TY - GEN
T1 - Effectiveness of Sodium Hydroxide Pretreatment in Enhancing Fission Product Removal During Alkaline Spent Fuel Reprocessing
AU - Abdelgader, Gosai A.M.
AU - Jiao, Caishan
AU - He, Mingjian
AU - Fang, Haofan
AU - Waeel Kamal, A. A.
N1 - Publisher Copyright:
© The Author(s), under exclusive license to Springer Nature Singapore Pte Ltd. 2026.
PY - 2026
Y1 - 2026
N2 - The majority of fission products in SNF can be separated during the oxidative dissolution in Na2CO3–H2O2 solutions through alkaline spent fuel reprocessing. However, since some fission products like Cs, Tc, Mo, and Te can be completely dissolved in a carbonate solution, they can not be removed during the extraction step of alkaline spent fuel reprocessing. In this work, a pretreatment step using sodium hydroxide solution has been investigated in order to remove these dissolved fission products in advance. At 25 °C, after 60 min, the results showed that the dissolution rates of Мо, Cs, Ba, Re, and Te were enhanced with the increase of NaOH concentration. However, the dissolution rate of UO2 and U3O8, respectively, decreased from 0.191% to 0.026% and from 0.223% to 0.035% at 85 °C, after 60 min. The dissolution rate of UO2 and U3O8 was decreased to 0.004% and 0.007%, respectively. Thus, the obtained results demonstrated that NaOH pretreatment for the SNF is considered an effective pre-step in selectively dissolving these fission products before moving to the oxidative dissolution step, as well as improving the reprocessing efficiency in order to minimize dissolution challenges during the different extraction stages.
AB - The majority of fission products in SNF can be separated during the oxidative dissolution in Na2CO3–H2O2 solutions through alkaline spent fuel reprocessing. However, since some fission products like Cs, Tc, Mo, and Te can be completely dissolved in a carbonate solution, they can not be removed during the extraction step of alkaline spent fuel reprocessing. In this work, a pretreatment step using sodium hydroxide solution has been investigated in order to remove these dissolved fission products in advance. At 25 °C, after 60 min, the results showed that the dissolution rates of Мо, Cs, Ba, Re, and Te were enhanced with the increase of NaOH concentration. However, the dissolution rate of UO2 and U3O8, respectively, decreased from 0.191% to 0.026% and from 0.223% to 0.035% at 85 °C, after 60 min. The dissolution rate of UO2 and U3O8 was decreased to 0.004% and 0.007%, respectively. Thus, the obtained results demonstrated that NaOH pretreatment for the SNF is considered an effective pre-step in selectively dissolving these fission products before moving to the oxidative dissolution step, as well as improving the reprocessing efficiency in order to minimize dissolution challenges during the different extraction stages.
KW - Alkaline spent fuel reprocessing
KW - Fission products
KW - Radioactive waste
KW - Sodium hydroxide
UR - https://www.scopus.com/pages/publications/105028884411
U2 - 10.1007/978-981-95-2732-8_14
DO - 10.1007/978-981-95-2732-8_14
M3 - 会议稿件
AN - SCOPUS:105028884411
SN - 9789819527311
T3 - Springer Proceedings in Physics
SP - 161
EP - 171
BT - Proceedings of the 32nd International Conference on Nuclear Engineering—Volume 4; ICONE 2025, Nuclear Fuel and Materials, Transportation and Fuel Cycle, and Reactor Physics II
A2 - Tan, Sichao
A2 - Xu, Weiqiang
A2 - Zhu, Yanyan
PB - Springer Science and Business Media Deutschland GmbH
T2 - 32nd International Conference on Nuclear Engineering, ICONE 2025
Y2 - 22 June 2025 through 26 June 2025
ER -