WO2004036595A1 - Method and apparatus for reprocessing spent fuel from light-water reactor - Google Patents

Method and apparatus for reprocessing spent fuel from light-water reactor

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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
French (fr)
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 JP2004544970A priority Critical patent/JPWO2004036595A1/en
Priority to AU2003273024A priority patent/AU2003273024A1/en
Publication of WO2004036595A1 publication Critical patent/WO2004036595A1/en

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Classifications

    • GPHYSICS
    • G21NUCLEAR PHYSICS; NUCLEAR ENGINEERING
    • G21CNUCLEAR REACTORS
    • G21C19/00Arrangements 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
    • G21C19/42Reprocessing of irradiated fuel
    • G21C19/44Reprocessing of irradiated fuel of irradiated solid fuel
    • YGENERAL 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
    • Y02TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02EREDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
    • Y02E30/00Energy generation of nuclear origin
    • Y02E30/30Nuclear fission reactors
    • YGENERAL 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
    • Y02TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
    • Y02WCLIMATE CHANGE MITIGATION TECHNOLOGIES RELATED TO WASTEWATER TREATMENT OR WASTE MANAGEMENT
    • Y02W30/00Technologies for solid waste management
    • Y02W30/50Reuse, 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.

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  • 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

An apparatus for reprocessing a spent fuel from a light-water reactor, which comprises an uranium oxide elution section (8) practicing an uranium oxide elution step of immersing a spent fuel (1) from a light-water reactor in an eluting molten salt (4) added with an uranium oxide supplying source as an eluting anode (2) together with an eluting cathode (3), followed by application of a voltage, to thereby elute uranium oxide (6) from the spent fuel (1) of a light-water reactor into the eluting molten salt (4) and precipitate it on the eluting cathode (3), and an electrolytic reduction section (16) practicing a step of immersing the residue (10) of the eluting anode (2) in a reducing molten salt (13) added with an oxygen supplying source as a reducing cathode (11) together with a reducing anode (12), followed by application of a voltage between the reducing anode (12) and the cathode (11), to thereby reduce the residue (10) and produce an alloy; and a method for reprocessing the spent fuel using the apparatus. The apparatus and the method allows the production of a fuel for a fast reactor through the processing of a spent fuel from a light water reactor without adding plutonium from outside.

Description

明 細 書  Specification
軽水炉使用済燃料の再処理方法およぴ装置  Light water reactor spent fuel reprocessing method and equipment
技術分野  Technical field
本発明は、 軽水炉使用済燃料の再処理方法および装置に関する。 更に詳述する と、 本発明は電解を利用した軽水炉使用済燃料の再処理方法および装置に関する。  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.
技術用語  Technical terms
本明細書中で 「軽水炉使用済燃料」 とはいわゆる軽水炉使用済燃料のほかに軽 水炉 MO X使用済燃料を含む概念である。  In this specification, “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.
背景技術  Background art
原子力発電所の使用済み酸化物燃料に含まれる核燃料物質を金属に還元して回 収する方法を開発するための試験として、 核分裂反応で生成される元素と酸化ゥ ランとから成る模擬使用済ぺレットを陰極にすると共に黒鉛を陽極にして塩化力 ルシゥム溶融塩中において電気分解を行う手法が実施されている (電解還元技術 に基づく酸化物燃料の乾式簡易再処理法 (I)一模擬使用済酸化物燃料の電解還元試 験—、 日本原子力学会 2002年春の年会予稿集第 III分冊、 P. 610) 。 これによると、 陰極の酸化ウランが金属ウランに還元されると共に、 希土類元素の一部とアル力 リ土類元素が塩中に溶出することが確認されている。  As a test to develop a method for recovering nuclear fuel materials contained in spent oxide fuel of nuclear power plants by reducing them to metal, a simulated spent fuel consisting of elements generated by fission reactions and lanthanum oxide was developed. A method has been implemented in which chloride is used as the cathode and graphite is used as the anode to perform electrolysis in molten molten salt. (Simplified dry dry reprocessing method of oxide fuel based on electrolytic reduction technology (I) Electrolytic reduction test of oxide fuel—The Atomic Energy Society of Japan, 2002 Annual Meeting, Volume III, p. 610). According to this, it has been confirmed that uranium oxide on the cathode is reduced to uranium metal, and that some of the rare earth elements and alkaline earth elements are eluted in the salt.
また、 酸化ゥランを陰極にすると共に白金を陽極にして塩化リチウム溶融塩中 において電気分解を行う手法が実施されている Electrochemical reduction o f metal oxides in molted salts 、 Light Metals 2002, ed., W. Schneider, TM S (The Minerals, Metals & Materials Society)、 P. 1075) 。 これによると、 陰極 の酸化ウランが金属ウランに還元される。  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.
し力 しながら、 上述した電解還元のみによる金属の還元回収方法では、 軽水炉 使用済燃料中のプルトニゥム含有量はゥランに対して通常 1 w t %程度に過ぎず、 また元々プルトユウムが添カ卩されているプルサ一マル燃料であっても使用済燃料 中のプルトニウム含有量はウランに対し 5 w t %程度に過ぎないことから、 これ らの電解還元方法を軽水炉使用済燃料に直接適用した場合に得られるウラン一プ ルトニゥム合金のプルトユウム含有量は約 5 w t %以下に抑えられてしまう。 こ のため、 ウラン一プノレトユウム合金のプルトニウム含有量が 1 0— 3 O w t %程 度必要である高速炉燃料を得るためには、 電解還元で回収したウラン一プ^^ト二 ゥム合金に外部からプルトニウムを添加しなければならず、 プルトニゥムを調達 できることが必須となりコストも高くなつてしまう。 However, in the above-described metal reduction and recovery method using only electrolytic reduction, 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%. 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.
発明の開示  Disclosure of the invention
そこで、 本発明は、 軽水炉使用済燃料を処理することにより外部からプルトニ ゥムを添加することなく高速炉用金属燃料を得ることができる軽水炉使用済燃料 の再処理方法およびこれを利用する装置を提供することを目的とする。  Accordingly, 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.
かかる目的を達成するため、 本発明の軽水炉使用済燃料の再処理方法は、 再処 理対象である軽水炉使用済燃料を脱被覆する脱被覆工程と、 脱被覆後の軽水炉使 用済燃料を溶出用陽極に保持して溶出用陰極と共に酸化ウラン供給源が添加され た溶出用溶融塩に浸し、 溶出用陽極および溶出用陰極に電圧を印加して軽水炉使 用済燃料から酸化ウランを溶出用溶融塩中に溶出させると共に溶出用陰極に析出 させる酸化ウラン溶出工程と、 溶出用陽極の残留物を還元用陰極に保持して還元 用陽極と共に酸素供給源が添加された還元用溶融塩に浸し、 還元用陽極おょぴ還 元用陰極に電圧を印加して残留物を還元する電解還元工程とを備えるようにして いる。  In order to achieve this object, 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. A uranium oxide elution step of eluting the salt into the salt and depositing it on the elution cathode, and holding the residue of the elution anode on the reducing cathode, immersing it in a reducing molten salt to which an oxygen source is added together with the reducing anode, An electrolytic reduction step of applying a voltage to the reducing anode and the reducing cathode to reduce the residue is provided.
また、 本発明の軽水炉使用済燃料の再処理装置は、 再処理対象である軽水炉使 用済燃料を保持する溶出用陽極と、 溶出用陰極と、 溶出用陽極および溶出用陰極 が浸される酸化ゥラン供給源が添加された溶出用溶融塩と、 該溶出用溶融塩を貯 留する溶出用容器と、 溶出用陽極およぴ溶出用陰極に電圧を印加して軽水炉使用 済燃料から酸化ウランを溶出用溶融塩中に溶出させる溶出用直流電源とを有する 酸化ウラン溶出部と、 溶出用陽極の残留物を保持する還元用陰極と、 還元用陽極 と、 還元用陽極およぴ還元用陰極が浸される酸素供給源が添加された還元用溶融 塩と、 該還元用溶融塩を貯留する還元用容器と、 還元用陽極および還元用陰極に 電圧を印加して残留物を還元する還元用直流電源とを有する電解還元部とを備え るようにしている。  Further, 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.溶出 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.
したがって、 まず酸化ウラン溶出工程において軽水炉使用済燃料からウランの みを取り除いて残存する酸化物中のプルトニウム含有量を相対的に高めてから、 例えば 1 0〜3 0 w t %程度に高めてから、 このプルトニウム含有量を高めた残 留物に対して電解還元工程を実行して合金を得ることができるので、 外部からプ ルトニゥムを添加することなく、 得られた合金をそのまま高速炉燃料の原料とし て使用することができる。 Therefore, firstly, in 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.
ここで、 酸化ウランの溶融塩中への溶出量は通電量で決まるので、 陽極に残留 するプルトニウムなどの酸化ウランを除くものの全て (マイナーァクチ二ド元素 (MA :ネプツニウム、 アメリシウム、 キュリウムなど) を含む) の含有量、 即ち これらと酸化ウランとの比率を自由に制御できる。  Here, since 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.
このように高速炉燃料の原料としては余剰の酸化ウランを先に取り除くことに よって、 電解還元工程で処理しなければならない軽水炉使用済燃料を大きく減容 することは、 再処理プロセス全体の経済性から見ても有利である。 しかも、 余剰 のゥランが化学的に安定な酸化物の形態で回収されることは、 余剰ゥランの貯蔵 という観点からも利点である。  By removing surplus uranium oxide first as a raw material for fast reactor fuel, 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. Moreover, 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.
しかも、 本発明の再処理方法によれば、 プルトニウムだけでなく、 長半減期核 種であるマイナーァクチ二ド元素も金属として回収することができるため、 環境 負荷低減の面からも優れている。 更に、 プルトニウムなどの含有量は酸化ウラン 溶出工程での通電量によって自由に制御できるので、 任意に成分調整された高速 炉用金属燃料として再生することができる。  Moreover, according to the reprocessing method of the present invention, 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.
また、 本発明の再処理方法によれば、 酸化ウラン溶出工程に先立ち脱被覆工程 を実行して燃料被覆管のジルカロイ合金を予め除去するようにしているので、 酸 化ゥラン溶出工程においてジルカ口ィ合金が優先的に溶解することにより酸化ゥ ラン溶出の電流効率が低下することを防止して、 陽極で軽水炉使用済燃料から高 い電流効率で酸化ウランを溶出させることができると共に、 ジルカロイ合金と混 ざらない純度の高い酸化ウランを選択的に溶出用陰極に析出させて容易に回収す ることができる。  Further, according to the reprocessing method of the present invention, 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. Preventing the current efficiency of eluting silane oxide from decreasing due to preferential dissolution of the alloy, 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.
しかも、 酸化ウラン溶出工程で酸化ウランを溶出用陰極に析出させて回収して いるので、 高純度の酸化ウランを回収することができる。 ここで、 酸化ウランは 純度が高い方が核分裂生成物やプルトニウムが含有されて純度が低いものよりも 放射能が低く取り扱いが容易で高速炉プランケットの原料としての利用あるいは 保管貯蔵に適している。 このため、 高純度の酸化ウランを得ることは取り扱いや 保管などの観点から好ましい。 Moreover, since uranium oxide is deposited on the elution cathode and recovered in the uranium oxide elution step, high-purity uranium oxide can be recovered. Here, 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. . For this reason, obtaining high-purity uranium oxide requires handling and It is preferable from the viewpoint of storage and the like.
また、 溶出用溶融塩には酸化ウラン供給源が、 還元用溶融塩には酸素供給源が 添加されている。 ここで、 酸化ウラン供給源としては塩化ゥラエルが、 酸素供給 源としては酸化カルシウムの使用が好ましい。 溶出用溶融塩への酸化ウラン供給 源の添加は、 操業初期から酸化ウラン供給源によって供給された酸化ウランを陰 極に析出させうるので、 陰極において溶出用溶融塩の成分である塩ィ匕リチウムが 還元されて金属リチウムが発生することがなく処理速度を高めることができる。 そして、 操業中には軽水炉使用済燃料から溶融塩中に溶け出す分だけ酸化ウラン が陰極に析出するため、 追加の酸化ゥラン供給源を用いることなく酸化ウランの 溶出処理を継続することができる。 また、 還元用溶融塩への酸素供給源の添加は、 操業初期から酸素供給源によって供給された酸化物イオンを電荷の担体として電 解還元処理を行なうことができ、 陽極から発生する気体は酸素や二酸化炭素にな り、 腐食性のある塩素ガスを発生させることがなく処理速度を高めることができ る。 そして、 操業中には溶出用陽極の残留物から放出される酸化物イオンが電荷 の担体として働くため、 追加の酸素供給源を用いることなく電解還元処理を継続 することができる。  The molten salt for elution contains a uranium oxide source, and the molten salt for reduction contains an oxygen source. Here, it is preferable to use 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. During operation, 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. In addition, the addition of 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. During operation, 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.
さらに、 本発明において、 溶出用溶融塩は塩化リチウム一塩化カリウム溶融塩 であることが好ましい。 この場合、 酸化ウランの溶出の操業温度を 5 0 0 °C程度 に抑えることができるので、 溶出用容器や管などの腐食を抑制することができ、 再処理装置の初期設備導入コス ト並びに保守 ·管理コス トを削減できる。 また、 核分裂生成物の内でアルカリ金属、 アルカリ土類金属、 2価の希土類元素は、 こ の塩化リチウム一塩化カリゥム溶融塩中に蓄積するため、 ゼォライトを用いて選 択的に溶融塩中から除去することが可能である。 よって、 溶融塩を長期間使用す ることができ、 塩廃棄物量が低減される。  Furthermore, in the present invention, the molten salt for elution is preferably a lithium chloride potassium monochloride molten salt. In this case, 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. In addition, among the fission products, 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.
また、 本発明において、 還元用溶融塩は塩化カルシウム溶融塩であることが好 ましい。 塩化カルシウム溶融塩では、 塩を分解することなく印加できる電位の上 限が大きいために (電位窓が広い) 、 希土類元素などの核分裂生成物の酸化物ま で、 溶出用陽極の残留物を全て金属に還元することができるために、 得られた合 金には酸素が残留しない。 したがって、 還元により得られる合金は、 酸化物を除 去する工程を付加することなく、 そのまま高速炉用燃料の原料として利用できる。 In the present invention, the molten salt for reduction is preferably a calcium chloride molten salt. In the case of 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.
図面の簡単な説明  BRIEF DESCRIPTION OF THE FIGURES
図 1は本発明の軽水炉使用済燃料の再処理装置の一実施形態を示す概略図であ る。 図 2は軽水炉使用済燃料の再処理方法の一実施形態を示すフロー図である。  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.
発明を実施するための最良の形態  BEST MODE FOR CARRYING OUT THE INVENTION
以下、 本発明の構成を図面に示す実施の形態の一例に基づいて詳細に説明する。 図 1に本発明の軽水炉使用済燃料の再処理装置の一実施形態を示す。 この再処理 装置 9は、 大きく分けて軽水炉使用済燃料 1から酸化ウランを溶出させる酸化ゥ ラン溶出工程を実施する酸化ウラン溶出部 8と、 酸化ウラン溶出工程で酸化ゥラ ンの一部が除去された後の P u , U, F P , MA等が混じり合った酸化物から成 る陽極残留物を金属に電解還元する電解還元工程を実施する電解還元部 1 6とか ら構成されている。  Hereinafter, the configuration of the present invention will be described in detail based on an example of an embodiment shown in the drawings. 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.
酸化ウラン溶出部 8は、 再処理対象である軽水炉使用済燃料 1を保持する溶出 用陽極 2と、 溶出用陰極 3と、 溶出用陽極 2および溶出用陰極 3が浸される溶出 用溶融塩 4と、 該溶出用溶融塩 4を貯留する溶出用容器 5と、 溶出用陽極 2およ び溶出用陰極 3に電圧を印加して軽水炉使用済燃料 1から酸化ウラン 6を溶出用 溶融塩 4中に溶出させると共に溶出用陰極 3に析出させる溶出用直流電源 7とを 有する。  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. And an elution DC power supply 7 for elution at the same time and elution at the elution cathode 3.
この酸化ウラン溶出部 8において、 溶出用溶融塩 4は、 塩化リチウム一塩化力 リウム溶融塩としている。 このため、 酸化ウラン 6の溶出の操業温度を 5 0 0 °C 程度に抑えることができ、 溶出用容器 5や管などの腐食を抑制することができる。 また、 溶出用溶融塩 4には酸化ウラン供給源として塩化ゥラニル (U 0 2 C 1 2) を予め溶解しておく。 これにより、 操業初期から金属リチウムの析出を防止して 酸化ウランの回収速度を高めることができ、 溶出用陰極 3において効率的に酸化 ウラン 6を回収することができる。 In the uranium oxide elution section 8, the elution molten salt 4 is a lithium chloride monochloride molten salt. For this reason, 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. Further, the elution molten salt 4 is previously dissolved Uraniru chloride (U 0 2 C 1 2) as uranium oxide source. As a result, the precipitation of metallic lithium can be prevented from the early stage of operation, and the recovery rate of uranium oxide can be increased. As a result, uranium oxide 6 can be efficiently recovered at the elution cathode 3.
また、 この酸ィヒウラン溶出部 8では、 溶出用陽極 2に保持される軽水炉使用済 燃料 1と溶出用陰極 3との間に印加される電圧は 1 V程度であるようにしている。 但し、 この値は溶出用溶融塩 4の種類や温度、 さらには各電極 2, 3の形状ゃ大 きさや間隔によつて異なるものとなる。 軽水炉使用済燃料の大部分は酸化ウラン であるため、 極めて小さい電流で電解をする場合には、 電圧をほとんどかけなく ても電解が進行することになるが、 現実には電流値すなわち回収速度を高めるた めに 1 V程度の電圧を印加する必要があると推測できる。 In addition, in the acid / uranium elution section 8, 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. However, 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.
そして、 溶出用陽極 2と溶出用陰極 3とに与える通電量を調整することにより、 軽水炉使用済燃料 1から溶出する酸化ウラン 6の量を変更することができる。 こ れによって、 軽水炉使用済燃料 1に残存する酸化物中のプルトニウム含有量を相 対的に高めることができる。 例えば 1 0— 3 0 w t %程度の目的値まで高めるこ とができる。  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. Thereby, 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%.
また、 溶出用陽極 2では塩溶解性核分裂生成物であるアルカリ金属、 アルカリ 土類金属、 2価の希土類元素、 ハロゲンなどが溶出用溶融塩 4中に溶出する。 こ れらはそのまま塩中にとどまり、 残留物 1 0から除去される。  At the elution anode 2, 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.
溶出用容器 5は黒鉛やパイログラフアイトなどの炭素製とする。 そして、 溶出 用陽極 2は黒鉛やパイログラファイトなどの炭素製のバスケット 1 8を有してい る。 このバスケット 1 8の内部に軽水炉使用済燃料 1が収容される。 そして、 パ スケット 1 8ごと溶出用溶融塩 4に漬けられることにより、 軽水炉使用済燃料 1 が溶出用溶融塩 4に漬けられる。 溶出用陰極 3は黒鉛やパイログラファイトなど の炭素製としている。 この溶出用陰極 3に還元により酸化ウラン 6が析出するよ うになる。  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.
他方、 電解還元部 1 6は、 酸化ゥラン溶出部 8で発生した溶出用陽極 2の残留 物 (陽極残留物) 1 0を保持する還元用陰極 1 1と、 還元用陽極 1 2と、 還元用 陽極 1 2および還元用陰極 1 1が浸される還元用溶融塩 1 3と、 該還元用溶融塩 1 3を貯留する還元用容器 1 4と、 還元用陽極 1 2および還元用陰極 1 1に電圧 を印加して残留物 1 0を還元する還元用直流電源 1 5とを有する。  On the other hand, 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. The molten salt 13 for reduction in which the anode 12 and the cathode 11 for reduction are immersed, the reduction container 14 for storing the molten salt for reduction 13, the anode for reduction 12 and the cathode for reduction 11 1 And a reduction direct-current power supply 15 for reducing the residue 10 by applying a voltage.
この電解還元部 1 6では、 還元用溶融塩 1 3は塩化カルシウム溶融塩としてい る。 そして、 還元用溶融塩 1 3には予め酸素供給源が添加されている。 これによ り、 操業初期は酸素供給源によって供給された酸化物イオンを電荷の担体として 電解還元処理を行なうことができ、 操業中には残留物 1 0から放出される酸化物 イオンが電荷の担体として働く。 よって、 追加の酸素供給源を用いる必要はない。 このように酸化物イオンを担体として残留物 1 0の電解還元を行うことができる ので、 残留物 1 0のプルトニウム等の燃料成分が還元用溶融塩 1 3に溶け出すこ とがないために回収率を高めることができる。 また、 残留物 1 0中の非溶出核分 裂生成物からも酸素は除去され燃料成分と共に回収することができる。 さらに、 還元用陽極 1 2から発生する気体は酸素や二酸化炭素になるので、 腐食性のある 塩素ガスが発生することを避けることができる。 In the electrolytic reduction section 16, 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. As a result, in the early stage of the operation, 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. Thus, there is no need to use an additional oxygen source. Thus, 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.
そして、 操業初期から酸素供給源によって供給された酸化物イオンを電荷の担 体として電解還元処理を行なう場合、 塩素ガスを発生させることがなく処理速度 を高めることができる。 還元用陽極 1 2で塩素ガスを発生させないためには、 操 業処理速度に相当する速度で酸化物イオンが還元用陽極 1 2に供給されなければ ならず、 そのためには酸素供給源の添加量を増やすことや還元用溶融塩 1 3中を 効果的に攪拌することが有効であるが、 一方、 還元用溶融塩 1 3中の酸化物ィォ ン濃度の増加は、 還元用陰極 1 1での酸素の溶出を抑制する影響を及ぼすため、 過剰に添カ卩しない方が良い。  When the electrolytic reduction treatment is carried out using the oxide ions supplied from the oxygen supply source from the initial stage of operation as a carrier of electric charge, the treatment speed can be increased without generating chlorine gas. In order not to generate chlorine gas at the reducing anode 12, 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.
操業初期の酸素供給源としては酸化カルシウムを使用している。 還元用溶融塩 1 3は、 塩化カルシウムを溶媒として 0 . 0 1〜5 w t %の酸化カルシウムを溶 解したものとしている。 酸化カルシウムを加えると塩化カルシウムの融点は元々 の 7 7 5 °Cから、 最大で (C a Oが 3 . 4 w t °/0の時) 7 5 0 °Cまで低下する。 さらに、 電解還元部 1 6では、 還元用陰極 1 1に保持される残留物 1 0と還元 用陽極 1 2との間に印加される実際の電圧を 3 V程度としている。 この電圧設定 で電解還元することにより、 表 1に示すような残留物 1 0に含有される塩に非溶 解性の F P酸化物を、 ウラン、 プルトニウムと共に完全に金属に還元することが できる。 本実施形態の電位差範囲で還元されない一部の F P、 即ちアルカリ金属、 アルカリ土類金属、 2価の希土類元素などについては塩に溶解性であるため、 酸 化ウラン溶出工程ですでに溶出用溶融塩中に溶解して分離されている。 そして、 これら塩に溶解した核分裂生成物は、 還元されないため陰極には析出せずに、 塩 中にとどまる。 Calcium oxide is used as an oxygen source at the beginning of the operation. 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 ). Further, in the electrolytic reduction section 16, the actual voltage applied between the residue 10 held on the reduction cathode 11 and the reduction anode 12 is about 3 V. By performing the electrolytic reduction at this voltage setting, the FP oxide insoluble in the salt contained in the residue 10 as shown in Table 1 can be completely reduced to metal together with uranium and plutonium. 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.
アルカリ金属、 アルカリ土類金属、 2価の希土類元素が溶出用溶融塩である塩 化リチウム 塩化カリウム溶融塩に溶解して除去され、 還元用溶融塩である塩化 カルシウム中に持ち込まれないことは廃棄物処理の点から有利である。 即ち、 ァ ルカリ金属のセシウムとアル力リ土類金属のスト口ンチウムは発熱性の核分裂生 成物であり、 これらが溶融塩中に蓄積して行くとある時点でその発熱により溶融 塩が使用できなくなる。 ところが、 塩化リチウム 塩化カリウム溶融塩では、 ゼ ォライトを用いて溶融塩中からセシウムとストロンチウムを除去する技術が開発 されている (金属燃料リサイクルプラントの設計評価 (その 4 ) 使用済塩処理と 廃棄物処理システム、 日本原子力学会 2001年秋の大会予稿集第 III分冊、 P. 820) こと力ゝら、 溶融塩を長期間使用できる。 これに対し、 塩化カルシウム溶融塩では セシウムとストロンチウムを選択的に除去できないことから溶融塩を短期間しか 使用できず、 塩廃棄物量が増大してしまう。 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. However, for lithium chloride and potassium chloride molten salts, a technology has been developed to remove cesium and strontium from the molten salt using zeolite. (Design evaluation of metal fuel recycling plant (Part 4) Spent salt treatment and waste Disposal system, Atomic Energy Society of Japan, Fall 2001 2001 Annual Meeting, Volume III, p. 820) On the other hand, in the case of calcium chloride molten salt, cesium and strontium cannot be selectively removed, so that molten salt can be used only for a short period of time, and the amount of salt waste increases.
使用済核燃料に含有される代表的な酸化物の Typical oxides contained in spent nuclear fuel
8 0 0 °Cにおける理論分解電圧 (ボルト) Theoretical decomposition voltage at 800 ° C (volts)
U02 PU2O3 Ce2Os NcLOs Y2O3 Zr02 M0O2 Ru02 U0 2 PU2O3 Ce 2 Os NcLOs Y2O3 Zr0 2 M0O2 Ru0 2
2.33 2.39 2.55 2.60 2.75 2.32 1.02 0.33 還元用容器 1 4はステンレス製、 低炭素鋼製、 あるいはチタンなどの特殊鋼製 であることが好ましい。 そして、 還元用陰極 1 1は炭素製、 若しくはステンレス 製、 低炭素鋼製、 ジルコニウムあるいはチタンなどの特殊鋼製のパスケット 1 7 の形態をとつている。 このバスケット 1 7の内部に残留物 1 0が収容される。 そ して、 バスケット 1 7ごと還元用溶融塩 1 3に漬けられることにより、 残留物 1 0が還元用溶融塩 1 3に漬けられる。 還元用陽極 1 2は白金製であることが好ま しい。 2.33 2.39 2.55 2.60 2.75 2.32 1.02 0.33 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 passet 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.
上述した軽水炉使用済燃料 1の再処理装置 9により軽水炉使用済燃料 1を処理 して酸化ウラン 6を回収すると共にウラン—プルトニウム合金を得る手順を、 図 2に示すフローチャートに沿って説明する。  The procedure for processing the spent fuel 1 of the light water reactor and recovering the uranium oxide 6 and obtaining the uranium-plutonium alloy by the reprocessing device 9 for the spent fuel 1 of the light water reactor will be described with reference to the flowchart shown in FIG.
まず、 軽水炉使用済燃料 1を解体および剪断する (ステップ 1 : S 1 ) 。 次い で、 この解体およぴ剪断された軽水炉使用済燃料 1のジルカ口ィ被覆管を取り除 く (脱被覆加熱工程 (ステップ 2 ·· S 2 ) ) 。 具体的には、 解体した軽水炉使用 済燃料 1の燃料棒にスリットを入れ、 大気中で約 5 0 0 °Cに加熱する。 これによ り U O 2が U 38に酸化されて体積が膨張して被覆管が押し広げられる。 そして、 軽い振動を与えることにより酸化物が被覆管から分離されて回収される。 更に、 必要に応じて回収された酸化物を更に約 1 0 0 o °cにまで温度を上げて加熱処理 してやれば、 アルカリ金属、 カルコゲン、 ハロゲン、 一部の貴金属などのうちの 揮発性の核分裂生成物 (F P ) が除去される。 燃料被覆管のジルカロイ合金を酸 化ウラン溶出工程の前に除去することにより、 酸化ウラン溶出工程において量的 に酸化ゥランに匹敵するジルカ口ィ合金が溶出して溶出用陰極 3に析出すること により酸ィ匕ウラン 6の析出を妨げたり、 ジルカロイと混じり合った状態でしか酸 化ウランを回収できなくなることを防止できる。 また、 酸化ウラン溶出工程以前 に一部の F Pを除去しておくことは溶出用溶融塩 4にアル力リ金属などの塩溶解 性の F Pが蓄積していくことを抑制する効果があるが、 酸化ウランの溶出および 陰極での回収という点からは必要ないので、 上述の約 1 0 0 0 °Cでの加熱処理は 場合によっては実施しなくとも良い。 U 308は必要に応じて水素を含む気流中で 反応させれば、 U 02に還元される。 First, the spent fuel 1 of the light water reactor is dismantled and sheared (Step 1: S 1). Next, 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)). Specifically, using 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 38. Then, the oxide is separated from the cladding tube and recovered by giving a slight vibration. Further, if necessary, 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. By removing the zircaloy alloy from the fuel cladding tube before the uranium oxide elution step, the zirconia alloy comparable to the uranium oxide elution is quantitatively eluted in the uranium oxide elution step and deposited on the elution cathode 3. It is possible to prevent precipitation of uranium oxide 6 and prevent uranium oxide from being recovered only when mixed with zircaloy. Also, 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.
次に、 ジルカロイ被覆管を取り除いた軽水炉使用済燃料 1を入れた炭素製パス ケット 1 8を陽極 2として炭素製陰極 3との間で塩化ゥラニルを溶解した溶融塩 化物中で電解処理を行う (酸化ウラン溶出工程 (ステップ 3 : S 3 ) ) 。 この酸 化ウラン溶出工程では、 再処理対象である軽水炉使用済燃料 1を溶出用陽極 2に 保持して溶出用陰極 3と共に酸化ウラン供給源が添加された溶出用溶融塩 4に浸 し、 溶出用陽極 2および溶出用陰極 3に電圧を印加して軽水炉使用済燃料 1から 酸化ウラン 6を溶出用溶融塩 4中に溶出させる。 ここでは、 酸化ウラン溶出部 8 を利用して溶出用溶融塩 4を溶出用容器 5に入れて 5 0 0 °C程度に加熱する。 そ して、 脱被覆加熱処理工程で得られた軽水炉使用済燃料 1を溶出用陽極 2のパス ケット 1 8に収容して溶出用溶融塩 4に浸す。 また、 溶出用陰極 3も溶出用溶融 塩 4に浸す。 陽極 2と陰極 3に溶出用直流電源 7を接続して約 I Vの電圧を印加 することにより、 軽水炉使用済燃料 1から酸化ウラン 6を溶出用溶融塩 4中に溶 出させる。 このとき、 溶出用陽極 2と溶出用陰極 3とに与える通電量を調整する ことにより、 軽水炉使用済燃料 1から溶出する酸化ウラン 6の量を変更して軽水 炉使用済燃料 1に残存する酸化物中のプルトニウム含有量を例えば 10〜 30 w t%程度に高める。 Next, electrolytic treatment is performed in a molten chloride in which peranyl chloride is dissolved between the carbon cathode 18 and the carbon cathode 3 using the carbon packet 18 containing the spent fuel 1 from the light water reactor with the zircaloy cladding removed ( Uranium oxide elution step (Step 3: S3)). In this uranium oxide elution process, 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. Here, 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. Then, 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. At this time, 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%.
溶出用陽極 2においては下記化学式 1に示す反応によって酸化ウランが塩中に 溶出する。  At the elution anode 2, uranium oxide is eluted into the salt by a reaction represented by the following chemical formula 1.
U02 → U02 2++ 2 e- … (1) U0 2 → U0 2 2+ + 2 e-… (1)
この溶出用陽極 2ではプルトニゥムゃマイナーァクチ二ド元素は溶出しない。 これは表 2に示すように大量に存在する酸化ゥラン 6の酸化電位が酸化ネプツユ ゥムゃ酸化プルトニウムに比較してマイナス側であるためである。  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.
表 2  Table 2
L i C 1 -KC 1共晶塩系における標準電極電位 Standard electrode potential in L i C 1 -KC 1 eutectic salt system
電位 (C 12ノ C 1—電位基準) Potential (C 1 2 no C 1 — potential reference)
uo2 2Vuo2 - 0. 59V (450°C) uo 2 2 Vuo 2 - 0. 59V (450 ° C)
Pd2VPd — 0. 52V (450°C) Pd 2 VPd — 0.52V (450 ° C)
Rh /Rh — 0. 51V (450°C)  Rh / Rh — 0.51V (450 ° C)
Ru /Ru — 0. 42V (450。C)  Ru / Ru — 0.42V (450.C)
Np 02 +/Np 02 — 0. 22V (45 OX:) Np 0 2 + / Np 0 2 — 0.22V (45 OX :)
P u 02 2+/P uO: + 0. 28V ( 500 °C) 尚、 この酸化ウラン回収電解においては、 溶出用陽極 2側で核分裂生成物の一 部、 即ちアルカリ金属、 アルカリ土類金属、 2価の希土類元素、 ハロゲンなどの 塩溶解性核分裂生成物が溶出用溶融塩 4中に溶出するが、 これらはそのまま塩中 に止まり、 回収される酸化ウラン 6や残留物 10から除去される。 P u 0 2 2+ / P uO : + 0. 28V (500 ° C) Incidentally, in this uranium oxide recovery electrolysis, part of the fission products in the elution anode 2 side, i.e. the alkali metals, alkaline earth metals And salt-soluble fission products such as divalent rare earth elements and halogens are eluted in the molten salt for elution 4, but remain in the salt as they are and removed from the recovered uranium oxide 6 and residues 10 .
さらに、 溶出用陰極 3においては下記化学式 2に示す反応によってゥラュルイ オンが還元されて酸化ウラン 6が析出する。  Further, in the elution cathode 3, perulion is reduced by a reaction represented by the following chemical formula 2, and uranium oxide 6 is precipitated.
U02 2 + + 2 e- → UOz ·■· (2) U0 2 2 + + 2 e- → UOz
そして、 溶出用溶融塩 4には予め U02C 12が溶解されているので、 溶出用陰 極 3において効率的に酸化ウラン 6を回収することができる。 これにより、 高純 度の酸化ゥラン 6を効率的に回収することができる。 Then, 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.
ここで、 酸化ウラン 6の回収の後に塩素ガスを導入することにより、 軽水炉使 用済燃料 1中のプルトニウムを溶出用溶融塩 4中に溶解させて電解法あるいは沈 T/JP2003/013255 Here, by introducing chlorine gas after the recovery of uranium oxide 6, the plutonium in the spent fuel 1 of the light water reactor is dissolved in the molten salt 4 for elution and electrolysis or precipitation. T / JP2003 / 013255
11  11
殿法により核分裂生成物を十分に分離した酸化プルトニゥムを回収することが考 えられる。 しかしながら、 この回収のためには塩素ガスを多量に使用しなければ ならず、 その腐食性を考慮すると設備の安全性を確保するためにコスト高になつ てしまう。 また、 軽水炉使用済燃料 1中の核分裂生成物量は高速炉使用済燃料に 比べて少ないので、 プルトニウムから無理に分離しなくても高速炉燃料として用 いることが可能である。 このため、 本実施形態のように軽水炉使用済燃料 1中か らプルトニウムを分離.回収せずそのまま合金として製品化することがコスト面 から好ましい。 It is conceivable to recover plutonium oxide from which fission products have been sufficiently separated by the method. However, a large amount of chlorine gas must be used for this recovery, and considering its corrosiveness, the cost is high to ensure the safety of the equipment. Also, the amount of fission products in LWR spent fuel 1 is smaller than that in FBR spent fuel, so it can be used as fast reactor fuel without forcible separation from plutonium. For this reason, it is preferable in terms of cost to separate and recover plutonium from the spent fuel 1 of the light water reactor and to commercialize it as an alloy without recovering it as in the present embodiment.
次に、 酸化ウランが一部 (例えば半分程度) 除かれて、 プルトニウムなどの酸 化ウランを除くものの全ての含有量が相対的に高められた陽極残留物 1 0を電解 還元工程 (ステップ 4 : S 4 ) に移して還元処理を行う。 この電解還元工程では、 溶出用陽極 2の残留物 1 0を還元用陰極 1 1に保持して還元用陽極 1 2と共に酸 素供給源が添加された還元用溶融塩 1 3に浸し、 還元用陽極 1 2およぴ還元用陰 極 1 1に電圧を印加して残留物 1 0を還元するようにしている。 ここでは、 電解 還元部 1 6を利用して還元用溶融塩 1 3を還元用容器 1 4に入れて約 8 0 0 で 溶融させる。 そして、 酸化ウラン溶出工程での残留物 1 0を還元用陰極 1 1のパ スケットに収容して還元用溶融塩 1 3に浸す。 また、 還元用陽極 1 2も還元用溶 融塩 1 3に浸す。 陽極と陰極に還元用直流電源 1 5を接続して約 3 Vの電圧を印 加することにより、 残留物 1 0の電解還元を行う。  Next, the uranium oxide is partially removed (for example, about half), and the content of the anode residue 10 except for uranium oxide such as plutonium, which is relatively high, is reduced to an electrolytic reduction step (Step 4: Move to S 4) to perform reduction treatment. In this electrolytic reduction step, 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. Here, 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. Then, 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.
これにより、 酸化カルシウムにより供給される酸化物イオンおょぴ還元用陰極 力 ら溶出する酸化物イオンが還元用溶融塩 1 3中を還元用陽極に移行する。 また、 残留物 1 0では固溶する酸素がイオン伝導体として作用する。  As a result, 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.
還元用陰極 1 1においては下記化学式 3に示す反応によってァクチ二ド酸化物 が金属に還元されて合金の製品になる。  In the reducing cathode 11, the oxidized oxide is reduced to a metal by a reaction represented by the following chemical formula 3 to be an alloy product.
M〇2 + 4 e - → M+ 2 0 2一 … (3 ) M〇 2 + 4 e-→ M + 2 0 2 1… (3)
(ただし、 Mは U , N p, P uなど)  (However, M is U, Np, Pu, etc.)
ここで、 核分裂生成物の中で非溶出 F Pは酸化ゥランゃ酸化プルトニゥムと同 時に還元することができる。 よって、 最終製品である合金にはウランおょぴプル トニゥムの他に、 非溶出 F Pやマイナーァクチニドを含む。 還元された金属はバスケット 1 7の引き上げにより還元用溶融塩 1 3から取り 出され、 必要に応じて付着塩を除去することにより新たな金属製の核燃料の原料 になる。 本実施形態により得られた合金はプルトニウム含有量を 1 0— 3 O w t %に増加しているため付着塩を除去するだけで、 そのまま高速炉燃料の原料とす ることができる。 また、 ウラン、 プルトニウム、 マイナーァクチエドのロスがな い条件で核分裂生成物の約 4 0 %を占める溶出 F Pが除染されている。 Here, 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. In addition, eluted FP, which accounts for about 40% of fission products, is decontaminated without loss of uranium, plutonium, and minor arctied.
さらに、 還元用陽極 1 2においては下記化学式 4に示す反応によって酸素ガス が発生する。  Further, oxygen gas is generated at the reducing anode 12 by a reaction represented by the following chemical formula 4.
2 0 2 - → 0 2 + 4 e— … (4 ) 2 0 2- → 0 2 + 4 e—… (4)
なお、 上述の実施形態は本発明の好適な実施の一例ではあるがこれに限定され るものではなく本発明の要旨を逸脱しない範囲において種々変形実施可能である。 例えば本実施形態では、 酸化ウラン供給源や酸素供給源を予め各溶融塩中に添加 しているが、 塩素ガスを取り扱える設備を備えている再処理装置においては、 予 め酸化ゥラン供給源や酸素供給源を添加しておく必要はない。 軽水炉使用済燃料 から酸化ウランを溶かし出すこと及び陽極残留物から酸素を溶かし出すことにお いて処理速度的にも何らの問題はなく、 操業初期に溶出用陰極に塩素ガスを供給 すること及ぴ還元用陽極から発生する塩素ガスを回収することを各々行えばよい。 しかしながら、 酸化ウラン供給源や酸素供給源を予め添加した各溶融塩を使用す る方が、 塩素ガスを取り扱える設備を備え無くて済み、 再処理装置の設備コスト や保守 ·管理コストを削減できる。  The above embodiment is an example of a preferred embodiment of the present invention, but the present invention is not limited to this, and various modifications can be made without departing from the gist of the present invention. For example, in this embodiment, a uranium oxide supply source and an oxygen supply source are added to each molten salt in advance. However, in a reprocessing apparatus equipped with a facility capable of handling chlorine gas, 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. However, the use of 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.
また、 上述した実施形態では酸化ウラン溶出工程において純度の高い酸化ゥラ ンを回収することにしているが、 高純度で酸化ウランを回収した後に、 必要に応 じて陽極の電位をよりプラス側に設定することにより、 核分裂生成物のパラジゥ ム、 ロジウム、 ルテニウムなどを酸化ウランと共に回収して、 陽極に残留するプ ルトニゥムから取り除くことができる。 これは表 2に示すようにこれらの核分裂 生成物の酸化電位が酸化ウランに比較して少しだけプラス側にあるためである。 また、 上述した実施形態では溶出用溶融塩 4として塩化リチウム一塩化力リゥ ム溶融塩を使用しているが、 これには限られずより高温での操業が可能で気相中 の水分の影響を受けにくい塩化ナトリゥム一塩化力リゥム溶融塩やそれに塩化セ シゥムや塩化リチウムを加えた塩を使用することができる。 Further, in the above-described embodiment, high-purity uranium oxide is recovered in the uranium oxide elution step. However, after recovering uranium oxide with high purity, the potential of the anode is set to the positive side if necessary. With this setting, fission product palladium, rhodium, ruthenium and the like can be recovered together with uranium oxide and removed from the plutonium remaining on the anode. This is because the oxidation potential of these fission products is slightly positive compared to uranium oxide, as shown in Table 2. In the above-described embodiment, the molten salt for elution 4 is a molten salt of lithium chloride monochloride. However, 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.
さらに、 上述した実施形態では還元用溶融塩 1 3として塩化カルシウム溶融塩 を使用しているが、 これには限られず融点を降下させるために塩化カルシウムに 塩化パリゥムゃ塩化カリゥムなどを加えた塩、 あるいはアルカリ金属塩化物から 成る塩を使用することができる。  Further, in the above-described embodiment, a calcium chloride molten salt is used as the reducing molten salt 13. However, 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.
そして、 上述した実施形態では操業初期の酸素供給源として酸化カルシウムを 使用しているが、 これには限られず酸化リチウムや酸化バリウムなどの還元用溶 融塩 1 3に可溶の酸化物を溶融塩の種類に応じて使用することができる。  In the above-described embodiment, calcium oxide is used as an oxygen supply source in the early stage of the operation. However, 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.
また、 上述した実施形態では還元用陽極 1 2は白金製としているが、 これには 限られず例えば炭素製にしても良い。 この場合、 還元用陽極 1 2の表面で酸化物 ィオンが炭素と反応して二酸化炭素や一酸化炭素になる。 この二酸化炭素や一酸 化炭素は気泡になって気相中に排出される。 この場合、 還元用陽極 1 2は消耗す るので定期的に交換するようにする。  Further, in the above-described embodiment, the reducing anode 12 is made of platinum, but is not limited thereto, and may be made of, for example, carbon. In this case, 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.
さらに、 上述した実施形態では電解還元部 1 6で残留物 1 0と還元用陽極 1 2 との間に印加される実際の電圧を約 3 Vとしているが、 装置の設計条件などの要 因によって電位はこれには限られない。  Further, in the above-described embodiment, 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. However, depending on factors such as the design conditions of the apparatus, The potential is not limited to this.

Claims

請 求 の 範 囲 The scope of the claims
1 . 再処理対象である軽水炉使用済燃料を脱被覆する脱被覆工程と、 脱被覆後 の前記軽水炉使用済燃料を溶出用陽極に保持して溶出用陰極と共に酸化ウラン供 給源が添加された溶出用溶融塩に浸し、 前記溶出用陽極および前記溶出用陰極に 電圧を印加して前記軽水炉使用済燃料から酸化ウランを前記溶出用溶融塩中に溶 出させると共に前記溶出用陰極に析出させる酸化ウラン溶出工程と、 前記溶出用 陽極の残留物を還元用陰極に保持して還元用陽極と共に酸素供給源が添加された 還元用溶融塩に浸し、 前記還元用陽極および前記還元用陰極に電圧を印加して前 記残留物を還元する電解還元工程とを備えることを特徴とする軽水炉使用済燃料 の再処理方法。  1. The de-coating process of de-coating the light water reactor spent fuel to be reprocessed, and the elution with the uranium oxide supply source added together with the elution cathode while holding the light water reactor spent fuel after de-coating on the elution anode. Uranium oxide is immersed in a molten salt for use, and a voltage is applied to the elution anode and the elution cathode to cause uranium oxide to elute from the light water reactor spent fuel into the elution molten salt and precipitate on the elution cathode. An elution step, holding the residue of the elution anode on a reduction cathode, immersing the residue in a reduction molten salt to which an oxygen supply source is added together with the reduction anode, and applying a voltage to the reduction anode and the reduction cathode And an electrolytic reduction step for reducing the residue.
2 . 前記溶出用溶融塩は塩化リチウム一塩化カリゥム溶融塩であることを特徴 とする請求の範囲第 1項記載の軽水炉使用済燃料の再処理方法。  2. The method for reprocessing spent fuel of a light water reactor according to claim 1, wherein the molten salt for elution is a lithium chloride potassium monochloride molten salt.
3 . 前記還元用溶融塩は塩化カルシウム溶融塩であることを特徴とする請求の 範囲第 1項記載の軽水炉使用済燃料の再処理方法。  3. The method according to claim 1, wherein the molten salt for reduction is a molten salt of calcium chloride.
4 . 前記酸化ウラン供給源は、 塩化ゥラ -ルであることを特徴とする請求の範 囲第 1項記載の軽水炉使用済燃料の再処理方法。 4. The method for reprocessing spent fuel of a light water reactor according to claim 1, wherein the uranium oxide source is perchloral chloride.
5 . 前記酸素供給源は酸化カルシウムであることを特徴とする請求の範囲第 1 項記載の軽水炉使用済燃料の再処理方法。  5. The method according to claim 1, wherein the oxygen supply source is calcium oxide.
6 . 再処理対象である軽水炉使用済燃料を保持する溶出用陽極と、 溶出用陰極 と、 前記溶出用陽極および前記溶出用陰極が浸される酸化ウラン供給源が添加さ れた溶出用溶融塩と、 該溶出用溶融塩を貯留する溶出用容器と、 前記溶出用陽極 および前記溶出用陰極に電圧を印加して前記軽水炉使用済燃料から酸化ウランを 前記溶出用溶融塩中に溶出させると共に前記溶出用陰極に析出させる溶出用直流 電源とを有する酸化ゥラン溶出部と、 前記溶出用陽極の残留物を保持する還元用 陰極と、 還元用陽極と、 前記還元用陽極おょぴ前記還元用陰極が浸される酸素供 給源が添加された還元用溶融塩と、 該還元用溶融塩を貯留する還元用容器と、 前 記還元用陽極およぴ前記還元用陰極に電圧を印加して前記残留物を還元する還元 用直流電源とを有する電解還元部とを備えることを特徴とする軽水炉使用済燃料 の再処理装置。 6. An elution anode for holding the spent fuel of the light water reactor to be reprocessed, an elution cathode, and the elution molten salt to which the elution anode and the uranium oxide source into which the elution cathode is immersed are added. And an elution container for storing the elution molten salt; and applying a voltage to the elution anode and the elution cathode to elute uranium oxide from the light water reactor spent fuel into the elution molten salt. A lanthanum oxide elution section having a direct current power supply for elution deposited on an elution cathode, a reduction cathode for holding the residue of the elution anode, a reduction anode, and the reduction anode and the reduction cathode A reducing molten salt to which an oxygen supply source into which the molten salt is immersed, a reducing container for storing the reducing molten salt, and a voltage applied to the reducing anode and the reducing cathode to apply the residual DC power supply for reduction A reprocessing device for a spent fuel in a light water reactor, comprising: an electrolytic reduction section having:
7 . 前記溶出用溶融塩は塩化リチウム一塩化力リゥム溶融塩であることを特徴 とする請求の範囲第 6項記載の軽水炉使用済燃料の再処理装置。 7. The reprocessing device for a spent fuel in a light water reactor according to claim 6, wherein the molten salt for elution is a lithium chloride monochloride lime molten salt.
8 . 前記還元用溶融塩は塩化カルシウム溶融塩であることを特徴とする請求の 範囲第 6項記載の軽水炉使用済燃料の再処理装置。  8. The light water reactor spent fuel reprocessing device according to claim 6, wherein the molten salt for reduction is a calcium chloride molten salt.
9 . 前記酸化ウラン供給源は、 塩化ゥラエルであることを特徴とする請求の範 囲第 6項記載の軽水炉使用済燃料の再処理装置。  9. The light water reactor spent fuel reprocessing device according to claim 6, wherein the uranium oxide supply source is perchlorate.
1 0 . 前記酸素供給源は酸化カルシウムであることを特徴とする請求の範囲第 6 項記載の軽水炉使用済燃料の再処理装置。  10. The light water reactor spent fuel reprocessing device according to claim 6, wherein the oxygen supply source is calcium oxide.
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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 (en) * 2011-10-13 2013-05-13 Toshiba Corp Treatment method of corium
KR101711808B1 (en) * 2015-12-23 2017-03-06 한국원자력연구원 A preparation method for high removal ratio of cesium compounds from fuel fragments for the electrolytic reduction process
KR20200008308A (en) * 2018-07-16 2020-01-28 한국원자력연구원 Reduction apparatus and method of radioactive metal oxide using circulating reaction
KR102160973B1 (en) * 2018-07-16 2020-09-29 한국원자력연구원 Reduction apparatus and method of radioactive metal oxide using circulating reaction
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

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