Beryllium is a low-density metal that is used in a number of industries, including the nuclear, automo­tive, aerospace, defense, medical, and electronics

industries, for various applications because it is exceptionally strong, is light in weight compared with other metals, has high heat-absorbing capability, and has dimensional stability in a wide range of temperatures.

Beryllium has been considered for many years as a primary candidate for protection of PFCs in toka — maks because it offers distinct advantages when com­pared with alternative materials such as carbon and tungsten. It has a low atomic number and is an excel­lent oxygen getter. The interaction of beryllium with tritium is also significantly weaker than that of carbon, leading to potentially reduced tritium inven­tory. Beryllium does not form stable hydrides above 300 °C, so there should be very little trapping expected in codeposited layers formed at such tem­peratures in the divertor after sputtering, although work is still underway to clarify this problem. How­ever, beryllium has a relatively high physical sputter­ing rate and a relatively low melting temperature and as such is more susceptible to melting damage that may occur in a tokamak during thermal transients. In addition, because of its toxicity, special precautions are needed for working with beryllium, either for manufacturing or research investigation purposes.

Beryllium has been used with success in various tokamaks in the past mainly because of its ability to getter oxygen and to improve plasma performance. In particular, its successful deployment in JET that started in 1989 and is continuing today with the installation of a completely new beryllium wall is the main rationale for the selection of beryllium as a plasma-facing material for the first wall of ITER, on the basis of a combination of plasma compatibility and design considerations.

This paper reviewed the properties of beryllium that are of primary relevance for plasma protection applications in magnetic fusion devices (i. e., PWIs, thermal and mechanical properties for power handling, fabricability and ease of joining, chemical reactivity, etc.) together with the available knowledge on performance and operation in existing fusion machines.

Special attention was given to beryllium’s erosion and deposition, formation of mixed materials, and its hydrogen retention and release characteristics. These phenomena have a profound impact on component design, machine operation, and safety. Extensive data on the behavior of Be with plasmas have been col­lected from existing tokamaks and simulators during the last two decades and this has enabled great strides to be made in our understanding of the PWI pro­cesses involved. However, there are many issues for which there are still uncertainties and we will only learn from operating the next two major experiments that foresee the use of large amounts of Be (JET and ITER). Much work remains to be done in this area and more machine operational time and diagnostics dedicated to PWIs are required. Initiatives on these fronts, together with modeling of the results, are essential to advance the understanding of PWIs. This includes (1) the possible surface damage (melt­ing) during transients such as ELMs and disruptions and its implications for operations and (2) the prob­lem of beryllium mixing with other armor materials and in particular the issue of codeposition of tritium with Be, which is eroded from the first wall and deposited at the divertor targets. Such material may also be locally redeposited into shadowed areas ofthe shaped ITER first wall. Both issues are part of
ongoing research, the initial results of which are being taken into account in the ITER design so that the influence of these two factors on ITER operation and mission are minimized. For example, ITER will very likely employ, ELM control systems based on pellets and RMP coils, disruption mitigation systems, and increased temperature baking of the divertor to release Tfrom Be-codeposits. Dust generation is still a process which requires more attention. Conversion from gross or net erosion to dust and the assessment of dust on hot surfaces need to be investigated.

At the time of writing this paper, the ITER first wall and shielding blanket is undergoing a major rede­sign effort to overcome some ofthe main shortcomings that were identified in the context of a recent design review scrutinizing the internal components.

Complex and interrelated materials, manu­facturing, and design issues were briefly reviewed in this paper together with the progress of the manufacturing technologies being used and tested to demonstrate the durability of the joints. A critical feature of the ITER first-wall design is the beryllium to copper alloy bond. The joints must withstand the thermal, mechanical, and neutron loads and the cyclic mode of operation, and operate under vacuum, while providing an acceptable design for lifetime performance and reliability. The availability of reli­able joining technologies has a large impact on the design of the PFCs and on the overall cost of these components.

The status of the available techniques presently considered to join the Be armor to the heat sink material of Cu alloys for the fabrication of Be-clad actively cooled components for the ITER first wall was discussed. During earlier ITER design phases, the feasibility of manufacturing reliable Be-CuCrZr joints was demonstrated. The results of the perfor­mance and durability heat flux tests conducted in the framework of the further ITER first-wall qualifica­tion program were described. This program has been launched and is in progress in the ITER parties in order to qualify the design and manufacturing routes. The integrity of this bond must be assured for reliable ITER performance whatever process is used to fabri­cate joints. The original procurement sharing that assigned the fabrication of first-wall panels up to six parties was seen as a risk and the number of parties supplying these critical components has now been reduced to three, Europe, the Russian Federation, and China.

The selection of specific grades of specific beryl­lium for the ITER first wall was described. The effects of neutron irradiation on the degradation of the properties of beryllium itself and on the joints were also analyzed. Some of the changes are impor­tant while others are not significant for the ITER conditions. Change of thermal conductivity and swelling are not important because of the low fluence. The bulk tritium retention in neutron irradiated Be is expected to be significantly less than tritium reten­tion in the codeposited layers. The most critical consequence of neutron irradiation under ITER con­ditions is embrittlement. This is typical of all grades of beryllium. The structural integrity of neutron irradiated brittle Be is a key issue. Embrittlement of neutron-irradiated Be could lead to increased ther­mal erosion and crack formation, which is also observed to occur for unirradiated beryllium under severe transient heat loads. These cracks could serve as thermal fatigue crack initiation sites and accelerate this type of damage. While this effect has not been extensively studied because of the difficulty of simu­lating disruptions in the laboratory, it may not be a critical issue as thermal fatigue cracks form after a few hundred cycles in most materials and they grow only to depths where the thermal stress level is above the yield stress.

On the basis of the information available from exist­ing fusion machines, we discussed the problems that are still at issue in the design and operation of ITER. This includes, in particular, the problem of erosion/ damage and the problem of up-take and control of tritium in the beryllium-based codeposited films. Finally, on the basis of these results some tentative and speculative consideration of the limited prospects that beryllium has in future reactors was offered.

The worldwide fusion energy research over the last four decades has developed a tremendous amount of knowledge on plasma physics and related technolo­gies. From this point of view, collecting the latest information from a wide range of studies is important in order to help the fusion community to recognize the critical issues and the status. That has been the intent of this chapter. (See also Chapter 4.17, Tungsten as a Plasma-Facing Material and Chapter 4.18, Carbon as a Fusion Plasma-Facing Material).

Acknowledgments

The views expressed in this publication are the sole responsibility of the authors and do not necessarily reflect the views of Fusion for Energy. Neither Fusion for Energy nor any person acting on behalf of Fusion for Energy is responsible for the use which might be made of the information in this publication.

The authors from the ITER Organization wish to acknowledge that this paper was prepared as an account of work by or for the ITER Organization. The Members of the Organization are the People’s Repub­lic of China, the European Atomic Energy Commu­nity, Republic of India, Japan, Republic of Korea, the Russian Federation, and the United States. The views and opinions expressed herein do not necessarily reflect those of the Members or any agency thereof. Dissemination of the information in this paper is governed by the applicable terms of the ITER Joint Implementation Agreement.