Short communication on the Influence of the temperature between 30 and 70°C on the hydration of SON68 nuclear waste glass in a vapour phase

International audienc

Experimental approach to study the alteration of MOX MIMAS fuels in an underwater storage situation

International audienceAfter irradiation in a nuclear reactor, the irradiated Mimas MOX (Mixed Oxyde) fuel assemblies are stored in a pool before reprocessing or geological disposal. This long-term storage requires considering an incidental scenario corresponding to the presence of a through defect in the zircaloy cladding, responsible for a containment breach and potential degradation of the defective rod.This defect leads to the direct interaction between the spent fuel and the pool water. The pool water is pure aerated water that can contain boric acid at a concentration of 2g/L used as neutron absorber with a pH ranging from 4 to 6.5. It is also submitted to an intense gamma radiation field (around 1 k Gray/h) due to the presence of many fuel assemblies. The temperature of the pool ranges from 40 to 80°C because of the residual power released by the radioactive decay of the fuel assemblies.These conditions are favorable to the oxidizing dissolution of the fuel under alpha and beta gamma radiolysis of water which produces hydrogen peroxide H2O2, the main oxidizing species. This alteration can lead to the release of radionuclides in solution as well as to the formation of (U,Pu)O2+x oxidized layers and to the precipitation of secondary phases such as studtite or schoepite. These phases having a density lower than that of the fuel can induce a worsening of the defect and thus impact the mechanical strength of the rods. It is therefore important to study these degradation mechanisms with the aim of recovering the rods after several years and decades of storage. In order to assess this problem, an experimental approach coupling the study of simulating materials and highly radioactive materials is developed. It aims to describe the alteration mechanisms of fuels in contact with water as well as the reaction kinetics of dissolution/precipitation. This approach consists in implementing leaching experiments on materials ranging from model compounds (U, Th)O2, to MOX Mimas fuel (U, Pu)O2 irradiated or not in reactor. The model materials are developed at the ICSM by using thorium as a redox free plutonium surrogate. Thorium has the advantage of being easily manipulated but also of having a stable +IV valence, moreover like PuO2, ThO2 crystallizes in a fluorite structure Fm-3m. Homogeneous materials (U,Th)O2 are synthesized by hydroxide route in a first step. They are used as a precursor for the synthesis of materials with a heterogeneous microstructure produced by powder mixing in order to mimic the heterogeneous character of the MOX Mimas. Moreover, after irradiation in a reactor, the chemical composition of MOX is modified by the production of fission products that are classified into four families ((I) volatile fission products, (II) metallic precipitates, (III) oxide precipitates and (IV) FPs in solid solution in the fuel [3]. Two types of FPs are also simulated in the (U,Th)O2 model materials by addition of lanthanides (category IV) and platinum group metal elements (category (II)) during the synthesis of [4].Leaching experiments for these model or plutonium-containing materials are performed using two approaches:-Under dynamic conditions at the ICSM on (U,Th)O2 simulant/surrogate materials. These multiparametric experiments allow to study the effect of fuel heterogeneity and chemistry as well as water chemistry (boric acid, [H2O2], pH) on the oxidative dissolution of the fuel.-Under static conditions in a shielded cell in the Atalante facility at Marcoule on highly radioactive materials. The gamma dose rate within an assembly will be reproduced experimentally using a 60Co source given the small amount of fuel involved for the leaching experiment.These alteration experiments will be monitored over time by analyzing the solutions chemically and radiometrically and using surface characterization tools, the main one being Raman spectroscopy. This technique allows monitoring the evolution of many key parameters such as the chemical composition of the solid/solution interface, uranium oxidation state and the nature of the precipitated phases. It is also an isotopically sensitive technique for studying oxidation processes at the atomic scale [5, 6]. All this work will improve our understanding of the mechanisms of fuel oxidation and secondary phase precipitation in order to develop mechanistic model coupling chemistry to transport of chemical species in a defective rod. It will also contribute to the analysis of data acquired on defective rods extracted from storage pools

Leaching of homogeneous MOx pellets (U0.73Pu0.27O2) in carbonated water and clayey groundwater with metal iron

International audienc

Alpha dose rate and alpha decay dose impacts on the residual alteration rate regime of HLW nuclear glasses

International audienceThe long-term behavior of high-level nuclear glass subjected to alpha radiation by long-life minor actinides must be investigated with respect to geological disposal. This study focuses on the effects of alpha radiation on the chemical reactivity of R7T7-type glasses with pure water, mainly on the residual alteration rate regime, by considering separately the alpha dose rate and the alpha decay dose. Old SON68 glasses doped with $^(238/239)$PuO2 or $^(244)$CmO2 were studied in order to simulate high alpha dose rates corresponding to an early water ingress and also high level of alpha decay doses corresponding to long-term disposal conditions. A part of the $^(238/239)$Pu-doped glass block was annealed to fully recover the irradiation induced damage accumulated since the glass fabrication and to dissociate the effect of the alpha dose rate from the one of alpha decay dose. The glasses were then leached under static conditions in argon atmosphere at 90°C for several years. The results showed that the residual alteration rate is not impacted by the alpha dose rate parameter on a wide range of the dose rate values expected under disposal conditions, even in the eventuality of an early water ingress. It means that a SON68-type glass remained poorly sensitive to the alpha particle energy deposition at the glass-water interface. However, the residual alteration rate of the damaged $^(238/239)$Pu-doped glass was enhanced compared to the one of the annealed glass. This result is in agreement with the one obtained on the $^(244)$Cm-doped glass, and also with the literature about simplified glasses externally irradiated, indicating that the ballistic effects of the recoil nuclei are thus responsible for this increase of the residual alteration rate. A link between the glass structure and its leaching behavior is evidenced on these radioactive glasses and the role of the reactive interface is highly suspected

Trivalent actinides and lanthanides incorporation in UMo glass-ceramics

International audienceThe management of fission products solutions generated by reprocessing UMo fuel, containing high amounts of Mo and P, requires the development of a specific confinement matrix, called UMo glass-ceramic. In this UMo matrix, the formation of crystallized phases during the cooling after melting was observed and led to the formation of phosphate, molybdate and zircon phases, embedded in a durable glassy borosilicate matrix. Furthermore, UMo reprocessing solutions are enriched in americium (241Am in particular, an alpha-emitting isotope with a lifetime of 432 years) that raises the question of the Am incorporation and location in the glass-ceramic. This issue is therefore to assess the long-term behavior of such a matrix in the prospect of geological disposal. The present study focused on two types of UMo-glass ceramics. One type of samples was doped with various amounts of Am and/or Nd to rate the comparative role of Nd versus Am in terms of incorporation, location and microstructure as Nd is foreseen as an Am surrogate. A second type of samples was doped with 244Cm, in order to specify the 244Cm behavior compared to Nd and Am, in the prospect to use this short-lived isotope for ageing experiments to assess its long term evolution under self-alpha irradiation.These matrices were characterized by gamma spectrometry and isothermal calorimetry to check for Am, Cm and Nd global homogeneity, and then by Raman spectroscopy, SEM, EPMA and XRD to determine the microstructure and the actinide/Nd distribution in the different phases. The microstructure of the samples was heterogeneous as expected, with the presence of phosphate, molybdate and zircon crystalline phases. A great enrichment in Am, Cm and Nd was observed in the phosphate and molybdate phases compared to the embedding glass and the zircons. A partition coefficient of around x 6 - 7 for Am and Nd was calculated in these two crystalline phases compared to the global doping. Finally, Nd behaved like trivalent actinides in the UMo matrix and could be thus considered as a good surrogate

Modeling of the oxidative leaching of U0.73Pu0.27O2 in synthetic clay groundwater with metallic iron

International audienc

Behavior of B in passivating gels formed on International Simple Glass in acid and basic pH

International audienceInternational Simple Glass altered for years in silica saturated conditions develops a passivating gel, which retains some B and Ca in its the inner region. These elements could impact the way the gel limits glass alteration. However, the retention mechanism, speciation, and diffusion properties of these elements are still unknown. First, we found that the presence of B and Ca in solution can dramatically slow down glass alteration. However, splitting the B and Ca into two different solutions with other glass elements did not give similar protection against glass alteration. Further experiments in basic and acid pHs revealed that i) water diffusion is not dramatically affected by the retention of B and Ca in passivating gels, and ii) slow diffusion and high retention of B is observed only in gels formed in basic pH. This mechanism could explain how passivating gels control the residual rate

Ten years after the NPP accident at Fukushima : review on fuel debris behavior in contact with water

International audienceFollowing the NPP accident, several hundred tons of heat-generating corium and fuel debris have been cooled permanently by millions of m3 of flowing. Knowledge on the interaction with water is cruc..

Protein Analysis of Surface Active Macromolecular Fractions Isolated From Various Champagne Wines

International audienc