Material Performance in Sodium

Li, Meimei; Natesan, K.; Chen, Wei‐Ying

doi:10.1016/b978-0-12-803581-8.11753-9

Cited by 3 publications

(3 citation statements)

References 64 publications

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“…Since the surface oxide layer can be readily reduced in liquid sodium, the primary corrosion concern for a liquid sodium environment is the dissolution of the alloy's metallic elements and an exchange of non-metallic elements between the alloy and the sodium [43]. Depending on the solubility of specific elements in sodium, the alloy's constituents can be dissolved uniformly or preferentially.…”

Section: Sodium-cooled Fast Reactorsmentioning

confidence: 99%

Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors

Jokisaari,

Chen,

Hartmann

et al. 2023

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The Advanced Materials and Manufacturing Technologies (AMMT) program within the Department of Energy, Office of Nuclear Energy (DOE-NE) has developed its current recommendation for its corrosion testing strategy to deploy additively manufactured (AM) materials in advanced nuclear reactors. Additive manufacturing technologies have developed rapidly in recent years, creating new opportunities and challenges for the nuclear industry. To adopt AM technologies, the corrosion performance of AM materials needs to be adequately evaluated.This document is divided into seven chapters designed to support the proposed corrosion research strategy. The strategy focuses on the testing of AM 316H stainless steel (SS) to align with the broader AMMT program. An overview of localized and uniform corrosion mechanisms and their synergy with other degradation modes is provided. Specifically, the corrosion resistance of 316H is largely provided by the presence of an intact chromium oxide film on the surface, and loss or damage of this film is generally responsible for corrosion degradation.Engineering concerns for corrosion behavior are discussed, including component failure mechanisms and corrosion codes and standards. While the standards are extensive, certain material testing standards do not exist that would be relevant for the AMMT program. Reactor-specific corrosion concerns for molten salt reactors, liquid metal-cooled reactors, and high-temperature gas reactors will guide corrosion testing. Key factors likely to influence corrosion properties of AM materials (especially 316H) include build porosity, the presence of unusual phases, residual stress, microsegregation within the cellular substructure, and surface roughness. Post-build treatments and process variability need to be assessed with multiple specimens that are microstructurally quantified to develop structureproperty-performance relationships with uncertainty quantification. Sufficient facilities exist, at least for initial program efforts focusing on characterization of unirradiated material in laboratory settings, across Argonne National Laboratory

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Section: Sodium-cooled Fast Reactorsmentioning

confidence: 99%

Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors

Jokisaari,

Chen,

Hartmann

et al. 2023

View full text Add to dashboard Cite

show abstract

“…Since the surface oxide layer can be readily reduced in liquid sodium, the primary corrosion concern for structural alloys is the dissolution of metallic elements into the liquid sodium and the exchange of non-metallic elements with it [2]. Depending on the solubility in sodium, alloy elements can dissolve uniformly or preferentially into liquid sodium, forming surface corrosion products that can be eroded or spalled into flowing sodium.…”

Section: Introductionmentioning

confidence: 99%

Progress Report of Alloy 709 Performance in Liquid Sodium

Chen,

Zeng,

Momozaki

et al. 2023

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With improved high-temperature properties, Alloy 709 shows great potential for advanced reactor applications. To understand the effects of sodium on A709, small tensile specimens were machined from two commercial heats and subjected to sodium exposure in two forced convection loops operated by Argonne. In parallel, thermal aging experiments were also conducted to isolate any sodium-specific effects. An increasing trend in yield strength and a decrease trend in ultimate tensile strength were observed with increasing exposure time at 600 and 650°C, suggesting microstructural changes at these temperatures. Nonetheless, the thermally aged and sodium exposed samples exhibited similar behaviors, implying no sodium-specific effects at these temperatures. In contrast, 316H SS exposed to sodium at 650°C showed a pronounced sodium effect. The divergent responses of A709 and 316H SS to sodium exposure are noteworthy and necessitate further investigation.A thermodynamic analysis was performed to understand the carburization-decarburization behavior of A709 in sodium environments. The alloy's equilibrium carbon activity was evaluated with M23C6, TiC, and NbC. The carburization-decarburization boundary was calculated with two methods developed previously. Preliminary results suggest that A709 will be carburized in SMT-1 and SMT-2, and decarburized in low-carbon-activity environments such as EBR-II at temperatures exceeding 620-650°C. Nonetheless, A709 should be less likely undergo decarburization compared to 316H SS in identical low-carbon-activity environments.

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“…Depending on the solubility in sodium, alloy elements can be dissolved uniformly or preferentially into liquid sodium, forming surface corrosion products that can subsequently be eroded or spalled into flowing sodium. The metallic element transfer usually establishes a go/no-go type of evaluation of an alloy for use as a structural material in a sodium environment [3]. The mass transfer of nonmetallic elements involves oxygen, carbon, and nitrogen in the alloy-sodium system.…”

Section: Introductionmentioning

confidence: 99%

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

Chen

Zeng

Chen

et al. 2022

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Two heats of G91 steel have been investigated in liquid sodium at 550-650°C to understand their corrosion behaviour, microstructural evolution, and tensile properties. Sodium exposure experiments of subsized, sheet-type tensile specimens of G91 were performed in Argonne's forced convection sodium materials testing loops, SMT-1 and SMT-2. Maximum exposure times of ~49,000 h at ~38,000 h, and ~20,000 h have been achieved at 550, 600, and 650°C, respectively. Thermal aging experiments of G91 steel were also conducted in parallel, and the results were compared with those of the sodium-exposed specimens to isolate the effect of thermal aging from the effect of sodium exposure. It was found that the yield stress and ultimate tensile strength of G91 were reduced by more than 50% after sodium exposures at 650°C due to decarburization. The reduction in tensile strength was accompanied by the dissolution of M23C6 carbides and excessive grain growth. The effect of sodium at 550 and 600°C was insignificant for exposure times up to ~49,000 h and ~38,000 h, respectively. The effect of sodium on the microstructural stability and tensile properties of G91 is largely dependent on the carbon mass transfer in the alloy-sodium system.

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Material Performance in Sodium

Cited by 3 publications

References 64 publications

Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors

Corrosion Testing Needs and Considerations for Additively Manufactured Materials in Nuclear Reactors

Progress Report of Alloy 709 Performance in Liquid Sodium

Effects of Sodium Exposure on the Tensile Properties of Grade 91 steel

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