Optimizing Neutron Yield for Active Interrogation

Neutrons are commonly used for many applications, including active interrogation and cancer therapy. One critical aspect for active interrogation efficiency is neutron yield, which is more important for successful resolution than the energy spectrum. The typical approach for improving neutron yield entails producing more neutrons, which has motivated multiple studies using the interaction of increasingly more powerful tabletop lasers with plastic targets to generate protons or deuterons that are absorbed by another target to create neutrons [1]. Alternatively, one may use lenses to focus the neutrons to increase yield rather than simply generating more neutrons with more powerful lasers [2]. Assessing either approach requires a comprehensive model simulating neutron generation and transport to optimize the target material, system geometry, and neutron yield. A complete model from laser source to neutron generation is beyond the scope of the current study, so this project focuses on simulating the interaction of deuterons with typical target materials, such as lithium or beryllium. We use the neutron transport code Monte Carlo N-Particles (MCNP), which applies the Monte Carlo method to track particles [3]. The simulations accurately reflected experimental results from several groups [4]. Future analyses will assess improvements in neutron yield and directionality through strategically incorporating neutron lenses

Advancing the agent methodology to include the higher order of neutron anisotropy with accelerated solutions

With the development of new core designs for generation IV reactors with their complexity and newer fuel designs, the need for consideration of neutron anisotropic scattering is becoming important for enchasing the economy and reliability of these designs. The theory and accurate modeling of neutron anisotropy is one of the most important problems of the transport solution to neutron Boltzmann equation. A number of methods based on careful theoretical developments, were established to numerically determine the effect of anisotropy; some of these methods are: the spherical harmonics method, the so-called function method (FN), the discrete ordinate method, and the Monte Carlo method. The AGENT methodology, based on the method of characteristics, currently the most accurate neutron transport method, represents the state-of-the-art advanced neutronics simulation tool available for 2D, 3D, and full core modeling. The higher order of anisotropic scattering (with no limitation of the number of expansion) is introduced into the AGENT code. An extensive analysis is performed to verify and validate this new model. It is shown that anisotropic scattering is important to be considered for complex geometries due to high angular dependence of neutron flux. The first principle in physics were used to explain the effects of anisotropic scattering (at the level on particle interactions), importance in including the higher moments in flux development for the core designs of high heterogonous structure promoting biased scattering (at the level of heterogeneous reactor assemblies in 2D and 3D). This inclusion of higher order of anisotropic scattering as expected increased the complexity of the mathematical model which in turn increased the computational time. An analysis of the computational time dependence on anisotropic scattering and the method of characteristics resolution parameters are analyzed with accurate predictions of scaling to larger geometries. Finally, an accelerated module was developed for speeding up the solution prediction for anisotropic method of characteristics. The accelerated module has the ability to predict criticality of a heterogeneous system very efficiently and accurately

Radiation shielding aspects for long manned mission to space: Criteria, survey study, and preliminary model

The prospect of manned space missions outside Earth's orbit is limited by the travel time and shielding against cosmic radiation. The chemical rockets currently used in the space program have no hope of propelling a manned vehicle to a far away location such as Mars due to the enormous mass of fuel that would be required. The specific energy available from nuclear fuel is a factor of 106 higher than chemical fuel; it is therefore obvious that nuclear power production in space is a must. On the other hand, recent considerations to send a man to the Moon for a long stay would require a stable, secured and safe source of energy (there is hardly anything beyond nuclear power that would provide a useful and reliably safe sustainable supply of energy). National Aeronautics and Space Administration (NASA) anticipates that the mass of a shielding material required for long travel to Mars is the next major design driver. In 2006 NASA identified a need to assess and evaluate potential gaps in existing knowledge and understanding of the level and types of radiation critical to astronauts' health during the long travel to Mars and to start a comprehensive study related to the shielding design of a spacecraft finding the conditions for the mitigation of radiation components contributing to the doses beyond accepted limits. In order to reduce the overall space craft mass, NASA is looking for the novel, multi-purpose and multi-functional materials that will provide effective shielding of the crew and electronics on board. The Laboratory for Neutronics and Geometry Computation in the School of Nuclear Engineering at Purdue University led by Prof. Tatjana Jevremović began in 2004 the analytical evaluations of different lightweight materials. The preliminary results of the design survey study are presented in this paper