Effect of Pressure on the Corrosion of Materials in High Temperature Water

Li, W.; Woo, O.T.; Guzonas, David; Li, J.; Huang, Xiao; Sanchez, Rainier Garcia; Bibby, C

doi:10.1007/978-3-319-48191-3_12

Cited by 4 publications

(3 citation statements)

References 9 publications

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“…However, higher weight gain and excessive oxidation were found on A286 samples exposed to high-pressure steam with lower pressure than SCW of the same temperature. Similar results were also seen in our earlier study of 310S stainless steel under these 2 conditions [27]. The lower oxidation rate of A286 under the SCW condition is believed to be associated with the formation of a continuous Cr 2 O 3 layer adjacent to the substrate metal, delaying further oxidation [22].…”

Section: Discussionsupporting

confidence: 87%

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

Huang¹,

Sanchez²

2016

CNL Nuclear Review

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Section: Discussionsupporting

confidence: 87%

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

Huang¹,

Sanchez²

2016

CNL Nuclear Review

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“…In our earlier FIBS/TEM study of oxide formation on 310S stainless steel after exposure to steam at 650°C, a spinel in the form of Fe(Cr) 2 O 4 had a lattice structure of a = b = c = 0.851 nm and α = β = γ = 90°. 40 In a study of IN625 in SCW, both NiCr 2 O 4 (inner layer) and NiFe 2 O 4 (outer layer) were reported after 1000 hours in SCW at 500°C. 27…”

Section: Discussionmentioning

confidence: 99%

Corrosion behaviour of IN625 in superheated steam at 800°C

Bsat

Huang

2016

Corrosion Engineering, Science and Technology

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Concerns with greenhouse gas emissions and the uncertainty of long-term supply of fossil fuels have resulted in renewed interest in nuclear energy as an essential part of the energy mix for the future. Canada is currently undertaking the design of Generation IV supercritical water-cooled reactor (SCWR) with improved higher thermodynamic efficiency and considerable plant simplification. The identification of appropriate materials for in-core and out-of-core components to contain the SCWR coolant is one of the major challenges for the design of Canadian SCWR. This study is carried out to evaluate the oxidation/corrosion behaviours of Inconel 625 (IN625) under superheated steam (SHS) at a temperature of 800°C (0.1 MPa) for up to 3000 hours. The results show that chromium-based oxide forms on the surface after exposure in SHS for 800, 2000 and 3000 hours. The oxide formation leads to initial weight increase followed by weight loss after extended exposure, likely due to chromium oxide dissolution. No localised oxide spallation was observed on all samples examined. Despite the slight weight reduction, a dense and adhered oxide layer, consisting primarily of Cr 2 O 3 with some spinel (Ni(Cr,Fe) 2 O 4 ), remained on IN625 after 3000 hours in SHS.

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“…Some of the candidate materials that can be used are stainless steels and high nickel-chromium alloys. The corrosion resistance of stainless steels to oxidation in SCW environment was recently studied [1][2][3][4][5][6][7][8]. These alloys exhibit the lowest weight gains among most of the candidate alloys that are under investigation [9,10].…”

Section: Introductionmentioning

confidence: 99%

Investigation of Oxidation Behavior of 304L and 310S Steels with Potential Application in Supercritical Water-cooled Nuclear Reactors

Galan¹,

Ducu²,

Fulger³

et al. 2020

Rev. Chim.

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Selecting proper candidate materials is one key issue for the development of the supercritical water-cooled nuclear reactor (SCWR). Designing or choosing the most fitting materials means better sustainability, economics and safety. As the supercritical water is a very aggressive corrosive media, corrosion becomes one challenging problem for the materials used in the SCWR. This paper involves the corrosion testing of two stainless steels (304L and 310S) and microstructure evaluation of samples after being exposed to supercritical water. The test parameters were set at the temperature of 550oC and the pressure of 25 MPa for up to 63 days. The samples were investigated using gravimetric corrosion test, optical microscopy, scanning electron microscopy coupled with energy dispersive X-ray spectroscopy. Results showed a lower corrosion performance, in terms of weight change and surface oxide formation, for 304L due to its low chromium content. 310S has excellent corrosion resistance because the chromium content is higher. The results obtained will be useful in future research of test protocol and development of alloys that could be used as reactor fuel cladding and other components in the SCWR.

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Effect of Pressure on the Corrosion of Materials in High Temperature Water

Cited by 4 publications

References 9 publications

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

Effect of Water or Steam Pressure on the Oxidation Behaviour of Alloy 625 and A286 at 625 °c

Corrosion behaviour of IN625 in superheated steam at 800°C

Investigation of Oxidation Behavior of 304L and 310S Steels with Potential Application in Supercritical Water-cooled Nuclear Reactors

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