Experimental Simulation of Neutron Irradiation Damage in Reactor Pressure Vessel Steels

2006 ◽  
Vol 324-325 ◽  
pp. 1189-1192
Author(s):  
Fahim Hashmi ◽  
Su Jun Wu ◽  
Huan Xi Li
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Reactor Pressure Vessel ◽  
Life Assessment ◽  
Experimental Simulation ◽  
Irradiation Damage ◽  
Irradiation Embrittlement ◽  
Surveillance Programs ◽  
Reactor Pressure

Degradation of reactor pressure vessel (RPV) steels due to neutron irradiation embrittlement is directly related to safety and life of the nuclear power plant (NPP). In order to ensure structural integrity and safe operation of NPP, surveillance programs are conducted to monitor and predict the changes in RPV materials. Availability of irradiated specimen from RPV or irradiation of specimens under simulated conditions of RPV for conducting fracture toughness tests remains a major problem in surveillance programs. In order to resolve this problem, various methods are adopted to experimentally simulate the effect of neutron irradiation on mechanical behavior of RPV steels using electron irradiation, thermal aging, strain hardening, combined quenching and hardening and pre-straining combined with heat treatment. This paper presents a review of the existing research on experimental simulation of neutron irradiation damage through various methodologies and discusses the future scope of their application in plant safety and life assessment of RPV’s.

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.

Assessment of Structural Integrity on Irradiated Steel Structure: Focusing on Long Column Type Reactor Pressure Vessel Supports

2019 ◽  
Author(s):  
Goeun Han ◽  
Sukru Guzey
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Element Model ◽  
Reactor Pressure Vessel ◽  
Primary Coolant ◽  
Type Reactor ◽  
Column Type ◽  
Unfavorable Condition ◽  
Reactor Pressure

Abstract The structural steel in a nuclear facility experiences significant degradation due to the accumulated neutron irradiation. Particularly, the long-column type reactor pressure vessel supports have been focused since they resist considerable loading to maintain the primary coolant system in their position and experience high neutron irradiation in low temperature, which is an unfavorable condition for the fracture toughness. This study implemented the API 579-1/ASME FFS-1, fitness-for-service (FFS) method to consider both irradiated mechanical properties and multiple loading cases. A three-dimensional (3D) finite element model of long column type reactor pressure vessel support was built for the linear analysis. The metallurgical properties of reactor pressure vessel support for assessment were estimated by empirical equations. This study provides the structural margin of long-column type reactor pressure vessel support by levels of the loads and levels of the neutron fluence.

Susceptibility to Neutron Irradiation Embrittlement of Heat-Affected Zone of Reactor Pressure Vessel Steels

2019 ◽  
Author(s):  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Yoosung Ha ◽  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
...  
Keyword(s):  
Neutron Irradiation ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Microstructural Analysis ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Irradiation Embrittlement ◽  
Element Analysis ◽  
Thermal Histories ◽  
Reactor Pressure

Abstract The heat-affected zone (HAZ) of reactor pressure vessel (RPV) steels is known to show large scatter in Charpy impact properties because it has inhomogeneous microstructure due to thermal histories of multi-pass welding for butt-welded joints. The correlation between mechanical properties and microstructure such as grain size, phase-fraction, martensite-austenite constituent, on the characteristics of HAZ of un-irradiated materials was investigated. Neutron irradiation was conducted at Japanese Research Reactor −3 (JRR-3) operated by JAEA. The neutron irradiation susceptibility was evaluated based on post-irradiation examinations consisting of mechanical testing and microstructural analysis. In the experiments, typical RPV steel plate and their weldment were prepared. Simulated HAZ materials that have representative microstructures such as coarse-grain HAZ (CGHAZ) and fine-grain HAZ (FGHAZ) were also prepared based on the thermal histories calculated by finite element analysis. For un-irradiated materials, a part of simulated HAZ materials showed a higher reference temperature of the master curve method than that of the base metal (BM). The irradiation hardening of HAZ was almost the same or lower than that of the BM, and the shift of reference temperature for HAZ materials was comparable with that of BM.

Preliminary Results On Ultrasonic Attenuation Detection Of Neutron Irradiation Embrittlement Of Nuclear Reactor Steel

1997 ◽  
Vol 503 ◽  
Author(s):  
A. L. Hiser ◽  
R. E. Green
Keyword(s):  
Fracture Toughness ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Ultrasonic Attenuation ◽  
Reactor Pressure Vessel ◽  
Irradiation Embrittlement ◽  
Preliminary Results ◽  
Nondestructive Determination ◽  
Reactor Pressure

ABSTRACTNeutron bombardment of reactor pressure vessel (RPV) steels causes reductions in fracture toughness in these steels, termed neutron irradiation embrittlement. Currently there are no accepted methods for nondestructive determination of the extent of the irradiation embrittlement nor the actual fracture toughness of the reactor pressure vessel. This paper provides preliminary results of an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels.

Modeling of Irradiation Embrittlement of Reactor Pressure Vessel Steels

1999 ◽  
Vol 122 (1) ◽  
pp. 60-66 ◽  
Author(s):  
S. Murakami ◽  
A. Miyazaki ◽  
M. Mizuno
Keyword(s):  
Yield Stress ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Void Nucleation ◽  
Reactor Pressure Vessel ◽  
Material Behavior ◽  
Uniaxial Tensile ◽  
Irradiation Embrittlement ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

A model to describe the change in the inelastic and fracture properties of reactor pressure vessel steels due to neutron irradiation in the ductile region (i.e., irradiation embrittlement) is developed. First, constitutive equations for unirradiated elastic-viscoplastic-damaged materials are developed within the framework of the irreversible thermodynamics theory. To take into account the effect of hydrostatic pressure on the nucleation and growth of microvoids, properly defined dissipation potential is used. Then, the effect of irradiation on the material behavior is incorporated into the proposed model as a function of neutron fluence Φ by taking into account the interaction between irradiation-induced defects and movable dislocations. As regards the damage strain threshold pD, the mechanism of void nucleation due to pile-up of dislocations at the inclusions in the material is proposed first under unirradiated-condition, and then the effect of irradiation on the mechanism is formulated. In order to demonstrate the validity of this model, it is applied to the case of uniaxial tensile loading of a low alloy steel A533B cl. 1 for the pressure vessel use of light-water reactors at 260°C. The resulting model can describe the increase in yield stress and ultimate tensile strength, the decrease in total elongation and strain hardening, and the strain rate dependence of yield stress due to neutron irradiation. [S0094-4289(00)00901-4]

Study on Applicability of Master Curve Methodology Using Miniature C(T) Specimen to a Reactor Pressure Vessel With Low Upper Shelf Energy

2018 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Masato Oshikiri ◽  
Kazuya Tsutsumi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Master Curve ◽  
Energy Use ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Shelf Energy ◽  
Irradiation Embrittlement ◽  
Reactor Pressure

Miniature compact tension (Mini-C(T)) specimen can be an effective tool by utilizing together with Master Curve (MC) methodology for fracture toughness evaluation of irradiated reactor pressure vessel (RPV) steels. Recently, Mini-C(T) specimen has been incorporated into the Japanese standard test method related to MC methodology, JEAC4216-2015 and several studies were found focusing on applicability of Mini-C(T) specimen to irradiated RPV materials. However, there exist some other issues to be resolved considering application to irradiated materials. One of them is violation against the limitation criteria for ductile crack growth (DCG) specified in the standards. In general, upper shelf energy (USE) of RPV materials tends to decrease as well as shift in Charpy transition temperature due to neutron irradiation embrittlement. It may cause reduction in resistance of material against DCG and this leads to the problem peculiar to low USE materials such that the limitation for DCG might be dominant rather than that for KJclimit. Therefore, it is possible to fail to obtain valid KJc data even within valid temperature range of MC methodology, i.e. −50°C ≤ T-To ≤ 50°C, for low USE materials using Mini-C(T) specimens. In this study, the RPV steel with USE lower than 68J was made simulating reduction of USE due to neutron irradiation. Fracture toughness tests were performed using Mini-C(T) specimens as well as the standard 1T-C(T) specimens. Based on the test results, the validity for DCG limitation was also evaluated for each datum by post-test observation of fracture surface. Additionally, effectiveness of added side grooves and double thickness of specimen was examined as a countermeasure for Mini-C(T) specimen.

Evolution of precipitation in reactor pressure vessel steel welds under neutron irradiation

2017 ◽  
Vol 488 ◽  
pp. 222-230 ◽  
Author(s):  
Kristina Lindgren ◽  
Magnus Boåsen ◽  
Krystyna Stiller ◽  
Pål Efsing ◽  
Mattias Thuvander
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Steel Welds ◽  
Reactor Pressure
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