Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel.
The selection of the material for the blanket first wall (FW) is one of the crucial aspects to overcome in the framework of the EUROfusion program towards DEMO. In this reactor, a loss-ofcoolant accident (LOCA) with simultaneous air ingress into the vacuum vessel would lead to temperatures of the...
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| Format: | info:eu-repo/semantics/doctoralThesis |
| Language: | eng |
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Servicio de publicaciones. Universidad de Navarra
2022
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| Online Access: | https://hdl.handle.net/10171/64528 |
| _version_ | 1793400423124566016 |
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| author | Sal Broco, E. (Elisa) Garcia-Rosales, C. (Carmen) |
| author_facet | Sal Broco, E. (Elisa) Garcia-Rosales, C. (Carmen) |
| author_sort | Sal Broco, E. (Elisa) |
| collection | DSpace |
| description | The selection of the material for the blanket first wall (FW) is one of the crucial aspects to
overcome in the framework of the EUROfusion program towards DEMO. In this reactor, a loss-ofcoolant accident (LOCA) with simultaneous air ingress into the vacuum vessel would lead to
temperatures of the in-vessel components above 1000 °C up to almost 1200 °C due to the decay
heat. Under such a scenario, the use of pure tungsten, which is currently the main candidate,
represents a potential safety risk because of its poor oxidation resistance that may result in a full
oxidation of the armor layer. Such tungsten oxides, which are volatile at the involved temperature
and radioactive after exposure of the FW to the fusion plasma, would be partially released to the
atmosphere.
A possible way to mitigate this important safety issue is the addition of oxide-forming alloying
elements to pure tungsten, since in case of LOCA with simultaneous air ingress, the alloying
elements will diffuse to the surface forming an adherent and stable protective scale, preventing
further tungsten oxidation. Under normal operation conditions, these alloying elements would be
eroded preferentially by sputtering from the fusion plasma, resulting in a pure tungsten layer
exposed to the plasma. Initially, thin films alloys of the systems W-Cr-Si, W-Cr-Ti and W-Cr-Y were developed at the
Max-Planck-Institute for Plasma Physics (IPP) Garching, Germany, by means of magnetron
sputtering. These W-alloys exhibited a reduction of the oxidation rate of several orders of
magnitude compared to pure tungsten due to the formation of a stable Cr2O3 protective scale when
exposed to air at 1000 °C. Although these thin films are not applicable because thicknesses of
several mm are required for the blanket FW of DEMO, they served as model systems for the
manufacturing of bulk materials by powder metallurgy. Previous work have shown the good
oxidation behavior of the W-10Cr-0.5Y alloy manufactured by mechanical alloying (MA) and
subsequent hot isostatic pressing (HIP). Besides, several authors demonstrated the improvement
of the mechanical properties of pure tungsten by the addition of Zr or ZrC.
In the present dissertation, W-Cr-Y(-Zr) alloys manufactured by MA and HIP have been studied
and tested under relevant DEMO operating conditions. After HIPing, the alloys exhibited a twophase microstructure since the W-Cr system has a miscibility gap below 1680 °C. The application
of a heat treatment at a temperature above the miscibility gap enabled to obtain a dense material
formed by a single metastable phase, which remained stable for long periods of time under the
maximum expected operating temperature. The addition of Zr to the W-Cr-Y system resulted in
an improvement of thermal shock resistance while maintaining similar oxidation resistance and
thermo-mechanical properties. The heat-treated W-10Cr-0.5Y alloy has been exposed to 0.19 to
0.26 dpa neutron irradiation at 600, 800 and 1000 °C, after which an increase of fracture strength
was recorded under all irradiation conditions. This increase was especially relevant after
irradiation at 1000°C, where a value of 1.7 GPa was achieved, being a factor of about 3 higher than
the one of the non-irradiated material. This significant increase in strength after 1000 °C irradiation
is thought to be mainly associated to the solid-solution decomposition of the initial metastable
single phase, resulting in an ultrafine-grained, vermicular shaped, two-phase microstructure. The
as-HIPed W-10Cr-0.5Y alloy has been also exposed to W ion irradiation for producing damage and
subsequently deuterium implanted at temperatures up to 250 °C. The amount of retained
deuterium after implantation at 250 °C was 1/3 lower than that of pure tungsten, being the presence
of Cr responsible for this reduction. Regarding technological aspects, joints between W-10Cr-0.5Yalloy and P91 steel have been produced using diffusion bonding by HIP, resulting in a high shear
strength of 354 MPa, comparable to the one obtained by brazing this alloy to Eurofer. These
strength values are among the highest found in the literature for joining pure tungsten to steel. |
| format | info:eu-repo/semantics/doctoralThesis |
| id | oai:dadun.unav.edu:10171-64528 |
| institution | Universidad de Navarra |
| language | eng |
| publishDate | 2022 |
| publisher | Servicio de publicaciones. Universidad de Navarra |
| record_format | dspace |
| spelling | oai:dadun.unav.edu:10171-645282022-10-21T01:04:33Z Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. Sal Broco, E. (Elisa) Garcia-Rosales, C. (Carmen) Nuclear fusion energy DEMO reactor Self-passivating alloys Tungsten Oxidation Joining feasibility Hydrogen retention Neutron irradiation The selection of the material for the blanket first wall (FW) is one of the crucial aspects to overcome in the framework of the EUROfusion program towards DEMO. In this reactor, a loss-ofcoolant accident (LOCA) with simultaneous air ingress into the vacuum vessel would lead to temperatures of the in-vessel components above 1000 °C up to almost 1200 °C due to the decay heat. Under such a scenario, the use of pure tungsten, which is currently the main candidate, represents a potential safety risk because of its poor oxidation resistance that may result in a full oxidation of the armor layer. Such tungsten oxides, which are volatile at the involved temperature and radioactive after exposure of the FW to the fusion plasma, would be partially released to the atmosphere. A possible way to mitigate this important safety issue is the addition of oxide-forming alloying elements to pure tungsten, since in case of LOCA with simultaneous air ingress, the alloying elements will diffuse to the surface forming an adherent and stable protective scale, preventing further tungsten oxidation. Under normal operation conditions, these alloying elements would be eroded preferentially by sputtering from the fusion plasma, resulting in a pure tungsten layer exposed to the plasma. Initially, thin films alloys of the systems W-Cr-Si, W-Cr-Ti and W-Cr-Y were developed at the Max-Planck-Institute for Plasma Physics (IPP) Garching, Germany, by means of magnetron sputtering. These W-alloys exhibited a reduction of the oxidation rate of several orders of magnitude compared to pure tungsten due to the formation of a stable Cr2O3 protective scale when exposed to air at 1000 °C. Although these thin films are not applicable because thicknesses of several mm are required for the blanket FW of DEMO, they served as model systems for the manufacturing of bulk materials by powder metallurgy. Previous work have shown the good oxidation behavior of the W-10Cr-0.5Y alloy manufactured by mechanical alloying (MA) and subsequent hot isostatic pressing (HIP). Besides, several authors demonstrated the improvement of the mechanical properties of pure tungsten by the addition of Zr or ZrC. In the present dissertation, W-Cr-Y(-Zr) alloys manufactured by MA and HIP have been studied and tested under relevant DEMO operating conditions. After HIPing, the alloys exhibited a twophase microstructure since the W-Cr system has a miscibility gap below 1680 °C. The application of a heat treatment at a temperature above the miscibility gap enabled to obtain a dense material formed by a single metastable phase, which remained stable for long periods of time under the maximum expected operating temperature. The addition of Zr to the W-Cr-Y system resulted in an improvement of thermal shock resistance while maintaining similar oxidation resistance and thermo-mechanical properties. The heat-treated W-10Cr-0.5Y alloy has been exposed to 0.19 to 0.26 dpa neutron irradiation at 600, 800 and 1000 °C, after which an increase of fracture strength was recorded under all irradiation conditions. This increase was especially relevant after irradiation at 1000°C, where a value of 1.7 GPa was achieved, being a factor of about 3 higher than the one of the non-irradiated material. This significant increase in strength after 1000 °C irradiation is thought to be mainly associated to the solid-solution decomposition of the initial metastable single phase, resulting in an ultrafine-grained, vermicular shaped, two-phase microstructure. The as-HIPed W-10Cr-0.5Y alloy has been also exposed to W ion irradiation for producing damage and subsequently deuterium implanted at temperatures up to 250 °C. The amount of retained deuterium after implantation at 250 °C was 1/3 lower than that of pure tungsten, being the presence of Cr responsible for this reduction. Regarding technological aspects, joints between W-10Cr-0.5Yalloy and P91 steel have been produced using diffusion bonding by HIP, resulting in a high shear strength of 354 MPa, comparable to the one obtained by brazing this alloy to Eurofer. These strength values are among the highest found in the literature for joining pure tungsten to steel. La selección del material para la primera pared del blanket es uno de los mayores retos en el marco de EUROfusion de cara a la construcción del futuro reactor de fusión DEMO. En este reactor, un accidente de pérdida de refrigerante (LOCA) con entrada de aire en la vasija de vacío daría lugar a un incremento de la temperatura en los componentes de la vasija, alcanzándose temperaturas superiores a 1000 °C y hasta casi 1200 °C debido al calor de desintegración. En esta situación, el uso de wolframio puro –a día de hoy material candidato para la primera pared de DEMO- representa un importante riesgo de seguridad debido a su baja resistencia a la oxidación, ya que daría lugar a la liberación de óxidos volátiles y radiactivos a la atmósfera. Una posible solución para mitigar este problema es la adición de elementos aleantes que en presencia de oxígeno a altas temperaturas difundan hacia la superficie de manera que formen una capa adherente y estable, impidiendo la oxidación del wolframio. En condiciones de operación normal, estos elementos aleantes serán erosionados preferentemente por sputtering por partículas procedentes del plasma de fusión, dando lugar a una capa de wolframio puro expuesta al plasma. Inicialmente, estas aleaciones se desarrollaron en forma de películas delgadas de los sistemas W-Cr-Si, W-Cr-Ti y W-Cr-Y en el Instituto Max-Planck de Física del Plasma (IPP), Garching, Alemania, mediante sputtering de magnetrón. Estas aleaciones de W mostraron una reducción de la tasa de oxidación de varios órdenes de magnitud en comparación con el wolframio puro, debido a la formación de una capa protectora estable de Cr2O3 cuando se exponían al aire a 1000 °C. A pesar de que estas películas delgadas no se pueden emplear para la primera pared del blanket de DEMO debido a que se requieren espesores de varios mm, sirvieron como modelo para la fabricación de material denso por la ruta pulvimetalúrgica. Trabajos anteriores han demostrado el buen comportamiento a la oxidación de la aleación W-10Cr-0.5Y fabricada mediante aleación mecánica (MA) y posterior prensado isostático en caliente (HIP). Además, varios autores han demostrado la mejora de las propiedades mecánicas del wolframio puro mediante la adición de Zr o ZrC. En la presente tesis, se han estudiado las aleaciones de W-Cr-Y(-Zr) fabricadas por MA y HIP y se han sometido a las condiciones de operación esperadas en DEMO. Tras HIP, las aleaciones mostraron una microestructura bifásica, ya que el sistema W-Cr tiene un gap de miscibilidad por debajo de 1680 °C. La aplicación de un tratamiento térmico a una temperatura superior a dicho gap permitió obtener un material denso formado por una única fase metaestable, que se mantuvo estable durante largos periodos de tiempo a la temperatura máxima de funcionamiento prevista. La adición de Zr al sistema W-Cr-Y dio lugar a una mejora de la resistencia al choque térmico, manteniendo una resistencia a la oxidación y unas propiedades termomecánicas similares. La aleación W-10Cr-0.5Y tratada térmicamente se expuso a una irradiación de neutrones de 0,19 a 0,26 dpa a temperaturas de 600, 800 y 1000 °C, tras lo cual se registró un aumento de la resistencia a la fractura en todas las condiciones de irradiación. Este aumento fue especialmente relevante tras la irradiación a 1000 °C, donde se alcanzó un valor de 1,7 GPa, siendo aproximadamente 3 veces superior al del material no irradiado. Se cree que este importante aumento de la resistencia tras la irradiación a 1000 °C está asociado principalmente a la descomposición en solución sólida de la fase única metaestable inicial, dando lugar a una microestructura bifásica de grano ultrafino y forma vermicular. La aleación W-10Cr-0.5Y tras HIP también ha sido expuesta a la irradiación de iones de W para producir daño y posteriormente ha sido implantada con deuterio a temperaturas de hasta 250 °C. La cantidad de deuterio retenido tras la implantación a 250 °C fue 1/3 inferior a la del wolframio puro, siendo la presencia de Cr la responsable de esta reducción. En cuanto a los aspectos tecnológicos, se han producido uniones entre la aleación W-10Cr-0.5Y y el acero P91 mediante unión por difusión vía HIP, dando como resultado una elevada resistencia al cizallamiento de 354 MPa, comparable a la obtenida mediante la soldadura de esta aleación con Eurofer. Estos valores de resistencia se encuentran entre los más altos encontrados en la literatura para la unión de wolframio puro con acero. 2022-10-20T06:39:57Z 2022-10-20T06:39:57Z 2022-07-12 2022-07-01 info:eu-repo/semantics/doctoralThesis https://hdl.handle.net/10171/64528 eng info:eu-repo/semantics/openAccess application/pdf Servicio de publicaciones. Universidad de Navarra |
| spellingShingle | Nuclear fusion energy DEMO reactor Self-passivating alloys Tungsten Oxidation Joining feasibility Hydrogen retention Neutron irradiation Sal Broco, E. (Elisa) Garcia-Rosales, C. (Carmen) Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title | Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title_full | Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title_fullStr | Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title_full_unstemmed | Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title_short | Self-passivating W-Cr-Y alloys as plasma-facing materials: Effect of Zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| title_sort | self-passivating w-cr-y alloys as plasma-facing materials: effect of zr addition on fusion relevant properties; study of hydrogen retention, neutron irradiation and joining feasibility to steel. |
| topic | Nuclear fusion energy DEMO reactor Self-passivating alloys Tungsten Oxidation Joining feasibility Hydrogen retention Neutron irradiation |
| url | https://hdl.handle.net/10171/64528 |
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