Rapid analysis method for the determination of 14C specific activity in irradiated graphite (original) (raw)
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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...
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...
Radioprotection, 2009
A high temperature combustion method was used to analyze the 14 C and 3 H activities in graphite and the dose assessment was carried out to determine the clearance in the conservative way. By this method, the 3 H and 14 C were simultaneously trapped in the nitric acid and carbosorb, respectively. Accordingly, the sample preparation time for the measurement was reduced to the half. The combustion temperature was more than 800 degrees in centigrade for obtaining total tritium and 14 C in the sample. The combustion ratio was about 99% on the graphite sample with the weight of 0.1 g. Minimum detectable activity was 0.05 Bq/g for the 14 C and 0.15 Bq/g for the 3 H at the same background counting time. The recoveries from the combustion furnace were around 100% and 90% in 14 C and 3 H, respectively. The radioactivity were 2, 530 ∼ 3,160 Bq/g in 14 C and 1, 700 ∼2,040 Bq/g in 3 H at this experiment. The experimental uncertainty was less than 6% in both radionuclides where the furnace recovery was dominant factor. An individual effective dose from beta and gamma radionuclides was estimated by consideration of the scenario of inhalation, ingestion and external exposure. 60 Co, the radioactivity of which was measured by using HPGe detector, had a predominant effect in estimating the effective dose. The estimation showed that the graphite wastes from the dismantled research reactor should be disposed of as a low level radioactive waste rather than clearance.
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...
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 ...
PloS one, 2016
Pile Grade A graphite was used as a moderator and reflector material in the first generation of UK Magnox nuclear power reactors. As all of these reactors are now shut down there is a need to examine the concentration and distribution of long lived radioisotopes, such as 14C, to aid in understanding their behaviour in a geological disposal facility. A selection of irradiated graphite samples from Oldbury reactor one were examined where it was observed that Raman spectroscopy can distinguish between underlying graphite and a surface deposit found on exposed channel wall surfaces. The concentration of 14C in this deposit was examined by sequentially oxidising the graphite samples in air at low temperatures (450°C and 600°C) to remove the deposit and then the underlying graphite. The gases produced were captured in a series of bubbler solutions that were analysed using liquid scintillation counting. It was observed that the surface deposit was relatively enriched with 14C, with samples...
Radiocarbon, 2018
Ignalina NPP contains two Units with RBMK-1500 reactors. After shutdown, several Unit 1 systems and equipment were radiologically characterized and dismantled. The highest volume of reactor structures is attributed to the graphite stack of the reactor core, radiological characterization of which has not yet been performed. The stack can be visualized as a vertical cylinder 8 m high and 14 m diameter, made up of 2488 columns where each column is made up from several graphite blocks. The total mass of the graphite stack blocks is about 1700 tonnes. Therefore, the main goal of work reported in this paper was to estimate the inventory of 14 C and other key radionuclides in the irradiated graphite by a combination of activity measurements and full 3D reactor graphite stack neutron activation modeling. Obtained results show that, based on the combination of modeling and measurement techniques, the total inventory of 14 C in graphite stack is estimated at 3.22 × 10 14 Bq at 9 years after Unit 1 reactor final shutdown. 14 C activity is the highest among the analyzed radionuclides; the second highest is 60 Co (~6 times lower).
Journal of Nuclear Materials, 2015
Recent studies have been performed to determine the effectiveness of thermal treatment as a method for removing 14 C contamination from irradiated graphite surfaces. Samples of two grades of irradiated nuclear graphite (NBG-18 and NBG-25) were thermally treated to determine the amount of 14 C contamination on irradiated graphite surfaces. The results of these analyses indicate that specific chemical forms of 14 C (namely, 14 CO and 14 CO 2) may be selectively removed based on the temperature used during thermal treatment. Characterization studies utilizing various surface analysis techniques (XPS, SIMS, SEM/EDS) were employed to investigate the chemical speciation, bond structure, and morphology of the surfaces of pre-and post-thermally treated irradiated graphite.