The PGET device is currently being developed for partial-defect verication purposes on nuclear fuel assemblies. It Comprises CdTe detector elements in a heavy tungsten-alloy collimator, for which collimator slit openings define the field-of-view. This study aims at calculating the spatial response function of this device for further deployment in tomographic reconstruction algorithms. In this work, the detector response for 2 dierent sources (662 keV from Cesium-137 and 1274 keV from Europium-154) was simulated using the MCNPX software package. In the simulations, energy windows used in measurements with the PGET device were deployed. The results show the expected characteristics with strong response for a source position directly in front of the collimator slit opening and decreasing response as the source is moved into the penumbra and umbra region. The uncertainty of the simulated response function was less than 3.5 % for both sources. Separate simulations were made to quantify contributions from septal penetration and scattering from the collimator material into the detector for the energy windows covering the full -energy peak. These contributions were found to be around3% for the source of Cesium-137 and 6% for the source of Europium-154.

Magister Scientiae - MSc / Gamma ray spectroscopy has provided enormous amounts of information on the behaviour and structure of atomic nuclei [SHA88, BEA92, EBE08]. Most of the major discoveries in experimental nuclear physics over the last five decades are strongly associated with improvements in detector technologies. Inorganic scintilators led to the discovery in 1963 of the first excited states of a rotational band based on the ground state of 162Dy. Improvements in peak-to-background ratios and detector resolutions obtained with germanium led to the first evidence of backbending which is associated with a two quasi-particle excitation in 162Dy [SHA88]. More recently the development of composite and highly-segmented Ge detectors has significantly increased the performance and power of detection systems. The Clover detector is such a detector system and is in use at iThemba LABS. This study concerns the evaluation of the particle transport code MCNPX 2.5.0 as a tool to model complex composite detectors such as the Clover. Lanthanum silicate (LSO) and Lead tungstate (PbWO) are also evaluated as possible suppressor shield materials. It is shown that reasonable agreement between experiment and simulation is found when the experiment is accurately reproduced. However, when complex detection modes are implemented in the detector based on the number of elements that fire, MCNPX cannot be used to model the detector performance exactly. Differences between simulated and experimental results are found in suppressed add-back mode. It is proposed that the discrepancies are due to limitations in implementation of the pulse-height and special anti-coincidence tally in MCNPX. LSO and PbWO are compared to BGO as suppressor shield materials. It is found that LSO is not an ideal material for a suppression shield. PbWO is shown to give performance values similar to that of BGO. The back-plug is shown to have no effect on the Peak-to-Total ratio but is effective at reducing the background at lower energies. / South Africa

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. / The manganese bath technique is the reference method for neutron source emission rates calibration. It is used to calibrate neutron sources using radionuclides (AmBe, PuBe, 252Cf,…) in terms of neutron emission rate under 4π sr. As a complement to this technique, the anisotropy of the source is measured using a rotating source holder and a neutron long counter. The neutron source to be measured is immersed in a manganese sulphate solution whereby the emitted neutrons are captured within the bath contents. In a typical configuration (a 1m diameter sphere and a concentrated solution), approximately half of the neutrons lead to the creation of 56Mn via the 55Mn(n, γ) capture reaction. The 56Mn radionuclide has a half-life of approximately 2.6 hours and the bath reaches saturation when the number of nuclei decaying is equal to the number of nuclei created per unit time. The neutron emission rate from the source can then be deduced from the 56Mn activity at saturation, assuming proper modelling of the nuclear reactions occuring in the bath. The manganese bath facility at LNE-LNHB has been recently refurbished in order to comply with appropriate safety and radioprotection regulations. This has lead to the upgrading of both the measurement methodology and the modelling of the bath, and a study on the development of a new detector for the on-line measurement of the manganese activity was started. This new detector uses the β-γ coincidence measurement method. The bêta channel consists of two photomultipliers tubes which allow the detection of Cerenkov light, and the gamma channel uses a solid scintillation detector. The advantage of this measurement method is that it allows the determination of the bath activity without any prior calibration, unlike the former method which uses a gamma-ray detector calibrated using a high activity manganese source. The principle of the Cerenkov-gamma coincidence measurement has been validated by a prototype of the detector and via modelling of the system using the stochastic transport code GEANT4. The final detector has also been made and the results obtained have been compared to those from a primary measurement method already in use at LNE-LNHB. Furthermore, a comparison of the results from modelling the manganese bath with GEANT4, MCNPX and FLUKA have been undertaken to find the most reliable code. This comparison lead to the identification of various weaknesses, particularly in GEANT4, and several uncertainty factors, such as the modeling of the neutron emission and the choice of the cross-section library. Finally, neutron source calibration has been carried out with the Cerenkov-gamma method and the correction factors given by the new modeling of the bath using MCNPX. These results have been complemented with a comparison with the former method simultaneously undertaken, after calibration of the detector in the bath using a 56Mn source irradiated in a nuclear reactor. At the end of this study, several improvements have been proposed, from which a number are currently under development at LNE-LNHB.

Submitted by Irene Nascimento (irene.kessia@ufpe.br) on 2017-06-09T18:11:57Z No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) Dissertação - final.pdf: 1733080 bytes, checksum: f7cb6d19b01ccf4cefdfe6261cb2a5eb (MD5) / Made available in DSpace on 2017-06-09T18:11:57Z (GMT). No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) Dissertação - final.pdf: 1733080 bytes, checksum: f7cb6d19b01ccf4cefdfe6261cb2a5eb (MD5) Previous issue date: 2016-01-14 / Capes / Este trabalho apresenta simulações de Monte Carlo com objeto com forma de meia-lua, densidade e material conhecidos. Foi colocado dentro de um tubo de aço como arranjo de teste para simulação pelo código MCNPX, a fim de ser validada pelas medidas em experimentos estáticos de transmissão de raios gama e com comparação com dados teóricos. O MCNPX requer uma geometria definida para cada trajetória de fótons. Isto oferece empecilhos para o uso do código em simulação de tomografia. A solução desse problema foi escrever um programa para criar os arquivos de entrada para cada trajetória. Trabalhando nesta sequência, foi criado um banco de dados para organização da grande quantidade de dados gerados. A instalação experimental utiliza um único par fonte-detector, com fonte radioativa isotópica de Césio-137 e o detector de Nal (Tl) de cintilação acoplado a um analisador de multicanal. A análise de erros foi feita com a métrica do RMSE para avaliação da qualidade das simulações. As simulações foram validadas pelos dados experimentais e também pelos dados analíticos com erros satisfatórios segundo a métrica utilizada. / This work presents Monte Carlo simulations with object shaped like a half moon with density and material knowns. It was placed inside a steel tube as a test arrangement for simulating the MCNPX code, for validation by static experiments measurement of gamma ray transmission and compared with theoretical data. The MCNPX requires a geometry defined for each photon trajectory. This offers trammels to use the code for tomography simulation. The solution of this problem was developing a program to create input files for each trajectory. Working in this sequence, a database was created to organize large amount of data has been generated. The experimental setup uses a single source-detector pair, with a radioactive isotope Cesium source 137 and the detector NaI (Tl) scintillation coupled to a multichannel analyzer. The error analysis was made with the RMSE metric for evaluating the quality of the simulations. The simulations were validated by experimental data and also by comparison with theoretical data with satisfactory errors according to the metric used.

Due to the eminent shortage of 3He, there exists a significant need to develop a new (or optimize an existing) neutron detection system which would reduce the dependency on the current 3He-based detectors for Domestic Nuclear Detection Office (DNDO) applications. The purpose of this research is to develop a novel methodology for optimizing candidate neutron detector designs using multivariate statistical analysis of Monte Carlo radiation transport code (MCNPX) models. The developed methodology allows the simultaneous optimization of multiple detector parameters with respect to multiple response parameters which measure the overall performance of a candidate neutron detector. This is achieved by applying three statistical strategies in a sequential manner (namely factorial design experiments, response surface methodology, and constrained multivariate optimization) to results generated from MCNPX calculations. Additionally, for organic scintillators, a methodology incorporating the light yield non-proportionality is developed for inclusion into the simulated pulse height spectra (PHS). A Matlab® program was developed to post-process the MCNPX standard and PTRAC output files to automate the process of generating the PHS thus allowing the inclusion of nonlinear light yield equations (Birks equations) into the simulation of the PHS for organic scintillators. The functionality of the developed methodology is demonstrated on the successful multivariate optimization of three neutron detection systems which utilize varied approaches to satisfying the DNDO criteria for an acceptable alternative neutron detector. The first neutron detection system optimized is a 3He-based radiation portal monitor (RPM) based on a generalized version of a currently deployed system. The second system optimized is a 6Li-loaded polymer composite scintillator in the form of a thin film. The final system optimized is a 10B-based plastic scintillator sandwiched between two standard plastic scintillators. Results from the multivariate optimization analysis include not only the identification of which factors significantly affect detector performance, but also the determination of optimum levels for those factors with simultaneous consideration of multiple detector performance responses. Based on the demonstrated functionality of the developed multivariate optimization methodology, application of the methodology in the development process of new candidate neutron detector designs is warranted.

Tissue equivalent proportional counters are frequently used to measure dose and dose equivalent in mixed radiation fields that include neutrons; however, detectors simulating sites 1?m in diameter underestimate the quality factor, Q, for low energy neutrons because the recoil protons do not cross the detectors. Proportional counters simulating different site-sizes can be used to get a better neutron dose equivalent measurement since the range and stopping power of protons generated by neutrons in the tissue-equivalent walls depend on the energy of the primary neutrons. The differences in the spectra measured by different size detectors will provide additional information on the incident neutron energy. Monte Carlo N-particle extended (MCNPX) code was used to simulate neutron transportation in proportional counters of different simulated tissue diameter. These Monte Carlo results were tested using two solid walled tissue equivalent proportional counters, 2mm and 10mm in diameter, simulating tissue volumes 0.1?m and 0.5?m in diameter, housed in a single vacuum chamber. Both detectors are built with 3mm thick tissue equivalent plastic (A-150) walls and propane gas inside for dose measurement. Using these two detectors, the spectra were compared to determine the underestimation of y for large detector, and thereby obtain more information of the incident neutron particles. Based on the MCNPX simulation and experimental results, we can see that the smaller detector produces a larger average lineal energy than the larger detector, which means the larger detector (0.5?m diameter tissue equivalent size) underestimates the Q value for the low energy neutron, therefore underestimates the effective dose. These results confirm the results of the typical analysis of lineal energy as a function of site size.

The ability of inspection agencies and facility operators to measure powders containing several actinides is increasingly necessary as new reprocessing techniques and fuel forms are being developed. These powders are difficult to measure with nondestructive assay (NDA) techniques because neutrons emitted from induced and spontaneous fission of different nuclides are very similar. A neutron multiplicity technique based on first principle methods was developed to measure these powders by exploiting isotope-specific nuclear properties, such as the energy-dependent fission cross sections and the neutron induced fission neutron multiplicity. This technique was tested through extensive simulations using the Monte Carlo N-Particle eXtended (MCNPX) code and by one measurement campaign using the Active Well Coincidence Counter (AWCC) and two measurement campaigns using the Epithermal Neutron Multiplicity Counter (ENMC) with various (α,n) sources and actinide materials. Four potential applications of this first principle technique have been identified: (1) quantitative measurement of uranium, neptunium, plutonium, and americium materials; (2) quantitative measurement of mixed oxide (MOX) materials; (3) quantitative measurement of uranium materials; and (4) weapons verification in arms control agreements. This technique still has several challenges which need to be overcome, the largest of these being the challenge of having high-precision active and passive measurements to produce results with acceptably small uncertainties.

Submitted by Alice Araujo (alice.caraujo@ufpe.br) on 2018-06-20T21:54:59Z No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) DISSERTAÇÃO Klebson Marques de Oliveira.pdf: 10087402 bytes, checksum: 7e0f1e88e66ab3923f0b43cc701c9f73 (MD5) / Made available in DSpace on 2018-06-20T21:54:59Z (GMT). No. of bitstreams: 2 license_rdf: 811 bytes, checksum: e39d27027a6cc9cb039ad269a5db8e34 (MD5) DISSERTAÇÃO Klebson Marques de Oliveira.pdf: 10087402 bytes, checksum: 7e0f1e88e66ab3923f0b43cc701c9f73 (MD5) Previous issue date: 2017-02-17 / Os tubos metálicos transportam petróleo e gás na superfície, bem como através de poços submarinos, levando matéria-prima para a indústria e transportam o produto final para o mercado. A integridade desses tubos é problema crítico para a operação da indústria desde a fabricação com os processos de soldagem até a manutenção com problemas de corrosão. Em alguns casos, o processo de soldagem não garante a estrutura esperada para determinado uso da junta, surgindo com isso descontinuidades, que podem ser estruturais, dimensionais ou propriedades inadequadas. Os tubos considerados aqui são hipoeutetóides, isto é, tubos de aço com menos de 0,8% de carbono. Com as medidas da transmissão gama foram determinados o coeficiente de atenuação e a densidade radiométrica do aço base e da solda. Para comparar os resultados, a tomografia gama foi simulada com o código MCNPX para transporte de fótons. Importando no MCNPX a geometria definida foi otimizado o tempo de computação para simular a tomografia gama. A imagem tomográfica do tubo de aço foi reconstruída com algoritmo MART(Técnica de Reconstrução Algébrica com Correção Multiplicativa) para uma melhor visualização dos resultados. Os dados coletados com a varredura gama foram analisados e a matriz de correlação quantifica o grau de homogeneidade decrescente da estrutura do aço base, da solda de referencia e do tubo teste, nessa ordem. Aplicando-se uma métrica à distância de Chebyshev para os dados da transmissão gama pode-se observar resultados consistentes com os apresentados pela matriz de correlação. Na sequência o modelo estatístico multivariado o cluster hierárquico foi utilizado para processar os resultados gerados com a distância Chebyshev, e o dendrograma, uma função do Matlab, mostra distância contra número de clusters. Na geometria do MCNPX, defeitos foram introduzidos com o objetivo de detectar um defeito de 1µm na tomografia simulada. Uma curva de calibração foi obtida definindo o limite de resolução no intervalo de 1µm à 50µm.Testes com algoritmo de reconstrução foram realizados para visualizar neste intervalo o defeito simulado. Os resultados do método não destrutivo são comparados com os testes metalograficos tomados como referência. Os dados obtidos nesses procedimentos foram armazenados no banco de dados TOMC. / The pipes carry oil and gas on the surface, as well as through subsea wells, bringing raw material to the industry and transporting the final product to the market. The integrity of these pipes is a critical problem for the operation of the industry from manufacturing with the welding processes to maintenance with corrosion problems. In some cases, the welding process does not guarantee the structure expected for a particular use of the joint, resulting in discontinuities, which may be structural, dimensional or inappropriate properties. The tubes considered here are hypoeutectoids, this is, steel tubes having less than 0,8% carbon. The attenuation coefficient and the radiometric density of the base steel and the weld were determined using the gamma transmission measurements. To compare the results, gamma tomography was simulated with the MCNPX code for photon transport. Importing into MCNPX the defined geometry was optimized computing time to simulate gamma tomography. Importing into MCNPX the defined geometry was optimized computing time to simulate gamma tomography. The tomographic image of the steel tube was reconstructed with MART(Algebraic Reconstruction Technique with Multiplicative Correction) algorithm for a better visualization of the results. The data collected with the gamma sweep were analyzed and the correlation matrix quantifies the degree of decreasing homogeneity of the base steel structure, the reference weld and the test tube, in that order. By applying a metric to the Chebyshev distance for the data of the gamma transmission one can observe results consistent with those presented by the correlation matrix. Following the multivariate statistical model, the hierarchical cluster was used to process the results generated with the distance Chebyshev. The dendrogram, a function of Matlab, shows distance against number of clusters. In the geometry of the MCNPX, defects were introduced with the objective of detecting 1µm defect in the simulated tomography. A calibration curve was obtained by setting the resolution limit in the range of 1 µm to 50µm. Tests with a reconstruction algorithm were performed to visualize the simulated defect in this interval. The results of the non-destructive method are compared with the metallographic tests taken as reference. The data obtained in these procedures were stored in the TOMC database.

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