DRAFT

DRAFT

Report to the

U.S. Department of Energy

Office of Fusion Energy Sciences

Possible Pathways for

Pursuing Burning Plasma Physics

and

Fusion Energy Development

July 17, 1998

Coordinated by C. Baker and prepared by many contributors

TABLE OF CONTENT

1.0 Introduction Page 2

1.1 Purpose and Background Page 2

1.2 Community Process Page 2

Summary Page 2

2.0 Reduced Cost ITER Integrated Step Page 10

2.1 Pathway for a Reduced Cost ITER Page 10

2.2 Rationale for Reduced Cost ITER Pathway Page 12

2.3 Technical Contributions of a Reduced Cost ITER Page 13

2.4 Fusion Development Pathway Implications

of a Reduced Cost ITER Page 16

2.5 Pros & Cons of a Reduced Cost ITER Strategy Page 17

2.6 Near Term Actions Page 18

2.7 Summary of Technical Appendices on

Reduced Cost ITER Page 18

3.0 Modular Program Pathway Page 21

3.1 Pathway Overview Page 21

3.2 Rationale for the Modular Program Strategy Page 23

3.3 Technical Contributions of the Modular Plan Page 24

3.4 Pathway Implications for the Modular Plan Option Page 36

3.5 Advantages and Concerns for the Modular Program Pathway Page 36

3.6 Near Term Actions for the Modular Strategy Page 37

4.0 Enhanced Concept Innovation Pathway Page 39

4.1 Pathway Overview Page 39

4.2 Rationale Page 41

4.3 Technical Contribution Page 42

4.4 Advantages and Concerns Page 46

4.5 Near Term Actions Page 46

POSSIBLE PATHWAYS FOR PURSUING BURNING

PLASMA PHYSICS AND FUSION ENERGY DEVELOPMENT

1.0 Introduction

1.1 Purpose and Background

This report has been prepared in response to a request from the U.S. Department of Energy's (DOE) Office of Fusion Energy Sciences (letter sent by Dr. A. Davies to Dr. C. Baker, January 28, 1998) to consider possible alternatives on reduced cost options for "next-step" devices. A central focus of next-step devices is the study of "burning" plasmas which explore the impact of substantial fusion energy production via the deuterium-tritium reaction.

An important part of the U.S. Fusion Energy Sciences Program is its participation in the International Thermonuclear Experimental Reactor (ITER) program. Taking into account the international situation and U.S. domestic issues, the ITER process is exploring reduced-cost options to the present ITER device. A Special Working Group, reporting to the ITER Council, has been formed to explore these issues on behalf of the ITER Parties, i.e. the European Union, Russian Federation, Japan and the U.S. This report, and its related activities, will aid the U.S. in the international process.

1.2 Community Process

This report is the result of a broad-based U.S. community effort to discuss, debate and work together on the crucial issues involved in considering next-step options. The main content of this report is based on three potential pathways identified at a broadly-attended community Forum for Next-Step Fusion Experiments (University of Wisconsin, Madison, April, 1998) organized principally by the University Fusion Associates and by the work of the ITER Steering CommitteeUS (ISCUS) on reduced cost ITER options. The Madison Workshop was followed by a smaller Workshop on Next-Step Options (University of California, San Diego, June, 1998) to focus on preparing this report. A broadly-announced Web Site was established to facilitate access to documents related to this process.

Summary

The mission of the Fusion Energy Sciences Program is to advance plasma science, fusion science and fusion technology - the knowledge base needed for an economically attractive fusion energy source. The policy goals that support this mission include:

• advance plasma science in pursuit of national science and technology goals; • develop fusion science, technology and plasma confinement innovations ; and • pursue fusion energy science and technology as a partner in the international effort.

A key aspect of the third goal in particular is the study of the physics of burning plasmas. (This is sometimes referred to as the "third leg" of the program). This report describes potential pathways towards this goal that have been identified within the U.S. fusion community (most notably at the Madison Forum for Next-Step Fusion Experiments, April, 1998). There is a strong linkage between this element of the program and the other "two legs" of the U.S. program, i.e., to advance plasma science and concept innovation. The pathways described herein depend directly on the continuation of a viable and strong base program with adequate resources. This base program provides the underlying science and technology critical to the development of fusion energy and the study of burning plasmas. This report is predicated on the presumption that the present base program funding level of somewhat more than $200M/yr. will be continued.

The purpose of this report is to describe pathways for the U.S. to pursue burning plasma physics and fusion energy development. The first pathway features continued participation in the ITER design effort, now focused on a reduced-cost device, "ITER-RC". This device would be designed to achieve the overall programmatic objective of the original ITER device, but with somewhat reduced baseline plasma performance goals within nominal physics assumptions and the possibility of achievement of the full ITER mission if advanced physics performance can be realized. The ITER-RC tokamak would integrate a moderate-to-high energy gain plasma with a steady-state system including superconducting magnets as wells as some nuclear technology testing. The second tokamak pathway would include two devices rather than the ITER-RC: a copper-coil device, capable of DT burning plasma experiments, and a steady-state, advanced device with superconducting coils operating predominately DD plasmas. (In previous extensive design studies, substantial effort has been devoted to the design of devices suitable for both these two pathways.) The third pathway chooses to delay the burning plasma and steady-state steps to focus resources on enhancing concept innovation to prepare for later, but hopefully more affordable steps to the burning plasma and fusion energy development stage. Besides the facilities described above, all three pathways will require, to differing degrees, additional facilities for fusion technology development, such as a point neutron source and/or a volume neutron source.

The ITER-RC pathway puts stronger emphasis on physics and technical integration, facing many important and difficult integration problems in the near term. The modular pathway separately addresses the physics issues of burning plasmas and steady-state and also addresses some issues of integration (e.g. remote maintenance in at DT environment, superconducting coils in a tokamak magnetic environment) prior to initiating an integration step. The ITER-RC pathway is the shortest pathway for fusion energy development but is also the pathway which would require the largest initial funding outlay. The separate devices in the modular approach may provide greater flexibility for concept innovation allowing a more advanced integration step to follow. It is important to note that both ITER-RC and the modular pathway contribute physics and technology information to other lines. The Enhance Concept Innovation pathway puts increased emphasis on concept improvement in the tokamak, related concepts, and concepts more distant from the tokamak, including inertial fusion energy (IFE) prior to initiating either the modular pathway or the integrated machine pathway. This pathway is already a key aspect of the base program and thus also contributes to the possibility of pathway 1 and 2. Each of these pathways are aimed at the same long-term goal but will arrive there with different time scales, costs and degrees of technical challenges and risks.

The Promise of Fusion

The benefits of Fusion R&D were clearly seen by the President's Committee of Advisors on Science and Technology (PCAST) in the 1995 Report of the Fusion Review Panel. Their stated views are increasingly valid today and are quoted below.

"The principal objective of the U. S. program of fusion energy research and development is to provide this country and the world with an abundant, safe, environmentally attractive, and cost-competitive new energy source. Achieving this objective would bring large benefits almost irrespective of how the energy future unfolds; and achieving it could be crucial of society finds it necessary, for environmental or political reasons, to reduce sharply the currently dominant role of fossil fuels in world energy supply."

"In the course of pursuing this energy goal, fusion R&D yields an immediate and continuous additional benefit by nourishing an important branch of basic science - plasma physics - and the technologies related to pursuing it. This field of research, for which nearly all of the funding comes from fusion energy R&D budgets, has been prolific in the production of insights and techniques with wide applications in other fields of science and in industry."

"Finally, for a variety of reasons, fusion energy R&D has evolved a higher degree of international scientific and technological cooperation than any other field of scientific or technological research. This cooperation - entailing not only extensive exchanges of personnel and information but also full-fledged international collaboration in design, construction, and operation of some of the largest experiments - is in itself a valuable model and precedent for internationalization of R&D in other fields. Such cooperation is likely to become increasingly important as the costs of cutting-edge R&D continue to grow in relation to the capacities of individual nations to pay for it."

Progress in Fusion

The research field of Fusion Energy Science has in the last two decades made major scientific advances. Supported by large investments in the late 70's and early 80's, facilities capable of producing and sustaining plasmas with fusion relevant parameters were built and successfully operated. The science of plasma measurement techniques was developed; today nearly all quantities of relevance needed to compare to theories are measured with sufficient precision. Plasma stability limits have been explored experimentally and the results are quantitatively predicted with high accuracy. The theoretical minimum in cross-field plasma ion transport has been reached in some circumstances. Plasmas with high degrees of recombination before reaching a material surface have been produced, fulfilling the simplest vision of magnetic confinement as using the magnetic field to prevent hot plasma from touching the material wall of the confining chamber. Methods to drive the current in low-to-moderate density plasmas with auxiliary means have been successfully employed; the efficiencies of these methods are in accord with theory and code modules exist to calculate the driven currents. The theoretically predicted self-driven or bootstrap current has been confirmed and has significantly enhanced prospects for steady-state operation. The result of these advances has been a confirmation of theory in most areas and a computational basis of plasma understanding sufficient for predictive projection of the performance of next step devices. These recent advances also show the commonalty of the issues for fusion concepts (for example, various toroidal concepts) and how advances in one concept (e.g., tokamaks) can be exploited in alternate configurations.

This advance in scientific understanding was made possible by supporting advances in fusion technologies. Magnetic coil systems and their associated feedback control systems were developed to stably confine the plasma equilibrium and produce many variations of plasma shape for optimization of plasma performance. Plasma heating technologies were developed and deployed at the tens of megawatt level; these systems were indispensable in the studies of plasma stability and current drive and also now are the basis of many important plasma measurement techniques. Superconducting coils have been used in magnetic confinement systems and pulse lengths exceeding two hours have been produced, clearly showing the potential for steady-state. Tritium fueling systems were implemented and safely operated resulting in the large scale production of fusion power (11 MW in the TFTR tokamak and 16 MW in the JET tokamak) and over 1 gigaJoule of fusion energy produced.

The Stages of Fusion Development

In order to discuss the status of fusion progress and future pathways, it is useful to introduce the five stages of fusion concept development as described in the 1996 report by the Alternative Concept Panel formed from the Science Committee of the Fusion Energy Advisory Committee. There are five development phases applicable generally to all fusion concepts:

• concept exploration

• proof of principle

• proof of performance

• fusion energy development

• fusion power plant deployment (begins with a DEMO plant).

The concept exploration phase is generally implemented on small experiments aimed at particular new innovations. The concept exploration phase can also include particular new developments on larger facilities.

A proof-of-principle program has as its main goal the resolution of key scientific issues in depth and on a broad front. A proof-of-principle level program is generally implemented in intermediate sized devices which are capable of investigating a complete set of key issues in depth, of producing plasma parameters approaching reactor conditions, of achieving extensive control capability and of using a comprehensive set of diagnostics. In the United States, examples of proof-of-principle level devices are the Alcator C-Mod and DIII-D tokamaks.

The defining feature of the proof-of-performance level phase is the need for plasma parameters needed to minimize the extrapolation to the following more costly fusion energy development steps. The clear examples of devices in this class are the tokamaks JET in the EU, JT-60U in Japan, and TFTR in the U.S. The DIII-D and Alcator C-mod are sufficiently capable devices technically to make some contributions at this program level.

The fusion energy development phase is mainly defined by devices which can study deuterium-tritium burning plasmas and integrate reactor relevant technology. Another goal that often enters at this stage is steady-state operation with its associated technology implications such as plasma power exhaust and superconducting magnets. These two sets of issues and their integration cannot usually be addressed in devices associated with the proof-of-performance level research. It is a strategic issue of critical importance that the scientific issues of burning plasmas and steady-state cannot be addressed at the proof-of-principle level, but require facilities of significant scale and cost.

Role of the Base Program

The scientific and technological progress in the program cited above has been and will continue to be derived from a strong and healthy Fusion Energy Science base program. This base program encompasses the embryonic concept exploration stage up through the proof-of-principle stage. The scientific progress to date gives good confidence that many concepts can and should be brought through the proof-of-principle stage. The efforts on concept improvement are also contained in the base program. Given the large number and diversity of fusion approaches, an essential element of fusion strategy is the view that sequentially and over a period of time selected concepts may be moved up through the five development phases. Although the advances in the field cited above were primarily made using the tokamak confinement device, these general advances engender the anticipation that a similar level of scientific maturity can be realized for a number of other fusion approaches. The Base Program mission to bring a number of fusion concepts through the proof-of-principle stage is an essential and enduring fusion strategy element that will be pursued as a component of any larger strategy. Which concept will make the ultimately best fusion power system is a question that will be answered over time. The immediate strategic issues revolve around which concepts to advance in what order and at what pace.

Fusion Development Steps

Most of the scientific and technology progress cited above has centered around the tokamak concept. This concept was seen in the late 70's as the concept most likely to be capable of producing high performance plasmas. The research done with the tokamak has borne out this early view. The fusion program strategy has been to advance the tokamak through the development stages at the most rapid possible pace. In the 1970's a sufficient proof-of-principle basis was developed to motivate the construction of the proof-of-performance level tokamaks. These tokamaks have achieved performance levels that give the required confidence for the tokamak program to move on to the fusion development phase. It has been the U.S. and international view for the past decade that burning plasma physics is the next frontier of fusion plasma physics, and we should pursue this science as soon as practical in a tokamak device. The cost of the devices for the fusion development phase has motivated an examination of whether the proof-of-principle basis arrived at in the 80's could be improved upon. This new thrust, generally called the Advanced Tokamak (AT) program, represents renewed research at the proof-of-principle level aimed at finding the upper bounds to the potential of the tokamak as a magnetic confinement system.

The essential strategic question for the fusion program at this time in regard to the tokamak is whether to proceed to build tokamak devices in which we have confidence in a basic level of performance in order to get on now to the issues of burning plasmas and steady-state that are not readily addressable with proof-of-principle level facilities or to accelerate the development of more advanced operating modes to incorporate in the fusion development steps to come. A further aspect of this strategic question is whether to accelerate the development of concepts alternate to the tokamak that might be advanced into the development phase.

Three Pathways for Pursuing Burning Plasma Physics and Fusion Energy Development.

There is a strong consensus in the international fusion scientific community that the tokamak is technically ready for the steps to burning plasma physics and steady-state operation. There are, however, a range of opinions (hence different pathways) about the most cost-effective and technically

sound approach at this time. This has led us to define three potential pathways:

1. Integrated: a single device like ITER-RC: a reduced-cost ITER-like tokamak;

2. Modular: two separate tokamak devices to demonstrate DT burning and long-pulse/steady-state operation; and

3. Enhanced Concept Innovation.

The cost and technical challenge of the ITER-RC step is considerable. The combined cost of a smaller copper-coil burning-plasma and steady-state superconducting device is probably comparable to, perhaps less than, the cost of an ITER-RC. However, when the cost of an ITER-like ‘integrating facility (which would have to follow these two devices) is included, the total cost of the modular pathway to a DEMO is probably larger than the ITER-RC pathway. The U.S., at present budget levels, cannot proceed alone down either of these two paths , so the decision as to which pathway will be pursued is necessarily an international one. Indeed, all three strategies will be heavily dependent on, and would benefit greatly from, international collaboration.

The three pathways differ in the number of remaining sequential development steps and the total time to a DEMO step, as depicted in Figure 1. Pathway 1 addresses the major physics and technology issues of fusion energy development in an integrated manner and provides the most timely pathway for the development of fusion energy to the demonstration stage. Pathway 2 addresses the major physics issues separately in less expensive devices and then addresses physics and technology integration in a subsequent advanced integration facility to arrive at the demonstration stage. Pathway 3 delays addressing the burning plasma physics and integration issues of fusion energy in the manner of either pathways 1 or 2 until other confinement concepts have been developed through the Proof-of-Principle and/or Proof-of-Performance steps. Table 1 provides a summary of the principle advantages of each pathway. The choice which is ultimately made will depend on national and international factors, as well as the technical issues outlined in this report. We have confidence, however, that whichever path is pursued, fusion can and will play an important role in the world’s energy future.

The main part of this report are the following sections which describe three pathways in detail:

1. Integrated: a single device like ITER-RC: focused on a reduced-cost version of ITER;

2. Modular: focused on separate-mission tokamak technology; and

3. Enhanced Concept Innovation: focused on concept innovation leading to the study of burning plasmas at a later time.

Each section describes the pathway and its rationale, implications and advantages/disadvantages. Some topics, such as the potential contribution of the Strategic Simulation Initiative and fusion Technology and materials issues, apply to all pathways but are placed for now in the chapter on the Modular Pathway. This report does not attempt to make value judgments of choice among those pathways.

Table 1: Candidate Pathways - Principal Advantages

Pathway	Advantages
(1) ITER-RC	Early study of integration of burning plasmas, long-pulse/ steady-state operation and fusion technology. Minimizes number of steps (and time) to tokamak-based, demonstration power plant. No additional integrating facility needed. Consistent with strategic plans of ITER Partners. Makes maximum use of the leveraged U.S. investment and results of the ITER-EDA.
(2) Modular	Early study of burning plasmas and long-pulse, steady-state operation. Reduces initial facility investment costs and provides optimization for separable missions. Provides further optimization before integration step, allowing perhaps a more advanced integration step to follow. Provides multiple options for location of major facilities.
(3) Enhanced Concept Innovation	Provides for enhanced concept improvement leading to possibly the development of less expensive, more attractive fusion concepts. Reduces near-term facility investment costs. Provides further opportunities and time to optimize concept(s) for burning plasma, integration, and demonstration. Stimulates breadth of plasma science development (This is an enduring base program value common to all pathways).

2.0 Reduced Cost ITER Integrated Step

2.1 Pathway for a Reduced Cost ITER

The "Reduced-Cost ITER" pathway exploits the existing situation in which, based on strong progress in world-wide tokamak research, the world program has decided that the tokamak concept is technically ready to proceed to a physics and technology integration step in which the next major physics issues of burning plasmas and steady-state will be explored. Implementation of the strategy requires that funding can be secured to exploit this opportunity to follow what appears to be today the most direct, economical and timely pathway for fusion energy development. The reduced-cost ITER strategy will accomplish most, perhaps all, of the ITER mission in a less costly experimental facility. The four ITER Parties have agreed to develop a reduced cost design during the EDA extension period, with the objective of reaching a construction agreement by the end of the three-year extension in July, 2001. If the reduced cost ITER is constructed, the next -step program would combine in a single major facility

• the creation and experimental investigation of self-heated burning plasmas,

• the demonstration of long-pulse advanced tokamak operating modes in burning plasmas,

• the integrated exploration of related tokamak plasma physics issues,

• the integration of fusion reactor-relevant technologies, and

• the integrated testing of fusion reactor components in a single major facility.

Figure 2.1. Pathway Roadmap. The tokamak program is at a decision point regarding readiness to proceed to two possible next steps: (a) an integrated device that achieves both the physics integration of burning plasmas and long-pulse operation and the physics/technology integration of burning plasmas and reactor-relevant technology, or (b) a program of separate burning plasma and long-pulse devices whose physics results would be combined in a subsequent integration device, which would also integrate the technology Other elements of the overall program strategy would include the innovative concepts program, the technology program, and the inertial fusion energy program, which proceed in parallel with the tokamak program.

The RC-ITER experimental program would be phased, with the early phases emphasizing physics explorations, the intermediate phases emphasizing the achievement of reliable quasi-steady-state operation with moderate neutron flux, and later phases emphasizing the accumulation of neutron fluence and operating time for fusion nuclear and materials science studies and for integrated component and system testing. The device is being designed with sufficient flexibility to operate under a variety of experimental scenarios and to allow modifications during operation that would take advantage of results from earlier phases. The nuclear testing capability of ITER, implemented mainly through the introduction of blanket test modules, would reduce the need for a Volume Neutron Source.

The reduced-cost ITER pathway has a high probability of sustaining the four-party ITER collaboration on a next-step tokamak. It is consistent with the collaboratively evolved strategy of the ITER project and with the fusion development strategies of our ITER partners. While the parties differ on some of the detailed specifications for the DEMO step following ITER, they envision ITER and a "point neutron source" facility (for lifetime materials testing) as providing the technical basis for the DEMO. Figure 2.1 illustrates the combination of near-term research programs feeding into the ITER step, which would lead to a tokamak DEMO when combined with the point neutron source; the base program of advanced tokamaks, other innovative concepts and technology programs would feed into the ITER and DEMO programs throughout the periods of design and operations. (Some of our ITER partners also see ITER as providing the basis for a non-tokamak DEMO, especially one using a closely-related confinement approach such as the stellarator.) The Inertial Fusion Energy Program is envisioned as progressing in parallel with the magnetic fusion energy program. Under currently limited US fusion budgets, the need to preserve a strong base program will constrain the US’s ability to contribute to ITER; however, the return on investment for the US’s contribution will be enormous due to the larger investments of the international partners.

The reduced-cost ITER will be designed with the flexibility required to accomplish a two-fold mission:

• a reduced physics and testing mission (e.g. Q 3 10, moderate pulse length of 300-1000 seconds, Gn = 0.5 MW/m2) when operating in the basic physics performance mode, and

• the full ITER mission (e.g., ignition, steady-state, Gn= 1.0 MW/m2) when operating in the enhanced physics performance mode to demonstrate the upside potential of tokamak power systems, but with increased technical risks.

The mission reductions inherent in the de-scoping to the reduced-cost ITER eliminate the commitment to ignition (reducing mandatory performance to Q=10), and reduce the nuclear testing capability due to reduced inductive pulse length and reduced neutron flux when operating in the basic performance mode. With the modified Q310 performance objective, there would be less physics margin at the pessimistic end of the range of physics performance, but ignition might still be possible at the optimistic end of the range. Investigation of thermal control of an ignited discharge would be deferred in the base physics performance mode, and staging would allow delaying specific upgrade capabilities until experiments on ITER itself establish the requirements. The technologies and engineering design solutions developed during the ITER EDA will be adapted for the reduced-cost ITER design, including such improvements and innovations as are consistent with the ongoing ITER R&D program.

The basic physics performance mode will use projections of the established ELMy H-mode database that are the basis for the ITER EDA design. The reduced-cost ITER will be designed to achieve the reduced mission based on ITER EDA physics rules and consequent cost/benefit design optimizations. However, the design will also include features that permit exploration of advanced tokamak (AT) modes, including shaping (k9531.6 and d9530.3, possibly within the constraint of a single null configuration), n=0 internal coils if needed for vertical position control, current profile control, flexible heating/current drive systems for pressure-profile/transport-profile control, and real-time profile diagnostics. The enhanced physics performance mode will be based on the database for advanced tokamak operation that will be established within the next few years.

It is envisioned that, during the design and construction phases, the US would be involved in design, diagnostics, and manufacture of high-technology tokamak components and systems, targeted at achievement of US science and technology goals while emphasizing dual-use activities which both benefit ITER and achieve US program goals outside ITER. In the operations phase, the US would be involved in the scientific and technological aspects of ITER’s experimental program, addressing the "third leg" of the US fusion program --- the pursuit of burning plasmas and technology through international collaboration.

2.2 Rationale for Reduced Cost ITER Pathway

The primary rationale for the reduced-cost ITER pathway is to exploit the present status of the world-wide tokamak program:

• burning plasma physics is the next frontier of fusion energy science;

• all Parties and pathways eventually require such an integrated physics/technology step to develop fusion energy;

• the tokamak concept is technically ready to proceed with this step; and

• the cost to the Parties will be reduced if we proceed on a cost-shared international basis.

The reduced-cost ITER would build on the existing international collaboration in design and R&D. This international collaboration allows the US to highly leverage its contribution to accomplish the next stage of magnetic fusion science exploration and magnetic fusion energy development in a highly cost-effective manner. The reduced-cost ITER is consistent with the fusion development plans of the international partners.

The tokamak concept is highly developed, and ongoing tokamak research is very promising with respect to future performance enhancements. The tokamak has already demonstrated reliable and sustained operation in a physics mode that extrapolates to achievement of a burning plasma in the reduced-cost ITER. The world tokamak program has achieved further enhancements in plasma performance for short periods during a pulse, and the tokamak program is presently focused on the sustainment of this enhanced performance for extended periods and on the achievement of high levels of performance in several parameters simultaneously. The reduced-cost ITER pathway will address essentially all the major next-step issues in tokamak physics and fusion technology and their integration in a single device. The reduced-cost ITER pathway will maximize use of the fusion-reactor-relevant technologies and design solutions that have been developed in the ITER EDA and will establish the knowledge base for the demonstration of fusion power in the most timely and cost effective manner.

While there are uncertainties about the levels of enhanced performance that could be achieved and sustained in the reduced-cost ITER, the design will incorporate those features that are expected to be needed to obtain optimized reactor-scale plasma performance. The planned flexibility in the design is intended to respond to the uncertainties and to provide a range of physics operating modes that will permit both broad scientific research at the forefront of tokamak plasma science and improved likelihood of mission success. The world’s tokamak research program is now focused on the resolution of high-priority physics R&D issues that relate to the design of ITER; in particular, an ITER Topical Group on Advanced Tokamaks has been established to resolve the shape issue and current-profile/pressure-profile control for the reduced-cost ITER, and there is increased attention of all Physics Expert Groups to AT needs.

2.3 Technical Contributions of a Reduced Cost ITER

As stated in the formal ITER Agreement, "The overall programmatic objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes". A Special Working Group (SWG) has been charged with proposing "technical guidelines for possible changes to the current detailed technical objectives and overall technical margins, with a view to establishing options(s) of minimum cost still satisfying the overall programmatic objective of the ITER Agreement". The formal report of the SWG addressing this charge has been approved by the ITER Council. The physics benefits of the reduced-cost ITER assumes that the device would:

• achieve extended burn in inductively driven plasmas with Q310 for a range of operating scenarios and with duration sufficient to achieve stationary conditions on the time-scales characteristic of plasma processes, and

• aim at demonstrating steady-state operation using non-inductive current drive with the ratio of fusion power to input power for current drive of at least 5.

While these levels of plasma performance are less than those targeted for the present ITER, as described in the Final Design Report (FDR), the SWG expresses the view of the four ITER Parties that this reduced level still satisfies the overall programmatic objectives of the ITER EDA Agreement. Operation in enhanced performance modes in the reduced-cost ITER might permit the achievement of ignition.

A device meeting these technical and plasma performance objectives would permit studies not only of reactor-scale burning plasma physics and long-pulse physics separately, but also their integration. While aspects of the relevant individual physics phenomena could probably be studied separately on somewhat smaller devices, the integrated combination of long pulse and reactor-scale burning plasmas together with the relevant technology is key to the mission of a device such as ITER.

2.3.1. Burning plasma physics, steady state physics, and advanced tokamak physics

Experiments in the reduced-cost ITER will explore the physics issues of "burning plasmas", in which the heating is dominated by alpha-particles created by the fusion reactions themselves, as distinct from an "ignited" ITER plasma, in which the heating is only by alpha particles. At Q=10, the power from the alpha particles would be two-thirds of the total heating power. Burning plasma physics issues will include new plasma-physical effects on the Alfvén eigenmodes made unstable by the presence within the plasma of a population of super-Alfvénic alpha particles. Although the relative population of super-Alfvénic particles is expected to be smaller than in present experiments, theoretical studies of energetic-particle modes in plasmas such as ITER’s predict that new phenomena will arise: for example, ITER-scale burning plasmas would be able to address nonlinear collective effects in which many toroidal Alfvén eigenmodes are unstable and drive fundamentally different loss mechanisms, such as stochastic diffusion due to overlapping resonances, whereas present-day (Q<1) DT plasmas can study only coherent single modes of instability and particle trapping in a resonant drift island. While detailed studies of these effects in a Q=10 (rather than ignited) ITER plasma remain to be carried out, it is clear that there will be very little difference between ignition and Q=10 in regard to the alpha particle population, so that a reduced-cost ITER plasma will be fully representative of a reactor plasma in this regard.

The reduced-cost ITER will also permit studies of the physics of very-long-pulse/steady-state plasmas, in which much of the plasma current is self-generated and which will require effective control of plasma purity and plasma-wall interactions. Achievement of a large fraction of self-generated current in a high-performance plasma will require sufficient plasma shaping, plasma stability at high normalized beta-values (or active stabilization of wall modes), and control of the profiles of quantities such as plasma current, pressure, density, and flows. The reduced-cost ITER design is aimed at incorporating increased configurational flexibility, as well as features specifically chosen so as to permit the achievement and study of steady-state operational modes. In addition, the reduced-cost ITER will address issues of steady-state control of plasma purity and plasma-wall interactions, including the physics of a "radiative divertor" designed for handling high power flow for long pulses, and would allow studies of novel plasma and atomic-physics effects, as well as advancing the materials science of surfaces subject to intense plasma interaction. Increasing triangularity of the core plasma affects the divertor magnetic configuration in a way that reduces the flexibility to accommodate uncertainties in divertor physics. The replaceable divertor cassette developed for the present ITER design provides an opportunity for a trade-off of core plasma shaping versus divertor flexibility. In addition, the reduced-cost ITER gives greater emphasis to AT modes; the SWG statement on Qcurrent_drive35 (the same as in the original ITER guidelines) documents the commitment to exploration of AT and steady-state physics and technology.

Since even the reduced-cost ITER has three times more gyro-radii in its plasma minor radius than does the largest present-day tokamak, it will allow studies of size-scaling of transport which should resolve issues of the dependencies on relevant dimensionless parameters. The different scalings of the edge and the core would be studied. For the first time, the core-plasma and the edge-plasma would be simultaneously in a reactor-like regime; in particular, the size-scalings of confinement in the core and in the edge may give rise to fundamentally different density limits than in present-day tokamaks. Transport barrier studies would utilize a variety of auxiliary heating and current-drive techniques to create, control, and sustain localized regions of significantly reduced transport.

In either the present or reduced-cost ITER, studies of size-scaling of stability would include both the scaling of MHD modes with number of gyro-radii at relevant collisionality and effects of significantly larger energies and plasma currents during plasma transients, including so-called "disruptions". Size-scaling of both ideal and non-ideal beta-limits would be addressed. Feedback stabilization of neoclassical tearing modes and profile control would be studied and utilized to overcome long-pulse beta limits.

At reactor-scale levels of plasma current (i.e., in either the present or reduced-cost ITER), disruptions may induce the new phenomenon of avalanching of runaway electrons by hard collisions, as well as competing mechanisms for the loss of energetic electrons by fluctuations and non-axisymmetries. Physics studies of mitigating power flows during disruptions would address the feasibility of reducing the plasma-wall loads, complementing the technology program’s work on handling these loads.

2.3.2 Physics Integration

Most importantly, the reduced-cost ITER would allow the integration of burning plasmas with long-pulse/steady-state operation. This integration will involve the following complex interplay of transport, stability, and an internal self-generated heat source:

• The evolution of plasma profiles would no longer be dominated by external heating; internal self-generated feedback loops would be prominent, with the self-heating affecting temperature and density profiles, that in turn modify transport and stability and hence affect the self-heating. The important time-scales, in ascending order, are the energy confinement time, the time for significant accumulation of thermalized alpha particles (helium ash), the "skin time" that characterizes the slow evolution of the current profile within the plasma, and the time-scale for plasma-wall interactions and for wall temperatures to reach equilibrium. Since transport and stability are dependent on the current profile, ITER’s pulse length (which is longer than the relevant skin time in both the present and reduced-cost designs) is an important measure of its ability to study the approach to steady-state profiles. • With strong self-heating, profile control will most effectively take the form of control of local transport, by creating transport barriers that serve as tools for profile control. Such studies would be key to developing an attractive tokamak reactor concept, where the alpha power could be utilized to create pressure profiles that drive consistent self-generated current profiles. As such, ITER would be key to the optimization of the tokamak concept, adapting the advanced tokamak features and techniques developed throughout the world program on non-burning plasmas and applying them to the control of an optimized self-heated burning-plasma configuration.

For these key studies, an ignited plasma is not essential: it is sufficient for the self-heating to be dominant -- as it would be in the Q=10 reduced-cost ITER. The emphasis on research into integration of burning-plasma physics and long-pulse physics in advanced tokamak modes provides an opportunity for developing the basis for an even more attractive tokamak reactor concept.

2.3.3 Technology Integration

The integration of technology has been a clearly stated programmatic objective of the ITER agreement, "... by demonstrating technologies essential to a reactor in an integrated system, and by performing integrated testing of the high-heat-flux and nuclear components required to utilize fusion energy for practical purposes." There is universal agreement that an integrated device will be required prior to a demonstration reactor -- the current argument is over whether to proceed with the integrated device on the present database, or to proceed first with a modular approach, returning to an integration step only after further concept improvement.

The technology integration objectives for a reduced-cost ITER should be adequately achievable even at baseline-physics neutron flux half that of the current ITER. The physics/technology integration objectives would be even better satisfied in the reduced-cost option, provided advanced tokamak operation can be achieved and sustained in steady state under conditions of strong self heating.

The compatibility of the "nuclear" components (divertor, limiters, blanket, tritium extraction, shield, etc.) with the tokamak environment will be an important driver for design and R&D. Making design choices that are consistent with nuclear technology and remote-maintenance requirements is an imperative for ITER, but would be very unlikely to happen in a cost-conscious design of a non-nuclear or very-low-fluence device.

In particular, studies aimed at mitigating the power flows during disruptions would address the influence of melting and erosion of significant amounts of wall material and the influx of this material into the plasma. Although the energy deposited in a disruption would be substantially smaller in the reduced-cost ITER than in the full ITER, the surface area on which it is deposited is also smaller, so that erosion and melting are likely to be similar. Co-deposition of tritium with disruption-induced or steady erosion of wall material can adversely affect safety-related tritium in-vessel inventory ceilings. Tritium recovery technology can be adequately demonstrated on a reduced-cost ITER.

The compatibility of the "standard tokamak" components (superconducting magnets, vacuum chamber and pumping ducts, heating and current drive, and diagnostics) with a nuclear environment will also be a design driver absent from non-nuclear devices; an example is the need to avoid paths which irradiate components intolerant to too many unimpeded neutron flights. The design choices will in turn be confirmed by operation in the nuclear environment.

Reality provides an over-arching discipline to design solutions. The emphasis on failure modes, reliability and maintainability of components necessary in a major integrated nuclear facility will drive design and R&D to a much greater extent than in non-nuclear devices. The high availability goals of the later engineering-oriented phase of an integrated device will also be a strong driver.

The nuclear testing role of ITER is fulfilled mainly through the installation of blanket test-modules, introduced through ports specifically allocated for this purpose. The reduced-cost ITER could have the same nuclear testing capability as the original ITER, if modestly enhanced performance can be achieved with advanced physics, but would have a reduced neutron flux and fluence capability under ELMy H-mode operation. Even the reduced capability (e.g., fluence of about 0.3 MW-yr/m2), when combined with a point neutron source, could provide a sufficient basis for a DEMO design.

2.4 Fusion Development Pathway Implications of a Reduced Cost ITER

The primary advantages of the reduced-cost ITER pathway to fusion development are threefold:

• it minimizes the time and number of steps needed before a magnetic fusion demonstration reactor can be built,

• it shares the costs and risks internationally, providing a large return on investment to any one Party, and

• it provides for the involvement of the world’s fusion experts and the Parties’ industries in a coordinated world-wide program to achieve ITER’s objectives.

The ultimate objective of a government-funded fusion energy development program is to bring one or more fusion reactor concepts to the stage at which the concept is sufficiently demonstrated to provide an assurance that electricity can be generated at an affordable cost with acceptable environmental consequences. This is generally agreed to require demonstration of: 1. reliable, controlled operation of a D-T fusion plasma under reactor-relevant conditions;

2. reliable operation, to some significant fraction of their anticipated lifetimes, of reactor-extrapolatable technologies, components and systems under fusion reactor conditions;

3. reliable operation of an integrated fusion reactor at availabilities (> 50%) that are extrapolatable to commercial requirements;

4. tritium fuel self-sufficiency;

5. net electrical power production at significant levels (> 100s of MW);

6. the safety of fusion reactors;

7. the feasibility of economically competitive fusion reactors; and

8. the feasibility of environmentally benign fusion reactors.

The collection of sequential and parallel research and development steps that lead to these demonstrations is the fusion development pathway for a given fusion concept. The culmination of the development pathway for a given fusion reactor concept is generally agreed to be a demonstration plant (DEMO) in which the simultaneous demonstration of most, if not all, of the above requirements is achieved.

The ITER mission was defined with the objective that ITER, its supporting R&D programs, and the nuclear and materials testing programs that would be carried out, both as part of ITER and in supporting national R&D programs, would provide the design basis for a DEMO with respect to requirements #1-6. It was envisioned that advanced physics research would be carried out, both as part of the ITER experimental program and in parallel with ITER on other devices, to develop a tokamak physics concept for the DEMO in support of requirement #7. The development of advanced blanket, structural and other materials in parallel with ITER is necessary to provide the design data base for a DEMO that can satisfy requirement #8, consistent with all the other requirements. These major elements, together with the base tokamak plasma physics and fusion engineering science research programs, constitute the "ITER" fusion development pathway. To the extent that the reduced-cost ITER is successful in accessing advanced tokamak operation to accomplish the full ITER mission, the "reduced-cost ITER" development pathway is identical with the "ITER" development pathway. The "reduced-cost ITER" development pathway is probably the shortest and least expensive development pathway for tokamaks.

2.5 Pros & Cons of a Reduced Cost ITER Strategy

The advantages of the reduced-cost ITER (ITER-RC) strategy include the following:

• The ITER-RC strategy exploits the advanced status of tokamak development worldwide, including the readiness to move to burning plasmas and system integration. The ITER-RC step will be a dramatic advance in the development of fusion and will move fusion research much closer to the production of practical fusion power.

• The ITER-RC strategy will permit early study of the integration of burning plasmas, long-pulse/steady-state plasmas, and fusion technology. It addresses the major next-step tokamak physics and technology issues and their integration in a single device.

• The ITER-RC strategy, because of its early integration of physics and technology, minimizes the time and number of steps needed before a demonstration magnetic fusion reactor can be built, and it makes possible the earliest possible implementation of magnetic fusion energy.

• The ITER-RC strategy has a sound physics basis, well-supported by ongoing world tokamak programs. With basic-level plasma performance, the fusion alpha power will be dominant; with enhanced performance, ignition may be achieved. The reduced-cost ITER should have about the same probability of accomplishing its reduced physics (Q = 10) and nuclear (Gn ª 0.5 MW/m2) missions under ELMy H-mode operation as the ITER EDA design was judged by the international fusion community to have of achieving the full ITER mission. Moreover, the reduced-cost ITER would have a possibility of accomplishing the full ITER mission under modestly advanced physics assumptions.

• The ITER-RC strategy brings the world’s technical and financial resources to the task, with internationally shared benefits and risks. The ITER project (including INTOR before it) represents almost 20 years of collaboration on the definition, design and supporting R&D for a next-step tokamak experiment by the international partners. This collaboration and its products are generally held in high regard by the involved governments; for example, in the G8 Communique following the recent Birmingham summit (May 15-17, 1998), world leaders stated, "We acknowledge successful cooperation on the pilot project of the International Thermonuclear Experimental Reactor (ITER) and consider it desirable to continue international cooperation for civil nuclear fusion development". The ITER-RC strategy is consistent with the collaboratively evolved strategy of the ITER project and with the fusion development strategies of the other ITER partners. The ITER-RC strategy is strongly favored by all of the non-US ITER parties.

• The ITER-RC strategy would maximize use of the technologies and engineering design solutions that have been developed already for ITER, thus utilizing fully the results of a six-year technology R&D program and a nine-year engineering design effort.

• With US participation, the ITER-RC strategy enables US industry to maintain parity in fusion technology with competitors in other Parties.

• The ITER-RC program provides valuable physics and technology information for other magnetic fusion concepts.

• The ITER-RC strategy demonstrates full-scale integrated reactor technology, much of which is generic to all toroidal fusion concepts.

• The ITER-RC strategy reduces the need for a Volume Neutron Source.

• The ITER-RC strategy fits well within the tritium availability window and benefits from civilian supplies of tritium.

• Successful ITER-RC operation will give the Parties the option to proceed with a fusion demonstration plant based on their respective energy needs and utility industry circumstances.

The disadvantages of the reduced-cost ITER strategy include the following: • The ITER-RC strategy must confront all of the major next-step physics and technology issues and their integration in a single device.

• The consequences of technical failure of a single device that addresses all next-step issues together would certainly be greater than the consequences of technical failure of a single less-expensive device that addresses a single issue.

• If ITER-RC’s physics performance does not meet expectations, then the credibility of magnetic fusion would be damaged.

• Ultimately, a better concept may emerge and, because of overall resource limitations, ITER-RC might have delayed implementation of this better concept.

• An international agreement on siting and cost-sharing is required.

2.6 Near Term Actions

Assuming that the ITER-RC strategy is adopted, world efforts over the next few years should be focused on the following tasks:

• the completion of the ITER EDA technology R&D projects;

• development of an attractive reduced-cost ITER design; and

• the development of the physics and engineering databases to support an advanced design.

As part of this effort, the US should:

• fully participate in the completion of its assigned role in the ITER EDA technology R&D projects, which also have intrinsic value for the US domestic program;

• participate vigorously in the design of the reduced-cost ITER, which accommodates and exploits AT features, since the US has strongly advocated incorporating advanced features in a reduced-cost ITER;

• focus significant capabilities of the US base tokamak experimental program and of the supporting theory and computational programs on the development of the physics basis needed to support an attractive reduced-cost ITER design; and

• direct appropriate parts of the technology program at cost reduction for key ITER components.

2.7 Summary of Technical Appendices on Reduced Cost ITER

2.7.1 Advanced Physics Assumptions

The world tokamak program is making steady progress in understanding Advanced Tokamak (AT) operating modes that have the promise of significantly enhanced performance and potentially lead to attractive fusion power plants (e.g., the ARIES-RS study, and the Galambos et al. 1995 Nuclear Fusion paper). However, the highest performance results to date are transitory and are not yet achieved with all the relevant dimensionless parameters simultaneously. Demonstration and understanding of long pulse AT discharges is the challenge being pursued by the international tokamak community; tokamak facilities are implementing rf current-profile control, plasma density control (e.g., divertor and fueling), and exploring ideas for internal transport barrier control (e.g., rf flow drive). Given the promise of these emerging AT modes, we believe that they should be a central ITER research objective and design driver. A key question is the level of advanced performance that reasonably can be expected to be established in the next 2-4 years. Our judgment is that confinement 50% better (i.e., HH = 1.5) than the present ITER ELMy H-mode database and stability 50% better (i.e., bN = 3.5) than the ITER EDA design basis are plausible. Such performance has already been sustained for several energy confinement times (duration ~ 5tE ~ 1 sec), but, for lack of current drive capability and other reasons, such performance has not been sustained for several current relaxation time scales. It is reasonable to design a reduced size ITER to achieve a reduced mission under present ITER physics projections (HH ª 1, bN £ 2.5) but with the capabilities (e.g., high plasma shaping, and current profile control) to utilize AT performance (HH ~ 1.5, bN ~ 3.5) modes to achieve ignition and the full nuclear mission.

2.7.2 Systems and Transport Studies

Any further substantial reduction in the ITER capital cost will be achieved only by reducing some of the mission requirements and/or adopting less conservative physics/engineering guidelines. Systems code studies have examined the options for reduced cost ITER designs. Using ITER EDA technology and physics/engineering guidelines, it should be possible to design a (Q = 10, R = 6.0-6.5 m, I = 12-14 MA, k £ 1.7, Gn ª 0.5-0.9 MW/m2) device which would be able to explore high-Q, steady-state and AT physics operation, which would have a significant, albeit somewhat reduced, nuclear testing capability and which would have a cost about 60-70% that of the ITER EDA design. Projected performance of such designs under modest AT physics assumptions, such as should be supported by the experimental database within 2-4 years, include ignition and the full nuclear mission capability. Using modest AT physics and more innovative engineering design guidelines results in even smaller size designs, with size and cost saturating at R ª 5 m and 50% of the ITER EDA design cost. The smaller devices have larger divertor heat loads than the ITER EDA design under AT operating conditions.

A series of 1-D transport simulations have been performed to assess the performance of a representative reduced-cost ITER design point. The results indicate that an R= 6 m (Q = 10) design based on the present ITER ELMy H-mode physics design guidelines is possible. Ignition and full nuclear mission capability are predicted for modest advanced physics assumptions (H97 3 1.3, bN 3 2.5).

2.7.3 Illustrative AT ITER Design Point

An ITER-like conceptual design has been developed at R = 5.6m with an estimated machine cost that is 45% that of the ITER EDA, when full advantage is taken of various mission and engineering implementation cost reductions. It has been demonstrated that the combination of reducing the fusion power to high-Q operation (10<Q<20) and reducing the shield thickness so that neutron-gamma heating is absorbed inertially in the first layer of the magnet can reduce the size of a next-step "ITER-like" machine to less than half the volume of the current ITER. In this inertial regime, TF insulation radiation allowables would be reached after 60,000 pulses of 300 second duration and overall magnet system refrigerator requirements can be decreased by a factor of four (from 80 kW to 20 kW). A full steady-state regime can be achieved at reduced power, or at full-power as a refrigerator upgrade option. The single most important cost reduction is the reduction in physics performance and plasma power, which reduces the cost of a new ITER to 70% of its baseline value. The most important engineering idea for cost reduction and the second most important overall is the reduction of the shield thickness by 34 cm and the radial build by 50 cm. The overall cost savings of adiabatic operation is 19.5%. Cost improvements resulting from the use of more recent conductors, the use of quench detectors as internal dump resistors, and more realistic scaling algorithms for previously fixed costs result in a total potential savings of 55 % relative to the ITER EDA cost.

2.7.4 Device Capability Requirements for Accessing AT Modes

Advanced tokamak operation is a subject of current research. As such, it is difficult to make definitive statements regarding design requirements for advanced tokamak operation in ITER. However, we can point out design features that are likely to be important for advanced tokamak operation. We list these design features in roughly the order in which they must be addressed during the design of a reduced cost ITER.

(1) Strong Plasma Shaping. That is high elongation (k95 3 1.6), high triangularity (d95 3 0.3), and/or lower aspect ratio (A 2 3). Plasma shaping is a strong driver to the design of a reduced cost ITER since the achievable plasma shaping will be determined by basic device parameters [shape of the plasma chamber, location of the divertor(s), location and capability of PF coils] which cannot be modified easily either after construction or even during the detailed design process.

(2) Internal Control Coils are desirable in that they allow higher elongation Active control of (n0) resistive wall modes and/or tearing modes might also be achieved with internal coils. Such a system would have to be included in the initial machine design.

(3) A real-time diagnostic capability for measuring temperature, density, current, and rotation profiles is required for AT operation. Hence, diagnosticians should be involved early in the design process to insure adequate diagnostic access.

(4) Central Heating and Current Profile Control. Auxiliary heating and current drive systems mainly impact the design of the ports. Neutral beams for current (or rotation) drive require tangential ports, while RF systems require horizontal ports. We commend the good example set in the ITER FDR design, which included many ports for the RF heating and current drive, each with a common interface suitable for any of the candidate systems.

(5) Pressure Profile Control is the key issue for advanced modes. Schemes for active control of the pressure profile that we are aware of involve controlling transport through control of the velocity profile.

(6) Rotation Control. Advanced operating modes may require overall plasma rotation for stabilization of the resistive wall mode and/or neoclassical tearing modes and the introduction of velocity shear to produce (and control) transport barriers. While some progress has been made (particularly with IBW rotation drive), there is still much to learn about rotation drive in tokamaks. A vigorous physics R&D program will be required. A common port interface will allow the system(s) for driving sheared rotation to be added after the physics R&D effort has defined the requirements for rotation drive.

(7) Central Fueling would allow control the density profile, and thereby the pressure profile. Unfortunately, we do not yet have any proven means of getting fuel to the center of a reactor-like plasma. R&D is required to support inside pellet launch and alternative schemes for central fueling (like compact toroids). The only impact on the device design (as opposed to the supporting R&D program) is a possible increase in demand for port space.

(8) Advanced Divertor Techniques to allow highly dissipative divertor and/or core plasma operation in regard to confinement quality, tolerable impurity levels, and density limits.. High performance core plasmas are likely to call for increased plasma triangularity and perhaps some form of double null operation, features that demand reexamining the divertor solutions that need to be employed.

The full text of thse appendices can be found at the Web Site address: http://nso.ucsd.edu

3.0 Modular Program Pathway

3.1 Pathway Overview

The major issues in fusion R&D can be described as: (1) the achievement and understanding of self -heated plasmas with high energy gain that have characteristics similar to those expected in a fusion energy source, (2) the achievement and understanding of sustained self-heated plasmas with characteristics (steady-state or high duty factor pulsed systems) similar to those expected in a competitive fusion system and (3) the development of the nuclear technologies needed for fusion energy sources. These general categories can be used to describe both the magnetic and inertial fusion R&D programs which have historically pursued a modular approach with the individual modules focused on the technical issues described above. The 1995 PCAST review of Magnetic Fusion recommended that the modular strategy be continued with programs and facilities specialized to address the ignition, steady-state and technology issues. This modular pathway, with burning plasma physics as the highest priority element, was the central recommendation of the Grunder FESAC Panel (January 1998) and was the option that was preferred by many of the fusion community researchers at a workshop on approaches to burning plasma physics held in Madison, Wisconsin (April 1998). The continuation of the modular approach for the next major steps in magnetic fusion enhances the likelihood of successfully realizing a viable fusion power source.

The proposed Modular Program Pathway to Magnetic Fusion (Fig. 2.1) would have four major initiatives aimed at: (1) developing innovations in steady-state advanced magnetic confinement configurations, (2) exploration, optimization and understanding of strongly burning plasmas, (3) development of technologies and materials needed to make magnetic fusion an economically and environmentally attractive energy source, and (4) a Strategic Simulation Initiative to facilitate the fundamental science understanding in each of the first three initiatives and to then serve as a mechanism to intellectually integrate the science of these initiatives.

Fig. 3.1 Modular Strategy for the Magnetic Fusion Program (the grey bars correspond to operation) The Steady-State Advanced Confinement Initiative would be addressed by extensions of ongoing research with advanced tokamaks (DIII-D, C-Mod, JET and JT-60 U), very long pulse superconducting tokamaks (Triam and Tore Supra), new superconducting tokamaks under construction (KSTAR, SST-1), long pulse stellarators (W-7AS), two new $1B class superconducting stellarators (LHD, W-7X) and new facilities in the spherical torus and advanced stellarator configurations. A possible major new facility in this Initiative is the JT-60 SU, which if constructed would be capable of addressing fully the physics of steady-state advanced tokamak physics at the reactor plasma scale of ARIES-RS. Many of these advanced toroidal configurations are able to take advantage of fundamental toroidal plasma physics that was first developed and understood using the pulsed tokamak as a research tool to access and understand fusion plasma conditions. It is probable that the Steady-State portion of the modular pathway which carries forward the superconducting tokamak development path will be implemented by the machines listed above that will be built outside the United States.

There are currently no facilities in the world magnetic fusion program capable of the study of high-energy-gain, burning plasma issues. TFTR and JET carried out successful initial experiments with weakly burning D-T plasmas that were limited in plasma duration in 1993-97. JET is scheduled to carry out another series of weakly burning D-T experiments near the end of 2002. The TFTR and JET experiments have not only produced D-T fusion plasmas with Lawson parameters (nitETi) within a factor of 10 of that required for ignition but most importantly confirmed that D-T experiments could be carried out safely in the laboratory. The magnetic fusion program is technically ready today to begin construction of a $1B scale Ignitor-like compact ignition tokamak. The major thrust of the proposed Modular Pathway is to build a burning plasma facility at the earliest possible time as recommended by the Grunder FESAC Panel. The objective for the burning plasma initiative is to achieve, explore, understand and optimize strongly burning plasmas in a toroidal magnetic configuration. An analysis using the present tokamak data base indicates that a compact tokamak configuration would achieve the desired burning D-T plasma performance (Q 3 10) for pulse duration (>> energy confinement time and ~ plasma current redistribution time) needed to satisfy the burning plasma physics objectives. An important characteristic of the compact tokamak is that ignition can be achieved in a physical size much smaller than the final power plant such as ARIES-RS. Therefore, the incremental construction cost of this facility might be minimized to ~ 2$1B with construction taking ~7 - 8 years. The generic toroidal burning plasma physics information from this initiative would provide a foundation for understanding burning plasmas in the advanced tokamak, advanced stellarator and spherical torus configurations.

The Strategic Simulation Initiative (SSI) is a key element of the Modular Strategy. First, the SSI will be a powerful capability in developing the fundamental physics understanding of the Steady-State Advanced Confinement Initiative (Advanced Tokamaks, Advanced Stellarators and Spherical Tori) and in the Burning Plasma Initiative which uses the pulsed tokamak to cost-effectively access burning plasma conditions. The major advantage of the SSI will be to intellectually integrate the fundamental burning plasma physics understanding from the Burning Plasma Initiative and the fundamental physics understanding from the Steady-State Advanced Magnetic Confinement Initiative that will allow the development of an optimized step forward in magnetic fusion, the Advanced Fusion Integration Facility. The SSI is expected, in fact, to play a key role in all three pathways discussed in this report.

The Fusion Technology and Materials Initiative would focus on the critical task of developing and testing advanced materials that would lead to an attractive fusion power plant. An essential capability needed in this area is an intense neutron source capable of irradiating candidate materials to power plant scale fluences. A conceptual design for the Point Neutron Source (PtNS) has been developed through an IAEA collaboration and is estimated to cost ~$0.8B. A volume neutron source would test larger size (~10m2) sub-components to prior to reactor scale integration and is expected to have a construction cost in the range of $1-2B. Such facilities will be needed in the other two pathways described in this report.

Because the costs of the various facilities all exceed the amount that would be available in the US fusion budget under the present constraints, international collaboration would be required to implement this modular strategy.

The Three Major Fusion Initiatives would be carried forward to a Magnetic Fusion Assessment Check point in ~2015 which would review the status of magnetic fusion and decide whether to (1) proceed forward to an Advanced Fusion Integration Facility, (2) extend the Modular phase or (3) move to another innovative confinement concept.

3.2 Rationale for the Modular Program Strategy

3.2.1 Hardware Integration Strategy

The fusion R&D program has used the modular approach for the first decades of research and has understood that these program modules would be integrated near the final stages of fusion development. However, fusion is still in the research phase at this time. Significant progress has been made in producing reactor plasma conditions for short durations in the laboratory that gives encouragement that a solution is possible, but the knowledge base does not exist at the present time to build an attractive fusion power system.

The most efficient approach to pursue fusion R&D objectives at this time is to focus on critical issues in each sub-area, and to develop the knowledge in each sub-area to near that needed for integration at the energy production scale. The advantages of this approach are:

• allows the flexibility needed in a research program,

• reduces cost and time for individual steps, and

• allows innovation to be incorporated earlier.

Systems studies which have been used to evaluate linkages between sub-areas and the Strategic Simulation Initiative will be utilized to provide a "virtual" integration of the research modules. Physical hardware integration of the three main research modules should be done only when needed to address issues in those sub-areas, or when the status is near reactor levels and integration is the main objective.

Fusion has a particular challenge at this time to not only demonstrate the scientific and technological feasibility of magnetic fusion, but to also develop economically and environmentally attractive fusion power systems. Keys to this are advanced magnetic confinement systems with high fusion gain, high power density and high duty cycle preferably steady-state plasmas, the corresponding enabling technology and the necessary nuclear technologies with attractive environmental characteristics such as low activation and the ability to withstand the neutron fluence. The Modular Program Pathway has focused program elements or initiatives that are targeted on addressing these issues.

3.2.2 Toroidal magnetic confinement systems have generic physics and technology issues, and the pulsed tokamak is an effective tool for developing the generic physics and technology.

Now that fusion plasmas have been produced in the laboratory using the pulsed tokamak configuration, the emphasis is broadening to develop features for improving the characteristics of toroidal magnetic confinement as an economic and environmentally attractive energy source. The advanced tokamak, the spherical torus and the advanced stellarator all emphasize features which are based on the fundamental science of toroidal plasmas developed by the conventional tokamak such as shaping of plasma and magnetic profiles for increased b, utilization of the self-produced bootstrap current to optimize the magnetic configuration, and sheared plasma flows to reduce losses due to plasma turbulence. The initiatives underway and proposed for spherical tori and stellarators emphasize the common features and generic nature of the physics and technology of toroidal magnetic systems. The stellarator will explore configurations with low recirculating power that are expected to avoid plasma disruptions. The spherical torus may offer a cheaper next step in the development path since it allows smaller burning plasma devices than can be built using superconducting coil technology. While the spherical torus and the advanced stellarator are presently not as well developed as the advanced tokamak, they are expected to benefit greatly from the tokamak knowledge and infrastructure base.

Two major issues for toroidal magnetic confinement are: (1) the scaling of confinement in alpha heated plasmas and (2) the effect of dominant alpha heating on the magnetic configuration, plasma energy confinement and potential alpha driven instabilities. The basic physics of these processes has been studied using neutral beams or radio-frequency waves to simulate the effects of alpha heating. Information on the scaling of confinement during strong alpha heating and the magnetic configuration parameters required for ignition is central to developing magnetic fusion. In addition, alpha heating depends on the local plasma parameters, which in turn depend on local plasma confinement and alpha heating. Understanding and controlling this complicated non-linear feedback loop is a critical issue for all advanced toroidal magnetic systems - advanced tokamak, spherical torus and advanced stellarator, and experiments with high gain plasmas are needed. The basic strategy for the Burning Plasma Physics Initiative is to continue to use the pulsed tokamak as a research tool to cost effectively access strongly burning plasmas and to address these fundamental burning plasma issues for all toroidal configurations.

Plasma heating, current drive, fueling, particle and power exhaust are also generic and common plasma technology issues for the advanced tokamak, spherical torus and advanced stellarator. Tritium retention and handling, remote maintenance and blanket technology are closely related nuclear technologies for all toroidal systems as well. Detailed systems studies of potential power plants based on the advanced tokamak, spherical torus and modular stellarator confirms that these toroidal systems are almost identical in their capital cost and cost of electricity (COE), and are very similar in other characteristics such as plasma volume and magnet energy as shown in Table I.

Table I. System studies of advanced tokamak, spherical torus and stellarator power plants.

Advanced Reactor Innovation Evaluation Study (ARIES)

Spherical Torus
(A = 1.6)
Modular Stellarator Advanced Tokamak

Power (Thermal), GW 3.8 2.3 2.6

Power (Net Elec), GW 1.0 1.0 1.0

Capital Cost, $B(1992$) 4.3 4.3 4.2

COE, mil/kWh (1992$) 84 75 76

Plasma Volume (m3) 581 735 349

Magnetic Energy, GJ 76 80 85

Plasma Current, MA 29 ~0 11.3

3.3 Technical Contributions of the Modular Plan

3.3.1 Burning Plasma Physics Initiative

The fusion program needs the capability to extend the frontiers of fusion plasma physics that will enable discoveries in previously unexplored parameter space that have the possibility to lead to more attractive fusion regimes. The coupling of advanced toroidal physics with strongly alpha-heated plasmas is a key issue for the development of attractive toroidal magnetic reactors whether they are classified as advanced tokamaks, spherical tori advanced stellarators or reversed field pinches. The achievement of an ignited (Q 3 10) plasma will allow these scientific objectives to be achieved.

Objectives for the Burning Plasma Physics Initiative

• Determination of the conditions required to achieve high Q energy producing plasmas.

• Control of high Q plasmas through modification of plasma profiles and external sources.

• Determination of the effects of fast alpha particles on plasma stability.

• Sustainment of high Q plasma - high power density exhaust of plasma particles and energy and alpha ash exhaust, some evaluation of alpha heating on bootstrap current profiles.

• Exploration of high Q burning plasma physics in some advanced configurations/operating modes that have the potential to lead to attractive fusion applications.

These objectives would be pursued with a phased operating program very similar to that proposed for the Reduced Cost ITER program. Phase 0 : Evaluate plasma regimes and confirm hardware capability using deuterium plasmas.

Phase I : Demonstrate, control and optimize strongly burning D-T plasmas (Q 3 10) for an extended duration. Assume base line ITER performance [ HH ~ 0.85*ITER 93-H(Elm-free), bN 2 2.5] in line with the present conventional tokamak data base.

Phase II: Demonstrate, control and optimize enhanced performance (e.g. gyroBohm) or advanced physics modes (e.g.,TPX/ARIES-RS) in strongly burning plasmas for an extended duration.

Phase III: Demonstrate controlled ignition (Q 3 10) and extended burn.

Coupled technology issues (e.g., plasma exhaust/plasma facing components, tritium handling and remote handling) will also be addressed at conditions approaching those anticipated in a fusion system with the exception of some steady-state requirements.

Physics Requirements for an Advanced Burning Plasma Experiment

The physics of a burning plasma can be explored if the parameters listed below are attained.

Q 3 10, Pa/PHeat > 66%, - alpha heating dominant but still easily controlled

Burn time 3 10 tas - alpha heating, fast alpha effects ( e.g., TAE)

3 10 tE - pressure profile evolution due to alpha heating

3 3 tHe - helium ash accumulation

3 3 tcr(tcurrent redistribution ) - evolution of bootstrap current

The initial D-T experiments on TFTR and JET confirmed the single particle confinement requirements for alpha particles and were able to detect weak alpha heating in agreement with expectations (TFTR-1995, JET-1997). At Q > 10, alpha heating will dominate the plasma heating and the effect on energy confinement and pressure profile can be determined. The alpha slowing down time is in the range of 0.1 to 0.5 s for the Burning Plasma experiments to be discussed and is sufficiently short so that the alpha distribution is in equilibrium. The energy confinement time ranges from 0.6 to 3 s for experiments to be discussed and is short compared to burn times anticipated. The alpha ash confinement time is expected to range from 4 to 10 tE or from 2.4 to 30s. The devices with shorter pulse lengths operate at higher densities which means the tE is shorter for the same ntE. The normal conductor devices under consideration are expected to have burn times of several helium ash transport times. The current redistribution due to alpha heating modifications of the bootstrap current profile is a key issue for advanced burning plasma experiments. This requirement is more difficult to satisfy and must be determined for each device and specific operating mode. The current redistribution time, tcr, is Å 22 ka2Te3/2 s where k is the elongation, a is the minor radius in meters, Te is the average electron temperature in 10 keV and tcr is in seconds. The larger devices tend to have longer burn times but tcr increases at roughly the same rate. A lower temperature obtained by operating at higher density (bounded by the Greenwald limit) allows tcr to be reduced. For baseline physics assumptions and full magnetic fields the typical pulse lengths correspond to ~ 1 tcr. Fortunately, as advanced tokamak performance is attained the plasma current and magnetic field can be reduced allowing a very substantial increase in pulse length for normal conductor devices so that pulse lengths of several tcr can be attained.

Possible Facilities for the Burning Plasma Initiative

A normal conductor burning plasma device has the advantage of providing high magnetic fields and plasma currents at a reduced size and cost relative to a superconducting system since a neutron shield is not require to protect the toroidal and poloidal coils thereby allowing the major radius to be reduced. A significant cost savings is also realized by using copper alloys and inertially cooled cryogenic technologies. The copper coil systems could also allow for stronger and more flexible plasma shaping which is desirable for advanced burning plasma experiments.

A number of copper coil burning plasma devices have been studied including Ignitor (1978-98), CIT (1986-89), BPX(1990-91), BPX-AT(1991-1998) which are precooled to cryogenic temperatures prior to the pulse, HLT(1990) which was actively cooled with LN and PCAST(1996) which was actively cooled with water. The general physical parameters of these devices are summarized and compared with ARIES-RS and ITER-EDA in the following table. Projections of D-T performance for these facilities using the same methodology as ITER RC is given in a later section.

Table II. Parameters of Burning Plasma Facilities

R
(m)
a
(m)
k A B
(T)
Wmag
(GJ)
Ip
(MA)
FlatTop
(s)
Pfus
(MW)
Cost
($B)

TFTR 2.5 0.9 1.0 2.8 5.6 1.5 2.7 2 11 ~1

JET 2.85 0.85 1.6 3.8 1.5 4.2 5 16 ~1

Ignitor 1.32 0.47 1.85 2.8 13 4 12 5 150 0.5

Spherical Torus 1.05 0.75 3.0 1.4 2.6 0.5 16 steady 800 0.8

BPX-AT 2.00 0.50 2.0 4.0 10 4 6.25 12 150 0.70

CIT 2.14 0.65 2.0 3.3 10 4 11 10 150 0.73

BPX 2.59 0.79 2.0 3.3 8.1 9 10.6 10 400 1.63

Gyro-Bohm JET 3.0 1.0 2.0 3 6 10 360 ~3

HLT 3.40 1.20 2.0 2.8 8 15 20 40 500 NA

PCAST 5.00 1.50 1.75 3.3 7 35 15.3 120 400 5.3

ARIES-RS 5.52 1.38 1.7 4.0 7.98 85 11.3 steady 2170 4.6

ITER-EDA 8.14 2.80 1.6 2.9 5.68 120 21 1000 1500 10.7

The cost estimates are in FY1995$ and are for the construction project not including site costs.

The construction costs of fusion confinement facilities are a strong function of physical size, plasma current and energy stored in the magnets as suggested in Table II. An empirical fit to the cost estimates for several devices yields cost ~ B0.87 R1.85 (J.Schmidt, 1995 PCAST Machine Study) and Aamodt (Madison Forum) has found a similar relation with cost Å 25BR2 . The costs for CIT and BPX are the cost estimates at the end of their Preliminary Design Phase escalated to FY1995$. The BPX-AT cost estimate is based on a Conceptual Study during the 1991 New Initiatives Task Force activity and was a detailed cost estimate based on scaling down from BPX. The Ignitor cost estimate of $0.5B is on the high end of various cost "estimates". Ignitor, CIT, BPX-AT and BPX were costed on the basis of siting at an existing large tokamak or equivalent site. The PCAST and ITER cost estimates do not include site costs outside the fusion facility and its direct facilities costs. The ARIES cost is the total direct cost for the 10th of a kind. Consideration should be given to special studies that look for new design features or manufacturing techniques to reduce the unit construction costs for the tokamak core. The general conclusion is that compact tokamak facilities (Ignitor, BPX-AT and CIT) with major radii 2 2m have construction costs in the 2$1B class, while larger superconducting devices necessarily have major radii 3 5m with costs in the $5B range.

Estimated Fusion Performance of Normal Conductor Burning Plasma Experiments

The performance of PCAST, Ignitor, CIT, BPX and BPX-AT was estimated using a zero-D model assuming ITER-93H (ELM-free) confinement scaling with alpha heating and fuel depletion due to alpha ash accumulation calculated self-consistently. The plasma profiles were taken to be the same with a flat density profile (an = 0.1) and modestly peaked temperature profile (aT = 1). The impurity levels were taken to be 3% Be and the alpha ash was assumed to have a confinement time tHe= 5 tE resulting in Zeff Å 1.5. These assumptions are the same as the modeling assumptions made for the Reduced Cost ITER with the exception that ITER-RC assumes tHe= 10 tE for the baseline performance mode and tHe= 5 tE for advanced performance mode. For BPX-AT at Q = 10 increasing tHe from 5 to 10 tE increases the required H93-Elmfree H factor by 10%. Some of these calculations are summarized in Fig. 3.2 below.

Fig. 3.2. Fusion power gain estimates for several burning plasma experiments.

The existing confinement data base from all tokamaks is centered about an ITER93(Elm-free) H factor of Å 0.85 while the confinement data from Alcator C-mod, a prototype for the compact ignition tokamaks, has ITER93(Elm-free) H factors of Å 1.2. Ignitor and CIT are projected to ignite for ITER93(Elm-free) H factors of Å 0.85 which is the same requirement as for ignition in the ITER-EDA. Other Ignitor-like devices with somewhat lower field and plasma current such as BPX-AT are projected to achieve Q 3 10 ITER93(Elm-free) H factors of Å 0.85 and would ignite at ITER93(Elm-free) H factors less than those achieved in Alcator C-mod. Note that the full field performance of Ignitor and CIT(2.1m) is nearly the same suggesting that Ignitor would ignite at ~70% of full field and current if the enhanced Alcator C-Mod performance could be attained in D-T. At this reduced field, the flattop of Ignitor would be extended from 5 seconds to 15s.

Compact Ignition Tokamaks

The compact tokamaks (e.g., Ignitor and BPX-AT) first advocated by Coppi with higher magnetic field and higher plasma densities have additional advantages with respect to beta limits, operating density limits, impurities and fast alpha particle limits and are well suited for studying burning plasma physics during the time scales of interest. Recent results from tokamak confinement experiments, in particular Alcator C-Mod, confirm the high field compact ignition tokamak design assumptions with regard to confinement, ICRF heating, power handling and impurity control. Alcator C-Mod with ELMing H-modes tends to operate with 20-30% higher confinement relative to ITER93H scaling than larger lower field lower density tokamaks and provides confidence that Ignitor-like compact tokamaks would achieve the performance required to achieve the burning plasma objectives.

Detailed engineering designs have been carried out for copper alloy coils cooled to cryogenic temperatures and significant operating experience has been obtained. Ignitor is designed to have copper alloy coils that are precooled to 30 °K with liquid hydrogen while BPX-AT, CIT, BPX designs have copper alloy coils that are precooled by LN to 77 °K. The pulse length is determined by the adiabatic temperature rise of the conductor/structure thermal mass during the pulse. A small reduction in the peak coil current allows the pulse length to be increased dramatically for the coil/cooling configuration. For example, reduction of the magnetic field in BPX-AT from 10 T to 7 T allows the magnetic field flat top to be extended from 12 s to 56 s. This feature of Ignitor-like compact tokamaks can be used to advantage for studying advanced tokamak regimes where improved confinement and b allow the plasma current and magnetic field to be reduced as shown in the example below for BPX-AT (Fig. 3.3). In order to exploit this capability the initial design would need to incorporate or allow upgrades for active cooling of internal divertor components (as in BPX-AT) and techniques to pump helium ash during the pulse.

Fig. 5 Capability of an inertially cooled Ignitor-like tokamak to produce pulses with long burn (magnetic field flat top) times that would allow several plasma current redistribution times for studying advanced modes. Compact helium ash pumping systems would have to be provided in the divertor for the long pulses available.

An important point is that the Ignitor-like compact tokamaks can explore a broad range of experimental operating space by density variations and by reduction in the magnetic field/plasma current since they have large margins with regard to density, MHD beta and TAE limits.

Alpha Physics Considerations for the Burning Plasma Physics Initiative

After alpha heating, the primary alpha physics issue concerns alpha-driven toroidal Alfvén eigen (TAE) mode instabilities which could cause the loss of energetic alpha particles before effective alpha heating had occurred. Ideally, the burning plasma experiment should be able to avoid these instabilities while achieving high Q and to then controllably approach the stability boundary to determine the physics constraints for future devices. The TAE instability occurs when the alpha particle speed, Va, is 3 the Alfvén speed, VAlfvén, and (R/a)—ba exceeds a threshold that depends on details of the plasma and magnetic profile (e.g., b and shear). It can be shown that ba depends on bTe5/2 and the temperature can be varied by adjusting the plasma density. The curves in Fig. 3.4 are density scans for 0-D calculations where Q was held constant (by adjusting H) while maintaining constant helium ash and impurity fractions. The Ignitor-like compact tokamaks can scan the same general range of TAE instability parameters space as the ITER-EDA and PCAST devices.

Fig. 3.4. Comparison of parameters that determine the instability boundary for alpha-driven toroidal Alfvén eigen (TAE) modes.

Intermediate Sized Ignition Experiment based on Gyro Bohm Scaling

A moderate size normal conductor tokamak (R = 3m, B ~ 6 T, Ip ~ 10 MA) experiment based on extrapolation of DIII-D/JET gyro-Bohm scaling experiments has been proposed. This physics mode is not considered to be an advanced tokamak mode, such as the reversed shear mode, and therefore enhanced performance is achieved without the need for strong active plasma profile control. The advantage is a more robust (i.e., reliable) plasma configuration with potentially fewer plasma disruptions. A concept study for a similar size tokamak has been carried out for a high-performance long-pulse tokamak (HLT) with parameters as shown in Table II. The HLT design, which is also inertially cooled, increases its adiabatic pulse length by using the thermal mass of an external liquid nitrogen reservoir which is initially subcooled to 63.5 °K prior to each pulse. The toroidal field coil and central solenoid are actively cooled during the pulse by circulating the LN through the cooling channels. This design allows the pulse length to be arbitrarily extended by enlarging the liquid nitrogen reservoir, but reduces the maximum attainable magnetic field to accommodate the cooling channels. Active LN cooling of this type is being evaluated as a possibility for extending the pulse length on compact tokamaks as well as moderate scale tokamaks. Further work is needed in a multiple machine program to test the gyro-Bohm scaling and to investigate pedestal scaling especially at higher bN.

Findings on the Burning Plasma Physics Initiative

1. The compact high field tokamak utilizing cryogenic normal conductors is a potential pathway to access Q 3 10 conditions and address burning plasma physics with a facility costing 2$1B. In addition, evaluation of burning plasma physics in an advanced configuration for up to several skin times is possible.

2. Technological issues associated with the longer pulses (e.g. helium pumping and other internal components) in a compact high field tokamak need more detailed analysis and should be updated to include recent results.

3. Intermediate size (JET-scale) normal conductor tokamaks offer interesting possibilities for burning plasma physics research with costs in the ~$1.5B range.

4. Larger burning plasma tokamaks utilizing superconducting coils while somewhat more capable have costs in the several $B range.

3.3.2 Steady-state Advanced Confinement Physics Program Initiative

Objectives of the Steady-state Advanced Confinement Physics Initiative

The development, exploration and detailed understanding of high confinement, high fusion power density and high duty-cycle (steady-state) plasmas is needed for the development of economically and environmentally attractive applications of fusion power. This initiative includes subprograms on steady-state advanced tokamaks, steady-state advanced stellarators and other Proof of Performance experiments that might emerge from the ongoing Base Fusion Science and Technology Program such as the Spherical Torus.

Steady-state Advanced Tokamak Program Initiative

Outstanding progress in exploring and understanding advanced plasma regimes for short pulse magnetically confined plasmas has been made by a number of specialized medium-sized and several large size pulsed tokamaks. The next frontier in the advanced tokamak configuration is to extend high performance advanced tokamak regimes to near steady-state conditions in fusion-relevant plasmas so the plasma physics (with the exception of actual alpha physics) of a magnetic fusion plasma can be understood and optimized. The strategy is to use H and D plasmas rather than D-T plasmas to increase experimental flexibility and to reduce costs associated with tritium handling and neutron activation that are present at burning plasma facilities.

The steady-state advanced tokamak program will be dedicated to the development of the scientific basis for a compact and continuously operating tokamak fusion reactor. It will explore techniques for optimizing steady-state plasma performance through active control of the current profile, the pressure profile, the radial electric field profile, transport barrier formation, of plasma-wall interactions, and by advanced plasma shaping. Key areas to be optimized are the averaged plasma pressure (or b) through wall stabilization, the plasma confinement through transport barriers, and the current drive efficiency. This will involve making efficient use of the self-driven bootstrap current to provide a substantial fraction of the total plasma current. The integration of optimized plasma performance and efficient continuous operation will be a key issue, as will be control of major plasma disruption. Advanced tokamak operation will be at reduced current levels compared to conventional tokamaks and this will reduce the impact of disruptions.

Requirements for the Next Step Steady-state Advanced Tokamak Initiative

The plasmas needed to resolve the steady-state advanced confinement issues should have physics phenomena similar to that projected for a fusion plasma. This requirement would be satisfied if the dimensionless plasma parameters r*, n* and b* are comparable to those in a strongly burning (e.g., Q > 10) D-T plasma. Kadomtsev (Nuclear Fusion 1975) has shown that plasmas can have the same r*, n* and b* if they have the same similarity parameter, Ba5/4. In addition, the pulse length must be sufficient to allow a thorough study and controlled modification of plasma current evolution in advanced tokamak modes (e.g., pulse length 3 10 tcr), and to allow the plasma wall/divertor interaction to come into equilibrium. Some aspects of alpha particle physics (e.g., TAE modes in reversed shear magnetic configurations) can be simulated using the injection of high power beams and RF power. The physics issues and criteria have been previously discussed in detail in the Tokamak Physics Experiment (TPX) conceptual design.(e.g., IAEA paper 1994)

In addition to having dimensionless physics parameters similar to those in a strongly burning plasma and having long pulses for current control and wall equilibration, there are several other physics issues for this initiative. An advanced flexible divertor configuration is required to study the interplay between edge conditions needed for enhanced confinement modes and the requirements for heat dispersal and limited recycling back t

	Spherical Torus (A = 1.6)	Modular Stellarator	Advanced Tokamak
Power (Thermal), GW	3.8	2.3	2.6
Power (Net Elec), GW	1.0	1.0	1.0
Capital Cost, $B(1992$)	4.3	4.3	4.2
COE, mil/kWh (1992$)	84	75	76
Plasma Volume (m3)	581	735	349
Magnetic Energy, GJ	76	80	85
Plasma Current, MA	29	~0	11.3

	R (m)	a (m)	k	A	B (T)	Wmag (GJ)	Ip (MA)	FlatTop (s)	Pfus (MW)	Cost ($B)
TFTR	2.5	0.9	1.0	2.8	5.6	1.5	2.7	2	11	~1
JET	2.85	0.85	1.6		3.8	1.5	4.2	5	16	~1
Ignitor	1.32	0.47	1.85	2.8	13	4	12	5	150	0.5
Spherical Torus	1.05	0.75	3.0	1.4	2.6	0.5	16	steady	800	0.8
BPX-AT	2.00	0.50	2.0	4.0	10	4	6.25	12	150	0.70
CIT	2.14	0.65	2.0	3.3	10	4	11	10	150	0.73
BPX	2.59	0.79	2.0	3.3	8.1	9	10.6	10	400	1.63
Gyro-Bohm JET	3.0	1.0	2.0	3	6		10		360	~3
HLT	3.40	1.20	2.0	2.8	8	15	20	40	500	NA
PCAST	5.00	1.50	1.75	3.3	7	35	15.3	120	400	5.3
ARIES-RS	5.52	1.38	1.7	4.0	7.98	85	11.3	steady	2170	4.6
ITER-EDA	8.14	2.80	1.6	2.9	5.68	120	21	1000	1500	10.7