Zeolites / Decontamination / Radioactive waste microfiltration / UltrafiltrationSummary. The purpose of the work was to improve the process for treatment of liquid radioactive waste containing complexing agents, which are generated during the decontamination operations.We performed some experiments using simulated waste solutions like secondary waste from the modified CANDEREM process (Canadian Decontamination and Remediation Process) and secondary waste from the modified CANDECON process (Canadian Decontamination Process).To improve efficiency and economics of the process it was proposed to treat the waste by combining the sorption of radionuclides on natural inorganic sorbents (zeolites) with membrane filtration.Standard procedures are applied to compare the sorption of radionuclides on different sorbent forms-determination of the ion exchange capacity, construction of sorption isotherms, determination of the distribution coefficients, and kinetics experiments.To check the influence of converting the sorbents to various cationic forms on their sorption properties, distribution coefficients of 137 Cs and 57 Co on natural zeolites from local deposits converted to NH 4 + , Na + or H + forms were determined. The results obtained show that the distribution coefficients of 137 Cs on the materials converted to Na + form are higher than for the remaining forms studied [1].The parameters of Langmuir, Freundlich and DubininRadushkevich adsorption isotherms have been determined using sorption data. The Dubinin-Radushkevich model shows better correlation between the theoretical and experimental data for 137 Cs sorption on natural zeolites from local deposits converted to NH 4 + and H + forms than Langmuir and Freundlich equations.Kinetic studies were carried out with various zeolite forms. The sorbents studied are natural zeolites from local deposits (Marsid-Romania).The batch sorption kinetics has been tested for pseudosecond order reaction. The pseudo-second order model fits the experimental data well for all of the systems studied.

The paper presents the experimental tests concerning the treatment by membrane techniques of radioactive aqueous waste. Solutions, which have been treated by using the bench-scale installation, were radioactive simulated secondary wastes from the decontamination process with modified POD. Generally, an increasing of the retention is observed for most of the contaminants in the reverse osmosis experiments with pre-treatment steps. The main reason for taking a chemical treatment approach was to selectively remove soluble contaminants from the waste. In the optimization part of the precipitation step, several precipitation processes were compared. Based on this comparison, mixed [Fe(CN)6]4-/Al3+/Fe2+ was selected as a precipitation process applicable for precipitation of radionuclides and flocculation of suspended solid. Increased efficiencies for cesium radionuclides removal were obtained in natural zeolite adsorption pre-treatment stages and this was due to the fact that volcanic tuff used has a special affinity for this element. Usually, the addition of powdered active charcoal serves as an advanced purifying method used to remove organic compounds and residual radionuclides; thus by analyzing the experimental data (for POD wastes) one can observe a decreasing of about 50% for cobalt isotopes subsequently to the active charcoal adsorption.. The semipermeable membranes were used, which were prepared by the researchers from the Research Center for Macromolecular Materials and Membranes, Bucharest. The process efficiency was monitored by gamma spectrometry.

The method consists in the conditioning of organic solvents solutions contaminated with tritium (acetone, ethylene alcohol, methanol, trichloromethane, toluene, white spirit) into a cement and proper solidification additives matrix. Portland cement is the most common type of hydraulic cement and is the original agent utilized for the solidification of low-level radioactive waste. The paper contents: • the characterization of conditioning matrix components: waste, cement, additives; • preliminary laboratory tests on waste, cement and additives; • mechanical resistance verifications of conditioning matrix (results, tables); • leaching tests for tritium (results, graphs). This method is very simple, with low cost of working materials and long shelf life of working materials. The operator exposure is negligible because there is not any vapor problem. The values of mechanical resistance are higher than the disposal minimal value. Rate of release of tritium from the waste form as a result of interaction with water is very low. The higher values of mechanical resistant and the low rate of tritium released from the waste form offer security for transportation and final disposal of the waste placed in cement.

The immobilization process of radioactive solvent wastes produced during nuclear power plants operation tends to be safe and technically feasible perhaps its most important advantage from the safety point of view is the low chamber temperature and lower explosion risk. Radioactive solvent wastes resulted from Cernavoda NPP decontamination operations consist of miscellaneous acetone, toluene, methanol, chloroform, trichloroethan, white spirit, ethylene glycol and water. Magnesium phosphate binding systems are new cement-materials which could be used for the immobilization of radioactive wastes for the set-retarding effect of solvent radioactive waste in other matrices (Portland cement, composite cement). The paper presents the influence of mineral additive on the properties of solvent waste form, mainly the setting time and the leach rate of tritium. Results of the experiments have shown that using magnesium phosphate binding systems-mineral additive, could do the solidification of solvent radioactive wastes by better results. It can thus be anticipated that the level of expectation towards magnesium phosphate binding systems will remain high and probably increase for the other organic radioactive wastes (oils, scintillation liquids).

Solvent wastes produced at Cernavoda NPP consist of miscellaneous acetone, toluene, methanol, chloroform, triclorethan, white spirit and ethylene glycol. These are normally LLW containing only relatively small quantities of beta/gamma emitting radionuclides and varying amounts of tritium with activity below E08Bq/l. This paper is a review of some current innovative work of Waste Management Facility from Institute for Nuclear Research Pitesti in the development of a viable solidification technology to convert solvent wastes into a stable monolithic form, which minimises the probability to release tritium in the environment during interim storage, transportation and final disposal. The paper presents the author’s research on immobilisation of solvent wastes by cementation using aluminium stearate additive. A quality assurance program should accompany the production of waste forms. The goal of all tests should be to obtain a license for a certain process from a competent authority. The process will be clean, which means there will be no secondary waste and low doses to the personnel will be achieved; the product quality will meet any National requirement and the reproducibility of the process meets any QA requirement.