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Filtering by: Subject Reactor fuel reprocessing Remove constraint Subject: Reactor fuel reprocessing Advisor Klein, Andrew C. Remove constraint Advisor: Klein, Andrew C.

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  • Monte Carlo burnup analysis code development and application to an incore thermionic space nuclear power system

    Creator:
    Abdul-hamid, Shahab A.
    Abstract:
    Lattice bum-up calculations in thermal reactors are complicated by the necessity for use of transport theory to represent fuel rods, control rods, and burnable absorbers, by many time-dependent variables which must be considered in the analysis, and by geometric complexity which introduces time-dependent, spatial-spectral variations. Representation of lattice structure in...
    Resource Type:
    Masters Thesis
    Full Text:
    is uniformly mixed in the fuel. MCNPBURN was benchmarked using a standard Pressurized Water Reactor

  • A model for waterside oxidation of zircaloy fuel cladding in pressurized water reactors

    Creator:
    Almarshad, Abdullah I. A.
    Abstract:
    A model has been developed to predict the long-term oxidation rate of Zircaloy-4 for ex-reactor (autoclave) and in-reactor (PWR) environments and operating conditions. A computer program has been written to solve the oxygen diffusion equation by employing a fully implicit finite difference method for a one dimensional cylindrical geometry. The...
    Resource Type:
    Dissertation
    Full Text:
    oxide formed on the waterside surface of 132 reactor-irradiated rods with various fuel burnups. The

  • Determination of a calculation bias in the MCNP model of the OSTR

    Creator:
    Kitto, Allyson K.
    Abstract:
    Oregon State University is home to a TRIGA® Mark II reactor. In October of 2008, the reactor began operating on low enriched uranium fuel. A model of the facility exists in MCNP, a Monte Carlo code that can be used for criticality calculations. Until now, a bias in the calculation...
    Resource Type:
    Masters Thesis
    Full Text:
    , the reactor began operating on low enriched uranium fuel. A model of the facility exists in MCNP, a

  • Modeling transient thermalhydraulic behavior of a thermionic fuel element for nuclear space reactors

    Creator:
    Al-Kheliewi, Abdullah S.
    Abstract:
    A transient code (TFETC) for calculating the temperature distribution throughout the radial and axial positions of a thermionic fuel element (TFE) has been successfully developed. It accommodates the variations of temperatures, thermal power, electrical power, voltage, and current density throughout the TFE as a function of time as well as...
    Resource Type:
    Dissertation
    Full Text:
    , where t is a reactor transient time in seconds, the fuel temperature does not exceed the melting

  • Fission Gas Bubble Behavior in Uranium Carbide

    Creator:
    Matthews, Christopher
    Abstract:
    The need for cheap reliable energy, while simultaneously avoiding uranium supply constraints makes uranium carbide (UC) fueled Gas Fast Reactors offer an attractive nuclear reactor design. In order to qualify the fuel, an enhanced understanding of the behavior of uranium carbide during operation is paramount. Due to a reduced re-solution...
    Resource Type:
    Dissertation
    Full Text:
    ) fueled Gas Fast Reactors offer an attractive nuclear reactor design. In order to qualify the fuel, an