JPH0493799A - Solidification treatment of radioactive waste - Google Patents

Abstract

PURPOSE:To enable producing a solidified body which is faborable for safety evalution by adding a quick lime or a slaked lime to powder waste body in order to solidify it, after addition of carbonate ion to the extent that exceeds the saturating solubility of a calcium carbonate, to the powder waste body. CONSTITUTION:Soluble carbonate ion having been existing in a condition of a sodium carbonate in dry powder waste body, dissolves, for the first time, as a sodium carbonate into water, when a plastic solidified body gets in contact with the water. At the same time, a quick lime or a slaked lime dissolves into water, reacts with the sodium carbonate to change itself to a calcium carbonate and to an almost insoluble condition into water. However, since the amount of the carbonate ion containing a radioactive nuclide, C-14, is extremely small, the carbonate ion reaches its solubility and there is some possibility for the ion to dissolve into water. Accordingly, a sodium carbonate is added to the extent well to exceed the solubility of the calcium carbonate, and therewith the carbonate ion containing the nuclide, C-14, co-percipitates with the added carbonate ion and can be changed to an almost insoluble one into water. Finally, by plastic solidification, leaching-out of the carbonate ion can be remarkably reduced.

Description

[Detailed Description of the Invention] [Objective of the Invention] (Field of Industrial Application) The present invention is aimed at converting radioactive waste generated from nuclear power plants using boiling water reactors (hereinafter referred to as BWR) into plastics. Concerning a method for solidifying radioactive waste.

(Prior art) Conventionally, radioactive waste fluid generated from BWR nuclear power plants is heated and concentrated, and in order to further reduce the amount generated, it is dried and stabilized with a plastic material such as unsaturated polyester resin. It has been solidified into plastic. This radioactive waste liquid treatment method is widely used, especially in BWR nuclear power plants, because of its high volume reduction rate. For example, in the case of concentrated waste liquid generated from a BWR nuclear power plant, the amount of waste generated is reduced to 1/6 to 1/10 compared to the case where the waste liquid is solidified with cement.

Incidentally, recently, as the final disposal of low-level radioactive waste generated from nuclear power plants becomes more concrete, safety evaluations are being conducted. Currently, in Japan, such low-level radioactive waste is scheduled to be disposed of in shallow geological formations.
It has become clear that long-lived radionuclides such as C-14 have an important meaning in safety evaluation, and their behavior at the disposal site plays an important role in evaluating the environmental impact Iip. C-14 is 0-17 (n, α
) It is produced by a reaction, and is mixed into radioactive waste fluid from the reactor system through the turbine system, and its inclusion in the waste fluid is unavoidable. It is thought that C-14 is normally oxidized by oxygen generated by radiolysis of reactor water and exists as carbonate ions. These carbonate ions are captured by the anion exchange resin in the condensate purification system of the reactor turbine system, and when this resin is regenerated with caustic soda, it is incorporated into the recycled waste liquid in the form of sodium carbonate, and becomes the main component of the concentrated waste liquid. Contaminated with some sodium sulfate.

The recycled waste liquid is concentrated, solidified, and processed into radioactive waste packages that can be finally disposed of.

Currently, these waste liquids are stabilized by the above-mentioned plastic solidification method. After the concentrated waste liquid is dried, sodium sulfate powder, which is the main component of the waste liquid, is immobilized using a plastic material, and C-14 dissolved in the form of sodium carbonate is also incorporated into this powder. By the way, the solubility of sodium carbonate is 7.1, so when plastic solidified bodies or pellets come into contact with water after final disposal, water-soluble sodium sulfate is dissolved from these solidified bodies, and along with this, carbonate ions are dissolved. The C-14 contained therein also dissolves. That is, C-14 does not remain in the plastic solidified body or the pellet solidified body, but dissolves outside the solidified body. Therefore, C-14 does not remain in the solidified body,
Transfers to the surrounding aqueous phase. In the safety evaluation test of solidified bodies, the ratio of the amount remaining in the solidified body and the amount transferred to the aqueous phase is expressed as a partition coefficient, and the amount remaining in the solidified body is large and the amount transferred to the aqueous phase is small. A state in which the distribution coefficient is large is advantageous in terms of safety evaluation.

(Problem to be Solved by the Invention) However, the distribution coefficient of the plastic solidified body is estimated to be almost zero, which is not preferable from a safety evaluation point of view. In a cement solidified body, C-14 is incorporated into the solidified body as an insoluble component such as calcium carbonate, so that almost no C-14 is transferred to the surrounding aqueous phase. Therefore, the solidified cement has a large distribution coefficient, which is extremely preferable in terms of safety evaluation.

By the way, although the plastic solidification method has a higher volume reduction property than cement solidification, there is an undesirable problem in that the distribution coefficient of long-lived nuclides such as C-14 is small.

The present invention was made to solve the above-mentioned problems, and it maintains the excellent volume reduction properties of the plastic solidification method.
It is an object of the present invention to provide a method for solidifying radioactive waste that has a large distribution coefficient for long-lived nuclides such as -14 and is advantageous in terms of safety evaluation.

[Structure of the invention]

(Means for Solving the Problems) The present invention involves adding carbonate ions exceeding the saturation solubility of calcium carbonate as sodium carbonate powder to radioactive waste powder obtained by drying radioactive concentrated waste liquid, and then adding quicklime or slaked lime. Next, an unsaturated polyester resin, a polymerization initiator, and an accelerator are added, mixed, and polymerized and cured.

(Function) When the concentrated waste liquid is subjected to plastic solidification treatment, the soluble carbonate ions that existed in the form of sodium carbonate in the dry waste powder are converted into solidified plastic when the solidified plastic comes into contact with water. When water enters, it initially dissolves in the water as sodium carbonate, but at the same time, quicklime or slaked lime dissolves in the water, reacts with the sodium carbonate, and is converted into calcium carbonate, which becomes almost insoluble in water. By the way, the solubility of calcium carbonate is 0.001
4, which is about 1/50,000 times smaller than the solubility of sodium carbonate. However, since the amount of carbonate ions containing C-14 is extremely small, even with such pretreatment, the carbonate may reach its solubility and dissolve into water. Therefore, it is necessary to add sodium carbonate in advance to an extent that sufficiently exceeds the solubility of calcium carbonate, and by doing so, carbonate ions containing C-14 will co-precipitate with the added carbonate ions, and will hardly dissolve in water. can be made into something. If the plastic is solidified in such a state, the elution of carbonate ions from the solidified body and the transfer to the aqueous phase will be extremely small, and therefore the partition coefficient will be large.

(Example) 25% of sodium sulfate dissolved in carbonate containing C-14
% aqueous solution was prepared as a simulated concentrated waste liquid generated from a BWR nuclear power plant, which was dried and powdered.

A plastic solidified body was manufactured using the simulated concentrated waste liquid dry powder prepared as described above. For the plastic solidified body, an unsaturated polyester resin equivalent to that used in the solidification process of the BWRJJK power plant was used as the plastic material. First, an organic peroxide as a polymerization initiator was added to the unsaturated polyester resin in an amount of about 1% based on the resin. Thereafter, dry powder of the dried simulated concentrated waste liquid was mixed at a ratio of resin to powder of 40:60. Sodium carbonate powder was added to this so that the concentration was 0.4% based on the simulated radioactive waste dry powder. Furthermore, 0.8% of slaked lime was added to the simulated concentrated waste liquid dry powder and mixed uniformly. After the mixing was completed, a polymerization accelerator organometallic compound was added at a rate of 0.5% based on the unsaturated polyester resin. After about one week, the unsaturated polyester resin became a fully cured plastic solid. This solidified material was shaved to produce a powder of the solidified material, which was immersed in an aqueous phase to measure the distribution coefficient. The measurement results are as follows.

(7)C-14 concentration in aqueous phase: 3.2 x 10-μ
(7)C-14 concentration in Ci/*R solidified body: 2.6
10- μCi/wlfi distribution coefficient (C-
14 concentration/C-14 concentration in aqueous phase) = 812 mffi/
g As described above, this example shows a large distribution coefficient due to a significant decrease in the amount of C-14 transferred into the aqueous phase. In other words, C-1
Since the retention capacity of No. 4 is high, the solidified material is advantageous in terms of safety evaluation. Note that in the conventional example, the distribution coefficient was as small as almost zero, and it is recognized that the distribution coefficient was significantly improved by this embodiment.

〔Effect of the invention〕

According to the present invention, even if a situation such as contact with water occurs after the solidified radioactive waste is finally disposed of, the carbonate ions in the solidified waste will not migrate to the aqueous phase. Therefore, C-14 contained in carbonate ions also remains in the solidified body without transferring to the aqueous phase, making it possible to produce a plastic solidified body with a large distribution coefficient and extremely advantageous in terms of safety evaluation. Of course, the radioactive concentrated waste liquid pretreated in this way is dried and processed into a solidified product with excess moisture completely removed.

The conventional volume reduction properties can be maintained, and the advantage of being a solidification treatment method that generates a small amount of waste can be maintained.

(8733) Agent: Yoshiaki Inomata, patent attorney (and 1 others)
given name)

Claims (1)

[Claims] Carbonate ions exceeding the saturation solubility of calcium carbonate are added as sodium carbonate powder to radioactive waste powder obtained by drying radioactive concentrated waste liquid, then quicklime or slaked lime is added, and then unsaturated polyester resin, polymerization initiator and A method for solidifying radioactive waste, characterized by adding an accelerator, mixing, and polymerizing and curing.