The main objective of this research is: (1) to develop a model and perform numerical simulations to evaluate the radiation field and the resulting dose to personnel and activation of materials and structures throughout the IRIS nuclear power plant, and (2) to confirm that the doses are below the regulatory limit, and assess the possibility to reduce the activation of the concrete walls around the reactor vessel to below the free release limit. IRIS is a new integral pressurized water reactor (PWR) developed by an international team led by Westinghouse with an electrical generation capacity of 335 MWe and passive safety systems. Its design differs from larger loop PWRs in that a single building houses the containment as well as all the associated equipment including the control room that must be staffed continuously. The resulting small footprint has positive safety and economic implications, and the integral layout provides additional shielding and thus the opportunity to significantly reduce the activation, but it also leads to significantly more challenging simulations. The difficulty in modeling the entire building is the fact that the source is attenuated over 10 orders of magnitude before ever reaching the accessible areas. For an analog Monte Carlo simulation with no acceleration (variance reduction), it would take many processor-years of computation to generate results that are statistically meaningful. Instead, to generate results for this thesis, the Standardized Computer Analyses for Licensing Evaluation (SCALE) with the package Monaco with Automated Variance Reduction using Importance Calculations (MAVRIC) will be used. This package is a hybrid methodology code where the forward and adjoint deterministic calculations provide variance reduction parameters for the Monte Carlo portion to significantly reduce the computational time. Thus, the first task will be to develop an efficient SCALE/MAVRIC model of the IRIS building. The second task will be to evaluate the dose rate and activation of materials, specifically focusing on activation of concrete walls around the reactor vessel. Finally, results and recommendations will be presented.

Measuring the farfield noise levels of full-scale rotor systems is not trivial and can be costly. Researchers prefer to perform small-scale experiments in the laboratory so that they can extrapolate the model scaled results to the larger scale. Typically Inverse Square Law (ISL) is used to extrapolate the sound pressure levels (SPL), obtained from model-scale experiments at relatively small distances to predict noise at much larger distances for larger scale systems. The assumption underlying this extrapolation is that the source itself can be treated as a point sound source. At what distance from a rotor system it can be treated as a point source has never been established. Likewise, many theoretical models of shielding by hard surfaces assume the source to be a point monopole source. If one is interested in shielding the noise of a rotor system by interposing a hard surface between the rotor and the observer, can the rotor system really be considered to be a monopole? If rotating noise sources are under consideration what is the effect of configuration and design parameters? Exploring the validity of point source assumption alluded to above for a rotor for farfield acoustic measurements with and without shielding form the backbone of the present work.

Heatshield design and analysis has traditionally been a decoupled process, the designer creates the geometry generally without knowledge about how the design variables affect the thermostructural response or how the system will perform under off nominal conditions. Heatshield thermal and structural response analyses are generally performed as separate tasks where the analysts size their respective components and feedback their results to the designer who is left to interpret them. The analysts are generally unable to provide guidance in terms of how the design variables can be modified to meet geometric constraints and not exceed the thermal or structural design specifications. In general, the thermal response analysis of ablative thermal protection systems has traditionally been performed using a one-dimensional finite difference calculation. The structural analyses are generally one, two, or three-dimensional finite element calculations. In this dissertation, the governing differential equations for ablative thermal response are solved in three-dimensions using the finite element method. Darcy' Law is used to model the flow of pyrolysis gas through the ablative material. The three-dimensional governing differential equations for Darcy flow are solved using the finite element method as well. Additionally, the equations for linear elasticity are solved by the finite element method for the thermal stress using temperatures directly from the thermal response calculations. This dissertation also links the analysis of thermal protection systems to their design. The link to design comes from understanding the variation in the thermostructural response over the range of the design variables. Material property sensitivities are performed and an optimum design is determined based on a deterministic analysis minimizing the design specification of bondline temperature subject to appropriate constraints. A Monte Carlo simulation is performed on the optimum design to determine the probability of exceeding the design specifications. The design methodology is demonstrated on the Orion Crew Exploration Vehicle's compression pad design.

There is an intense interest in the availability of radioactive isotopes that could be developed into nuclear batteries. Promethium-147 is one of the isotopes of interest for use in nuclear batteries as well as in other compact low power applications. Pm-147 is a pure beta (β-) emitter with a half-life of 2.62 years. For this research, Pm-147 was produced from enriched Nd-146 via the neutron capture method in the Hydraulic Tube facility of HFIR at the Oak Ridge National Laboratory. Radioisotopic impurities produced via the neutron capture method have significant effects on its potential final use for nuclear battery applications. This research provides information on the co-production levels of the radioisotopic impurities in the samples containing Pm-147 and their effects on the required shielding. Gamma spectroscopy analysis served as the primary method in the evaluation of the impurities. Previous research had identified the presence of these impurities but it had not studied them in detail.

The collision probability approach to neutron transport can be used to obtain the energy-dependent neutron spectrum in nuclear reactor systems as well as other quantities of interest. This method makes the approximation that the neutron distribution is constant within homogeneous regions, or cells, in the system. This assumption restricts geometries that can be modeled by the collision probability approach. The geometry modeled is typically an infinite lattice of two homogeneous cells: a fuel pin cylinder and the coolant that surrounds it. The transport of neutrons between the homogeneous cells is done using probabilities describing the chance that a neutron having a collision in one cell has its next collision in another cell. These collision probabilities can be cast in terms of escape and transmission probabilities for each cell. Some methods exist that extend the collision probability approach to systems composed of more than two homogeneous cells. In this work, we present a novel collision probability method, based on previous work by Schneider et al. (2006a), for an arbitrary number of cells. The method operates by averaging the transmission probabilities across cells of the same shape, and thus assumes a certain level of homogeneity across all cells. When using multigroup cross sections, which the collision probability approach requires, it is necessary to consider the effect that a system's geometry and composition has on those multigroup cross sections. The cross sections must be computed in a way that accounts for the resonance self-shielding that may reduce the reaction rates in the resonance region. The process of developing self-shielded cross sections in a heterogeneous system utilizes an escape cross section. We compute this escape cross section using the same collision probabilities used to obtain the energy spectrum. Results are presented for simple two-cell systems, and preliminary results for four-cell simulations are also given. An extension to the method is provided that accounts for the fact that in thermal systems the assumption of homogeneity is not always valid. / text

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iThemba LABS is a particle accelerator facility housing a radionuclide production facility that uses a 66 MeV proton beam to produce radionuclides for medical and industrial use. Ionising radiation is produced by a variety of sources at Themba LABS. Ionising is a health hazard. High doses can cause acute radiation syndrome, i.e. "radiation sickness". Lower doses cannot cause acute symptom, but carry a risk of radiation-related cancer. Ionising radiation is also detrimental to materials, and can damage polymers and lubricants e.g. Shielding is used to reduce radiation levels to values that should be safe for the intended level of human occupancy. Shield performance is vital to human health and the life expectancy of polymers and lubricants, so that quality management in shield design is very important. However, until 2003, there has been no formal system at iThemba LABS to evaluate and improve all radiation shield designs and layouts from a radiation protection perspective. This study deals with evaluating and redesigning radiation shielding in the radionuclide production facility of iThemba LABS. There are several designs and layouts in the radionuclide production building of iThemba LABS that lead to unnecessary exposure of personnel to ionising radiation. The shielding in these areas are sub-standard. Performance criteria for radiation shields are developed. Inadequate radiation shields are identified. The identified inadequate shields are: the processing hotcells, the target store room and the hotcells in the radiopharmaceutical dispensing laboratory, Point-Kernel radiation shielding calculations are done to specify materials and material thickness that will adequately protect workers against ionising radiation. / Thesis (MSc. ARST) North-West University, Mafikeng Campus, 2004

Thesis (M.S.); Submitted to Tel Aviv Univ. (IL); 1 Dec 2003. / Published through the Information Bridge: DOE Scientific and Technical Information. "JLAB-PHY-03-219" "DOE/ER/40150-2651" Ran Shneor. 12/01/2003. Report is also available in paper and microfiche from NTIS.

Made available in DSpace on 2014-06-11T19:27:13Z (GMT). No. of bitstreams: 0 Previous issue date: 2008-02-01Bitstream added on 2014-06-13T19:35:06Z : No. of bitstreams: 1 moreira_af_me_ilha.pdf: 2273819 bytes, checksum: 1e78bc9d82277d9e226607de07b8a4a9 (MD5) / Coordenação de Aperfeiçoamento de Pessoal de Nível Superior (CAPES) / O objetivo deste trabalho é verificar a influência da atmosfera protetora nas características geométricas do cordão de solda, na susceptibilidade a formação de porosidades e inclusões, na microestrutura, na dureza e nas taxas de deposição e diluição do metal de solda obtido através das soldagens GMAW e FCAW, com arame sólido AWS E70S-6 e tubular AWS E309LT-1, respectivamente, visando otimizar o processo de recuperação por soldagem de turbinas Francis de aço carbono danificadas por erosão cavitacional. Para tanto, foram feitas deposições “bead on plate” com arames de 1,2 mm de diâmetro sobre chapas de aço carbono ASTM A-36. Utilizou-se como gás de proteção: Argônio puro, CO2 puro, misturas binárias Argônio/CO2, Argônio/O2 e uma mistura ternária Argônio/CO2/O2. No caso da soldagem FCAW, a taxa de deposição parece não ter estrita relação com o tipo de gás empregado, parecendo ser mais dependente da corrente de soldagem. Já parâmetros como largura, altura, penetração, taxa de diluição, nível de porosidades/inclusões e fração de ferrita parecem sofrer influência direta da composição química do gás de proteção empregado. No caso da soldagem GMAW, a taxa de deposição, largura, altura, penetração, taxa de diluição e nível de porosidades/inclusões demonstraram estar diretamente relacionados ao tipo de gás de proteção empregado. Não foi observada uma significativa variação nos valores de dureza do metal de solda obtidos com o processo FCAW devido a mudança na composição química do gás de proteção. Já os picos de dureza encontrados na ZTA para esse processo podem ser decorrentes da formação de carbonetos de cromo. No caso da soldagem GMAW, o que se pode observar foi uma queda na dureza desde o metal de solda até o metal de base. Os elevados valores de dureza do metal de solda para esse tipo de soldagem... / The goal of this study is to verify the influence of shielding gas composition in the geometric aspects of weld metals like width, height and depth, besides dilution and deposition rates. Besides this analyses were made upon the volumetric fraction of porosities/inclusions and upon the ferrite fraction to the austenitic weld metal. Additionally was withdrawn microhardness Vickers profiles from weld metal up base metal. Weld metals have been made with “bead on plate” deposition of wires AWS E309LT-1 and AWS E70S-6, both with 1,2 mm diameter, upon carbon steel ASTM A-36 plates by FCAW and GMAW welding process, respectively. The shielding gases employed were: pure Argon, pure CO2, binary mixtures of Argon/CO2, Argon/O2 and shielding gas containing Argon/CO2/O2. Related to the FCAW process, the deposition rate didn’t show a direct relationship with the chemical composition of shielding gas, being more dependent of the welding current. Parameters like width, height, depth, dilution rate, porosity/inclusion levels and ferrite fraction showed to keep a direct relation to the chemical composition of the shielding gas. To the GMAW process, deposition rate, width, height, depth, dilution ratio and porosity/inclusion levels maintained a direct relation to the chemical composition of shielding gas. It wasn’t observed a significative variation upon microhardness values of the weld metals produced with FCAW process because of the change in the shielding gas employed. In the GMAW welding, it was observed an abatement on the microhardness profile from the weld metal upto the base metal because, of, probably, the quenching increase in this area by the dilution of a metal with a high quantity of Manganese manganese coming from wire composition.