that the materials developments needed for fusion are similar to materials developments undertaken for many new technical systems, from those used in space and aircraft to materials for high speed transport and new power generating electric turbines.

DT fusion power plants built of the proper materials have all the environmental advantages one seeks in an energy source for the future, and they are achievable. Fusion plants based on any other fuel cycle such as D-D or D-He3 have by comparison no clear path to success that can presently be defended. It therefore is a matter of making a success of the enormous progress achieved on the path to DT fusion, or essentially returning to a blind path.

2. What is the benefit to ITER of the current U.S. Tokamak programs? Is the continued operation of these necessary for the ITER design effort, or do you already have sufficient data?

There are several high leverage issues for the ITER project that are currently being addressed in the U.S. tokamak research program. These issues are:

1. Divertor physics and design

2. Tokamak disruption control

3. Tokamak performance improvement
4. Tokamak steady-state operation

Experiments presently being conducted on the U.S. mainline tokamaks are focused to provide verification of ITER design issues, or confirmation of fusion physics models that are being used to predict ITER fusion performance. The results of these experiments and the verification of the ITER physics models, will determine some of the ITER design requirements. Compared to our present knowledge base, this information will allow a higher physics margin for attaining the ITER goals, or a higher engineering margin for device reliability. Both lead to overall higher performance, reliability, and higher availability of the ITER device.

The ITER-relevant experiments presently being conducted by the mainline U.S. tokamak programs are as follows:

Divertor Physics and Design

The DIII-D and the Alcator C-Mod will be testing divertor design scenarios. The divertor is the system that removes the particle energy from the reacting plasma and is presently the most challenging physics and technology issue in the ITER project. The DIII-D program will be simulating some of the ITER operating conditions as far as divertor heat load, and in the future will be testing new divertor concepts designed to significantly reduce the divertor heat load. The outcome of these experiments will have direct impact on the ITER divertor design and materials R&D. The Alcator C-Mod device is similarly testing divertor concepts is a slightly different geometry and using different divertor materials from the DIII-D device. The combination of these two tokamaks allows the U.S. to have greater flexibility and diversity in simulating future ITER divertor requirements and, thereby, verifying an ITER divertor design and materials choice in the shortest time period.

Tokamak Disruption Control

Alcator C-Mod, DIII-D, and TFTR, are acquiring information relevant to the advance identification of the precursor signals that lead to plasma disruptions. These unplanned disruptions are responsible for some of the large electromagnetic forces that the ITER device will see and are also responsible for some of the material's requirements for the plasma divertor. The disruption data from these operating tokamaks should result in a better understanding of the genesis of plasma disruptions, thereby allowing us to identify operating techniques to reduce the frequency and severity of disruptions. Applying these techniques to ITER operation can reduce some of the ITER engineering design requirements and will result in significantly greater reliability and availability of the ITER system.

Tokamak Performance Improvement and Steady-State Operation

The Alcator C-Mod and the DIII-D tokamaks are conducting experiments investigating new ways of controlling the plasma current and current profile through non-inductive current drive techniques. These experiments will identify a path for improved ITER plasma performance through current profile control, and will demonstrate the radio frequency heating and current drive technologies that will be utilized on ITER to realize long-pulse or quasi steadystate operation in the ITER device. These current drive and profile control techniques are deemed to be critical, not only to ITER, but to optimizing future tokamak fusion power demonstration reactor performance and realizing steadystate tokamak operation.

The TFTR device, along with the Joint European Torus, will be the first devices to experiment with high power self heating from a deuterium/tritium reaction. Experience with this self heating of the fusion produced alpha particles will give us our first evidence of an alpha particle heated plasma which will be the typical operating regime of the ITER device. It is worth noting that TFTR and JET are the only devices in the world that are presently capable of and authorized to conduct deuterium/tritium experiments prior to ITER operation.

In summary, we have sufficient data to proceed with the design and the engineering of the ITER device. We do not yet have sufficient data to confirm that the ITER design and materials selection is correct. That verification will come out of the continued implementation of the non-inductive current drive, current profile control, and advanced divertor experiments on the DIII-D and Alcator C-Mod tokamak devices.

3. It appears that from your schedule that ITER is planned to be well underway before the TPX comes on line. What is the real benefit of the TPX to ITER?

The DOE schedule for TPX puts first plasma operation in March 2000. If the transition from the Engineering Design Activities to the construction of ITER occurs with no delay in 1998, the earliest possible date for first plasma operation in ITER will be 2005. Experience gained during the construction of TPX in the areas of superconducting magnets, cryogenic systems, low-activation materials, robotics for remote maintenance, plasma-facing components, and computer controls will dramatically enhance the capability of U.S. industry to contribute to ITER construction. Thus, the TPX will give U.S. Industry experience in design and

construction of major fusion components during a period which our competitors in Europe and Japan have been actively building new devices and major upgrades.

The problem of handling high heat fluxes to the divertor, while maintaining good impurity control, is a key technical issue facing ITER. The TPX will be unique in its ability to develop and test advanced solutions to the divertor heat-flux and impurity-control problems, in full steady state. While TPX is scheduled to operate after ITER construction has just begun, experiments on TPX can still influence ITER divertor operations as well as the eventual upgrades to the divertor system. The low activation levels in TPX will allow for divertor iterations and modifications that would be very costly in ITER. Thus, divertor experiments on TPX will confirm the most effective approach to future divertor design upgrades on ITER.

The technical objectives adopted for the ITER EDA dictate that the ITER design must accommodate upgrades to allow tests of steady-state operation by noninductive current drive. The TPX will be unique in the world program in regard to its capability for steady-state current drive in a shaped, divertor tokamak. Thus, TPX will provide crucial information for decisions on current-drive techniques to be employed in the later phases of ITER; the flexibility being designed into TPX to allow testing of multiple current-drive techniques would be prohibitively expensive to implement on ITER. The most attractive steady-state operating regimes can be transferred quite directly from TPX to ITER.

In summary, the capability of TPX for flexible experimentation will provide very valuable guidance in optimizing the operation of ITER and in the choice of future upgrades to ITER systems.

In addition to areas in which TPX and ITER have common technical features, each has its own unique mission-ITER to demonstrate controlled ignition, integration of nuclear technologies and extended burn of reactor-grade plasmas in the 1000-megawatt range-TPX to develop the scientific basis for an economical, smaller, continuously operating reactor. Together, the results from the two facilities will allow a more advanced fusion reactor demonstration by 2025 than would be possible on the basis of ITER alone.