This paper presents a new procedure to optimize the geometric parameters of a n-type coaxial HPGe detector. It is based on a statistical technique called “Design of Experiments” (DoE). This technique ...aims to identify the most influential parameters and to determine the optimal configuration. In this work, The effects of each parameter on the detector responses have been investigated by a fractional factorial design. Only the most influential factors contributing to the detector response have been selected. Precise modeling of these factors was then performed using a full factorial design. Based on the results obtained from this design, the full energy peak efficiencies according to the geometric parameters were modeled by a multiple-linear regression. These models have been statistically validated by analysis of variance (ANOVA). The optimal combination of the geometric parameters has been identified using the desirability function approach, which is a useful tool to optimize multi-response problems. A verification test was performed to validate the results obtained. It was observed that the relative deviation found between experimental and simulated values was less than 5%.
•Optimization of geometrical dimensions of an HPGe detector.•Application of Monte Carlo simulation and “Design of Experiments” technique.•Investigating influence of each detector parameter on the FEPE.•Achieving good agreement between the measured and the simulated results.
Personal and environmental radiation monitoring services are widely used through luminescent techniques. In this paper, we practiced performance testing on thermoluminescent and optically stimulated ...luminescent dosimeters by assessing their homogeneity, linearity, energy, and angular dependence tests. The IEC and ICRP requirements were used to compare the performance response of dosimeters. Based on the experimental results, we realized that both detectors comply with the international criteria. The homogeneity percentage was 8.9% and 13.7% for TL and OSL detectors, respectively. The percentage deviation of the linearity test does not exceed 10% for both dosimeters except for the TL dosimeters at low irradiation dose. For the angular dependence, deviations were less than 2% for TLDs and 5% for OSLDs. These detectors display mean values of the relative energy response of −15.29% and −6.51% for OSL and TL detectors. Generally, TL materials manifested low sensitivity to radiation dose levels. On the other hand, the OSLDs demonstrated a more pronounced under-response to energy beam qualities than TLDs. Regarding COV tests, TL and OSL dosimeters have passed the c2 test.
•TL and OSL dosimeters characteristics were tested.•Nine radiation qualities and eleven angles of incidence were used to test homogeneity, non-linearity, energy, and angular dependence.•The Hp(10) measurement accuracy was evaluated by both the ICRP trumpet curve analysis and IEC 62387 covariance test.
This work aims to establish some X-ray qualities recommended by the International Standard Organization (ISO) using the half-value layer (HVL) and Hp(10) dosimetry approaches. The HVL values of the ...following qualities N-60, N-80, N-100, N-150 and N-250 were determined using various attenuation layers. The obtained results were compared to those of reference X-ray beam qualities and a good agreement was found (difference less than 5% for all qualities). The GAMOS (Geant4-based Architecture for Medicine-Oriented Simulations) radiation transport Monte Carlo toolkit was employed to simulate the production of X-ray spectra. The characteristics HVLs, mean energy and the spectral resolution of simulated spectra have been calculated and turned out to be conform to the ISO reference ones (difference less than the limit allowed by ISO). Furthermore, the conversion coefficients from air kerma to personal dose equivalent for simulated and measured spectra were fairly similar (the maximum difference less than 4.2%).
In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle ...MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), U235(nν¯) and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.
The main objective of this study is to assess the neutronic modeling and calculations of the nuclear heating reactor NHR-5. To this end, the neutronic parameters of NHR-5 reactor underwent a ...comparative and validation protocols/methods through the analysis of the finite multiplication factor keff and some neutronic parameters including D,νΣf,Σa as well as the power and flux distributions, using the evaluated nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0 based on 172 energy groups.
In this study, the collision probability approach is used to simulate and calculate the neutronic parameters in NHR5 reactor using the deterministic core diffusion code DONJON5 and the lattice transport code DRAGON5. The group constants of the reactor components were produced by DRAGON5. The DONJON5 code was then used to calculate the effective multiplication factor, excess reactivity, and power and flux distributions by introducing these group constants. The results are compared to those of the experiment to validate the calculation scheme. The results of the simulations reveal that the computed neutronic parameters consistently produce reasonable and consistent results of flux and power distributions, excess reactivity and all other parameters when compared to the experiment values. Therefore, the reactor models of NHR5 developed by DRAGON and DONJON codes were good in predicting the effective multiplication factor as well as the studied neutronic parameters. In this study, we being able to show the potential validation of the reactor physics lattice transport code DRAGON5 and the core diffusion code DONJON5, as well as the nuclear data libraries ENDF/B-VII.1, JEFF3.1 and JENDL4.0.
•Neutronic Modeling and calculations of the Nuclear Heating Reactor NHR-5.•Validation of the deterministic transport code DRAGON5 and diffusion code DONJON5.•Calculation and analysis of excess reactivity as well as the power and flux distributions by deterministic codes.•To rely on DRAGON5 and DONJON5 codes for NHR-5 calculations.•The good consistency of the results ensures that a thermal-hydraulic analysis will be performed for NHR reactor.
This work aims to establish some X-ray qualities recommended by the International Standard Organization (ISO) using the half-value layer (HVL) and Hp(10) dosimetry approaches. The HVL values of the ...following qualities N-60, N-80, N-100, N-150 and N-250 were determined using various attenuation layers. The obtained results were compared to those of reference X-ray beam qualities and a good agreement was found (difference less than 5% for all qualities). The GAMOS (Geant4-based Architecture for Medicine-Oriented Simulations) radiation transport Monte Carlo toolkit was employed to simulate the production of X-ray spectra. The characteristics HVLs, mean energy and the spectral resolution of simulated spectra have been calculated and turned out to be conform to the ISO reference ones (difference less than the limit allowed by ISO). Furthermore, the conversion coefficients from air kerma to personal dose equivalent for simulated and measured spectra were fairly similar (the maximum difference less than 4.2%).
In order to study the sensitivity and the uncertainty of the Moroccan research reactor TRIGA Mark II, a model of this reactor has been developed in our ERSN laboratory for use with the N-Particle ...MCNP Monte Carlo transport codes (version 6). In this article, the sensitivities of the effective multiplication factor of this reactor are evaluated using the ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0 libraries and in 44 energy groups, for the cross sections of the fuel (U-235 and U-238) and the moderator (H-1 and O-16). However, the quantification of the uncertainty of the nuclear data is performed using the nuclear code NJOY99 for the generation and processing of covariance matrices. On the one hand, the highest uncertainty deviations, calculated using the ENDFB-VII.1 and JENDL4.0 evaluations, are 2275, 386 and 330 pcm respectively for the reactions U235(n, f), $ U_{235}(n\bar{\nu})$ and H1(n, γ). On the other hand, these differences are very small for the neutron reactions of O-16 and U-238. Regarding the neutron spectra, in CT-mid plane, they are very close for the three evaluations (ENDF/B-VII.0, ENDF/B-VII.1 and JENDL-4.0). These spectra present two peaks (thermal and fission) around the energies 0.05 eV and 1 MeV.
When a radiotracer is injected into a patient’s body as part of a nuclear medicine investigation, the entire body is exposed to the ionizing radiation emitted, which can cause biological damage. ...Therefore, it is important to predict the internal radiation dose to properly balance the advantages of radiological examinations. Currently, various Monte Carlo tools, such as MCNP, Geant4, and GATE, are available to estimate internal radiation dosimetry-related quantities, such as
S
values (
S
) and specific absorbed fractions (SAF). Such codes make physics easier for physicists who are experienced with computer programming; however, programming and/or simulation inputs remain a time-consuming and intensive task. In this study, we present a newly developed Geant4-based code for internal dosimetry calculations, namely “DoseCalcs”. To assess the performance of the geometrical methods and computational capabilities of our developed tool, we used the GDML, TEXT, STL, and C++ methods to model the ORNL adult phantom, and a voxel-based structure to construct the ICRP adult male. SAFs in the ORNL and ICRP adult male phantoms for eight discrete mono-energetic photons with energies ranging from 0.01 to 2 MeV are calculated with DoseCalcs and compared to ORNL and OpenDose reference data. The two phantoms showed good agreement with both references, which indicates the accuracy of DoseCalcs for subsequent use in estimating internal dosimetry quantities using a variety of geometrical methods.
Knowledge of SAF at different energies is crucial for internal dosimetry. For this purpose, a set of calculated SAF values for a mouse voxelized phantom's selected organs are presented below. Values ...of SAF were calculated for mono-energetic photons and electrons with energy varying from 10 keV to 4 MeV using the Monte Carlo simulation via GATE/GEANT4 code (GEANT4 Application for Emission Tomography). The heart, liver, lungs, kidneys, and spleen were considered as the source organs from which the particles were released. Then, the estimated results were compared to those calculated in a previous study using EGS4 code. It is indicated that the obtained results are in good agreement with the reference values for all energies of photons and electrons, with discrepancies less than 9% and 5% for self-irradiation and cross-irradiation, respectively.
•Photon and Electron SAFs were calculated in some organs of Digimuse phantom using GATE/GEANT4 code.•The ICRU publication number 44 served as the reference for the organ compositions and densities used in this study.•Photon SAFs were compared to results obtained with the EGS 4 code.•Between the two series of measurements, there was a very good accord.
Current legislation mandates the inspection and calibration of operational survey radiation monitoring instruments used in nuclear medicine, radiotherapy departments, and other fields utilizing ...ionizing radiation sources. To comply with national and international radiation protection standards, Morocco's National Secondary Standard Dosimetry Laboratory provides reliable calibration results with high accuracy and covers various measurement ranges using attenuators provided by the automated Gamma G10 irradiator or validated beam qualities produced by the X-ray irradiator type X80–320 kV.
This study aims to develop a digital graphical user interface using Python programming language, designed for calibrating radiation protection measuring instruments . The interface is intended to facilitate all operations and calculations related to determining calibration factors and measurement uncertainties in accordance with the ISO 4037 standard, ensuring minimal processing time and minimizing potential error sources . The interface enables calculations to be recorded, as well as the establishment and electronic archiving of the calibration certificate and the report in PDF format using the Hypertext Preprocessor FPDF library (PHP FPDF). With the development of this interface, multiple instruments can be processed per day with high accuracy, streamlining the calibration process and improving efficiency.
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The importance of compliance with international standards to ensure the quality and reliability of measurements in radiation protection was examined.
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Description of X-ray and Gamma-ray irradiators designed for the calibration of radiation protection measuring instruments within the Secondary Dosimetry Calibration Laboratory (SSDL) which is a member of the WHO/IAEA network within the National Center for Radiation Protection of Morocco
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Graphical User Interface using python for the calibration of photon measurement instruments for radiation protection purposes was developped.
Image, graphical abstract
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