conclude from the comparison between calculated and measured void volume distributions that several mechanisms redistribute void volume. These are crack formation, crack healing, migration of sinter pores and fission gas bubbles, gas swelling, evaporation-condensation phenomena in the region of the central void, irradiation-induced sintering and increase in diameter of the cladding.     

Thermal conductivities of irradiated UO2 and (U, Gd)O2 pellets

Thermal diffusivities of UO₂ and (U, Gd)O₂ pellets irradiated in a commercial reactor (maximum burnups: 60 GWd/t for UO₂ and 50 GWd/t for (U, Gd)O₂) were measured up to about 2000 K by using a laser flash method. The thermal diffusivities of irradiated UO₂ and (U, Gd)O₂ pellets showed hysteresis phenomena: the thermal diffusivities of irradiated pellets began to recover above 750 K and almost completely recovered after annealing above 1400 K. The thermal diffusivities after recovery were close to those of simulated soluble fission products (FPs)-doped UO₂ and (U, Gd)O₂ pellets, which corresponded with the recovery behaviors of irradiation defects for UO₂ and (U, Gd)O₂ pellets. The thermal conductivities for irradiated UO₂ and (U, Gd)O₂ pellets were evaluated from measured thermal diffusivities, specific heat capacities of unirradiated UO₂ pellets and measured sample densities. The difference in relative thermal conductivities between irradiated UO₂ and (U, Gd)O₂ pellets tended to become insignificant with increasing burnups of samples.     

Thermophysical properties of (U,Y)O2

We prepared polycrystalline pellets of (U,Y)O₂, containing YO_1.5 up to 11 mol.%. We performed indentation tests on the pellets, and evaluated the Young’s modulus and hardness. We measured the heat capacity and the thermal diffusivity, and evaluated the thermal conductivity. We succeeded in evaluating the effect of Y content on the thermophysical properties of (U,Y)O₂. We revealed that the Young’s modulus, hardness, and thermal conductivity of (U,Y)O₂ decreased with increasing the Y content.     

Electrical conductivity of UO2-ThO2 solid solutions

The electrical conductivities of UO_2+x. ThO₂ and their solid solutions, in thermodynamic equilibrium with the gas phase, were measured as a function of temperature, and of oxygen partial pressure in the temperatnre range 800 to 1200°C. The slope of the plot log α versus 1/$\text{T}$ for UO_2+x and UO₂-rich solid solutions exhibits a single region, whereas in the ThO₂-rich solid solutions it exhibits two regions. The pressure dependence of the conductivity (σ) in the UO₂-rich solid solutions can be represented by σ ∝ [O_i] ∝ in the range of $\text{0.01 < x < 0.1}$. Here, O_i is an interstitial oxygen and the partial pressure of oxygen, and it varies with the ThO₂ content. At greater deviation from stoichiometry ($\text{x ? 0.1}$) the presence of U₄O₉ or (Th U)₄O₉ phases influences the conductivity data. In ThO$\text{2}$ or ThO₂-rich solid solutions. P-type conduction at high oxygen pressures is interpreted as arising from the incorporation of excess oxygen into oxygen vacancies.     

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.     

(U,Gd)O2芯块试验与工业生产

文章介绍在宜宾核燃料元件厂(YFP)生产线上进行(U,Gd)O2芯块工业规模的生产试验及产品合格性鉴定,对试验结果进行讨论和评价.结果表明YFP的(U,Gd)O2芯块生产线完全具备工业生产能力,并实现(U,Gd)O2芯块制造的国产化.     

Electrical conductivity of near-stoichiometric (U,Gd)O2 solid solutions

The electrical conductivity of near-stoichiometric (U, Gd)O₂ solid solutions containing gadolinia up to 14 mol% was measured by four probes dc technique. The addition of gadolinia to urania enhanced the p-type electrical conductivity in the temperature range from room temperature to 1000 K. The electrical conductivity was able to be interpreted in terms of hopping of localized holes between U⁵⁺ and U⁴⁺ ions in the solid solutions. The experimental results presented in this study were consistent with the behavior predicted by the adiabatic small polaron theory.     

Phase diagram analysis of (U,Pu)O2−x sub-system

The high plutonium, hypo-stoichiometric fuel exists as two phase system at low temperatures. The partial phase diagram of (U,Pu)O_2−x with two coexisting cubic phases was extensively investigated in this work using theoretical models. The critical temperature of the miscibility gap varies with Pu/M and O/M of the system. Based on the similar miscibility gap behaviour observed in PuO_2−x system and the experimental data available on the phase boundaries of (U,Pu)O_2−x for various Pu/M, some semi-empirical relationships and solution models were developed. With the help of these relationships, ternary isothermal sections of the miscibility gap, O/M at different temperatures and the critical temperature of the miscibility gap of (U,Pu)_2−x for different Pu/M values were calculated. These calculated values were compared with the available literature data.     

Extension of vapour pressure measurements of nuclear fuels (U,Pu)O2 and UO2 to 7000 K for fast reactor safety analysis

A new high-energy laser technique, including fast temperature recording in the microsecond range, was developed for measuring the vapour pressure of fast breeder uranium-plutonium oxide fuels up to 7000 K. In the ternary system, the pressure of uranium-plutonium oxide above its melting point was determined to be $\text{log p (atm) = 7.966 ? (28137/T)}$, yielding a high-temperature heat of evaporation of $\text{ΔH}{}_{\mathrm{evap}}\text{= 128.7 kcal/mol}$. Measurement of the UO₂ partial pressure over the solid gave $\text{log p (atm) = 9.365 ? (32436/T)}$, resulting in a heat of sublimation of $\text{ΔH}{}_{\mathrm{sub}}\text{= 148.4 kcal/mol}$. The difference of these heats yields the heat of fusion $\text{ΔH}{}_{\mathrm{f}}\text{= 19.7 kcal/mol}$, which is in good agreement with the literature value of 19.4 kcal/mol. In the binary UO₂ system, the pressure above the melting point was determined to be $\text{log p (atm) = 7.7 ? (27900/T)}$,giving a heat of evaporation of $\text{ΔH}{}_{\mathrm{evap}}\text{= 127.6 kcal/mol}$. An assessment of literature data for below the melting point yielded $\text{log p (atm) = 8.846 ? (31506/T)}$, and a heat of sublimation of $\text{ΔH}{}_{\mathrm{sub}}\text{= 144.1 kcal/mol}$. The resulting heat of fusion, $\text{ΔH}{}_{\mathrm{f}}\text{= 16.5 kcal/mol}$, is only slightly below the published value of $\text{ΔH}{}_{\mathrm{f}}\text{= 17.7 kcal/mol}$.