Patent Application
20090294299
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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.
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:
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.
The first embodiment of spent fuel reprocessing method according to the present invention will be described below by referring to
Thereafter, Pu is electrolytically reduced to trivalent in an electrolysis/valence adjustment step 4.
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/cm2 and 40 mA/cm2. 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.
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.
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.
U, Pu, minor actinides and rare earth group elements are collected as oxalic acid precipitate 7 in the oxalic acid precipitation step 6.
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.
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.
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.
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
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.
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.
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.
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.
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 UO2 13. The high purity UO2 13 can be used as oxide fuel for light water reactors.
Now, the second embodiment of spent fuel reprocessing method according to the present invention will be described below by referring to
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.
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.
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.
Subsequently, moisture is completely removed from the oxides 16 while drawing oxygen by vacuum. Thereafter, as shown in
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 Li2O into a molten salt of LiCl, dissolving MgO into a molten salt of MgCl2 or dissolving CaO into a molten salt of CaCl2.
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 CO2 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.
At this time, the oxides are reduced to become metals at the cathode in a manner as expressed by the formulas shown below.
UO2+4e−→U+2O2−
PuO2+4e−→Pu+2O2−
On the other hand, oxygen gas is produced at the anode in a manner as expressed by the formula shown below.
2O2−→O2+4e−
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.
| Number | Date | Country | Kind |
|---|---|---|---|
| 2008-143431 | May 2008 | JP | national |
