Gamma Shielding Experiment Simulation utilizing MCNPX Code

Experimental investigation requires materials, radiation sources, and test arrangements with a high monetary financial plan. Furthermore, radiation exposure involves people during the experiment. On the contrary, the simulation technique for examining radiation interactions is radio-logically safer, less timeconsuming, cost-effective, and applicable for all desired radiation sources. Through 48.86 mCi 662 keV Caesium-137 gamma-ray source; shielding experiment as well as simulation of it with MCNPX were performed for three shielding materials Lead, Copper, and Aluminum. These materials were placed in front of the gamma source and the emergent radiation was counted in a Geiger- Muller detector to understand the attenuation quality of these materials to each other. These courses of action were simulated utilizing the MCNPX code version 2.7.0 and the results likewise gave and looked at that of the experiment. There are huge similarities of shielding behavior between MCNPX simulation and experiments for the three absorbing materials. The modeled geometry of this MCNPX simulation could be used for future approaches of new designs and structures of radiation shielding, especially where no analogous experimental data exist Journal of Engineering Science 12(2), 2021, 11-21

Nondestructive Evaluation of Hot Cell Liner Bricks by Radiometric Testing

In nuclear facilities, controlling radiation exposure to operating personnel within limits specified by the regulatory body is mandatory; it is required to maintain the radiation level in occupancy areas “as low as reasonably achievable” (ALARA). Highly radioactive sources, materials, and components are enclosed in concrete cells, known as hot cells. Masterslave manipulators (MSMs) are used for the remote handling of equipment located in the hot cell. The MSM is fitted into the hot cell with liner bricks made of lead. The liner brick assembly is a lead poured structure with a steel lining, which gives an effective longitudinal shielding thickness of 250 mm when fitted into the cell with the MSM. Conventional nondestructive testing (NDT) methods such as radiographic testing (RT) and ultrasonic testing have limitations with respect to thickness and coarse-grained structures. Gamma radiometry, which is based on the principle of differential attenuation of radiation, has been developed and adopted in similar applications (Suresh Kumar et al. 2011). It is a challenge to optimize the gamma source strength required for testing, taking into account the sensitivity of the radiation detector and shielding thickness of the component under testing. This paper focuses on the choice of the radiation source and its strength, the experimental approach, the theoretical modeling undertaken for dose-rate estimation, and the actual measurement of radiation during the radiometric testing of MSM liner bricks.

Bayesian algorithm to estimate position and activity of an orphan gamma source utilizing multiple detectors in a mobile gamma spectrometry system

To avoid harm to the public and the environment, lost ionizing radiation sources must be found and brought back under the regulatory control as soon as possible. Usually, mobile gamma spectrometry systems are used in such search missions. It is possible to estimate the position and activity of point gamma sources by performing Bayesian inference on the measurement data. The aim of this study was to theoretically investigate the improvements in the Bayesian estimations of the position and activity of a point gamma source due to introduction of data from multiple detectors with angular variations of efficiency. Three detector combinations were tested—a single 123% HPGe detector, single 4l NaI (Tl) detector and a 123% HPGe with 2x4l NaI (Tl) detector combination—with and without angular efficiency variations for each combination resulting in six different variants of the Bayesian algorithm. It was found that introduction of angular efficiency variations of the detectors did improve the accuracy of activity estimation slightly, while introduction of data from additional detectors lowered the signal-to-noise ratio threshold of the system significantly, increasing the stability and accuracy of the estimated source position and activity, for a given signal-to-noise ratio.

The Mini Labyrinth – a Simple Benchmark for Radiation Protection and Shielding Analysis

The Mini Labyrinth experiment is a simple neutron and gamma shielding experiment developed at STU, inspired by the ALARM-CF-AIR-LAB-001 ICSBEP benchmark experiment. The STU Mini Labyrinth is approximately ten times smaller and consists of NEUTRONSTOP shielding blocks. This paper describes the second version of the Mini Labyrinth experiment and presents the results of the neutron and gamma fields simulation and measurement. The PuBe neutron source with the emission rate of 1.0E7 n.s-1 was utilized in the experiment. The measurement of gamma ambient dose equivalent H*(10) and neutron count rates is performed by the Thermo Scientific RadEye portable survey meter. The simulation part was carried out using the state-of-the-art MCNP6 and SCALE6 MONACO stochastic calculation tools taking into account the detailed geometry of the labyrinth and a combined neutron and gamma source of particles. The comparisons were performed between codes and an experiment, based on the dose rate in the unique detection positions and using a 2D map of neutron and photon fluxes. The propagation of cross-section uncertainties was investigated through the shielding analysis. Partial agreement between codes and measurement was achieved, however serious discrepancies near the PuBe source were identified.

COMPARISON OF CALCULATED AND EXPERIMENTAL METHODS FOR DETERMINING THE ATTENUATION MULTIPLICITY OF A SHIELDED GAMMA-RAY SOURCE IN SPECTROMETRIC ANALYSIS

Проведено сравнение экспериментального и расчетного методов определения кратности ослабления экранированного гамма-источника для проб малого объема. Установлено, что для определения кратности ослабления проб для образцов малого объёма достаточно применять только расчетный метод. Experimental and computational methods for determining the attenuation multiplicity of a shielded gamma source for a small sample are compared. We can use the calculation method to determine the attenuation multiplicity of small volume samples.