IASCC Susceptibility and Slow Tensile Properties of Highly-irradiated 316 Stainless Steels

Fukuya, Koji; Nakano, Morihito; Fujii, K.; Torimaru, Tadahiko

doi:10.1080/18811248.2004.9715532

Cited by 49 publications

(37 citation statements)

References 14 publications

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“…4b and 4c) have shown that radiation-hardening could be caused by the formation of a high number density of Frank loops and PD clusters [17][18][19][20][21]. However, to ensure that the yield stress evolution is only due to the Frank loops formation, a chemical characterization of the microstructure was performed.…”

Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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Abstract:Even if temperature, pressure and chemistry of the cooling water are not very high and aggressive, materials used in PWRs (Pressurized Water Reactors) are exposed to different degradation mechanisms. One of the main goals of the research programs in this field is to develop physical model of the mechanisms down to the atomic scale. Such approach needs a clear description and understanding of the degradation mechanisms at the same small scale. This paper illustrates the benefits of different microscopies and of their last improvements up to the promising possibilities of monochromated and aberrations corrected TEM/STEM. A specific focus is placed on four different degradation mechanisms observed in austenitic stainless steel: irradiation ageing, corrosion fatigue, stress corrosion cracking and corrosion.

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Section: State Of the Artmentioning

confidence: 99%

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Legras¹,

Volgin²,

Radiguet³

et al. 2017

JMSE-B

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“…These features were very similar to those observed on the fracture surface after low-strain-rate testing in the same water environment. 10) On the side surfaces of the specimens, a small number of subcracks were observed near the fracture surface, as shown in Fig. 3.…”

Section: Resultsmentioning

confidence: 83%

“…This configuration was the same as that used for low-strain-rate tensile tests. 10) IASCC susceptibility, mechanical properties and microstructures of the as-irradiated tube materials were thoroughly examined in the previous studies. [9][10][11] UCL tests were conducted at 320 C in the simulated PWR primary water (500 ppm B, 2 ppm Li, 30 ccH 2 /kgH 2 O).…”

Section: à8mentioning

confidence: 99%

“…10) IASCC susceptibility, mechanical properties and microstructures of the as-irradiated tube materials were thoroughly examined in the previous studies. [9][10][11] UCL tests were conducted at 320 C in the simulated PWR primary water (500 ppm B, 2 ppm Li, 30 ccH 2 /kgH 2 O). Stress was gradually loaded to the specimens at a fast strain rate of 3:3 Â 10 À4 /s, to avoid specimen fracture due to IASCC during dynamic straining before reaching target applied stresses.…”

Section: à8mentioning

confidence: 99%

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IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions

Nishioka¹,

Fukuya²,

Fujii³

et al. 2008

J. Nucl. Sci. Technol.

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Uniaxial constant load tests in 320 C simulated PWR primary water were conducted for cold-worked SUS316 stainless steels irradiated to 30-73 dpa. IASCC failure occurred within $100 h at applied stresses of 470-752 MPa. The time-to-failure decreased with increasing neutron dose and applied stress. A compilation of literature data obtained using ring compression tests indicated that the results of ring compression tests for IASCC initiation were similar to those of uniaxial tensile tests. The border stress for IASCC failure decreased with increasing dose and saturated at $400 MPa at $30 dpa. This behavior was similar to that of radiation-induced material changes, such as hardness or grain boundary segregation.

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“…Irradiated stainless steels are susceptible to intergranular cracking even in an inert gas environment, and irradiation-assisted stress corrosion cracking (IASCC) often takes place in a high-temperature water environment. 3,4) Recent studies suggested that the localized deformation near grain boundaries plays an important role in intergranular cracking and IASCC. 5,6) The spacing and inclination angle of surface steps to the loading direction (hereinafter termed ''step angle'') are representative parameters for characterizing localized plastic deformation.…”

Section: Introductionmentioning

confidence: 99%

Angular Distribution of Slip Steps by Three-Dimensional Polycrystalline Model for Stainless Steel

Kamaya¹,

Fukuya²,

Kitamura

2009

Journal of Nuclear Science and Technology

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Localized deformation plays an important role in the initiation of irradiation-assisted stress corrosion cracking, and it can be characterized by the spacing and inclination of slip steps observed on the surface. In order to examine the contribution of mechanical factors (stress) to slipping, the angular distribution of slip steps was evaluated by using a three-dimensional polycrystalline model generated by Voronoi tessellation. An elastic finite element analysis was performed to derive the resolved shear stress (RSS) acting on each slip system. Random crystallographic orientations assigned to each grain caused the microstructural inhomogeneity of local stress due to anisotropic elastic properties. The angular distribution obtained was compared with the experimental results observed in irradiated and deformed stainless steel. It was shown that mechanical factors dominate the slipping behavior of the irradiated stainless steel and that not only RSS but also normal stress acting on the slip plane affects the crystallographic slip. The angular distribution was also evaluated from the maximum Schmid factor of individual grains, in which the inhomogeneous local stress was not taken into account. The angular distributions obtained using RSS and the Schmid factor were almost the same. This suggests that the inhomogeneity of local stress negligibly affects the angular distribution of slip steps, although the change in the local stress in the polycrystalline material is significant.

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IASCC Susceptibility and Slow Tensile Properties of Highly-irradiated 316 Stainless Steels

Cited by 49 publications

References 14 publications

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

Using Microscopy to Help with the Understanding of Degradation Mechanisms Observed in Materials of Pressurized Water Reactors

IASCC Initiation in Highly Irradiated Stainless Steels under Uniaxial Constant Load Conditions

Angular Distribution of Slip Steps by Three-Dimensional Polycrystalline Model for Stainless Steel

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