Deuterium release measurements in the Be phase of JET and determination of tritium content in the exhaust gas

Sartori, R.; Saibene, G.; Goodall, Dominic; Usselmann, E.; Coad, P.; Holland, D.F.

doi:10.1016/0022-3115(90)90117-6

Cited by 40 publications

(26 citation statements)

References 2 publications

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“…The contribution of C from the divertor has been assessed [73] and the total amount of gas required to increase the plasma density is about three times higher than in the carbon-dominated machine but the amount of gas recovered after the pulse is also increased by the same factor. This is similar behaviour to that already observed with the Be limiter in JET [21]. However, although the carbon concentration in the SOL is in the range of only $0.5-1%, the retention in the inner leg region is still dominated by carbon co-deposition, whilst the D retention in the outer leg is dominated by W trapping [72].…”

Section: Retention In Tokamaks With Metallic Wallssupporting

confidence: 82%

“…This is the case for both limiter and divertor machines [17][18][19][20]. This behaviour is enhanced with beryllium, as demonstrated in JET with carbon and beryllium limiters [21] and also in AUG with the full tungsten coverage [22]. Similar results are also observed with the wall conditioning based on the D recovery by He GDC after the discharge, as shown in AUG (45% and 70% of W coverage) [23] and in TS [24][25][26][27].…”

Section: Short and Long-term Retentionmentioning

confidence: 86%

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Fuel retention in tokamaks

Loarer

2009

Journal of Nuclear Materials

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Section: Retention In Tokamaks With Metallic Wallssupporting

confidence: 82%

Section: Short and Long-term Retentionmentioning

confidence: 86%

Fuel retention in tokamaks

Loarer

2009

Journal of Nuclear Materials

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“…They showed, essentially, no long-term retention within the measurement accuracy of about 10%. This is much lower than the 40% retention obtained in L-Modes in the limiter configuration (1987)(1988) in JET [91].…”

Section: Tritium Retentioncontrasting

confidence: 53%

Inverse Theorems for Multipoint Padé Approximants

Chin'¹

1992

Math. USSR Sb.

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High density and high confinement operation in ELMy H-mode is confirmed at or above the normalized parameters foreseen for the ITER operating point (H 98(y,2) ∼ 1, n/n GW ∼ 1, β N > 1.8 at q 95 ∼ 3). The scaling of the ELMy H-mode with β N could be more favourable than that predicted by the IPB98(y,2) scaling. In ELMy H-mode, ion cyclotron current drive (ICCD) control of large sawteeth stabilized by fast particle has been demonstrated and the underlying neo-classical tearing modes (NTMs) and sawtooth physics is being refined. At high-density, Type I ELMy H-modes show trends that would lead to marginally acceptable ELMs on ITER. Type II ELM regime has been produced, though under very restrictive conditions. Type III ELMy operation with radiation fractions up to 95% has been demonstrated by seeding of N 2 in H-modes and could extrapolate to Q = 10 ITER operation, albeit at high current (17 MA). The mitigation of Type I ELMs, nevertheless, remains a challenge. Considerable progress has been obtained in internal transport barrier (ITB) plasmas, with operation at central densities close to the Greenwald density or/and low toroidal rotation or/and high triangularity. Demonstrations of full current drive and successful simultaneous real time control of safety factor and temperature profiles have been achieved in ITB plasmas. Physics of resistive wall modes (RWMs) has been compared with theory, showing favourable scaling for ITER. High β N ∼ 2.8 operation of hybrid modes (also called improved H-modes) has been obtained with dominant neutral beam heating. Hybrid modes with dominant ion cyclotron resonance heating (ICRH) have also been achieved.Trace tritium experiments yielded valuable information on particle transport in H-mode, ITB and hybrid regimes. In Type I ELMy plasmas, successful tests of the conjugate-T ICRH scheme have been achieved as well as lower hybrid coupling at ITER-relevant 10-11 cm distances. Reduced D and T fuel retention has been observed, which could relate to operation with vertical targets in the divertor and/or lower (ITER-like) vessel temperature. It is confirmed that erosion occurs predominantly on the main chamber surfaces, with possible benefits for T retention in ITER, although consequences for the metallic first wall lifetime need to be assessed. Disruption and ELM studies indicate that transient power deposition could be less constraining than expected for the ITER divertor, but more challenging for the metallic first wall. Alpha particle tomography and direct observation of alpha particle slowing down have been made possible by γ -spectroscopy. Measurements of Alfvén cascades have been improved by a new interferometric technique. Promising tests of ITER relevant neutron counting detectors have been conducted.

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“…Review: Plasma-material interactions required by a factor of 4 [691,692] and allowed the attainment of a very high ion temperature [100]. A comparison of wall deuterium retention and plasma fuelling requirements in the ASDEX-Upgrade, DIII-D and TdeV tokamaks is reported in Ref.…”

Section: (January 2001)mentioning

confidence: 99%

“…The percentage of tritium retained (i.e., not recovered in the exhaust) was higher than that of D in both JET [692] and JT-60U [731,732]. Analysis of tritium in PFCs and exhaust gases shows that 50% or more of tritium produced by D(D,p)T nuclear reactions in JT-60U, during the period between July 1991 and October 1993, remained inside the vessel [733].…”

Section: Jt-60umentioning

confidence: 99%

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

et al. 2001

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The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

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Deuterium release measurements in the Be phase of JET and determination of tritium content in the exhaust gas

Cited by 40 publications

References 2 publications

Fuel retention in tokamaks

Fuel retention in tokamaks

Inverse Theorems for Multipoint Padé Approximants

Plasma-material interactions in current tokamaks and their implications for next step fusion reactors

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