Ignition Tokamak Research Articles

New horizontal and vertical field coils with optimised location for robust decentralized plasma position control in the IGNITOR tokamak

The paper deals with the improvement of the poloidal system of the IGNITOR tokamak, and the design and simulation of the robust decentralised control systems of the plasma’s vertical and horizontal position. The results were obtained on a linear model that is constructed from the circuit equations for the coils, plasma, and vacuum vessel (VV) currents, and from the force balance equations for the plasma. Two new coils that are located close to the VV have been introduced in the IGNITOR poloidal system; specifically, a Horizontal Field Coil (HFC) that is designated for the plasma vertical position control and a Vertical Field Coil (VFC) that is used for the horizontal position control. For the first time in the design of D-shaped tokamaks poloidal systems, the location of the HFC was optimised to achieve the maximum size of the plasma controllability region in the vertical direction and the location of the VFC was determined from the optimisation of the plasma position response to the voltage step function. Although the ideal position of the VFC with fast response time is on the equatorial plane, its location usually is occupied by the other devices such as ports, diagnostic system, and heating systems, so the separated two-section VFC is used. Careful calculation shows the performance is degraded by only 20 % in this layout. For the improved IGNITOR poloidal system, the new robust decentralised plasma position control systems were designed by means of two robust approaches: H∞-optimisation and Linear Matrix Inequalities (LMI). The results of the simulation of the control systems with two types of actuator models as the multiphase thyristor rectifier and the voltage inverter in Pulse- Width Modulation (PWM) mode for the new IGNITOR poloidal system are presented.

Current Status and Objectives of Modernizing the Engineering, Physical, and Energy Infrastructure of the SFT Facility for the Implementation of the Ignitor Project

The Ignitor tokamak nuclear fusion research project is one of the most ambitious initiatives undertaken under the long-term scientific cooperation program between the Russian Federation and the Italian Republic. Currently, negotiations are underway to develop the engineering design of the Ignitor tokamak to be sited at JSC SRC TRINITI (Troitsk Institute for Innovation and Fusion Research) in Troitsk, Moscow. The Ignitor tokamak differs essentially from other tokamaks currently under development. It will have an ultra-powerful magnetic field of 13 T, in which a powerful plasma current of 11 MA will flow at a pulse discharge of about 10 s. The fusion reaction will be ignited mainly by ohmic heating. On the basis of the inspection of practically all the basic components of the energy and supporting infrastructure of the TRINITI Strong-Field Tokamak (SFT) experimental test facility, including steady-state and pulsed power supply systems, vacuum, cryogenic, and fuel systems, and diagnostic equipment, as well as on the analysis of the technical requirements for the tokamak support systems, the proposals for modernizing the TRINITI SFT experimental test facility for the implementation of the Ignitor Project were developed.

Concept of the Fuel Cycle of the IGNITOR Tokamak

In 2015, the development of the conceptual design of IGNITOR, a Russian–Italian experimental thermonuclear tokamak, was completed, and the parameters of all the systems, including the fuel system, were determined. In this work, an integrated design of arrangement of all the tritium-containing gas and secondary water streams expected in the operation of the IGNITOR tokamak was proposed on the basis of analyzing the designs of the fuel cycles of the large thermonuclear facilities JET, TFTR, and ITER. This design includes the storage of deuterium and tritium in a system of gas cylinders at a pressure of 0.4 MPa, the purification of the exhaust gas mixture from the plasma chamber using a “hot getter,” the separation of the deuterium–tritium isotope mixture by displacement gas chromatography, the purification of the process gas and air streams in a wet scrubber, and the detritiation of the water streams by chemical isotope exchange of hydrogen with water. The main technical parameters of equipment in the proposed design were estimated.

Scope of modification of the TRINITI site fuel cycle complex for the IGNITOR project tasks: System of isotopes storage

The IGNITOR tokamak project is one of main directions of scientific collaboration between Russia and Italy. The project is entering the stage of technical design for the location of the machine at the TRINITI site in Troitsk of the Moscow region. The IGNITOR machine differs considerably from other machines based on the tokamak concept by using a super strong magnetic field (13 T) and plasma current (11 MA). It will operate with short pulses (of approximately 10 s in length) and will not have a tritium breeding blanket. The requirements of the tritium fuel cycle are high because three tritium pulses per day are foreseen with a total amount of approximately 10 g to be processed daily. Fuelling will be based on gas injection. It necessary to provide a full tritium processing cycle including the storage and supply of the tritium, plasma exhaust purification, separation of the hydrogen isotopes, detritiation of gaseous and water streams, and tritium recovery from the plasma facing components. This paper considers tritium storage and delivery system as part of the fuel cycle.

Status and tasks of TRINITI site infrastructure modernization for the Ignitor project

The tokamak Ignitor is one of the main long-term projects involving scientific cooperation between Russian Federation and Italian Republic. Currently, negotiations are underway to develop the Ignitor Technical Design Project with on-site placement at site of the experimental bench complex "Tokamak with Strong Field" (complex TSP) of the Joint Stock Company "State Research Centre of Russian Federation TRINITI" (TRINITI) Moscow, Troitsk. The Ignitor project differs significantly from present-day experimental fusion tokamak reactors, having a strong magnetic field (13 T), in which the pulse discharge (about 10 s) flows a powerful discharge current (11MA). Ohmic heating is the main mechanism for ignition of the fusion reaction. During inspection of the current state of equipment and communications of all the main elements of the power and engineering infrastructure of the complex TSP, including stationary and pulsed power supply systems, the vacuum, cryogenic, and fuel systems and diagnostic complex were analyzed. Based on a comparison of the requirements of the Ignitor project to the site location infrastructure, involving from published sources technical characteristics and requirements for livelihood systems of the Ignitor tokamak the technical proposals for modernization of the complex TSP under IGNITOR project task were prepared.

Smaller & Sooner: Exploiting High Magnetic Fields from New Superconductors for a More Attractive Fusion Energy Development Path

The current fusion energy development path, based on large volume moderate magnetic B field devices is proving to be slow and expensive. A modest development effort in exploiting new superconductor magnet technology development, and accompanying plasma physics research at high-B, could open up a viable and attractive path for fusion energy development. This path would feature smaller volume, fusion capable devices that could be built more quickly than low-to-moderate field designs based on conventional superconductors. Fusion’s worldwide development could be accelerated by using several small, flexible devices rather than relying solely on a single, very large device. These would be used to obtain the acknowledged science and technology knowledge necessary for fusion energy beyond achievement of high gain. Such a scenario would also permit the testing of multiple confinement configurations while distributing technical and scientific risk among smaller devices. Higher field and small size also allows operation away from well-known operational limits for plasma pressure, density and current. The advantages of this path have been long recognized—earlier US plans for burning plasma experiments (compact ignition tokamak, burning plasma experiment, fusion ignition research experiment) featured compact high-field designs, but these were necessarily pulsed due to the use of copper coils. Underpinning this new approach is the recent industrial maturity of high-temperature, high-field superconductor tapes that would offer a truly “game changing” opportunity for magnetic fusion when developed into large-scale coils. The superconductor tape form and higher operating temperatures also open up the possibility of demountable superconducting magnets in a fusion system, providing a modularity that vastly improves simplicity in the construction, maintenance, and upgrade of the coils and the internal nuclear engineering components required for fusion’s development. Our conclusion is that while tradeoffs exist in design choices, for example coil, cost and stress limits versus size, the potential physics and technology advantages of high-field superconductors are attractive and they should be vigorously pursued for magnetic fusion’s development.

Electromagnetic disruption analysis in IGNITOR

The present paper reports a detailed analysis of disruptions in the IGNITOR tokamak. Two different numerical models are used (one axisymmetric, the other three-dimensional), in order to validate the results and to highlight specific effects due to different assumptions. The results show a good agreement between the two codes, when similar assumptions are made, and highlight the potential significance of three-dimensional effects.

On the thermodynamics of IGNITOR plasma

Independent assessment of the feasibility of controlled nuclear fusion in the proposed IGNITOR tokamak is difficult because of the uncertainties concerning energy transport in the plasma. We discuss one IGNITOR scenario with the help of well-known general constraint on transport in weakly collisional, axisymmetric, toroidal, low-β turbulent plasma. No particular ad hoc model for transport coefficients is invoked. As a result, the relevance of the Coppi–Mazzuccato–Gruber model is questioned. Moreover, the crucial role of temperature pedestal at the boundary is highlighted.

The zero-waste option for nuclear fusion reactors: Advanced fuel cycles and clearance of radioactive materials

This paper deals with the radioactive waste issue for fusion reactors, proposing an innovative solution (the “zero-waste” option), which could be a clear advantage of fusion power versus fission, in view of its ultimate safety and public acceptance. Its goal is the Clearance (declassification to non-radioactive materials) of all reactor components, after a sufficient period of interim decay, according to the recommendations recently issued by IAEA. Even if feasible in theory, a zero-waste option for fusion reactors using the Deuterium–Tritium fuel cycle will be difficult to obtain in practice: a relevant amount of radioactive materials from reactor decommissioning – even if recyclable within the nuclear industry – should be disposed of as low-level waste. As a further step towards the zero-waste option, the features of fusion reactors based on alternative advanced fuel cycles have been examined, to assess whether that goal could be reached for such devices. Fusion reactors with advanced Deuterium–Helium-3 (DHe) fuel cycle have quite outstanding environmental advantages. Compact ignition tokamaks can be designed in order to achieve DHe ignition. Ignitor, a compact ignition experiment aimed at studying DT plasmas, may also be used in that direction. The extrapolation of Ignitor technologies towards a larger and more powerful experiment using advanced fuel cycles (Candor) is described. Results show that Candor does reach the zero-waste option. A fusion power reactor based on the DHe cycle could be the ultimate correct response to the environmental requirements for future nuclear power plants.

Burning plasma projections using drift-wave transport models and scalings for the H-mode pedestal

The GLF23 and multi-mode core transport models are used along with models for the H-mode pedestal to predict the fusion performance for the International Thermonuclear Experimental Reactor, Fusion Ignition Research Experiment, and IGNITOR tokamak designs. Simulations using combinations of core and pedestal models have also been compared with experimental data for H-mode profiles in DIII-D, JET, and Alcator C-Mod. Power-independent (ballooning mode limit) and power-dependent pedestal scalings lead to very different predictions when used with the core models. Although the two drift-wave transport models reproduce the core profiles in a wide variety of tokamak discharges, they differ in their projections to burning plasma experiments for the same pedestal parameters. Differences in the core transport models in their response to the ion temperature gradient (i.e. their stiffness) and impact of the power dependence of the H-mode pedestal on fusion performance predictions are discussed.

Feasibility Study of a Compact Ignition Tokamak Based upon GyroBohm Scaling Physics

The design of a reduced size (R = 4.45 m, BT = 5.04 T) ignition tokamak (Q = ∞) with superconducting coils using a standard ELMing H-mode plasma appears to be feasible. This effective size (BT2/3R5/6) is smaller than current proposals for Q = 10 burning (D-T) plasma experiments. The good confinement required for ignition with this small effective size is obtained by operating along a gyroBohm scaling path starting from the existing tokamak database at high beta (β = 4.1%) so that the loss power from core transport exceeds the H-mode threshold power. Using a design that can achieve a high normalized current (Ip/aBT = 1.63) also helps to decrease the size of the machine. The design of this relatively compact ignition tokamak satisfies reasonable engineering constraints on the superconducting toroidal field coils and central solenoid, and allows for a sufficiently long burn time for the plasma current to relax to its final state.

Activation analysis for the Ignitor tokamak

The purpose of the Ignitor experiment is to produce DT plasma regimes where ignition can take place. From the neutronics point of view, after a first period of machine set up and tuning in aneutronic operation, a second phase of 2.5 MeV neutron production with pure deuterium plasmas at increasing density is anticipated. After that, a transition phase to increasing percentage of tritium will be followed by several years of operations with 50% of tritium leading to short but intense 14-MeV neutron emission. To deal with the concerns due to the strong neutron emission, a detailed description of the actual Ignitor machine is implemented in the MCNP-4B Monte Carlo code to calculate the neutron fluxes in all machine components, taking in account the streaming through the ports. These fluxes are then used as input for the FISPACT-97 code for the analysis of the activation at the end of life and at intermediate times for safety assessment purposes. The resulting dose rate outside the device is also studied. The results point out that remote handling is needed for repair/maintenance of inner components following the DD operations, and that the access in the hall near the device could be restricted from the beginning of DT operations. The requirements on additional shield are discussed.

Absorbed dose calculations for the Ignitor tokamak magnet coils insulator

The Ignitor experiment is an advanced compact high magnetic field tokamak with cryogenically cooled (30 K) normal conductor magnets. Its main purpose is to produce deuterium–tritium plasma regimes where ignition can take place. From the neutronics point of view, the routine operations with 50% of tritium will lead to a very short (4 s) but intense emission of 14 MeV neutrons (up to 3×10 19 n/s). The capability of the boron-free glass reinforced epoxy resin presently adopted for the insulation of the coils conductor to withstand the severe irradiation condition could be a key issue in the performance, reliability and lifespan of Ignitor. In order to better know the radiation environment in which the insulators will operate, calculations of the absorbed dose are performed using the parallelized version of MCNP-4B Monte Carlo code. A detailed 3D geometry description of the tokamak, an accurate representation of the neutron source distribution together with the more recent version of transport cross-section libraries EFF (European Fusion File) are used in this work. The variation of calculated dose depending on the insulator position inside the device, on the neutron energy spectra and on the total number of neutrons produced over the scheduled machine lifetime are discussed in the present paper.

Ignitor neutronics and activation

Abstract The neutron and gamma fluxes in all the main components of the projected Ignitor tokamak are calculated and discussed. The results are obtained using the 3-D Monte Carlo code mcnp with a detailed description of the device geometry and an accurate representation of the neutron source. The neutron-induced radioactivity, contact dose rates, and radiation damage of the relevant parts of the device are then derived. The consequences on the life and maintenance of device components, the exposure of personnel working in tokamak environments, and the disposal of components after shutdown are described.

Passive Plasma Vertical Stabilization in a Single-Turn Tokamak Configuration

The plasma vertical stability in a single-turn tokamak configuration is analyzed. The stabilization effects of the vacuum vessel and poloidal field magnet are studied numerically with rigid and magnetohydrodynamic models. An analytic mode dispersion relation is derived to estimate the effect of the single-turn toroidal field magnet on the plasma vertical stability. The typical growth time of the mode is found to be >1 s. The stability advantages of the single-turn configuration for a high-current tokamak plasma and the differences among the three models used are discussed. A single-turn tokamak configuration seems to be appropriate for a fusion ignition experiment in that it simplifies plasma control and makes feasible the control of high-current, elongated tokamak ignition plasma.

Neutron transport studies for the Ignitor neutron diagnostics

As a preliminary step for the project of the neutron diagnostics in the Ignitor tokamak, the overall neutron field in the device has been calculated with the help of the MCNP code, which solves, with the Monte Carlo method, neutron and photon transport problems in arbitrary three-dimensional geometries. The poloidal distribution of the fluxes on the first wall and the related energy spectra have been calculated to estimate the backscattered contribution to signals at spectrometers and collimated detectors for measurements with spatial resolution in the plasma. The average fluxes on the top and the lateral cryostat surfaces and the local fluxes on the equatorial plane outside the cryostat wall are also calculated in order to determine the neutron counters characteristics and expected performances.

Potential for D-3He Experiments in Next-Generation Tokamaks

In this paper the potential for D-{sup 3}He experiments in the proposed Compact Ignition Tokamak (CIT) and International Thermonuclear Experimental Reactor (ITER) tokamak test devices is examined. In CIT, an energy multiplication Q of {approximately}0.3 can be obtained with an injection power of {approximately}100 MW. Without modifications to ITER, except for the change of fuel, it is found that Q of the order of 0.3 to 0.5 can be obtained. Breakeven with D-{sup 3}He requires modification to the device to increase the elongation to 2.4, reduce the major radius to 5.6 m, and increase the magnetic field at the plasma from 4.9 to 5.6 T. Operation with a small amount of tritium seeding can reduce the auxiliary power required to achieve breakeven and leads to Q = 2 in an unmodified device.

Ohmic Ignition of Neo-Alcator Tokamak with Adiabatic Compression.

Ohmic ignition condition on axis of the DT tokamak plasma heated by minor radius and major radius adiabatic compression is studied assuming parabolic profiles for plasma parameters, elliptic plasma cross section, and Neo-Alcator confinement scaling.It is noticeable that magnetic compression reduces the necessary total plasma current for Ohmic ignition device.Typically in compact ignition tokamak of the minor radius of 0.47m, major radius of 1.5m and on-axis toroidal field of 20T, the plasma current of 6.8MA is sufficient for compression plasma, while that of 11.7MA is for no compression plasma. Another example with larger major radius is also described.In such a device the large flux swing of Ohmic transformer is available for long burn.Application of magnetic compression saves the flux swing and thereby extends the burn time.

Shielding Design Optimization for the Compact Ignition Tokamak Test Cell Building

Shielding optimization for the Compact Ignition Tokamak baseline test cell building design is performed to minimize the cost. The design is based on iterative calculations using one-dimensional SN ...

Alpha particle destabilization of the toroidicity-induced Alfvén eigenmodes

The high-frequency, low mode number toroidicity-induced Alfvén eigenmodes (TAE) [Phys. Fluids 29, 3695 (1986)] are shown to be driven unstable by the circulating and/or trapped α particles through the wave–particle resonances. Satisfying the resonance condition requires that the α-particle birth speed vα≥vA/2‖m−nq‖, where vA is the Alfvén speed, m is the poloidal mode number, and n is the toroidal mode number. To destabilize TAE modes, the inverse Landau damping associated with the α-particle pressure gradient free energy must overcome the velocity space Landau damping due to both the α particles and the core electrons and ions. The growth rate was studied analytically with a perturbative formula derived from the quadratic dispersion relation, and numerically with the aid of the nova-k code. Stability criteria in terms of the α-particle beta βα, α-particle pressure parameter (ω*/ωA) (ω* is the α-particle diamagnetic drift frequency), and (vα/vA) parameters will be presented for the Tokamak Fusion Test Reactor (TFTR) [Proceedings of the Thirteenth International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Crystal City, VA, 1990 (International Atomic Energy Agency, Vienna, in press)], Compact Ignition Tokamak (CIT) [Phys. Scr. T16, 89 (1987)], and the International Thermonuclear Experimental Reactor (ITER) [ITER Documentation Series, No. 21 (International Atomic Energy Agency, Vienna, 1991)]. The volume-averaged α-particle beta threshold for TAE instability also depends sensitively on the core electron and ion temperature. Typically the volume-averaged α-particle beta threshold is in the order of 10−4. Typical growth rates of the n=1 TAE mode can be in the order of 10−2ωA , where ωA=vA/qR . Other types of global Alfvén waves are stable in deuterium–tritium (D–T) tokamaks due to toroidal coupling effects.