Applicability of Miniature Compact Tension Specimens for Fracture Toughness Evaluation of Highly Neutron Irradiated Reactor Pressure Vessel Steels

2018 ◽  
Vol 140 (5) ◽  
Author(s):  
Yoosung Ha ◽  
Tohru Tobita ◽  
Takuyo Ohtsu ◽  
Hisashi Takamizawa ◽  
Yutaka Nishiyama
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Compact Tension ◽  
Specimen Size ◽  
Reactor Pressure Vessel ◽  
Margin Of Error ◽  
Toughness Evaluation ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure ◽  
Selection Of

The applicability of miniature compact tension (Mini-C(T)) specimens to fracture toughness evaluation of neutron-irradiated reactor pressure vessel (RPV) steels was investigated. Three types of RPV steels neutron-irradiated to a high-fluence region were prepared and manufactured as Mini-C(T) specimens according to Japan Electric Association Code (JEAC) 4216-2015. Through careful selection of the test temperature by considering previously obtained mechanical properties data, valid fracture toughness, and reference temperature (To) was obtained with a relatively small number of specimens. Comparing the fracture toughness and To values determined using other larger specimens with those determined using the Mini-C(T) specimens, To values of both unirradiated and irradiated Mini-C(T) specimens were found to be the acceptable margin of error. The scatter of 1T-equivalent fracture toughness values of both unirradiated and irradiated materials obtained using Mini-C(T) specimens did not differ significantly from the values obtained using larger specimens. The correlation between the Charpy 41 J transition temperature (T41J) and the To values agreed very well with that of the data in the literature, regardless of specimen size and fracture toughness of the materials before irradiation. Based on these findings, it was concluded that Mini-C(T) specimens can be applied to fracture toughness evaluation of neutron-irradiated materials without significant specimen size dependence.

Fracture Toughness Evaluation of Reactor Pressure Vessel Steels by Master Curve Method Using Miniature Compact Tension Specimens

2015 ◽  
Vol 137 (5) ◽  
Author(s):  
Tohru Tobita ◽  
Yutaka Nishiyama ◽  
Takuyo Ohtsu ◽  
Makoto Udagawa ◽  
Jinya Katsuyama ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Master Curve ◽  
Loading Rate ◽  
Compact Tension ◽  
Master Curve Method ◽  
Toughness Evaluation ◽  
Curve Method ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure

We conducted fracture toughness testing on five types of commercially manufactured steel with different ductile-to-brittle transition temperatures. This was performed using specimens of different sizes and shapes, including the precracked Charpy-type (PCCv), 0.4T-CT, 1T-CT, and miniature compact tension specimens (0.16T-CT). Our objective was to investigate the applicability of 0.16T-CT specimens to fracture toughness evaluation by the master curve method for reactor pressure vessel (RPV) steels. The reference temperature (To) values determined from the 0.16T-CT specimens were overall in good agreement with those determined from the 1T-CT specimens. The scatter of the 1T-equivalent fracture toughness values obtained from the 0.16T-CT specimens was equivalent to that obtained from the other larger specimens. Furthermore, we examined the loading rate effect on To for the 0.16T-CT specimens within the quasi-static loading range prescribed by ASTM E1921. The higher loading rate gave rise to a slightly higher To, and this dependency was almost the same for the larger specimens. We suggested an optimum test temperature on the basis of the Charpy transition temperature for determining To using the 0.16T-CT specimens.

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.

Applicability of Miniature C(T) Specimen to Fracture Toughness Evaluation for the Irradiated Japanese Reactor Pressure Vessel Steel

2017 ◽  
Author(s):  
Takuji Sugihara ◽  
Takatoshi Hirota ◽  
Hiroyuki Sakamoto ◽  
Kentaro Yoshimoto ◽  
Kazuya Tsutsumi ◽  
...  
Keyword(s):  
Fracture Toughness ◽  
Weld Metal ◽  
Pressure Vessel ◽  
Charpy Impact ◽  
Reactor Pressure Vessel ◽  
Base Metals ◽  
Toughness Evaluation ◽  
Fracture Toughness Evaluation ◽  
Reactor Pressure ◽  
Mc Method

Fracture toughness evaluations for irradiated reactor pressure vessel (RPV) steels are essential in the structural integrity assessment of RPVs. In Japanese pressurized water reactor (PWR) plants, fracture toughness tests are conducted for irradiated RPV steels through the surveillance tests and fracture toughness data are obtained. Lately, the Master Curve (MC) approach has become the main stream in fracture toughness evaluation. However, there can be the case that the number of fracture toughness data is not enough for the MC method in some Japanese PWR plants because of limited numbers of fracture toughness specimens contained in the surveillance capsules. On the other hand, for the Japanese PWR plants, a surveillance capsule generally contains a lot of Charpy impact specimens which miniature C(T) (Mini-C(T)) specimens with a size of 4×10×9.6mm can be taken from. Therefore, it is planned that additional fracture toughness tests are performed using Mini-C(T) specimens after the Charpy impact tests to obtain sufficient fracture toughness data for the MC method. Applicability of the Mini-C(T) specimen to the MC evaluation has been studied in a series of international round robin test programs coordinated by Central Research Institute of Electric Power Industry (CRIEPI). In these programs and the related studies, it was demonstrated that the reference temperature (To) can be determined by the Mini-C(T) specimens without any specific difficulties for the unirradiated RPV base metals. In addition, CRIEPI has recently reported on the basis of their studies that the fracture toughness tests could be successfully performed on the typical unirradiated RPV weld metal and the valid To can be determined with the data obtained from the weld metal as well as base metals. However, few papers reported applicability of the Mini-C(T) specimen to the MC evaluation for irradiated RPV steels. In this study, fracture toughness tests using Mini-C(T) specimens were conducted on the irradiated Japanese Industrial Standards (JIS) SFVQ1A steel (equivalent to ASME A508 Cl.3 steel). The Mini-C(T) specimens were machined out from some broken halves of Charpy impact specimens used in a surveillance test of an actual Japanese PWR plant by a wire cut electric spark machine followed by fatigue precracking. After the fracture toughness tests, the evaluation was performed on the obtained fracture toughness data according to the MC method. The effect of specimen size on To was studied and applicability of the Mini-C(T) specimen was discussed by comparing the existing results of fracture toughness tests using the 1/2T-C(T) specimens conducted in the surveillance test. In addition, the issues to obtain valid To for irradiated materials were discussed.

The Effect of a Low Constraint Geometry on Measured T0 Values for a Nuclear Reactor Pressure Vessel Ferritic Steel

2018 ◽  
Author(s):  
Geena K. Rait ◽  
Catrin M. Davies ◽  
Stephen J. Garwood
Keyword(s):  
Fracture Toughness ◽  
Pressure Vessel ◽  
Nuclear Reactor ◽  
Compact Tension ◽  
Reactor Pressure Vessel ◽  
Single Edge ◽  
Single Edge Notch ◽  
Edge Notch ◽  
Low Constraint ◽  
Reactor Pressure

Current requirements for assessing the fracture toughness of reactor pressure vessel (RPV) ferritic steels are potentially overly conservative due to the employment of high constraint geometries such as compact tension (C(T)) or single edge notch bend, SEN(B), specimens for material testing. These high constraint conditions are not representative of the actual conditions experienced by the RPV in service. If this conservatism could be reduced, more appropriate predictions for RPV lifetime extension could become a possibility. In this study, a known low constraint geometry, single edge notch tension, SEN(T), has been tested alongside the higher constraint SEN(B) specimen in order to compare measured T0 and fracture toughness values for both cases. Finite element analyses have also been conducted for both geometries in order to measure T-stress and calculate Q values thereby allowing quantification of the level of constraint for both geometries. Eight SEN(B) and eight SEN(T) specimens were tested with dimensions 24 × 254 × 96 mm and 20 × 20 × 200 mm, respectively. Testing was conducted at sub-zero temperatures, as close to the T0 as possible, in accordance with the guidelines presented in ASTM E1921-17a. Contrary to expected behaviour the SEN(T) specimen indicated a higher (less negative) T0 then the SEN(B) specimen. The reason for these results are explored in this paper.

Constraint Effect On Fracture Behavior of Underclad Crack in Reactor Pressure Vessel

2021 ◽  
Author(s):  
Masaki Shimodaira ◽  
Tohru Tobita ◽  
Hisashi Takamizawa ◽  
Jinya Katsuyama ◽  
Satoshi Hanawa
Keyword(s):  
Fracture Toughness ◽  
Surface Crack ◽  
Pressure Vessel ◽  
Structural Integrity ◽  
Compact Tension ◽  
Reactor Pressure Vessel ◽  
Local Approach ◽  
Integrity Assessment ◽  
Plastic Constraint ◽  
Reactor Pressure

Abstract According to JEAC4206-2016, 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 postulated underclad crack (UCC) near the inner surface of the RPV during a pressurized thermal shock event. Previous analytical studies show that the plastic constraint for UCC is lower than that for surface crack. Consequently, the apparent KJc for UCC is expected to be higher than that for surface crack. In this study, we performed three-point bending fracture toughness tests and finite element analyses (FEAs) for RPV steel containing a UCC or a surface crack to quantitatively investigate the effect of cladding on the plastic constraint and subsequent KJc evaluation. From the tests, we found that the apparent KJc for the UCC was considerably higher than that for the surface crack. Such a high KJc could be explained by the lower plastic constraint parameters, such as T-stress and Q-parameter, of the UCC compared with those for the surface crack. Additionally, local approach analysis showed that the KJc for the UCC was significantly higher than the master curve estimated from the fracture toughness tests using compact tension specimens.

Sign in / Sign up

Export Citation Format

Share Document