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Search theses and dissertations gathered from participating repositories worldwide. Every result links back to the library that holds it. No account is needed.
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Showing 1 to 20 of 77 for “"Neutron transport"”.
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Quasi-static Methods in Neutron Transport
… advancement of the research in the field of the neutron kinetics of innovative nuclear reactors. As modern computer frames are becoming more and more performing and because the treatment of innovative nuclear reactor systems requires high accuracy results, the neutron transport theory is …
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Nonlinear acceleration methods for even-parity neutron transport
Convergence acceleration methods for even-parity transport are being developed that have the potential to speed up transport calculations and provide a natural avenue for an implicitly coupled multiphysics code. An investigation was performed into the acceleration properties of the introduction of …
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High-Energy Neutron Transport Through Tungsten and Iron
Made available in DSpace on 2014-12-14T06:26:08Z (GMT). No. of bitstreams: 1 8004188.pdf: 5825708 bytes, checksum: 7e7b6496b41591a9809501a51685b4d6 (MD5) Previous issue date: 1979
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PROBABILISTIC METHOD IN RADIATIVE TRANSFER AND NEUTRON TRANSPORT PROBLEMS
… time-dependent and time independent transfer and transport problems in plane-parallel atmospheres. In the present thesis, a scheme has been outlined to develop a probabilistic model of photon movement, capable of interpreting the transfer problems in spherical geometry. As the basic inspiration of …
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Calculating reaction rate derivatives in Monte Carlo neutron transport/
… There is great interest in solving the neutron transport with Monte Carlo methods due to their extremely high fidelity, but Monte Carlo methods are too slow to run in an iterative brute-force search of the reactor parameter space. This thesis discusses the derivation, implementation, and …
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Accelerating Monte Carlo neutron transport with variable fidelity nuclear data
Monte Carlo (MC) neutron transport simulations can be performed with several types of nuclear data representations; the most common ones are continuous energy (CE) and multi-group (MG) data. CE is the highest fidelity option, since it allows each cross section resonance to be represented precisely. …
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GPU-Oriented Algorithms for Continuous Energy Monte Carlo Neutron Transport
… take place in continuous energy Monte Carlo (MC) neutron transport simulations. These simulations play a crucial role in designing fission, fusion, and security systems that may play a pivotal role in the transition to a decarbonized world. This work contributes to adapting continuous energy MC …
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A low order acceleration scheme for solving the neutron transport equation
… equations, and has been applied to the neutron transport problems for many years. The MOC method requires many transport iterations to solve large heterogeneous LWR reactor problems with high dominance ratio, and effective acceleration schemes are necessary to make MOC method practical. …
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Development and analysis of high order neutron transport-depletion coupling algorithms
The coupling of depletion and neutron transport together creates a particularly challenging mathematical problem. Due to the stiffness of the ODE, special algorithms needed to be developed to minimize the number of transport simulations required. In addition, for stochastic transport, both the time …
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The Phase Space Time Evolution method applied to multigroup neutron transport
… method was initially developed for one speed neutrons. This discussion considers the alterations performed in transforming the one speed case to a more general multigroup code. Also the complications which arose in this transformation are discussed. The multigroup formulation of the PSTE …
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On the Spectrum of Neutron Transport Equations with Reflecting Boundary Conditions
… is devoted to investigating the time dependent neutron transport equations with reflecting boundary conditions. Two typical geometries --- slab geometry and spherical geometry --- are considered in the setting of <I>L^p</I> including <I>L^1</I>. Some aspects of the spectral properties of the …
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Computational methods for efficient nuclear data management in Monte Carlo neutron transport simulations
… and utilizing nuclear data in Monte Carlo neutron transport code simulations. Using the OpenMC code, profiling studies are conducted in order to determine the types of nuclear data that are used in realistic reactor physics simulations, as well as the frequencies with which those data are …
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A novel equivalence method for high fidelity hybrid stochastic-deterministic neutron transport simulations
… Downstream highly parallelized 3D deterministic transport can then use those cross sections to compute accurate solutions at the full core scale. This hybrid computation scheme makes the most of both worlds to achieve fast and accurate reactor physics simulations. Among the few remaining …
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A lattice boltzmann/s n framework for coupled heat transfer and neutron transport problems
… dissertation is to study the coupling between neutron transport and heat conduction equations with temperature-dependent thermal conductivity and temperature-dependent neutron cross sections. The neutron transport equation is solved using the discrete ordinates method (SN ); and the heat …
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Numerical stability of Monte Carlo neutron transport and isotopic depletion for nuclear reactor analysis
Coupling Monte Carlo neutron transport with isotopic depletion is known to produce non-physical results for large reactor geometries. This thesis begins with a survey of the stable methods used to avoid this and highlights some of their drawbacks. Chapter 2 introduces the known phenomenon of …
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Reactor agnostic multi-group cross section generation for fine-mesh deterministic neutron transport simulations
A key challenge for full-core transport methods is reactor agnostic multi-group cross section (MGXS) generation. Monte Carlo (MC) presents the most accurate method for MGXS generation since it does not require any approximations to the neutron flux. This thesis develops novel methods that use MC to …
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Full core 3D neutron transport simulation using the method of characteristics with linear sources
The development of high fidelity multi-group neutron transport-based simulation tools for full core Light Water Reactor (LWR) analysis has been a long-standing goal of the reactor physics community. While direct transport simulations have previously been far too computationally expensive, advances …
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Modeling and simulation of The Transient Reactor Test Facility using modern neutron transport methods
The Transient Reactor Test Facility (TREAT) has regained the interest of the nuclear engineering community in recent years. While TREAT's design makes it uniquely suited to transient fuel testing, it also makes the reactor very challenging to model and simulate. In this thesis, we build a Monte …
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