Tritium removal from JET-ILW after T and D–T experimental campaigns

Nuclear Fushion

article

D. Matveev¹  D. Douai²  T. Wauters³  A. Widdowson⁴  I. Jepu⁴  M. Maslov⁴  S. Brezinsek¹  T. Dittmar¹  I. Monakhov⁴  P. Jacquet⁴  P. Dumortier⁵  H. Sheikh⁴  R. Felton⁴  C. Lowry⁴  D. Ciric⁴  J. Banks⁴  R. Buckingham⁴  H. Weisen⁶  L. Laguardia⁷  G. Gervasini⁷  E. de la Cal⁸  E. Delabie⁹  Z. Ghani⁴  J. Gaspar^1,10  J. Romazanov¹  M. Groth^1,11  H. Kumpulainen^1,11  J. Karhunen^1,11  S. Knipe⁴  S. Aleiferis⁴  T. Loarer²  A. Meigs⁴  C. Noble⁴  G. Papadopoulos⁴  E. Pawelec^1,12  S. Romanelli⁴  S. Silburn⁴  E. Joffrin²  E. Tsitrone²  F. Rimini⁴  C.F. Maggi⁴ 

[1] Forschungszentrum Jülich GmbH, Institut für Energie- und Klimaforschung – Plasmaphysik;CEA Cadarache;ITER Organization;UK Atomic Energy Authority, Culham Science Centre;Laboratory for Plasma Physics, Koninklijke Militaire School—Ecole Royale Militaire;Swiss Plasma Center, Station 13;Institute for Plasma Science and Technology;Laboratorio Nacional de Fusión;Oak Ridge National Laboratory;Aix-Marseille University;Department of Applied Physics, Aalto University;University of Opole, Institute of Physics

关键词: JET-ILW;    tritium retention;    tritium removal;    wall cleaning;    DTE2;   

DOI  :  10.1088/1741-4326/acf0d4

来源: Institute of Physics Publishing Ltd.

【 摘 要 】

After the second Deuterium–Tritium Campaign (DTE2) in the JET tokamak with the ITER-Like Wall (ILW) and full tritium campaigns that preceded and followed after the DTE2, a sequence of fuel recovery methods was applied to promote tritium removal from wall components. The sequence started with several days of baking of the main chamber walls at 240 °C and at 320 °C. Subsequently, baking was superimposed with Ion-Cyclotron Wall Conditioning (ICWC) and Glow Discharge Conditioning (GDC) cleaning cycles in deuterium. Diverted plasma operation in deuterium with different strike point configurations, including a Raised Inner Strike Point (RISP) configuration, and with different plasma heating—Ion Cyclotron Resonance Frequency (ICRF) and Neutral Beam Injection (NBI)—concluded the cleaning sequence. Tritium content in plasma and in the pumped gas was monitored throughout the experiment. The applied fuel recovery methods allowed reducing the residual tritium content in deuterium NBI-heated plasmas to about 0.1% as deduced from neutron rate measurements. This value is well below the requirement of 1% set by the maximum 14 MeV fusion neutron budget allocated in the ensuing deuterium plasma campaign. The quantified tritium removal over the course of the experiment was left( {13.4 pm 0.7} right) imes {10^{22}}atoms or left( {0.67 pm 0.03} right)g with ∼58% attributed to baking, ∼12.5% to ICWC, ∼26% to GDC, and ∼3.5% to first low power RISP plasmas. The experimentally estimated amount of removed tritium is in good agreement with long-term tritium accounting by the JET tritium reprocessing plant, in which the unaccounted amount was reduced by 0.71g after the cleaning experiment.

【 授权许可】

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