Dosimetry and methodology of gamma irradiation for degradation studies on solvent extraction systems

Abstract The recycling of minor actinides from dissolved nuclear fuels by hydrometallurgical separation is one challenging strategy for the management of spent fuel. These future separation processes will likely be based on solvent extraction processes in which an organic solvent system (extractant and diluent) will be contacted with highly radioactive aqueous solutions. To establish a separation between different elements in spent nuclear fuel, many extractants have been studied in the past. A particular example is N,N,N′,N′ -tetraoctyl diglycolamide (TODGA), which co-extracts lanthanides and actinides from nitric acid solutions into an organic phase ( e.g. TODGA in n -dodecane). The radiolytic stability of these extractants is crucial, since they will absorb high doses of ionizing radiation during their usage. Worldwide, different gamma irradiation facilities are employed to expose extractants to ionizing radiation and gain insight in their radiation stability. The facilities differ in many ways, such as their environment (pool-type or dry), configuration and gamma sources (often 60 Co or spent nuclear fuel). In this paper, a dosimetric assessment is made using different dosimeter systems in a pool-type irradiation facility, which has the advantage to be flexible in its arrangement of 60 Co sources. It is shown that Red Perspex dosimeters can be used to accurately characterize this high dose rate gamma irradiation field ( approx. 13.6 kGy h −1 ), after comparison with alanine, Fricke and ceric-cerous dosimetry in a lower dose rate gamma irradiation field ( approx. 0.5 kGy h −1 ). A final validation of the whole chain of techniques is obtained by reproduction of the dose constants for TODGA in n -dodecane.