La méthode de mesure de référence du débit d'émission de sources neutroniques se fonde sur la technique du bain de manganèse. Elle est destinée à étalonner des sources de neutrons utilisant des radionucléides (241AmBe, 239PuBe, 252Cf,...) en termes de débit d'émission neutronique sous 4π sr. Ce dispositif est complété par un banc de mesure de l'anisotropie d'émission utilisant un support rotatif et un compteur long de type BF3. La source à mesurer est immergée dans une solution de sulfate de manganèse et les neutrons émis sont capturés par les constituants du bain. Dans une configuration classique (sphère de bain de manganèse de 1 m de diamètre et solution concentrée), environ la moitié de ces neutrons conduisent à la création de 56Mn par réaction (n, γ) sur 55Mn. Le radionucléide 56Mn a une période radioactive d'environ 2,6 heures et le bain de manganèse atteint son activité de saturation en 56Mn quand le nombre d'atomes radioactifs créés par unité de temps devient égal au nombre d'atomes se désintégrant pendant ce même temps. Le débit d'émission de la source peut alors être déduit de l'activité en 56Mn de la solution à saturation, via une modélisation ad hoc des réactions nucléaires se produisant dans le bain. Cette installation a été récemment rénovée au LNE-LNHB afin de respecter les règles de sécurité et de radioprotection en vigueur. Cette rénovation a été l'occasion de moderniser et de remettre à niveau les méthodes de mesure et de modélisation du bain et d'entreprendre une étude sur le développement d'un détecteur original pour la mesure directe en ligne de l'activité du manganèse. Ce détecteur est fondé sur la méthode de mesure par coïncidences β-γ. La voie bêta est constituée de deux photomultiplicateurs permettant de détecter l'émission de lumière due à l'effet Cerenkov et la voie gamma utilise un détecteur à scintillateur solide. L'intérêt de cette méthode de mesure est qu'elle permet d'avoir accès à l'activité du bain sans nécessiter d'étalonnage préalable, contrairement à la méthode classique qui utilise un compteur gamma et nécessite la fabrication d'une source de haute activité. Le principe de mesure a été validé à l'aide d'un prototype de détecteur et d'une modélisation effectuée à l'aide du code de calcul stochastique GEANT4. Le détecteur définitif a été réalisé et les mesures obtenues ont été comparées à celles données par une méthode primaire présente au laboratoire. Par ailleurs, des modélisations du bain de manganèse effectuées sous GEANT4, MCNPX et FLUKA, ont été comparées afin de choisir le code le plus fiable. Cette comparaison a permis d'identifier des lacunes notamment dans le code GEANT4 ainsi que des facteurs d'incertitude nécessitant une attention particulière, tels que la modélisation de l'émission neutronique et le choix des sections efficaces. Enfin, un étalonnage de source neutronique a été réalisé grâce à la méthode Cerenkov-gamma et aux facteurs correctifs donnés par la nouvelle modélisation du bain sous MCNPX. Ces mesures ont été complétées dans le cadre d'une comparaison comprenant également des mesures par l'ancienne méthode après étalonnage du couple bain/détecteur par irradiation d'une cible de manganèse en réacteur. Au terme de cette étude, plusieurs voies d'améliorations ont été proposées, dont certaines font déjà l'objet de travaux au LNHB.

Ce travail vise à démontrer la faisabilité d'un dosimètre opérationnel neutrons basé sur la technologie CMOS. Le capteur utilisé (MIMOSA-5) doit pour cela être transparent aux γ et pouvoir détecter les neutrons sur une large gamme d'énergie en gardant évidemment une bonne efficacité de détection. La réponse du système de détection, constitué du capteur CMOS adjoint d'un matériau convertisseur (polyéthylène pour les neutrons rapides, 10B pour les neutrons thermiques), a été confrontée à des si-mulations Monte Carlo effectuées avec MCNPX et GEANT4. Un travail de validation de ces codes a préalablement été effectué pour justifier leur utilisation dans le cadre de notre application. Les expériences nous permettant de caractériser le capteur ont été menées au sein de l'IPHC ainsi qu'à l'IRSN/LMDN (Cadarache). Les résultats de l'exposition du capteur à des sources de photon pures et à un champ mixte n/γ (source 241AmBe) montrent la possibilité d'obtenir un système transparent aux γ par application d'une coupure appropriée sur le dépôt d'énergie (aux alentours de 100 keV). L'efficacité de détection associée est très satisfaisante avec une valeur de 10-3, en très bon accord avec MCNPX et GEANT4. La réponse angulaire du capteur a été étudiée par la suite à l'aide de la même source. La dernière partie de cette étude traite de la détection des neutrons thermiques (de l'ordre de l'eV). Les expériences ont été menées à l'IRSN sur une source de 252Cf modérée à l'eau lourde. Les résultats ob-tenus ont montré une très bonne efficacité de détection (allant jusqu'à 6×10-3 pour un convertisseur do-pé au 10B) en très bon accord avec GEANT4.

For efficient beta detection in a mixed beta gamma field, Monte Carlo simulation models have been built to optimize the thickness of a plastic scintillator, used in whole body monitor. The simulation has been performed using MCNP/X code and different thicknesses of plastic scintillators ranging from 150 to 600 um have been used. The relationship between the thickness of the scintillator and the efficiency of the detector has been analyzed. For 150 m thickness, an experimental investigation has been conducted with different beta sources at different positions on the scintillator and the counting efficiency of the unit has been measured. Evaluated data along with experimental ones have been discussed. A thickness of 300 um to 500 um has been found to be an optimum thickness for better beta detection efficiency in the presence of low energy gamma ray. / UOIT

The Oak Ridge Isotope Generation and Depletion – Automatic Rapid Proccessing (ORIGEN-ARP) deterministic code has been extensively utilized for determining nuclide concentrations at various specific burnup values for a variety of nuclear reactor designs. Given nuclide concentrations or ratios, such calculations can be used in nuclear forensics and nuclear non-proliferation applications to reverse-calculate the type of reactor and specific burnup of the fuel from which the nuclides originated. Recently, Los Alamos National Laboratory has released a version of its probabilistic radiation transport code, MCNPX 2.6.0, which incorporates a fuel burnup feature which can also determine, via the probabilistic Monte Carlo method, nuclide concentrations as a function of fuel burnup. This dissertation compares the concentrations of 46 nuclides significant to nuclear forensics analyses for different reactor types using results from the ORIGEN-ARP and the MCNPX 2.6.0 codes. Three reactor types were chosen: the Westinghouse 17x17 Pressurized Water Reactor (PWR), the GE 8x8-4 Boiling Water Reactor (BWR), and the Canadian Deuterium Uranium, CANDU-37, reactor. Additionally, a sensitivity study of the different reactor parameters within the MCNPX Westinghouse 17x17 PWR model was performed. This study analyzed the different nuclide concentrations resulting from minor perturbations of the following parameters: assembly rod pitch, initial moderator boron concentration, fuel pin cladding thickness, moderator density, and fuel temperature. / text

Submitted by Almir Azevedo (barbio1313@gmail.com) on 2018-05-09T13:11:49Z No. of bitstreams: 1 dissertação mestrado ien 2018 Tamara Porfiro Teixeira.pdf: 807837 bytes, checksum: b01215afaee9c4a8b22c3267552cfcbc (MD5) / Made available in DSpace on 2018-05-09T13:11:49Z (GMT). No. of bitstreams: 1 dissertação mestrado ien 2018 Tamara Porfiro Teixeira.pdf: 807837 bytes, checksum: b01215afaee9c4a8b22c3267552cfcbc (MD5) Previous issue date: 2018-02 / Este trabalho apresenta uma metodologia para predição de incrustação concêntricas e excêntricas em tubulações utilizadas na indústria de petróleo off-shore. A aproximação é baseada nos princípios de densitometria gama e redes neurais artificiais. Foi desenvolvido um modelo de estudo preliminar visando definir as composições do duto e incrustações. Para isso, foi avaliada a influência na transmissão gama de dutos com quatro tipos diferentes de aços utilizados em plataformas de petróleo, bem como a influência das principais formações inorgânicas de depósitos. A divergência da fonte radioativa também foi considerada nessa avaliação, com aberturas de colimação de 2 mm a 7 mm, com passos de 2,5 mm. Após a definição da composição do duto e incrustação, foi definida uma geometria de medição por meio do código MCNP-X para calcular a espessura da incrustação por meio de equações analíticas independentemente dos fluidos presentes no duto (água salgada, gás e óleo). A geometria representativa utiliza um duto composto por Ferro, com incrustação inorgânica formada por sulfato de bário (BaSO4). Modelos de incrustações concêntricas foram simulados e os dados obtidos foram utilizados para treinamento e validação de uma rede neural artificial, bem como modelos de incrustações excêntricas. O sistema de detecção simulado consistiu em uma geometria com feixe estreito com 2 mm de abertura de colimação, compreendendo uma fonte de raios gama (137Cs) e detectores NaI(Tl) 2x2” posicionados adequadamente ao redor do sistema duto-incrustação-fluido para o cálculo da espessura de incrustação considerando o feixe transmitido e o espalhado. O espalhamento Compton foi considerado nos casos de incrustações com formação excêntrica para auxílio na determinação e localização das espessuras máximas de incrustação. Os modelos teóricos foram desenvolvidos usando o código matemático MCNPX e utilizados para o treinamento, teste e validação das redes neurais artificiais. A metodologia proposta foi capaz de predizer as espessuras de incrustações concêntricas e excêntricas com resultados satisfatórios para esses dois tipos de formações inorgânicas. / This work presents a methodology for predicting concentric and eccentric scales in pipelines used in the offshore oil industry. The approximation is based on the principles of gamma densitometry and artificial neural networks. A preliminary study model was developed to define the compositions of the duct and scale. In order to do so, the influence of pipeline transmission with four different types of steel used in oil platforms was evaluated, as well as the influence of the main inorganic deposit formations. The divergence of the radioactive source was also considered in this evaluation, with collimation openings of 2 mm to 7 mm, with steps of 2.5 mm. After defining the composition of the duct and scale, a measurement geometry was defined by means of the MCNP-X code to calculate the scale thickness by means of analytical equations, independent of the fluids present in the duct (salt water, gas and oil). The representative geometry uses a duct composed of iron, with inorganic scale formed by barium sulfate (BaSO4). Concentric scale models were simulated and the data obtained were used for training and validation of an artificial neural network, as well as eccentric scale models. The simulated detection system consisted of a narrow-beam geometry with a 2 mm collimation aperture, comprising a gamma ray source (137Cs) and 2x2 "NaI (Tl) sensors suitably positioned around the duct-scale-fluid system for calculation of the scale thickness considering the transmitted beam and the scattered beam. Compton scattering was considered in cases of eccentric scale to aid in the determination and location of maximum scale thicknesses. The theoretical models were developed using the mathematical code MCNP-X and used for training, testing and validation of artificial neural networks. The proposed methodology was able to predict the concentric and eccentric scale thicknesses with satisfactory results for these two types of inorganic formations.

Proton therapy uses proton beams with energies of 70 – 230 MeV to treat cancerous tumours very effectively, while preserving surrounding healthy tissues as much as possible. During nuclear interactions of these protons with matter, secondary neutrons can be produced. These neutrons can have energies ranging up to the maximum energy of the protons and can thus be particularly difficult to attenuate. In fact, the rooms of a proton therapy facility are generally surrounded by concrete walls of at least ~2 m in thickness, in order to protect the members of the staff and the public from the stray radiation. Today, the design of the shielding walls is generally based on Monte Carlo simulations. Amongst the numerous parameters on which these simulations depend, some are difficult to control and are therefore selected in a conservative manner. Despite these conservative choices, it remains important to carry out accurate neutron dose measurements inside proton therapy facilities, in order to assess the effectiveness of the shielding and the conservativeness of the simulations. There are, however, very few studies in literature which focus on the comparison of such simulations with neutron measurements performed outside the shielding in proton therapy facilities. Moreover, the published measurements were not necessarily acquired with detectors that possess a good sensitivity to neutrons with energies above 20 MeV, while these neutrons actually give an important contribution to the total dose outside the shielding. A first part of this work was dedicated to the study of the energy response function of the WENDI-2, a rem meter that possesses a good sensitivity to neutrons of more than 20 MeV. The WENDI-2 response function was simulated using the Monte Carlo code MCNPX and validation measurements were carried out with 252Cf and AmBe sources as well as high-energy quasi-monoenergetic neutron beams. Then, WENDI-2 measurements were acquired inside and outside four rooms of the proton therapy facility of Essen (Germany). MCNPX simulations, based on the same conservative choices as the original shielding design simulations, were carried out to calculate the neutron spectra and WENDI-2 responses in the measurement positions. A relatively good agreement between the simulations and the measurements was obtained in front of the shielding, whereas overestimates by at least a factor of 2 were obtained for the simulated responses outside the shielding. This confirmed the conservativeness of the simulations with respect to the neutron fluxes transmitted through the walls. Two studies were then carried out to assess the sensitivity of the MCNPX simulations to the defined concrete composition and the selected physics models for proton and neutron interactions above 150 MeV. Both aspects were found to have a significant impact on the simulated neutron doses outside the shielding. Finally, the WENDI-2 responses measured outside the fixed-beam treatment room were also compared to measurements acquired with an extended-range Bonner Sphere Spectrometer and a tissue-equivalent proportional counter. A satisfactory agreement was obtained between the results of the three measurement techniques. / Doctorat en Sciences / info:eu-repo/semantics/nonPublished

No description available.

Independent yield distributions of high-energy neutron-induced fission are of importance to achieve a good understanding of fission. Even though the mass and charge yield distributions of thermal neutron-induced fission are well known, there are few experimental data for high-energy neutron-induced fission. In addition to basic research on the fission process, independent yield distributions of high-energy neutron-induced fission play a key role in the development of Generation IV fast nuclear reactors. To facilitate measurements of independent fission yields of high-energy neutron-induced fission, a dedicated ion guide and a proton-neutron converter were developed and put to use in experiments at the isotope separator facility IGISOL in Jyväskylä. In parallel, a simulation model of the system was developed in order to optimize the collecting efficiency of fission products in the ion guide. The model uses the Monte Carlo code MCNPX to simulate the neutron production, the fission model code GEF to simulate the fission process, and GEANT4 for ion transportation. In order to benchmark the simulation model, metal foils were inserted in the ion guide with the purpose of collecting fission products. At the same time, nickel, cobalt and indium foils were located between the pn-converter and the ion guide to record the neutron flux from the pn-converter. After the beam was turned off, and after several days of cooling, g-ray spectroscopy measurements of the foils were conducted using a well shielded HPGe detector. Based on the identified g-ray transitions in the spectroscopy data, the productions of corresponding fission products and neutron activation products were calculated, and then used to benchmark the transportation and collection of fission products, as well as neutron production, in the simulations. The conclusion from the benchmark is that the transportation of fission products in the helium gas, as simulated by GEANT4, agrees very well with the measurement, while the transportation of fission products in the uranium targets agrees with the measurement within 10%. The neutron flux at the high-energy part of the neutron spectrum is overestimated by about 40%.Thanks to the benchmark it has been shown that the predictive power of the model is satisfactory and sufficient for the purpose of modeling the ion guide. Furthermore, the parameters involved in the simulations, such as neutron production, distance between the neutron source and the ion guide, volume of the ion guide and so on, play an important role in the optimization of the setup. However, the lower than expected fission rate suggests that the optimization on these parameters may not be enough to achieve a sufficiently high intensity of fission products, especially for nuclei far from the stability line. To achieve a sufficiently high intensity, an electric field guidance, similar to the RF structure of the CARIBU gas catcher presented in G.Savard et al. Nucl. Inst. Meth. B, 376: 246, 2016, to collect fission products is considered.

La mesure de la radioactivité dans les filtres utilisés pour la surveillance de l’aérocontamination de l’air présente une difficulté métrologique majeure. En effet, l’absorption des rayonnements a dans le médium filtrant et la masse d’aérosols accumulés biaisent la réponse nucléaire. Ce travail de thèse porte sur la détermination de facteurs de correction de la dégradation de la radioactivité mesurée dans les filtres de surveillance. Dans un premier temps, des filtres radioactifs représentatifs des prélèvements atmosphériques ont été préparés à l’aide du banc d’essais nucléaire ICARE. L’étude expérimentale sur les filtres de référence a permis d’avoir une base de données pout la détermination des facteurs de correction dans diverses conditions de filtration. Dans un second temps, ce travail a conduit une nouvelle méthode numérique mise au point pour déterminer les facteurs de correction. Il s’agit de coupler des simulations de filtration des particules d’aérosol à l’aide de GeoDict, permettant de calculer des écoulements dans les milieux poreux et des simulations de parcours de particules a dans la matière à l’aide de MCNPX. Le bon accord obtenu, en comparant les réponses des spectres en énergie et des facteurs de correction numériques et expérimentaux, a permis de valider le modèle numérique / The measurement of radioactivity in the filters of airborne radioactive surveillance is a major difficulty metrology. Indeed, the absorption of a radiation in the filter media and the mass of aerosols accumulated distort the nuclear counters response. This thesis work focuses on the determination of correction factors for the radioactivity loss in the survey filters. In a first step, radioactive filters representing the atmospheric samples have been prepared using the nuclear test bench ICARE. The experimental study on reference filters provided a database to determine correction factors for various filtration conditions. The second step of the work proposed a new numerical method developed to determine the correction factors. It consists of coupling GeoDict for particles filtration simulations and MCNPX simulations for a transport in matter. The good agreement obtained by comparing the numerical and experimental correction factors has permitted to validate the numerical model

Numerous scenarios exist whereby radioactive particulates are transported between spatially separated points of interest. An example of this phenomenon is, in the aftermath of a Radiological Dispersal Device (RDD) detonation, the resuspension of radioactive particulates from the resultant fallout field. Quantifying the spatial distribution of radioactive particulates allow for the calculation of potential radiation doses that can be incurred from exposure to such particulates. Presently, there are no simulation techniques that link radioactive particulate transport with subsequent radiation field determination and so this thesis develops a coupled Computational Fluid Dynamics (CFD) and Monte Carlo (MC) Radiation Transport approach to this problem. Via particulate injections, the CFD simulation defines the spatial distribution of radioactive particulates and this distribution is then employed by the MC Radiation Transport simulation to characterize the resultant radiation field. GAMBIT/FLUENT are employed for the CFD simulations while MCNPX is used for the MC Radiation Transport simulations. / UOIT