17th IEEE/NPSS Symposium on Fusion Engineering

Summary of the 17th IEEE/NPSS Symposium on Fusion Engineering, San Diego, California,

October 6-10, 1997

Introduction, (Mark Tillack, UCSD)

During the past 33 years, the Institute of Electrical and Electronics Engineers (IEEE) Nuclear and Plasma Systems Society (NPSS) Symposium on Fusion Engineering (SOFE) has been a leading technical forum for exchange of information on the engineering and technology of fusion energy. The 17th Symposium was held October 6-9, 1997 in San Diego, California. This report includes a brief summary of each technical session written by the session chairpersons. The Symposium proceedings, including full papers for each presentation, will be published by the IEEE.

Eleven plenary speakers provided the backbone of the symposium:

M. Keilhacker: Deuterium-Tritium Experiments in JET and Their Implications for ITER
M. Cambell: Recent Accomplishments in ICF
D. M. Meade TFTR Retrospective
A. Sakasai: High Performance and Steady-State Experiments on JT-60U
P. I. Petersen: Results from the DIII-D Tokamak
J. H. Irby: The Alcator C-Mod Tokamak and Recent Results
O. Motojima: Review of Engineering Progress in LHD Project
V. Erckmann: The W7-X Project: Scientific Basis and Technical Realization
R. Aymar: Status and Prospects of ITER
J. Paisner: The National Ignition Facility for Inertial Confinement Fusion
R. W. Conn: From ITER and NIF to an Attractive Commercial Energy Source

Twelve oral sessions included both invited and contributed lectures:

O1: Tokamak Engineering and Operations
O2: Safety & Environment and Power Plants
O3: New Machines
O4: Tritium Systems and Tritium Control
O5: Plasma-Facing Component Technology
O6: Heating and Current Drive Systems
O7: Plasma Control and Power Systems
O8: Inertial Fusion Engineering: Targets, Drivers & Chambers
O9: Magnet Systems
O10: Physics and Advanced Devices
O11: Next Generation ICF Devices
O12: Data Acquisition and Diagnostics

The majority of the presentations were provided in 23 poster sessions:

P1: Magnet Systems - I
P2: Safety & Environment - I
P4: Divertor and Plasma-Facing Component Engineering - I
P5: Data Acquisition and Diagnostics
P6: Magnet Systems - II
P7: Nuclear Engineering
P8: Materials Engineering
P9: Tritium Systems
P10: Vacuum Systems
P11: Plasma-Material Interactions and In-Vessel Tritium Control
P12: Safety & Environment - II
P13: Divertor and Plasma-Facing Component Engineering - II
P14: Electromagnetics and Electromechanics
P15: Assembly, Fabrication and Maintenance
P16: Plasma Control and Control Systems
P17: New and Proposed Machines
P18: Inertial Fusion Engineering
P19: Plasma Physics
P20: Blanket, Shield and Vacuum Vessel Engineering
P21: Power Plant Studies
P22: Neutral Beam Injection and Fuelling Systems
P23: Power Systems

The Technical Program Chairman for the Symposium was Mark Tillack (UCSD), who was supported by co-chairmen Ken Schultz (GA), Phil Heitzenroeder (PPPL), and Wayne Meier (LLNL). Session chairmen are identified below, together with the sessions they chaired.

Plenary Sessions

Plenary Session L1, Ken Schultz (GA) and Valerie Chuyanov (ITER-US)

The opening plenary session of the Symposium established the ecumenical tone that was evident throughout the meeting by having presentations on recent exciting events in both the magnetic side and the inertial side of the world fusion program.

At the first lecture Dr. Martin Keilhacker, Director of the JET Joint Undertaking, presented the latest results of JET's D-T experiments and their implications for ITER. In D-T experiments performed during the last several weeks before the Symposium, JET established a new record for fusion power production with 13 MW of fusion power and up to 14 MJ of fusion energy per pulse. It produced a total of 340 MJ of fusion eneergy in 100 full D-T pulses and achieved energy amplification of Q = 0.6. This series of experiments enabled the JET staff to study the scaling of important physical phenomena with ion mass. The results were overall quite favorable for ITER.

It was found that the power threshold of the H-mode has a strong favorable dependence on ion mass, ~ 1/A. It the same time, the energy confinement is practically independent of ion mass. While it would be beneficial for ITER if there were a positive dependence of confinement on mass, this effect is more than compensated by a stronger than expected dependence of confinement on density. Energy confinement scaling was found to be close to Gyro-Bohm, but without mass dependence. Studies of the JET plasma edge parameters have shown that the edge gradients scale as T^1/4. This is consistent with confinement scaling as (a.r)^1/2, which is also favorable for ITER. About half of the heating power was provided by fusion alphas and the effects of this heating were observed. No negative effects connected with Alfvén eigenmodes driven by alpha particles have been observed.

These positive results bode well for the fusion program and for the recent proposal to extend JET operation through 2003 with extensive studies of open and closed divertor geometry followed by another program of D-T experiments.

The second presentation was given by Dr. Michael Cambell, Associate Director of Lawrence Livermore National Laboratory (LLNL) for Laser Programs, summarizing recent accomplishments in the Inertial Confinement Fusion (ICF) Program. He began with an overview of the National Ignition Facility (NIF) project that just broke ground at LLNL. This 1.8 MJ, 500 TW laser is the culmination of 25 years of R&D on ICF at LLNL, Los Alamos National Laboratory (LANL) and the University of Rochester Laboratory for Laser Energetics. It is designed to achieve ignition and modest energy gain. Both indirect drive and direct drive implosion experiments will be possible in this facility which will be an important element in DOE's "science based stockpile stewardship" program, and also will allow significant advances in R&D on inertial fusion energy applications.

The science and technology needed for the NIF have been developed on experiments such as the Beamlet at LLNL which is a full scale prototype of one of the 192 NIF beamlines, and through an extensive program of industrial facilitization to develop the capacity to produce the quantities of large, very high quality laser components the NIF will need. The scientific basis for ignition has been developed over the years through extensive programs of experimentation and analysis of the physics of laser-plasma coupling, implosion symmetry control and hydrodynamic stability on the Nova and Omega lasers and other ICF program facilities. Important developments have included recognition of the need for and the techniques to achieve laser beam smoothing by mechanisms of introducing incoherence.

The target designs for ignition include both direct and indirect drive candidates with both beryllium and plastic capsules, so there are multiple options for ignition, giving a robust program. The heart of all target designs will be a spherical shell of frozen D-T, made by techniques being developed by LLNL, LANL and General Atomics. The NIF will develop the physics basis for inertial fusion energy (IFE) applications, and will permit some IFE technology development, but NIF will only shoot a few targets each day and IFE must shoot a few every second.

To develop a rep-rated laser for IFE application LLNL is now constructing the Mercury laser, a 0.1 kW, 10 Hz diode-pumped solid state laser (DPSSL) that should achieve efficiency of 10% or more. The DPSSL also will allow application of the "chirped pulse amplification" technique, wherein laser pulses are stretched, amplified and recompressed to obtain petawatt power laser pulses. These intense pulses offer the possibility of igniting a cold compressed target at much lower energy than "conventional" ICF, raising the possibility for very high gain and a low cost path for IFE development. Further, these intense pulses offer exciting potential for materials processing and biomedical application.

Plenary Session L2, Michael Williams (PPPL) and Phillip Heitzenroeder (PPPL)

Dr. Dale Meade of the Princeton Plasma Physics Laboratory presented the opening talk of this session: "TFTR - A Twenty Year Retrospective". He began with a reminder of the general political environment twenty years ago when TFTR was being constructed. Memories of gas rationing and heating oil shortages were still strong, energy was a major public issue and fusion budgets were strong. This contrasts with the present situation, where fuel supplies appear to be plentiful, the budget deficit is a major issue and fusion budgets have been declining. The major fusion issues twenty years ago were confinement scaling, MHD stability, impurity control, fusion power production and large scale engineering and operation of D-T devices. During those twenty years, the widely used Lawson criteria n_iT_it_E used as a measure of fusion progress has increased by a factor of 1000 -- from 0.01 x 10²⁰ to 10 x 10²⁰m^-3 s keV.

TFTR demonstrated the maturity of fusion engineering and helped provide confidence that a large-scale tokamak can be designed to meet specifications, achieve cost and schedule goals and be operated safely and reliably. TFTR operated with D-T for more than 3 years, producing over 1090 D-T discharges and 1.5 GJ of fusion energy. TFTR's highest power plasmas produced 10.7 MW of fusion power with fusion power densities in the core exceeding those expected for ignited plasmas producing 1500 MW in ITER. In spite of operating in a D-T environment with its attendant activation and strict safety requirements, TFTR had >85% availability and the neutral beams were operated above their original design ratings (40 MW vs. 33 MW), as were the TF coils (6 T vs. 5.2 T). Early D-T experiments demonstrated a favorable "isotopic" scaling with D-T plasmas and confirmed the ITER ICRF heating mode using second harmonic tritium resonance. TFTR's "ITER-like" plasma core conditions, coupled with comprehensive D-T diagnostics, allowed detailed measurements of alpha particle energy distributions which were used to test calculations of alpha particle dynamics. TFTR made the first direct observations of alpha heating and alpha driven instabilities in a tokamak. It also performed detailed studies of plasma turbulence and transport, sheared plasma flow, and sheared magnetic fields that demonstrated the limitations of the traditional empirical scaling models and led to the development of new models of plasma transport based on marginal stability and sheared plasma flow. In looking towards the future, Dr. Meade feels that these experiments and theoretical models indicate the possibility of increasing the fusion power production in TFTR to over 20 MW with new opportunities for studies of strong alpha heating with only a modest extension of TFTR operating regimes.

Dr. A Sakasai of the Japan Atomic Energy Research Institute followed with his talk entitled "High Performance and Steady State Experiments on JT-60U". The objectives of JT-60 research are to establish the physics basis for a steady state tokamak fusion reactor and contribute to ITER physics R&D. JT-60U has made significant progress in support of these objectives in key areas such as improved confinement, radiative divertor, and non-inductive current drive. Improved confinement regimes with reversed shear and high b_p H-modes were successfully developed in 1996. Triple products as high as 1.5 x 10²¹ m^-3 s keV and a world record temperature of 45 keV were achieved.

Dr. Sakasai noted that the reversed shear configuration is one of the leading scenarios for steady state tokamak operation. The JT-60 U reversed shear experiments produced a Q_DT^eq of 1.05 with a record stored energy level of 10.9 MJ at a plasma current of 2.8 MA. Negative ion based neutral beam (NINB) experiments also were performed in 1996 in support of the goal of developing methods for steady state, non-inductive tokamak operation. The beamline is designed to operate at 0.5 MeV with a total power of 10 MW. In the initial experiments, 2.5 MW at 350 keV achieved a NB driven current of 0.28 MA with a current drive efficiency of ~8 x 10¹⁸ m^-2 A/W. The beam at this energy achieved a high neutralization efficiency of 60%. TAE mode excitation was observed during NINB injection.

Halo current measurements indicated I_h/ I_p= 0.05 - 0.25 with a toroidal peaking factor (TPF) ranging from 1.4 - 3.6 and the product I_h/ I_px TPF < 0.52. A W-shaped pumped divertor similar to that proposed for ITER was brought into operation in May 1997. Its pumping speed is 13 m³/s. Experiments are underway to examine radiative divertor operation with high recycling. Future plans include construction of a pellet injector capable of continuous pellet injection at 20 Hz and 1 km/s, full current drive at high plasma currents by high power NINB injection and high performance radiative divertor experiments.

Plenary Session L3, Enzo Bertolini (JET) and Keith Thomassen (LLNL)

The main features and recent results of the two major tokamaks still operational in the USA, were presented in this session.

Peter Petersen gave a talk on behalf of the DIII-D Team. Several new systems, upgrades and repairs were reviewed. A new divertor was installed at the top of the vessel to pump high triangularity plasmas, and a new view for the motional Stark effect diagnostics also was installed. Two gyrotrons, rated 1 MW each, became operational. Steerable mirrors allow the gyrotron to heat the plasma from the center to the edge. The plasma control has been upgraded to `isoflux' control and exploits the new real-time plasma equilibrium capability. The faulty ohmic heating coil has been repaired and the circuit will be reconfigured to make use of the additional 2.5 Vs. Remote experimental operations of the machine were done in collaboration with LLNL.

New physics results show that ExB rotational shear causes a transport barrier in the plasma for the ions and the electrons. VDE's have been studied, and a toroidal peaking of 4 is seen. Halo currrent can be 30% of the plasma current. Argon pellets can lower the force and heatloads on the vessel. Additional ECH power (from 2 to 10 MW) is planned for controlling the plasma profiles. Divertor development will include pumping in the private flux area, study of a more closed divertor, and dumping of a double null advanced tokamak shape.

Jim Irby summarized the salient features of the Alcator C-Mod machine, its construction and engineering features, and its missions in divertors, transport, and advanced tokamak physics. Recent results from disruption, H-mode, and electron cyclotron discharge cleaning also were presented. C-Mod has been operated at 8 T, near its design goal of 9 T, and at currents to 1.5 MA. It is planned to operate up to 2.5 MA in the future, and the rf power will increase from 4 MW to 12 MW. It operates over the wide density range of n = 0.2 - 15 x 10²⁰m^-3.

One of the major contributions of the C-Mod machine to the tokamak physics and engineering program is on disruptions and the effects of eddy and halo currents, which can be rather dramatic at its highest fields. The structures are already highly stressed at toroidal field currents of 250 kA, with forces up to 10⁸N. Consequently, an extensive set of halo diagnostics have been installed and they first discovered the asymmetry (peaking factors of 1.5 - 2) and scaling of the halo currents (proportional to the square of the plasma current and inversely with the toroidal field).

In a study of the H-mode transition they note the importance of the plasma edge and pedestals, and have implemented several new high spatial resolution diagnostics there. In addition to the existing tangential XUV and poloidal (38 channel) soft x-ray arrays, they plan to add a tangential interferometer and modification of the core Thomsen system. Their studies have greatly extended the empirical scaling law of the L-H mode transition, to values of power per unit of surface of 0.6 MW/m² and density-field product to 25 x 10²⁰ T-m^-3.

Using a 2.45 GHz ECH source of about 3 kW, the cylindrical resonant surface is swept from the inside to the outside of the vessel for discharge cleaning of the walls. This results, after about 10 days of continuous operation, in a clean vessel. New studies to determine the plasma parameters of this ECH generated plasma have begun, and it is found that the temperature and density profiles (T_e ~ 6-10 eV, n_e ~ 10¹⁶ m^-3) are rather broad, so that sweeping the resonance (which will be very difficult in large superconducting machines such as ITER) may not be necessary.

Finally, the planned upgrades of C-Mod over the next two years were described. In the divertor region, the inner plate will be modified, a bypass flap added, and full cryopumping installed in the outer divertor. With a prototype pump they have already reduced the recycling and core impurity concentration by a factor of 2. Structurally, the divertor will be made compatible with full 9 T field and 2.5 MA current. There are many planned rf upgrades as well, with a new 4-strap, 4 MW ICRF antenna constructed by PPPL for single port operation, and a 4 MW lower hybrid system at 4.6 GHz. Later, another 4 MW ICRF antenna will be added for a total of 12 MW of rf power.

With these new capabilities in field, current, power, and diagnostics, the C-Mod program should be able to significantly extend the parameter ranges of study in their various program missions.

Plenary Session L4, Tom Simonen (GA) and Fred Puhn (ITER-US)

O. Motojima opened the session with his talk entitled: "Review of Engineering Progress in LHD Project". The LHD project in Toki, Japan is based on past experience in Japanese original heliotron research. LHD is approximately 90% complete at this time, and plasma operation is planned to start in March 1998. It will be the first superconducting helical type device. The machine has a major radius of 3.9 m, a minor radius of ~0.6 m, a field of 3 Tesla (in Phase I), and ultimately will be equipped with 40 MW of heating power. The goals of the program are to perform physics experiments that will extrapolate to breakeven conditions. The operational program will include demonstrating advanced toroidal operation, confinement improvement, currentless operation, and use of many generic fusion technologies.

In this paper, he summarized the engineering progress developed during the LHD device construction which were required to establish the mission of LHD experiments. The LHD uses NbTi superconducting coils. The helical coil is a continuous winding incorporating about 36 km of conductor, cooled by pool boiling liquid helium. It was fabricated using a specially designed helical coil winding machine. LHD also uses circular poloidal field coils which are cooled by forced flow liquid helium. The total cold mass of the LHD is about 900 tonnes.

Holding tight tolerances during construction was a critical requirement, and this has been accomplished successfully using precise and rigid tooling. The helical coil has a tolerance of +/-2.1 mm on the minor radius and -1.5 mm on the major radius of the winding. The PF coils have a tolerance of +/-1.4 mm on the radius and +/-1.5 mm on the height. The welded vacuum vessel had a more relaxed tolerance of 10 mm.

The first cooldown is planned in January 1998. The cooldown scenario is carefully programmed to keep the temperature differences in the system below 50K. This results in an initial cooldown time of about one month. The LHD is expected to produce an equivalent Q of about 0.1 and a temperature of about 10 kev. The program is designed to lead up to a Demo reactor in about 2025.

The second talk, entitled "The W7-X Project: Scientific Basis and Technical Realization," was presented by V. Erckmann. The W7-X stellarator is in an early stage of construction in Greifswald Germany. The design is based on experience with previous stellarators operated in Germany. The goals of the program are steady-state operation, good confinement, a reactor relevant beta of ~5%, and development of a divertor. W7-X will have a major radius of 5.5 m, a minor radius of 0.55 m, and 5 magnetic field periods in the coil system.

W7-X is unique in that it uses modular superconducting coils to provide the magnetic fields. There are 50 twisted coils of 5 different types in the array. There are also 20 planar coils used for experimental purposes. The NbTi superconductor is cooled by forced flow liquid helium. The cold mass is about 350 tonnes. The field at the center of the plasma is 3 Tesla. The conducting cable incorporates an aluminum jacket, which is aged at elevated temperature after winding to develop its full structural strength. A prototype coil is now being built, and will be tested at FZK. In addition, a test section of the machine is being built, including the cryostat, vacuum vessel, coils, and ports. This test section will verify the manufacturing tolerances, assembly procedures, and liquid helium operation.

An important part of the experimental research program is the development of an efficient divertor configuration. The initial divertor will use a number of plates which interact with "natural" magnetic islands near the edge of the plasma. Simulations indicate that the heat deposited on the plates is accommodated for various operating modes. The divertors are design to handle 10 MW steady state heat load. Cryopumping of the divertor is included in the design.

Plasma heating will be supplied by ECRH, neutral beam injection, and ICRF. The ECRH system provides steady state power at 140 GHz. The ECRH system will use 10 units of 1 MW each to provide plasma start-up, profile shaping, and bootstrap current compensation. The microwave transmission is an optical system with steerable launchers. The neutral beam will be the type used on ASDEX-upgrade, and will start out at 5 MW. The machine can be later upgraded with more neutral beam power and also possibly negative ion beams.

W7-X will be located in a totally new facility at Greifswald. The building will be completed in 1999, and the machine will be complete in 2004. The program is intended to show the reactor capability of the stellarator.

Plenary Session L5, Charles Baker (UCSD and ITER/US)

The final plenary session of the conference included three invited papers covering the two major facilities of fusion R&D, i.e., the International Thermonuclear Experimental Reactor (ITER) and the National Ignition Facility (NIF), and a look ahead towards future, attractive fusion power systems.

The ITER Project Director, Dr. Robert Aymar, presented the status and prospects of ITER. He reviewed ITER's principal mission, namely to establish the scientific and technological feasibility of magnetic fusion energy, which is to be accomplished via an international project involving the European Union, Japan, Russian Federation and the United States. A comprehensive design has been developed and will be presented in a Final Design Report in 1998. This report will include ITER's physics basis, system and component design descriptions, cost estimates, construction schedules and site requirements. Dr. Aymar also described the status of R&D activities carried out by the ITER Home Teams. These activities are focused on seven main projects including the central solenoid model coil, toroidal field model coil, a section of the vacuum vessel, first wall/shield modules, divertor cassette and two remote maintenance demonstrations for the shield modules and divertor cassettes. Numerous papers were presented at the conference describing many details of the ITER Project.

Dr. Jeffrey Paisner provided an overview of the NIF Project. The NIF device is based on a 1.8 MJ, 500 TW laser which is designed to achieve ignited pellet burns at modest energy gains. The NIF facility is under construction at the Lawerence Livermore National Laboratory and is scheduled to be completed in 2003. The NIF laser is a neodymium-glass system working at the third harmonic. The NIF laser and target building will occupy an area of 125 x 170 m². Key components of the laser system include amplifiers, cavity filters, a master oscillator, beam transport systems and the final optics assembly. NIF is designed for both direct and indirect drive pellet implosions. The total project cost during construction is $1.2B.

Finally, Prof. Robert Conn provided a perspective of the current status and future outlook of the US fusion program based on recent deliberations of the Energy R&D Panel of PCAST: the President's Committee of Advisors on Science and Technology. In 1996, funding needs to continue the US fusion program with its previous set of objectives became divergent with actual allocations from Congress. "Restructuring" of the program resulted, with stronger emphasis on fusion sciences and innovations which could lead to more attractive end products. The PCAST report attempts to place fusion together with other energy R&D programs in a context which can provide sustainable funding. These other programs include end-use efficiency, renewables, fossil improvements, and advanced fission.

Energy research is being conducted now in a radically changing national and global marketplace. Deregulation, together with increased concern over the prospect of global climate change, are affecting energy portfolio R&D decisions. While on the one hand, the fusion program will continue to support basic scientific research in high-temperature plasma physics, it also must find an appropriate strategy to support its energy objectives. For that purpose, we must address clearly the relationship between developing an attractive fusion product, the cost of an energy R&D pathway, the changing marketplace, and the issue of global climate change.

Oral Sessions

O1: Tokamak Engineering and Operations, Osamu Motojima (NIFS)

Dr. A. von Halle reported on the final operations of the Tokamak Fusion Test Reactor (TFTR). In April 1997, TFTR completed its final operating period, bringing to close a highly successful phase of research in plasma science. The device produced over 80,000 high power plasmas since 1982 with the objectives of studying the plasma physics of large tokamaks, gaining experience in the solution of engineering problems associated with large fusion systems, and demonstrating fusion energy production from the burning, on a pulsed basis, of deuterium and tritium in a magnetically confined toroidal plasma system. In 1993, TFTR became the first magnetic fusion device to study plasmas using nearly equal concentrations of deuterium and tritium. Since that time, over 1000 D-T experimental shots and over 23,000 D-D shots have been carried out demonstrating new regimes of plasma confinement, proof of alpha heating, and reactor level fusion power densities by producing a plasma which yielded over 10 MW of fusion power at a corresponding central fusion power density of ~2.8 MWm^-3. The TFTR technical systems routinely operated at or beyond the original design criteria throughout the period, maintaining an impressive machine availability of >85%. Safe operation in D-T has been demonstrated, with over 950 kCi of tritium processed within the constraints of a 50 kCi site limit and a 20 kCi machine limit.

Dr. D. Stork reported on "JET Engineering Development toward D-T Operations in an ITER-like Machine Configuration". The Joint European Tours (JET) has recently begun a series of experiments with deuterium-tritium plasmas (DTE1). These are the first plasma experiments with a 50:50 D-T mixture in a tokamak with a divertor. Extensive technical work was carried out to ensure that the JET machine and its major subsystems were able to carry out an extended period of D-T operation. The JET Active Gas Handing System has supplied around 40 g of tritium to the Torus and Neutral Beam Injectors (NBI) and has reprocessed batches of the Torus and NBI. The injection of tritium beams at up to 155 kV energy and up to total powers of 11.3 MW has taken place. Extensive deterministic analyses of Design Basis Accidents (DBAs) establish changes required to protection systems for D-T operation and to satisfy the regulatory authorities. In the one serious fault which has been encountered in DTE1, the monitoring systems proved capable of identifying a tiny leak in the TNBI system (about 7 orders of magnitude below the Design Basis LOCA). Operations were suspended safely.

Dr. E. Bertolini reported on "Current Engineering Issues and Future Upgrading of the JET Tokamak". Flexibility and suitable stress margins were included in the original design, so as to allow modifications and upgrading of the machine to follow the evolving requirements of the physics issues, i.e., toroidal coils, vacuum vessel, additional heating power, and power supply upgrades.

- The plasma current has been increased up to 7.0 MA, the X-point magnetic configuration has been established up to 5.0 MA, and the first wall was progressively covered with CFC or beryllium tiles.

- In this machine configuration, Q_DT^equivalent>1 was achieved and the first ever controlled thermonuclear experiments in D-T produced 1.7 MW of fusion power (Nov. 1991).

- The need for active control of the impurities requires an axisymmetric pumped divertor, which was installed without replacing any of the major components of the machine.

Further progress in performance can be achieved by increasing the toroidal magnetic field to 4.0 T. If JET will be extended beyond 1999 and funds will be made available, additional heating power could be substantially increased and new machine configurations could be tested. The 15 years of operational history indicate that JET is an experimental tool that can be rejuvenated, to make progress in the understanding of fusion physics for several years to come.

Dr. S. Ciattaglia reported on the Frascati Tokamak Upgrade (FTU). About 3500 pulses were performed in the past two years (mostly in D₂) with a toroidal field between 2.5 and 8 Tesla and a plasma current between 0.35-1.3 MA. The main research activities were concentrated on plasma MHD studies with different heating scenarios. Lower hybrid waves current drive (LHCD) and electron cyclotron resonance heating (ECRH) were used both alone and together to investigate synergistic effects. Strong effort was devoted to high density (1x10²⁰m^-3) current drive experiments and investigating the possibility of obtaining and sustaining shear reversal configurations. After a 3-month shutdown planned at the beginning of 1998 to complete commissioning and debugging of all four ECRH lines, the main program during the next two years will be largely devoted to experiments with the three rf systems (LHW, IBM and ECRH). In particular, LHW heating and CD at high density and high plasma current (10²⁰m^-3/0.5-1 MA) were demonstrated, combining LHW+ECRH and LHW+IBW injection. Major physics issues are central heating at high density (~10²⁰m^-3), transient transport studies, and MHD mode control through ECRH. The long-term programme envisages studying advanced scenarios (high-beta, large bootstrap fraction, shaped plasma).

Dr. G. Martin reported on Enhanced Performance for Long Pulses on Tore Supra. Among the large tokamaks in operation around the world, Tore Supra has the unique feature of a superconducting magnet which provides a permanent toroidal field. Its main axis of research is therefore concentrated on the control of multi-MW plasmas during long duration, with steady-state as an ultimate goal. Both lower hybrid and fast waves are used to drive the current. Enhanced performance related to current profile shaping has been intensively studied. A world record in injected energy was achieved in 1996 with 280 MJ during a two minute shot. Very long evolution times are then possible in plasma wall interaction physics.

On the basis of these results, an enhancement in the capability to handle large power and to control the particles over long duration is planned. This is the main motivation for the Composants Internes Et Limiteur (CIEL) project, which consists mainly of upgrading the first wall components:

- A new toroidal belt limiter in the lower of the vessel, which is made from carbon fibre composite (CFC) brazed on copper tubes.

- A set of pumps to evacuate all types of gas species through the toroidal throat of this limiter, allowing improved density control.

- A new water-cooled radiation screen to cover as much as possible of the inner vessel, to avoid uncooled parts interacting with the plasma.

O2: Safety & Environment and Power Plants, L. Topilski (ITER-US)

"EASY-97: A Multipurpose Activation and Transmutation Code System", by R. A. Forrest. The European Activation System (EASY) is a complete tool for the calculation of activation in materials exposed to neutrons up to 20 MeV. EASY consists of the inventory code FISPACT and the European Activation File (EAF), which contains various libraries on nuclear data. The EAF-97 library contains about 12,500 excitation functions involving 766 different targets from ¹H to ²⁵⁷Fm, in the energy range 10^-5 eV to 20 MeV. Major effort has gone into verification, validation and quality assurance of EASY-97. This, together with EASY's explicit treatment of uncertainties, is of great importance as the need to license ITER approaches. For the longer term, EASY's facilities to analyze pathways of activation product production is of importance in the optimization of low activation fusion materials.

"Design-Point Determination for the Commercial Spherical Tokamak" by R. L. Miller. This study was done for the ARIES Team, which provides conceptual design projections for future commercial fusion power plants, characterizing their environmental and economic aspects as driven and constrained by presently understood physics and reasonably extrapolated technology. Interest in the Spherical Tokamak stems from its high beta and high bootstrap current. As a result of this study, parameters for a 1-GWe ARIES-ST Commercial-Power-Plant, including the total capital cost and cost of electricity, were presented for 5 different aspect ratios. Tradeoffs among system power density, recirculating power, plant availability, technical and operational complexity, and coolant and structural material choices are resolved in the context of their impacts on capital and operating costs. The Spherical Tokamak power-plant projection is somewhat problematic due to issues such as toroidal field coil centerpost, recirculating power fraction, high power density, etc.

"Identification of Postulated Accident Sequences in ITER", by N. P. Taylor. For ITER accident analysis to be complete, it must be based on a comprehensive list of fault conditions, or accident initiating events, covering all conceivable hazards arising in the plant. Two independent and complementary approaches have been employed, illustrated in this paper by a portion of a global fault tree and a sample event tree. These approaches are the component-level (bottom-up) and the top-down approaches. The former is based on the application of systematic methods which seek to catalogue all potential faults in the ITER systems, and to consider the conceivable consequences of these faults. The focus is on the failure of individual components, and based on the design in as much detail as available. The top-down approach starts at the plant level, and takes a global view of the potential hazards and the safety functions which provide protection. By considering the abnormal events which could have to occur to realize these hazards, a list of accident initiators is again produced, in terms of system or in some cases component faults. Using this approach, the resulting catalogue of primary initiating events (for ITER accident analysis) and event sequences have been checked carefully to ensure that each is covered directly and indirectly, by one of the reference accidents chosen for detail analysis. This has confirmed that the consequences of all identified accidents are enveloped by the assessed consequences in one or more of the analyses. Thus, the outcome of these studies, taken together with the results of reference accident analyses, gives confidence that the ITER engineering design will achieve its safety targets.

"Implementation of the ITER Confinement Function", by D. A. Dilling. This paper summarizes the ITER confinement functional requirements, describes the confinement strategy for the ITER tokamak, and shows how the design meets the objectives. All radionuclide sources are confined by at least two barriers. The first barrier for the ITER plasma and in-vessel deposits is the vacuum vessel (VV) and the ex-vessel portions of the primary heat transfer system (PHTS) of the plasma facing components. The second barrier starts with the cryostat vessel (CV) and is extended to include the heat transfer vaults. Thus, two critical components are high quality vacuum vessels, which must be intact for the machine to operate independent of safety considerations. Features to prevent or limit releases via various penetrations of the VV are tailored to the characteristics of the potential hazards associated with each system, as in a chemical plant. This strategy includes first and second barriers in various heating and current drive systems interfacing with the plasma. The primary heat transfer coolant is contained by the cooling systems and by the PHTS vaults that form the second barrier. Outside the second barrier, some volumes are assigned confinement-related functions and are equipped with heating, ventilation, and air-conditioning (HVAC) systems which treat potential releases. These features work together to assure that the potential impact on public health and safety is within acceptable limits for all credible challenging events. The paper also includes limited information on the confinement of sources during maintenance, in the tritium plant, and in the hot cell building.

"Analysis for the Safety Case for JET D-T Operation", by A. C. Bell. Approval has been given by appropriate safety and regulatory authorities for D-T operation of JET with up to 20 g of tritium in torus systems, and for the generation of 14 MeV neutrons within a limit of 2.5 x 10²⁰. This permits the DTE1 series of experiments to be performed. The results of the safety case analysis were presented, concentrating on the methodology of accident sequence identification. This includes elimination of low consequences events; the source terms, dose factors and consequences for key design basis and beyond design basis accidents (such as LOVA, in-vessel LOCA, ex-vessel LOCA, LOFA, etc.); and the assessed probabilities of such events.

O3: New Machines, Ronald Stambaugh (General Atomics)

The papers in this session covered several initiatives for new fusion devices around the world.

The KSTAR Tokamak (presented by J. Kim). This paper described the KSTAR (Korea Superconducting Tokamak Advanced Research) project, the major effort of the Korean National Fusion Program to design, construct, and operate a steady-state capable superconducting tokamak. The project is led by Korea Basic Science Institute and shared by national laboratories, universities, and industry along with international collaboration. It is in the conceptual design phase and aims for first plasma by mid 2002. The major design parameters of KSTAR are major radius 1.8 m, minor radius 0.5 m, toroidal field 3.5 T, plasma current 2 MA, elongation 2.0, triangularity 0.8, and double null poloidal divertor. The toroidal and poloidal coil magnets are superconducting. The device is initially configured for a 20 second pulse and then to be upgraded to 300 second operation with non-inductive current drive. The auxiliary heating and current drive system consists of neutral beams, ICRF, lower hybrid, and ECRF. Deuterium operation is planned with full radiation shielding.