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[en] Complete text of publication follows. A Hungarian AMS graphite production facility was established in ATOMKI in 2005. After the first promising results, the complete vacuum system was upgraded and the hydrogen reduction based graphitization protocol was revised and improved. Also the whole lab building was completely renewed implementing an enhanced air filtration and conditioning system. Several sets of blanks and known-activity graphite samples were processed and tested in cooperation with the NSF Arizona 14C AMS facility in Tucson, Arizona, US. The gas handling line is made of stainless steel with Swagelok valves and fittings. Using a newly installed Pfeiffer turbo molecular pump (TMU 071 P) we have a minimum pressure of < 1 x 10-3 mbar at the connections for gas ampoules and graphitisation reaction rigs. Graphite targets were prepared by reduction of carbon dioxide gas sample using Fe catalyst and hydrogen gas. The Fe powder we used was less than 325 meshes, 98 % purity (Alfa Aesar). The reaction rig consists of a Hy-Lok plug valve, a Swagelok Vacuum Ultra-Torr Union Tee fitting, a quartz tube designed to limit carbon contamination, and a glass reusable water trap tube. The volume of the reaction rig is ∼ 7.0 cm3. Before the beginning of the graphitization process iron powder was activated by prereduction in 0.7 bar H2 atmosphere at 450 deg C for 90 min. The H2:CO2 ratio was fixed at 2:1 according to our earlier results, which suggest that this ratio avoids hydrocarbon formation during graphitisation. The initial pressure of a carbon dioxide gas sample was 300 - 500 mbar. The temperature of the water trap during the iron powder activation and the graphitization process was -5 deg C. Minimum graphitization time was 300 minutes, however a typical sample was processed overnight. The background of the gas handling line and the graphite target production system were tested by graphitization of C-14 free, old borehole carbon dioxide gas (δ13C = -3.4 ± 0.2 h PDB, purity 4.5, Linde AG, Repcelak, Hungary). We also checked the reliability of the system by graphitizing gas produced from an intercalibration sample VIRI B (consensus 14C age value: 2820 yr BP). Most of the blank graphite targets produced in ATOMKI with hydrogen reduction gave good background results (14C age > 50,000 yrs BP) in comparison with zinc reduced graphite blanks that are routinely produced in the NSF Arizona Lab. Although all of the graphite samples from ATOMKI were processed in a completely different manner than is usual at the NSF Arizona Lab, the mean value of the VIRI B samples (2870 ± 30 yr BP) measured by the NSF Arizona AMS is consistent with measured by gas proportional counting technique in ATOMKI (2810 ± 30 yr BP) and with the consensus value for this sample (2820 yr BP) as published in the first report of the VIRI project
[en] The simple method of graphite target development, first presented at AMS-7 has been further developed and refined. Experiments have been performed to study the range of possible reaction conditions and the effect of these on the nature of the graphite generated. The results from these experiments have been used to make the method as robust as possible with a high success rate, a quick reaction time and very simple apparatus requirements. This paper covers the details of the experiments, the conclusions drawn from them, and the technique now employed for routine graphite sample preparation at ORAU
[en] There exists more than 250,000 tonnes of irradiated nuclear graphite in the world, primarily as a result of the development of graphite-moderated power-reactor systems, (Figure 1). Only a very small number of such plants have been dismantled and, for most cases, the final destiny of the irradiated graphite (“i-graphite”) remains unresolved. Future high-temperature reactor programmes, such as the Chinese HTR-PM development, will produce more graphite and carbonaceous wastes from both structural components and the fuel pebbles (which are approximately 96% carbonaceous). The problem of dismantling irradiated graphite reactor stacks, possibly distorted through neutron damage and in some cases degraded further by radiation-chemical attack by gaseous coolants, and then finding the appropriate treatments and final destiny of the material, has exercised both the International Atomic Energy Agency and the European Union for more than 25 years, seeking to address the different issues and available disposal solutions in different IAEA Member States. This paper reviews the current ‘state of play’.
[en] Highlights: • Structural evolution during pyrolysis of novolac resin containing H3BO3 or B2O3. • The catalytic graphitization depends on BOC formation and cleavage. • Composition, bond strength and crystallization controlled the carbons' reactivity. • Better oxidation resistance can be attained without the carbon's crystallization.
[en] Use of ultrasound as an intensified non-destructive decontamination technique for processing graphite limits its reusability beyond a few number of decontamination cycles due to the exfoliation of graphite due to cavitation effects. A recent article established that the use of platinum nanoparticles in the leachant reduces the erosion of graphite substrate due to cavitation. It presents an improved way of sonochemical recovery of ceria using a mixture of nitric acid, formic acid and hydrazinium nitrate in the presence of platinum nanoparticles and ionic liquid. The platinum nanoparticles catalyst in ionic liquid prevented the generation of the carbon residue due to the combined effect of denitration and reduced sonication. The presence of the catalyst showed a fivefold increase in dissolution kinetics of ceria as well as absence of graphite erosion, facilitating better chances of graphite recycling than the decontamination without the catalyst. The catalytic approach offers a better recycle strategy for graphite with reduced exfoliation and NOx generation due to denitration, making it a more sustainable decontamination process. Since ceria is used as a surrogate for plutonium oxide, the results can be extended to decontaminate such deposits clearly establishing the utility of the presented results in the nuclear industry. (author)