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

A dual neutron and gamma detection system has been developed for online measurements. The system consists of a single crystal mounted on a photomultiplier tube to detect simultaneously gamma radiation as well as thermal neutrons. A compact data acquisition system has been used for neutron and gamma discrimination. The system has been tested with different gamma energies and with an Am-Be neutron source at the University of Ontario Institute of Technology neutron facility. This thesis presents the characteristics of the developed detector, and experimental data carried out in different experiments in different fields. / UOIT

In this thesis, the parallel performance of the finite element-spherical harmonics (FE-PN) method implemented in the general-purpose radiation transport code EVENT is studied both analytically and empirically. EVENT solves the coupled set of space-angle discretized FE-PN equations using a parallel block-Jacobi domain decomposition method. As part of the analytical study, the thesis presents complexity results for EVENT when solving for a 3D criticality benchmark radiation transport problem in parallel. The empirical analysis is concerned with the impact of the main algorithmic factors affecting performance. Firstly, EVENT supports two solution strategies, namely MOD (Moments Over Domains) and DOM (Domains Over Moments), to solve the transport equation in parallel. The two strategies differ in the way they solve the multi-level space-angle coupled systems of equations. The thesis presents empirical evidence of which of the two solution strategies is more efficient. Secondly, different preconditioners are used in the Preconditioned Conjugate Gradient (PCG) inside EVENT. Performance of EVENT is compared when using three preconditioners, namely diagonal, SSOR(Symmetric Successive Over-Relaxation) and ILU. The other two factors, angular and spatial resolutions of the problem affect both the performance and precision of EVENT. The thesis presents comparative results on EVENTs performance as these two resolutions are increased. From the empirical performance study of EVENT, a bottleneck is identified that limits the improvement in performance as number of processors used by EVENT is increased. In some experiments, it is observed that uneven assignment of computational load among processors causes a significant portion of the total time being spent in synchronization among processors. The thesis presents two indicators that identify when such inefficiency occur; and in such a case, a load rebalancing strategy is applied that computes a new partition of the problem so that each partition corresponds to equal amount of computational load.

A general finite difference discretization of the time-dependent radiation transport equation is developed around the framework of an existing steady-state three dimensional radiation transport solver based on the slice-balance approach. Three related algorithms are outlined within the general finite difference scheme: an explicit, an implicit, and a semi-implicit approach. The three algorithms are analyzed with respect to the discretizations of each element of the phase space in the transport solver. The explicit method, despite its small computational cost per time step, is found to be unsuitable for many purposes due to its inability to accurately handle rapidly varying solutions. The semi-implicit method is shown to produce results nearly as reliable as the fully implicit solver, while requiring significantly less computational effort.

Linear discontinuous (LD) spatial discretization of the transport operator can generate negative angular flux solutions. In slab geometry, negativities are limited to optically thick cells. However, in multi-dimension problems, negativities can even occur in voids. Past attempts to eliminate the negativities associated with LD have focused on inherently positive solution shapes and ad-hoc fixups. We present a new, strictly non-negative finite element method that reduces to the LD method whenever the LD solution is everywhere positive. The new method assumes an angular flux distribution, e , that is a linear function in space, but with all negativities set-to- zero. Our new scheme always conserves the zeroth and linear spatial moments of the transport equation. For these reasons, we call our method the consistent set-to-zero (CSZ) scheme. CSZ can be thought of as a nonlinear modification of the LD scheme. When the LD solution is everywhere positive within a cell, psi csz = psi LD. If psi LD < 0 somewhere within a cell, psi csz is a linear function psi csz with all negativities set to zero. Applying CSZ to the transport moment equations creates a nonlinear system of equations which is solved to obtain a non-negative solution that preserves the moments of the transport equation. These properties make CSZ unique; it encompasses the desirable properties of both strictly positive nonlinear solution representations and ad-hoc fixups. Our test problems indicate that CSZ avoids the slow spatial convergence properties of past inherently positive solutions representations, is more accurate than ad-hoc fixups, and does not require significantly more computational work to solve a problem than using an ad-hoc fixup. Overall, CSZ is easy to implement and a valuable addition to existing transport codes, particularly for shielding applications. CSZ is presented here in slab and rect- angular geometries, but is readily extensible to three-dimensional Cartesian (brick) geometries. To be applicable to other simulations, particularly radiative transfer, additional research will need to be conducted, focusing on the diffusion limit in multi-dimension geometries and solution acceleration techniques.

Made available in DSpace on 2014-10-09T12:45:02Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:57:10Z (GMT). No. of bitstreams: 1 07020.pdf: 10603359 bytes, checksum: ba346d79b2ac338721d3936457ab850b (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

A instalação do IPEN para pesquisas em BNCT (Terapia por Captura de Nêutrons em Boro) utiliza o canal de irradiação número 3 do reator IEA-R1, no qual tem-se um campo misto de radiação nêutrons e gama. As pesquisas em andamento necessitam que o campo de radiação, na posição de irradiação de amostra, tenha na composição os nêutrons térmicos maximizados e os componentes de nêutrons epitérmicos, rápidos e radiação gama minimizados. Este trabalho foi desenvolvido com o objetivo de avaliar se o campo de radiação atual na instalação é adequado aos trabalhos em BNCT. Para cumprir com este objetivo, uma metodologia para dosimetria de nêutrons térmicos e radiação gama em campos mistos de altas doses, que não era disponível no IPEN, foi implantada no Centro de Engenharia Nuclear do IPEN, utilizando dosímetros termoluminescentes TLDs 400, 600 e 700. Para as medidas de fluxo de nêutrons térmicos e epitérmicos foram utilizados detetores de ativação de ouro aplicando a técnica de razão de cádmio. Um simulador antropomórfico cilíndrico composto de discos de acrílico foi desenvolvido e testado na instalação e para obter valores teóricos do fluxo de nêutrons e a dose ao longo do simulador antropomórfico foi utilizado o código computacional DOT 3.5. Na posição correspondente a aproximadamente metade do comprimento do cilindro do simulador antropomórfico, foram obtidos os seguintes valores: fluxo de nêutrons térmicos (2,52 ± 0,06).108n/cm2s, epitérmicos (6,17 ± 0,26).107n/cm2s, dose absorvida devido a nêutrons térmicos de (4,2 ± 1,8)Gy e devido a radiação gama (10,1 ± 1,3)Gy. Os valores obtidos mostram que os fluxos de nêutrons térmicos e epitérmicos são adequados para os estudos em BNCT, porém, a dose devido a radiação gama está elevada, indicando que a instalação deve ser aprimorada. / IPEN facility for researches in BNCT (Boron Neutron Capture Therapy) uses IEA-R1 reactor\'s irradiation channel number 3, where there is a mixed radiation field neutrons and gamma. The researches in progress require the radiation fields, in the position of the irradiation of sample, to have in its composition maximized thermal neutrons component and minimized, fast and epithermal neutron flux and gamma radiation. This work was developed with the objective of evaluating whether the present radiation field in the facility is suitable for BNCT researches. In order to achieve this objective, a methodology for the dosimetry of thermal neutrons and gamma radiation in mixed fields of high doses, which was not available in IPEN, was implemented in the Center of Nuclear Engineering of IPEN, by using thermoluminescent dosimeters TLDs 400, 600 and 700. For the measurements of thermal and epithermal neutron flux, activation detectors of gold were used applying the cadmium ratio technique. A cylindrical phantom composed by acrylic discs was developed and tested in the facility and the DOT 3.5. computational code was used in order to obtain theoretical values of neutron flux and the dose along phantom. In the position corresponding to about half the length of the cylinder of the phantom, the following values were obtained: thermal neutron flux (2,52 ± 0,06).108n/cm2s, epithermal neutron flux (6,17 ± 0,26).107.106n/cm2s, absorbed dose due to thermal neutrons (4,2 ± 1,8)Gy and (10,1 ± 1,3)Gy due to gamma radiation. The obtained values show that the fluxes of thermal and epithermal neutrons flux are appropriate for studies in BNCT, however, the dose due to gamma radiation is high, indicating that the facility should be improved.

Made available in DSpace on 2014-10-09T12:45:02Z (GMT). No. of bitstreams: 0 / Made available in DSpace on 2014-10-09T13:57:10Z (GMT). No. of bitstreams: 1 07020.pdf: 10603359 bytes, checksum: ba346d79b2ac338721d3936457ab850b (MD5) / Tese (Doutoramento) / IPEN/T / Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP

Water is a major component of living organisms, which can be 70-85% of the weight of cells. For this reason, water is a main target of ionizing radiations and plays a central role in radiobiology. Heavy ions, electrons and photons interact with water molecules; mainly by ionization and excitation. Neutrons interact with water molecules by elastic interactions, which generate recoil ions that will create ionizations and excitations in water molecules. These fast events (~10[superscript -12] s) lead to the formation of Reactive Oxygen Species (ROS). The ROS, in particular the hydroxyl radical (¨OH), interact with neighbour molecules such as proteins, lipids and nucleic acids by chemical interaction. Microbeams can irradiate selectively either the external membrane, the cytoplasm and the cell nucleus. These studies have shown that cell survival is greatly reduced when the nucleus is irradiated, but that this is not the case when cytoplasm or cell membrane is irradiated. Thus, DNA is a very sensitive site to ionizing radiation and ROS. For this reason, DNA has long been considered the most important molecule to explain radiobiological effects such as cell death. However, this concept has been challenged recently by new experimental results that have shown that cells which have not been directly in contact with radiation are also affected. This is called the bystander effect. Further studies have shown that a group of cells and their environment reacts collectively to radiation. A hypothesis put forward to explain this radiobiological phenomenon is that a irradiated cell will secrete signalling molecules that will affect non-irradiated cells. The implicated phenomenon and molecules are poorly understood at this moment. The purpose of this work is to improve our comprehension of the phenomenon in the microsecond that follows the irradiation. To these ends, a new Monte-Carlo simulation program of water radiolysis by photons has been generated. For photons of energy <2 MeV, they interact with water mainly by Compton and photoelectric effects, which create energetic electrons in water. The created electrons are then followed by our existing programs to simulate the radiolysis of water by photons. Similarly, a new code has been built to simulate the neutrons interaction with water. This code simulates the elastic collisions of a neutron with water molecules and calculates the number and energy of recoil protons and oxygen ions. The main part of this Ph.D. work was the generation of a non-homogeneous Monte-Carlo Step-By-Step (SBS) simulation code of non-homogeneous radiation chemistry. This new program has been used successfully to simulate radiolysis of water by ions of various LET, pH, ion types ([superscript 1]H[superscript +], [superscript 4]He[superscript 2+], [superscript 12]C[superscript 6+]) and temperature. The program has also been used to simulate the dose-rate effect and the Fricke and Ceric dosimeters. More complex systems (glycine, polymer gels and HCN) have also been simulated.

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