Thursday, December 15, 2011 11:00 am – 12:00 pm
Room 610, M&M Building
Dr. Delwin C. Mecham
CWI (Idaho National Engineering Laboratory)
The Calcine Disposition Project (CDP) of the Idaho Cleanup Project (ICP) has the responsibility to retrieve, treat, and dispose of the calcine stored at the Idaho Nuclear Technology and Engineering Center (INTEC) located at the Idaho National Laboratory in Southeast Idaho. Calcine is the product of thermally treating, or “calcining”, liquid high-level or sodium-bearing nuclear waste produced at INTEC from 1963 to 1998 during the reprocessing of spent nuclear fuel (SNF). The CDP is currently completing the design of the Hot Isostatic Pressure (HIP) treatment process for the calcine to produce a volume- reduced, monolithic, glass-ceramic waste form suitable for transport and disposition.
Conceptual design for the CDP requires the design of a large scale HIP can
which maintains containment of calcine during the HIP treatment cycle. The HIP can must be filled with calcine and additives and sealed remotely. The HIP cans will undergo approximately 50% volume reduction at a temperature of 1000-1250°C and a pressure of 50-100MPa. The HIP can’s main function is to provide primary containment of the radioactive calcine material during and after the HIP treatment process. Development of a virtual testing program using high fidelity modeling techniques is required due to the prohibitive
cost of full-scale testing using actual HLW calcine. This paper describes current design, analysis, and testing of HIP cans and the design for filling and sealing HIP Cans. The basic HIP technology is summarized and the remote HIP can fill and seal design is presented. Simulation models are developed to establish a virtual testing program using Finite Element Analysis (FEA). Software packages COMSOL and ABAQUS are being used to analyze the thermal and structural response of HIP cans during the HIPing process. The software packages increase the understanding of can deformation and allow for HIP can virtual testing before large scale testing of the HIP cans. This decreases the number of physical HIP can tests needed during the development of a HIP can design. The models utilize a macroscopic representation of the granular material “constitutive model” for the material inside the can and a non-linear representation of the stainless steel. Initial small scale testing of HIP cans has been performed to benchmark the FEA analysis and provide validation of the constitutive models used. Analytic results, test data, and comparisons between them are presented.
Dr. Del Mecham has forty years of experience in the development and application of thermal-hydraulic computer codes for nuclear reactor safety analysis; planning and management of large-scale thermal-hydraulic experiments. Dr. Mecham has developed and managed irradiation testing programs as well as participated on national and international research technical advisory boards; program development and technical management. Dr. received his PhD in Mechanical Engineering from Utah State University and is a Registered Professional Engineer in the State of Idaho. Dr. Mecham serves in the Industrial Advisory Committee for Mechanical Engineering at Utah State and holds an Adjunct Professor position at the University of Idaho.