Effect of Neutron Irradiation on the Mechanical Properties of an A508 CL2 and 15Kh2NMFA Irradiated in the NOMAD_3 Rig in the BR2 Cooling Water

2021 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Cooling Water ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Significant Loss ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Transition Curves ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR are between 290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 260°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing happens below approximately 230°C. The NOMAD_3 irradiation consisted in total of 24 Charpy sized samples from an A508 Cl.2 forging and a 15Kh2NMFA material. They were irradiated to three various fluences between 1.55 and 7.90 × 1019 n/cm2 (E > 1MeV) at approximately 100°C. The hardening of the A508 Cl.2 was between 260 and 400 MPa which was much higher than the NOMAD_0 properties which were irradiated at approximately 280°C. The tensile tests of irradiated materials are all characterized by a significant loss of work hardening capacity leading to plastic flow localization promptly after the yield strength is reached. This affects also the shape of the Charpy impact transition curves. The radiation embrittlement derived from Charpy impact tests, ΔT41J, is up to 156°C for the highest fluence. For this irradiation, the embrittlement to hardening ratio was also around 0.43 +/−0.2°C/MPa as it was found in the previous campaign NOMAD_0. This paper discusses the tensile, hardness and impact properties of the NOMAD_3 irradiation campaign. It is compared to the NOMAD_0 with respect to effect of irradiation temperature and annealing recovery.

Effect of Neutron Irradiation on the Mechanical Properties of an A508 Cl.2 Forging Irradiated in a BAMI Capsule

2020 ◽  
Author(s):  
Inge Uytdenhouwen ◽  
Rachid Chaouadi
Keyword(s):  
Fracture Toughness ◽  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Operating Temperature ◽  
Specimen Size ◽  
Irradiation Temperature ◽  
Irradiation Damage ◽  
Temperature Window ◽  
Preliminary Irradiation ◽  
Reactor Pressure

Abstract The typical operating temperatures of a nuclear reactor pressure vessel in a PWR is typically between ∼290°C and 300°C. However, many BWRs and some PWRs operate at slightly lower temperatures down to 275°C. Most of the literature and neutron irradiation damage is therefore focused on those irradiation temperatures. It is well-known that the lower the irradiation temperature, the more neutron irradiation damage occurs, because no appreciable annealing occurs at approximately 230°C. The NOMAD-0 irradiation campaign at the BR2 was a preliminary irradiation specifically designed to determine the appropriate irradiation conditions that result in specific irradiation damage levels of an A508 Cl.2 grade at lower temperatures than the usual PWR operating temperature window. The BAMI capsules with controlled He gas gap were used for this irradiation. To avoid temperature gradients from the outside to the center of the cylindrical blocks, the latter were limited in size. Only tensile and fracture toughness data could therefore be obtained with mini-tensile and mini-CT specimens. The results show that no influence of the temperature gradient could be found on the tensile and fracture toughness properties. The specimen size and geometry on the fracture toughness results was analyzed and discussed. The effect of irradiation temperature on the tensile properties and the transition temperature as determined by the master curve approach on the mini-CT samples is discussed.

Effects of Neutron Irradiation on the Mechanical and Electromagnetic Properties of Reactor Pressure Vessel Steels

2017 ◽  
Author(s):  
Li Chengliang ◽  
Shu Guogang ◽  
Chen Jun ◽  
Liu Yi ◽  
Liu Wei ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Power Plants ◽  
Hysteresis Loops ◽  
Reactor Pressure Vessel ◽  
Electromagnetic Properties ◽  
Irradiation Damage ◽  
Indentation Hardness ◽  
Rpv Steel ◽  
Reactor Pressure

The effect of neutron irradiation damage of reactor pressure vessel (RPV) steels is a main failure mode. Accelerated neutron irradiation experiments at 292 °C were conducted on RPV steels, followed by testing of the mechanical, electrical and magnetic properties for both the unirradiated and irradiated steels in a hot laboratory. The results showed that a significant increase in the strength, an obvious decrease in toughness, a corresponding increase in resistivity, and the clockwise turn of the hysteresis loops, resulting in a slight decrease in saturation magnetization when the RPV steel irradiation damage reached 0.0409 dpa; at the same time, the variation rate of the resistivity between the irradiated and unirradiated RPV steels shows good agreement with the variation rates of the mechanical properties parameters, such as nano-indentation hardness, ultimate tensile strength, yield strength at 0.2% offset, upper shelf energy and reference nil ductility transition temperature. Thus, as a complement to destructive mechanical testing, the resistivity variation can be used as a potentially non-destructive evaluation technique for the monitoring of the RPV steel irradiation damage of operational nuclear power plants.

Effects of Neutron Irradiation and Thermal Aging on Mechanical Properties of ODS Ferritic Steels for Advanced Nuclear Systems

2007 ◽  
Vol 561-565 ◽  
pp. 1773-1776 ◽  
Author(s):  
Ryuta Kasada ◽  
Hang Sik Cho ◽  
Naoyuki Okuda ◽  
Akihiko Kimura
Keyword(s):  
Neutron Irradiation ◽  
Thermal Aging ◽  
Charpy Impact ◽  
Oxide Dispersion Strengthened ◽  
Tensile Tests ◽  
Total Elongation ◽  
Ferritic Steels ◽  
Dislocation Loops ◽  
Irradiation Hardening ◽  
Nuclear Systems

Effects of neutron irradiation and thermal aging on the tensile properties and Charpy impact properties of oxide dispersion strengthened (ODS) ferritic steels for advanced nuclear systems were investigated and discussed with the results of microstructural observation. After the neutron irradiation in JMTR, significant hardening after irradiation at 290 and 400 °C as well as thermal aging at 500 °C, while no effect was observed after irradiation at 600 °C. While the irradiation hardening was not accompanied by a reduction of total elongation in tensile tests, Charpy impact energy at room temperature was reduced after the irradiation. The hardening after the irradiation at 400 °C and the aging at 500 °C was probably due to the formation of Cr-rich phases. The irradiation hardening observed in the ODS ferritic steels irradiated at 290 °C was well explained by the formation of dislocation loops.

Effect of neutron irradiation on the microstructure of the stainless steel electroslag weld overlay cladding of nuclear reactor pressure vessels

2013 ◽  
Vol 443 (1-3) ◽  
pp. 266-273 ◽  
Author(s):  
T. Takeuchi ◽  
Y. Kakubo ◽  
Y. Matsukawa ◽  
Y. Nozawa ◽  
Y. Nagai ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Weld Overlay ◽  
Reactor Pressure

Evaluation of Irradiation Embrittlement of the Weld Metals in the Chinese RPV

2020 ◽  
Author(s):  
Yupeng Cao ◽  
Yinbiao He ◽  
Yifeng Huang ◽  
Binxi Wang ◽  
Yan Yu ◽  
...  
Keyword(s):  
Mechanical Properties ◽  
Nuclear Power ◽  
Degradation Mechanism ◽  
Base Material ◽  
Charpy Impact ◽  
Tensile Tests ◽  
Irradiation Damage ◽  
Limiting Factor ◽  
Irradiation Embrittlement ◽  
Weld Metals

Abstract Reactor pressure vessel (RPV) operates under high temperatures and pressures and is exposed to relatively high neutron radiation. RPV is considered to be irreplaceable, which is the most limiting factor for the lifetime of a nuclear power plant. As the most severe ageing degradation mechanism in RPV materials, irradiation embrittlement is a major issue affecting the integrity through the service life of a RPV. Our previous paper (ASME PVP2019-93615[1]) introduced our project for assessment of irradiation embrittlement of the materials for the Chinese RPVs to verify the 60-year design life, in which the specimens made of the RPV base material manufactured in China, the SA-508 Gr.3 Cl.1 forging, and the different types of weld metals were irradiated in the high fluence engineering test reactor (HFETR). The paper analyzed extent irradiation damage of the forging in terms of mechanical properties. As another part of the project, this paper concentrates on the evaluation of the weld metals in the same project. Tensile tests, Charpy impact tests and fracture toughness tests by master curve approach were carried out for the three types of weld metals subjected to different irradiation fluences (2.6E19n/cm2, 8.9E19n/cm2). Comparison of the mechanical properties of the irradiated and the unirradiated materials is made. The irradiation resistance of the weld metals in our project is also compared with the data in the literatures.

scholarly journals IMITATING NEUTRON IRRADIATION DAMAGE USING A COMBINATION OF ION BEAMS

2018 ◽  
Vol 14 ◽  
pp. 42
Author(s):  
Václav Šísl ◽  
Martin Ševeček
Keyword(s):  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Ion Irradiation ◽  
Ion Beams ◽  
Third Party ◽  
Irradiation Damage ◽  
Material Structure ◽  
Monte Carlo Code ◽  
The One ◽  
Induced Defects

There is a strong motivation for using ion beams to imitate neutron irradiation damage, mainly in order to reduce costs and time demands linked to neutron irradiation experiments. The long-term goal of the authors is to create an ion irradiation methodology, which could be employed in the development process of innovative nuclear fuel materials. This methodology will be based on combining of a set of ion beams in such a way that the final distribution of irradiation-induced defects in the material structure is similar to the one which would have been introduced by neutrons in a nuclear reactor. The first part of the methodology is represented by an optimization tool described here. The tool uses a third party Monte Carlo code SRIM to simulate ion transport in a target and to determine the distribution of radiation damage. Subsequently, a custom genetic optimization algorithm is applied to a set of damage distribution profiles to find their optimal combination.

Effects of neutron irradiation on microstructures and hardness of stainless steel weld-overlay cladding of nuclear reactor pressure vessels

2014 ◽  
Vol 449 (1-3) ◽  
pp. 273-276 ◽  
Author(s):  
T. Takeuchi ◽  
Y. Kakubo ◽  
Y. Matsukawa ◽  
Y. Nozawa ◽  
T. Toyama ◽  
...  
Keyword(s):  
Stainless Steel ◽  
Neutron Irradiation ◽  
Nuclear Reactor ◽  
Pressure Vessels ◽  
Steel Weld ◽  
Weld Overlay ◽  
Stainless Steel Weld ◽  
Reactor Pressure

Evolution of Impact Properties of 16MND5 Forgings for Nuclear Reactor Pressure Vessel during Thermal Aging at 500°C

2019 ◽  
Vol 795 ◽  
pp. 54-59
Author(s):  
Rui Si Xing ◽  
Xu Chen ◽  
Dun Ji Yu
Keyword(s):  
Thermal Aging ◽  
Aging Temperature ◽  
Nuclear Reactor ◽  
Charpy Impact ◽  
Impact Properties ◽  
Brittle Transition ◽  
Microstructure Observation ◽  
Thermal Embrittlement ◽  
Ductile To Brittle Transition ◽  
Reactor Pressure

Effects of thermal aging on tensile and Charpy impact properties in 16MND5 steel was investigated, which were aged at 500°C for 0 h, 1000 h, 3000 h, 5000 h. A significant decrease in the yield stress and ultimate tensile strength was observed after thermal aging, while the elongation exhibited a slight decrease follow by an increase aged for 5000 h. What's more, the ductile-to-brittle transition temperature (DBTT) showed a remarkable increase with the prolongation of thermal aging duration. These facts indicate thermal aging caused embrittlement of the steel, which was further investigated by microstructure observation of SEM. The results show cleavage fracture after thermal aging. Furthermore, experimental results at 350°C thermal aging temperature originated from the previous literature were used to analysis the effect of thermal aging temperature. Thus, thermal embrittlement should be taken seriously.

Sign in / Sign up

Export Citation Format

Share Document