Chemistry of inorganic compounds – Treating mixture to obtain metal containing compound – Radioactive metal
Reexamination Certificate
1999-03-01
2002-10-29
Griffin, Steven P. (Department: 1754)
Chemistry of inorganic compounds
Treating mixture to obtain metal containing compound
Radioactive metal
C423S003000, C423S006000, C588S010000
Reexamination Certificate
active
06471922
ABSTRACT:
BACKGROUND OF THE INVENTION
1. Field of the Invention
The present invention relates to the removal of actinides from spent salt generated from a molten salt oxidation reactor. In particular, this method removes uranium and thorium from spent salt to facilitate recycling of the salt for reuse in the reactor or disposal of the salt as non-hazardous waste.
2. Description of Related Art
Molten salt oxidation (MSO) is a thermal process that is capable of destroying organic constituents of energetic materials, hazardous wastes, and mixed wastes (i.e., wastes containing both organic and radioactive materials). In this process, combustible waste and air are introduced into a bath of molten carbonate salt (typically sodium carbonate), where the organic constituents of the waste materials are oxidized to carbon dioxide, nitrogen, and water. Inorganic products resulting from the reaction of the molten salt with the halogens, sulfur, phosphorous, metals, and radionuclides introduced into the salt bath must be removed to prevent the excessive build-up of inorganic products in the sodium carbonate. The excess build-up of these products in the carbonate salt would result in a dramatic drop in the efficiency of the system and would greatly increase the amount of toxic off-gases produced.
The carbonate salt serves both as a chemical reagent and as an acid scrubber to neutralize any acidic by-products produced during the waste destruction process. As the carbonate content in the salt decreases, the efficiency of the process decreases. At a certain point, the salt is removed from the reactor and the hazardous constituents are separated from the salt.
Because many of the metals and radionuclides captured in the salt are hazardous, the spent salt removed from the reactor would create a large secondary waste stream without further treatment. Thus, there is a need for a spent salt clean-up and recovery system to segregate these materials and minimize the amount of secondary waste. Once the hazardous constituents have been isolated, they can then be encapsulated for final disposal. This invention describes a separation strategy developed for actinide removal from mixed spent salt.
SUMMARY OF THE INVENTION
The present invention is a method for removing actinide contaminants from the spent salt of a molten salt oxidation (MSO) reactor. Removal of these contaminants enables the secondary waste stream generated by the MSO operation to be kept at a minimum. Once the contaminants are removed, the spent salt may either be re-used in the MSO process or disposed of as non-hazardous waste. If the salt still contains a high amount of carbonate (greater than about 20% by weight), it will be recycled into the MSO reactor. If the salt contains a low amount of carbonate (less than about 20% by weight), it no longer serves a useful purpose in the MSO process and will therefore be disposed of. Although this invention may be applied to other actinides, the processes for removing uranium and thorium from MSO spent salt will be emphasized. (Actinides are defined as elements 89-103 of the periodic table.)
To begin removal of the contaminants, the spent salt is cooled to ambient temperature, removed from the reactor, ground up, analyzed, and dissolved in water. If thorium is present, the clean-up procedure followed is not dependent on the pH of the dissolved salt solution or the concentration of carbonate present. An alkali hydroxide such as sodium hydroxide is added to the salt solution, causing the thorium present to form the insoluble precipitate thorium oxide (ThO
2
). If uranium is also present and the salt contains more than about 20% carbonate, then the uranium will also precipitate as an alkali diuranate. The solution is filtered and yields an actinide contaminated filter cake that is dried and packaged for disposal as radioactive waste. Alternatively, the cake can be mixed with ceramic powder to form stabilized pellets after calcination and sintering.
If uranium is present and the salt contains less than about 20% carbonate, then the uranium exists in the hexavalent uranyl state and forms the uranyl tricarbonate complex [UO
2
(CO
3
)
3
]
−4
. Any thorium must first be removed by the process described above. The remaining solution must then be neutralized by the addition of acid (e.g., HCl) to facilitate the removal of uranium. A reducing agent, such as an alkali sulfide or an alkali dithionite, is added to reduce the oxidation state of the uranium from U(VI) to U(IV). Once the oxidation state is reduced, uranium precipitates as uranium oxide (UO
2
), which is filtered and disposed of as radioactive waste.
Generally, the precipitation and filtration steps remove about 90% of the thorium and/or uranium that is present in the original spent salt solution. The filtered salt solution is treated differently depending on whether the salt is to be disposed of or reused. If the salt is destined for re-use, it is merely dried using a spray dryer and returned to the MSO reactor. If the salt is to be disposed of, a further clean-up step is necessary. The additional clean-up is accomplished by sending the solution through a commercially available ion exchange column (such as Diphonix™), which yields salt solutions that contain less than 0.1 ppm thorium and/or uranium.
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patent: 4436704 (1984-03-01), Krennrich et al.
patent: 4675166 (1987-06-01), Joubert
patent: 4879006 (1989-11-01), Turner
patent: 4954293 (1990-09-01), Cailly et al.
patent: 4968504 (1990-11-01), Rourke
patent: 5656009 (1997-08-01), Feng et al.
Removal of Uranium from Spent Salt from the Molten Salt Oxidation Process, Leslie Summers et al., UCRL-ID-126857, Mar. 1997 (Published Oct. 1997), Lawrence Livermore National Laboratory, pp. 1-11.
Adamson Martyn G.
Hipple David L.
Hsu Peter C.
Summers Leslie J.
Von Holtz Erica H.
Griffin Steven P.
Ildebrando Christina
Lee Ann M.
The Regents of the University of California
Thompson Alan H.
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