Determination of mass attenuation coefficients of Th, U, Np and Pu for oxygen Kα x rays using an electron microprobe

Mass attenuation coefficients (MACs) of Th, U, Np, and Pu for oxygen X-rays have been experimentally determined using an electron microprobe. The MACs were obtained by measuring relative X-ray intensities emitted from ThO2, UO2, NpO2, and PuO2 targets, for incident electron energies from 5 to 30 keV, and processing them with the help of the computer program XMAC. The accuracy of the measured MACs is estimated to be better than 5%. Results are compared with MAC tabulations commonly used in electron probe microanalysis as well as with theoretical photoionization calculations. It is concluded that the MACs implemented in the Monte Carlo simulation program PENELOPE which are based on the photoionization cross-section calculations of Sabbatucci & Salvat [(2016). Theory and calculation of the atomic photoeffect. Rad Phys Chem121, 122–140], provide the best agreement with our measurements. The use of different MAC schemes for the analysis of mixed actinide oxide materials is discussed.JRC.G.I.3-Nuclear Fuel Safet

Study of the redistribution of U, Zr, Nb, Tc, Mo, Ru, Fe, Cr, and Ni between oxide and metallic phases in the matrix of a multiphase Chernobyl hot-particle extracted from a soil sample of the Western Plume

A “hot particle” found 6 km west of the Chernobyl nuclear power plant 4 years after the Chernobyl severe nuclear accident was analysed by scanning electron microscopy and electron probe micro-analysis. The matrix of the particle consists of relics of partly molten UO2 nuclear fuel and two different phases of solidified U–Zr–O melt (U0.77Zr0.23O2 and U0.67Zr0.33O2). The particle also contains a unique metallic inclusion of a size of 30×22 μm. The inclusion is non-homogeneous and in some parts shows a dendrite-like structure. It consists of about 38 wt.% Fe, about 10 wt.% U, Mo, and Nb, about 5 wt.% Ru, Zr, Ni, and Cr, and small amounts of Tc (2 wt.%) and Si (0.4 wt.%). The presence of partly molten nuclear fuel suggests a local temperature exceeding 2850 °C. The metallic inclusion most likely formed when steel, fuel, and cladding reacted together and molten steel incorporated U, Zr, Nb, Tc, Mo, and Ru from molten fuel and cladding during a very fast high-temperature process. Fast quenching of the metallic and the oxide melt left no time for Tc and Mo to evaporate. Molten Zr was partly oxidised and acted as a buffer for O which caused the reduction of a fraction of the U. The data of this study support the previously reported supercritical nature of the Chernobyl explosion.JRC.G.I.3-Nuclear Fuel Safet

Study of a “hot” particle with a matrix of U-bearing metallic Zr: Clue to supercriticality during the Chernobyl nuclear accident

This paper is dedicated to the 30th anniversary of the severe nuclear accident that occurred at the Chernobyl NPP on 26 April 1986. A detailed study on a Chernobyl "hot" particle collected from contaminated soil was performed. Optical and electron microscopy, as well as quantitative x-ray microbeam analysis methods were used to determine the properties of the sample. The results show that the particle (~ 240 x 165 μm) consists of a metallic Zr matrix containing 2-3 wt. % U and bearing veins of an U,Nb admixture. The metallic Zr matrix contains two phases with different amounts of O with the atomic proportions (U,Zr,Nb)0.73O0.27 and (U,Zr,Nb)0.61O0.39. The results confirm the interaction between UO2 fuel and zircaloy cladding in the reactor core. To explain the process of formation of the particle, its properties are compared to laboratory experiments. Because of the metallic nature of the particle it is concluded that it must have formed during a very high temperature (> 2400 °C) process that lasted for only a very short time (few microseconds or less); otherwise the particle should have been oxidised. Such a rapid very high temperature process indicates that at least part of the reactor core could have been supercritical prior to an explosion as it was previously suggested in the literature.JRC.G.I.3-Nuclear Fuel Safet

High-Temperature Heat Capacity of Gd-pyrochlore (Gd2Ti2O7)

The heat capacity of gadolinium pyrochlore (Gd2Ti2O7) was measured by enthalpy increment measurements in the range of (585 to 1485) K utilizing drop calorimetry on a sinthtic Gd-pyrochlore and the high-temperature dependence of the heat capacity has been determined.JRC.E.3-Materials researc

Contrasting immobilization behavior of Cs+ and Sr2+ cations into a titanosilicate matrix

ETS-10 titanosilicate was tested as an adsorbent for the removal of Cs+ and Sr2+ cations from radioactive waters considering both, the ion exchange and the behavior of the loaded adsorbent during thermal conditioning. The studies indicate that ETS -10 has a high and similar affinity for both Cs+ and Sr2+ cations reaching the IECETS-10 at a concentration of about 50 meq/L. Thermal treatment of Cs+- and Sr2+-exchanged ETS-10 materials results in melting at approximately 700°C. The temperature for the melting increases with the initial Cs+ and Sr2+ concentration and is higher for Sr 2+ than for Cs+ exchanged ETS-10. Recrystallization occurs only in the presence of Sr2+ as evidenced by the exothermic effects between 800 and 900 °C. After calcination of Cs+- and Sr2+-exchanged ETS-10 in air at 800 °C two types of materials were obtained: an amorphous glass material with homogeneous Cs+ distribution and a strontium fresnoites glass-ceramic material.JRC.DG.E.2-Hot cell

Microbeam analysis of irradiated nuclear fuel

Microbeam analysis is widely used in the nuclear power industry. It is used to characterise the as-fabricated fuel, for routine post-irradiated examination and for research into the mechanisms of phenomena that limit the energy production of the fuel. The techniques most commonly used are wavelength-dispersive electron probe microanalysis, scanning electron microscopy and secondary ion mass spectrometry. Other microbeam analysis techniques that have been successfully applied to irradiated nuclear fuel are transmission and replica electron microscopy, X-ray fluorescence and micro X.-ray diffraction. Specific examples illustrating the past and present use of microbeam analysis in nuclear research establishments are presented with emphasis on the unique results they provide. As an aid to understanding, some basic facts about nuclear fuel rods and their irradiation are first given. This is followed by a description of features that set apart the microbeam analysis of high radioactive materials from standard practice.JRC.E.2-Hot cell

Inert Matrix Fuel

In a strict sense the term inert matrix fuel (IMF) refers to any nuclear fuel containing a low activation matrix as carrier for the fissile material. Since the early days of nuclear technology, this idea has been investigated, originally with the goal to improve fuel properties or to save uranium resources. However, currently, the term IMF is strongly associated with plutonium fuel that does not contain uranium to obtain the highest efficiency for destruction of excess plutonium (separated civil plutonium or plutonium from dismantled weapons) in a single irradiation campaign. This chapter focuses on this application. The term IMF is also used in the context of uranium-free fuels for transmutation of minor actinides (MA), although in many cases this is not appropriate as the fissile content is too low for fuel purposes and it is better to call this type of materials targets. Transmutation fuels and targets are described in detail in 3.05, Actinide Bearing Fuels and Transmutation Targets. A variety of materials has been proposed as inert matrix for nuclear fuel, ceramics such as MgO, ZrO2, or CeO2, refractory materials such as graphite or SiC, and metals such as stainless steel, zirconium, or molybdenum, depending on the type of application and the type of reactor foreseen. In thermal reactors, only a few matrices can be envisaged as a result of the very tight neutron economy. MgO and ZrO2 have been the focus of research during many years. This chapter principally deals with IMFs based on these materials. In this context, one can distinguish two fuel types: solid solutions (SS), such as the zirconia-based fuels ((Zr,Y,Pu)O2x), and composite fuels such as ((Zr,Y,Pu)O2x (ss)þMgO (matrix)). The composite fuels can be split into two groups: microdispersion fuels, which have usually fissile inclusions smaller than 25 mm and macrodispersed fuels, with fissile inclusions larger than 100 mm. The main difference is that in irradiated microdispersion or SS type of fuels the fission fragment damage is distributed homogeneously in the fuel matrix, while in the macrodispersed fuel the fission fragment damage is concentrated in a small volume shell around the fissile inclusions. All designs (SS, micro, or macro) have their advantages and disadvantages. Although some aspects of the composite fuels will be addressed here, the reader is referred to 3.10, Composite Fuel (cermet, cercer) for a comprehensive discussion of composite fuels.JRC.E-Institute for Transuranium Elements (Karlsruhe

Corium formation from reactor components and how its properties affect later stages of a severe nuclear accident.

The early stages of a severe accident are characterised by a loss of cooling capability in the reactor core and subsequent rise in temperature of the fuel & cladding to well in excess of typical operating temperatures. This results in various interactions and reactions between the UO2 fuel, Zircaloy cladding, moderator and reactor internal structures to produce a high temperature aggressive liquid, or corium, capable of dissolving the remaining reactor internal materials (in-vessel scenario), or even penetrate the reactor vessel wall and react with the basemat of multiphase concrete (ex-vessel scenario). Research on these various and complex high-temperature systems is clearly needed on a broad front. This includes both the determination of single compound thermodynamic properties and large-scale integral tests to test complex corium phases. Experimental data can support and validate modelling of properties of simpler and more complex systems. At JRC-ITU small scale testing of typical high temperature systems for the in-vessel scenario, as well as examination of irradiated fuel/cladding interactions has been and is carried out in the context of the main international research programmes on severe accidents. This paper includes data on the UO2-PUO2 phase diagram and the U-Zr-Fe-O system melting ranges. These will be compared with other results of corium structure analysis stemming from major irradiated fuel examinations such as Phébus PF project or the TMI-2 post-accident investigation programmes. This comparison can help elucidate the underlying mechanisms of the key reactions occurring during a severe accident in a nuclear power plant, enable better prediction of the likely progression and outcome of the event, and also help improve the overall accuracy of the severe accident codes.JRC.E.2-Safety of Irradiated Nuclear Material

Electron Microprobe Examination of Metallic Fuel for Minor Actinides Transmutation in Fast Reactor

Advanced nuclear reactors and closed nuclear fuel cycles are being considered worldwide as an important option to ensure sustainable nuclear energy supplies to satisfy future demands while minimizing the long-term radiotoxicity of high level waste. Spent fuel reprocessing and the subsequent recycling of Pu as fuel and transmutation of Minor Actinides (MA) Np, Am, Cm in fast reactors are necessary steps to achieve this goal. The METAPHIX programme is a collaboration between the Central Research Institute of Electric Power Industry (CRIEPI, Japan) and the Institute for Transuranium Elements (ITU, a Joint Research Centre of the European Commission) with the support of the Commissariat à l'Energie Atomique et aux Energies Alternatives (CEA, France) devoted to the study of MA-containing fast reactor metal fuels. The final objective of this project is to investigate safety and effectiveness of a closed nuclear fuel cycle based on MA separation and irradiation in metallic fuel using a fast reactor. In this frame, three assemblies containing nine Na-bonded experimental pins of metallic alloy fuel prepared at ITU were loaded in the Phenix reactor in 2003 and irradiated for up to 11 cycles, corresponding to a maximum burn-up of ~10 at%. This paper presents results of the first Electron Probe Micro Analysis (EPMA) performed on a ternary fuel sample.JRC.E.2-Hot cell

Fission product behaviour in minor actinide-bearing metal fuel

EDS analysis was performed on irradiated MA-bearing metal fuel samples. Fuel alloy is segregated into a few phases of different Zr peak intensity. RE-rich precipitates include Zr, Y, Mo, MAs and/or noble metal elements. Single Zr phase is formed in the periphery of the fuel.JRC.E.2-Safety of Irradiated Nuclear Material