Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Bibliography: leaves 145-147. / by Wee Tee Loh. / M.S.

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.

The delayed neutron spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. In this study, a sensitivity analysis was performed to study the sensitivity of the reactor power and power density to uncertainties in the delayed neutron spectra during a rod ejection accident. The generalised methodology, developed by Cacuci et. al., was used to derive a set of sensitivity derivatives. This method is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. A two-energy multigroup and two precursor group model was formulated for the INFCE reference design MOX-fuelled LMFBR. The accidents studied were central control rod ejections with ejection times of 2, 10, and 30 seconds. The power and power density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed neutron precursor group, resulting from the fission of U-238, producing neutrons in the first energy group. It was found, for example, that for a rod ejection time of 30 seconds, an uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy was only 1.6%. The versatility and accuracy of Cacuci’s methodology has been demonstrated. The results of the sensitivity analysis indicates the need for improved delayed neutron spectral data in order to reduce the uncertainties in the accident analyses. The model can be extended by using more energy groups, more precursor groups, and more spatial dimensions. Other important responses that may be studied are the linear power density, linear heat rate, and reactivity worths. / Ph. D.

Thesis (M.S.)--Massachusetts Institute of Technology, Dept. of Nuclear Engineering, 1980. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / by King-Wo Thomas Chiu. / M.S.

Thesis. 1979. Ph.D. cn--Massachusetts Institute of Technology. Dept. of Nuclear Engineering. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND SCIENCE. / Includes bibliographical references. / Ph.D.cn

The SABR fusion-fission hybrid concept for a fast burner reactor, which combines the IFR-PRISM fast reactor technology and the ITER tokamak physics and fusion technology, is adapted for a fusion-fission hybrid reactor, designated SABrR. SABrR is a sodium-cooled 3000 MWth reactor fueled with U-Pu-10Zr. For the chosen fuel and core geometry, two configurations of neutron reflector and tritium breeding structures are investigated: one which emphasizes a high tritium production rate and the other which emphasizes a high fissile production rate. Neutronics calculations are performed using the ERANOS 2.0 code package, which was developed in order to model the Phenix and SuperPhenix reactors. Both configurations are capable of producing fissile breeding ratios of about 1.3 while producing enough tritium to remain tritium-self-sufficient throughout the burnup cycle; in addition, the major factors which limit metal fuel residence time, fuel burnup and radiation damage to the cladding material, are modest.

Thesis (Ph.D.)--Massachusetts Institute of Technology, Dept. of Mechanical Engineering, 1982. / MICROFICHE COPY AVAILABLE IN ARCHIVES AND ENGINEERING. / Includes bibliographical references. / by Alfred Walters Reed. / Ph.D.

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