Promising Methods of Spent Ion-Exchange Resins Treatment (original) (raw)
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Treatment and Conditioning of Spent Ion Exchange Resin from Nuclear Power Plant
There are a number of liquid processes and waste streams at nuclear facilities (i.e. nuclear power plants, fuel reprocessing plants, nuclear research centers, etc.) that require treatment for process chemistry control reasons and/or the removal of radioactive contaminants. These processes may be for reactor primary coolants, the cleanup of spent fuel pools, liquid radioactive waste management systems, etc. One of the most common treatment methods for such aqueous streams is the use of ion exchange, which is a well developed technique that has been employed for many years in both the nuclear industry and in other industries. Nuclear power plant process water systems have typically used organic ion exchange resins to control system chemistry to minimize corrosion or the degradation of system components and to remove radioactive contaminants. Organic resins are also used in a number of chemical decontamination or cleaning processes for the regeneration of process water by reagents and ...
ICONE19-43160 Development of Spent Ion Exchange Resin Processing in Nuclear Power Stations
The Proceedings of the International Conference on Nuclear Engineering (ICONE)
Commercial operation of the Tsuruga nuclear power station Unit 1, owned by the Japan Atomic Power Company (JAPC), will be terminated in 2016. For safe decommissioning of the station, technologies for processing stored radioactive wastes such as spent Ion EXchange resin (IEX) and Filter Sludge (FS) have been jointly developed by JAPC and JGC Corporation. The Wet-Oxidation process (WetOx) was applied for decomposition of the spent IEX and FS, and Super Cement (SC) solidification was chosen for immobilization of the WetOx residue. Pilot scale tests for the WetOx treatment have been successfully conducted with simulated wastes. The WetOx residue is a concentrate with Suspended Solid (SS) and sodium sulfate, and was solidified by SC in lab scale cementation apparatus. The compressive strength of the solidified waste was confirmed to meet the desired level for radioactive waste disposal.
Immobilization of Spent Ion Exchange Resin Arising From Nuclear Power Plants: An Introduction
J. Pakistani Mater. Sci, 2009
Ion exchange resins are used for purification of radioactive waste waters in the nuclear industry. The process involves removal of radioactive nuclides and other hazardous contaminants that could potentially harm the equipment or corrode reactor fuel rods. These resins have to be replaced periodically with clean ones to continue the purification process and dispose of the spent resins. Disposal often becomes uneconomic because of the large volume of the resin produced and the relatively few technologies capable of economically stabilizing this waste. Various methods to treat liquid radioactive waste in the reactor fuel pond using ion exchange resins have been reviewed in this paper. The potential of verification to immobilize and contain the spent ion exchange resin for long term disposal using novel borosilicate glass has been discussed.
Integrated S ystem for Spent Ion Exchange Resin Processing in Nuclear Power Stations - 11164
2011
The Tsuruga nuclear power station Unit 1 owned by the Japan Atomic Power Company (JAPC) will terminate its commercial operation in 2016. For a safe decommissioning of the station, JAPC and JGC Corporation have investigated the processing of stored radioactive wastes such as spent ion exchange resin (IEX) and filter sludge (FS). A unique wet-oxidation process operating at 100 C under atmospheric pressure will be applied for decomposing organic substances in the waste. The treated waste containing radioactive species will undergo cementation using our Super Cement (SC) solidification process. For the treatment of high activity waste, the combined process of wet oxidation and SC solidification provides advantages over conventional incineration and direct solidification methods; for example, it can be operated environmentally safely under milder operating conditions with simple equipment and has the capability to reduce the volume of the waste material. It is expected that the combined...
Experimental Study of Thermochemical Treatment of Spent Ion- Exchange Resins
2004
A thermochemical treatment technique was investigated for application to the volume reduction of spent ion-exchange resins of nuclear power plants. The thermochemical treatment technique uses powder metal fuel to incinerate the resins. Insignificant amounts of soluble species were found in the slag from the thermochemical treatment of H-type cation exchange resins. The slag could be solidified firmly with magnesium-potassium-phosphate ceramics. The volume ratio of initial resin to the final ceramics form was about 10. Leaching rates of radionuclides from the ceramics were low. Thus the thermochemical treatment technique was confirmed to applicable to the volume reduction of spent ion-exchange resins.
Nuclear Technology, 2010
Incineration of spent ion exchange resin was simulated using the ChemSheet chemical calculation program. The simulation of the incineration was modeled for typical spent resin produced by pressurized water reactors (PWRs) and boiling water reactors (BWRs) in Finland. The objective of the study was to find the volume and mass reduction and the chemical compounds formed during incineration. The simulation showed that active elements did not play any role in incineration owing to small amount of Cs, Co, etc. The ash contained metal oxidesmainly hematite, iron oxide Fe 2 O 3. Other products of the incineration were water, carbon dioxide, sulfuric acid, and nitrogen oxides. The volume reductions 1/100 and 1/14 of the spent resin were obtained for PWRs and BWRs, respectively. The annual ash production from incineration was calculated to be 408 kg and 746 kg for the currently operating Finnish PWR and BWR plants in Loviisa and Olkiluoto, respectively.
Pyrolysis of Spent Ion Exchange Resins - 12210
2012
Organic ion exchangers (IEX) play a major and increasing role in the reactor coolant and other water purification processes. During their operation time they receive significant amounts of radioactivity, making their disposal, together with their organic nature, as medium active waste challenging. Processes applied so far do not eliminate the organic matter, which is unwanted in disposal facilities, or, if high temperatures are applied, raise problems with volatile radionuclides. NUKEM Technologies offers their well introduces process for the destruction of spent solvent (TBP), the pebble bed pyrolysis, now for the treatment of spent IEX (and other problematic waste), with the following benefits: the pyrolysis product is free of organic matter, and the operation temperature with approx. 500 deg. C keeps Cs radionuclides completely in the solid residue. (authors)
2015
In this work we are interested in the physical, chemical and radiological characterization of Ion Exchange Resins (IER) (MBD-15). These are polymeric beads used in the purification of nuclear reactor water circuits. The physicochemical characterization of the latter is to determine the moisture content and the adsorbed metal elements, and to study the evolution of pH and conductivity as a function of the stirring time. As for the radiological characterization involves the quantification and identification of radionuclides adsorbed by the resin. The results obtained show that the studied resin has moisture content in the vicinity of 63.07%, the presence of some metal ions such as B, Al, Fe, Si and Sb, and optionally a very low radioactivity (K40).
Journal of Nuclear Science and Technology, 2006
A series of separation experiment was performed in order to study a multi-functional spent fuel reprocessing process based on ion-exchange technique. The tertiary pyridine-type anion-exchange resin was used in this experiment and the mixed oxide fuel highly irradiated in the experimental fast reactor ''JOYO'' was used as a reference spent fuel. As the result, 106 Ru+ 125 Sb, 137 Cs+ 155 Eu+ 144 Ce, plutonium, americium and curium could be separated from the irradiated fuel by only three steps of ion-exchange. The decontamination factor of 137 Cs and trivalent lanthanides ( 155 Eu, 144 Ce) in the final americium product exceeded 3:9Â10 4 and 1:0Â10 5 , respectively. The decontamination factor for the mutual separation of 243 Cm and 241 Am was larger than 2:2Â10 3 for the americium product and, moreover, the content of 137 Cs, trivalent lanthanides and 243 Cm included in 241 Am product did not exceed 2 ppm.
Permissible radionuclide loading for organic ion exchange resins from nuclear power plants
1983
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