Acid Thorex process

Fig. 12.9 Two-stage acid Thorex process for highly irradiated fuels. Numbers in the frames indicate stage number, whereas numbers on the lines indicate flow volumes relative to the feed volume (DOD and FP are dodecane and hssion products). 

Fluoride and aluminum were also included in the feed because they would be needed in an actual dissolution step for ThC (fluoride catalyzes the dissolution, and aluminum counteracts the corrosiveness of the fluoride). For efficient Th extraction, the acidity of the feed solution should be in the range of 2 to 3 mol/L in HNO3. This is a significant departure from the acid Thorex process which uses an acid-deficient feed solution and is reported to. achieve improved decontamination from fission products (8). However, acid-deficient feed solutions were considered undesirable in our flow sheet because Pu hydrolyzes and tends to polymerize at low acidity (9). The effect of the higher feed acidity used here on fission product decontamination has not yet been established but will be assessed in later experiments. 

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. 

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 second campaign, in 1970, was described in detail by Jackson and Walser [Jl] of the Atlantic Richfield Hanford Company. This used less aluminum nitrate than the Savannah River campaign and is closer to the acid Thorex process presently favored. A summary of the Hanford operation will be given in Sec. 5.5. 

In the early 1970s Kiichler and associates [K6, K7] of Farbwerke Hoechst adapted the acid Thorex process to fuel irradiated to bumups to 100,000 MWd/MT, such as are expected from the HTGR, AVR, and THTR. They made laboratory runs on spiked synthetic fuel simulating chemically high-burnup fuel. They also made hot-laboratory runs at the Kernfor-schungsanlage Jiilich on 1 kg/day of fuel irradiated to burnups up to 54,000 MWd/MT. The flow sheet demonstrated in these hot-laboratory runs is described in Sec. 5.6. 

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. 

Two-Stage Acid Thorex Process for High Bumup Fuel... 

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. 

Figure 10.22 First stage of two-stage acid Thorex process for high-bumup fuel. From Kuchler etal [K7].)...

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. 

Figure 10.26 Distribution coefficients of principal metal nitrates in acid Thorex process at low concentration.

R3. Rainey, R. H., and J. G. Moore Laboratory Development of the Acid Thorex Process for Recovery of Consolidated Edison Thorium Reactor Fuel, Report ORNL-3155, May 11, 1962. 

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