Slow strain rate tensile tests on irradiated austenitic stainless steels in simulated light water reactor environments

Chen, Y.; Rao, A. S.; Alexandreanu, B.; Natesan, K.

doi:10.1016/j.nucengdes.2013.08.003

Cited by 13 publications

(8 citation statements)

References 9 publications

Supporting

Mentioning

Contrasting

Order By: Relevance

“…The low cracking susceptibility in the current SSRT tests seems to be consistent with field observations that cracking is reported for internals at much higher doses in PWRs than in BWRs. While IG cracking has been observed extensively in the previous SSRT tests in BWR water, 26,43 brittle fracture areas are less frequently seen in PWR environments under similar test conditions. An example of this difference is given in Fig.…”

Section: Fracture Morphologymentioning

confidence: 80%

“…Mixed-mode TG and cleavage cracking are the most common brittle features identified on the fracture surfaces. While mixed-mode TG cracking is also observed in BWR water tests, 26,43 cleavage fracture is only observed among PWR water tests. Because cleavage cracking represents a form of low-energy fracture that is not normally associated with ductile materials such as austenitic SSs, the observation of cleavage indicates severe embrittlement among these samples.…”

Section: Fracture Morphologymentioning

confidence: 86%

“…For the GBE and non-GBE 690 specimens, the fracture surfaces are almost entirely covered by dimples at both ~5 and 9.6 dpa (Figs. [41][42][43][44]. Spherical and elongated dimples can be seen even at sample surfaces (Figs.…”

Section: Ssrt Tests In Pwr Water On Gbe Alloysmentioning

confidence: 99%

See 2 more Smart Citations

Technical Letter Report of Slow-Strain-Rate Tensile Tests on Irradiated Stainless Steels in a PWR Environment

Chen¹,

Soppet²,

Alexandreanu³

et al. 2013

Self Cite

View full text Add to dashboard Cite

Slow-strain-rate tensile (SSRT) tests have been performed on irradiated specimens in a simulated pressurized water reactor (PWR) environment. The samples are miniature tensile specimens of various austenitic stainless steels (SSs) with different thermal-mechanical treatments commonly used for reactor core internal components. Neutron irradiation was carried out in the BOR-60 reactor, a sodium-cooled, fast breeder reactor in Russia, at ~320°C. The damage doses of the specimens are 5, 10, and 48 dpa (displacements per atom). All irradiated materials show significant irradiation hardening and loss of ductility in the SSRT tests. The yield strengths of cold-worked are higher than that of solution-annealed samples at all doses up to 48 dpa. While the irradiation hardening seems to saturate between ~5 and ~10 dpa, the loss of ductility continues to increase above 10 dpa. Strain softening is also observed for all irradiated materials above 5 dpa. Fractographic examinations show that ductile dimple fracture is the dominant morphology for all SSRT tests in the PWR environment. Small areas of transgranular, mixedmode and cleavage fractures are seen on most fracture surfaces in PWR water tests. Intergranular cracking is also observed in 48-dpa Types 316 and 347 SSs. Cracking susceptibility of the tested materials was evaluated with fracture morphology and time to failure. In general, high-doses cold-worked SSs are more susceptible to transgranular cleavage cracking in the PWR environment. Solution-annealed Type 347 SS is susceptible to intergranular cracking at 45 dpa in the PWR environment.

show abstract

Section: Fracture Morphologymentioning

confidence: 80%

Section: Fracture Morphologymentioning

confidence: 86%

See 1 more Smart Citation

Technical Letter Report of Slow-Strain-Rate Tensile Tests on Irradiated Stainless Steels in a PWR Environment

Chen¹,

Soppet²,

Alexandreanu³

et al. 2013

Self Cite

View full text Add to dashboard Cite

show abstract

“…3 [9,10]. These tensile specimens were irradiated in the high-flux research reactor SM at a temperature close to 300 °C at two different positions.…”

Section: Methodsmentioning

confidence: 99%

TEM Foil Preparation from Irradiated Metallic Materials: A Practical Approach

Namburi¹,

Bublíkova²,

Rosnecky³

et al. 2015

JESE-B

View full text Add to dashboard Cite

This paper focuses on work related to post irradiation examination of 300-series austenitic stainless steel taken from reactor vessel internals of PWR. High neutron irradiation dose in NNP's leads to a degradation of microstructure of the material in a nano-metric scale. Hence, it is important to characterize the irradiated materials to understand the physical basis of the degradation mechanisms. Microstructural characterization of neutron-irradiated materials by TEM requires enhanced sample preparation methodologies, which commonly needs general improvements regarding particular experiment to be performed. In this study, the authors have developed methodology specialized in 1 mm TEM thin foil preparation from a deformed shank of a broken miniaturized tensile specimen. TEM foil size in current studies is smaller than standard because of the small shank diameter and high radioactivity of the studied material. The reduction of the TEM foil radioactivity to minimum is crucial to perform EDX chemical analysis and to increase the EDX detector lifetime. This paper describes whole process from bulk sample handling, including remote-controlled material cutting in shielded hot-cells and disc polishing in glow-boxes, up to the final procedure of electrolytic-polishing of electron transparent 1 mm TEM foils. Eventually, results of TEM microanalysis of radiation-induced defects were present.

show abstract

“…It corresponds to an increased susceptibility to stress corrosion cracking in irradiated materials. This is a complex phenomenon that involves simultaneous action of neutron irradiation, applied stress, and corrosive environment [4]. As pointed out in [4], irradiation hardening and embrittlement are key irradiation effects that have significant impact on IASCC behavior of austenitic stainless steels.…”

Section: Introductionmentioning

confidence: 99%

Full-field polycrystal plasticity simulations of neutron-irradiated austenitic stainless steel: A comparison between FE and FFT-based approaches

Shawish

Vincent

Moulinec

et al. 2020

Journal of Nuclear Materials

View full text Add to dashboard Cite

We compare two full-field approaches-a crystal plasticity finite element method (CP-FEM) and crystal plasticity fast Fourier transform-based (CP-FFT) method-for a specific crystal plasticity law introduced for neutron-irradiated austenitic stainless steel SA304L currently used in nuclear reactor vessel internals. This particular law is employed to identify and quantify possible advantages and drawbacks of the two approaches when used in the large-scale simulations to predict the effect of irradiation damage (e.g., crack initiation) in stainless steel microstructures. A comparison is performed in a polycrystalline context for different periodic Voronoi microstructures deformed under tension. Special emphasis is put on studying the performance of the two approaches in terms of mesh convergence analysis using aggregate models with different spatial discretizations. In the CP-FEM approach, the performance of linear as well as quadratic tetrahedral meshes is investigated. A similar performance between the CP-FEM and CP-FFT methods is demonstrated in a smaller 2-grain aggregate. However, a slower mesh convergence is observed for the CP-FEM method when comparing tensile responses of a larger 100-grain polycrystal. A drop in the convergence rate is much more pronounced on linear than on quadratic tetrahedral meshes. As a consequence, largest (average) grain boundary stresses are shown to be overestimated with the linear mesh CP-FEM approach, thus raising a concern of possible over-conservatism employed in the CP-FEM prediction of crack initiation in irradiated stainless steels. On the contrary, the mesh convergence of the CP-FFT approach is found to be practically independent of the applied macroscopic strain and also aggregate size. Therefore, for such steels, the CP-FFT approach seems to be better justified.

show abstract

Slow strain rate tensile tests on irradiated austenitic stainless steels in simulated light water reactor environments

Cited by 13 publications

References 9 publications

Technical Letter Report of Slow-Strain-Rate Tensile Tests on Irradiated Stainless Steels in a PWR Environment

Technical Letter Report of Slow-Strain-Rate Tensile Tests on Irradiated Stainless Steels in a PWR Environment

TEM Foil Preparation from Irradiated Metallic Materials: A Practical Approach

Full-field polycrystal plasticity simulations of neutron-irradiated austenitic stainless steel: A comparison between FE and FFT-based approaches

Contact Info

Product

Resources

About