Inservice property degradation is known to occur in fuel clads of reactors systems currently in operation and under design_ Irradiation, corrosion, diffusion induced mass transfer, and a host of mechanical influences constitute the complex environmental variables responsible for the degradation. The degradation could occur in the bulk of the clad or through its thickness depending on the component of the environment and the reactor operating history. Synergistic influences of more than one component of the environment must also be considered. From the mechanics viewpoint, the degraded alloy is a material with nonuniform properties. Thus, the basic stress-strain relations require modifications which are functions of the reactor operating history. The nature of the balance equations of stress equilibrium changes, especially if the degradation occurs through the thickness of the material. Prescription of the complete stress-strain relationship is required for an elastic-plastic type of analysis at the beginning of each new time step in modeling the performance of the material. Knowledge of the metallurgical and mechanical nature of the degraded alloy and the spatial variation of the properties which result is a prerequisite for the modeling. Evidence from available literature is presented to illustrate this problem. The study involves the degradation of the 316 type stainless steel considered for use in Liquid Metal Fast Breeder Reactors where sodium is used as the coolant. Nonuniform changes in properties of the steel have been found to occur due to the thermal, thermochemical, and irradiation environment to which it is exposed. Variations in imparity concentrations (such as carbon in the steel) of several orders of magnitude compared to the original values have been observed under controlled sodium exposures. At temperatures relevant to the reactor system migration of impurities by diffusion, compound formations and carburization/decarburization behavior have been observed to occur. Mechanical property measurements such as tensile and yield strengths made under these conditions indicate that thermal and thermochemical influences can result in variations in the above properties quite different from the original material. Modified formulations of the elastic and elastic-plastic analysis of the degraded fuel-clad are presented in two dimensions. The elastic and plastic parameters relating to the properties of the degraded material are represented by spatially varying functions as opposed to being treated as constants which is the conventional case. The changes in the mathematical nature of the constitutive equations are demonstrated by sample illustrations and solutions involving continuous changes in the elastic moduli through-the-thickness of the clad. Recommendations for the establishment of improved Reactor Research Development Standards are made based on the studies. / Ph. D.

A nonlinear model is developed for the xenon induced flux oscillation problem that occurs in nuclear power plants. The model is based on Galerkins' method of weighted residuals applied to multigroup diffusion theory. A similar linear model is developed by the same methods in order to consider the effects of the nonlinearities of the system. The effects of multi- and single-energy group considerations are also examined. Finally, the effects of various number of basis functions used to approximate the flux, iodine, and xenon concentrations is determined. A partial listing of the computer program XORA, developed from the nonlinear and linear models, is given along with representative input and output from this program. / Ph. D.

A subcritical heterogeneous nuclear reactor was designed and constructed for use in the nuclear engineering program at Virginia Polytechnic Institute. Natural uranium fuel and reactor grade graphite were used as building materials. Considering that no two samples of moderator have the same characteristics, the resulting parameters measured on this reactor are in good agreement with published data. The reactor performs the primary function as a versatile laboratory instrument that can measure as many reactor parameters as possible. / Master of Science

Theoretial calculations of critical mass, temperature coefficient, and transient behavior have been made for proposed uranium dioxide and urania-thoria fuel cores for the Virginia Polytechnic Institute UTR-10 Reactor. / Master of Science

The problem of installing a control rod in the V. P. I. subcritical reactor is complicated by the limiting two foot clearance between the top of the reactor and the ceiling. Within this space it is necessary to insert and operate an effective seven foot, solid, straight rod along the vertical axis of the reactor. The problem is overcome by cutting the rod in six inch sections at an angle of 45 degrees and installing an internal pivoting arm which allows the sections to break 60 degrees with respect to one another. The rod is wrapped on a drum which takes the shape of a hexagonal helix when withdrawn from the pile. When inserted, an effective solid, straight rod is presented to the neutron flux. The control rod is motor driven and remotely operated. Provision is made for changing the absorber material in the control rod to permit the investigation of materials having various neutral absorbing properties. Tests were performed using cadmium as the absorber material to determine the effect of the control rod on the neutron flux. Results indicate that the control rod, although flexible, is effectively a solid, straight rodo An equation is obtained for expressing the control rod effect as a function of pile level and control rod position. / Master of Science

Fluidization, a relatively new unit operation by which a solid and gas or liquid may be contacted, is being widely developed in the field of catalytic cracking of petroleum because of its characteristic reduction of temperature gradients within the reaction mass. Basic research of the heat transfer properties of fluidized systems has lagged far behind industrial applications. It was the purpose of this investigation to evaluate the effect of temperature driving force and mass superficial air velocity on the coefficient of heat transfer at the bed wall of an externally heated, fluidized bed of Ottawa sand at wall temperature of 200, 400, and 600 °F , and average mass superficial air velocities of 82.5, 123.2, 170.3, and 217.5 pounds per hour-square foot. The tests were carried out under steady state conditions of air flow and bed wall temperature. A complete heat and material balance, including evaluation of heat losses, was made for each test. The boundary coefficients based on the internal area and temperature of the pipe wall were calculated. The effects of mass superficial air velocity and wall temperature on the boundary coefficient of heat transfer and on bed and bed wall temperature gradients were studied. From observations made it was noted that the fluidization range of the Ottowa sand bed began at a mass superficial air velocity of 91.0 pounds per hour-square foot and ended at 210.0 pounds per hour-square foot, the velocity at which slugging occurred throughout the bed. The horizontal temperature gradient across the bed increased with increasing bed wall temperature, increasing from a minimum of 0 °F at 200 °F wall temperature to 6 °F at 600 °F . The rate of heat flux to the air stream passing through the fluidized bed increased with mass air flow rate at constant bed wall temperature. The minimum heat flux was 84 Btu per hour and occurred at 200 °F and 82.5 pounds per hour-square foot, while the maximum was 1172 Btu per hour and occurred at 600 °F and 217.5 pounds per hour-square foot. The coefficient of heat transfer increased with bed wall temperature, reaching maximum values of 9.55, 13.40, 13.31, and 13.30 Btu per hour-square foot-°F at 600 °F and at mass superficial air velocities of 82.5, 123.2, 170.3, and 217.5 pounds per hour-square foot, respectively. / Master of Science

A Sigma Pile was constructed to serve as a laboratory facility at V.P.I. Reproducible foil positions are possible with the positioning plates installed on the pile. Withdrawal of the two outer blocks in each foil channel is facilitated by a coupling mechanism. Five source positions permit operation with reduced harmonics. The neutron distribution in the pile was proven to be symmetric by horizontal and vertical traverses. By means of a diffusion length measurement, which gave an "L” of 54.5 ± 2 cm, and a Fermi Age measurement, which gave a "r” of 357 ± 5 cm², the parameters of the pile were shown to be in good agreement with published data on graphite. / Master of Science

Thesis (M.B.A.)--Boston University / PLEASE NOTE: Boston University Libraries did not receive an Authorization To Manage form for this thesis or dissertation. It is therefore not openly accessible, though it may be available by request. If you are the author or principal advisor of this work and would like to request open access for it, please contact us at open-help@bu.edu. Thank you. / There are a number of nuclear reactor types. It will be a purpose of this thesis to examine both a power reactor and a research reactor to determine what inherent measures, as well as applied measures, safety controls, and techniques are used in attempting to make nuclear reactor operations safe. Whereas nuclear reactors are extremely expensive and applicable to only a limited number of industries, radioisotopes, the unstable form of various elements, offers to a much larger industrial family a useful, inexpensive, and simple tool. A second purpose of this thesis will be to observe, describe, and evaluate the use and application of radioisotopes as they are currently being employed in industry, with particular reference to safety features. / 2999-01-01

The purpose of this thesis is to design a robust test facility for a small space nuclear power system and model its physical behavior under different scenarios. The test facility will be used to simulate a 1-10kWe nuclear reactor, its electrical generation, and heat removal capabilities. This simulator will be used to explore, test and understand the steady-state and transient operation capabilities of small space nuclear power systems. Currently, the system is planned to operate on a variable, electrical heat source directly connected to heat pipes. The heat pipes are to be stainless steel with a water working fluid. These heat pipes will then be connected to a power conversion simulator or actual power conversion technologies. The power conversion simulator is connected to a radiator using a water based heat pipe network using fins and connecting plates in a cylindrical geometry. Modeling of the facility was performed using two different analysis programs, STELLA and SolidWorks. STELLA was used as a lumped sum heat transport code, and SolidWorks was used as a more accurate system to verify the validity of STELLA's results. Both programs were used to analyze startup, heat pipe failures, and loss of power conversion with the end goal of finding safe operational transient scenarios for the transient test facility. / Graduation date: 2013

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