The interest to immobilize 137 Cs can pursue different aims. When it is separated from spent nuclear fuels, it is possible to reduce the thermal loading of the host glass and increase the waste loading capacity of the matrix. Also, the separated 137 Cs can be used to produce radioactive sealed sources useful in industrial and medical applications. If reprocessing is not envisaged after the generation of electrical energy, the spent nuclear fuel is the most important radioactive waste generated in the nuclear fuel cycle. In this case, aqueous solutions obtained from the chemical dissolution of spent nuclear fuels are high-level wastes. Short-lived fission products, such as 90 Sr and 137 Cs, constitute the main source of heat generation from beta decay, causing self-heating of the glassy matrix. It is well known that vitrification is the main used technology, capable to immobilize these wastes during hundreds of years generating a waste form with proved structural stability, thermal shock resistance, and high chemical durability. In this article we present in an analogous way the immobilization of an specific amount of stable cesium in a porous Si-rich glass matrix through adsorption and sintering. Sintering requires lower temperature than that required in immobilization by melting. The leaching behavior of the waste form obtained was studied from the procedure described by the MCC-1 test method. Considering 137 Cs only for simulations, this work also includes the thermal evolution calculation of a simulated silica glass block loaded with 2 wt.% of 137 Cs.