Decontamination & decommissioning
Grappling with graphite20 February 2009
The properties of graphite, and therefore the resulting reactor decommissioning and waste management requirements, can vary significantly from one reactor to another. By Christiane Bruynooghe and Michel Bièth
Graphite management is a key issue when it comes to the decommissioning of older nuclear reactors. Recently it was one focus of a two-year European Commission Tacis project (R2.04/03) carried out to assist Russia. The project was implemented through a consultant consortium comprising French utility EDF and Energiewerke Nord GmbH (EWN) of Germany, with part of the work subcontracted to VNIIAES (All-Russian Scientific Research Institute for Nuclear Power Plant Operations).
As part of the project, a three-day seminar was organised to discuss: dismantling programmes for graphite moderated reactors and the characterisation, treatment, packaging, and long-term storage of irradiated graphite. A tour of the Bugey 1 reactor in Lyon – currently undergoing decommissioning – was also arranged. This 540MWe gas-cooled graphite-moderated reactor, owned by EDF, has been shutdown since 1994.
The project had two main objectives: to provide feedback on the dismantling of large irradiated graphite structures; and the long-term storage of the resulting material. Its main conclusions, including those on graphite characterisation, packaging, and treatment are presented below.
Radiochemical analysis of the graphite in Bugey 1 (which did not experience fuel leakages) was carried out by EDF. Measurements were taken from sample graphite cores, extracted from graphite bricks in the reactor by drilling off fuel channels. The results show that tritium (3H) is the dominant radionuclide followed by 14C, which unlike many other radionuclides follows a cosine-like activity curve.
The specific activity of 36Cl was measured at different axial heights in eleven channels. Analysis of this data showed a widespread distribution of the radionuclide (half-life 3.01x105 years).
The specific activities of 94Nb and 93Mo were also measured. These long-life radionuclides, with half-lives of around 20,000 and 4000 years, respectively will have an impact on long-term dose around any final repository.
A very accurate measurement of the specific activity of 94Nb was recently carried out at a CEA laboratory.
An attempt at a similarly accurate measurement for 10Be (half-life 1.51 x 106 years) is now in progress.
Results exhibit wide spreads: 14C most often follows fluence, 36Cl most often does not, and correlations cannot be derived. In areas that experienced flat flux typical nuclide concentrations across the area may vary by a factor of ten, with 36Cl varying even more.
The following conclusions were drawn from these investigations. First, sampling of graphite waste for characterisation should involve at least forty core samples drilled from each graphite stack and cover its whole height and radius. Second, extrapolation from one reactor to another is not recommended since stack material as well as operating conditions can vary.
Comparison with calculations
Specific activity calculations were carried out from neutron flux maps by CEA (Neutron Transport code Tripoli 4) and by using activity code (Darwin-PEPIN2). For these calculations the concentrations of nitrogen and chlorine impurities were taken into account. Nevertheless, the calculation/measurement or C/M ratios for the analysed samples were far from one. The C/M ratio ranged from 0.4 to 5.0 for 14C, from 0.01 to 2.0 for 36Cl and for tritium reached values of over 200 (range 4-220).
The discrepancies seen for 14C and 36Cl can be partly corrected through impurity concentration adjustments in the calculation. But the values obtained for 3H show that the calculation greatly overestimates tritium concentration in the graphite stack, probably because it doesn’t appropriately consider leakage. The C/M ratio for the fission product 137Cs was calculated to be in the range 5-8.
These results show the difficulty of building radiochemical inventories through mesh-by-mesh calculation of activities. The calculations make the assumption that impurities are evenly distributed in graphite. However it was recognised that this basic assumption is not based on fact and therefore rationales are needed to account for unexpected distribution of impurities through contamination, degassing, corrosion etc. Furthermore, radiological activities should be assessed in areas that are out of drilling reach (such as the reflector, biological screen) and the dispersion of radionuclides needs to be analysed using statistical methods.
The radiological character of graphite strongly influences the design of the packaging for its final disposal. Prototype packaging was developed for the Bugey 1 graphite during the project. As the repository characteristics have not been finalised, packaging was designed to comply with Andra (the French national agency for radioactive waste management) standards for low and medium level short life waste.
The thickness of the package walls was determined by the mechanical resistance (for the bottom layers of packages) or to meet radioprotection requirements (for top and side layers). To meet the mechanical criteria a wall thickness of 20cm was required. The radioprotection requirement – a maximum dose rate of 2mSv/h at package contact – was satisfied with a wall 30cm thick.
Radioprotection criteria are essentially dictated by the presence of 60Co. Its maximum specific activity by 2013 (planned opening date for Andra’s graphite disposal facility) in the uncontaminated graphite is estimated to be between 104 and 105Bq/g. The 30cm thick concrete walls ensure the required radioprotection up to a specific activity of 2x105Bq/g. Long-lived radionuclides (eg 36Cl, 14C) dictate the confinement rules while tritium releases have the greatest environmental impact.
Contaminated graphite from St Laurent 1 & 2 displayed higher figures: 2x105Bq/g and 5x105Bq/g, respectively, again showing that properties are reactor dependent.
Leaching tests were carried on graphite cores and a large number of parameters were shown to affect the results including: graphite porosity; radionuclide localisation and chemical speciation; water chemistry; temperature and pressure. This makes further theoretical simulation questionable.
The physical and mechanical properties of graphite were also investigated. Some 425 graphite core samples drilled from Bugey 1 were tested to determine mechanical and compressive strengths. The large number of samples enabled a ‘strength versus weight loss’ relationship to be determined. It showed a decrease in compressive strength with irradiation.
The applicability of this relationship to other reactors was tested, but different behaviour was observed. No decrease in compressive strength was found when seven samples from St Laurent 2 were tested. This can be explained by the fact that the Bugey 1 reactor operated with a slightly higher flux – especially in the fast neutron range – and with a higher coolant pressure than the other five reactors of EDF’s gas-cooled reactor fleet. This led to huge differences in weight loss (up to 40% at Bugey 1 compared with 8% for St Laurent 2) and loss of compressive strength (up to 78% for Bugey 1 compared with none for St Laurent 2).
Finally, the risk of graphite fire and dust explosion was investigated. It was concluded that this risk is minimal and can be ignored – a result of general significance since it is not linked to graphite of a specific reactor.
Concern Energoatom investigated methods of sorting and treating graphite, in an attempt to lower repository costs.
Graphite waste contaminated by fuel leakage is classified as high-level waste, containing fissile materials. In accordance with the Russian regulatory requirements, such waste must be immobilised ie converted to a stable physical and chemical form. A particular feature of such waste is the substantially high ratio of graphite volume (or mass) to fissile material volume (or mass) – in the range of 103 to 104. In order to reduce the volume of waste, Concern Energoatom has chosen to separate graphite contaminated by fuel leakage from the rest of the graphite. Characterisation of the contaminated graphite was carried out under a contract with the NIKIET.
The physical and chemical forms of fuel contaminants in graphite can range from metals to oxides, hydroxides and other more complicated compounds. The medium used for decontaminating the graphite should retain as much of the radionuclide content as possible, particularly the actinides which contribute most to the radiological risk. The proposed treatment is to use Molten Salt Oxidation (MSO) which according to NIKIET’s experience achieves: radionuclide and heavy metal retention up to 99.9%; relatively low operating temperatures (750-900oC) when compared with combustion of graphite in a flame furnace; and reprocessing of graphite without preliminary crushing. Furthermore, actinide oxidation and subsequent vitrification of molten salt containing materials from fuel leakages could both be carried out at the same MSO plant.
It is estimated that by using this method the amount of radioactive graphite waste (HLW-LL) from Russia’s two AMB reactors could be reduced from 4600m3 to just 5m3. Significant savings could therefore be made in repository costs as untreated graphite waste must be disposed of in a deep geological repository, at an estimated cost nine times higher than that for near surface disposal.
Areas of possible further cooperation between the consultant consortium and the Russians have been identified. They include: design of MSO plant and gas cleaning system; development of robotic equipment for collection of graphite waste and separation of high-level and medium-level waste; development of devices for loading of high-level graphite waste into the MSO plant; and development of a process for carbon dioxide collection.
Thanks to the project, information was shared about a methodology for the systematic characterisation of graphite. Numerous test results have shown that the radiological and mechanical properties of graphite are very reactor dependent. Unless fresh graphite composition and reactor operating conditions were very similar, which is mostly not the case, theoretical simulation (based on sample data applied to the whole core geometry) is only acceptable for a specific reactor and is not transferable from one reactor to another. This result must be taken into account when preparing for decommissioning and designing packaging.
Further research needs identified by the consultant consortium mainly cover graphite characterisation issues such as: radionuclide distribution within graphite stacks and radionuclide migration out of the graphite. Also, as the mechanical strength of Bugey 1 graphite was much lower than in the other five French GCRs, its stack needs further investigation. Since this behaviour is probably related to irradiation conditions this result may have implications in assessing the mechanical condition of RBMK graphite stacks. With regard to actinide separation from contaminated graphite, several areas of possible further collaboration have been identified, all concerning the Molten Salt Oxidation technology.
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