Engineering Commons^™

Modeling A Nuclear Research Reactor And Radiation Dose Estimation In An Accident Scenario, Abdulaleem Abdulmajeed Bugis

Doctoral Dissertations

“A detailed, flexible three-dimensional (3D) model of the Missouri S&T Reactor (MSTR) with a heterogeneous core geometry was developed using the Standardized Computer Analyses for Licensing Evaluation (SCALE). A Graphical User Interface (GUI) was developed for SCALE, which allows the user to generate an input file automatically. The SCALE model was validated with a Monte Carlo N-Particle Transport (MCNP) model of the MSTR. The validation process was based on the criticality calculations using the Criticality Safety Analysis Sequence (CSAS6). Three geometrical models were examined. The SCALE model that has the full detailed geometry showed a good agreement with the MCNP …

Beta-Delayed Neutron Data And Models For Scale, Kemper Dyar Talley

Doctoral Dissertations

Recent advancements in experimental and theoretical nuclear physics have yielded new data and models that more accurately describe the decay of fission products compared to historical data currently used for many applications. This work examines the effect of the adopting the Effective Density Model theory for beta-delayed neutron emission probability on calculations of delayed-neutron production and fission product nuclide concentrations after fission bursts as well as the total delayed neutron fraction in comparison with the Keepin 6-group model. We use ORIGEN within the SCALE code package for these calculations. We show quantitative changes to the isotopic concentrations for fallout nuclides …

Automated Doppler Broadening Of Cross Sections For Neutron Transport Applications, Shane William Daniel Hart

Doctoral Dissertations

This dissertation discusses the research and development of new cross section temperature handling techniques for the SCALE computer code, which is developed and maintained at Oak Ridge National Laboratory. In particular, methods will be added to the KENO Monte Carlo code.

Areas of interest include: neutron scattering off of heavy isotopes in the epithermal energy range, implementation of Doppler pre-broadening of continuous energy one-dimensional cross-section data, implementation of interpolation on the continuous energy two-dimensional cross sections, and implementation of the direct S(A,B) [S alpha beta] method for thermal neutron scattering and interpolation on that data.

Accurate cross section scattering off …

Optimization Of Transcurium Isotope Production In The High Flux Isotope Reactor, Susan Hogle

Doctoral Dissertations

The Radiochemical Engineering Development Center at Oak Ridge National Laboratory is the world's leader in production of californium-252. This and other heavy actinides are produced by irradiation of mixed curium/americium targets in the High Flux Isotope Reactor. Due to the strong dependence of isotopic cross sections upon incoming neutron energy, the efficiency with which an isotope is transmuted is highly dependent upon the neutron flux energy spectrum and intensities. There are certain energy ranges in which the rate of fissions in feedstock materials can be minimized relative to the rate of (n,γ) absorptions. This work shows that by perturbing the …

Spatially-Dependent Reactor Kinetics And Supporting Physics Validation Studies At The High Flux Isotope Reactor, David Chandler

Doctoral Dissertations

The computational ability to accurately predict the dynamic behavior of a nuclear reactor core in response to reactivity-induced perturbations is an important subject in the field of reactor physics. Space-time and point kinetics methodologies were developed for the purpose of studying the transient-induced behavior of the Oak Ridge National Laboratory (ORNL) High Flux Isotope Reactor’s (HFIR) compact core. The space-time simulations employed the three-group neutron diffusion equations, which were solved via the COMSOL partial differential equation coefficient application mode. The point kinetics equations were solved with the PARET code and the COMSOL ordinary differential equation application mode. The basic nuclear …