Utilizing a polymer-based radiation shield offers lightweight, low cost, non-toxic compared to lead and solution for eliminating generated secondary neutrons. Incorporating silicon (i.e., one of the most abundant elements) in new applications, such as shielding, would have an impact on the economy and industry. In this study, seven potential shielding materials, composed of silicon, silicon carbide, and boron carbide embedded ethylene vinyl acetate (EVA) copolymers, are proposed. The shielding performance of these composite materials, including the attenuation coefficients (µ), the mass attenuation coefficients (µ_m), the half value layer (HVL), the mean free path (MFP), and the radiation protection efficiency (RPE) were examined using photon beams. Measured µ_m were verified against the calculated values. The averaged agreement was within ±7.4% between the experimental measurements and the theoretical calculation results. The HVL and MFP measured values for the polymer composites were lower than that for the pure EVA polymer, indicating the fillers in the polymers enhanced the shielding performance. The EVA + SiC (30%) and EVA + Si (15%) + B₄C (15%) composites required the lowest thickness to stop 50% of the incident photons. The evaluation of experimental results of the RPE revealed that the polymer composites containing SiC (30%), Si (15%) + B₄C (15%), or SiC (15%) + B₄C (15%) succeeded in blocking 90–91% of X-rays at nearly 80 keV. However, a thicker shield of the proposed composite materials or combined layers with other high-Z materials could be used for higher energies.

A CS-30 cyclotron has been in operation at King Faisal Specialist Hospital and Research Center (KFSHRC) since 1982. The CS-30 cyclotron has been used to produce medical radioisotopes for positron emission tomography (PET) and single-photon emission computed tomography (SPECT). Some of the nuclear reactions of radionuclide production are associated with the intense release of a wide range of fast neutrons. In this work, we investigated the radionuclides produced from neutron interactions with the cyclotron facility walls. Activation isotopes were determined by performing gamma ray spectrometry utilizing a high-purity germanium (HPGe) detector. The major radionuclides found were ¹⁵²Eu, ¹⁵⁴Eu, ¹³⁴Cs, ⁶⁵Zn and ⁶⁰Co. Activation isotope accumulation had increased the dose rate inside the facility. The surface dose rates were measured at all of the surrounding walls. The maximum surface dose rate was found to be 1.2 µSv/h, which is much lower than the permissible occupational exposure of 15 µSv/h based daily 5 work hours.

Utilization of nuclear research reactors is of high importance for education and training, research and development, and many other applications. However, less effective utilization encountered in research reactors is mainly due to limitations in power levels and related experimental facilities. Such limitations, however, have led different global owners of research reactors to consider upgrading the power levels of their reactors to accommodate the increase in utilization demands. To consider upgrading the power levels of research reactors without replacing major components, a pair of essential analyses must be performed, namely the neutronic evaluation of nuclear fission and thermal-hydraulic evaluation for heat removal from the reactor core. In this work, a conceptual upgrade to the core design and configuration of MSTR, or Missouri University of Science and Technology Reactor (200 kilowatts (kW)), is demonstrated. The conceptual design of the MSTR high-power configuration (MSTR-HPC) aims to achieve high neutron flux and demonstrate the power level and core configuration with greater flexibility and adaptability while not exceeding safety limits. The conceptual design of the MSTR-HPC involves uprating the power level to 2 megawatts (MW), reconfiguring the core, changing the fuel meat type, inclusion of a flux trap (FT) facility, and others. In addition, the conceptual design of MSTR-HPC includes three in-core irradiation facilities, namely FT, bare rabbit tube (BRT), and cadmium rabbit tube (CRT). The neutronic evaluation of the MSTR-HPC was carried out using the Monte Carlo N-particle Code (MCNP), version 6. In addition, the thermal-hydraulic behavior of MSTR-HPCs’ hot-channel has been assessed by means of ANSYS Fluent to evaluate the satisfaction of thermal-hydraulic safety requirements of the conceptual design. The results obtained have shown that the conceptual MSTR-HPC has demonstrated a maximum neutron flux obtained higher than that obtained in the current MSTR core by two orders of magnitude. The conceptual design of MSTR-HPC with composite BeO/graphite reflector blocks was able to sustain critically and operate continuously at full power for 61 days. In regards to the hottest fuel plate of MSTR-HPC, the results have shown that the determined temperature for the fuel plate regions was below the safety limits. In addition, at full operation power of the MSTR-HPC, the mass flow rate of 39.86 kg/s (10.644 gallon/s) was sufficient for removing the generated heat. In conclusion, the conceptual design of the MSTR-HPC has demonstrated its flux enhancement capabilities while maintaining safety limits, which are of high importance in enhancing reactor utilization for a larger window of time.

Changes in the thermal power of a nuclear research reactor will lead to changes in experimental, irradiation, and testing conditions. Consequently, reactor core parameters are inevitably susceptible to changes. One such parameter is gamma heating (GH), which results from gamma interaction with materials. In this work, a gamma thermometer was used to measure GH over the course of 7 operational days and nights. In addition, the Monte Carlo reactor physics code Serpent-2 was used to evaluate the sensitivity of common detection methods for monitoring reactor core parameters such as neutron fluxes, GH, and gamma flux under the following conditions: reactor core power variation, reactor core fuel shuffling, and detector vicinity fuel assembly shuffling. The GH values obtained through measurements and calculations were linearly proportional to the reactor power. In addition, the Serpent-2 code for the McMaster nuclear reactor showed that despite maintaining the reactor power core at the same level, the fuel burnup distribution could alter the studied parameters.