Deciphering the role of dispersoids in dispersion-strengthened tungsten alloys on the mechanical properties and irradiation- driven response under nuclear fusion reactor-relevant conditions

Lang, Eric Joseph

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Description

Title

Deciphering the role of dispersoids in dispersion-strengthened tungsten alloys on the mechanical properties and irradiation- driven response under nuclear fusion reactor-relevant conditions

Author(s)

Lang, Eric Joseph

Issue Date

2020-09-21

Director of Research (if dissertation) or Advisor (if thesis)

Allain, Jean Paul

Doctoral Committee Chair(s)

Ruzic, David

Committee Member(s)

• Krogstad, Jessica
• Zhang, Yang
• Stubbins, James

Department of Study

Nuclear, Plasma, & Rad Engr

Discipline

Nuclear, Plasma, Radiolgc Engr

Degree Granting Institution

University of Illinois at Urbana-Champaign

Degree Name

Ph.D.

Degree Level

Dissertation

Keyword(s)

• Tungsten
• fusion

Abstract

Tungsten is the material of choice for plasma-facing components in future plasma- burning fusion reactors because of its high melting point, high sputter threshold, and low hydrogenic species retention. However, tungsten is an intrinsically brittle material, displaying no room temperature ductility and only exhibiting non-brittle failure at temperatures at high temperatures. In addition to its limited ductility, tungsten’s high melting point and low recrystallization temperature pose complications during fabrication and limit its temperature operating window in a future fusion reactor. Traditional synthesis routes tend to result in non-fully dense samples with coarse-grained microstructures. As a consequence, there is a desire for a fine-grained, fully-dense tungsten material that exhibits enhanced ductility. Tungsten is embrittled by impurity oxygen atoms residing at grain boundaries. It is theorized that by microalloying tungsten with transition metal carbides that capture the oxygen atoms, the impurity distribution can be altered to beneficially impact the mechanical properties. Additionally, altering the interface concentration and type can increase the irradiation tolerance of tungsten. Through the advent of advanced powder processing techniques such as spark plasma sintering, dense, fine-grained tungsten samples can be developed with these microalloyed microstructures. Spark plasma sintering is a powder compaction technique that provides high pressure and heating rates, allowing for a lower final temperature and hold time for compaction. In this work, spark plasma sintering is employed to develop tungsten materials alloyed with tantalum carbide, titanium carbide, or zirconium carbide, subsequently referred to as dispersion-strengthened tungsten. Samples are fabricated with varying compositions of added carbides (from 0.5-10 wt.%), and the sintering process results in >90% dense samples with grains <10 μm in size. Control of the microstructure and impurity distribution within the matrix is evident through the second phase addition, showing this technique can be used for advanced tungsten synthesis. The behavior of these dispersion-strengthened tungsten materials under reactor-like thermal and irradiation conditions is investigated to understand the plasma-material interaction properties and near-surface mechanical response of these materials to fusion-relevant irradiation conditions. Under high temperature exposure, the enhanced recrystallization inhibition of dispersion-strengthened tungsten materials is shown up to 1800oC and the long-term annealing properties exhibit increased incubation time for recovery, as a high number of tungsten-dispersoid interfaces is presented to pin grain boundaries and enhance mechanical properties. However, deuterium irradiation results in surface blistering, while the dispersoid composition and distribution in the bulk samples has been shown to increase deuterium transport and retention. Multiscale helium irradiations show no enhanced surface sputtering compared to pure tungsten nor preferential sputtering, indicating surface compositional stability for stable properties under irradiation. Diminished surface nanostructuring under helium irradiation due to altered helium trapping and decreased bubble formation within the microstructure is demonstrated. The altered helium bubble dynamics is attributed to the dispersoid composition and the W-dispersoid interface chemistry, which help mitigate the detrimental effects of helium on the tungsten microstructure. This work shows the promise of dispersion-strengthened tungsten materials to prevent the detrimental effects of helium irradiation of pure tungsten, while also preventing recrystallization of tungsten under high temperature exposure. Compared to pure tungsten in a fusion environment, these results indicate a wider temperature operating window and show greater helium irradiation resistance, demonstrating enhanced thermo-mechanical and irradiation properties that can benefit future fusion reactor plasma-facing materials development.

Graduation Semester

2020-12

Type of Resource

Thesis

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Copyright 2020 Eric Lang

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