R. Chaouadi - Profile on Academia.edu (original) (raw)
Papers by R. Chaouadi
Cryogenics, 1992
YBa2Cu307-x films were grown by a low pressure plasma spraying technique and atmospheric plasma s... more YBa2Cu307-x films were grown by a low pressure plasma spraying technique and atmospheric plasma spray on polycrystalline nickel substrate using yttrium-stabilized zirconia as a buffer layer. After post-annealing in oxygen at high temperature, zero resistance is reached above 77 K. YBa2Cu307_x films with thicknesses between 120 and 420 #m were prepared and their critical currents measured. The a.c. magnetic shielding properties were determined in the high frequency region, 100 kHz to 3 MHz. The relations between the screening properties, the critical currents and the microstructure of the thin films are discussed.
FP7 Project LONGLIFE: Overview of results and implications
Nuclear Engineering and Design, 2014
ABSTRACT
RPV steel embrittlement: Damage modeling and micromechanics in an engineering perspective
Fracture Toughness Characterization of an A508-Type Weld Metal With the Mini-Ct Geometry Before and After Irradiation
Nomad: Non-Destructive Evaluation System for the Inspection of Operation Induced Material Degradation in Nuclear Power Plants
Contributions of Ni-content and irradiation temperature to the kinetic of solute cluster formation and consequences on the hardening of VVER materials
Journal of Nuclear Materials
Reactor Pressure Vessel Surveillance Programs in Belgium
International Review of Nuclear Reactor Pressure Vessel Surveillance Programs, 2018
Effect of step cooling and P-segregation to grain boundaries on the tensile and fracture toughness properties of A533B plate and A508 forging steels
Journal of Nuclear Materials, 2021
Abstract Although available data do not report significant thermal ageing of reactor pressure ves... more Abstract Although available data do not report significant thermal ageing of reactor pressure vessel materials, this should be reconsidered in the perspective of long term operation where exposure time is significantly higher than the available reported experimental data. One of the element that is reported to promote thermal ageing degradation is phosphorus. The thermal ageing embrittlement susceptibility is assessed by submitting reactor pressure vessel steels to a step cooling heat treatment known to induce P-segregation to grain boundaries resulting in intergranular rather than transgranular cleavage fracture. Thermal ageing susceptibility was experimentally assessed on two typical reactor pressure vessel steels with P-content of 0.017% and 0.006%. The experimental data including tensile, Charpy impact and fracture toughness tests, show a very consistent picture of the various mechanical properties. Upon step cooling, intergranular fracture is promoted and P-content has a major effect on the thermal ageing susceptibility. However, quantitative assessment shows that the effects of step cooling on the mechanical properties of both steels in their as received condition are not sufficiently high to induce thermal ageing embrittlement and therefore their susceptibility to thermal ageing is low. Consequently, given the low P-content of most Belgian RPV materials, it can be concluded that they will be not prone to thermal ageing in the perspective of long term operation.
Use of Broken Charpy V-notch Specimens from a Surveillance Program for Fracture Toughness Determination
Journal of ASTM International, 2006
An Energy-Based Crack Extension Formulation for Crack Resistance Characterization of Ductile Materials
Journal of Testing and Evaluation, 2004
Simulation of Irradiation Effects in Reactor Pressure Vessel Steels: the Reactor for Virtual Experiments (REVE) Project
Journal of Testing and Evaluation, 2002
Í\l3 Stephanie Jumel, Christophe Domain, ' Jacky Ruste, Jean-Claude Vati Duy... more Í\l3 Stephanie Jumel, Christophe Domain, ' Jacky Ruste, Jean-Claude Vati Duysen, ' Charlotte Becquart,2 Alexandre Le gris,1 Philippe Pareige,3 Alain Barbu* Eric Van Waller Rachid Chaouadi* Marc Hou,6 G, Robert Odette,'7 Roger E. Sfollerà and Brian D. Wtrtti* ...
Prediction of radiation induced hardening of reactor pressure vessel steels using artificial neural networks
Journal of Nuclear Materials, 2011
ABSTRACT In this paper, we use an artificial neural network approach to obtain predictions of neu... more ABSTRACT In this paper, we use an artificial neural network approach to obtain predictions of neutron irradiation induced hardening, more precisely of the change in the yield stress, for reactor pressure vessel steels of pressurized water nuclear reactors. Different training algorithms are proposed and compared, with the goal of identifying the best procedure to follow depending on the needs of the user. The numerical importance of some input variables is also studied. Very accurate numerical regressions are obtained, by taking only four input variables into account: neutron fluence, irradiation temperature, and chemical composition (Cu and Ni content). Accurate extrapolations in term of neutron fluence are obtained.
Tensile and low-cycle fatigue properties of solution annealed type 316L stainless steel plate and TIG-weld exposed to 5 dpa at low-temperature (42°C)
Journal of Nuclear Materials, 2000
The austenitic stainless steel type AISI 316L was selected as the main structural material of the... more The austenitic stainless steel type AISI 316L was selected as the main structural material of the next-step International thermonuclear experimental reactor (ITER) fusion device, i.e., the first wall, blanket modules, and vacuum vessel components. Although this steel was extensively investigated under different aspects, most results concern irradiation temperatures above 300°C. In the present work, tensile and fatigue specimens were irradiated in the BR2 materials testing reactor at 42°C up to a maximum neutron fluence of 8×1021 n/cm2 (E>0.1 MeV), corresponding to 5 dpa. The European reference AISI 316L in the solution annealed condition and the TIG-metal deposit were tested in the baseline and irradiated conditions. The tensile specimens were tested at 25°C, 250°C and 450°C, while the low-cycle fatigue tests were performed at room temperature. The tensile test results obtained in this work are consistent with published data: substantial radiation hardening combined with some reduction of elongation. No specimen orientation effect could be evidenced. The amount of hardening decreases with increasing test temperature. By contrast, the low-cycle fatigue data show no or little effect of irradiation, independent from irradiation and testing conditions. No major difference was found between the plate and the weld metal.
Mechanical properties of the European reference RAFM steel (EUROFER97) before and after irradiation at 300 °C
Journal of Nuclear Materials, 2004
ABSTRACT EUROFER97 is the European candidate RAFM steel for use as a structural material in fusio... more ABSTRACT EUROFER97 is the European candidate RAFM steel for use as a structural material in fusion energy systems. It is presently under investigation by several European laboratories, within the long term programme of EFDA (European Fusion Development Agreement). This paper presents the outcome of the mechanical characterization of this steel that has been carried out at SCK•CEN (Belgian Center for Nuclear Studies), in the unirradiated condition and after irradiation during three different campaigns in the BR2 reactor (300 °C, doses between 0.3 and 1.6 dpa). Tensile, Charpy impact, and fracture toughness specimens have been irradiated and tested, in order to obtain an experimental assessment of the main effects of neutron exposure on tensile and toughness properties; namely, irradiation hardening (increase of mechanical resistance and loss of ductility) and irradiation embrittlement (shift of ductile-to-brittle transition temperature and degradation of upper shelf energy). Comparisons with another well-known IEA reference RAFM steel, F82H, are provided.
Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo–5%Re
Journal of Nuclear Materials, 2000
Due to their good resistance at high temperature, good thermal conductivity and swelling resistan... more Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation-induced embrittlement. This paper aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo–5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40°Cand450°C up to a fast neutron fluence of 3.5×1020n/cm2 (0.2 dpa). Tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain.
Journal of Nuclear Materials, 2009
This paper provides a physically-based engineering model to estimate radiation hardening of 9%Cr-... more This paper provides a physically-based engineering model to estimate radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades but incorporating a number of assumptions and simplifications [Trinkaus, J. Nucl. Mater. 318 ]. As a result, the kinetics of the damage accumulation kept fixed, its amplitude is fitted on one experimental condition. The model was rationalized on an experimental database that mainly consists of $9%Cr-steels irradiated in the range of 50-600°C up to 50 dpa and with a He-content up to 5000 appm. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Despite the large experimental scatter, inherent to the variety of the material and irradiation as well as testing conditions, the obtained results are very promising.
Journal of Nuclear Materials, 2005
Copper is known to play an important role in irradiation hardening and embrittlement of RPV mater... more Copper is known to play an important role in irradiation hardening and embrittlement of RPV materials. This is particularly true for old vessels. Indeed, while the Cu-content is low (<0.1%) in modern RPV materials, it often exceeded 0.15% in older vessels. Within the RADAMO irradiation program aiming to provide a reliable and extensive (chemistry, heat treatments, fluence, irradiation temperature) databank to investigate irradiation-induced hardening of RPV materials, we irradiated steels and welds with copper contents ranging from 0.06 to 0.31% at 300 and 265°C. Experiments on re-irradiation after annealing were also performed to investigate the re-irradiation kinetics. It is found that copper plays a role in the very early stage of irradiation but saturates quite rapidly. The peak hardening is in agreement with the ageing data. Considering a two-component model, the linear superposition law provides the most appropriate one to rationalize the experimental data including re-irradiation path.
Neutron flux and annealing effects on irradiation hardening of RPV materials
Journal of Nuclear Materials, 2011
Effect of crack length-to-width ratio on crack resistance of high Cr-ODS steels at high temperature for fuel cladding application
Journal of Nuclear Materials, 2013
Confirmatory investigations on the flux effect and associated unstable matrix damage in RPV materials exposed to high neutron fluence
Journal of Nuclear Materials, 2013
Cryogenics, 1992
YBa2Cu307-x films were grown by a low pressure plasma spraying technique and atmospheric plasma s... more YBa2Cu307-x films were grown by a low pressure plasma spraying technique and atmospheric plasma spray on polycrystalline nickel substrate using yttrium-stabilized zirconia as a buffer layer. After post-annealing in oxygen at high temperature, zero resistance is reached above 77 K. YBa2Cu307_x films with thicknesses between 120 and 420 #m were prepared and their critical currents measured. The a.c. magnetic shielding properties were determined in the high frequency region, 100 kHz to 3 MHz. The relations between the screening properties, the critical currents and the microstructure of the thin films are discussed.
FP7 Project LONGLIFE: Overview of results and implications
Nuclear Engineering and Design, 2014
ABSTRACT
RPV steel embrittlement: Damage modeling and micromechanics in an engineering perspective
Fracture Toughness Characterization of an A508-Type Weld Metal With the Mini-Ct Geometry Before and After Irradiation
Nomad: Non-Destructive Evaluation System for the Inspection of Operation Induced Material Degradation in Nuclear Power Plants
Contributions of Ni-content and irradiation temperature to the kinetic of solute cluster formation and consequences on the hardening of VVER materials
Journal of Nuclear Materials
Reactor Pressure Vessel Surveillance Programs in Belgium
International Review of Nuclear Reactor Pressure Vessel Surveillance Programs, 2018
Effect of step cooling and P-segregation to grain boundaries on the tensile and fracture toughness properties of A533B plate and A508 forging steels
Journal of Nuclear Materials, 2021
Abstract Although available data do not report significant thermal ageing of reactor pressure ves... more Abstract Although available data do not report significant thermal ageing of reactor pressure vessel materials, this should be reconsidered in the perspective of long term operation where exposure time is significantly higher than the available reported experimental data. One of the element that is reported to promote thermal ageing degradation is phosphorus. The thermal ageing embrittlement susceptibility is assessed by submitting reactor pressure vessel steels to a step cooling heat treatment known to induce P-segregation to grain boundaries resulting in intergranular rather than transgranular cleavage fracture. Thermal ageing susceptibility was experimentally assessed on two typical reactor pressure vessel steels with P-content of 0.017% and 0.006%. The experimental data including tensile, Charpy impact and fracture toughness tests, show a very consistent picture of the various mechanical properties. Upon step cooling, intergranular fracture is promoted and P-content has a major effect on the thermal ageing susceptibility. However, quantitative assessment shows that the effects of step cooling on the mechanical properties of both steels in their as received condition are not sufficiently high to induce thermal ageing embrittlement and therefore their susceptibility to thermal ageing is low. Consequently, given the low P-content of most Belgian RPV materials, it can be concluded that they will be not prone to thermal ageing in the perspective of long term operation.
Use of Broken Charpy V-notch Specimens from a Surveillance Program for Fracture Toughness Determination
Journal of ASTM International, 2006
An Energy-Based Crack Extension Formulation for Crack Resistance Characterization of Ductile Materials
Journal of Testing and Evaluation, 2004
Simulation of Irradiation Effects in Reactor Pressure Vessel Steels: the Reactor for Virtual Experiments (REVE) Project
Journal of Testing and Evaluation, 2002
Í\l3 Stephanie Jumel, Christophe Domain, &amp;amp;#x27; Jacky Ruste, Jean-Claude Vati Duy... more Í\l3 Stephanie Jumel, Christophe Domain, &amp;amp;#x27; Jacky Ruste, Jean-Claude Vati Duysen, &amp;amp;#x27; Charlotte Becquart,2 Alexandre Le gris,1 Philippe Pareige,3 Alain Barbu* Eric Van Waller Rachid Chaouadi* Marc Hou,6 G, Robert Odette,&amp;amp;#x27;7 Roger E. Sfollerà and Brian D. Wtrtti* ...
Prediction of radiation induced hardening of reactor pressure vessel steels using artificial neural networks
Journal of Nuclear Materials, 2011
ABSTRACT In this paper, we use an artificial neural network approach to obtain predictions of neu... more ABSTRACT In this paper, we use an artificial neural network approach to obtain predictions of neutron irradiation induced hardening, more precisely of the change in the yield stress, for reactor pressure vessel steels of pressurized water nuclear reactors. Different training algorithms are proposed and compared, with the goal of identifying the best procedure to follow depending on the needs of the user. The numerical importance of some input variables is also studied. Very accurate numerical regressions are obtained, by taking only four input variables into account: neutron fluence, irradiation temperature, and chemical composition (Cu and Ni content). Accurate extrapolations in term of neutron fluence are obtained.
Tensile and low-cycle fatigue properties of solution annealed type 316L stainless steel plate and TIG-weld exposed to 5 dpa at low-temperature (42°C)
Journal of Nuclear Materials, 2000
The austenitic stainless steel type AISI 316L was selected as the main structural material of the... more The austenitic stainless steel type AISI 316L was selected as the main structural material of the next-step International thermonuclear experimental reactor (ITER) fusion device, i.e., the first wall, blanket modules, and vacuum vessel components. Although this steel was extensively investigated under different aspects, most results concern irradiation temperatures above 300°C. In the present work, tensile and fatigue specimens were irradiated in the BR2 materials testing reactor at 42°C up to a maximum neutron fluence of 8×1021 n/cm2 (E>0.1 MeV), corresponding to 5 dpa. The European reference AISI 316L in the solution annealed condition and the TIG-metal deposit were tested in the baseline and irradiated conditions. The tensile specimens were tested at 25°C, 250°C and 450°C, while the low-cycle fatigue tests were performed at room temperature. The tensile test results obtained in this work are consistent with published data: substantial radiation hardening combined with some reduction of elongation. No specimen orientation effect could be evidenced. The amount of hardening decreases with increasing test temperature. By contrast, the low-cycle fatigue data show no or little effect of irradiation, independent from irradiation and testing conditions. No major difference was found between the plate and the weld metal.
Mechanical properties of the European reference RAFM steel (EUROFER97) before and after irradiation at 300 °C
Journal of Nuclear Materials, 2004
ABSTRACT EUROFER97 is the European candidate RAFM steel for use as a structural material in fusio... more ABSTRACT EUROFER97 is the European candidate RAFM steel for use as a structural material in fusion energy systems. It is presently under investigation by several European laboratories, within the long term programme of EFDA (European Fusion Development Agreement). This paper presents the outcome of the mechanical characterization of this steel that has been carried out at SCK•CEN (Belgian Center for Nuclear Studies), in the unirradiated condition and after irradiation during three different campaigns in the BR2 reactor (300 °C, doses between 0.3 and 1.6 dpa). Tensile, Charpy impact, and fracture toughness specimens have been irradiated and tested, in order to obtain an experimental assessment of the main effects of neutron exposure on tensile and toughness properties; namely, irradiation hardening (increase of mechanical resistance and loss of ductility) and irradiation embrittlement (shift of ductile-to-brittle transition temperature and degradation of upper shelf energy). Comparisons with another well-known IEA reference RAFM steel, F82H, are provided.
Analysis of tensile and fracture toughness results on irradiated molybdenum alloys, TZM and Mo–5%Re
Journal of Nuclear Materials, 2000
Due to their good resistance at high temperature, good thermal conductivity and swelling resistan... more Due to their good resistance at high temperature, good thermal conductivity and swelling resistance, molybdenum alloys are amongst the candidates for divertor structural materials. However, little is known about their tensile and fracture toughness behaviour, in particular after irradiation-induced embrittlement. This paper aims to investigate the tensile and fracture toughness properties of two molybdenum alloys, namely TZM and Mo–5%Re. Tensile and compact tension specimens were irradiated in the BR2 reactor at 40°Cand450°C up to a fast neutron fluence of 3.5×1020n/cm2 (0.2 dpa). Tests were performed on both precracked and notched specimens. Results show a drastic decrease of the ductility due to irradiation, but only a slight decrease of the fracture toughness in the lower shelf domain.
Journal of Nuclear Materials, 2009
This paper provides a physically-based engineering model to estimate radiation hardening of 9%Cr-... more This paper provides a physically-based engineering model to estimate radiation hardening of 9%Cr-steels under both displacement damage (dpa) and helium. The model is essentially based on the dispersed barrier hardening theory and the dynamic re-solution of helium under displacement cascades but incorporating a number of assumptions and simplifications [Trinkaus, J. Nucl. Mater. 318 ]. As a result, the kinetics of the damage accumulation kept fixed, its amplitude is fitted on one experimental condition. The model was rationalized on an experimental database that mainly consists of $9%Cr-steels irradiated in the range of 50-600°C up to 50 dpa and with a He-content up to 5000 appm. The test temperature effect is taken into account through a normalization procedure based on the change of the Young's modulus and the anelastic deformation that occurs at high temperature. Despite the large experimental scatter, inherent to the variety of the material and irradiation as well as testing conditions, the obtained results are very promising.
Journal of Nuclear Materials, 2005
Copper is known to play an important role in irradiation hardening and embrittlement of RPV mater... more Copper is known to play an important role in irradiation hardening and embrittlement of RPV materials. This is particularly true for old vessels. Indeed, while the Cu-content is low (<0.1%) in modern RPV materials, it often exceeded 0.15% in older vessels. Within the RADAMO irradiation program aiming to provide a reliable and extensive (chemistry, heat treatments, fluence, irradiation temperature) databank to investigate irradiation-induced hardening of RPV materials, we irradiated steels and welds with copper contents ranging from 0.06 to 0.31% at 300 and 265°C. Experiments on re-irradiation after annealing were also performed to investigate the re-irradiation kinetics. It is found that copper plays a role in the very early stage of irradiation but saturates quite rapidly. The peak hardening is in agreement with the ageing data. Considering a two-component model, the linear superposition law provides the most appropriate one to rationalize the experimental data including re-irradiation path.
Neutron flux and annealing effects on irradiation hardening of RPV materials
Journal of Nuclear Materials, 2011
Effect of crack length-to-width ratio on crack resistance of high Cr-ODS steels at high temperature for fuel cladding application
Journal of Nuclear Materials, 2013
Confirmatory investigations on the flux effect and associated unstable matrix damage in RPV materials exposed to high neutron fluence
Journal of Nuclear Materials, 2013