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Utilization of HTR reflector graphite as embedding matrix for radioactive waste
Abstract
The reflector graphite of an HTR reactor has to be handled as radioactive waste after the operational period of the reactor. However the waste management of irradiated graphite from Magnox reactors shows, that waste management of even low contaminated graphite could be expensive and requires special retrieval, treatment and disposal technologies for safe long term storage as low or medium radioactive waste.However the reflector graphite could be transferred into long term stable embedding matrix for high level radioactive waste especially for HTR fuel elements. This can be achieved by closing the pore system of the graphite with a stable inorganic binder, e.g. glass.First investigations proved the sealing of the pore system and the potential for embedding HTR fuel pebbles.
... 19 In situ entombment. 20 Waste volume reduction & emission to atmosphere. 21 Making use of graphite as inert filler, removing the need for some encapsulation. ...
The European Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project sought to develop best practices in the retrieval, treatment, and disposal of irradiated graphite including other irradiated carbonaceous waste such as structural material made of graphite, nongraphitized carbon bricks, and fuel coatings. Emphasis was given on legacy irradiated graphite, as this represents a significant inventory in respective national waste management programs. This paper provides an overview of the characteristics of graphite irradiated during its use, primarily as a moderator material, within nuclear reactors. It describes the potential techniques applicable to the retrieval, treatment, recycling/reuse, and disposal of these graphite wastes. Considering the lifecycle of nuclear graphite, from manufacture to final disposal, a number of waste management options have been developed. These options consider the techniques and technologies required to address each stage of the lifecycle, such as segregation, treatment, recycle, and ultimate disposal in a radioactive waste repository, providing a toolbox to aid operators and regulators to determine the most appropriate management strategy. It is noted that national waste management programs currently have, or are in the process of developing, respective approaches to irradiated graphite management. The output of the Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste project is intended to aid these considerations, rather than dictate them.
Tri-Structural Isotropic (TRISO) coated particle-fuel is a key component in designs for future high temperature nuclear reactors. This study investigated the suitability of three soda lime silicate glass compositions, for the encapsulation of simulant TRISO particle fuel. A cold press and sinter (CPS) methodology was employed to produce TRISO particle–glass composites. Composites produced were determined to have an aqueous durability, fracture toughness and Vickers’ hardness comparable to glasses currently employed for the disposal of high level nuclear wastes. Sintering at 700 °C for 30 min was found to remove all interconnected porosity from the composite bodies and oxidation of the outer pyrolytic carbon layer during sintering was prevented by processing under a 5% H2/N2 atmosphere. However, the outer pyrolytic carbon layer was not effectively wetted by the encapsulating glass matrix. The aqueous durability of the TRISO particle–glass composites was investigated using PCT and MCC-1 tests combined with geochemical modelling. It was found that durability was dependent on silicate and calcium solution saturation. This study provides significant advancements in the preparation of TRISO particle encapsulant waste forms. The potential for the use of non-borosilicate sintered glass composites for TRISO particle encapsulation has been confirmed, although further refinements are required.
The significance of syngenetic and epigenetic graphite occurrences from the Variscan high-temperature/low-pressure Aracena Metamorphic Belt is discussed in the framework of the tectono-thermal evolution of this southern zone of the Iberian Massif. Syngenetic graphite is associated with both low- to medium-grade metamorphic rocks (La Umbría series, Precambrian in age) and high-grade, granulite facies rocks (the Precambrian Fuente del Oro series and a Cambrian calc-silicate series). Epigenetic, fluid-deposited occurrences correspond to overgrowths on existing metamorphic graphite grains and vein-type mineralization. Two types of graphitized particles with remarkable differences in reflectance, anisotropy and size can be distinguished in the Precambrian metapelites of the La Umbría series. Large, >150 6.74 Å), and the anomalously high temperatures of the DTA exothermic peak (close to 600 °C) of graphite with respect to that inferred from mineral assemblages in these rocks. The presence of graphite-rich quartzites and gneisses within the Fuente del Oro series and the calc-silicate series is evidence of sedimentation under reducing conditions in a continental shelf. The characteristics of graphite reflect the high-grade metamorphic conditions attained in the southern area of the Aracena Metamorphic Belt. Pervasive flow of fluids related to a major Variscan extensional event resulted in overgrowths on the pre-existing graphite in the gneisses and quartzites of the calc-silicate series, as evidenced by the heterogeneous isotopic composition of graphite single crystals in these rocks. A later stage of graphite precipitation is represented by scarce vein-type occurrences in mafic granulites that document channelled flow of fluids.
One back-end option for spent HTR fuel elements proposed for future HTR fuel cycles in the EC is an open fuel cycle with direct disposal of conditioned or non-conditioned fuel elements. This option has already been chosen in Germany due to the political decision to terminate the use of HTR technology. First integral leaching investigations at Research Centre Juelich on the behaviour of spent HTR fuel in salt brines, typical of accident scenarios in a repository in salt, proved that the main part of the radionuclide inventory cannot be mobilised as long as the coated particles do not fail. However, such experiments will not lead to a useful model for performance assessment calculations, because a failure of the coatings by corrosion will not occur during experimental times of a few years. In order to get a robust and realistic model for the long-term behaviour in aqueous phases of host rock systems, it is necessary to understand the barrier function of the different parts of an HTR fuel element, i.e. the matrix graphite, the different coating materials, and the fuel kernel.
Zugleich: Aachen, Techn. Hochsch., Fak. f. Maschinenwesen, Diss. M. Hrovat 1973.
Inventory of reactor graphite and treatment for disposal
- J Fachinger
Fachinger, J., 2008. Inventory of reactor graphite and treatment for disposal. In: RADWAP 2008, 27–31 October 2008, Würzburg, Germany.
Aqueous phase ingression into non-irradiated graphite. FZJ-ISR internal report
- J Fachinger
- P Diekmann
Fachinger, J., Diekmann, P., 2005. Aqueous phase ingression into non-irradiated graphite. FZJ-ISR internal report.