In order to evaluate the impact of nuclear data on the reactor safety assessment, eigenvalue sensitivity and uncertainty analysis of the Moroccan TRIGA Mark II research reactor are evaluated in this ...paper. Using the SCALE package, a nominal criticality calculation was implemented in this work to assess the accuracy of the developed TRIGA SCALE model. The aim has been reached by comparing several neutronic parameters results with those obtained by the MCNP TRIGA model which has been performed in previous studies.
To generate the sensitivity coefficients, two models of the TRIGA reactor were compared, including the averaged and detailed fuel models, based on the SCALE /TSUNAMI-3D sequence and MCNP/KSEN card. Comparison between those models was performed to evaluate and increase the efficiency of the sensitivity and uncertainty calculations. The obtained results show a good agreement between the TRIGA models studied in this work. Furthermore, the results of this analysis indicate that the uncertainty in the effective multiplication factor (keff) is mainly awarded to uncertainties on the ν‾,χ, fission reactions of 235U as well as the moderator scattering and the clad absorption.
•Neutronic modeling and calculation of the C5G7 Benchmark.•Analysis of the C5G7 Benchmark using the Characteristics Methods (MOC).•Study the effect of differents MOC parameters on the results.•Study ...the effect of type and number of angular quadrature sets on the results.•The Monte Carlo code OpenMC was used to verify and validate the results obtained by DRAGON5 code.
This work uses the Characteristics Methods (MOC) implemented in the lattice DRAGON5 code to analyze the 2D C5G7 MOX Benchmark. This method was performed using the MCCGT: module of DRAGON5, this module includes parameters that affect the results. So, a sensitivity study was necessary to find the optimum parameters. The eigenvalues calculated by DRAGON5 code agree well with the reference Monte Carlo code MCNP, and also with the OpenMC code with a percent error of just 0.001% and 0.004% respectively. Compared to the Monte Carlo code OpenMC, the maximum percentage error of assembly power, the maximum and the average percentage error of pin power are 0.07%, 0.85% and 0.13%, respectively.
Based on the accuracy of the results and on the computational cost, the sensitivity study of various MOC parameters reveals that the 2D C5G7 benchmark problem requires: between 16 and 32 for the azimuthal angles, more than 10 cm-1 for the ray density and 8 × 8 for the special mesh.
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,β) which is a function of momentum transfer and energy transfer parameters α and β 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.
•To study the possible existing phases of zirconium alloys in the fuel and cladding materials of a nuclear reactor.•The calculation of densities of states of phonons the ABINIT code.•The generation of new thermal scattering law S(α,β) libraries using the LEAPR module of NJOY2016 code.•Comparing the effect of our new thermal scattering law data coupled with the effectiveness of considering the different phases of ZrHx on the neutron scattering cross sections.
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.
This research analysis is mainly devoted to enhancing the safe and optimum use of the Center des Etudes Nucléaires de la Maâmora (CENM) TRIGA MARK II research reactor. To serve this purpose, various ...integral neutronic responses, such as the effective multiplication factor k
eff
, the effective delayed neutron fraction β
eff
, the neutron flux distributions at the beam port entrances and the pneumatic transfer system bottom, the pin power peaking factors, the total excess reactivity, the control rod worth, the shutdown margin, and the worth of 11 fuel elements taken from different TRIGA core positions are calculated in order to evaluate the accuracy and the reliability of the developed TRIGA SCALE reactor model. The aim has been fulfilled by comparing the TRIGA SCALE results with those obtained by the MCNP TRIGA model, as well as with some recent experimental measurements from 2021. In general, all the obtained results reveal a good consistency between the SCALE and MCNP TRIGA models studied in this paper. The results analysis indicates also that the B-2 fuel element (Ring B) is the hottest rod among the 101 fuel rods existing in the TRIGA reactor core, which releases a maximum power of 31.67 kW. Furthermore, the total control rod worth, the total core excess reactivity, and the shutdown margin results are also closer to the experimental measurements.
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.
•Simulation of the Moroccan TRIGA Mark-II research reactor.•WIMSD-5B code and CITATION have been utilized to model the core.•Calculations are performed by use of ENDFB-VII.1, JENDL-4.0 and JEFF-3.1.1 libraries.•Calculation results are compared to experiment and MCNP simulation results.
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.
•A seasonal pattern of 7Be and 210Pb was identified in Granada, Spain.•Back trajectory analysis was used to identify the origin and pathway of air masses.•Potential sources regions of 7Be and 210Pb activity were investigated.•Granada is affected by Mediterranean Sea, Atlantic Ocean, and Sahara.•Mediterranean basin is the main source common of 210Pb and 7Be.
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).
•A new Deterministic package called NTP-ERSN is developed.•An open-source code with a pedagogical purpose to solve the steady-state multi-group NTE.•This package is provided with GUI.•NTP-ERSN is easy to use for user.•NTP-ERSN is a simple framework to add and test new algorithms.
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 calculations of the TRIGA MARK II research reactor.•Validation of the deterministic transport code DRAGON5 and diffusion code DONJON5.•Calculation and analysis of control rod worth, excess reactivity as well as power peaking factors by deterministic codes.•The good consistency of the results ensures that a thermal-hydraulic analysis will be performed for TRIGA reacteor.•To rely on DRAGON5 and DONJON5 codes for TRIGA Mark-II calculations.
