Thermal desorption of hydrogen isotopes from the JET Be plasma facing components

TY - JOUR

T1 - Thermal desorption of hydrogen isotopes from the JET Be plasma facing components

AU - JET contributors

AU - Avotina, Liga

AU - Jepu, Ionut

AU - Baron-Wiechec, Aleksandra

AU - Kresina, Michal

AU - Widdowson, Anna

N1 - Publisher Copyright: © 2020 The Royal Swedish Academy of Sciences. Copyright: Copyright 2020 Elsevier B.V., All rights reserved.

PY - 2020/1/1

Y1 - 2020/1/1

N2 - Understanding of fuel retention and release processes from materials for ITER (International Thermonuclear Experimental Reactor) is important from fundamental and technological aspects. Detailed information regarding fuel retention and release characteristics will allow global fuel inventories to be estimated in fusion devices as well as indicate the requirements for development detritiation methods and the re-use of tritium. Selected beryllium (Be) tiles were extracted from the JET vacuum vessel after each experimental campaign period with the ITER-like wall (ILW); so called ILW1 (2011-2012), ILW2 (2013-2014) and ILW3 (2015-2016). Desorption of hydrogen isotopes of samples taken from the inner wall guard limiter (IWGL), outer poloidal limiter (OPL) and dump plate (DP) tiles were analysed by means of thermal desorption spectrometry (TDS). The results presented compare data across ILW1, ILW2 and ILW3 and show the long term trends of fuel retention in Be limiter tiles. For all three campaigns the level of retention correlates with erosion and deposition that takes place during plasma operations Deuterium retention varies from 0.01-1 × 1018 atoms cm-2. Deuterium release takes place in several stages, related to different types of traps which can be within the co-deposit layer and/or below the surface in the Be bulk, with the main release stages around 700-760, 850-900 and 1020 K. The level of tritium in Be was found to be 104 times lower than deuterium for ILW campaigns 1-3.

AB - Understanding of fuel retention and release processes from materials for ITER (International Thermonuclear Experimental Reactor) is important from fundamental and technological aspects. Detailed information regarding fuel retention and release characteristics will allow global fuel inventories to be estimated in fusion devices as well as indicate the requirements for development detritiation methods and the re-use of tritium. Selected beryllium (Be) tiles were extracted from the JET vacuum vessel after each experimental campaign period with the ITER-like wall (ILW); so called ILW1 (2011-2012), ILW2 (2013-2014) and ILW3 (2015-2016). Desorption of hydrogen isotopes of samples taken from the inner wall guard limiter (IWGL), outer poloidal limiter (OPL) and dump plate (DP) tiles were analysed by means of thermal desorption spectrometry (TDS). The results presented compare data across ILW1, ILW2 and ILW3 and show the long term trends of fuel retention in Be limiter tiles. For all three campaigns the level of retention correlates with erosion and deposition that takes place during plasma operations Deuterium retention varies from 0.01-1 × 1018 atoms cm-2. Deuterium release takes place in several stages, related to different types of traps which can be within the co-deposit layer and/or below the surface in the Be bulk, with the main release stages around 700-760, 850-900 and 1020 K. The level of tritium in Be was found to be 104 times lower than deuterium for ILW campaigns 1-3.

KW - Beryllium

KW - Hydrogen isotopes

KW - Jet

KW - Thermal desorption

UR - http://www.scopus.com/inward/record.url?scp=85081369544&partnerID=8YFLogxK

U2 - 10.1088/1402-4896/ab3c38

DO - 10.1088/1402-4896/ab3c38

M3 - 会议文章

AN - SCOPUS:85081369544

SN - 0031-8949

VL - 2020

JO - Physica Scripta

JF - Physica Scripta

IS - T171

M1 - 014009

T2 - 17th International Conference on Plasma-Facing Materials and Components for Fusion Applications, PFMC 2019

Y2 - 20 May 2019 through 24 May 2019

ER -