CN108269621B - Method for constructing magnetic field configuration of tokamak mixed divertor - Google Patents
Method for constructing magnetic field configuration of tokamak mixed divertor Download PDFInfo
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- CN108269621B CN108269621B CN201611255820.0A CN201611255820A CN108269621B CN 108269621 B CN108269621 B CN 108269621B CN 201611255820 A CN201611255820 A CN 201611255820A CN 108269621 B CN108269621 B CN 108269621B
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21B—FUSION REACTORS
- G21B1/00—Thermonuclear fusion reactors
- G21B1/05—Thermonuclear fusion reactors with magnetic or electric plasma confinement
- G21B1/057—Tokamaks
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- G—PHYSICS
- G21—NUCLEAR PHYSICS; NUCLEAR ENGINEERING
- G21B—FUSION REACTORS
- G21B1/00—Thermonuclear fusion reactors
- G21B1/11—Details
- G21B1/13—First wall; Blanket; Divertor
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- Y—GENERAL TAGGING OF NEW TECHNOLOGICAL DEVELOPMENTS; GENERAL TAGGING OF CROSS-SECTIONAL TECHNOLOGIES SPANNING OVER SEVERAL SECTIONS OF THE IPC; TECHNICAL SUBJECTS COVERED BY FORMER USPC CROSS-REFERENCE ART COLLECTIONS [XRACs] AND DIGESTS
- Y02—TECHNOLOGIES OR APPLICATIONS FOR MITIGATION OR ADAPTATION AGAINST CLIMATE CHANGE
- Y02E—REDUCTION OF GREENHOUSE GAS [GHG] EMISSIONS, RELATED TO ENERGY GENERATION, TRANSMISSION OR DISTRIBUTION
- Y02E30/00—Energy generation of nuclear origin
- Y02E30/10—Nuclear fusion reactors
Abstract
The invention belongs to the technical field of magnetic confinement fusion design, and particularly relates to a method for constructing a magnetic field configuration of a tokamak hybrid divertor. The method for constructing the magnetic field configuration of the Tokamak hybrid divertor comprises the steps of arranging a first polar field coil, a second polar field coil and a third polar field coil on a strong field side, wherein the distances between the geometric center positions of the first polar field coil, the second polar field coil and the third polar field coil and a first X point are respectively 1-1.5 a, 1.5-3 a and 1.5-3 a, and a is the small radius of a plasma. The invention solves the technical problems that the magnetic field configuration of the divertor constructed by the existing method for constructing the magnetic field configuration of the divertor can cause the heating area of the target plate to be smaller, and the heat flow of the target plate faces major challenges under the high-heating operation condition. The magnetic field configuration structure of the hybrid divertor has the capabilities of simultaneously reducing the heat load of the target plates of the inner divertor and the outer divertor and enhancing the particle control of the target plates, and improves the compatibility of the operation of the divertor and the operation of core high-heating plasma.
Description
Technical Field
The invention belongs to the technical field of magnetic confinement fusion design, and particularly relates to a method for constructing a magnetic field configuration of a tokamak hybrid divertor.
Background
The divertor is one of the most important key components of the magnetically confined tokamak device, and is used for removing heat which enters the edge region from the core plasma and flows to the divertor, and simultaneously removing 'helium ash' particles which are cooled down by alpha particles generated by fusion reaction in the core region, so as to ensure the cleanness of the core plasma and maintain the continuity of the fusion reaction. In addition, the divertor also controls the entry of impurity particles generated or actively implanted in the edge region into the main plasma region.
With the increase of the plasma operation parameters of the Tokamak experimental device and the increase of auxiliary heating, in the divertor configuration, the energy flowing out of the separating surface into the scraping layer and the divertor area flows to the divertor along the magnetic lines and is deposited at a small divertor target plate, and experimental and theoretical calculation results show that the radial energy attenuation length of the heat flow at the outer and middle planes is only related to the intensity of the limiting magnetic field of the outer and middle planes, the attenuation length of an International Thermonuclear Experimental Reactor (ITER) is only about 1 millimeter, and the heat flow of the future Kamak fusion device at the outer and middle planes is only about 1 millimeterThe width is also in the order of millimetres, with an ITER case, the total heating power (auxiliary heating and alpha particle heating) is up to 150MW, the power radiated by the core plasma is removed, the energy flowing into the edge is up to 87MW, and if estimated according to an energy decay length of 1mm, the highest thermal load of the ITER target plate will reach 60MW/m even if the magnetic surface at the target plate is widened by a factor of 20 (smaller target plate tilt angle)2Significantly exceeding the current divertor target plate surface material durability and divertor target plate cooling requirements, a significant challenge for target plate design for both the inner and outer divertors to reduce the target plate thermal load to the current engineering acceptable threshold of 10MW/m2The energy allowed to flow into the edge and reach the target plate is only 15MW, about 90% of heating power is required to be radiated before reaching the target plate, or the magnetic surface at the position of the target plate is required to be widened, so that the heat flow of the target plate is effectively dispersed, and the aim of relieving the heat load of the target plate is fulfilled.
In order to alleviate the thermal load of the target plate, the injected impurities inevitably flow into the main plasma region, posing a serious threat to the high parameter operation of the core combustion plasma, especially under higher fusion power and heating power conditions. If the fusion reactor in the future realizes the fusion power of 3GW, the power flowing into the divertor far exceeds the ITER device, too high heat load can also generate more impurities, and the design and the like of the divertor of the fusion reactor are challenged, so that the design and the research of the divertor of the fusion reactor are always the key points.
The divertor configuration is one of the most critical factors in the divertor design and research, and for the problem of heat flow of the divertor target plate, the research of heat flow of the divertor target plate by improving the magnetic field configuration, i.e. the design and research of the advanced divertor, has been internationally carried out, and the first proposal is the CUSP divertor and the X divertor, which add one or two pairs of polar field coils near the target plate of the divertor, and generate an additional X point (i.e. the second X point) near the target plate, and the divertor target plate will cover the second X point of the magnetic field, and the CUSP divertor and the X divertor realize the increase of the heat flow distribution width by means of flow expansion, so as to reduce the heat flow amplitude, the distance between the two coils is relatively close, the wetting area of the target plate cannot be infinitely increased, and the position of the wetting area can be controlled, and besides, the design space for the divertor is insufficient near the target plate of the divertor, the design of the coil also brings many difficulties, especially the divertor configuration design of the superconducting tokamak, and the second X point of the configuration cannot be fully utilized to relieve the heat flow of the target plate.
The super X divertor is an improvement aiming at a CUSP divertor and an X divertor, and through the optimized arrangement of a poloidal field coil, the striking point of a divertor target plate is moved to the outer side as much as possible, and the magnetic surface expansion as much as possible is generated near the target plate, so that the wetting area of plasma is increased. Therefore, the super X divertor configuration can further increase the wetting surface of the target plate, the connection length of the plasma to the divertor target plate and the length of the divertor leg (the length from the first X point to the target plate) by increasing the radial position of the target plate, thereby maximizing the capacity of improving the heat discharge of the divertor. The realization of super X divertor configuration has proposed the extremely high requirement to the engineering design, especially coil design under superconductive tokamak device, secondly inside and outside divertor all designs super X divertor configuration, and the space has received the restriction, if in inside and outside divertor, only one department adopts super X divertor configuration, the higher heat load of opposite side relative to conventional divertor can appear, if improve through realizing the double zero divertor, can't be unanimous often in the face of upper and lower coil current, form the discharge of quasi-double zero, the heat load of its interior target plate is still very high.
The snowflake divertor changes a first-order X point on the original standard divertor into a second-order X point, the second-order X point changes 4 branches of the standard X point into 6 branches, the configuration has a very large extremely low polar field area near the second-order X point, effectively realizes magnetic surface expansion, increases the wetting area of plasma and the connection length from the outermost middle plane to a divertor target plate, and the weak field area can also cause the particle loss enhancement of the area near the X point, through the high polar direction specific pressure area with the polar direction magnetic field close to zero, the plasma can generate strong convection diffusion, then flows along the four legs to the target plate, but the target plate of the snowflake divertor is too close to the main plasma region, although the thermal load can be reduced, the temperature of particles reaching the target plate is high and the particles are dispersed, and effective control of the particles, particularly control of the impurity particle density, is not possible.
Disclosure of Invention
The technical problems to be solved by the invention are as follows: the divertor magnetic field configuration constructed by the existing divertor magnetic field configuration construction method can enable the heating area of the target plate to be smaller, and the heat flow of the target plate faces significant challenges under the high-heating operation condition.
The technical scheme of the invention is as follows:
a method for constructing the magnetic field configuration of a tokamak mixed divertor comprises the following steps:
the method comprises the following steps that a first polar direction field coil, a second polar direction field coil and a third polar direction field coil are arranged on a strong field side, the distances between the geometric center positions of the first polar direction field coil, the second polar direction field coil and the third polar direction field coil and a first X point are respectively 1-1.5 a, 1.5-3 a and 1.5-3 a, wherein a is a small radius of plasma;
the distance between the first polar field coil and the second polar field coil is 0-1.0 a, and the distance between the second polar field coil and the third polar field coil is 0-1.0 a;
a fourth polar field coil, a fifth polar field coil and a sixth polar field coil are arranged on the weak field side, and the distances between the geometric center positions of the fourth polar field coil, the fifth polar field coil and the sixth polar field coil and the first X point are 1-2.5 a, 1.5-2.5 a and 2-3 a respectively;
the distance between the fourth polar field coil and the fifth polar field coil is 0-1.0 a, and the distance between the fifth polar field coil and the sixth polar field coil is 1-2.0 a;
the current direction of the first poloidal field coil is the same as the plasma current direction, the magnitude of the current is 0.2-0.5 Ip, and Ip is plasma current; the current direction of the second polar field coil is opposite to the current direction of the plasma, and the magnitude of the current is 1.0-4.0 Ip; the current direction of the third axial field coil is the same as the current direction of the plasma, and the magnitude of the third axial field coil is 1.0-5.0 Ip; the current direction of the fourth axial field coil is opposite to the current direction of the plasma, and the magnitude of the current is 0.2-2.0 Ip; the current direction of the fifth poloidal field coil is the same as the current direction of the plasma, and the magnitude of the fifth poloidal field coil is 1.0-2.0 Ip; the current direction of the sixth poloidal field coil is opposite to the current direction of the plasma, and the magnitude of the sixth poloidal field coil is 0.1-0.5 Ip.
The invention has the beneficial effects that:
the magnetic field configuration of the mixed divertor constructed by the method of the invention fully combines the structural characteristics of a Tokamak magnetic confinement fusion reactor device and the requirements of the magnetic field structure of the divertor and the physical operation of the divertor, a plasma balance configuration of a second X point is generated at one side of a scraping layer near a target plate of the inner divertor, a snowflake reduction (left) divertor configuration is formed by optimizing the distance between the second X point and the first X point, the characteristics of a snowflake divertor are possessed, the plasma wetting area and the connection length of the inner target plate are enhanced, and the goals of reducing the thermal load of the target plate of the inner divertor and improving the particle control are realized under the condition of no excessive space; meanwhile, under the condition of not being influenced by the second X point of the inner divertor, in the outer divertor area, at the position which is as far as possible from the first X point and is radially far from the outer divertor target plate with a larger radius, a balance configuration of the third X point is formed, so that the outer divertor also has the characteristics of larger target plate magnetic surface expansion coefficient, longer divertor leg length, longer magnetic line connection length and the like, has the characteristics of a super X divertor, and meets the requirements of reducing the heat load of the inner divertor target plate and enhancing the particle control of the target plate.
The magnetic field configuration structure of the hybrid divertor has the capabilities of simultaneously reducing the heat load of the target plates of the inner divertor and the outer divertor and enhancing the particle control of the target plates, and improves the compatibility of the operation of the divertor and the operation of core high-heating plasma.
Drawings
FIG. 1 is a schematic diagram of the distribution of balanced potential poloidal field coils of a hybrid divertor;
FIG. 2 is a schematic diagram of the equilibrium configuration of a hybrid divertor;
FIG. 3 is a schematic illustration of a target plate in a hybrid divertor configuration;
FIG. 4 is a schematic thermal load profile of a target plate in a hybrid divertor;
FIG. 5 is a schematic thermal load profile of an outer target plate of the hybrid divertor;
wherein, 1-the first polar field coil, 2-the second polar field coil, 3-the third polar field coil, 4-the fourth polar field coil, 5-the fifth polar field coil, 6-the sixth polar field coil.
Detailed Description
The following describes a method for constructing a magnetic field configuration of a tokamak hybrid divertor according to the present invention in detail with reference to the accompanying drawings and examples.
The position shape of the hybrid divertor mainly combines the structural characteristics of a Tokamak magnetic confinement fusion reactor device, namely the characteristics that the space on the strong field side is compact and the space on the weak field side is relatively abundant, the plasma balance position shape of a second X point is generated on one side of a scraping layer near a target plate of the inner divertor by the re-optimization distribution of a polar field coil of the device, the position shape of the (left) divertor with snowflakes is formed by optimizing the distance between the second X point and the first X point, the plasma wetting area and the connection length of the inner target plate are enhanced, and the targets of reducing the heat load of the target plate of the inner divertor and improving the particle control are realized under the limited space condition of the inner divertor; meanwhile, under the condition of not being influenced by the second X point of the inner divertor, in the outer divertor area, at the position which is as far as possible from the first X point and is radially far from the outer divertor target plate with a larger radius, the balance configuration of the third X point is formed, so that the outer divertor also has the characteristics of larger target plate magnetic surface expansion coefficient, longer divertor leg length, longer magnetic line connection length and the like, and has the characteristic of a super X divertor. Therefore, the design requirements of reducing the heat load of the target plates of the inner and outer divertors and enhancing the particle control of the target plates are met.
FIG. 1 shows a distribution diagram of a hybrid divertor balanced configuration coil, combining the elongation ratio of the configuration of the divertor and the X point position determined by the triangle variation parameter, on the side of strong field, the distances between the geometric center positions of the first, second and third poloidal field coils 1, 2 and 3 and the first X point are 1-1.5 a, 1.5-3 a and 1.5-3 a, respectively, where a is the small radius of the plasma. The distance between the first polar field coil 1 and the second polar field coil 2 is 0-1.0 a, and the distance between the second polar field coil 2 and the third polar field coil 3 is 0-1.0 a. The space of the outer divertor is relatively rich with respect to the inner divertor space, and on the weak field side, the geometric center positions of the fourth polar field coil 4, the fifth polar field coil 5, and the sixth polar field coil 6 are at distances of 1 to 2.5a, 1.5 to 2.5a, and 2 to 3a from the first X point, respectively, the distance between the fourth polar field coil 4 and the fifth polar field coil 5 is 0 to 1.0a, and the distance between the fifth polar field coil 5 and the sixth polar field coil 6 is 1 to 2.0 a.
The current magnitude of the first poloidal field coil 1, the second poloidal field coil 2 and the third poloidal field coil 3 is closely related to the distance between the coils and the center of the main plasma and the position of a required second X point, the current direction of the first poloidal field coil 1 is the same as the plasma current direction, the magnitude is 0.2-0.5 Ip, and Ip is plasma current. The current direction of the second polar field coil 2 is opposite to the current direction of the plasma, and the magnitude of the current is 1.0-4.0 Ip. The current direction of the third axial field coil 3 is the same as the current direction of the plasma, and the magnitude of the current direction is 1.0-5.0 Ip. The current direction of the fourth axial field coil 4 is opposite to the current direction of the plasma, and the magnitude of the current is 0.2-2.0 Ip. The current direction of the fifth poloidal field coil 5 is the same as the plasma current direction, and the magnitude of the current is 1.0-2.0 Ip. The current direction of the sixth polar direction field coil 6 is opposite to the current direction of the plasma, and the magnitude of the current is 0.1-0.5 Ip.
The size of the coil is mainly determined by the poloidal coil current required for balancing the configuration, the current density limit borne by the coil and the spatial position of the coil.
In the case of a mixed divertor configuration designed as a double zero configuration, the upper and lower coil configurations can be symmetrically distributed, and in the case of a mixed divertor configuration with a single lower zero, the lower coil configuration shown in FIG. 1 can be used. Setting current parameters of the coils according to coil distribution and coil turns and parameters such as plasma current magnitude, specific pressure, internal inductance and the like, calculating a balance configuration under a free boundary condition, solving a Grad-Shafranov equation by a program, continuously debugging the current value of each coil, optimizing the parameters of the balance configuration and the magnetic surface positions of the second and third X points, and forming the balance configuration of the mixed bias filter configuration.
FIG. 2 is a schematic diagram showing the position of the hybrid divertor, wherein a second X point is formed on the separating surface side of the inner divertor region, and the distance between the second X point and the first X point is 0-0.4 a, so that the inner divertor region has a larger weak polar field area and forms a snowflake-reduced (left) position; for the outer divertor, a third X point is formed on the separating surface side at the first X point, the outer divertor region, the third X point being at a distance greater than a from the first X point, and the third X point being at a position R in the radial direction>(R0+ a) wherein R0The plasma has a large radius, so that the plasma has large divertor leg length, connection length and large magnetic surface expansion to form a super X outer divertor configuration.
In order to increase the effective heating area of the target plate and maximally reduce the heat load of the target plate, as shown in fig. 3, the distance between the second X-point magnetic surface of the inner divertor and the magnetic surface of the separating surface on the outer midplane should be less than the width of the heat flow channel, the heat flow is mainly concentrated within the heat flow width and reaches the divertor area, the divertor target plate is located near the second X-point, so that the included angle between the magnetic force line of the magnetic field and the divertor target plate is not less than 1 degree, thereby increasing the area of the target plate bearing the heat flow, efficiently reducing the heat load of the target plate, and particularly the highest heat load at the divertor target plate.
In order to reduce the heat flow of the inner target and effectively reduce the heat load of the outer divertor target plate, as shown in fig. 3, the distance between the third X point magnetic surface and the magnetic surface of the separating surface on the outer midplane should be less than the width of the heat flow channel, the heat flow is mainly concentrated within the heat flow width, the divertor target plate is located near the third X point, so that the angle between the magnetic lines of force of the magnetic field and the divertor target plate is not less than 1 degree, thus, from the leg length of the first X point and the connection length of the magnetic lines of force, the enlarged target plate area bearing the heat flow is also provided, and the high-efficiency reduction of the heat load of the target plate, especially the highest heat load at the divertor target plate, is realized.
The heat load profile of the target plate in the hybrid divertor shown in FIG. 4, for the same core boundary conditions, for the conventional divertor, the highest heat load is located in the region close to the separation plane and concentrated in a very small target plate area, for the snowflake filter configuration, the highest heat load of the target plate is less than 30% of that of the conventional divertor, regardless of the influence of the second X point on the radial expansion transport, the profile exhibits a flatter profile, the highest heat load being located on the inner target plate at a distance of about 20cm from the separation plane.
The heat load profile of the outer target plate of the hybrid divertor as shown in fig. 5, for the conventional divertor, most of the energy will be deposited in a very small target plate area near the separation plane, resulting in a very high target plate heat load, for the outer super X divertor, especially, a large divertor leg length and connection length, effectively reducing the particle temperature reaching the target plate, increasing the particle density of the target plate, especially the recombination loss of particles (in the case of impurities, not only the radiant energy loss can be increased, but also the impurities can be effectively prevented from entering the main plasma region), easily resulting in the off-target operation of the outer divertor, achieving effective control of the particles, the maximum heat load of the target plate being less than 10% of the conventional divertor, the profile exhibiting a flatter distribution, the temperature of the particles also being reduced to the case of only a few electron volts (much lower than the particle temperature under the conventional divertor), off-target of the divertor is achieved.
Claims (1)
1. A method for constructing the magnetic field configuration of a tokamak mixed divertor is characterized by comprising the following steps of:
a first polar direction field coil (1), a second polar direction field coil (2) and a third polar direction field coil (3) are arranged on the strong field side, the distances between the geometric center positions of the first polar direction field coil (1), the second polar direction field coil (2) and the third polar direction field coil (3) and a first X point are respectively 1-1.5 a, 1.5-3 a and 1.5-3 a, wherein a is a small radius of plasma;
the distance between the first polar field coil (1) and the second polar field coil (2) is 0-1.0 a, and the distance between the second polar field coil (2) and the third polar field coil (3) is 0-1.0 a;
a fourth polar field coil (4), a fifth polar field coil (5) and a sixth polar field coil (6) are arranged on the weak field side, and the distances between the geometric center positions of the fourth polar field coil (4), the fifth polar field coil (5) and the sixth polar field coil (6) and the first X point are respectively 1-2.5 a, 1.5-2.5 a and 2-3 a;
the distance between the fourth polar field coil (4) and the fifth polar field coil (5) is 0-1.0 a, and the distance between the fifth polar field coil (5) and the sixth polar field coil (6) is 1-2.0 a;
the current direction of the first poloidal field coil (1) is the same as the plasma current direction, the magnitude of the current is 0.2-0.5 Ip, and Ip is plasma current; the current direction of the second polar field coil (2) is opposite to the current direction of the plasma, and the magnitude of the current is 1.0-4.0 Ip; the current direction of the third axial field coil (3) is the same as the current direction of the plasma, and the magnitude of the third axial field coil is 1.0-5.0 Ip; the current direction of the fourth axial field coil (4) is opposite to the current direction of the plasma, and the magnitude of the current is 0.2-2.0 Ip; the current direction of the fifth poloidal field coil (5) is the same as the current direction of the plasma, and the magnitude of the current direction is 1.0-2.0 Ip; the current direction of the sixth poloidal field coil (6) is opposite to the current direction of the plasma, and the magnitude of the sixth poloidal field coil is 0.1-0.5 Ip.
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| CN112992386B (en) * | 2019-12-12 | 2022-07-26 | 核工业西南物理研究院 | Method for constructing magnetic field configuration of tokamak reverse triangular divertor |
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