CRP(2002-2006): Tritium Inventory in Fusion Reactors

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Contents

Introduction

Current designs for nuclear fusion reactors call for the use of tritium and deuterium as the fuel for the energy producing fusion reactions. The use of tritium (T) must be carefully controlled due to its radioactivity. Its safety aspects will attract intense public scrutiny. An additional problem that did not arise in deuterium (D) fuelled experiments is the issue of fuel economy. Experiments in existing fusion machines indicate that tritium retention, while not a problem for plasma operations, would be unacceptable in a fusion reactor due to fuel economy reasons.

Tritium retention and the control of the tritium inventory depends strongly on the choice of plasma-facing materials (PFM) and their operational conditions (e.g. temperature, flux of impinging particles), plasma edge conditions, and geometry effects. The quantification of the inventory depends on an accurate knowledge of processes for both retention and release of tritium from the machine materials. To date there are still large uncertainties in quantifying the in-vessel tritium inventory of future devices such as ITER. They arise mainly from the plasma edge physics parameters, which are anticipated to strongly affect the erosion, deposition and co-deposition patterns and rates. Moreover, mixed-materials effects, arising from the simultaneous use of different plasma-facing materials introduce significant uncertainties.

Designs for next generation fusion machines such as ITER employ several plasma facing materials (PFM) selected for their suitability to regions of the vessel with different power and flux characteristics. Among the materials under consideration are carbon based materials, beryllium, tungsten as well as steels and other materials. A complication is the realization that material mixing will be important in devices that use more than one PFM. It is often the case that the mixed material has different properties than the pure material (e.g. T-retention, erosion yields, thermal conductivity).


Scope of the IAEA CRP Project[1] - Tritium Inventory in Fusion Reactors

There are a number of important issues for the interaction of tritium with graphites and carbon-based materials. These materials have a number of desirable properties in the construction of tiles for the PFC of a fusion reactor. There are, however, serious concerns regarding the retention of tritium by the carbon-based materials. Major areas to be investigated include the absorption of tritium by neutron-irradiated carbon, the diffusion of tritium in carbon, and the outgassing of tritium from graphite tiles.

In addition topics such as laser detritiation of carbon-based tiles should be investigated as a means of releasing tritium. Recent experiments with a scanning laser have shown that the thermal response of codeposits is different from that of manufactured graphite. The experiments also revealed microscopic hot spots due to the granular inhomogeneous morphology of codeposits. It is therefore important to study codeposits to obtain reliable predictions of the temperature rise of codeposits in tokamaks under high heat flux.

In addition to carbon, there are other materials under consideration for PFC design. Beryllium is the primary candidate of the ITER first-wall. Much study has already gone into beryllium, and results are encouraging. The recent successful use of high-Z metal PFC in working tokamaks has also been welcome news in this respect. There are still some important areas for research for Be and tungsten. Issues for Be concern D-retention and release behaviour of C/Be mixed layers, including effects of thermal oxidation on release of hydrogen. Further study must be conducted to better characterise deposition (rates and patterns) of Be films on C as a function of the substrate temperature and relative D+/Be+ fluxes. Further work on W includes investigations on blister cracking under high hydrogen fluxes, thermal response of blisters to heat flux, effect of neutron irradiation on the tritium retention, permeation of tritium as well as tritium retention of tungsten mixed with other materials such as Be and C.

Further research could also be done on steel, especially concerning the permeation issue. There is also some possibility of considering using a liquid metal such as lithium for continual removal of tritium from the machine. Other types of mixed materials may also be of importance. These specific issues need to be addressed in order to have some confidence of successfully modeling the total tritium inventory of a fusion reactor machine. This CRP should be able to begin a serious investigation of these problems and make available an extremely useful database on a number of these processes.


CRP Research Results

A joint paper[2] and a APID volume[3] by CRP participants were produced after the completion of the CRP. Presentations on the following topics can be found in the IAEA A+M Homepage[1].

Research Activities Related to Tritium Retention

Conclusions

  • Choice of plasma facing materials (PFM) arguably the highest risk factor for ITER
  • Erosion and tritium retention properties determine the selection of PFM
  • Tritium inventory risks of proposed ITER wall materials:
    • Tungsten demonstrates the lowest tritium-inventory risk
    • Carbon presents greatest risk; tritium removal techniques are essential
    • Beryllium presents major risk through co-deposits in the presence of oxygen; tritium removal techniques need also to be developed and applied for Be
    • Limited understanding of mixed material effects increases tritium-inventory risks


Recommendations

  • Focus more R&D on tritium removal from Be and BeO co-deposits with C and W
  • ITER design needs to allow change of first wall materials, due to the concern of unacceptable tritium inventory with current materials
  • ITER should explore the possibility of using high (400 C or more) temperature for tritium removal and for reduction of tritium inventory
  • It would be desirable to design a room temperature co-deposit collector in the divertor that is heatable to >700 C * for subsequent hydrogen release and removal


References

  1. 1.0 1.1 http://www-amdis.iaea.org/CRP/
  2. C.H. Skinner, A.A. Haasz, V.Kh. Alimov, N. Bekris, R.A. Causey, R.E.H. Clark, J.P. Coad, J.W. Davis, R.P. Doerner, M. Mayer, A. Pisarev, J. Roth, T. Tanabe, Recent Advances on Hydrogen Retention in ITER's Plasma-Facing Materials: Beryllium, Carbon, and Tungsten, Fusion Science and Technology 54 (2008) 891-945
  3. Atomic and Plasma-Material Interaction Data for Fusion, IAEA, Vienna, Austria, volume 15 (in press)
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