Thesis (M.S.)--University of Michigan, 1982.

Thesis (M.S.)--University of Michigan, 1977.

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 a core is further complicated by fuel materials and loading patterns which can be non-symmetric, and by the type of material used as the moderator. The incore thermionic reactor system developed under the Advanced Thermionic Initiative (ATI) is an example of such a reactor. In this design, the fuel is highly enriched uranium dioxide and the moderating medium is zirconium hydride. The traditional bum-up and fuel depletion analysis codes have been found to be inadequate for these calculations, largely for the reasons mentioned above and because the neutron spectra assumed for the codes such as LEOPARD and ORIGEN do not even closely fit that for a small, thermal reactor using ZrH as moderator. A more sophisticated codes such as the transport lattice type code WIMS is suitable for the terrestrial commercial reactors. However it lacks some materials, such as ZrH, needed in special applications and it is not capable of performing calculations with highly enriched fuel. Thus a new method which could accurately calculate the neutron spectrum and the appropriate reaction rates within the Thermionic Fuel Elements (TFE) is needed to be developed. The method developed utilizes and interconnects the accuracy of the Monte Carlo Neutron/Photon (MCNP) method to calculate reaction rates for the important isotopes, and a time dependent depletion routine to calculate the temporal effects on isotope concentrations within the TFEs. This required the modification of the MCNP itself to perform the additional task of accomplishing burn-up calculations. The modified version called, MCNPBURN, evolved to be a general dual purpose code which can be used for standard calculations as well as for burn-up. The of burnable absorber Gadolinium which adds complications both in the physical model and the numerical analysis requires frequent spatial and spectral reevaluations as a function of burn-up. This difficulty is overcome by the application of MCNPBURN by assuming that the burnable poison is uniformly mixed in the fuel. MCNPBURN was benchmarked using a standard Pressurized Water Reactor fuel element against the LEOPARD and WIMS computer codes. The results from MCNPBURN show good agreement with LEOPARD and WIMS. The maximum difference between MCNPBURN and either of the two codes was approximately 9%. The differences can be accounted for by the appropriate treatment of the accumulated fission product. Application of the MCNPBURN for the ATI reactor core, which consists of 165 TFEs and operates at 375 kW of thermal power, showed a system lifetime greater than the projected lifetime of 7 years at full power. The average efficiency is about 5.86% and the change in the overall efficiency over the life time is 0.2%. The percentage of fuel mass burned is estimated to be about 3.6% of the initial mass. Another calculation includes the influence of burnable poisons mixed in the peak pins to flatten the overall core radial power distribution was performed. The efficacy of this change is quite apparent in reducing the power effectively in the peak pins though it may give rise in power elsewhere in the core. / Graduation date: 1994

No description available.

Thesis (Ph.D.)--University of California, Berkeley, 1963. / "UC-80 Reactor Technology" -t.p. "TID-4500 (18th Ed.)" -t.p. Includes bibliographical references (p. 203-207).

A fuel burnup model of a Pressurized Water Reactor (PWR) was developed based on one-group diffusion theory and used simple thermal cross sections. A computer program which simulates the depletion of the core of a PWR was written based on this model. The basic idea was to develop a fuel depletion program which could be readily understood by nuclear engineering students. Thus, accuracy was sacrificed for the sake of simplicity. The model was based upon a typical PWR with three concentric regions in the radial direction of differing fuel enrichment. Each of the regions was homogenized and the concentrations of the isotopes in each region were considered constant over a time interval. The isotopes considered were U-235, Pu-239, U-238, Xe-135, I-135, Sm-149, Pm-149 and the lumped burnable poison isotope. The flux was approximated by the sum of two trigonometric functions. The magnitude and shape of the flux were determined by holding power constant, constraining system to be critical and varying the soluble boron concentration to find the fla~test possible positive flux. A flux magnitude computed in this manner was compared to a similar flux magnitude given in a Final Safety Analysis Report. The concentrations of the isotopes were determined from the differential equations describing the rate of change of the concentrations. The behavior of the isotopes over core life was graphed and wherever possible compared to graphs from other sources. The concentrations calculated for U-235, U-238 and Pu-239 after 450 days were compared to the concentrations of the same isotopes calculated by a zero dimensional three-group model. The percentage difference between the concentrations determined by the two models varied from about 69% for Pu-239 to 1% for U-238. / Master of Science

Typescript. / Digitized by Kansas Correctional Industries

No description available.

Thesis (Ph. D.)--University of Missouri-Columbia, 2006. / The entire dissertation/thesis text is included in the research.pdf file; the official abstract appears in the short.pdf file (which also appears in the research.pdf); a non-technical general description, or public abstract, appears in the public.pdf file. Title from title screen of research.pdf file viewed on (March 2, 2007) Vita. Includes bibliographical references.

In the deep burn research of Very High Temperature Reactor (VHTR), it is desired to make an accurate estimation of absorption cross sections and absorption rates in burnable poison (BP) pins. However, in traditional methods, multi-group cross sections are generated from single bundle calculations with specular reflection boundary condition, in which the energy spectral effect in the core environment is not taken into account. This approximation introduces errors to the absorption cross sections especially for BPs neighboring reflectors and control rods. In order to correct the BP absorption cross sections in whole core diffusion calculations, energy spectrum reconstruction (ESR) methods have been developed to reconstruct the fine group spectrum (and in-core continuous energy spectrum). Then, using the reconstructed spectrum as boundary condition, a BP pin cell local transport calculation serves an imbedded module within the whole core diffusion code to iteratively correct the BP absorption cross sections for improved results. The ESR methods were tested in a 2D prismatic High Temperature Reactor (HTR) problem. The reconstructed fine-group spectra have shown good agreement with the reference spectra. Comparing with the cross sections calculated by single block calculation with specular reflection boundary conditions, the BP absorption cross sections are effectively improved by ESR methods. A preliminary study was also performed to extend the ESR methods to a 2D Pebble Bed Reactor (PBR) problem. The results demonstrate that the ESR can reproduce the energy spectra on the fuel-outer reflector interface accurately.