Patent Number: 049816160
Section: description

DESCRIPTION OF THE PREFERRED EMBODIMENT An embodiment of the invention will now be described with reference to the drawing. The FIGURE is a view showing an embodiment of the spent fuel treatment method of this invention, in which (1) represents a dissolving tank, (2) a solvent extraction process, (3) a plutonium nitrate solution and uranyl nitrate solution, (4) a freeze-vacuum drying apparatus, (5) a nitrate, (6) a condensate, (7) a denitrification process, (8) a roasting reduction process, (9) a product, (10) a spent solvent, (11) a freeze-vacuum drying apparatus, (12) TBP, DBP, etc., (13) n-dodecan, (14) a vacuum distillation apparatus, (15) DBP, etc., (16) TBP, (17) a preparation process, (18) an incinerator, (19) liquid waste, (20) a freeze-vacuum drying apparatus, (21) residue, (22) water and nitric acid, (23) storage or solid waste treatment system, (24) a preparation process, (25) a utilization process, and (26) an emission process. In the drawing, nuclear fuel scrap which contains impurities generated at a fuel manufacturing plant or the like is supplied to (1) the dissolving tank along with a nitric acid solution, heated there and dissolved. Then uranium and plutonium solutions are sent to the solvent extraction process (2) after preparation. Solvents consisting of TBP, n-dodecan, etc., and the nitric acid solution are employed to effect separation into plutonium nitrate and uranyl nitrate solutions (3), spent solvent (10) and liquid waste (19). The plutonium nitrate and uranyl nitrate solutions (3) are separated into nitrates (5) and condensate (6) by the freeze-vacuum drying process (4). The condensate (6) is fed to the freeze-vacuum drying apparatus (4). Meanwhile, the nitrates (5) are sent to the denitrification process (7). After microwave heating, for example, for conversion to oxide, powder is prepared as needed by the roasting reduction process (8) employing a roasting reduction furnace or the like. The result is the product (9). Spent solvent (10) is separated into TBP, DBP, etc. at (12) and into n-dodecan (13) by freeze-vacuum drying apparatus (11). TBP, OBP (12) are separated into DBP, etc. (15) and TBP (16) by the vacuum distillation apparatus (14). DBP, etc. (15) is sent to the incinerator (18). Meanwhile, TBP (16) and n-dodecan (13) are blended in the preparation process (17) and the result is sent to the solvent extraction process (2) after preparation by the further addition of TBP, n-dodecan and so on as necessary. Liquid waste (19) is sent to the freeze-vacuum drying apparatus (20) and separated into residue (21) consisting of plutonium, uranium and americium impurities and the like, and into water and nitric acid (22). For recovery, residue (nitrates) (21) is sent to storage at process (23) or to a solid waste treating system. At the preparation process (24), water and nitric acid (22) are prepared by either concentration or dilution by means of adding water or nitric acid as necessary. The result is used at the process (25) and is also sent to, e.g., the dissolving tank (1), the solvent extraction tank (2) or another process, such as an off-gas scrubbing process, not shown. If there is a surplus, this can be released at the process (26). In the embodiment described above, the freeze-vacuum dry apparatus is employed at three points, namely (4), (11) and (20). However, if the system is operated with storage tanks provided, a single freeze-vacuum drying apparatus would of course be quite satisfactory. In accordance with the present invention, TBP, DBP and the like and n-dodecan can be separated by using a freeze-vacuum drying method in a solvent cleansing process, TBP and DBP can be separated by using a vacuum evaporation method in the solvent cleansing process, and the use of sodium can be eliminated. As a result, the amount of liquid radioactive waste is reduced, it is possible to abbreviate treatment, the amount of sludge produced is reduced and neutralization and filtration are unnecessary. By treating the liquid radioactive waste using a freeze-vacuum drying process having a high decontamination efficiency, most of the radioactive substance can be recovered as residue, the recovered solution can be reutilized, liquid waste can be reduced and liquid waste treatment simplified. Furthermore, plutonium and uranium solutions are recovered as nitrates by the freeze-vacuum drying method, and these solutions are rendered into oxides by thermal decomposition, thereby obtaining a powdered oxide product. As many apparently widely different embodiments of the present invention can be made without departing from the spirit and scope thereof, it is to be understood that the invention is not limited to the specific embodiments thereof except as defined in the appended claims.