The essential functions of the reprocessing of spent fuel elements is to separate and plutonium from one another and both of them from the radioactive fission products. For this purpose, the PUREX process (Plutonium and Uranium Recovery by Extraction), based on extractive separation, has become accepted worldwide. It is currently used in all modern reprocessing plants.

The Purex process was developed between 1945 and 1949 in the USA for military purposes and since 1954 has been operated industrially in more than 10 countries in reprocessing plants of various sizes. Decades of experience with this process exist in the USA, Great Britain and France. In the Federal Republic of Germany operational experience has been acquired over 20 years with the reprocessing plant at Karlsruhe. The Purex process is also used in Japan and Russia.

At the end of the 1970's the civil reprocessing plant in the USA was closed down for political reasons. The spent fuel elements have been stored since then at interim storage sites. The only reprocessing plants operated in Western Europe, those in France and Great Britain, have successively expanded their capacities and have currently a joint capacity of 2300 t heavy metal per year.

With the plants UP 2-800 and UP 3 (each with a capacity of 800 t/a) in France and Thorp (with 700 t/a) in Great Britain, there is sufficient capacity in Europe to reprocess the waste of 100 nuclear power stations and thereby to cover the total demand for the next 30 to 40 years. Japan disposes over an operating pilot plant and an industrial plant has been in the planning phase for a number of years.

The Purex process is in principle also suitable for the reprocessing of fuel from fast breeder reactors, as shown by the results from development work in France, Great Britain and the Federal Republic of Germany. The important differences compared with the reprocessing of light-water reactor fuel elements are a ca. ten-fold greater plutonium content and much shorter spent fuel element cooling times before reprocessing (these should only be ca. 6 months to 1 year).

Purex process: The actual reprocessing process begins with the cutting up of the fuel elements taken from entry basins. This can be carried out in two process variants: cutting up in ca. 5 cm long pieces with rod shears, whereby initially the head pieces are separated off and the individual fuel rods withdrawn from the rod bundle, or direct cutting with hydraulic bundle shears.

In the second step the nuclear fuel is selectively dissolved in hot nitric acid in an apparatus with criticality avoidance features, the zirkaloy cladding tubing not being dissolved. The dissolution process itself can also be carried out in two process variants: cutting of the fuel elements in a given quantity of acid in the dissolver or cutting and then adding acid. The second process variant has the advantage that the gases liberated during the dissolution (nitrogen(Ⅱ) oxide, ⁸⁵Krypton, ¹³¹, tritium, ¹²⁹) are continuously liberated. The cladding is left behind in the dissolution process in so-called dissolver baskets. All of the process steps proceed by remote control in bunker-type rooms with meter-thick concrete walls (hot cells) to prevent exposure to the intensively radioactive radiation.

The nitric acid solution from the dissolution of the fuel rod contents is filtered [poly(propene) fleece] or centrifuged, to remove suspended solids (zirconium- or molydenum- compounds and ruthenium and palladium alloys). The thus obtained fuel solution contains uranium, plutonium and the radioactive fission products. It is, after its composition is adjusted to the extraction conditions (3 molar in nitric acid and 240 to 300 g/L uranium) subjected to multi-cyclic extraction with tributylphosphate (dissolved in dodecane). Uranium and plutonium pass into the organic phase and are thereby separated from the fission products, which remain in the aqueous phase.

In the case of large throughputs, pulse-type sieve plate columns or mixer-settlers are used as extraction apparatuses both for this process step and for the later extraction steps.

In the next step uranium and plutonium are separated from one another by adding hydrazine, whereupon the uranium present as a uranium(Ⅳ) salt forms a complex with hydrazine which remains in the organic phase, but the plutonium present in the organic phase as a plutonium(Ⅳ) salt is reduced to plutonium(Ⅲ), which is insoluble in the organic phase and therefore goes into the aqueous phase. A separation of the uranium from the plutonium is thereby achievable. Recent development work has shown that the plutonium(Ⅳ) can be electrolytically reduced to plutonium(Ⅲ) in situ, which results in more efficient separation.

In the next process step the uranium is stripped with 0.01 M nitric acid into the aqueous phase.

Separation of the fuel solution into three aqueous solutions containing uranium, plutonium and the fission products respectively has thereby been achieved in the first extraction cycle.

In two further extractive uranium purification cycles, each consisting of extraction and stripping, the uranium solution is further purified to remove residual plutonium, neptunium and technetium.

In two extractive plutonium purification cycles plutonium is separated from small quantities of coextracted fission products. The plutonium(Ⅲ) is oxidized with nitrogen(Ⅳ) oxide to plutonium(Ⅳ), which is extracted with tributylphosphate. This oxidation can also take place anodically.

With the help of this multicyclic extraction the contamination of uranium and plutonium with fission products is reduced to 0.1 to 1 ppm. The residual concentration of plutonium in uranium may not exceed 10 ppb, since the uranium must be able to be processed without protective measures. The recovery efficiency for uranium and plutonium is 98 to 99%.

After the purification cycles, the dilute plutonium nitrate solution is concentrated to ca. 250 g/L and the uranium nitrate solution to ca. 450 g/L.

Tributylphosphate is recycled, it being scrubbed, e.g. with sodium carbonate, to remove associated interfering impurities such as dibutylphosphate produced by radiation, before being reused.

The Purex process is carried out at temperatures up to 130 °C, at, or slightly below, atmospheric pressure and uses aqueous dissolution and extraction processes, which are tried and tested in the chemical industry. In addition, it has proved possible to limit the places in the plant with high radiation activity to a few areas.