Uranium Thorex process

Table 12.10 Performance of Two-Stage Thorex Process and a Uranium Purification Cycle...

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 The former methods have been historically important but have now been superseded by solvent extraction with TBP.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. 

More recently a flowsheet has been developed which employs 30% TBP/OK as the solvent.349-446 447 This involves the use of an acid feed to the first cycle to assist in zirconium decontamination and suppress hydrolysis. An acid-deficient partition cycle then follows in which the U-Th separation is effected. A pilot plant (JUPITER) has been constructed at Julich in Germany to process Th02/U02 fuel using this flowsheet. Although a complete separation of thorium, uranium and FPs is possible using TBP in the Thorex process,448 alternative approaches... 

In order to make use of thorium as a nuclear resource for power generation, development of efficient separation processes are necessary to recover 233U from irradiated thorium and fission products. The THORium uranium Extraction (THOREX) process has not been commercially used as much as the PUREX process due to lack of exploitation of thorium as an energy resource (157,180). Extensive work carried out at ORNL during the fifties and sixties led to the development of various versions of the THOREX process given in Table 2.6. The stable nature of thorium dioxide poses difficulties in its dissolution in nitric acid. A small amount of fluoride addition to nitric acid is required for the dissolution of more inert thorium (181). 

Improvements in Thorium-Uranium Separation in the Acid-Thorex Process... 

The Acid-Thorex process has been used in recent years to recover 233U from neutron irradiated thoria targets. (] M This process uses n-tributyl-phosphate (TBP) in normal paraffin hydrocarbon (NPH) as the extractant and the relative uranium and thorium solubilities in each phase are adjusted by control of the nitric acid concentration. The Acid-Thorex process is the primary candidate for use in proposed aqueous thorium fuel cycles. In this process, uranium is separated from thorium through exploitation of the difference in equilibrium distributions since no usable valence change is available to aid in this separation. 

It is also of interest to note that the effect of DBP on zirconium separation from thorium in the Acid-Thorex system is different than zirconium separation from uranium in the Purex system.(Figure k) The Purex data are from reference 6 and the Acid-Thorex data are from General Atomic Company pilot plant studies. The thorium probably forms a stronger DBP complex than does uranyl ion and, therefore, the amount of uncomplexed DBP available for raising the equilibrium distribution of zirconium would be less in the Acid-Thorex process. 

In the Acid-Thorex process, fluoride ion should be added to the thorium partitioning solution (1BX) to decrease thorium transfer to the uranium stripping column, particularly where highly radioactive feeds are used. This fluoride ion addition then decreases the precipitation of thorium-DBP in the uranium stripping column. Also, the partition cycle should be the first cycle in the Acid-Thorex process to allow separation of thorium from DBP. 

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. 

At the end of irradiation in such reactors, fuel consists of a mixture of thorium, uranium containing fissile isotopes, and fission products. Figure 3.33 showed a fuel-cycle flow sheet for an HTGR. The Thorex process has been developed for recovering the uranium and thorium from such fuel cycles, freeing them from fission products and separating them from each other. The Thorex process will be described in this section. When the fuel being irradiated contains appreciable the plutonium thus formed requires that a combination of the Thorex and Purex processes be used. 

As in the Purex process, the Thorex process uses a solution of TBP in hydrocarbon diluent to extract the desired elements, uranium and thorium, from an aqueous solution of nitrates. Thorium nitrate however, has a much lower distribution coefficient between an aqueous solution and TBP than uranium or plutonium. To drive thorium into the TBP, the Thorex process as first developed at the Knolls Atomic Power Laboratory [HI] and the Oak Ridge National Laboratory [G14] added aluminum nitrate to the thorium nitrate in dissolved fuel. This had the disadvantage of increasing the bulk of the high-level wastes, which then contained almost as many moles of metallic elements as the original fuel. To reduce the metal content of the waste, the Oak Ridge National Laboratory in the late 1950s [Rl, R2] developed the acid Thorex process, in which nitric acid is substituted for most of the aluminum nitrate in the first extraction section. The nitric acid is later evaporated from the wastes, as in the Purex process. 

To simulate recovery of uranium and thorium from irradiated 6 percent uranium, 94 percent thorium fuel from the first loading of Consolidated Edison Company s Indian Point 1 nuclear power plant, Oak Ridge National Laboratory [R3] made small-scale experiments on application of the acid Thorex process to fuel containing the appropriate amounts of uranium and thorium, with tracer quantities of the principal fission products. Spent uranium-thorium fuel from the Indian Point 1 plant was subsequently processed by Nuclear Fuel Services, Inc., at West Valley, New York, for recovery of uranium, but without separation of thorium from fission products. No account of this separation has been published. 

The other full-scale applications of the Thorex process have been to separation of from thorium irradiated at the U.S. Atomic Energy Commission s production reactors at Savannah Rivet and Hanford. As the object of these irradiations was to produce of high isotopic purity for use in the first core of the LWBR, the bumup to which the fuel was exposed was low, and the concentrations of uranium and fission products in the irradiated thorium were much lower than will exist in power reactor fuel irradiated to full bumup. Nevertheless, the successful separation of uranium and thorium from each other and from fission products is significant confirmation of the workability of the Thorex process. 

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. 

Kiichler and associates [K6, K7] of Farbwerke Hoechst have investigated the modifications necessary in the acid Thorex process to enable it to handle (I) the high concentration of fission products present in fuel with the burnups of up to 100,000 MWd/MT expected in fuel from the HTGR, AVR, and THTR, and (2) uranium concentrations of up to 20 percent in thorium, which may be used in these reactors when fissile uranium is diluted with U to deter its use as a nuclear explosive. They found two difficulties with the acid Thorex process flow sheets previously used at Oak Ridge [B14] and Hanford [Jl] ... 

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. 

Adequate data on distribution coefficients of uranyl nitrate between 30 v/o TBP and aqueous solutions of thorium nitrate and nitric acid are not available. Examination of concentrations of coexistent phases in Thorex process mixer-settler runs reported in references [Rll], [01], and [02] indicate that the distribution coefficient of uranium Dy when present at uranium concentrations below 0.02 M in Thorex systems at thorium concentrations above 0.1 Mis given approximately by... 

Table 10.19 Decontamination factors observed by Kiichler in hot-cell run with two-stage Thorex process followed by thM uranium cycle...

However, column separation performance in the Hanford Thorex campaign correlates better with a Z)u/ Th rstio of 14 (Prob. 10.5). Because the distribution coefficient of thorium is so much less than that of uranium, the Thorex process requires a much higher organic/aqueous flow ratio than the Purex process. 

Figure 10.26 compares the low-concentration distribution coefficients of uranium, thorium, plutonium, protactinium, and the principal fission products. The spread between thorium and fission-product zirconium is greatest between 1 and 2 M HNO3, the range used in the decontamination step of the acid Thorex process. Because the distribution coefficient of protactinium is close to that of thorium, it is necessary to remove protactinium or complex it with fluoride or phosphate ion to prevent its extraction with thorium. 

Interest now is centered on the thorium cycle (23) and laboratory studies have continued to investigate both an adaptation of the Thorex process to CANDU fuel and the application of the amine process to recovering uranium-233 from irradiated thorium. The program to develop and fully demonstrate the thorium fuel cycle has been outlined, and would require about 25 years to complete. However, the current research level will not be expanded until a decision can be taken by the Canadian Government when the information from the current International Nuclear Fuel Cycle Evaluation has been assessed. 

One of the major issues of the Thorex process is the creation of a third phase between thorium and TBP if the thorium concentration in the solvent is too high. Furthermore, the partition of uranium and thorium is more difficult than the partition of uranium and plutonium. No change of the thorium oxidation state is required, but the separation of thorium from uranium in the IB contactor must be obtained entirely by a rather delicate adjustment of salting strength inside the contactor. It appears that the Thorex process has been variable in performance. Decontamination from ruthenium has varied and has been particularly poor when short-cooled thorium-based fuel was processed. 

Fig. 6-16. Thorex process, uranium isolation and third cycle flowsheet.

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. 

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