Hydrogen permeation from F82H wall of ceramic breeder pebble bed: The effect of surface corrosion

Mukai, Keisuke; Kenjo, Shunsuke; Iwamatsu, Naoto; Bakr, Mahmoud; Chikada, Takumi; Yagi, Juro; Konishi, Satoshi

doi:10.1016/j.ijhydene.2021.11.225

Cited by 11 publications

(5 citation statements)

References 44 publications

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“…The permeation coefficient at 800 • C for NITE SiC f /SiC was found to be K p = 4.2 × 10 −16 mol/s/m/Pa [7] compared to K p = 5.2 × 10 −13 mol/s/m/Pa 0.5 for monolithic CVD-SiC [40]. These values are, however, not directly comparable, because at low hydrogen partial pressure, the permeation flux is dependent on ∼Pa limited by surface reaction, compared to close to ∼Pa 0.5 at high partial pressure, where diffusion is rate controlling [42]. The inflection point is at around 4 kPa.…”

Section: A Sic F /Sicmentioning

confidence: 88%

“…SiC f /SiC cannot be welded, and current designs at Kyoto University use liquid phase sintering (LPS) or a combination of screw joints with reaction sintering (RS) at around 1450 • C. It is noted that known problems, such as tritium permeation and retention in graphite, which is a possible reflector for the SCYLLA blanket (see previously), may be avoided by the high hermeticity of SiC f /SiC. Such problems will be resolved in KF's plans to utilize the experimental setup from [42] to further evaluate SiC f /SiC permeability with LiPb, including with samples that have been irradiated.…”

Section: A Sic F /Sicmentioning

confidence: 99%

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Overview of Kyoto Fusioneering’s SCYLLA© (“Self-Cooled Yuryo Lithium-Lead Advanced”) Blanket for Commercial Fusion Reactors

Pearson¹,

Baus²,

Konishi³

et al. 2022

IEEE Trans. Plasma Sci.

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This article outlines Kyoto Fusioneering's (KF's) initial engineering and development activities for its self-cooled lithium lead-type blanket: Self-Cooled Yuryo Lithium-Lead Advanced (SCYLLA©). We provide details on overall design, including an initial tritium breeding ratio (TBR) assessment via neutronics analysis, as well as the status of SCYLLA©-relevant R&D. This includes silicon carbide composite (SiC f /SiC) manufacturing techniques, tritium extraction, materials compatibility, and heat transfer, which are being explored via collaboration with Kyoto University. Results of previous work in relation to this R&D are presented. Permeability coefficients indicate a promising property of SiC f /SiC tritium hermeticity at high temperatures. Tritium extraction technology via vacuum sieve tray (VST) is shown to be demonstrated at engineering scale. A local TBR of up to 1.4 can be achieved with the SCYLLA© configuration. Fabrication methods for various SiC f /SiC components including the blanket module, heat exchanger, and flow path components are provided. A tritium compatible high-temperature SiC f /SiC heat exchanger is discussed. Commercial viability and reactor adaptability are considered as a theme throughout. Finally, KF's plans to build a facility for demonstration reactor relevant testing of a SCYLLA© prototype in the mid-2020s, which will provide a significant step toward commercial fusion energy, are presented.

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Section: A Sic F /Sicmentioning

confidence: 88%

Section: A Sic F /Sicmentioning

confidence: 99%

Overview of Kyoto Fusioneering’s SCYLLA© (“Self-Cooled Yuryo Lithium-Lead Advanced”) Blanket for Commercial Fusion Reactors

Pearson¹,

Baus²,

Konishi³

et al. 2022

IEEE Trans. Plasma Sci.

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“…Figure 1 a shows the proposed facility layout, including the IECF vacuum vessel. The output radiation from the IECF system is volumetric around the centre of the cathode [ 32 , 33 ]. The core of the radiation source is assembled at the centre of a cubic room of 3 × 3 × 3 m 3 , 60 cm from the floor, and 130 cm from each side with 200 cm from the top.…”

Section: Methodsmentioning

confidence: 99%

Assessment of Five Concrete Types as Candidate Shielding Materials for a Compact Radiation Source Based on the IECF

et al. 2023

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A radiation source based on the inertial electrostatic confinement fusion (IECF) system is being developed for multidisciplinary research applications. The radiation outputs from the IECF system are 2.45 MeV fast neutrons and the associated co-generated X-rays with an energy less than 3 MeV. A radiation shielding study has been performed on five types of concrete to define the most efficient material for the shielding design of the system. The proposed materials were ilmenite-magnetite concrete (IMC), ordinary concrete-1 (OC-1), barite-containing concrete (BC), ordinary concrete-2 (OC-2), and serpentine-containing concrete (SC). A numerical model was applied to determine the effective removal cross-section coefficients (∑Rt) for the fast neutrons and the total mass attenuation coefficients (µm), the half-value layer (HVL), the mean free path (MFP), the effective atomic number (Zeff), and effective electron density (Neff) for photons inside the materials. The model considered the radiation source energy and the material properties of the concrete types. The results revealed that the serpentine-containing concrete exhibited the highest ∑Rt with 12 cm of concrete thickness needed to attenuate an incident neutron flux to 1/100 of its initial value. In addition, the BC shows the highest µm with a 38 cm concrete thickness needed to attenuate the 3 MeV energy X-ray flux to 1/100 of its initial value. This study suggests that a 40 cm thickness of SC or BC adequately shields the radiation generated from an IECF system with a maximum particle production rate of up to 1 × 107 n/s.

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“…The chemical compatibility of RAFM steel with solid breeders at the operation temperatures has been studied [4][5][6][7]. Li 2 TiO 3 statically and chemically reacted with Indian RAFM steel (Fe-9.02Cr-1.46W) [4] and RAFM steel ARAA (Fe-9Cr-1.2W) [5], and formed oxide layers.…”

Section: Introductionmentioning

confidence: 99%

“…Li 2 TiO 3 statically and chemically reacted with Indian RAFM steel (Fe-9.02Cr-1.46W) [4] and RAFM steel ARAA (Fe-9Cr-1.2W) [5], and formed oxide layers. Li 2+x TiO 3+x with Li 2 ZrO 3 also chemically reacted with F82H and formed oxide layers [6].…”

Section: Introductionmentioning

confidence: 99%

Effect of Temperature on Fretting Corrosion Behaviors between Li<sub>2</sub>TiO<sub>3 </sub>Pebble and F82H

MASAKI

Kondo

KIM

et al. 2022

Plasma and Fusion Research

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Lithium titanate (Li 2 TiO 3 ) pebble is a candidate tritium breeder of solid breeder blanket systems of fusion reactors. The oscillation of coolant tubes can be induced by the coolant flow. Fretting corrosion is caused between the Li 2 TiO 3 pebbles and the coolant tubes which are made of reduced activation ferritic martensitic steel F82H. The purpose of the present study is to clarify the fretting behaviors at the temperatures of blanket conditions. The Li 2 TiO 3 pebble produced by sol-gel method was pushed onto the surface of the oscillating F82H plate in the fretting tests which were performed for 10 min in an air atmosphere up to 573 K. The fretting scars of the Li 2 TiO 3 pebble and the F82H plate were observed and analyzed by SEM/EDX and 3D laser scanning microscope. The fretting wear was mitigated at the temperatures of 373 K and 473 K due to the formation of the oxide layer, which might reduce the friction. The pebble was partially destructed by the fretting motion in the test performed at 573 K. The fretting wear of the pebble and the F82H plate was mitigated when the pebble was not fixed on the holder since the pebble could vibrate together with the oscillating plate.

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Hydrogen permeation from F82H wall of ceramic breeder pebble bed: The effect of surface corrosion

Cited by 11 publications

References 44 publications

Overview of Kyoto Fusioneering’s SCYLLA© (“Self-Cooled Yuryo Lithium-Lead Advanced”) Blanket for Commercial Fusion Reactors

Overview of Kyoto Fusioneering’s SCYLLA© (“Self-Cooled Yuryo Lithium-Lead Advanced”) Blanket for Commercial Fusion Reactors

Assessment of Five Concrete Types as Candidate Shielding Materials for a Compact Radiation Source Based on the IECF

Effect of Temperature on Fretting Corrosion Behaviors between Li<sub>2</sub>TiO<sub>3 </sub>Pebble and F82H

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