14C selective extraction from French graphite nuclear waste by CO2 gasification (original) (raw)
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Chemical Characterization and Removal of Carbon-14 from Irradiated Graphite II - 13023
2013
Approximately 250,000 tonnes of irradiated graphite waste exists worldwide and that quantity is expected to increase with decommissioning of Generation II reactors and deployment of Generation IV gas-cooled, graphite moderated reactors. This situation indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from some nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research presented last year and updated here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. A nuclear-grade graphite, NBG-18, and a high-surface-area graphite foam, POCOFoam{sup R}, were exposed to liquid nitrogen (to increase the quantity of C-14 precursor) and neutron-irradiated (10{sup 13} neutrons/cm{sup 2}/s). Finer grained NBG-25 w...
Chemical Characterization and Removal of C-14 from Irradiated Graphite-12010
2012
Quantities of irradiated graphite waste are expected to drastically increase, which indicates the need for a graphite waste management strategy. Of greatest concern for long-term disposal of irradiated graphite is carbon-14 (C-14), with a half-life of 5730 years. Study of irradiated graphite from nuclear reactors indicates C-14 is concentrated on the outer 5 mm of the graphite structure. The aim of the research described here is to identify the chemical form of C-14 in irradiated graphite and develop a practical method by which C-14 can be removed. Characterization of pre- and post-irradiation graphite was conducted to determine bond type, functional groups, location and concentration of C-14 and its precursors via the use of surface sensitive characterization techniques. Because most surface C-14 originates from neutron activation of nitrogen, an understanding of nitrogen bonding to graphite may lead to a greater understanding of the formation pathway of C-14. However, no single te...
Investigation of reactor graphite processing with the carbon-14 retention
MRS Proceedings, 2002
The system C -Al -TiO 2 has been demonstrated to be a strong candidate for the processing of irradiated reactor graphite waste with the retention of biologic hazardous carbon-14 in chemically and thermal stable corundum-carbide ceramics. The corundum-carbide ceramics is obtained from the powdered precursors blend through self-sustaining thermochemical reactions. Investigations of the system C -Al -TiO 2 were carried out both theoretically and experimentally. The refining thermodynamic calculations of the phase composition of resulting end product were performed for a wide variety of components content in the system being investigated. Aluminium oxycarbides production was taken into account in the calculations. Thermodynamic functions of aluminium oxycarbides Al 4 O 4 C and Al 2 OC have been calculated for this purpose using currently available literature evidences and own assessments of missing data. On the basis of thermodynamic simulation the proportions of the source substances were determined, which result in the aluminium oxycarbides production. These simulation results have been supported by XRD-analysis of produced specimens. The experimental processing of reactor graphite was conducted by the use of self-sustaining reactions in C -Al -TiO 2 powder blends. Test specimens were produced by mass ranging from 0.1 to 3 kg in the argon atmosphere. Various techniques were used to characterize the produced specimens. The compressive strength of specimens of corundum-carbide matrices produced ranges from 7 to 13 MPa. The leaching rates of Cs-137 and Sr-90 from specimens ranged between 10 -4 and 10 -5 g/(cm 2 .day) respectively. The carry-over of the carbon combined in carbon monoxide from the reacting mixtures during exothermic process may run up to 1% wt. that appropriates roughly to less than 0.01% wt. of the carbon-14 in the irradiated reactor graphite.
Carbon, 2015
Graphite has been used in gas-cooled nuclear reactors as a neutron moderator. The dismantling of nuclear reactors will generate significant amounts of graphite waste. Neutron irradiation is responsible for 14 C formation in graphite, and it leads to severe structural and nanostructural degradations. At high neutron fluence, nanoporous turbostratic carbon is formed from original lamellar graphite. This phase is supposed to be especially 14 C enriched. An original 14 C extraction process was proposed: to ''decontaminate'' graphite waste from 14 C by selectively gasifying such degraded areas, without entirely consuming the graphite waste. To specify the operating conditions, milled graphite was used as a nonradioactive analogue. Raman microspectrometry and transmission electron microscopy techniques show that neutron irradiation and milling lead to similar multiscale organization, and especially nanoporous carbon formation. Thermogravimetry experiments were then carried out between 800 and 1100°C, at a CO 2 pressure of 0.1 MPa. To determine the best temperature range allowing a nanoporous component selective gasification, Raman microspectrometry analysis was coupled with transmission electron microscopy observations on the residues obtained for each gasification temperature. The 950-1000°C temperature range is the most efficient allowing a complete elimination of degraded areas supposed to be representative of nuclear graphite waste 14 C-rich areas.
14C leaching and speciation studies on Irradiated graphite from vandellós I Nuclear Power Plant
2018
The understanding of the 14C behavior in waste packages could lead, in the Spanish context, to a revision of the management strategies for radioactive waste and a revaluation of the near surface repository devoted to the disposal of waste containing this radionuclide in high concentrations. To achieve this objective, and in the context of the EU project Carbon-14 Source Term (CAST), the authors of the work presented in this paper have performed leaching experiments with irradiated graphite considering two different scenarios. One, in which the leaching solution simulates some of the expected conditions in a repository where a granite/bentonite mixture has been used as backfill material, and the other, using deionized water as a high efficiency chemical removal agent and for comparison purposes. The analytical approach to measure the release rate and speciation of 14C from irradiated graphite samples in the aqueous and gaseous phase is also described. The main results obtained shows ...
PROCESSING OF GRAPHITE WITH CARBON14 RETENTION
2000
Waste graphite containing fragments of nuclear fuel and fission products is produced mainly as a result of operation of uranium−graphite reactors. Retention of radionuclides, including carbon-14, after disposal is an important goal of treatment procedure for such waste. Conversion of waste graphite into a stable waste form acceptable for long term storage and disposal has been previously considered from a
14C and Other Radionuclides in Impermeable Graphite Material Waste form Long Term Behavior
Radiocarbon, 2018
ABSTRACTThe radiocarbon (14C) content of irradiated graphite is the most important problem for the management of Spanish irradiated graphite (Vandellós I NPP) as L&ILW, due to this material exceeding the maximum 14C inventory for the C.A. El Cabril repository. Therefore, the encapsulation of graphite in an impermeable matrix and making an appropriate waste form are indicated as potential management options to be studied. The conversion of the graphite to a long-term stable glass matrix, called IGM (impermeable graphite matrix), uses a long-term stable inorganic binder which additionally encloses the graphite pore system. The world’s first IGM samples made with irradiated graphite have been manufactured in CIEMAT facilities. The durability of the matrix is investigated in leaching experiments in deionized water and granitic bentonite water. The results show that ∼0.05% of 14C is leached. A species of organic carbon was found as formate and oxalate (∼10–1 mg/L). CO was detected as vol...
Rapid analysis method for the determination of 14C specific activity in irradiated graphite
2018
14C is one of the limiting radionuclides used in the categorization of radioactive graphite waste; this categorization is crucial in selecting the appropriate graphite treatment/disposal method. We propose a rapid analysis method for 14C specific activity determination in small graphite samples in the 1±100 μg range. The method applies an oxidation procedure to the sample, which extracts 14C from the different carbonaceous matrices in a controlled manner. Because this method enables fast online measurement and 14C specific activity evaluation, it can be especially useful for characterizing 14C in irradiated graphite when dismantling graphite moderator and reflector parts, or when sorting radioactive graphite waste from decommissioned nuclear power plants. The proposed rapid method is based on graphite combustion and the subsequent measurement of both CO2 and 14C, using a commercial elemental analyser and the semiconductor detector, respectively. The method was verified using the liquid scintillation counting (LSC) technique. The uncertainty of this rapid method is within the acceptable range for radioactive waste characterization purposes. The 14C specific activity determination procedure proposed in this study takes approximately ten minutes, comparing favorably to the more complicated and time consuming LSC method. This method can be potentially used to radiologically characterize radioactive waste or used in biomedical applications when dealing with the specific activity determination of 14C in the sample.