Um método SN híbrido direto para cálculos de sistemas combustível-moderador em geometria unidimensional

Desenvolvemos nesta dissertação um método híbrido direto para o cálculo do fator de desvantagem e descrição da distribuição do fluxo de nêutrons em sistemas combustível-moderador. Na modelagem matemática, utilizamos a equação de transporte de Boltzmann independente do tempo, considerando espalhamento linearmente anisotrópico no modelo monoenergético e espalhamento isotrópico no modelo multigrupo, na formulação de ordenadas discretas (SN), em geometria unidimensional. Desenvolvemos nesta dissertação um método híbrido direto para o cálculo do fator de desvantagem e descrição da distribuição do fluxo de nêutrons em sistemas combustível-moderador. Na modelagem matemática, utilizamos a equação de transporte de Boltzmann independente do tempo, considerando espalhamento linearmente anisotrópico no modelo monoenergético e espalhamento isotrópico no modelo multigrupo, na formulação de ordenadas discretas (SN), em geometria unidimensional. Descrevemos uma análise espectral das equações de ordenadas discretas (SN)a um grupo e a dois grupos de energia, onde seguimos uma analogia com o método de Case. Utilizamos, neste método, quadraturas angulares diferentes no combustível (NC) e no moderador (NM), onde em geral assumimos que NC > NM . Condições de continuidade especiais que acoplam os fluxos angulares que emergem do combustível (moderador) e incidem no moderador (combustível), foram utilizadas com base na equivalência entre as equações SN e PN-1, o que caracteriza a propriedade híbrida do modelo proposto. Sendo um método híbrido direto, utilizamos as NC + NM equações lineares e algébricas constituídas pelas (NC + NM)/2 condições de contorno reflexivas e (NC + NM)/2 condições de continuidade para determinarmos as NC + NM constantes. Com essas constantes podemos calcular os valores dos fluxos angulares e dos fluxos escalares em qualquer ponto do domínio. Apresentamos resultados numéricos para ilustrar a eficiência e a precisão do método proposto.

Part I: Total yield measurements for the ^(21)Ne(α,n)^)(24)Mg reaction. Part II: A study of the ^(12)C(He,p)^(14)N reaction

PART I

The total cross-section for the reaction ²¹Ne(α, n)²⁴Mg has been measured in the energy range 1.49 Mev ≤ E_cm ≤ 2.6 Mev. The cross-section factor, S(O), for this reaction has been determined, by means of an optical model calculation, to be in the range 1.52 x 10¹² mb-Mev to 2.67 x 10¹² mb-Mev, for interaction radii in the range 5.0 fm to 6.6 fm. With S(O) ≈ 2 x 10¹² mb-Mev, the reaction ²¹Ne(α, n)²⁴Mg can produce a large enough neutron flux to be a significant astrophysical source of neutrons.

PART II

The reaction¹²C(³He, p)¹⁴N has been studied over the energy range 12 Mev ≤ E_lab ≤ 18 Mev. Angular distributions of the proton groups leading to the lowest seven levels in ¹⁴N were obtained.

Distorted wave calculations, based on two-nucleon transfer theory, were performed, and were found to be reliable for obtaining the value of the orbital angular momentum transferred. The present work shows that such calculations do not yield unambiguous values for the spectroscopic factors.

Caracterização de um sistema digital de aquisição de imagens radiográficas utilizando nêutrons térmicos e raios gama para a inspeção de componentes mecânicos.

Ensaio não destrutivo é uma ferramenta essencial quando um equipamento, dispositivo ou componente não pode ser submetido a procedimentos destrutivos ou invasivos devido a razões de segurança, alto custo ou outras restrições físicas ou logísticas. Dentro deste quadro radiografias por transmissão com raios gama e nêutrons térmicos são técnicas singulares para inspecionar um objeto e desvendar sua estrutura interna devido à capacidade de atravessar uma vasta gama de materiais utilizados na indústria. Grosso modo, raios gama são mais atenuados por materiais pesados enquanto nêutrons térmicos são mais atenuados por materiais mais leves, tornando-as ferramentas complementares. Este trabalho apresenta os resultados obtidos na inspeção de vários componentes mecânicos, através da radiografia por transmissão com nêutrons térmicos e raios gama. O fluxo de nêutrons térmicos de 4,46x105 n.cm-2.s-1 disponível no canal principal do reator de pesquisa Argonauta do Instituto de Engenharia Nuclear foi usado como fonte para as imagens radiográficas com nêutrons. Raios dekeV emitidos pelo 198Au, também produzido no reator, foram usados como fonte de radiação para radiografias . Imaging Plates, especificamente produzidos para operar com nêutrons térmicos ou com raios X, foram empregados como detectores e dispositivos de armazenamento e captação de imagens para cada uma dessas radiações. Esses dispositivos exibem varias vantagens quando comparados ao filme radiográfico convencional. Com efeito, além de maior sensibilidade e serem reutilizáveis não são necessários câmaras escuras e processamento químico para a revelação. Em vez disso, ele é lido por um feixe de laser que libera elétrons armadilhados na rede cristalina durante a exposição à radiação, fornecendo uma imagem final digital. O desempenho de ambos os sistemas de aquisição de imagens, assim constituído, foi avaliado com respeito à sensibilidade, resolução espacial, linearidade e range dinâmico, incluído uma comparação com sistemas radiográficos com nêutrons empregando filmes e folhas de gadolínio como conversor de nêutrons em partículas carregadas. Além desta caracterização, diversos equipamentos e componentes foram radiografados com ambos os sistemas visando-se avaliar suas capacidades de desvendar a estrutura interna desses objetos e detectar estruturas e estados anormais. Dentro desta abordagem, uma neutrongrafia detectou a presença de material cerâmico remanescente empregado como molde no processo de fabricação nos canais de refrigeração de uma aleta do estator de uma turbina tipo turbo-fan, que deveria estar livre desse material. O reostato danificado de um sensor de pressão automotivo, foi identificado por neutrongrafia, embora nesse caso a radiografia também conseguiu realizar essa tarefa com melhor resolução, corroborando assim as curvas de resolução espacial obtidas na caracterização dos dois sistemas. A homogeneidade da distribuição do material encapsulado em uma gaxeta explosiva de chumbo utilizada na indústria aeroespacial foi igualmente verificada por neutrongrafia porque esse metal é relativamente transparente para nêutrons, mas suficientemente opaco para o explosivo rico em hidrogênio. Diversos outros instrumentos e componentes tais como variômetro, altímetro, bússola aeronáutica, injetor automotivo de combustível, foto-camera, disco rígido de computador, motor de passo, conectores eletrônicos e projéteis foram radiografados com ambos os sistemas visando avaliar suas habilidades em desvendar diferentes peculiaridades em função do agente interrogador.

A preliminary study of target multiplicity for ADSRs

The Accelerator Driven Subcritical Reactor (ADSR) concept is based on the coupling of a particle accelerator to a subcritical reactor core by means of a neutron spallation target interface. This paper investigates the benefits of multiple spallation targets in ADSRs. The motivation behind this is, firstly, to improve the overall reliability of the accelerator-reactor system, and, secondly, to evaluate other potential advantages such as lower beam power requirements. The results show that a system containing two or three spallation targets, coupled to independent accelerators, offers better neutronic performance. This is demonstrated through the increased effective multiplication factor (keff) in the two- and three-target configurations and a more uniform neutron flux distribution. A multiple-target ADSR also proves effective in mitigating the impact of frequent beam interruptions, a pressing issue that needs to be addressed for the ADSR concept to advance. Assuming no simultaneous beam shutdowns, the two- and three-target configurations reduce the risk of fuel cladding failure due to thermal cyclic fatigue. © 2013 Elsevier B.V. All rights reserved.

Detailed design of242mAm breeding in pressurized water reactors

There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

Further investigating the determination of phosphorus in plants by INAA using bremsstrahlung measurement

Phosphorus is an essential element for plants and animals, playing a fundamental role in the production of biochemical energy. Despite its relevance, phosphorus is not commonly determined by instrumental neutron activation analysis (INAA), because (32)P does not emit gamma-rays in its decay. There are alternative methods for the determination of phosphorus by INAA, such as the use of beta counting or the measurement of bremsstrahlung originated from the high energy beta particle from (32)P. Here the determination of phosphorus in plant materials by measuring the bremsstrahlung production was further investigated, to optimize an analytical protocol for minimizing interferences and overcoming the poor specificity. Eight certified reference materials of plant matrices with phosphorus ranging between 171 and 5,180 mg kg(-1) were irradiated at a thermal neutron flux of 9.5 x 10(12) cm(-2) s(-1) and measured with a HPGe detector at decay times varying from 7 to 60 days. Phosphorus solutions added to a certified reference material at three levels were used for calibration. Counts accumulated in the baseline at four different regions of the gamma-ray spectra were tested for the determination of phosphorus, with better results for the 100 keV region. The Compton scattering contribution in the selected range was discounted using an experimental peak-to-Compton factor and the net areas of all peaks in the spectra with energies higher than 218 keV, i.e. Compton edge above 100 keV. Amongst the interferences investigated, the production of (32)P from sulfur, and the contribution of Compton scattering should be considered for producing good results.

A search for point sources of EeV neutrons

A thorough search of the sky exposed at the Pierre Auger Cosmic Ray Observatory reveals no statistically significant excess of events in any small solid angle that would be indicative of a flux of neutral particles from a discrete source. The search covers from -90 degrees to +15 degrees in declination using four different energy ranges above 1 EeV (10(18) eV). The method used in this search is more sensitive to neutrons than to photons. The upper limit on a neutron flux is derived for a dense grid of directions for each of the four energy ranges. These results constrain scenarios for the production of ultrahigh energy cosmic rays in the Galaxy.

K-41(n, gamma)K-42 thermal and resonance integral cross section measurements

We measured the K-41 thermal neutron absorption and resonance integral cross sections after the irradiation of KNO3 samples near the core of the IEA-R1 IPEN pool-type research reactor. Bare and cadmium-covered targets were irradiated in pairs with Au-Al alloy flux-monitors. The residual activities were measured by gamma-ray spectroscopy with a HPGe detector, with special care to avoid the K-42 decay beta(-) emission effects on the spectra. The gamma-ray self-absorption was corrected with the help of MCNP simulations. We applied the Westcott formalism in the average neutron flux determination and calculated the depression coefficients for thermal and epithermal neutrons due to the sample thickness with analytical approximations. We obtained 1.57(4) and 1.02(4) b, for thermal and resonance integral cross sections, respectively, with correlation coefficient equal to 0.39.

Diodenlaserbasierte Resonanzionisations-Massenspektrometrie zur Spektroskopie und Ultraspurenanalyse an Uranisotopen

In dieser Arbeit wird die Erweiterung und Optimierung eines Diodenlasersystems zur hochauflösenden Resonanzionisationsmassenspektrometrie beschrieben. Ein doppelinterferometrisches Frequenzkontrollsystem, welches Absolutstabilisierung auf ca. 1 MHz sowie sekundenschnelle Frequenzverstimmungen um mehrere GHz für bis zu drei Laser parallel ermöglicht, wurde optimiert. Dieses Lasersystem dient zwei wesentlichen Anwendungen. Ein Aspekt waren umfangreiche spektroskopische Untersuchungen an Uranisotopen mit dem Ziel der präzisen und eindeutigen Bestimmung von Energielagen, Gesamtdrehimpulsen, Hyperfeinkonstanten und Isotopieverschiebungen sowie die Entwicklung eines effizienten, mit kommerziellen Diodenlasern betreibbaren Anregungsschemas. Mit diesen Erkenntnissen wurde die Leistungsfähigkeit des Lasermassenspektrometers für die Ultraspurenanalyse des Isotops 236U, welches als Neutronendosimeter und Tracer für radioaktive anthropogene Kontaminationen in der Umwelt verwendet wird, optimiert und charakterisiert. Anhand von synthetischen Proben wurde eine Isotopenselektivität von 236U/238U=4,5(1,5)∙10-9 demonstriert.

Characterisation and optimisation of the new Prompt Gamma-ray Activation Analysis (PGAA) facility at FRM II

At the research reactor Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRM II) a new Prompt Gamma-ray Activation Analysis (PGAA) facility was installed. The instrument was originally built and operating at the spallation source at the Paul Scherrer Institute in Switzerland. After a careful re-design in 2004–2006, the new PGAA instrument was ready for operation at FRM II. In this paper the main characteristics and the current operation conditions of the facility are described. The neutron flux at the sample position can reach up 6.07×1010 [cm−2 s−1], thus the optimisation of some parameters, e.g. the beam background, was necessary in order to achieve a satisfactory analytical sensitivity for routine measurements. Once the optimal conditions were reached, detection limits and sensitivities for some elements, like for example H, B, C, Si, or Pb, were calculated and compared with other PGAA facilities. A standard reference material was also measured in order to show the reliability of the analysis under different conditions at this instrument.

Diffusion Issues of Heat and Light Species in Laser Fusion Devices

There are several heat and mass diffusion problems which affect to the IFC chamber design. New simulation models and experiments are needed to take into account the extreme conditions due to ignition pulses and neutron flux

A proposed parameterization of interface discontinuity factors depending on neighborhood for pin-by-pin diffusion computations for LWR

There exists an interest in performing full core pin-by-pin computations for present nuclear reactors. In such type of problems the use of a transport approximation like the diffusion equation requires the introduction of correction parameters. Interface discontinuity factors can improve the diffusion solution to nearly reproduce a transport solution. Nevertheless, calculating accurate pin-by-pin IDF requires the knowledge of the heterogeneous neutron flux distribution, which depends on the boundary conditions of the pin-cell as well as the local variables along the nuclear reactor operation. As a consequence, it is impractical to compute them for each possible configuration. An alternative to generate accurate pin-by-pin interface discontinuity factors is to calculate reference values using zero-net-current boundary conditions and to synthesize afterwards their dependencies on the main neighborhood variables. In such way the factors can be accurately computed during fine-mesh diffusion calculations by correcting the reference values as a function of the actual environment of the pin-cell in the core. In this paper we propose a parameterization of the pin-by-pin interface discontinuity factors allowing the implementation of a cross sections library able to treat the neighborhood effect. First results are presented for typical PWR configurations.