Evaluation of tritium retention in plasma facing components during JET tritium operations

Abstract

An assessment of the tritium (T) inventory in plasma facing components (PFC) during JET T & DT operations is presented based on the most comprehensive ex-situ fuel retention data set on JET PFCs from the 2015-2016 ILW3 operating period is presented. The global fuel retention is 4.19 x 1023 D atoms, 0.19% of injected fuel. The inner divertor remains the region of highest fuel retention (46.5%). The T inventory in PFCs at the end of JET operations is calculated as 7.48 x 1022 atoms and is informative for accountancy, clean-up efficacy and waste liability assessments. The T accumulation rate at the upper inner divertor during JET deuterium-tritium (DT) operations has been used to assess the requirements and frequency of operation of a new laser induced desorption diagnostic to be installed on JET for the final DT experiments in 2023.

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

“…While baking, GDC and ICWC mainly targeted the main chamber wall retention, the RISP plasma configuration was designed to provide increased particle and heat loads to the upper part of the inner divertor (Tile 1, figure 1(a)). Indeed, the highest co-deposition of fuel with Be eroded from the main chamber limiters was measured post-mortem in this location after the preceding JET-ILW campaigns, contributing about 50% to the total long-term fuel inventory [14]. It was demonstrated by measurements and modeling that an increase of the surface temperature up to at least 800 • C is needed to achieve a significant fuel release from such co-deposits, especially for experimentally observed layer thicknesses above 10 µm [15][16][17].…”

“…For a 40 µm thick layer (i.e. the largest thickness of Be layers measured post-mortem on the top inner divertor [14]), less than 50% of fuel is removed in the case of initial D/Be = 1% and only about 15% is removed for initial D/Be = 10%. All in all, for moderately thick layers of <10 µm with initial D content <5% as observed for JET-ILW samples from Tile 1, solely the heating of divertor surfaces by plasma during RISP pulses can lead to significant fuel removal from co-deposited layers by thermal outgassing (>65%).…”

“…Taking into account longer breaks in operation between campaigns (full T → D-T → full T) and occasional routine GDC cycles, a significantly lower retention fraction can be expected. As a lower bound, the longterm D retention in JET-ILW of 0.19% [14] can be taken for reference. This will result in 0.48 g of retained T, which is below the value of 0.67 g removed by the cleaning measures presented here.…”

Application of Ion Beam Analysis in Studies of First Wall Materials in Controlled Fusion Devices

“…In the exposed plate, one detects gettered oxygen (20-40%) and co-deposited H, D, C, N. Hydrogen is clearly detected. Its content is greater than that of D, because the campaign was finished with 300 discharges fueled with H [59,61,62]. 3 He-based NRA is extremely efficient in D studies on PFC surfaces.…”

Dust Monitors in JET with ITER-like Wall for Diagnosis of Mobilized Particles and Co-Deposited Layers

“…The JET tokamak with the ITER-like wall (JET-ILW) [ 1 ] is a large-scale test bed for plasma operation with walls of the same material configuration as will be implemented in ITER [ 2 ]: beryllium (Be) on the main chamber wall and tungsten (W) in the divertor. The JET-ILW research programme has included broad studies of PFC (erosion and fuel inventory) [ 3 , 4 , 5 , 6 , 7 , 8 , 9 , 10 ] and various aspects of dust formation mechanisms, quantities, in-vessel location, morphology, fuel retention, classification (ITER-relevant or only JET-specific), mobilization and generation under water impact [ 11 , 12 , 13 , 14 , 15 , 16 , 17 ]. Lessons and previously developed collection methods from other devices have been taken into account [ 18 , 19 , 20 , 21 , 22 ].…”