JP3214366B2 - Reprocessing of spent nuclear fuel - Google Patents

Description

DETAILED DESCRIPTION OF THE INVENTION

[0001]

The present invention relates to a method for reprocessing spent nuclear fuel, and more particularly to a method suitable for selectively recovering uranium or plutonium after dissolving spent nuclear fuel in a molten salt. .

[0002]

2. Description of the Related Art After dissolving spent nuclear fuel in molten salt,
Alternatively, a method for reprocessing a nuclear fuel in which uranium and / or plutonium is recovered by electrolyzing the molten salt while dissolving the same is described in, for example, Nuclear Power Industry Vol. 34 No. 12-4.
It is described on pages 6 to 60. In this method, a spent metal nuclear fuel or a spent oxide fuel reduced to a metal by pretreatment is dissolved in a solution cadmium, and a molten chloride is applied to an upper portion thereof. Next, the cathode of the metal electrode is inserted into the molten chloride, and current is passed through using liquid cadmium as the anode. Uranium in liquid cadmium is easily ionized and migrates into the molten chloride, and is further reduced at the metal cathode and deposited as a metal on the cathode. At this time, since plutonium forms an intermetallic compound with cadmium, it does not easily move into the molten salt, and since a very large potential difference is required to deposit plutonium on the metal electrode, uranium can be separated from plutonium. After recovering uranium, a liquid cadmium cathode is inserted into the molten chloride and an electric current is passed. In the cathode cadmium, plutonium is deposited on the cadmium cathode with a small potential difference so that plutonium forms an intermetallic compound. That is, in this process, uranium and plutonium can be recovered from spent metal nuclear fuel.
The feature of this treatment method is that since uranium in liquid cadmium is easily ionized, other metals are hardly dissolved in the molten salt, and the purity of the uranium recovered is high. Another method of reprocessing nuclear fuel is to recover uranium and / or plutonium by dissolving spent nuclear fuel in molten salt or by electrolyzing the molten salt while dissolving it. Proceedings, 391 pages, Presentation No. Q3 and 1995 Atomic Energy Society of Japan Annual Proceedings, 539 pages, Publication No. K41, and Atomic Energy Society of Japan 199
It is described in the 5th meeting of the conference proceedings, page 697, announcement number J39. In this method, a spent oxide fuel is dissolved in a molten salt using chlorine gas, and a target oxide of uranium and / or plutonium is selectively recovered by a cathode by electrolysis. A developed method of this method is shown in the Annual Meeting of the Atomic Energy Society of Japan, 1996 Annual Meeting, p. 603, Publication No. L61, in which uranium and / or Alternatively, the plutonium oxide is recovered. When uranium is dissolved by chlorine or anodic oxidation, uranium is dissolved in the molten salt as uranyl ions having an oxidation number of 6. The chemical formula of the uranyl ion is UO ₂ ²⁺ , which contains two oxygen atoms in the ion.

[0003]

## STR1 ## UO ₂ ²⁺ (in the molten salt) + 2e ~ (cathode) → UO ₂ (solid). On the other hand, fission products such as samarium are present in the molten salt in a form that does not contain oxygen. Supply is required.

[0004]

4Sm ³⁺ (in molten salt) + 6e ~ (cathode) + 3O ₂ (in molten salt or gas phase) → 2Sm ₂ O ₃ (solid) ... (Formula 2) Therefore, the oxygen partial pressure of the gas phase is constant. The conversion of fission products to oxides can be prevented, and uranium can be selectively recovered.

[0005]

According to the above invention, uranium and / or plutonium can be recovered from spent metal nuclear fuel and spent oxide nuclear fuel.
However, in the reprocessing method in which uranium or plutonium is recovered as a metal, uranium ions that can be present in the molten salt have an oxidation number of 3, and are easily converted to oxides by bonding with oxygen. For this reason, it is necessary to limit the oxygen concentration in the gas phase to 10 ppm or less. In the method of recovering uranium or plutonium as an oxide, the oxygen concentration in the gas phase is limited to a certain value or less so that fission products are not mixed into uranium or plutonium. This limit value depends on the concentration of chlorine gas. Most of the elements in spent fuel are dissolved by chlorine or anodic dissolution, so radioactive fission products and rare earth element chlorides that adversely affect fuel properties are added to the molten salt attached to the recovered oxide. Mixed. In order to prevent these chlorides from being converted into oxides and being mixed with the recovered uranium, it is necessary to limit the oxygen concentration in the gas phase to a certain value or less even in the salt separation process from the recovered uranium. .

[0006]

In the above prior art, it is necessary to control the oxygen concentration in the gaseous phase to a certain value or less, which complicates the structure of the apparatus and the processing facility. In the present invention, when dissolving the spent oxide nuclear fuel in the molten salt, the uranium is selectively dissolved by controlling the oxygen concentration to a certain value or more, preferably in an open state to the atmosphere, to recover uranium and plutonium. Prevention of contamination of fission products.

The action of oxygen on uranium and plutonium according to the present invention will be described below with reference to FIGS. 2 and 3.
When an oxide such as samarium is anodically dissolved in a molten salt,

[0008]

2Sm ₂ O ₃ (solid) + 3C (graphite) -6e ~ (cathode) → 4Sm ³⁺ (in molten salt) + 3O ₂ (in molten salt and in gas phase) A reaction of the following formula occurs. At this time, if the oxygen concentration in the molten salt is high, the anodic reaction hardly occurs. Therefore, Atomic Energy Society of Japan 1
In the method of dissolving oxide fuel described in the 994 Autumn Meeting Preprints, page 391, Publication No. Q3, etc., graphite is used as a crucible material. With graphite,

[0009]

The oxygen concentration in the molten salt is suppressed to a low level by the equilibrium reaction of 3C (graphite) + 3O ₂ (in the molten salt or in the gas phase) = CO ₂ (in the molten salt or in the gas phase). FIG. 2 shows the anodic dissolution potential of each oxide at the oxygen concentration at which the equilibrium of Chemical formula 4 is established. The lower the lysis reaction in this figure, the more easily the dissolution reaction occurs. In FIG.

[0010]

UO ₂ → UO ₂ ²⁺ + 2e ~ (Formula 5) and

[0011]

The reaction of UO ₂ → U ⁴⁺ + O ₂ 4e ~ (Formula 6) occurs at potentials higher than -0.68 V and -0.53 V, respectively. In order to dissolve UO ₂ , it is sufficient that either the reaction of Chemical formula 5 or Chemical formula 6 takes place.
If the voltage is set to 0.68 V or more, UO ₂ can be dissolved. At this time, PuO ₂ , Sm ₂ O ₃ , Np ₂ O ₃ , C
Since the potential at which anodic dissolution of m ₂ O ₃ , Am ₂ O _{3 and the} like is lower than 0.68 V, these oxides also dissolve together with UO ₂ . In contrast to this conventional method, in the present invention, dissolution is preferably performed in an open-to-atmosphere state. FIG. 3 shows the anodic dissolution potential at an oxygen concentration corresponding to gas-phase oxygen in the atmosphere open state. The reaction of releasing oxygen as shown in Chemical formulas 4 and 6 is difficult to occur due to the high oxygen concentration, and the anodic dissolution potential is high. On the other hand, since the uranium dissolution reaction of Chemical formula 5 does not release oxygen, the anodic dissolution potential does not change even if the oxygen concentration increases. As a result, in FIG. 3, the anodic elution potential corresponding to the uranium dissolution reaction of Chemical Formula 5 is the lowest potential. Thus, when the anodic potential is set between the anodic dissolution potential corresponding to the uranium dissolution reaction of Chemical Formula 5 and the anodic dissolution potential of Am ₂ O ₃ ,
UO ₂ dissolves anodicly, but PuO ₂ , Np ₂ O ₃ , Cm
Oxides of transuranium elements such as ₂ O ₃ and Am ₂ O ₃ and Sm ₂ O ₃
Oxides of rare earth elements such as do not dissolve. That is, UO ₂ can be selectively dissolved, and if cathodic reduction is performed after or simultaneously with dissolution, UO ₂ can be recovered at the cathode. When the oxygen partial pressure of the gas phase decreases, the anodic dissolution potential of the oxygen release reaction as shown in Chemical formula 4 decreases, and it becomes difficult to selectively dissolve UO ₂ . In order to dissolve UO ₂ quickly enough, the anode potential needs to be about 0.2 V higher than −0.68 V and −0.48 V. The anodic dissolution potential of Am ₂ O ₃ is -0.48 V
Is obtained from the thermodynamic formula

[0012]

Pc / (0.21 atm) = exp {(− 0.48 − (− 0.32)) × 4 × F / (RxT)} (Equation 1), which is about 0.0001 atm. . That is, if there is an oxygen partial pressure of about 1 / 10,000 atmosphere or more, UO ₂ can be selectively dissolved and collected in the cathode.

After recovering UO ₂ and setting the anodic potential between 0.02 V and −0.1 V under an oxygen concentration equivalent to opening to the atmosphere, the anodic dissolution potential is higher than 0.02 V.
O ₂ does not dissolve, but oxides of transuranium elements other than plutonium such as Np ₂ O ₃ , Cm ₂ O ₃ , Am ₂ O ₃ and Sm ₂
Oxides of rare earth elements such as O ₃ have an anode dissolution potential of −0.1 V
Dissolves because it is lower. Therefore, PuO ₂ selectively remains in the form of an anode, and high-purity plutonium can be recovered.

[0014]

BEST MODE FOR CARRYING OUT THE INVENTION

[Embodiment 1] A configuration of a spent nuclear fuel reprocessing apparatus in which a spent nuclear fuel reprocessing method according to one embodiment of the present invention is executed will be described with reference to FIG. This spent nuclear fuel reprocessing device comprises a reactor 1, a molten salt for electrolysis stuck in the reactor 2,
A dish-shaped anode 3 for dissolving spent fuel 3, a cathode 4 for depositing uranium oxide 4, a scraper 5 for peeling off deposited uranium 5, a recovery container 6 for storing scraped uranium oxide 6, a power supply 7 for electrolysis 7, It comprises a reference electrode 8 and a potentiometer 9 for measuring the potential difference between the reference electrode and the anode. The reference electrode 8 is calibrated so as to show a zero potential when the potential of the chlorine / chloride ion electrode at 1 atm is measured, or the chlorine / chloride ion electrode at 1 atm is used as a reference electrode. Spent oxide nuclear fuel 1
0 is loaded on the anode 3 from the upper part of the reaction tank. The gas phase of the reaction tank is opened to the atmosphere, or an inert gas containing oxygen is passed so that the oxygen partial pressure becomes 1 / 10,000 atmosphere or more. The potential difference between the anode and the reference electrode is read by a potentiometer 9, and a potential difference is applied between the cathode and the anode by the power supply 7 so that the potential of the anode becomes -0.48V. Uranium oxide contained in the spent oxide fuel dissolves in the anode and precipitates on the disk-shaped cathode 4.
Oxides of a transuranium element such as plutonium and oxides of a rare earth element such as samarium do not dissolve at an anode potential of -0.48 V since the oxygen partial pressure is at least 1 / 10,000 atmosphere. Therefore, when all of the uranium oxide contained in the spent oxide fuel is dissolved, no current flows. When the current stops flowing, the uranium oxide 11 deposited by the scraping device is transferred to the recovery container 6 while rotating the cathode 11. After scraping, the recovery container is taken out of the reaction tank and uranium is recovered. Next, the cathode is replaced before the plutonium is recovered. This new cathode is for depositing an oxide of a transuranium element other than plutonium and an oxide of a rare earth element such as samarium. If the cathode is not replaced, these elements will be mixed into the next recovered uranium. After replacing the cathode,
The potential difference between the anode and the reference electrode is read by a potentiometer 9 and
A power source 7 applies a potential difference between the cathode and the anode so that the potential of the anode is between 0.02 V and -0.1 V. An oxide of a transuranium element other than plutonium and an oxide of a rare earth element such as samarium contained in the spent oxide fuel are anodically dissolved and deposited on the replaced cathode. Anode potential is 0.02
At V and -0.1 V, the plutonium does not dissolve and remains on the anode. Take out the anode and collect the plutonium. In addition, the cathode is replaced with a cathode 4 for uranium recovery in preparation for the next uranium recovery.

As described above, according to the embodiment of the present invention, uranium and plutonium can be separately recovered from spent oxide fuel.

[Embodiment 2] The structure of a spent nuclear fuel reprocessing apparatus in which a spent nuclear fuel reprocessing method according to a second embodiment of the present invention is executed will be described with reference to FIG. The apparatus of the embodiment shown in FIG. 4 has a configuration in which a uranyl chloride input device 12 and a uranyl chloride concentration sensor 13 are added to the apparatus of the configuration of FIG. In the electrolysis of the molten salt, a substance having the same charge amount as the substance dissolved at the anode is deposited at the cathode. When recovering uranium in Example 1, uranium oxide UO ₂ is converted to uranyl ion UO ₂ ²⁺ and dissolved at the cathode. On the other hand, in the cathode, U
O ₂ ²⁺ is converted to oxide UO ₂ . UO ₂ generated at the anode
^{Since 2+} does not reach the cathode instantaneously,
For example

[0017]

## STR00007 ## Electrolysis of the molten salt itself occurs as shown in the following formula: KCl + 2e.fwdarw.K (metal) + Cl.about.
Therefore, when dissolving UO ₂ at the anode, UO ₂
It is desirable to add uranyl ion UO ₂ ²⁺ to the molten salt in advance so that ₂ is deposited. In the present embodiment, uranyl chloride is added to the molten salt by the uranyl chloride charging device 11, and the uranyl chloride concentration sensor 12 confirms that a required amount of uranyl ion is present.

According to this embodiment, in addition to the effects of the first embodiment, the electrolysis of the molten salt itself is prevented.

[0019]

According to the present invention, the dissolution is carried out with the anodic potential set lower than the anodic dissolution potential of plutonium oxide and higher than the anodic deposition potential of samarium oxide. By dissolving the transuranium element and leaving plutonium, plutonium can be recovered. Since these operations can be performed in the open air, the structure of the apparatus and the processing facility can be simplified and the cost is reduced compared to the conventional reprocessing method using a molten salt performed in a chlorine gas or inert gas atmosphere. Sex.

[Brief description of the drawings]

FIG. 1 is a block diagram of a spent nuclear fuel reprocessing device according to one embodiment of the present invention.

FIG. 2 is an explanatory diagram showing an anodic dissolution potential of an oxide at an oxygen concentration in which graphite coexists, which is a conventional reprocessing method.

FIG. 3 is an explanatory diagram showing the anodic dissolution potential of an oxide at an oxygen concentration under open air, which is the reprocessing method of the present invention.

FIG. 4 is an explanatory view of a spent nuclear fuel reprocessing device according to a second embodiment of the present invention.

[Explanation of symbols]

DESCRIPTION OF SYMBOLS 1 ... Reaction tank, 2 ... Molten salt, 3 ... Anode, 4 ... Cathode for uranium recovery, 5 ... Scraping device, 6 ... Collection vessel, 7 ... Power supply for electrolysis, 8 ... Reference electrode, 9 ... Potentiometer, 10 ... Spent oxide nuclear fuel, 11 ... Cathode deposit of uranium oxide, 12
... Uranyl chloride dropping device, 13 ... Uranyl ion sensor.

Claims (4)

(57) [Claims] In a reprocessing method of a nuclear fuel for recovering uranium and / or plutonium by electrolyzing the molten salt after or while dissolving the spent oxide nuclear fuel in the molten salt, the anode potential is changed. By dissolving the fission products and transuranium elements other than plutonium by setting the lower than the anodic dissolution potential of plutonium oxide and higher than the anodic deposition potential of samarium oxide, by leaving plutonium A method for reprocessing spent nuclear fuel, comprising recovering plutonium. Wherein Oite to claim 1, prior to dissolving or adding a compound of uranyl chloride or uranyl ion in the molten salt in advance, by allowed to generate uranyl chloride or Rauniruion, damage of the molten salt by electrolysis How to prevent reprocessing of spent nuclear fuel. 3. The fuel cell according to claim 1 , wherein the spent oxide atomic fuel is used as an anode to dissolve the fuel by anodic oxidation dissolution to determine the residual amounts of fission products and transuranium elements other than plutonium from the anode potential. And a reprocessing method for determining the point at which the dissolution reaction should be stopped. After dissolving wherein spent oxide nuclear fuel in a molten salt, or while dissolved in reprocessing of nuclear fuel to recover uranium and or plutonium by electrolyzing the molten salt, the uranium After recovering uranium, the spent fuel from which uranium has been recovered is recovered by plutonium according to the method of claim 1 or 3 ,
Reprocessing of nuclear fuel to recover uranium and plutonium separately.