Compacted graphite iron is the material of choice for engine cylinder heads of heavy-duty trucks. Compacted graphite iron provides the best possible compromise between optimum mechanical properties, compared to flake graphite iron, and optimum thermal conductivity, compared to spheroidal graphite iron. The vermicular-shaped graphite particles, however, act as stress concentrators, and, as a result of delamination from the metal matrix, they are responsible for crack initiation during the thermomechanical fatigue cycles occurring through engine startup and shutdown cycles. Scratch tests driven over the matrix and into the graphite particles were performed in order to characterize the strength of the metal-graphite interface. Samples extracted from a cylinder head in as-cast condition were compared to samples subjected to a heat-treatment at 700 °C for 60 h. The former samples were composed of a primarily pearlitic matrix and graphite particles (~11.5 vol %), whereas, after annealing, a certain pearlite fraction decomposed into Fe and C, producing a microstructure with graphite-ferrite interfaces, exhibiting a partially spiky morphology. The scratch test revealed that the ferrite-graphite interfaces with spiky nature exhibited a stronger resistance to delamination compared to the ferrite-graphite interfaces with smooth morphology. One reason for the high interface strength is the mechanical interlocking between graphite spikes and ferrite, increasing the contact area between the two phases.@en

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.@en

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.@en

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.@en

Six tungsten grades were irradiated in the Belgian material test reactor (BR2) and characterized by Vickers hardness tests in order to investigate the irradiation-induced hardening. These tungsten grades included: Plansee (Austria) ITER specification tungsten, ALMT (Japan) ITER specification tungsten, two products from KIT (Germany) produced by powder injection molding (PIM) and strengthened by 1% TiC and 2% Y2O3 dispersed particles, and rolled tungsten strengthened by 0.5% ZrC from ISSP (China). The materials were irradiated face-to-face at three temperatures equal to 600 °C, 1000 °C, and 1200 °C to the dose of ∼1 dpa. The Vickers hardness tests under 200 gf (HV0.2) were performed at room temperature. The Vickers hardness increases as the irradiation temperature increases from 600 to 1000 °C for all materials, except for the ZrC-reinforced tungsten, for which the increase of hardness does not depend on irradiation temperature. The irradiation-induced hardness decreases after irradiation at 1200 °C. This is a result of defect annealing enhanced by thermally activated diffusion. However, even at 1200 °C, the impact of neutron irradiation on the hardness increase remains significant; the hardness increases by ∼30 to 60% compared to the non-irradiated value. In the case of TiC-strengthened material, the irradiation hardening progressively raises with irradiation temperature, which cannot be explained by the accumulation of neutron irradiation defects solely.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

A large campaign of characterization of the ductile to brittle transition temperature (DBTT) and microstructure has been performed on several commercial and lab-scale pure tungsten grades, potassium doped tungsten alloys, and particle reinforced tungsten grades (with particles of TiC, Y2O3, or ZrC), all integrated in a large-scale neutron irradiation campaign. The DBTT is deduced based on miniaturized three-point bending tests to provide reference data for the assessment of the irradiation effects on the tungsten alloys. This miniaturized geometry is designed to minimize the operational cost of neutron irradiation, to speed up post-irradiation examination, and to reduce the amount of nuclear waste. The resulting DBTT ranges from around −15 up to 450 °C, depending on the material. The potassium doped tungsten alloys have the lowest DBTT, followed by rolled ZrC reinforced tungsten grade, commercial pure tungsten grades, lab-scale pure tungsten grades, and other particle reinforced tungsten grades. The crack plane orientation and microstructure with respect to grain shape and grain boundaries significantly affect the DBTT for forged/rolled tungsten products with elongated grains. The L-T orientation has a lower DBTT compared to the T-L orientation. Moreover, the DBTT difference in the L-T and T-L orientation raises with increasing the grain aspect ratio. An attempt is made to establish a relationship between the density of low and high angle grain boundaries and DBTT value. The obtained relationship is discussed in the frame of mechanical processing (i.e., rolling or forging) to optimize the DBTT by optimized manufacturing. The results are compared to recent computational predictions of the DBTT in tungsten.@en

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten. The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method. The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.@en

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten. The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method. The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.@en

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten. The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method. The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.@en

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten. The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method. The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.@en

The strength and ductility of two tungsten products, developed for application in nuclear fusion environment, are studied before and after neutron irradiation using uniaxial tensile tests. The first product is a commercially pure tungsten, produced by AT&M company according to ITER specification, and the second one is reinforced with zirconium carbide (W–ZrC) particles. The addition of ZrC particles leads to a reduction of the ductile to brittle transition temperature (DBTT) in non-irradiated conditions down to 50–100 °C without loss of strength and of other attractive properties of tungsten. The neutron irradiation was performed in the range 625–700 °C up to 1.125 dpa. The tests were performed to screen the shift of the DBTT as well as to characterize the evolution of the strength and ductility at the irradiation temperature. In addition, a series of interrupted tensile tests were performed in order to determine the variation of the yield strength as a function of temperature using an original single specimen test method. The neutron irradiation causes the reduction of the total elongation of both tungsten products. The DBTT range, which was evaluated from the tensile test results, of W–ZrC lies in the 300–500 °C range (while it is ∼100 °C in non-irradiated state). The DBTT range of pure tungsten is between 400 and 575 °C i.e. higher than that of W–ZrC. The irradiation hardening, measured at ∼600 °C, which is close to the irradiation temperature, leads to an increase of the proof stress by a factor of two in both studied grades. Despite that the irradiation induced hardening, both products retain a total elongation of about 10% prior to fracture. W–ZrC exhibits a similar total elongation at 500 °C, thus maintaining a significant ductility resource, while pure W becomes brittle at 500 °C and below.@en

The effect of neutron irradiation on the fracture toughness of two commercially pure tungsten materials processed according to ITER specifications has been investigated for three doses: 0.08 dpa, 0.44 dpa, and 0.67 dpa at 600 °C. The choice of this temperature was motivated by its technological importance due to the risk of irradiation-induced embrittlement. The temperature of 600 °C is below the void swelling peak temperature (∼800 °C) and, at the same time, well above the ductile to brittle transition temperature (DBTT) of the reference material (∼300 °C). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to limit the transmutation of rhenium and osmium close to the rates expected in a fusion environment. The results of the mechanical tests performed up to 600 °C show that the fracture toughness decreases with the increase in the irradiation dose for both tungsten products. The fracture surfaces of the non- A nd irradiated specimens were systematically analysed to determine the evolution of the failure mechanisms.@en

The effect of neutron irradiation on the fracture toughness of two commercially pure tungsten materials processed according to ITER specifications has been investigated for three doses: 0.08 dpa, 0.44 dpa, and 0.67 dpa at 600 °C. The choice of this temperature was motivated by its technological importance due to the risk of irradiation-induced embrittlement. The temperature of 600 °C is below the void swelling peak temperature (∼800 °C) and, at the same time, well above the ductile to brittle transition temperature (DBTT) of the reference material (∼300 °C). Neutron irradiation was performed in the BR2 material test reactor inside the fuel channel in order to limit the transmutation of rhenium and osmium close to the rates expected in a fusion environment. The results of the mechanical tests performed up to 600 °C show that the fracture toughness decreases with the increase in the irradiation dose for both tungsten products. The fracture surfaces of the non- A nd irradiated specimens were systematically analysed to determine the evolution of the failure mechanisms.@en