Thorex solvent-extraction process

Np, and fission products. The Thorex solvent extraction process is generally used to reprocess spent Th-based fuels. As in the Purex process, the solvent is TBP diluted in an appropriate mixture of aliphatic hydrocarbons. Figure 12.9 shows the Thorex process flow sheet used by Kuchler et al. [41] for reprocessing high-burn-up thorium fuel. 

The residue of BISO particles is dissolved in mixed HNO3 and HF and then separated by the Thorex solvent extraction process (Chap. 10) into a decontaminated U-rich uranium fraction, a thorium fraction containing 1.9-year radioactive Th, and fission-product wastes. 

A small bleed stream is concentrated in the slurry letdown system and delivered to a chemical processing plant where the uranium and thorium are recovered by a thorex solvent extraction process. The chemical plant is designed for a small throughput and low over-all decontamination factors. Although the rates of flow to the auxiliary systems are small compared with the 18,000,000 Ib/hr rate of circulation in the primary system, these auxiliary systems contribute the major part of the complexity of the plant and a large fraction of its cost. 

Thorex [Thorium extraction] A process for separating the products from the nuclear breeder reaction in which uranium-233 is produced by the neutron bombardment of thorium-232. It uses solvent extraction into tri-n-butyl phosphate. Developed at the Oak Ridge National Laboratory, TN, in the early 1960s. See also Butex, Purex, Redox. 

The uranium and thorium ore concentrates received by fuel fabrication plants still contain a variety of impurities, some of which may be quite effective neutron absorbers. Such impurities must be almost completely removed if they are not seriously to impair reactor performance. The thermal neutron capture cross sections of the more important contaminants, along with some typical maximum concentrations acceptable for fuel fabrication, are given in Table 9. The removal of these unwanted elements may be effected either by precipitation and fractional crystallization methods, or by solvent extraction. The former methods have been historically important but have now been superseded by solvent extraction with TBP. The thorium or uranium salts so produced are then of sufficient purity to be accepted for fuel preparation or uranium enrichment. Solvent extraction by TBP also forms the basis of the Purex process for separating uranium and plutonium, and the Thorex process for separating uranium and thorium, in irradiated fuels. These processes and the principles of solvent extraction are described in more detail in Section 65.2.4, but the chemistry of U022+ and Th4+ extraction by TBP is considered here. 

Solvent extraction can be carried out in pulsated extraction columns, in mixer-settlers or in centrifuge extractors. Organic compounds such as esters of phosphoric acid, ketones, ethers or long-chain amines are applied as extractants for U and Pu. Some extraction procedures are listed in Table 11.11. The Purex process has found wide application because it may be applied for various kinds of fuel, including that from fast breeder reactors. The Thorex process is a modification of the Purex process and has been developed for reprocessing of fuel from thermal breeders. 

Solvent extraction with TBP has become the standard procedure for purifying thorium, just as for uranium. Processes used in different countries differ, however, in details such as the solvent used to dilute TBP, its concentration, and the means used to strip thorium and coextracted uranium from TBP. Table 6.20 summarizes the main features of processes used for purification of thorium on an industrial scale in the principal thorium-producing countries. Wylie [W5] gives more detail on early pilot-plant thorium-purification runs. Most of the published U.S. work on thorium purification on an industrial scale deals with irradiated thorium rather than natural this will be described under the Thorex process, in Sec. 5 of Chap. 10. 

Culler and Blanco [CIS] have summarized other aqueous processes that have been studied for processing power reactor fuels not readily handled by the standard Purex or Thorex processes. Many of these require reagents other than nitric acid to dissolve either the cladding or the fuel, but finally use solvent extraction with TBP to separate and purify fissile materials. Details of these other processes are given in references cited by Culler and Blanco [Cl8]. 

Codecontamination and partition cycle. Because the codecontamination and partition cycle is the critical step in the acid Thorex process, it will be described in more detail. In this cycle, shown in Fig. 10.21, most of the fission products were separated from the uranium and thorium, which were then separated from each other. The four solvent extraction units, HA, IBX, IBS, and 1C, were pulse columns with dimensions given in the figure. 

To avoid these difficulties they reduced the thorium content of solvent extraction feed to 1.15 Af and developed a two-stage acid Thorex process. In this process thorium and uranium were coextracted from an acid feed to separate them from most of the fission products and then stripped back into the aqueous phase. By this means fission products were removed to such an extent that the Thorex process with acid-deficient feed could be used in the second stage without causing them to precipitate. 

There are two breeder reactor fuel cycles. One involves the irradiation of U/ Pu oxide fuel with fast neutrons and is at the prototype stage of development. The other involves the irradiation of Th/ U oxide fuel with thermal neutrons and is at the experimental stage. Fuel from the U/ Pu cycle may be reprocessed using Purex technology adapted to accommodate the significant proportion of plutonium present in the fuel. Increased americium and neptunium levels will also arise compared with thermal reactor fuel. The Th/ U fuel may also be reprocessed using solvent extraction with TBP in the Thorex (Thorium Recovery by Extraction) process. In this case the extraction chemistry must also take account of the presence of Pa arising as shown in Scheme 2. 

As far as reprocessing in the U/Pu fuel cycle is concerned, several chemical separation techniques have been proposed and developed in the past few decades. The most efficient process to date remains the PUREX process (Plutonium and Uranium Recovery by Extraction). This process uses nitric acid HNO3 and organic solvents to dissolve and extract selectively U and Pu, resulting in two separate product streams (U on one side and Pu on the other side of the process chain). As far as reprocessing in the Th/ U fuel cycle is concerned, THOREX (Thorium Oxide Recovery by Extraction) technology must be used, also based on dissolution in nitric acid and solvent extraction (however, with special care for the extraction of Pa, for the separa-tion of U and U, and for the dissolution of thorium dioxide in pure nitric acid). 

Fig. 6-15. Thorex process, solvent extraction co-decontamination flowsheet.

Even though in the THOREX process 233U can be preferentially recovered from irradiated thorium fuel by using an extraction flowsheet based on 5% TBP n-dodecane as the extractant, further lowering of the concentration of TBP in the solvent has certain advantages in terms of reduced co-extraction of thorium and fission products (195, 196). Ramanujam et al. reported a sequential precipitation technique... 

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