CRP(2012-2016): Data for Erosion and Tritium Retention in Beryllium Plasma-Facing Materials

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Contents

Summary

There is very active interest at present in the properties of beryllium as a wall material exposed to plasma in a fusion reactor environment. The planned plasma-facing materials for nuclear operation in ITER are beryllium and tungsten: beryllium for most of the vacuum vessel and tungsten for the regions of highest heat load. Since August 2011 the Joint European Torus (JET) device is operating with a similar Be-W "ITER-Like Wall". Beryllium has in its favour good heat conductivity, strong gettering capability, high tolerance as a plasma impurity and low nuclear activation. On the other hand, erosion and tritium retention are issues of concern. The central issues are erosion under regular heat and particle loads from the plasma, melting and ablation under extreme (pulsed) loads, tritium retention, and ways to extract trapped tritium. It must be taken into account that the material and surface properties are highly variable as a result of interaction with impurities (primarily C, N, O, Ne and Ar), implantation of H and He, redeposition of eroded Be and resolidification of melt layers.

The CRP on "Data for erosion and tritium retention in beryllium plasma-facing materials" is intended to enhance the knowledge base on fundamental particle-material interaction processes involving beryllium in the fusion plasma environment. The key processes to be studied in the CRP are physical and chemical sputtering by H, He and Be, which release beryllium impurities into the plasma, trapping and reflection of hydrogen (H, D, T) on beryllium surfaces, the transport of hydrogen in beryllium and means to extract trapped tritium. The CRP will emphasize data for the relevant mixed materials, especially Be-(H,D,T,He), Be-C, Be-N, Be-O and ternary and higher mixtures, and data for the principal plasma impurities as projectiles. The most important projectiles are therefore H, D, T, He, Be, C, N, O, Ne and Ar. The CRP brings together experimentalists and computational theorists that are engaged in studies of plasma-material interaction with beryllium and related mixed materials and of hydrogen migration in solid beryllium.

The behaviour of the eroded material in the plasma belongs to our CRP on "Light Element Atom, Molecule and Radical Behaviour in the Divertor and Edge Plasma Regions". Macroscopic surface processes such as melting and ablation under intense (pulsed) heat loads are not emphasized in the present CRP. Materials issues such as neutron damage, fabrication and structural properties are outside the scope of the CRP.

Background Situation Analysis

Overview

In the magnetic confinement approach to fusion energy production, deuterium-tritium plasma at a temperature of around 15 keV (about 150 million K) is trapped in a toroidal magnetic field inside a vacuum vessel. The confinement is not perfect and the plasma interacts with the vacuum vessel wall, causing erosion of the wall and introducing impurities into the plasma by the mechanisms of physical and chemical sputtering and, exceptionally, localized melting and ablation. Also tritium may be trapped in the wall material, making it unavailable for fusion. All these issues are a serious concern for the realization of economically viable fusion energy production.

In present-day fusion experiments the regions most exposed to plasma are generally covered with graphite or carbon-fibre composites, but because of its propensity to absorb tritium carbon is not considered suitable for use in a reactor. The presently favoured plasma-facing materials for nuclear operation in ITER are beryllium and tungsten: beryllium for most of the vacuum vessel and tungsten for the regions of highest heat load. Because of its toxicity beryllium has generally been avoided in fusion research to-date, but in ITER and in a reactor the radioactive environment demands remote handling anyway and the toxicity is not a particular concern. During 2010-2011 a new “ITER-Like” Be-W vacuum vessel wall was installed on the Joint European Torus (JET) experiment, which is also equipped for tritium operation and remote handling, and plasma experiments on that machine have resumed in Fall 2011. Therefore there is very active interest at present in properties of beryllium as a wall material in the fusion environment. The central issues are erosion under regular heat and particle loads from the plasma, melting and ablation under extreme (pulsed) loads, tritium retention, and ways to extract tritium from the wall.

Scope of the CRP

In order to delineate the scope of the proposed CRP on data for erosion and tritium retention in beryllium plasma-facing materials a Consultants’ Meeting was held at IAEA Headquarters on 30-31 May 2011. Four experts on experimental and computational studies of plasma-wall interaction with beryllium discussed with us the scope and priorities of the proposed CRP and related activities of the Unit in the area of plasma-material interaction. The consultants strongly endorsed the intent to organize this CRP. They advised us to emphasize mixed materials (Be-C, Be-N, Be-O and surfaces impregnated with H, D, T and He) and to include the principal impurity ions as projectiles (C, N, O, Ne, Ar in addition to H, He and Be). They recommended to focus on basic particle-material interaction processes, leaving aside macroscopic surface processes such as melting and ablation. Possible participants in the CRP were identified too. The report of this Consultants’ Meeting is available [1].

A good overview of activities world-wide on plasma-material interaction with beryllium surfaces can be found through the contributions to two recurring major international meetings: the International Conference on Plasma Surface Interactions in Controlled Fusion Devices (“PSI”) and the International Workshop on Plasma-Facing Materials and Components for Fusion Applications (“PFMC”). The 19th PSI meeting was held in May 2010 and the 13th PFMC in May 2011. A list of contributions to these two meetings that concerned beryllium surfaces is provided here [2].

Experimental work on plasma interaction with beryllium is limited because of the hazardous nature of the material. Three facilities stand out. During 2010-2011 a beryllium main wall was installed on the Joint European Torus (JET), the largest operating tokamak world-wide, and experiments on JET resumed in Fall 2011 with special interest in plasma-wall interaction and boundary plasma physics. JET should produce a wealth of data over the course of the proposed CRP. The PISCES-B experiment at University of California at San Diego is a linear plasma device for simulation of plasma-wall interaction. PISCES-B is fully enclosed to allow operation with beryllium and its primary focus is the investigation of plasma-wall interaction with beryllium and of the behaviour of beryllium in plasma. QSPA-Be at the A. A. Bochvar Research Institute of Inorganic Materials in Moscow is a plasma-gun experiment also fully qualified for operation with beryllium; this facility is particularly suited for study of high heat and particle loads.

Theoretical and computational work on plasma-material interaction with beryllium is of increasing importance in connection with the experimental programme on JET and the construction of ITER. The existing database at the IAEA Atomic and Molecular Data Unit is based upon calculations employing the so-called binary collision approximation (BCA). However, there are several projects around the world that involve more comprehensive molecular dynamics simulations. Molecular dynamics simulations rely on interaction potentials that must be fitted to or calibrated against first principles electronic structure calculations. Transport of hydrogen in the material, and also the propagation of vacancies in the solid structure, rely on further first principles-based simulations. BCA calculations, molecular dynamics simulations, construction of interaction potentials and quantum-based studies of hydrogen transport in the material are all represented within the CRP.

[1] Data Needs for Erosion and Tritium Retention in Beryllium Surfaces, Summary Report of the Consultants’ Meeting, IAEA Headquarters, Vienna, Austria, 30–31 May 2011. http://www-amdis.iaea.org/publications/INDC/INDC_NDS-592.pdf

[2] Bibliography of recent conference contributions concerning plasma-material interaction with beryllium surfaces. http://kc.iaea.org/livelink/llisapi.dll/open/31460005

Nuclear Component

Fusion energy relies on the nuclear fusion of hydrogen isotopes deuterium and tritium. In the magnetic confinement approach a D-T plasma must be maintained at high temperature for a long time. Interaction of the hot plasma with material boundaries causes erosion of the wall, influx of impurities into the plasma and loss of tritium fuel as it gets trapped in the wall material. These are all critical issues for fusion energy production.

CRP Overall Objective

To increase capabilities of Member States to undertake fusion plasma modelling and simulation of present and future experiments and reactor designs through improved data for plasma-material interaction processes involving beryllium surfaces, and thereby to contribute to the development of fusion energy generation.

Specific Research Objectives

  • To inventorize existing data collections for plasma-material interaction with beryllium surfaces, including sputtering, erosion, reflection and trapping of incident particles and including mixed surfaces (Be-C, Be-N, Be-O) and surfaces impregnated with H and/or He.
  • To evaluate existing experimental and theoretical data, identify differences, conflicts and gaps, and make recommendations about best existing data for plasma-material interaction with beryllium and beryllium compounds.
  • To produce new measured and calculated data for particle and plasma interaction with beryllium and beryllium compounds.
  • To assemble existing and new data into a coherent database and knowledge base for use in fusion plasma modelling.

Expected Research Outputs

  • A meeting report in the INDC (Nuclear Data Section) series will be produced after each Research Coordination Meeting.
  • Scientific articles will be produced by participants in the CRP.
  • The Knowledge Base section of the web pages of the IAEA A+M Data Unit will be kept up-to-date to reflect the work of the CRP.
  • ALADDIN and other databases of the A+M Data Unit will be augmented with data produced or, where appropriate, evaluated in the CRP.
  • A final report of the CRP will be published in the IAEA Atomic and Plasma-Material Interaction Data for Fusion (APID) series.

CRP Expected Research Outcomes

  • Data produced or assembled in this CRP will be used in the interpretation of diagnostics of the divertor and edge plasma in fusion experiments and in plasma modelling for the design of future fusion reactor experiments.
  • Data produced or assembled in this CRP will be used in modelling and simulation of surface composition dynamics for beryllium plasma-facing materials.
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