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.

Toughness and Embrittlement of RPV Steels Using Ultrasonic Measurements

2004 ◽  
Author(s):  
Allen L. Hiser
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Ultrasonic Attenuation ◽  
Reactor Pressure Vessel ◽  
Irradiation Embrittlement ◽  
Nondestructive Determination ◽  
Rpv Steel ◽  
Suitable Parameter ◽  
Reactor Pressure

Neutron 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 summarizes the findings from an effort addressing the use of ultrasonic attenuation as a suitable parameter for nondestructive determination of irradiation embrittlement in RPV steels. The materials examined in this study include one heat of RPV steel that was heat treated to induce changes in its fracture toughness, several heats of RPV steel irradiated to assess neutron embrittlement changes in fracture toughness, and a matrix of RPV steels (in the unirradiated condition) with a range of as-fabricated fracture toughness levels.

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.

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.

On the Failure Mechanisms in Reactor Pressure Vessel with Austenitic Cladding

2015 ◽  
Vol 784 ◽  
pp. 492-499
Author(s):  
Jan Štefan ◽  
Jan Siegl ◽  
Miloš Kytka ◽  
Milan Brumovský
Keyword(s):  
Fracture Toughness ◽  
Crack Propagation ◽  
Neutron Irradiation ◽  
Pressure Vessel ◽  
Failure Mechanisms ◽  
Reactor Pressure Vessel ◽  
Fractographic Analysis ◽  
The Relationship ◽  
Reactor Pressure

The austenitic cladding of the WWER pressure vessel is made from two different layers with different fracture toughness values. Based on the fractographic analysis of the tested specimens in the initial, as well as in the irradiated conditions, it was found that individual failure micromechanisms take place during the crack propagation. The obtained results were used to find the relationship between the failure micromechanism changes and the fracture toughness values, as well as to assess the effect of neutron irradiation on the failure micromechanisms.

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]

Alternative Reference Temperature Based on Master Curve Approach in Japanese Reactor Pressure Vessel Steel

2013 ◽  
Author(s):  
Takatoshi Hirota ◽  
Takashi Hirano ◽  
Kunio Onizawa
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Reference Temperature ◽  
Ferritic Steels ◽  
Pressure Vessel Steel ◽  
Reactor Pressure

Master Curve approach is the effective method to evaluate the fracture toughness of the ferritic steels accurately and statistically. The Japan Electric Association Code JEAC 4216-2011, “Test Method for Determination of Reference Temperature, To, of Ferritic Steels” was published based on the related standard ASTM E 1921-08 and the results of the investigation of the applicability of the Master Curve approach to Japanese reactor pressure vessel (RPV) steels. The reference temperature, To can be determined in accordance with this code in Japan. In this study, using the existing fracture toughness data of Japanese RPV steels including base metals and weld metals, the method for determination of the alternative reference temperature RTTo based on Master Curve reference temperature To was statistically examined, so that RTTo has an equivalent safety margin to the conventional RTNDT. Through the statistical treatment, the alternative reference temperature RTTo was proposed as the following equation; RTTo = To + CMC + 2σTo. This method is applicable to the Japan Electric Association Code JEAC 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components” as an option item.

Mechanical and Microstructural Characterization of Heat-Affected Zone Materials of Reactor Pressure Vessel

2010 ◽  
Author(s):  
Jinya Katsuyama ◽  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
Kunio Onizawa
Keyword(s):  
Mechanical Properties ◽  
Fracture Toughness ◽  
Grain Size ◽  
Base Metal ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Reactor Pressure Vessel ◽  
Irradiation Embrittlement ◽  
Reactor Pressure

In order to monitor the neutron irradiation embrittlement of the reactor pressure vessel (RPV) steels for the safe operation of light-water reactors, surveillance specimens of representative materials, i.e. base metal, weld metal and heat affected zone (HAZ), are installed in the RPV during reactor operation according to the regulation. Among these materials, HAZ specimens exhibit a relatively large scatter in Charpy impact properties because of the microstructural inhomogeneity due to multi-pass welding. ASTM E185 and JSME S NC1 stipulate the exception of HAZ specimens from surveillance test. However, the technical basis on the exception has not been established. Therefore, we have started a research on the irradiation embrittlement in HAZ material of RPV steels. Typical RPV steel plates with different impurity levels and their weldments were fabricated to characterize the microstructures and subsequent mechanical properties of typical HAZ materials. Simulated HAZ materials were also made by applying several heat treatments representative of HAZ. Finite element analysis was conducted to draw maps of distributions of grain size and phase-fraction. Using simulated HAZ materials with different grain size and phase before irradiation, mechanical properties such as hardness, Charpy impact property and fracture toughness were measured in comparison with those of base metals and actual HAZ materials. Through the comparison, it was indicated that mechanical properties such as fracture toughness in some simulated HAZ materials were different from base metal and dependent significantly on the metallurgical structure, particularly phase but prior austenitic grain size. Higher fracture toughness in CGHAZ (Coarse-Grain HAZ) materials compared to base metal is due to mixed structure of martensite and lower-bainite phases. Upper-bainite phase in FGHAZ (Fine-Grain HAZ) and base materials causes lower fracture toughness than CGHAZ materials.

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