Thermal desorption of hydrogen isotopes from JET beryllium plasma facing components

Thermal desorption of hydrogen isotopes from JET beryllium plasma facing components

Thermal desorption of hydrogen isotopes from JET beryllium plasma facing components 150 150 UKAEA Opendata
UKAEA-CCFE-CP(20)112

Thermal desorption of hydrogen isotopes from JET beryllium plasma facing components

Understanding of fuel retention and release processes from plasma facing components (PFCs) of ITER like wall materials is important from fundamental and technological aspects [1]. Detailed information about fuel retention and release characteristics in plasma facing components from JET will allow global fuel inventories to be estimated in fusion devices as well as inform requirements for the development of PFC detritiation methods and re-use of tritium.

Selected beryllium (Be) PFCs 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 inner wall guard limiter (IWGL), outer poloidal limiter (OPL) and dump plate (DP) tiles were analysed by means of thermal desorption spectrometry (TDS). Annealing was performed in vacuum with a heating rate 10 K/min up to 1050 K with a dwell time of 2 hours. The release of hydrogen isotopes during annealing was measured with quadrupole mass spectrometer.

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. This is most noticeable for the midplane IWGL tile which has two main zones – an erosion zone in the central part of tile and deposition zones at each end of the tile where minimal plasma interaction occurs. Deuterium and tritium concentrations are higher in the deposition areas than in erosion area. Retention is also found to be lower at the inner midplane limiter and dump plate compared with the outer midplane limiter which experiences lower heat flux during operations. Typical deuterium concentrations are 0.1-1 x 1018 atoms/cm2. Deuterium release takes place in several stages, related with different types of the traps in the tile surface, with the main release stages around 700, 760, 850 and 900 and 1020 K. The maximal deuterium release rate from beryllium is around 760 K, which may be related with detrapping from ion induced sites [2].

Low levels of tritium are also released during analysis. Although not fully calibrated for tritium, the results indicate a level 104 times lower than deuterium. Tritium can be from DD reaction and also due to off-gassing from previous DT operations in JET.

[1] A. Baron-Wiechec, K.Heinola, J.Likonen et al., Fusion Engineering and Design, 133, 135-141 (2018)

[2] M. Reinelt, Ch. Linsmeier, Journal of Nuclear Materials, 390-391 (2009) 568-571

Collection:
Conference
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Publisher:
Conference:
17th International Conference on Plasma-Facing Materials and Components for Fusion Applications (PFMC-17), Eindhoven, The Netherlands, 20-24 May 2019
Published date:
16/10/2021