Johannes Fachinger

CARBOWASTE: New EURATOM Project on ‘Treatment and Disposal of Irradiated Graphite and Other Carbonaceous Waste’

A new European Project has been launched in April 2008 under the 7th EURATOM Framework Programme (FP7-211333), with a duration of four years, addressing the ‘Treatment and Disposal of Irradiated Graphite and other Carbonaceous Waste (CARBOWASTE)’. The objective of this project is the development of best practices in the retrieval, treatment and disposal of irradiated graphite & carbonaceous waste-like structural material e.g. non-graphitised carbon bricks and fuel coatings (pyrocarbon, silicon carbide). It addresses both legacy waste as well as waste from future generations of graphite-based nuclear fuel. After defining the various targets for an integrated waste management, comprehensive analysis of the key stages from in-reactor storage to final disposal will then be undertaken with regard to the most economic, environmental and sustainable options. This will be supported by a characterisation programme to localize the contamination in the microstructure of the irradiated graphite and so more to better understand their origin and the release mechanisms during treatment and disposal. It has been discovered that a significant part of the contamination (including 14 C) can be removed by thermal, chemical or even microbiological treatment. The feasibility of the associated processes will be experimentally investigated to determine and optimise the decontamination factors. Reuse of the purified material will also be addressed to close the ‘Graphite Cycle’ for future graphite moderated reactors. The disposal behaviour of graphite and carbonaceous wastes and the improvement of suitable waste packages will be another focus of the programme. The CARBOWASTE project is of major importance for the deployment of HTR as each HTR module generates (during a 60 years operational lifetime) about 5,000 to 10,000 metric tonnes of contaminated graphite containing some Peta-Becquerel of radiocarbon. It is strongly recommended to take decommissioning and waste management issues of graphite-moderated reactors already into account when designing new HTR concepts.Copyright

Investigation of the Pyrolytic Degradation of Ion Exchange Resins by Means of Foil Pulse Pyrolysis Coupled with Gas Chromatography/Mass Spectrometry

Abstract Cationic (LEWATIT S100) and anionic ion exchangers (LEWATIT M500MB) were thermally decomposed in amounts of 15–60 μg in a foil pulse pyrolyzer. The gaseous products were gas-chromatographically separated and identified by mass spectrometry. Cationic resins release mainly sulfur dioxide and benzene at temperatures up to 500°C. At higher temperatures, degradation products like ethylbenzene, styrene, hydrogen sulfide, benzene, and toluene are increasingly observed. Their ratio strongly depends on the applied temperatures. After fractionated pyrolysis of a single sample with rising temperatures, mainly benzene, hydrogen sulfide, and carbon disulfide are produced at 1000°C, leaving behind a residue of pyrolysis coke. Anionic resins generate mainly trimethylamine and methyl chloride up to 400°C. With rising temperatures, the formation of styrene, p-methylstyrene, and p-ethylstyrene dominates. Pyrolysis of anionic resins ends at about 900°C without leaving behind significant amounts of residue. Kinetic ...

Impermeable graphite: A new development for embedding radioactive waste

The natural occurrence of graphite proves its geological stability over long time periods and therefore it could be considered as a matrix for embedding radioactive waste. However its porous structure affects the possible use of graphite as long term stable waste matrix for final disposal. Aqueous phases can penetrate the pore system and radionuclides adsorbed on the surface can be leached. Furthermore slow corrosion in aquatic phases can be induced by high irradiation dose rates in the range of 10−5 to 10−7 gm−2 d−1 . Therefore radiation induced corrosion process cannot be neglected in geological time scales. These problems can be solved with a graphite material with a closed pore system. A graphite composite material with an inorganic binder has been developed with a density > 99.7% of theoretical density and a negligible porosity. An initial calculation predicts that the life time of the graphite will be at least 2 orders of magnitude higher than porous graphite. This material represents a long term stable leaching resistant matrix applicable for the embedding of irradiated graphite (i-graphite). Natural graphite can be added to improve the compaction behavior and mechanical properties. Additional applications could be the embedding of other radioactive wastes in this matrix.Copyright

Production of an Impermeable Composite of Irradiated Graphite and Glass by Hot Isostatic Pressing as a Long Term Leach Resistant Waste Form

Around 250,000 tons of irradiated graphite (i-graphite) exists worldwide and can be considered as a current waste or future waste stream. The largest national i-graphite inventory is located in UK (∼ 100,000 tons) with significant quantities also in Russia and France [5]. Most of the i-graphite remains in the cores of shutdown nuclear reactors including the MAGNOX type in UK and the UNGG in France. Whilst there are still operational power reactors with graphite cores, such as the Russian RBMKs and the AGRs in UK, all of them will reach their end of life during the next two decades.The most common reference waste management option of i-graphite is a wet or dry retrieval of the graphite blocks from the reactor core and the grouting of these blocks in a container without further conditioning. This produces large waste package volumes because the encapsulation capacity of the grout is limited and large cavities in the graphite blocks could reduce the packing densities. Packing densities from 0.5 to 1 tons per cubic meter have been assumed for grouting solutions.Furthermore the grout is permeable. This could over time allow the penetration of aqueous phases into the waste block and a potential dissolution and release of radionuclides. As a result particularly highly soluble radionuclides may not be retained by the grout.Vitrification could present an alternative, however a similar waste package volume increase may be expected since the encapsulation capacity of glass is potentially similar to or worse than that of grout.FNAG has developed a process for the production of a graphite-glass composite material called Impermeable Graphite Matrix (IGM) [3]. This process is also applicable to irradiated graphite which allows the manufacturing of an impermeable material without volume increase. Crushed i-graphite is mixed with 20 vol.% of glass and then pressed under vacuum at an elevated temperature in an axial hot vacuum press (HVP). The obtained product has zero or negligible porosity and a water impermeable structure. Structural analysis shows that the glass in the composite has replaced the pores in the graphite structure. The typical pore volume of a graphite material is in the range of 20 vol.%. Therefore no volume increase will occur in comparison with the former graphite material.This IGM material will allow the encapsulation of i-graphite with package densities larger than 1.5 ton per cubic meter. Therefore a huge volume saving can be achieved by such an alternative encapsulation method. Disposal performance is also enhanced since little or no leaching of radionuclides is observed due to the impermeability of the material.NNL and FNAG have proved that IGM can be produced by hot isostatic pressing (HIP) which has several advantages for radioactive materials over the HVP process.• The sealed HIP container avoids the release of any radionuclides.• The outside of the waste package is not contaminated.• The HIP process time is shorter than the HVP process time.• The isostatic press avoids anisotropic density distributions.• Simple filling of the HIP container has advantages over the filling of an axial die.Copyright

Corrosion Behavior of Spent MTR Fuel Elements in a Drowned Salt Mine Repository

Spent MTR fuel from German Material Test Reactors will not be reprocessed, but stored in a final salt repository in the deep geologic underground. Fuel elements will be placed in POLLUX containers, which are assumed to resist the corrosive attack of an accidentally formed concentrated salt brine for about 500 years. After a container failure the brine would contact the fuel element, corrode the aluminum plating and possibly leach radionuclides from the fuel. A source term for the calculation of radionuclide mobilization results from the investigation of the behavior of MTR fuel in this scenario, which has to be considered for the long- term safety analysis of a deep mined rock salt repository. Experiments with the different plating materials show that the considered aluminum alloys will not resist the corrosive attack of a brine solution, especially in the presence of iron, under the conditions in a drowned salt mine repository. Although differences in the corrosion rates of about two orders of magnitude were observed when applying different parameter sets, the deterioration must be considered to be almost instantaneous in geological terms. Radionuclides are mobilized from irradiated MTR fuel, when the “meat” of the fuel element becomes accessible to the brine solution. It seems, however, that the radionuclides are effectively trapped by the aluminum hydroxide formed, as the activity concentrations in the brine solution soon reach a constant level with the progressing corrosion of the cladding aluminum. In the presence of iron a more significant initial release was observed, but also in this case an equilibrium activity seems to be reached as a consequence of radionuclide trapping.