Rishit Gupta

Using OpenMC to Model Shielding for Nuclear Fission Power Space Crafts

Started June 12, 2024

Abstract or project description

Global carbon emissions are increasing every year, and we need clean energy alternatives. This research focuses on modeling a simplified shielding structure for nuclear fission powered aircraft using OpenMC, a simulation software used to simulate neutron transport in nuclear reactors . A Constructive Solid Geometry (CSG) spherical model was used to represent the shielding, and the reactor structure, making high-level assumptions about the model.Nuclear fission-powered aircrafts, demonstrated in the 1950s were used for propulsion, but this was put to a stop due to the nonproliferation and global security concerns. Furthermore, there was not enough shielding to protect the passengers and avionics from harmful radiations, and the government no longer had a need for these aircrafts so they defunded this field. This is important because as we approach the future, CO2 emission rates increase year by year, ultimately worsening climate change. New methods of clean energy and sustainability must be discovered and one method rwould be through using nuclear reactors however these reactors do produce radioactive waste. However, due to recent technologies that can reprocess the radioactive waste, which is recycled and produced into nuclear fuel. Making fission powered aircrafts a sustainable way to travel. Additionally it is a much more efficient form of travel as nuclear fuel can produce a significant amount of energy. Allowing it to operate for a longer period of time without the need of refueling, resulting in much longer haul flights.While this is a much more efficient and cleaner method of air travel, there are safety measures that must be taken to ensure that this type of reactor propulsion technology is safe to operate without portraying any risks to those onboard or any damage to avionics. This paper models a simple geometry using OpenMC, using uranium-238 as our reactor fuel source, and polyethylene and lead as our shielding material. The goal from testing in OpenMC is to assess the effectiveness of shielding that can minimize radiation being emitted from the reactor.