14C leaching and speciation studies on Irradiated graphite from vandellós I Nuclear Power Plant (original) (raw)
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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...
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...
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...
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...
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 ...
14C in Radioactive Waste for Decommissioning of the Ignalina Nuclear Power Plant
Radiocarbon, 2013
Radiocarbon is one of the most significant radionuclides affecting the safety margins of near-surface repositories for the disposal of low-and intermediate-level, short-lived radioactive waste, arising from the operation and decommissioning of nuclear power plants (NPPs). One of the goals of the present study was to characterize radioactive waste from Ignalina NPP (Lithuania) (storage tanks TW18B01 and TW11B03) from the spent ion-exchange resins/perlite stream to determine the 14 C-specific activity of inorganic and organic carbon compounds. The approach applied is based on classical radiochemical separation methods, including acid-stripping techniques and wet oxidation with subsequent catalytic combustion. The suitability of the method for 14 C-specific activity determination in ion-exchange resin samples with a minimum detectable activity of 0.5 Bq/g by liquid scintillation counting (LSC) was demonstrated. The extraction efficiency of inorganic and organic carbon compounds based on model samples with known 14 C activity was estimated. The fraction of 14 C associated with organic compounds ranged from 42% to 63% for storage tank TW18B01 and from 30% to 63% for storage tank TW11B03. The specific activity of inorganic 14 C was estimated as 12.6 Bq/g with a relative standard deviation (RSD) of 29% for storage tank TW18B01, and 177.5 Bq/g with a RSD of 35% for storage tank TW11B03. Based on volume and density data, the total 14 C activity for radioactive waste stored in tanks TW18B01 and TW11B03 was estimated as 3.59E + 10 Bq (±32%) and 4.15E + 11 Bq (±28%), respectively.
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.
14C Exposure from Disposal of Radioactive Waste Compared to 14C Exposure from Cosmogenic Origin
Radiocarbon, 2018
The potential 14 C (carbon-14, radiocarbon) flux from disposal of 14 C containing waste into air is compared with the natural 14 C emanation rate from soil in order to put the 14 C hazard potential from disposal of this waste in perspective with the 14 C exposure from cosmogenic origin. Chemical corrosion of neutron irradiated metals, steel and Zircaloy, is bounded by diffusion of water through a thermodynamically stable metal-oxide layer and dissolution of this metal-oxide in a nuclear plant. Many countries process radioactive waste for disposal using cementitious materials, an acknowledged end-point management technique for this waste. The metal-oxides are also stable when these waste forms are embedded in cementitious materials. The 14 C release rate from this Zircaloy at these alkaline and reducing conditions is comparable to the natural 14 C emanation rate from soil into air. Neutron irradiated graphite and spent ion exchange resins are chemically inert and therefore other release mechanisms need to be assumed. Radiolytic corrosion is used to determine the 14 C release rate from this graphite. Moreover, ion exchange-with ingressing anionic species that have a higher affinity than contained anionic 14 C-is proposed as a release mechanism for these resins.