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In reactor core design, a gap exists between the manual calculation of few-group constants and the many-group calculation, by large computer programs. A method is needed by which group constants may be calculated easily and quickly. The FASPEC program is designed to reduce the amount of manual calculation and to complement the large program by reducing the number of times the large program must be run to achieve desired results. The program calculates group constants from 940 microgroups, collapsing to any user-specified number of macrogroups up to 47. FASPEC is based on group-averaged flux calculations by a solution of the Infinite medium neutron transport equation. Flux contributions from inelastic scatter are included while those from neutron up-scatter are not. The energy spectrum considered is from 10 MeV to 0.625 eV. Required input is the atomic number density of each isotope, the number of macrogroups desired and the upper and lower microgroup numbers of each macrogroup. Input is facilitated by prompting in each case. Cross section look-up tables were provided by the Very Improved Monte Carlo code (VIM) for a mid-range Infinite hexagonal lattice. Self-shielding effects are included indirectly. A brief user's guide is provided. Group constants calculated and stored for either terminal display or printed output are group number, lowest energy of the group, macroscopic removal cross section, macroscopic absorption cross section, diffusion coefficient, flux, macroscopic fission cross section, v, the average number of neutrons emitted per fission, and vΣ_f. / Master of Science

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Thesis (M. S.)--Mechanical Engineering, Georgia Institute of Technology, 2008. / Committee Chair: Willem F.G. Van Rooijen; Committee Member: Ghiaasiaan, Seyed M; Committee Member: Weston M. Stacey

Digitized by Kansas Correctional Industries

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The Liquid Salt Cooled Reactor (LSCR) provides several potential benefits compared to pressurized water-cooled reactor systems. These include low operating pressure of the liquid salt coolant, the high burnup tolerance of the fuel, and the high operating temperatures which leads to increases in efficiency. However, due to inherently low heavy metal loading, the fuel cycle design presents specific challenges. In order to study options for optimizing the fuel design and fuel cycle, SCALE6.1 was used to create simplified models of the reactor and look at various parameters. The primary parameters of interest included packing factor and fuel enrichment. An economic analysis was performed on these results by developing a simple fuel cycle cost (FCC) model that could be used to compare the different options from an economic standpoint. The lithium enrichment of the FLiBe coolant was also investigated. The main focus was to understand the practical limitations associated with the Li-7 enrichment and whether it could be used for beneficial purposes. The main idea was to determine whether a lower-than-equilibrium enrichment could be used at reactor start up so that the Li-6 isotope acts as a burnable absorber. The results for the lithium enrichment study showed that the enrichment converges over time, but the amount of time required to reach steady state is much too long and the FLiBe coolant could not be utilized for reactivity control as a burnable absorber. The results found through this research provide reasonable guidelines for expected costs and narrow down the types of configurations that should be considered as fuel design options for the LSCR. Additionally, knowledge was gained on methods for modeling the system not only accurately but also efficiently to reduce the required computing power and time.

The United States' fleet of Light Water Reactors (LWRs) produces a large amount of spent fuel each year; all of which is presently intended to be stored in a fuel repository for disposal. As these LWRs continue to operate and more are built to match the increasing demand for electricity, the required capacity for these repositories grows. Georgia Tech's Subcritical Advanced Burner Reactor (SABR) has been designed to reduce the capacity requirements for these repositories and thereby help close the back end of the nuclear fuel cycle by burning the long-lived transuranics in spent nuclear fuel. SABR's design is based heavily off of the Integral Fast Reactor (IFR). It is important to understand whether the SABR design retains the passive safety characteristics of the IFR. A full safety analysis of SABR's transient response to various possible accident scenarios needs to be performed to determine this. However, before this safety analysis can be performed, it is imperative to model all components of the reactivity feedback mechanism in SABR. The purpose of this work is to develop a calculational model for the fuel bowing reactivity coefficients that can be used in SABR's future safety analysis. This thesis discusses background on fuel bowing and other reactivity coefficients, the history of the IFR, the design of SABR, describes the method that was developed for calculating fuel bowing reactivity coefficients and its validation, and presents an example of a fuel bowing reactivity calculation for SABR.

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.

CIVINS / The United States Navy has recognized the rising age of its nuclear reactors. With this increasing age comes increasing importance of backup generators. In addition to the need for decay heat removal common to all (naval and commercial) nuclear reactors, naval vessels with nuclear reactors also require a backup means of propulsion. All underway Navy nuclear reactors are operated with diesel generators as a backup power system, able to provide emergency electric power for reactor decay heat removal as well as enough electric power to supply an emergency propulsion mechanism. While all commercial nuclear reactors are required to incorporate muhiple backup generators, naval submarine nuclear plants feature a single backup generator. The increasing age of naval nuclear reactors, coupled with the dual reqmrements of a submarine's solitary backup generator, makes the study of submarine backup generators vital. / CIVINS / US Navy (USN) author