{"id":{"repo_id":"uiuc","oai_identifier":"oai:www.ideals.illinois.edu:2142/109476"},"canonical_url":"https://search.dev.ndltd.org/etd/uiuc/oai:www.ideals.illinois.edu:2142/109476","repository":{"repo_id":"uiuc","name":"University of Illinois - Urbana-Champaign","base_url":"https://www.ideals.illinois.edu/oai-pmh"},"display":{"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","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.","abstract_html":"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 &gt;90% dense samples with grains &lt;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.","abstract_has_math":false,"creators":["Lang, Eric Joseph"],"institution":"University of Illinois at Urbana-Champaign","degree_name":"Ph.D.","degree_level":"Dissertation","degree_discipline":"Nuclear, Plasma, Radiolgc Engr","degree_department":null,"school":null,"contributors":["Allain, Jean Paul","Ruzic, David","Krogstad, Jessica","Zhang, Yang","Stubbins, James"],"advisors":[],"committee_chairs":[],"committee_members":[],"year":2021,"date_issued":"2021-03-05T21:40:35Z","date_published":"2021-03-05T21:40:35Z","updated_at":"2026-07-22T22:24:50Z","subjects":["Tungsten","fusion"],"languages":["en"],"rights":["Copyright 2020 Eric Lang"],"rights_urls":[],"identifier_entries":[]},"links":{"outbound_url":"http://hdl.handle.net/2142/109476","outbound_label":"Handle","outbound_source":"dc:identifier"},"metadata_groups":[{"id":"people","label":"People","entries":[{"key":"dc:contributor","label":"Contributor","values":["Allain, Jean Paul","Ruzic, David","Krogstad, Jessica","Zhang, Yang","Stubbins, James"]},{"key":"dc:creator","label":"Author","values":["Lang, Eric Joseph"]}]},{"id":"academic_context","label":"Academic Context","entries":[{"key":"dc:date","label":"Dc Date","values":["2021-03-05T21:40:35Z","2023-03-05T21:43:00Z","2020-09-21","2020-12"]},{"key":"dc:type","label":"Dc Type","values":["text","Thesis"]},{"key":"thesis:degree_discipline","label":"Discipline","values":["Nuclear, Plasma, Radiolgc Engr"]},{"key":"thesis:degree_level","label":"Degree Level","values":["Dissertation"]},{"key":"thesis:degree_name","label":"Degree Name","values":["Ph.D."]},{"key":"thesis:institution_name","label":"Thesis Institution Name","values":["University of Illinois at Urbana-Champaign"]}]},{"id":"subjects_keywords","label":"Subjects and Keywords","entries":[{"key":"dc:subject","label":"Dc Subject","values":["Tungsten","fusion"]}]},{"id":"language_rights","label":"Language and Rights","entries":[{"key":"dc:language","label":"Dc Language","values":["en"]},{"key":"dc:rights","label":"Dc Rights","values":["Copyright 2020 Eric Lang"]}]},{"id":"identifiers","label":"Identifiers","entries":[{"key":"dc:identifier","label":"Identifier","values":["http://hdl.handle.net/2142/109476"]}]},{"id":"additional","label":"Additional Metadata","entries":[{"key":"dc:description","label":"Description","values":["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.","Submission published under a 24 month embargo labeled 'U of I Access', the embargo will last until 2022-12-01","The student, Eric Lang, accepted the attached license on 2020-09-16 at 22:07.","The student, Eric Lang, submitted this Dissertation for approval on 2020-09-16 at 22:15.","This Dissertation was approved for publication on 2020-09-21 at 13:39.","DSpace SAF Submission Ingestion Package generated from Vireo submission #15813 on 2021-03-04 at 16:18:50","Made available in DSpace on 2021-03-05T21:40:35Z (GMT). No. of bitstreams: 2 LANG-DISSERTATION-2020.pdf: 15145127 bytes, checksum: 1c1282c49d89273fffca4d5b650fa6ef (MD5) LICENSE.txt: 4206 bytes, checksum: 54e900cb82ffc4898f9f2f4b06e4a5c7 (MD5) Previous issue date: 2020-09-21","Embargo set by: Seth Robbins for item 117180 Lift date: 2023-03-05T21:40:52Z Reason: Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","Embargo set by: Seth Robbins for item 117180 Lift date: 2023-03-05T21:43:00Z Reason: Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","U of I Only"]},{"key":"dc:format","label":"Dc Format","values":["application/pdf"]},{"key":"dc:title","label":"Title","values":["Deciphering the role of dispersoids in dispersion-strengthened tungsten alloys on the mechanical properties and irradiation- driven response under nuclear fusion reactor-relevant conditions"]}]}],"canonical_facts":{"dc:contributor":["Allain, Jean Paul","Ruzic, David","Krogstad, Jessica","Zhang, Yang","Stubbins, James"],"dc:creator":["Lang, Eric Joseph"],"dc:date":["2021-03-05T21:40:35Z","2023-03-05T21:43:00Z","2020-09-21","2020-12"],"dc:description":["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.","Submission published under a 24 month embargo labeled 'U of I Access', the embargo will last until 2022-12-01","The student, Eric Lang, accepted the attached license on 2020-09-16 at 22:07.","The student, Eric Lang, submitted this Dissertation for approval on 2020-09-16 at 22:15.","This Dissertation was approved for publication on 2020-09-21 at 13:39.","DSpace SAF Submission Ingestion Package generated from Vireo submission #15813 on 2021-03-04 at 16:18:50","Made available in DSpace on 2021-03-05T21:40:35Z (GMT). No. of bitstreams: 2 LANG-DISSERTATION-2020.pdf: 15145127 bytes, checksum: 1c1282c49d89273fffca4d5b650fa6ef (MD5) LICENSE.txt: 4206 bytes, checksum: 54e900cb82ffc4898f9f2f4b06e4a5c7 (MD5) Previous issue date: 2020-09-21","Embargo set by: Seth Robbins for item 117180 Lift date: 2023-03-05T21:40:52Z Reason: Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","Embargo set by: Seth Robbins for item 117180 Lift date: 2023-03-05T21:43:00Z Reason: Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","Author requested U of Illinois access only (OA after 2yrs) in Vireo ETD system","U of I Only"],"dc:format":["application/pdf"],"dc:identifier":["http://hdl.handle.net/2142/109476"],"dc:language":["en"],"dc:rights":["Copyright 2020 Eric Lang"],"dc:subject":["Tungsten","fusion"],"dc: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"],"dc:type":["text","Thesis"],"thesis:degree_discipline":["Nuclear, Plasma, Radiolgc Engr"],"thesis:degree_level":["Dissertation"],"thesis:degree_name":["Ph.D."],"thesis:institution_name":["University of Illinois at Urbana-Champaign"]},"updated_at":"2026-07-22T22:24:50Z"}