14C and Other Radionuclides in Impermeable Graphite Material Waste form Long Term Behavior (original) (raw)
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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 ...
C Leaching and Speciation Studies on Irradiated Graphite from Vandellós I Nuclear Power Plant
2018
The understanding of the C 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 C from irradiated graphite samples in the aqueous and gaseous phase is also described. The main results obtained shows that...
2012
Spain has to manage about 3700 tons of irradiated graphite from the reactor Vandellos I as radioactive waste. 2700 tons are the stack of the reactor and are still in the reactor core waiting for retrieval. The rest of the quantities, 1000 tons, are the graphite sleeves which have been already retrieved from the reactor. During operation the graphite sleeves were stored in a silo and during the dismantling stage a retrieval process was carried out separating the wires from the graphite, which were crushed and introduced into 220 cubic containers of 6 m 3 each and placed in interim storage. The graphite is an intermediate level radioactive waste but it contains long lived radionuclides like 14 C which disqualifies disposal at the low level waste repository of El Cabril. Therefore, a new project has been started in order to investigate two new options for the management of this waste type. The first one is based on a selective decontamination of 14 C by thermal methods. This method is ...
Impermeable Graphite: A New Development for Embedding Radioactive Waste
ASME 2010 13th International Conference on Environmental Remediation and Radioactive Waste Management, Volume 1, 2010
Graphite is a geological stable material proven by its natural occurrence. However its porous structure affects the possible use of graphite as long term stable waste matrix for final disposal because slow corrosion in aquatic phases can be induced by high irradiation dose rates in the range of 2 kGy/hr and resulting in corrosion rates of 10-5 to 10-7 gm-2 d-1. The porous structure is accompanied by a large surface area and therefore radiation induced corrosion processes cannot be neglected for final disposal. Furthermore aqueous phases will penetrate into the pore system and radionuclides adsorbed on the surface will be dissolved. 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 therefore a negligible porosity. An initial calculation predicts that the lifetime of the graphite will be at least 2 orders of magnitude greater than porous graphite. This material represents a long term stable leach resistant matrix for the embedding of irradiated graphite (i-graphite). Granulated i-graphite is mixed with natural graphite and an inorganic binder and pressed into a block. Other radioactive wastes can be embedded in this matrix, e.g. coated particles of spent fuel from high temperature reactors (HTR) reactors.
Characterisation and treatment of irradiated graphite waste
Decommissioning of the UK's Magnox and Advanced Gas Cooled reactors, research reactors and plutonium production reactors will produce approximately 90,000 tonnes of graphite waste, this by far the greatest volume produced worldwide from one country. The management of this graphite waste will require complex planning and consideration due to the volume and nature of the material, therefore radioactive graphite core dismantling and the management of radioactive graphite waste is an important issue in the UK. The current UK plan for graphite disposal is that this material will be packaged as Intermediate Level Waste (ILW) for deep geological repository disposal. It has not yet been shown that the current baseline represents the optimum solution in terms of safety, cost and protection of the environment. In order to evaluate the possibility of reducing costs, it is important to consider alternative decommissioning methods. To make informed decisions of how best to dispose of large volumes of irradiated graphite waste, it is necessary to understand fully its microstructural and radioisotopic character and the consequent effectiveness of the various proposed preparative treatment options. The aim of this research project is to develop and demonstrate advanced immobilisation a spectrum of graphite wastes, in order to reduce this volume of waste ILW to that of Low Level Waste.
MRS Proceedings, 2014
ABSTRACTAn overview is given of an International Atomic Energy Agency Coordinated Research Project (CRP) on the treatment of irradiated graphite (i-graphite) to meet acceptance criteria for waste disposal. Graphite is a unique radioactive waste stream, with some quarter-million metric tons worldwide eventually needing to be disposed of. The CRP has involved 24 organizations from 10 Member States. Innovative and conventional methods for i-graphite characterization, retrieval, treatment and conditioning technologies have been explored in the course of this work, and offer a range of options for competent authorities in individual Member States to deploy according to local requirements and regulatory conditions.
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...
Management of Radioactive Waste Containing Graphite: Overview of Methods
Energies, 2020
Since the beginning of the nuclear industry, graphite has been widely used as a moderator and reflector of neutrons in nuclear power reactors. Some reactors are relatively old and have already been shut down. As a result, a large amount of irradiated graphite has been generated. Although several thousand papers in the International Nuclear Information Service (INIS) database have discussed the management of radioactive waste containing graphite, knowledge of this problem is not common. The aim of the paper is to present the current status of the methods used in different countries to manage graphite-containing radioactive waste. Attention has been paid to the methods of handling spent TRISO fuel after its discharge from high-temperature gas-cooled reactors (HTGR) reactors.
2012
Worldwide , more than 250,000 tonnes of irradiated graphite have arisen through commercial nuclear-power operations and from military production reactors. Whilst most nations responsible for the generation of this material have in mind repository disposal alongside other radwaste, the lack of progress in this regard has led in some cases to difficulties where, for example, the site of an existing graphite-moderated reactor is required for re-utilisation. In any case, graphite as a radwaste stream has unique chemical and physical properties which may lend itself to more radical and innovative treatment and disposal options, including the recovery of useful isotopes and also recycling within the nuclear industry. Such aspects are important in making the case for future graphite-moderated reactor options (for example, High-Temperature Reactors planned for simultaneous power production and high-grade heat sources for such applications as hydrogen production for road fuel). A number of initiatives have taken place since the mid 1990s aimed at exploring such alternative strategies and, more recently, improving technology offers new options at all stages of the dismantling and disposal process. A new IAEA Collaborative Research Program aims to build upon the work already done and the knowledge achieved, in order to identify the risks and uncertainties associated with alternative options for graphite disposal, along with cost comparisons, thus enabling individual Member States to have the best-available information at their disposal to configure their own programs.