New open-source software for gamma-ray spectra analysis.

Many commercially available software packages have been created to analyze gamma-ray spectra, but their source code has generally not been shared, although some users may wish to add or modify certain functionality, which is impossible without access to the source code. This study therefore presents a new open-source software package for the analysis of gamma-ray spectra. The name of the software is GSA (Gamma-ray Spectra Analysis), the source code of which is freely available through the GitHub website (https://github.com/LAHCEN-EL-AMRI/Gamma-Spectra-Analysis). The main function of this initial version of the software is to locate peaks, calculate areas, and identify corresponding radionuclides. A future version will complement this by measuring the concentrations of radionuclide elements. The software was validated by comparing its analysis results with those generated by three other software programs, namely Genie 2000, Maestro, and FitzPeaks. All the formulas used are explained in this work, which could be useful for researchers or students looking to create their own software packages for analyzing gamma-ray spectra.

Modeling and simulating the induced effect on the steel collimator plug used in the PGAA facility of the Moroccan TRIGA Mark-II reactor under different neutron irradiation levels.

The main objective of this study is to assess the behavior of the steel collimator plug (steel plug) dedicated to the Prompt Gamma Neutron Activation Analysis (PGAA) facility in the neutron beam tube (NB1) of the 2 MW Moroccan TRIGA Mark-II research reactor. The main function of this steel plug is to reduce the neutron and gamma beam cross section from 15 cm to 5 cm in diameter. This steel plug plays a crucial role in reactor safety because it replaces the original neutron beam plug while also stopping the entire incoming neutron beam. Three aspects were therefore involved in this study, including: i) the released heat induced by both neutron and gamma radiation, ii) the swelling effect caused by both radiation and temperature increases in the steel and iii) the radioactivity induced by neutron radiation in the steel plug. An MCNP6.2 model for the TRIGA Mark-II reactor was used to calculate the neutron spectrum inside the NB1 beam tube at the inlet part of the steel plug. A gamma spectrum of a 900 MW PWR reactor was used as input to account for the gamma radiation's effects on the steel plug. In order to select a convenient steel, two investigations were carried out for two grades of steel, namely mild steel (E235) and 304L stainless steel (SS304L). The results were calculated using an in-house FORTRAN-based program and validated using COMSOL, SPECTER codes, and "Neutron Activation and Scattering Calculator" tools. The results revealed that both the E235 and 304L steels are convenient from a safety point of view, although the E235 steel is preferred at the decommissioning stage.

Processing of JEFF-3.3 and ENDF/B-VIII.0 and testing with critical benchmark experiments and TRIGA Mark II research reactor using MCNPX.

A comparative study has been performed with the latest evaluated nuclear data libraries JEFF-3.3 and ENDF/B-VIII.0. The study has been conducted through the benchmark calculations for 120 criticality problems and the TRIGA Mark II research reactor with the libraries processed using NJOY21 for MCNPX Monte Carlo transport code. The criticality benchmarks assemblies, taken from the ICSBEP benchmark, cover Uranium (highly enriched uranium, intermediate enriched uranium, low enriched uranium, and233U) and Plutonium fuel systems in a various metal forms, and under a various spectral conditions. The Moroccan TRIGA Mark II research reactor calculation is used to look into the predictive capability of those nuclear data libraries and then to compare the accuracy of the predicted results with the experimental data published elsewhere. Actually, the purpose of this study is to investigate some neutronic and kinetic parameters of those benchmarks for both libraries. The former consist of effective multiplication factor, heat distribution, neutron flux distribution, effective delayed neutron fraction (ßeff), prompt removal lifetime (τr) and the mean neutron generation time (Λ). The results show that the calculated effective multiplication factor, heat distribution, neutron flux distribution, and the kinetic parameters are in good agreement with references. However, it is found that the computed values are strongly depending on the nuclear data set used in calculations.