JPH0735894A - Reprocessing method for spent nuclear fuel - Google Patents

Abstract

PURPOSE:To provide a reprocessing method for spent nuclear fuel which realizes simultaneous separation of uranium, plutonium, and transuranium elements from a spent nuclear fuel. CONSTITUTION:A spent nuclear fuel is dissolved into nitric acid and the solution of nitric acid is brought into contact with tributyl phosphate under reductive conditions thus extracting uranium into the tributyl phosphate and plutonium into the waste liquid while leaving transuranium elements and fission products.

Description

Detailed Description of the Invention

[0001]

BACKGROUND OF THE INVENTION 1. Field of the Invention The present invention relates to a method for reprocessing spent nuclear fuel, and more specifically to separating and recovering uranium, plutonium and transuranium elements from spent nuclear fuel by a solvent extraction method and separating only fission products. The present invention relates to a reprocessing method for processing.

[0002]

2. Description of the Related Art Spent nuclear fuel produced in a nuclear power plant is mainly composed of unburned uranium, plutonium produced from uranium during combustion, and transuranic elements, although there are differences depending on the composition, burnup and cooling period of the fuel. Americium, curium, etc. are present, and other fission products such as strontium, zirconium, cesium, ruthenium, etc. are present. Such spent nuclear fuel was finely cut, then treated by a solvent extraction method called the Purex method to recover uranium and plutonium, and reused for nuclear reactors. In this method, as shown in FIG. 3, spent nuclear fuel is dissolved in 2-4 normal nitric acid and brought into contact with tributyl phosphate to extract uranium and plutonium (referred to as co-decontamination), and a trace amount is mixed in this organic solvent. The fission products and transuranic elements are removed by washing with nitric acid, and a reducing agent is added to the organic solvent to separate plutonium from uranium as a non-extractable chemical form. As a result, transuranium elements such as americium and curium are treated as high-level radioactive liquid waste together with fission products.

However, transuranic elements generally have extremely long half-lives, and therefore extremely careful safety is required for preservation. Therefore, development is being made to reduce the burden of such waste management. For example, if only transuranic elements are separated from high-level radioactive liquid waste (referred to as group separation) and mixed again with uranium or plutonium and used as fuel in light water reactors or fast breeder reactors, they are converted into short-lived nuclides. And disappears (called the group separation extinction method). However, this method includes a process of recovering transuranic elements from a high-level radioactive liquid waste and a process of mixing transuranic elements with uranium or plutonium, and thus must be said to be complicated.

[0004]

Therefore, if the transuranium element can be extracted together with plutonium (referred to as a batch group separation method), the effect is great in terms of economic efficiency, safety, and protection of nuclear materials. From such a viewpoint, when uranium and plutonium are simultaneously extracted and separated from the spent nuclear fuel into the organic solvent, the transuranic element is extracted into the organic solvent side, and then the uranium is separated by combining the plutonium and the transuranic element. Methods have been examined but have not been realized. Therefore, the present invention provides a spent nuclear fuel reprocessing method that overcomes the above problems and realizes a method for collectively separating uranium, plutonium, and transuranium elements from spent nuclear fuel.

[0005]

[Means for Solving the Problems] In order to achieve the above object, the first invention is to dissolve a spent nuclear fuel in nitric acid and then bring the nitric acid solution into contact with tributyl phosphate under reducing conditions. It is characterized in that uranium is extracted into the tributyl phosphate and plutonium, transuranium element and fission products remain in the extraction waste liquid.

In the second aspect of the present invention, after dissolving spent nuclear fuel in nitric acid, the nitric acid solution is brought into contact with tributyl phosphate (1) under reducing conditions to extract uranium into the tributyl phosphate. , Plutonium in the extraction waste liquid (1),
The transuranic element and the fission product are left, and the acid concentration of the extraction waste liquid (1) is adjusted to 13 N or more by nitric acid,
Contacting with tributyl phosphate (2) to extract plutonium and transuranium element into the tributyl phosphate (2),
It is characterized in that fission products are left in the extraction waste liquid (2).

[0007]

The function of the spent nuclear fuel depends on the composition, burnup and cooling period of the fuel, but in the case of general LWR fuel, uranium, plutonium and other fission products and transuranic elements are in a weight ratio of 100: 1 :. 4 exists. The nitric acid that dissolves the spent nuclear fuel may be a normal product of 2 to 4N. Tributyl phosphate may be dissolved in dodecane or kerosene and usually used as a 20 to 60% solution. A mixer, a pulse column, or the like can be used to bring the nitric acid solution and tributyl phosphate into contact with each other, but is not particularly limited as long as it is a liquid-liquid contact type extractor.

It is necessary to create reducing conditions at the time of liquid-liquid contact, and examples thereof include addition of an inorganic reducing agent such as uranus nitrate and hydroxylamine nitrate. If these substances are added in an amount 3 to 8 times the stoichiometric amount, plutonium is reduced from tetravalent to trivalent, and it becomes difficult to dissolve it in tributyl phosphate. It is also possible to use Al, Zn, Mg or the like as a reducing metal, but this is not preferable because it also increases impurities.
The reduction of plutonium may be performed electrolytically.

As shown in the method of the present invention in FIG. 1, uranium is extracted into tributyl phosphate (1) by the extraction treatment under the above-mentioned reducing conditions, and plutonium, transuranium element, fission products, etc. are extracted. It remains in the waste liquid (1). At this time, the reducing action becomes efficient by dividing and adding the reducing agent to a plurality of stages of the mixer settler or a plurality of positions of the pulse column.

When the saturation of uranium with respect to tributyl phosphate is maintained at 80% or more, the amount of impurities mixed in the tributyl phosphate solvent (1) is reduced to one fifth of that of the Purex method.
It can be a degree. Further, it is effective to wash with the reducing agent and 1 to 3 N nitric acid in order to remove impurities from the tributyl phosphate solvent (1).

Next, FIG. 2 shows a method for treating the extraction waste liquid (1) containing plutonium, transuranium element and fission product. Nitric acid was added to the extraction waste liquid (1) to adjust the acid concentration to 13
It must be above the regulation. By this treatment, transuranium elements such as americium and curium and plutonium are oxidized from trivalent to tetravalent and are easily extracted by tributyl phosphate. On the other hand, fission products can be left in the extraction effluent (2). Since a small amount of fission products may be mixed in the tributyl phosphate solvent (2), it may be washed with nitric acid of 13 N or more and this washing waste liquid is added and mixed in the extraction waste liquid (1).

Further, in order to separate and recover uranium, plutonium and transuranium elements from the tributyl phosphate solvents (1) and (2) obtained in the above step, it is possible to obtain them by contacting with dilute nitric acid and back-extracting. it can. Tributyl phosphate after separation can be reused in the method of the present invention. The recovered uranium and plutonium and transuranic elements can be denitrified, concentrated and then mixed at an appropriate mixing ratio and reused as fuel for a nuclear reactor.

[0013]

【Example】

Example 1 The flow sheet used in this example is shown in FIG. First, tributyl phosphate diluted with normaldodecane to 30 vol% was supplied to the 10th stage of a compact mixer set with 10 stages in total at 500 ml / hour, and the washing liquid-nitric acid and HNO ₃ 2N- was supplied at 40 ml / hour from the 1st stage. From the 3rd and 7th stages, the reducing agent-uranas nitrate, U
(NO ₃ ) ₄ 0.5 mol / l, hydrazine (stabilizer of Uranus nitrate) N ₂ H ₅ NO ₃ 0.5 mol / l, nitric acid HNO _{3 2} N- ₂ ml / hr and 8 ml / hr, respectively.
Supplied by. About 1 hour after the start of operation, a simulated solution simulating a solution of spent nuclear fuel-uranyl nitrate UO ₂
(NO ₃ ) ₂ 1 mol / l, cerium nitrate (simulating substance of plutonium) Ce (NO ₃ ) ₄ 0.01 mol / l,
Gadolinium Nitrate (simulating substance of fission product) Gd (N
O ₃ ) ₃ 0.5 mol / l, neodymium nitrate (simulating substance of transuranium element) Nd (NO ₃ ) ₂ 0.5 mol / l,
HNO ₃ 3N-nitric acid was supplied to the fifth stage at 200 ml / h and operated for an additional 4 hours, and it was confirmed that almost steady operation was reached due to changes in the component concentrations in the organic solvent and the extraction waste liquid flowing out from the outlet of the mixer-settler. . As a result, the uranium concentration in the tributyl phosphate solvent (1) was about 0.41 mo.
The concentration of uranium in the nitric acid aqueous solution of the extraction waste liquid (1) was 0.0005 mol / l or less, and almost 100% of uranium was recovered. The cerium concentration in the tributyl phosphate solvent is 10 ^-5 mol / l or less, both neodymium and gadolinium are 10 ^-6 mol / l or less, and almost 100% of the components other than uranium are contained in the nitric acid aqueous solution of the extraction waste liquid (1). It was found to remain. Table 1 shows the balance of each component.

[0014]

[Table 1]

Example 2 In this example, the behavior of uranium and plutonium in the flow sheet shown in FIG. 5 was analyzed by using the calculation code DYNAC. This calculation code DYN
AC can accurately estimate the physicochemical behavior in the pulse column extraction device in the tributyl phosphate-uranium-plutonium-nitric acid system. In this example, a simulated solution-uranyl nitrate, UO ₂ (NO ₃ ) was added to the top of the first column of the two column pulse columns (effective height 8 m, column diameter 35 cm).
₂ 1.05 mol / l, plutonium nitrate, Pu (NO
₃ ) ₄ 0.012 mol / l, nitric acid, HNO ₃ 3 normal-
At 600 liters / hour from the bottom of the first tower with organic solvent
Tributyl phosphate 30%, normal dodecane 70% -at 1500 liters per hour, the washing liquid from the top of the second column-nitric acid, HNO ₃ 2 normal-at 100 liters per hour, the center of the first column and the bottom of the second column. Reducing agent to a height of 1m-
Uranus nitrate, U (NO ₃ ) ₄ 0.42 mol / l, hydrazine, N ₂ H ₅ NO ₃ 0.5 mol / l, nitric acid, H
NO ₃ 2 regulation-5 liters per hour and 3 per hour respectively
Supplied with 0 liter. Table 2 shows the balance of uranium and plutonium contained in the aqueous nitric acid solution of the extraction waste liquid (1) and the tributyl phosphate solvent (1) in the steady operation state obtained as a result of the calculation. From Table 2, it was found that almost 100% of uranium was recovered in the tributyl phosphate solvent and almost 100% of plutonium was recovered in the nitric acid aqueous solution of the extraction waste liquid (1).

[0016]

[Table 2]

Example 3 The flow sheet used in this example is shown in FIG. First, the washing solution-nitric acid, HNO ₃ 14.4N-from the first stage of the eight-stage small mixer-settler was used for 2 hours per hour.
0 ml and tributyl phosphate diluted with normaldodecane to 30 vol% were fed from the eighth stage at 100 ml / h. Approximately 2 hours after the start of operation, a simulated plutonium-containing solution of the extraction waste liquid (1) prepared in 13N nitric acid-cerium nitrate (plutonium simulating substance) Ce (NO ₃ ) ₄
0.01 mol / l, Gadolinium nitrate (simulator of fission product) Gd (NO ₃ ) ₃ 0.4 mol / l, neodymium nitrate (simulator of transuranium element) Nd (NO ₃ ).
₂ 0.4 mol / l, nitric acid and HNO ₃ 13N- were supplied from the 4th stage at 100 ml / h, and the operation was continued for another 4 hours. The tributyl phosphate solvent (2) and the extraction waste liquid (2) flowing out from the outlet of the mixer settra were extracted. It was confirmed that almost steady operation was reached due to the change in component concentration. As a result, almost 100% of cerium and neodymium were recovered in the tributyl phosphate solvent (2), and 99% or more of gadolinium was recovered in the nitric acid aqueous solution of the extraction waste liquid (2). Table 3
Figure 3 shows the balance of each component obtained.

[0018]

[Table 3]

[0019]

The present invention provides a reprocessing method for spent nuclear fuel that realizes a method for collectively separating uranium, plutonium and transuranium elements from spent nuclear fuel. It is possible to extract transuranic elements together with plutonium, which is economical and safe. The effect is great in terms of sex and protection of nuclear materials.

[Brief description of drawings]

FIG. 1 is a process diagram for extracting or remaining uranium and plutonium, fission products, and transuranium elements from a nitric acid aqueous solution of spent nuclear fuel.

FIG. 2 is a process diagram for extracting plutonium and transuranium elements from an aqueous nitric acid solution containing plutonium, fission products, and transuranium elements to leave fission products.

FIG. 3 shows a process diagram of a Purex method for an aqueous nitric acid solution of spent nuclear fuel.

FIG. 4 is a process diagram of Example 1.

5 is a process diagram of Example 2. FIG.

FIG. 6 is a process diagram of Example 3.

Claims (2)

[Claims] 1. A spent nuclear fuel is dissolved in nitric acid, and then this nitric acid solution is brought into contact with tributyl phosphate (1) under reducing conditions to extract uranium into the tributyl phosphate, and the extraction waste liquid (1) is extracted. ) Plutonium, transuranium element and fission product remain in the process). 2. The spent nuclear fuel is dissolved in nitric acid, and then this nitric acid solution is brought into contact with tributyl phosphate (1) under reducing conditions to extract uranium into the tributyl phosphate, and the extraction waste liquid (1 ), Plutonium, transuranium element and fission products remain, and the extraction waste liquid (1) is adjusted to have an acid concentration of 13 N or more with nitric acid, and then contacted with tributyl phosphate to obtain tributyl phosphate (2). 2. A method for reprocessing spent nuclear fuel, which comprises extracting plutonium and transuranium elements into the extraction waste solution and leaving fission products in the extraction waste liquid (2).