C Leaching and Speciation Studies on Irradiated Graphite from Vandellós I Nuclear Power Plant (original) (raw)
Related papers
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 ...
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...
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...
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.
The Location of Radioisotopes in British Experimental Pile Grade Zero Graphite Waste
SECURING THE SAFE PERFORMANCE OF GRAPHITE REACTOR CORES, 2008
The UK has approximately 90’000 tonnes of irradiated graphite waste accumulated since the 1940s from over 40 nuclear reactors. In order to make an informed decision as to how to best deal with this waste, information on the activation and location of impurities contained within the graphite porous structure is required. In addition possible mechanisms that may lead to the release of these isotopes must also be well understood, not only to assess the possibility of release after disposal but also to consider it may be possible to “clean” the graphite using thermal or chemical treatment thus significantly reducing the activity. The activities of isotopes contained within nuclear graphite may be theoretically calculated from the trace elemental impurities present within virgin graphite material and the cross sectional areas of these elements. This combined with reactor operational conditions provides background to the isotopic inventory currently accepted. However, other isotopes may arise from impurities trapped in the porous graphite during reactor operation. These activated impurities will need to be accounted. This paper presents microstructural and radiochemical techniques used to quantify the isotopic location and distribution within the graphite. These impurities have been characterised in terms of location and retention using high resolution techniques such as Scanning Electron Microscopy, Raman, micro X-ray Tomography and Energy Dispersive X-ray Spectroscopy.
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.
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 ...
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...
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.