This paper focuses on the development of a new computational model of the CNESTEN’s TRIGA Mark II researchreactor using the 3D continuous energy Monte-Carlo code TRIPOLI-4 (T4). This new model was ...developed toassess neutronic simulations and determine quantities of interest such as kinetic parameters of the reactor,control rods worth, power peaking factors and neutron flux distributions. This model is also a key tool used toaccurately design new experiments in the TRIGA reactor, to analyze these experiments and to carry out sensitivityand uncertainty studies. The geometry and materials data, as part of the MCNP reference model, were usedto build the T4 model. In this regard, the differences between the two models are mainly due to mathematicalapproaches of both codes. Indeed, the study presented in this article is divided into two parts: the first part dealswith the development and the validation of the T4 model. The results obtained with the T4 model were comparedto the existing MCNP reference model and to the experimental results from the Final Safety Analysis Report(FSAR). Different core configurations were investigated via simulations to test the computational model reliabilityin predicting the physical parameters of the reactor. As a fairly good agreement among the results wasdeduced, it seems reasonable to assume that the T4 model can accurately reproduce the MCNP calculated values.The second part of this study is devoted to the sensitivity and uncertainty (S/U) studies that were carried out toquantify the nuclear data uncertainty in the multiplication factor keff . For that purpose, the T4 model was used tocalculate the sensitivity profiles of the keff to the nuclear data. The integrated-sensitivities were compared to theresults obtained from the previous works that were carried out with MCNP and SCALE-6.2 simulation tools anddifferences of less than 5% were obtained for most of these quantities except for the C-graphite sensitivities.Moreover, the nuclear data uncertainties in the keff were derived using the COMAC-V2.1 covariance matriceslibrary and the calculated sensitivities. The results have shown that the total nuclear data uncertainty in the keff isaround 585 pcm using the COMAC-V2.1. This study also demonstrates that the contribution of zirconium isotopesto the nuclear data uncertainty in the keff is not negligible and should be taken into account when performingS/U analysis.
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 National Center for Energy, Sciences and Nuclear Techniques (CNESTEN)'s Training Research and Isotope Production General Atomics (TRIGA) Mark II is a pool-type light water moderated and cooled ...research reactor operating at a maximum steady state thermal power of 2 MW. The reactor was designed to be used as a training facility for reactor operators, neutron activation analysis, isotopes production, and for implementing different reactor physics experiments. This article deals with the numerical and experimental characterization of reaction rates (RRs) in different irradiation channels inside the CNESTEN's TRIGA Mark II research reactor, located in Rabat/Morocco. The main objective of this study is to validate the calculated neutron RRs against the measured ones and to prove that the new TRIPOLI-4 model of the reactor is capable to reproduce the measured quantities. Therefore, the measurements were carried out using the neutron activation technique and gamma spectrometry measurements. Preliminary simulations were performed with TRIPOLI-4 transport Monte Carlo code to establish the experimental design and set up for activation foils experiments. The selected set of foils with known characteristics were irradiated, at different power levels, inside the irradiation facilities of the TRIGA reactor. The resulting activities were evaluated via <inline-formula> <tex-math notation="LaTeX">\gamma </tex-math></inline-formula> spectrometry measurements. Normalized calculated and measured RRs were compared, and a good agreement was shown for most nuclides, which indicates that the new detailed TRIPOLI-4 model of the TRIGA reactor can accurately predict the relative experimental RRs values. Further work is ongoing to analyze absolute RR values, as well as to carry measurements in other irradiation channels.
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.
The National Center for Nuclear Energy, Sciences and Technology (CNESTEN) located in Rabat, Morocco, operates a 2 MW TRIGA Mark II research reactor. This type of reactor is specially designed to ...effectively implement the various fields of nuclear research such as neutron activation analysis, neutron radiography, detectors testing, radioisotopes production as well as education and training. In the last few years, a collaboration between the French Atomic Energy and Alternative Energies Commission (CEA) and the CNESTEN was established to expand the utilization of the TRIGA computational model by carrying out new in-situ measurements in order to characterize neutron and photon fields within and beyond the TRIGA reactor core. These new measurements will consolidate the knowledge of neutron and photon fluxes in different irradiation and instrumentation channels. The results of these experiments will also be used to extend the experimental validation of the new developed TRIPOLI-4® computational model of the reactor and to quantify the uncertainties and biases. This paper focuses on the neutron and photon flux characterization of two irradiation channels inside the TRIGA reactor core. Neutron measurements are ensured by activation dosimetry whereas photon measurements are ensured by thermo-luminescent detectors (TLD400 – CaF
2
: Mn). Based on these techniques, the experiments were carried out during an experimental campaign conducted in June 2022. Preceding the implementation of these measurements, two experimental devices were specifically designed and manufactured to ensure a reproducible positioning of the detectors in the selected irradiation channels. This paper presents the experimental results analysis and the associated uncertainty quantification. These results will be then compared to the calculation ones obtained by the computational models of the TRIGA reactor.
This paper presents a case study of applying Genetic Algorithm (GA) coupled with Monte Carlo N-Particle Transport (MCNP) and PARET codes for a thermal-hydraulic and safety analysis to optimize the ...fuel reload for the TRIGA Mark II Moroccan research reactor. Based on the radial distribution of the 238U burnup ratio inside the reactor core, the five most burned fuel elements were replaced by others fresh fuel elements (12 % wt of uranium) using the Multi-Objective Genetic Algorithms (MOGA) method. Three aspects for the fuel reload optimization were considered in this study including 1) maximization of the effective multiplication factor (Keff), 2) minimization of maximum Centre Fuel Temperature (CFT) and 3) maximization of the Departure from Nuclear Boiling Ratio (DNBR).
The GA programming process developed in this work was adapted to handle the constraints concerning the safety limits for the successive core configurations (CCs) automatically generated by the code. MOGA method works with an elitist selection based on the Binary Tournament Selection (BTS) method, a modified two-point crossover and a simple mutation operator. The results obtained indicate that the MOGA can successfully find an optimal CC with a Keff of 1.03498, a maximum CFT of 554 °C and a DNBR of 2.94 when five fresh fuel elements are inserted. The variation of neutron fluxes with respect to radial distance for the best CC and the fresh core was illustrated.
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•Optimum core configurations were generated for the Moroccan TRIGA Mark II research reactor.•Calculations were performed using a global reactor calculation, formed by MCNP and PARET codes.•Genetic Algorithms were used to optimize three parameters at once.•Keff and the safety constraints on the CFT and DNBR were considered for optimization.•The optimal core configuration was determined and compared with the fresh core map.
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.
•A new irradiation channel (IC) was inserted in the reactor core.•The neutronic and thermal hydraulic parameters were analyzed.•Three new configurations was proposed for this study.•The introducing of the IC don’t affect to the safety parameters.
The CNESTEN (National Center for Energy Sciences and Nuclear Technology, Morocco) operates a TRIGA Mark II reactor, which can reach a thermal maximum power at steady state of 2 MW. In reactors ...devoted to research and experiments, it is mandatory to characterize the neutron and photon fields in the irradiation positions. Together with a computational model of the core, it ensures the ability to reach the requested uncertainties when performing experiments, such as detectors testing, irradiation for hardening or nuclear data measurements.
The neutron field of different irradiation positions has been characterized by dosimetry techniques and compared to the MCNP full model of the reactor. Preliminary photon propagation calculations are also performed with this model, but up to now, no experimental validation of the results exists. The aim of the newly set collaboration between CEA and CNESTEN is to characterize the gamma field of these positions. The first position investigated is the part of the NB1 tangential channel closest to the core.
Among gamma measurements techniques, and according to the constraints arising from using this channel, it was chosen to use thermos- and optically stimulated luminescent detectors.
This paper presents the experiments carried out in September 2018 as well as their results. Three detectors types were used: TLD400 (CaF
2
:Mn), TLD700 (
7
LiF:Mg,Ti) and OSLD (Al
2
O
3
:C). Measurements were performed in several steps: background measurements, transient measurements (divergence phase + SCRAM), and irradiation at steady state. In the end, these measurements will provide a dose as well as a gamma flux value for this position.
•PARET model for the Moroccan TRIGA MARK II reactor has been developed.•Transient behavior under reactivity insertion has been studied based on PARET code.•Power factors required by PARET code have ...been calculated by using MCNP5 code.•The dependence on time of the main thermal-hydraulic parameters was calculated.•Results are largely far to compromise the thermal design limits.
A three dimensional model for the Moroccan 2MW TRIGA MARK II reactor has been developed for thermal-hydraulic and safety analysis by using the PARET/ANL and MCNP5 codes. This reactor is located at the nuclear studies center of Mâamora (CENM), Morocco. The model has been validated through temperature measurements inside two instrumented fuel elements located near the center of the core, at various power levels, and also through the power and fuel temperature evolution after the reactor shutdown (SCRAM). The axial distributions of power factors required by the PARET code have been calculated in each fuel element rod by using MCNP5 code. Based on this thermal-hydraulic model, a safety analysis under the reactivity insertion phenomenon has been carried out and the dependence on time of the main thermal-hydraulic parameters was calculated. Results were compared to the thermal design limits imposed to maintain the integrity of the clad.
Computation time constitutes an important and a problematic parameter in Monte Carlo simulations, which is inversely proportional to the statistical errors so there comes the idea to use the variance ...reduction techniques. These techniques play an important role in reducing uncertainties and improving the statistical results. Several variance reduction techniques have been developed. The most known are Transport cutoffs, Interaction forcing, Bremsstrahlung splitting and Russian roulette. Also, the use of a phase space seems to be appropriate to reduce enormously the computing time. In this work, we applied these techniques on a linear accelerator (LINAC) using the MCNPX computer Monte Carlo code. This code gives a rich palette of variance reduction techniques. In this study we investigated various cards related to the variance reduction techniques provided by MCNPX. The parameters found in this study are warranted to be used efficiently in MCNPX code. Final calculations are performed in two steps that are related by a phase space. Results show that, comparatively to direct simulations (without neither variance-reduction nor phase space), the adopted method allows an improvement in the simulation efficiency by a factor greater than 700.
► We have developed a computation scheme dedicated to LINAC simulation based on MC method. ► Several VRT have been used in order to reduce uncertainties. ► The VRT improve the statistical results in short computation time. ► The MCNPX code offers several variance reduction techniques.
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