The performance of thick B4C coatings on graphite divertor tiles in the DIII-D tokamak

Buzhinskij, O.I.; Barsuk, V.A.; Opimach, I.V.; Otrozhenko, V.G.; Trazhenkov, A.I.; West, W.P.; Brooks, N.H.

doi:10.1016/s0022-3115(96)00053-0

Cited by 13 publications

(5 citation statements)

References 7 publications

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“…The formation of Be-carbide was also observed during PISCES plasma experiments at temperatures >300°C [3,4]. As reported earlier [5,6], differently structured C-films with a varying content of H-isotopes can be observed in tokamaks.…”

Section: Introductionmentioning

confidence: 56%

Study of hydrogen isotope interaction with beryllium and carbon surfaces under their simultaneous exposure to stationary and powerful pulsed plasma

Guseva

Gureev

Danelyan

et al. 2004

Journal of Nuclear Materials

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Section: Introductionmentioning

confidence: 56%

Study of hydrogen isotope interaction with beryllium and carbon surfaces under their simultaneous exposure to stationary and powerful pulsed plasma

Guseva

Gureev

Danelyan

et al. 2004

Journal of Nuclear Materials

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“…RGT samples with a 100 μm crystalline B 4 C coating were tested in the T-10 and T-11M tokamaks as limiters [8][9][10], in the DIII-D tokamak as divertor tiles, together with a DiMES sample located in the divertor region [11][12][13][14]; they were also tested under electron and ion beams in the PLADIS and MK-200 plasma facilities [12]. The energy density of the plasma in the PLADIS plasma gun facility was 1. coating, whereas damage of the coating was found on the MPG-8 graphite samples under similar tests and smaller loads [11].…”

Section: Results Of the Experiments And Their Considerationmentioning

confidence: 99%

Application of RGT graphite in fusion devices

Buzhinskij

Otroshchenko

West

2007

Plasma Devices and Operations

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The application of fine-grained recrystallized RGT graphite modified by titanium is considered. Thermally stabilized fine-grained RGT graphite modified by 10 at.% Ti has a density of 2.1 g cm −3 , a heat conductivity of up to 800 W m −1 K −1 , a low porosity and a small retention of hydrogen isotopes. The paper presents the results of full-scale tests on RGT graphite samples carried out in the T-10 and T-11M tokamaks in Russia, in the DIII-D tokamak in the USA, and in the MK-200 and PLADIS plasma facilities. The tested samples made from RGT graphite were found to be sufficiently stable, in particular, under irradiation for 600 ms by plasma with an electron temperature of 60 eV, an electron density of 2.5 × 10 14 cm −3 and a power of the parallel heat flux of 5 kW cm −2 , which are typical for the working modes and disruptions of plasma discharges in tokamaks.

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“…These fine-grained graphites have a high density of about 1.8 g cm −3 and a thermal conductivity of about 100-140 W m −1 K −1 . The research carried out on mock-ups [4,5] in real conditions of operating tokamaks [6][7][8][9][10][11] reveals the considerable modification of the irradiated surface after exposure (figures 1(a) and (b)). The surface modification essentially depends on the load value, on its duration, on the temperature and, in particular, on the presence of oxygen molecules in the vacuum volume [4].…”

Section: Methodsmentioning

confidence: 99%

Modification of the graphite tile surface in tokamaks

Buzhinskij

Begrambekov²,

Otroshchenko

2005

Plasma Devices and Operations

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The paper presents the results of experimental research on the surface modification of graphite protective elements from different graphite materials of MPG-8, RGT, ATJ, CFC, etc., types exposed in the T-10, T-11M, DIII-D, NSTX and Tore Supra tokamaks, as well as after plasma and ion irradiation in devices simulating wall boundary conditions in tokamaks. It is revealed that, in all cases of exposure, significant modification of the carbon-graphite materials surface occurs as a result of the plasma and ion beam. In addition, homogeneous pyrolitic layers with hexagonal planes oriented parallel to the irradiated surface are produced on the surface of protective elements. Thus, the physicomechanical characteristics of the plasma-facing surface degrade sharply. Application of a protective crystalline boron carbide coating reduces considerably the negative consequences of this phenomenon.

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The performance of thick B4C coatings on graphite divertor tiles in the DIII-D tokamak

Cited by 13 publications

References 7 publications

Study of hydrogen isotope interaction with beryllium and carbon surfaces under their simultaneous exposure to stationary and powerful pulsed plasma

Study of hydrogen isotope interaction with beryllium and carbon surfaces under their simultaneous exposure to stationary and powerful pulsed plasma

Application of RGT graphite in fusion devices

Modification of the graphite tile surface in tokamaks

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