Hazardous or toxic waste destruction or containment – Destruction or containment of radioactive waste – By fixation in stable solid media
Reexamination Certificate
2001-04-24
2003-02-11
Silverman, Stanley S. (Department: 1754)
Hazardous or toxic waste destruction or containment
Destruction or containment of radioactive waste
By fixation in stable solid media
C588S011000, C588S012000, C588S014000, C588S019000, C588S253000, C588S255000, C588S256000
Reexamination Certificate
active
06518477
ABSTRACT:
FIELD OF THE INVENTION
The present invention relates generally to the remediation of radioactive waste and, more particularly, to the remediation of non-homogeneous radioactive tank waste to greatly reduce the waste volume and to produce end products which meet federal regulatory compliance standards for disposal. Millions of gallons of radioactive waste, the result of plutonium production, are currently buried in single-shell tanks that have exceeded their design life. The tremendous variety of components mixed in these tanks has complicated attempts at remediation of the waste. The remediation method of the present invention utilizes well-known, cost effective, EPA approved treatment processes to substantially reduce the mass/volume of the waste while resulting in end products that meet federal regulatory compliance standards. A thermal desorption-type apparatus adapted for effectively performing the method of the present invention is also disclosed.
BACKGROUND OF THE INVENTION
The production of plutonium for the nation's nuclear defense program has resulted in the storage of millions of gallons of multi-component radioactive waste. This waste is stored in underground tanks at several cites in the United States. Many of the tanks are older, single-shell tanks that have exceeded their design life by over three decades. It is believed that many of the tanks have leaked significant amounts of waste into the ground. The release of this waste may cause radionuclides to reach groundwater. Risks to the environment will increase as more radioactive waste is released from these tanks. A permanent solution, the immobilization of the waste so that the hazardous components of the waste cannot escape into the environment is required.
Conventional methods of remediating radioactive waste are ineffective for handling this waste due to the number of various components, both low activity and high activity, in these tanks. Further, the waste components can vary from tank to tank. Conventional methods are not easily adapted to handle out-of-specification wastes. Specifically, the presence of polychlorinated biphenyls (PCBs) often complicates processing. The detection of a variant in the waste often results in a complete overhaul of the waste management system, costing time and money.
Conventional methods are also often very expensive. In conventional radioactive waste immobilization processes, pretreatment steps, material handling, employee exposure to radiation, volume of waste generated and the complexity of the operation have major impacts on the cost of the process.
Finally, the complexity of the tank waste often leads to less than satisfactory disposal end products. Sodium aluminum silicate glass is conventionally utilized for storage of low activity waste (LAW) and borosilicate glass is conventionally utilized for storage of high level waste (HLW). Both of these disposal methods provide excellent products that meet federal regulatory compliance standards when the proper components are vitrified therein.
Unfortunately, even miniscule amounts of corresponding glass poisons greatly decrease the efficiency and effectiveness of these glasses. Borosilicate glass contaminates include sodium, phosphorus, iron, nickel, and chromium. Removal of these contaminates from the waste is necessary to ensure maximum insolubility to prevent leaching. Sodium aluminum silicate glass is compromised by large organic loadings, PCB's, fluorine, chlorine, and sulfur. Conventional methods of waste remediation do not provide satisfactory separation of the various components of the tank waste to prevent these contaminates from entering the melter.
It is therefore an object of the present invention to provide a radioactive waste remediation method that results in a significant reduction in the total volume and mass of waste which must be immobilized for disposal, thereby reducing disposal and storage costs.
It is a further object of the present invention to provide a method that allows the waste to be pumped straight from the tank, virtually free of any pretreatment steps, thereby reducing costs while minimizing handling and maximizing safety.
It is still a further object of the present invention to provide a method that uses proven industrial grade processes and results in improved separation of specific types of waste to facilitate treatment and disposal.
It is yet a further object of the present invention to provide a method that allows the virtual elimination of sodium from the waste stream prior to vitrification into borosilicate glass.
It is another object of the present invention to provide a method that is simplified and streamlined with a minimum number of steps and a minimum of additions to the waste, but allows for easy adaptation to variations in the tank waste components.
It is yet another object of the present invention to provide an apparatus that is especially well suited to perform the method of the present invention.
SUMMARY OF THE INVENTION
A method for remediating non-homogeneous radioactive waste to significantly reduce the waste mass/volume and to result in products that meet federal regulatory compliance standards is disclosed. The tank waste may include liquid and solid/sludge LAW as well as solid/sludge HLW. The HLW may include low boiling organic material, volatile metals, and heavy metal/transuranic components. In a preferred embodiment, the LAW liquids present in the tank are decanted from the LAW/HLW solids/sludge. The solid/sludge waste is isolated in a thermal desorption-type reaction vessel under reduced pressure and an inert atmosphere to limit or eliminate explosive reactions.
The thermal desorption is performed at a pre-determined and carefully controlled ramp of various combinations of temperatures and pressures. The heating ramp includes at least three distinct temperature phases. The first phase is a temperature necessary to vaporize the low boiling organic components, after which the vaporized organic components are removed from the reaction vessel for off-gas treatment prior to conventional disposal means.
The second phase in the predetermined heating ramp is a temperature necessary to vaporize the volatile metals, after which the vaporized volatile metals are removed from the reaction vessel for off-gas treatment. This treated waste can then be immobilized by conventional methods or by immobilizing in a radiation-shielding polymer.
In the third phase the temperature in the reaction vessel is raised to a temperature necessary to cause pyrolysis of the remaining waste, primarily heavy metal/transuranics. Pyrolysis results in the formation of gaseous nitrogen oxides and leaves a metal oxide ash residue. The gaseous nitrogen oxides are removed from the reaction vessel for off gas treatment and disposal.
The metal oxide ash is then removed from the reaction vessel for treatment. The following procedure for producing products that meet federal regulatory compliance standards is presently preferred. The metal oxides are washed with water to remove water-soluble metal oxides, including sodium, strontium, technetium and cesium. The LAW liquids that were previously decanted from the tank are then added to the wash solution for treatment.
The wash solution is then filtered to remove any solids. Carbon dioxide is then bubbled through the filtered wash solution to precipitate the strontium as strontium carbonate, and hydrazine hydrate is added to reduce any technetium that is present. The wash solution is then decanted from the precipitate, and the precipitate is added to the removed solids and dried for disposal by vitrification into borosilicate glass.
The sodium is removed from the wash solution by diafiltration and reverse osmosis. The sodium is then recovered by drying and is disposed of as sodium carbonate.
The cesium and technetium are then removed from the sodium free wash solution by utilizing a zeolite. The zeolite is dried and disposed of by vitrification into borosilicate glass.
Under this process, only {fraction (1/10)} of the original volume or mass of waste ends u
Blackwell Sanders Peper Martin LLP
Hanford Nuclear Services, Inc.
Nave Eileen E.
Silverman Stanley S.
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