JPH07280985A - Method for co-extraction of uranium, plutonium and neptunium - Google Patents

Abstract

PURPOSE:To extract and collect neptunium coexisting in a nitric acid solution of spent nuclear fuel, along with uranium and plutonium, in the co- decontamination process employing purex wet process. CONSTITUTION:A nitric acid solution of spent nuclear fuel, wherein at least 10wt.% of total plutonium exists in the form of hexavalent plutonium, is brought into contact with a TBP extraction agent. When the amount of hexavalent plutonium is not sufficient in the nitric acid solution, the nitric acid solution is heated in the range between 70 deg.C and the boiling point of the nitric acid solution thus achieving a predetermined amount of hexavalent plutonium.

Description

Detailed Description of the Invention

[0001]

TECHNICAL FIELD The present invention relates to a purex wet reprocessing method for reprocessing spent nuclear fuel, in which neptunium, which is one of the transuranium elements present in the spent nuclear fuel, is added together with uranium and plutonium. It relates to a method of extracting and recovering.

[0002]

BACKGROUND OF THE INVENTION The main purpose of the Purex wet reprocessing method is to separate and recover uranium and plutonium contained in spent nuclear fuel from fission products. In this case, neptunium, which is one of the transuranium elements also contained in the spent nuclear fuel, is not targeted for recovery, and part of it is separated from fission products along with uranium and plutonium. Alternatively, it is separated from uranium and plutonium together with fission products and discharged to the high-level liquid waste side.

In carrying out the Purex wet reprocessing method, the spent nuclear fuel is first dissolved in nitric acid to adjust the solution composition, and then the mixture is set in a multistage extractor such as a mixer / settler, a centrifugal extractor or a pulse column. % TBP
Contact with (tributyl phosphate) extractant. As a result, uranium and plutonium in the nitric acid solution of spent nuclear fuel are extracted in the extractant phase (organic phase), and fission products remain in the nitric acid solution phase (aqueous phase) and are removed as high-level waste liquid.
This step is called a co-decontamination step in the Purex wet reprocessing method. However, since neptunium is a long-lived nuclide, it is a problem in waste management to discharge this together with fission products to the high-level liquid waste side.

The above co-decontamination process will be further described with reference to FIG. A multi-stage extractor is generally composed of an extraction stage group or extraction section and a cleaning stage group or cleaning section. When a liquid feed (containing uranium, plutonium, neptunium, and fission products) consisting of a nitric acid solution of spent nuclear fuel is introduced into the feed stage in the middle part of the extractor, the TBP extractant is countercurrently contacted in the extractor to feed the liquid. Uranium (adjusted to easily extractable hexavalent) and plutonium (adjustable to extractable tetravalent) are extracted by the extractant, and fission products remain in nitric acid solution and discharged as high-level waste liquid. To be done. The extractant loaded with uranium and plutonium is further sent to the washing section and comes into countercurrent contact with the washing solution (nitric acid) to further remove impurities such as fission products.

The behavior of neptunium in this co-decontamination process is not stable and is discharged to the high-level waste liquid side,
It is mixed in the uranium / plutonium loading extractant side. Neptunium mixed in the loaded extractant side becomes an impurity, and its handling becomes troublesome. This is because the valence of neptunium in the nitric acid solution maintains the equivalence of pentavalent and hexavalent, and depending on the condition, the pentavalent or hexavalent may increase. That is, since the pentavalent neptunium is not extracted by the extractant, it is discharged to the high-level waste liquid side, and the hexavalent neptunium is extracted by the extractant together with uranium and plutonium.

The equilibrium relationship between pentavalent and hexavalent neptunium in the extractor of the extractor can be expressed by the following equation. Since neptunium hexavalent (NpO ₂ ²⁺ ) is extracted to the extractant, the formula (1) proceeds to the right side, and therefore, the neptunium pentavalent hexavalent oxidation occurs in the extraction part.

The neptunium hexavalent extracted in the extractant in the extraction section of FIG. 1 is then carried to the washing section, and nitrite (NO ₂ ) generated by decomposition of nitric acid in the extraction section is also extracted in the extractant. Carried to the cleaning department. Therefore, neptunium hexavalent and nitrous acid coexist at a certain concentration in the cleaning section,
The reduction reaction of neptunium hexavalent to pentavalent by nitrous acid occurs, and equation (1) proceeds to the left side, and neptunium pentavalent is discharged to the high-level waste liquid side.

In order to prevent the reduction of neptunium hexavalent to pentavalent in the cleaning section, the amount of nitrous acid carried to the cleaning section should be minimized. Therefore, in the method proposed in Japanese Patent Publication No. 52-26318, a method of introducing hydrazine, for example, as a nitrous acid decomposing agent from the feed liquid supply stage into the extractor cleaning section in order to reduce the nitrite concentration in the cleaning section. Has been adopted. As a result, the concentration of nitrite in the cleaning section is reduced to suppress the reduction of heptonite to pentavalent neptunium, and the recovery rate of neptunium existing in the liquid supply to the extractant is increased to about 90%. There is.

[0009]

However, in the above-mentioned prior art, hydrazine, which is a nitrous acid decomposing agent, must be supplied to the cleaning section while delicately adjusting the nitrous acid concentration that changes intricately depending on the conditions. However, it is not always a satisfactory method for practical use. For example, when hydrazine is supplied in a large excess, plutonium is reduced to trivalent due to the reducing action of hydrazine, plutonium cannot be efficiently recovered, and the original purpose of the Purex wet reprocessing method is to reduce plutonium. It will also have an adverse effect on recovery.

Therefore, the present invention provides a spent nuclear fuel in the co-decontamination step of the Purex wet reprocessing method, which does not require the delicate adjustment of the nitrite concentration using a nitrite decomposing agent as in the prior art. The purpose of the present invention is to provide an improved method capable of efficiently extracting and recovering neptunium coexisting in the nitric acid solution together with uranium and neptunium.

[0011]

[MEANS FOR SOLVING THE PROBLEMS] That is, the method for co-extracting uranium, plutonium and neptunium according to the present invention comprises contacting a nitric acid solution of spent nuclear fuel with a TBP extractant to extract neptunium in the nitric acid solution together with uranium and plutonium. In the method of extracting into a phase, a nitric acid solution in which at least 10% by weight or more of the total amount of plutonium contained in the nitric acid solution of the spent nuclear fuel is present as plutonium hexavalent is contacted with a TBP extractant. is there.

Plutonium in the nitric acid solution of spent nuclear fuel is present in a large excess as compared with neptunium. For example, in the case of a typical liquid metal fast breeder reactor, Pu / Np = about 4
00, Pu / Np = for a typical pressurized water reactor
It is about 12. Further, the redox potential of neptunium and the redox potential of plutonium have very similar values. Therefore, under the condition that at least 10% by weight or more of the total amount of plutonium in the nitric acid solution is oxidized to be hexavalent, it means that substantially all the coexisting neptunium is oxidized to be hexavalent.

On the other hand, the valence of plutonium in the nitric acid solution has an equilibrium relationship with the valence of plutonium as shown in the following equation (2). From this equation (2), plutonium hexavalent is 4
It can be seen that the value is reduced. Therefore, in a system in which a large excess of plutonium hexavalent is present, the plutonium hexavalent acts as a decomposing agent for nitrous acid. According to various experiments, the present inventors have found that if at least 10% by weight or more of the total amount of plutonium in the nitric acid solution is present as a hexavalent plutonium, it acts as a decomposing agent for nitrous acid as shown in formula (2). The present invention has been completed by finding that the amount of hexavalent plutonium is sufficient, and therefore the reduction reaction of neptunium hexavalent to pentavalent by nitrous acid can be reliably prevented.

Conventionally, plutonium in a nitric acid solution is brought into contact with an extracting agent after being adjusted to be tetravalent by using a valency adjusting agent such as sodium nitrite. The part is contacted with the extractant in a hexavalent state. Plutonium hexavalent has a lower extractability than plutonium tetravalent, but according to the present invention, since plutonium hexavalent is reduced to tetravalent by nitrous acid according to the formula (2), plutonium has a high extraction efficiency. Can be maintained.

When the nitric acid solution of the spent nuclear fuel contains more than a predetermined amount of plutonium hexavalent, it is possible to use this nitric acid solution as a liquid supply to the extractor without adjusting the valence of plutonium. it can. On the other hand, when the plutonium hexavalent value in the nitric acid solution of the spent nuclear fuel is less than 10% of the total amount of plutonium, the nitric acid solution can be easily oxidized to the plutonium hexavalent value by heating at 70 ° C. or higher. Almost no oxidation occurs at temperatures below 70 ° C. The higher the heating temperature, the faster the oxidation rate, which is advantageous in terms of processing capacity.However, when the heating temperature is higher than the boiling point of the nitric acid solution, the material corrosion rate of the heating container increases and the life of the heating container is shortened. Connect From such a viewpoint, it is preferable that the heating temperature be within a temperature range from 70 ° C. to the boiling point of the nitric acid solution. When a nitric acid solution in which the total amount of plutonium is hexavalent by heating is left at room temperature,
As shown in the graph of FIG. 2, due to the influence of nitrous acid generated by the decomposition of nitric acid, the hexavalent plutonium in the nitric acid solution is spontaneously reduced and the ratio thereof decreases. Therefore, it is necessary to perform the extraction operation with the TBP extractant before the ratio of the hexavalent plutonium in the nitric acid solution becomes lower than 10%.

[0016]

EXAMPLES From the eighth stage of a mixer-settler type multi-stage extractor having a total of 16 stages consisting of an eight-stage extraction section and an eight-stage washing section, a nitric acid solution A having the composition shown in Table 1 to D (simulated solution of spent nuclear fuel nitric acid solution) was supplied as a feed liquid, and was brought into countercurrent contact with the TBP extractant (TBP diluted to 30% with n-dodecane) supplied from the first-stage extraction section. On the other hand, a cleaning liquid (3M nitric acid) was supplied from the 16th stage cleaning section to make countercurrent contact with the loaded extractant sent from the extraction section to the cleaning section.

The feed liquids A, B, and C in Table 1 were prepared by heating a nitric acid solution to about 80 ° C. to oxidize the total amount of plutonium contained therein to hexavalent, and then allowed to stand as it was to spontaneously reduce the hexavalent plutonium. Plutonium 6
It was prepared so that the residual value ratio was 40% or 20%.

The feed solution D was prepared by heating a nitric acid solution to about 80 ° C. to oxidize the total amount of plutonium contained therein to hexavalent, and then adding sodium nitrite (NaNO ₂ ) as a valence adjuster to add hexavalent plutonium. Was prepared by reducing the total amount of the above to tetravalent.

[0019]

TABLE 1 supply fluid composition A B C D Pu (g / l) 39 39 45 20 Pu 6+ / T-Pu 40% 20% 20% 0% U (g / l) 140 140 110 81 HNO 3 (M ) 5.00 5.00 5.61 2.92 Np (mg / l) 35 35 63 63 28 Np (%) 100 100 100 100 Flow rate (ml / h) 100 100 100 100 150

Table 2 shows the analysis results of the composition of the loaded extractant obtained from the washing section of the 16th stage by the above operation in the multistage extractor, and Table 2 shows the composition of the high-level waste liquid discharged from the extracting section of the 1st stage. The analysis results are shown in Table 3, respectively.

[0021]

TABLE 2 loaded extraction agent composition A B C D Pu (g / l) 12 12 20 21 U (g / l) 43 44 28 67 HNO 3 (M) - - - - Np (mg / l) 12 12 25 3.4 Np (%) 100 100 100 14.1 ^a) Flow rate (ml / h) 300 300 250 250 170

[0022]

Table 3 High-level liquid waste composition A B C D Pu (g / l) ND b) ND ND ND U (g / l) ND ND ND ND HNO 3 (M) 4.95 5.09 4.55 3. 26 Np (mg / l) ND ND ND 9.6 Np (%) 0 0 0 46.7 ^a) Flow rate (ml / h) 196 196 200 205 Note a) Mass balance cannot be achieved due to accumulation in the extractor. b) ND: Not detected.

As can be seen from Table 1, Table 2 and Table 3,
Liquid A with plutonium hexavalent coexisting with neptunium
In B and C, no accumulation of neptunium in the extractor was observed, and it can be seen that the total amount of neptunium was extracted by the extractant together with uranium and plutonium, and was not discharged to the high-level waste liquid side. On the other hand, in the liquid feed D, since the liquid feed was brought into contact with the extractant under the condition that the hexavalent plutonium did not exist in the liquid feed, neptunium was reduced to pentavalent nitrite by the decomposition of nitric acid in the extractor. I will end up. As a result, 1 of neptunium
Only 4.1% is extracted by the extractant, 46.7% is discharged to the high-level waste liquid side, and the rest is accumulated in the extractor.

[0024]

As described above, according to the present invention,
Neptunium is effective together with uranium and plutonium in countercurrent contact between the nitric acid solution and the TBP extractant by allowing more than a predetermined amount of plutonium existing in the nitric acid solution of spent nuclear fuel to coexist with neptunium in a plutonium hexavalent state. Moreover, it can be stably extracted by the extractant and can be recovered at a high recovery rate.

As a result, since neptunium is not substantially discharged to the high-level liquid waste side, 60 to 60% of neptunium is discharged.
It is possible to reduce the problem of waste management that has been posed in the co-decontamination process of the conventional Purex wet reprocessing method in which 70% or more is discharged to the high-level liquid waste side.

Further, in order to adjust the amount of plutonium hexavalent in the nitric acid solution, it suffices to simply heat the nitric acid solution. Therefore, a valence adjuster or the like conventionally used for adjusting the plutonium valence is used. You don't have to.

[Brief description of drawings]

FIG. 1 is an explanatory view showing the concept of a co-decontamination process of a Purex wet reprocessing method.

FIG. 2 is a graph showing spontaneous reduction of plutonium hexavalent in a nitric acid solution over time.

 ─────────────────────────────────────────────────── ─── Continuation of the front page (72) Inventor Junya Tomohiro 4-3 Muramatsu, Tokai-mura, Naka-gun, Ibaraki Prefecture Inside the Tokai Plant, Reactor and Nuclear Fuel Development Corporation (72) Akio Togashi, Tokai-mura, Naka-gun, Ibaraki Prefecture Muramatsu No. 4 33 Tokai Plant, Reactor and Nuclear Fuel Development Corp.

Claims (3)

[Claims] 1. A method of contacting a nitric acid solution of a spent nuclear fuel with a TBP extractant to extract neptunium in the nitric acid solution into an extractant phase together with uranium and plutonium, and a total amount of plutonium contained in the nitric acid solution of the spent nuclear fuel. A method for coextracting uranium, plutonium and neptunium, which comprises contacting a nitric acid solution containing at least 10% by weight of plutonium as hexavalent with a TBP extractant. 2. A method of contacting a nitric acid solution of spent nuclear fuel with a TBP extractant to extract neptunium in the nitric acid solution into an extractant phase together with uranium and plutonium, by heating the nitric acid solution of the spent nuclear fuel. A coextraction method for uranium, plutonium and neptunium, characterized in that at least 10% or more by weight of plutonium contained in the solution is plutonium hexavalent, and the nitric acid solution is brought into contact with a TBP extractant. 3. The co-extraction method according to claim 2, wherein the nitric acid solution is heated within a temperature range from 70 ° C. to the boiling point of the nitric acid solution.