Cementation of Radioactive Waste from a Pwr with Calcium Sulfoaluminate Cement

Spent radioactive ion-exchange resin (SIER) and evaporation concentrates are radioactive wastes that are produced at by pressurized water reactor (PWR) nuclear power stations. Borate, which is used as a retardent for cement, is also present as a moderator in a PWR, therefore, borate will be present in both ion-exchange resins and evaporation concentrates. In this study the use of Calcium sulfoaluminate cements (SAC) as encapsulation medium for these waste streams was investigated. The study involved the manufacturing of different cement test samples with different amounts of SAC cement, waste resins (50% water content) and admixtures. In order to reduce hydration heat during 200 L solidification experiments, different admixtures were investigated. Initial results based on compressive strength tests and hydration temperature studies, indicated that zeolite was the best admixture for the current waste form. Experiments indicated that the addition of resin material into the current cement matrix reduces the hydration heat during curing Experimental results indicated that a combination of SAC (35 wt. %), zeolite (7 wt. %) mix with 42 wt. % resins (50% water content) and 16 wt. % of water forms a optimum cured monolith with low hydration heat. The microstructures of hydrated OPC, SAC and SAC with zeolite addition were studied using a Scanning Electron Microscopy (SEM). SEM results indicated that the SAC matrices consist of a needle type structure that changed gradually into a flake type structure with the addition of zeolite. Additionally, the presence of zeolite material inside the SAC matrix reduced the leaching rates of radionuclides significantly. In a final 200 L grouting test, measured results indicated a hydration temperature below 90C withno thermal cracks after solidified. The influence of radiation on the compressive strength and possible gas generation (due to radiolysis) on cement waste forms containing different concentrations ion exchange resin was studied using a Co-60 source. Results indicated that 106 Gy radiation had no influence on the compressive strength of the different matrixes. However, results indicated that radiation (105 Gy) of encapsulated resin matrixes have resulted in gas formation. The different gas compositions were analyzed and t 3.5% (max) hydrogen was measured. The presence of hydrogen formation can be limited by reducing the concentration of spent resin in a matrix intended for long term storage or disposal in High-Integrity-Containers. Calculations confirmed that the cumulative amount of solidified spent radioactive ion exchange resin encapsulated in China will not generate high concentrations of hydrogen. In summarizing, the results from this study indicated that t SAC is one of the preferential encapsulation cement for ion exchange resins and that a resin loading of 75(vol %) (wet resin) is achievable. It is therefore recommended that the performance requirements for the solidified of radioactive waste form could be amended and that new guidelines for performance characterization should be established. The biodegradation of solidified resin waste is unknown and could be a safety concern and therefore future studies must investigate this aspect. Modeling regarding the leaching of radionuclide from solidified resin waste should be encouraged.