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First assessment of a digestion method applied to recover plutonium from refractory residues after dissolving spent SFR MOX fuel in nitric acid

First assessment of a digestion method applied to recover plutonium from refractory residues after dissolving spent SFR MOX fuel in nitric acid

Abstract: Given the initial plutonium content in SFR MOX fuel, its quantitative recovery is one of the main objectives for the global dissolution process. A silver (II) oxidative digestion step was studied in the Atalante facility to assess its possible application to the treatment of undissolved residues obtained after dissolving spent SFR MOX fuel in nitric acid. This process was first optimized on plutonium dioxide powders, and then pretested on irradiated LWR MOX fuel residues. The digestion step permitted the recovery of up to 99% of the residual plutonium contained in the undissolved tiny particles, with some slight differences depending on the position of the pin part from which the dissolution residue was obtained. Local burnup and chemical composition were found to be influential. The quantity of residues after digestion was also significantly reduced thanks to this complementary treatment, with the final residues consisting mainly of metallic compounds including ruthenium, molybdenum, rhodium or palladium.
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First assessment of a digestion method applied to recovery of plutonium from refractory residues after dissolving spent sfr mox fuel in nitric acid

First assessment of a digestion method applied to recovery of plutonium from refractory residues after dissolving spent sfr mox fuel in nitric acid

4. Conclusion Higher plutonium concentrations in SFR MOX fuel require specific process steps to prevent plutonium retention in residues at the front end of the recycling cycle. More aggressive dissolution conditions can lead to cladding corrosion with consequences on the waste to deal with downstream. A complementary digestion method was first tested on plutonium dioxide powder, then validated on irradiated LWR MOX fuel residues and applied to solid residues obtained after dissolving irradiated pin parts from a PHENIX NESTOR-3 assembly in nitric media.
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Overview of French RandD on SFR MOX fuel fabrication

Overview of French RandD on SFR MOX fuel fabrication

manufacturing of MOX pellets for one fuel bundle, designed for a prototypical irradiation. IV.A. Main results and lessons learnt A first test was held at MELOX in 2015 to study of the effect of pelletizing conditions, the thermal sintering cycle (carried out under Ar-H2 (4%v) humidified at 360 vpm H2O) and the type of recycled scraps (2h grinded sintered pellets) coming from LWR or SFR type fuel. The first batch meets almost all the requirements including the plutonium distribution. Optimized sintering treatment, comprising a sharp humidification cycle, have been studied and performed to avoid any defects in the microstructure and obtain the specified O/M ratio. The incentive of the 2016 test was to study the influence of manufacturing parameters such as grinding on a larger scale (on the production mill) and the robustness of sintering treatment. The pore spectra obtained from the microstructure characterization range between 1 to a few tens of μm. Dense islands or porosity have been observed, similar to those observed in the pellets of the 1999 Phenix campaign of ATPu. The Pu distribution meets the requirement whatever is the sintering conditions. However, over-milled type microstructure have been sometimes observed due to an excess of energy released in the powder in the production mill.
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A comprehensive study of the dissolution of spent sfr mox fuel in boiling nitric acid (the phenix nestor-3 tests)

A comprehensive study of the dissolution of spent sfr mox fuel in boiling nitric acid (the phenix nestor-3 tests)

Three dissolution experiments were carried out on 30 mm long pieces of irradiated materials after shearing three distinct 120 mm sections of a (U,Pu)O2 fissile NESTOR-3 pin (bottom, middle i.e. full-flux zone, and upper). Dissolution experiments were carried out in boiling nitric acid (8 M) for 6 hours to produce a feed solution concentrated to about 180 g/L of HM. The results showed that the undissolved Pu rate is between 0.09% for the full-flux zone portion and 0.21% for the upper zone, the least irradiated. These results are in accordance with the feedback on SFR fuel treatment, i.e. an undissolved Pu rate value on average close to 0.1%.
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Study of irradiated and non-irradiated MOX fuel reprocessing

Study of irradiated and non-irradiated MOX fuel reprocessing

[2] N. Reynier-Tronche, E. Buravand, E. Esbelin, L. Huyghe, B. Catanese, S. Grandjean, Pu dissolution yield of a spent SFR MOX fuel as a function of axial position in the reactor (PHENIX NESTOR-3 tests), Plutonium Future September 2018, San Diego, USA. [3] E. Buravand, N. Reynier-Tronche, B. Catanese, P. Huot, L. Huyghe, E. Esbelin, B. Arab-

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NEA SFR subassembly benchmark sensitivity/uncertainty propagation with depletion

NEA SFR subassembly benchmark sensitivity/uncertainty propagation with depletion

mastered. Thus the confidence level of the safety demonstration relies on best-estimate methods as well as exhaustive knowledge of the associated uncertainties. Within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of the OECD/NEA, an international collaboration was conducted in 2009 (SFR-Task Force expert group) to assess the core performance characteristics and reactivity feedback coefficients of several Sodium-cooled Fast Reactor (SFR) concepts with various fuel forms such as oxide, carbide and metal alloy. Four numerical benchmark cases were initially developed with large and medium core sizes in order to perform:
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ZPR Core Representativity of SFR Reactivity Effects During Core Meltdown

ZPR Core Representativity of SFR Reactivity Effects During Core Meltdown

With the development of the ASTRID sodium fast reactor (SFR) industrial demonstrator, current core design, such as the low void fraction CFV desing, presents particularities that potentially deserve an experimetal validation, in terms of core physics [1]. In the current project several accidental scenarios have been investigated in the CFV design context, in particular separated degraded upper and lower fissile columns, followed by fuel crossing the fertile zone to form a large fissile melted pool. Large uncertainties on the behavior of the fuel could lead to local power excursions. These phenomenona can be experimentally studied in Zero Power Reactors (ZPRs), such as MASURCA fast critical assembly (to start again in 2019), or in the future ZEPHYR facility [2] (to be build at CEA Cadarache around 2028). The main idea is to utilize the available MASURCA fuel stockpile (rodlets and plates) to model the various states of SFR degraded core (MOX fuel and sodium for nominal state and metal fuel for various stages of SCA). However, as SCA occurs at high temperatures, the experimental facility of a ZPR is operated at room condition. Therefore, an innovative methodology must be developed to ensure proper representativity between hypothetical case and
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Dependence of transuranic content in spent fuel on fuel burnup

Dependence of transuranic content in spent fuel on fuel burnup

For the analysis in this work, the Studsvik Core Management System (CMS) code CASMO-4 is utilized to numerically analyze the transuranic content in spent fuel of particular burnu[r]

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Sodium Fast Reactor: an Asset for a PWR UOX/MOX Fleet

Sodium Fast Reactor: an Asset for a PWR UOX/MOX Fleet

IV.A. Efficiency and kinetic of Pu multirecycling in SFR Before studying the Pu multirecycling in both LWR and SFR, a preliminary study was conducted to evaluate the efficiency and the kinetics of Pu multirecycling in SFR. To do so, Pu vectors with different isotopic contents have been selected from the previous scenario and irradiated several time in a 600MWe SFR. Each time the Pu is irradiated up to 87GWd/t during 4x400 EFPD. The different Pu vectors come from the scenario 2 results and correspond to Pu in UOX, MOX 1, MOX 2 and MOX 3 spent fuels. They are detailed in TABLE III . The beginning date of this scenario, without any relevance here, is arbitrarily fixed at 2020.
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Americium transmutation in a scenario of progressive SFR deployment

Americium transmutation in a scenario of progressive SFR deployment

Reactors and fuel cycle assumptions The current French fleet is composed of 58 PWR, loaded with UOX fuels, ERU (Enriched Reprocessed Uranium) fuels and MOX fuels (a third of the fleet is loaded with 30% of MOX fuels). Those 58 PWR are commissioned between 1978 and 2002 and will be decommissioned until 2062. EPRs are commissioned at a rate of 2 EPR every 1.5 years up to 2051 and then 1 EPR every 1.5 years. The CFV V1 core concept [9], developed by the CEA, is the one considered for the SFR fleet. For the SFR cores, two power levels are considered: 1GWe and 1.45GWe. Both EPR and CFV cores characteristics are detailed in Table 1.
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Nuclear data propagation with burnup : impact on SFR reactivity coefficients

Nuclear data propagation with burnup : impact on SFR reactivity coefficients

sensitivities relative to cross sections. It was widely used in the pass to assess uncertainty level and to identify main contributors (isotopes and relevant cross sections) in order to design dedicated experiments, to improve measurements and/or develop theoretical nuclear models. As the most penalizing core configuration involves irradiated fuel and uncertainty propagation with burnup is needed to get a comprehensive handling of the problematic. To do so, the coupled Boltzman/Bateman sensitivity approach has been implemented in the ERANOS code package used in order to reach this goal. It enables us to identify the impact of the irradiation process, its main changes and contributors and to estimate the global change on reactivity effects such as feedback coefficients.
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Impact of the thermal scattering law of H in H$_2$O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

Impact of the thermal scattering law of H in H$_2$O on the isothermal temperature reactivity coefficients for UOX and MOX fuel lattices in cold operating conditions

3 Interpretation of the MISTRAL programs with the Monte-Carlo TRIPOLI4 ® 3.1 Description of the MISTRAL con figurations The MISTRAL experimental programs were designed in the late nineties to evaluate the feasibility of using 100% MOX fuel in light water reactors. The different core con figurations were tested in the EOLE reactor of CEA Cadarache (France). Many relevant neutronic parameters were mea- sured during the MISTRAL programs such as critical mass, geometrical buckling, spectral indices, conversion factor, isothermal temperature coef ficient, single absorber worth, soluble boron worth and effective delayed neutron fraction. The present work focuses on the isothermal temperature reactivity coef ficient measured in the MISTRAL-1, MISTRAL-2 and MISTRAL-3 con figurations ( Fig. 2 ). A detailed description of the experiments can be found in reference [16] .
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EVALUATION OF ALTERNATIVE FLUIDS FOR SFR INTERMEDIATE LOOPS

EVALUATION OF ALTERNATIVE FLUIDS FOR SFR INTERMEDIATE LOOPS

Interaction with structures The interactions between the fluid and the material of the circuits and components lead to particular choices for the material. This item includes general corrosion (oxidation or dissolution), localized corrosion (grain boundary corrosion, liquid metal embrittlement…), mass transfer (plugging, fouling), mechanical properties and long term modelling. The general corrosion must not induce an unacceptable loss of the mechanical or heat transfer properties of the material. For example, if an oxide is formed on the surface of the material, the increase in its thickness during the lifespan of the reactor should not induce a decrease in the thermal transfer more than about 20%. In case of mass transfer, the quantity of corrosion products should not reasonably exceed dozens of kilograms per year (equivalent of the maximum expected in the primary circuits of the SFR). The best known corrosion resistant “industrial” material for each fluid has been selected for the evaluation. As they vary from one fluid to another, the mechanical properties of the materials were also compared, as well as their resistance to water or sodium corrosion (respectively in the steam generator SG or in the intermediate heat exchanger IHX). The availability of the reactor can be affected by the corrosion, the mass transfer or the loss of mechanical properties.
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Modélisation des effets de l'endommagement dans les milieux hétérogènes viscoélastiques - simulation du comportement des combustibles MOX.

Modélisation des effets de l'endommagement dans les milieux hétérogènes viscoélastiques - simulation du comportement des combustibles MOX.

 Le MOX est un combustible utilisé dans les Réacteurs à eau pressurisée français (REP) de 900 MWe.  C’est un matériau hétérogène au sens de la distribution des teneurs en plutonium : matériau triphasé (thèse Oudinet, 2003).

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Méthodes de traitement du signal pour l'analyse quantitative de gaz respiratoires à partir d’un unique capteur MOX

Méthodes de traitement du signal pour l'analyse quantitative de gaz respiratoires à partir d’un unique capteur MOX

4.6. Bilan En conclusion, nous arrivons à inverser le modèle et à estimer les concentrations d’un mélange de deux gaz de manière supervisée. Ce travail d’estimation des concentrations de gaz dans un mélange permet aussi d’explorer les performances des méthodes. Dans ce cas supervisé, c’est-à-dire lorsque les coefficients du modèle de mélange sont connus au moment de l’estimation des sources, les concentrations d’acétone et d’éthanol dans un mélange de deux gaz sont du bon ordre de grandeur, avec une erreur faible, puisque le rapport signal sur bruit est d’environ 26 dB sur données simulées. Grâce à ce modèle, l’inversion et l’estimation des concentrations de manière supervisée à partir des données expérimentales se fait correctement, preuve que le modèle quadratique est adapté à la réponse des capteurs MOX. Ce travail permet alors de valider le modèle linéaire quadratique. Cependant, à cause de problèmes d’inversibilité dont nous avons discuté, des contraintes sont à ajouter pour l’estimation sur l’ensemble du domaine de concentrations. En effet, sur données simulées, pour des concentrations supérieures à 6 ppm, l’inversion se déroule très bien. Pour des concentrations plus faibles, l’instabilité augmente car on recherche des concentrations plus faibles devant le niveau de bruit des mesures. L’indétermination augmente également comme le montre l’analyse de la diversité sur cette zone de concentrations (Figure 3.10). De ce fait, il est nécessaire d’ajouter une régularisation. Aussi, comparé aux autres modèles, notamment le modèle bilinéaire, l’estimation des concentrations est meilleure, avec un RSB 3 fois plus élevé à faibles concentrations. De plus, ce travail valide également le protocole d’expérimentation et notamment le mode de fonctionnement des capteurs. Le modèle déterminé avec un seul capteur MOX permet effectivement de bien différencier deux gaz dans un mélange, grâce au mode de fonctionnement à deux températures.
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Development of a continuous O/M measurement technics during sintering for MOX application

Development of a continuous O/M measurement technics during sintering for MOX application

4. Conclusions and outlook By coupling a dilatometer and a zirconia probe, it was possible to identify the different redox phenomena and to assess the evolution of the O/M ratio of the oxides at each time of the densification process. It was shown that under certain conditions, the thermodynamic equilibrium between gas and fuel is long to be reached so that it can be difficult to control and predict O/M ratio, especially for SFR- type fuel (sintered in reductive conditions). Thanks to this equipment, it is possible to recommend the best sintering atmosphere and thermal cycle to obtain a high density and an O/M ratio close to the target value.
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Validation of mox core reactivity versus flux variation feedback on delayed neutron data

Validation of mox core reactivity versus flux variation feedback on delayed neutron data

ABSTRACT The reactivity of commercial Light Water Reactors (LWR) is given by the reactimeter, which computes the core reactivity from the flux variation measurement. The correspondence is obtained from the kinetics equations with a large 6-10% uncertainty due to the delayed neutron (DN) data. This paper presents the first direct validation of the relationship between the measured flux variation and the MOX core reactivity, thanks to the boron reactivity worth measurements in EOLE-MISTRAL2 experiment. The kinetics relationship based on JEFF3 DN data is satisfactory. On the contrary, using B-VII DN data leads to an under-prediction by -8.3% ± 2.7% of MOX core reactivity. Considering the satisfactory prediction within 2% of the measured β eff value, we can
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MOx Powders Calculation ImprovementCriticality Calculations in the context of OECD NEA benchmark

MOx Powders Calculation ImprovementCriticality Calculations in the context of OECD NEA benchmark

Nine cases are proposed with three PuO 2 contents, and three Pu isotopic vectors (criticality-conservative, realistic, and Pu weapon-grade). Six additional cases are also proposed (with different powder moisture rates). For those fifteen cases, the MOx fissile medium is a sphere surrounded by a 20cm-large water reflector. For criticality studies, one should notice that a reflected sphere is the most penalizing configuration (reduced leakage).

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In-situ experiments at elevated temperatures -an invaluable aid in studying MOX

In-situ experiments at elevated temperatures -an invaluable aid in studying MOX

Les 15 èmes Journées Scientifiques de Marcoule 9 – 10 juin 2015 Document propriété du CEA – Reproduction et diffusion externes au CEA soumises à l’autorisation de l’émetteur In-situ experiments at elevated temperatures - an invaluable aid in studying MOX.

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Modeling of fuel permeation in multilayer automotive plastic fuel tanks

Modeling of fuel permeation in multilayer automotive plastic fuel tanks

/ La version de cette publication peut être l’une des suivantes : la version prépublication de l’auteur, la version acceptée du manuscrit ou la version de l’éditeur. For the publisher’[r]

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