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.

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

2012 ◽  
Vol 134 (3) ◽  
Author(s):  
Jinya Katsuyama ◽  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
Kunio Onizawa
Keyword(s):  
Mechanical Properties ◽  
Grain Size ◽  
Base Metal ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Postweld Heat Treatment ◽  
Microstructural Characterization ◽  
Reactor Pressure Vessel ◽  
Reactor Pressure

In order to provide the technical basis for the judgment of the needs of surveillance specimens of heat-affected zone (HAZ) in reactor pressure vessel (RPV) steels, we performed a research on the characterization of metallurgical and mechanical properties of the HAZ in RPV steels. The distributions of grain size and phases were drawn as a map based on temperature histories around HAZ obtained from welding simulation. Referring to the HAZ map, typical HAZ materials were made by simulating temperature histories including postweld heat treatment (PWHT) within the HAZ. Metallurgical and mechanical characteristics for those HAZ materials were compared with those of base metal. From the results, it is clear that mechanical properties of HAZ materials depend not on the prior austenitic grain size but mostly on the phases. Concerning on the fracture toughness in HAZ, HAZ materials close to weld metal indicated higher toughness than that of base metal due to mixed structure of martensite and lower-bainite, while HAZ materials close to base metal were equivalent or slightly lower toughness than that of base metal due to the same phase as base metal of upper-bainite.

Fracture Toughness Evaluation of Heat-Affected Zone Under Weld Overlay Cladding in Reactor Pressure Vessel Steel

2018 ◽  
Author(s):  
Yoosung Ha ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Satoshi Hanawa ◽  
Yutaka Nishiyama
Keyword(s):  
Fracture Toughness ◽  
Base Metal ◽  
Heat Affected Zone ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Weld Overlay ◽  
Reactor Pressure Vessel Steel ◽  
Reactor Pressure

An evaluation of the fracture toughness of the heat-affected zone (HAZ), which is located under the weld overlay cladding of a reactor pressure vessel (RPV), was performed. Considering inhomogeneous microstructures of the HAZ, 0.4T-C(T) specimens were manufactured from the cladding strips locations, and Mini-C(T) specimens were fabricated from the distanced location as well as under the cladding. The reference temperature (To) of specimens that were aligned with the middle section of a cladding strip (HAZMCS) was ∼12°C higher than that of specimens that were aligned with cladding strips at the overlap (HAZOCS). To values of partial area in the HAZ were obtained using Mini-C(T) specimen. The To values obtained near the side of the cladding were ∼13°C higher than those away from the cladding. To values of HAZ for both 0.4T-C(T) and Mini-C(T) specimens were significantly lower than that of the base metal at a quarter thickness by 40°C–60°C. Compared to the literature data that indicated fracture toughness at the surface without overlay cladding and base metal of a quarter thickness in a pressure vessel plate, this study concluded that the welding thermal history showed no significant effect on the fracture toughness of the inner surface of RPV steel.

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.

Applicability of Fracture Toughness Curves Developed for Japanese Pressure Vessel Steels to Structural Integrity Evaluation

2015 ◽  
Author(s):  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Charpy Impact ◽  
Compact Tension ◽  
Transition Curve ◽  
Pressurized Water Reactor ◽  
Irradiation Embrittlement ◽  
Pressurized Water ◽  
Reactor Pressure

Surveillance tests have been conducted on Japanese Pressurized Water Reactor (PWR) plants for more than 40 years to monitor irradiation embrittlement of reactor pressure vessel (RPV) beltline materials. Fracture toughness specimens are contained as well as tensile and Charpy impact specimens in a surveillance capsule and utilized for structural integrity evaluation. Therefore, a lot of fracture toughness data have been obtained by fracture toughness tests using such as Compact Tension (CT) and Wedge Opening Loading (WOL) specimens. More than one thousand data have been accumulated for both unirradiated and irradiated materials until 2013. Additionally, in terms of fracture toughness, Master Curve (MC) concept has been widely used for fracture toughness transition curve expression of ferritic steels. Considering such a situation, the new fracture toughness curves using Tr30, which denotes Charpy V-notch 30ft-lb transition temperature, as an indexing parameter were developed based on MC concept depending on product form for Japanese RPV steels in 2014. In this study, applicability of the newly developed curves of Japanese RPV steels to structural integrity evaluation is investigated. Especially, this paper focused on conservatism of the curves and the adequate margin to be added in evaluation of RPV integrity employing statistical methodology.

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.

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.

Evaluation of Brittle Crack Arrest Toughness for Highly-Irradiated Reactor Pressure Vessel Steels

2021 ◽  
Author(s):  
Keiko Iwata ◽  
Kuniki Hata ◽  
Tohru Tobita ◽  
Takatoshi Hirota ◽  
Hisashi Takamizawa ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Impact Test ◽  
Charpy Impact ◽  
Crack Arrest ◽  
Charpy Impact Test ◽  
Reactor Pressure Vessel ◽  
Reference Temperature ◽  
Arrest Toughness ◽  
Reactor Pressure

Abstract The crack arrest fracture toughness, KIa, values for highly-irradiated reactor pressure vessel (RPV) steels are estimated according to the linear relationship between crack arrest toughness reference temperature, TKIa, and the temperature corresponding to a fixed arrest load, equal to 4 kN, TFa4kN, obtained by instrumented Charpy impact test. The relationship between TKIa derived from the instrumented Charpy impact test and fracture toughness reference temperature, To, was expressed as an equation proposed in a previous study. The coefficients in the equation could be fine-tuned to obtain a better fitting curve using the present experimental data and previous KIa data. The KIa curve for RPV;A533B class 1 steels irradiated up to 1.3 × 1020 n/cm2 (E > 1 MeV) was compared with a KIR curve defined in JEAC4206-2016. The KIR curve was always lower than the 1%ile curve of KIa for these irradiated RPV steels. This result indicates that the conservatism of the method defined in JEAC4206-2016 to evaluate KIa using the KIR curve is confirmed for highly-irradiated RPV steels.

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