Pål Efsing - Academia.edu (original) (raw)

Papers by Pål Efsing

Electron backscatter diffraction data sets used in this work. The files named 00, 03, 06, 09, 12 ... more Electron backscatter diffraction data sets used in this work. The files named 00, 03, 06, 09, 12 and 15.ctf are from the reference specimens, and the filenames represent the approximate nominal plastic strains applied. The files CRDM1, CRDM2 and DP.ctf represent the weld metal to base metal interface region.

Defect distributions have been documented by optical metallography, scanning electron microscopy ... more Defect distributions have been documented by optical metallography, scanning electron microscopy and electron backscatter diffraction in alloy 152 and 52 mockups welds, alloy 52 and 52M overlay mockups and an alloy 52M inlay. Primary defects were small isolated grain boundary cracks except for more extensive cracking in the dilution zone of an alloy 52 overlay on 304SS. Detailed characterizations of the dilution zone cracks were performed by analytical transmission electron microscopy identifying grain boundary titanium-nitride precipitation associated with the intergranular separations.

Springer eBooks, 2011

New stress corrosion studies are presented from initiation tests and crack growth measurements on... more New stress corrosion studies are presented from initiation tests and crack growth measurements on Alloy 600 in primary PWR environment. The effects of hydrogen, lithium and boron have been studied. Initiation tests have been performed at a high hydrogen content of 70 cc/kg. A longer initiation time was measured compared to intermediate levels but shorter than for low hydrogen concentration of 5 ml/kg. Initiation tests at a Li content of 6 ppm have also been performed. The time to initiation was longer than for 3.5 ppm and about the same as for 2.2 ppm. In addition crack growth measurements using different Li/B contents to simulate a reactor cycle and study any variations with the different composition has been performed. The results show a weak influence of the primary environment. A correlation with the boron concentration was noticed and for high boron contents also a weak dependence on the lithium concentration. The results show that the beginning of cycle conditions usually used for PWSCC studies creates conservative crack growth rates.

Because the reactors of the Swedish reactor program were erected over a limited period of time, t... more Because the reactors of the Swedish reactor program were erected over a limited period of time, there are significant similarities regarding used materials and manufacturing methods between the dif ...

ASTM International eBooks, May 19, 2014

As part of an on-going effort to verify the long-term fitness for service of the PWR-plants at th... more As part of an on-going effort to verify the long-term fitness for service of the PWR-plants at the Swedish Ringhals site, weld material relevant for the two most modern units has been irradiated in the OECD Halden Materials test reactor (MTR) for up to three cycles of operation. The dose level achieved for each cycle is approximately equivalent to 20 years of operation during Light Water Reactor (LWR) conditions. The purpose of the test was three-fold. The first objective was to study the effect of the dose rate, the flux, -level on these kind of materials in order to verify or to discard the use of MTR-irradiated materials as part of the model building to understand the evolution of the mechanical behaviour under LWR conditions. The second objective was to enhance the available database of post-irradiation mechanical properties for analyses purposes, such as reactor pressure-temperature limit curves and defect tolerance analyses. Finally, the third objective was to produce ample amount of relevant irradiated material, enabling a comprehensive microscopy analysis of the evolution of the structure in the material to establish the occurrence frequency and type of precipitates and agglomerates, and if possible to study the occurrence of late blooming phases in high Ni and Mn bearing materials. This study will concentrate on the two first objectives. From the study, it appears clear that with these materials, it is possible to enhance the flux to speed up the irradiation induced degradation and still produce results that fall well in line with data extracted from the normal surveillance programs of the reactors. The flux effect as analysed from the mechanical property data appears to be negligible, if any.

Engineering Fracture Mechanics, Jun 1, 2019

A probabilistic model for the cumulative probability of failure by cleavage fracture with a mater... more A probabilistic model for the cumulative probability of failure by cleavage fracture with a material related length scale is further developed in this study. A new generalized effective stress measure is proposed, based on a normal stress decomposition of the stress tensor, capable of describing a state of normal stress in the range from the mean stress to the maximum principal stress. The effective stress measure associated with a material point is evaluated from the stress tensor averaged over the material related length scale. The model is shown to be well capable to predict both the fracture toughness at loss of both in-plane and out-of-plane constraint by model application to two different datasets from the open literature. The model is also shown to be well capable of predicting the probability of failure of discriminating experiments on specimens containing semi-elliptic surface cracks. A comparison where the master curve methodology is used to predict the probability of failure of the experiments is also included.

Under some circumstances nuclear fuel cladding tubes made from zirconium based alloys may develop... more Under some circumstances nuclear fuel cladding tubes made from zirconium based alloys may develop long axial cracks. The formation of these cracks is mainly thought to be connected with the oxiditi ...

Fracture toughness testing is normally performed in air on specimens provided with a transgranula... more Fracture toughness testing is normally performed in air on specimens provided with a transgranular pre-crack generated in air by fatigue loading. However, stress corrosion cracks in nuclear power plants are usually intergranular and in contact with reactor coolant. Fracture toughness data used in e.g., flaw tolerance analyses are generated in air with transgranular pre-cracks. Since the effects of the fracture mode of the pre-crack and the reactor coolant on the fracture toughness are not known in detail, it is important to investigate if the data used today are sufficiently conservative. Compact tension (CT) specimens of Alloy X-750 with thickness (B) 9.3 mm and width (W) 18.6 mm were tested under various conditions with the objective to investigate the possible effects of an intergranular pre-crack as well as BWR coolant on the fracture toughness. Three specimens were tested under constant stress intensity (K) in simulated BWR normal water chemistry (NWC) in order to generate an intergranular pre-crack. One specimen was removed from the autoclave and then fracture toughness tested in air at 288 oC. The other specimens remained in the autoclave in the presence of simulated BWR coolant during the fracture toughness test. For comparison, specimens with a transgranular pre-crack were tested in air at 288 oC. Neither the fracture mode, nor the BWR coolant appeared to have any adverse effects on the fracture toughness in these tests.

Modelling of IGSCC mechanism through coupling of a potential-based cohesive model and Fick’s seco... more Modelling of IGSCC mechanism through coupling of a potential-based cohesive model and Fick’s second law

In many nuclear power plants, areas of susceptible material in the reactor systems are replaced o... more In many nuclear power plants, areas of susceptible material in the reactor systems are replaced or mitigated. Many of the areas where the nickel-based weld metal Alloy 182 have been used, are not r ...

International Journal of Nuclear Energy, Aug 3, 2014

The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 ... more The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 has been questioned in recent years. As a step towards understanding its relevancy for weld deformed Alloy 690 in operating plants, Alloy 690 base metal and heat affected zone (HAZ) microstructures of three mockup components have been studied. All mockups were manufactured using commercial heats and welding procedures in order to attain results relevant to the materials in the field. Thermodynamic calculations were performed to add confidence in phase identification as well as understanding of the evolution of the microstructure with temperature. Ti(C,N) banding was found in all materials. Bands with few large Ti(C,N) precipitates had negligible effect on the microstructure, whereas bands consisting of numerous small precipitates were associated with locally finer grains and coarser M 23 C 6 grain boundary carbides. The Ti(C,N) remained unaffected in the HAZ while the M 23 C 6 carbides were fully dissolved close to the fusion line. Cold deformed solution annealed Alloy 690 is believed to be a better representation of this region than cold deformed TT Alloy 690.

Engineering Fracture Mechanics, Mar 1, 2022

A thermally aged low alloy steel is investigated in terms of its fracture toughness and microstru... more A thermally aged low alloy steel is investigated in terms of its fracture toughness and microstructural evolution and compared to a reference. The main purpose of the study is to investigate the ef ...

Journal of Nuclear Materials, Feb 1, 2017

h i g h l i g h t s Hardness testing is combined with post irradiation annealing at 330, 360 and ... more h i g h l i g h t s Hardness testing is combined with post irradiation annealing at 330, 360 and 390 C. Unstable matrix defects is studied in a reactor pressure vessel steel. Comparison between surveillance material and accelerated irradiation. No evidence of unstable matrix defects, i.e. not present in studied material. Difference in hardness recovery between irradiation conditions found at 390 C.

Journal of Nuclear Materials, May 1, 2017

Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation... more Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58 %) low Cu (0.04 %) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance

Journal of Nuclear Materials, Feb 1, 2009

The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating p... more The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290°C and 70 dpa at 315°C were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during postirradiation slow strain rate testing in PWR water conditions.

Journal of Nuclear Materials, Oct 1, 2018

SANS experiments were performed on a high Ni weld surveillance sample from the Ringhals NPP and t... more SANS experiments were performed on a high Ni weld surveillance sample from the Ringhals NPP and the Maximum Entropy method was applied to determine the most probable size distribution of irradiation-induced scattering features. The results were shown to be consistent with atom probe observations. The sensitivity of the data analyses with respect to constraints such as the limited experimentally available Q range was explored. The calculated volume fraction and the mean volume-weighted diameter of the precipitates were found to be relatively insensitive to Q max (the maximum scattering vector) greater than ∼0.40 Å −1. However, use of a lower Q max results in a shift of the size distribution to larger diameters and a reduced particle number density. Simulations demonstrated that the experimentally observed decrease in the A-ratio at higher Q values is consistent with the presence of vacancies or higher Mn contents in smaller features. Importantly, features which are experimentally unresolvable do not add to the apparent volume fraction of the features which are resolved.

Journal of Nuclear Materials, Aug 1, 2023

Electron backscatter diffraction data sets used in this work. The files named 00, 03, 06, 09, 12 ... more Electron backscatter diffraction data sets used in this work. The files named 00, 03, 06, 09, 12 and 15.ctf are from the reference specimens, and the filenames represent the approximate nominal plastic strains applied. The files CRDM1, CRDM2 and DP.ctf represent the weld metal to base metal interface region.

Defect distributions have been documented by optical metallography, scanning electron microscopy ... more Defect distributions have been documented by optical metallography, scanning electron microscopy and electron backscatter diffraction in alloy 152 and 52 mockups welds, alloy 52 and 52M overlay mockups and an alloy 52M inlay. Primary defects were small isolated grain boundary cracks except for more extensive cracking in the dilution zone of an alloy 52 overlay on 304SS. Detailed characterizations of the dilution zone cracks were performed by analytical transmission electron microscopy identifying grain boundary titanium-nitride precipitation associated with the intergranular separations.

Springer eBooks, 2011

New stress corrosion studies are presented from initiation tests and crack growth measurements on... more New stress corrosion studies are presented from initiation tests and crack growth measurements on Alloy 600 in primary PWR environment. The effects of hydrogen, lithium and boron have been studied. Initiation tests have been performed at a high hydrogen content of 70 cc/kg. A longer initiation time was measured compared to intermediate levels but shorter than for low hydrogen concentration of 5 ml/kg. Initiation tests at a Li content of 6 ppm have also been performed. The time to initiation was longer than for 3.5 ppm and about the same as for 2.2 ppm. In addition crack growth measurements using different Li/B contents to simulate a reactor cycle and study any variations with the different composition has been performed. The results show a weak influence of the primary environment. A correlation with the boron concentration was noticed and for high boron contents also a weak dependence on the lithium concentration. The results show that the beginning of cycle conditions usually used for PWSCC studies creates conservative crack growth rates.

Because the reactors of the Swedish reactor program were erected over a limited period of time, t... more Because the reactors of the Swedish reactor program were erected over a limited period of time, there are significant similarities regarding used materials and manufacturing methods between the dif ...

ASTM International eBooks, May 19, 2014

As part of an on-going effort to verify the long-term fitness for service of the PWR-plants at th... more As part of an on-going effort to verify the long-term fitness for service of the PWR-plants at the Swedish Ringhals site, weld material relevant for the two most modern units has been irradiated in the OECD Halden Materials test reactor (MTR) for up to three cycles of operation. The dose level achieved for each cycle is approximately equivalent to 20 years of operation during Light Water Reactor (LWR) conditions. The purpose of the test was three-fold. The first objective was to study the effect of the dose rate, the flux, -level on these kind of materials in order to verify or to discard the use of MTR-irradiated materials as part of the model building to understand the evolution of the mechanical behaviour under LWR conditions. The second objective was to enhance the available database of post-irradiation mechanical properties for analyses purposes, such as reactor pressure-temperature limit curves and defect tolerance analyses. Finally, the third objective was to produce ample amount of relevant irradiated material, enabling a comprehensive microscopy analysis of the evolution of the structure in the material to establish the occurrence frequency and type of precipitates and agglomerates, and if possible to study the occurrence of late blooming phases in high Ni and Mn bearing materials. This study will concentrate on the two first objectives. From the study, it appears clear that with these materials, it is possible to enhance the flux to speed up the irradiation induced degradation and still produce results that fall well in line with data extracted from the normal surveillance programs of the reactors. The flux effect as analysed from the mechanical property data appears to be negligible, if any.

Engineering Fracture Mechanics, Jun 1, 2019

A probabilistic model for the cumulative probability of failure by cleavage fracture with a mater... more A probabilistic model for the cumulative probability of failure by cleavage fracture with a material related length scale is further developed in this study. A new generalized effective stress measure is proposed, based on a normal stress decomposition of the stress tensor, capable of describing a state of normal stress in the range from the mean stress to the maximum principal stress. The effective stress measure associated with a material point is evaluated from the stress tensor averaged over the material related length scale. The model is shown to be well capable to predict both the fracture toughness at loss of both in-plane and out-of-plane constraint by model application to two different datasets from the open literature. The model is also shown to be well capable of predicting the probability of failure of discriminating experiments on specimens containing semi-elliptic surface cracks. A comparison where the master curve methodology is used to predict the probability of failure of the experiments is also included.

Under some circumstances nuclear fuel cladding tubes made from zirconium based alloys may develop... more Under some circumstances nuclear fuel cladding tubes made from zirconium based alloys may develop long axial cracks. The formation of these cracks is mainly thought to be connected with the oxiditi ...

Fracture toughness testing is normally performed in air on specimens provided with a transgranula... more Fracture toughness testing is normally performed in air on specimens provided with a transgranular pre-crack generated in air by fatigue loading. However, stress corrosion cracks in nuclear power plants are usually intergranular and in contact with reactor coolant. Fracture toughness data used in e.g., flaw tolerance analyses are generated in air with transgranular pre-cracks. Since the effects of the fracture mode of the pre-crack and the reactor coolant on the fracture toughness are not known in detail, it is important to investigate if the data used today are sufficiently conservative. Compact tension (CT) specimens of Alloy X-750 with thickness (B) 9.3 mm and width (W) 18.6 mm were tested under various conditions with the objective to investigate the possible effects of an intergranular pre-crack as well as BWR coolant on the fracture toughness. Three specimens were tested under constant stress intensity (K) in simulated BWR normal water chemistry (NWC) in order to generate an intergranular pre-crack. One specimen was removed from the autoclave and then fracture toughness tested in air at 288 oC. The other specimens remained in the autoclave in the presence of simulated BWR coolant during the fracture toughness test. For comparison, specimens with a transgranular pre-crack were tested in air at 288 oC. Neither the fracture mode, nor the BWR coolant appeared to have any adverse effects on the fracture toughness in these tests.

Modelling of IGSCC mechanism through coupling of a potential-based cohesive model and Fick’s seco... more Modelling of IGSCC mechanism through coupling of a potential-based cohesive model and Fick’s second law

In many nuclear power plants, areas of susceptible material in the reactor systems are replaced o... more In many nuclear power plants, areas of susceptible material in the reactor systems are replaced or mitigated. Many of the areas where the nickel-based weld metal Alloy 182 have been used, are not r ...

International Journal of Nuclear Energy, Aug 3, 2014

The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 ... more The stress corrosion cracking (SCC) resistance of cold deformed thermally treated (TT) Alloy 690 has been questioned in recent years. As a step towards understanding its relevancy for weld deformed Alloy 690 in operating plants, Alloy 690 base metal and heat affected zone (HAZ) microstructures of three mockup components have been studied. All mockups were manufactured using commercial heats and welding procedures in order to attain results relevant to the materials in the field. Thermodynamic calculations were performed to add confidence in phase identification as well as understanding of the evolution of the microstructure with temperature. Ti(C,N) banding was found in all materials. Bands with few large Ti(C,N) precipitates had negligible effect on the microstructure, whereas bands consisting of numerous small precipitates were associated with locally finer grains and coarser M 23 C 6 grain boundary carbides. The Ti(C,N) remained unaffected in the HAZ while the M 23 C 6 carbides were fully dissolved close to the fusion line. Cold deformed solution annealed Alloy 690 is believed to be a better representation of this region than cold deformed TT Alloy 690.

Engineering Fracture Mechanics, Mar 1, 2022

A thermally aged low alloy steel is investigated in terms of its fracture toughness and microstru... more A thermally aged low alloy steel is investigated in terms of its fracture toughness and microstructural evolution and compared to a reference. The main purpose of the study is to investigate the ef ...

Journal of Nuclear Materials, Feb 1, 2017

h i g h l i g h t s Hardness testing is combined with post irradiation annealing at 330, 360 and ... more h i g h l i g h t s Hardness testing is combined with post irradiation annealing at 330, 360 and 390 C. Unstable matrix defects is studied in a reactor pressure vessel steel. Comparison between surveillance material and accelerated irradiation. No evidence of unstable matrix defects, i.e. not present in studied material. Difference in hardness recovery between irradiation conditions found at 390 C.

Journal of Nuclear Materials, May 1, 2017

Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation... more Reactor pressure vessel steel welds are affected by irradiation during operation. The irradiation results in nanometre cluster formation, which in turn affects the mechanical properties of the material, e.g. the ductile-to-brittle transition temperature is shifted to higher levels. In this study, cluster formation is characterised in high Ni (1.58 %) low Cu (0.04 %) steel welds identical to Ringhals R4 welds, using atom probe tomography in both surveillance material and in material irradiated at accelerated dose rates. Clusters containing mainly Ni and Mn, but also some Si and Cu were observed in all of the irradiated materials. Their evolution did not change drastically during irradiation; the clusters grew and new clusters were nucleated. Hence, both the cluster number density and the average size increased with irradiation time. Some flux effects were observed when comparing the high flux material and the surveillance material. The surveillance

Journal of Nuclear Materials, Feb 1, 2009

The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating p... more The radiation-induced microstructure of a cold-worked 316SS flux thimble tube from an operating pressurized water reactor (PWR) was examined. Two irradiated conditions, 33 dpa at 290°C and 70 dpa at 315°C were examined by transmission electron microscopy. The original dislocation network had completely disappeared and was replaced by fine dispersions of Frank loops and small nano-cavities at high densities. The latter appear to be bubbles containing high levels of helium and hydrogen. An enhanced distribution of these nano-cavities was found at grain boundaries and may play a role in the increased susceptibility of the irradiated 316SS to intergranular failure of specimens from this tube during postirradiation slow strain rate testing in PWR water conditions.

Journal of Nuclear Materials, Oct 1, 2018

SANS experiments were performed on a high Ni weld surveillance sample from the Ringhals NPP and t... more SANS experiments were performed on a high Ni weld surveillance sample from the Ringhals NPP and the Maximum Entropy method was applied to determine the most probable size distribution of irradiation-induced scattering features. The results were shown to be consistent with atom probe observations. The sensitivity of the data analyses with respect to constraints such as the limited experimentally available Q range was explored. The calculated volume fraction and the mean volume-weighted diameter of the precipitates were found to be relatively insensitive to Q max (the maximum scattering vector) greater than ∼0.40 Å −1. However, use of a lower Q max results in a shift of the size distribution to larger diameters and a reduced particle number density. Simulations demonstrated that the experimentally observed decrease in the A-ratio at higher Q values is consistent with the presence of vacancies or higher Mn contents in smaller features. Importantly, features which are experimentally unresolvable do not add to the apparent volume fraction of the features which are resolved.

Journal of Nuclear Materials, Aug 1, 2023