Radiation Shielding Concrete for Spent Nuclear Fuel Interim Dry Storage

Abstract

The demand for nuclear energy, which is marketed as zero-emission clean energy and a secure source of electricity, is continuously rising globally. However, associated radioactive waste such as spent nuclear fuel (SNF) presents a serious health and waste management hurdle. Interim dry storage methods have emerged as a remedial measure for the nuclear industry. This includes the concrete-based dry storage cask which is the focus of this study. The evolution and design requirements of SNF dry storage casks and interim storage facilities were first reviewed, and the existing design and operation challenges were identified. Accordingly, four comprehensive studies were undertaken: 1) quantify the effects of minerals and other additives added to enhance radiation shielding concrete on the concrete physical and mechanical properties; 2) develop a user-friendly computational software, named MRCsC, for estimating fast neutron macroscopic removal cross-section, ΣR (in cm-1), for shielding materials; 3) investigate experimentally and analytically the effects of adding powder boron carbide on Portland cement hydration kinetics and concrete compressive strength and radiation shielding properties; 4) study analytically the effectiveness of aggregates such as barite and celestite on concrete radiation shielding properties, and model the feasibility of developed concrete mixes as an overpack in a sandwich-design storage cask using the software OpenMC. Overall, the results reveal that sandwich-design concrete cask meets all the requirements for SNF dry storage. The addition of powder boron carbide, up to 50% by weight of cement, delayed the cement initial setting time by approximately 6 h, yielded a 15% increase in cumulative heat of hydration and a 28% increase in concrete compressive strength after 3 days, and increased the thermal neutron absorption (Σabs) close to 62000%. Concrete mixes with barite and celestite, as coarse aggregates replacement, significantly improve concrete radiation shielding efficiency and feasibility of sandwich-design storage cask. The OpenMC model results revealed over 60% decrease in the total dose rate.