During the decommissioning of nuclear power plant (NPP), massive amount of concrete wastes is generated, which are non-radioactive and radioactive. The concrete is a representative construction material which affords reliable structural stability, good formability, and trustful integrity. Also, its reasonable neutron absorbing property allows the various application for many components, including building construction material, bio-shield concrete, etc. Due to the noted aspects of concrete, the radiological concrete characterization is classified as an important process for development of effective strategy for concrete management, in terms of process management and financial control during the decommissioning. The characterization of bio-shield concrete is important in waste management. The understanding and characterization of activation depth is essential for the determination of waste management strategy, classification of bio-shield concrete, and process development of decommissioning. On the other hand, concrete for construction application requires the evaluation of surface contamination of them. The concrete for containment building and its structure is rarely activated, but surface contaminated. In this paper, the reactor data from representative PWR reactors in the US is studied. The experience on Yankee Rowe, Maine Yankee, and Connecticut Yankee NPPs are systematically studied. The Yankee Rowe was a 4-loop PWR of Westinghouse design with 185 MWe. The Main Yankee was a 3- loop PWR of Combustion Engineering design with 864 MWe. The Connecticut Yankee was a 4-loop Westinghouse type with 560 MWe. The characterization studies on bio-shield concrete will be discussed in this paper, including activation depth, primary nuclides, etc.

This study evaluates the radioactivity of concrete waste that occurs due to large amounts of decommissioned nuclear wastes and then determines the surface dose rate when the waste is packaged in a disposal container. The radiation assessment was conducted under the presumption that impurities included in the bio-shielded concrete contain the highest amount of radioactivity among all the concrete wastes. Neutron flux was applied using the simplified model approach in a sample containing the most Co and Eu impurities, and a maximum of 9.8×104 Bq·g−1 60Co and 2.63×105 Bq·g−1 152Eu was determined. Subsequently, the surface dose rate of the container was measured assuming that the bio-shield concrete waste would be packaged in a newly developed disposal container. Results showed that most of the concrete wastes with a depth of 20 cm or higher from the concrete surface was found to have less than 1.8 mSv·hr−1 in the surface dose of the new-type disposal container. Hence, when bio-shielded concrete wastes, having the highest radioactivity, is disposed in the new disposal container, it satisfies the limit of the surface dose rate (i.e., 2 mSv·hr−1) as per global standards.