WO2004036595A1 - Procede et appareil permettant de retraiter un combustible epuise provenant d'un reacteur a eau ordinaire - Google Patents

Procede et appareil permettant de retraiter un combustible epuise provenant d'un reacteur a eau ordinaire

Info

Publication number
WO2004036595A1
WO2004036595A1 PCT/JP2003/013255 JP0313255W WO2004036595A1 WO 2004036595 A1 WO2004036595 A1 WO 2004036595A1 JP 0313255 W JP0313255 W JP 0313255W WO 2004036595 A1 WO2004036595 A1 WO 2004036595A1
Authority
WO
WIPO (PCT)
Prior art keywords
elution
molten salt
anode
cathode
spent fuel
Prior art date
Application number
PCT/JP2003/013255
Other languages
English (en)
Japanese (ja)
Inventor
Yoshiharu Sakamura
Masaki Kurata
Masatoshi Iizuka
Tadashi Inoue
Original Assignee
Central Research Institute Of Electric Power Industry
Priority date (The priority date is an assumption and is not a legal conclusion. Google has not performed a legal analysis and makes no representation as to the accuracy of the date listed.)
Filing date
Publication date
Application filed by Central Research Institute Of Electric Power Industry filed Critical Central Research Institute Of Electric Power Industry
Priority to AU2003273024A priority Critical patent/AU2003273024A1/en
Priority to JP2004544970A priority patent/JPWO2004036595A1/ja
Publication of WO2004036595A1 publication Critical patent/WO2004036595A1/fr

Links

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
    • 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
    • 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

Definitions

  • the present invention relates to a method and an apparatus for reprocessing spent fuel of a light water reactor. More specifically, the present invention relates to a method and apparatus for reprocessing spent fuel of a light water reactor using electrolysis.
  • light water reactor spent fuel is a concept that includes light water reactor MOX spent fuel in addition to so-called light water reactor spent fuel.
  • Electrochemical reduction of metal oxides in molted salts a method of performing electrolysis in lithium chloride molten salt using platinum oxide as a cathode and platinum as an anode, has been implemented, Light Metals 2002, ed., W. Schneider, TM S (The Minerals, Metals & Materials Society), P. 1075). According to this, uranium oxide on the cathode is reduced to uranium metal.
  • the plutonium content in the spent fuel of a light water reactor is usually only about 1 wt% with respect to peran, and plutonium was originally added to the fuel.
  • the plutonium content in the spent fuel is only about 5 wt% of uranium, even if it is used as a fuel, and it can be obtained when these electrolytic reduction methods are applied directly to the spent fuel in light water reactors.
  • the plutonium content of the uranium-platinum alloy is kept below about 5 wt%. For this reason, the plutonium content of the uranium-pnoletium alloy is about 10-3 wt%.
  • plutonium In order to obtain the required fast reactor fuel, plutonium must be added from outside to the uranium-pump alloy recovered by electrolytic reduction, and it is essential to be able to procure plutonium, which increases costs. It will be connected.
  • the present invention provides a method for reprocessing spent fuel of a light water reactor, which can obtain a metal fuel for a fast reactor without adding plutonium from the outside by processing the spent fuel of the light water reactor, and an apparatus utilizing the same.
  • the purpose is to provide.
  • the light water reactor spent fuel reprocessing method of the present invention comprises a de-coating step of de-coating the light water reactor spent fuel to be re-processed, and eluting the light water reactor spent fuel after the de-coating.
  • Uranium oxide from the spent fuel of the light water reactor by applying a voltage to the elution anode and the elution cathode, and immersing it in the molten salt to which the uranium oxide supply source is added, together with the elution cathode.
  • An electrolytic reduction step of applying a voltage to the reducing anode and the reducing cathode to reduce the residue is provided.
  • the light water reactor spent fuel reprocessing device of the present invention includes an elution anode for holding the light water reactor spent fuel to be reprocessed, an elution cathode, and an oxidation in which the elution anode and the elution cathode are immersed.
  • a Molten salt for elution to which the orchid supply source is added an elution container for storing the molten salt for elution, and a voltage applied to the anode for elution and the cathode for elution to remove uranium oxide from the spent fuel of the light water reactor
  • a uranium oxide eluting unit having a direct current power source for elution into the molten salt for elution, a reducing cathode holding the residue of the eluting anode, a reducing anode, a reducing anode and a reducing cathode.
  • a molten salt for reduction to which an oxygen supply source to be immersed is added, a reducing container for storing the molten salt for reduction, a direct current for reduction for applying a voltage to the reducing anode and the reducing cathode to reduce the residue And an electrolytic reduction unit having a power supply. It is.
  • the uranium oxide elution step only the uranium is removed from the spent fuel of the light water reactor and the plutonium content in the remaining oxide is relatively increased, and then, for example, is increased to about 10 to 30 wt%.
  • This plutonium content remains high Since an alloy can be obtained by performing an electrolytic reduction step on the distillate, the obtained alloy can be used as it is as a raw material for a fast reactor fuel without adding plutonium from the outside.
  • the amount of uranium oxide eluted into the molten salt is determined by the amount of electricity, it excludes all uranium oxide such as plutonium remaining on the anode (including all minor elements (MA: neptunium, americium, curium, etc.) ), That is, the ratio of these to uranium oxide can be freely controlled.
  • the amount of spent fuel in the light water reactor that must be treated in the electrolytic reduction step is greatly reduced, which means that the economics of the entire reprocessing process It is also advantageous from the viewpoint.
  • the recovery of the excess perrane in the form of a chemically stable oxide is advantageous from the viewpoint of storage of the excess perrane.
  • plutonium not only plutonium but also a minor-actide element, which is a long half-life nuclide, can be recovered as a metal, which is excellent in terms of reducing environmental load. Furthermore, since the content of plutonium and the like can be freely controlled by the amount of electricity supplied in the uranium oxide elution step, it can be regenerated as a metal fuel for fast reactors whose components are arbitrarily adjusted.
  • the zircaloy alloy in the fuel cladding tube is removed in advance by performing the de-coating step prior to the uranium oxide elution step.
  • uranium oxide can be eluted with high current efficiency from the spent fuel of the light water reactor at the anode, and zircaloy alloy and High-purity uranium oxide that does not mix can be selectively deposited on the elution cathode and easily recovered.
  • uranium oxide is deposited on the elution cathode and recovered in the uranium oxide elution step, high-purity uranium oxide can be recovered.
  • uranium oxide with higher purity has lower radioactivity and is easier to handle than those with lower purity containing fission products and plutonium, and is suitable for use as a raw material for fast reactor planks or for storage and storage. .
  • obtaining high-purity uranium oxide requires handling and It is preferable from the viewpoint of storage and the like.
  • the molten salt for elution contains a uranium oxide source, and the molten salt for reduction contains an oxygen source.
  • the molten salt for reduction contains an oxygen source.
  • Peryl chloride as a uranium oxide supply source and to use calcium oxide as an oxygen supply source.
  • the addition of the uranium oxide supply source to the elution molten salt can precipitate uranium oxide supplied from the uranium oxide supply source from the early stage of operation on the negative electrode. Can be reduced and metallic lithium is not generated, thereby increasing the processing speed.
  • uranium oxide is deposited on the cathode as much as it dissolves into the molten salt from the spent fuel of the light water reactor, so that the uranium oxide elution process can be continued without using an additional uranium oxide supply source.
  • an oxygen supply source to the molten salt for reduction can be carried out by performing an electrolytic reduction treatment using oxide ions supplied from the oxygen supply source from the beginning of the operation as a charge carrier, and the gas generated from the anode is oxygen.
  • the process speed can be increased without generating corrosive chlorine gas.
  • the oxide ions released from the residue of the elution anode function as charge carriers, so that the electrolytic reduction treatment can be continued without using an additional oxygen supply source.
  • the molten salt for elution is preferably a lithium chloride potassium monochloride molten salt.
  • the operating temperature for dissolving uranium oxide can be suppressed to about 500 ° C, so that corrosion of the dissolution containers and pipes can be suppressed, and the initial equipment costs and maintenance of the reprocessing equipment can be reduced. ⁇ Management costs can be reduced.
  • alkali metals, alkaline earth metals, and divalent rare earth elements accumulate in the molten salt of lithium chloride and potassium chloride, so that they can be selectively removed from the molten salt using zeolite. It is possible to remove it. Therefore, the molten salt can be used for a long time, and the amount of salt waste is reduced.
  • the molten salt for reduction is preferably a calcium chloride molten salt.
  • molten calcium chloride since the upper limit of the potential that can be applied without decomposing the salt is large (the potential window is wide), all the residue of the elution anode up to oxides of fission products such as rare earth elements is removed. Since it can be reduced to metal, no oxygen remains in the resulting alloy. Therefore, the alloy obtained by reduction excludes oxides. It can be used as it is as a raw material for fast reactor fuel without any additional process.
  • FIG. 1 is a schematic view showing an embodiment of a reprocessing device for a spent fuel of a light water reactor according to the present invention.
  • FIG. 2 is a flowchart showing an embodiment of a method for reprocessing spent fuel of a light water reactor.
  • FIG. 1 shows one embodiment of a reprocessing device for spent fuel of a light water reactor according to the present invention.
  • This reprocessing device 9 is roughly divided into a uranium oxide elution section 8 that performs an oxidized lanthanum elution step to elute uranium oxide from the spent fuel 1 of the light water reactor, and a part of the uranium oxide is removed in the uranium oxide elution step.
  • It is composed of an electrolytic reduction section 16 for performing an electrolytic reduction step of electrolytic reduction of an anode residue composed of an oxide mixed with Pu, U, FP, MA and the like after the reduction to metal.
  • the uranium oxide elution section 8 is composed of an elution anode 2 for holding the spent fuel 1 of the light water reactor to be reprocessed, an elution cathode 3, and an elution molten salt 4 into which the elution anode 2 and the elution cathode 3 are immersed.
  • a voltage is applied to the elution container 5 for storing the molten salt 4 for elution, the anode 2 for elution and the cathode 3 for elution, and uranium oxide 6 is dissolved from the spent fuel 1 in the light water reactor.
  • an elution DC power supply 7 for elution at the same time and elution at the elution cathode 3.
  • the elution molten salt 4 is a lithium chloride monochloride molten salt.
  • the operating temperature for the elution of uranium oxide 6 can be suppressed to about 500 ° C., and corrosion of the elution container 5 and pipes can be suppressed.
  • the elution molten salt 4 is previously dissolved Uraniru chloride (U 0 2 C 1 2) as uranium oxide source.
  • Uraniru chloride U 0 2 C 1 2
  • the precipitation of metallic lithium can be prevented from the early stage of operation, and the recovery rate of uranium oxide can be increased.
  • uranium oxide 6 can be efficiently recovered at the elution cathode 3.
  • the voltage applied between the spent fuel 1 for light water reactor held by the elution anode 2 and the elution cathode 3 is about 1 V.
  • this value differs depending on the type and temperature of the molten salt for elution 4 and the shape, size, and spacing of the electrodes 2 and 3.
  • Most of the spent fuel in light water reactors is uranium oxide Therefore, when electrolysis is performed with an extremely small current, electrolysis proceeds without applying a voltage, but in reality, a voltage of about 1 V is applied to increase the current value, that is, the recovery rate. It can be assumed that it is necessary to apply.
  • the amount of uranium oxide 6 eluted from the light water reactor spent fuel 1 can be changed by adjusting the amount of electricity supplied to the elution anode 2 and the elution cathode 3.
  • the plutonium content in the oxide remaining in the spent fuel 1 of the light water reactor can be relatively increased. For example, it can be increased to a target value of about 10 to 30 wt%.
  • alkali metal, alkaline earth metal, divalent rare earth element, halogen, and the like which are salt-soluble fission products, are eluted into the elution molten salt 4. These remain in the salt as is and are removed from residue 10.
  • the dissolution container 5 is made of carbon such as graphite or pyrographite.
  • the dissolving anode 2 has a basket 18 made of carbon such as graphite or pyrographite.
  • the spent fuel 1 of the light water reactor is stored in the basket 18.
  • the spent fuel 1 is soaked in the molten salt 4 for dissolution by immersing the entire packet 18 in the molten salt 4 for dissolution.
  • the elution cathode 3 is made of carbon such as graphite or pyrographite.
  • Uranium oxide 6 is deposited on the elution cathode 3 by reduction.
  • the electrolytic reduction section 16 is composed of a reduction cathode 11 holding the residue (anode residue) 10 of the elution anode 2 generated in the oxidized peran oxide elution section 8, a reduction anode 12, and a reduction anode 12.
  • a reduction direct-current power supply 15 for reducing the residue 10 by applying a voltage.
  • the molten salt for reduction 13 is a calcium chloride molten salt.
  • An oxygen supply source is added to the reducing molten salt 13 in advance.
  • the electrolytic reduction treatment can be performed using the oxide ions supplied from the oxygen supply source as a charge carrier, and during the operation, the oxide ions released from the residue 10 are charged. Work as a carrier.
  • the electrolytic reduction of the residue 10 can be performed using the oxide ion as a carrier. Therefore, since the fuel component such as plutonium of the residue 10 does not dissolve into the reducing molten salt 13, the recovery rate can be increased.
  • Oxygen is also removed from non-eluted nuclear fission products in the residue 10, and can be recovered together with the fuel component. Further, since the gas generated from the reducing anode 12 is oxygen or carbon dioxide, the generation of corrosive chlorine gas can be avoided.
  • the treatment speed can be increased without generating chlorine gas.
  • oxide ions must be supplied to the reducing anode 12 at a rate corresponding to the operation processing speed. It is effective to increase the concentration of oxides in the molten salt for reduction 13 and to increase the concentration of oxide ions in the molten salt 13 for reduction. It is better not to add too much sashimi because it has the effect of suppressing the elution of oxygen.
  • the molten salt for reduction 13 is prepared by dissolving 0.01 to 5 wt% of calcium oxide using calcium chloride as a solvent.
  • the addition of calcium oxide lowers the melting point of calcium chloride from the original 775 ° C to a maximum of 75 ° C (when CaO is 3.4 wt ° / 0 ).
  • the actual voltage applied between the residue 10 held on the reduction cathode 11 and the reduction anode 12 is about 3 V.
  • Some FPs that are not reduced within the potential difference range of the present embodiment that is, alkali metals, alkaline earth metals, divalent rare earth elements, etc. are soluble in salts, and are already melted for elution in the uranium oxide elution step. Dissolved in salt and separated. The fission products dissolved in these salts are not reduced and do not deposit on the cathode, but remain in the salts.
  • Alkali metals, alkaline earth metals, and divalent rare earth elements are removed by dissolving in the molten salt for dissolution, lithium chloride and potassium chloride, and are not discarded if they are not introduced into the reduced salt, calcium chloride.
  • This is advantageous in terms of material processing. That is, The alkali metals cesium and arsenic earth metal stotium are exothermic fission products, and when these accumulate in the molten salt, the exotherm at some point will render the molten salt unusable.
  • a technology has been developed to remove cesium and strontium from the molten salt using zeolite.
  • the reduction vessel 14 is preferably made of stainless steel, low carbon steel, or special steel such as titanium.
  • the reduction cathode 11 is in the form of a past 17 made of carbon, stainless steel, low carbon steel, or special steel such as zirconium or titanium.
  • the residue 10 is stored inside the basket 17. Then, by immersing the entire basket 17 in the molten salt for reduction 13, the residue 10 is immersed in the molten salt 13 for reduction.
  • the reduction anode 12 is preferably made of platinum.
  • the spent fuel 1 of the light water reactor is dismantled and sheared (Step 1: S 1).
  • the disassembled and sheared zirconate cladding of the spent fuel 1 of the light water reactor is removed (de-cladding heating step (step 2 ⁇ S 2)).
  • de-cladding heating step step 2 ⁇ S 2
  • a dismantled light water reactor A slit is made in the fuel rod of spent fuel 1 and it is heated to about 500 ° C in the atmosphere.
  • This I Ri UO 2 is the volume is oxidized is spread the covering expands tube U 3 ⁇ 8. Then, the oxide is separated from the cladding tube and recovered by giving a slight vibration.
  • the recovered oxide is further heated to about 100 ° C and heated to obtain volatile fission of alkali metals, chalcogens, halogens, and some noble metals.
  • the product (FP) is removed.
  • removing some FP before the uranium oxide elution step has the effect of suppressing the accumulation of salt-soluble FP, such as Alkyri metal, in the molten salt for elution 4.
  • the heat treatment at about 100 ° C. described above may not be necessary in some cases, because it is not necessary from the viewpoint of elution of uranium oxide and recovery at the cathode.
  • U 3 0 8 will be reacted in an air stream containing hydrogen if necessary, it is reduced to U 0 2.
  • Uranium oxide elution step (Step 3: S3)
  • the spent fuel 1 of the light water reactor to be reprocessed is held on the elution anode 2 and immersed together with the elution cathode 3 in the elution molten salt 4 to which a uranium oxide supply source has been added, to elute.
  • the uranium oxide 6 is eluted from the spent fuel 1 of the light water reactor into the molten salt 4 for elution by applying a voltage to the anode 2 for elution and the cathode 3 for elution.
  • the molten salt 4 for elution is put into the container 5 for elution using the uranium oxide elution section 8 and heated to about 500 ° C.
  • the spent fuel 1 of the light water reactor obtained in the de-coating heat treatment step is accommodated in the passivation 18 of the anode 2 for elution and immersed in the molten salt 4 for elution.
  • the elution cathode 3 is also immersed in the elution molten salt 4.
  • the uranium oxide 6 is dissolved from the spent fuel 1 in the light water reactor into the molten salt 4 for elution by connecting the elution DC power supply 7 to the anode 2 and the cathode 3 and applying a voltage of about IV.
  • the amount of uranium oxide 6 eluted from the LWR spent fuel 1 was changed by adjusting the amount of electricity supplied to the elution anode 2 and the elution cathode 3 to reduce the amount of light water.
  • the plutonium content in the oxide remaining in the spent nuclear fuel 1 is increased to, for example, about 10 to 30 wt%.
  • uranium oxide is eluted into the salt by a reaction represented by the following chemical formula 1.
  • the elution anode 2 does not elute the element of plutonium minor activity. This is because, as shown in Table 2, the oxidation potential of lanthanum oxide 6 present in a large amount is on the minus side as compared with neptum oxide and plutonium oxide.
  • perulion is reduced by a reaction represented by the following chemical formula 2, and uranium oxide 6 is precipitated.
  • the elution molten salt 4 because pre U0 2 C 12 is dissolved, can be recovered efficiently oxidized uranium 6 in eluting negative electrode 3. As a result, highly pure peroxidized 6 can be efficiently recovered.
  • Step 4 Move to S 4
  • the residue 10 of the elution anode 2 is retained on the reduction cathode 11 and immersed together with the reduction anode 12 in a reduction molten salt 13 to which an oxygen supply source has been added, and A voltage is applied to the anode 12 and the reducing anode 11 to reduce the residue 10.
  • the molten salt for reduction 13 is put into the container for reduction 14 using the electrolytic reduction section 16 and melted at about 800.
  • the residue 10 from the uranium oxide elution step is accommodated in a packet of the reducing cathode 11 and immersed in the reducing molten salt 13.
  • the anode for reduction 12 is also immersed in the molten salt for reduction 13.
  • the residue 10 is electrolytically reduced by connecting a DC power source 15 for reduction to the anode and the cathode and applying a voltage of about 3 V.
  • the oxide ions supplied from the calcium oxide and the oxide ions eluted from the reducing cathode force migrate through the reducing molten salt 13 to the reducing anode. Further, in the residue 10, oxygen which forms a solid solution acts as an ion conductor.
  • the oxidized oxide is reduced to a metal by a reaction represented by the following chemical formula 3 to be an alloy product.
  • non-eluted FP among fission products can be reduced at the same time as lanthanum oxide and plutonium oxide. Therefore, the final product alloy contains non-eluting FP and minor actinide in addition to uranium pull tonium.
  • the reduced metal is taken out of the molten salt for reduction 13 by lifting the basket 17, and becomes a raw material for a new metal nuclear fuel by removing the attached salt as necessary. Since the plutonium content of the alloy obtained according to the present embodiment has been increased to 10-3 O wt%, it can be used as a raw material for fast reactor fuel only by removing the attached salt.
  • eluted FP which accounts for about 40% of fission products, is decontaminated without loss of uranium, plutonium, and minor arctied.
  • oxygen gas is generated at the reducing anode 12 by a reaction represented by the following chemical formula 4.
  • a uranium oxide supply source and an oxygen supply source are added to each molten salt in advance.
  • a uranium oxide supply source and an oxygen There is no need to add a source. There is no problem in terms of processing speed in dissolving uranium oxide from light water reactor spent fuel and in dissolving oxygen from anode residue. Supplying chlorine gas to the elution cathode in the early stage of the operation.
  • the chlorine gas generated from the reducing anode may be collected.
  • each molten salt to which a uranium oxide supply source and an oxygen supply source are added in advance eliminates the need for equipment that can handle chlorine gas, and can reduce the equipment costs of reprocessing equipment and the maintenance and management costs.
  • the molten salt for elution 4 is a molten salt of lithium chloride monochloride.
  • the present invention is not limited to this, and it is possible to operate at a higher temperature and to reduce the influence of moisture in the gas phase. Sodium chloride monochloride Salt added with shim or lithium chloride can be used.
  • a calcium chloride molten salt is used as the reducing molten salt 13.
  • the present invention is not limited to this, and in order to lower the melting point, a salt obtained by adding calcium chloride, potassium chloride, or the like to calcium chloride, Alternatively, a salt composed of an alkali metal chloride can be used.
  • calcium oxide is used as an oxygen supply source in the early stage of the operation.
  • the present invention is not limited to this, and oxides soluble in the reducing molten salt 13 such as lithium oxide and barium oxide are melted. It can be used according to the type of salt.
  • the reducing anode 12 is made of platinum, but is not limited thereto, and may be made of, for example, carbon.
  • oxide ions react with carbon on the surface of the reducing anode 12 to become carbon dioxide or carbon monoxide. This carbon dioxide or carbon monoxide is discharged as gas bubbles into the gas phase. In this case, the reducing anode 12 is worn out and should be replaced periodically.
  • the actual voltage applied between the residue 10 and the reducing anode 12 in the electrolytic reduction section 16 is set to about 3 V.
  • the potential is not limited to this.

Landscapes

  • Physics & Mathematics (AREA)
  • Engineering & Computer Science (AREA)
  • Plasma & Fusion (AREA)
  • General Engineering & Computer Science (AREA)
  • High Energy & Nuclear Physics (AREA)
  • Electrolytic Production Of Metals (AREA)

Abstract

L'invention concerne un appareil permettant de retraiter un combustible épuisé provenant d'un réacteur à eau ordinaire. Cet appareil comprend une section d'élution d'oxyde d'uranium (8) effectuant une étape d'élution d'oxyde d'uranium consistant à immerger le combustible épuisé (1) provenant d'un réacteur à eau ordinaire dans un sel fondu d'élution (4) ajouté à une source d'alimentation d'oxyde d'uranium utilisée comme anode d'élution (2) conjointement à une cathode d'élution (3), puis appliquant une tension de manière à éluer l'oxyde d'uranium (6) du combustible épuisé (1) du réacteur à eau ordinaire dans le sel fondu d'élution (4) et le précipitant sur la cathode d'élution (3), ainsi qu'une section de réduction électrolytique effectuant une étape consistant à immerger le résidu (10) de l'anode d'élution dans un sel fondu de réduction (13) ajouté à une source d'alimentation d'oxygène utilisée comme cathode de réduction (11) conjointement à une anode de réduction (12), puis appliquant une tension entre l'anode de réduction et la cathode (11), de manière à réduire ainsi le résidu (10) et à produire un alliage ; ainsi qu'un procédé de retraitement du combustible épuisé au moyen de l'appareil. Celui-ci et le procédé permettent de produire un combustible destiné à un réacteur rapide par le biais du traitement du combustible épuisé provenant d'un réacteur à eau ordinaire sans ajout de plutonium de l'extérieur.
PCT/JP2003/013255 2002-10-16 2003-10-16 Procede et appareil permettant de retraiter un combustible epuise provenant d'un reacteur a eau ordinaire WO2004036595A1 (fr)

Priority Applications (2)

Application Number Priority Date Filing Date Title
AU2003273024A AU2003273024A1 (en) 2002-10-16 2003-10-16 Method and apparatus for reprocessing spent fuel from light-water reactor
JP2004544970A JPWO2004036595A1 (ja) 2002-10-16 2003-10-16 軽水炉使用済燃料の再処理方法および装置

Applications Claiming Priority (2)

Application Number Priority Date Filing Date Title
JP2002301588 2002-10-16
JP2002/301588 2002-10-16

Publications (1)

Publication Number Publication Date
WO2004036595A1 true WO2004036595A1 (fr) 2004-04-29

Family

ID=32105023

Family Applications (1)

Application Number Title Priority Date Filing Date
PCT/JP2003/013255 WO2004036595A1 (fr) 2002-10-16 2003-10-16 Procede et appareil permettant de retraiter un combustible epuise provenant d'un reacteur a eau ordinaire

Country Status (3)

Country Link
JP (1) JPWO2004036595A1 (fr)
AU (1) AU2003273024A1 (fr)
WO (1) WO2004036595A1 (fr)

Cited By (7)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JP2013088117A (ja) * 2011-10-13 2013-05-13 Toshiba Corp 炉心溶融物の処理方法
KR101711808B1 (ko) * 2015-12-23 2017-03-06 한국원자력연구원 세슘화합물의 제거율이 높은 전해환원공정용 연료파편 제조 방법
US9799414B2 (en) 2010-09-03 2017-10-24 Atomic Energy Of Canada Limited Nuclear fuel bundle containing thorium and nuclear reactor comprising same
US10176898B2 (en) 2010-11-15 2019-01-08 Atomic Energy Of Canada Limited Nuclear fuel containing a neutron absorber
KR20200008308A (ko) * 2018-07-16 2020-01-28 한국원자력연구원 순환 반응을 이용한 방사성 금속 산화물의 환원 장치 및 방법
US10950356B2 (en) 2010-11-15 2021-03-16 Atomic Energy Of Canada Limited Nuclear fuel containing recycled and depleted uranium, and nuclear fuel bundle and nuclear reactor comprising same
US11894154B2 (en) 2022-02-02 2024-02-06 Curio Solutions Llc Modular, integrated, automated, compact, and proliferation-hardened method to chemically recycle used nuclear fuel (UNF) originating from nuclear reactors to recover a mixture of transuranic (TRU) elements for advanced reactor fuel to recycle uranium and zirconium

Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH1073693A (ja) * 1996-08-30 1998-03-17 Hitachi Ltd 使用済原子燃料の再処理方法
JPH10111388A (ja) * 1996-10-07 1998-04-28 Toshiba Corp 使用済み燃料の再処理方法
JPH11218597A (ja) * 1998-02-03 1999-08-10 Hitachi Ltd 使用済燃料の再処理方法
WO2001041152A2 (fr) * 1999-12-03 2001-06-07 British Nuclear Fuels Plc Production d'actinides
JP2002071865A (ja) * 2000-09-04 2002-03-12 Hitachi Ltd 原子炉用混合酸化物燃料の製造方法

Patent Citations (5)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
JPH1073693A (ja) * 1996-08-30 1998-03-17 Hitachi Ltd 使用済原子燃料の再処理方法
JPH10111388A (ja) * 1996-10-07 1998-04-28 Toshiba Corp 使用済み燃料の再処理方法
JPH11218597A (ja) * 1998-02-03 1999-08-10 Hitachi Ltd 使用済燃料の再処理方法
WO2001041152A2 (fr) * 1999-12-03 2001-06-07 British Nuclear Fuels Plc Production d'actinides
JP2002071865A (ja) * 2000-09-04 2002-03-12 Hitachi Ltd 原子炉用混合酸化物燃料の製造方法

Non-Patent Citations (2)

* Cited by examiner, † Cited by third party
Title
KURATA MASAKI ET AL.: "L7 denkai kangen gijutsu ni motozuku sankabutsu nenryo no kanshiki kan'i saishoriho (1) - Mogi shiyozumi sankabutsu nenryo no denkai kangen shiken", ATOMIC ENERGY SOCIETY OF JAPAN, vol. 40, (SEPARATE VOL. III), 5 March 2002 (2002-03-05), pages 610, XP002977651 *
OTA KOICHI ET AL.: "L8 denkai kangen gijutsu ni motozuku sankabutsu nenryo no kanshiki kan'i saishoriho (2) - Kaishubutsu o mochiita kinzoku nenryo kosokuro no nenryo sosei hyoka oyobi nensho tokusei kaiseki", ATOMIC ENERGY SOCIETY OF JAPAN, vol. 40, (SEPARATE VOL. III), 5 March 2002 (2002-03-05), pages 611, XP002977652 *

Cited By (8)

* Cited by examiner, † Cited by third party
Publication number Priority date Publication date Assignee Title
US9799414B2 (en) 2010-09-03 2017-10-24 Atomic Energy Of Canada Limited Nuclear fuel bundle containing thorium and nuclear reactor comprising same
US10176898B2 (en) 2010-11-15 2019-01-08 Atomic Energy Of Canada Limited Nuclear fuel containing a neutron absorber
US10950356B2 (en) 2010-11-15 2021-03-16 Atomic Energy Of Canada Limited Nuclear fuel containing recycled and depleted uranium, and nuclear fuel bundle and nuclear reactor comprising same
JP2013088117A (ja) * 2011-10-13 2013-05-13 Toshiba Corp 炉心溶融物の処理方法
KR101711808B1 (ko) * 2015-12-23 2017-03-06 한국원자력연구원 세슘화합물의 제거율이 높은 전해환원공정용 연료파편 제조 방법
KR20200008308A (ko) * 2018-07-16 2020-01-28 한국원자력연구원 순환 반응을 이용한 방사성 금속 산화물의 환원 장치 및 방법
KR102160973B1 (ko) * 2018-07-16 2020-09-29 한국원자력연구원 순환 반응을 이용한 방사성 금속 산화물의 환원 장치 및 방법
US11894154B2 (en) 2022-02-02 2024-02-06 Curio Solutions Llc Modular, integrated, automated, compact, and proliferation-hardened method to chemically recycle used nuclear fuel (UNF) originating from nuclear reactors to recover a mixture of transuranic (TRU) elements for advanced reactor fuel to recycle uranium and zirconium

Also Published As

Publication number Publication date
JPWO2004036595A1 (ja) 2006-02-16
AU2003273024A1 (en) 2004-05-04

Similar Documents

Publication Publication Date Title
Herrmann et al. Separation and recovery of uranium metal from spent light water reactor fuel via electrolytic reduction and electrorefining
Souček et al. Pyrochemical reprocessing of spent fuel by electrochemical techniques using solid aluminium cathodes
Souček et al. Exhaustive electrolysis for recovery of actinides from molten LiCl–KCl using solid aluminium cathodes
JP3549865B2 (ja) 使用済核燃料中の希少元素ｆｐの分離回収方法およびこれを利用した原子力発電−燃料電池発電共生システム
EP1481401B1 (fr) Cellule electrochimique pour la production de metal
JP2008134096A (ja) 使用済み酸化物原子燃料の還元装置及びリチウム再生電解装置
JPH03123896A (ja) アクチニド回収
Galashev Recovery of actinides and fission products from spent nuclear fuel via electrolytic reduction: Thematic overview
WO2004036595A1 (fr) Procede et appareil permettant de retraiter un combustible epuise provenant d'un reacteur a eau ordinaire
JP4487031B2 (ja) 使用済酸化物燃料の乾式再処理方法
EP1240647B1 (fr) Production d'actinides
JP3763980B2 (ja) 使用済み酸化物燃料の還元装置およびその還元方法
EP1393324B1 (fr) Production d'actinide
US5096545A (en) Plutonium recovery from spent reactor fuel by uranium displacement
JP3519557B2 (ja) 使用済み燃料の再処理方法
Jeong et al. An experimental study on an electrochemical reduction of an oxide mixture in the advanced spent-fuel conditioning process
JP3872873B2 (ja) 使用済み燃料の溶融塩電解再処理方法
JP5017069B2 (ja) 使用済燃料の再処理方法
RU2499306C1 (ru) Способ очистки облученного ядерного топлива
JP2005315790A (ja) 使用済み酸化物燃料の再処理方法
Stevenson Development of a novel electrochemical pyroprocessing methodology for spent nuclear fuels
JPWO2003063178A1 (ja) 使用済み酸化物燃料の電解還元方法および乾式簡易再処理方法
WO2011144937A1 (fr) Nouveau procédé de retraitement
EP1570114A2 (fr) Separation de metaux
Zhou et al. Safeguards for pyroprocessing

Legal Events

Date Code Title Description
AK Designated states

Kind code of ref document: A1

Designated state(s): AE AG AL AM AT AU AZ BA BB BG BR BY BZ CA CH CN CO CR CU CZ DE DK DM DZ EC EE EG ES FI GB GD GE GH GM HR HU ID IL IN IS JP KE KG KP KR KZ LC LK LR LS LT LU LV MA MD MG MK MN MW MX MZ NI NO NZ OM PG PH PL PT RO RU SC SD SE SG SK SL SY TJ TM TN TR TT TZ UA UG US UZ VC VN YU ZA ZM ZW

AL Designated countries for regional patents

Kind code of ref document: A1

Designated state(s): GH GM KE LS MW MZ SD SL SZ TZ UG ZM ZW AM AZ BY KG KZ MD RU TJ TM AT BE BG CH CY CZ DE DK EE ES FI FR GB GR HU IE IT LU MC NL PT RO SE SI SK TR BF BJ CF CG CI CM GA GN GQ GW ML MR NE SN TD TG

121 Ep: the epo has been informed by wipo that ep was designated in this application
DFPE Request for preliminary examination filed prior to expiration of 19th month from priority date (pct application filed before 20040101)
WWE Wipo information: entry into national phase

Ref document number: 2004544970

Country of ref document: JP

122 Ep: pct application non-entry in european phase