Ultra-compact strong-field spherical tokamak for fusion energy

Abstract

The invention provides an attractive, fast and cost-effective alternative method for conventional nuclear fusion reactors. An ultra-compact spherical Tokamak which combines three key features intended to accelerate fusion energy as feasible energy sources, the features being a spherical Tokamak, a negative triangle plasma shape and a high temperature superconductor coil.

Description

[0001] Object of the invention

[0002] The present invention relates to a fusion reactor, in particular to a tokamak fusion reactor based on a spherical tokamak combined with a negative triangular plasma shape, which operates under a high toroidal field.

[0003] The object of the present invention is to provide a fast and cost-effective alternative for building a commercial fusion power plant that provides for achieving superior confinement mechanisms, ensuring enhanced plasma performance as well as efficient power consumption solutions and self-sufficient tritium production. Background Art

[0004] Utilizing fusion reactions for controlled and cost-effective energy production presents considerable challenges, primarily due to the complexity involved in confining high-temperature plasmas with fusion-relevant properties. One promising approach to nuclear fusion power generation devices is the use of toroidal magnetic fields. The most advanced magnetic topology for confining fusion-burning plasmas is the so-called tokamak device (J. Wesson, Tokamaks, 4 th ed., Oxford University Press, ISBN: 9780199592234). A conventional tokamak is a large-aspect-ratio tokamak (A=R / a, where R is the primary and secondary plasma radii) that confines the plasma in a helical magnetic field, primarily formed by an externally applied toroidal magnetic field and a poloidal field generated by externally induced currents in the plasma. Large conventional tokamaks such as JET, JT60-U, and TFTR have been successfully used to develop the physical foundation and baseline scheme for ITER, the Pinnacle experimental fusion reactor designed to demonstrate the scientific and technological feasibility of fusion as a future energy source with net energy gain. However, the scalability of ITER requires alternative power and particle removal technologies, which are critical to the success of tokamak-based fusion power plants.

[0005] Spherical tokamaks (STs) offer a more compact and more economic alternative path to fusion reactors compared to their conventional tokamak counterparts. Patent application WO2015155531A1 describes a compact nuclear fusion reactor for use as a neutron source or energy source. The reactor comprises a toroidal plasma chamber and a magnet system arranged to produce a magnetic field for confining plasma in the chamber. With a lower aspect ratio (A ~ 2), STs exhibit high confinement levels and more efficient use of the magnetic field and improved stability enabling normalized plasma pressures up to 40% when compared to their conventional tokamak counterparts. (D.C. Robinson, 1999, Plasma Physics and Controlled Fusion, 41, A143). However, the high power densities achievable in STs impose even more stringent constraints on their plasma-facing components (PFCs) compared to conventional tokamaks, making alternative power and particle exhaust technologies necessary for ST-based fusion pilot plants (FPPs). Moreover, due to their compactness, standard STs face important challenges in accommodating the coil system, neutron shielding and tritium breeding blanket in the central pile.

[0006] While the ITER baseline scenario, the high confinement mode (H-mode) with tolerable plasma-wall interactions, has been demonstrated in most of the current tokamaks, its applicability to future fusion power plants remains uncertain. The main challenge is to demonstrate that fusion plasmas can self-heat and sustain through thermonuclear reactions in an economic tokamak configuration. Unfortunately, H-mode tokamak fusion plasmas tend to generate magnetohydrodynamic (MHD) fluctuations that hinder steady-state operation, which is essential for an attractive FPP. MHD fluctuations, in particular edge-localized modes (ELMs), have been the main challenge to the integrity of large burning plasma devices in current tokamaks. The development of core-edge integrated scenarios that combine high performance and ELM-free dynamic exhaust solutions is a top priority in major tokamak research programs worldwide.

[0007] While the robust ELM control technique based on externally applied resonant magnetic perturbations (RMP) has been successful for ITER, such external actuators can not be applicable to FPPs. Recent experiments in TCV and DIII-D tokamaks have shown that H-mode like plasmas with negative triangularity (NT) shape can offer an attractive solution for the ideal core-edge integration scenario in future tokamak-based FPPs (A. Pochelon et al., 1999, Nucl. Fusion 39 1807, Y. Camenen et al., 2007, Nucl. Fusion 47 510, M. E. Austin et al., 2019, Phys. Rev. Lett. 122 115001).

[0008] Experiments performed in TCV have shown that the core plasma confinement is significantly enhanced during the transition from positive to negative triangularity (A. Pochelon et al., 1999, Nucl. Fusion 39 1807). However, this beneficial effect is reduced at high collisional levels. SUMMARY

[0009] According to a first aspect of the present invention, there is provided a spherical tokamak nuclear fusion reactor with high magnetic field and plasma shape characterized by negative triangularity. The high confinement properties observed in negative triangularity L-mode plasmas of conventional tokamaks and the naturally improved core confinement and stability observed in low aspect ratio tokamaks provide for the establishment of an excellent confinement mechanism, ensuring improved plasma performance and an efficient power exhaust solution.

[0010] High temperature superconductors (HTS) exhibit a transformative capability to operate at elevated magnetic fields with reduced size and cost compared to conventional materials such as copper coils and low temperature superconductors. The synergy of HTS technology with negative triangularity and spherical tokamak plasmas is a distinctive feature that is essential for the advancement of the small scale fusion power plant provided by the present invention. The integration of these three key technologies lays the foundation for an advanced super-optimal confinement mechanism, avoiding the edge localized modes, providing a natural solution to the power exhaust challenge and self-sufficient tritium breeding of the super compact fusion reactor. Self-sufficient tritium breeding requires a tritium breeding blanket in the center cell of the reactor, which cannot be installed with positive triangularity plasmas in a compact ST. This approach of the present invention provides a significant step forward in the potential to realize a compact and efficient fusion energy system.

[0011] The ultra-compact spherical tokamak of the present invention uses high temperature superconductor coils. The high temperature superconductor coils are used to achieve higher conductor current density and, therefore, higher magnetic field at higher operating temperature, which exceeds conventional materials such as copper and low temperature superconductors.

[0012] The negative triangularity plasma is another key feature of the tokamak design of the present invention. The plasma shape is configured to have negative triangularity, a unique feature that provides an integrated core-edge power exhaust solution with high confinement for high power density reactors of static plasmas with a boundaryless local mode. The negative triangularity naturally places the divertor legs at a larger radius and provides additional space in the central solenoid to accommodate a robust solenoid, neutron shielding, and tritium breeding blanket.

[0013] The present invention provides a spherical tokamak nuclear fusion reactor with an advanced central solenoid. The design combines a spherical tokamak featuring an advanced central solenoid based on an hourglass shape that provides more space for neutron shielding and tritium breeding on the central column.

[0014] The toroidal magnetic field in a tokamak is generated by toroidal field (TF) coils, while the central solenoid induces the plasma current responsible for the poloidal magnetic field. This combination creates a helicity of the magnetic field that confines the plasma. Conventional TF coils are typically D-shaped copper or superconducting magnets. In conventional tokamaks, the large size required to achieve high fusion energy (fusion power scales with β N 2 B t 4 R 3 、B t toroidal magnetic field, β N scaled by normalized plasma pressure) leads to increased construction costs.

[0015] In STs, the high power density that can be achieved presents additional challenges to their plasma-facing components (PFCs), requiring alternative power and particle exhaust technologies. The limited space of the central column is one of the biggest challenges of STs. To address the limitations in the design of the central solenoid in STs, the present invention utilizes a negative triangularity plasma shape to optimize the available space of the central column by enabling an hourglass shape for the central solenoid.

[0016] The optimized HTS central solenoid with flared ends allows for higher magnetic field, enabling increased fusion power in a compact device with large plasma β N .

[0017] Negative Triangularity Plasma Spherical Tokamak

[0018] A spherical tokamak with negative trigonometric plasma is an attractive solution for tokamak-based fusion power plants, offering an integrated solution for the plasma core and edge, as it provides a high confinement regime with a manageable power load on the plasma-facing components (PFC).

[0019] The synergistic effects of high core plasma confinement in NT-shaped plasmas, the observed reduction in perturbations, and the unique characteristics of low aspect ratio devices may lead to a new ultra-optimal high confinement regime in NT-shaped high-field STs. According to the present invention, the reduced aspect ratio combined with the NT-shaped plasma has the potential to create favorable conditions for providing an excellent high confinement regime with reduced fluctuations, compatible with boundary solutions for power consumption and self-sufficient tritium breeding.

[0020] Superior core plasma confinement of NT-shaped plasma in ST

[0021] The combination of the improved confinement properties observed in NT-shaped plasmas of conventional tokamaks and the naturally improved core confinement observed in low aspect-ratio devices can lead to a new ultra-improved high-confinement scenario in NT-shaped STs.

[0022] While NT-shaped plasmas have recently been observed to suppress perturbations and associated transport in TCV and DIII-D tokamaks, the benefits of low aspect ratios in suppressing electrostatic and electromagnetic microinstabilities have been reported many times in the past (S. Kaye et al., 1999, Fusion Technology 36, 17).

[0023] In a linear physics description, a reduced aspect ratio can reduce the precession frequency of trapped particles due to the so-called finite beta reversal effect, thereby enhancing the favorable region of good orbit-averaged curvature (C) (C. M. Roach et al., 1995, Plasma Physics and Controlled Fusion, 37, 679). However, it should be emphasized that trapped particle driven modes, such as trapped electron modes (TEMs), still play a key role in setting the turbulent transport in a spherical tokamak, as the trapped particle fraction increases with decreasing aspect ratio. Thus, in a TEM-dominated regime, the turbulent transport will decrease with decreasing aspect ratio. Furthermore, spherical tokamaks are typically characterized by large toroidal rotation, and the associated strong E x B flow shear can strongly suppress long-wavelength microturbulence (> ion Larmor radius) driven by ion temperature gradient (ITG) and / or long-wavelength trapped electron mode instability. Thus, ion heat transport in a spherical tokamak can approach the neoclassical level (J. E. Kinsey et al., 2007, Physics of Plasmas 14 102306), although electron heat transport is almost always dominated by turbulent transport. The combination of NT-shaped plasmas in a low-aspect-ratio device can lead to a new, superimproved confinement regime with low fluctuation levels.

[0024] STs typically achieve higher toroidal plasma beta (β t ) than conventional tokamaks, making them particularly attractive for exploring the potential of negative triangularity plasmas to enhance confinement. While conventional tokamaks are typically limited by the Troyon limit (β N = 3), STs can achieve β t values up to 40%, where β N reaches 6 (D. C. Robinson, 1999, Plasma Physics and Controlled Fusion, 41, A143).

[0025] The present invention utilizes NT-shaped plasmas to facilitate power exhaust in spherical tokamaks. In the case of tokamak-based FPPs, efficient power and particle exhaust presents a significant challenge, particularly for future ST-based FPPs with increased power density. NT-shaped plasmas offer a compelling solution, including the absence of edge-localized modes (ELMs), a larger divertor footprint, and it is also easier to incorporate a radiative divertor in an L-mode regime to mitigate power exhaust, assuming that no power needs to cross the separatrix. These attributes help address key challenges on the path to designing a compact ST-based FPP.

[0026] Furthermore, the NT-shaped plasma provides a solution to the main drawback of the ST, which is the limited space on the central column for basic shielding to mitigate high nuclear power loads. According to one aspect of the present invention, the adaptability of the central column design from a simple straight cylinder to an hourglass shape, achieved by the NT-shaped plasma, provides increased space for neutron shielding and tritium breeding. The spherical tokamak of the present invention allows for higher toroidal magnetic fields at higher operating temperatures, thus reducing costs, whereby the fusion power output is increased by using high temperature superconductors (HTS). This polyhedral approach of the present invention reveals the potential of the NT-shaped plasma in optimizing the performance and feasibility of future ST-based FPPs. BRIEF DESCRIPTION OF DRAWINGS

[0027] To complement the description being made and to help a better understanding of the features of the invention, according to a preferred example of a practical embodiment of the invention, a set of drawings is attached as an integral part of the description, in which the following features are represented, in an illustrative and non-limiting manner:

[0028] Figure 1 - Two graphs are shown comparing the fusion power of a conventional tokamak and a spherical tokamak as a function of size and magnetic field.

[0029] Figure 2 - A cross-sectional view of a compact spherical tokamak according to one embodiment of the present invention is shown.

[0030] Figure 3 - A cross-sectional view of a compact spherical tokamak according to one embodiment of the present invention is shown. DETAILED DESCRIPTION

[0031] The present invention presents a super-compact spherical tokamak (1) that combines three key features aimed at accelerating fusion energy as a viable energy source, which are a spherical tokamak, a negative-triangle-shaped plasma shape, and high-temperature superconductor coils to achieve high magnetic fields in a super-compact reactor, while reducing operating costs.

[0032] In Figure 1 , two graphs are shown to compare the fusion power and fusion power gain (Q) of a conventional and a spherical tokamak as a function of size, where the major radius is R and the toroidal magnetic field is B t . As Figure 1 shown, the ST can achieve the same fusion power and fusion power gain as a conventional tokamak in a significantly smaller machine.

[0033] In Figure 2 and Figure 3 , cross-sectional views of the compact spherical tokamak (1) of the present invention are shown from different angles.

[0034] According to one embodiment of the present application, a compact spherical tokamak (1) comprises a toroidal vacuum chamber (8) and a magnetic field system configured to confine a high temperature plasma (2), wherein the plasma (2) is configured to operate in a negative triangularity (NT) plasma shape, the major radius of the plasma (2) is 3 m or less, the current of the plasma (2) is 14 MA or less, the plasma (2) has a toroidal pitch ratio of 2.2 or less, the magnetic field system comprises a solenoid (4), a toroidal field coil (3) and a poloidal field coil (5), wherein the toroidal field coil (3) comprises a high temperature superconductor (HTS) material, the magnetic field in use comprises a toroidal component of 10 T or less, preferably 6 T.

[0035] Fusion alpha particles and high energy neutrons are produced by fusing deuterium and tritium ions. Fusion of deuterium-tritium is carried out in a ratio of 50:50, preferably at least 70:30, to produce neutrons at a rate of at least 1 x 10 17 neutrons per second, preferably at least 5 x 10 18 neutrons per second.

[0036] The central column of the tokamak (1) houses the solenoid (4), as shown in Figure 2 The tokamak (1) of the present application uses a solenoid (4) which is constructed from HTS material and adopts an advanced shape. Instead of using a regular shape, such as a cylindrical shape, for the solenoid (4), an hourglass shape is used which is suitable for the negative triangularity shape of the plasma (2). This provides more space for shielding and tritium breeding on the central column, thus addressing the most critical issue of spherical tokamaks. The plasma (2) is configured to operate in a negative triangularity (NT) shape. The NT shape is obtained by the poloidal field coil (5) which is placed outside the vacuum chamber (8). The poloidal field coil (5) provides flexibility in the operation of single null and double null plasmas.

[0037] The tokamak (1) can be equipped with an advanced divertor (6) to reduce the heat and particle load per unit area on the walls of the vacuum chamber (8). Having a negative triangularity plasma (2) places the divertor (6) at a larger radius compared to positive triangularity, thus providing a natural tool for moderation of the heat load of the divertor (6) and providing additional space in the central dump to accommodate the solenoid (4), neutron shielding and tritium breeding blanket. The tritium breeding blanket is located behind the first wall (7) of the vacuum chamber (8).

[0038] The magnetic field system of the tokamak (1) comprises a toroidal field coil (3) which is suitable for the negative triangularity plasma shape and has an inverted D design. The toroidal field coil (3) is made of a material comprising high temperature superconductor (HTS), preferably cooled to 20 K or higher in use.

[0039] The tokamak (1) of the invention can have wave heating and / or neutral beam heating systems with different injection angles, optimized for plasma heating and current drive. This configuration delivers less than 100 MW, preferably less than 10 MW, of power input to the plasma. And the tokamak (1) operates at a gain factor Q > 1, preferably Q > 10, more preferably Q > 50, of fusion energy.

[0040] The plasma of the tokamak (1) can be kept in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds. The plasma (2) is characterized by a ratio of plasma to magnetic pressure (plasma beta) of more than 5%, preferably more than 10%.

[0041] According to an embodiment of the invention, a method of generating fusion energy by operating a compact spherical tokamak (1) with a negative triangularity plasma (2), the tokamak (1) comprising a toroidal vacuum chamber (8) and a magnetic field system comprising a solenoid (4), a poloidal field coil (5) and a toroidal field coil (3), the toroidal field coil (3) being made of a material comprising HTS configured to confine the high temperature plasma (2), the method comprising: generating a magnetic field with a toroidal component of 10 T or less, preferably 6 T; confining the plasma (2) with a major radius of 3 m or less and a toroidal to poloidal aspect ratio of 2.2 or less; operating a plasma (2) current of 14 MA or less; emitting fusion alpha particles and neutrons; and confining the alpha particles in the plasma (2). Fusion deuterium-tritium is performed with a ratio of 50:50, preferably at least 70:30, to produce neutrons at a rate of at least 1 x 10 17 neutrons per second, preferably at least 5 x 10 18 neutrons per second. The plasma (2) can be kept in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds. The tokamak (1) of the method can have a divertor (6) and a solenoid (4) with an hourglass glass shape.

Claims

1. A compact spherical tokamak (1) comprising an annular vacuum chamber (8) and a magnetic field system configured to confine a high temperature plasma (2), wherein: the plasma (2) is configured to operate with a negative triangularity (NT) plasma shape, the major radius of the plasma (2) is 3 m or less and the plasma current is 14 MA or less; the plasma (2) has a toroidal aspect ratio of 2.2 or less; the magnetic field system comprises a solenoid (4), a toroidal field coil (3) and a poloidal field coil (5), wherein the toroidal field coil (3) comprises a high temperature superconductor (HTS) material; the magnetic field in use comprises a toroidal component of 10 T or less, preferably 6 T.

2. The tokamak (1) according to claim 1, wherein the plasma comprises deuterium and tritium ions to produce fusion alpha particles and high energy neutrons.

3. The tokamak (1) according to claim 1, wherein, the solenoid (4) has an hourglass shape.

4. The tokamak (1) according to claim 1, wherein the poloidal field coil (5) is disposed outside the annular vacuum chamber (8).

5. The tokamak (1) of claim 1, further comprising a divertor (6) for reducing the heat and particle load per unit area on the walls of the vacuum chamber (8).

6. The tokamak (1) according to claim 1, wherein the high temperature superconductor material is cooled to 20 K or more in use.

7. The tokamak (1) according to claim 1, wherein the plasma (2) can be maintained in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds.

8. The tokamak (1) of any preceding claim, further comprising a neutral beam heating system having different injection angles optimised for plasma heating and current drive, the current drive delivering a power input to the plasma of less than 100 MW, preferably less than 10 MW.

9. The tokamak (1) of claim 1, operating at a fusion energy gain factor Q > 1, preferably Q > 10, more preferably Q > 50.

10. The tokamak (1) according to claim 1, wherein, the plasma (2) has a ratio of plasma pressure to magnetic pressure greater than 5%, preferably greater than 10%.

11. A method of generating fusion energy by operating a compact spherical tokamak (1) having a negative triangularity plasma (2), the tokamak (1) comprising an annular vacuum chamber (2) and a magnetic field system comprising a solenoid (4), a poloidal field coil (5) and a toroidal field coil (3) made from a material comprising a high temperature superconductor configured to confine a high temperature plasma, the method comprising the steps of: generating a magnetic field having a toroidal component of 10 T or less, preferably 6 T; confining the plasma (2) with a major radius of 3 m or less and a toroidal aspect ratio of 2.2 or less; operating a plasma current of 14 MA or less; emitting fusion alpha particles and neutrons and confining the alpha particles in the plasma (2).

12. The method of claim 11, further comprising maintaining the plasma (2) in a steady state for more than 10 seconds, preferably more than 100 seconds, more preferably more than 1000 seconds.

13. The method of claim 11, comprising generating said neutrons at a rate of at least 1 x 10 17 neutrons / second, preferably at least 5 x 10 18 neutrons / second, by deuterium-tritium fusion at a ratio of 50:50, preferably at least 70:30.

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