Masters Theses

Date of Award

12-2020

Degree Type

Thesis

Degree Name

Master of Science

Major

Nuclear Engineering

Major Professor

Nicholas R. Brown

Abstract

Modular High Temperature Gas-Cooled Reactors (mHTGRs) have numerous applications other than electricity production and exhibit strong safety characteristics, but have challenges associated with the design. Graphite, primarily used as moderator material within an mHTGR, exhibits unfavorable material changes under irradiation and contributes to the low power density of an mHTGR. These issues have led to the investigation and development of alternative moderators to be utilized in mHTGRs, including beryllium- and hydride-based concepts with compositions selected for favorable moderating power and the potential for an improved in-service lifetime as compared to graphite.

The proposed moderators are fabricated as two-phase composites with magnesium oxide, MgO, as the radiation-stable host matrix and beryllium metal, Be, beryllium oxide, BeO, or zirconium hydride, ZrHx=1 as the entrained moderating phase. Here, the reactor performance and safety characteristics of these moderator concepts are evaluated relative to a graphite reference using a Ft. Saint Vrain-style fuel block. The cycle length, discharge burnup, natural resource utilization, neutron flux spectra, moderating power, moderating ratio, critical size, moderator and fuel temperature feedback, fuel cycle cost, spent nuclear fuel and high-level waste radioactivity per unit energy generated, and environmental impact per unit energy generated are assessed. These advanced moderators are also assessed from a reactor safety standpoint for Design Basis Accidents (DBAs) including Pressurized Loss of Forced Cooling and Depressurized Loss of Forced Cooling accidents for a 350-megawatt thermal prismatic-type mHTGR. Preliminary calculations of microcore neutronics and safety performance is demonstrated to illustrate a framework for designing a microcore.

The results demonstrate that the advanced moderators have the potential for comparable or enhanced cycle performance to that of the graphite reference case with significantly improved performance for an optimized moderator-to-fuel ratio design. The full core thermal-hydraulic analysis of DBAs shows that the high volumetric heat capacity of the beryllium-based moderator grants them a greater margin to fuel failure in these analyses than a conventional graphite-moderated system, but the lower thermal conductivity of the beryllium-based moderators leads to longer times at elevated temperatures. Microcore neutronics calculations show the effectiveness that the advanced moderators have on cycle length and safety evaluation of the design shows comparable performance to that of graphite.

Download
Files over 3MB may be slow to open. For best results, right-click and select "save as..."

Share

COinS