Index Catalog // ScholarsArchive@OSU

An advanced nodal discretization for the quasi-diffusion low-order equations

Creator:: Nes, Razvan
Abstract:: The subject of this thesis is the development of a nodal discretization of the low-order quasi-diffusion (QDLO) equations for global reactor core calculations. The advantage of quasi-diffusion (QD) is that it is able to capture transport effects at the surface between unlike fuel assemblies better than the diffusion approximation. We...
Resource Type:: Masters Thesis
Full Text:: scale 44 4.2. UO2-MOX configuration 45 4.3. Fast neutron flux in UO2 (3%)-MOX (12%) configuration 47

Neutron transport benchmarks for binary stochastic multiplying media : planar geometry, two energy groups

Creator:: Davis, Ian M. (Ian Mack)
Abstract:: Benchmark calculations are performed for neutron transport in a two material (binary) stochastic multiplying medium. Spatial, angular, and energy dependence are included. The problem considered is based on a fuel assembly of a common pressurized water nuclear reactor. The mean chord length through the assembly is determined and used as...
Resource Type:: Masters Thesis
Full Text:: random size. Neutron flux distributions are numerically computed using a discretized form of the

Burnup dependent neutron flux and heat source distribution in sphere-pac mixed carbide fuels

Creator:: Robinson, Cheryl A.
Abstract:: Through a joint effort by the Swiss Federal Institute for Reactor Research and Oregon State University, the fuels modeling code SPECKLE is being developed to model sphere-pac mixed carbide nuclear fuel. Major parameters, such as fission gas release, fuel restructuring phenomena, and the temperature and porosity distributions in the fuel...
Resource Type:: Masters Thesis
Full Text:: Burnup Dependent Neutron Flux and Heat Source Distribution

An assessment of a mesh-based method for slab-geometry transport in stochastic materials

Creator:: Bassett, Brody R.
Abstract:: There exists a variety of methods to solve the radiation transport equation in stochastic materials, including the Levermore-Pomraning method and the atomic mix method, each of which is designed to replicate the results given by benchmarks with materials rearranged in a randomized fashion. While the atomic mix method is very...
Resource Type:: Masters Thesis
Full Text:: proposed to approximate the brute force solution, the atomic mix model and the Levermore-Pomraning method

Exchange interaction studies in magnetic semiconductors by neutron scattering

Creator:: Wiren, Zachary Quincy
Abstract:: Theories predict that making a dilute magnetic semiconductor strongly p -type would allow it to remain ferromagnetic at room temperature. This is of intrest as a ferromagnetic semiconductor would allow for a great many applications in the field of spintronics. Neutron scattering experiments were undertaken to better understand the antiferromagnetic...
Resource Type:: Dissertation
Full Text:: Magnetic Semiconductors by Neutron Scattering Abstract approved: Tomasz M. Giebultowicz Theories predict

Solution of the 2D quasi-diffusion low-order equations using a coupled nodal/finite volume discretization in Cartesian and hexagonal geometry

Creator:: Nes, Razvan
Abstract:: An important improvement in the area of reactor core neutronic modeling is the development and use of the methods based on "quasi-diffusion" (QD) low-order equations. This family of methods takes into account the transport exactly using "functionals" computed by solving transport equations, and is amenable to solution with a variety...
Resource Type:: Dissertation
Full Text:: neutron flux, fast and thermal. The QD functionals are computed using the moments of the angular

Use of BC-523a liquid scintillator for simultaneous neutron spectroscopy and gamma counting with the implementation of a neutron history reconstruction algorithm

Creator:: Frey, Wesley D.
Abstract:: Real-time, high efficiency neutron spectroscopy has, historically, been a persistent challenge in the field of radiation detection and, for the most part, has gone unsolved. The most promising method to resolve this challenge, is the boron-capture technique using an organic scintillation system (BC-523a). Detectors that utilize this method possess an...
Resource Type:: Dissertation
Full Text:: to the negative flux multiplied by the neutron total cross section: d drmΣ−= φφ . (7) This

A comparative study of nodal course-mesh methods for pressurized water reactors

Creator:: Bukar, Kyari Abba
Abstract:: Several computer codes based on one and two-group diffusion theory models were developed for SHUFFLE. The programs were developed to calculate power distributions in a two-dimensional quarter core geometry of a pressurized power reactor. The various coarse-mesh numerical computations for the power calculations yield the following: the Borresen's scheme applied...
Resource Type:: Masters Thesis
Full Text:: model have quite attractive features. A Comparative Study of Nodal Coarse-Mesh Methods for

Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

Creator:: Betts, Curt M.
Abstract:: The solution of coupled neutronic/thermal hydraulic nuclear reactor calculations requires the treatment of the nonlinear feedback induced by the thermal hydraulic dependence of the neutron cross sections. As a result of these nonlinearities, current solution techniques often diverge during the iteration process. These instabilities arise due to the low level...
Resource Type:: Masters Thesis
Full Text:: convergence criterion (outer flux profile iteration) 49. Thermal neutron flux profile variations due to

ScholarsArchive@OSU

An advanced nodal discretization for the quasi-diffusion low-order equations

Neutron transport benchmarks for binary stochastic multiplying media : planar geometry, two energy groups

Burnup dependent neutron flux and heat source distribution in sphere-pac mixed carbide fuels

An assessment of a mesh-based method for slab-geometry transport in stochastic materials

Exchange interaction studies in magnetic semiconductors by neutron scattering

Solution of the 2D quasi-diffusion low-order equations using a coupled nodal/finite volume discretization in Cartesian and hexagonal geometry

Use of BC-523a liquid scintillator for simultaneous neutron spectroscopy and gamma counting with the implementation of a neutron history reconstruction algorithm

A comparative study of nodal course-mesh methods for pressurized water reactors

Numerical techniques for coupled neutronic/thermal hydraulic nuclear reactor calculations

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