Spent radioactive ion-exchange resin (SIER) is a long-standing issue for radioactive waste management safety. Performance of radioactive waste form is featured with extra long-term and non-repairable. Calcium sulfoaluminate cement (SAC) was used in radioactive spent resins solidification in China. A prescription of X SAC cement + 0.5X waste resins (50%water hold) + 0.35X water was obtained first. In order to control the temperature rise caused by hydration of cement in 200L solidification matrix, various supplementary materials were tried. Based on compressive strength tests and center temperature rise, super powered zeolite was selected. In addition, more resins were added to reduce the center temperature rise. A superior combination was obtained as SAC 35wt.%, zeolite 7wt.% to mix 42wt.% of resins (50%water hold) with 16wt.% of water. The microstructures of hydrated Ordinary Portland Cement (OPC), SAC and SAC with different zeolite addition were compared by means of Scanning Electron Microscopy (SEM). From the SEM pictures, the structures of the needles or spines can be seen in SAC matrices and the needles structure of SAC change into flake structure gradually with more zeolite added. The simulated leaching tests showed that inclusion of zeolite in SAC reduced the leaching rates of radionuclides significantly. From 200L matrix test, the centre temperature curve was measured, and the highest temperature was lower than 90°C. No thermal cracks were found in the final solidified products. The effect of radiation on compressive strength and radiolysis gas generation was studied for cement solidified form of various content of ion exchange resin with Co-60 irradiator. Variation of compressive strength, as well as the compressive strength of the waste form both with and without irradiation all within the standards requirement under irradiation of 106 Gy. However, the data obtained for ion exchange resin shows that hydrogen generation under irradiation of 105 Gy reached up to 3.5% of the total gas generated. This implies that the radioactivity of spent ion exchange resin shall be limited for long term storage and disposal with High-Integrity-Container. Calculation demonstrates that cement solidification of spent radioactive ion exchange resin existing in China so far should not result in radiation stability concern. It is concluded that SAC is one of the preferential binding material for ion exchange resins, the resin loading can be up to 75 (vol%) (wet resin). It is recommended that the performance requirement for cement solidified radioactive form shall be amended and guidelines for performance characterization in certain detail should be established. Biodegradation of cement solidified resin waste would be a safety concern and shall be investigated. Modeling of leaching should be promoted.

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