PSCF and CWT methods as a tool to identify potential sources of 7Be and 210Pb aerosols in Granada, Spain.

This research is focused on studying the preferred source regions and the pathways of the air masses with high particulate concentrations impacting on the activity concentrations of 7Be and 210Pb aerosols in Granada atmosphere. For this purpose, three different source-receptor methods have been used: Cluster Analysis, Potential Source Contribution Function (PSCF), and Concentration Weighted Trajectory (CWT). Air filter samples were weekly collected and analysed in Granada university (Spain 37.177N, 3.598 W, 687m a.s.l.) during 12 years (2006-2017) for the activity concentration of 7Be, and during 5 years (2010-2014) for the one of 210Pb. The time series of the collected data indicate that the concentration of both radiotracers present a cyclical and seasonal pattern, in association with their origins and atmospheric conditions. Clustering analysis showed that the air masses arriving to Granada can be classified as: (1) tropical continental air masses coming from the Mediterranean Sea, (2) tropical and warm polar maritime air masses produced over the Atlantic Ocean, and (3) continental air masses originated over Europe and Northern Africa. The PSCF and CWT methods confirmed that the main source areas of 7Be are located in the Atlantic coast of southern Morocco, and Northern Africa. On the other hand, southern France and the Algerian desert were found to be the main region sources of 210Pb. In addition, the Mediterranean Basin has been postulated as a strong source region for 7Be and 210Pb. Furthermore, the PSCF and CWT models show that the regions with larger 7Be/210Pb ratios are located in the Atlantic Ocean, due to frequent stratospheric intrusions specially during the winter months.

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.

First principles calculation for generating new thermal neutron scattering data for Zirconium hydride.

Lattice parameters of materials have the same magnitude as the energy of thermal neutrons in reactors, which directly affects the neutron cross section and its energy. While they are thermalized, incident neutrons can lose or gain energy during their interactions with materials components. Since several decades, methods and models were developed in the aim to generate nuclear data sub-libraries required in correcting neutrons interactions cross sections at thermal energies. However, very few experimental works were dedicated to this field. In this paper we focus our efforts on reviewing the theoretical models and their adequacy in describing thermal scattering events in the aim of proposing new formalisms to calculate the density of states (DOS) and phonon responses of zirconium hydride material, which constitutes an important moderator of neutrons in TRIGA reactors fuel elements. Generally the effects of thermal scattering are provided in nuclear data evaluations by a thermal sub-library ENDF file 7. Data in file 7 are described by the known thermal scattering law S(α,ß) which is a function of momentum transfer and energy transfer parameters α and ß respectively. The thermal scattering law has been used to calculate the double differential cross sections and the corresponding results are presented. Although the comparison with other models shows satisfactory results, no previously personalized use of data may be the raison of its usefulness in some cases and not in others.

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.

An analytical fit and EGSnrc code (MC) calculations of personal dose equivalent conversion coefficients for mono-energetic electrons.

This study aims for calculating a new set of the personal dose equivalent conversion coefficients Hp(d)/Φ when d = 0.07, 3, and 10 mm, for mono-energetic electron beams ranged from 0.06 MeV to 50 MeV for Hp(0.07), from 0.7 MeV to 50 MeV for Hp(3) and from 2 MeV to 50 MeV for Hp(10), which has incident on ICRU slab phantom. Additionally, we have calculated the conversion coefficients of a dose to the lens eye of into a new cylindrical phantom of the ORAMED project. The cylindrical phantom was proposed for calculating the eye lens dose equivalent as a cylinder much better approximates the form of a head than a slab. The ICRU tissue of density of 1 g/cm3 which is consists of 4-elemental has been used in slab and cylindrical phantoms. The calculations were carried out with EGSnrc code Monte Carlo (MC) and a new an analytical fit is applied to the data. Our results are found in a good agreement with those published by Veinot and Hertel (2012)) , with a local difference less than 1.5%. We have concluded that new analytical fits provide a convenient approach for determining conversion coefficients for discrete incident such as MC.

The discrete ordinate method for angular flux calculations in slab geometry.

In many deterministic methods, concerning the resolution of the neutron transport equation, a more global and practical representation of the angular flux is needed to provide us with useful and complete information on the neutron population in a reactor core. The purpose of this paper is to provide a reference quality calculation for the angular flux. The discrete ordinate method (S N ) is processed with a new matrix form which is used to model an isotropic and anisotropic multiplicative system in a region, multi-region for one energy group in a cartesian geometry. The obtained results are compared to those obtained by MCNP6 code.

Modeling and simulation of a TRIGA MARK-II research reactor using WIMSD-5B and CITATION codes.

The main objective of this study is to analyse neutronic safety parameters of the Moroccan TRIGA Mark-II research reactor using the WIMSD-5B and CITATION computer codes. New 172-group libraries of multi-group constants for the lattice code WIMSD-5B have been generated for all isotopes presented in the TRIGA reactor core by processing nuclear data from ENDFB-VII.1, JENDL-4.0 and JEFF-3.1.1 using NJOY99. The lattice code WIMSD-5B was employed to generate multi-group cross sections in the suitable format that will be used by the 3-dimensional diffusion code CITATION. This later was used to calculate various neutronic safety parameters of the TRIGA Mark-II research reactor, such as reactivity excess and neutron fluxes profiles. The results of these calculations are compared to the results of Monte Carlo calculation based on MCNP code. A good agreement is achieved and the current computation scheme will be adopted for our further coupling neutronic/thermal-hydraulic study of the Moroccan TRIGA reactor.

NTP-ERSN: A new package for solving the multigroup neutron transport equation in a slab geometry.

The package, called NTP-ERSN (N eutron T ransport P ackage from the R adiations and N uclear S ystems G roup), is an open-source code written in FORTRAN90 for a pedagogical purpose to solve the steady-state multigroup neutron transport equation. This package is based on three classical methods, namely the collision probability (CP) method, the discrete ordinates (SN) method and the method of characteristics (MOC). These methods are employed to obtain scalar and angular flux distributions in homogeneous and heterogeneous slab geometry with isotropic and anisotropic scattering. The source code algorithms are very simple to be comprehensive by engineering students. In addition, NTP-ERSN is a simple framework to add and test new algorithms. On the other hand, a graphical user interface written in Python programing language has been developed to simplify the use of NTP-ERSN. Numerical results are given to illustrate the NTP-ERSN code's accuracy. Finally, the present software can be useful as an academic tool for teaching reactor physics. It is freely available for download on GitHub (https://github.com/mohamedlahdour/NTP-ERSN).

7Be spatial and temporal pattern in southwest of Europe (Spain): Evaluation of a predictive model.

This study presents a comprehensive statistical analysis of the cosmogenic radioisotope 7Be measured in surface air in ten stations over Spain for a 9 years period (from January 2006 to December 2014). Besides the analysis of 7Be seasonal and inter-annual variability, 7Be frequency distributions and its correlations with meteorological variables observed in the 10 sampling sites were investigated. The second part of this paper focuses on a feasibility study for the application of the Artificial Neural Networks (ANNs) to predict monthly 7Be activity concentrations using meteorological variables, PM10 concentrations and the sunspot number as input parameters. Notwithstanding the low correlations found between 7Be and input parameter, the performance of the ANNs, as evaluated by the relevant statistical parameters, demonstrates their capability to correctly predict 7Be monthly activities in the 10 Spanish sampling sites.

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 analysis of the influence of germanium dead layer thickness on the HPGe gamma detector experimental efficiency measured by use of extended sources.

We have carried out a study to figure out the influence of crystal inactive-layer thickness on gamma spectra measured by an HPGe detector. The thickness of this dead layer (DL) is not known (no information about it was delivered by the manufacturer) due to the existence of a transition zone where photons are increasingly absorbed. To perform this analyses a virtual model of a Canberra HPGe detector was produced with the aid of MCNPX 2.7 code. The main objective of this work is to produce an optimal modeling for our GPGe detector. To this end, the study included the analysis of the total inactive germanium layer thickness and the active volume that are needed in order to obtain the smallest discrepancy between calculated and experimental efficiencies. Calculations and measurements were performed for all of the radionuclides included in a standard calibration gamma cocktail solution. Different geometry sources were used: a Marinelli and two other new sources represented as S(1) and S(2). The former was used for the determination of the active volume, whereas the two latter were used for the determination of the face and lateral DL, respectively. The model was validated by comparing calculated and experimental full energy peak efficiencies in the 50-1900keV energy range. the results show that the insertion of the DL parameter in the modeling is absolutely essential to reproduce the experimental results, and that the thickness of this DL varies from one position to the other on the detector surface.

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).

Seasonal 7Be concentrations in near-surface air of Granada (Spain) in the period 1993-2001.

Near-surface air samples for assessment of seasonal 7Be levels were taken weekly at Granada, Spain, (37 degrees 10'50"N-3 degrees 35'44"W, altitude 670 m) in the period 1993-2001. The 7Be levels measured reflect strong seasonal trends. Concentrations during autumn and/or winter showed a minimum (ca. 1.5 mBq/m3), and in each year a fallout maximum was found in summer (ca. 7.6 mBq/m3). Multiple regression of 7Be concentrations with temperature and rainfall, and with sunspot number explained about 71% of the variance in the 7Be data. The data reveal strong variations in the mean annual 7Be concentrations due to variations in the flux of cosmic galactic primary radiation caused by the 11-year sunspot cycle. A difference of a factor of 2.6 was found between the highest level (5.8 mBq/m3) obtained in 1996 (beginning of the solar cycle 23) and the lowest (2.6 mBq/m3) obtained in 2000 (maximum of the solar cycle 23).