Evaporation of UO2 Containing Pu Nuclides Formed during Irradiation

Evaporation of UO₂ containing Pu nuclides formed during irradiation is discussed on the basis of data from the mass spectrometry of UO₂ specimen irradiated to about 8,000 MWD/T burnup. The principal findings are that: (1) the ratio between Pu⁺ and U⁺ ion currents obtained from the vaporization-ionization process of the UO₂ specimen was greater than the corresponding ratio of atoms present in the specimen; (2) in temperature dependency, the U⁺ ion current followed that of the UO₂ vapor pressure, but no such correspondence was seen between the Pu⁺ ion current and the PuO₂ vapor pressure.     

Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

Measured isotopic compositions of UO₂ and MOX fuel samples taken from irradiated light water reactor fuel assemblies were analyzed by CASMO5 coupled with a JENDL-4.0 base library to assess the uncertainties in the calculated isotopic compositions on heavy and fission product nuclides. The burnup calculations for the analysis were performed based on a single-assembly model taking into account the detail fuel assembly specifications and irradiation histories. For the MOX fuel samples, a multiple-assembly model was also adopted taking into account the effect of the surrounding UO₂ fuel assemblies. The average and standard deviation of the biases (C/E ? 1's (here C and E are calculated and measured results, respectively)) were calculated for each nuclide separately on the PWR and BWR UO₂ fuel samples. The averaged biases for ²³⁵U, ²³⁶U, ²³⁹Pu, ²⁴⁰Pu, ²⁴¹Pu and ²⁴²Pu were 2.7%, ?0.9%, 0.3%, 0.7%, ?2.4% and ?1.7% for PWR UO₂ samples, and 6.7%, ?1.5%, 2.5%, ?0.6%, 0.4% and ?0.1% for BWR UO₂ samples, respectively. The biases with the single-assembly model on the MOX fuel samples showed large positive values of ²³⁹Pu, and application of the multiple-assembly model reduced the biases as reported in our previous studies.     

Helium behavior in oxide nuclear fuels: First principles modeling

UO₂ and (U, Pu)O₂ solid solutions (the so-called MOX) nowadays are used as commercial nuclear fuels in many countries. One of the safety issues during the storage of these fuels is related to their self-irradiation that produces and accumulates point defects and helium therein.We present density functional theory (DFT) calculations for UO₂, PuO₂ and MOX containing He atoms in octahedral interstitial positions. In particular, we calculated basic MOX properties and He incorporation energies as functions of Pu concentration within the spin-polarized, generalized gradient approximation (GGA) DFT calculations. We also included the on-site electron correlation corrections using the Hubbard model (in the framework of the so-called DFT + U approach). We found that PuO₂ remains semiconducting with He in the octahedral position while UO₂ requires a specific lattice distortion. Both materials reveal a positive energy for He incorporation, which, therefore, is an exothermic process. The He incorporation energy increases with the Pu concentration in the MOX fuel.     

Determination of fissile fraction in MOX (mixed U + Pu oxides) fuels for different burnup values

When spent Light Water Reactor fuels are processed by the standard Purex method of reprocessing, plutonium (Pu) and uranium (U) in spent fuel are obtained as pure and separate streams. The recovered Pu has a fissile content (consisting of ²³⁹Pu and ²⁴¹Pu) greater than 60% typically (although it mainly depends on discharge burnup of spent fuel). The recovered Pu can be recycled as mixed-oxide (MOX) fuel after being blended with a fertile U makeup in a MOX fabrication plant. The burnup that can be obtained from MOX fuel depends on: (1) isotopic composition of Pu, which is closely related to the discharge burnup of spent fuel from which Pu is recovered; (2) the type of fertile U makeup material used (depleted U, natural U, or recovered U); and (3) fraction of makeup material in the mix (blending ratio), which in turn determines the total fissile fraction of MOX. Using the Non-linear Reactivity Model and the code MONTEBURNS, a step-by-step procedure for computing the total fissile content of MOX is introduced. As was intended, the resulting expression is simple enough for quick/hand calculations of total fissile content of MOX required to reach a desired burnup for a given discharge burnup of spent fuel and for a specified fertile U makeup. In any case, due to non-fissile (parasitic) content of recovered Pu, a greater fissile fraction in MOX than that in fresh U is required to obtain the same burnup as can be obtained by the fresh U fuel.     

Thermal conductivity degradation analyses of LWR MOX fuel by the quasi-two phase material model

The temperature measurements of mixed oxide (MOX) and UO₂ fuels during irradiation suggested that the thermal conductivity degradation rate of the MOX fuel with burnup should be slower than that of the UO₂ fuel. In order to explain the difference of the degradation rates, the quasi-two phase material model is proposed to assess the thermal conductivity degradation of the MIMAS MOX fuel, which takes into account the Pu agglomerate distributions in the MOX fuel matrix as fabricated. As a result, the quasi-two phase model calculation shows the gradual increase of the difference with burnup and may expect more than 10% higher thermal conductivity values around 75 GWd/t. While these results are not fully suitable for thermal conductivity degradation models implemented by some industrial fuel manufacturers, they are consistent with the results from the irradiation tests and indicate that the inhomogeneity of Pu content in the MOX fuel can be one of the major reasons for the moderation of the thermal conductivity degradation of the MOX fuel.     

Electrochemical Reduction of (U,Pu)O2 in Molten LiCl and CaCl2 Electrolytes

The electrochemical reduction of UO₂-PuO₂ mixed oxides (MOX) was performed in molten LiCl at 923 K and CaCl₂ at 1,123 K to evaluate the behavior of the plutonium quantitatively and to define the optimum conditions for the electrochemical reduction of those materials.

In LiCl, excess deposition of lithium metal can be avoided and the MOX was smoothly reduced at ?0.65 V vs. Bi-35 mol% Li reference electrode. The reduction ratio calculated from the mass change of the samples taken during the electrochemical reduction and the ratio evaluated by gas-burette method were in good agreement. The cathodic current efficiency remained 30–50% mainly due to the deoxidation of tantalum cathode basket. Although dissolution of plutonium and americium into the electrolyte was found by the chemical analysis, the dissolved amount was negligible and had no immediate influence on the feasibility of the electrochemical reduction process.

In CaCl₂, reduction of the MOX occurred in whole range of the tested cathode potential (?0.15 V to ?0.40 V vs. Ca-Pb reference electrode). The cathodic current efficiency was around 30%. Although the MOX was completely reduced at ?0.25 V, the reduction was interrupted by formation of the surface barrier made of the reduced material and the vacancy between the reduced and the non-reduced areas at ?0:30 V. Plutonium and americium dissolved also into the CaCl₂ electrolyte to slightly higher concentrations than those observed in LiCl electrolyte. The analyses for the reduction products showed that the amount of those actinides lost from the cathode was much larger than that found in the electrolyte, probably due to the formation of mixed oxide precipitate.     

Structural studies of the phase separation in the UO2–PuO2–Pu2O3 ternary system

In the oxygen hypo-stoichiometric range of (U_1?yPu_y)O_2?x mixed oxide MOX fuels, the U–Pu–O phase diagram is known to exhibit a large biphasic domain depending on the Pu content. However, the phase equilibria are still to be fully described as various representations are proposed in the literature.In the present work, we notify new insights into the phase separation occurring in the UO₂–PuO₂–Pu₂O₃ domain at room temperature. Our microstructural and X-ray diffraction results are compared to the different representations reported in the literature. We provide, for the first time in the hypo-stoichiometric domain, an indisputable experimental observation of a triphasic region at high Pu content, composed of two fluorite-type structures and of one α-Pu₂O₃ sesquioxyde type structure. These results are in contradiction with previous experimental representations of the U–Pu–O ternary system.     

ADAGIO technique: From UO2 fuels to MOX fuels

The amount of gas at the grain boundaries plays an important role in the fuel transient behaviour during accident conditions, such as a loss-of-coolant accident (LOCA) or a reactivity-initiated accident (RIA). Direct experimental determination of the grain boundary gas inventory has been performed for MOX fuel irradiated in an EDF pressurised water reactor (PWR) using the ADAGIO technique (ADAGIO is a French acronym meaning ‘Discriminatory Analysis of Accumulated Inter-granular and Occluded Gas’). The ADAGIO protocol applied to a MOX MIMAS fuel produced inter-granular gas fraction results that were consistent with those reached with other methods of evaluation i.e. electron probe microanalysis (EPMA). Furthermore, a new methodology for the numerical treatment of ⁸⁵Kr release kinetics which was developed for UO₂ was applied to MOX fuels. The corresponding results evidenced two types of release kinetics. These kinetics were attributed to the inter-granular bubbles of the UO₂ matrix and the bubbles located in the restructured zones, i.e. Pu agglomerates.     

Feedback on neutron capture cross sections of 238Pu and 241Am from analyses of measured isotopic compositions of irradiated LWR fuels and MOX core physics experiments

The atomic fractions of ²³⁸Pu and ²⁴¹Am in MOX fuels recycled in light water reactors are 1% to 2% and not significant compared with those of major Pu isotopes. On the other hand, recent evaluated nuclear data libraries, such as JENDL-4.0 and JEFF-3.2, give noticeably different thermal and epithermal neutron capture cross sections for ²³⁸Pu and ²⁴¹Am. The thermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am in JEFF-3.2 are 31% and 9% larger than those of JENDL-4.0, respectively. This paper shows the effect of the differences in the neutron cross sections on analysis results of two different integral experiments. The first is the isotopic compositions of ²³⁸Pu on UO₂ and MOX fuels irradiated in BWR and PWR, and the second is the critical experiments of the water moderated cores fully loaded with MOX fuels. The former was analyzed by using the continuous energy Monte Carlo burnup calculation code MVP-BURN and the latter by the continuous energy Monte Carlo calculation code MVP. The comparisons between the calculated and measured results indicate that the most likely thermal and epithermal neutron capture cross sections of ²³⁸Pu and ²⁴¹Am should be around at the middle between those of JEFF-3.2 and JENDL-4.0.     

Modeling and parametric studies of the effect of inhomogeneity on fission gas release in LWR MOX fuel

To analyze the effect of an inhomogeneous mixture of an PuO₂ powder on fission gas release in MOX fuel, a model has been developed using the assumption that gas release mechanism in Pu-rich particles is identical with that in UO₂ fuel. A parametric study was performed to see the respective effect of the number density, size and fraction of Pu retained in the Pu-rich particles on gas release in MOX fuel. The model shows that, for the condition of all the other remaining parameters being fixed, more gas is released in a MOX fuel for lower number density of, smaller size of, and larger fraction of Pu retained in, the Pu-rich particles. However, there exists some condition or combination of parameters for which the effect of inhomogeneity on gas release is negligible depending on the characteristics of MOX fuel. Comparison with measured data for OCOM MOX fuel shows that the present model can predict the level of gas release in MOX fuel once the release mechanism in the Pu-rich particles is known.     

An Apparatus for Continuously Measuring Radon-222 Exhalation from Ground

High burnup MOX and UO₂ test rods were prepared from the fuel rods irradiated in commercial BWRs. Each test rod was equipped with a fuel center thermocouple and reirradiated in the Halden boiling water reactor (HBWR) in Norway. The burnups of MOX and UO₂ test rods reached about 84GWd/tHM and 72GWd/t, respectively. Fuel temperature was measured continuously during the re-irradiation tests. Thermal conductivity change in high burnup fuel was evaluated from the results of comparison between the measured fuel temperature and the data calculated by using the fuel analysis code FEMAXI-6. The comparison results suggested that the thermal conductivity of MOX fuel pellets is comparable to that of UO₂ fuel pellets in the high burnup region around 80 GWd/t. It is probable that the impurity effect of Pu atoms gradually diminishes with increasing burnup because other factors that affect pellet thermal conductivity, such as the accumulation effect of soluble fission products and irradiation-induced defects in crystal lattice, become dominant in a high burnup region.     

The phase state at high temperatures in the MOX-SiO2 system

Influence of impurity Si on microstructure in a plutonium and uranium mixed oxide (MOX), which is used for fast breeder reactor fuel, was investigated, and phase state in 25% SiO₂ - (U_0.7Pu_0.3)O₂ was observed as a function of oxygen chemical potential. Compounds composed of Pu and Si with other elements were observed at grain boundaries of the MOX parent phase in the specimens after annealing. These compounds were not observed in the grain interior and the MOX phase was not affected significantly by impurity Si. It was found that the compounds tended to form more observably with decreasing O/M ratio and with increasing annealing temperatures.     

Determination of Li-6/Li-7 Isotopic Ratio by Means of Thermogravimetric Analysis for Lithium Sulphate Monohydrate

Plutonium concentrations and burnup at Pu spots were calculated in U-Pu mixed oxide (MOX) fuel pellets for light water reactors with the neutron transport and burnup calculation code VIMBURN. The calculation models were suggested for Pu spots and U matrices in a heterogeneous MOX fuel pellet. The calculated Pu concentrations and burnup at Pu spots were compared with the PIEs data in a MOX pellet (38.8 MWd/kgHM). The calculated Pu concentrations agreed by 5–18% with the measured ones, and the calculated burnup did by less than 10% with the estimated one with the measured Nd concentrations. Commercial PWR types of MOX fuels were also analyzed with the calculation code and the models. Burnup at Pu spot increased as the distance was greater from the radial center of a MOX fuel pellet. Burnup at Pu spots in the peripheral region became 3–5 times higher than pellet average burnup of 40 MWd/kgHM. The diameters (20–100 μm) of Pu spots were not found a significant factor for burnup at Pu spots. In the outer half volume region (outer than r/r _o=0.7) of a MOX fuel pellet, burnup at Pu spots exceeded 70MWd/kgHM (the threshold burnup of microstructure change in UO₂ fuel pellet) at pellet average burnup of 1430 MWd/kgHM.     

Neutron cross section sensitivity analysis of UO2 and MOX core physics experiments on light water reactors

The perturbation theory based on the transport calculation has been applied to study sensitivity of neutron multiplication factors (k_eff's) to neutron cross sections used for the reactivity analysis of UO₂ and MOX core physics experiments on light water reactors. The studied cross sections were neutron capture, fission and elastic scattering cross sections, and a number of fission neutrons, ν. The obtained sensitivities were multiplied to relative differences in the cross sections between JENDL-4.0 and JENDL-3.3 in order to estimate the reactivity effects. The results show that the increase in k_eff, 0.3%Δk/kk′, from JENDL-3.3 to JENDL-4.0 for the UO₂ core is mainly attributed to the decreases in the capture cross sections of ²³⁸U. On the other hand, there are various contributions from the differences in the cross sections of U, Pu, and Am isotopes for the MOX cores. The major contributions to increase in k_eff are decreases in the capture cross sections of ²³⁸U,²³⁸Pu, ²³⁹Pu, and those to decrease in k_eff are decreases in ν of ²³⁹Pu and increases in the capture cross sections of²⁴¹Am. They compensate each other, and the difference in k_eff between JENDL-3.3 and JENDL-4.0 is less than 0.1%Δk/kk′ and relatively small.     

Analysis of measured isotopic compositions of high-burnup PWR MOX and UO2 fuels in the MALIBU program

The measured isotopic compositions of fuel samples taken from high-burnup spent PWR MOX and UO₂ assemblies in the MALIBU program has been analyzed by lattice physics codes. The measured isotopes were U, Np, Pu, Am, and Cm isotopes and about 30 major fission product nuclides. The codes used in the present study were a continuous-energy Monte Carlo burnup calculation code (MVP-BURN) and a deterministic burnup calculation code (SRAC) based on the collision probability method. A two-dimensional multi-assembly geometrical model (2 × 2 model) was mainly adopted in the analysis in order to include the fuel assemblies adjoining the relevant fuel assembly, from which the samples were taken. For the MOX sample, the 2 × 2 model significantly reduces the deviations of the calculated results from the measurements compared with a single assembly model. The calculation results of MVP-BURN in the 2 × 2 model reproduce the measurements of U, Np, and Pu isotopes within 5% for the MOX sample of 67 GWd/t. The deviations of their calculated results of U, Np, and Pu isotopes from the measurements are less than 7% for the UO₂ sample of 72 GWd/t.     

Nuclear Design Study on a Super High-burnup Fuel Assembly for Pressurized Water Reactors Using Transuranium

A conceptual design study was carried out on a super high-burnup mixed-oxide (MOX) fuel assembly (SHB FA) for pressurized water reactors (PWRs) using transuranium (TRU). This study aims to avoid the surplus plutonium (Pu) accumulation and to reduce the accumulation of long-lived radioactive minor actinides (MAS) by utilizing the currently existing PWRs under the condition that the Japanese program to develop fast breeder reactors (FBRs) is tend to delay. For this purpose, an SHB FA with discharged burnup of ?80 GWd/t was investigated by utilizing MAS positively as both burnable absorbers and fissile suppliers and loading high-content Pu. It is possible to load the SHB FAs in a current PWR together with UO₂ FAs and to use 2.5 times as much amount of Pu as that in a standard 1/3 MOX core. Moreover, it is found to be possible to reduce the total number of fresh FAs further from that of a high-burnup (55 GWd/t in maximum) UO₂ (4.9 wt%) core and also to reduce the accumulation of MAS in the nuclear fuel cycle significantly.     

Effect of anisotropic scattering in neutronics analysis of BWR assembly

The anisotropic scattering effect to keff is studied for UO₂ and MOX fueled BWR assemblies. The anisotropic scattering effect increases the assembly k_∞ by 0.44% Δk for the UO₂ assembly with 0% void fraction, and by 0.21% Δk for the MOX assembly with 0% void fraction. This is because the anisotropic scattering effect flattens the intra-assembly thermal flux, and the absorption rate in the surrounding water gap is decreased, but the absorption rates in the MOX fuel rods are increased compared to the UO₂ rods. Therefore, the total decrease in absorption rates in the UO₂ assembly is relatively large, and the k_∞ is increased in the UO₂ assembly. The dependence of the anisotropic scattering effect on the void fraction is investigated, and the significant difference of 0.62% Δk/k is found for the 0% and the 80% void fractions. The BWR assemblies with Gd rods are also considered. Furthermore, the usefulness of the transport cross section is investigated, and it is found that the transport cross section gives reasonable anisotropic scattering effect, though not satisfactory.     

Determination of the composition and instability constants of complex Pu+3oxalate ions

The solubility of Pu₂(C₂O₄)₃ · 9H₂O in aqueous solutions of K₂C₂O₄ of various concentrations (0.01–2.4 moles /liter) has been determined at constant ionic strength of the solution at 20. It was found that Pu⁺³ complexes are formed in these solutions. It was found from the results of Pu₂(C₂O₄)₃ · 9H₂O solubility determinations that in the region of K₂C₂O₄ concentrations studied the following complex ions are formed [Pu(C₂O₄)₂]^?, [Pu (C₂O₄)₃]^?3 and [Pu (C₂O₄)₄]^?5, the total instability constants of which are 4.9 · 10^?10; 4.10 · 10^?10 and 11.9 · 10^?11 respectively. The solubility of Pu₂(C₂O₄)₃ · 9H₂O in aqueous (NH₄)₂C₂O₄ solutions has also been determined in the range of ammonium oxalate concentrations from 0.07 to 0.7 mole/liter at 70 °. It is shown that the composition of the complex ions under these conditions corresponds to [Pu(C₂O₄)₂]^?, [Pu(C₂O₄)₃]^?3 and [Pu(C₂O₄)₄]^?5. The calculated total instability constants of these complex ions are 11.6 · 10^?9; 5.6 · 10^?9 and 2.5 · 10^?9 respectively. The heats of formation of complex Pu⁺³ oxalate ions have been calculated for the reaction Pu⁺³ + nC₂O₄ ^?2 ?[Pu(C₂O₄)_n]^3?2n Δ¯Q for the [Pu(C₂O₄)₂]^? ion is 1300 cal., for [Pu(C₂O₄)₃]^?3, 1200 cal., and for [Pu(C₂O₄)₄]^?5, 1300 cal.     

Thermal conductivity evaluation of high burnup mixed-oxide (MOX) fuel pellet

The thermal conductivity formula of fuel pellet which contains the effects of burnup and plutonium (Pu) addition was proposed based on the Klemens’ theory and reported thermal conductivities of unirradiated (U, Pu) O₂ and irradiated UO₂ pellets. The thermal conductivity of high burnup MOX pellet was formulated by applying a summation rule between phonon scattering parameters which show the effects of plutonium addition and burnup. Temperature of high burnup MOX fuel was evaluated based on the thermal conductivity integral which was calculated from the above-mentioned thermal conductivity formula. Calculated fuel temperatures were plotted against the linear heat rates of the fuel rods, and were compared with the fuel temperatures measured in a test reactor. Since both values agreed well, it was confirmed that the proposed thermal conductivity formula of MOX pellets is adequate.