International Thermonuclear Experimental Reactor Design Research Articles

Analytic expressions of neoclassical transport coefficients for fusion-born alpha particles in a tokamak reactor

Analytic approximate expressions of neoclassical transport coefficients for energetic alpha particles are presented. The energy scattering effect, as well as the usual slowing-down and pitch angle scattering effects, is retained. The formulas are valid for arbitrary aspect ratios and agree well with the previous results, which numerically evaluated the coefficients using an eigenfunction expansion technique. Application of the newly developed formulas to the International Thermonuclear Experimental Reactor Design Activity (ITER EDA) [J. C. Wesley and ITER Joint Central Team, Phys. Plasmas 4, 2642 (1997)] plasma condition is also given.

Experimental investigation of AC losses and effective time constant in the high current cable

The report is dedicated to the results of AC-loss measurements in the ITER (International Thermonuclear Experimental Reactor) Nb 3Sn strand and ITER design subsize cable samples. The test facility described below includes the set of superconducting magnet systems furnished with boil-off calorimetric inserts and calorimetric schemes including gas-flow meters. The calorimetric method and the magnetization loop recording method were used simultaneously for the AC-loss measurements. The coincidence between the results of both measuring techniques was confirmed. The original calorimetric scheme makes it possible to measure the released energy during one pulse of the magnet field. The results of AC loss measurements in ITER subsize conductor samples manufactured in Russia are given as an example.

Extruder system for high-throughput/steady-state hydrogen ice supply and application for pellet fueling of reactor-scale fusion experiments

Pellet injection systems for the next-generation fusion devices, such as the proposed International Thermonuclear Experimental Reactor (ITER), will require feed systems capable of providing a continuous supply of hydrogen ice at high throughputs. A straightforward concept in which multiple extruder units operate in tandem has been under development at the Oak Ridge National Laboratory. A prototype with three large-volume extruder units has been fabricated and tested in the laboratory. In experiments, it was found that each extruder could provide volumetric ice flow rates of up to ∼1.3 cm3/s (for ∼10 s), which is sufficient for fueling fusion reactors at the gigawatt power level. With the three extruders of the prototype operating in sequence, a steady rate of ∼0.33 cm3/s was maintained for a duration of 1 h. Even steady-state rates approaching the full ITER design value (∼1 cm3/s) may be feasible with the prototype. However, additional extruder units (1–3) would facilitate operations at the higher throughputs and reduce the duty cycle of each unit. The prototype can easily accommodate steady-state pellet fueling of present large tokamaks or other near-term plasma experiments.

ITER breeding blanket design for the enhanced performance phase

The International Thermonuclear Experimental Reactor (ITER) breeding blanket is designed to breed the necessary tritium for ITER operation during the enhanced performance phase by replacing the shielding blanket of the basic performance phase. Similar to the shielding blanket, it has to remove the majority of the fusion power generated by the plasma and to protect the vacuum vessel and the toroidal field coils from excessive nuclear heating and radiation damage. It has to produce a net tritium breeding ratio of more than 0.8 to satisfy the technical objectives of the enhanced performance phase. For compatibility with the ITER design and to satisfy the blanket functional requirements, a water-cooled modular solid breeder blanket with a beryllium neutron multiplier has been selected. Lithium zirconate is the reference breeder material based on the current database. Enriched lithium is used to enhance the tritium breeding capability, to reduce the radial blanket thickness, to decrease the breeder material volume, to lower the breeder thermal stresses, and to enhance the shielding capability. Similar to the shielding blanket, the breeding blanket uses Type 316LN-IG austenitic steel structural material. Both forms of beryllium material, porous and pebbles, are used at different blanket locations based on design requirements. This paper is concerned with the design analyses and design selections, including beryllium form, breeder material, tritium breeding, and heat transfer across the beryllium–steel interface. Also, the required research and development tasks for the ITER breeding blanket are summarized.

Advanced solutions for beryllium and tungsten plasma-facing components

Beryllium and tungsten are candidate plasma-facing armour materials for the International Thermonuclear Experimental Reactor (ITER). These armours are proposed for areas with low heat flux (⩽5 MW m −2); however, in the divertor, surface melting during abnormal events may occur. This paper reports the progress made in developing novel approaches to solving the difficulties posed in designing with these armours. A Be monoblock brazed to an oxygen free high conductivity (OFHC) 10 mm ID Cu tube using InCuSil `ABA' braze alloy has survived 130 cycles of 10–11 MW m −2 for 6 s, with surface temperatures of 1250°C. No visible surface cracking occurred. The same monoblock was then exposed to several cycles of 20–22 MW m −2 for 8 s, creating a 2 mm deep molten layer. High cycle fatigue was then performed. The test results are detailed in this paper. Comparison between experimental and theoretical results are made. W and Cu have a large mismatch in their thermal expansion coefficients and two designs are proposed that minimise the interface stresses. These are: a `brush'-like structure with rectangular fibres set in a Cu substrate using the `active metal casting' (AMC) technique; and thin monoblocks (or lamellae) brazed or active metal cast onto a Cu tube. Analyses of the lamellae concept for steady-state heat loads of 5 MW m −2 are presented. Fatigue analyses show that both solutions are theoretically viable (∼10 4 cycles). A `brush' mock-up has been manufactured and progress on its testing is reported. Results of all tests and their relevance to the ITER design are discussed.

Effect of seismic isolation on the Tokamak in ITER

The support structure of the International Thermonuclear Experimental Reactor (ITER) is suggested to have appropriate stiffness that accommodates both thermal distortion of the system and a maximum ground acceleration of 0.2 G (1 G=9.8 m s −1) due to earthquakes. For a site with earthquakes more severe than 0.2 G, the seismic isolation design is a possible candidate for keeping the seismic input low enough. The present design of ITER assumes that only the Tokamak pit portion is isolated, due to the whole building being very large and complex. In this study, dynamic analyses of the whole Tokamak building with a base-isolated Tokamak pit were carried out, and the effect of the isolation of the pit structure was evaluated. As a possible input level, a 0.4 G maximum ground acceleration was assumed. Particular attention was paid to the relative displacement between isolated and non-isolated portions, due to the smaller relative displacement being desirable for the design of pipes across the isolation gap.

Investment in Tokamak Fusion

The debate over the future of the International Thermonuclear Experimental Reactor (ITER) (J. Glanz, “Requiem for a heavyweight at meeting on fusion reactors,” News & Comment, 8 May, p. [818][1]), unfortunately, does not go far enough. The real issue is not how much money should be invested in the next large tokamak, but whether any further investment in tokamak confinement is warranted at this time. The tokamak has been the main approach to magnetic confinement fusion since its inception almost 50 years ago. During the intervening half century, great progress has been made in understanding the physics of toroidal confinement and in translating that understanding into improvements and innovations in tokamak design. Although tokamak design is still based on empirical scaling laws, confidence in these laws has been strengthened by a wealth of experimental data. Numerous reviews of the ITER design have concluded that if the machine is built to the ITER design specifications, there is little doubt that it can achieve its scientific goal of a sustained thermonuclear burn. This statement reflects both the triumph and the tragedy of fusion research, because it also implies that if a tokamak is significantly smaller than the ITER design, it will not achieve a sustained thermonuclear burn and thus will not provide the basis for a power-producing reactor. The scientific community needs to reexamine the premise on which the public was originally asked to support fusion research, namely, that it would lead to the development of a practical, power-producing technology. In light of today's knowledge, it is highly unlikely that further development of the tokamak will lead to that outcome. [1]: /lookup/doi/10.1126/science.280.5365.818

Partners Will Rethink Fusion Project

ITER T he search for cheaper versions of a massive fusion project will begin next month, when partners in the $10 billion International Thermonuclear Experimental Reactor (ITER) meet to discuss the project's fate. The impending decision to consider cost-saving alternatives to the current design, revealed last week by U.S. fusion officials and confirmed by ITER director Robert Aymar, marks a major shift in direction for the troubled program. The proposed doughnut-shaped reactor is designed to contain a self-sustaining thermonuclear burn that could lead to advances in plasma science and eventually fusion power plants. But budget constraints among the four partners—Europe, Japan, the United States, and Russia—recently forced postponement of construction by 3 years. That delay has sparked widespread concern about the project's future, particularly among U.S. researchers ( Science , 2 January, [p. 20][1]). U.S. Department of Energy (DOE) officials said on 22 January that the partners, after weeks of intense discussions, have informally agreed to set up a panel to examine less expensive designs while keeping the partnership intact. The decision is expected to be ratified in mid-February in San Diego. “It's a done deal,” says Anne Davies, DOE fusion chief. The partners plan to consider “broader options” and appoint a panel to examine how to reduce costs, Davies told a DOE advisory committee last week. “It's a very small number of words, but a crucial change,” she said about the pending agreement. Aymar confirmed the plan to examine cost-saving measures. He said one way to save money would be to abandon some technological advances, including those favored by the United States. “The only way to reduce ITER's design cost is by cutting some of the objectives,” he added. He estimated, however, that the changes would trim the price by no more than $1 billion or $2 billion. European officials declined to comment. “There is no official position yet, though I've heard rumors,” says Regis Saison, a spokesperson for the European fusion program based in Brussels. But he warns that cutting ITER's costs will inevitably mean creating a less capable machine. “Then the question is whether it could still be called ITER,” he adds. Japanese officials are equally circumspect. “It's been assumed that a discussion of streamlining would probably come up at the ITER meeting in February in San Diego,” says Masaharu Shiozaki, deputy director of the Science and Technology Agency's Office of Fusion Energy. “But at this point, I don't think we know what kind of agreement might result.” However, U.S. officials aren't waiting. Charles Baker, the U.S. home team leader and an engineer at the University of California, San Diego, will begin gathering input from the U.S. fusion community in late February and hopes to finish by July. The timing, Davies says, “is going to be tight, and [the review] is going to be intense.” [1]: /lookup/doi/10.1126/science.279.5347.20

Enhanced probabilistic decision analysis for radiological confinement barriers of the International Thermonuclear Experimental Reactor

To ensure a defence-in-depth approach, several radiological confinement barriers surrounding a tokamak plant can be employed. A methodology using probabilistic risk assessment (PRA) techniques is a useful tool for evaluating the performance of each confinement barrier within the context of a limited allowable risk of accidental radioactivity releases. Such a methodology was developed and applied to the confinement strategy for the International Thermonuclear Experimental Reactor (ITER). Accident sequence models were constructed for each of the confinement barriers to evaluate the probabilities of events leading to radioactive releases from the corresponding confinement barrier. The current ITER design requirements set radioactive release and dose limits for individual event sequences grouped in categories by frequency. To limit the plant's overall risk and account for event uncertainties in both frequency and consequence, an analytical form for a limit line is derived here as a complementary cumulative frequency (CCF) of radioactive releases to the environment. By comparing the releases from each confinement barrier against the limit line, a decision can be made about the number of barriers required to comply with the design requirements. The first barrier is the vacuum vessel (VV) and the primary heat transfer systems. The second confinement barrier consists of the cryostat vessel (CV) and the heat transfer system vaults. In case the outer building is needed to act as a third barrier for ITER, a decision model using the multi-attribute utility theory was constructed to help the designer choose the best type of tokamak building. The decision model allows for performing sensitivity analysis on relevant parameters and for design features of new options for the ITER tokamak building.

A probabilistic methodology for the design of radiological confinement of tokamak reactors

A methodology using probabilistic risk assessment techniques is proposed for evaluating the design of multiple confinement barriers for a fusion plant within the context of a limited allowable risk. The methodology was applied to the reference design of the International Thermonuclear Experimental Reactor (ITER). Accident sequence models were developed to determine the probability of radioactive releases from each confinement barrier. The current ITER design requirements, that set environmental radioactive release limits for individual event sequences grouped in categories by frequency, is extended to derive a limit on the plant overall risk. This avoids detailed accounting for event uncertainties in both frequency and consequence. Thus, an analytical form for a limit line is derived as a complementary cumulative frequency of permissible radioactive releases to the environment. The line can be derived using risk aversion of the designer's own choice. By comparing the releases from each confinement barrier against this limit line, a decision can be made about the number of barriers required to comply with the design requirements. A decision model using multi-attribute utility function theory was constructed to help the designer in choosing the type of the tokamak building while considering preferences for attributes such as construction cost, project completion time, technical feasibility and public attitude. Sensitivity analysis on some of the relevant parameters in the model was performed.

Safety Analysis of ITER EDA Design by GEMSAFE

General Methodology of Safety Analysis and Evaluation for Fusion Systems (GEMSAFE) was applied to the International Thermonuclear Experimental Reactor (ITER) design in the stage of Engineering Design Activities (EDA) to identify Design Basis Events (DBEs) and the related safety features, which were compared with those of the ITER design in the stage of Conceptual Design Activities (CDA). As a result, 18 DBEs for the EDA design were selected in comparison with 25 DBEs for the CDA design. DBEs related to the fuel area were categorized in higher event category than those of the CDA design due to the increase of the mobile tritium contained in some components. It was necessary to reduce the inventory of the tritium absorbed in the tokamak dust in the EDA design as well as in the CDA design. Some measures were recommended to reduce mobile tritium dissolved in the coolant in the single cooling loop due to the increase of this estimated inventory.

The Conversion of a Room Temperature NAK Loop to a High Temperature MHD Facility for LI/V Blanket Testing

The Vanadium/Lithium system has been the recent focus of ANL`s Blanket Technology Pro-ram, and for the last several years, ANL`s Liquid Metal Blanket activities have been carried out in direct support of the ITER (International Thermonuclear Experimental Reactor) breeding blanket task area. A key feasibility issue for the ITER Vanadium/Lithium breeding blanket is the Near the development of insulator coatings. Design calculations, Hua and Gohar, show that an electrically insulating layer is necessary to maintain an acceptably low magneto-hydrodynamic (MHD) pressure drop in the current ITER design. Consequently, the decision was made to convert Argonne`s Liquid Metal EXperiment (ALEX) from a 200{degrees}C NaK facility to a 350{degrees}C lithium facility. The upgraded facility was designed to produce MHD pressure drop data, test section voltage distributions, and heat transfer data for mid-scale test sections and blanket mockups at Hartmann numbers (M) and interaction parameters (N) in the range of 10{sup 3} to 10{sup 5} in lithium at 350{degrees}C. Following completion of the upgrade work, a short performance test was conducted, followed by two longer multiple-hour, MHD tests, all at 230{degrees}C. The modified ALEX facility performed up to expectations in the testing. MHD pressure drop and test section voltage distributions were collected at Hartmann numbers of 1000.

The ITER Project

Letters from: [ Paul H. Rutherford ][1] [ Charles C. Baker ][1] Thomas H. Stix et al . take issue with the present course of the International Thermonuclear Experimental Reactor (ITER) initiative (Letters, 16 Feb., p. 891). ITER is a fusion research facility based on the highly successful “tokamak” approach to plasma confinement, now being designed under an international agreement executed in 1992 by the European Union, Japan, the Russian Federation, and the United States. ITER seeks to achieve, for the first time ever in controlled fusion research, a self-heated “burning” deuterium-tritium plasma. Plasma burn has been widely endorsed as a critically important and appropriate next step for fusion. Stix et al . question combining this step with three other “giant steps”: large size, long pulse, and superconducting magnets. The size of the ITER plasma is dictated by the requirement to provide sufficient energy confinement to reach the burning state; this is determined from size-scaling relationships derived from experiments on tokamaks with plasma volumes spanning two orders of magnitude; the largest of these tokamaks already operate at plasma densities and temperatures similar to those expected in ITER. Confinement in a tokamak is determined primarily by the magnitude of the plasma current: ITER's plasma current is about four times larger than that in the Joint European Torus. The step in pulse length is derived from the additional physics requirement to sustain the burn over times long compared with those characteristic of the evolution of plasma profiles. The step from copper to superconducting magnets has been found to be the most practical way of achieving the required pulse lengths. Superconducting magnets have already been used in tokamaks, including one of the size of Princeton's successful Tokamak Fusion Test Reactor. The long pulse and the superconducting magnets are also essential for ITER to advance fusion engineering objectives. The engineering features of ITER are generic to any fusion reactor; in ITER, they can be tested together in an integrated facility, made possible only by international partnership. Stix et al . also question the capability of the ITER design to achieve its stated technical objectives. Recognizing the need for broad technical consensus, a Technical Advisory Committee (TAC) was formed which, on a continuing basis, reports to the ruling ITER Council on the adequacy of the ITER design and its physics basis. The TAC's 16 members are drawn equally from all four partners, but they serve as individuals, independent of both the design team and their respective governmental agencies. Last July, the TAC completed a review of the “Interim Design” and, with unanimity, concluded that the design can fully meet ITER's mission and technical objectives. # {#article-title-2} The letters from Stix et al ., Ernesto Mazzucato, and William E. Parkins (16 Feb., p. 891) have two common themes, namely, that we should postpone or even cancel the ITER project and that we do not know how to make an attractive fusion reactor. Stix et al . and Mazzucato correctly note the tremendous progress that has been made in fusion research and that we are in a period where our expectations of continued progress are high. This is used as an argument to suggest that a commitment to the construction of ITER be postponed. I would argue just the opposite. We need to build on our excellent progress and take advantage of the unique opportunity offered by participation in ITER to move forward with fusion science and energy development. Projects like ITER are always faced with the idealistic argument that we should wait for better results. Such logic would have postponed most major undertakings that have lead to significant scientific and technological progress. The letters raise valid technical issues (including improved plasma confinement, control of disruptions, and engineering and materials development). These issues have been clearly recognized by fusion researchers, in general, and ITER, in particular. They are being responsibly addressed in the ITER research and development and base program activities. [1]: /lookup/doi/10.1126/science.272.5259.179e

Overview of DT results from TFTR

Experiments with plasmas having nearly equal concentrations of deuterium and tritium have been carried out on TFTR. To date (September 1995), the maximum fusion power has been 10.7 MW, using 39.5 MW of neutral beam heating, in a supershot discharge and 6.7 MW in a high beta P discharge following a current ramp-down. The fusion power density in the core of the plasma has reached 2.8 MW/m3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER). The energy confinement time tau E is observed to increase in DT, relative to D plasmas, by 20% and the n1(0).T1(0). tau E product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter H mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high beta P discharges. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP simulations assuming classical confinement. Measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from helium gas puffing experiments. The loss of energetic alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha particle driven instabilities has yet been observed. ICRF heating of a DT plasma, using the second harmonic of tritium, has been demonstrated. DT experiments on TFTR will continue both to explore the physics underlying the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor

Fuel Source Isotopic Tailoring and Its Impact on International Thermonuclear Experimental Reactor Design, Operation, and Safety

Isotopic tailoring of the deuterium and tritium density profiles in fusion reactors can lead to reduced tritium inventory in plasma facing components and, therefore, improved safety considerations. The isotopic tailoring concept consists of utilizing a tritium-rich pellet source for core fueling and a deuterium-rich gas source for edge fueling. Because of the improved particle confinement associated with the deeper tritium core fueling component, comparable core densities of deuterium and tritium can be maintained even when the edge deuterium fuel source is much larger than the core tritium fuel source. The fuel composition of the edge and scrape-off plasmas as well as the isotope fractions in plasma facing components reflect the total throughput of all makeup fuel and are therefore deuterium-rich. This innovative fueling concept results in about a factor of two reduction in tritium inventory of the plasma facing components. The higher tritium burn fraction allows a significant reduction in tritium gas flows into and out of the vacuum vessel and, for fusion reactors, implies lower required tritium breeding ratios. 10 refs., 3 figs., 2 tabs.

Review of deuterium–tritium results from the Tokamak Fusion Test Reactor

After many years of fusion research, the conditions needed for a D–T fusion reactor have been approached on the Tokamak Fusion Test Reactor (TFTR) [Fusion Technol. 21, 1324 (1992)]. For the first time the unique phenomena present in a D–T plasma are now being studied in a laboratory plasma. The first magnetic fusion experiments to study plasmas using nearly equal concentrations of deuterium and tritium have been carried out on TFTR. At present the maximum fusion power of 10.7 MW, using 39.5 MW of neutral-beam heating, in a supershot discharge and 6.7 MW in a high-βp discharge following a current rampdown. The fusion power density in a core of the plasma is ≊2.8 MW m−3, exceeding that expected in the International Thermonuclear Experimental Reactor (ITER) [Plasma Physics and Controlled Nuclear Fusion Research (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 239] at 1500 MW total fusion power. The energy confinement time, τE, is observed to increase in D–T, relative to D plasmas, by 20% and the ni(0) Ti(0) τE product by 55%. The improvement in thermal confinement is caused primarily by a decrease in ion heat conductivity in both supershot and limiter-H-mode discharges. Extensive lithium pellet injection increased the confinement time to 0.27 s and enabled higher current operation in both supershot and high-βp discharges. Ion cyclotron range of frequencies (ICRF) heating of a D–T plasma, using the second harmonic of tritium, has been demonstrated. First measurements of the confined alpha particles have been performed and found to be in good agreement with TRANSP [Nucl. Fusion 34, 1247 (1994)] simulations. Initial measurements of the alpha ash profile have been compared with simulations using particle transport coefficients from He gas puffing experiments. The loss of alpha particles to a detector at the bottom of the vessel is well described by the first-orbit loss mechanism. No loss due to alpha-particle-driven instabilities has yet been observed. D–T experiments on TFTR will continue to explore the assumptions of the ITER design and to examine some of the physics issues associated with an advanced tokamak reactor.

ITER: the first experimental fusion reactor

The International Thermonuclear Experimental Reactor (ITER) project is a multiphased project, at present proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the USA and the Government of Russia (“the parties”). The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test technologies for a demonstration fusion power plant. During the extended performance phase of ITER, it will demonstrate the characteristics of a fusion power plant, producing more than 1500 MW of fusion power. The objective of the engineering design activity (EDA) phase is to produce a detailed, complete and fully integrated engineering design of ITER and all technical data necessary for the future decision on the construction of ITER. The ITER device will be a major step from present fusion experiments and will encompass all the major elements required for a fusion reactor. It will also require the development and the implementation of major new components and technologies. The inside surface of the plasma containment chamber will be designed to withstand temperature of up to 500 °C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures, and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stage of development of testing. The main systems of ITER are the superconducting magnet coils and their support, the vacuum vessel and the shield-blanket, the heating and fueling systems, the cryostat, the power supplies and the buildings. During operation the ITER device will sustain a controlled fusion burn for periods of more than 1000 s. Numerous components, experimental packages and test modules will have to be remotely installed and removed from the ITER device in order to test materials properties, component characteristics and material lifetimes. The status of the ITER design as of July 1994 is presented below.

ITER: the first experimental fusion reactor

The International Thermonuclear Experimental Reactor (ITER) Project is a multi-phased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement between the European Atomic Energy Community, the Government of Japan, the Government of the United States, and the Government of the Russian Federation. The project is based on the tokamak, a Russian invention which has been brought to a high level of development and progress in all major fusion programs throughout the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for commercial energy production and to test the technologies for a demonstration fusion power plant. During extended performance operaton ITER will be capable of producing more than 1000 MW (electric) of fusion power, an amount of power that is comparable with one of today's electricity generating plants. The objective of the Engineering Design Activities (EDA) phase is to produce the detailed, complete, and fully integrated engineering design of the tokamak and all technical data necessary for the construction of ITER. The ITER project will be a major step from present fusion experiments and will progress towards a fusion reactor. It will also require the development and implementation of major new components and technologies. The inside surfaces of the plasma containment chamber will be designed to withstand temperatures of up to nearly 500°C, although normal operating temperatures will be substantially lower. Materials will have to be carefully chosen to withstand these temperatures and a high neutron flux. In addition, other components of the device will be composed of state-of-the-art metal alloys, ceramics and composites, many of which are now in the early stages of development and testing. The main systems of ITER are the superconducting magnet coils and their support, the vacuum vessel and the shield/blanket, the heating, fueling systems, the cryostat, the power source and the buildings. During operation the ITER device will sustain a controlled fusion burn for periods of greater than 1000 s. Numerous components, experimental packages, and test modules will have to be remotely installed and removed from the ITER device in order to test materials properties, component characteristics and material lifetimes. This paper reviews the present status of the ITER design, its components and technologies.

LOCA Temperature/Hydrogen Generation Parametric Study for ITER TAC4 Design

A parameter study of the transient thermal response of the recent International Thermonuclear Experimental Reactor design was performed for unmitigated loss of coolant accidents (LOCAs). Inboard component temperatures and hydrogen production levels remained low for an unmitigated blanket cooling system LOCA. Adding an unmitigated LOCA of the vacuum vessel cooling system to the blanket cooling system LOCA scenario results in much higher first wall temperatures and the production of flammable and, eventually, detonatable amounts of hydrogen in the plasma chamber.

Industrial roles in the international thermonuclear experimental reactor (ITER) project

The International Thermonuclear Experimental Reactor (ITER) represents a major step toward realizing the benefits of magnetic fusion energy. Four international parties, the European Communities, Japan, the Russian Federation and the United States are collaborating on the design of ITER and the development of the relevant technology. Because the long range goal is to construct a demonstration reactor, and ultimately to develop fusion power on a commercial basis, there is worldwide emphasis on enhancing industrial involvement throughout the Engineering Design Activities (EDA).