Estimating the Airborne Release Fraction (ARF) during fuel debris (FD) retrieval is important for the safe decommissioning of the Fukushima Daiichi Nuclear Power Plant (1F) because it supports exposure assessment and the design of mitigation measures. Dust generation during dry mechanical cutting depends not only on operating conditions but also strongly on the properties of the cut material, including size, geometry, mechanical behavior and multiphase heterogeneity. Therefore, representative simulants are required to obtain reliable datasets. In the CRIEPI-led Dust Acquisition project, this paper focuses on the design and kilogram-scale fabrication of uranium-bearing FD simulants representative of major in-vessel and ex-vessel debris conditions at 1F. Five simulants were manufactured at CEA using thermite reactions, direct induction heating or combined processing, depending on thermochemical constraints and target compositions. A representative manufacturing example is presented to illustrate composition selection and optimization of thermite reaction in order to obtain complete melting and coherent blocks at the kilogram scale. As a result of this project and previously manufactured compositions in URASOL IRID-JAEA/CLADS project, some of the five FD representative blocks with multiphase (U-Zr-O oxide matrices with metallic inclusions and MCCI-like silicate-rich phases) and non-uniform aspect have been selected. Main features of the blocks are presented. They provide realistic FD for generation of aerosols by dry disc cutting on FUJISAN2 facility, designed within a collaboration between CRIEPI, ONET, CEA, and IRSN (now ASNR). Physical and chemical characterization of aerosols techniques are shortly presented.

Hf1-xYxO2-x/2 ceramics have been recently suggested as a solid electrolyte for potentiometric sensors monitoring oxygen in liquid sodium. High density ceramics can be produced by an innovative dropwise wet-process using hydroxide coprecipitation allowing a homogeneous mixture of hafnium and yttrium under specific stoichiometric conditions in the presence of ammonia. Different parameters such as the synthesis duration (2-26 min), concentration (2-10 M) and excess of ammonia (200-1000 %) used were varied to determine the optimized synthesis parameters for obtaining a very dense material. After synthesis, the precursor obtained was mainly characterized by BET and then calcined under air at 700 degrees C to obtain the corresponding oxide. The latter was further characterized by BET, SEM, EDS and TGA before being shaped and sintered at 1600 degrees C. The relative density of the densest ceramics reached 97.6 +/- 0.5 % of the theoretical density, rarely obtained by wet chemistry routes.

Fuel debris retrieval is one of the important challenges in view of Fukushima Daiichi decommissioning. Indeed, hundreds of tons of fuel debris will have to be cut and collected in the 3 units that have been subject to core meltdown in March 2011. A large R&D effort has been supported by the Japanese stakeholders in support of this retrieval. In this context, CEA has used its severe accident and decommissioning research expertise to launch research activities in support of fuel debris retrieval. A first series of works aims to acquire knowledge about these fuel debris: experiments have been carried out to simulate Molten Core Concrete Interaction or to fabricate fuel debris simulants or depleted uranium-containing prototypes. These samples have been thoroughly analyzed and their properties have been determined. A second line of research deals with fuel debris cutting. CEA has successfully applied and improved its laser cutting technology to fuel debris. Mechanical cutting has also been studied. One of the important safety issues related to debris cutting is the generation of radioactive aerosols and particles. Dedicated research programs have been carried out to characterize these releases and study mitigation techniques.

Optimizing melt spreading in the aftermath of a core disruptive accident is crucial for achieving sufficient melt cooling to maintain reactor containment integrity. Two approximate to 30 kg-scale experiments performed at the VULCANO facility explore the spreading of high-temperature molten corium-concrete mixtures over ceramic and sacrificial concrete substrates. Imaging of the melt front propagation revealed a 7% increase in spreading length and a 30% increase in maximum front velocity during spreading over sacrificial concrete, despite a reduced mass partaking in spreading due to increased holdup within the crucible. Infrared imaging of the melt indicated surface temperatures around 45 degrees C lower during spreading on sacrificial concrete, resulting in a roughly three-fold increase in melt viscosity. The enhanced viscosity and reduced mass during the VE-U9-concrete test imply an increased spreadability on sacrificial concrete greater than the observed 7% increase in spreading length. This enhanced spreadability on sacrificial concrete could be explained by the apparent gliding motion of the melt, consistent with reduced friction at the melt-substrate interface. Reduced friction at the melt-substrate interface is best explained by a diphasic film of molten concrete and gaseous concrete decomposition products acting as a lubricant between the melt and solid substrate.

The effect of several impuritites on the life-cycle of yttria-doped thoria electrolyte was evaluated in this study. All the oxides exhibited the single-phase fluorite-type structure expected for such materials, while the variation of the unit cell parameter attested for the incorporation of the impurities in the lattice. Doping with 0.2 – 1 wt.% of Al, Si or Zr only slightly affected the morphology of the powders, as well as the microstructure of the pellets obtained after sintering, even if a deleterious effect, leading to the fragilization of the samples, was observed for high silicon contents. The presence of silicon and zirconium in the ceramic further modified its electrical properties. The grain conductivity was always found to decrease. Impedance spectroscopy also revealed a strong modification of the signal assigned to grain boundaries, which evidenced that impurities partially migrated during the sintering step. Finally, corrosion tests undertaken in liquid sodium at 773 K showed that the pellets were fragilized by the presence of silicon or high zirconium contents, possibly through the formation of Na2MO3 ternary oxides at the grain boundaries. From a general point of view, silicon seems to be the most harmful impurity that should be especially checked during the fabrication processes.

The lithium niobate used in an ultrasonic transducer for liquid metal fast nuclear reactors is prone to a reduction leading to a rapid short circuit of the sensor in operating conditions. Oxygen consumption on stainless steel sensor’s walls was suspected to be the cause of the reduction. Impedance spectroscopy monitoring around 600 °C under different O2 partial pressures shows that the reduction limit is above the equilibrium pressure of iron oxides. As the housing material (304L) cannot be easily changed, various coating solutions enhancing the passivation of the steels were investigated. Sol–gel lanthanum coatings show the best behavior when tested by TGA at 800 °C. However, in situ monitoring of the oxygen consumption by an oxygen sensor close to operating condition shows that it was not enough protective.

Within the development of SFR nuclear reactors, yttrium-doped thoria has been suggested by previous researchers as an electrolyte in sensors monitoring oxygen in liquid sodium. In this paper, Th1-xYxO2-x/2 electrical properties were assessed by impedance spectroscopy before their chemical durability was evaluated in liquid sodium. Pellets always kept their mechanical integrity although sodium penetration was evidenced through the modification of the cleavage surfaces. Post-immersion impedance spectroscopy showed a decrease of grain boundary contribution, even if the global conductivity was not significantly affected. These good performances could be strongly affected by the presence of impurities since silicon-doping increased the sodium penetration.

In the framework of JAEA-CEA collaboration, experimental studies have been conducted for estimating the material characteristics of corium debris representative of the Fukushima Daiichi nuclear damaged plants. A test has been performed in the VULCANO facility in CEA Cadarache to simulate the concrete corium interaction (CCI) with prototypic corium (using depleted uranium) and concrete of Fukushima Daiichi 1F1 Nuclear Plants. This paper presents the Post Test Analyses on 9 samples representative of the CCI during this test: in the corium pool, in the crusts and at the vertical and horizontal interfaces with the concrete. Analyses have been performed by SEM/EDS, X-Ray Diffraction, complete dissolution and ICP, micro-hardness measurements of the main phases. The pool is very porous, its composition is homogeneous but metallic blocks have been observed at the interfaces. The major phases encountered are uranium rich and zirconium rich oxides forming nodules from micrometers to millimeters size, chromium-iron rich precipitates of several micrometers, metallic Fe–Ni droplets and chromium-silicon rich filaments in a matrix, likely vitreous, rich in concrete elements: Si, Al, Ca, but containing up to 12 cations. The matrix is the softer oxide phase, when the Cr rich precipitates are the harder. The analyses are consistent with the estimated macroscopic ablation ratio, but do not still explain the important axial ablation observed for this specific basaltic concrete. The different phases formation, distribution and solidification path are discussed. First comparisons are proposed with the former CCI tests with European concretes. These results give helpful insights for the future dismantling of the plant and for a deeper understanding of the CCI process for basaltic concrete.

Characterization of fuel debris is required to develop fuel debris removal tools for decommissioning Fukushima Daiichi nuclear power plant (1F). Especially, knowledge about the characteristics of molten core-concrete interaction (MCCI) product is needed because of the limited information available at present. Samples from a large-scale MCCI test performed under quenching conditions, VULCANO VW-U1 were analyzed to evaluate the characteristics of the surface of MCCI product. Four samples were selected from test sections at different locations. As a result, the characteristics of the samples were found to be similar. Several corium phases, such as cubic-(U,Zr)O-2 and tetragonal ZrO2, were detected by X-ray diffraction (XRD), but concrete-based phases, such as the crystalline SiO2 phase, were not detected by XRD because the quantity of the SiO2 phase was too small to be measured. The Vickers hardness of each phase in these samples was higher than that of previously analyzed samples in another VULCANO test campaign, VBS-U4. Based on a comparison between MCCI product generated under quenching condition, such as VW-U1, and gently cooled MCCI product, such as VBS-U4, the MCCI product generated under quenching condition is more homogeneous, and its hardness is higher than that of the gently cooled MCCI product.

The French Atomic and alternative Energy Commission (CEA) aims to reuse its sodium fire (carbonate base) extinguishing powder after long term storage (stock from the dismantlement of its old sodium facilities). As the composition of the powder appears to change during the storage, the efficiency on the extinction as a function of the physicochemical properties was questioned. Small sodium fire extinction experiments were carried out with powders of different compositions. The results demonstrated a dominant role of water of crystallization on the extinction. Two steps are proposed for the extinction mechanism that includes: (1) the formation of liquid sodium hydroxide and (2) the melting of carbonate mixture at eutectic composition. The sodium hydroxide behaves as a protective layer and insulates the sodium surface from prolonged contact with oxygen. Consequently, it provides rapid decrease of temperature, unlike the slow melting of carbonates eutectic and its porous layer formed due to its higher viscosity. The presence of trona (aging product) does not alter the extinction capacity of the powder. To extrapolate the results to large fires, 35 g of water of crystallization are necessary to extinguish 1 m(2) of sodium pool fire. Finally, the particle size appears to be a non-significant parameter to the quality of extinction except for the spreading performance.

The characteristics of fuel debris are required to develop tools to remove it from severely damaged nuclear power plants, but the knowledge on molten core concrete interaction (MCCI) product is limited.The VULCANO MCCI test campaign, VBS-U4, was selected for the analysis as the conditions are similar to MCCI at Fukushima daiichi nuclear power plant.Physicochemical phenomena during MCCI were estimated from the macrostructure of the samples.Moreover, heterogeneous microstructure of the oxide region and homogeneous microstructure of the metallic region was confirmed.Additionally, the obvious difference of Vickers hardness between oxide and metallic phase is evaluated.

Sintering of Th1-xYxO2-x/2 ceramics (x=0.01, 0.08, 0.15 and 0.22), planned to be used as solid electrolytes in oxygen sensors for sodium-cooled fast nuclear reactors, was investigated. High densification state (i.e. up to 98% TD) was reached after 4h of heat treatment at 1600°C and beyond. In addition, ESEM observations showed a major effect of yttrium on grain size due to solute drag effects. Sintering maps were plotted for all the samples and evidenced different stages driven by densification and grain growth. Grain growth was found to be strongly slowed down for x >0.01, resulting in high values of relative density correlated to submicrometric grain size. Also, activation energies related to densification and grain growth were evaluated around 450 and 500–650kJmol−1, respectively. These results led to deliver guidelines for the formulation and sintering of Th1-xYxO2-x/2 ceramics in prospect of their use as a solid electrolyte.

The OSCAR-Na code has been developed to calculate t he mass transfer of corrosion products and related contamination in the primary circuit of sod ium fast reactors (SFR). Indeed, even if fuel cladd ing corrosion appears to be very limited, the contamina tio of the reactor components plays an important r ole in defining the design, the maintenance and the decomm issioning operations for SFR. The modeling is based on the solution/precipitation of the different elements of the steel. These elem nts dissolve mainly at the hot surfaces, and precipitat e on the cold surfaces, and then induce the shiftin g of the metal/sodium interface (bulk corrosion or bulk depo sit). The diffusion in the steel is also taken into account and allows calculating the preferential release of the most soluble elements (nickel, chromium, and mangan ese). The code uses a numerical method for solving the di ffusion equation in the steel and the complete mass b lance in sodium for all elements, allowing the calculatio n f the metal/sodium interface shifting and of the flux of each element through this interface. Code validation has already been carried out agains t PHENIX contamination on heat exchanger surfaces f or the main radionuclides. This paper presents the continu ation of the validation process against experimenta l results obtained on the STCL sodium loop, with well control led experimental conditions. Different parameters o f the model are adjusted to match concentration profiles in the metal and elementary releases measured at 60 4 °C. These parameters are the solubility in the sodium a nd the diffusion coefficient in the steel for each element, as well as the oxygen-enhanced iron dissolution rate. The new values are compared to those published in t he literature and discussed. Moreover the modeling all ows reproducing the effects of oxygen concentration , and sodium velocity which were varied in the experiment s.

Highly reactive and nanosized Th1-x Yx O2-x/2 or Ce0.8 Ln0.2 O1.9 mixed oxides were prepared through the initial precipitation of hydroxide precursors which were further dried under vacuum. Whatever the chemical system investigated, the characterization of the powdered samples evidenced a rapid aging process leading to hydrated oxides. The thermal behavior of these samples was further investigated and first showed a two-step dehydration process, with the successive departure of adsorbed and constitutive water, both yielding a drastic drop of the powders' reactivity (i.e. decrease of the specific surface area). Sintering experiments were then undertaken by starting directly from raw powders and revealed very rapid densification kinetics. Highly densified pellets (above 95 %TD) with a fine grain microstructure were obtained after only 1 hour of heat treatment at 1600 °C. This easy and versatile process of precipitation, that can be followed by direct densification of the powders, then appears as a promising option for the elaboration of homogenous ceramic electrolytes.

The French Atomic and alternatives Energy Commission (CEA) developed an effective powder capable of extinguishing sodium fire. However, ever since the dismantling of several old sodium installations facilities, CEA has at its disposal an important stock of unused powder batches, some of which have not been kept in suitable conditions. The idea of reutilizing these powders initiates the question about their efficiency to extinguish a sodium fire after long term storage. The various physicochemical analyses that have been carried out to characterize these stocked powder batches highlight the presence of lithium sodium carbonate (LiNaCO3), which is not a product of aging, and trona that, contrarily, formed because of aging during storage. This study presents the first experimental results of small sodium pool fire extinction using the powder samples previously characterized. The presence of trna did not affect significantly the extinction. Moreover, it did not affect the fusion temperature of the powders due to its decomposition at low temperature. On the contrary, the powder having high percentage of LiNaCO3 encountered a difficulty in extinguishing the fire. This happened because of their difficulty to flow, which enhanced by the external condition of the tests. These first results show that this facility allows us to test the extinguishing powders on sodium fires, with temperature monitoring at different positions and video recording. Improvements suggested from these preliminary tests should permit to discriminate the extinguishing efficiency of the powders as a function of their chemical composition.

In sodium fast reactors (SFR), dissolved oxygen in sodium can be monitored via potentiometric sensors with an yttria-doped thoria electrolyte. Yttria-doped ceria (YDC) was chosen as a surrogate material to validate the process of such sensors. The material must exhibit high density and a fine grain microstructure to be resistant to the corrosion by liquid sodium and thermal shocks. Thus, the oxalic co-precipitation route was chosen to avoid milling steps that could bring impurity incorporation which is suspected to induce grain boundary corrosion in sodium. The powder and sintered pellets examination show that the synthesis conditions are of primary importance on the process yield, the oxalate powder microstructure and, eventually, on the ceramic density and microstructure. The impurity content was limited by controlling the synthesis, calcination, and sintering steps.

The preparation of Th1-xYxO2-x/2 ceramics, to be used as electrolyte in oxygen sensors for sodium-cooled nuclear reactors, was successfully undertaken from oxalate precursors. Such method was found to provide quantitative precipitation of the cations into (Th, Y)(C2O4)(2)center dot 2H(2)O solid solutions up to x = 0.15 while a polyphase system was obtained for x = 0.22. The corresponding oxides were obtained through heat treatment in air at 500 degrees C and characterized by the means of PXRD, SEM and statistical X-EDS measurements. The conditions for the densification of Th1-xYxO2-x/2 ceramics were further determined by dilatometry (T = 1575 degrees C, t = 8 h) resulting in densification rates up to 99%. Finally, a first estimation of the electrical properties of the solids was undertaken by impedance spectroscopy. Electrical conductivity was found to increase linearly with the incorporation of Y3+ content while the associated values of activation energy decreased, with a minimum value of 1.1 eV for Th0.85Y0.15O1.925. (C) 2016 Elsevier B.V. All rights reserved.

A code named OSCAR-Na has been developed to calculate the mass transfer of corrosion products in the primary circuit of sodium fast reactors (SFR). It is based on a solution/precipitation model, including diffusion in the steel (enhanced under irradiation), diffusion through the sodium boundary layer, equilibrium concentration of each element, and velocity of the interface (bulk corrosion or deposition). The code uses a numerical method for solving the diffusion equation in the steel and the complete mass balance in sodium for all elements. Corrosion and deposition rates are mainly determined by the iron equilibrium concentration in sodium and its oxygen-enhanced dissolution rate. All parameters of the model have been assessed from a literature review, but iron solubility had to be adjusted. A simplified primary system description of PHENIX French SFR was able to assess the correct amounts and profiles of contamination on heat exchanger surfaces for the main radionuclides.