CORE WORKING GROUP OPINION PAPER TO STIMULATE PRE-SNOWMASS DISCUSSION (S.J. Zinkle and M.C. Billone, co-chairs)
July 7, 1999 draft

  1. Introduction

The appropriate selection of materials is a key factor in realizing the full potential of fusion energy. The performance of first-wall and divertor structural materials has a significant impact on fusion economics, environmental issues and safety. In addition, numerous non-structural materials (e.g., plasma-facing, ceramic and liquid breeders, magnetic, coolant, insulator, optical, etc.) are required to successfully design inertial- and/or magnetic-fusion energy power plants. Materials issues are specifically mentioned in the supporting objectives of four of the five elements of the Office of Fusion Energy Sciences (OFES) Technology Program: Enabling Technologies, Advanced Technologies, Advanced Materials, and IFE Chamber/Target Technologies (cf. C.C. Baker, The US Technology Program, Version 5, December 4, 1998, http://www.fusionscience.org/).

The emphasis on materials R&D is most clearly visible in the OFES Advanced Materials Program. The currently-defined goal of this program is “ to develop structural materials that will permit fusion to be developed as a safe, environmentally acceptable and economically competitive energy source ”. Materials R&D is also highlighted in the Enabling Technologies (develop high-performance low-cost superconducting magnets, understand plasma-materials interactions and develop reliable plasma-facing components, etc.), Advanced Technologies (perform R&D to establish knowledge base...), and IFE Chamber/Target Technologies Program (assess chamber and final optic materials development requirements, etc.).

There are a number of important questions with regard to the scope and direction of the current materials programs. Are the materials programs addressing the needs of the engineering design communities in their efforts to develop attractive and competitive fusion power systems? Is there adequate interfacing between the materials and plasma sciences communities to address issues such as the electromagnetic effects of ferritic steels in a magnetically-confined reactor? Within the materials programs, what is the proper balance and timing of activities in the areas of basic materials studies and modeling, development of engineering databases, and component testing?

Key Issues

Are there materials performance issues which would effectively elevate or eliminate certain blanket and divertor design concepts? Most blanket designs involve integrated structure/coolant/breeder systems. If a particular coolant, structural or breeder material is determined to be unfavorable, it can impact the entire blanket design. For example, if there are mechanistic reasons why SiC/SiC composites cannot maintain a high enough thermal conductivity in a radiation environment, then a number of designs may be rendered impractical. On the other hand, if high-performance SiC/SiC composites can be developed, the attractiveness of these high thermal-efficiency concepts increases significantly.

What are the state-of-the-art materials science developments that may have a profound impact on fusion energy? Examples of these are: stir-friction welding for field construction and in-situ repair of refractory materials; new non-structural materials such as KU-1 quartz, free-standing CVD diamond wafers and high T C superconductors; creep-resistant oxide-dispersion-strengthened copper and ferritic-steel alloys which allow higher temperature and higher thermal-efficiency operation; and advances in computational materials sciences which allow “materials by design”.

What are the technical bases for the current materials R&D programs? What steps can be taken to ensure optimal interaction of the various materials-related portions of the Fusion Energy Sciences Program? What interaction checkpoints are needed? What resources and time scales are needed to develop particular structural and non-structural materials? Is there sufficient leveraging with international programs?

In the following section, possible steps to resolve some of these key issues are proposed as a starting point for the Snowmass discussions. Summaries of the various materials issues are contained in an accompanying appendix.



  1. Discussion of key fusion materials issues

At the present time, fusion materials R&D is distributed among several different elements of the Fusion Energy Sciences program. This can create inefficiencies if avenues for effective interaction are not maintained.




What is the appropriate technical basis for deciding the “thrust areas” for fusion materials; how can the fusion program most effectively incorporate any new developments in materials as they arise (e.g., high temperature superconductors, rapid prototyping concepts, etc.)




What steps can be taken to ensure optimal technical interaction among the various materials-related portions of the Fusion Energy Sciences Program with the design efforts? Currently there are formal interfaces between the Advanced Materials Program and the ARIES and APEX design efforts in the area of structural materials. Structural materials questions arising within the design teams are presented to the designated materials expert assigned to the design team. These questions are presented to the appropriate experts within the Advanced Materials Program for resolution. The response of the materials experts is then communicated to the design team in design-appropriate terms. New developments regarding database expansion and alloy development are communicated to the design teams through the designated interface person. However, in design efforts such as ARIES and APEX, numerous questions arise with regard to non-structural materials. Currently, there is no formal way to address these non-structural materials issues. Possible ideas to improve this situation are listed below







Are there materials performance issues which would effectively elevate or eliminate certain blanket and divertor design concepts?


Discussion on possible fusion materials R&D activities for the next ten years





Appendix: Position papers for the various fusion materials.

1.0 CURRENT RESEARCH AND DEVELOPMENT PROGRAM ON STRUCTURAL MATERIALS (E.E. Bloom, A.F. Rowcliffe et al.)

The Importance of Materials to the Development of Fusion Power

Development of fusion as a power source will depend upon finding solutions to very challenging problems in engineering and technology. Which approach to confinement, magnetic or inertial, provides the most attractive energy source? Can fusion be developed to capture its potential as a safe and environmentally attractive energy source? What concepts for tritium breeding and energy conversion are feasible and offer the potential for economic, safe, and environmentally attractive fusion power? Developing an understanding of the expected performance of materials in the very demanding service environment of fusion power systems, formulating improved materials, and addressing feasibility issues that relate to specific concepts is essential in obtaining answers to these and similar fundamental questions that will determine the direction of fusion power development. Materials developed for other applications, e.g., aerospace, high- temperature fossil energy systems, light water reactors, even liquid metal breeder reactors, will not meet the myriad challenges presented by fusion systems.

Why Place Emphasis on Structural Materials for the Blanket and First Wall


Numerous materials, with widely varying properties and characteristics, will be required in fusion power systems. The research emphasis in the Advanced Materials Program has been placed on the development of structural materials. In DT fueled fusion power systems the blanket system serves two essential functions: tritium fuel is produced from lithium contained in the blanket, and a large fraction of the energy produced in the burning plasma is converted to heat, extracted, and sent to a power conversion system via a coolant circulating through the blanket. The blanket system must be robust against failure and have a lifetime and temperature capability consistent with competitive economics. The blanket system is also a major contributor to the performance of fusion in key areas of safety and environmental impact. At this time all material systems have fundamental feasibility questions. To a large extent the properties of the structural material will dictate the blanket concept which in turn determines fundamental aspects of the design of the fusion power system. For example, vanadium alloys are possibly the only structural materials that could be used in a concept employing liquid lithium as the coolant/breeder. Thus if vanadium alloys with satisfactory performance cannot be developed, concepts employing liquid lithium as the coolant/breeder may not be viable. Stated simply, the fundamental characteristics of the blanket structural material dictate critical aspects of the power system design.

Performance Requirements


In addition to meeting requirements related to economic performance, the structural materials must also meet performance requirements related to maximizing both the environmental attractiveness and safety of fusion power systems. Achieving maximum economic performance will be constrained by the need to restrict alloying elements and impurities in order to lower and control the levels of radioactivity induced by the interaction of neutrons with the structural materials. The adoption of strict compositional requirements makes it possible to develop materials that will not require eventual long-term geological disposal and raises the possibility of recycling some materials to further minimize environmental impact. From safety considerations, a different set of materials requirements arise to minimize the levels of heat arising from radioactive decay and chemical reactions, to limit the dispersability of radioactivity and to minimize the biological hazard potential.

Materials performance requirements may be considered under three broad categories that directly impact the economics of fusion power. Firstly, structural materials must be capable of operating at high temperatures as well as operate over a wide temperature range in order that high power conversion efficiencies can be achieved. Secondly, structural materials must be capable of transmitting high heat fluxes; higher thermal loads permit higher power densities, which translates directly into smaller, lower-cost systems. Thirdly, high reliability and long component lifetimes are necessary to achieve high system availability.

These challenging requirements must be maintained in the face of a hostile environment that encompasses severe levels of radiation damage, dynamic and static mechanical loading, varying temperatures and heat loads, and exposure to gaseous or liquid coolants, tritium breeding materials, and the hydrogen plasma. It is generally accepted that in high performance power systems, the primary wall material must withstand an average neutron wall load of at least 2-3 MW/m 2 with a lifetime requirement of ~15 MWy/m 2; such an exposure would result in every atom in the structural material being displaced from its lattice site approximately 150 times.


Brief Summary of How We Got to This Point


Selection of material systems for development of structural materials for fusion power blanket systems is based upon the general performance targets discussed in the previous paragraph. The determination of material systems with the potential of meeting these performance targets is made through interaction between Advanced Design Studies, Plasma Chamber Technologies, and Advanced Materials Program tasks. Prior to the establishment of definitive safety and environmental attractiveness goals in the early 1980s, the program considered and eliminated Mo, Ni, Ti, and Al based alloys. The basic reason for elimination of these alloy systems was: for Mo, low temperature irradiation embrittlement and fabricability; for Ni, high-temperature grain boundary embrittlement; for Ti, excessive tritium inventory and permeability; and for Al, limited temperature capability. Niobium based alloys were eliminated on the basis of environmental impact considerations, e.g., any alloy containing more than a few parts ppm Nb could not be disposed of by shallow land burial. Molybdenum, Ni, and Al alloys would also be eliminated on the basis of environmental impact considerations. For designs utilizing bare metal walls, austenitic stainless steels are eliminated because of inadequate thermal-physical properties. Three material systems remain that are judged to have potential of being developed as fusion power system structural materials: SiC composites, vanadium-based alloys and advanced ferritic steels. Copper alloys, because of their excellent thermal and electrical conductivity, are critically important in near-term applications and will most likely find special applications in fusion power systems. Opportunities for consideration of new material systems may arise in the future as a result of advances within the broad field of materials science, or as the overall waste management strategy for fusion is re-evaluated.

Focus of Present Research


The current advanced materials program is focused primarily on three groups of materials with attractive low activation properties: (a) advanced ferritic steels, (b) vanadium alloys, and (c) silicon carbide composites. Research and development activities in these three systems are linked through a theory and modeling program that is closely integrated with the experimental program and that focuses strongly on crosscutting phenomena. Each group of materials has a unique set of characteristics.

The Advanced Ferritic Steels are furthest along the development path with a well-developed ferritic steel technology for nuclear, fossil, and other applications; the Advanced Materials Program has developed low activation versions of these commercially deployed materials with equivalent properties. These materials are resistant to radiation-induced swelling and helium embrittlement and are compatible with a range of aqueous, gaseous, and liquid metal coolants. There are feasibility issues associated with loss of creep strength at temperatures above 550°C, the potential for radiation-induced degradation of flow and fracture properties below 400°C and the possibility of design difficulties due to ferromagnetic properties. Current research is focused on: (a) collaborative irradiation experiments with Europe and Japan to understand the micromechanics of fracture processes, (b) development of methodologies to account for the effects of radiation-induced changes in properties on component integrity, and (c) approaches to improve creep performance and thus improve high temperature capability (e.g., oxide dispersion-strengthened alloys).

Because of their favorable combination of physical properties, compatibility with pure lithium, and relatively high creep strength, Vanadium Alloys based on the V-Cr-Ti system have the greatest potential of the three materials systems for high temperature operation in a liquid lithium-cooled blanket system. The current research is focused on feasibility issues which include (a) the development of self-healing insulator coatings to mitigate magnetohydrodynamic effects on coolant flow, (b) long-term effects of interstitial impurity pickup from the operating environment, and (c) welding methods. The lower operating temperature is limited by irradiation-induced degradation of fracture properties which occurs at temperatures below about 425°C. Work is in progress to understand the micromechanics of fracture in the transition regime and to develop methodologies to account for these radiation-induced property changes in assessing component performance. Thermal creep studies in representative environments are in progress to establish the upper operating temperature limits for V-Cr-Ti alloys.

The development of SiC/SiC Composite Materials presents the most difficult challenges of the three material systems. However, there are potentially high payoffs in terms of very low radioactivity and after heat and high operating temperatures. The primary feasibility issues involved in the development of SiC/SiC composite materials are: a) understanding the effects of neutron irradiation on the behavior of SiC fiber/interphase/SiC matrix structures, b) the limited technology base on the production, joining, hermetic sealing of composite materials, and c) the adverse impact of radiation-induced loss of thermal conductivity on allowable heat fluxes. The current program is aggressively addressing the effects of neutron irradiation through a series of reactor experiments conducted collaboratively with researchers in Europe and Japan. Based upon our current knowledge and understanding of radiation effects, advanced composite materials with improved properties are being fabricated in conjunction with Japanese researchers and US private industry partners through the SBIR program.

Time Scale for Development of Structural Materials


The process of materials development may be envisioned in terms of three overlapping and interconnected steps.

Initially, Feasibility Studies are undertaken to identify material systems that have the potential to meet physical, mechanical, and compatibility performance requirements combined with the potential to meet environmental and safety criteria. This stage is analogous to the Concept Exploration Stage in plasma physics studies. Research is undertaken to explore relationships between composition, microstructure and properties in order to evaluate pathways to achieve required performance.

When it is established that a material system has the basic characteristics required of a structural material, it enters the Materials Developmental stage. This stage is analogous to the Proof of Principle Stage in plasma physics studies. Experiment, theory, and modeling are combined in the process of defining specific compositions and structures that will yield the desired properties. The development process focuses initially on improving those properties that clearly limit performance. As development proceeds, however, all aspects of materials performance must be addressed including primary production, fabrication, welding and joining, chemical compatibility with coolants and tritium breeders, and the complete range of mechanical properties. Neutron irradiation affects most physical, chemical, and mechanical properties and thus becomes an overarching consideration.

The Materials Engineering phase provides the extensive materials property data base required for design, licensing, construction and operation of a large complex fusion power system. This stage is analogous to the Proof of Performance Stage in plasma physics studies. All aspects of materials performance must be considered including response to synergistically interacting environmental, radiation and metallurgical variables. The product of this phase is a specification for producing materials in the required product forms and a validated database on properties and structural assessment methods to support design, construction, and licensing.

The time scale of this complete process is measured in tens of years and is closely integrated with the advanced technology and advanced design components of the overall fusion program. A comparable scenario involving a simpler design and radiation environment was the development of advanced cladding and duct alloys for the U.S. liquid metal fast breeder reactor program. That activity spanned a period of approximately 25 years in reaching well into the materials engineering phase. There are several recent positive interactions between the Advanced Materials and advanced design programs. For example, recent experiments established that the lower operating temperature for vanadium alloys is ~400°C due to radiation embrittlement effects. The design study team modified their coolant inlet temperatures to account for this effect, and this design change highlighted the need for thermal creep data at high temperatures (creep experiments are currently in progress). Similar iterative interactions have recently occurred for SiC/SiC composites (impact of irradiation on thermal conductivity).

The Importance of a Fusion Neutron Source


The foremost challenge in developing materials for use in the blanket system of fusion power reactors is to understand and limit the degradation of mechanical and physical properties that occurs in materials when exposed to neutron irradiation. The intense flux of high-energy neutrons in the fusion reactor blanket region will create irradiation damage in the form of atomic displacements and transmutations. In the required lifetime of the conventional fusion blanket system the level of displacement damage in the structural material will exceed 150 dpa (displacements per atom), a level much greater than encountered in light water reactors. Totally unprecedented levels of the gases helium and hydrogen will be produced by transmutations. Essentially all of the mechanical and many of the physical properties of materials are altered by irradiation damage at these levels. At present, irradiation experiments in fission reactors are the primary means of investigating the effects of irradiation on the properties of materials. This approach has significant limitations because the neutron energy and flux are lower than anticipated in fusion reactors. The approach has been to investigate regions of damage space in which the damage levels are low and in which displacement damage generally dominates the material response. We can investigate many of the underlying mechanisms and explore approaches to overcome problems encountered in the accessible parameter space. However we can neither explore phenomena nor develop materials with suitable properties over the entire parameter space encountered in service. To accomplish this a fusion materials neutron source will ultimately be required. A suitable neutron source must reproduce the irradiation damage, particularly in terms of gaseous transmutation products and atomic displacements. It must have sufficient neutron flux and fluence capabilities to allow accelerated testing and to achieve end-of-life damage levels. The international fusion materials community has proposed an accelerator facility, the International Fusion Materials Irradiation Facility (IFMIF) that will meet these basic requirements. Development and qualification of structural materials in some type of fusion neutron source will be required before embarking on the design of a high-fluence DT fusion power system.

The Central Role of International Collaboration


International collaborations have been the hallmark of the Advanced Materials Program for two decades. Formal bilateral collaborations that involve joint planning, funding, testing and data analysis, and reporting of materials research and development have been in existence with Japan Atomic Energy Research Institute and Monbusho (Japanese Universities) for over 15 years. These collaborations have been crucially important to the vitality of the U.S. Fusion Materials Program. Through these two collaborations, the U.S. program receives approximately $2.5 M/y for the Japan share of joint experiments. An important bilateral collaboration exists with the Russian Federation (RF) that provides access to the RF fast reactor BOR-60 for radiation effects research. The international fusion materials research and development effort is coordinated through the IEA Implementing Agreement on a Programme of Research and Development on Fusion Materials. Canada, Japan, the European Union, Switzerland, the Peoples Republic of China, the Russian Federation, and the United States are participants. Fusion materials research activities coordinated through the IEA include SiC composites, vanadium alloys, advanced ferritic steels, beryllium technology, ceramic insulators, conceptual design of IFMIF, test specimen miniaturization technology, irradiation damage theory and modeling. Within the national laboratories and universities the fusion materials program is generally embedded in the organizational structures (divisions, departments, etc.) in which other materials R&D is carried out. This co-location provides extensive interactions with materials scientists and engineers working on materials problems ranging from basic to applied.

Leveraging the Broad Field of Materials Science Research


Fusion materials research is an extremely small fraction of the broad field of materials research. Our strategy is to focus on those areas of technological need that are unique to fusion and to leverage other areas of materials science research to the maximum extent possible. Examples of areas of materials science in which we rely predominantly on other programs for developments include: advanced techniques for structural characterization such as analytical electron microscopy, atom probe-field ion microscopy, synchrotron x-ray facilities, development of advanced materials synthesis and processing techniques, and basic studies of the relationship between structure, composition and properties of metals and ceramics. Funding for research in these and related areas comes primarily from DOE-Basic Energy Sciences, NSF, and DOD.

The Prospects for Success


The development of materials that will enable fusion to be realized as an economically competitive, safe, and environmentally attractive energy source is a major challenge in materials science. In few if any other applications are the demands that are placed on a structural material so all encompassing. In many applications only one or two mechanical or physical properties are essential to performance. In the case of fusion blanket structural materials, essentially all mechanical and physical properties along with chemical compatibility with coolants and breeding media play critical roles. Irradiation damage to structural materials in the high-energy neutron spectrum of a fusion reactor is unique; it affects most mechanical, physical and chemical properties, and cannot be completely duplicated in currently available fission reactor facilities.

We are not faced with a matter of characterizing and possibly extending the performance of existing materials. Rather, we must develop totally new materials tailored for service in the fusion environment while capitalizing on our ability to control the characteristics of the induced radio-activation This requires a sustained, focused, and scientifically sound materials research and development program, and we have the core of such a program currently in place. In materials science we have the ability to characterize and manipulate the structure of materials at the atomic level. We are developing understanding and models of the physical processes that control properties over length scales from the interatomic to macroscopic and time scales from nanoseconds to years. Today we have in technological use metals and ceramics with properties that were once thought impossible to achieve. Nothing inherently dictates that the development of materials that satisfy the goals of economic, safe, and environmentally attractive fusion cannot be accomplished.

For More Information

  1. “Advanced Materials Program, A Whitepaper/Roadmap” by E. E. Bloom, N. M. Ghoniem, R. H. Jones, R. J. Kurtz, G. R. Odette, A. F. Rowcliffe, D. L. Smith, and F. W. Wiffen can be located at http://www-ferp.ucsd.edu.

  1. “A Whitepaper Proposing an Integrated Program of Theoretical, Experimental, and Database Research for the Development of Advanced Fusion Materials” by R. E. Stoller, G. R. Odette, and H. L. Heinisch can be located at http://www-ferp.ucsd.edu.

3. Fusion Energy Sciences Advisory Committee Reports on Review of Fusion Materials Research Program, Review of the Proposed Proof-of-Principle Programs, Review of the Possible Pathways for Pursuing Burning Plasma Physics, and Comments on the ER Facilities Roadmap , DOE/ER-0727, July 1998.

  1. Proceedings of the Eighth International Conference on Fusion Reactor Materials (ICFRM-8), Sendai, Japan, October 26-31, 1997, published in J. Nucl. Mater. 258-263 (1998) eds., C. Kinoshita, T. Muroga, H. Matsui, A. Kohyama, and D. S. Gelles.

2.0 IMPACT OF MATERIALS SELECTION ON SAFETY AND LONG-TERM ACTIVATION IN FUSION POWER PLANTS (D. A. Petti, K.A. McCarthy, et al.)

Neutron-induced transmutation of materials in a D-T fusion power plant gives rise to long-term activation products. This requires special attention to ensure that the attractive safety and environmental characteristics of fusion power are not degraded. By careful design choices it should be possible to produce power from fusion without generating an environmental burden on future generations. To achieve this goal it is necessary to optimize fusion power plant design to minimize both the level of activation and the total volume of active material that might ultimately be categorized as waste requiring disposal. Materials selection is central to this optimization.

The use of low activation materials in regions of high neutron flux is an important part of the strategy. However, the minimization of total volume of active material can be achieved only by providing adequate neutron shielding of massive components such as the vacuum vessel and magnet structure. Scoping studies have examined the influence of blanket materials choices on activation of these items, and concluded that low activation materials with the best performance as blanket structure may not be optimum from the global plant perspective.

These studies have enabled some underlying trends to be identified, leading to recommendations for a strategy in which low activation materials choices are made with a broader view of the plant long-term activation performance.

(See Snowmass Chamber Science & Technology CQ5 draft opinion paper for further details, http://www.fusion.ucla.edu/snowmass/questions/chamber.html.)

3.0 PREDICTING THE BEHAVIOR OF MATERIALS IN SOLID BREEDER BLANKETS (M. Billone, A. Ying, S. Malang, et al.)

Solid Breeder blankets are composed of structural material, ceramic breeder, neutron multiplier, and coolant. For power plants, the main candidate materials are: Ferritic/ martensitic steels, oxide dispersion strengthened (ODS) steels or SiC f/SiC-composites as structural material, Li-ceramics such as Li 2O, Li 4SiO4, Li 2TiO3, and Li 2ZrO3 as breeder material, beryllium as neutron multiplier, and helium as coolant. Possible forms of the ceramic breeder and beryllium multiplier are pellets, plates or pebble beds. In most blanket designs pebble beds are employed in order to avoid excessively high thermal stresses in the breeder material and unpredictable thermal resistances in gaps between beryllium plates and cooling plates, and to facilitate tritium release from breeder and multiplier.

Thermal and mechanical behavior of all the materials involved is strongly dependent on the operating temperature and can change decisively by the high fluence neutron and high Li burnup irradiation in a breeding blanket. Higher temperatures can lead to plastic deformation/creep of breeder and multiplier. The neutron irradiation can cause swelling, irradiation-induced creep, and hardening of all materials employed. All these changes can have a strong impact on the heat transport between breeder, multiplier, and cooling plates as well as on the mechanical interactions between pebble beds and cooling plates. In general, the thermal performance of high conductivity materials such as Be is more sensitive to pebble-to-pebble and pebble-to-wall mechanical interactions than are low conductivity lithium ceramics.

On the other hand, it has to be ensured that all limits with regard to maximum temperature, stresses and strains, deformations, and tritium inventory in the entire blanket are not exceeded during the lifetime of a blanket. This asks for suitable out-of-pile experiments as well as high fluence irradiations of all materials involved in order to provide a sufficient data base for models required for the prediction of the thermal-mechanical behavior of the pebble beds and the blanket structure.

This section will address the status of the knowledge with regard to material properties including the irradiation impact, and the state-of-the-art in modeling the behavior of the blanket under realistic loads in typical geometries.

4.0 LIQUID BREEDER/COOLANT MATERIALS (R. Moir, N. Morley, et al.)

The idea of using thick liquids as a FW/Blanket changes the requirements on structural materials.  Neutron damage, volume and surface heating and primary loads are all altered from standard solid wall design concepts and present new challenges to the materials community.  Several opportunities for materials research and development for liquid wall concepts are outlined below.

  1. Backing Wall for Thick Liquid Layer Concepts

For Thick Liquid Walls you need a backing wall of structural material following 40 to 70 cm of liquid (lithium, Sn-Li, or flibe). The backing wall is ~ 1-3 cm in thickness. With Li and Sn-Li, we will likely need insulator coatings too. With flibe we do not need an insulator.

What structural material should we use in the backing wall? The flux, helium production and dpa are about two orders of magnitude lower than in a solid first wall (depending on the design). However, we want the lifetime to be that of the plant lifetime. We also want higher wall load. So, we still need ~ 40 to 100 dpa (depending on design) capability in the backing wall. The helium to dpa ratio is significantly lower than in a solid first wall. Hopefully, this will make it possible to operate the backing wall at a somewhat higher temperature than in a solid first wall. Also, this might make it easier to extrapolate the fission irradiation data with more confidence.  One relevant question here is the operating temperature of the backing wall. We suggest that we first look at moderate temperatures (600-650°C) and later we can examine the possibility of designs with higher temperatures.

So, the backing wall requirements are about 40 to 100 dpa (for plant lifetime, range depends on the liquid and its thickness), 600-650°C, and the rate of transmutation is considerably reduced relative to normal solid first walls. The minimum operating temperature does not seem to be an issue because the temperature rise in the liquid is generally about 50°C.

So, the question is what does the materials community recommend for structural material for each of the three candidate liquids? Is there a fundamental change in thinking that comes about due to this set of back-wall requirements such that austenitic steels should be considered (304 stainless is considered by the IFE community)? Is this a good choice? How about Ferritic Steel, ODS Steels, or Vanadium? Other suggestions?

  1. Backing Wall for Thin Liquid Layer Concepts

Here there is only 2 cm of liquid in front of the structural backing wall. So, relative to a traditional solid first wall, we just eliminate the thermal stress due to surface heat flux. There is nuclear heating. The operating temperature is also about 600-650°C. The question here is what is the preferred choice for structural material for the backing wall under these conditions?

  1. Nozzles and other Near-Plasma Structures

The nozzles will use structural materials, so will RF antennae and other plasma support structures. There are no design details yet of the operating conditions. Nozzles will be at a somewhat lower temperature than the Backing Wall, but have higher primary stress (a few atmospheres) and a higher nuclear heating rate. Reliability is an important consideration here. Should we assume the same structural material as the backing wall? Better suggestions?

  1. Other Areas

There are other areas that have material issues. They are not yet well defined and the input of the materials community is greatly needed. We should keep an eye on them as designs evolve. Eventually, I would like us to consider the question of the vacuum vessel (this is really a generic issue, not specific to liquid walls). The 1 appm helium limit for rewelding adopted in ITER is a serious constraint for fusion designs. 

Please feel free to consult and comment on the Opinion Paper on Chamber Technology Q1 - Liquid Walls for IFE and MFE available at

http://www.fusion.ucla.edu/snowmass/questions/chamber.html

5.0 PLASMA-FACING MATERIALS (M.A. Ulrickson, et al.)

(See Snowmass Chamber Science & Technology CQ2 draft opinion paper for further details, http://www.fusion.ucla.edu/snowmass/questions/chamber.html)

6.0 IFE MATERIALS (L.L. Snead, et al.)

As with magnetic confinement concepts, the range of material property requirements for inertial confinement concepts is extreme. However, due to the significant effort and study which have been put into the MFE design studies such as ITER, the critical design-limiting properties are fairly well understood. For the case of IFE, the myriad challenges are not as clearly understood. Presently, there are two competing IFE concepts being put forward: 1) Laser Driven Option, and 2) Heavy Ion Driver Option. While both concepts are similar in that they implode a D-T containing target at several Hertz, the handling of the heat and neutron fluxes following the fusion event is quite different. However, the factors which ultimately limit the design of fusion machines (whether ITER-like MFE concepts or laser or ion driven ICF concepts) can be generalized into the following areas: 1) Extreme surface and volumetric heat loads, 2) Radiation damage to optical/diagnostic systems, 3) Radiation damage to structural materials, 4) Plasma/wall interaction (i.e. sputtering), 5) Tritium retention, and 6) Safe, Economic and Environmental acceptability. In addition to these areas, the fundamental IFE materials question of the development of economically viable target materials can be thought of as a materials issue. While some of these areas can be mitigated through intelligent design, this often increases the magnitude of some other problem. A brief outline of the two competing IFE designs is given below to illustrate trade-offs in materials design challenges.

Laser Driven Option

An example of a laser-driven IFE plant is the SOMBRERO design with dry walls of carbon fiber carbon matrix composite (CFC), or other composite, surrounding the D-T target. The SOMBRERO CFC cavity is about 8 m in diameter and 20-m high. Inside the cavity is Xe to stop the particles and gamma rays from reaching the walls. The wall is cooled by high velocity helium or helium/particulate coolant. The cavity temperature is 1000-1400°C and would receive about 15 displacements per atom (dpa) per year. The assumption is that the cavity lifetime is 3-5 years. In order to place the KrF laser beam onto the target, SOMBRERO called for 60 penetrations through the wall.

Key materials feasibility issues for this design have been recently identified as follows:

  1. Chamber Lifetime Uncertainty. X-ray and debris damage to the first wall must be prevented; neutron damage life of first wall and blanket structures must be acceptably long, probably at least 1 year depending on replacement time; possible erosion of coolant channels by flowing granular coolant/breeder. It should be noted that the SOMBRERO study was never completed and the first wall lifetime was based on optimistic assumptions regarding the neutron-damaged graphite composite liner.

  1. Final Optics Survivability. Damage from laser light, x-rays and debris must be prevented; neutron damage life must be acceptably high (also greater than 1 year); optics must be mechanically stable against gas shocks even after attenuation up the beam line.

3) Safety and Environmental. An improved power plant design needs to meet the “No-Public-Evacuation-Plan” criteria; tritium containment and tritium inventory concerns (especially in the graphite) must be resolved; end-of life materials processing (recycling and radioactive waste disposal) must be acceptable.

Heavy Ion Beam Driver Option


Heavy-ion fusion is possible in principle using a wide variety of targets and chambers, including direct-drive, indirect-drive, and fast-ignition targets, and dry, wetted-wall, and thick-liquid wall chamber concepts. However, recently the combination of indirect-drive target geometry, and liquid protected chambers are being highlighted. The design studies performed to date on these flowing-liquid-wall IFE concepts have not had sufficient engineering detail to allow exploration of all of the significant questions (and opportunities) related to such a design. However, it is clear that the liquid wall concept offers several advantages, including: a) the reduction of lifetime dpa for the high-volume structural material components from tens to a few dpa, possibly reducing the radiation damage issues, b) through moderation of the fusion neutrons the need for a fusion neutron source for materials studies is mitigated, c) plasma particle erosion of first wall components is reduced and possibly eliminated.

Key materials feasibility issues for this design have been recently identified as follows:

  1. Final-Focus/Chamber Interface. Physical accommodation and shielding of final focus magnet arrays consistent with chamber solid-angle limits, required number of beams, magnet dimensions and neutron shielding.

  1. Safety and Environmental. Need improved power plan design that meets the “No-Public-Access-Plan” criteria and minimization of waste. Requires chamber/plant design improvements, improved data on important radionuclide release fractions (including hohlraum materials.) Study of the radiological issues with the use of high volumes of liquid need to be studied. The benefit of reduced flux to metallic structural components offers the possibility of reduced high level waste if low activation materials are employed.

  1. Material interaction. Corrosion and mass transport in a flowing liquid wall design is a radiological, safety and lifetime concern.

  1. “Neutron Damage to Structural Materials.” At present, there is no proof of principle experiment or detailed design available for the thin or thick flowing metal walls. However, some type of flow-directors in high-flux inlet and exit regions will be needed. The effect of neutron streaming through beam ports on the vacuum vessel accumulated dose needs additional evaluation.


  1. Fiber Optic Survivability

  1. MAGNET MATERIALS (R.W. Wooley, L. Bromberg, et al.)—the following is pasted from the Snowmass Technology PQ2 prospectus by J. Minervi and R.W. Wooley, see http://www.fusion.ucla.edu/snowmass/questions/Qplasma.html


Magnetic fields are required for containment and control of the plasma in magnetic confinement fusion (MCF). Many configurations rely on both dc and pulsed magnetic fields for plasma initiation, confinement, ohmic heating, inductive current drive, plasma shaping, equilibrium and stability control. Inertial Confinement Fusion (ICF) may also utilize magnetic fields, for example, in beam focusing quadrupole fields in the Heavy Ion Driver (HID) concept or Helmholtz coils in KrF lasers. These magnets may use either resistive conductors or superconductors. The preponderance of past and present magnetic fusion devices use normal resistive magnets, but almost all of the large fusion machines being built outside the United States (e.g. LHD, Wendelstein VII-X, KSTAR, SST, and HT7-U) use superconducting magnets.

Most design concepts for power producing commercial power plants depend on superconducting magnets for efficient production of magnetic fields. The attraction of superconductivity is the ability to carry very high current density with zero dc power dissipation. Superconductors will dissipate energy in a changing magnetic field, but overall power losses, including refrigeration power required to maintain the magnets in the superconducting state (typically 4K-8K temperature range for Low Temperature Superconductors (LTS)), are extremely small compared with resistive magnets. This advantage grows with increasing magnetic fields and magnetic field volume, or where relatively long pulse or steady state operation is required.

There are also design concepts for power producing commercial power plants based on dc (steady) resistive magnets or combined resistive and superconducting magnets (e.g.- ARIES-ST), particularly for confinement concepts promising high beta. The attractions of dc resistive magnets over superconducting magnets include: 1) potentially increased compatibility with DT fusion neutron and gamma radiation environment, 2) the possibility of lower initial capital cost for the magnets, 3) elimination of capital expenses for superconducting magnet support systems such as quench protection energy dump systems, or such as cryogenic cooling systems if room temperature magnet operation is used, but partially offset by a water treatment/cooling system, 5) demountable joints, which simplify both the initial assembly and the maintenance of magnet components in a radiation environment, while also allowing the efficient centralized industrial manufacturing of magnet modules small enough to be shipped to the power plant site on roads. Although resistive magnets do consume part of the fusion power produced, their recirculating power fraction disadvantage is minimized by increasing magnetic field strength and increasing plasma volume. The desirability of this approach increases with increasing plasma beta.

Magnets require a high level of electrical, mechanical and structural engineering design and technology, advanced materials, as well as supporting cryogenic technology if they are superconducting or cooled with a cryogen (e.g.- liquid nitrogen). At present, only the Low Temperature Superconductors with critical temperatures of ~10K for the ductile alloy NbTi or ~18K for the brittle compound Nb3Sn are in use or planned for future devices. The newer High Temperature Superconductors (HTS) with critical temperatures of order 90K and above are expected to be used primarily for low-loss magnet current leads in the short term. Longer-term application of HTS depends on progress in development of materials in long lengths with significantly improved critical current densities. The primary area for use would be in magnetic fields of 0.5-20 Tesla and at higher operating temperatures than typical LTS. An essential issue is that the HTS must be available at similar cost (including refrigeration cost impact).

Some of the key technical issues for dc (steady) resistive magnets designed for use in commercial power plants are similar to and simplified from the case using superconducting magnets: 1) steady active cooling optimization within mechanical stress limits, 2) low cost, 3) radiation compatibility, 4) reliability/availability/maintainability. The relative importance of these issues depends on the concepts, such as plasma beta, size and field strength.

Significant success in development of magnet technology for fusion applications has been achieved over the past two decades, but no large fusion programs using superconductors or resistive conductors are now underway in the U.S. Countries that have been more aggressive in introducing superconducting technology with working tokamaks and stellerators include France, Russia, China and Japan. Germany (Wendelstein VII-X), Korea (KSTAR), India (SST-1) and China (HT-7U) all have significant programs underway for introducing new superconducting steady state stellerators and tokamaks in the next few years. The last large operating fusion experiments with resistive conductors were produced in Europe and Japan whereas smaller scale devices were built in the US (i.e.- D-IIID and Alcator C-MOD).

For the U.S. fusion program, the main issues for magnet technology revolve around cost and reliability. The United States should be developing the elements of magnet technology that are specifically focused on the experimental needs of the magnetic and inertial fusion physics programs and that will substantially lower the projected cost of the experiment or of fusion power in the future. There are primarily three ways in which magnet technology can lower the cost of experiments and fusion power production: 1) by lowering the cost of the magnet components and/or assembly processes, 2) by reducing the size of the magnet systems, so that the cost of other fusion subsystems may be reduced, and 3) by providing magnet performance which substantially increases or optimizes the physics performance, e.g. increased magnetic field or some special magnetic field configuration.

The ‘goodness’ parameters for a magnet system need to be defined for specific applications, but are obvious for some requirements. For instance, a strong benefit results from higher magnetic field, since fusion power is proportional to B4. Another benefit would result from the ability to absorb higher nuclear flux and fluence in insulation systems or to have increased superconductor stability in order to reduce the size of the radiation shield protecting the magnet system and consequently reduce the machine size and cost as a whole. Such shield size reduction might be limited in a DT fusion power plant because most neutrons are needed for power production and tritium breeding and so must be intercepted in any case by the blanket.

In general, assembly and maintenance access is less constrained if the magnets can operate at higher current density and/or higher stress levels, thus reducing the size of the hardware producing the field and the surrounding interface systems, or if resistive magnets are used that are jointed and demountable. In the longer term, the cost of superconducting magnets and their supporting systems may be reduced if their operating temperature can also be increased (e.g.-higher temperature superconductors) or if they can be constructed using less expensive materials. Therefore, magnet technology should strive toward operating at higher fields, higher current densities, higher stress levels and higher temperature levels. Each of these improvements implies smaller devices for lower cost field production. In parallel, design margins must be sufficiently understood and applied so as not to sacrifice reliability.

Thus, besides reducing cost, the thrust of a superconducting magnet technology improvement should be towards increasing the limitations on design margin (e.g.-high stability. low AC losses, and rapid quench detection in superconductors) and improving materials (e.g.-higher superconductor critical properties, and high shear strength/high radiation resistance insulation) to enhance performance or reduce size. Steady dc resistive magnet technology should also strive to operate at the highest fields and stress levels that are compatible with demountability and maintainability and require improved materials properties for high strength/high conductivity alloys or laminates of copper/steel for pulsed coils, and also high shear strength/high radiation resistance insulation.