A composite panel for a toxic material encapsulation system, comprising a reinforcing structure extending within and integrally formed with a non-biodegradable thermoplastic polymer.

The invention relates to an encapsulation composition for the storage or the confinement of waste which is toxic to health and/or the environment, comprising a resin composition containing at least one epoxy resin, and a hardening composition containing at least one polyamidoamine and at least one aromatic polyamine, said encapsulation composition having an aromaticity rate which is equal to, or higher than, 35%. The invention also relates to the use of said composition for encapsulating said waste.

The present invention is a method for eliminating radioactive iodine using a hydrophilic resin that adsorbs radioactive iodine, wherein the hydrophilic resin is at least one selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin and has a hydrophilic segment and, in the principal chain and/or a side chain in the structure thereof, has a tertiary amino group or has a tertiary amino group and polysiloxane segment. By means of the present invention, a novel method for eliminating radioactive iodine is provided that is simple and low-cost, furthermore does not require an energy source such as electricity, moreover can take in and stably immobilize the eliminated radioactive iodine within a solid, and is capable of reducing the volume of radioactive waste as necessary.

The present invention is a method for eliminating radioactive iodine using a hydrophilic resin that adsorbs radioactive iodine, wherein the hydrophilic resin is at least one selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin and has a hydrophilic segment and, in the principal chain and/or a side chain in the structure thereof, has a tertiary amino group or has a tertiary amino group and polysiloxane segment. By means of the present invention, a novel method for eliminating radioactive iodine is provided that is simple and low-cost, furthermore does not require an energy source such as electricity, moreover can take in and stably immobilize the eliminated radioactive iodine within a solid, and is capable of reducing the volume of radioactive waste as necessary.

A composition and method for spray-applying a two-part, self-setting composition containing a dopant that provides a hazard shielding component and is particularly adapted for delivering the components of the composition at a temperature that promotes their spray application as well as a self-setting reaction. The method includes selecting a self-setting compound that is adapted for curing in place once applied, the self-setting compound including at least one dopant material; and applying the compound to a hazard to be encapsulated such as a radiological, lead, asbestos, or PCB. Alternately, a self-curing compound includes a multi-part compound which, upon a mixing of the parts, chemically reacts and cures, and at least one dopant material dispersed into at least one of the parts, wherein the dopant material is selected for providing radiation shielding upon application of the compound.

A method for removing radioactive cesium and/or iodine from a radioactive substance in liquid and/or a solid matter using a hydrophilic resin composition comprising a hydrophilic resin and a metal ferrocyanide compound, wherein the hydrophilic resin includes at least one hydrophilic resin selected from the group consisting of a hydrophilic polyurethane resin, a hydrophilic polyurea resin, and a hydrophilic polyurethane-polyurea resin each having at least a hydrophilic segment, and a metal ferrocyanide compound is dispersed in the hydrophilic resin composition in a ratio of at least 1 to 200 mass parts relative to 100 mass parts of the hydrophilic resin.

Methods are provided for separating radioisotopes from spent ion exchange resins used to process nuclear waste. In some embodiments, the resin is contacted with an aqueous solution including one or more acids to elute the radioisotopes into an enriched solution. In some embodiments, the resin is contacted with an aqueous solution including one or more salts, such as a sulfate or nitrate salt solution, to elute the radioisotopes into an enriched salt solution. The enriched solution can include at least 80%, or at least 95%, of the activity of radioisotopes such as Sr-90, Co-60, Ni-63, Cs-137, C-14, Co-58, and/or Mn-54 originally present in the resin. In some embodiments, a mixed bed resin is separated into cation and anion components, and the anion resin is treated to release C-14 as carbon dioxide gas which is captured and purified to obtain a C-14 product.

Methods are provided for separating radioisotopes from spent ion exchange resins used to process nuclear waste. In some embodiments, the resin is contacted with an aqueous solution including one or more acids to elute the radioisotopes into an enriched solution. In some embodiments, the resin is contacted with an aqueous solution including one or more salts, such as a sulfate or nitrate salt solution, to elute the radioisotopes into an enriched salt solution. The enriched solution can include at least 80%, or at least 95%, of the activity of radioisotopes such as Sr-90, Co-60, Ni-63, Cs-137, C-14, Co-58, and/or Mn-54 originally present in the resin. In some embodiments, a mixed bed resin is separated into cation and anion components, and the anion resin is treated to release C-14 as carbon dioxide gas which is captured and purified to obtain a C-14 product.

Methods are provided for separating radioisotopes from spent ion exchange resins used to process nuclear waste. In some embodiments, the resin is contacted with an aqueous solution including one or more acids to elute the radioisotopes into an enriched solution. In some embodiments, the resin is contacted with an aqueous solution including one or more salts, such as a sulfate or nitrate salt solution, to elute the radioisotopes into an enriched salt solution. The enriched solution can include at least 80%, or at least 95%, of the activity of radioisotopes such as Sr-90, Co-60, Ni-63, Cs-137, C-14, Co-58, and/or Mn-54 originally present in the resin. In some embodiments, a mixed bed resin is separated into cation and anion components, and the anion resin is treated to release C-14 as carbon dioxide gas which is captured and purified to obtain a C-14 product.