Loading...

Study on Immobilization of the Spent Ion Exchange Resins of Tehran Research Reactor in Borosilicate Glass

Rastgoo, Pouria | 2023

88 Viewed
  1. Type of Document: M.Sc. Thesis
  2. Language: Farsi
  3. Document No: 56088 (46)
  4. University: Sharif University of Technology
  5. Department: Energy Engineering
  6. Advisor(s): Samadfam, Mohammad; Yadollahi, Ali; Sepehrian, Hamid
  7. Abstract:
  8. Considering the operation of the current nuclear reactors and the country's policy to achieve 10,000 megawatts of nuclear power in the horizon of 1420, as well as the construction of research reactors, we will face a huge amount of radioactive waste in the coming years. Meanwhile, spent ion exchange resins constitute a large amount of low and intermediate level (LILW) solid radioactive waste produced from the nuclear industry. Therefore, appropriate precautionary measures should be taken for the immobilization and disposal of these radioactive wastes in order to ensure the sustainable development of the nuclear industry and the protection of the environment and human health. In this study, the immobilization of spent ion exchange resins of Tehran research reactor in borosilicate glass matrix was investigated on a laboratory scale. For this purpose, after characterizing the spent ion exchange resins of the Tehran research reactor, the simulated waste was prepared by doping non-radioactive cesium and cobalt elements for conducting experiments. First, thermal pretreatment of the resins was done in order to facilitate the immobilization process in the glass matrix at a temperature of 150 oC. Then, By using the combined design of experiment method, the effect of various parameters such as loading amount of IEX waste (20-40 %wt), borosilicate glass ferrite (60-80 %wt) and the melting temperature (1100-1200 oC) on the characteristics of glass waste forms was investigated. Characterization of the prepared waste forms was done by XRD and SEM-EDAX analyses. The chemical stability of glass waste forms was also evaluated by standard PCT leaching test method for cesium and cobalt elements. The investigations showed that the samples prepared at a temperature lower than 1175 oC and those prepared with a loading of more than 35 %wt have a crystalline phase and therefore are not chemically and structurally homogeneous. Also, in the samples with loading higher than 30 %wt, phase separation (white color containing insoluble sulfate) was observed on the surface of the glass. The formation of Inclusions was also observed in samples with a loading of more than 30 %wt and samples prepared at a temperature lower than 1175 oC and loading of more than 20 %wt. The density of all the prepared waste forms was measured in the range of 3.591-3.634 g/cm3 and the percentage of volume reduction of all waste forms was found in the range of 72.11-89.69%. Based on the obtained results, with the increase in temperature and the amount of waste loading, the density and volume reduction factor of the waste form increases. Investigations showed that glass waste form containing 30 %wt of resin waste prepared at a melting temperature of 1200 oC provides the best conditions for waste immobilization. In optimal conditions, the normalized leaching rate (NLR) of cesium and cobalt elements was found about 7.43×10-5 and 6.93×10-5 g/m2day, respectively, and the volume reduction factor was measured as 86.61. Experiments were continued with real radioactive resin samples in optimal conditions previously obtained for the immobilization of simulated ion exchange resins. Based on the obtained results, the samples were in very favorable conditions and the gamma analysis results confirmed the successful immobilization of spent IEX resins in borosilicate glass matrix
  9. Keywords:
  10. Ion Exchange Resines ; Cesium Sorption ; Cobalt ; Immobilization in Glass ; Borosilicate Glass ; Simulated Non-Radioactive Waste ; Chemical Durability ; Tehran Research Reactor

 Digital Object List

 Bookmark

No TOC