JP4487031B2 - Method for dry reprocessing of spent oxide fuel - Google Patents

Description

  The present invention relates to a method for recovering uranium and plutonium as oxides from spent oxide fuel produced in a nuclear reactor. More specifically, the fuel pin fragment is converted into a molybdic acid molten salt or a tungstic acid molten salt as a solvent. The spent oxide fuel that was separated into the hull (cladding tube piece, etc.) and dissolved in the fuel at the same time without de-coating in advance by charging and dissolving the fuel component in an inert gas atmosphere The present invention relates to a dry reprocessing method.

  Since oxide fuel is currently used in commercial nuclear reactors, a method for reprocessing spent nuclear fuel that can treat oxide fuel and recover uranium and plutonium as a mixed oxide is required. ing. For this purpose, dry reprocessing methods using oxide electrolysis have been developed.

  In the conventional method, a salt having a composition of NaCl-2CsCl is used as a solvent for dissolving the spent oxide fuel. This salt is heated to about 650 ° C. and melted, and spent oxide fuel is supplied and dissolved therein, and a mixed oxide of uranium and plutonium is recovered by electrolysis. Here, when the spent oxide fuel is dissolved in the NaCl-2CsCl molten salt, chlorine gas is used.

For example, in Non-Patent Document 1, as a “salt cycle method”, “Oxide fuel is dissolved in a chloride mixed molten salt into which Cl ₂ —HCl gas is blown, and dissolved uranium and plutonium are 500 to 700 ° C. It is recovered as an oxide by cathodic electrodeposition ".

  However, since this dissolution reaction is a gas-solid reaction that occurs between gaseous chlorine gas and solid spent oxide fuel, the dissolution is slow and requires a long time for dissolution (dissolution takes 4 to 6 hours). ) There is a disadvantage that is not economical. In addition, since corrosive chlorine gas is used, there is a problem that the constituent materials of the processing apparatus are corroded.

  As a method for solving such a problem, the present inventors first used a molybdate or tungstate that has been heated and melted as a solvent, added a melting aid to the solvent, and used it while blowing an oxygen-containing gas. A method has been developed in which spent oxide fuel is dissolved and uranium / plutonium mixed oxide is recovered by electrolysis (see Japanese Patent Application No. 2005-200590). The flow is shown in FIG.

  Although this method can eliminate the drawbacks of the prior art as described above, there are still some problems to be solved. One of them is that a decoating step for removing the spent oxide fuel from the cladding tube in advance is necessary prior to the dissolution of the fuel. Fuel pins used in commercial reactors have a structure in which a cladding tube made of stainless steel, Zircaloy, or the like is filled with oxide fuel, so that the fuel pins are mechanically crushed (for example, sheared) prior to fuel dissolution. And the oxide fuel needs to be separated (decoated) from the cladding tube piece. However, with the current decoating technology, many fuel components remain and adhere to the cladding tube, so it is necessary to add a hull cleaning process, etc. to obtain a high fuel recovery rate, which complicates the process and increases costs. Invited.

In addition, this method has a problem that the dissolution rate is relatively slow because it involves an oxidation reaction with a slow reaction when the fuel is dissolved. Furthermore, since Cs, which is a high heating element included in the spent oxide fuel and is a factor of high level waste, dissolves in the solvent (molten salt), there is a problem that a large amount of solvent becomes high level waste.
"Chemical Engineering of Fuel Reprocessing and Radioactive Waste Management" Nuclear Chemical Engineering Volume 4 Manson Benedict et al., Translated by Kiyose Kyohei, December 1983, published by Nikkan Kogyo Shimbun, p.16

  The problem to be solved by the present invention is to eliminate the use of highly corrosive harmful gas in the dissolution reaction of spent oxide fuel, and it is not necessary to perform decoating in advance prior to the dissolution of the fuel. In addition, the dissolution rate is fast, and the amount of high-level waste generated can be greatly reduced.

  The present invention relates to a dry reprocessing method of spent oxide fuel in which spent oxide fuel is dissolved in a solvent and uranium / plutonium mixed oxide is precipitated and recovered by electrolysis. A crushing process for mechanically crushing used fuel pins, and a fuel pin fragment into a molybdic acid molten salt or tungstic acid molten salt used as a solvent to dissolve fuel components in an inert gas atmosphere By performing a hull separation / fuel dissolution process for performing hull separation simultaneously with fuel dissolution, an oxidation process for oxidizing the fuel component dissolved in the solvent, and an electrolytic process for the solvent after the oxidation process. An electrolysis process for depositing and collecting plutonium mixed oxide on a cathode, and a dry reprocessing method for spent oxide fuel.

  According to the dry reprocessing method of spent oxide fuel according to the present invention, it is not necessary to perform a decoating process in advance in a separate process prior to fuel dissolution, and fuel component dissolution and hull separation can be processed simultaneously. The problem of low fuel recovery rate in the decoating process can be solved, and the process is simplified. In addition, since dissolution in molten salt does not involve an oxidation reaction that is slow in reaction, the dissolution rate is increased, and the dissolution is completed in an extremely short time, thereby greatly shortening the operation time of the process. Further, since the method of the present invention is dissolved in an inert gas atmosphere, the noble metal FP and the high heat generating element such as Cs that is a cause of high-level waste are not dissolved in the molybdate or tungstic acid molten salt of the solvent. Thus, separating them with the hull can avoid a large amount of solvent becoming a high level waste. In addition, since a highly corrosive gas is not required in the dissolution reaction, it goes without saying that corrosion of the processing apparatus constituent material does not occur and the life of the processing apparatus can be greatly extended.

  FIG. 1 shows a flowchart of a method for dry reprocessing of spent oxide fuel according to the present invention. A fuel pin used in a commercial nuclear reactor has a structure in which an oxide fuel is filled inside a cladding tube. A cladding tube consists of metal materials, such as stainless steel and Zircaloy, for example. First, the spent fuel pin is mechanically crushed as in the prior art. Specifically, it is preferable to perform a shearing process (shearing process) using a biaxial shearing machine or the like.

  In the conventional dry reprocessing method, the fuel pin fragment is previously decoated before proceeding to the next dissolution step, but in the method of the present invention, the process proceeds directly to the dissolution step. That is, the fuel pin fragment is directly put into a molybdate or tungstic acid molten salt used as a solvent, and dissolved while passing an inert gas through the solvent, thereby performing hull separation simultaneously with fuel dissolution (hull separation / Fuel dissolution process). In this step, only the fuel component is dissolved and the hull (cladding tube piece and the like) remains as it is, and the noble metal FP (fission product) and the high-temperature heating element Cs and the like are not dissolved. Therefore, these hulls and dissolved residues are removed.

Among spent oxide fuels, there are uranium, alkali metal elements, noble metal elements, rare earth elements, and transuranium elements such as plutonium produced in a nuclear reactor. In the method of the present invention, as a solvent for dissolving the spent oxide fuel, for example, sodium molybdate (Na ₂ MoO ₄ , Na ₂ Mo ₂ O _7, etc.) or sodium tungstate (Na ₂ WO ₄ , Na ₂ W ₂ O _7). Etc.). The melting point of these compounds is about 700 ° C., but it is desirable to heat to 750 ° C. or higher in order to promote the dissolution reaction. The reaction formula in which uranium dioxide and plutonium dioxide contained in the spent oxide fuel are dissolved in molten sodium molybdate under an inert gas atmosphere is as follows.
UO ₂ + 2Na ₂ Mo ₂ O ₇ → Na ₄ U (MoO ₄ ) ₄
(Na ₄ U (MoO ₄ ) ₄ is the same expression as U (MoO ₄ ) ₂ + 2Na ₂ MoO ₄ )
PuO ₂ + 2Na ₂ Mo ₂ O ₇ → Na ₄ Pu (MoO ₄ ) ₄
(Na ₄ Pu (MoO ₄ ) ₄ is the same as Pu (MoO ₄ ) ₂ + 2Na ₂ MoO ₄ )
Uranium and plutonium dissolve in the molten salt without changing the valence as tetravalent, so there is no need for oxidation aids, etc., and there is no slow oxidation reaction, so the dissolution rate is greatly increased. improves. At this time, the solvent dissolves only the fuel component without dissolving the structural material such as the cladding tube, so that hull separation and fuel dissolution can be achieved simultaneously. Further, since noble metal elements and high heating elements Cs contained in the spent fuel are not dissolved, they are separated from the fuel together with the hull. Then, a uniform melt in which transuranium elements such as uranium and plutonium and some rare earth elements are dissolved in the molten salt can be obtained.

Subsequently, the fuel dissolved in the solvent is oxidized (oxidation process). Oxygen gas or air is blown into the solvent in which the fuel is dissolved to oxidize the fuel components. This changes the valence of uranium and plutonium from tetravalent to hexavalent. At this time, since uranium and plutonium are already dissolved in the solvent, the oxidation reaction proceeds promptly. Then, the uranium / plutonium mixed oxide is deposited and recovered on the cathode by subjecting the solvent after the oxidation treatment to electrolytic treatment (electrolysis step). The uranium / plutonium (TRU) mixed oxide can be selectively deposited on the cathode and recovered by immersing the anode and the cathode in a solvent and performing electrolysis near the redox potential of UO ₂ . This is a reprocessed product.

(Hull separation and fuel dissolution)
A hull separation and fuel dissolution test was conducted using the test apparatus shown in FIG. The test apparatus has a structure in which a sealable test vessel 12 is installed in a furnace 10 and an alumina crucible 14 is provided therein. A solvent (molten sodium molybdate) 16 is placed in a crucible 14, and uranium dioxide and a cladding tube material piece (material name: PNC1520 (a cladding tube material used in the fast reactor "Monju" The component was the same stainless steel as SUS316))) and the dissolution treatment was performed. An inert gas was blown from the gas blowing pipe 20 and the off gas was discharged from the off gas discharge pipe 22 to the off gas processing system. The solvent used was a sodium molybdate molten salt represented by Na ₂ Mo ₂ O ₇ and was heated to 750 ° C.

As a result, it was confirmed that the clad tube material piece did not melt, and that only the entire amount of uranium oxide was dissolved in molybdate in about 5 minutes. The dissolution reaction at this time is as follows.
UO ₂ + 2Na ₂ Mo ₂ O ₇ → U (MoO ₄ ) ₂ + 2Na ₂ MoO ₄
The dissolution rate was 5 mg / cm ² / h or more. Incidentally, the dissolution rate of the UO ₂ of 60 ° C. nitrate 6~7N assuming the wet reprocessing is 5mg / cm ² / h, dissolution rate have about 10 times when compared with the experimental results have been obtained.

(Oxidation)
After the clad tube material piece was pulled up and removed, an oxidation test was conducted. For this purpose, the test apparatus shown in FIG. 2A was used as it was. Oxygen gas was blown from the gas blow pipe 20 and the off gas was discharged from the off gas discharge pipe 22 to the off gas treatment system. According to the XPS (X-ray photoelectron spectroscopy) analysis result of the molten salt after the oxidation treatment, it was confirmed that the peak due to the tetravalent U was decreased and the peak due to the hexavalent U was increased. . The oxidation reaction at this time is as follows.
U (MoO ₄ ) ₂ + O ₂ + 2NaMoO ₄ → UO ₂ MoO ₄ + 2Na ₂ Mo ₂ O ₇

(electrolytic)
After the clad tube material piece was pulled up and removed, an electrolytic test was conducted using the test apparatus shown in FIG. The test apparatus is basically the same as that used in the hull separation / fuel dissolution test, and has a structure in which a sealable test vessel 12 is installed in a furnace 10 and a silica crucible 24 is provided therein. It is. In the crucible 24, a solvent (sodium molybdate molten salt) 26 dissolved as hexavalent uranium is contained. The anode 30 and the cathode 32 were inserted into the molten salt, and electrolytic treatment was performed. If necessary, oxygen gas was appropriately blown from the gas blowing pipe 20 and the off gas was discharged from the off gas discharge pipe 22 to the off gas processing system. A platinum wire was used for the anode 30 and a platinum plate was used for the cathode 32. A platinum wire was inserted as the reference electrode 34. Electrolysis was performed at a molten salt heating temperature of 700 ° C. and a current density of 50 mA / cm ² . When the cathode was pulled up after the electrolytic treatment, electrolytic deposits were observed. As a result of XRD (X-ray diffraction method) analysis of this electrolytic deposit, it was confirmed that UO2 can be recovered on the cathode by electrolytic treatment.

The flowchart of the dry reprocessing method of the spent oxide fuel which concerns on this invention. Explanatory drawing which shows the example of a test apparatus. The flowchart which shows an example of a prior art.

Explanation of symbols

DESCRIPTION OF SYMBOLS 10 Furnace 12 Sealed container 14,24 Crucible 16,26 Solvent 18 Sample 20 Gas blowing pipe 22 Gas exhaust pipe 30 Anode 32 Cathode

Claims (1)

In the dry reprocessing method of spent oxide fuel in which spent oxide fuel is dissolved in a solvent and uranium / plutonium mixed oxide is precipitated and recovered by electrolysis,
A crushing process that mechanically crushes spent fuel pins filled with oxide fuel in a metal cladding tube, and fuel pin crush pieces are put into molybdate or tungstic acid molten salt used as a solvent for inertness A hull separation / fuel dissolution step for performing hull separation simultaneously with fuel dissolution by dissolving the fuel component in a gas atmosphere, an oxidation step for oxidizing the fuel component dissolved in the solvent, and a solvent after the oxidation treatment And an electrolytic process for depositing and recovering a uranium / plutonium mixed oxide on the cathode by subjecting to electrolysis to a dry reprocessing method of spent oxide fuel.

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