Constraint Effect on Fracture Mechanics Evaluation for an Under-Clad Crack in a Reactor Pressure Vessel Steel

2020 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Satoshi Hanawa
Keyword(s):  
Fracture Toughness ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Constraint Effect ◽  
Rpv Steel ◽  
Low Constraint ◽  
Reactor Pressure

Abstract For structural integrity assessment of the reactor pressure vessel (RPV) in JEAC 4206-2016, it is required that the fracture toughness (KJc) be higher than the stress intensity factor at the crack tip of a postulated under-clad crack (UCC) near the inner surface of RPV steel under the pressurized thermal shock event. Previous analytical studies showed a low constraint effect at the crack tip of an UCC, compared with that of a normal surface crack. Such a low constraint effect may increase the apparent KJc. In this study, we performed three-point bending (3PB) fracture toughness tests and finite element analysis (FEA) for RPV steel containing an UCC or a surface crack to quantitatively investigate the effect of cladding on the KJc. The FEAs considering the anisotropic property of the cladding successfully reproduced the load vs. load-line displacement curves obtained from the tests. We found that the apparent KJc for the UCC was considerably higher than that for the surface crack. FEA also showed that the constraint effect for the 3PB test specimen with the UCC was lower than that for the specimen with the surface crack owing to the cladding. Thus, a low constraint effect from an UCC may increase the apparent KJc.

Evaluation on Constraint Effect of Reactor Pressure Vessel Under Pressurized Thermal Shock

2013 ◽  
Author(s):  
Naoki Ogawa ◽  
Kentaro Yoshimoto ◽  
Takatoshi Hirota ◽  
Shohei Sakaguchi ◽  
Toru Oumaya
Keyword(s):  
Fracture Toughness ◽  
Thermal Shock ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Constraint Effect ◽  
Elastic Plastic ◽  
Pressurized Thermal Shock ◽  
Reactor Pressure

In recent years, the integrity of reactor pressure vessel (RPV) under pressurized thermal shock (PTS) accident has become controversial issue since the larger shift of RTNDT in some higher fluence surveillance data raised a concern on RPV integrity. Under PTS condition, the combination of thermal stress due to a temperature gradient and mechanical stress due to internal pressure causes considerable tensile stress inside the wall of RPV. Currently, RPV integrity is assessed by comparing stress intensity factor on a crack tip under PTS condition and a reference toughness curve based on the fracture toughness data of irradiated compact specimens. Since PTS loading is large enough to cause plastic deformation, a crack tip behavior on the inner surface of RPV can be explained by elastic-plastic fracture mechanics using the J-integral. In this study, 3D elastic plastic finite element analyses were performed to assess the crack tip behavior on surface of a RPV under Loss of coolant Accident, which causes one of the most severe PTS condition. In order to quantify the constraint effect on a surface crack, J-Q approach was applied. The constraint effect of a surface crack was compared with a compact specimen and its influence on the fracture toughness was assessed. As a result, the difference of constraint effect was clearly obtained. And it is recommended to consider constraint effects in the evaluation of structural integrity of RPV under PTS.

Effect of Plastic Constraint and Cladding on Semi-Elliptical Shaped Crack in Fracture Toughness Evaluation for a Reactor Pressure Vessel Steel

2021 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Yasuto Nagoshi ◽  
Kai Lu ◽  
Jinya Katsuyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Fracture Toughness Test ◽  
Integrity Assessment ◽  
Plastic Constraint ◽  
Shaped Crack ◽  
Reactor Pressure

Abstract In the structural integrity assessment of a reactor pressure vessel (RPV), the fracture toughness (KJc) should be higher than the stress intensity factor at the crack tip of a semi-elliptical shaped under-clad crack (UCC), which is prescribed in JEAC4206-2016. However, differences in crack depth and existence of cladding between the postulated crack and fracture toughness test specimens would be affected to the plastic constraint state and KJc evaluation. In this study, we performed fracture toughness tests and finite element analyses to investigate the effect of plastic constraint and cladding on the semi-elliptical shaped crack in KJc evaluation. The apparent KJc value evaluated at the deepest point of the crack exceeded 5% fracture probability based on the Master Curve method estimated from C(T) specimens, and the conservativeness of the current integrity assessment method was confirmed. Few initiation sites were observed along the tip of semi-elliptical shaped crack other than the deepest point. The plastic constraint state was also analyzed along the crack tip, and it was found that the plastic constraint at the crack tip near the surface was lower than that for the deepest point. Moreover, it was quantitatively showed that the UCC decreased the plastic constraint. The local approach suggested higher KJc value for the UCC than that for the surface crack, reflecting the low constraint effect for the UCC.

Fracture Toughness Evaluation of Neutron-Irradiated Reactor Pressure Vessel Steel Using Miniature-C(T) Specimens

2017 ◽  
Author(s):  
Yoosung Ha ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Yutaka Nishiyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Reactor Pressure Vessel Steel ◽  
Rpv Steel ◽  
Toughness Evaluation ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure

The applicability of miniature-C(T) (Mini-C(T)) specimens to fracture toughness evaluation was investigated for neutron-irradiated reactor pressure vessel (RPV) steel. By carefully selecting the test temperature, valid fracture toughness and reference temperature (To) were determined successfully with a relatively small number of specimens. The value of To determined using irradiated Mini-C(T) specimens was in good agreement with that determined using irradiated pre-cracked Charpy-type (PCCv) specimens. In addition, the scatter of the 1T-equivalent fracture toughness values obtained using the irradiated Mini-C(T) specimens was not significantly different from that obtained using the irradiated PCCv and other larger unirradiated specimens. The To values determined using Mini-C(T) specimens agree very well with the correlation between the Charpy 41J transition temperature and the To of commercially manufactured RPV steels reported in the past.

Reactor Surveillance Test and Fracture Mechanics Evaluation of Irradiation Embrittlement in Reactor Pressure Vessel Steels

1980 ◽  
Vol 102 (4) ◽  
pp. 317-326
Author(s):  
Hideaki Takahashi ◽  
Kiyoshi Saito ◽  
Tetsuo Shoji ◽  
Kazuhiro Date ◽  
Masahiko Suzuki
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Crack Tip ◽  
Reactor Pressure Vessel ◽  
Evaluation Procedure ◽  
Pressure Vessel Steel ◽  
Irradiation Embrittlement ◽  
Tip Resistance ◽  
Controlled Fracture ◽  
Reactor Pressure

With special reference to a nuclear reactor surveillance test, a new evaluation procedure for the fracture toughness from Charpy Vee-notch data is developed. This procedure utilizes a recrystallization-etch technique to determine the crack tip energy dissipation (Wp) within an intense strain region or the dissipation rate (dWp/da). These two parameters serve to characterize the crack tip resistance to cleavage-controlled fracture initiation. The effects of specimen geometry, strain rate, temperature and notch acuity on the cleavage-controlled fracture toughness transition are explained using a critical value of Wp or dWp/da and a modified rate parameter. A feasibility of the new surveillance test procedure for evaluating the irradiation embrittlement of reactor pressure vessel steel, such as SA533B-1, is here verified experimentally, utilizing the Charpy or small compact tension specimen irradiated in a test reactor.

Fracture Toughness Characterization of Midland Beltline Weld After High Dose of Irradiation

2006 ◽  
Author(s):  
Mikhail A. Sokolov ◽  
Randy K. Nanstad
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Neutron Fluence ◽  
National Laboratory ◽  
Reactor Pressure Vessel ◽  
High Dose ◽  
Pressure Vessel Steel ◽  
Vessel Steel ◽  
Oak Ridge ◽  
Reactor Pressure

The Heavy-Section Steel Irradiation (HSSI) Program at Oak Ridge National Laboratory includes a task to investigate the shape of the fracture toughness master curve for reactor pressure vessel steel highly embrittled as a consequence of irradiation exposure, and to examine the ability of the Charpy 41-J shift to predict the fracture toughness shift. As part of this task, a low upper-shelf WF-70 weld obtained from the beltline region of the Midland Unit 1 reactor pressure vessel was characterized in terms of static initiation and Charpy impact toughness in the unirradiated and irradiated conditions. Irradiation of this weld was performed at the University of Michigan Ford Reactor at 288°C to neutron fluence of 3.4×1019 neutron/cm2 in the HSSI irradiation-anneal-reirradiation facility. This reusable facility allowed the irradiation of either virgin or previously irradiated material in a well-controlled temperature regime, including the ability to perform in-situ annealing. This was the last capsule irradiated in this facility before reactor shut down. Thus, the Midland beltline weld was irradiated within the HSSI Program to three fluences — 0.5×1019; 1.0×1019; and 3.4×1019 neutron/cm2. It was anticipated that it would provide an opportunity to address fracture toughness curve shape and Charpy 41-J shift compatibility issues at different levels of embrittlement, including the highest dose considered to be in the range of the current end of life fluence. It was found that the Charpy 41-J shift practically saturated after neutron fluence of 1.0×1019 neutron/cm2. The transition fracture toughness shift after 3.4×1019 neutron/cm2 was only slightly higher than that after 1.0×1019 neutron/cm2. In all cases, transition fracture toughness shifts were lower than predicted by the Regulatory Guide 1.99, Rev. 2 equation.

Nondestructive SANS Study of Reactor Pressure Vessel Steel after Irradiation

2006 ◽  
Vol 321-323 ◽  
pp. 1667-1670
Author(s):  
Young Soo Han ◽  
Eun Joo Shin ◽  
Baek Seok Seong ◽  
Chang Hee Lee ◽  
Duck Gun Park
Keyword(s):  
Neutron Scattering ◽  
Small Angle ◽  
Pressure Vessel ◽  
Small Angle Neutron Scattering ◽  
Reactor Pressure Vessel ◽  
Pressure Vessel Steel ◽  
Rpv Steel ◽  
Induced Defects ◽  
Irradiation Induced Defects ◽  
Reactor Pressure

The irradiation induced defects of irradiated reactor pressure vessel(RPV) steel were investigated by a small angle neutron scattering. The degradation of the mechanical properties of RPV steels during an irradiation in a nuclear power plant is closely related to the irradiation induced defects. The size of these defects is known to be a few nanometers, and the small angle neutron scattering technique is regarded as the best non destructive technique to characterize the nano sized inhomogeneities in bulk samples. The RPV steel was irradiated in the HANARO reactor in KAERI. The small angle neutron scattering experiments were performed at SANS instrument in the HANARO reactor. Both unirradiated and irradiated RPV steels were measured and the SANS data of both steels were compared. The nano sized irradiation induced defects were quantitatively analyzed by SANS. The type of defects was also analyzed based on the SANS results, and the effect of the chemical composition of the RPV steel on the irradiation induced defects was discussed.

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