Masters Theses

Abstract

"Renewed interest in nuclear safety has resulted in development of accident-tolerant materials, such as cladding. Although developed for light water reactors, their use in more advanced reactors would save time and development costs for other new materials. This paper performed a Multiphysics investigation into the power profile, thermodynamics, heat transfer, and tensor mechanics of four candidate materials (Haynes 230, Zircaloy-4, FeCrAl, and SiC-SiC), using the highly benchmarked Kilo power design as an example of a high-temperature, highly enriched, sodium-cooled microreactor environment.

From neutronic code MCNP, no significant difference in power profile was found among the four. Using finite element software framework MOOSE, temperature convergence and pipe rupture analysis using hoop stress were performed. SiC-SiC displayed the most resistance to temperature change, although more noteworthy was Haynes 230 (the default material in Kilo power), the best performing candidate during stress analysis. The high temperatures had the most impact on these results, greatly affecting Zircaloy-4 and FeCrAl’s material limits, while SiC-SiC’s inelastic, ceramic nature was non-conducive to the stress loads it experienced.

Also investigated was the possibility of variance in the material properties. The impact of Young’s Modulus and Poisson’s Ratio on maximum hoop stress were measured with uncertainty quantification and sensitivity analysis using polynomial chaos expansion and Sobol’ indices, respectively. No significant sensitivities were found, although Zircaloy- 4 was close to performing under stress limits, and further investigation into its properties is recommended.

A recommendation is put forth for more research into the high temperature properties of further existing materials for a more robust conclusion on their applicability to advanced reactors, although Haynes 230 is recommended as the current best in the tested areas" -- Abstract, p. iii

Advisor(s)

Alajo, Ayodeji Babatund

Committee Member(s)

Alam, Syed B.
Kumar, Dinesh

Department(s)

Nuclear Engineering and Radiation Science

Degree Name

M.S. in Nuclear Engineering

Publisher

Missouri University of Science and Technology

Publication Date

Summer 2024

Pagination

xii, 75 pages

Note about bibliography

Includes_bibliographical_references_(pages 71-73)

Rights

©2024 Alexander Niles Foutch , All Rights Reserved

Document Type

Thesis - Open Access

File Type

text

Language

English

Thesis Number

T 12383

Electronic OCLC #

1460027143

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