MCNP Study of Concrete Activation Due to Beam Port Penetration of the Illinois TRIGA Reactor

The purpose of this paper is to investigate the radioactivity of the concrete biological shield in the University of Illinois Advanced TRIGA research reactor for decommissioning and disposal. The Illinois reactor was modeled using the Monte Carlo N-Particle transport code (MCNP). The MCNP calculated flux in the core was compared to experimental flux measurements determined by foil activation. The MCNP calculated flux in the beam ports was compared to TRIGA reactor simulated flux by other studies. Activity of the concrete was determined using the MCNP calculated flux measurements. The highest concrete activity was found for the piercing beam port region. Radionuclides of interest are Ca-41, Fe-55, Mn-54, Co-60, Zn-65 and Cs134. The classification of the activity of the concrete cells with respect to disposal is set forth in the Code of Federal Regulations. Co-60 was the radionuclide with activity levels that determined waste classification in most concrete cells. It was found that all concrete in the Illinois reactor would reach class A activity levels ten years from shutdown.