Sodium-cooled fast reactor

Ohshima, Hiroyuki; Kubo, S.

doi:10.1016/b978-0-08-100149-3.00005-7

Cited by 26 publications

(9 citation statements)

References 26 publications

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“…Liquid sodium has been utilized as an HTF for thermal energy transport for many decades [43], [43], and molten sodium is known for its good compatibility with most structural alloys, stainless steels in particular, up to around 650°C (see Figure 42,left). In hot liquid sodium systems, satisfactory performance of structural alloys depends in part on the alloy compatibility (and species present) with sodium and the phase (i.e., thermal) stability of the alloy [44].…”

Section: Alloy Compatibility With Sodium -Literature Reviewmentioning

confidence: 99%

CSP Gen3: Liquid-Phase Pathway to SunShot

Turchi

Gage

Martinek

et al. 2021

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Section: Alloy Compatibility With Sodium -Literature Reviewmentioning

confidence: 99%

CSP Gen3: Liquid-Phase Pathway to SunShot

Turchi

Gage

Martinek

et al. 2021

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“…The US Navy briefly operated a sodium-cooled submarine propulsion reactor, then switched the ship back to a PWR after about two years of operation. Sodium is of general industrial interest as a heat transfer and storage medium, including for concentrated solar power [Oshima 2016].…”

Section: 22mentioning

confidence: 99%

Initial Evaluation of Fuel-Reactor Concepts for Advanced LEU Fuel Development

Mariani

Nelson

Blomquist

et al. 2020

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The National Nuclear Security Administration (NNSA) Office of Defense Nuclear Nonproliferation R&D has commenced a project to evaluate and develop low-enriched uranium (LEU) fuel systems, materials, and fabrication technologies that could be used for compact, advanced LEU (aLEU) fuel, power-producing, high-burnup reactor applications. Potential applications are intended to be off-grid power sources for ship-borne systems and transportable systems to fixed locations, e.g., to a remote base or to a seawater desalination service location. Reactors with these qualities and fuel enrichments between 5 and 20% are atypical. The aLEU Fuel for Nonproliferation Applications Project addresses this technology gap.In the initial stages, the team is assessing key fuel performance criteria, reactor performance criteria, and selection criteria. The project team will identify and leverage the relevant aspects of the mature knowledge base for deployed reactors, both thermal and fast reactor systems, and innovative design features that have been proposed and evaluated to varying degrees but not produced. This project assumes a uranium fuel system at an enrichment of 19.75%. The team will not consider alternate fissile or fertile fuel options. The target power rating is 350 MWt, or approximately 100 MWe with rapid load following capability from 0 to 100% power with an average power rating of approximately 20 MWe. The project assumes a single-batch core lifetime of 30 years. Fuel materials can be monolithic or dispersion, metallic, or ceramic, in rod, plate, or particle bed configurations.Heat transfer from the fuel to the working fluid is necessary for power generation. Fuel geometry (e.g., plate, rod, other) and coolant dictate the heat transfer parameters. Limitations on core size and thermal mass make gas-cooled systems more challenging than liquid cooled. The team considered water and liquid metals first, since molten salt systems would require extensive development and demonstration of corrosion control. The linkage of coolant and reactor types affects not only performance metrics but also drive numerous requirements of the fuel-cladding system. Neutron moderation, its presence or absence, is another significant consideration. Companion reactor analyses depend on, as well as, inform the material studies of the aLEU fuel project.To first order, reactor cooling limits the core power density. Uranium loading and fission controls limit the core life. For a given power rating, coolant, and fuel geometry, a more compact core is possible with higher uranium densities. Recognizing the relevance of uranium density, the project will consider only fuels with a density equal to or greater than UO 2, such as U-xMo alloys and UC, for further evaluation. The project is considering fuels deployed as monoliths (e.g., pellets), dispersion, or as inert matrix. The team will survey fuel systems for applicability with respect to power production, dimensional stability, technology readiness level as well as other criteria. However, demonstration of ...

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“…A challenge for developing advanced nuclear systems, such as fusion reactors, Generation IV fission reactors, and accelerator-driven spallation (ADS) devices, is the deep understanding of the complicated irradiation damage mechanisms in irradiation resistant materials [1,2,3,4]. Reduced-activation ferritic/martensitic (RAFM) steels are considered as prime candidate structural materials for advanced nuclear systems, owing to their excellent mechanical properties, microstructural stability, and irradiation resistance [5,6,7].…”

Section: Introductionmentioning

confidence: 99%

Evolution of Dislocation Loops Induced by Different Hydrogen Irradiation Conditions in Reduced-Activation Martensitic Steel

et al. 2018

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Hydrogen can be induced in various ways into reduced-activation ferritic/martensitic (RAFM) steels when they are used as structural materials for advanced nuclear systems. However, because of the fast diffusion of hydrogen in metals, the effect of hydrogen on the evolution of irradiation-induced defects was almost neglected. In the present work, the effect of hydrogen on the evolution of dislocation loops was investigated using a transmission electron microscope. Specimens of reduced-activation ferritic/martensitic (RAFM) steels were irradiated with hydrogen ions to 5 × 1020 H+ • m−2 at 523–823 K, and to 1 × 1020 H+ • m−2 − 5 × 1020 H+ • m−2 at 723 K. The experimental results reveal that there is an optimum temperature for dislocation loop growth, which is ~723 K, and it is greater than the reported values for neutron irradiations. Surprisingly, the sizes of the loops produced by hydrogen ions, namely, 93 nm and 286 nm for the mean and maximum value, respectively, at the peak dose of 0.16 dpa under 723 K, are much larger than that produced by neutrons and heavy ions at the same damage level and temperature. The results indicate that hydrogen could enhance the growth of loops. Moreover, 47.3% 12 a0 <111> and 52.7% a0 <100> loops were observed at 523 K, but 12 a0 <111> loops disappeared and only a0 <100> loops existed above 623 K. Compared with the neutron and ion irradiations, the presence of hydrogen promoted the formation of a0 <100> loops.

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Sodium-cooled fast reactor

Cited by 26 publications

References 26 publications

CSP Gen3: Liquid-Phase Pathway to SunShot

CSP Gen3: Liquid-Phase Pathway to SunShot

Initial Evaluation of Fuel-Reactor Concepts for Advanced LEU Fuel Development

Evolution of Dislocation Loops Induced by Different Hydrogen Irradiation Conditions in Reduced-Activation Martensitic Steel

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