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Transactions of the American Nuclear Society, 114, ISSN:0003-018X

Particle bed arrangements (e.g. sphere-pac and vipac) are considered as alternative fuel-forms for nuclear fission of actinides. The fuel material is potentially UO2, MOX or an Inert Matrix such as yttria-stabilized zirconia. A disadvantage of ceramic particle fuel is its low thermal conductivity at startup. The macroscopic structure of a particle bed (e.g. spherical particles in sphere-pac) hinders heat transport prior to sintering. To remedy this situation, a novel cer-met concept is proposed and studied in this work. In the sphere-pac model with two well differing size fractions, one material can be distributed evenly into the other one. In this case, the ceramic material (e.g. yttria stabilized zirconia) and the metallic component (e.g. zirconium) are the large and small size fractions respectively. Calculations for this initial fuel configuration show a significant enhancement of the thermal conductivity. Sintering occurs following startup and especially at the center of the fuel. The macroscopic sphere-pac arrangement then transforms into a porous-pellet-like structure. Sintering experiments have been performed to investigate these effects. In the cer-met arrangement the restructuring was enhanced. Therefore heat transport improves even as the metallic component becomes oxidized.

We demonstrate that irradiation may enhance the plasticity in metallic glasses by increasing the free-volume content via micropillar compression experiments on an ion-irradiated bulk metallic glass (BMG). Results show that irradiation decreases the flow stress and enhances the shear band formation by lowering the magnitude of stress serrations in plastic flow regime. These results highlight that amorphous alloys can mitigate the deleterious affects of severe ion irradiation as compared to their crystalline counterparts. (C) 2009 Acta Materialia Inc. Published by Elsevier Ltd. All rights reserved.

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. In this frame, the thermal conductivity (TC) of cladding prototype sections are measured both before and after neutron irradiation with the purpose of better understanding the effects of neutron irradiation on the pyrolytic carbon interphase linking the SiC fibres to the SiC matrix.

Oxide dispersed strengthened (ODS) ferritic-martensitic steels are investigated as possible structural material for the future generation of High Temperature Gas Cooled Nuclear Reactors. The Ni based austenitic ODS superalloys are not considered, because of the Ni presence, which is unfavorable under neutron irradiation. ODS-steels are considered to replace other high temperature materials for tubing or structural parts. Interestingly, ODS is also considered as material being used in future fusion applications. The oxide particles serve for interfacial pinning of moving dislocations. Therefore the creep resistance is improved. In case of the usage of these materials in reactor, the behavior under irradiation must be further clarified. In this paper the effects induced by He implantation are investigated. The induced swelling is measured and the mechanical behavior of the irradiated surface is investigated. These first tests are performed at room temperature, where a clear swelling and hardening could be observed. -------- Окисно диспергированные упрочненные (ОДУ) ферритно-мартенситные стали изучались как возможные конструкционные материалы для нового поколения высокотемпературных ядерных реакторов с газовым охлаждением. Аустенитные ОДУ-суперсплавы на базе никеля не рассматривались из-за присутствия никеля, который нежелателен при нейтронном облучении. ОДУ-стали рассматривались как возможные заменители других высокотемпературных материалов для изготовления трубопроводов или других композиционных узлов. Интересно, что ОДУ рассматривается так же, как возможный кандидат для использования в термоядерных устройствах. Окисные частицы служат как межфазные ловушки для закрепления движущихся дислокаций. Поэтому сопротивление ползучести увеличивается. В случае использования этих материалов в реакторе их поведение под облучением должно изучаться более тщательно. В предлагаемой работе исследуются эффекты, обусловленные внедрением Не. Измеряется обусловленное распухание и механические характеристики облученной поверхности. Эти первые испытания были выполнены при комнатной температуре, когда можно явно наблюдать распухание и упрочнение. ------ Окисно дисперговані зміцнені (ОДЗ) феритно-мартенситні сталі досліджуються як можливі конструкційні матеріали для нового покоління високотемпературних ядерних реакторів з газовим охолодженням. Аустенітні ОДЗ-сплави на основі никелю не вивчаються завдяки никелю, присутність якого під дією опромінення небажана. ОДЗ-сталі разглядаються як можливі кандидати на заміну інших високотемпературних матеріалів для вигoтовлення трубопроводів або інших конструкційних вузлів. Цікаво, що ОДЗ-матеріали розглядаються також з точки зору їх можливого використання для майбутнього застосування в термоядерних пристроях. Окисні частки служать як міжфазні пастки для закріплення дислокацій, що рухаються. У разі використання ціх материалів в реакторі їх поведінка під опроміненням повинна вивчатись більш ретельно. В роботі досліджуються ефекти, зумовлені проникненням Не. Вимірюється зумовлене розпухання та механічні характеристики опроміненої поверхні. Ці перші дослідження були виконані при кімнатній температурі, коли можно чітко спостерігати розпухання та зміцнення.

New experiments conducted at pH 9 and 298 K clearly indicate that addition of bicarbonate ions in aqueous solutions increases the solubility of both cubic and monoclinic zirconia by several orders of magnitude. In pure water the solubility of zirconia proved to be very low (about 10−10 to 10−8 M), while in the presence of 0.05 M NaHCO3 it increased to 10−6– 10−5 M, indicating the formation of strong carbonate complexes. Due to the limited number of available data and the restricted experimental conditions (fixed pH) it was not possible to identify the stoichiometry of the complex(es). However, up to a total carbonate concentration of 0.05 M, our data are consistent with the formation of a mononuclear tetracarbonate or pentacarbonate complex. At higher carbonate concentrations basic Zr carbonates are possibly formed. The formation constant fitted for the pentacarbonate complex is at the upper limit of a range previously estimated with the help of data on analogous actinide-carbonate complexes. Although our results are not sufficient to determine reliable thermodynamic data for the complexation of Zr with carbonate, they can be used to set limits to the solubility of inert matrix spent fuel in radioactive waste storage environments. Our measurements indicate a Zr solubility in the order of 10−6 to 10−5 M for typical repository conditions, which is still 2–3 orders of magnitude below silica concentrations measured from the dissolution of nuclear waste glasses.

In the frame of the Generation IV initiative, different structural material candidates are investigated at the Paul Scherrer Institute. These are oxide dispersion strengthened (ODS) steels, intermetallic materials and ceramic composite materials. The response of the material to different potential loads (irradiation, temperature …) is addressed in a multi-scale approach, both, modelling wise and also experimentally. The investigation of each scale delivers at least a qualitative understanding of possibly evolving damage in the material and also delivers a validation of the corresponding scale on the modelling side. From the experimental side, the lower end of the scale, the atomistic and structural level, can be investigated by conventional techniques, as for example transmission electron microscopy (TEM) and X-ray diffraction (XRD). However, the use of synchrotron radiation techniques offers an ideal, complementary way to investigate the material structure and other properties. This paper presents applications in the field of the ODS research, where the structural behaviour of the nanoscopic dispersoids can selectively be investigated, although only being present with roughly 5 wt % in the matrix. A study showing the structural behaviour of these oxide particles as a function of irradiation illustrates the potential of the extended X-ray absorption fine structure (EXAFS) technique. Using X-ray magnetic circular dichroism (XMCD), which is a difference-signal of two X-ray absorption spectra recorded for positive and negative helicities of the beam, the magnetic structure and some magnetic parameters, can be resolved. An example shows, how this can be applied to understand (Fe,Cr) systems, which is the base alloy of the investigated ODS steel. The results deliver an important crosscheck for modelling. Beside the presentation of these techniques, this paper shows how beamline techniques can serve nuclear research, with possibly activated materials. At the Paul Scherrer Institute, a sample holder for highly active materials has been developed, and has already served for EXAFS measurements at the Swiss Light Source (SLS). The set-up of this sample holder is briefly presented here.

In today's applied light water reactor (LWR) technology, the fissile material is embedded in a ceramic matrix, pressed, and sintered to pellets, which are then filled into the cladding tube of fuel pins that are assembled to a fuel bundle. This is the most widespread and well-known concept, which is also mostly adapted for the present fast breeder reactor (FBR) technology. Many alternative fuel forms have, however, been researched, seeking simplified fabrication routes and other advantages. When fissile isotopes are coming from spent fuel that is chemically separated (reprocessed), particle fuel with its direct filling of fuel particles into the fuel pin offers several advantages. Two major types of particle fuel are discussed here: Sphere-pac and Vipac fuel.

Future nuclear power plants like the very high temperature reactor or the gas cooled fast reactor need materials able to comply with extreme operating conditions (temperature, cooling gas, radiation). Advanced materials like oxide dispersion strengthened (ferritic) steels (ODS) or intermetallics are considered as candidates for such plants. The ferritic ODS alloy PM 2000 and a creep resistant TiAl intermetallic were He-implanted at different temperatures. Swelling behaviour as well as damage development were studied. Irradiation induced creep was investigated for TiAl at temperatures of 350°C and 500°C.

The gas cooled fast reactor (GFR) is a helium-cooled fast spectrum reactor operating within a closed fuel cycle. It combines the advantages of fast reactors, in terms of a more sustainable use of uranium resources and waste minimisation, with the wider applicability of high temperature gas reactors, in terms of high efficiency electricity generation and the co-generation of high-quality process heat. Other advantages like the absence of threshold effect due to phase changing, the optical transparency and chemical inertness of the Helium coolant are also acknowledged. Within the European Union, GFR is one of the three fast reactors proposed for development to the demonstration stage within the European Sustainable Nuclear Industry Initiative (ESNII). On a wider global scale, GFR is one of the six systems proposed for further development within the Generation IV International Forum (GIF). In this respect, France, Switzerland, Japan and the European Union (through EURATOM) are signatories to the 'System Arrangement', the instrument through which the international research efforts are coordinated. This paper presents the current status of the development of the GFR system. The status of the GFR programme in each of the signatory countries is summarised including the intended contribution of the newly launched EURATOM 7. Framework Programme project - GoFastR. France has provided the bulk of the effort on conceptual design, safety assessment and fuel development. Switzerlanssessment and fuel development. Switzerland makes significant contributions to the GFR system in the areas of core physics, uncertainty analysis, deterministic safety assessment and fuel development. Historically Japan has been very active in the development of the GFR system. Within the Generation IV GFR system, Japan contributes to the development of fuel and core materials

The GFR (Gas cooled Fast Reactor) combines the advantages of a fast neutron spectrum with those of high temperatures. Given the high core outlet temperatures exceeding 800 C, it is clear that conventional metallic core structures and fuel cladding will be unable to meet the demanding requirements. Thus, only refractory metals or ceramic components can be considered for these purposes, as temperature excursions above 1600 C have to be foreseen. The design and development of an innovative refractory fuel in an advanced ceramic cladding remains a fundamental goal of the GFR system. Today, the focus is on SiC composite as structural and cladding material, with carbide (MC) fuel taking first priority over both oxide and nitride fuels as backups. Compatibility tests have shown that SiC and mixed metal carbide fuels (MC) react, necessitating the introduction of a protective liner made of W or other refractory material. 2 fuel element design option are considered: the plate type fuel and the pin type fuel. The basic design of plate type fuel is based on two ceramic plates, which enclose a honeycomb structure containing cylindrical pellets made of the mixed carbide fuel. Pin type fuel will be similar to current fast reactor designs

Synchrotron radiation techniques represent powerful tools to characterize materials down to the nanometer level. This paper presents a survey of the state-of-the-art synchrotron-based techniques which are particularly well-suited for investigating materials properties. Complementary X-ray absorption techniques such as extended X-ray absorption fine structure (EXAFS), X-ray magnetic circular dichroism (XMCD), photoemission electron microscopy (PEEM) are used to address the individual local atomic structure and magnetic moments in Fe–Cr model systems. The formation of atomic clusters/precipitates in such systems is also investigated by means of scanning transmission X-ray microscopy (STXM). Such advanced analytical techniques can not only offer valuable structural and magnetic information on such systems, they can also serve for validating computational calculations performed at different time and length scales which can help improve materials lifetime predictions.

During the last two decades, silicon carbide based ceramic composites (SiC/SiC) have become a candidate material for nuclear fuel cladding, first for advanced nuclear systems, such as the GFR, and recently also as potential replacement material for nuclear fuel cladding in light water reactors. This work is contributing to the General Atomics/Westinghouse led CARAT project. This paper focuses on the consequences of the microstructural changes occurring at the pyrolytic (PyC) carbon interphases linking silicon carbide matrix and fibres in the composite. The experimental approach is two-fold: a macro-scale approach deals with the measurement of the thermal conductivity of short sections of prototype SiC/SiC cladding tubes, whereas the effects of radiation damage on the PyC interlayer are studied by energy filtered transmission electron microscopy (EFTEM) as well as spectrum imaging in STEM (ESI). The results from both faces of the experimental work are brought to gether in an effective medium model based on the work of Markworth and Youngblood et al. A detailed outlining of the radial heat flow experimental apparatus developed to this end has been presented at the 2016 ANS meeting in New Orleans.

The performances of oscillating differential scanning calorimetry (ODSC) and laser-flash technique in determining the thermal conductivity and the specific heat of zirconia-based materials, analogous to a potential nuclear fuel, were compared. The tested materials were (Zr1−x−y−z,Yx,Ery,Mez)O2−(x+y)/2, with Me=Ce or Th. The measured specific heats were around 0.4 J K−1 g−1 and the thermal conductivities ranged from 2 to 3 W K−1 m−1. The ODSC measurements resulted in cp values. The thermal conductivity was derived from two complementary measurements, one with a thin and the other with a thick sample. The laser-flash technique directly delivered the thermal diffusivity of the sample; consequently, the specific heat capacity cp has to be known for determining the thermal conductivity. The ODSC measurements were affected by the position of the sample on the support. This, consequently, influenced the reproducibility of the measurements. The reproducibility of the scans by laser flash was excellent. Thermal conductivity decreased with increase in the stabilizer (Y, Er) concentration. This trend was justified on the basis of a model including concentration and size of the oxygen vacancies.

Based on closed packed microspheres produced by internal gelation, the Sphere-pac concept provides a safer and cleaner process and better remote production capability compared to the traditional production of nuclear fuels. Whereas classical reactor fuel rods are fed with UO2 pellets, this advanced concept is suggested for minor actinide bearing fuels. This paper focuses on the thermal analysis of the microspheres during their preparation by microwave-assisted internal gelation (MIG). Internal gelation is a sol-gel process initiated by a temperature shift within aqueous droplets. Alternatively this critical temperature increase takes place by contact heat transfer from a hot silicon-oil bath. Microwave heating of the droplets during their free fall provides a simpler and cleaner process, where the experimental parameters have to be carefully optimized. A matlab model is developed to study the thermal behaviour of a droplet during processing inside a microwave resonant cavity. This finite elements model takes into account the most relevant parameters (size of droplet, thermal properties of the solution, heat and mass transfer …) and calculates the temperature profile inside the droplet. The model is validated with experimental data from literature. The influence of each parameter on the temperature profile is investigated and the optimal microwave power can then be determined and applied to the production unit.