MCNP Modeling of the UMass Lowell Research Reactor Gamma Irradiation Facility Michael Ducey Department of Chemical and Nuclear Engineering
Introduction The goal of this project was to develop and validate a detailed Monte Carlo N-Particle Transport1 (MCNP) model of the Gamma Irradiation Facility located in the University of Massachusetts Lowell Research Reactor (UMLRR). The Gamma Irradiation Facility is an offshoot room of the UMLRR, where materials are irradiated by a Cobalt 60 source. Cobalt 60 emits two primary gammas at 1.17 and 1.33 MeV. Knowledge of the gamma irradiation field is needed for both planning and subsequent analysis of any sample irradiation. With a working MCNP model of this facility, the gamma flux as a function of distance from the source can be determined at any particular time. MCNP is a computer code that allows a user to simulate both neutron and photon particle transport. MCNP uses statistical probabilities to track the paths that these particles take throughout a given geometry. This evaluation technique is much different from other deterministic methods like the diffusion equation, which looks at averaged cross sections in homogenized zones to determine the flux distribution throughout the system. In contrast to that solution technique, MCNP looks at the interaction of every radiation particle at every junction. Some examples of this for photons would be simulating the possibility of a particle being involved in a Compton scatter, pair production, or photoelectric effect at a given location. Further, MCNP will use probabilities to calculate the scatter angle and the energy that the photon has after these interactions, and continues to do this until the photon has deposited all of its energy. As can be seen from this brief description, MCNP is a powerful tool that can be used to supplement experimental methods, or validate deterministic modeling techniques. 22 UMLJUR
For the purposes of this project, MCNP was used to create a working model of the irradiation facilities at UML. These facilities are used to expose different materials to a photon source and test the material’s response to this irradiation. The photon source that is used in this facility is Cobalt 60 which emits gamma particles, or high energy photons. Cobalt 60 is an unstable isotope of cobalt that readily decays to Nickel 60 by photon decay. Intricate knowledge of both the number of photons present and the energy with which these photons possess, are vital to properly test the exposure to materials. We refer to the flux (the number of particles per cm2 per second) to give us insight into the “quantity” of photons present, and the dose rate (the amount of radiation absorbed per second) to give us insight into the energy deposition associated with a photon field. These terms will be used regularly in the following sections. Establishing a working MCNP model would reduce the need for periodical radiation measurements. Currently, experimental techniques are used to determine photon fluxes and dose rates. This requires irradiating a known sample and “back tracking” information about the photon field from there. Here, MCNP could be used to supplement this existing procedure and would be able to quickly gauge experimental accuracy. In addition, reactor staff could have a preexisting knowledge of the photon flux and energies a head of time and plan accordingly. With a working MCNP model, a user could input the decay corrected source data in a post processing code from MATLAB, and flux profiles as a function of distance and dose rates could be generated. Clearly, this would be a valuable asset to the UML Irradiation Facilities.