Comparison of OSL and TL dosimetry systems against IEC and ICRP standards.

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.

Validation of deterministic code OpenNTP for the analysis of the C5G7-1D benchmark with different configurations.

This study was conducted with the aim of verifying that the OpenNTP (Open Neutron Transport Package) code is capable of reproducing high-performance and exact calculation results. Therefore, a thorough analysis of the One-dimensional C5G7 benchmark with different configuration of control rods was performed. This benchmark was chosen because of its strong heterogeneity. For the modelling of the pin cell geometry, different sets of spatial meshes are used to study the sensitivity of the multiplication factors on the spatial and angular discretization. In addition, the influence of the control rods has been studied by analysing the scalar fluxes, the power in each pin cell and the power of the assemblies. It was found that assembly power distribution errors were significant for the Unrodded case of the C5G7 benchmark. The calculation results in the present paper have a good agreement with the reference values, which demonstrates that the OpenNTP code can solve neutronics problems accurately.

Air-kerma to Hp(0.07), Hp(3), and Hp(10) conversion coefficients for monoenergetic photons.

This work aims for calculating a new set of the personal dose equivalent conversion coefficients Hp(d)/Kair when d = 0.07, 3, and 10 mm, for monoenergetic photons beams ranged from 0.015 to 10 MeV, which have incident on ICRU slab phantom for Hp (0.07) and Hp (10) and cylindrical phantom for Hp (3). The new cylindrical phantom of the ORAMED project was proposed for calculating the eye lens dose equivalent as a cylinder much better approximates the form of a head than a slab. The calculations were carried out with MC and an analytical fit is applied to the data. Our results are found in a good agreement with those previously published, with a local difference less than 1.5%. We have concluded that new analytical fits provide a suitable method for determining conversion coefficients for discrete incident such as MC.

Neutronic study of the 2-MW TRIGA MARK-II research reactor by the deterministic codes DRAGON5 & DONJON5.

The main objective of this work is to perform a neutronic study of the 2 MW TRIGA MARK-II research reactor of the National Centre of Sciences, Energy and Nuclear Techniques (CNESTEN), Rabat, Morocco and then validate the results by comparing the experimental values and those published for an ordinary 2 MW TRIGA MARK II research reactor. The core diffusion code DONJON5 and the lattice code DRAGON5 were coupled to perform a full model of the TRIGA core and their consistency and accuracy were established by benchmarking the TRIGA experiments. In this study, the nuclear data libraries ENDF/B-VII.1 and JEFF3.1 based on 172 energy groups were used. The group constants of all the reactor components were generated using DRAGON5 code and the collision probability method. These group constants were used then in the DONJON5 core code to calculate the multiplication factor, core excess reactivity, total and integral control rods worth as well as power peaking factors. Good agreement found between the calculated and measured results.

Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

Monte Carlo calculation for the development of a BNCT neutron source (1eV-10KeV) using MCNP code.

Different materials have been studied in order to produce the epithermal neutron beam between 1eV and 10KeV, which are extensively used to irradiate patients with brain tumors such as GBM. For this purpose, we have studied three different neutrons moderators (H(2)O, D(2)O and BeO) and their combinations, four reflectors (Al(2)O(3), C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation showed that the best obtained assembly configuration corresponds to the combination of the three moderators H(2)O, BeO and D(2)O jointly to Al(2)O(3) reflector and two filter Cd+Bi optimize the spectrum of the epithermal neutron at 72%, and minimize the thermal neutron to 4% and thus it can be used to treat the deep tumor brain. The calculations have been performed by means of the Monte Carlo N (particle code MCNP 5C). Our results strongly encourage further studying of irradiation of the head with epithermal neutron fields.

Monte Carlo calculation for the development of a BNCT neutron source (1eV – 10 KeV) using MCNP code

Different materials have been studies in order to produce the epithermal neutron beam between 1eV - 10 KeV, which extensively used to irradiate patients with brain tumors such as (GBM). For this purpose we have studied three different neutrons moderators (H2O, D2O and BeO) and their combinations, four reflectors (Al2O3, C, Bi, and Pb) and two filters (Cd and Bi). Results of calculation show, That the best obtained assembly configuration correspond to the combination of the three moderators H2O, BeO and D2O jointly to Al2O3 reflector and two filter Cd +Bi optimize the spectrum of the epithermal neutron at 72 percent, and minimize the thermal neutron to 4 percent and thus it can be used to treat the deep brain tumor. Our calculations have been performed by means of the Monte Carlo N- particle code MCNP 5C.

Seasonal variability in 7Be depositional fluxes at Granada, Spain.

Measurement of 7Be depositional fluxes at Granada, Spain (37 degrees 10'50''N-3 degrees 35'44''W, altitude 670 m) in the period 1995 through 1998 indicates substantial variations between the four seasons and also between corresponding seasons in different years, ranging from 23.6 to 242 Bq m(-2) per season. A strongly positive correlation with precipitation is shown, which explains about 70% of the variations in the 7Be depositional fluxes over the 16 seasons studied. The depositional 7Be flux is on the average highest in the fall and lowest in the summer. The study shows that precipitation primarily controls the 7Be depositional flux and plays a dominant role in the removal of 7Be from the troposphere. The average annual 7Be depositional flux at Granada amounts to 469+145 Bq m(-2).