U.S. ITER Forum Contributed Abstracts

US Participation in the ITER Test Blanket Module and Technology Testing Programs

A critical element in the ITER mission since its inception has been testing integrated blanket modules in special ports. Among the principal objectives of the ITER Test Blanket Module (TBM) program are 1) demonstrating the principles of tritium self-sufficiency in practical systems, and 2) providing the technology necessary to install breeding capabilities to supply ITER with the tritium necessary for operation in its extended phase of operation. Adequate tritium supply has been a central issue for the operation of ITER and the development of fusion energy. TBMs will be inserted in ITER from Day 1 of its operation and will provide the first EXPERIMENTAL data on the feasibility of the DT cycle for fusion. A decision on the types of TBMs allowed in ITER is scheduled in 2005. It is critical that the US immediately re-initiate participation in the ITER TBM program to ensure that the US-favored concepts and interests are included and that the US has access to information from much larger blanket programs of our international partners in Europe and Japan. The US has been a leader in the science and engineering of technology testing on ITER and other fusion devices and has many unique capabilities to contribute to the ITER TBM program. This presentation for the ITER Forum and a white paper are being prepared by the community to summarize the many key aspects of the ITER TBM Program and US participation.

U.S. Research Capabilities in Beryllium Manufacturing

Beryllium is to be used as the armor material for ~80% of the plasma facing components for ITER. Manufacture and in-situ or hot cell repair of damaged surfaces can be done using plasma spray technologies. Los Alamos National Laboratory has a dedicated research facility for the manufacture and testing of beryllium articles. Highlights of cooperative research relating to plasma sprayed beryllium mock-ups will be presented. In addition, future research topics will be presented. Disclaimer: The Los Alamos National Laboratory strongly supports the academic freedom and a researcherís right to publish; therefore, the Laboratory as an institution does not endorse the viewpoint of a publication or guarantee its technical correctness.

See White Paper and Presentation.

Superconductor for ITER

The quantity of niobium tin strand required for ITER will be compared to other markets for superconductors, and US industry's strong competitive position in these markets will be documented. Recent advances in niobium tin performance for NMR spectroscopy and high energy physics will be described. Technical issues and alternative routes for producing material meeting target ITER specifications will be outlined. The presentation will show the benefit of development activity and pilot production prior to the start of ITER construction in order to optimize processes and enhance yields.

See U.S. ITER Activity Form and Presentation.

The US ITER Role in Magnet Technology

See Presentation.

Integrated Predictive Modeling of ITER --- Basic Issues

Integrated modeling simulations of ITER are used to predict the expected fusion power production in ITER discharges and to provide the plasma profiles needed for more detailed studies of ELM's, toroidal AlfvÈn modes, sawtooth oscillations, and other macroscopic instabilities. Theory-based models are used for both the core and boundary conditions so that the input variables and assumptions used in the simulations correspond, as much as possible, to parameters that are controlled by experimentalists. The simulation protocol has been tested against experimental data from H-mode discharges in DIII-D and JET. Issues have been identified that need to be resolved. It is found that different models for the pedestal height and core transport, which match current experimental data about equally well on the average, predict different levels of performance for ITER. The dependence of the pedestal height and core transport models on heating power, in particular, needs to be resolved. The height of the pedestal cannot increase indefinitely with heating power, as suggested by recent empirical models for the pedestal. The pressure gradient in the pedestal must eventually be limited by macroscopic instabilities, which are sensitive to current density in the pedestal and the plasma shape. In addition, models are being developed for use in integrated modeling codes to predict the frequency and radial extent of the Edge Localized Modes that occur periodically near the edge of most H-mode plasmas. Integrated predictive modeling simulations will help to maximize the US investment in the ITER project.

US Diagnostics Proposals for ITER

This would include a basic description of ITER diagnostics and how the US can and should contribute. It will also identify the needs, requirements and process by which people can and should get involved. It will illustrate the unique features of fielding diagnostics on ITER, and some of the basic aspects of the environment expected on ITER. The benefits and interests will be detailed as much as possible and how they connect with the physics program.

See white-paper Opportunities for US Diagnostic Participation in ITER and Presentation.

U. S. Background in ITER Fueling Systems and Future Contributions

U. S. BACKGOUND IN ITER FUELING SYSTEMS AND FUTURE CONTRIBUTIONS S. K. Combs(a), L. R. Baylor(a), P. W. Fisher(a), M. J. Gouge(a), P. B. Parks(b), D. A. Rasmussen(a), and G. L. Schmidt(c) (a) Oak Ridge National Laboratory, Oak Ridge, Tennessee (b) General Atomics, San Diego, California (c) Plasma Physics Laboratory, Princeton University, Princeton, New Jersey Fueling systems are essential components for fusion energy research, including standard gas fueling and pellet fueling. Gas fueling systems are relatively straightforward and have always been included in the initial construction and operations of fusion experiments. In contrast, the pellet fueling systems have typically been added to the experiments later to extend the plasma density limit and provide a perturbative tool for particle transport studies. With pellet fueling systems now available, practically every fusion experiment is now equipped with a pellet injector (e. g., DIII-D, Alcator C-Mod, and MST in the U. S. and JET, Tore Supra, JT-60U, LHD, and ASDEX-U abroad). With the larger plasma volume of ITER, pellet fueling will be more important than ever to achieve deeper core fueling and high fueling efficiency, especially for tritium. The US Fusion Science and Technology Program has sponsored a Plasma Fueling Research and Development Program (with participation led by ORNL) for over 25 years; it is generally recognized as the preeminent such program in the world. It carries out R&D in direct support of present US experiments (DIII-D, MST, NSTX, and ET) as well as doing research on advanced concepts for future experiments both nationally and internationally. For example, the US program has provided pellet fueling systems for JET, TFTR, DIII-D, Tore Supra, PLT, PDX, ISX, ATF, MST, and Gamma-10. During the ITER CDA and EDA (1989 ? 1998), the U. S. was responsible for ITER fueling system design and R&D. During that period, some key accomplishments included (1) tritium inject!or design and testing with ITER-size (up to 8 mm) DT and T pe!llets and cumula

See white-paper U. S. BACKGOUND IN ITER FUELING SYSTEMS AND FUTURE CONTRIBUTIONS and Presentation.

U.S Industry Interests in ITER Construction

Forty US industries were surveyed with respect to their areas of interest in ITER construction. Prioritized results will be presented, along with a written report.

See Presentation.

Data acquisition, data management and remote participation for ITER

We outline a vision for data systems and remote collaboration for the ITER project and propose that the U.S. take primary responsibility for defining and implementing software for this purpose. The importance and cost of this device requires that it operate at the highest possible level of scientific productivity. In this sense, it is useful to think of ITER as the largest and most expensive scientific instrument ever built for fusion research. It is our assertion, that for experiments as complex as those carried out in this field, scientific productivity is inextricably linked to the capability and usability of their data and computing systems. Such an effective infrastructure is required both for the success of the entire ITER project and will maximize the value of ITER to the U.S. program as well. To this end a white paper has been prepared, "A Discussion on data acquisition, data management and remote participation for ITER" by researchers at MIT,GA and PPPL. Outlined in this paper is an approach for creating the software infrastructure to satisfy the project's requirements and to maximize the value to the U.S. of ITER participatifon. It also emphasizes that the experience and accomplishments of the U.S. fusion program in these areas and our established links to those conducting relevant computer science research places our community at the forefront to carry out these tasks. We expect that software created for ITER will expand the boundaries of such technology and will likely be applicable to a broad range of scientific disciplines.

See white paper: A DISCUSSION ON DATA ACQUISITION, DATA MANAGEMENT AND REMOTE PARTICIPATION FOR ITER.

See US ITER Activity Form and Presentation.

US Active Spectroscopy Diagnostics for ITER

The purpose of the accompanying White Paper is to review the capabilities of the United States (US) to provide a diagnostic package for ITER, which would include all of the active spectroscopy diagnostics for ITER. These are the diagnostics, which utilize the heating or a diagnostic neutral beam to perform their measurements. This proposal does not include the diagnostic neutral beam for ITER; however, the US certainly has the expertise to take on the beam, as well, if so desired.

The US presently has a leading role worldwide in diagnostic development, detector technology, and the physics for active spectroscopy diagnostics in fusion plasmas. A 20 year US effort in the area of Active Spectroscopy has developed this technology to state-of-the-art detection systems which are currently being employed on US Tokamaks such as DIII-D, Alcator C-Mod, NSTX, and more recently on JET in Europe. The US program has historically been at the forefront of development and implementation of active spectroscopy diagnostics for fusion research. The examples of such active spectroscopy diagnostics include current profile measurements (Motional Stark Effect - MSE), radially resolved measurements of ion temperature, toroidal rotation, Helium ash content, fuel mixture, impurity density measurements, and particle transport coefficients (Charge Exchange Recombination Spectroscopy - CXRS), as well as core plasma density and temperature fluctuations (Beam Emission Spectroscopy - BES). This set of diagnostic measurements would provide a wealth of detailed information to address the many burning plasma physics issues to be addressed by ITER and provide the necessary data to continue the development of our strong US theory and modeling programs.

These diagnostics (MSE, CXRS, and BES) all require a diagnostic beam which is used in conjunction with the measurements, but they also require a very detailed knowledge of the details of the beam and its geometry. Each of the diagnostics also utilize the same types spectrometers, detectors, and similar light collection optics. These common factors for these 3 diagnostics make it an ideal package for US participation. The US currently has a number of unique spectrometer designs with high photon throughput and detectors with quantum efficiencies of > 95%. These spectrometers and detector systems are currently faster and higher detection efficiencies than are currently available in Europe and Japan. Utilizing traditional spectrometers and optics for these active spectroscopy diagnostics, it is estimated that some 70 spectrometers and detectors would be needed. Utilizing several innovations in US spectrometer designs with multiple entrance slits, this number could be reduced by about 50%. These spectrometer innovations are already being implemented by ORNL and PPPL, in a Helium Ash measurement on JET which is currently being planned. Maintaining the absolute calibrations of these diagnostics is a labor intensive effort and reducing the number of active spectrometer systems to be calibrated is a key consideration.
US Participation in this diagnostic package would involve the integration of many institutional contributions. Expertise in developing and operating these active spectroscopy diagnostics is currently being applied on a number of present day tokamak experiments by University of Wisconsin, Princeton Plasma Physics Laboratory, Oak Ridge National Laboratory, General Atomics, and Nova Photonics, Inc. All of these research groups also have joint ongoing international research programs with our European and Japanese ITER partners in these diagnostic areas which should further assure the success of this effort. This diagnostic package also provides many opportunities to strengthen the link between university facilities, national laboratories, and industrial partners.

See white-paper US ITER - Active Spectroscopy Diagnostics and Presentation.

US Involvement in ITER Plasma Control

The United States (US) has a high level of capability in the development of the technology and physics for control of fusion plasmas, in large part resulting from pioneering advanced tokamak research which requires high performance control. Many capabilities present in US laboratories are unique among ITER Parties, and can address critical needs of the ITER project, while complementing the capabilities of the other Partiesí Home Teams and within the Joint Central Team (JCT). The ITER plasma cntrol problem includes unprecedented constraints on accuracy and reliability, as well as novel integration and supervisory issues arising from its highly optimized nature and burning plasma mission. Plasma shape, position, and axisymmetric stability must be regulated with a relative accuracy which is a factor of 10 greater than presently operating tokamaks, in concert with MHD suppression, precise operating point regulation, supervisory action, response to off-normal events, and guaranteed protection of many machine components operating close to performance margins. This highly-coupled, multi-system integrated control problem requires accurate model-based control design and sophisticated simulations in order to provide reliable high performance control. The US Home Team played a major role in the ITER Engineering Design Activity (EDA) plasma control design and analysis effort, solving many of these coupled control problems in close collaboration with the other ITER Parties and the JCT. Since that time, US control design resources have continued to grow through its ongoing efforts in the control-intensive US AT program.

See Presentation.

U.S. Opportunities for Integrated Modeling of ITER --- The Fusion Simulation Project

The Integrated Simulation and Optimization of Fusion Systems (ISOFS) FESAC Subcommittee completed its formal activity in December, 2002, and the ISOFS final report was adopted as a final report. The main recommendation of this report is that a major initiative be undertaken referred to as the Fusion Simulation Project (FSP), the purpose of which will be to make a significant advance within five years towards the ultimate objective of fusion simulation: to predict reliably the behavior of plasma discharges in a toroidal magnetic fusion device on all relevant time and space scales. The long term [15 year] goal is in essence the capability for carying out 'virtual experiments' of a burning magnetically confined plasma, implying predictive capability over many energy-confinement times, faithful representations of the salient physics processes of the plasma, and inclusion of the interactions with the external world (sources, control systems and bounding surfaces). This activity, if it goes forward as planned, represents an opportunity for the US to take the lead in this rich and exciting area. To paraphrase Dr. Raymond Orbach (Director, DOE Office of Science), "Such a capability is absolutely essential for realizing our nation's goal of commercially viable fusion power in a realistic timeframe". We will discuss the status, scope, and proposed organization for this activity.

ITER and edge plasma studies

Edge plasma bridges hot core plasma and material wall and plays crucially important roles in key ingredients affecting ITER design: core plasma confinement and plasma-wall interactions. Both of these ingredients involve very complex and strongly coupled physics/chemistry processes, such as turbulence, strongly intermittent transport, and wall erosion and re-deposition effects. Many of these processes are not understood well enough yet which makes difficult to choose important elements of ITER design. I will advocate for establishing international cross-disciplinary "Edge Group" focused on critical issues of the ITER edge plasma and its effects on ITER design, and where, I believe, the US team can and should play a leading role.

Integrated Predictive Modeling of ITER --- Optimization

The optimization of ITER performance has been examined using theory-based models in predictive integrated modeling simulations. The simulation protocol, which uses a combination of models for the H-mode pedestal and core plasma, has been tested against experimental data from DIII-D and JET. Optimization of the baseline H-mode scenario of ITER has been studied by varying plasma parameters such as the plasma density, auxiliary heating power, and impurity concentration. The sensitivity of the fusion power production to the profile of auxiliary heating power will also be investigated. In addition, advanced tokamak scenarios, which involve the production and control of internal transport barriers, might be used to produce significant improvements in ITER performance over the baseline H-mode scenarios. In the limited modeling studies of internal transport barriers in tokamak discharges that have been carried out so far, it has been found that the height and position of internal transport barriers depend on magnetic shear, flow shear, and finite beta effects. More exotic phenomena, such as current holes, which also produce strong internal transport barriers, need to be studied with theory-based integrated modeling simulations. Optimizing the performance of ITER is an essential element in the design of ITER and in planning the experimental campaign to be carried out.

Proposed simulations addressing important physics issues of ITER operation

During the ITER-EDA, the Georgia Tech Fusion Research Center performed simulations to evaluate ITER operating scenarios with a radiative mantle. Our contributions helped to identify the impurity-seeded radiative mantle as a significant component of the ITER power exhaust solution. As the US rejoins ITER, we believe that our modeling capabilities and computational tools will be useful in the performance and operational analysis of ITER and in the planning of the ITER operational and experimental programs. Our collaboration with DIII-D during the past 5 years gave us the opportunity to benchmark our computational tools by comparison with DIII-D data. We are in a strong position to carry out simulations in the areas of multi-species, multi-charge state impurity transport, toroidal plasma rotation, neutral transport in the divertor and core regions and density limit analysis. All these areas are very important in the analysis of the ITER operating space and its performance and in scenario modeling.

Prospects and Design Requirements for Advanced Tokamak Operation of ITER

The period 1998-2003 has been an exciting in terms of discovering and documenting a variety of promising Advanced Tokamak modes and their control. The question then becomes: Can the present ITER design exploit and investigate these new possibilties ? This contribution briefly describes potential investigations of AT operation of ITER FEAT and identifies the upgrades needed to carry them out. Of particular importance are the upgrades and access that must be built into the machine during its construction so as not to preclude subsequent investigations. Among the items discussed are: Negative ion neutral beams for heating and current drive, positive ion neutral beams for rotation control, pancake solenoid, ECH for heating and/or current drive and ITB initiation, lower hybrid, fast waves for electron heating, minority ICRF, ITBs, and current drive, actuator coils for resistive wall mode feedback and dynamic error field control, TAE mode antennas to measure frequency and damping, inside pellet launch, long divertor legs, cold-trap for unburned tritium recovery, superconducting ac loss limitations, first wall heat flux limit that permits exploitation of high beta, high fusion power modes. Overall, the US Advanced Tokamak investigations should develop an integrated plan that assures promising AT operations are not precluded while fulfilling the overarching goal.

See Presentation.

Safety, Environmental and In-vessel Tritium Opportunities Afforded by ITER

The design and construction of ITER will bring with it a host of safety and environmental issues, many of which are relevant to reactor scale fusion facilities. Furthermore, any environmental and safety regulatory decisions made by the host country of ITER will set a precedent world-wide for the construction and operation of any future fusion machine. The US has a unique opportunity to help resolve some of those issues in ways that are benificial to the US fusion program. In the paper and presentation, the key safety and environmental issues that ITEr will face will be outined and those that are of greatest concern to the US will be discussed. This will form a recommended set of safety and environmental activities that the US should be involved with during ITER negotiations, construction and operation.

Overview of Recent Activities of the ITPA Topical group on Steady State Operation and Energetic Particles

Since the formation of the ITPA in 2001, there have been two meetings thus far of the international working group on Steady State Operation and Energetic Particles. This international working group is chaired by Dr. Claude Gormezano, Euratom / C.R. Frascatti. In this presentation I will briefly summarize the results of the first two meetings and will discuss the near term and longer range plans for the group.

See Presentation.

ITER Information Plant (ITER IP)

SAIC and GA propose to build a centralized multi-component information system for ITER. Such a system would make it possible to reap maximum benefits from the scientific and technological achievements of the project. We refer to this system as ITER Information Plant (ITER IP). The ITER IP will bring the greatest return to the U.S. for a relatively modest investment. The party to win the ITER Information System bidding will get an exceptional opportunity to access and maintain the knowledge accumulated in all ITER components. SAIC and GA propose to join expertise in information technology and fusion, thus bringing complementary strengths to this complex task. We are submitting a White Paper discussing details of the ITER Information Plant.

See Presentation and White Paper.

Neutronics and Nuclear Analysis Requirements for the ITER Final Design

ITER is still undergoing major design changes that require detailed nuclear analysis to determine the effects of radiation on safety and component performance. Work carried out during the EDA and beyond addressed a numerous problems some of which were unique to particular phase of the reactor design. As the ITER progresses toward and construction, neutronics and nuclear analyses will remain an essential part of the design process. Calculations to support the design and placement of diagnostics in the major ports are far from complete and analyses are necessary to estimate radiation damage to diagnostic system components such as mirrors, fiber optics, critical electronics, etc. Radiation leakage through holes and other penetrations in these systems must be fully assessed to establish activation levels in and near diagnostic equipment where frequent access will be necessary. Analyses are also required as the reactor building design progresses to determine operation and shutdown radiation levels for establishing personnel access and maintenance criteria. Experience shows that neutronics and radiation environment assessments continue through final design and construction phases of nuclear facilities (e.g., TFTR, JET, SNS, etc.). It is essential to institute a strong nuclear analysis capability early in the final design and early construction of a nuclear facility. Considerable money savings can be realized as the result of early investigations of radiation and activation levels that can be easily remedied early rather than by retro fitting and post construction re-design.

Potential US contributions to serious plasma material interaction issues in ITER.

ITER's ability to address burning plasma physics depends on the successful resolution of serious plasma material interaction (PMI) issues. The change in ITER pulse length and duty cycle is larger than the change in any plasma parameter and poses severe challenges in tritium management and material erosion. Predictions of tritium retention are uncertain due to lack of data, uncertain effects of mixed materials, and lack of code validation in detached plasmas. For ITER with carbon plasma facing materials it is clear that fast and efficient tritium removal will be necessary, however no relevant method to remove tritium has ever been tested on a tokamak. Diagnostic mirrors will become coated with deposits and cleaning methods will be essential but are unproven. Compliance with regulatory dust inventories will be required but relevant dust diagnostics have not been demonstrated. These issues fall in-between the physics and technology communities and have hitherto suffered from lack of support. The US can contribute to PMI areas, however a new organizational initiative is needed to help fund, focus and prioritize R&D (along the lines of the EFDA Task Force on Plasma-Wall Interaction). The ITER physics program will be 'at risk' until PMI solutions have been demonstrated on existing tokamaks.

Active MHD Spectroscopy Diagnostics for ITER

One area of interest to the U.S. for ITER is to participate in the design, construction, and operation of Active MHD Spectroscopy diagnostics. Such diagnostics operate successfully on JET and Alcator C-Mod to actively determine the properties of Toroidal Alfven Eigenmodes, Neoclassical Tearing Modes, and other instabilities. The signals can then be used in a feedback control algorithm to maintain high performance operation just below the point where the instabilities become dangerous. The time evolution of the radial distribution of fast ion fusion products can also be measured with this diagnostic. Through Alfven cascades, information on the q profile evolution and its relation to Internal Transport Barriers can also be determined. While the technology is relatively straightforward so that little R&D is required to develop this diagnostic, the physics information to be obtained from participation in this diagnostic will lead to a fruitful research program throughout the operational phase. This work continues a long standing collaboration with JET and the CRPP-Lausanne in this area and efforts will be made to divide up the responsibilities for this diagnostic between the U.S. and European parties.

See Presentation.

US involvement in ITER Design Integration and Assembly Activities

During the CDA and EDA phases of ITER, the US team had an active involvement in design integration and assembly activities, and should pursue a proportional role in these activities again. Design integration includes coordination of physical and functional interfaces, global analyses, preparation and dissemination of specifications and CAD models, etc. This coordination is necessary on all projects, but is even more critical on a multi-institution, multi-discipline, multi-national project like ITER. Assembly activities are normally a subset of design integration, since assembly involves all the physical interfaces among components and systems. Design integration roles in which the US should consider re-involvement include global neutronics analyses, global electromagnetic analyses (e.g. modeling and effects of disruption and seismic events, updating the ITER Structural Design Criteria and advancing the adoption of a Fusion design code (e.g. ASME Section III, Division 4), updating the ITER Materials Database, and translation and dissemination of CAD models. Assembly roles that the US should consider may include robotic welding of the vacuum vessel sector-to-sector and port extension field joints, initial blanket installation, and associated in-vessel metrology. A brief description of the various design integration and assembly opportunities as well as benefits to ITER and the US fusion program will be discussed.

See Presentation.

US Contribution to ITER Remote Handling Technology

A WHITE PAPER IS BEING SUBMITTED WITH THE SAME TITLE. U.S. Contribution to ITER Remote Handling Technology T.W. Burgess, M.M. Menon, P.T. Spampinato During the CDA and EDA phases of ITER, the U.S. contribution for remote handling technology consisted of: a) providing personnel assignments to the Joint Central Team with responsibility for remote handling R&D and design, b) developing a laser based in-vessel metrology system, c) developing and testing vacuum vessel cutting and welding equipment, d) developing the ITER Remote Handling Design Guide and Radiation Hardness Design Manual, e) testing radiation hardened electrical connectors for remote handling, and f) design of the port assembly handling and transport systems. Since that time, development of the metrology system continued at ORNL and precision range measurements were performed in TFTR and elaborate mapping of plasma facing surfaces was done in NSTX. An SBIR grant is supporting development of the next-generation reactor-relevant metrology system. In addition, ORNL has been responsible for developing all remote handling systems and remote maintenance designs for FIRE. The experience gained from ITER has been applied to projects in the US that are presently under construction. Since 1998, the Remote Systems Group at ORNL has played a major role in the development and testing of maintenance equipment for the Spallation Neutron Source (SNS). This support includes a) remote handling systems design and procurement, b) target process systems design and procurement, and c) developmental testing of critical components and operations using small-scale mockups to test tools and full size mockups to demonstrate equipment performance and procedures. Remote Systems also provides R&D and design support to the SNS component designers to ensure compatibility with the remote handling systems, and has developed procurement specifications for robotic manipulators, shield windows, and viewing s!ystems, to name a few. In addition, Remote Systems is respons!ible for develop

See white-paper U.S. Contribution to ITER Remote Handling Technology and Presentation.

See Activity Form.

Development of soft X-ray to VUV diagnostics for ITER

D. Stutman
John Hopkins University

The Johns Hopkins Plasma Spectroscopy Group proposes to investigate new ideas and tools for soft X-ray to VUV diagnostic on ITER, focusing on the following:
i) Development and testing of new ideas for the extraction and detection of soft Xray to VUV light in the burning plasma environment
ii) Design and prototyping of robust and easily replaceable soft X-ray to VUV spectroscopic and imaging devices
iii) Development of new atomic data and modeling tools for the analysis of the measurements obtained with these devices

See white paper: Development of soft X-ray to VUV diagnostics for ITER.

See Presentation.

See US ITER Activity Form.

WHY THE US SHOULD BUILD THE ITER ION CYCLOTRON SYSTEM

Ion cyclotron (IC) heating and current drive systems are ubiquitous in fusion energy research. They are installed on the three major US tokamaks as well as a number of university-level experimental facilities, and are planned for proposed near-term US experiments (NCSX, QPS,Ö). IC systems are in use on most present international fusion experiments (e.g., JET, Tore Supra, JT-60U, W7-AS, W7-X, ASDEX-U). Ion cyclotron systems are a major component of the heating and current drive system designs for all the burning-plasma experiments (ITER, FIRE, and IGNITOR) studied at the Snowmass meeting in July 2002. The US Office of Fusion Energy Science has sponsored an IC Research and Development Program (with participation by GA, MIT, ORNL, PPPL and others) for many years. It carries out R&D in direct support of present US experiments (NSTX, C-Mod, DIII-D, ET) as well as doing research on advanced concepts for future experiments both nationally and internationally; for example, the US program is building and testing a high-power prototype of an advanced, ìITER-likeî antenna in collaboration with the JET program. During the ITER CDA and EDA (1989 ñ 1998) the US was a major contributor to the design and R&D for the ITER IC system. The ITER ion cyclotron system offers new challenges. The antenna must operate in a nuclear environment and withstand heat loads and disruption forces beyond present-day designs. It must operate for long pulse lengths and be highly reliable, delivering nearly full power to a plasma with properties that will change during a pulse. A development and testing program will be required to validate the proposed ITER antenna design, and to modify it if needed. The US should take the lead in the design, R&D, and construction of the ITER ion cyclotron system, particularly the antenna. The US lead in this program will maintain the present capability of cutting-edge research, and will further the cause of understanding the power l!imits in antennas and in developing improved IC systems. It i!s likely to lead

See white-paper US ITER Ion Cyclotron System White Paper and Presentation.

US Contributions to the ITER ECH System

The ITER FEAT ECH system consist of twenty-four 170 GHz and three 120 GHz megawatt gyrotrons with controls; power supplies; a transmission line system; an equatorial launcher and three upper port launchers. This system will be used for heating, current drive and NTM mode stabilization. The US ECH technology community includes CPI, General Atomics, MIT, Princeton PPPL, Univ. MD, Univ. WI, Calabazas Creek Research and Diversified Technologies, Inc. The US team has world leadership capability in many areas of ECH technology, including the most advanced concepts for transmission lines and steerable launchers and the only continuous (CW) gyrotron operating experience. The US could supply the entire ITER ECH system, or the US could work in partnership with the other ITER parties. This presentation will discuss US ECH capabilities and possible scenarios for supplying all or some of the ITER ECH system. The US has a major industrial capability in ECH technology. US participation in the ITER ECH system is considered critical in maintaining a strong US capability in the ECH area. This capability leads to a greater ability to carry out ECH heating and plasma control experiments within the domestic US fusion program, on both large and small machines. It also supports the US community by providing multiple, valuable spinoffs in the form of high frequency microwave systems for industrial heating, radar, communications and scientific applications. A White Paper will be submitted.

See white paper: US ITER Electron Cyclotron System White Paper .

See US ITER Activity Form.

US Participation in ITER Plasma Facing Components

Finding high performance plasma facing components (PFCs)with long life is essential for a burning plasma experiment and for future devices on the path to fusion energy. The realization of affordable and practical fusion energy relies on suitable PFCs. Understanding plasma materials interactions continues to be a fruitful research area. US participation in supporting research and fabrication of the ITER divertor components will provide a rich array of plasma, materials, engineering, and atomic physics research opportunities and stimulate the development of technology that is essential for the fusion energy development path. Participation in construction of the divertor will reinvigorate the moribund fusion industrial talent necessary for a healthy fusion program. We will briefly discuss the US role in the ITER divertor during the Engineering Design Activity. Possible roles for the US in R&D leading up to divertor construction and during divertor construction will be presented. The benefits to the US fusion effort will be enumerated. The need for an active PFC program in a healthy fusion research program will be discussed. A white paper will be prepared.

US Background in ITER Tritium Plant and Potential for Further Contributions

The US should consider expressing considerable interest in contributing to the upcoming ITER Tritium Plant work. During the last US involvement in ITER, the US was a major contributor to the ITER tritium plant R&D and design effort, taking a number of lead and supporting roles. US US participation was viewed as being particularly important due to the TSTA (Tritium Systems Test Assembly) experience which was the highest-throughput, continuous, integrated fusion fuel processing system in the world. This experience remains very unique to this today. ITER will be a substantial extension beyond the TSTA knowledge-base (10x T throughput, 10x T inventory, 0.1x T cycle time). This will be a significant challenge, and the US should have play a key role in the Tritium Plant construction. The US strength in this area continues to be recognized at various international meetings. Constructing a Tritium Plant that will operate safely is of paramount importance for ITER's (and fusion's) overall success. Other programs have stumbled because of tritium. US fusion fuel processing has had an outstanding safety record, and bringing this practical experience to ITER is crucial. Doing the Tritium Plant right is also of imperitive since tritium is a very limited resource. The present embodiment of ITER will consume half of world's non-weapons tritium. Generating more tritium borders on being prohibitively expensive. For ITER about 5% of the world supply will be in-process at a given time, and it will be critical that it is handled wisely. Tritium will be needed not only for ITER, but also for other parts of the US 35-year fusion development plan including Demo, IFE (NIF and other experiments) and materials testing (CTF). Thus, experience gained by working on the ITER tritium plant will have broad benefits to the US fusion program. All these factors combine to make a compelling case that the US should participate in a substantial way in the IT!ER Tritium Plant construction.

Application of MeV symmetric neutralized ion beams to magnetically confined fusion plasma

Alfred Wong
UCLA

The use of symmetric neutralized ion beams of MeV energies for heating, current drive, diagnosis, and phase space control of magnetically confined fusion plasmas is discussed. Symmetric neutralized ion beams are composed of positive and negative ions of equal charge-to-mass ratios (e.g. D+ and D-). Given sufficient density, they are capable of propagating across a transverse magnetic field and can therefore accomplish many of the tasks of a conventional neutral beam. RFQ accelerators driven at GHz frequencies yield compact beam modules (~1 m length) with beam energy > 1 MeV and beam power density > 10 MW / cm^2. The compact size allows beam modules to be placed in a variety of locations and injection angles, including on the inner wall of a toroidal vessel. 2-D and 3-D PIC simulation of the beam propagation is ongoing, as are experiments on beam production.

See US ITER Activity Form, and White Paper.

Intense Diagnostic Neutral Beam for ITER

LANL has proposed that R&D for an intense diagnostic neutral beam, to enable use of neutral beam-based diagnostics (such as CHERS, MSE, Helium ash measurements, and possibly BES) be performed as soon as possible. In the previous decade, substantial OFES diagnostic resources were invested in feasibility experiments at LANL, but R&D funding stopped in the early 1990's. The goal of this proposal is to re-build and characterize a rep-rated short pulse ion source, with the intent of testing it in a neutral beam line in ~4 years on a medium-scale tokamak, prior to eventual implementation on ITER. The problem in a large, hot machine such as ITER, is that existing 100keV/amu neutral beams are so severely attenuated by the time they reach the plasma core, that they no longer provide sufficient signal/noise ratio for the "standard" neutral beam-based optical diagnostics. In addition to the attenuation problem, the background interfering bremsstrahlung radiation is also muchhigher in dense, or large hot plasmas. Our proposed positive ion beam source is ~10kA in a 1 microsecond pulse, at 30 Hz. This is to be compared to the present baseline ITER Diagnostic Neutral Beam which is rated at 20 A and nearly CW operation. A white paper will be submitted.

See white-paper Intense Diagnostic Neutral Beam for ITER.

See ITER Activity Form.

New approaches for ITER operational regime, its stability control and fueling.

L.E.Zakharov, S.I.Krasheninnikov, P.Parks, R.Majeski The peaked temperature profile, unavoidable in fusion powered plasmas with relatively low edge temperatures, is the souorce of long standing uncertainties in predicting the stability of the plasma core. At least, two kinds of instabilities, i.e., ballooning modes and m=1 internal disruptions, are very sensitive to the value of (q0-1) in the center. With a peaked temperature profiles q0 drops below 1 and makes the stability of the plasma core unpredictable. Moreover, the recent discovery of the possibility of a helical, island held equilibrium suggests a long duration phase of the discharge with q0 < 1, which will finally be terminated by a disruption. At present, the only feasible way of eliminating the temperature peaking is creation of low recycling conditions at the plasma edge (e.g., by insertion of Li coated panels between ITER plasma and the wall.Also, in order to unload the divertor plates from the high-temperature impact, the active blob formation by a localized gas "fueling" from the low field side is suggested. Because of the "unfavorable" curvature effects, such blobs, while cooling the SOL,cannot go inside the high-temperature plasma and, in fact, will transport the plasma energy to the LiWall panels. The high edge temperature is the most favorable for plasma fueling by Diamagnetic "Hot Dogs" (DHD), which can be launched from the high field side of ITER. The DHD fueling mechanism does not rely on the speed of gas jet, which is converted at the plasma boundary into the fully ionized, over pressured diamagnetic object, which expands along the field line with the sound speed and is accelerated to a significant fraction of the sound speed along the major radius toward the plasma center.DHD, which requires the low recycling plasma, is potentially the most effective way of controlling pwoer deposition and the density profile in the reactor relevant plasmas.