In nature, thorium exists mainly in a monoisotopic form with no fissile isotope for its use as fuel in nuclear reactors. Though a fissile isotope of uranium, 233U is formed by neutron irradiation of fertile thorium in reactors; the subsequent reprocessing of irradiated thorium for recovery, purification, and further handling of 233U product (accompanied by 232U) has remained a challenge because of the complex radiological problems associated with irradiated thorium. During irradiation in the reactor, 233Pa with a 27 day half-life is formed by (n, γ) reaction of 232Th. Its complete decay to 233U is to be ensured prior to reprocessing for maximum recovery. The (n, 2n) reactions encountered during the irradiation of Th give rise to long-lived 231Pa and rather short-lived 232U (~70 years). The 232U and its hard beta gamma emitting short-lived daughters in the separated 233U and the 229Th and 228Th in thorium contribute to the radiation dose of these products. TBP has been the most widely used extractant in the nuclear industry for hydro metallurgical and reprocessing applications and hence, TBP-based THOREX process, in its various forms has become the natural choice to treat irradiated thorium in plants meant for PUREX process, on a campaign basis. Different flow sheets have been used and fine-tuned to meet the specific processing requirements of irradiated Th from different reactor systems based on the type of thoria target/fuel and the cladding under treatment, their irradiation, and cooling history and the end objectives, namely, separation and purification of Th/233U/233Pa in short cooled fuels, or Th and 233U in long cooled fuels/targets or 233U alone from the fuel target matrix and the final product decontamination factor aimed at. Thus the process has evolved over the years from the experience of individual plants that had conducted these pioneering earlier campaigns. This chapter details the major steps involved in this generic process and the several possible variations in the process flow sheet that can be utilized to meet the different end objectives. The different head end treatment options, fluoride ion catalyzed nitric acid dissolution of the fuel, the different TBP-based cycles for the extraction, individual separation and purification of Th/233U, extraction behavior of Th/233U/233Pa and fission products during this step, the solvent management in the cycles, third phase formation tendencies of Th with TBP, the solvent degradation and its consequences in the process are some of the topics covered. The global status on thorium utilization along with country-wise practices and experiences on thorium reprocessing, and recent developments in this domain have been briefly reviewed. Just prior to 233U product reconversion and fuel fabrication, as of now, a chemical separation of the longer lived daughters of 232U (like 228Th and 224Ra) is required to control the personnel exposure during these steps. The activities of short-lived, but strong gamma emitters in the chain, 208Tl and 212Bi, depend on 228Th, the nuclide with longest half-life in the chain. Techniques based on ion exchange, solvent extraction, or precipitation routes are followed for purification of 233U product from 228Th and for removing the residual thorium accompanying product. These are summarized along with the practices followed at different facilities to treat bulk quantities of aged 233U and their recycle during fuel fabrication. Third-phase formation encountered in TBP extraction of Th and Pu has been discussed in detail as it has serious repercussions on process performance and safety. In this context, many homolog neutral organophosphorus esters in the TBP family and alkyl amides have been tested as substitutes for TBP and some of them are reported to have better performance in terms of Th/U separation factors and higher thorium loading. Their status is presented. Major areas that need further developmental efforts for the process to succeed on industrial scale have been identified.

Aqueous Reprocessing by THOREX Process

Abstract The effectiveness of the solvent wash reagents such as sodium carbonate and hydrazine carbonate were compared using the lean solvent obtained during reprocessing of high plutonium content fast reactor spent nuclear fuel. The concentrations of the reagents were optimized based on the removal of retained metal ions from the lean solvent and 1.5 M hydrazine carbonate was found to be optimal and superior to sodium carbonate to be adopted as the reagent for primary cleanup of the spent solvent. Both the reagents were found to possess excellent room temperature stability and their performance was not affected upon gamma irradiation which was assessed in terms of capacity to remove di-butyl phosphate from degraded TBP/n-DD system.

Comparison of the performance of solvent wash reagents used for the primary cleanup of degraded PUREX solvent

.................................................................................................................1 Acknowledgements ................................................................................................2 List of original publications .....................................................................................3 Abbreviations .........................................................................................................4

Ion Exchange in Nuclear Fuel Reprocessing : Zirconium Phosphate Materials for the Separation of Trivalent Actinides and Lanthanides

Solvent wash studies for the removal of di-butyl phosphate from spent solvent under simulated PUREX condition

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W.W. Schulz

Papers

Macroreticular ion exchange resin cleanup of purex process tbp solvent.