This work provides an in-depth analysis of the extensive research and development activities on americium-based ceramics for space applications, particularly as heat source for radioisotope power generation. Our pioneering efforts focus on synthesizing and characterizing various americium ceramics with fluorite, monazite, perovskite, zircon, and pyrochlore structures, and assessing their potential for use in Radioisotope Power Systems (RPSs). This study identifies uranium-stabilised cubic americium oxide as the best candidate among the ceramic forms analysed, due to its superior stability and performance under extreme conditions relevant to space missions. The review emphasises the unique facilities and methodologies employed, including remote-handling techniques and advanced material characterization, to overcome the challenges posed by the high radiation dose and specific activity of Am-241 when working with gram quantities.

In search for chemically stable americium compounds with high power densities for radioisotope sources for space applications, AmVO3 and AmVO4 were prepared by a solid-state reaction. We present here their crystal structure at room temperature solved by powder X-ray diffraction combined with Rietveld refinement. Their thermal and self-irradiation stabilities have been studied. The oxidation states of americium were confirmed by the Am M5 edge high-resolution X-ray absorption near-edge structure (HR-XANES) technique. Such ceramics are investigated as potential power sources for space applications like radioisotope thermoelectric generators, and they have to endure extreme conditions including vacuum, high or low temperatures, and internal irradiation. Thus, their stability under self-irradiation and heat treatment in inert and oxidizing atmospheres was tested and discussed relative to other compounds with a high content of americium.

Americium (Am) is a strong contributor to the long-term radiotoxicity of high-level waste from nuclear fuels. Transmutation of long-lived nuclides like 241Am by irradiation in nuclear reactors is therefore an option for the reduction of radiotoxicity and heat production of waste volumes to be stored in a repository. The MARINE irradiation experiment is the latest in a series of European experiments on americium transmutation (e.g. EFTTRA-T4, EFTTRA-T4bis, HELIOS, MARIOS, SPHERE) performed in the High Flux Reactor (HFR) in Petten (The Netherlands). The development and irradiation of MARINE was carried out in the framework of the collaborative research project PELGRIMM of the EURATOM 7th Framework Programme (FP7). Dismantling was completed and post-irradiation examinations (PIE) were started within the Dutch national research programme PIONEER. Destructive PIE is foreseen within the Euratom H2020 funded project PATRICIA. The main objective of the MARINE experiment is to study the in-pile behaviour of uranium oxide fuel containing 13% of americium and to compare the behaviour of sphere-pac versus pellet fuel, in particular the role of microstructure and temperature on fission gas and helium release dynamics on fuel swelling. The MARINE experiment was irradiated for 359 Full Power Days in the HFR in 2016 and 2017. This paper discusses results from irradiation, i.e. power and temperature history and transmutation rates as well as preliminary results from post irradiation examinations, assessing a.o. clad strains and helium and fission gas release and first ceramographic observations, putting a preliminary upper bound on fuel swelling.

This paper details the first dedicated production of homogeneous nanocrystalline particles of mixed actinide oxide solid solutions containing americium. The target compositions were U0.75Pu0.20Am0.05O2, U0.90Am0.10O2 and U0.80Am0.20O2. After successful hydrothermal synthesis and chemical characterisation, the nanocrystals were sintered and their structure and behaviour under self-irradiation were studied by powder XRD. Cationic charge distribution of the as-prepared nanocrystalline and sintered U0.80Am0.20O2 materials was investigated applying U M4 and Am M5 edge high energy resolution XANES (HR-XANES). Typical oxidation states detected for the cations are U(iv)/U(v) and Am(iii)/Am(iv). The measured crystallographic swelling was systematically smaller for the as-synthesised nanoparticles than the sintered products. For sintered pellets, the maximal volumetric swelling was about 0.8% at saturation, in line with literature data for PuO2, AmO2, (U,Pu)O2 or (U,Am)O2.

The European Space Agency is funding the research and development of 241Am-bearing oxide-fuelled radioisotope power systems (RPSs) including radioisotope thermoelectric generators (RTGs) and European Large Heat Sources (ELHSs). The RPSs’ requirements include that the fuel’s maximum temperature, Tmax, must remain below its melting temperature. The current prospected fuel is (Am0.80U0.12Np0.06Pu0.02)O1.8. The fuel’s experimental heat capacity, Cp, is determined between 20 K and 1786 K based on direct low temperature heat capacity measurements and high temperature drop calorimetry measurements. The recommended high temperature equation is Cp(T/K) = 55.1189 + 3.46216 × 102 T − 4.58312 × 105 T−2 (valid up to 1786 K). The RTG/ELHS Tmax is estimated as a function of the fuel thermal conductivity, k, and the clad’s inner surface temperature, Ti cl, using a new analytical thermal model. Estimated bounds, based on conduction-only and radiation-only conditions between the fuel and clad, are established. Estimates for k (80–100% T.D.) are made using Cp, and estimates of thermal diffusivity and thermal expansion estimates of americium/uranium oxides. The lowest melting temperature of americium/uranium oxides is assumed. The lowest k estimates are assumed (80% T.D.). The highest estimated Tmax for a ‘standard operating’ RTG is 1120 K. A hypothetical scenario is investigated: an ELHS Ti cl = 1973K-the RPSs’ requirements’ maximum permitted temperature. Fuel melting will not occur.

Knudsen Effusion Mass Spectrometry is an important research technique to determine the partial pressures of vapour species at equilibrium over different materials at a given temperature. The vapour pressure is one of the fundamental properties of a material and needed for any modelling involving temperatures where it is not negligible. Many research groups worldwide use Knudsen Effusion mass spectrometry techniques to determine vapour pressures of different materials. They usually develop their instrument and software themselves. The software, which is presented in the frame of this paper, targets to provide a modular LabVIEW software framework for the control and data acquisition of Knudsen Effusion Mass Spectrometry instruments that can easily accommodate any type of instrumentation.

The current paper reports the results of gamma spectroscopic burn-up determination and KuFA safety testing at JRC Karlsruhe on spherical high-temperature reactor fuel elements, which were fabricated by the Institute of Nuclear and New Energy Technology of the Tsinghua University, Beijing. The fuel elements were irradiated in the High Flux Reactor, Petten, in the frame of the HFR-EU1 and HTR-PM campaigns and transported to JRC Karlsruhe for post-irradiation examination and accident testing in the KuFA device in the frame of a bilateral safety research study. Burn-up determination was performed on seven fuel elements based on the quantitative measurement of their Cs-137 inventories using an established gamma spectroscopy set-up in the JRC hot cell facilities. Accident testing was conducted on three HTR-PM fuel elements in several phases at simulated accident temperatures between 1620 degrees C and 1770 degrees C for 150 h to simulate hypothetical depressurization and loss-offorced circulation accidents. The release of Kr-85 was measured during the tests in a cold trap and solid fission product release was determined by gamma spectroscopic analyses of exchangeable cold plates in a low background environment. After successful completion of the KuFA tests the fuel elements are currently undergoing further post-irradiation examinations including profile disintegration as well as ceramography and scanning electron microscopy of individual coated particles.

AmPO4 was prepared by a solid-state reaction method, and its crystal structure at room temperature was solved by powder X-ray diffraction combined with Rietveld refinement. The purity of the monazite-like phase was confirmed by spectroscopic (high-resolution solid-state 31P NMR and Raman) and microscopic (SEM-EDX and TEM) techniques. The thermal and self-irradiation stability have been studied. The compound is stable under argon and air atmosphere at least up to 1773 K. It remains crystalline under self-irradiation for circa two months, with a crystallographic volume swelling of ∼1.5%, and then is amorphizing over a year. However, microcrystals are present in the amorphous material even after a two year period of time. All these characteristics are discussed in relation to the potential application of AmPO4 as a stable form of Am in radioisotope power sources for space exploration and of behavior of the monazites under irradiation.

Serial | Major Trends in Energy Policy and Nuclear Power Research in Support of European Radio isotope Power System Development at the European Commission’s Joint Research Centre in Karlsruhe ı Daniel Freis, Jean-François Vigier, Karin Popa and Rudy J.M. Konings S E R IA L | M A JO R T R E N D S I N E N E R G Y P O L IC Y A N D N U C L E A R P O W E R 19 8 Serial | Major Trends in Energy Policy and Nuclear Power

In some of the currently developed Sodium Fast Reactor (SFR) designs the core configuration will berearranged to include internal spent fuel storage positions, with the benefit of avoiding an externalsodium pool. However, in-vessel storage of failed fuel pins can lead to a direct contact betweencoolant and fuel, i.e. in a scenario where the cladding was breached, leading to their potential inter-action in case of oxide based fuel forms (UO2 and MOX). The reaction product of sodium and oxidefuel is generally denoted as Na3MO4 (where M=U, Pu), and characterized by unfavorable physical properties, which might result in fuel swelling and/or pulverization, with the consequence of fissileisotopes or fission products dissemination into the primary system. The understanding of the corro-sion mechanism and kinetics between liquid sodium and oxide based nuclear fuel becomes then ofprior importance for establishing the feasibility of an internal storage of failed fuel pins.In this contribution, we will present the results of out-of-pile tests that we perform to provide a basicknowledge of defective fuel pin behavior in contact with liquid sodium. With the aim to determinethe physical mechanism involved in the sodium-fuel reaction, we firstly focused on the behavior of UO2 corrosion by sodium.In order to cover the internal storage scenario, isothermal experiments were performed inside cap-sules with stagnant liquid sodium and UO2 at 800°C. To establish the influence of the grain orienta-tion on the growth process of the reaction product, well-oriented ( and ) single crystalswere used and finally to extend the behavior to the fuel pellet, a polycrystalline UO2 sample wastested as well. The corrosion product was analyzed by XRD, SEM-EDX and RAMAN. It consists ina homogeneous layer, whose morphology depends on the crystallographic orientation of the UO2 corroded grain or single crystal. However, the thickness of the corrosion layer does not seem todepend on the crystallographic orientation of the UO2 corroded grain that was unexpected.A tentative interpretation of UO2 corrosion by sodium is proposed. This interpretation will be usedfor further study on the modelling of MOX-sodium interaction. There the influence of Pu on the corrosion process will be addressed and added to the UO2-Na model.

The European Space Agency funded programme into the research and development of European radioisotope power systems (RPSs) began in 2008. Three RPS technologies are under development, namely, radioisotope heater units, radioisotope thermoelectric generators, and Stirling generators. Americium ( 241 Am) was selected as the `fuel', which provides radiogenic heat to the RPSs. An essential aspect of the programme is the ability to create an americium oxide fuel form, namely discs or pellets, that meet a range of requirements e.g. intact bodies with relatively high relative densities that allow for He-outgassing. Research with surrogates for americium oxides is essential for investigating the range of variables that influence the ability to achieve this whilst limiting the research with the highly radioactive material. In this study, americium oxide surrogates (e.g. Nd 2 O 3 ) have been created using two different techniques (continuous oxalate precipitation and calcination, and sol-gel and calcination) with the objective of creating particles with differing morphology. Owing to the polymorphism of Nd 2 O 3 , X-ray diffraction is conducted to assess crystal structure phase changes in the powder material to inform sintering studies. The surrogate powders are cold-pressed and sintered to assess the impact on pellet properties e.g. density and integrity. The surrogate fuel study highlights the importance of assessing the impact of particle shape and crystal structure on the ability to meet fuel form requirements, and will inform future research with the americium oxide fuel.

Both AmAlO3 and PuAlO3 perovskites have been synthesized and characterized using powder X-ray diffraction (XRD), Raman spectroscopy, Fourier transform infrared spectroscopy (FT-IR), and 27Al magic angle spinning nuclear magnetic resonance spectroscopy (MAS NMR). AmAlO3 perovskite showed a rhombohedral configuration (space group R3̅c) in agreement with previous studies. The effect of americium α-decay on this material has been followed by XRD and 27Al MAS NMR analyses. In a first step, a progressive increase in the level of disorder in the crystalline phase was detected, associated with a significant crystallographic swelling of the material. In a second step, the crystalline AmAlO3 perovskite was progressively converted into amorphous AmAlO3, with a total amorphization occurring after 8 months and 2 × 1018 α-decays/g. For the first time, PuAlO3 perovskite was synthesized with an orthorhombic configuration (space group Imma), showing an interesting parallel to CeAlO3 and PrAlO3 lanthanide analogues. High-temperature XRD was performed and showed a Imma → R3̅c phase transition occurring between 473 and 573 K. The thermal behavior of R3̅c PuAlO3 was followed from 573 to 1273 K, and extrapolation of the data suggests that cubic plutonium perovskite should become stable at around 1850 K (R3̅c → Pm3̅m transition).

Radioisotope power systems utilising americium-241 as a source of heat have been under development in Europe as part of a European Space Agency funded programme since 2009. The aim is to develop all of the building blocks that would enable Europe to launch and operate deep space and planetary missions in environments where use of solar power or alternative power generation technologies is challenging. Although some technical and policy work activity predate the ESA programme, the maturity of the technology has now reached a level that it can be incorporated in mission studies and roadmaps targeting the period from the mid 2020s onwards. This paper describes the state of the art in European radioisotope thermoelectric generators and radioisotope heater units. This paper includes: the evolution of the technical programme in detail; descriptions of the design; evolution of RTG and RHU devices from laboratory prototypes to more advanced fully functional systems; and experimental data obtained to date. This paper also outlines the technical challenges and multidisciplinary skills required to develop what is a world leading, original, significant and transformative technology solution for planetary science and exploration missions from the mid 2020s onwards.

Thorium-uranium oxide pellets with high densities were prepared at the Canadian Nuclear Laboratories (CNL) by co-milling, pressing, and sintering at 2023 K, with UO2 mass contents of 0, 1.5, 3, 8, 13, 30, 60 and 100%. At the Joint Research Centre, Karlsruhe (JRC-Karlsruhe), thorium-uranium oxide pellets were prepared using the spark plasma sintering (SPS) technique with 79 and 93 wt. % UO2. The thermal diffusivity of (Th1-xUx)O-2 (0 <= x <= 1) was measured at CNL and at JRC-Karlsruhe using the laser flash technique. ThO2 and (Th, U)O-2 with 1.5, 3, 8 and 13 wt. % UO2 were found to be semi-transparent to the infrared wavelength of the laser and were coated with graphite for the thermal diffusivity measurements. This semi-transparency decreased with the addition of UO2 and was lost at about 30 wt. % of UO2 in ThO2. The thermal conductivity was deduced using the measured density and literature data for the specific heat capacity. The thermal conductivity for ThO2 is significantly higher than for UO2. The thermal conductivity of (Th, U)O-2 decreases rapidly with increasing UO2 content, and for UO2 contents of 60% and higher, the conductivity of the thorium-uranium oxide fuel is close to UO2. As the mass difference between the Th and U atoms is small, the thermal conductivity decrease is attributed to the phonon scattering enhanced by lattice strain due to the introduction of uranium in ThO2 lattice. The new results were compared to the data available in the literature and were evaluated using the classical phonon transport model for oxide systems. (c) 2018 Elsevier B.V. All rights reserved.

The PELGRIMM project was a FP7-European project (funded by the European Commission, Grant Agreement No. 295664), carried out from 2012 to 2017. It was devoted to the investigation of spherepacked and pelletized fuel forms for Minor Actinide transmutation in homogeneous and heterogeneous recycling modes. PELGRIMM aimed at constituting a new step in the long term process in the assessment of Minor Actinide-bearing fuel, initiated within the European projects ACSEPT, F-BRIDGE, CP-ESFR and FAIRFUELS. The present paper provides an overview of the main technical outcomes gained within PELGRIMM. The developments of alternative processes in order to simplify synthesis routes and to limit secondary waste streams for Minor Actinide-bearing fuel preparation are detailed. The first results of behaviour under irradiation of spherepacked and pelletized fuel forms are provided from Post-Irradiation Examinations on (U,Pu,Am)O2 and (U,Am)O2 fuels respectively irradiated during SPHERE and MARIOS experiments, along with the description of the latest irradiation experiment, MARINE. In parallel, the capabilities of existing models and calculation codes have been improved to describe Minor Actinide-bearing fuel behaviour under irradiation in a more reliable way, and their predictive results have been compared to available Post-Irradiation Examinations. Finally, to start linking fuel behaviour with core neutronic problematics, a preliminary design of a Sodium-cooled Fast Reactor core loaded with spherepacked (U,Pu,Am)O2 fuels was built and correlated preliminary safety assessments have been performed.