Thorex waste

The Thorex wastes are not allowed to exceed 140 °F this minimizes corrosion of the stainless steel tanks. A series of three independent cooling coils has proved to be more than adequate for this purpose. 

The off gases from the Thorex waste pass through a caustic scrubber to remove the oxides of nitrogen prior to being combined with the Purex waste off-gases. This step is necessary since ammonia is evolved from the Purex waste and the combination of gaseous ammonia with oxides of nitrogen forms ammonium nitrate which sublimes on the filter media... 

The composition of the Thorex waste with the limiting speciflcations is shown in Table IV. This waste containing the thorium along wdth the fission products is being maintained at about 120 °F using one of three installed coils. Both Purex and Thorex waste systems are being main-... 

Other changes in the extractability of the actinides are performed using complexing agents. The distribution coefficients of Pu(lV) with various citric acid concentrations is demonstrated for the system TBP/HNO3 in table 3 and figure 3. The range of the HNO concentrations corresponds to that of the waste effluents from the coextraction steps of the THOREX-process. 

The actinides as trace components will not take any penulty on the already tested waste management procedures (denitration, calcination, vitrification), provided that Np and Pu can be retained in the aqueous waste stream. Without changes of the THOREX flow-sheet, the 1st feed solution contains Np(V) and Pu(lV). Consequently, Np ends up in the waste stream, while Pu following the heavy metal, can be withdrawn with the aqueous phase of the 2nd coextraction step with reducing agents and thus combined with Np (hatched arrow) for the further waste treatment. 

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. 

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. 

Halaszowich, St., et al. Interim Storage and Solidification for Thorex-Type Fission-Product Solutions, Proceedings of the Management of Radioactive Waste from Fuel Reprocessing, Paris, 1972, Report CONF-721107, Mar. 1973, p. 705. 

---