ITER R&amp;D: Vacuum Vessel and In-Vessel Components: Shield Blanket Module

Dänner, W.; Cardella, A.; Ioki, K.; Mattas, R.F.; Ohara, Y.; Strebkov, Yu.S.

doi:10.1016/s0920-3796(01)00210-1

Cited by 5 publications

(3 citation statements)

References 19 publications

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“…Graphite heaters are used in the FIWATKA facility 84) in Karlsruhe (Germany) and in the CEF 1-2 thermal hydraulic facility in Brasimone (Italy). Both facilities can achieve a heat load of 0.8 MW/ m 2 , 85) whereas, this value represents the absorbed heat load in the FIWATKA facility (water calorimetry) and the incident heat load at CEF 1-2. 86) For temperature and stress evaluation FE-calculations are performed because the possibility of direct diagnostics is not given.…”

Section: Static Heat Loads By Infrared Heatermentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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The current ITER design employs beryllium, carbon fiber reinforced composite and tungsten as plasma facing materials. Since these materials are exposed to high heat fluxes during the operation, it is essential to perform high heat flux tests for R&D of ITER components. Static heat loads corresponding to cycling loads during normal operation, are estimated to be up to 20 MW/m 2 in the divertor targets and around 0.5 MW/m 2 at the first wall in ITER. For the static high heat flux testing, tests in electron beam facilities, particle beam facilities, IR heater and in-pile tests have been performed. Another type, more critical heat loads, which have high power densities and short durations, corresponding to transient events, i.e. plasma disruption, vertical displacement events (VDEs) and edge localized modes (ELMs) deliver considerable heat flux onto the plasma facing materials. For this purpose, tests in electron beam (short pulses), plasma gun and high power laser facilities have been carried out. The present work summarizes the features of these facilities and recent experimental results as well as the current selection of ITER plasma facing components.

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Section: Static Heat Loads By Infrared Heatermentioning

confidence: 99%

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

2005

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“…According to Behrisch et al [13], the first walls were irradiated with the charge-exchange neutrals and ions with energies of $600 and $300 eV (3kT e + 2kT i ), respectively. The total flux of these particles was on the order of 10 19 -10 20 atoms/m 2 s. During operation, since blanket modules are cooled by water at a temperature of 373-423 K [14], the temperature of the first wall would be higher than that of the cooling water. Therefore, the experimental conditions match well with the ITER-FEAT first wall conditions.…”

Section: Methodsmentioning

confidence: 99%

Hydrogen behavior in damaged tungsten by high-energy ion irradiation

Fukumoto

Kashiwagi

Ohtsuka

et al. 2009

Journal of Nuclear Materials

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“…The heat deposition will not be uniform and reactor components need to withstand localized heat fluxes that may be several times higher than the average value. Although high heat flux components [12] for special situations like divertors are being designed to withstand 10-20 MW m −2 by making the material between the plasma and the coolant thinner, their lifetime may be reduced. It is important to bear in mind that one of the potential advantages claimed for D- 3 He fuels is that first walls and structures should last the entire reactor lifetime.…”

Section: Power Flux To the Wallmentioning

confidence: 99%

The feasibility of using D–³He and D–D fusion fuels

Stott¹

2005

Plasma Phys. Control. Fusion

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A careful study of the conditions required to burn D-3 He and D-D fuels in a fusion reactor, with realistic models for bremsstrahling and synchrotron radiation losses, shows that the low reactivity of D-3 He and D-D fusion reactions severely restricts the choice of fuel mixtures that can be brought to ignition and requires very low levels of impurities and alpha particle ash, with plasma temperature, density, beta and energy confinement time that are far beyond the capability of any known magnetic confinement system. The fuel mixtures of D-3 He and D-D that can be brought to ignition produce large fluxes of neutrons and there are serious problems with fuel cycles and reserves.

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ITER R&D: Vacuum Vessel and In-Vessel Components: Shield Blanket Module

Cited by 5 publications

References 19 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Hydrogen behavior in damaged tungsten by high-energy ion irradiation

The feasibility of using D–³He and D–D fusion fuels

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ITER R&D: Vacuum Vessel and In-Vessel Components: Shield Blanket Module

Cited by 5 publications

References 19 publications

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

ITER Relevant High Heat Flux Testing on Plasma Facing Surfaces

Hydrogen behavior in damaged tungsten by high-energy ion irradiation

The feasibility of using D–3He and D–D fusion fuels

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Product

Resources

About

The feasibility of using D–³He and D–D fusion fuels