Thermal hydraulic analysis of hydride fuels in BWR's

Abstract

This thesis contributes to the hydride nuclear fuel project being completed by UC Berkeley and MIT to assess the possible benefits of using hydride fuel in light water nuclear reactors (LWR's). More specifically, this thesis deals with the thermal hydraulic analysis of BWR reactors. Several papers and theses have already been written for this project, mainly focusing on PWR reactors. The primary goal of this thesis is to find the optimal fuel rod lattice pitch and diameter such that a reactor can safely operate at the highest possible power. This fuel geometry is found out of hundreds of possible choices by using a script to automate a parametric study. A similar process was completed by an MIT graduate student for PWR reactors. While this thesis demonstrates the ability to use such a method for thermal hydraulic BWR analysis, there are some shortcomings which are mainly due to the difficulty of obtaining proprietary information about BWR nuclear reactors. All results hold equally for uranium dioxide as well as hydride fuel since the design limits imposed, critical heat flux, maximum flow velocity and pressure drop constrain only pin array geometry and diameter. It is shown that applicable uranium oxide and hydride fuel limits are both met within the constraints imposed by these three limits which were applied.

(cont.) The final analysis of this report shows a possible reactor power improvement of order 30% but this is based on several analysis selections which introduce error and/or a degree of unrealism into the analysis. First the EPRI critical heat flux correlation was used versus a more appropriate critical power correlation Second the expedient of using a fixed mass flux was adopted which caused the hot channel exit quality to change with power changes. This was done since the means to keep the ratio of reactor power to mass flow rate constant which would have maintained constant exit quality over the geometry map explored by scripting could not be developed in the time available for this work.., Hence definite conclusions on achievable BWR core power over the range of geometries investigated are not available and hence warrant further investigation. More importantly the accomplishment of this thesis is the demonstration that the scripted methodology described in this paper can be used to assess thousands of different reactor parameters in order to optimize reactor power.