Abstract: A mechanical decladder for a spent nuclear fuel rod-cut includes an opening and closing unit configured to open and close an outlet of a basket into which the nuclear fuel rod-cut is loaded, a supporter on which the nuclear fuel rod-cut discharged from the outlet of the basket is seated, a hydraulic cylinder module configured to move the nuclear fuel rod-cut seated on the supporter; and a cutting module for slitting the nuclear fuel rod-cut while the nuclear fuel rod-cut is being moved by the hydraulic cylinder module. The opening and closing unit opens and closes the outlet of the basket in conjunction with a movement of the hydraulic cylinder module.

Abstract: A mechanical decladder for a spent nuclear fuel rod-cut includes an opening and closing unit configured to open and close an outlet of a basket into which the nuclear fuel rod-cut is loaded, a supporter on which the nuclear fuel rod-cut discharged from the outlet of the basket is seated, a hydraulic cylinder module configured to move the nuclear fuel rod-cut seated on the supporter; and a cutting module for slitting the nuclear fuel rod-cut while the nuclear fuel rod-cut is being moved by the hydraulic cylinder module. The opening and closing unit opens and closes the outlet of the basket in conjunction with a movement of the hydraulic cylinder module.

Abstract: Disclosed herein is an apparatus for recovering residual salt from the reduced uranium metal. The apparatus comprising: an evaporating chamber accommodating mixed molten salt or a reduced uranium metal; a heating furnace surrounding the evaporating chamber to heat the mixed molten salt in the evaporating chamber; an insulator disposed over the evaporating chamber to block heat generated from the evaporating chamber, and including an evaporating pipe in a center thereof to move vapor generated from the evaporating chamber; a receiver disposed over the insulator to collect powder formed by condensing and solidifying vapor passing through the evaporating pipe; and a condenser disposed over the receiver to prevent the vapor passing through the evaporating pipe from leaking out of the apparatus.

Abstract: Disclosed herein is an apparatus for recovering residual salt from the reduced uranium metal. The apparatus comprising: an evaporating chamber accommodating mixed molten salt or a reduced uranium metal; a heating furnace surrounding the evaporating chamber to heat the mixed molten salt in the evaporating chamber; an insulator disposed over the evaporating chamber to block heat generated from the evaporating chamber, and including an evaporating pipe in a center thereof to move vapor generated from the evaporating chamber; a receiver disposed over the insulator to collect powder formed by condensing and solidifying vapor passing through the evaporating pipe; and a condenser disposed over the receiver to prevent the vapor passing through the evaporating pipe from leaking out of the apparatus.

Abstract: Provided are an apparatus and method for manufacturing a solidified salt that is easy to treat in size and shape by using a vacuum transfer and a dual vessel. In the apparatus for quantitative solidification of a molten salt, a molten salt is introduced into a first vessel, and a second vessel is disposed inside the first vessel and quantitatively supplied with the molten salt. A molten-salt transferring unit quantitatively transfers the molten salt from the first vessel to the second vessel by vacuum pressure. A valve controls a discharge of the molten salt from the second vessel. A mold receives the molten salt from the second vessel and solidifies the molten salt. Accordingly, the molten salt can be stably discharged to the mold by quantitatively transferring the molten salt at vacuum pressure within the dual vessel, and a predetermined size and shape of the solidified salt can be manufactured, thereby processing and collecting the solidified salt safely.

Abstract: Provided are an apparatus and method for manufacturing a solidified salt that is easy to treat in size and shape by using a vacuum transfer and a dual vessel. In the apparatus for quantitative solidification of a molten salt, a molten salt is introduced into a first vessel, and a second vessel is disposed inside the first vessel and quantitatively supplied with the molten salt. A molten-salt transferring unit quantitatively transfers the molten salt from the first vessel to the second vessel by vacuum pressure. A valve controls a discharge of the molten salt from the second vessel. A mold receives the molten salt from the second vessel and solidifies the molten salt. Accordingly, the molten salt can be stably discharged to the mold by quantitatively transferring the molten salt at vacuum pressure within the dual vessel, and a predetermined size and shape of the solidified salt can be manufactured, thereby processing and collecting the solidified salt safely.

Abstract: Disclosed are a method of reducing spent oxides nuclear fuel to nuclear-fuel metal, in which metal oxides are reduced to metals using an electrochemical reduction device with LiCl—Li2O salt as an electrolyte, a cathode electrode assembly used in the method, and a reduction device including the cathode electrode assembly. The method is advantageous in that the process of reducing the spent oxide nuclear fuel to the nuclear-fuel metal and another process of recovering Li are united to simplify the whole processes, direct use of high oxidative Li metals is excluded to secure safety, and conversion efficiency of the spent oxide nuclear fuel is 99% or more.

Abstract: Disclosed are a method of reducing spent oxides nuclear fuel to nuclear-fuel metal, in which metal oxides are reduced to metals using an electrochemical reduction device with LiCl—Li2O salt as an electrolyte, a cathode electrode assembly used in the method, and a reduction device including the cathode electrode assembly. The method is advantageous in that the process of reducing the spent oxide nuclear fuel to the nuclear-fuel metal and another process of recovering Li are united to simplify the whole processes, direct use of high oxidative Li metals is excluded to secure safety, and conversion efficiency of the spent oxide nuclear fuel is 99% or more.