Review
Hydrogen isotope retention in beryllium for tokamak plasma-facing applications

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Abstract

Beryllium has been used as a plasma-facing material to effect substantial improvements in plasma performance in the Joint European Torus (JET), and it is planned as a plasma-facing material for the first wall (FW) and other components of the International Thermonuclear Experimental Reactor (ITER). The interaction of hydrogenic ions, and charge-exchange neutral atoms from plasmas, with beryllium has been studied in recent years with widely varying interpretations of results. In this paper we review experimental data regarding hydrogenic atom inventories in experiments pertinent to tokamak applications and show that with some very plausible assumptions, the experimental data appear to exhibit rather predictable trends. A phenomenon observed in high ion-flux experiments is the saturation of the beryllium surface such that inventories of implanted particles become insensitive to increased flux and to continued implantation fluence. Methods for modeling retention and release of implanted hydrogen in beryllium are reviewed and an adaptation is suggested for modeling the saturation effects. The TMAP4 code used with these modifications has succeeded in simulating experimental data taken under saturation conditions where codes without this feature have not. That implementation also works well under more routine conditions where the conventional recombination-limited release model is applicable. Calculations of tritium inventory and permeation in the ITER FW during the basic performance phase (BPP) using both the conventional recombination model and the saturation effects assumptions show a difference of several orders of magnitude in both inventory and permeation rate to the coolant.

Introduction

Beryllium has for several years been considered a primary candidate for plasma-facing components (PFCs) in tokamaks because of its low atomic number and excellent oxygen gettering capabilities [1]. Both of these features result in a lower Zeff for the plasma. After incorporating beryllium as a FW material in the Joint European Torus (JET), first by evaporating beryllium over other internal surfaces and then by using beryllium tiles in the vessel bottom facing the X-point, an improvement in plasma performance was seen as manifested by increases in deuterium ion density, deuterium ion temperature, and energy confinement time 2, 3. The addition of beryllium also resulted in a more than doubling of the fueling rate requirement as well as a marked increase in the post-shot gas release [4].

This paper concerns one key issue in the evaluation of materials for PFC applications, namely, tritium uptake, retention, and permeation. Quantitative predictions for these processes are important because they affect safety assessments, fuel economy, and tokamak plasma operational performance. In general, the factors that influence uptake and retention include specific material type, temperature, defect microstructure, surface properties, hydrogen solubility, diffusivity and surface recombination 5, 6.

Our specific purpose is to establish the scientific framework for a realistic prediction of tritium uptake and retention in Be that might be used as a PFC material in next-step tokamaks like ITER. This task is accomplished in two ways: (1) by critically reviewing the experimental data base from the perspective of ITER-relevant conditions such as particle flux, energy, and fluence and (2) by presenting an appropriate model to simulate tritium uptake and retention in Be, under ITER-relevant conditions. Our intent is to bring together all relevant information in one publication that pertains to assessment of tritium uptake and retention in Be used in tokamak PFC applications.

In Section 2, we first consider the utilization of beryllium in the design of tokamaks, focusing particularly on the ITER. This design is presently considered, by many, to be representative of fusion power-producing machines that will be developed, though admittedly, it is still embryonic for a commercial power plant design. In Section 2.1the design features of ITER Be components are considered, and in Section 2.2we review expected operational conditions.

Section 3provides a detailed compilation and review of the experimental data base for hydrogen uptake and retention in Be. We consider, in Section 3.1, experimental data that were measured for pure Be, including fully-dense, consolidated powder metallurgy (CPM) Be, ingot metallurgy Be foils, and plasma-sprayed Be. Recent results are included from experiments with ion-beam systems, linear plasma simulators, and tokamak plasma environments. Section 3.2presents the results of experiments to investigate hydrogen uptake in C/Be mixed material layers. The retention of hydrogen during co-deposition of sputtered Be and hydrogen were investigated in experiments that are reviewed in Section 3.3. Microstructural evolution in hydrogen-implanted beryllium is discussed in Section 3.4, and the effects of neutron irradiation on tritium production and retention is discussed in Section 3.5.

In Section 4we present the results of different strategies for modeling the uptake and retention of hydrogen in Be. Both a conventional diffusion/recombination model and a newly developed saturation model are applied. The results of model calculations for ITER-relevant conditions are presented and compared with the experimental data base.

Finally, Section 5summarizes the key findings of this work. Observations are presented regarding the experimental data review and the development of an uptake/retention model that is consistent with the experimental data. Factors that influence the uptake and retention of tritium in Be are emphasized and guidance is proposed to enable realistic model prediction of tritium inventories for ITER-relevant conditions. In addition, areas requiring further research are identified.

Section snippets

Beryllium utilization as a plasma-facing material

The ITER tokamak is envisioned to be the next major step in the world's fusion program from the present generation of tokamaks, designed to study fusion plasmas with the reactor relevant range of plasma parameters (see key features and parameters in Table 1Table 2). ITER engineering design has been completed to provide ITER with the capability to achieve sustained ignition, and extended-duration fusion burn in deuterium–tritium (DT) plasmas with reactor-relevant engineering features that

Review of hydrogen retention data

In this section we review experimental measurements that have been made to understand the processes of hydrogen retention in beryllium. Deuterium and tritium ions and charge-exchange neutral atoms will implant into plasma-facing surfaces. Most of these atoms will return to the plasma but some will diffuse to a greater or lesser extent into the bulk of the beryllium. Traps inhibit release of these isotopes. Return of implanted atoms to the plasma has long been believed to be governed by

Models for predicting tritium retention in Be PFC materials

Conventional wisdom, applicable in a wide variety of experiments, has been that hydrogenic atoms or ions implanting into a metal will diffuse both ways from the implantation zone. Atoms that reach the surface may recombine with other diffusing hydrogenic species or with receptor atoms such as oxygen present on the surface to form molecules that are then released to the gaseous state above the surface [8]. Traps present in the material can retard diffusion and enhance retention of the hydrogenic

Discussion and conclusions

The experimental data reviewed here clearly show the following

  • 1.

    Under ITER-like implantation conditions (>1020 D + T/m2 s, E > 100 eV), hydrogenic atom concentrations saturate near the surface of beryllium causing damage and resulting in strongly enhanced rate of return of plasma ions and atoms to the surface.

  • 2.

    The quantity of hydrogenic atoms retained in the near-surface region (well beyond the implantation depth) is dependent on the temperature of the specimen and to a lesser extent on the energy of

Acknowledgements

This work was performed for the US Department of Energy, Office of Energy Research under DOE Idaho Operations Office Contract DE-AC07-94ID13223 and under other contracts at the respective institutions of the authors. The authors gratefully acknowledge the many helpful conversations with others active in this area of investigation, both within the US and Canada and in other countries, that have helped to clarify understanding and expand on information available in the literature.

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