Thesis (Ph. D.)--Oregon State University, 2009. / Printout. Includes bibliographical references (leaves 331-360). Also available on the World Wide Web.

Increased conversion ratios and burnup can be achieved by mechanically changing the fuel-to-water volume ratio of a reactor over the core lifetime. As the fuel-to-water ratio decreases, the neutron spectrum softens, thereby increasing core reactivity. Proposed mechanical spectral shift reactors utilize this concept. RFD-1, a 1-dimensional, 4-group code was developed to compute fuel burnup cycles for spectral shift reactors. The code calculates burnup for a triangular core lattice having a beginning fuel to water ratio as high as 1.30. Core shutdown occurs at a fuel to water ratio of 0.50. The microscopic cross sections were obtained through use of the VIM code and tabulated for use in RFD-1 as a function of fuel to water ratio and burnup time. The fission product group cross sections were developed using the VIM and TOAFEW codes. The flexibility of RFD-1 allows the user to study a wide variety of possible core configurations. Results of RFD-1 show that increased conversion and burnup, using lower initial enrichments than that of standard Pressurized Water Reactors, result for mechanical spectral shift designs. The next step is to study specific spectral shift designs in greater detail. The RFD-1 code could be improved primarily through refinements in its cross section data tables. / Master of Science

Using the fuel burnup as objective function, it is desired to determine the initial distribution of the fuel in a reactor in order to obtain the maximum energy possible, for which, without changing a fixed initial fuel mass, the results for different initial fuel and control poison configurations are analyzed and the corresponding running times compared. One-dimensional, two energy-group theory is applied to a reflected cylindrical reactor using U-235 as fuel and light water as moderator and reflector. Fissions in both fast and thermal groups are considered. The reactor is divided into several annular regions, and the constant flux approximation in each depletion step is then used to solve the fuel and fission-product poisons differential equations in each region. The computer code OPTIME was developed to determine the time variation of core properties during the fuel cycle. At each depletion step, OPTIME calls ODMUG, [12] a criticality search program, from which the spatially-averaged neutron fluxes and control poison cross sections are obtained. A uniform initial fuel distribution was chosen as a benchmark and the results for several different fuel configurations were analyzed. Two different initial control poison distributions were investigated for each fuel configuration: a uniform and a fuel proportional distribution. Using an iterative approach fuel was moved from the low burnup regions toward the high burnup regions; reactor running times were in this way increased from 9000 to 11,500 hours in the fuel proportional control poison distribution case and from 9000 to 11,000 hours in the uniform control poison distribution case. Beyond this point not only did the running time not increase, but no criticality was reached. / Ph. D.

Improving computer technology and the desire to more accurately model the heterogeneity of the nuclear reactor environment have made the use of Monte Carlo depletion codes more attractive in recent years, and feasible (if not practical) even for 3-D depletion simulation. However, in this case statistical uncertainty is combined with error propagating through the calculation from previous steps. In an effort to understand this error propagation, four test problems were developed to test error propagation in the fuel assembly and core domains. Three test cases modeled and tracked individual fuel pins in four 17x17 PWR fuel assemblies. A fourth problem modeled a well-characterized 330MWe nuclear reactor core. By changing the code's initial random number seed, the data produced by a series of 19 replica runs of each test case was used to investigate the true and apparent variance in k-eff, pin powers, and number densities of several isotopes. While this study does not intend to develop a predictive model for error propagation, it is hoped that its results can help to identify some common regularities in the behavior of uncertainty in several key parameters.

Nuclear nonproliferation research and forensics have a need for improved software solutions, particularly in the estimates of the transmutation of nuclear fuel during burnup and depletion. At the same time, parallel computers have become effectively sized to enable full core simulations using highly-detailed 3d mesh models. In this work, the capability for modeling 3d reactor models is researched with PENBURN, a burnup/depletion code that couples to the PENTRAN Parallel Sn Transport Solver and also to the Monte Carlo solver MCNP5 using the multigroup option. This research is computationally focused, but will also compare a subset of results of experimental Pressurized Water Reactor (PWR) burnup spectroscopy data available with a designated BR3 PWR burnup benchmark. Also, this research will analyze large-scale Cartesian mesh models that can be feasibly modeled for 3d burnup, as well as investigate the improvement of finite differencing schemes used in parallel discrete ordinates transport with PENTRAN, in order to optimize runtimes for full core transport simulation, and provide comparative results with Monte Carlo simulations. Also, the research will consider improvements to software that will be parallelized, further improving large model simulation using hybrid OpenMP-MPI. The core simulations that form the basis of this research, utilizing discrete ordinates methods and Monte Carlo methods to drive time and space dependent isotopic reactor production using the PENBURN code, will provide more accurate detail of fuel compositions that can benefit nuclear safety, fuel management, non-proliferation, and safeguards applications.