Chemistry of inorganic compounds – Treating mixture to obtain metal containing compound – Radioactive metal
Reexamination Certificate
1999-10-07
2001-07-10
Griffin, Steven P. (Department: 1754)
Chemistry of inorganic compounds
Treating mixture to obtain metal containing compound
Radioactive metal
C423S002000, C423S008000, C423S021500
Reexamination Certificate
active
06258333
ABSTRACT:
BACKGROUND OF THE INVENTION
The present invention relates to extracting agents and methods for selectively recovering certain radionuclides, rare earths and actinides from radioactive wastes. More specifically, the invention relates to extracting agent compositions comprising complex organoboron compounds, substituted polyethylene glycol and, a neutral organophosphorus compound in a diluent. The invention also provides a method of using extracting agents to recover actinides and certain radionuclides from radioactive waste.
Liquid wastes from nuclear fuel reprocessing are extremely hazardous and expensive to dispose of. Cesium-137, strontium-90, transuranium elements and technetium, which are present in these wastes are of particular concern. The transuranium elements are typically alpha-emitters with very long half-lives, and cesium and strontium are the major heat generators of the waste and produce gamma and beta radiation. Technetium, as the pertechnetate ion, is very mobile in the environment and has an extremely long half life. To increase safe handling of the majority of the waste, and to significantly reduce the storage and/or disposal cost of the waste, it is desirable to partition the waste into two fractions, one containing the majority of the highly radioactive components, and one containing the bulk of the non-radioactive portion of the waste.
Nuclear waste exists in numerous forms and locations world-wide. The largest inventory of highly radioactive materials is produced from the reprocessing of spent nuclear fuel. The fission process produces a number of undesirable, highly-radioactive elements which accumulate in the nuclear fuel. For the reuse or recycling of the unused fissionable material left in the fuel, normally either uranium-235 or plutonium-239, a separation process is employed to partition the fissionable material from the undesirable fission products. This is normally accomplished by leaching or dissolution of a portion or all of the spent nuclear fuel material, followed by chemical separation. Early chemical separation processes were based on precipitation, where, for example, BiPO
4
was used to coprecipitate plutonium for weapons-grade plutonium production. More recently, and by far the most common, solvent extraction processes utilizing tri-n-butyl phosphate are used to chemically separate uranium and/or plutonium from solutions resulting from dissolution or leaching of spent nuclear fuel. The remaining acidic liquid waste, containing the highly radioactive fission products and trace transuranic elements, has been accumulated and stored in various forms around the world for the past 45 years.
Facilities in the United Kingdom, France, Japan, Russia, and China currently use the PUREX process (plutonium-uranium extraction) to recover and purify uranium and/or plutonium. The United States utilized this process (or variations of this process) for commercial-fuel reprocessing at the West Valley Plant, in upstate New York in the 1970's, and for reprocessing of weapons-grade plutonium at the Hanford Site in eastern Washington State, aluminum driver fuel at Savannah River Site in South Carolina, and naval fuel at the Idaho Chemical Processing Plant at the Idaho National Engineering and Environmental Laboratory (INEEL) in eastern Idaho. The reprocessing activities at Hanford were discontinued in the late 1980's and operations at Savannah River and the INEEL were discontinued in the early 1990's. There is currently no active nuclear fuel reprocessing facility in the United States. There is however, a significant legacy of nuclear waste in storage from previous reprocessing activities. The majority of this waste was neutralized with caustic to facilitate storage in carbon steel vessels (Hanford and Savannah River). At the Idaho Chemical Processing Plant, this waste was calcined in a fluidized-bed calciner at 500° C., producing a granular solid. This solid calcine is stored in stainless steel bins inside concrete vaults. Currently about 4100 m
3
of highly radioactive calcine is stored at the INEEL, and about 1.2 million gallons of acidic liquid waste.
Currently, separate technologies are required for removing actinides and fission products from the wastes, and often times, separate processes may be required for specific radionuclides such as cesium, strontium and technetium.
The invention describes novel extraction processes that will readily meet current safety standards and that will effectively separate the above-mentioned radioactive elements from typical nuclear reprocessing wastes. The liquid waste can be effectively decontaminated to meet low level waste standards in one simultaneous solvent extraction process. This process offers the advantage of completing the decontamination of the waste in a single process using a novel solvent, which will significantly reduce capital and operating costs.
U. S. Pat. No. 4,749, 518 (Davis) teaches a method for reprocessing nuclear waste by extracting cesium and strontium with crown compounds and cation exchangers.
U. S. Pat. No. 5,603,074 (Miller et al.) teaches a method of recovering cesium and strontium from an aqueous solution using a cobalt dicarbollide derivative.
U. S. Pat. No. 5,666,641 (Abney et al.) discloses a method of recovering cesium and strontium from an aqueous solution with polymeric materials and derivatives of cobalt dicarbollide.
U. S. Pat. Nos. 5,666,642 and 5,698,169 (Hawthorne et al.) teach the extraction of cesium and strontium from aqueous solutions, including fission product waste, using substituted metal dicarbollide ions.
To date, known technologies have required separate techniques for separating actinides and fission products. However, Applicants have surprisingly discovered a novel extracting agent solvent system for the selective and simultaneous recovery of radionuclides and actinides. Prior art methods were all very cost intensive and involved two or three separate processes. The process of the invention is cost effective and separates the actinides and fission products in one process.
SUMMARY OF THE INVENTION
The present invention provides for an extracting agent composition and a method for the selective sequential recovery of radionuclides, rare earths, and actinides using a novel solvent system.
One embodiment of the invention relates to an extracting agent composition for extracting specific radionuclides, rare earths, and actinides from a liquid radioactive waste comprising a complex organoboron compound, an unsubstituted or substituted polethylene glycol, and a neutral organophosphorus compound.
Another embodiment of the invention relates to an extracting agent composition for extracting specific radionuclides, rare earths, and actinides from a liquid radioactive waste comprising a chlorinated cobalt dicarbollide, polyethylene glycol-400, diphenyl-dibutylcarbamoylmethylenephosphine oxide in a diluent such as phenyltrifluoromethyl sulfone.
Another embodiment of the invention relates to a method for simultaneously recovering radionuclides, specifically cesium and strontium, rare earths, and actinides from liquid radioactive waste comprising contacting the liquid radioactive waste with a solution of a complex organoboron compound, an unsubstituted or substituted polyethylene glycol, a neutral organophosphorus compound and a diluent.
Another embodiment of the invention relates to a method for simultaneously recovering radionuclides, specifically cesium and strontium, rare earths, and actinides from liquid radioactive waste comprising contacting the liquid radioactive waste with a solution of a chlorinated cobalt dicarbollide, polyethylene glycol-400, diphenyl-dibutylcarbamoylmethylene phosphine oxide, in a diluent such as phenyltrifluoromethyl sulfone.
Additional advantages of the invention will be set forth in part in the description which follows, and in part will be obvious from the description, or may be learned by practice of the invention. The advantages of the invention will be realized and attained by means of the elements and combinations particularly po
Babain Vasiliy A.
Brewer Ken N.
Romanovskiy Valeriy Nicholiavich
Smirnov Igor V.
Todd Terry A.
Caress Virginia B.
Gottlieb Paul A.
Griffin Steven P.
Ildebrando Christina
The United States of America as represented by the United States
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