WO2004036595A1

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

Info

Publication number: WO2004036595A1
Application number: PCT/JP2003/013255
Authority: WO
Inventors: Yoshiharu Sakamura; Masaki Kurata; Masatoshi Iizuka; Tadashi Inoue
Original assignee: Central Research Institute Of Electric Power Industry
Priority date: 2002-10-16
Filing date: 2003-10-16
Publication date: 2004-04-29
Also published as: JPWO2004036595A1; AU2003273024A1

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

Specification

Light water reactor spent fuel reprocessing method and equipment

Technical field

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.

Technical terms

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.

Background art

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.

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.

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.

Disclosure of the invention

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

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.

BRIEF DESCRIPTION OF THE FIGURES

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.

BEST MODE FOR CARRYING OUT THE INVENTION

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.

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.

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.

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.

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%.

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.

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.

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.

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.

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.

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.

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.

Typical oxides contained in spent nuclear fuel

Theoretical decomposition voltage at 800 ° C (volts)

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 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.

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.

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 ₃ 〇 _8. 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.

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%.

At the elution anode 2, uranium oxide is eluted into the salt by a reaction represented by the following chemical formula 1.

U0 ₂ → U0 ₂ ²⁺ + 2 e-… (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.

Table 2

Standard electrode potential in L i C 1 -KC 1 eutectic salt system

Potential (C 1 ₂ no C 1 — potential reference)

_{^{_{uo 2 2 Vuo 2 - 0. 59V}}} (450 ° C)

Pd ² VPd — 0.52V (450 ° C)

Rh / Rh — 0.51V (450 ° C)

Ru / Ru — 0.42V (450.C)

Np 0 ₂ ⁺ / Np 0 ₂ — 0.22V (45 OX :)

_{^{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 .

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.

U0 ₂ ^{2 +} + 2 e- → UOz

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.

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

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.

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.

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.

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 1… (3)

(However, M is U, Np, Pu, etc.)

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.

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)

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.

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.

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.

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.

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.

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

The scope of the claims

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. 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. The method according to claim 1, wherein the molten salt for reduction is a molten salt of calcium chloride.

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. The method according to claim 1, wherein the oxygen supply source is calcium oxide.

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. 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. 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. The light water reactor spent fuel reprocessing device according to claim 6, wherein the uranium oxide supply source is perchlorate.

10. The light water reactor spent fuel reprocessing device according to claim 6, wherein the oxygen supply source is calcium oxide.