Reprocessing

It must first be remembered that, although a once-through cycle is an option, the use of thorium-based fuels generally assumes reprocessing of spent fuel to achieve full potential of the Th/U-233 cycle. The recovered U-233 would then be used as seed material for another cycle. Reprocessed thorium would be recycled to save natural thorium resources, since, once again, thorium is only a fertile material. Therefore, it is clear that to take full advantage of the thorium fuel cycle, it is highly desirable to retrieve the fissile U-233 recovered by reprocessing thorium spent fuel and to recycle as much as possible.

Starting in the late 1940s, the US reprocessed approximately 900 metric tons of irradiated thorium fuels to recover about 1500 kg of U-233. Other countries also recovered U-233 from thorium-based fuels. India has done so recently.

Early experience with commercial reprocessing of thorium-based fuel was limited in the US as was reprocessing in general at that time. The only US reprocessing facility to ever operate on a commercial basis was the Nuclear Fuel Services (NFS), West Valley, NY Plant. It was permanently shut down in 1972 after six years of operation. The initial core from Indian Point 1 was fabricated with HEU/thorium fuel, and it was reprocessed at the NFS Plant in late 1968. The amount of heavy metal contained in the fuel was 16 Mt. Approximately 1.1 Mt of enriched uranium containing 7 wt% U-233, 58 wt% U-235 and other uranium isotopes was recovered, and then shipped to Oak Ridge National Laboratory. It was stored there for over 15 years in liquid form, and finally processed to produce a stable oxide form. A potential stabilization process for other U-233 bearing materials was therefore demonstrated.910

Reprocessing of thorium-based fuel is somewhat more challenging than that of uranium-based fuels mainly because the dissolution of thorium metal and thorium oxide is not as straightforward as with uranium. Developed by ORNL, THOREX is a hydrometallurgic process, a derivative of the Purex process, to recover thorium and uranium from thorium-based fuel (reference 3, section 6.2). The mechanical head-end steps are similar to those of uranium-based fuels (for they are similar fuels). However, the dissolution of irradiated thorium-based fuel is slow in nitric acid. Therefore a small amount of hydrofluoric acid must be added as a catalyst to improve the dissolution process. The presence of fluoride ions causes corrosion of the stainless steel equipment (such as dissolver tanks) since fluoride has aggressive chemical properties. Consequently, appropriate buffering agents to prevent corrosion are generally needed, which complicates the design of the equipment and increases the overall reprocessing cost. Aluminium nitrate can also be added to the aqueous dissolver solution to reduce corrosion. A downside of this is that the aluminium nitrate passes through the plant to be added to the fission products and thereby increases waste production. Because of this and other factors it is expected that the THOREX process will generate 50-70% more vitrified waste than PUREX.3 Other differences from PUREX arise from the presence of sulphates, phosphates and fluorides in the reprocessing plant raffinates. These could result in considerable corrosion issues in the process of vitrifying waste materials, where high temperatures are required.

Another issue in the back-end of the thorium fuel cycle, already mentioned in Section 8.3.2, arises from the rather long half-life of Pa-233 (27 days), the precursor to U-233. As a matter of fact, the increase in reactivity resulting from the transformation of Pa-233 into U-233 after reactor shutdown must not only be taken into account in the design and operation of the reactor but also in the design of handling and storage facilities needed prior to reprocessing and dedicated to spent fuel. Practically, the cooling time before reprocessing must last at least nine months (10 times the Pa-233 half-life) or more to allow virtually complete decay of Pa-233 to U-233 (such cooling time may also be needed for other reasons, such as because of decay heat). In the THOREX process, the remaining Pa-233 is passed into the fission product waste, as would Pa-231, which is an alpha-emitting isotope in the thorium burn-up chain (produced by the (n,2n) reaction on Th-232). Studies were carried out in the 1960s to develop a process for extraction of protactinium from a nitric solution but no simple solution was found at the time. With a half-life of 30 000 years, its radiological impact could be significant for the long-term safety of disposal.

Non-aqueous processes, alternatives to THOREX, were also studied in the past, like processes of volatilization of fluorides or electrorefining, but they did not lead to any convincing result.

Finally, it should be noted that reprocessing would be even more difficult with HTR-type fuels of whatever composition, since the particle coatings and graphite matrix are chemically resistant and troublesome to break down mechanically.

In conclusion, if it were thought necessary to make the thorium cycle more attractive by reprocessing thorium fuel to recover and recycle U-233, this would require a significant R&D programme to work out a viable industrial process.