Fusion Core Experiment Research Articles

II. Device Description and General Physics Requirements

To accomplish the dual goals set forth in the mission statement - determination of burning plasma physics and demonstration of fusion power production - requires a tokamak with special characteristics. A conceptual design for such a facility has been developed by the CIT/BPX Project team over a period of about 5 years. The process has drawn extensively upon the world tokamak physics data base as well as engineering experience gained on actual machines like Tokamak Fusion Test Reactor (TFTR) and the Alcators and on design studies for machines like the Long-Pulse Ignited Test experiment (LITE) and the Tokamak Fusion Core Experiment (TFCX). The Joint European Torus (JET) and Doublet III-D (DIII-D) experiments in particular have had a significant influence on the physics base. The resulting design incorporates features that have proven successful in tokamak experiments, combined with new features that are needed in the regime of high-power-density deuterium-tritium (D-T) fusion plasmas.

Tiber II as a precursor to an international thermonuclear experimental reactor

The Tokamak Ignition/Burn Experimental Reactor (TIBER) [1] was pursued in the U.S. as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak [2]. While the coppercoil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT) [3], the superconducting TIBER [4] with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing.

Consequences of resistive disruptions on vacuum vessel components

A computer code, DSTAR, has recently been developed to quantitify the surface erosion and induced forces that can occur during major tokamak plasma disruptions. A disruption analysis has been performed with DSTAR for the Tokamak Fusion Core Experiment (TFCX) device. The limiters and inboard first wall were assumed to be clad with beryllium. Disruption simulations were performed with and without these structures present, to determine their electromagnetic influence. The conclusion is drawn that disruption simulations that do not include both the thermal and electromagnetic response of the vacuum vessel will not result in an accurate erosion prediction.

Determination of current densities for tokamak superconducting toroidal field coils

A major goal of designing a tokamak is to minimize the size of the device and achieve lowest cost. Two key factors influencing the size of the device employing superconducting magnets are toroidal field (TF) winding current density and its nuclear heat load withstand capability. Lower winding current density requires larger radial build of the winding pack. Likewise, lower allowable nuclear heating in the winding requires larger shield thickness between the plasma and TF coils. In order to achieve a low-cost device, it is essential to maximize the winding's current density and nuclear heating withstand capability. A methodology for determining optimum current density is developed by using the Tokamak Fusion Core Experiment (TFCX) as an example. A winding current density of 3500 A/cm2 is determined to be optimal at a peak field of 10 T and peak nuclear heat load limit of 1 mW/cm3. This study is based on employment of Nb3Sn cable-in-conduit conductors cooled with forced-flow helium.

Plasma Performance of Tokamak Fusion Core Experiment and Joint European Torus with Sawtoothing

The plasma performance with sawtoothing is assessed for two tokamak reactor experiments, the proposed Tokamak Fusion Core Experiment (TFCX) and the joint European Torus (JET). TFCX is ignitable with sawtoothing and impurities assuming neo-Alcator plus Gorbunov-Mirnov-Strelkov scalings for plasma confinement. JET is also ignitable at optimistically high currents (greater than or equal to7 MA).

Low-Temperature Plasma Near a Tokamak Reactor Limiter

Analytic and two-dimensional computational solution for the plasma parameters near a toroidally symmetric limiter are illustrated for the projected parameters of a tokamak fusion core experiment (TFCX). The temperature near the limiter plate is below 20 eV, except when the density 10 cm inside the limiter contact is 8 x 10/sup 13/ cm-/sup 3/ or less and the thermal diffusivity in the edge region is 2 x 10/sup 4/ cm/sup 2//S or less. Extrapolation of recent experimental data suggests that neither of these conditions is likely to be met near ignition in TFCX, so as a low plasma temperature near the limiter should be considered a likely possibility.

The ignition studies project at princeton

During the past year the Ignition Studies Project (ISP) at Princeton has been investigating design options and performing systems studies on science oriented DT ignition experiments. The goal has been to identify promising approaches leading to a device that could test the basic viability of the burning-plasma regime at a relatively early time and moderate cost. This is a substantial departure from the Tokamak Fusion Core Experiment (TFCX) which had both a physics and an engineering mission. This paper describes some of the approaches under consideration and highlights some of the tradeoffs that need to be understood for the development of an ignition experiment.

Determination of current densities for tokamak superconducting toroidal field coils

A major goal of designing a tokamak is to minimize the size of the device and achieve lowest cost. Two key factors influencing the size of the device employing superconducting magnets are toroidal field (TF) winding current density and its nuclear heat load withstand capability. Lower winding current density requires larger radial build of the winding pack. Likewise, lower allowable nuclear heating in the winding requires larger shield thickness between the plasma and TF coils. In order to achieve a low-cost device, it is essential to maximize the winding's current density and nuclear heating withstand capability. A methodology for determining optimum current density is developed by using the Tokamak Fusion Core Experiment (TFCX) as an example. A winding current density of 3500 A/cm 2 is determined to be optimal at a peak field of 10 T and peak nuclear heat load limit of 1 mW/cm 3. This study is based on employment of Nb 3Sn cable-in-conduit conductors cooled with forced-flow helium.

Cryogenic Hydrogen Isotope Distillation for the Fusion Fuel Cycle

Cryogenic distillation is an attractive method for the hydrogen isotopic separations required in fusion fuel cycles. The theoretical and practical aspects of designing and constructing such systems are well established. Practical considerations in the design of systems are presented and applied to the Isotope Separation System (ISS) at the Tritium Systems Test Assembly (TSTA), as well as systems of distillation columns that might be used for a reactor such as the Tokamak fusion Core Experiment (TFCX) and the recovery of breeding blanket product.

Tritium safety issues for TFCX

Estimated tritium releases from the Tokamak Fusion Core Experiment are compared to the expected limits. A reaction kinetics model is described that predicts the conversion of tritium to the oxide form in free space. An analysis of the required capacity of the Emergency Tritium Cleanup System is also presented. The conclusions of this work are expected to be applicable to other experimental fusion devices that are now being considered.

TFCX Current Density and Nuclear Heating Limitations for Superconducting Tokamaks

A major goal of the Tokamak Fusion Core Experiment (TFCX) study was to minimize the size of the device and achieve lowest cost. Two key factors influencing the size of the device employing superconducting magnets are toroidal field (TF) winding current density and its nuclear heat load withstand capability. Lower winding current density requires larger radial build of the winding pack. Likewise, lower allowable nuclear heating in the winding requires larger shield thickness between the plasma and coil. To achieve a low-cost device, it is essential to maximize the winding's current density and nuclear heating withstand capability. To meet this objective, the TFCX design specification adopted as goals a nominal winding current density of 3500 A/cm/sup 2/ with 10-T peak field at the winding, peak nuclear heat load limits of 1 mW/cm/sup 3/ for the nominal design and 50 mW/cm/sup 3/ for an advanced design. This study developed justification for these current density and nuclear heat load limits.

Tokamak Fusion Core Experiment (TFCX) Special-Purpose Remote Maintenance Systems

A key element in the preconceptual design of the Tokamak Fusion Core Experiment (TFCX) was the development of design concepts for special-purpose remote maintenance systems. Included were systems for shield sector replacement, vacuum vessel sector and toroidal field coil replacement, limiter blade replacement, protective tile replacement, and general-purpose maintenance. This paper addresses these systems as they apply to the copper toroidal field (TF) coil version of the TFCX.

Impurity Control and Vacuum Pumping System Design and Analysis for Next-Generation Tokamaks

Impurity control system design and performance studies were performed in support of the Tokamak Fusion Core Experiment (TFCX) preconceptual design. Efforts concentrated on the pumped limiter and vacuum pumping system design configuration, thermal/mechanical and erosion lifetime performance of the limiter protective surface, and helium ash removal performance. Analysis results indicate that the limiter/vacuum pumping system design provides marginally adequate helium ash removal. Difficulties in providing adequate helium ash removal for more compact or higher fusion-power-density devices are addressed. Erosion, primarily by disruption-induced vaporization and/or melting, limits the protective surface lifetime to about one calendar year or only about 60 full power hours of operation. In addition to evaluating impurity control system performance for nominal TFCX conditions, these studies attempt to focus on the key plasma physics and engineering design issues that should be addressed in future research and development programs.

Impact of a Poloidal Divertor in Ignition Tokamak Design

System design studies were performed to assess the effect of assuming a poloidal divertor instead of a limiter as a means of impurity control for ignition tokamak configurations. Results show that for the nominal Tokamak Fusion Core Experiment (TFCX) device with superconducting TF coils, a feasible poloidal divertor configuration can be obtained without increasing the major radius. In the TFCX nominal copper TF coil device, however, field limits at the PF coils are exceeded when the effects of asymmetry associated with a poloidal divertor are included. It was found that a 12% increase in the major radius of this device is necessary to simultaneously satisfy the plasma-shaping requirements of a poloidal divertor and the magnetics constraints at the superconducting PF coils.

Tokamak Fusion Core Experiment Maintenance Study

The recently completed Tokamak Fusion Core Experiment (TFCX) design project was carried out to investigate potential next generation tokamak concepts. An important aspect of this project was the early development and incorporation of remote maintainability throughout the design process. This early coordination and incorporation of maintenance aspects to the design of the device and facilities would assure that the machine could ultimately be maintained and repaired in an efficient and cost effective manner. To meet this end, a rigorously formatted engineering trade study was performed to determine the preferred configuration for the TFCX reactor based primarily on maintenance requirements. The study indicated that the preferred design was one with an external vacuum vessel and torrodial field coils that could be removed via a simple radial motion. The trade study is presented and the preferred TFCX configuration is described.

Tokamak Fusion Core Experiment Facilities Functional and Operational Requirements

A team of national laboratory, university, and industrial participants completed the preconceptual design for the Tokamak Fusion Core Experiment (TFCX), a long pulse, plasma ignition machine, and required support facilities. Functional and Operational Requirements (F and ORs) for the TFCX support facilities were developed as the basis for the preconceptual design, ensuring that adequate housing and site would be provided to support the tokamak machine and auxiliary systems. This paper presents partial F and ORs developed for the base case TFCX machine, the nominal superconducting option (liquid helium-cooled magnets), and describes the resulting preconceptual design.

Issues Relating to the Siting of Tritium-Fueled Fusion Experiments

A preconceptual design study and safety analysis of the Tokamak Fusion Core Experiment (TFCX) was conducted in 1984 for the Department of Energy. This paper summarizes the calculations and comparisons related to TFCX siting and environmental issues such as radiological doses to the public living near the facility. Included are discussions of (a) routine and accidental releases of tritium, (b) routine releases of activated air, (c) direct radiation (including ''skyshine''), and (d) seismic criteria. Other potential issues are also discussed including the amount of tritium that might be retained in the graphite armor in the torus, the possible severity of magnet accidents, and the extent of damage due to plasma disruptions. The conclusions drawn from these calculations should be applicable to some of the other planned ignited core experiments that have operating parameters similar to those of TFCX.

Fusion Reactor and Test Cell Configuration

Remote maintenance and repair considerations will play a major role in determining the design and layout of next-generation fusion devices, including the buildings and facilities supporting the nuclear operation of those devices. The reactor/test cell described in this paper represents a “nominal copper” TFCX (Tokamak Fusion Core Experiment) concept that emphasizes an integrated approach to the design of the experimental device and its supporting facilities. This configuration is unique with respect to other TFCX concepts in that it promotes the use of a modular design with readily replaceable stand-alone sectors, integrates the machine and test cell structures, incorporates a facility arrangement that results in a smaller overall facility footprint, reduces the transportation path of activated hardware, and allows access to three sides of the test cell facilities.

Shield Design for Next-Generation, Low-Neutron-Fluence, Superconducting Tokamaks

A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components.

Design of Aggressive Superconducting TFCX Magnet Systems

An investigation of several fundamental limits of machine design indicate that a machine fitting the specifications of the Tokamak Fusion Core Experiment (TFCX) can be built with both a superconducting toroidal field (TF) coil set and a plasma major radius of less than 3.2 m. This small size is achieved by accepting a peak nuclear heat load of 50 (kW)(m/sup -3/) in the TF coin inner leg while operating at a 10-T maximum field with a current density of 35 (A)(mm/sup -2/) in the winding pack. This performance, high by traditional standards, is justified based on developments in forced flow conductor technology using Nb/sub 3/Sn composite superconductors.