Method and apparatus of treating waste from nuclear fuel handling facility

A waste treatment apparatus treats radioactive contaminated waste from a nuclear fuel material handling facility to decontaminate the radioactive contaminated waste by using an electrolytic molten salt, and reuses the electrolytic molten salt so that any effluents are not produced. Radioactive contaminated waste (10) from a nuclear fuel material handling facility is subjected to electrolysis by a molten salt electrolysis unit (20) to decontaminate the waste (10). The used salt (16) used for decontaminating the waste (10) is filtered to separate nuclear fuel materials (19) from the used salt (16). The filtered salt (18) is reused by the molten salt electrolysis unit (20). The salt adhering to the decontaminated waste (12) is recovered by an evaporating unit (59), and the recovered salt (15) is reused by the molten salt electrolysis unit (20).

BACKGROUND OF THE INVENTION
 1. Field of the Invention
 The present invention relates to a method of treating electrically
 conductive waste contaminated with nuclear fuel materials and disposed
 from a nuclear fuel handling facility, and an apparatus for carrying out
 the method. More specifically, the present invention relates to a method
 of treating contaminated metallic waste produced when a nuclear fuel
 handling facility is dismantled, such as waste steel materials
 contaminated with nuclear fuel materials, or an adsorbent used for
 adsorbing nuclear fuel materials mounted in a nuclear fuel handling
 facility, and an apparatus for carrying out the method.
 2. Description of the Related Art
 FIG. 20 is a typical view of an apparatus for carrying out an electrolytic
 polishing process generally used for decontaminating waste contaminated
 with radioactive substances, such as nuclear fuel materials, (hereinafter
 referred to as "radioactive contaminated waste") by electrolysis. As shown
 in FIG. 20, a radioactive contaminated waste 3 is held by a holding device
 2 and is immersed in an electrolytic solution contained in an electrolytic
 vessel 1 of a stainless steel. The radioactive contaminated waste 3
 functions as an anode. A cathode 4 is immersed in the electrolytic water
 solution 5. When the radioactive contaminated waste 3 is a stainless steel
 waste, a phosphoric acid solution is used as the electrolytic water
 solution 5 serving as a bath. When the radioactive contaminated waste 3 is
 a carbon steel material, a sulfuric acid solution is used as the
 electrolytic water solution 5. The holding device 2 and the cathode 4 are
 connected to a dc power supply 6. When a dc voltage is applied across the
 holding device 2 and the cathode 4 by the dc power supply 6, the
 radioactive contaminated waste 3 functions as an anode. A surface layer of
 the radioactive contaminated waste 3 dissolves in the electrolytic water
 solution 5 simultaneously with coming off of radioactive contaminants
 adhering to the radioactive contaminated waste 3. Part of substances came
 off the radioactive contaminated waste 3 remains in the electrolytic water
 solution 5 and the rest is precipitated in sludge 8 on the bottom of the
 electrolytic vessel 1. Hydrogen 7 is produced on the cathode 4 of a
 stainless steel.
 Generally, when decontaminating a radioactive contaminated waste by the
 electrolytic polishing process using the electrolytic water solution 5 as
 a bath, current is unable to flow uniformly over the entire surface of the
 radioactive contaminated waste when the radioactive contaminated waste has
 a complicated shape because the resistance of the bath is high.
 Consequently, the decontaminating effect of the electrolytic polishing
 process is reduced for some portions of the radioactive contaminated
 waste. If a high current is supplied to the bath to enhance electrolytic
 processing speed, heat is generated in the bath due to the high resistance
 of the bath. Hydrogen 7 produced on the cathode 4 during the electrolytic
 polishing process cause problems in safety. It is difficult to remove
 radioactive substances accumulated in the waste electrolytic water
 solution 5, particularly, radioactive substances dissolved in the waste
 electrolytic water solution 5 from the waste electrolytic water solution
 5. The waste electrolytic solution 5 cannot be reused and becomes an
 additional radioactive contaminated waste. Thus, the total amount of
 radioactive contaminated waste increases.
 The present invention has been made to solve those problems and it is
 therefore an object of the present invention to provide a method of
 treating waste from a nuclear fuel handling facility, capable of easily
 decontaminating a contaminated waste having a complicated shape, not
 discharging any effluent, capable of repeatedly using an electrolytic
 solution and not producing additional waste.
 Another object of the present invention is to provide an apparatus for
 carrying out the foregoing method.
 SUMMARY OF THE INVENTION
 According to a first aspect of the present invention, a method of treating
 electrically conductive waste contaminated with nuclear fuel materials
 from a nuclear fuel handling facility comprises a molten salt electrolysis
 process for removing the nuclear fuel materials adhering to a surface of
 the waste by immersing the waste in a molten salt to dissolve a surface
 layer of the waste electrochemically in the molten salt; and a filtering
 process for filtering the molten salt used in the molten salt electrolysis
 process to extract the nuclear fuel materials removed from the surface of
 the waste and accumulated in the molten salt from the molten salt. The
 molten salt filtered in the filtering process is reused in the molten salt
 electrolysis process.
 Preferably, the method further comprises an evaporation process for
 removing the molten salt adhering to a surface of the waste processed by
 the molten salt electrolysis process and taken out of the molten salt by
 heating the waste so that the molten salt adhering thereto evaporates. The
 molten salt recovered in the evaporation process is reused in the molten
 salt electrolysis process.
 Preferably, the method further comprises a cleaning process for removing
 the molten salt adhering to the waste processed by the molten salt
 electrolysis process and taken out of the molten salt by a cleaning
 liquid, and an evaporative drying process for drying the molten salt
 contained in the cleaning liquid by evaporating the cleaning liquid used
 in the cleaning process. The molten salt recovered in the evaporative
 drying process is reused in the molten salt electrolysis process, and the
 cleaning liquid evaporated in the evaporative drying process is reused in
 the cleaning process.
 Preferably, in the molten salt electrolysis process, the molten salt and
 the waste immersed in the molten salt are moved relative to each other to
 remove the nuclear fuel materials from the surface of the waste.
 Preferably, in the molten salt electrolysis process, the waste is contained
 in a basket serving as an electrode for an electrolysis and the basket is
 vibrated in the molten salt.
 Preferably, in the molten salt electrolysis process, the waste is contained
 in a basket serving as an electrode for an electrolysis and the basket is
 rotated in the molten salt.
 Preferably, in the molten salt electrolysis process, the molten metal is
 spouted against the waste immersed in the molten salt.
 Preferably, a liquid metal, which is in a liquid phase at a temperature
 high enough to maintain the molten salt in a molten state, is placed in
 the molten salt as an electrode for the molten salt electrolysis process.
 Preferably, when the nuclear fuel materials are oxides, the method further
 comprises a reducing process for reducing the nuclear fuel materials to
 metals before subjecting the waste to the molten salt electrolysis
 process.
 Preferably, in the reducing process, the nuclear fuel materials are reduced
 to metals by making the nuclear fuel materials react with a reducing
 agent.
 Preferably, the reducing process comprises immersing the waste contaminated
 with the nuclear fuel materials in a reducing molten salt, supplying a
 reducing agent into the reducing molten salt, applying a voltage that will
 not cause a decomposition of the reducing molten salt across an anode and
 a cathode immersed in the reducing molten salt to regenerate the reducing
 agent reacted with the nuclear fuel materials.
 Preferably, the reducing process comprises immersing the waste contaminated
 with the nuclear fuel oxides in a reducing molten salt, reducing the
 nuclear fuel oxides to metals by applying a voltage across an anode and a
 cathode immersed in the reducing molten salt for an electrolytic
 reduction.
 According to a second aspect of the present invention, a method of treating
 an electrically conductive waste contaminated with nuclear fuel materials
 from a nuclear fuel handling facility comprises a reducing process for
 reducing the nuclear fuel materials to metals; a thermal melting process
 for producing a molten salt by heating and melting the metals produced by
 reducing the nuclear fuel materials and the waste; and a molten salt
 electrolysis process for recovering the metals produced by reducing the
 nuclear fuel materials and contained in the molten salt by applying a
 voltage across an anode and a cathode immersed in the molten salt so that
 the metals produced by reducing the nuclear fuel materials are deposited
 on the cathode.
 Preferably, a chloride or a hydride having a same kind of cation as that of
 the molten salt is added to the molten salt to lower the melting point of
 the molten salt so that an operating temperature of the molten salt in the
 molten salt electrolysis process is lowered.
 Preferably, the method further comprises a cleaning process for separating
 the nuclear fuel materials from the waste by cleaning the nuclear fuel
 materials deposited on the cathode in the molten salt electrolysis process
 and the waste with a cleaning liquid to dissolve the waste in the cleaning
 liquid; and an oxidation process for converting the nuclear fuel materials
 separated from the waste by the cleaning process into oxides by oxidizing
 the nuclear fuel materials; wherein the waste is an adsorbent used in the
 nuclear fuel material handling facility.
 Preferably, the method further comprises an evaporative drying process for
 drying the adsorbent contained in the cleaning liquid by evaporating the
 cleaning liquid used in the cleaning process. The cleaning liquid
 evaporated by the evaporative drying process is reused in the cleaning
 process.
 According to a third aspect of the present invention, an apparatus for
 treating an electrically conductive waste contaminated with nuclear fuel
 materials from a nuclear fuel handling facility comprises a molten salt
 electrolysis unit for removing the nuclear fuel materials adhering to a
 surface of the waste by immersing the waste in a molten salt to dissolve a
 surface layer of the waste electrochemically in the molten salt; a
 filtering unit for filtering the molten salt used by the molten salt
 electrolysis unit to extract the nuclear fuel materials removed from the
 surface of the waste and accumulated in the molten salt from the molten
 salt, and a molten salt return line for returning the molten salt filtered
 by the filtering unit to the molten salt electrolysis unit.
 Preferably, the apparatus further comprises an evaporation unit for
 removing the molten salt adhering to a surface of the waste processed by
 the molten salt electrolysis unit and taken out of the molten salt by
 heating the waste so that the molten salt adhering thereto evaporates, and
 a molten salt return line for returning the molten salt removed from the
 surface of the waste by the evaporation unit to the molten salt
 electrolysis unit.
 Preferably, the apparatus further comprises a cleaning unit for removing
 the molten salt adhering to the waste processed by the molten salt
 electrolysis unit and taken out of the molten salt by a cleaning liquid,
 and an evaporative drying unit for drying the molten salt contained in the
 cleaning liquid by evaporating the cleaning liquid used by the cleaning
 unit, a molten salt return line for returning the molten salt recovered by
 the evaporative drying unit to the molten salt electrolysis unit, and a
 cleaning liquid return line for returning the cleaning liquid evaporated
 by the evaporative drying unit to the cleaning unit.
 Preferably, the molten salt electrolysis unit is provided with a driving
 mechanism for moving the molten salt and the waste immersed in the molten
 salt relative to each other.
 Preferably, the molten salt electrolysis unit is provided further with a
 basket capable of containing the waste and serving as an electrode for an
 electrolysis, and the driving mechanism vibrates the basket in the molten
 salt.
 Preferably, the molten salt electrolysis unit is provided further with a
 basket capable of containing the waste and serving as an electrode for an
 electrolysis, and the driving mechanism rotates the basket in the molten
 salt.
 Preferably, driving mechanism includes a spouting means for spouting the
 molten salt against the waste immersed in the molten salt.
 Preferably, the molten salt electrolysis unit is provided with an electrode
 formed from a liquid metal which is immersed in the molten salt and is in
 a liquid phase at a temperature high enough to maintain the molten salt in
 a molten state.
 Preferably, when the nuclear fuel materials are oxides, the apparatus
 further comprises a reducing unit for reducing the nuclear fuel materials
 to metals.
 According to a fourth aspect of the present invention, an apparatus for
 treating an electrically conductive waste contaminated with nuclear fuel
 materials from a nuclear fuel handling facility comprises a reducing unit
 for reducing the nuclear fuel materials to metals, a thermal melting unit
 for producing a molten salt by heating and melting the metals produced by
 reducing the nuclear fuel materials and the waste, and a molten salt
 electrolysis unit for recovering the metals produced by reducing the
 nuclear fuel materials and contained in the molten salt by applying a
 voltage across an anode and a cathode immersed in the molten salt so that
 the metals produced by reducing the nuclear fuel materials are deposited
 on the cathode.
 Preferably, the waste is an adsorbent used for adsorbing the nuclear fuel
 materials in the nuclear fuel handling facility, and the apparatus further
 comprises a cleaning unit for separating the nuclear fuel materials from
 the waste by cleaning the nuclear fuel materials deposited on the cathode
 of the molten salt electrolysis unit and the waste with a cleaning liquid
 to dissolve the waste in the cleaning liquid, and an oxidation unit for
 converting the nuclear fuel materials separated from the waste by the
 cleaning unit into oxides by oxidizing the nuclear fuel materials.
 Preferably, the apparatus further comprises an evaporative drying unit for
 drying the adsorbent contained in the cleaning liquid by evaporating the
 cleaning liquid used by the cleaning unit, and a cleaning liquid return
 line for returning the cleaning liquid recovered by the evaporative drying
 unit to the cleaning unit.

DESCRIPTION OF THE PREFERRED EMBODIMENTS
 A waste treatment apparatus in a first embodiment according to the present
 invention and a waste treatment method to be carried out by the same waste
 treatment apparatus will be described hereinafter.
 Nuclear fuel handling facilities include uranium mining facilities, uranium
 refining facilities, conversion plants, enrichment plants, nuclear fuel
 processing plants, nuclear reactors, reprocessing plants, waste disposal
 facilities, and transportation facilities for transporting nuclear fuel
 materials between those facilities and plants.
 Waste from nuclear fuel handling facilities includes various steel
 materials that are produced when nuclear fuel handling facilities are
 dismantled, and adsorbents which are used for arresting nuclear fuel
 materials in nuclear fuel handling facilities. The waste treatment
 apparatus in the first embodiment is suitable for treating contaminated
 metal wastes, such as contaminated steel materials, or contaminated metal
 waste cut into small pieces by a pretreatment process.
 Nuclear fuel materials include uranium, uranium ores, uranium oxides,
 uranium chloride, uranium fluoride, uranium hydride, uranium nitrate and
 uranium sulfate.
 Referring to FIGS. 1 and 2, the waste treatment apparatus has a molten salt
 electrolysis unit 20 for decontaminating radioactive contaminated waste 10
 from a nuclear fuel handling facility by molten salt electrolysis using a
 molten salt 24. The salt 24 adheres to the decontaminated waste 12
 decontaminated by the molten salt electrolysis unit 20. The salt 24
 adhering to the decontaminated waste 12 is separated from the waste 12 by
 an evaporation unit 59. The evaporation unit 59 melts and evaporates the
 salt 24 by heating the salt 24 adhering to the waste 12 at a temperature
 not lower than its melting point to separate the salt 24 from the waste
 12. The evaporation unit 59 is a known evaporation device used in chemical
 engineering. The waste treatment apparatus has a recovered salt return
 line 53 for returning the recovered salt 15 separated from the
 decontaminated waste 12 and recovered by the evaporating unit 59 to the
 molten salt electrolysis unit 20. The recovered salt return line 53 may be
 of either a transfer pipe type or a conveyor type.
 The waste treatment apparatus has a filtering unit 54 for filtering the
 used salt 16 used by the molten salt electrolysis unit 20 to filter out
 nuclear fuel materials 19 from the used salt 16 to provide the filtered
 salt 18. The filtering unit 54 may be a filtering device generally used in
 chemical engineering and capable of separating the nuclear fuel materials
 19 and the filtered salt 18 by subjecting the used salt 16 to filtration.
 The waste treatment apparatus has a filtered salt return line 55 for
 returning the filtered salt 18 to the molten salt electrolysis unit 20.
 The filtered salt return line 55 may be of either a transfer pipe type or
 a conveyor type.
 Referring to FIG. 2, the molten salt electrolysis unit 20 has an
 electrolytic vessel 20a made of a low-carbon steel, an anode basket 21
 which is a mesh structure of a low-carbon steel or a stainless steel,
 placed in the electrolytic vessel 20a, and a driving device 96. The basket
 21 is driven for rotation in a molten salt 24 contained in the
 electrolytic vessel 20a by the driving device 96 to promote electrolytic
 reaction by moving the radioactive contaminated waste 10 contained in the
 basket 21 relative to the molten salt 24.
 The anode basket 21 containing the radioactive contaminated waste 10
 contaminated with nuclear fuel materials is immersed in the molten salt
 24. A cathode 23 of a low-carbon steel is immersed in the molten salt 24.
 A dc power supply 25 has a positive electrode and a negative electrode
 connected to the anode basket 21 and the cathode 23, respectively. In FIG.
 2, indicated at 26 is a cathodic deposit and at 27 is sludge.
 The molten salt 24 is an electrolyte prepared by melting one of chemical
 compounds including an alkali metal chloride, an alkaline earth metal
 chloride, an alkali metal fluoride, an alkaline earth metal fluoride, a
 chloride or fluoride of an element included in the component elements of
 the waste 10, or a mixture of some of those chemical compounds, and
 keeping the molten salt at a temperature not lower than its melting point.
 Referring to FIG. 3, the evaporating unit 59 has a melting crucible 70 for
 heating the decontaminated waste 12 decontaminated by the molten salt
 electrolysis unit 20 and soiled with the salt 24, and an induction heating
 coil 71 surrounding the melting crucible 70. The decontaminated waste 12
 contained in the melting crucible 70 is heated. Consequently, the
 decontaminated waste 12 melts into molten waste 72 and the salt 24
 adhering to the decontaminated waste 12 evaporates in a gas phase. The
 salt 24 in a gas phase flows in the direction of the arrows 73 and is
 recovered to obtain the recovered salt 15 in a liquid phase.
 A method of treating the radioactive contaminated waste 10 from a nuclear
 fuel handling facility to be carried out by the waste treatment apparatus
 in the first embodiment shown in FIGS. 1 to 3 will be described with
 reference to FIGS. 1 to 4.
 Referring to FIG. 4, a molten salt electrolysis process 11 puts the
 radioactive contaminated waste 10 from the nuclear fuel handling facility
 in the anode basket 21 and immerses the anode basket 21 in the molten salt
 24 contained in the electrolytic vessel 20a of the molten salt
 electrolysis unit 20. A current is supplied through the radioactive
 contaminated waste 10 functioning as an anode, and the cathode 23 to
 dissolve electrochemically a surface layer of the radioactive contaminated
 waste 10 contaminated with nuclear fuel materials in the molten salt 24 to
 provide decontaminated waste 12. When a dc voltage is applied across the
 anode basket 21 and the cathode 23 by the dc power supply 25, the
 radioactive contaminated waste 10 functions as an anode, and the surface
 layer of the radioactive contaminated waste 10 dissolves in the molten
 salt 24. Consequently, the nuclear fuel materials adhering to the surface
 of the contaminated waste 10 fall into the molten salt 24, and sludge of
 the nuclear fuel materials deposits on the bottom of the electrolytic
 vessel 20a of the molten salt electrolysis unit 20. Ions of the component
 metals of the radioactive contaminated waste 10 are reduced and cathodic
 deposit 26 deposits on the cathode 23.
 The decontaminated waste 12 is soiled with the salt 24 used by the molten
 salt electrolysis process 11. The salt 24 adhering to the decontaminated
 waste 12 is removed from the decontaminated waste 12 by an evaporation
 process 13 using the evaporating unit 59. The evaporation process 13 heats
 the decontaminated waste 12 at a temperature not lower than the melting
 point of the salt 24 in an environment of the atmospheric pressure or a
 reduced pressure to evaporate the salt 24 from the decontaminated waste
 12. Thus clean waste 14 is obtained. The recovered salt 15 is returned
 through the recovered salt return line 53 to the molten salt electrolysis
 unit 20 and is reused for the molten salt electrolysis process 11. Thus,
 the salt 15 is removed from the decontaminated waste 12 to obtain the
 clean waste 14. In the evaporation process 13, the decontaminated waste 12
 can be melted to reduce the same to a metal ingot by heating the
 decontaminated waste 12 at a temperature higher than its melting point
 during or after the removal of the salt 24 from the decontaminated waste
 12.
 The used salt 16 used in the molten salt electrolysis process 11 contains
 sludge of the nuclear fuel materials 19 removed from the radioactive
 contaminated waste 10. A filtering process 17 filters out the sludge from
 the used salt 16 by the filtering unit 54. The filtered salt 18 thus
 filtered by the filtering unit 54 is returned through the filtered salt
 return line 55 to the molten salt electrolysis unit 20 and is reused for
 the molten salt electrolysis process 11.
 A molten salt electrolysis unit 20 in a first modification of the molten
 salt electrolysis unit 20 shown in FIG. 2 will be described with reference
 to FIG. 5, in which parts like or corresponding to those of the molten
 salt electrolysis unit 20 shown in FIG. 2 are designated by the same
 reference characters and the description thereof will be omitted. The
 molten salt electrolysis unit 20 shown in FIG. 5 is provided with a liquid
 metal 28 instead of the solid cathode 23 shown in FIG. 2. The liquid metal
 28 serves as a cathode. The liquid metal 28 is in a liquid phase at the
 temperature of the melting point of the molten salt 24. The liquid metal
 28 is contained in an electrically insulating ceramic pot 29, and the
 ceramic pot 29 containing the liquid metal 28 is immersed in the molten
 salt 24. A cathode wire 30 has one end dipped in the liquid metal 28 and
 the other end connected to the dc power supply 25. The liquid metal 28 may
 be stirred by a stirring device to promote the mixing of the cathodic
 deposit deposited on the surface of the liquid metal 28 with the liquid
 metal 28. The cathode wire 30 is extended through an electrically
 insulating ceramic tube 31 to insulate the same from the molten salt 24.
 Ions of the component metals of the radioactive contaminated waste 10 are
 reduced on the surface of the liquid metal 28 and the cathode deposit is
 deposited on the surface of the liquid metal 28.
 A molten salt electrolysis unit 20 in a second modification of the molten
 salt electrolysis unit 20 shown in FIG. 2 will be described with reference
 to FIG. 6, in which parts like or corresponding to those of the molten
 salt electrolysis unit 20 shown in FIG. 2 are designated by the same
 reference characters and the description thereof will be omitted. The
 molten salt electrolysis unit 20 shown in FIG. 6 is provided with
 actuators 36 and 37 for vibrating the anode basket 21. The anode basket 21
 is held by an anode basket holding bar 38. The actuator 36 vibrates the
 anode basket holding bar 38 in vertical directions, and the actuator 47
 vibrates the same in horizontal directions. The actuators 36 and 37 are
 used selectively to vibrate the anode basket holding bar 38 at an optional
 frequency in horizontal directions, vertical directions or in both
 vertical and horizontal directions to promote the separation of the
 nuclear fuel material from the surface of the radioactive contaminated
 waste 10.
 A molten salt electrolysis unit 20 in a third modification of the molten
 salt electrolysis unit 20 shown in FIG. 2 will be described with reference
 to FIG. 7, in which parts like or corresponding to those of the molten
 salt electrolysis unit 20 shown in FIG. 2 are designated by the same
 reference characters and the description thereof will be omitted. The
 molten salt electrolysis unit 20 shown in FIG. 7 is provided with a
 cleaning device for cleaning the surface of the radioactive contaminated
 waste 10 in the molten salt 24. The cleaning device has a molten salt
 suction pipe 40, a molten salt jetting pipe 41 provided with a molten salt
 jetting nozzle 42, and a pump 39. The molten salt suction pipe 40 and the
 molten salt jetting pipe 41 are connected to the inlet port and the outlet
 port of the pump 39, respectively. The pump 39 operates to suck the molten
 salt 24 through the molten salt suction pipe 40 and to clean the
 radioactive contaminated waste 10 contained in the anode basket 21 by
 jetting the molten salt 24 through the molten salt jetting nozzle 42
 against the radioactive contaminated waste 10. In FIG. 7 the arrows 43
 indicate the flow of the molten salt 24.
 As apparent from the foregoing description, the waste treatment apparatus
 in the first embodiment decontaminates the radioactive contaminated waste
 10 contaminated with the nuclear fuel materials by the molten salt
 electrolysis unit 20, removes the salt 24 adhering to the decontaminated
 waste 12 by heating the decontaminated waste 12 in the environment of the
 atmospheric pressure or a reduced pressure to evaporate the salt 24 by the
 evaporating unit 59. Thus, the salt 24 adhering to the decontaminated
 waste 12 can easily be removed from the decontaminated waste 12 to obtain
 the clean waste 14. The recovered salt 15 recovered by the evaporating
 unit 59 can be returned through the recovered salt return line 53 to the
 molten salt electrolysis unit 20 to reuse the same. The used salt 16 is
 filtered and the filtered salt 18 can be returned through the filtered
 salt return line 55 to the molten salt electrolysis unit 22 to reuse the
 same.
 A waste treatment apparatus in a second embodiment according to the present
 invention for treating radioactive contaminated waste from a nuclear fuel
 handling facility will be described hereinafter. The waste treatment
 apparatus in the second embodiment is a modification of the waste
 treatment apparatus in the first embodiment. Parts of the waste treatment
 apparatus in the second embodiment like or corresponding to those of the
 waste treatment apparatus in the first embodiment are designated by the
 same reference characters and the description thereof will be omitted.
 Referring to FIG. 8, the waste treatment apparatus in the second embodiment
 is provided with a cleaning unit 56 instead of the evaporating unit 59 of
 the waste treatment apparatus in the first embodiment, and is provided
 additionally with an evaporative drying unit 57 and a cleaning liquid
 return line 58. The cleaning unit 56 cleans the decontaminated waste 12
 with a cleaning liquid, such as water. The recovered cleaning liquid
 recovered by the evaporative drying unit 57 is returned through the
 cleaning liquid return line 58 to the cleaning unit 56. The recovered salt
 15 recovered by the evaporative drying unit 57 is returned through the
 recovered salt return line 53 to the molten salt electrolysis unit 20 to
 reuse the same.
 Referring to FIG. 9, the cleaning unit 56 has a filter 74 for filtering the
 cleaning liquid, and a pump 75 for spraying the filtered cleaning liquid
 on the decontaminated waste 12 decontaminated by the molten salt
 electrolysis unit 20.
 A waste treatment method using the waste treatment apparatus shown in FIGS.
 8 and 9 will be described with reference to FIGS. 8 to 10. As shown in
 FIG. 10, the waste treatment method has a cleaning process 32 instead of
 the waste treatment method shown in FIG. 4. The cleaning process 32 cleans
 the decontaminated waste 12 decontaminated by the molten salt electrolysis
 process 11 of the salt 24 adhering to the decontaminated waste 12 with a
 cleaning liquid containing at least one of liquids including water, a
 nitric acid solution, a sulfuric acid solution and a hydrochloric acid
 solution. The used cleaning liquid 33 containing the salt 24 and
 discharged from the cleaning unit 56 is subjected to evaporation by the
 evaporative drying unit 57 to recover the salt 24 by evaporative drying.
 The recovered salt 15 is returned to the molten salt electrolysis unit 20
 to reuse the same in the molten salt electrolysis process ll. The cleaning
 liquid 35 recovered by the evaporative drying process 34 is returned
 through the cleaning liquid return line 58 to the cleaning unit 56 to
 reuse the same in the cleaning process 32.
 As apparent from the foregoing description, the waste treatment apparatus
 in the second embodiment is capable of readily removing the salt 24
 adhering to the decontaminated waste 12 by the cleaning unit 56 after the
 radioactive contaminated waste 10 contaminated with the nuclear fuel
 materials has been decontaminated by the molten salt electrolysis unit 20.
 The recovered salt 15 recovered by the evaporative drying unit 57 is
 returned through the recovered salt return line 53 to the molten salt
 electrolysis unit 20 and can be reused. The used salt 16 used by the
 molten salt electrolysis unit 20 is filtered by the filtering unit 54 to
 recycle the filtered salt 18. The filtered salt 18 is returned through the
 filtered salt return line 55 to the molten salt electrolysis unit 20 and
 can be reused in the molten salt electrolysis process 11. The cleaning
 liquid 35 recovered by the evaporative drying unit 57 is returned through
 the cleaning liquid return line 58 to the cleaning unit 56. Thus, the
 cleaning liquid 35 can efficiently be reused and hence additional
 effluents are not produced.
 A waste treatment apparatus in a third embodiment according to the present
 invention for treating radioactive contaminated waste from a nuclear fuel
 handling facility, and a waste treatment method to be carried out by the
 same waste treatment apparatus will be described hereinafter. The waste
 treatment apparatus in the third embodiment is a modification of the waste
 treatment apparatus in the second embodiment. Parts of the waste treatment
 apparatus in the third embodiment like or corresponding to those of the
 waste treatment apparatus in the second embodiment are designated by the
 same reference characters and the description thereof will be omitted.
 Referring to FIG. 11, the waste treatment apparatus in the third embodiment
 has a reducing unit 60 disposed on the upstream side of the molten salt
 electrolysis unit 20. When the nuclear fuel materials adhering to the
 waste 10 are uranium ore or oxides, the reducing unit 60 reduces the
 nuclear fuel materials prior to the treatment of the waste 10 by the
 molten salt electrolysis unit 20 for the efficient molten salt
 electrolysis of the waste 10.
 Referring to FIG. 12, the reducing unit 60 has a reaction vessel 45
 containing a molten salt 47, a meshed waste container 46 placed in the
 reaction vessel 45 to contain the waste 10, and a stirring device 48
 inserted in the waste container 46. A reducing agent 49 is supplied into
 the reaction vessel 45. The reducing agent 49 is lithium (Li), magnesium
 (Mg) or calcium (Ca). Preferably, the reducing agent 49 is Li. The waste
 10 from a nuclear fuel material handling facility is put into the waste
 container 46. The reducing agent 49, such as Li, comes into direct contact
 with the waste 10 for reducing reaction.
 A waste treatment method to be carried out by the waste treatment apparatus
 in the third embodiment shown in FIGS. 11 and 12 will be described with
 reference to FIGS. 11 to 13. The waste treatment method comprises a
 reducing process 44 in addition to the processes of the waste treatment
 method to be carried out by the waste treatment apparatus in the second
 embodiment. The waste treatment method carries out the reducing process 44
 by the reducing unit 60 before the molten salt electrolysis process 11.
 The reducing process 44 reduces the nuclear fuel materials adhering to the
 waste 10 to metals through the direct interaction of the reducing agent 49
 and the nuclear fuel materials. The waste 10 thus treated by the reducing
 process 44 is subjected to processes entirely the same as those of the
 waste treatment method shown in FIG. 10.
 FIG. 14 shows a reducing unit 60 in a first modification of the reducing
 unit 60 shown in FIG. 12. The reducing unit 60 in the first modification
 comprises, in addition to the components of the reducing unit 60 shown in
 FIG. 12, a reducing agent regenerating device for regenerating the
 reducing agent. The reducing agent regenerating device comprises a cathode
 50 inserted in the waste container 46, an anode 52 (carbon electrode)
 inserted in the reaction vessel 45, and a power supply 51 for applying a
 voltage across the cathode 50 and the anode 52. Suppose that the reducing
 agent is Li. Lithium oxide (Li.sub.2 O) is produced by the reduction
 reaction of Li and the nuclear fuel materials adhering to the waste 10,
 and Li.sub.2 O disperses in the waste container 46. Part of the Li.sub.2 O
 is converted into Li and O at the cathode 50. Part of the thus regenerated
 Li is used for reduction and the rest disperses in the waste container 46.
 The Li dispersed in the waste container 46 does not contribute to
 reduction and hence efficient reduction cannot be achieved. A voltage that
 will not decompose a molten salt 46 contained in the reaction vessel 45,
 for example about 3 V, is applied across the cathode 50 and the anode 52
 by the power supply 51. Consequently, Li, i.e., the reducing agent 49,
 supplied into the reaction vessel 45 penetrates the waste container 46
 gradually, comes into contact with the nuclear fuel materials adhering to
 the waste 10 and reducing reaction progresses. Oxygen (O) generated when
 Li is regenerated at the cathode 50 disperses outside the waste container
 46. The stirring device 48 disposed in the waste container 46 promotes the
 dispersion of O and the supply of O to the anode 52. The following
 electrode reactions occur at the electrodes during the foregoing
 processes.
 ##STR1##
 After the completion of the reducing reaction, the waste container 46 is
 raised and pulled out of the molten salt 47 contained in the reaction
 vessel 45.
 FIG. 15 shows a reducing unit 60 in a second modification of the reducing
 unit 60 shown in FIG. 12. The reducing unit 60 shown in FIG. 15 has a
 cathode 61 and an anode 62 immersed in a molten salt 47. A voltage is
 applied across the cathode 61 and the anode 62 by a power supply 73 to
 reduce oxides dispersed in a molten salt 47 to metals by electrolytic
 reduction. A reducing reaction progresses in the molten salt 47 contained
 in a reaction vessel 45 of the reducing unit 60. Oxides, i.e., nuclear
 fuel materials, are reduced to metals U and TRU at the cathode 61, and O
 generated at the cathode 61 disperses outside a waste container 46. A
 stirring device 48 disposed in the waste container 46 promotes the
 dispersion of O and promotes the supply of O to the anode 62 (carbon
 electrode). The following reactions occur at the electrodes during the
 foregoing processes.
 ##STR2##
 After the completion of the reducing reaction, the waste container 46 is
 raised and pulled out of the molten salt 47 contained in the reaction
 vessel 45.
 A waste treatment apparatus in a fourth embodiment according to the present
 invention for treating radioactive contaminated waste from a nuclear fuel
 handling facility, and a waste treatment method to be carried out by the
 same waste treatment apparatus will be described hereinafter. The waste
 treatment apparatus in the fourth embodiment is suitable for treating
 radioactive contaminated waste when the radioactive contaminated waste is
 an adsorbent, such as NaF, and the nuclear fuel materials adhering to the
 adsorbent are fluorides, such as UF.sub.6, UF.sub.4 and UO.sub.2 F.sub.2.
 Referring to FIG. 16, the waste treatment apparatus comprises a reducing
 unit 60 for reducing waste 100, a thermal melting unit 64 connected to the
 reducing unit 60, a molten salt electrolysis unit 65 connected to the
 thermal melting unit 64, a cleaning unit 66 connected to the molten salt
 electrolysis unit 65, a evaporative drying unit 67 connected to the
 cleaning unit 66 and an oxidizing unit 68 connected to the cleaning unit
 66.
 The reducing unit 60 reduces radioactive contaminated waste 100. The
 thermal melting unit 64 heats and melts the reduced waste 101 provided by
 reducing the radioactive contaminated waste 100 by the reducing unit 60.
 The molten salt electrolysis unit 65 subjects a molten salt 102, i.e., the
 molten waste provided by the thermal melting unit 64 to electrolysis.
 Thus, the molten waste prepared by melting the reduced waste 101 produced
 by reducing the radioactive contaminated waste 100 by the reducing unit 60
 is used as the molten salt 102 for electrolysis. The cleaning unit 66
 separates nuclear fuel materials (uranium metal) and an adsorbent (NaF)
 contained in a cathodic deposit 76 deposited on the cathode of the molten
 salt electrolysis unit 65. The evaporative drying unit 67 processes a used
 cleaning liquid 77 used by the cleaning unit 66 for evaporative drying to
 recover the adsorbent (NaF) dissolved in the used cleaning liquid 77. A
 cleaning liquid 78 recovered by evaporation is returned through a
 recovered cleaning liquid return line 79 to the cleaning unit 66 and is
 reused. The nuclear fuel materials (uranium metal) 80 separated from the
 adsorbent by the cleaning unit 66 is oxidized by the oxidizing unit 68,
 and oxides (Uranium oxide) 81 thus produced by the oxidizing unit 68 are
 collected.
 As shown in FIG. 17, the molten salt electrolysis unit 65 comprises a
 reaction vessel 85 for containing the molten salt 102 prepared by melting
 the reduced waste 101, an anode 82 and a cathode disposed in the reaction
 vessel 85, and a power supply 84 for applying a voltage across the anode
 82 and the cathode 83.
 A waste treatment method to be carried out by the waste treatment apparatus
 in the fourth embodiment shown in FIGS. 16 and 17 will be described with
 reference to FIGS. 16 to 18. A reducing process 86 processes the
 radioactive contaminated waste 100, i.e. , the adsorbent (NaF)
 contaminated with the nuclear fuel materials, such as UF.sub.6, UF.sub.4
 and UO.sub.2 F.sub.2, to reduce, for example, UF.sub.6 (uranium
 hexafluoride) to UF.sub.4 (uranium tetrafluoride). More concretely, a
 reducing gas, such as hydrogen gas, argon gas or phosgene gas, is spouted
 against adsorbent particles to reduce the nuclear fuel materials.
 A thermal melting and salt-processing process 87 heats and melts the
 reduced waste 101, i.e., the adsorbent containing the reduced nuclear fuel
 materials by the thermal melting unit 64 and adds a fluoride or a hydride
 having the same cations as those of the reduced waste 101 to the molten
 waste 101 to produce a molten salt 102 having a low melting point. The
 chloride having the same cations as those of the reduced waste 101 is, for
 example, NaCl. When NaF and NaCl are mixed, a eutectic of NaF--NaCl having
 a melting point of 600.degree. C. is produced. The melting point of this
 eutectic is lower than the melting point of 992.degree. C. of NaF by
 390.degree. C.
 In a molten salt electrolysis process 88, the anode 82 and the cathode 83
 of the molten salt electrolysis unit 65 are immersed in the molten salt
 102, and a voltage is applied across the anode 82 and the cathode 83 to
 reduce UF.sub.4 to uranium metal. A cathodic deposit 76 containing uranium
 metal, NaF and NaCl is deposited on the cathode 83. The cathodic deposit
 76 is recovered from the cathode 83. A cleaning process 89 cleans the
 cathodic deposit 76 with a cleaning liquid, such as water to separate
 uranium metal from other components of the cathodic deposit 76. An
 oxidizing process 90 oxidizes the thus recovered uranium metal to uranium
 oxide by the oxidizing unit 68. The uranium oxide is stable in the
 atmosphere. An evaporative drying process 91 heats and evaporates the used
 cleaning liquid 77 containing NaF by the evaporative drying unit 67 to
 recover the NaF. An evaporated cleaning liquid 78 is returned to the
 cleaning unit 66 and is reused.
 As shown in FIG. 19, a waste treatment apparatus in a modification of the
 waste treatment apparatus shown in FIG. 16 heats and melts the radioactive
 contaminated waste 100 before the reducing process 86 to produce a molten
 salt 92, blows a reducing gas 93, such as hydrogen gas, argon gas or
 phosgene gas, through a nozzle 95 into the molten salt 92 contained in a
 vessel 94 to reduce the nuclear fuel materials within the molten salt 92.
 The nuclear fuel materials may be reduced by electrolytic reduction by
 immersing an anode and a cathode in the molten salt 92 and applying a
 voltage across the anode and the cathode. A chloride or the like is added
 to the molten salt after reduction and the reduced molten salt 92 is
 subjected to electrolysis by the molten salt electrolysis unit 65.
 As apparent from the foregoing description, according to the present
 invention, electrically conductive waste contaminated with nuclear fuel
 materials from a nuclear fuel handling facility is immersed in a molten
 salt, the waste is connected to an anode and a surface layer of the waste
 is dissolved electrochemically in the molten salt. Thus, the nuclear fuel
 materials adhering to the waste can easily be removed.
 Since the electrical resistance of the molten salt is very low as compared
 with that of an electrolytic water solution, an electric current flows
 uniformly over the surface of the waste. Consequently, the waste having a
 complicated shape, which is difficult to decontaminate by conventional
 techniques, can surely be decontaminated. Since the electrical resistance
 of the molten salt is low, a large current can be supplied through the
 molten salt without entailing abnormal heat generation to increase the
 process speed. The molten salt electrolysis process is safe because
 hydrogen is not generated at the cathode when the molten salt is used for
 the electrolysis.
 The sludge of the nuclear fuel material accumulated in the molten salt can
 satisfactorily be separated from the molten salt by filtration because the
 surface tension of the molten salt is lower than that of an aqueous
 solution. The nuclear fuel material dissolved in the molten salt can be
 recovered in a cathodic deposit. The molten salt can be reused even if
 some nuclear fuel material dissolved in the molten salt remains in the
 molten salt.
 The waste, such as the absorbent used in the nuclear fuel material handling
 facility, can be easily treated by reducing the nuclear fuel materials to
 metals, heating and melting the metals and the waste, and recovering the
 metals on the cathode in the molten salt electrolysis.
 Although the invention has been described in its preferred form with a
 certain degree of particularity, obviously many changes and variations are
 possible therein. It is therefore to be understood that the present
 invention may be practiced otherwise than as specifically described herein
 without departing from the scope and spirit thereof.