Effects of Microstructure and Thermal Aging on Neutron Irradiation Embrittlement of 2[fraction one-quarter]Cr-1Mo Steel for a Nuclear Pressure Vessel

2008 ◽  
pp. 98-98-13
Author(s):  
M Suzuki ◽  
K Fukaya ◽  
T Kodaira ◽  
T Oku
Keyword(s):  
Neutron Irradiation ◽  
Thermal Aging ◽  
Pressure Vessel ◽  
Irradiation Embrittlement

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.

Industry Practice for the Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels in Japan

2009 ◽  
Author(s):  
Norimichi Yamashita ◽  
Masanobu Iwasaki ◽  
Koji Dozaki ◽  
Naoki Soneda
Keyword(s):  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Correlation Method ◽  
Surveillance Data ◽  
Surveillance Program ◽  
Irradiation Embrittlement ◽  
Term Operation ◽  
Pressure Vessel Steels ◽  
Reactor Pressure

Neutron irradiation embrittlement of reactor pressure vessel steels (RPVs) is one of the important material ageing issues. In Japan, almost 40 years have past since the first plant started its commercial operation, and several plants are expected to become beyond 40 years old in the near future. Thus, the safe operation based on the appropriate recognition of the neutron irradiation embrittlement is inevitable to ensure the structural integrity of RPVs. The amount of the neutron irradiation embrittlement of RPV steels has been monitored and predicted by the complementally use of surveillance program and embrittlement correlation method. Recent surveillance data suggest some discrepancies between the measurements and predictions of the embrittlement in some old BWR RPV steels with high impurity content. Some discrepancies of PWR RPV surveillance data from the predictions have also been recognized in the embrittlement trend. Although such discrepancies are basically within a scatter band, the increasing necessity of the improvement of the predictive capability of the embrittlement correlation method has been emphasized to be prepared for the future long term operation. Regarding the surveillance program, on the other hand, only one original surveillance capsule, except for the reloaded capsules containing Charpy broken halves, is available in some BWR plants. This situation strongly pushed establishing a new code for a new surveillance program, where the use of the reloading and reconstitution of the tested specimens is specified. The Japan Electric Association Code, JEAC 4201–2007 “Method of Surveillance Tests for Structural Materials of Nuclear Reactors,” was revised in December, 2007, in order to address these issues. A new mechanism-guided embrittlement correlation method was adopted. The surveillance program was modified for the long term operation of nuclear plants by introducing the “long-term surveillance program”, which is to be applied for the operation beyond 40 years. The use of the reloading, re-irradiation and reconstitution of the tested Charpy/fracture toughness specimens is also specified in the new revision. This paper reports the application and practice of the JEAC4201–2007 in terms of the prediction of embrittlement and the use of tested surveillance specimens in Japan.

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.

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