JP5483867B2 - Method for recovering metallic fuel material from spent fuel and method for reprocessing spent fuel - Google Patents

Description

  The present invention relates to a method for recovering metallic fuel material from spent fuel and a method for reprocessing spent fuel, and more particularly to a method for recovering metallic fuel material from spent fuel suitable for obtaining metallic fuel, and spent fuel. The present invention relates to a fuel reprocessing method.

Plutonium (mainly plutonium 239) is generated by neutron absorption of uranium 238 in the core of a light water reactor using uranium as a nuclear fuel material. The spent fuel assembly loaded in the light water reactor is taken out of the light water reactor and reprocessed. There is a fast breeder reactor as a nuclear reactor that generates power by using plutonium extracted by reprocessing of the spent fuel assembly as a nuclear fuel material and producing more plutonium than the loaded plutonium. As this fast breeder reactor, there is a reactor of a type using a metal fuel containing uranium and plutonium as nuclear fuel materials.
The spent fuel assembly taken out from the fast breeder reactor loaded with the fuel assembly containing the metal fuel is reprocessed by the molten salt electrolysis method, and the recovered Pu is molded into the metal fuel for reuse. (For example, see Masafumi Koyama et al., “Dry-type reprocessing technology”, Denchu Ken review No. 37, pages 26-37, (2000)) However, most of the currently operated nuclear reactors are light water reactors using oxide nuclear fuel as nuclear fuel material. The spent fuel assembly removed from the light water reactor contains an oxide nuclear fuel containing about 1% Pu. For this reason, an example of producing metal fuel used for a fuel assembly initially loaded in a fast breeder reactor to which metal fuel is applied from oxide nuclear fuel is Takeshi Usami et al. “Application of dry reprocessing technology to oxide fuel. ”Denchu Research Review No. 37, pages 40-46, (2000). In this method for producing a metal fuel, uranium oxide in a spent fuel assembly of a light water reactor is reduced with lithium, and the produced metal uranium is recovered.

  As a method of reprocessing spent fuel, uranium and plutonium contained in spent nuclear fuel materials are converted to fluorides, and uranium and a mixture of uranium and plutonium are separated and recovered using the difference in volatility. It has been proposed (Japanese Patent Laid-Open No. 2000-284089). In this reprocessing method, most of uranium contained in the spent nuclear fuel material is volatilized and separated by fluorination using a first fluorinating agent. Thereafter, the remaining uranium and plutonium are recovered by volatilization using a second fluorinating agent, and the recovered uranium and plutonium are oxidized. Thereby, a mixed oxide of uranium and plutonium is generated.

Japanese Patent Application Laid-Open No. 2004-233066 also describes a method for reprocessing spent fuel by fluorination. This reprocessing method causes spent nuclear fuel material and fluorine gas to react and volatilize in the form of UF ₆ and PuF ₆ . These mixed gases are supplied to a Pu collection trap filled with granular UO ₂ F ₂ , and PuF ₆ is adsorbed to UO ₂ F ₂ in the form of PuF ₄ and separated. UF6 ₆ is converted to UO _{2, UO} 2 _{F 2} and PuF ₄ is converted to a mixed oxide of UO ₂ and PuO _2.

  Japanese Patent Application Laid-Open No. 2002-257980 describes a spent fuel reprocessing method for obtaining a mixed oxide nuclear fuel of uranium and plutonium from spent nuclear fuel material by the Purex method.

Another spent fuel reprocessing method using a molten salt electrolysis method is described in JP-A-2003-43187. Oxide nuclear fuel, which is a used nuclear fuel material, is put into a molten salt filled in a container, and chlorine gas is blown into the molten salt. UO ₂ and PuO ₂ contained in the oxide nuclear fuel are dissolved as chlorides in the molten salt. An anode and a cathode are attached in the molten salt. When energized between the anode and the cathode, UO ₂ and PuO ₂ are deposited on the cathode. A method for reprocessing spent fuel using the molten salt electrolysis method is also described in Japanese Patent Application Laid-Open No. 2000-284089.

JP 2000-284089 A JP 2004-233066 A JP 2002-257980 A JP 2003-43187 A Masafumi Koyama et al., “Dry reprocessing technology”, Denchu review No. 37, pages 26-37, (2000) Takemi Usami et al., "Application of dry reprocessing technology to oxide nuclear fuel" 37, pages 40-46, (2000)

  The inventors examined a method for reprocessing the oxide nuclear fuel contained in the spent fuel assembly taken out from the light water reactor in order to obtain a metal fuel used in the fuel assembly loaded in the fast breeder reactor. As a result, the inventors have come to the conclusion that the molten salt electrolysis method may be used to obtain the metal fuel from the spent oxide nuclear fuel.

  However, the amount of uranium contained in the metal fuel present in the new fuel assembly loaded in the fast breeder reactor core is 3 to 4 times the amount of plutonium contained in the metal fuel, but was removed from the light water reactor. The amount of uranium contained in the oxide nuclear fuel in the spent fuel assembly is about 100 times that of plutonium contained in the oxide nuclear fuel. For this reason, most of the excess uranium must be removed from the spent oxide nuclear fuel before the molten salt electrolysis process.

  The inventors described in Japanese Patent Application Laid-Open Nos. 2000-284089, 2004-233066, and 2002-257980 in order to remove excess uranium before the molten salt electrolysis step for obtaining the metal fuel. As shown in the figure, it has been thought that the spent oxide nuclear fuel is previously fluorinated. By fluorinating the spent oxide nuclear fuel in advance, most of the uranium can be converted into a fluoride and volatilized and removed. Further, as described in JP-A-2000-284089 and JP-A-2004-233066, the remaining uranium and plutonium fluorides are converted into respective oxides. The obtained oxides of uranium and plutonium are put into the molten salt in the container, and molten salt electrolysis is performed on these oxides. As a result, a metal fuel containing metal uranium and metal plutonium can be generated. Since excess uranium can be removed from the spent oxide nuclear fuel, the apparatus for performing molten salt electrolysis on uranium and plutonium oxides can be made compact. Furthermore, the time required to obtain a metal fuel by molten salt electrolysis can be shortened.

  The inventors have found a new problem that it is necessary to further simplify the process of recovering the metal fuel material as the raw material of the metal fuel from the spent oxide nuclear fuel.

  An object of the present invention is to provide a method for recovering metal fuel material from spent fuel and a method for reprocessing spent fuel that can simplify the process of recovering metal fuel material from spent oxide nuclear fuel. It is in.

A feature of the present invention that achieves the above-described object is that a spent oxide nuclear fuel taken out of a nuclear reactor is reacted with fluorine to produce a nuclear fuel fluoride,
Of this nuclear fuel fluoride, part of the uranium fluoride is removed,
The remaining nuclear fuel fluoride and oxide are dissolved in the fluoride molten salt,
As the oxide, an oxide having a cation common to the fluoride molten salt and different from the oxide nuclear fuel is used, and a first electrode and a cathode which are anodes immersed in the fluoride molten salt The purpose is to deposit a metal fuel substance dissolved in a fluoride molten salt on the second electrode by energizing a certain second electrode.

Nuclear fuel fluoride and oxide are dissolved in fluoride molten salt, and the anode first electrode and cathode second electrode immersed in the fluoride molten salt are energized and dissolved in the fluoride molten salt. Since the metal fuel substance is deposited on the second electrode, a conversion process for converting the produced nuclear fuel fluoride into an oxide is unnecessary as a pretreatment for fluoride molten salt electrolysis. For this reason, the process of the method for recovering the metal fuel material from the spent oxide nuclear fuel can be simplified. Since the oxide dissolved in the fluoride molten salt has an cation common to the fluoride molten salt and is different from the oxide nuclear fuel, generation of fluorine gas in the vicinity of the first electrode as the anode is prevented. The corrosion of the first electrode can be prevented, and the composition of the fluoride molten salt can be maintained.

  According to the present invention, the process of recovering the metal fuel material from the spent oxide nuclear fuel can be simplified.

  The inventors have newly found a step of a method for reprocessing spent oxide nuclear fuel, that is, fluorination treatment of spent oxide nuclear fuel, and the resulting oxides of uranium and plutonium fluorides. And the flow of treatment of molten salt electrolysis of uranium and plutonium oxides. As a result, the inventors further simplified the reprocessing process of spent oxide nuclear fuel by performing molten salt electrolysis of each uranium and plutonium fluoride obtained by fluorination of the oxide nuclear fuel. Reached the idea that In order to realize this, it has been newly found that it is necessary to add (a) a fluoride molten salt as a molten salt used in molten salt electrolysis and (b) an oxide to the fluoride molten salt. It was.

As the fluoride molten salt, for example, a mixture of LiF and BeF ₂ is used, and as the oxide to be added, for example, a mixture of Li ₂ O and BeO is used. As a result, it has become possible to carry out molten salt electrolysis of uranium and plutonium fluorides obtained by fluorination treatment of oxide nuclear fuel. Therefore, it is not necessary to convert uranium and plutonium into oxides of oxides, which is a pretreatment for molten electrolysis. In the method for recovering metallic fuel material from spent fuel, the spent fuel from spent oxide nuclear fuel to metallic fuel The process of collecting the substance can be simplified.

  Examples of the present invention reflecting the above examination results will be described below.

  A method for recovering a metal fuel material from spent fuel according to embodiment 1 which is a preferred embodiment of the present invention will be described with reference to FIG.

  The spent fuel assembly loaded in the core of the light water reactor is taken out from the reactor pressure vessel of the light water reactor and stored in the fuel storage pool for a predetermined period. This spent fuel assembly is removed from the fuel storage pool and transferred from the nuclear power plant where the light water reactor is installed to the nuclear fuel reprocessing facility. In the nuclear fuel reprocessing facility, the nuclear fuel material contained in the spent fuel assembly, that is, the oxide nuclear fuel is reprocessed.

  After the oxide nuclear fuel is taken out from the fuel rod by decoating and pulverized, the spent oxide nuclear fuel is subjected to fluorination treatment (fluorination treatment step 1). When 1 ton of nuclear fuel is loaded into the core of the light water reactor, it decreases to 0.994 ton as shown in Table 1 at the end of one operation cycle due to nuclear fission during the operation of the light water reactor. The spent oxide nuclear fuel contained in the spent fuel assembly taken out from the light water reactor has the composition shown in the spent oxide nuclear fuel column of Table 1, and includes uranium and uranium as main components. And transuranium elements such as plutonium, neptunium, americium, and curium produced from the above, and fission products (hereinafter referred to as FP). In the column of spent oxide nuclear fuel in Table 1, the weight, substance amount (number of moles) and radioactivity of each element contained in about 1 ton of spent oxide nuclear fuel are shown.

In the fluorination treatment step 1, the oxide nuclear fuel is reacted with fluorine and separated into gaseous uranium hexafluoride (UF ₆ ) and solid converted fluoride. The UF ₆ generated in the fluorination treatment step 1 is sent to the UF ₆ treatment step 7 for purification and converted to UO ₂ . This UO ₂ is loaded into a fuel rod of a new fuel assembly for a light water reactor.

In the fluorination treatment step 1, a flame furnace (or fluidized bed) is used. After the oxide nuclear fuel is supplied into the flame furnace, fluorine is supplied into the flame furnace, and the fluorination treatment of the oxide nuclear fuel is performed in the flame furnace. By controlling the flow rate of fluorine (F ₂ ) supplied into the flame furnace, the ratio of volatilization of uranium in the fluorination treatment step 1 can be adjusted between 80% and 98%.

By adjusting the rate at which uranium volatilizes, the amount of uranium remaining in the solid converted fluoride can be adjusted. Table 1 shows the composition of the converted fluoride obtained by subjecting the spent oxide nuclear fuel having the composition described in Table 1 to the fluorination treatment and volatilizing 98% of uranium. In the column of conversion fluoride. Each compound contained in the spent oxide nuclear fuel before the fluorination treatment is converted into each compound shown in the form column in the column of converted fluoride in Table 1 after the fluorination treatment. Most of the uranium is removed as UF ₆ during the fluorination treatment, so it is reduced to 2% of the amount of uranium contained in the spent oxide nuclear fuel. In addition, since H, Br, Kr, I, and Xe contained in the spent oxide nuclear fuel are gases, they are removed in the fluorination treatment step 1 and converted into the conversion fluoride that is a product of the fluorination treatment step 1. Not included in chemicals. Since the fluorides of Se, Nb, Tc, Mo, Ru, and Sb are volatile, only about 10% remains in the converted fluoride.

  Since the fluorination treatment step using the fluidized bed is a batch treatment, when a necessary amount of uranium has volatilized, the supply of fluorine into the fluidized bed is stopped and the residue is taken out from the fluidized bed. By performing such an operation, even when a fluidized bed is used, the ratio of volatilization of uranium can be adjusted.

After the fluorination treatment step 1 described above is completed, a fluoride molten salt electrolysis step 3 is performed. The converted fluoride produced in the fluorination treatment step 1 is supplied into the electrolytic cell used in the fluoride molten salt electrolysis step 3. The fluoride molten salt electrolysis apparatus used in the fluoride molten salt electrolysis step 3 has an electrolytic bath filled with fluoride molten salt, and the first electrode as the anode and the second electrode as the cathode are in the electrolytic bath. Is provided. The first electrode and the second electrode are immersed in the fluoride molten salt (see FIG. 5 described later). The converted fluoride is charged into the fluoride molten salt. As the fluoride molten salt, for example, a mixture of LiF and BeF ₂ is used. Instead of the mixture of LiF and BeF _2, a mixture containing LiF, NaF and KF as a fluoride molten salt may be used.

By using the fluoride molten salt, uranium existing in the form of UO ₂ F ₂ and neptunium existing in the form of NpO ₂ F ₂ contained in the converted fluoride are converted into the fluoride molten salt by the reaction shown below. Dissolve.

UO ₂ F ₂ (solid) → UO ₂ ²⁺ (in molten salt) + 2F ⁻ (in molten salt) (1)
NpO ₂ F ₂ (solid) → NpO ₂ ²⁺ (in molten salt) + 2F ⁻ (in molten salt) (2)
Since the other converted fluoride produced in the fluorination treatment step 1 exists in the form of _An F _m , for example, Pu, La, Sr and Cs are dissolved in the fluoride molten salt by the following reaction formula.

PuF ₄ (solid) → Pu ⁴⁺ (in molten salt) + 4F ⁻ (in molten salt) (3)
LaF ₃ (solid) → La ³⁺ (in molten salt) + 3F ⁻ (in molten salt) (4)
SrF ₂ (solid) → Sr ²⁺ (in molten salt) + 2F ⁻ (in molten salt) (5)
CsF (solid) → Cs ⁺ (in molten salt) + F ⁻ (in molten salt) (6)
When the first electrode and the second electrode are energized to electrolyze the fluoride molten salt in the electrolytic cell, the following reactions (7) to (10) occur in the fluoride molten salt, and at the cathode Uranium metal is deposited on a certain second electrode. That is, while current between the electrodes, the fluoride molten salt is caused reaction for the high solubility of unlike chloride molten salt O ^2- ions (8) is, uranium, U ³⁺ from U ⁴⁺ Via uranium metal.

UO ₂ ²⁺ (in molten salt) + 2e ⁻ (cathode) → UO ₂ (near the cathode) (7)
UO ₂ (near the cathode) → U ⁴⁺ (in molten salt) + 2O ^2- (in molten salt) (8)
U ⁴⁺ (in molten salt) + e ⁻ (cathode) → U ³⁺ (in molten salt) (9)
U ³⁺ (in molten salt) + 3e ⁻ (cathode) → U (deposited on the cathode) (10)
When formulas (7) to (10) are put together, the reaction occurring on the second electrode becomes the reaction of formula (11).

UO ₂ ²⁺ (in molten salt) + 6e ⁻ → U (deposited on cathode) + 2O ²⁻ (in molten salt) (11)
NpO ₂ F ₂ is deposited on the cathode as Np in the same manner as UO ₂ F ₂ . The fluoride present in the form of _An F _m is deposited on the cathode by the following reaction.

Pu ⁴⁺ (in molten salt) + 4e− (cathode) → Pu (deposited on the cathode) (12)
La ³⁺ (in molten salt) + 3e− (cathode) → La (deposited on the cathode) (13)
Sr ²⁺ (in molten salt) + 2e− (cathode) → Sr (deposited on the cathode) (14)
Cs ⁺ (in molten salt) + e − (cathode) → Cs (deposited on the cathode) (15)
Although it was stated that electrolysis of the fluoride molten salt was previously performed, when electrolysis is performed as it is, the reaction of the formula (16) occurs in the first electrode that is the anode with respect to the reaction of the formula (11), Fluorine gas is generated near the first electrode. For this reason, the first electrode is easily corroded.

6F ⁻ (molten salt) → 3F ₂ + 6e ⁻ (anode) (16)
In this embodiment, in order to prevent the generation of fluorine in the vicinity of the first electrode, the oxide is dissolved in advance in the fluoride molten salt in the electrolytic cell. As the oxide added to the fluoride molten salt, an oxide having a cation common to the fluoride molten salt is preferably used. This is because, by using an oxide common to the fluoride molten salt, for example, as shown in the following formula (19), the fluoride produced by reacting with excess fluorine in the fluoride molten salt is produced. It becomes the same composition as fluoride molten salt. For this reason, the composition of the fluoride molten salt can be maintained. In this embodiment, since a mixture of LiF and BeF ₂ is used as the fluoride molten salt, a mixture of Li ₂ O and BeO is used as the oxide to be added. By using a mixture of Li ₂ O and BeO, LiF and BeF ₂ that are the composition of the fluoride molten salt used in this example are generated based on the reaction of the formula (19). In this example, a carbon anode is used as the first electrode, and the reaction of the formula (11) is promoted. Therefore, the anodic reaction at the first electrode is expressed by the equations (17) and (18).
Li ₂ O + BeO → 2Li ⁺ (molten salt) + Be ²⁺ (molten salt) + 2O ²⁻ (molten salt) (17)
3O ²⁻ (molten salt) + 3 / 2C (anode material) → 3/2 CO ₂ + 6e ⁻ (anode) (18)
In the reaction of formula (18), carbon dioxide is generated from the first electrode. However, depending on the temperature of the fluoride molten salt, the generated carbon dioxide may be decomposed to produce carbon monoxide.

In the reaction of the formula (11) in which uranium metal is recovered from UO ₂ F ₂ by fluoride molten salt electrolysis, two O ²⁻ ions are generated, and thus the reaction of the formula (18) occurs at the first electrode. Therefore, Li ₂ O and BeO may be added to the mixture of LiF and BeF ₂ that is a fluoride molten salt in an amount that produces 1 mole of O ²⁻ ions per mole of UO ₂ F ₂ . The overall reaction is represented by the following formula:

1/2 (Li ₂ O + BeO) + UO ₂ F ₂ + 3 / 2C
→ Li ⁺ +1/2 Be ²⁺ + 2F ⁻ + U + 3 / 2CO ₂ (19)
Equation (11) shows what kind of substance is accompanied with the recovered uranium metal. Among the reactions represented by the equations (7) to (10), the reaction that requires the second electrode to be the most negative is the reaction of the equation (10). Since there is little data on the potential of the precipitation reaction with the fluoride molten salt, the following description will be made using the potential of the chloride molten salt.

The potential of the precipitation reaction differs between the fluoride molten salt and the chloride molten salt, but the order of the potential at which each ion precipitates (precipitation potential) does not change between the two. That is, the potential E1 at which U ³⁺ ions become uranium metal and the potential E2 at which La ³⁺ ions become La metal differ depending on the type of molten salt, but the potential E2 is more negative than the potential E1 even if the molten salt is different. That doesn't change.

Table 2 summarizes the compounds dissolved in the molten chloride salt in descending order of precipitation potential. Each element from line number 1 to line number 19 has a positive deposition potential than uranium, and is easier to precipitate than uranium. For this reason, when uranium is deposited on the second electrode, each element from row number 1 to row number 19 is also deposited on the second electrode. In Table 2, the elements after row number 21 are more difficult to deposit than uranium because the deposition potential is more negative than uranium, and when uranium is deposited on the second electrode, it is contained in the molten salt in the electrolytic cell. Se and Tc do not have deposition potential data, but conservatively, when uranium is deposited on the second electrode, Se and Tc are also deposited on the second electrode. Of course, plutonium is also deposited on the second electrode.

  In the fluoride molten salt electrolysis step 3 in the present embodiment, the potential difference between the first electrode and the second electrode provided in the electrolytic cell is the deposition potential in the fluoride molten salt among the elements necessary for the metal fuel to be produced. The first electrode and the second electrode are energized so as to be equal to the deposition potential of the lowest element (uranium). As a result, an element that is equal to or higher than the deposition potential of the element having the lowest deposition potential is deposited on the second electrode as the cathode.

  Since 98% of the uranium contained in the spent oxide nuclear fuel in the fluorination treatment step 1 was removed, the ratio of metal plutonium (enrichment) to metal uranium deposited on the second electrode and metal plutonium is as follows. 2 is 0.32. That is, according to this embodiment, after removing excess uranium from the spent oxide nuclear fuel, the remaining uranium and plutonium can be converted into a metal fuel having a plutonium enrichment of 0.25 or more. . The second electrode to which the recovered metal fuel material (metal elements from row number 1 to row number 20 shown in Table 2 such as metal uranium and metal plutonium) is attached is the metal fuel material of this embodiment. The plant that carries out the recovery method is transported to another plant that carries out the purification step 4 installed at a different location.

  Table 3 summarizes how much each of the metal elements from row number 1 to row number 20 shown in Table 2 are collected together when trying to collect Pu at a rate of 6.93 kg / month. Is. In Table 3, MA is a minor actinide, RE is a rare earth, AM is an alkali metal, AEM is an alkaline earth metal, and NM is a noble metal. In Table 2, MA is Am, Cm and Np, RE is Tb, Dy and Sm, and NM is each element from row number 1 to row number 11. AEM and NM are elements after the row number 21 shown in Table 2, and remain in the fluoride molten salt without being deposited on the second electrode.

Since Li ₂ O and BeO are added to the fluoride molten salt for the purpose of performing the reaction of the formula (14), the amount of the fluoride molten salt in the electrolytic cell increases. Therefore, after the fluoride molten salt electrolysis, the increased amount of fluoride molten salt (excess fluoride molten salt) is extracted from the electrolytic cell together with the FP and sent to the waste salt treatment step 6.

Table 4 shows the composition of P0798 simulated glass, which is a model glass of vitrified high-level waste, in terms of mole fraction. The glass base material includes an acidic oxide base material that forms a network structure such as SiO ₂ , and a basic oxide base material that cuts the network and lowers the melting point. The acidic oxide base material contains boron that decomposes fluoride.

Excess fluoride molten salt (LiF and BeF ₂ ) containing FP, Li ₂ O and BeO extracted from the electrolytic cell is supplied to the high-level radioactive waste treatment apparatus. In a high-level radioactive waste treatment apparatus, FP and excess fluoride molten salt are injected into a solidification container, and further, an acidic oxide base material and a basic oxide base material are injected into the solidification container. Stir. Thereafter, the solidification container is sealed, and the solidification container is stored in the storage chamber in a sealed state.

An example of the mass balance of FP, Li, Be, etc. in this example will be described below. Table 5 shows the number of electrons involved in the reaction of the converted fluoride shown in Table 2, the amount of O ²⁻ consumed at the first electrode (anode) at the time of fluoride molten salt electrolysis, and the spent oxide nuclear fuel of about 1 t. This is a summary of the amount of substances that accumulate in the fluoride molten salt each time the is treated.

A summary of the mass balance of FP, Li and Be is shown in FIG. Of this FP, 1.6 × 10 ⁻¹ kmol FP is introduced into the fluoride molten salt in the fluoride molten salt electrolysis step 3 from about 1 t of spent oxide nuclear fuel generated in the light water reactor. .12 × 10 ⁻² kmol is used for performing the purification step 4 of Example 2 (or Example 3) described later together with the metal fuel material (metal uranium, metal plutonium, etc.) recovered in the fluoride molten salt step 3 To the next plant. The remaining 9.84 × 10 ⁻² kmol FP is discharged from the fluoride molten salt electrolysis step 3 to the waste salt treatment step 6.

In the fluoride molten salt electrolysis step 3, 2.78 × 10 ⁻¹ kmol of O ²⁻ ions are consumed by recovery of the metal fuel material (metal uranium, metal plutonium, etc.). In order to supplement this O ²⁻ ion, Li ₂ O and BeO are added to the fluoride molten salt by 1.39 × 10 ⁻¹ kMol respectively. As a result of fluoride molten salt electrolysis, these oxides change to LiF and BeF ₂ which are the main components of fluoride molten salt, respectively. The amount of the molten fluoride salt is 2.78 × 10 ⁻¹ kmol by LiF and 1.39 × 10 ⁻¹ kmol by BeF ₂ , and the total amount is 4.17 × 10 ⁻¹ kmol. The increased amount of this fluoride molten salt (4.17 × 10 ⁻¹ kmol) is sent to the waste salt treatment step 6.

In order to vitrify the radioactive waste produced by the waste salt treatment step 6, according to Table 4, about 65% acidic glass matrix and about 30% basic glass with respect to 3.5% FP A parent drug is required. For an FP of 9.84 × 10 ⁻² kmol, an acidic glass matrix of 1.83 kmol and a basic glass matrix of 8.79 × 10 ⁻¹ kmol are required. Since Li and Be discharged from the electrolytic cell after completion of the fluoride molten salt electrolysis step 3 act as a basic glass matrix, the basic glass mother to be supplied into the electrolytic cell in the fluoride molten salt electrolysis step 3 The agent is 4.62 × 10 ⁻¹ kmol.

When the fluorinated conversion product generated in the fluorination treatment step 1 is converted into a metal fuel material using a new fluoride molten salt that does not contain FP, the concentration of FP in the fluoride molten salt is low. The amount of FP sent to the waste salt treatment step 6 is less than 9.84 × 10 ⁻² kmol. However, when the concentration of FP is increased by repeating the treatment of the fluoride molten salt electrolysis process 3, 9.84 × 10 ⁻² kmol of FP is regularly discharged from the electrolytic cell to the waste salt treatment process 6. Sent. The relationship between the number of treatments in the fluoride molten salt electrolysis step 3 and the amount of FP sent from the electrolytic cell to the waste salt treatment step 6 is shown in FIG. The fluoride molten salt contains ₂ kmol LiF and 1 kmol BeF2 as the initial composition. For this reason, in this embodiment, F ⁻ is 4 kmol. In the chloride molten salt electrolysis method, the concentration of uranium and plutonium in the chloride molten salt is about 1 to 5%. In the case where 1 ton spent oxide nuclear fuel is processed in the fluoride molten salt electrolysis step 3, uranium fluoride contained in the fluoride conversion product generated from the oxide nuclear fuel and dissolved in the fluoride molten salt; Since the amount of plutonium fluoride is 1.17 × 10 ⁻¹ kmol in total from the column of “Substance (kMol)” of the converted fluoride in Table 1, the amount of fluoride molten salt required is About 2 to 10 kmol. Therefore, the aforementioned 4 kmol in this embodiment is an appropriate amount.

  According to the present embodiment, the converted fluoride and oxide produced by removing most of the uranium in the fluorination treatment step 1 are supplied into the fluoride molten salt in the electrolytic cell, and fluoride molten salt electrolysis is performed. As a result, before the fluoride molten salt electrolysis, the conversion treatment to oxides of conversion fluorides such as uranium fluoride and plutonium fluoride generated in the fluorination treatment step 1 becomes unnecessary. For this reason, the process of the recovery method of the metal fuel substance from spent fuel can be simplified.

  In this embodiment, fluoride molten salt electrolysis is performed, so that spent oxide nuclear fuel generated in a light water reactor can be dissolved in fluoride molten salt, and metal uranium and metal plutonium contained in the oxide nuclear fuel can be dissolved. It can be easily recovered.

  Since the present embodiment has the fluorination treatment step 1 as the pretreatment of the fluoride molten salt electrolysis step 3, most of uranium contained in the spent oxide nuclear fuel can be removed in advance. it can. For this reason, the fluoride molten salt electrolysis apparatus used in the fluoride molten salt electrolysis step 3 can be made compact, and the amount of fluoride molten salt used can also be reduced.

  In this embodiment, since the oxide is added to the fluoride molten salt, generation of fluorine gas in the vicinity of the first electrode as the anode can be prevented, and carbon dioxide gas is generated in the vicinity of the first electrode. be able to. For this reason, corrosion of the first electrode can be prevented and the life of the first electrode can be extended. In particular, in this example, since an oxide having a cation common to the fluoride molten salt is used, for example, as shown in the equation (19), it reacts with excess fluorine in the fluoride molten salt. The resulting fluoride has the same composition as the fluoride molten salt. For this reason, the composition of the fluoride molten salt can be maintained. When an oxide having no cation common to the fluoride molten salt is used, the composition of the fluoride molten salt is changed.

  In this embodiment, since a carbon electrode is used as the first electrode that is the anode, carbon dioxide (or carbon monoxide) is generated from the first electrode that is the anode. For this reason, generation of fluorine gas from the first electrode can be prevented.

A spent fuel reprocessing method according to the second embodiment, which is another embodiment of the present invention, to which the method for recovering metallic fuel material from the spent fuel according to the first embodiment is applied will be described with reference to FIG. The spent fuel reprocessing method of the present embodiment is the same as that of the recovery method of the metallic fuel material from the spent fuel of the first embodiment, that is, the fluorination treatment step 1, the UF ₆ treatment step 2, and the fluoride melting. A refinement process 4 and a metal fuel production process 5 are added to the salt electrolysis process 3 and the waste salt treatment process 6. Since the fluorination treatment step 1, the UF ₆ treatment step 2, the fluoride molten salt electrolysis step 3 and the waste salt treatment step 6 in the present embodiment perform the same treatments as those steps in the first embodiment, description thereof is omitted. The refining process 4 and the metal fuel manufacturing process 5 will be described.

  In the purification step 4, for example, Masafumi Koyama et al., “Dry-type reprocessing technology”, Denchu Research Review No. 37, pages 26-37, (2000), the recovered metal fuel material is purified by dry reprocessing using chloride molten salt electrolysis. By performing dry reprocessing using this chloride molten salt electrolysis, a metal fuel having a composition in which all NM having a high radioactivity level is removed from the metal fuel material having the composition shown in Table 3 can be obtained. The obtained metal fuel has the composition shown in Table 6.

  Table 6 shows the composition of the metal fuel used for manufacturing the new core fuel assembly. The metal fuel having the composition shown in Table 6 is disclosed in Masafumi Koyama et al. 37, pages 26-37, (2000), page 37, and FIG. 4-5-1.

  Table 7 shows the composition of the metal fuel shown in Table 6 rewritten with the amount of Pu as 1. The new core fuel assembly having the metal fuel having the composition shown in Table 6 contains 3.82 times U of Pu. In the fluorination treatment step 1, U is removed so that the amount of U contained in the converted fluoride supplied to the fluoride molten salt electrolysis step 3 is 3.82 times or less of the amount of Pu.

  The metal fuel obtained in the refining step 4 (metal fuel having a composition in which NM is 0 in Table 3) is a nuclear fuel material filled in each fuel rod included in the new core fuel assembly used in the fast breeder reactor. Become. When the composition of the metal fuel obtained in the refining process 4 and the composition of the metal fuel required in the new core fuel assembly (Table 6) are compared, the metal fuel obtained in the refining process 4 has U and Zr of Table 6 Less than them. In the metal fuel obtained in the refining step 4, Pu, MA and RE satisfy the specifications shown in Table 6.

In the metal fuel production process 5, the fuel rod is produced using a metal fuel production apparatus. That is, a plurality of fuel pellets are manufactured by using the metal fuel obtained in the purification step 4 and the metal uranium and metal zirconium added so as to have the amounts of U and Zr shown in Table 6. These fuel pellets are filled into a cladding tube, and then the cladding tube is sealed to produce a fuel rod. As the metal uranium to be added, for example, metal uranium in which a part of UF ₆ generated in the fluorination treatment step 1 is used is used. It is also possible to use metal uranium produced based on depleted uranium generated by uranium enrichment as additional metal uranium.

A plurality of fuel rods are bundled with a plurality of fuel spacers and attached to the lower tie plate, a bundle of fuel rods is passed through the wrapper tube, and the lower end of the wrapper tube is attached to the lower tie plate. In this way, a new core fuel assembly loaded in the fast breeder reactor core is manufactured.
Also in this embodiment, each effect obtained in the first embodiment can be obtained. In the present embodiment, in which the method for recovering metallic fuel material from spent fuel according to the first embodiment is applied, the process of the spent fuel reprocessing method can be simplified.

  A spent fuel reprocessing method according to embodiment 3, which is another embodiment of the present invention, to which the method for recovering metallic fuel material from spent fuel according to embodiment 1 is applied will be described with reference to FIGS. To do. The spent fuel reprocessing method of this embodiment implements the steps shown in FIG. The fluoride molten salt electrolysis step 3 in this example will be described in detail with reference to FIG. 5 and the purification step 4 will be described in detail with reference to FIG.

In the fluoride molten salt electrolysis step 3 of the present embodiment, a fluoride molten salt electrolysis apparatus 10 shown in FIG. 5 is used. The fluoride molten salt electrolyzer 10 includes an electrolytic cell 11, a first electrode (electrode for electrolysis) 13 as an anode, and a second electrode (first recovery electrode) 14 as a cathode. For example, an annular weir 12 is installed on the bottom surface of the electrolytic cell 11. The second electrode 14 is disposed inside the weir 12 in the electrolytic cell 11. The first electrode 13 faces the second electrode 14 and is disposed above the weir 12 in the electrolytic cell 11. The first electrode 13 and the second electrode 14 are not in contact with the weir 12. For example, a carbon electrode is used as the first electrode 13, and an iron electrode is used as the second electrode 14. A fluoride molten salt 15, which is a mixture of LiF and BeF ₂ , is filled in the electrolytic cell 11, and the liquid level of the fluoride molten salt 15 is located above the upper end of the weir 12. The first electrode 13 and the second electrode 14 are immersed in the fluoride molten salt 15. The converted fluoride 17 obtained in the fluorination treatment step 1 is put into a region outside the weir 12 in the electrolytic cell 11. The converted fluoride 17 is dissolved in the fluoride molten salt 15.

A mixture of Li ₂ O and BeO, which are oxides, is added to the fluoride molten salt 15. As in Example 1, the potential of the second electrode 14 provided in the electrolytic cell is the element (uranium) having the lowest precipitation potential in the fluoride molten salt among the elements necessary for the metal fuel to be produced. A voltage is applied between the second electrodes 14 so as to be equal to the deposition potential. An element that is equal to or higher than the deposition potential of the element having the lowest deposition potential is deposited on the second electrode 14. The composition of the metal fuel material deposited on the second electrode 14 is as shown in Table 3. Carbon dioxide gas is generated from the first electrode 13.

  After the fluoride molten salt electrolysis step 3 is finished, the purification step 4 is performed to remove the impurity metal deposited on the second electrode 14 (Te of row number 1 to Cd of row number 11 shown in Table 2). The Before executing this purification step 4, the first electrode 13 used in the fluoride molten salt electrolysis step 3 is removed and a new first electrode (second recovery electrode) 16 is attached to the electrolytic cell 11 (see FIG. 5). . The first electrode 16 is also immersed in the fluoride molten salt 15.

  The metal fuel material deposited on the second electrode 14 in the fluoride molten salt electrolysis step 3 has the largest amount of uranium. In the purification step 4, in the state where no voltage is applied to the first electrode 16 and the second electrode 16, the potential of the second electrode 14 is close to the deposition potential of uranium. Therefore, if a slight potential difference is applied between the first electrode 16 and the second electrode 14 so that the first electrode 16 becomes negative, the potential of the second electrode 14 becomes slightly more positive than the deposition potential of uranium. For this reason, uranium deposited on the second electrode 14 starts to dissolve in the fluoride molten salt 15. Since the first electrode 16 is more negative than the second electrode 14, uranium dissolved in the fluoride molten salt 15 starts to deposit on the first electrode 16. As the potential difference applied to the first electrode 16 and the second electrode 14 is increased, the potential of the second electrode 14 gradually becomes positive and becomes higher than the deposition potential of Np. At this time, Np deposited on the second electrode 14 is dissolved in the fluoride molten salt 15, and the dissolved Np is deposited on the first electrode 16. Thus, by adjusting the potential difference applied between the first electrode 16 and the second electrode 14 so that the potential of the second electrode 14 is more positive than the deposition potential of Tb and more negative than the deposition potential of Cd. Each element having a deposition potential of Tb to U that has been deposited on the second electrode 14 is dissolved in the fluoride molten salt 15 and deposited on the first electrode 16. At this time, each element having a deposition potential from Te to Cd remains deposited on the second electrode 14. In the purification step 4 of the present embodiment, the first electrode 16 becomes a cathode and the second electrode 14 becomes an anode. By the above, the refinement | purification process 4 is complete | finished and the metal fuel which has the same composition as the time of completion | finish of the refinement | purification process 4 of Example 1 can be obtained. Each element remaining in the second electrode 14 in the purification step 4 may be recovered by removing the second electrode 14 from the electrolytic cell 11 and washing it if necessary.

  After the purification step 4 is completed, the first electrode 16 is removed from the electrolytic cell 11 and the first electrode 16 is transported to the metal fuel production apparatus in the metal fuel production step 5. A new core fuel assembly is manufactured by the metal fuel manufacturing apparatus using the metal fuel deposited on the first electrode 16 in the same manner as in the first embodiment.

  In the present embodiment, each effect produced in the second embodiment can be obtained. In this example, the first electrode 13 is replaced with the first electrode 16 while the fluoride molten salt 15 is filled in the electrolytic bath 11 of the fluoride molten salt electrolysis apparatus 10 used in the fluoride molten salt electrolysis step 3, By applying a potential difference using the first electrode 16 as a cathode and the second electrode 14 as an anode, the purification step 4 can be performed using the fluoride molten salt electrolysis apparatus 10. That is, in this embodiment, the fluoride molten salt electrolysis step 3 and the purification step 4 can be performed by the fluoride molten salt electrolysis device 10, so the fluoride molten salt electrolysis step 3 is performed by the fluoride molten salt electrolysis device, The process of the spent fuel reprocessing method can be further simplified as compared with the first embodiment in which the purification process 4 is performed by a chloride molten salt electrolysis apparatus. In this embodiment, it is not necessary to use a different molten salt electrolysis apparatus as in the first embodiment.

  The fluoride molten salt electrolysis apparatus 10 used in the fluoride molten salt electrolysis step 3 and the purification step 4 is an example, and the metal deposited on the second electrode 14 by the fluoride molten salt electrolysis is anodically dissolved into another electrode. Any device capable of being deposited may be used. For example, in the fluoride molten salt electrolysis apparatus 10, the first electrode 13 is replaced with the first electrode 16, but a fluoride molten salt electrolysis apparatus in which the first electrodes 13, 16 and the second electrode 14 are provided in the electrolytic cell 11 is used. It is also possible. In this case, the fluoride molten salt electrolysis process 3 adjusts the potential difference between the first electrode 13 and the second electrode 14 as described above, and the purification process 4 adjusts the potential difference between the first electrode 16 and the second electrode 14 as described above. May be adjusted.

  The principle of the third embodiment is summarized as follows. A fluoride molten salt in which spent oxide nuclear fuel is dissolved is a plurality of elements to be recovered (referred to as group A elements), a plurality of elements that are more easily dissolved in the fluoride molten salt than elements in group A (group B elements) Element) and a plurality of elements (referred to as elements of group C) that are more difficult to dissolve in the fluoride molten salt than the elements of group A. When a voltage is applied between these electrodes so that the first electrode becomes the anode and the second electrode becomes the cathode, and the group A element is deposited from the fluoride molten salt to the second electrode as the cathode, the group C These elements are also deposited on the second electrode, and the group B elements are not deposited on the second electrode. When a voltage is applied between these electrodes so that the second electrode becomes the anode and the other first electrode becomes the cathode, each element of group A adhering to the second electrode is dissolved in the fluoride molten salt. And then deposited on another first electrode. However, each element of Group C remains attached to the second electrode and does not dissolve in the fluoride molten salt. In other words, once the metal to be recovered from the metal deposited on one electrode is moved to another electrode by electrolysis, each element of the element group to be recovered contained in the spent oxide nuclear fuel is in a metal state. Can be separated.

It is explanatory drawing which shows the process of the recovery method of the metal fuel substance from the spent fuel of Example 1 which is one suitable Example of this invention. It is explanatory drawing which shows the material balance of the fission product in the Example shown in FIG. It is a characteristic view which shows the relationship between the frequency | count of a process of a fluoride molten salt electrolysis process, and the quantity of the fission product sent to a waste salt treatment process from an electrolytic cell. It is explanatory drawing which shows the process of the spent fuel reprocessing method of Example 2 which is another Example of this invention. It is explanatory drawing of the fluoride molten salt electrolysis process in the reprocessing method of the spent fuel of Example 3 which is another Example of this invention. It is explanatory drawing of the refinement | purification process in the reprocessing method of the spent fuel of Example 3. FIG.

Explanation of symbols

DESCRIPTION OF SYMBOLS 1 ... Fluorine treatment process, 2 ... UF ₆ treatment process, 3 ... Fluoride molten salt electrolysis process, 4 ... Purification process, 5 ... Metal fuel manufacturing process, 6 ... Waste salt treatment process, 10 ... Fluoride molten salt electrolysis apparatus 11 ... Electrolyzer, 13, 16 ... 1st electrode, 14 ... 2nd electrode, 15 ... Fluoride molten salt, 17 ... Conversion fluoride.

Claims (7)

Reacting spent oxide nuclear fuel removed from the reactor with fluorine to produce nuclear fuel fluoride,
Removing a part of the uranium fluoride from the nuclear fuel fluoride,
Dissolving the remaining nuclear fuel fluoride and oxide in a fluoride molten salt;
As the oxide, an oxide having a cation common to the fluoride molten salt and different from the oxide nuclear fuel is used, and a first electrode and a cathode which are anodes immersed in the fluoride molten salt A method for recovering a metal fuel material from spent fuel, comprising energizing a second electrode to deposit a metal fuel material dissolved in the fluoride molten salt on the second electrode. The method for recovering a metallic fuel material from spent fuel according to claim 1, wherein an electrode containing carbon is used as the first electrode. Reacting spent oxide nuclear fuel removed from the reactor with fluorine to produce nuclear fuel fluoride,
Removing a part of the uranium fluoride from the nuclear fuel fluoride,
Dissolving the remaining nuclear fuel fluoride and oxide in a fluoride molten salt;
As the oxide, an oxide having a cation common to the fluoride molten salt and different from the oxide nuclear fuel is used,
The first electrode as the anode and the second electrode as the cathode immersed in the fluoride molten salt are energized to deposit the metal fuel substance dissolved in the fluoride molten salt on the second electrode; and A method for reprocessing spent fuel, comprising purifying the metal fuel material deposited on the second electrode. The spent fuel reprocessing method according to claim 3 , wherein the metal fuel material is refined using a chloride molten salt electrolyzer having an electrolytic cell filled with chloride molten salt. Refining the metal fuel material,
In a state in which the metal fuel material deposited on the second electrode immersed in the fluoride molten salt, the third electrode are immersed before Symbol fluoride molten salt, and the second electrode is an anode the The second electrode and the third electrode are energized so that the three electrodes become cathodes, and a plurality of substances used for manufacturing the metal fuel among the substances contained in the metal fuel substance are deposited on the third electrode. The spent fuel reprocessing method of Claim 3 performed by this. The metal fuel material deposited on the second electrode is less soluble in the fluoride molten salt than the first metal material used in the production of the metal fuel and the first metal material. When an unnecessary second metal material is included and the fluoride molten salt includes a third material that is more easily dissolved in the fluoride molten salt than the first metal material,
In a state in which the metal fuel material deposited on the second electrode immersed in the fluoride molten salt, and a third electrode is immersed in the prior SL fluoride molten salt,
The first metal material deposited on the second electrode is dissolved in the fluoride molten salt and the second metal material deposited on the second electrode is not dissolved in the fluoride molten salt; Then, the second electrode and the third electrode are energized by applying a potential at which the first metal material dissolved in the fluoride molten salt is deposited on the third electrode and the third material is not deposited on the third electrode. The spent fuel reprocessing method according to claim 3 , wherein the metal fuel material is refined by being generated during the period. The spent fuel reprocessing method according to any one of claims 3 to 6 , wherein an electrode containing carbon is used as the first electrode.

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