JP3159887B2 - Reprocessing of spent nuclear fuel - Google Patents

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 generated from a nuclear power plant, and more particularly to a method for reprocessing spent nuclear fuel containing minor actinoids such as neptunium and americium.

[0002]

2. Description of the Related Art Spent nuclear fuel generated from a nuclear power plant includes fission product nuclides (hereinafter referred to as FP as appropriate),
Unreacted uranium and transuranium elements including plutonium are included. Of these, uranium and plutonium can be used as nuclear fuel, so they are chemically separated and recovered from fission product nuclides and other transuranium elements. This chemical process is used fuel reprocessing, for example, N 2
uclear Chemical Engineerin
g Second Edition, pages 466 to 514 (M. Benedic)
The Purex method described in McGraw-Hill Co., Ltd.) is currently the mainstream. In this Purex method, after the spent nuclear fuel is dissolved with nitric acid,
VI-valent uranium and IV-valent plutonium migrate (extract) from an aqueous nitric acid solution into an organic extractant called tributyl phosphate (TBP), which separates them from fission product nuclides (co-decontamination) . Thereafter, the plutonium is reduced to a valence of III and migrates again from the extractant to the aqueous nitric acid solution (back extraction), whereby the plutonium is separated (distributed) from uranium.

At this time, since the main purpose of the Purex method is to obtain pure uranium and plutonium, other transuranium elements such as neptunium and americium are not recovered, and are reprocessed together with fission-produced nuclides. It shifts to a raffinate called high level waste liquid.
As described above, the transuranium element that is excluded from the recovery target in the Purex method and remains in the raffinate is a minor actinoid, and typical examples include neptunium and americium. In an aqueous solution, neptunium exists in V valence and americium in III valence. In this valence state, TBP
Is not extracted. These minor actinoids have very long half-lives, for example, neptunium-237 of 2.14 million years, americium-241 of 432.6 years, and americium-243 of 7370 years, and are α-emitters. The burden of radioactive waste management has been reduced by further separating and separately treating and disposing of other short-lived fission product nuclides contained in the reprocessed high-level waste liquid. For example:

(1) Separation of groups A group of fission-producing nuclides having a short half-life and a group of minor actinoids having a long half-life are separated from each other. Proc. GLOBAL '93, vol. 1, p58
8-594 (1993; M. Kubota et al) This known technique separates a minor actinoid, which has been transferred to a high-level waste liquid side, from most fission product nuclides using an extractant. At this time, americium is separated from the fission product nuclide together with the rare earth element of the homologous element at the most stable valence of III, and then further separated from the rare earth element.

(2) Group separation in the reprocessing process This is a combination of the above-mentioned group separation of (1) with fuel reprocessing, and for example, there is the following known technique. Proc. GLOBAL '93, vol. 2, p1
008-1014 (1993; JM Adnet e)
t al) This known technique involves the separation of uranium, plutonium, neptunium (VI) and americium (III) from most fission product nuclides, followed by the separation of americium and its accompanying rare earth elements from uranium, plutonium and neptunium. Finally, the valence of americium is adjusted to IV valence or VI valence to separate it from the rare earth element of III valence.

[0006] Japanese Patent Application Laid-Open No. 6-186389 discloses a known technique in which a step of oxidizing neptunium (V) to neptunium (VI) is provided on the high-level waste liquid side in the co-decontamination step of uranium and plutonium. A step of selectively reducing only neptunium (VI) to neptunium (V) between the uranium and plutonium distribution step to separate uranium, plutonium and neptunium from spent nuclear fuel. is there.

[0007]

However, the above-mentioned known technology has the following problems. That is, as in the known art, once the minor actinoid that has migrated to the high-level waste liquid is separated, the burden of long-term waste management can be reduced, but other processing systems that separate the minor actinoid from fission product nuclides are separately provided. Required, complicating the processing steps. Further, in the above-mentioned known technology, it is possible to separate americium from fission-producing nuclides such as strontium and cesium, but when americium is separated at a valence of III, a special solvent is used. In addition, a trivalent rare earth element accompanies americium. Therefore, a step of further separating the americium and the rare earth element separated from the fission product nuclide is required, and the processing step becomes complicated. Furthermore, in the above-mentioned known technology, uranium and plutonium do not coexist in the step of oxidizing neptunium (V), which is a minor actinoid, to neptunium (VI). That is, what is once taken out to the waste liquid side is returned, and the simplification (for example, one stream) of the treatment process is not sufficient. In addition, it is impossible to apply to, for example, americium other than neptunium. Further, in the above-mentioned known technique, in the valence adjustment step before the neptunium oxidation step, plutonium (I
Adjust to V) and finally purify plutonium alone with uranium. Therefore, there is a problem that extremely strict attention must be paid particularly to the handling of pure plutonium from the viewpoint of preventing nuclear proliferation. Also in the known technique, there is a similar problem since the purification of plutonium alone is performed.

An object of the present invention is to recover a minor actinoid containing neptunium and americium while avoiding the single extraction of plutonium, reducing the burden on handling, and sufficiently simplifying the process by one stream. It is an object of the present invention to provide a method for reprocessing spent nuclear fuel.

[0009]

According to the present invention, there is provided a dissolution process for dissolving spent nuclear fuel into a solution containing uranium, plutonium, minor actinoids, and fission product nuclides. A valence adjusting step of adjusting the valence of uranium and plutonium contained in the solution; a co-decontamination step of separating the uranium and plutonium whose valence has been adjusted from the fission product nuclide; and In a method for reprocessing a spent nuclear fuel having an element containing uranium and plutonium separated from nuclides, and a distribution step of distributing the element containing minor actinoids to an element containing uranium, the valence adjustment step includes: The valence of uranium and plutonium is adjusted to VI valence, and the valence of the minor actinoid is reduced to IV valence and VI valence. At least one, and in the co-decontamination step, a partitioning target element containing uranium (VI) and plutonium (VI) and a minor actinoid that is at least one of IV and VI are separated from the fission product nuclide. Separating, and in the partitioning step, while keeping uranium among the elements to be partitioned at VI valence, converting the minor actinoid to III valence
Reduce plutonium to at least one of the V valences and III
By reducing the element to be distributed into a mixture of a minor actinoid having at least one of valence III and V and plutonium having valence III, and uranium having valence VI. A reprocessing method is provided.

Preferably, in the reprocessing method for spent nuclear fuel, in the valence adjusting step, the valences of uranium and plutonium are adjusted to VI valence, and both the valences of neptunium and americium are set to VI valence. A method is provided for reprocessing spent nuclear fuel, the method comprising adjusting.

Preferably, in the reprocessing method for spent nuclear fuel, in the valence adjusting step, a peroxodisulfate, a xenonate, a bismuth (V) acid, a fluorooxysulfate, Also, there is provided a method for reprocessing nuclear fuel, wherein at least one of ozone is used.

[0012]

[0013] More preferably, in the reprocessing method for spent nuclear fuel, the co-decontamination is performed in the co-decontamination step in a state in which a stabilizer for a VI-valent actinoid ion is present. The present invention provides a method for reprocessing spent nuclear fuel.

[0014]

Preferably, in the reprocessing method for spent nuclear fuel, in the valence adjusting step, the valence of uranium and plutonium is adjusted to VI valence, and the valence of neptunium is adjusted to VI valence. A method for reprocessing spent nuclear fuel, comprising adjusting the valence of americium to IV valence.

[0016]

[0017]

More preferably, in the reprocessing method for spent nuclear fuel, the valence adjusting step includes an oxidizing agent for promoting oxidation for adjusting valence, the valence of which is I, II or III. A method for reprocessing spent nuclear fuel, comprising using a compound of soluble silver and an acid having at least one of valences.

[0019]

[0020]

[0021]

Further preferably, in the reprocessing method of the spent nuclear fuel, in the distribution process, the solvent extraction method, ion exchange method, by at least one of precipitation separation method, uranium of said distribution target element A method for reprocessing a nuclear fuel, comprising reducing a minor actinoid to at least one of valence III and V while reducing plutonium to valence III while maintaining the valence of VI.

More preferably, in the method for reprocessing spent nuclear fuel, at least one of hydroxylamine nitrate, hydrazine, nitrous acid, and iso-butyraldehyde is used as a valence adjusting agent for distribution in the distribution step. A method for reprocessing nuclear fuel is provided, wherein one is used.

[0024] To achieve the above object, according to the present invention, uranium by dissolving the spent nuclear fuel, plutonium, a melting step of the solution containing minor actinides and fission products nuclide, the solution A valence adjusting step of adjusting the valence of uranium and plutonium contained in, a co-decontamination step of separating the valence adjusted uranium and plutonium from the fission product nuclide, and separated from the fission product nuclide In a reprocessing method of spent nuclear fuel having a distribution step of distributing an element containing uranium and plutonium to an element containing a minor actinoid and an element containing uranium, in the valence adjustment step, the minor actinoid and plutonium are used. A method for reprocessing spent nuclear fuel, characterized in that .

In order to further achieve the above object, according to the present invention, there is provided a dissolving step for dissolving spent nuclear fuel into a solution containing uranium, plutonium, minor actinoids, and fission product nuclides; A valence adjusting step of adjusting the valence of the element contained in, and a reprocessing method of the spent nuclear fuel having a co-decontamination step of separating the element to be separated from the fission product nuclide, after the dissolving step and Provided before the valence adjusting step, the uranium valence is maintained at VI valence, the minor actinide valence is maintained at at least one of III valence and V valence, and the plutonium valence is maintained at III valence. A uranium recovery step of separating and recovering the VI-valent uranium from the mixture of the plutonium and the minor actinoid by reducing In the valence adjustment step,
Adjusting the valence of the plutonium to VI valence,
The valence of the minor actinoid is adjusted to at least one of IV valence and VI valence, and in the co-decontamination step, the VI
Valence plutonium, and the distribution target element including at least one of the IV valence and VI valence minor actinoid,
A method for reprocessing spent nuclear fuel is provided, wherein the method is separated from the fission product nuclide.

[0026]

In general, actinoids such as uranium, plutonium, neptunium and americium are generally liable to form a complex at IV valence and VI valence, but are difficult to form a complex at III valence and V valence. For this reason, it is easy to separate the valences IV and VI by solvent extraction or ion exchange, and it is difficult to separate the valences III and V from ions having low valences. In the present invention, in order to utilize this property, first, in the valence adjustment step, the valence of uranium and plutonium is adjusted to VI valence, and the valence of the minor actinoid is adjusted to IV valence or VI valence. Oxidizes the minor actinide in the presence of uranium and plutonium, ie, changes neptunium from V valence to VI valence and americium from III valence to IV valence or VI valence. Therefore, as compared with the case where the minor actinoid once taken out to the waste liquid side is oxidized and returned, one-stream operation is possible and the processing steps can be simplified. Then, in the co-decontamination step, elements to be distributed such as VI-valent uranium, VI-valent plutonium, and IV- or VI-valent minor actinoids are separated from fission-produced nuclides. That is, for example, in solvent extraction, actinide ions of IV valence or VI valence generally have a large partition coefficient and easily migrate from an aqueous solution to an organic solvent, but fission product nuclides are common extractants,
For example, tributyl phosphate has a small partition coefficient and is non-extractable, so that uranium, plutonium, and minor actinoids can be co-decontaminated from fission product nuclides.
At this time, the rare earth accompanies the fission-produced nuclide, or once it is VI-valent uranium, VI-valent plutonium, IV-valent
Even when entrained in a mixture of VI-valent minor actinoids, rare earths can be easily separated from this mixture thereafter. Here, when neptunium is separated in a trivalent state as in the prior art, a mixture of trivalent americium and a rare earth element is separated from the others, and then the neptunium is further separated.
A final separation step for separating I-valent americium and rare earth from each other is required. However, this final separation step becomes unnecessary, so that the number of steps can be reduced by one and the process can be simplified. it can. Further, unlike in the case of separating the minor actinoid once transferred to the waste liquid side of the fission product nuclide, another processing system for separating the minor actinoid from the fission product nuclide is not required separately. That is, the process can be simplified. In the distribution process, uranium among the elements to be distributed is
By reducing the minor actinoid to III or V valence (= non-extractable valence) and reducing plutonium to III valence (= non-extractable valence) while maintaining the VI valence (extractable valence) Then, the element to be distributed is separated into a mixture of a minor actinide having a valence of III or V, a plutonium having a valence of III, and uranium having a valence of VI. As a result, the minor actinoid can be recovered together with the plutonium, so that the recovery of the minor actinoid can be performed while avoiding the purification of plutonium alone, and the burden of handling can be reduced as compared with the case where pure plutonium is purified alone.

In the valence adjusting step, by using a valence adjusting agent such as peroxodisulfate, xenonate, bismuth (V) ate, fluoroxysulfate and ozone, uranium, plutonium and minor actinoids can be used. Everything can be adjusted to VI value. At this time, by performing the co-decontamination in a state of stabilizer VI valent actinides ion exist, it is possible to perform reliable and stable co-decontamination. Further, in atomic valence adjustment step, as an oxidizing aid promotes oxidation for valence adjustment, I value · II valent · II
By using the compounds of the I-valent soluble silver and acid,
It can help the valence regulator to oxidize plutonium from IV to VI, neptunium from V to VI, and americium from III to IV or VI . The of et, in the distribution process, hydroxylamine nitrate, hydrazine, nitrous acid, or iso - and solvent extraction method of the butyraldehyde and the like and the valence control agent for distribution, ion exchange method, using the sedimentation separation method or the like Thereby, it is possible to realize a step of reducing the minor actinide to III or V and reducing plutonium to III while maintaining uranium at VI.

[0028]

DESCRIPTION OF THE PREFERRED EMBODIMENTS Embodiments of the present invention will be described below with reference to the drawings. A first embodiment of the present invention will be described with reference to FIGS. In this embodiment, uranium, plutonium, neptunium and americium are adjusted to valence VI using oxidizing agent ammonium peroxodisulfate and oxidizing agent silver nitrate, and then co-decontamination is performed by a solvent extraction method using TBP. Things.

FIG. 1 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. Hereinafter, this procedure is referred to as a shearing step 1, a dissolving step 2, a valence adjusting step 3,
The co-decontamination step 4 and the distribution step 5 will be described in order. Spent nuclear fuel containing minor actinides such as uranium, plutonium, and neptunium americium,
First, in the shearing step 1, the sheet is cut into a predetermined length by the shearing device 6. Thereafter, the process proceeds to the dissolving step 2, and about 3 mol / l of nitric acid 41 is added and dissolved in the dissolving tank 7.
At this time, the radioactive gas waste generated in these steps is discharged from the main exhaust tower via the off-gas treatment system 15. In addition, for the shearing / dissolving / off-gas processing method and apparatus configuration in the shearing step 1 and the dissolving step 2, a known method / configuration adopted in a current spent fuel reprocessing process can be used.

Next, the process proceeds to the valence adjustment step 3, and the solution is supplied to the denitration / valence adjustment tank 8. Then, after formic acid 18 is added by a batch operation, the formic acid 18 is heated by a heating means (not shown) and denitrated until the nitric acid concentration becomes about 0.2 mol / l. This avoids local polymerization of IV-valent plutonium and lowers the nitric acid concentration without increasing the volume of the solution or the amount of waste generated. As a result of such denitration, molybdenum and zirconium contained in the solution as fission-producing nuclides precipitate in the denitration / valence adjustment tank 8 as a compound called zirconium molybdate. A part of plutonium having a valence of IV is coprecipitated with the plutonium or undergoes hydrolysis polymerization to precipitate. Note that there is no practical problem in the steps up to this point, since the results have already been obtained in the past by the Talpeak method, which is a type of group separation method. Further, since the plutonium precipitated at this time is later recovered in the solution, separation of the solution and the precipitate is not particularly performed. Further, the insoluble residue generated in the dissolving solution in the dissolving step 2 is removed together with other precipitates generated by denitration in the sediment removal device 14 described later, so that separation by solvent extraction as in the current reprocessing method is performed. No removal of insoluble residue is performed before the operation is performed.

Next, ammonium peroxodisulfate 19 as a valence adjusting agent is added to the solution in the denitration / valence adjusting tank 8 containing the precipitate by batch operation, and heated by a heating means (not shown). At this time, the concentration of ammonium peroxodisulfate is 0.35 mol / m in total of the amount necessary for oxidizing americium having a high redox potential and the amount of the coexisting substance consuming ammonium peroxodisulfate.
l is the required amount. These are kept at about 70 degrees for about 20 minutes, and uranium, plutonium, neptunium, and americium are adjusted to VI value.

Here, the principle of the adjustment of valence to VI valence will be described. In general, uranium has a VI value in aqueous solution,
Plutonium has the highest valence of IV, neptunium has the highest valence of V, and americium has the highest valence of III, and exists in these valence states without active valence adjustment.
When trying to adjust the valence of minor actinoids such as uranium, plutonium, and neptunium / americium to a valence of VI in a nitric acid solution, from the comparison of the oxidation-reduction potential,
It is most difficult to adjust americium from III to VI. In other words, if americium is changed from III valence to VI valence, the valence adjustment of other elements can be achieved at the same time. Then, the present inventors, in order to confirm the difficulty of valence adjustment of americium in nitric acid solution,
Experiments were conducted in which 0.1 M, 1 M, 1.5 M, and 2 M ammonium peroxodisulfate ((NH 4 ⁺ ) ₂ S ₂ O ₈ ) were added to various concentrations of nitric acid, and the yield of americium (VI) was measured. The result is shown in FIG. As shown in FIG. 2, the lower the nitric acid concentration and the larger the amount of ammonium peroxodisulfate added, the more likely it is to obtain VI-valent americium in high yield. That is, for example, when the nitric acid concentration is 0.2
In nitric acid up to mol / l, the americium is adjusted to a VI value with a maximum of 0.1 mol / l ammonium peroxodisulphate in a yield of 100%. And the nitric acid concentration is 0.2
It can be seen that the yield decreases as the amount increases from mol / l.

However, as shown, even in such a high concentration of nitric acid, the yield can be increased by increasing the concentration of ammonium peroxodisulfate. For example, by adding 1.5 mol / l ammonium peroxodisulfate in nitric acid having a concentration of 1 mol / l, about 90% of americium can be adjusted to VI value. At this time, since the concentration of silver ions does not affect the yield of VI valence, it is understood that silver obtained as a fission product nuclide can be used as an oxidation aid.

Considering based on the above experimental results, in this example, as described above, since the nitric acid concentration is 0.2 mol / l, 100% of americium is completely changed from III valence to VI valence. It can be seen that migration is possible. At the same time, uranium, which is originally VI-valent, is maintained at the same VI-valence, and neptunium is oxidized to VI-valent.
In addition, plutonium with IV valence contained in the precipitate
It is oxidized to a valency and leached into the solution. As described above, the silver ions contained in the nuclear fuel and present as fission product nuclides in the solution act as an oxidation aid at this time to accelerate the oxidation reaction. After the completion of the reaction, concentrated nitric acid 20 is added by batch operation, and the nitric acid concentration is increased again to 1 mol / l or more.

In this valence adjusting step, ruthenium of the fission product nuclide is oxidized to VIII and volatilized, and is removed from the off-gas in the off-gas treatment system 15 by a contact / reduction method or the like.

The lysate having undergone the valence adjustment step 3 as described above is sent to the filtration device 14 where the precipitate is removed, and then sent to the measuring tank 9. After being measured in the measuring tank 9, the supply tank 10 is intermittently supplied to the supply tank 10.
Is temporarily stored until a predetermined amount is reached.

The solution having reached a predetermined amount is sent to the co-decontamination step 4 and supplied to the extractor 11. And this extractor 11
Inside, 100% tributyl phosphoric acid (TBP) 21 is added via a solvent washing system 17, and uranium, plutonium,
Neptunium and americium are solvent extracted.

Here, this tributyl phosphoric acid (TBP)
The principle of solvent extraction by will be described. In order to confirm the difficulty of solvent extraction in tributylphosphoric acid (TBP), the inventors of the present invention have made representatives of neptunium (VI), americium (VI), and fission-producing nuclides whose valences have been adjusted by the above procedure. An experiment was conducted to measure the dependence of the partition coefficient of neodymium on the concentration of tributyl phosphate (TBP). Generally, using tributyl phosphate (TBP)
During solvent extraction of VI-valent actinides, the partition coefficient tends to increase with nitric acid concentration and tributyl phosphate (TBP) concentration. Here, particularly, since the nitric acid concentration was fixed at a substantially constant value (= 1 mol / l), the experiment was performed by changing the concentration of only tributylphosphoric acid (TBP). The result is shown in FIG.

As shown in FIG. 3, neptunium (V
It can be seen that all of I), americium (VI) and neodymium (III) show a higher partition coefficient as the concentration of tributyl phosphate (TBP) is higher. For example, in 100% tributylphosphoric acid (= 3.65 mol / l), the partition coefficients of neptunium and americium both exceed 1, and
Neodymium distribution coefficient is 1/100 of Neptunium (VI)
Hereinafter, it is about 1/50 of americium (VI). Therefore, by performing extraction with such tributylphosphoric acid (TBP), the desired minor actinoid can be separated from neodymium, which is one of the rare earth elements originally considered most difficult to separate. It can be seen that a distribution coefficient can be obtained.

Considering the above experimental results, in this embodiment, since 100% tributyl phosphate (TBP) is added as described above, uranium, plutonium, It turns out that neptunium and americium can be separated well.

At the time of extraction in the extractor 11, the sulfate ion of ammonium peroxodisulfate added in the denitration / valency adjusting tank 8 remains in the solution. (Uranium, plutonium, neptunium, americium). Further, since the specific gravity of tributylphosphoric acid (TBP) is 0.9724 (25 ° C.), which is close to that of water, there is a concern that phase separation may be deteriorated if tributylphosphoric acid (TBP) is used at 100% without dilution. However, since the solution to be used in the present embodiment mainly contains uranium and has a large specific gravity, this problem hardly occurs, and sound phase separation can be performed. Furthermore, if the extractor 11 and the later-described setters of the washer 12 and the distributor 13 are centrifugally separated, the phase separation becomes more sound, and there is no practical problem.

As described above, the solvent is extracted in the extractor 11, and the raffinate of the fission product nuclide containing the rare earth element is discharged as a high-level waste liquid 26 and sent to a high-level waste liquid treatment system (not shown). Finally, processing such as vitrification is performed. Uranium, plutonium, neptunium and americium extracted in the organic phase are sent to the washer 12. Then, after the nitric acid 22 is added and washed in the cleaning device 12, the nitric acid 22 is supplied to the distributor 13 and the process proceeds to the distribution process 5.

In the distribution step 5, a nitric acid aqueous solution 23 containing a reducing agent (eg, hydroxylamine nitrate, hydrazine, nitrous acid, iso-butyraldehyde, etc.) is supplied to and brought into contact with the solution in the distributor 13. As a result, while maintaining uranium at VI valence, plutonium is converted from VI valence to III
Neptunium from VI to V, Americium
Since it is possible to reduce from VI valence to III valence, a mixture 25 of plutonium, americium and neptunium can be back-extracted while uranium 24 remains in the organic phase of the solution.

According to the present embodiment configured as described above,
In the valence adjustment step 3, in the state where uranium and plutonium coexist, americium and neptunium can be oxidized using tributylphosphoric acid (TBP), which is used in the current reprocessing method, Purex method, and whose properties are well known. Therefore, as compared with the case where the minor actinoid once taken out to the waste liquid side is oxidized and returned, one-stream operation is possible, and the processing steps can be simplified. Then, in the co-decontamination step 4, the rare earth element is separated along with the fission product nuclides, so that the number of separation steps can be reduced by one, thereby simplifying the steps. Further, the process can be simplified as compared with the case where the minor actinoid once transferred to the waste liquid side of the fission product nuclide is separated. And, in the distribution step 5, the minor actinoid (neptunium / americium) having a long half-life can be recovered together with plutonium, so that the minor actinoid can be recovered while avoiding the purification of plutonium alone,
The burden of handling can be reduced as compared with a case where pure plutonium is purified alone, and the danger of nuclear proliferation can be avoided. Also, recovery of uranium and plutonium useful as nuclear fuel and separation of long-lived minor actinoids from short-lived fission-produced nuclides can be performed in the same process, thereby simplifying waste disposal. Further, compared with the third embodiment described later, there is also an effect that the amount of nonvolatile components contained in the oxidizing agent (= ammonium peroxodisulfate) can be reduced.

In the first embodiment, denitration is used as a method for lowering the concentration of nitric acid in the solution. However, the concentration of nitric acid can be reduced by dilution. In the first embodiment, ammonium peroxodisulfate is used as a valence adjusting agent. However, the present invention is not limited to this, and other peroxodisulfates, xenonates, bismuth (V) salts, and fluorooxysulfates are used. Uranium, plutonium, neptunium, and americium may be adjusted to a VI value using ozone or the like. In these cases, a similar effect is obtained. Uranium, plutonium, neptunium, and americium may be adjusted to a VI value by a known electrochemical method, and the same effect is obtained in this case. Furthermore, in the above-described embodiment, silver ions originally contained in the solution of nitric acid (that is, silver nitrate) were used as an oxidation aid. However, the present invention is not limited to this. For example, silver sulfate, silver phosphate, silver oxide, etc. A compound of an acid and silver may be used. In these cases, a similar effect is obtained.

A second embodiment of the present invention will be described with reference to FIG. This example is an example using another type of extractant in the co-decontamination step. The same reference numerals are given to members, processes and the like equivalent to those in the first embodiment.

Generally, among the minor actinoids, an extractant capable of extracting americium / curium and the like in a valence state of III without adjusting the valence can extract actinide ions of IV and VI under the same conditions. However, in this case, the rare earth element, which is one of the fission product nuclides, is extracted at the same time as the trivalent americium.
Separation step for separating valence americium and rare earth element is required separately. This embodiment solves this problem by adjusting the valence of americium to VI valence after the dissolution step.

FIG. 4 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. Uranium, plutonium, americium, and neptunium at a nitric acid concentration of 0.2 mol / l using silver ions as an oxidizing aid in the denitration and valence adjusting tank 8 in the shearing step 1, the dissolving step 2, and the valence adjusting step 3 The procedure up to the adjustment to the value is the same as that of the first embodiment, and the description is omitted.

After the adjustment of the valence to the VI valence, in the first embodiment, nitric acid 20 (see FIG. 1) was added to increase the nitric acid concentration. Concentration 0.
The mixture is sent to the co-decontamination step 4 via the precipitate removing device 14, the measuring tank 9, and the supply tank 10 as it is at 2 mol / l. The extractor 11 in the co-decontamination step 4 is supplied with di-isodecyl phosphate (DIDPA) 221 which is a typical acidic organic phosphorus-based extractant capable of extracting a trivalent actinoid ion from a nitric acid aqueous solution with a large partition coefficient. Is done. This di-isodecyl phosphate (DIDPA) 221 has a cation exchange mechanism, and actinide ions having VI and IV valences have a large partition coefficient in a wide acid concentration range in di-isodecyl phosphate (DIDPA). With. Therefore, it is not necessary to adjust the acid concentration as in the first embodiment before the extraction. As a result, uranium (VI), plutonium (VI), neptunium (VI), americium (VI)
Is separated from the fission product nuclide and extracted into the organic phase. At this time, the rare earth element which is one of the fission product nuclides is also extracted into the organic phase.

However, in this case, the partition coefficient of III-valent ions, for example, americium, curium and rare earth elements before adjusting the valence, shows a large value with nitric acid having a nitric acid concentration of about 0.1 mol / l. It has the property that the value decreases with increasing. Therefore, by washing this organic phase with a 3 to 6 mol / l nitric acid aqueous solution 222 in the washer 12, the rare earth element once extracted into the organic phase can be back-extracted into the aqueous phase. Subsequent distribution step 5 includes the first
The description is omitted because it is the same as that of the embodiment.

According to the present embodiment configured as described above,
In the co-decontamination step 4, the rare earth is once extracted together with uranium, plutonium, neptunium, and americium into the organic phase, but thereafter, the rare earth alone can be easily back-extracted and separated therefrom. Conventionally, after separating a mixture of trivalent americium and rare earth from the others, a final separation step of further separating these trivalent americium and rare earth from each other was required. The number of steps can be reduced by one to eliminate the need, and the steps can be simplified. In other respects, the same effects as in the first embodiment can be obtained. In addition to this, it is not necessary to adjust the acid concentration again after adjusting the valence with ammonium peroxodisulfate 19 in the denitration / valence adjusting tank 8, so that the amount of waste generated due to the oxidizing agent can be reduced. There is. Further, there is an effect that curium among minor actinoids can be separated from other fission product nuclides together with rare earth elements.

A third embodiment of the present invention will be described with reference to FIG. This embodiment is an embodiment in which the concentration of nitric acid in the denitration / valence adjusting tank 8 in the valence adjusting step is different. That is, uranium, plutonium, neptunium and americium whose valence has been adjusted in a relatively high concentration of nitric acid are extracted and separated at the same concentration without adjusting the nitric acid concentration. Members and processes equivalent to those in the first and second embodiments are denoted by the same reference numerals.

FIG. 5 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. The spent nuclear fuel in the shearing step 1 and the dissolving step 2
Up to the procedure of dissolving with about 3 mol / l nitric acid in
The description is omitted because it is the same as that of the embodiment.

After dissolving the spent fuel, in the first embodiment, the process immediately proceeds to the valence adjusting step 3 and the dissolved liquid is supplied to the denitration / valency adjusting tank 8, but in the present embodiment, the finer 327 is used.
After removing the undissolved residue in, it is sent to the denitration / valence adjusting tank 8. This is because the molar concentration of nitric acid in the valence adjustment step 3 described later is relatively high, and no precipitation occurs in the denitration / valence adjustment tank 8, and the downstream side of the denitration / valence adjustment tank 8 as in the first embodiment. This is because the insoluble residue must be independently removed since no precipitation removing device 14 is provided.

Formic acid 18 is added to the solution in the denitration / valency adjusting tank 8 by a batch operation in the same manner as in the first embodiment, and the concentration of nitric acid becomes about 1 mol / l, which is higher than in the first embodiment. Denitration.

Next, the temperature of the denitrated solution was set to about 7
The temperature is adjusted to 0 ° C., and as in the first embodiment, ammonium peroxodisulfate 19 as a valence adjusting agent is added by a batch operation to adjust uranium, plutonium, neptunium and americium to valence VI. At this time, the required amount of ammonium peroxodisulfate is 1.75 mol / l, which is larger than in the first embodiment. By maintaining at about 70 ° C. for about 20 minutes, uranium, plutonium, neptunium and americium are adjusted to VI value. At this time, since the nitric acid concentration was 1 mol / l, the yield of americium (VI) was about 90%, as shown in FIG. It is contained in the lysate as it is. Thereafter, the dissolving solution is not added with nitric acid 20 (see FIG. 1) as in the first embodiment, and the nitric acid concentration is not increased. After passing through a similar measuring tank 9, it is temporarily stored in a supply tank 10.

The solution that has reached a predetermined amount in the supply tank 10 is
It is sent to the co-decontamination step 4 and supplied to the extractor 11. Then, in the extractor 11, as in the first embodiment, 100% tributyl phosphate (TB) is passed through the solvent washing system 17.
P) 21 is added and uranium, plutonium, neptunium and americium are solvent extracted. However,
The solvent extraction behavior at this time is slightly different from that of the first embodiment. This will be described with reference to FIG.

FIG. 6 shows neptunium (VI), americium (VI), and valence-adjusted by the above-described procedure, similarly to the experiment by the inventors of the present invention whose results are shown in FIG. 3 in the first embodiment. It is a figure showing the experiment result which the present inventors performed in the experiment which measures the dependence of the distribution coefficient of neodymium which is a representative fission product nuclide to the concentration of tributyl phosphate (TBP).

As shown in FIG. 6, similarly to FIG. 3, both neptunium (VI) and americium (VI) show a higher partition coefficient as the concentration of tributyl phosphate (TBP) is higher. 100% tributyl phosphoric acid (3.65 mol / l)
Then, the partition coefficients of neptunium and americium are both 1
Is over. At this time, the distribution coefficient of neodymium is 1/100 or less of that of neptunium (VI), and americium (V
Since it is about 1/100 of I), the desired minor actinoid can be separated from neodymium, which is one of the rare earth elements, and a large distribution coefficient can be obtained as in the first embodiment. Understand.

Considering the above experimental results, in this embodiment, 100% of tributylphosphoric acid (TBP) is added in the same manner as in the first embodiment. ,plutonium,
Neptunium can be well separated and quantitatively extracted and recovered. However, for Americium,
As described above, although about 90% has shifted to VI valence, the remaining about 10% of americium exists in a valence of III. Since the partition coefficient of this III-valent americium is almost the same as that of neodymium shown in FIG. 6, it is not extracted into the organic phase as in the case of VI-valent americium, and is divided into fission product nuclides containing rare earth elements. They will behave together.

Therefore, in the extraction in the extractor 11 as described above, the raffinate 42 containing the fission product nuclide containing the rare earth element and the americium having a valence of III is discharged. Therefore,
In order to quantitatively recover the americium remaining in the raffinate 42, the raffinate 4 after the first extraction operation has been completed.
2 is returned to the denitration / valence adjustment tank 8 again and circulated, and the valence adjustment and extraction are repeated. At this time, the raffinate 42 from the extractor 11 contains tributyl phosphate (TB) from the organic phase.
Since P) 21 is dissolved and mixed, it is removed by diluent washing which is also performed in the current reprocessing. That is, the solubility in water is low and tributyl phosphoric acid (TB
A cleaning solvent (for example, n-dodecane) 30 of an organic substance that dissolves P) is dissolved in a solution cleaning device 2 through a cleaning solvent treatment system 29.
8 and the raffinate 42 is supplied to and brought into contact with the solvent liquid washer 28, and tributyl phosphoric acid (TBP) is supplied to the raffinate 4.
Remove from 2. This prevents the oxidation reaction from being hindered when returning to the denitration / valence adjustment tank 8 and adjusting the valence again. The washed raffinate 43 is sent to the denitration / valency adjusting tank 8, and the above-described method is repeated again. Finally, all unrecovered III-valent americium is adjusted to VI-valent and the extractor 11 Recovered in the organic phase within the raffinate 4
2 is discharged from the extractor 11 in a form containing only fission product nuclides containing rare earth elements. The discharged raffinate 42 is further supplied to the high-level waste liquid 2
It is discharged as 6 and sent to a high-level waste liquid treatment system (not shown), and finally, a treatment such as vitrification is performed.

The uranium, plutonium, neptunium, and americium finally extracted into the organic phase as described above are sent to the washing device 12, where the nitric acid 22 is added and washed as in the first embodiment. Then, it is supplied to the distributor 13 and moves to the distribution step 5. At this time, as described above, the ratio of the contained elements differs between the organic phase after the first extraction in the extractor 11 and the organic phase after the extraction from the relationship of the valence adjustment efficiency, Since the ratio of americium becomes larger later, the organic phase from the washer 12 after the co-decontamination step 4 has been completed is distributed to the distributor 13 in the next distribution step.
Before being sent to the storage tank 331, it is stored in the storage tank 331 until it reaches a predetermined amount.

The subsequent distribution step 5 is the same as in the first embodiment, and a description thereof will be omitted.

According to this embodiment, the same effect as that of the first embodiment can be obtained.

A fourth embodiment of the present invention will be described with reference to FIG. This embodiment is an embodiment using another type of extractant in the co-decontamination step in the third embodiment. First
The same reference numerals are given to members, processes, and the like equivalent to those of the first to third embodiments.

FIG. 7 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. The procedure from the shearing step 1, the dissolving step 2, the valence adjusting step 3, the subsequent measurement in the measuring tank 9 and the storage in the supply tank 10 is the third step.
The description is omitted because it is the same as that of the embodiment.

The solution from the supply tank 10 is subjected to the co-decontamination step 4
To be supplied to the extractor 11. In the extractor 11,
As in the case of the above-mentioned DIDPA, n-octyl-N, N-diisobutylcarbamoylmethylene phosphonic acid (CMPO) 421 capable of extracting a trivalent actinoid ion from a nitric acid aqueous solution with a large partition coefficient is supplied. As a result, uranium (VI), plutonium (VI), neptunium (VI), and americium (VI) whose valence has been adjusted are separated from fission product nuclides that become high-level waste liquid 26 and extracted into the organic phase. However, at this time, about 10% of americium (III) whose valence is not adjusted is contained in the organic phase, and rare earth elements are also extracted in the organic phase. Then, this organic phase is supplied to the rare earth decontamination device 412 and washed with diluted nitric acid 422, whereby the rare earth element once extracted into the organic phase is back-extracted into the aqueous phase 44 together with americium (III). This is guided to a solvent liquid washer 28 in the same manner as in the third embodiment, and is brought into contact with a cleaning solvent for cleaning. After that, it is sent to the denitration / valency adjusting tank 8, and the above-described method is repeated again.

As a result, finally, the unrecovered III-valent americium is completely adjusted to VI-valent in the extractor 1.
In the aqueous phase 44 from the rare-earth decontamination device 412, the aqueous phase 44 contains only the rare-earth element and is discharged from the solution cleaning device 28 as a rare-earth waste liquid 45. By repeating this extraction operation and valence adjustment, americium can be quantitatively extracted as VI valence.

Thus, VI-valent uranium / plutonium /
Rare earth decontamination device 412 containing Americium / Neptunium
Is sent to the storage tank 331. Thereafter, the distribution step 5 is the same as in the third embodiment, and a description thereof will be omitted.

According to this embodiment, the same effect as that of the third embodiment can be obtained. In addition to this, there is an effect that curium among minor actinoids can be separated from other fission product nuclides together with rare earth elements.

A fifth embodiment of the present invention will be described with reference to FIG. In the present example, in the valence adjustment step, uranium, plutonium, and neptunium are made to have a VI valence by using ozone as an oxidizing agent, silver nitrate as an oxidation aid, and potassium phosphotungstate as an ion stabilizer (= complexing agent). On the other hand, americium is adjusted to an IV value, and in the subsequent co-decontamination step, co-decontamination is performed by a solvent extraction method using an amine-based extractant. Members and processes equivalent to those in the first to fourth embodiments are denoted by the same reference numerals.

FIG. 8 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. The spent nuclear fuel in the shearing step 1 and the dissolving step 2
Up to the procedure of dissolving with about 3 mol / l nitric acid in
The description is omitted because it is the same as that of the embodiment. After dissolving the spent fuel into a solution, the insoluble residue is removed by a finer 327 as in the third embodiment, and the solution is sent to a valence adjusting tank 532.

In the valence adjusting tank 532, ozone 33 is blown into the solution, and uranium, plutonium and neptunium are adjusted to valence VI in advance. And to this potassium phosphotungstic acid _{(K 10 P 2 W 17 O} 61)
After the addition of 34, ozone 33 is blown again to adjust the americium to IV value.

Here, in general, americium is converted from III-valent to I-valent.
When adjusting to V value, the standard oxidation-reduction potential (=
(Formula potential) is as high as 2.34 V, so it is necessary to coexist a complexing agent for stabilizing americium (IV). That is, for example, when a complexing agent, which is a complexing agent selective to an IV-valent ion and has a property of stabilizing the IV valence, is added, such as various heteropolyphosphates or phosphate ions, complex formation with americium occurs. Is performed to lower the formula weight potential, which makes it easier to adjust the valence. In particular, when potassium phosphotungstate is used, the formal potential under conditions that allow americium to quantitatively form a complex with phosphotungstate ions can be reduced to 1.52V. Can be adjusted. This potential is also sufficient to adjust neptunium from V valence to VI valence. In this case, the valence adjustment includes periodate, permanganate, cerium (IV) compound, peroxodisulfate, xenonate, bismuth (V), fluorooxysulfate, in addition to ozone. And the like, and a known electrochemical method may be used.

The conditions under which americium forms a complex quantitatively, that is, the ratio of phosphotungstate ions to americium varies depending on the acid concentration. As the acid concentration increases, excess phosphotungstate ions are required. Show a tendency to become In the present embodiment, since the valence adjustment is carried out at 3 mol / l without adjusting the acid concentration of the solution, the concentration of phosphotungstate ions is required to be at least four times that of americium. Further, in the present embodiment, ozone 33 is blown before adjusting the valence of americium, and the valence of plutonium is adjusted from the IV valence to the VI valence, so that the phosphotungstate ion is consumed other than the americium. , And the amount of potassium phosphotungstate 34 added thereafter is reduced as much as possible.

The solution whose valence has been adjusted is temporarily stored in the supply tank 10 through the measuring tank 9 without adjusting the concentration of nitric acid as in the first embodiment, and at 3 mol / l at the time of dissolution. .

The solution that has reached a predetermined amount in the supply tank 10 is
It is sent to the co-decontamination step 4 and supplied to the extractor 11. In the extractor 11, trioctylamine (T) which is an amine-based extractant having an anion exchange mechanism and extracting actinide ions having a valence of IV and VI is provided through a solvent washing system 17.
OA) 35. Thereby, uranium, plutonium, and neptunium of VI valence become an anion complex coordinated with nitrate ion, and americium of IV valence becomes an anion complex coordinated with phosphotungstate ion. At this time, fission product nuclides containing rare earths are
Since it cannot exist as an anion complex, only uranium, plutonium, neptunium, and americium are extracted to the organic phase side, and fission product nuclides including rare earths are discharged as high-level waste liquid 26.

The reason for using an extractant having such an anion exchanger is that americium existing in a valence state of IV valence forms an anion complex having a negative charge with phosphotungstate ion. And VI
This is because uranium, plutonium, and neptunium existing in a valence state of valence form a complex with nitrate ions, and thus extraction by an anion exchanger is considered desirable.

The uranium, plutonium, neptunium and americium extracted into the organic phase by the extractor 11 as described above are sent to the washing device 12. And washing machine 1
After the nitric acid 22 is added and washed in 2, it is supplied to the distributor 13 and moves to the distribution step 5.

In the distribution step 5, a nitric acid aqueous solution 23 containing a reducing agent (for example, hydroxylamine nitrate, hydrazine, nitrous acid, iso-butyraldehyde, etc.) is supplied to and brought into contact with the solution in the distributor 13. Thus, while maintaining uranium at VI value, plutonium changes from VI value to III value, neptunium changes from VI value to V value, and americium changes to IV value.
Since the valence can be reduced from valence to valence III, the mixture 25 of plutonium, americium, and neptunium can be back-extracted while the uranium 24 remains in the organic phase of the solution.

According to this embodiment configured as described above, the same effect as that of the first embodiment can be obtained. In addition, there is an effect that reprocessing can be performed without adjusting the nitric acid concentration of the solution of the spent nuclear fuel at all.

In the above embodiment, potassium phosphotungstate was used as a complexing agent. However, the present invention is not limited to this, and other phosphates, heteropolyacid compounds and the like may be used. In these cases, a similar effect is obtained. Also,
When the amount of phosphotungstate added in the valence adjusting step 3 is insufficient for americium, americium having a valence of IV generates VI valence by a disproportionation reaction. Can be recovered by the following method. In this case, it is also possible to use the tributylphosphoric acid (TBP) described in the first and third embodiments, thereby reducing the amount of waste generated by the complexing agent. There is.

A sixth embodiment of the present invention will be described with reference to FIG. This embodiment is an embodiment in which an anion exchange method is applied in the co-decontamination step and the distribution step, and a lysis solution, a co-decontamination solution, and an eluate are continuously passed through several stages of ion exchange columns. Members and processes equivalent to those in the first to fifth embodiments are denoted by the same reference numerals.

FIG. 9 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to the present embodiment. The procedure of shearing by the shear 6, dissolution in the dissolving tank 7, denitration / valency adjustment in the denitration / valence adjusting tank 8, and removal of insoluble residue / precipitate in the precipitation removing device 14 are the same as those in the second embodiment. Description is omitted because there is.

In the precipitation removing device 14, the solution separated from the precipitate of zirconium molybdate is supplied to a supply tank 64.
Sent to 6. Then, from the supply tank 646, a certain amount of the solution is supplied by a valve operation to the ion exchange column 639-1,
639-2, 639-3.

Then, the ion exchange columns 639-1, 63
Among 9-2 and 639-3, 4 to 8 mol / l nitric acid as the co-decontamination liquid stored in the co-decontamination liquid storage tank 637 is passed by a valve in order from the one that has passed through the dissolution liquid. Thereby, a co-decontamination operation is performed and fission product nuclides are eluted. That is, since almost no fission product nuclides containing rare earths are adsorbed to the anion exchanger from the aqueous nitric acid solution, VI-valent actinoids having an adsorptivity to the ion exchanger, ie, uranium (VI) / plutonium (VI) ) ・ Americium (VI) ・ Neptunium (V
Fission product nuclides can be separated and eluted from I). This eluate is sent from the high-level waste liquid receiving tank 640 to the treatment system as a high-level waste liquid 626, where a process such as vitrification is performed.

Thereafter, the ion-exchange columns 639-1, 639 are used.
Among the 39-2 and 639-3, the minor actinoid eluent storage tank 63 is sequentially arranged from the one after the co-decontamination operation.
A few mol / l aqueous solution of nitric acid containing a reducing agent stored in 8 is passed through a valve. As a result, plutonium and americium are reduced to valence III, and neptunium is reduced to valence V, and ion-exchange columns 639-1, 639-2, 6
Eluted from 39-3, separated from uranium adsorbed on ion-exchange columns 639-1, 639-2, 639-3 retained as VI value and introduced into plutonium / neptunium / americium storage tank 625 Can be.

Then, the ion exchange columns 639-1, 639
Of the 39-2 and 639-3, the dilute nitric acid stored in the uranium eluent storage tank 647 is passed through the valve in the order from the end of the above separation, whereby the ion exchange columns 639-1, 639-2, U-valent uranium adsorbed on 639-3 is eluted and collected in the uranium receiving tank 624.

Further, the ion exchange columns 639-1, 639-2, and 639 after the uranium recovery operation are completed.
-3, the solution is received again from the supply tank 646, and the above operation is repeated to continuously reprocess the solution.

According to this embodiment configured as described above, the same effect as that of the first embodiment can be obtained. In addition to this, there is an effect that both co-decontamination and distribution steps can be performed with one column, and an effect that it is not necessary to readjust the acid concentration of the solution after adjusting the valence.

In the first to sixth embodiments, in the co-decontamination step 4, co-decontamination was performed by using a solvent extraction method or an ion exchange method. However, the present invention is not limited to this.
Co-decontamination may be performed by applying a known precipitation separation method. In this case, the same effect can be obtained. In the first to sixth embodiments, the dissolved solution of the spent nuclear fuel dissolved in the dissolving step 2 is used in the valence adjusting step 3.
After adjusting the valence of uranium, plutonium, and neptunium to VI valence and the valence of americium to IV valence or VI valence, in the co-decontamination step 4, uranium, plutonium, neptunium, and americium are converted into fission containing rare earth elements. The uranium is separated from the nuclides and, finally, in a partitioning process
By reducing plutonium and americium to valence III and neptunium to valence V while maintaining the valence, a mixture of americium, neptunium and plutonium was separated from uranium. On the other hand, by further applying the above idea, the dissolved liquid of the spent nuclear fuel dissolved in the dissolving step 2 is first treated with plutonium and americium while maintaining the uranium at a VI value, as in the distribution step. III value
By reducing neptunium to a valence of V, a mixture of americium / neptunium / plutonium and uranium are separated and uranium is recovered first, and then, in a valence adjustment step 3, the valence of plutonium and neptunium is converted to a valence of VI. , The valence of Americium is adjusted to IV valence or VI valence, and then, in the co-decontamination step 4, plutonium, neptunium and americium can be separated from fission product nuclides including rare earths. In this case, the same effects as those of the first to sixth embodiments can be obtained.

[0092]

According to the present invention, the minor actinoid is oxidized in the valence adjusting step in a state where uranium and plutonium coexist, so that the minor actinoid once taken out to the waste liquid side is oxidized and returned. Streaming is possible, and processing steps can be simplified. In the co-decontamination step, the rare earth element can be easily separated after being accompanied by the fission product nuclide or once with the mixture of uranium, plutonium, and minor actinoid. Therefore, the number of separation steps can be reduced by one, and the steps can be simplified. Further, the process can be simplified as compared with the case where the minor actinoid once transferred to the waste liquid side of the fission product nuclide is separated. And since the minor actinide with long half-life can be recovered together with plutonium in the distribution process, the recovery of minor actinoids can be performed while avoiding the purification of plutonium alone, and the handling burden is reduced compared to the case where pure plutonium is purified alone. And the danger of nuclear proliferation can be avoided. Also, recovery of uranium and plutonium useful as nuclear fuel and separation of long-lived minor actinoids from short-lived fission-produced nuclides can be performed in the same process, thereby simplifying waste disposal.

[Brief description of the drawings]

FIG. 1 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a first embodiment of the present invention.

FIG. 2 is a graph showing the dependence of the yield of americium on the concentration of a nitric acid solution and the concentration of ammonium peroxodisulfate.

FIG. 3 shows distribution coefficients of neptunium (VI), americium (VI) and neodymium (III) in tributyl phosphate (TB).
It is a figure which shows the dependence on P) density | concentration.

FIG. 4 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a second embodiment of the present invention.

FIG. 5 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a third embodiment of the present invention.

FIG. 6 shows the distribution coefficient of tributylphosphoric acid (TB) for neptunium (VI), americium (VI) and neodymium (III).
It is a figure which shows the dependence on P) density | concentration.

FIG. 7 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a fourth embodiment of the present invention.

FIG. 8 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a fifth embodiment of the present invention.

FIG. 9 is a conceptual diagram showing a procedure of a method for collecting spent fuel according to a sixth embodiment of the present invention.

[Explanation of symbols]

REFERENCE SIGNS LIST 1 shearing step 2 dissolving step 3 valence adjusting step 4 co-decontamination step 5 distribution step 6 shearing device 7 dissolving tank 8 denitration / valency adjusting tank 9 measuring tank 10 supply tank 11 extractor 12 washing machine 13 distributor 14 filtration treatment Apparatus 15 Offgas treatment system 16 Main exhaust tower 17 Solvent cleaning system 18 Formic acid 19 Ammonium peroxodisulfate 20 Concentrated nitric acid 21 Tributylphosphoric acid 22 Nitric acid as a cleaning solution 23 Nitric acid aqueous solution containing a reducing agent 24 Uranium 25 Mixture of plutonium, neptunium and americium 26 High-level waste liquid 28 Dissolver washing machine 29 Washing solvent treatment system 30 Washing solvent 33 Ozone 34 Potassium phosphotungstate 35 Trioctylamine 41 About 3 mol / l nitric acid 42 Extraction liquid 43 Washed extraction liquid 44 Water phase 45 Rare earth waste liquid 221 Di-isodecyl phosphate 2 22 3-6 mol / l nitric acid aqueous solution 327 Clarifier 331 Storage tank 412 Rare earth decontamination device 421 n-octyl-N, N-diisobutylcarbamoylmethylene phosphoric acid 422 Dilute nitric acid 532 Valence adjustment tank 624 Uranium receiving tank 625 Plutonium・ Neptunium / Americium receiving tank 626 High-level waste liquid 637 Co-decontamination liquid storage tank 638 Minor actinoid eluent storage tank 639-1 Anion exchange column 639-2 Anion exchange column 639-3 Anion exchange column 640 High-level waste liquid receiving tank 646 Supply tank 647 Uranium eluent storage tank

 ────────────────────────────────────────────────── ─── Continuing on the front page (72) Akira Sasahira, Inventor 7-2-1, Omika-cho, Hitachi City, Ibaraki Prefecture Energy Research Laboratory, Hitachi, Ltd. (72) Inventor Fumio Kawamura 7-2, Omika-cho, Hitachi City, Ibaraki Prefecture No. 1 Hitachi, Ltd. Energy Research Laboratory (56) References JP-A-8-114696 (JP, A) JP-A-7-246301 (JP, A) JP-A-7-159589 (JP, A) 4-218797 (JP, A) JP-A-7-280992 (JP, A)

Claims (10)

(57) [Claims] 1. A dissolving step of dissolving spent nuclear fuel into a solution containing uranium, plutonium, minor actinoids and fission product nuclides, and adjusting the valence of uranium and plutonium contained in the solution. A valence adjusting step, a co-decontamination step of separating valence adjusted uranium and plutonium from the fission product nuclide, and an element containing uranium and plutonium separated from the fission product nuclide, an element containing a minor actinoid And a distribution step of distributing to the element containing uranium, and the reprocessing method of the spent nuclear fuel, wherein in the valence adjustment step, while adjusting the valence of the uranium and plutonium to VI valence, the minor actinoid The valence is adjusted to at least one of IV valence and VI valence, and in the co-decontamination step, uranium having VI valence and VI And plutonium, the distribution target elements including a minor actinides is at least one of the IV-valent · VI value, separated from the fission products nuclide, in the distribution process, the uranium of the distribution object element VI
By reducing the minor actinoid to at least one of valence III and V and reducing the plutonium to valence III while maintaining the valence, the element to be distributed is converted to valence III / V
Minor actinides and / or II
A method for reprocessing nuclear fuel, comprising separating a mixture of I-valent plutonium and VI-valent uranium. 2. The method for reprocessing a spent nuclear fuel according to claim 1, wherein in the valence adjusting step, the valences of uranium and plutonium are adjusted to VI valence, and both valences of neptunium and americium are adjusted. A method for reprocessing spent nuclear fuel, which is adjusted to a VI value. 3. The method for reprocessing spent nuclear fuel according to claim 2, wherein in the valence adjusting step, peroxodisulfate, xenonate, bismuth (V), fluoroxysulfate are used as valence adjusters. A method for reprocessing nuclear fuel, comprising using at least one of salt and ozone. 4. The method for reprocessing spent nuclear fuel according to claim 3, wherein in the co-decontamination step, co-decontamination is performed in a state in which a stabilizer for a VI-valent actinoid ion is present. Characteristic method of reprocessing spent nuclear fuel. 5. The method for reprocessing a spent nuclear fuel according to claim 1, wherein in the valence adjusting step, the valence of the uranium and plutonium is adjusted to a VI valence, and the valence of neptunium is converted to a VI valence. A method for reprocessing spent nuclear fuel, comprising adjusting the valence of americium to IV valence. 6. A reprocessing method of a spent nuclear fuel according to claim 2 or 5, wherein in the valence adjustment step, as an oxidizing aid promotes oxidation for valence adjustment, valence I
A method for reprocessing a spent nuclear fuel, comprising using a compound of soluble silver and an acid having at least one of valences, valences of II and III. 7. The method for reprocessing spent nuclear fuel according to claim 1, wherein in said partitioning step, at least one of a solvent extraction method, an ion exchange method, and a precipitation separation method is used. A method for reprocessing a nuclear fuel, comprising reducing a minor actinide to at least one of III valence and V valence while reducing uranium to VI valence and reducing plutonium to III valence. 8. The method for reprocessing spent nuclear fuel according to claim 7 , wherein, in said distributing step, hydroxylamine nitrate, hydrazine, nitrous acid, and iso-butyraldehyde are used as valence adjusting agents for distribution. At least one of them
A method for reprocessing nuclear fuel, characterized by using one of the following. 9. A dissolving step of dissolving spent nuclear fuel into a solution containing uranium, plutonium, minor actinoids, and fission product nuclides, and adjusting the valence of uranium and plutonium contained in the solution. A valence adjusting step, a co-decontamination step of separating valence adjusted uranium and plutonium from the fission product nuclide, and an element containing uranium and plutonium separated from the fission product nuclide, an element containing a minor actinoid And a distribution step of distributing the element containing uranium to the spent nuclear fuel, wherein in the valence adjustment step, the minor actinoid and plutonium are adjusted to the same valence. How to reprocess spent nuclear fuel. 10. A dissolving step of dissolving spent nuclear fuel into a dissolving solution containing uranium, plutonium, minor actinoids and fission product nuclides, and adjusting the valence of elements contained in the dissolving solution. A reprocessing method for spent nuclear fuel, comprising: a co-decontamination step of separating an element to be separated from the fission product nuclide, wherein the uranium is provided after the dissolving step and before the valence adjusting step. Is maintained at VI valence, the valency of the minor actinoid is maintained at at least one of III valence and V valence, and the plutonium valence is reduced to III valence, thereby obtaining a mixture of the plutonium and the minor actinoid. Further comprising a uranium recovery step of separating and recovering the VI-valent uranium from the uranium, wherein in the valence adjustment step, the valence of the plutonium VI
While adjusting the valence of the minor actinoid to at least one of IV valence and VI valence, in the co-decontamination step, the VI valent plutonium and the IV
A method for reprocessing spent nuclear fuel, comprising separating an element to be distributed containing at least one of a valence and a VI valence and a minor actinoid from the fission product nuclide.