A PROCESS FOR TREATMENT OF A SPENT, RADIOACTIVE, ORGANIC ION EXCHANGE RESIN

Abstract: A spent, radioactive, organic ion exchange resin is converted into a stable inorganic product having a considerably reduced volume in the following way. The radioactive ion exchange resin is mixed with a salt and an inorganic sorbent for radioactive nucleides liberated by the salt, the mixture is dried and incinerated, whereupon the ash is solidified in cement.

Description: A PROCESS FOR TREATMENT OF A SPENT, RADIOACTIVE, ORGANIC ION EXCHANGE RESIN TECHNICAL AREA The present invention relates to a process for the treatment of a spent, radioactive, organic ion exchange res to reduce the volume thereof and to obtain a stable final product. In this context ion exchange resin primarily mean a cationic exchange resin but also an anionic exchange res and an exchange resin of the mixed bed type, containing cation exchanger as well as anion exchanger, can be advant geously treated in accordance with the invention. The inve tion primarily relates to the treatment of such ion exchan resins which have been utilized to purify cooling water in nuclear reactor, and the water in a pool for the storage o spent nuclear fuel. TECHNICAL BACKGROUND It is previously known to solidify a spent ion excha resin in cement or bitumen. However, by such a measure the volume is heavily increased. Furthermore, in the case of solidification in cement, the stability against leaching i not very good. In the case of solidification in bitumen th fire hazards thereof is a problem. Moreover, it is previously known, for instance from Swedish patent specification No. 8101801-2, that the volum of a spent ion exchange resin can be reduced by an incine¬ ration thereof. According to said Swedish patent specifica ion the incineration residue is then heated to sintering or melting, a stable product being obtained thereby. The measure of cementing the incineration residue has been con sidered improper due to the bad stability against leaching which has been observed when solidifying a non-incinerated ion exchange resin in cement. fu OroMPI Λr- WHO DISCLOSURE OF THE INVENTION It has now been found that in an unexpectedly simple it is possible to reduce the volume of the spent ion exch resin as well as to prepare a cement matrix wherein the ra active nucleides are bound in a stable way. The process according to the invention is characterized by mixing the ion exchange resin partly with a salt, to liberate radioac substances from the ion exchange resin, partly with an inorganic sorbent for the radioactive substances thus libe ted, then drying and incinerating the mixture, and solidi¬ fying in cement the residue from the incineration. The salt may be added to the aqueous ion exchanger i a solid form or as an aqueous solution thereof. The salt i preferably added in such a quantity that the ion exchanger will be saturated. The cation of the salt should effective elute active ions, such as Cs-ions, wich are sorbed on the ion exchanger. In order to obtain such an elution it is possible to utilize several common water-soluble salts, suc as calcium nitrate or aluminium nitrate. However, according to the invention it is preferable to use water-soluble salts, the anions of which tend to liberate active nucleides, such as cobolt, zinc, through the formation of complexes, for instance salts of phosphor acid, citric acid, tartaric acid, oxalic acid, formic acid propionic acid. It has turned out that such complex-formin anions do not disturb the subsequent process steps, i.e. the incineration and cementation operations, and that said organic acids are eliminated in the incineration step. As cations of the salt calcium and aluminium are preferred. These salts are conducive to a favourable course of incine ration. The explanation thereto seems to be that after their sorption on the ion exchanger the salts make said ion exchanger rather heavy, which facilitates the incinera ion. Furthermore, these salt reduce the tendency to an agg.o eration of the ion exchange resin grains, which results i a larger contact surface towards the incineration air and more rapid incineration. Salts of calcium and aluminium % make the incineration residue more compatible with the cement matrix, and accordingly the solidification in ceme will be facilitated. The inorganic sorbent should be added in such an am that it completely sorbs the liberated radioactive nuclei Preferably the sorbent has a particle size of 10-100 μm. During the incineration operation the sorbent will retain radioactive nucleides, such as Cs-137, by converting them into stable compounds having low vapour pressures at high temperatures. Furthermore the sorbent imparts to the fina product a good stability against leaching of radioactive nucleides from the cement matrix, which effect is especia pronounced for Cs-137. As said sorbent we prefere to utilize titanates or titanium hydroxide, zirconates or zirconium hydroxide or zirconium phosphate, aluminates or aluminium hydroxide, alumino silicates such as bentonite or a natural or synthetic zeolite, or a mixture of two or more of these sorbents. The ion exchange resin, the salt and the sorbent ar preferably admixed at a temperature of 20-70 C, and the aquous admixture is preferably dried at 90-120 C. The dried admixture is preferably incinerated at 500-900 C, preferably at about 800°C, suitably in air that has been enriched to an oxygen content of 30-40 % by volume. The residue from the incineration is mixed with cement and water. The water content of the mixture is preferably bet 10 and 20 % by weight. The percentage of the residue from incineration should be at most 120 _ of the weight of the cement. In connection with the invention cement preferabl means Portland cement, but also similar aqueous-hardening binders The cement mixture is now cast in a mould, wherei it is allowed to harden, and the hardened body is allowed to dry. Our examinations show that the volume of the final end product can be reduced up to 1/10 as compared to a direct solidification of a spent ion exchange resin in cement. It has also been found that the stability against SN leaching is increased at least ten times as compared to sa direct cementation . EXAMPLE A spent radioactive organic ion exchange resin conta ed inter alia 10 kBq of Cs-137 per gram of resin. The resi had a dry solidscontent of 50 by weight and was of the mixed-bed type, the ratio of cationic exchanger:anionic exchanger being 1:1. 100 grams of said resin were mixed with 25 grams of calcium formate and 4 grams of bentonite. The mixture was dried at 110°C and incinerated at 700°C in air that had been enriched on oxygen. An incineration residue of 15 grams was then obtained. This was mixed with 15 grams° ortland cement and 6 grams of water and from the 3 mixture there was cast a cube having a volume of 20 cm . After said cube had hardened leaching tests showed that Cs-137 was leached at room temperature with a rate of abou 10 "5 g/cm 2 -d.

Claims: CLAIMS 1. A process for treatment of a spent,radioactive, organic icn exchange resin to reduce the volume thereof an to obtain a stable end product, characterized by mixing th ion exchange resin with a salt, to liberate radioactive substances from said ion exchange resin, as well as with a inorganic sorbent for the radioactive substances thus libe rated, then drying and inciderating said mixture and soli¬ difying the residue from the incineration in cement. 2. A process according to claim 1, characterized i that the salt is added in such a quantity that the ion exchange resin will be essentially saturated. 3. A process according to any one of claims 1 and 2, characterized in that the salt is a salt of aluminium o calcium. 4. A process according to any one of claims 1 and 2, characterized in that the salt is a salt of phosphoric acid, citric acid, tartaric acid, oxalic acid, formic aci or propionic acid. 5. A process according to any one of the preceding claims, characterized in that the sorbent is a titanate or titanium hydroxide, a zirconate or a zirconium hydroxide o zirconium phosphate, an aluminate or an aluminium hydroxid an alumino silicate such as bentonite or a natural or synthetic zeolite, or a mixture of two or more of these sorbents. 6. A process according to any one of the preceding claims, characterized in that the dried mixture is incine¬ rated at a temperature of 500-900 C. 7. A process according to claim 6, characterized i that the dried mixture is incinerated in oxygen-enriched air. OMPI e/A * ,. WIPO

Inventors: HULTGREN AOKE VALDEMAR, HULTGREN, AOKE, VALDEMAR
Applicants: STUDSVIK ENERGITEKNIK AB
Application year: 1984
Main classification: G21F9
IPC codes: G21F9