GB2461370A - Spent nuclear fuel reprocessing methods - Google Patents
Spent nuclear fuel reprocessing methods Download PDFInfo
- Publication number
- GB2461370A GB2461370A GB0909309A GB0909309A GB2461370A GB 2461370 A GB2461370 A GB 2461370A GB 0909309 A GB0909309 A GB 0909309A GB 0909309 A GB0909309 A GB 0909309A GB 2461370 A GB2461370 A GB 2461370A
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- United Kingdom
- Prior art keywords
- precipitate
- electrolysis
- oxalic acid
- cathode
- uranium
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- Granted
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- 238000000034 method Methods 0.000 title claims abstract description 40
- 239000002915 spent fuel radioactive waste Substances 0.000 title claims abstract description 28
- 238000012958 reprocessing Methods 0.000 title claims abstract description 23
- MUBZPKHOEPUJKR-UHFFFAOYSA-N Oxalic acid Chemical compound OC(=O)C(O)=O MUBZPKHOEPUJKR-UHFFFAOYSA-N 0.000 claims abstract description 120
- 235000006408 oxalic acid Nutrition 0.000 claims abstract description 40
- 238000005868 electrolysis reaction Methods 0.000 claims abstract description 36
- 239000002244 precipitate Substances 0.000 claims abstract description 36
- 150000003839 salts Chemical class 0.000 claims abstract description 36
- 229910052770 Uranium Inorganic materials 0.000 claims abstract description 35
- 229910052768 actinide Inorganic materials 0.000 claims abstract description 29
- 150000001255 actinides Chemical class 0.000 claims abstract description 28
- 229910052778 Plutonium Inorganic materials 0.000 claims abstract description 25
- 239000000047 product Substances 0.000 claims abstract description 23
- JFALSRSLKYAFGM-UHFFFAOYSA-N uranium(0) Chemical compound [U] JFALSRSLKYAFGM-UHFFFAOYSA-N 0.000 claims abstract description 22
- 150000001805 chlorine compounds Chemical class 0.000 claims abstract description 19
- 238000000605 extraction Methods 0.000 claims abstract description 19
- 239000000446 fuel Substances 0.000 claims abstract description 19
- 239000000203 mixture Substances 0.000 claims abstract description 14
- GRYLNZFGIOXLOG-UHFFFAOYSA-N Nitric acid Chemical compound O[N+]([O-])=O GRYLNZFGIOXLOG-UHFFFAOYSA-N 0.000 claims abstract description 13
- 229910017604 nitric acid Inorganic materials 0.000 claims abstract description 13
- 229910052783 alkali metal Inorganic materials 0.000 claims abstract description 12
- OYEHPCDNVJXUIW-UHFFFAOYSA-N plutonium atom Chemical compound [Pu] OYEHPCDNVJXUIW-UHFFFAOYSA-N 0.000 claims abstract description 12
- 229910052784 alkaline earth metal Inorganic materials 0.000 claims abstract description 11
- VEXZGXHMUGYJMC-UHFFFAOYSA-N Hydrochloric acid Chemical compound Cl VEXZGXHMUGYJMC-UHFFFAOYSA-N 0.000 claims abstract description 10
- 150000001340 alkali metals Chemical class 0.000 claims abstract description 8
- 150000001342 alkaline earth metals Chemical class 0.000 claims abstract description 7
- 230000003647 oxidation Effects 0.000 claims abstract description 7
- 238000007254 oxidation reaction Methods 0.000 claims abstract description 7
- WZECUPJJEIXUKY-UHFFFAOYSA-N [O-2].[O-2].[O-2].[U+6] Chemical compound [O-2].[O-2].[O-2].[U+6] WZECUPJJEIXUKY-UHFFFAOYSA-N 0.000 claims abstract description 5
- 239000003795 chemical substances by application Substances 0.000 claims abstract description 5
- 239000003960 organic solvent Substances 0.000 claims abstract description 5
- 229910000439 uranium oxide Inorganic materials 0.000 claims abstract description 5
- 239000011261 inert gas Substances 0.000 claims abstract description 4
- 230000004992 fission Effects 0.000 claims description 17
- 239000000243 solution Substances 0.000 claims description 16
- 238000003916 acid precipitation Methods 0.000 claims description 15
- BASFCYQUMIYNBI-UHFFFAOYSA-N platinum Chemical group [Pt] BASFCYQUMIYNBI-UHFFFAOYSA-N 0.000 claims description 14
- 238000004090 dissolution Methods 0.000 claims description 9
- 239000000706 filtrate Substances 0.000 claims description 9
- 230000002829 reductive effect Effects 0.000 claims description 9
- CURLTUGMZLYLDI-UHFFFAOYSA-N Carbon dioxide Chemical compound O=C=O CURLTUGMZLYLDI-UHFFFAOYSA-N 0.000 claims description 8
- 230000018044 dehydration Effects 0.000 claims description 8
- 238000006297 dehydration reaction Methods 0.000 claims description 8
- 239000003758 nuclear fuel Substances 0.000 claims description 8
- 239000001301 oxygen Substances 0.000 claims description 8
- 229910052760 oxygen Inorganic materials 0.000 claims description 8
- -1 oxygen ions Chemical class 0.000 claims description 8
- 229910052781 Neptunium Inorganic materials 0.000 claims description 7
- 230000009467 reduction Effects 0.000 claims description 6
- 238000005660 chlorination reaction Methods 0.000 claims description 5
- LFNLGNPSGWYGGD-UHFFFAOYSA-N neptunium atom Chemical compound [Np] LFNLGNPSGWYGGD-UHFFFAOYSA-N 0.000 claims description 5
- MYMOFIZGZYHOMD-UHFFFAOYSA-N Dioxygen Chemical compound O=O MYMOFIZGZYHOMD-UHFFFAOYSA-N 0.000 claims description 4
- 229910002092 carbon dioxide Inorganic materials 0.000 claims description 4
- 229910001882 dioxygen Inorganic materials 0.000 claims description 4
- 239000007789 gas Substances 0.000 claims description 4
- 238000010008 shearing Methods 0.000 claims description 4
- 229910021607 Silver chloride Inorganic materials 0.000 claims description 3
- 239000001569 carbon dioxide Substances 0.000 claims description 3
- 229910052709 silver Inorganic materials 0.000 claims description 3
- 239000004332 silver Substances 0.000 claims description 3
- HKZLPVFGJNLROG-UHFFFAOYSA-M silver monochloride Chemical compound [Cl-].[Ag+] HKZLPVFGJNLROG-UHFFFAOYSA-M 0.000 claims description 3
- 229910000287 alkaline earth metal oxide Inorganic materials 0.000 claims description 2
- 239000003929 acidic solution Substances 0.000 claims 1
- 229910000272 alkali metal oxide Inorganic materials 0.000 claims 1
- 238000000151 deposition Methods 0.000 claims 1
- 239000002198 insoluble material Substances 0.000 claims 1
- 230000001376 precipitating effect Effects 0.000 claims 1
- 230000008569 process Effects 0.000 description 9
- 229910052751 metal Inorganic materials 0.000 description 8
- 239000002184 metal Substances 0.000 description 8
- 229910052761 rare earth metal Inorganic materials 0.000 description 7
- FAPWRFPIFSIZLT-UHFFFAOYSA-M Sodium chloride Chemical compound [Na+].[Cl-] FAPWRFPIFSIZLT-UHFFFAOYSA-M 0.000 description 5
- 150000002739 metals Chemical class 0.000 description 5
- KDLHZDBZIXYQEI-UHFFFAOYSA-N palladium Substances [Pd] KDLHZDBZIXYQEI-UHFFFAOYSA-N 0.000 description 5
- SNRUBQQJIBEYMU-UHFFFAOYSA-N dodecane Chemical compound CCCCCCCCCCCC SNRUBQQJIBEYMU-UHFFFAOYSA-N 0.000 description 4
- 239000010808 liquid waste Substances 0.000 description 4
- 239000010948 rhodium Substances 0.000 description 4
- 229910052713 technetium Inorganic materials 0.000 description 4
- XLYOFNOQVPJJNP-UHFFFAOYSA-N water Substances O XLYOFNOQVPJJNP-UHFFFAOYSA-N 0.000 description 4
- 229910052695 Americium Inorganic materials 0.000 description 3
- VEXZGXHMUGYJMC-UHFFFAOYSA-M Chloride anion Chemical compound [Cl-] VEXZGXHMUGYJMC-UHFFFAOYSA-M 0.000 description 3
- 229910052685 Curium Inorganic materials 0.000 description 3
- 150000002500 ions Chemical class 0.000 description 3
- 238000005259 measurement Methods 0.000 description 3
- 229910052750 molybdenum Inorganic materials 0.000 description 3
- 229910052763 palladium Inorganic materials 0.000 description 3
- 238000000746 purification Methods 0.000 description 3
- 229910052703 rhodium Inorganic materials 0.000 description 3
- 229910052707 ruthenium Inorganic materials 0.000 description 3
- STCOOQWBFONSKY-UHFFFAOYSA-N tributyl phosphate Chemical compound CCCCOP(=O)(OCCCC)OCCCC STCOOQWBFONSKY-UHFFFAOYSA-N 0.000 description 3
- XKRFYHLGVUSROY-UHFFFAOYSA-N Argon Chemical compound [Ar] XKRFYHLGVUSROY-UHFFFAOYSA-N 0.000 description 2
- IJGRMHOSHXDMSA-UHFFFAOYSA-N Atomic nitrogen Chemical compound N#N IJGRMHOSHXDMSA-UHFFFAOYSA-N 0.000 description 2
- MHAJPDPJQMAIIY-UHFFFAOYSA-N Hydrogen peroxide Chemical compound OO MHAJPDPJQMAIIY-UHFFFAOYSA-N 0.000 description 2
- 230000002378 acidificating effect Effects 0.000 description 2
- 239000007864 aqueous solution Substances 0.000 description 2
- 230000008859 change Effects 0.000 description 2
- 238000000926 separation method Methods 0.000 description 2
- OKTJSMMVPCPJKN-UHFFFAOYSA-N Carbon Chemical compound [C] OKTJSMMVPCPJKN-UHFFFAOYSA-N 0.000 description 1
- 229910052684 Cerium Inorganic materials 0.000 description 1
- UFHFLCQGNIYNRP-UHFFFAOYSA-N Hydrogen Chemical compound [H][H] UFHFLCQGNIYNRP-UHFFFAOYSA-N 0.000 description 1
- 229910052779 Neodymium Inorganic materials 0.000 description 1
- CBENFWSGALASAD-UHFFFAOYSA-N Ozone Chemical compound [O-][O+]=O CBENFWSGALASAD-UHFFFAOYSA-N 0.000 description 1
- KJTLSVCANCCWHF-UHFFFAOYSA-N Ruthenium Chemical compound [Ru] KJTLSVCANCCWHF-UHFFFAOYSA-N 0.000 description 1
- LXQXZNRPTYVCNG-UHFFFAOYSA-N americium atom Chemical compound [Am] LXQXZNRPTYVCNG-UHFFFAOYSA-N 0.000 description 1
- 229910052786 argon Inorganic materials 0.000 description 1
- QVGXLLKOCUKJST-UHFFFAOYSA-N atomic oxygen Chemical compound [O] QVGXLLKOCUKJST-UHFFFAOYSA-N 0.000 description 1
- 229910052799 carbon Inorganic materials 0.000 description 1
- 238000006243 chemical reaction Methods 0.000 description 1
- 238000002474 experimental method Methods 0.000 description 1
- 239000011521 glass Substances 0.000 description 1
- 239000001257 hydrogen Substances 0.000 description 1
- 229910052739 hydrogen Inorganic materials 0.000 description 1
- 239000000155 melt Substances 0.000 description 1
- 229910052757 nitrogen Inorganic materials 0.000 description 1
- 229910052697 platinum Inorganic materials 0.000 description 1
- 150000002910 rare earth metals Chemical group 0.000 description 1
- MHOVAHRLVXNVSD-UHFFFAOYSA-N rhodium atom Chemical compound [Rh] MHOVAHRLVXNVSD-UHFFFAOYSA-N 0.000 description 1
- 239000000126 substance Substances 0.000 description 1
- GKLVYJBZJHMRIY-UHFFFAOYSA-N technetium atom Chemical compound [Tc] GKLVYJBZJHMRIY-UHFFFAOYSA-N 0.000 description 1
- 239000002699 waste material Substances 0.000 description 1
Classifications
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21C—NUCLEAR REACTORS
- G21C19/00—Arrangements for treating, for handling, or for facilitating the handling of, fuel or other materials which are used within the reactor, e.g. within its pressure vessel
- G21C19/42—Reprocessing of irradiated fuel
- G21C19/44—Reprocessing of irradiated fuel of irradiated solid fuel
- G21C19/46—Aqueous processes, e.g. by using organic extraction means, including the regeneration of these means
-
- C—CHEMISTRY; METALLURGY
- C25—ELECTROLYTIC OR ELECTROPHORETIC PROCESSES; APPARATUS THEREFOR
- C25C—PROCESSES FOR THE ELECTROLYTIC PRODUCTION, RECOVERY OR REFINING OF METALS; APPARATUS THEREFOR
- C25C1/00—Electrolytic production, recovery or refining of metals by electrolysis of solutions
- C25C1/22—Electrolytic production, recovery or refining of metals by electrolysis of solutions of metals not provided for in groups C25C1/02 - C25C1/20
-
- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21F—PROTECTION AGAINST X-RADIATION, GAMMA RADIATION, CORPUSCULAR RADIATION OR PARTICLE BOMBARDMENT; TREATING RADIOACTIVELY CONTAMINATED MATERIAL; DECONTAMINATION ARRANGEMENTS THEREFOR
- G21F9/00—Treating radioactively contaminated material; Decontamination arrangements therefor
- G21F9/04—Treating liquids
- G21F9/06—Processing
-
- C—CHEMISTRY; METALLURGY
- C25—ELECTROLYTIC OR ELECTROPHORETIC PROCESSES; APPARATUS THEREFOR
- C25C—PROCESSES FOR THE ELECTROLYTIC PRODUCTION, RECOVERY OR REFINING OF METALS; APPARATUS THEREFOR
- C25C3/00—Electrolytic production, recovery or refining of metals by electrolysis of melts
- C25C3/34—Electrolytic production, recovery or refining of metals by electrolysis of melts of metals not provided for in groups C25C3/02 - C25C3/32
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/30—Nuclear fission reactors
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02P—CLIMATE CHANGE MITIGATION TECHNOLOGIES IN THE PRODUCTION OR PROCESSING OF GOODS
- Y02P10/00—Technologies related to metal processing
- Y02P10/20—Recycling
-
- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02W—CLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
- Y02W30/00—Technologies for solid waste management
- Y02W30/50—Reuse, recycling or recovery technologies
Landscapes
- Engineering & Computer Science (AREA)
- Physics & Mathematics (AREA)
- General Engineering & Computer Science (AREA)
- High Energy & Nuclear Physics (AREA)
- Chemical & Material Sciences (AREA)
- Plasma & Fusion (AREA)
- Chemical Kinetics & Catalysis (AREA)
- Electrochemistry (AREA)
- Materials Engineering (AREA)
- Metallurgy (AREA)
- Organic Chemistry (AREA)
- Electrolytic Production Of Metals (AREA)
- Manufacture And Refinement Of Metals (AREA)
- Electrolytic Production Of Non-Metals, Compounds, Apparatuses Therefor (AREA)
Abstract
A spent fuel reprocessing method comprises: dissolving the spent fuel in a nitric acid solution; reducing the plutonium within the solution to a trivalent state; collecting uranium oxide by bringing the fuel into contact with an organic solvent and extracting hexavalent uranium by means of an extraction agent; and causing the minor actinides and fissure products remaining in the nitric acid solution to precipitate together as oxalic acid precipitate. In a first embodiment, the oxalic acid precipitate is converted into chlorides by adding hydrochloric acid; anhydrous chlorides are then produced by dehydrating the chlorides in a flow of inert gas; before the anhydrous chlorides are dissolved into molten salt. Alternatively, in a second embodiment (see Figure 3), the oxalic acid precipitate is dehydrated and converted into precipitate oxides in an oxidation atmosphere; before being immersed in a mixture of molten salts of chlorides of alkali metals or alkaline earth metals. Uranium, plutonium and minor actinides are subsequently collected by electrolysis.
Description
SPENT FUEL REPROCESSING METHOD
FIELD OF THE INVENTION
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.
DESCRIPTION OF THE RELATED ART
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.
SUMMARY OF THE INVENTION
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 sait electrolysis step of dissolving the anhydrous chlorides into molten salt and collecting uranium, plutonium and minor actinides at the cèthode 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 oxjdes 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.
BRIEF DESCRIPTION OF THE DRAWINGS
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: FIG. I is a flowchart of spent fuel reprocessing method according to a first embodiment of the present invention; 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; 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; 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 -1 00 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; FIG. 5 is a flowchart of spent fuel reprocessing method according to a second embodiment of the present invention; and 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 PREFERRED EMBODIMENTS
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] The first embodiment of spent fuel reprocessing method according to the present invention will be described below by referring to FIGS. I and 2.
FIG. I 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.
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.
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.
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/cm2.
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. 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 -1 00 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/cm2 and 40 mA/cm2 for the cathode potential of -100 mV.
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 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.
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 -10 -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 U02 13.
The high purity U02 13 can be used as oxide fuel for light water reactors.
[Second Embodiment] 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.
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.
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 -11 -and an electrolysis/reduction step 17 instead of the chlorination step 8, the dehydration step 40 and the molten salt electrolysis step 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 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.
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 Li20 into a molten -12 -salt of LiCI, dissolving MgO into a molten salt of MgCI2 or dissolving CaO into a molten salt of CaCI2.
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.
U02 + 4e -+ U + 202 Pu02 + 4e -Pu + 202 On the other hand, oxygen gas is produced at the anode in a manner as expressed by the formula shown below.
202 -02 + 4e [Other Embodiment] 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.
-13 -
Claims (7)
- CLAIMS1. 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 -14 -chlorides into molten salt and collecting uranium, plutonium and minor actinides at the cathode by electrolysis.
- 2. 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 -15 -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.
- 3. The method according to claim 2, wherein the electrolysis/reduction step is conducted by containing the precipitate oxides in a stainless-steel-made cathode basket, immersing the cathode basket in the molten salts and connecting the cathode to the cathode basket.
- 4. The method according to claim 2 or 3, wherein the mixture of molten salts is a mixture of molten salts obtained by dissolving Li20 in molten salt of LiCI, a mixture of molten salts obtained by dissolving MgO in molten salt of MgCI2 or a mixture of molten salts obtained dissolving CaO in molten salt of CaCI2.
- 5. The method according to any one of claims I to 4, further comprising: a fission product collection step of putting the filtrate left without precipitating in the oxalic acid precipitation step into a cathode chamber, putting a cathode made of an insoluble material -16 -into the cathode chamber, putting acidic solution into an anode chamber separated from the cathode chamber by a diaphragm for electrolysis and depositing and collecting fission products of the platinum group remaining in the filtrate at the cathode.
- 6. The method according to any one of claims I to 5, wherein the electrolysis/valence adjustment step is conducted at or lower than -100 mV relative to a silver/silver chloride electrode operating as reference electrode.
- 7. The method according to any one of claims I to 6, wherein the electrolysis/valence adjustment step is conducted with a cathode current density between 20 mA/cm2 and 40 mA/cm2 relative to a silver/silver chloride electrode operating as reference electrode.-17 -
Applications Claiming Priority (1)
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| JP2008143431A JP5193687B2 (en) | 2008-05-30 | 2008-05-30 | Spent fuel reprocessing method |
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| JP (1) | JP5193687B2 (en) |
| CN (1) | CN101593566B (en) |
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| RU (1) | RU2403634C1 (en) |
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| FR2960690B1 (en) * | 2010-05-27 | 2012-06-29 | Commissariat Energie Atomique | PROCESS FOR PROCESSING NUCLEAR FUELS USING NO PLUTONIUM REDUCING EXTRACTION OPERATION |
| FR2971948B1 (en) * | 2011-02-28 | 2013-03-29 | Commissariat Energie Atomique | PROCESS FOR PRECIPITATION OF ONE OR MORE SOLUTES |
| JP5758209B2 (en) * | 2011-06-14 | 2015-08-05 | 株式会社東芝 | Spent fuel reprocessing method |
| CN102412002B (en) * | 2011-09-06 | 2013-10-30 | 中国原子能科学研究院 | N2O4Device for oxidation regulation of plutonium valence state |
| FR2980468B1 (en) * | 2011-09-26 | 2014-01-24 | Commissariat Energie Atomique | PROCESS FOR THE PREPARATION OF AN OXYHALIDE AND / OR ACTINIDE OXIDE (S) AND / OR LANTHANIDE (S) FROM A MEDIUM COMPRISING AT LEAST ONE MELT SALT |
| JP5784476B2 (en) * | 2011-12-09 | 2015-09-24 | 株式会社東芝 | Uranium recovery method |
| JP5944237B2 (en) * | 2012-06-15 | 2016-07-05 | 株式会社東芝 | Method for recovering nuclear fuel material |
| FR2992330B1 (en) * | 2012-06-26 | 2014-08-08 | Commissariat Energie Atomique | PROCESS FOR SEPARATING AT LEAST ONE FIRST E1 CHEMICAL ELEMENT OF AT LEAST ONE SECOND E2 CHEMICAL ELEMENT INVOLVING THE USE OF A MEDIUM COMPRISING A SPECIFIED MELT SALT |
| KR101316925B1 (en) * | 2012-10-08 | 2013-10-18 | 한국수력원자력 주식회사 | Treatment method of spent uranium catalyst |
| RU2561065C1 (en) * | 2014-03-31 | 2015-08-20 | Открытое акционерное общество "Радиевый институт имени В.Г. Хлопина" | METHOD OF PRODUCING COMBINED SOLUTION OF U AND Pu |
| CN104008785B (en) * | 2014-06-16 | 2016-08-17 | 中国工程物理研究院核物理与化学研究所 | A kind of metal mold spentnuclear fuel after-treatment device |
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| CN104562089B (en) * | 2014-10-17 | 2017-03-22 | 中国原子能科学研究院 | Method for preparing initial molten salt system in molten salt electrolysis dry after-treatment process |
| GB2545934A (en) * | 2016-01-02 | 2017-07-05 | Richard Scott Ian | Single stage reprocessing of spent nuclear fuel |
| CN108389641B (en) * | 2017-12-28 | 2019-07-26 | 中国科学院近代物理研究所 | A kind of preparation facilities and preparation method of nuclear fuel bead |
| CN109324070B (en) * | 2018-08-08 | 2021-04-02 | 中国原子能科学研究院 | Passive neutron analysis method for uranium plutonium content in waste cladding |
| JP7074615B2 (en) * | 2018-08-27 | 2022-05-24 | 株式会社東芝 | Neutron supply device and neutron supply method |
| CN109402413B (en) * | 2018-10-30 | 2020-11-03 | 中国工程物理研究院核物理与化学研究所 | Method for recovering palladium in fission product of spent fuel element |
| CN111188084B (en) * | 2020-01-09 | 2021-04-02 | 中国原子能科学研究院 | Replaceable integral electrode suitable for hot chamber application and electrode replacing device |
| GB2606640A (en) * | 2020-10-14 | 2022-11-16 | China Nuclear Power Technology Res Inst Co Ltd | Dry aftertreatment method for spent fuel employing plasma |
| CN113684504B (en) * | 2021-07-27 | 2022-12-09 | 西安交通大学 | Electrolytic refining waste molten salt treatment method for spent fuel dry-process post-treatment |
| CN114752783B (en) * | 2022-04-22 | 2023-10-03 | 厦门稀土材料研究所 | High-efficiency Sr separation 2+ And Cs + Is a method of (2) |
| CN116665942B (en) * | 2023-05-29 | 2024-01-23 | 西安交通大学 | Spent fuel nuclide pre-separation method |
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Also Published As
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| CN101593566B (en) | 2012-08-29 |
| CN101593566A (en) | 2009-12-02 |
| RU2403634C1 (en) | 2010-11-10 |
| FR2931989A1 (en) | 2009-12-04 |
| GB2461370B (en) | 2010-10-20 |
| US20090294299A1 (en) | 2009-12-03 |
| JP2009288178A (en) | 2009-12-10 |
| JP5193687B2 (en) | 2013-05-08 |
| GB0909309D0 (en) | 2009-07-15 |
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