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.

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.

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.

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.

Master Curve Approach for Some Japanese Reactor Pressure Vessel Steels

2008 ◽  
Author(s):  
Minoru Tomimatsu ◽  
Takashi Hirano ◽  
Seiji Asada ◽  
Ryoichi Saeki ◽  
Naoki Miura ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Power Plant ◽  
Pressure Vessel ◽  
Nuclear Power ◽  
Master Curve ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steels ◽  
The World ◽  
Reactor Pressure ◽  
Plant Components

The Master Curve Approach for assessing fracture toughness of reactor pressure vessel (RPV) steels has been accepted throughout the world. The Master Curve Approach using fracture toughness data obtained from RPV steels in Japan has been investigated in order to incorporate this approach into the Japanese Electric Association (JEA) Code 4206, “Method of Verification Tests of the Fracture Toughness for Nuclear Power Plant Components”. This paper presents the applicability of the Master Curve Approach for Japanese RPV steels.

Evaluation of Fracture Toughness of a SA580 Gr. 3 Reactor Pressure Vessel Using a Bimodal Master Curve Approach

2019 ◽  
Vol 141 (6) ◽  
Author(s):  
Jong-Min Kim ◽  
Seok-Min Hong ◽  
Min-Chul Kim ◽  
Bong-Sang Lee
Keyword(s):  
Fracture Toughness ◽  
Cooling Rate ◽  
Pressure Vessel ◽  
Master Curve ◽  
Random Parameter ◽  
Random Inhomogeneity ◽  
Reactor Pressure Vessel ◽  
Inhomogeneous Materials ◽  
Material Inhomogeneity ◽  
Reactor Pressure

Abstract The standard master curve (MC) approach has a major limitation in that it is only applicable to homogeneous datasets. In nature, steels are macroscopically inhomogeneous. Reactor pressure vessel (RPV) steel has different fracture toughness with varying distance from the inner surface of the wall due to the higher cooling rate at the surface (deterministic material inhomogeneity). On the other hand, the T0 value itself behaves like a random parameter when the datasets have large scatter because the datasets are for several different materials (random inhomogeneity). In this paper, four regions, the surface, 1/8 T, 1/4 T, and 1/2 T, were considered for fracture toughness specimens of Korean Standard Nuclear Plant (KSNP) SA508 Gr. 3 steel to provide information on deterministic material inhomogeneity and random inhomogeneity effects. Fracture toughness tests were carried out for the four regions at three test temperatures in the transition region and the microstructure of each region was analyzed. The amount of upper bainite increased toward the center, which has a lower cooling rate; therefore, the center has lower fracture toughness than the surface so reference temperature (T0) is higher. The fracture toughness was evaluated using the bimodal master curve (BMC) approach. The results of the BMC analyses were compared with those obtained via a conventional master curve analyses. The results indicate that the bimodal master approach considering inhomogeneous materials provides a better description of scatter in the fracture toughness data than a conventional master curve analysis does.

Crack Arrest Test Results of Unirradiated and Irradiated German RPV Steels

2013 ◽  
Author(s):  
Florian Obermeier ◽  
Julia Barthelmes ◽  
Elisabeth Keim ◽  
Hieronymus Hein ◽  
Hilmar Schnabel ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Lower Bound ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Crack Arrest ◽  
Reactor Pressure Vessel ◽  
Test Method ◽  
Fracture Toughness Test ◽  
Test Results ◽  
Reactor Pressure

In the CARISMA[1] and CARINA[2] projects comprehensive tensile, Charpy-impact and fracture toughness tests were performed for unirradiated and irradiated original reactor pressure vessel (RPV) steel specimens from German pressurized water reactors (PWR) up to neutron fluences in the range of 60 operational years and beyond. In addition, crack arrest fracture toughness tests were performed to demonstrate the crack arrest behavior of the materials. To determine the crack arrest properties of ferritic steels, the designated test method according to ASTM E1221 [3] was used. However, in particular for irradiated reactor pressure vessel materials with higher irradiation embrittlement, the prescribed standard test specimen does not always provide adequate test results. During starter notch preparation annealing effects occurred in the heat affected zone (HAZ) of the brittle weld of the starter notch causing crack arrest in the HAZ after unstable crack initiation. Therefore a modified test method to perform crack arrest tests with so called duplex specimens was investigated. In this paper this modified method and the test results of five base and four weld metals with a fluence up to 4,69E+19 cm−2 (E >1 MeV) are discussed. The available test results show that the duplex specimen is an appropriate alternative to the standard compact crack arrest (CCA) specimen. The measured KIa fracture toughness data are enveloped by the “lower bound” of the ASME KIa-curve indexed with RTNDTj or TKIa but not all data are enveloped by indexing the “lower bound” curve with RTT0 like described in the ASME Code Case N-629 [4]. Furthermore correlations of the crack arrest test results with Charpy-impact and fracture toughness test results will be shown.

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.

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