Tritium Implantation Research Articles

Modelling neutron damage effects on tritium transport in tungsten

A damage-induced hydrogen trap creation model is proposed, and parameters for tungsten are identified using experimental data. The methodology for obtaining these parameters using thermo-desorption analysis spectra data is outlined. Self-damaged and optionally annealed tungsten samples have undergone TDS analysis, which has been analysed to identify the properties of extrinsic traps induced by the damage and to determine how they evolve with damage and annealing temperature. A parametric study investigated the impact of the damage rate and temperature on tritium inventories in tungsten. Tritium transport simulations have been performed with FESTIM considering a 1D model of a 2 mm sample of tungsten with damage rates and temperatures varying from 0– 102 dpa/fpy and 600–1300 K, respectively. The results show that after 24 h simultaneous exposure to neutron damage at 102 dpa/FPY and tritium implantation at 700 K, tritium inventories can increase by up to four orders of magnitude when damaged at a rate of 102 dpa/FPY compared to undamaged, defect-free tungsten, increasing further to five orders of magnitude after one full power year. The time taken to reach retention saturation shows the need for kinetic models of trapping properties on time scales relevant to reactor operation. The trap-creation model parameterisation procedure can be used to investigate neutron damage effects on other fusion-relevant materials such as EUROFER, Inconel and more.

Possible mitigation of tokamak plasma–surface interactions using thermionic divertor plates with inverse sheaths

Recent studies demonstrate that when a plasma-facing surface emits a sufficient flux of electrons, it will form an inverse sheath. Here, we consider a possibility of using thermionic target plates with inverse sheaths as an innovative divertor operating scenario. We derive an electron heat flux boundary condition for inverse sheaths and show that for given power exhaust into a tokamak scrape-off-layer, an inverse sheath leads to a much lower target plasma electron temperature than a conventional sheath. Low enough target plasma temperatures for radiative divertor detachment could therefore be achieved using inverse sheaths instead of the usual need to inject neutral atoms that compromise the core plasma. Other advantages of inverse sheath detachment over conventional sheath operating scenarios include (a) ion impact energies are as low as possible, minimizing sputtering and tritium implantation, (b) surface recombination heat flux is reduced due to ion flux reduction, and (c) arcs are inhibited due to the sign of the surface electric field. This paper outlines the basic properties of inverse sheath detachment and considers the feasibility of implementation. We offer recommendations for future modeling efforts needed to better understand the effects of thermionic emission in tokamaks and whether inverse sheaths present a viable divertor solution.

Effects of heat treatment on trapping and release of tritium from He pre-irradiated tungsten

The effects of heating on retention of tritium implanted into tungsten pre-irradiated with helium have been examined by β-ray-induced X-ray spectrometry (BIXS), which gives information about changes in the amount and depth profile of tritium retained on/in materials. Tritium implantation and isochronal heating of the helium pre-irradiated and un-irradiated tungsten samples were repeated as a series of experiments, and changes in the X-ray spectra were measured in each heating cycle. Quite different decreasing behavior between Ar(Kα) and W(Mα) X-ray intensities for both samples was observed in the heating process, and it was seen that the majority of tritium implanted into the helium pre-irradiation sample was trapped by the radiation defects rather than in helium bubbles.

Quantitative Measurement of Surface Tritium by β-Ray-Induced X-Ray Spectrometry (BIXS)

ABSTRACTTo make a nondestructive measurement of tritium retained on/in materials surfaces, conversion efficiency of β-rays to characteristic X-rays in an argon atmosphere has been examined. For this purpose, various tritium-containing graphite plates were prepared at first by ion implantation. After the tritium implantation, measurements of an X-ray spectrum from the graphite plates were carried out in the argon atmosphere. A good linear relation was observed between the intensity of Ar(Kα) characteristic X-rays and the total amount of tritium deter-mined by full-combustion. The apparent conversion efficiency was determined as 4.15x10−6 counts/s/Bq. To determine the intrinsic conversion efficiency for argon atoms, relevant correction factors such as geometrical efficiency, absorption of X-rays, effects of a tritium depth profile and a photoelectric effect were experimentally evaluated through numerical calculations. Taking into account these correction factors, the intrinsic conversion efficiency was determined to be 3.1x10−4 photons/β-particle.

Modeling Tritium Processes in Plasma-Facing Beryllium

In this paper we present techniques and recommended parameters for modeling tritium implantation, trapping and release, and permeation, in beryllium-clad structures adjacent to the plasma. Among the features that should be considered are the effects of surface films, the mobility of beryllium through those films, damage caused by ion implantation, especially in regions where pitting may be expected, and bubble formation. Tritium transport parameters recommended are based on fits with experimental data and available theory. Estimates of inventories in ITER using these parameters are also given. 31 refs., 2 figs., 1 tab.

Measurement of Diffusion Coefficient of Tritium in Alpha-Phase of Zirconium Based on Glow Discharge Implantation Method

A tritium implantation method based on a glow discharge of tritium gas was applied to the measurement of diffusion coefficient of tritium in alpha-phase of zirconium. The diffusion experiment was carried out in a temperature region from 673 K to 873K, and the diffusion coefficient of tritium was determined as follows, D/(cm 2 /s) = 3.14x10 -3 exp? The present data were in good agreement with our previous data which were determined with an usual implantation method based on a nuclear reaction 6 Li(n,a)T.

Implantation-Driven Tritium Permeation in Vanadium Alloy Fusion Reactors

The safety concerns resulting from tritium implantation into vanadium and vanadium alloys have been studied. Tests were conducted to measure the physical constants that determine permeation and inventory in these materials. Surface recombination coefficients, enhancement factors due to plasma sputtering, and the diffusion constants were measured. To determine the implications of these results, application was made to the TITAN-I fusion reactor design. For this design, the tritium permeation rate and inventory buildup in the first wall were calculated for a range of surface conditions on the coolant side. 7 refs., 8 figs., 1 tab.

Chemisorption and implantation of hydrogen isotopes in lithium-based fusion blanket ceramics and their thermal release: A study by nuclear reaction analysis

A brief summary is presented on tritium implantation techniques and depth profiling analysis by elastic recoil detection (ERD) and T(d, α)n nuclear reaction analysis (NRA), as applied to the studies of lithium ceramic materials to be used as tritium-breeding blankets in fusion reactors. New data about thermal behaviour of hydrogen and tritium near the surface of both poly- and monocrystalline Li 2O, and ceramic γ-LiAlO 2, are discussed. Ion beam techniques are especially suitable for studying near-surface diffusion and chemisorption of tritium, as well as for modelling the influence of irradiation damage on tritium transport phenomena.

The Trapping and Release of Injected Hydrogen Isotopes in Vitreous Silica

Abstract The behavior in vitreous silica of hydrogen isotopes, which had been introduced by 80-keV D+-ion implantation, thermal D2 doping, or recoil tritium implantation (2.7 MeV), was studied by means of FT-IR spectroscopy, ESR spectroscopy, and radioassay with a liquid scintillation counter. The hydrogen isotopes are trapped by forming hydroxyl bonds (OD, OT). In the D+-ion implantation, OD bonds are formed mainly by the rupture of the ≡Si–O–Si≡ bonds. In the thermal D2 doping, however, OD bonds are formed mainly by the isotope-exchange reaction between the preexisting OH bonds and D2 molecules. When vitreous silica implanted by recoil tritium is heated to about 1150 K, HTO and HT are released. The OD bonds produced by D+ implantation decrease upon annealing at 1150 K. The OD bond-rupture is controlled by the diffusion of hydrogen isotopes trapped as OD(OT) bonds to the surface of vitreous silica. The mechanism of the diffusion of the hydrogen isotopes through vitreous silica was explained in terms of a repetition of the chemical reaction of ≡SiOD(OT)+≡SiOSi≡→≡SiOSi≡+SiOD(OT). The diffusion coefficient (DTHTO)) of tritium according to the above mechanism was obtained as: D T(HTO)=2.3×10−4exp[−1.7×102(kJ mol−1)⁄RT]cm2s−1

Implantation Measurements to Determine Tritium Permeation in First Wall Structures

A principal safety concern for a D-T burning fusion reactor is release of tritium during routine operation. Tritium implantation into first wall structures, and subsequent permeation into coolants, is potentially an important source of tritium loss. This paper reports on an experiment in which an ion accelerator was used to implant deuterium atoms in a stainless steel disk to simulate tritium implantation in first wall structures. The permeation rate was measured under various operating conditions. These results were used in the TMAP computer code to determine potential tritium loss rates for fusion reactors.