Abstract:
A spent fuel reprocessing method has a dissolution step of dissolving the spent fuel in nitric acid solution, an electrolysis/valence adjustment step of reducing Pu to trivalent, maintaining the pentavalent of Np, a uranium extraction step of collecting UO 2  by bringing the fuel into contact with organic solvent and extracting hexavalent U by means of an extraction agent, an oxalic acid precipitation step of causing MA and the fissure products remaining in the nitric acid solution to precipitate together as oxalic acid precipitate, a chlorination step of converting the oxalic acid precipitate into chlorides by adding hydrochloric acid to the oxalic acid precipitate, a dehydration step of synthetically producing anhydrous chlorides by dehydrating the chlorides in a flow of Ar gas, and a molten salt electrolysis step of dissolving the anhydrous chlorides into molten salt and collecting U, Pu and MA at the cathode by electrolysis.

Description:
CROSS REFERENCES TO RELATED APPLICATIONS 
       [0001]    This application is based upon and claims the benefits of priority from the prior Japanese Patent Applications No. 2008-143431, filed in the Japanese Patent Office on May 30, 2008, the entire content of which is incorporated herein by reference. 
       BACKGROUND OF THE INVENTION 
       [0002]    The present invention relates to a spent fuel reprocessing method comprising a step of collecting uranium (U), plutonium (Pu) and minor actinides (MA) from spent oxide nuclear fuel. 
         [0003]    The Purex process is a known typical process for reprocessing spent fuel produced from nuclear power plants so as to refine and collect useful substances contained in the spent fuel in order to reutilize them as fuel and isolate unnecessary fission products. Spent fuel contains alkali metal (AM) elements, alkaline-earth metal (AEM) elements and platinum group elements as fission products (FP) besides transuranic elements (TRU) such as uranium and plutonium. 
         [0004]    In the Purex process, spent fuel is dissolved in nitric acid solution and subsequently fission products are isolated in a first extraction step. Thereafter, U and Pu are separated from each other in a separation step and respectively subjected to a U purification process and a Pu purification process. Then, the Pu solution and the U solution are put together and subjected to mixture and denitration so that it is not possible to collect Pu alone. 
         [0005]    Since U and Pu are temporarily separated from each other in the separation step of the Purex process, nuclear non-proliferability is not absolutely secured. 
         [0006]    Therefore, there is a demand for a reprocessing process realized by partly modifying the Purex process so as to make it impossible to collect Pu alone and realize a high degree of nuclear non-proliferability. 
         [0007]    Meanwhile, the high level liquid waste produced from a Purex process contains U and Pu to a small extent and minor actinides (Np: Neptunium, Am: Americium, Cm: Curium, etc.) to a large extent. The aqua-pyro process is known to collect such transuranic elements (Pu and minor actinides) in combination by applying a technique of oxalic acid precipitation—conversion to chloride—molten salt electrolysis to high-level liquid waste (Patent Documents 1 and 2: Japanese Patent No. 2,809,819 Publication and Japanese Patent No. 3,319,657 Publication). The entire content of which is incorporated herein by reference. With the aqua-pyro process, Pu and U are made to accompany minor actinides and collected in combination with each other. In other words, Pu is not collected alone by itself. 
         [0008]    In view of the above-identified problem of the prior art, it is therefore the object of the present invention to provide a spent fuel reprocessing method that can isolate most of the uranium contained in spent fuel solution and collect it as light water reactor fuel and, at the same time, can collect Pu and minor actinides with U so as to make them utilizable as metal fuel for a fast reactor in order to ensure a high degree of nuclear non-proliferability. 
       BRIEF SUMMARY OF THE INVENTION 
       [0009]    In order to attain the object, according to an aspect of the present invention, there is provided a spent fuel reprocessing method comprising: a disassembly/shear step of disassembling and shearing spent oxide nuclear fuel; a dissolution step of dissolving the fuel subjected to the disassembly/shear step in nitric acid solution; an electrolysis/valence adjustment step of reducing plutonium to trivalent, maintaining the pentavalent of neptunium for the fuel subjected to the dissolution step; a uranium extraction step of collecting uranium oxide by bringing the fuel subjected to the electrolysis/valence adjustment step into contact with organic solvent and extracting hexavalent uranium by means of an extraction agent; an oxalic acid precipitation step of causing the minor actinides and the fissure products remaining in the nitric acid solution after the uranium extraction step to precipitate together as oxalic acid precipitate by means of an oxalic acid precipitation method; a chlorination step of converting the oxalic acid precipitate into chlorides by adding hydrochloric acid to the oxalic acid precipitate; a dehydration step of synthetically producing anhydrous chlorides by dehydrating the chlorides in a flow of reductive inert gas; and a molten salt electrolysis step of dissolving the anhydrous chlorides into molten salt and collecting uranium, plutonium and minor actinides at the cathode by electrolysis. 
         [0010]    According to another aspect of the present invention, there is provided a spent fuel reprocessing method comprising: a disassembly/shear step of disassembling and shearing spent oxide nuclear fuel; a dissolution step of dissolving the fuel subjected to the disassembly/shear step in nitric acid solution; an electrolysis/valence adjustment step of reducing plutonium and neptunium respectively to trivalent and pentavalent for the fuel subjected to the dissolution step; a uranium extraction step of collecting uranium oxide by bringing the fuel subjected to the electrolysis/valence adjustment step into contact with organic solvent and extracting hexavalent uranium by means of an extraction agent; an oxalic acid precipitation step of causing the minor actinides and the fissure products remaining in the nitric acid solution after the uranium extraction step to precipitate together as oxalic acid precipitate by means of an oxalic acid precipitation method; an oxidation/dehydration step of dehydrating the oxalic acid precipitate and subsequently converting it into precipitate oxides in an oxidation atmosphere; and an electrolysis/reduction step of immersing the precipitate oxides in a mixture of molten salts obtained by dissolving alkali metal oxides in molten salts of chlorides of alkali metals or a mixture of molten salts obtained by dissolving alkaline-earth metal oxides in molten salts of chlorides of alkaline-earth metals, bringing the precipitate oxides into contact with a cathode to drawn out oxygen ions in the precipitate oxides, removing the oxygen ions as oxygen gas or carbon dioxide gas at the side of an anode in the molten salts and collecting uranium, plutonium and minor actinides in the precipitate oxides at the cathode. 
         [0011]    Thus, according to the present invention, it is possible to isolate most of the uranium from spent fuel solution and collect it as light water reactor fuel, while it is possible to collect Pu and minor actinides with U so as to make them utilizable as metal fuel for a fast reactor. As Pu is not collected alone by itself and Pu and minor actinides are collected with U, the present invention can ensure a high degree of nuclear non-proliferability. 
     
    
     
       BRIEF DESCRIPTION OF THE DRAWINGS 
         [0012]    The above and other features and advantages of the present invention will become apparent from the discussion hereinbelow of specific, illustrative embodiments thereof presented in conjunction with the accompanying drawings, in which: 
           [0013]      FIG. 1  is a flowchart of spent fuel reprocessing method according to a first embodiment of the present invention; 
           [0014]      FIG. 2  is a schematic sectional elevational view of an apparatus that can be employed for an electrolysis/valence adjustment step of spent fuel reprocessing method according to the first embodiment of the present invention; 
           [0015]      FIG. 3  is a graph showing some of the results of measurement of the initial value of the electrode potential and that of the current density in the electrolysis/valence adjustment step of spent fuel reprocessing method according to the first embodiment of the present invention; 
           [0016]      FIG. 4  is a graph showing some of the results of measurement of the change with time of the current density when the electrolysis potential is held to −100 mV relative to a reference electrode, which is a silver/silver salt electrode, in the electrolysis/valence adjustment step of spent fuel reprocessing method according to the first embodiment of the present invention; 
           [0017]      FIG. 5  is a flowchart of spent fuel reprocessing method according to a second embodiment of the present invention; and 
           [0018]      FIG. 6  is a schematic sectional elevational view of an apparatus that can be employed for the electrolysis/reduction step of spent fuel reprocessing method according to the second embodiment of the present invention. 
       
    
    
     DETAILED DESCRIPTION OF THE INVENTION 
       [0019]    Now, the present invention will be described by referring to the accompanying drawings that illustrate preferred embodiments of spent fuel reprocessing method according to the present invention. 
       First Embodiment 
       [0020]    The first embodiment of spent fuel reprocessing method according to the present invention will be described below by referring to  FIGS. 1 and 2 . 
         [0021]      FIG. 1  is a flowchart of spent fuel reprocessing method according to a first embodiment of the present invention. Referring to  FIG. 1 , first, spent oxide fuel  1  is disassembled and sheared in a disassembly/shear step  2 . Subsequently, all the spent oxide fuel is dissolved by nitric acid in a dissolution step  3 . At this time, U exists in a hexavalent state whereas Pu exists in a tetravalent state. 
         [0022]    Thereafter, Pu is electrolytically reduced to trivalent in an electrolysis/valence adjustment step  4 .  FIG. 2  is a schematic sectional elevational view of an apparatus that can be employed for the electrolysis/valence adjustment step  4  of the first embodiment. More specifically, a cathode chamber  27  and an anode chamber  28  are separated from each other by means of a diaphragm  50  in the apparatus. Catholyte  24  is stored in the cathode chamber  27  and a cathode  25  and a reference electrode  30  are dipped in the catholyte  24 . Anolyte  51  is stored in the anode chamber  28  and an anode  26  is dipped in the anolyte  28 . The cathode  25  and the anode  26  are connected to a power source  29 . The cathode  25  and the reference electrode  30  are connected to a potentiometer  31 . The reference electrode  30  may typically be a silver/silver chloride electrode. The cathode chamber  27  is provided with an agitator  52  for agitating the catholyte  24 . 
         [0023]    With the above-described arrangement, Pu can be reduced to trivalent, while maintaining Np as pentavalent by limiting the cathode potential to not higher than −100 mV or confining the cathode current density within a range between not less than 20 mA/cm 2  and 40 mA/cm 2 . The U that is partly reduced to tetravalent is employed to reduce Pu from tetravalent to trivalent. Then, U itself is oxidized to become hexavalent. 
         [0024]      FIG. 3  is a graph showing some of the results obtained by an experiment, which illustrates the correlation between the cathode potential and the current density observed in an electrolysis/valence adjustment step  4 . It is experimentally proved that the cathode potential can be made equal to −0.1 V (−100 mV) by making the current density not less than about 20 mA/cm 2 . 
         [0025]    Since the U is mostly hexavalent, only hexavalent U can be extracted into tributyl phosphate (TBP)—30% dodecane solution by using such a solution in a U extraction step  5 . Trivalent ions of Pu and pentavalent ions of Np remain in the aqueous solution with the tetravalent ions of part of U. 
         [0026]      FIG. 4  is a graph showing some of the results of measurement of the change with time of the current density when the electrolysis potential is held to −100 mV relative to the reference electrode, which is a silver/silver salt electrode, in the electrolysis/valence adjustment step  4  and the U extraction step  5 .  FIG. 4  shows that the cathode current density is within a range between 20 mA/cm 2  and 40 mA/cm 2  for the cathode potential of −100 mV. 
         [0027]    Thereafter, oxalic acid is added to the aqueous solution that is left after the U extraction step  5  to produce oxalic acid precipitate  7  in an oxalic acid precipitation step  6 . The oxalic acid precipitate  7  contains Pu and minor actinides such as Np, Am and Cm, some of rare earth elements (RE) and some of alkaline-earth metal elements. Of the fission products (FP), alkali metal elements and platinum group elements do not precipitate but are dissolved in the filtrate. 
         [0028]    U, Pu, minor actinides and rare earth group elements are collected as oxalic acid precipitate  7  in the oxalic acid precipitation step  6 . 
         [0029]    In a chlorination step  8 , hydrochloric acid is added to the oxalic acid precipitate  7  and dissolved at not higher than 100° C. and subsequently hydrogen peroxide is added thereto in order to decompose the oxalic acid into water and carbon dioxide. The U, the Pu and the minor actinides in the oxalic acid precipitate  7  are converted into chloride  9  in this chlorination step  8 . 
         [0030]    Thereafter, the moisture in the hydrochloric acid solution is evaporated and removed in a dehydration step  40  and subsequently the moisture is completely eliminated at about 200° C. in a flow of reductive inert gas (e.g., argon and nitrogen). As a result, chlorides (anhydrous chlorides)  41  of U, Pu and minor actinide are produced. 
         [0031]    Then, U, Pu and metals of minor actinides that can be used as fast reactor fuel can be collected in combination with each other as the produced anhydrous chlorides  41  are electrolyzed in a molten salt electrolysis step  10 . 
         [0032]    Now, a platinum group fission product collection step  14  of collecting platinum group fission products from the oxalic acid precipitate  7  obtained in the oxalic acid precipitation step  6  will be described below by referring to  FIGS. 1 and 2 . An apparatus having a structure same as the apparatus shown in  FIG. 2  that is employed in the electrolysis/valence adjustment step and the U extraction step may be used in the platinum group fission product collection step  14 . For example, the same apparatus may be used or another apparatus having the same structure or a similar structure may be used. 
         [0033]    The oxalic acid precipitate  7  contains Pu and minor actinides such as Np, Am and Cm, some of rare earth elements and some of alkaline-earth metal elements. Of the fission products, alkali metal elements and platinum group elements are not precipitated by oxalic acid and are dissolved in the filtrate (catholyte)  24 . The filtrate  24  that melts the fission products is put into the cathode chamber  27  and the insoluble cathode  25  is immersed in the filtrate  24  for electrolysis in the platinum group fission product collection step  14 . 
         [0034]    As a voltage is applied to the anode  26  and the cathode  25  from the power source  29 , Pd(Palladium), Ru(Ruthenium), Rh(Rhodium), Mo and Tc(Technetium) that are platinum group fission products are deposited and collected out of the fission products contained in the filtrate  24  in the cathode chamber  27 . On the other hand, acidic anolyte  51  is put into the anode chamber  28 . Since alkali metal elements such as Cs and alkaline-earth metal elements such as Sr that are contained in the filtrate, which is catholyte  24 , remain in the filtrate, they can be separated from the platinum group fission products. 
         [0035]    An applied voltage is observed by measuring the potential difference between the reference electrode  30  and the cathode  25  that are immersed in the cathode chamber  27  for the by means of the potentiometer  31 . It is important to control the potentials so as to deposit Pd, Ru, Rh, Mo and Tc that are platinum group fission products without generating hydrogen. 
         [0036]    Thus, the load of producing nuclear waste glass can be reduced because Pd, Ru, Rh, Mo and Tc that are platinum group fission products do not move into the high level liquid waste. Additionally, the rate of producing high level liquid waste can also be reduced. 
         [0037]    The hexavalent U that is extracted by TBP—30% dodecane in the U extraction step  5  is washed with nitric acid in a U purification step  11  and subsequently converted into an oxide in a denitration step  12  so as to be collected as high purity UO 2    13 . The high purity UO 2    13  can be used as oxide fuel for light water reactors. 
       Second Embodiment 
       [0038]    Now, the second embodiment of spent fuel reprocessing method according to the present invention will be described below by referring to  FIGS. 5 and 6 . The parts of this embodiment that same as or similar to those of the first embodiment are denoted respectively by the same reference symbols and will not be described repeatedly. 
         [0039]      FIG. 5  is a flowchart of spent fuel reprocessing method according to a second embodiment of the present invention.  FIG. 6  is a schematic sectional elevational view of an apparatus that can be employed for the electrolysis/reduction step of the second embodiment. 
         [0040]    The sequence down to the oxalic acid precipitation step  6 , where the oxalic acid precipitate  7  containing U, Pu, minor actinides and rare earth elements are collected, is same as that of the first embodiment. 
         [0041]    This second embodiment has an oxidation/dehydration step  15  and an electrolysis/reduction step  17  instead of the chlorination step  8 , the dehydration step  40  and the molten salt electrolysis step  10  of the first embodiment. 
         [0042]    More specifically, the oxalic acid precipitate  7  collected in the oxalic acid precipitation step  6  is heated to remove moisture, while ozone or acidic gas is blown into it, in the oxidation/dehydration step  15  to produce oxides (precipitate oxides)  16  of U, Pu, minor actinides and rare earth elements. 
         [0043]    Subsequently, moisture is completely removed from the oxides  16  while drawing oxygen by vacuum. Thereafter, as shown in  FIG. 6 , the oxides  16  are put into a stainless-steel-made cathode basket  19  and loaded in a molten salt electrolytic cell  22 . The cathode basket  19  containing the oxides  16  of U, Pu, minor actinides and rare earth elements is connected to the cathode and an insoluble anode  20  typically made of platinum or grassy carbon is placed in position. As a voltage is applied to the cathode basket  19  and the anode  20  in a molten salt  21 , oxygen ions are drawn out from the oxides of U, Pu and minor actinides in the cathode basket  19  to reduce them to make them become metals so that metals  18  of U, Pu and minor actinides can be collected. 
         [0044]    The oxides  16  are put into the stainless-steel-made cathode basket  19  in a mixture of molten salts. The mixture of molten salts is preferably prepared by dissolving an oxide of an alkali metal or an alkaline-earth metal into a molten salt of chloride of an alkali metal or an alkaline-earth metal. More specifically, a mixture of molten salts is preferably prepared by dissolving Li 2 O into a molten salt of LiCl, dissolving MgO into a molten salt of MgCl 2  or dissolving CaO into a molten salt of CaCl 2 . 
         [0045]    After putting the oxides  16  into the cathode basket  19  in the mixture of molten salts, oxygen ions in the oxides  16  are drawn out and the drawn out oxygen ions are removed at the anode as oxygen gas or CO 2  gas. Since alkali metal elements such as Cs, alkaline-earth metal elements such as Sr and rare earth elements such as Ce and Nd that are fission products are dissolved in the molten salts from the cathode basket  19  so that they can be separated from metals of U, Pu and minor actinide metals  18 . 
         [0046]    At this time, the oxides are reduced to become metals at the cathode in a manner as expressed by the formulas shown below. 
         [0000]      UO 2 +4e − →U+2O 2−   
         [0000]      PuO 2 +4e − →Pu+2O 2−   
         [0047]    On the other hand, oxygen gas is produced at the anode in a manner as expressed by the formula shown below. 
         [0000]      2O 2− →O 2 +4e −   
       Other Embodiment 
       [0048]    The embodiments of the spent fuel reprocessing method in accordance with the present invention explained above are merely samples, and the present invention is not restricted thereto. It is, therefore, to be understood that, within the scope of the appended claims, the present invention can be practiced in a manner other than as specifically described herein.