Corrosion of high burn-up structured UO2 fuel in presence of dissolved H2

The influence of high burn-up structured material on UO₂ corrosion has been studied in an autoclave experiment. The experiment was conducted on spent fuel fragments with an average burn-up of 67 GWd/tHM. They were corroded in a simplified groundwater containing 33 mM dissolved H₂ for 502 days. All redox sensitive elements were reduced. The reduction continued until a steady-state concentration was reached in the leachate for U at 1.5 × 10⁻¹⁰ M and for Pu at 7 × 10⁻¹¹ M. The instant release of Cs during the first 7 days was determined to 3.4% of the total inventory. However, the Cs release stopped after release of 3.5%. It was shown that the high burn-up structure did not enhance fuel corrosion.     

Applicability of CeO2 as a surrogate for PuO2 in a MOX fuel development

The applicability of cerium oxide, as a surrogate for plutonium oxide, was evaluated for the fabrication process of a MOX (mixed oxide) fuel pellet. Sintering behavior, pore former effect and thermal properties of the Ce–MOX were compared with those of Pu–MOX. Compacting parameters of the Pu–MOX powder were optimized by a simulation using Ce–MOX powder. Sintering behavior of Ce–MOX was very similar to that of Pu–MOX, in particular for the oxidative sintering process. The sintered density of both pellets was decreased with the same slope with an increasing DA (dicarbon amide) content. Both the Ce–MOX and Pu–MOX pellets which were fabricated by an admixing of 0.05 wt% DA and sintering in a CO₂ atmosphere had the same average grain size of 11 μm and a density of 95%T.D. The thermal conductivity of the Pu–MOX was a little higher than that of the Ce–MOX at a lower temperature but both conductivities became closer to each other above 900 K. Cerium oxide was found to be a useful surrogate to simulate the Pu behavior in the MOX fuel fabrication.     

Reduction of UO2 by H2

The reactivity of H₂ towards UO₂²⁺ has been studied experimentally using a PEEK coated autoclave where the UO₂²⁺ concentration in aqueous solution containing 2 mM carbonate was measured as a function of time at p_H2∼40 bar. The experiments were performed in the temperature interval 74-100 °C. In addition, the suggested catalytic activity of UO₂ on the reduction of UO₂²⁺ by H₂ was investigated. The results clearly show that H₂ is capable of reducing UO₂²⁺ to UO₂ without the presence of a catalyst. The reaction is of first order with respect to UO₂²⁺. The activation energy for the process is 130 ± 24 kJ mol⁻¹ and the rate constant is k_298K=3.6×10⁻⁹ l mol⁻¹ s⁻¹. The activation enthalpy and entropy for the process was determined to 126 kJ mol⁻¹ and 16.5 J mol⁻¹ K⁻¹, respectively. Traces of oxygen were shown to inhibit the reduction process. Hence, the suggested catalytic activity of freshly precipitated UO₂ on the reduction of UO₂²⁺ by H₂ could not be confirmed.     

Simulations of H2O2 concentration profiles in the water surrounding spent nuclear fuel

A simple mathematical model describing the hydrogen peroxide concentration profile in water surrounding a spent nuclear fuel pellet as a function of time has been developed. The water volume is divided into smaller elements, and the processes that affect hydrogen peroxide concentration are applied to each volume element. The model includes production of H₂O₂ from α-radiolysis, surface reaction between H₂O₂ and UO₂ and diffusion. Simulations show that the surface concentration of H₂O₂ increases fairly rapidly and approaches the steady-state concentration. The time to reach steady-state is sufficiently short to be neglected compared to the times of interest when simulating spent fuel dissolution under deep repository conditions. Consequently, the steady-state approach can be used to estimate the rate for radiation-induced spent nuclear fuel dissolution.     

H2 inhibition of radiation induced dissolution of spent nuclear fuel

In order to elucidate the effect of noble metal clusters in spent nuclear fuel on the kinetics of radiation induced spent fuel dissolution we have used Pd particle doped UO₂ pellets. The catalytic effect of Pd particles on the kinetics of radiation induced dissolution of UO₂ during γ-irradiation in containing solutions purged with N₂ and H₂ was studied in this work. Four pellets with Pd concentrations of 0%, 0.1%, 1% and 3% were produced to mimic spent nuclear fuel. The pellets were placed in 10 mM aqueous solutions and γ-irradiated, and the dissolution of was measured spectrophotometrically as a function of time. Under N₂ atmosphere, 3% Pd prevent the dissolution of uranium by reduction with the radiolytically produced H₂, while the other pellets show a rate of dissolution of around 1.6 × 10⁻⁹ mol m⁻² s⁻¹. Under H₂ atmosphere already 0.1% Pd effectively prevents the dissolution of uranium, while the rate of dissolution for the pellet without Pd is 1.4 × 10⁻⁹ mol m⁻² s⁻¹. It is also shown in experiments without radiation in aqueous solutions containing H₂O₂ and O₂ that ?-particles catalyze the oxidation of the UO₂ matrix by these molecular oxidants, and that the kinetics of the catalyzed reactions is close to diffusion controlled.     

Dynamic Monte-Carlo modeling of thermal desorption of H2 from H irradiated graphite

Thermal desorption of hydrogen molecules from H⁺ irradiated graphite is studied using dynamic Monte Carlo simulation. The purpose of this study is to understand the experimentally observed phenomena that the thermal desorption of H₂ from the graphite exhibits sometimes single desorption peak, sometimes double peaks, and even three desorption peaks under certain circumstances. The study result reveals that the fluence of pre-implanted H⁺, the concentration of trap sites, porosity, and mean crystallite volume are important parameters in determining the number of desorption peaks. It is found that low implantation fluence and high concentration of trap sites easily lead to the occurrence of single desorption peak at around 1000 K, and high implantation fluence and low concentration of trap sites favor the occurrence of double desorption peaks, with a new desorption peak at around 820 K. It is also found that small porosity of graphite and large crystallite volume benefit the occurrence of single desorption peak at around 1000 K while large porosity of graphite and small crystallite volume facilitate the occurrence of double desorption peaks, respectively, at around 820 and 1000 K. In addition, experimentally observed third desorption peak at lower temperature is reproduced by simulation with assuming the graphite containing a small concentration of solute hydrogen atoms.     

An advanced model for intragranular bubble diffusivity in irradiated UO2 fuel

An advanced model for intragranular bubble diffusivity in irradiated UO₂ fuel is developed. Three various (surface, volume and gas-phase) mechanisms for the gas-filled bubbles diffusivity are reconsidered. It is shown that the bubble mobility by the volume diffusion mechanism can be strongly enhanced under irradiation conditions. Influence of the two-phase interface kinetics at a bubble surface on the volume diffusion, evaporation/condensation and surface migration mechanisms that can strongly suppress diffusivity of small nanometre bubbles, is additionally studied using a non-linear adsorption law derived for the van-der-Waals gas in the bubbles. The improved model is implemented in the MFPR code and validated against measurements of the small bubbles mobility.     

Specific low temperature release of Xe from irradiated MOX fuel

A particular low temperature behaviour of the ¹³¹Xe isotope was observed during release studies of fission gases from MOX fuel samples irradiated at 44.5 GWd/tHM. A reproducible release peak, representing 2.7% of the total release of the only ¹³¹Xe, was observed at ∼1000 K, the rest of the release curve being essentially identical for all the other xenon isotopes. The integral isotopic composition of the different xenon isotopes is in very good agreement with the inventory calculated using ORIGEN-2. The presence of this particular release is explained by the relation between the thermal diffusion and decay properties of the various iodine radioisotopes decaying all into xenon.     

Post-irradiation examination of AlFeNi cladded U3Si2 fuel plates irradiated under severe conditions

Three full size AlFeNi cladded U₃Si₂ fuel plates were irradiated in the BR2 reactor of the Belgian Nuclear Research Centre (SCK·CEN) under relatively severe, but well defined conditions. The irradiation was part of the qualification campaign for the fuel to be used in the future Jules Horowitz reactor in Cadarache, France. After the irradiation, the fuel plates were submitted to an extensive post-irradiation campaign in the hot cell laboratory of SCK·CEN. The PIE shows that the fuel plates withstood the irradiation successfully, as no detrimental defects have been found. At the cladding surface, a multilayered corrosion oxide film has formed. The U-Al-Si layer resulting from the interaction between the U₃Si₂ fuel and the Al matrix, has been quantified as U(Al,Si)_4.6. It is found that the composition of the fuel particles is not homogenous; zones of USi and U₃Si₂ are observed and measured. The fission gas-related bubbles generated in both phases show a different morphology. In the USi fuel, the bubbles are small and numerous while in U₃Si₂ the bubbles are larger but there are fewer.     

Controlled oxidation experiments of simulated irradiated UO2 fuel in relation to thermochemical modelling

A thermochemical model to describe the chemical state of irradiated nuclear fuel has been advanced and validated by comparison to results of experiments on naturally-enriched UO₂ with additions of selected simulated fission products. These experiments involved controlled oxidation in Ar/H₂O/H₂ gas mixtures conducted at the Atomic Energy of Canada Limited-Chalk River Laboratories. A coulombic titration technique provided measurements of moles of oxygen acquired by the samples in relation to oxygen potential. Emphasis was placed on the role of molybdenum in buffering the oxidation of fuel.This treatment is expected to be especially useful when integrated into fuel performance codes that make use of thermodynamics as boundary conditions in heat and mass transfer computations.     

Preliminary study of spent UO2 fuel corrosion in the presence of bentonite

The corrosion of spent UO₂ fuel in the presence of a dilute suspension (1.5%) of bentonite in synthetic groundwater has been studied. No significant changes in the uranium concentrations no indications of increased corrosion due to changes in solution chemistry, or due to sorption was found when bentonite was introduced to the system. The measured uranium concentrations were (3 ± 2) × 10⁻⁶M in the presence as well as in the absence of bentonite. The concentrations of plutonium and cationic fission products in the aqueous phase were lowered considerably, by up to two orders of magnitude in the case of plutonium due to sorption onto the bentonite.     

Non destructive determination of PuO2 content in MOX fuel pins for fast reactors using Passive Gamma Scanning

Use of Passive Gamma Scanning for non destructive evaluation of PuO₂ content in mixed oxide (MOX) fuels for fast reactors is demonstrated. Experiments have been carried out on MOX fuel pins for the hybrid core of Fast Breeder Test Reactor having nominal PuO₂ content of 44% and MOX pins having nominal PuO₂ content of 21% for the Prototype Fast Breeder Reactor. A comparison of results obtained using a conventional NaI(Tl) detector and that using a through well shaped detector is also presented.     

Some observations of density and porosity changes in UO2 fuel irradiated in water-cooled reactors

A technique has been developed for the hot-cell measurement of the apparent density of irradiated UO₂ fuel after extraction from a fuel pin. A single determination is accurate to $\text{± 3 \%}$ at the 95 % confidence limit. The method has been applied to fuel irradiated in thermal neutron fluxes in the Winfrith SGHWR and in the Halden BWR. Material has been examined at ratings of 1–51 W/g and in the burn-up range $\text{0.09–5.79 × 10}{}^{20}\text{fissions/cm}$. It is concluded that pellets with peak temperatures below 1100°C densify during irradiation, but at higher temperatures the pellets begin to swell. Fuel micrography has shown that the densification is principally due to the loss of micropores with a temperature dependency given by an activation energy of 5200 cal/mol. Above 1000°C the densification is masked by the formation and growth of intergranular fission gas bubbles, whose volume may exceed that of the manufactured pores which have sintered. In solid fuel pellets central swelling did not balance densification in the cooler rim until the fuel centre temperature exceeded 1700°C.     

NHO3氧化去除Np—Pu反萃液中的H2C2O4

研究了用ＮＨＯ３氧化去除ＴＲＰＯ流程反萃Ｎｐ－Ｐｕ的Ｈ２Ｃ２Ｏ４反萃液中Ｈ２Ｃ２Ｏ４的条件。７．５ｍｏｌ．Ｌ＾－１ＨＮＯ３－０．３ｍｏｌ．Ｌ＾－１Ｈ２Ｃ２Ｏ４混合液于９０℃下蒸发１３０ｈ和１００℃下蒸馏回流６ｈ,Ｈ２Ｃ２Ｏ４可完全分解去除;混合液中添加适量催化剂ＭｎＣＯ３,于１００℃下蒸发或蒸馏回流,Ｈ２Ｃ２Ｏ４分解加速,１－１．５ｈ内Ｈ２Ｃ２Ｏ４完全分解。蒸发或蒸馏回流过程中产生的ＨＮＯ２把Ｎｐ     

Interactions of slow H, H2, and H3 with thin carbon foils

Interaction processes resulting from the transit of incident 2–30 keV H⁺, H⁺₂ and H⁺₃ through 1.2 to 2.5 μg cm⁻² carbon foils are investigated by examining the charge state and angular scatter distributions of atomic and molecular species that exit the foils. A comparison of the scatter distributions of exiting H⁺₂ and H⁰ from incident H⁺₂ and H⁺₃ show that the atomic components of transmitted molecules scatter independently from foil atoms. For a given foil thickness, the measured fractions of H⁺₂ from incident H⁺₂ and H⁺₃ are inversely proportional to the square of the angular scatter half-width.     

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.     

Evaluation of high plutonia (44% PuO2) MOX as a fuel for fast breeder test reactor

Uranium plutonium mixed oxide (MOX) containing up to 30% plutonia is the conventional fuel for liquid metal cooled fast breeder reactor (LMFBR). Use of high plutonia (>30%) MOX fuel in LMFBR had been of interest but not pursued. Of late, it has regained importance for faster disposition of plutonium and also for making compact fast reactors. Some of the issues of high plutonia MOX fuels which are of concern are its chemical compatibility with liquid sodium coolant, dimensional stability and low thermal conductivity. Available literature information for MOX fuel is limited to a plutonium content of 30%. Thermodynamic assessment of mixed oxide fuels indicate that with increasing plutonia oxygen potential of the fuel increases and the fuel become more prone to chemical attack by liquid sodium coolant in case of a clad breach. In the present investigation, some of these issues of MOX fuel have been studied to evaluate this fuel for its use in fast reactor. Extensive work on the out-of-pile thermo-physical properties and fuel-coolant chemical compatibility under different simulated reactor conditions has been carried out. Results of these studies were compared with the available literature information on low plutonia MOX fuel and critically analyzed to predict in reactor behaviour of this fuel containing 44% PuO₂. The results of these out-of-pile studies have been very encouraging and helped in arriving at a suitable and achievable fuel specification for utilization of this fuel in fast breeder test reactor (FBTR). As a first step of test pin irradiation programme in FBTR, eight subassemblies of the MOX fuel are undergoing irradiation in FBTR.     

The creep of UO2 fuel doped with Nb2O5

The creep of UO₂ containing small additions of Nb₂O₅ has been investigated in the stress range 0.5–90 MN/m² at temperatures between 1422 and 1573 K. The functional dependence of the creep rate of five dopant concentrations up to 0.8 mol% Nb₂O₅ has been examined and it was established that in all the materials the secondary creep rate could be represented by the equation /.εkT = Aσⁿexp(?Q/RT), where /.ε is the steady state creep rate per hour, Q the activation energy and A and n are constants for each material. It was observed that Nb₂O₅ additions can cause a dramatic increase in the steady state creep rate as long as the niobium ion is maintained in the Nb⁵⁺ valence state. Material containing 0.4 mol% Nb₂O₅ creeps three orders of magnitude faster than the pure material.Analysis of the results in terms of grain size compensated viscosity suggest that, like “pure” UO₂, the creep rate of Nb₂O₅ doped fuel is diffusion-controlled and proportional to the reciprocal square of the grain size. A model is developed which suggests that the increase in creep rate results from suppression of the U⁵⁺ ion concentration by the addition of Mb⁵⁺ ions, which modifies the crystal defect structure and hence the uranium ion diffusion coefficient.     

Corrosion of UO2 and ThO2: A quantum-mechanical investigation

The addition of Th to U-based fuels increases resistance to corrosion due to differences in redox-chemistry and electronic properties between UO₂ and ThO₂. Quantum-mechanical techniques were used to calculate surface energy trends for ThO₂, resulting in (1 1 1) < (1 1 0) < (1 0 0). Adsorption energy trends were calculated for water and oxygen on the stable (1 1 1) surface of UO₂ and ThO₂, and the effect of model set-up on these trends was evaluated. Molecular water is more stable than dissociated water on both binary oxides. Oxidation rates for atomic oxygen interacting with defect-free UO₂(1 1 1) were calculated to be extremely slow if no water is present, but nearly instantaneous if water is present. The semi-conducting nature of UO₂ is found to enhance the adsorption of oxygen in the presence of water through changes in near-surface electronic structure; the same effect is not observed on the insulating surface of ThO₂.     

On stability of spatial distributions of crystal structure defects in irradiated high burnup UO2 fuel

Conditions of Kinoshita instability development of point defects and dislocation spatial distributions in the crystal structure of UO₂ fuel are studied. As a result of the instability development, spatially non-uniform regions with increased dislocation density are formed. Closed-form expressions of instability increment and spatial scale are derived. Parameters of the instability for irradiation conditions of high burnup UO₂ fuel are obtained by means of numerical simulation. Instability development time is shown to be inversely proportional to fission rate and it increases as dislocation density decreases. Calculated values of instability spatial scale and increment are in accordance with the size of fine grains and their formation rate in the peripheral zones of high burnup LWR fuel pellets.