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The fission product release of two irradiated mixed oxide (MOX) fuel variants from the high burn-up disk irradiation test IFA-655 in the Halden reactor was investigated by Knudsen Effusion Mass Spectrometry (KEMS). The fresh samples were provided and characterised within the NFIR 1 1 NFIR: https://www.epri.com/portfolio/programs/108031 program. Both fresh fuel variants contained 16 wt.% plutonium but differed principally by the distribution of the plutonium: in one it was heterogeneously dispersed, in the other one homogeneously. The fission products release (Kr, Xe, I, Cs, Te, Sr, Nd) measured by KEMS as a function of annealing temperature is reported. The gas release (including helium) has been measured quantitatively. The interpretation of the release measurements is supported by microstructure and elemental analyses (SEM, TEM, EPMA). A correlation has been made with results from thermal diffusivity measurements obtained on the same fuels. [ABSTRACT FROM AUTHOR]

UO2 samples doped with 6, 11, 22 mol% lanthanum were examined before and after air oxidation. To verify the formation of uranium-lanthanum-mixed oxide solid solutions, powder X-ray diffraction ( XRD) analyses of the crystalline phases in the materials were carried out. The presence of oxygen vacancies in the La-doped UO2 samples was identified by Raman spectrometry. It was evidenced by changes induced in the Raman spectra by air oxidation. This latter was carried out either by increasing the Raman laser power or by thermally treating the samples at 500 K for 370 h. In addition, oxidation behavior differences of pure and La-doped UO2 samples were reported by comparing XRD and Raman results of the samples before and after air oxidation. It was shown that the concentration of the M4O9 (M: U, La) phase increased with increasing content of La, whereas inhibition for the formation of M3O8 phase was observed. [ABSTRACT FROM AUTHOR]

The author discusses the study on the release of fission products from the fuel during a nuclear reactor accident. Tackled is the release of the radioactive gases in the environment as well as the destruction of the reactor cores through the Three Mile Island accident in Pennsylvania. Also explored are health and environmental effects during a nuclear accident.

Abstract: Significant amounts of spent uranium dioxide nuclear fuel are accumulating worldwide from decades of commercial nuclear power production. While such spent fuel is intended to be reprocessed or disposed in geologic repositories, out-of-reactor radiation damage from alpha decay can be detrimental to its structural stability. Here we report on an experimental study in which radiation damage in plutonium dioxide, uranium dioxide samples doped with short-lived alpha-emitters and urano-thorianite minerals have been characterized by XRD, transmission electron microscopy, thermal desorption spectrometry and hardness measurements to assess the long-term stability of spent nuclear fuel to substantial alpha-decay doses. Defect accumulation is predicted to result in swelling of the atomic structure and decrease in fracture toughness; whereas, the accumulation of helium will produce bubbles that result in much larger gaseous-induced swelling that substantially increases the stresses in the constrained spent fuel. Based on these results, the radiation-ageing of highly-aged spent nuclear fuel over more than 10,000years is predicted. [Copyright &y& Elsevier]

During its operating life in the core of a nuclear reactor nuclear fuel is subjected to significant restructuring processes determined by neutron irradiation directly through nuclear reactions and indirectly through the thermo-mechanical conditions established as a consequence of such reactions. In today''s light water reactors, starting after ∼4 years of operation the cylindrical UO2 fuel pellet undergoes a transformation that affects its outermost radial region. The discovery of a newly forming structure necessitated the answering of important questions concerning the safety of extended fuel operation and still today poses the fascinating scientific challenge of fully understanding the microstructural mechanisms responsible for its formation. [Copyright &y& Elsevier]

The transmutation of minor actinides (in particular, Np and Am), which are among the main contributors to spent fuel α-radiotoxicity, was studied in the SUPERFACT irradiation. Several types of transmutation UO2-based fuels were produced, differing by their minor actinide content (241Am, 237Np, Pu), and irradiated in the Phénix fast reactor. Due to the high content in rather short-lived alpha-decaying actinides, both the archive, but also the irradiated fuels, cumulated an alpha dose during a laboratory time scale, which is comparable to that of standard LWR fuels during centuries/millenaries of storage. Transmission Electron Microscopy was performed to assess the evolution of the microstructure of the SUPERFACT archive and irradiated fuel. This was compared to conventional irradiated spent fuel (i.e., after years of storage) and to other 238Pu-doped UO2 for which the equivalent storage time would span over centuries. It could be shown that the microstructure of these fluorites does not degrade significantly from low to very high alpha-damage doses, and that helium bubbles precipitate. [ABSTRACT FROM AUTHOR]

The article analyzes the emerging trend in the development of nuclear reactors using materials that can handle very high energy and particle fluxes. A recent symposium on nuclear materials dealt with ways of predicting the behavior of material when subjected to high irradiation, causes of damage to metals and storage of spent fuel. It concludes that the development of next generation nuclear reactors requires a synergy between experimental data and simulations based on a multi scale approach.

Abstract: Studying self-irradiated materials is an ideal means to investigate the effect of the damage on material structure and to better understand the behavior of irradiated nuclear fuels. In this context, X-ray diffraction, X-ray absorption spectroscopy and transmission electron microscopy have been used to investigate self-irradiation damaged AmO2. Combining these techniques allows studying the microstructure and the variation of the fluorite structure at both short-range and long-range order. Thus, the increase of both interatomic distances and lattice parameter was shown, as well as the presence of nanometer sized He bubbles and dislocation loops. As confirmed by the observed high-level of crystallinity, the fluorite structure exhibits a high radiation tolerance, which is confirmed by the low increase of the lattice parameter. This could be explained by a self-annealing mechanism of the created defects at room temperature. [Copyright &y& Elsevier]

We present here for the first time the production of 3‐ to 5‐nm–sized thoria nanopowders by decomposition of Th‐hydroxide under hot compressed water and their consolidation by high‐pressure spark plasma sintering into dense nanostructured ThO2 with grain size of 50 nm. [ABSTRACT FROM AUTHOR]

Oxalate precipitation is a powerful technique for actinide oxide preparation at either laboratory or industrial scales. In this study we focus on the low temperature decomposition of Th(C 2 O 4 ) 2 ⋅2H 2 O and (N 2 H 5 ) 2 U 2 (C 2 O 4 ) 5 ⋅ n H 2 O into nanograined ThO 2 and UO 2 powders, which will be used later as precursors for the generation of materials emulating the nuclear fuel high burn up structure (HBS). The evolution with temperature of the nanoparticles properties has been investigated using several solid state analytical techniques (transmission and scanning electron microscopy, room and high temperature powder X-ray diffraction, Raman spectroscopy, thermogravimetry). Oxide powders with a high degree of crystallinity and grain size from ∼10 nm were prepared. Their purity was examined with special focus on the presence of carbonates. It was found that carbonate content in the final powder did not exceed 3.4 wt.%, for the powder treated at 600 °C for 1 h. [ABSTRACT FROM AUTHOR]

The long-term behavior of high-level nuclear glass subjected to alpha/beta radiation by long-life radionuclides must be investigated with respect to geological disposal. This study focuses on the effects of alpha and beta radiations on the chemical reactivity of R7T7 glass with pure water, mainly on the residual alteration rate regime. Glass specimens doped with 0.85 wt% 239PuO2 (α emitter) and 0.24 wt% 99TcO2 (β emitter) that simulate alpha and beta dose rate corresponding to long-term disposal conditions are leached under static conditions in argon atmosphere at 90°C, in initially pure water and at a high surface-area-to-volume ratio (S/V = 20/cm). The alteration rate is monitored by the release of glass alteration tracer elements (B, Na, and Li). Radiation effects on the leached glass and its gel network are characterized by SEM and TEM analyses. Plutonium and technetium releases are also measured by radiometry, and their chemical oxidation state is assessed by measuring the pH and reduction-oxidation potential of the leachates. The results do not highlight any significant effect of alpha/beta radiation on the residual alteration of these doped glasses. These observations are consistent with solid characterizations, which show that a protective layer can be formed under such irradiation fields. Under our experimental conditions (Eh~380 mV/SHE, pH90°C = 8-8.6), very low concentrations of soluble plutonium are measured in the leachate, indicating strong plutonium retention, whereas technetium performs as a soluble element and is not retained in the altered layer. [ABSTRACT FROM AUTHOR]

Abstract: During in-pile, irradiation-induced damage occurs in nuclear fuel and results in a deterioration of its properties, which can affect the margin to melt of the fuel. Damage also occurs in fresh fuel through the self-irradiation process, and thus provides a convenient means to investigate changes in the material. A uranium–plutonium mixed nitride fuel made over 25years ago, and stored in the ITU archives has been retrieved. Coupling EXAFS and TEM has shown that this material was still well-crystallized. However, an increase of 0.3% of the lattice parameter was found. As shown by the EXAFS, the U–N and Pu–N as well as the metal–metal distances are similarly affected. However, no significant modification of both anion and cation sublattice was found. Although no defect clustering was found by EXAFS, the presence of nanometric helium bubbles was demonstrated by TEM as well as nanometric disordered domains. [Copyright &y& Elsevier]

Reducing the uncertainties associated with extrapolation to very long term of corrosion data obtainable from laboratory tests on a relatively young spent nuclear fuel is a formidable challenge. In a geologic repository, spent nuclear fuel may come in contact with water tens or hundreds of thousands of years after repository closure. The corrosion behavior will depend on the fuel properties and on the conditions characterizing the near field surrounding the spent fuel at the time of water contact. This paper summarizes the main conclusions drawn from multiyear experimental campaigns performed at JRC-ITU to study corrosion behavior and radionuclide release from spent light water reactor fuel. The radionuclide release from the central region of a fuel pellet is higher than that from the radial periphery, in spite of the higher burnup and the corresponding structural modifications occurring at the pellet rim during irradiation. Studies on the extent and time boundaries of the radiolytic enhancement of the spent fuel corrosion rate indicate that after tens or hundreds of thousands of years have elapsed, very small or no contribution to the enhanced corrosion rate has to be expected from a radiolysis. A beneficial effect inhibiting spent fuel corrosion due to the hydrogen overpressure generated in the near field by iron corrosion is confirmed. The results obtained so far point toward a benign picture describing spent fuel corrosion in a deep geologic repository. More work is ongoing to further reduce uncertainties and to obtain a full description of the expected corrosion behavior of spent fuel. [ABSTRACT FROM AUTHOR]

R7 T7-type glass samples doped with different concentrations of 244 Cm and stored for different time periods to produce different levels of accumulated radiation damage were analyzed by differential scanning calorimetry to better understand the structural evolution of R7T7 nuclear glass due to the accumulation of α radiation damage. The stored energy and fictive temperature were observed to increase as a function of the α decay dose. The stored energy saturates at about 1018 α decays·g−1, whereas the fictive temperature continues to increase and saturates at about 6-7 × 1018 α decays·g−1. This discrepancy was interpreted in terms of different effects of energy lost during α decay on the glass structure due to electronic stopping and nuclear stopping. This study demonstrates that the stored energy is affected by both phenomena, whereas the fictive temperature variation depends mainly on elastic collisions. The increase of fictive temperature with the cumulative α dose is direct evidence that the energy lost by recoil nuclei induces the formation of a new structure similar to a fast-quenched glass. [ABSTRACT FROM AUTHOR]

In this article the authors discuss the role of carbon nanotubes (CNTs) in the field of nuclear waste management. They argue that the qualities of CNTs, such as the ability of industries to produce them in mass quantities, advances in catalysis, and continuous growth processes make them a suitable and commercially viable product. They present a brief overview of nuclear waste management and an in-depth analysis of CNTs in the nuclear industry and research field.

Aerosols were produced from a MOX (Mixed Oxide of uranium and plutonium) sample by laser heating up to 4000 K under atmospheric conditions. They were collected with a cascade impactor as function of their aerodynamic size, and studied by SEM, EDS, TEM, EELS, SAED and ICPMS. The uranium and plutonium concentrations were related to the aerosol morphology, size and formation temperature. A clear decrease of the Plutonium concentration was observed as function of the Aerodynamic Equivalent Diameter (AED). Some bigger aerosols enriched in Plutonium were found which originated from solidification of ejected droplets of the liquid phase at the surface of the sample. • In sub micrometric aerosols produced from laser heated MOX the plutonium concentration decrease with aerosol size. • The specific properties of UO 2 and PuO 2 affects the formation of aerosols leading to elemental enrichment in the aerosols. • The collected EELS data on the smallest aerosols <20 nm shows that they are composed of pure uranium oxide. [ABSTRACT FROM AUTHOR]

Fundamental knowledge on intrinsic plutonium colloids is important for the prediction of plutonium behaviour in the geosphere and in engineered systems. The first synthetic route to obtain salt-free intrinsic plutonium colloids by ultrasonic treatment of PuO2 suspensions in pure water is reported. Kinetics showed that both chemical and mechanical effects of ultrasound contribute to the mechanism of Pu colloid formation. In the first stage, fragmentation of initial PuO2 particles provides larger surface contact between cavitation bubbles and solids. Furthermore, hydrogen formed during sonochemical water splitting enables reduction of Pu(IV) to more soluble Pu(III), which then re-oxidizes yielding Pu(IV) colloid. A comparative study of nanostructured PuO2 and Pu colloids produced by sonochemical and hydrolytic methods, has been conducted using HRTEM, Pu LIII-edge XAS, and O K-edge NEXAFS/STXM. Characterization of Pu colloids revealed a correlation between the number of Pu-O and Pu-Pu contacts and the atomic surface-to-volume ratio of the PuO2 nanoparticles. NEXAFS indicated that oxygen state in hydrolytic Pu colloid is influenced by hydrolysed Pu(IV) species to a greater extent than in sonochemical PuO2 nanoparticles. In general, hydrolytic and sonochemical Pu colloids can be described as core-shell nanoparticles composed of quasi-stoichiometric PuO2 cores and hydrolyzed Pu(IV) moieties at the surface shell. [ABSTRACT FROM AUTHOR]