Waste glass made from high-level liquid waste (HLLW) is the main high-level waste produced in spent fuel reprocessing plants. Our understanding of the radiation characteristics of such waste packages is relatively in-depth, and we have well-established radiation protection procedures for their hoisting and transfer operations. Under the strategy of HLLW partitioning (the TRPO process developed in China is taken as an example in this study) aiming at the optimization of waste management, HLLW is separated into several streams of waste/product with different properties, which facilitates the implementation of refined waste management and isotope recovery in reprocessing plants. Among the waste streams, the raffinate after TRPO extraction, i.e., Sr- and Cs-containing stream, has the highest radioactivity and thermal power. It contains most of the intermediate- and short-lived heat-emitting radionuclides in the HLLW (mainly Cs-137/Ba-137m and Sr-90/Y-90), but with very little residual long-lived α nuclides. At present, no quantitative study has been conducted on the radiological properties of the waste glass made from the Sr- and Cs-containing stream, and their differences from “conventional” waste glasses (i.e., direct vitrification of HLLW). This paper uses the Monte Carlo method to model the waste glass. It calculates the dose rates of several locations on the surface of, and in the surrounding space of the waste glass package. It analyzes the impact of spent fuel burn-up and source items (waste streams before and after HLLW partitioning) on dose rates. The calculation results provide a basis for formulating radiation protection plans for different types of waste glass packages during hoisting and transfer operations, as well as for the shielding design of transfer tools. This enables the vitrification facility to better ensure the radiation safety of operators while maintaining economic efficiency.