JP2002131472A - Reprocessing method for spent nuclear fuel oxide - Google Patents

Abstract

PROBLEM TO BE SOLVED: To provide a reprocessing method wherein the life of an apparatus can be made long, uranium and plutonium are recovered efficiently and generated crucible wastes can be recycled as a radioactive waste processing material. SOLUTION: The reprocessing method for a spent nuclear fuel oxide contains a chlorination dissolving process wherein the spent nuclear fuel oxide containing the uranium and the plutonium is supplied to a chloride fused salt as a reaction medium stored in a quartz crucible and a first mixed gas composed of chlorine, carbon monoxide and an inert gas is blown into the fused salt so as to dissolve the whole amount of the uranium and the plutonium, a PuO2 sedimentation and recovery process wherein a second mixed gas composed of oxygen, chlorine and an inert gas is blown into the fused salt and the plutonium dissolved in the fused salt is precipitated as a plutonium oxide and a UO2 electrolytic recovery process wherein a third mixed gas composed of oxygen, chlorine and an inert gas is blown into the fused salt, a current is made to flow across a cathode and an anode installed inside the fused salt and the uranium dissolved in the fused salt is precipitated as a urianium oxide.

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

The present invention relates to a method for recovering spent nuclear fuel oxide in a dry process. More specifically,
The present invention relates to a method for recovering spent nuclear fuel oxide, which recovers useful metals such as uranium and plutonium as oxides from spent nuclear fuel oxide.

[0002]

2. Description of the Related Art In the reprocessing of spent nuclear fuel discharged from a nuclear power plant, in order to separate and recover unburned fissile material and newly purified fissile material, a wet process called the Purex method is used. The law is taking place. In this method, uranium and plutonium are separated and recovered by using solvent extraction due to differences in solubility in an organic solvent. However, since the organic solvent used is expensive and deteriorates due to strong radiation, the spent nuclear fuel must be treated after being left for a long period of time.

On the other hand, in recent years, a reprocessing technique called a dry method utilizing the principle of electrolysis has been used in the Research Institute of Atomic Reactors (hereinafter referred to as RIAR) and the Argonne National Laboratory (Argonne) in the United States. National Laboratory (hereinafter referred to as ANL). In this dry method, spent nuclear fuel is converted into chloride, and the chloride is electrochemically purified in a molten salt bath in which the chloride is melted. The reprocessing by the dry method has attracted attention from the viewpoint of economical efficiency because it does not use water as compared with the conventional wet method, so that the equipment can be made compact and no waste liquid is generated. FIG. 2 shows a process chart of the molten salt electrolysis method (hereinafter, referred to as RIAR method) performed by RIAR.
In the RIAR method, a pyrographite crucible is first prepared in a reaction vessel, and a molten salt is stored in the crucible. This crucible also functions as an anode in a UO ₂ electrolytic recovery step described later. Supplying spent nuclear fuel oxide into the stored molten salt,
Subsequently, chlorine fuel is chlorinated by blowing chlorine gas into the molten salt to dissolve uranium as an oxychloride compound and plutonium as chloride in the molten salt (chlorination dissolution step). Next, electrolysis is performed to precipitate and recover uranium in the molten salt as uranium oxide on the electrode (UO ₂ electrolytic recovery step). Next, a mixed gas of chlorine and oxygen is blown into the molten salt from which uranium has been recovered to precipitate and recover plutonium oxide (PuO ₂ precipitation recovery step). In this method, uranium and plutonium oxides purified directly from spent nuclear fuel oxides can be obtained.

Specifically, a molten chloride salt as a reaction medium is stored in a crucible made of pyrographite, a spent nuclear fuel oxide is charged into the molten salt, and chlorine gas is blown into the molten salt. Uranium oxide is converted into an oxychloride compound and chloride, and plutonium oxide is converted into chloride, thereby dissolving it in a molten salt. The main reaction formulas are shown in the following formulas (1) to (4).

UO ₂ (s) + Cl ₂ (g) → UO ₂ Cl ₂ (l) (1) UO ₂ (s) + C (s) + 2Cl ₂ (g) → UCl ₄ (l) + CO ₂ (g) ↑ …… (2) PuO ₂ (s) + 2Cl ₂ (g) → PuCl ₄ (l) + O ₂ (g) ↑ …… (3) PuO ₂ (s) + C (s) + 2Cl ₂ (g) → PuCl ₄ (l) + CO ₂ (g) ↑ (4) where (s) indicates a solid, (l) indicates a liquid, and (g) indicates a gas.

[0006] However, the reaction of the above formula (3) is thermal
Reaction is difficult to proceed
Of uranium and plutonium in wastes
In order to achieve this, a reaction according to the above formula (4), ie, a carbon source is required
Becomes In the case of the RIAR method, the reaction of the above formula (4) proceeds
Carbon required to cause the corrosion of the crucible material
In. Also, by the reaction of the above formula (2), the recovery odor
Unpleasant UCl _FourFormation and crucible corrosion progress
You. Therefore, pyrolysis is caused by the dissolution reaction of nuclear fuel oxide.
Corrosion of the fight crucible limits the life of the equipment.
Limited. The reaction of the above formulas (2) and (4)
With carbon eluted from the crucible surface or carbon on the crucible surface
Solid UO_TwoOr PuO_TwoIs chlorine gas Cl_TwoIn the presence of
A solid-solid-gas reaction in which a reaction occurs only upon contact
Therefore, the reaction rate is very slow. Also plutonium
In order to recover by the precipitation method, the back
Fission products other than plutonium
ts, hereinafter referred to as FPs. ) Also converts all oxides to chloride
Since it is necessary to change the equation, the equations (2) and (4)
Like the reaction, a carbon source from the crucible is required. So
Therefore, the life of the crucible is further reduced. Also follow
Uranium forms UO for recovery of uranium_TwoCl_TwoForm of
It is preferable to use the above-mentioned formulas (1) to (4).
UO in reaction_TwoCl_TwoIt is extremely important to keep the form in salt
Difficult. Therefore, this form of control cannot be performed well.
And the following UO_TwoElectrolytic recovery process and PuO_TwoPrecipitation collection
There was also a problem that the process could not be performed well. Furthermore,
It is consumed by the reaction of the above formulas (2) and (4)
Pots are ultimately generated as carbon-based waste,
It was necessary to examine waste treatment and disposal measures.

On the other hand, in Japan, a method of simultaneously performing dissolution and electrolytic recovery of uranium without using chlorine gas has been proposed (JP-A-6-324189). In this method, the molten salt phase is concentrically partitioned by an electrically insulating annular partition wall and the metal fuel component concentration in the molten metal phase and the metal fuel component concentration in the molten salt phase are controlled to predetermined concentrations,
The anode dissolution of the spent metal fuel and the electrolytic deposition on the solid cathode surface are simultaneously performed by one voltage applying means.

[0008]

However, Japanese Patent Laid-Open No. 6-3 / 1994
In the method disclosed in Japanese Patent No. 24189, the spent nuclear fuel oxide is not chlorinated because chlorine gas is not supplied, and is expected to exist as a precipitate. For this reason, although uranium can be recovered, undissolved oxides are expected to accompany precipitation in the subsequent plutonium, and only extremely low-purity plutonium can be recovered. In addition, the above RIA
Since a crucible made of pyrographite is used for the reaction vessel as in the case of the R method, there is also a problem that the crucible is corroded by the dissolution reaction of the nuclear fuel oxide, and the life of the apparatus is limited.

It is an object of the present invention to provide a method for reprocessing spent nuclear fuel oxide, which prolongs the life of the apparatus and efficiently recovers uranium and plutonium. Another object of the present invention is to provide a method for reprocessing spent nuclear fuel oxide, which can recycle generated crucible waste as radioactive waste disposal material.

[0010]

The invention according to claim 1 is
As shown in FIG. 1, a spent nuclear fuel oxide containing uranium and plutonium is supplied to a chloride molten salt, which is a reaction medium stored in a quartz crucible, and chlorine, carbon monoxide and inert gas are contained in the molten salt. A chlorination dissolution step 11 in which a first mixed gas composed of gas is blown to dissolve the entire amount of uranium and plutonium, and a second mixed gas composed of oxygen, chlorine and an inert gas is blown into the molten salt to be dissolved in the molten salt. PuO ₂ precipitation / recovery step 12 for depositing plutonium as plutonium oxide; blowing a third mixed gas comprising oxygen, chlorine and an inert gas into the molten salt and applying a current between the cathode and the anode installed in the molten salt And a UO ₂ electrolytic recovery step 13 in which uranium dissolved in the molten salt is deposited on the cathode as uranium oxide. In the invention according to claim 1, the chlorination dissolution step 1
In 1, since the first mixed gas uses carbon monoxide together with chlorine, it is not necessary to use a material serving as a carbon source for the crucible. Therefore, a silicon-based material having corrosion resistance to chlorine can be used, and corrosion of the device can be avoided. In addition, the reaction in the chlorination dissolution step proceeds by solid-gas reaction, so that the reaction occurs much more smoothly than the solid-gas reaction of the RIAR method. When a large amount of insoluble residue such as a platinum group element is generated, the insoluble residue precipitate is collected after the chlorination dissolution step. In the PuO ₂ precipitation / recovery step 12, a second mixed gas whose oxygen partial pressure has been adjusted to a high level is blown, and plutonium is precipitated and recovered as PuO ₂ . In the UO ₂ electrolytic recovery step 13, a third mixed gas with oxygen-rich chlorine is blown in to selectively generate uranium oxychloride compound UO ₂ Cl ₂ and reduce it to UO ₂ by electrolysis.

[0011] The invention according to claim 2, an invention according to claim 1, as shown in FIG. 1, UO ₂ electrolytic recovery step 1
Subsequent to 3, the reprocessing method includes a phosphate precipitation recovery step 14 in which sodium phosphate is further added to the molten salt to precipitate impurities contained in the molten salt as phosphate. In the invention according to claim 2, in the phosphate precipitation recovery step, phosphoric acid is added to the molten salt to precipitate impurities contained in the molten salt in the form of phosphate.

[0012]

Next, an embodiment of the present invention will be described with reference to FIG. Dissolution of spent nuclear fuel oxide (chlorination dissolution process 11)
First, a quartz crucible is prepared as a reaction vessel, and a molten salt, for example, a eutectic salt of NaCl-KCl is stored in the quartz crucible. Next, the spent nuclear fuel oxide is supplied into the molten salt, and then the first mixed gas composed of chlorine, carbon monoxide and an inert gas is blown into the molten salt, whereby uranium contained in the nuclear fuel oxide is discharged. And plutonium (formula (1) above and formulas (5) to (9) below).

UO ₂ (s) + 2CO (g) + 2Cl ₂ (g) → UCl ₄ (l) + 2CO ₂ (g) ↑ (5) PuO ₂ (s) + 2CO (g) + 2Cl ₂ (g) → PuCl ₄ (l) + 2CO ₂ (g) ↑ …… (6) PuO ₂ (s) + 2CO (g) + 3 / 2Cl ₂ (g) → PuCl ₃ (l) + 2CO ₂ (g ) ↑ …… (7) PuO ₂ (s) + UCl ₄ (l) + Cl ₂ (g) → PuCl ₄ (l) + UO ₂ Cl ₂ (l) …… (8) PuO ₂ (s) + UCl ₄ (l) + 1 / 2Cl ₂ (g) → PuCl ₃ (l) + UO ₂ Cl ₂ (l) (9) Here, the first mixed gas contains 0.1 mol of chlorine per mole of carbon monoxide.
Chlorine and carbon monoxide are mixed at a ratio of 5 to 1.5 moles, and the mixed gas is further mixed with 5% per 100% by volume of the inert gas.
It is a gas prepared by mixing with an inert gas at a ratio of about 90% by volume. The reaction rate can be controlled by changing the ratio of the inert gas. A for inert gas
r, N _{2 and the} like.

In this step, the reaction is carried out by heating to 400 to 800 ° C. Preferably it is 600-700 degreeC. 40
If the temperature is lower than 0 ° C., it is difficult to select a salt having a low melting point, and disadvantages such as a slow reaction are caused. When the temperature exceeds 800 ° C., the molten salt evaporates. The reaction temperature is adjusted to the dissolution rate, and the composition ratio of the first mixed gas is adjusted by the change in the oxidation order due to the heat treatment during the decoating treatment of the spent nuclear fuel oxide.
In addition, since this dissolution reaction is an exothermic reaction, by increasing the dilution ratio with an inert gas, it is possible to control the exotherm and cause a slow reaction. Therefore, it is also possible to change the dilution ratio according to the change in the exothermicity of the spent nuclear fuel oxide. The carbon dioxide generated by the above equations (5) to (7) is discharged out of the system. Since the reaction in the chlorination dissolution step is a solid-gas reaction between chlorine and carbon monoxide (gas) and spent nuclear fuel oxide (solid), the conventionally used spent nuclear fuel oxide (solid) and crucible The reaction rate is faster than the solid-gas reaction of carbon source (solid) and chlorine gas (gas).

PuO ₂ Precipitation Recovery (PuO ₂ Precipitation Recovery Step 12) A second mixed gas consisting of oxygen, chlorine and an inert gas is blown into the molten salt to precipitate plutonium dissolved in the molten salt as plutonium oxide.

PuCl ₄ (l) + O ₂ (g) → PuO ₂ (s) + 2Cl ₂ (g) ↑ (10) Here, the second mixed gas contains 0 to 5 moles of oxygen per mole of chlorine. It is a gas prepared by mixing oxygen and chlorine at a ratio and further mixing this mixed gas with an inert gas at a ratio of 5 to 90% by volume per 100% by volume of the inert gas. This second mixed gas is set to have a higher oxygen partial pressure than a third mixed gas described later. By increasing the oxygen partial pressure, the above equation (1)
The reaction of 0) is allowed to proceed. In this step, 400 to 700
The reaction is carried out by heating to ° C. Preferably 500-700 ° C
It is. If the temperature is lower than 400 ° C., problems such as difficulty in selecting a salt having a low melting point and a low deposition rate occur. When the temperature exceeds 700 ° C., salts and dissolved chlorinated substances evaporate. In addition, the precipitation of plutonium oxide can increase the crystal diameter of the precipitate by increasing the dilution ratio with an inert gas at a high reaction temperature, and can reduce the particle size by decreasing the dilution ratio with a low temperature.

The electrolytic recovery of UO ₂ (UO ₂ electrolytic recovery step 13) oxygen into the molten salt by blowing a third gas mixture consisting of chlorine and inert gas, molten salt to form the oxychloride compounds of uranium Uranium dissolved in a molten salt is precipitated on the cathode as uranium oxide by flowing an electric current between the cathode and the anode provided therein.

UCl ₄ (l) + O ₂ (g) → UO ₂ Cl ₂ (l) + Cl ₂ (g) ↑ (11) First, a third mixed gas is blown into the molten salt to obtain the above formula ( 1
As shown in 1), UO ₂ Cl ₂ which is a uranium oxychloride compound is generated. Here, the third mixed gas is
Oxygen and chlorine are mixed at a rate of 0.1 to 2 moles of oxygen per mole of chlorine, and the mixed gas is further mixed with an inert gas at a rate of 5 to 90% by volume per 100% by volume of the inert gas. Gas. Examples of the inert gas include Ar and N ₂ . The mixing ratio of oxygen and chlorine is changed depending on the concentration of the residual impurities in consideration of the electrolytic current. Further, similarly to the first mixed gas in the chlorination dissolution step described above,
By changing the ratio of the inert gas, the above equation (1)
The rate of the reaction shown in 1) can be controlled. In this step, the reaction is performed by heating to 400 to 700 ° C. Preferably it is 500-700 degreeC. If the temperature is lower than 400 ° C., it is difficult to select a salt having a low melting point. When the temperature exceeds 700 ° C., salts and dissolved chlorinated substances evaporate.

Next, an electrolytic reaction represented by the following formulas (13) to (15) is caused by placing a cathode and an anode in the molten salt and flowing an electric current. High-density graphite is used for the anode, and high-density graphite or pyrographite is used for the cathode. UO ₂ Cl ₂ before current is applied is ionized in the molten salt as shown in the following formula (12). The following formula (13) shows the reaction at the cathode, the following formula (14) shows the reaction at the anode, and the following formula (15) is an electrolytic reaction formula combining formulas (12) to (14).

UO ₂ Cl ₂ → UO _{2 2} ⁺⁺ 2Cl ⁻ … (12) UO _{2 2} ⁺⁺ 2e ⁻ → UO ₂ (s)… (13) 2Cl ⁻ → Cl ₂ (g) g + 2e ⁻ (14) UO ₂ Cl ₂ (l) → UO ₂ (s) + Cl ₂ (g) ↑ (15) The chlorine generated by the above formula (15) is discharged out of the system. As described above, uranium is converted into an oxychloride compound, and uranium dissolved in the molten salt is precipitated on the cathode as uranium oxide by electrolyzing the uranium oxychloride compound. The precipitated uranium oxide UO ₂ is removed and collected together with the electrode. In the PuO ₂ precipitation recovery step and the UO ₂ electrolysis recovery step, first, preliminary electrolysis is performed, and a noble metal FP such as Nb, Zr, Rh, Pd, or Rn is used.
It is also possible to concentrate s in UO ₂ and then carry out each step. By this method, highly pure plutonium and uranium can be recovered.

Phosphate Precipitation Recovery Step 14 Sodium phosphate is added to the molten salt to precipitate impurities in the form of phosphate. This impurity phosphate is neither soluble in the molten salt nor in the aqueous solution. 400 in this step
The reaction is carried out by heating to ８００800 ° C. Preferably 600-
700 ° C.

Processing of Waste Crucible In the reprocessing method of the present invention, since the material of the crucible is quartz having corrosion resistance to chlorine, it is hardly corroded. When the crucible made of quartz reaches the end of its life, it can be ground as a glass frit raw material of a high-level vitrified substance by adding a glass additive, so that waste can be recycled.

[0023]

As described above, according to the present invention, a spent nuclear fuel oxide containing uranium and plutonium is supplied to a molten chloride salt as a reaction medium stored in a quartz crucible in a chlorination melting step. Then, a first mixed gas comprising chlorine, carbon monoxide and an inert gas is blown into the molten salt to dissolve the entire amount of uranium and plutonium, and oxygen, chlorine and inert gas are dissolved in the molten salt in the PuO ₂ precipitation recovery step. Plutonium dissolved in the molten salt is precipitated as plutonium oxide by blowing a second mixed gas of gas, and a third mixed gas of oxygen, chlorine and an inert gas is blown into the molten salt in the UO ₂ electrolytic recovery step. At the same time, an electric current is passed between the cathode and the anode installed in the molten salt to precipitate uranium dissolved in the molten salt as uranium oxide on the cathode, thus prolonging the life of the device, And the plutonium can be efficiently recovered, and since the quartz crucible is used for the reaction vessel, the generated crucible waste can be recycled as a radioactive waste disposal material.

[Brief description of the drawings]

FIG. 1 is a diagram showing a process order of a method for reprocessing spent nuclear fuel oxide according to the present invention.

FIG. 2 is a diagram showing a process sequence of a conventional method for reprocessing spent nuclear fuel oxide.

[Explanation of symbols]

11 chlorination dissolution process 12 PuO ₂ precipitation recovery process 13 UO ₂ electrolysis recovery process 14 phosphate precipitation recovery process

Claims (6)

[Claims] 1. A reaction medium stored in a quartz crucible.
Spent containing uranium and plutonium in chloride molten salt
Supplying a nuclear fuel oxide and chlorine, monoacid in the molten salt;
The first gas mixture consisting of activated carbon and inert gas
Chlorination to dissolve all of the uranium and plutonium
A dissolving step (11), a second step comprising oxygen, chlorine and an inert gas in the molten salt;
Plutonium dissolved in the molten salt by blowing a mixed gas
That precipitates as plutonium oxide _TwoSettling times
Collecting step (12), a third step comprising oxygen, chlorine and an inert gas in the molten salt.
A cathode installed in the molten salt while blowing the mixed gas
A current is passed between the anode and the anode to dissolve the molten salt in the cathode.
UO that precipitates uranium as uranium oxide_Twoelectrolytic
Reprocessing method of spent nuclear fuel oxide including recovery step (13)
Law. 2. A phosphate precipitation-recovering step (13) in which sodium phosphate is further added to the molten salt to precipitate impurities contained in the molten salt as phosphate, following the UO ₂ electrolytic recovery step (13). The reprocessing method according to claim 1, wherein the method further comprises (14). 3. The first mixed gas is obtained by mixing chlorine and carbon monoxide at a ratio of 0.5 to 1.5 moles of chlorine per mole of carbon monoxide, and further mixing the mixed gas with 100% by volume of an inert gas.
The reprocessing method according to claim 1, wherein the reprocessing method is prepared by mixing the inert gas with an inert gas at a ratio of 5 to 90% by volume. 4. The method according to claim 1, wherein the second mixed gas contains no oxygen per mole of chlorine.
2. The method according to claim 1, wherein oxygen and chlorine are mixed at a ratio of about 5 to 5 mol, and the mixed gas is further mixed with an inert gas at a rate of 5 to 90% by volume per 100% by volume of the inert gas. Processing method. 5. The third mixed gas is obtained by mixing oxygen and chlorine at a ratio of 0.1 to 2 moles of oxygen per mole of chlorine, and further mixing the mixed gas with 5 to 90% by volume of inert gas 100% by volume. The reprocessing method according to claim 1, wherein the reprocessing method is prepared by mixing with an inert gas at a ratio of: 6. The chlorination and dissolution step is performed at 400 to 800 ° C.
uO ₂ precipitate recovery step 400 to 700 ° C., re-processing method according to claim 1, wherein the UO ₂ electrolytic recovery step is performed respectively in the temperature range of 400 to 700 ° C..