Increasing energy gain in magnetically confined plasma by increasing edge temperature: super-XT divertor

Abstract

An annular confinement plasma vessel having a substantially symmetric magnetically confinement plasma region wherein the plurality of magnetic field coils are configured to provide at least one X-point and direct plasma particles from the magnetically confinement region to the divertor target; and wherein the total magnetic field strength (including all components of the magnetic field) at the divertor target is lower than the total magnetic field strength (including all components of the magnetic field) at a location between the divertor target and the X point on the last closed flux surface of the SOL closest to the divertor target. When the mean free path of the neurons is longer than the width of the SOL, the two critical functions can be separated: a) withstanding high heat flux, and b) pumping plasma particles to maintain low density.

Description

CROSS-REFERENCE TO RELATED APPLICATIONS

[0001] This application claims priority to U.S. Serial No. 18 / 102,698, filed on January 28, 2023, and Provisional Application Serial No. 63 / 304,310, filed on January 28, 2022. Technical Field

[0002] The present invention generally relates to magnetically confined plasmas for producing nuclear fusion reactions. Background Art

[0003] The magnetic confinement (MC) plasma must maintain temperatures on the order of 100 million degrees. This, for temperatures used in the art, is in units of about 10 kiloelectron volts (keV). These temperatures are necessary to produce intense thermonuclear fusion reactions. This is required for fusion to become a useful energy source, and for this, the fusion energy produced must be many times greater than the energy required to keep the MC plasma hot. The ratio of energy produced to energy input is called the fusion gain. (This quotient is usually denoted by Q.)

[0004] As is well known in the art, energy gain depends strongly on the so-called energy confinement time - typically the time that a MC plasma will lose its high temperature state. High energy-confinement time is crucial for fusion devices. And even for fusion devices where energy gain is not the primary goal, e.g., fusion neutrons are used for transmutation, high confinement makes the device much cheaper to operate (by reducing expensive energy input).

[0005] Many consider the MC concept to be the leading concept for achieving energy gains greater than one. Getting energy gains is not easy: after decades of effort and billions of dollars in many different countries, energy confinement is still too small for the energy gain. Two major magnetic confinement experiments are being built to achieve Q>1 in the near future - ITER (an international effort by governments), and SPARC (through the private company Commonwealth Fusion Systems). However, both of these devices are large and expensive, and their likelihood of achieving sufficient energy confinement for high energy gains is far from certain. Overview of the Prior Art

[0006] The present invention is a device that greatly improves the energy confinement of a wide variety of MC plasmas, making high energy gains possible. Equally important, our invention will enable high fusion gains to be achieved in much smaller and cheaper equipment.

[0007] Devices using MC plasmas use the following principle, which is well known in the art. A strong magnetic field guides the hot particles so that they travel in circles within an annular region without hitting the material walls. This prevents the plasma from rapidly losing energy to the walls. It also prevents the walls from being destroyed by the thermonuclear plasma, which must be much hotter than any material can tolerate for some time.

[0008] In the art, the understanding of the process of controlling confinement has been greatly improved over the years. The applicant of this patent application has published dozens of peer-reviewed scientific papers on this topic and has been a core participant in establishing a scientific understanding of this topic. Based on this understanding, if the temperature at the edge of the MC region is increased, the energy confinement in the MC region will be greatly improved. If the density at its edge is reduced relative to the core (under appropriate conditions), it will also be improved. Lower plasma edge density often occurs at the same time as higher edge temperature. Low density is almost as important as high temperature in order to improve confinement. For our purposes here, it is intuitively obvious that increasing the temperature at the edge of the MC region will make it easier to achieve high temperature at its center, which is enough. A deeper scientific understanding reveals that this is more strongly true than simple intuition suggests: even a moderate increase in boundary temperature can reduce energy losses by a disproportionate amount. And lower edge density can also reduce energy losses to an astonishing degree. This is the way our invention is used to improve energy confinement.

[0009] However, when the temperature of the MC region boundary is high and the edge density is low, several major and potentially disqualifying problems can arise. These problems ultimately come from the interaction of the plasma with the wall, which the plasma must eventually contact. The present invention is a method of creating a high temperature boundary and a low density boundary at the same time and overcoming the problem of plasma interaction with the wall. Therefore, the present invention ensures that the energy confinement of the MC region is greatly improved in a feasible, smaller device. The applicant of this patent application has also published many peer-reviewed scientific papers that are related to the interaction of plasma with the wall, especially the so-called divertor, where the interaction of the plasma with the wall is most intense.

[0010] Let's focus our attention on this region near the edge. As is well known in the art, the MC region has a finite extent. This means that the region where the particles go in a circle without hitting the wall has a finite extent. In the art, the boundary of this region is called the "Last Closed Flux Surface," or LCFS. Outside the LCFS, the magnetic field guides the particles to intersect the wall at a fairly short distance. This is very different from the inner region. Introducing another term from the art: the plasma in the region outside the LCFS is called the scraped layer (SOL). And there is another term here too: the region near where the plasma interacts with the wall is called the divertor. And another term is that the location on the wall where the plasma hits the material is called the divertor target. Figure 1 shows a representative geometry from the prior art, including the location of the terms used here.

[0011] One of the applicants for this patent application is one of the inventors listed in a previously issued patent, and the issued patent incorporates a divertor they invented, known as a Super-X divertor. Figure 1BThe actual geometry of the tokamak MAST-U experiment is shown in FIG. MAST-U is the main experiment in the UK, which is specifically designed to test the Super-X divertor. Note that the divertor target is in a large chamber. The function of the chamber is to capture neutral atoms and molecules (called neutral particles) so that they collide repeatedly and interact repeatedly with the plasma, as well as repeatedly interact with the walls. The interaction of these neutral particles removes energy from the plasma and transfers energy from the neutral particles to the walls, making the neutral particles have low energy, and dispersing the energy to the walls of the chamber. All these interactions work together to minimize damage to the walls. This is an excellent strategy for protecting the walls, but it comes with a high cost: the SOL has a relatively high density and a relatively low temperature. This gives the core MC plasma boundary conditions that are far from its optimal confinement conditions. (Recall that the SOL near the MC plasma is also essentially the boundary of the MC plasma.) Compared with the boundary conditions of high temperature and low density, low confinement will reduce the fusion gain of the MC plasma. The divertor disclosed in this patent adopts a very different strategy. It aims to create a high temperature, low density SOL, and thus create much higher confinement for MC plasmas. We call the divertor that implements this strategy the Super-XT divertor. In order to achieve the desired high temperature and low density SOL conditions, the interaction of neutral particles with the plasma must be reduced. When there is a high temperature plasma, the neutral particles also have high energy, and they damage the wall instead of protecting it. Therefore, in order to avoid damaging the wall (and producing impurities that contaminate the plasma due to wall sputtering), the interaction of neutral particles with the wall must also be reduced. This is the reverse strategy of the Super-X divertor in the MAST-U device. And therefore, making the plasma compatible with the wall (i.e. avoiding damaging the wall or contaminating the MC plasma) is necessary for a very different strategy. The fact that "Super-XT" has only one additional letter does not mean that it is obvious how to use the Super-X and adapt it to a completely opposite or reverse strategy - far from it. One certainly cannot carry over any previous element of the Super-X divertor to the Super-XT without a careful and non-obvious analysis of whether it conforms to a completely new strategy. And while this analysis reveals that some aspects of the Super-X concept are useful for new inventions, many other new aspects are needed, which we describe below.

[0012] In order to understand the ingenuity of the present invention, we first describe and discuss a series of problems from SOL plasma-wall interaction. This will help us understand: 1) why existing technologies have not succeeded in achieving high energy confinement by setting a high temperature boundary for the MC region, and 2) why and how the present invention will solve these problems and open an effective path for fusion.

[0013] Heat flow is one of the fundamental aspects of MC plasmas, and we start here. As in basic thermodynamics, heat goes from hot to cold. Starting in the center of the MC region, the heat moves slowly toward the colder boundary (LCFS). The heat then continues its journey through the SOL plasma toward the much colder walls. On its journey within the SOL, the heat is carried by the particles. By moving along the magnetic field, the SOL plasma particles rush to the walls as fast as possible. The characteristic time of this "travel" is very short: it is several orders of magnitude shorter than the time it takes for the heat to traverse the MC region.

[0014] As expected, the rapid departure of heat from the SOL typically causes the SOL to be much colder than the center of the MC region. The SOL is at the boundary of the MC region, and when the SOL gets cold, the boundary gets colder too. And as we mentioned before, cold boundaries also cause heat to leak out of the bulk region of the MC significantly faster.

[0015] The problem that these very basic considerations lead to is that, in the standard approach in the art, large and expensive equipment is required to obtain high confinement and large energy gains; ITER and SPARC are examples.

[0016] If the temperature of the LCFS can be raised, the confinement of the MC region will be greatly improved. Reducing the density of the LCFS relative to the MC region as a whole can also give considerable improvements to its confinement.

[0017] It is difficult to increase the temperature of the SOL because it is in contact with the relatively cold walls. But there is a way to have high temperatures in the SOL - by greatly reducing the density of the SOL. In its journey through the SOL, the heat will be carried to the wall by very few particles, and therefore, each particle must carry a lot of energy. This is equivalent to saying that the temperature of the SOL is very high.

[0018] Since the SOL and LCFC are adjacent, the strategy in the SOL also combines the two strategic elements required to increase the MC plasma confinement - increasing the temperature of the LCFS and reducing its density.

[0019] Despite these advantages of MC plasmas, SOL plasmas with their high temperature and low density present other serious challenges due to their interaction with the walls. The problems arise primarily in the divertor region where the interaction is most intense. As we will see, some aspects of the Super-X divertor are indeed key elements of the present invention. But a number of new aspects are also required.

[0020] To understand the need for new inventions, we must first describe all of the new and potentially disqualifying problems that arise from high temperature, low density SOLs.

[0021] The wall will be unacceptably damaged in three ways: 1) the heat flux will melt the wall struck by the SOL, 2) the high energy particles in the SOL will knock atoms off the wall surface, unacceptably eroding the wall, and 3) this eroded material will also be redeposited elsewhere in the machine, causing unacceptable accumulation of material where it does not belong, and thus causing failure.

[0022] Before describing in more detail, we first introduce some other standard terms in the field. The material wall against which most of the heat strikes is called the divertor target. If standard techniques are applied to a high temperature and low density SOL, the heat flux will likely melt through the wall, even with the best cooling practical within the standard techniques. This may take only seconds. If this problem can be avoided in some way, then on longer time scales, erosion and redeposition processes will destroy the components facing the plasma. This will take on the order of days, which is too short a time to simply replace such a component. Replacement of large solid components is estimated to take several months for analysis of fusion designs in the field.

[0023] The interaction of high temperature SOL with the wall will cause more damage than the wall: as mentioned above, it will also unacceptably degrade the MC plasma by contamination with impurities. This occurs because the atoms of the wall are knocked out by the high temperature SOL plasma. This is physically unavoidable because these atoms are confined to materials with an intensity of the order of eV, but high energy SOL particles have energies of a higher order. Therefore, the collision of SOL particles with the wall leads to the well-known process of "sputtering": knocking some of the atoms of the wall out of the surface. These unacceptable numbers of sputtered wall atoms can reach the MC plasma through two possible pathways. One is that the atoms are ionized in the SOL plasma. Then, they are affected by the magnetic field and travel through the SOL to the MC boundary, where they will be absorbed. Another pathway will be that some of the sputtered particles are not ionized, and remain neutral, and travel to the MC region along a straight line that is not affected by the magnetic field. After reaching the MC plasma, they will be ionized and absorbed.

[0024] It is well known in the art that once impurities are absorbed into the MC plasma, they may cause unacceptable results. One is that the impurities will greatly dilute the energy producing fuel in the MC region and thereby greatly reduce energy production. Even more seriously, the impurities can cause high energy X-ray photon radiation to be generated after being bombarded by high energy electrons in the plasma. (The physical process is similar to that in ordinary medical X-ray tubes, where materials are bombarded with electrons to produce X-radiation.) The energy of the MC plasma will be converted into these X-ray photons. The energy loss caused by this radiation can be very large, even with only a small impurity fraction, and this is known to lead to an overall collapse of the MC plasma temperature.

[0025] Another way impurities can be generated is by simply overheating the divertor target, making the normal evaporation amount larger.Once they are vaporized, atoms can enter the MC plasma through the same two pathways.

[0026] By the above process, impurities will build up to unacceptable levels within a matter of seconds, which is too fast for practical devices.

[0027] One particular impurity is special because it is created by the fusion reactions within the MC plasma itself: so-called helium "ash," which is a byproduct of the fusion reactions. This ash must be removed from the plasma as quickly as possible when it is created, or it will accumulate in the plasma and dilute the fusion fuel, ultimately substantially reducing the fusion rate. This accumulation is sometimes referred to in the art as stalling the fusion reaction due to the helium ash.

[0028] In summary, five serious problems arise when the hot, low-density SOL plasma interacts with the divertor target, which we call the "SOL problem": 1) Unacceptable heat flux on the wall, resulting in unacceptable melting or evaporation of the wall 2) Unacceptable impurity accumulation in MC plasma due to sputtering and evaporation 3) Unacceptable erosion of the divertor target by sputtering or evaporation 4) Unacceptable redeposition of eroded material in Problem #3 5) Unacceptable helium exhaust

[0029] In the following, we will continue to use the name "SOL problem" as an abbreviation for the five problems listed above.

[0030] It is important to realize that if any of these issues are not addressed in a device, then the device will not be a practical energy source (or neutron source).

[0031] However, it is also important to distinguish between two different types of equipment for MC plasma in the commercial field. Some of the above issues are more strongly applicable to a certain type compared to another type. The first type of equipment has a high duty cycle, that is, they operate in a significant time partition. Equipment for producing useful energy gain or neutron production belongs to this type. The second type of equipment has a low duty cycle. They can be operated with pulses in the range of a few seconds or minutes, and then not operated for a long period of time. Equipment with a low duty cycle is usually research equipment, which is a prototype for developing the first type of equipment. SPARC built by Commonwealth Fusion Systems and ST40 currently operated by Tokamak Energy Co., Ltd. are examples of some but not all commercial equipment with short duty cycles. And ITER is a research device built according to an international agreement between governments.

[0032] All the SOL issues mentioned above will have serious negative effects on the first type of devices (i.e., high duty cycle devices). However, low duty cycle devices are mainly affected by the first two SOL issues and may be affected by the fifth SOL issue.

[0033] The present invention avoids all of these problems.Before we present its detailed working, we will summarize the prior art attempts to solve the SOL problem; the need for the present invention will become abundantly clear.

[0034] By far, the most common approach to solving the SOL problem in existing technologies has been to move away from SOLs with high temperatures and low densities, as the associated problems are considered too daunting to solve. In fact, both ITER and SPARC have chosen the opposite approach—creating SOLs with as low temperatures and as high densities as practical.

[0035] It is generally believed in the art that operating in this conventional manner usually results in a certain degree of constraint degradation. In the conventional approach, SOL is not seen as a possible means by which better core performance can be designed, but rather as a source of intractable problems, so that people are willing to tolerate some degradation of core performance. This is very different from the approach in the present invention. Let us consider how five SOL problems are solved in the conventional way.

[0036] The high heat flux of the divertor target (problem #1) has been extensively analyzed and is recognized as a serious problem. One element of the solution in the conventional approach is to use highly refractory metals (e.g., tungsten or molybdenum) as divertor targets. But this is recognized to be insufficient by itself. The conventional solution must be enhanced by creating conditions - maintaining the lowest possible plasma temperature near the divertor target, e.g., about 10 eV or less - to disperse the energy in the SOL plasma through enhanced radiation near the target. These temperatures are significantly lower than the KeV range contemplated by the present invention.

[0037] Since such high levels of radiation are not possible with the high temperature, low density SOL of the present invention, managing heat flux by other means becomes a key requirement for success.

[0038] Because of the low temperature, the standard method greatly reduces sputtering. Impurity accumulation in the MC plasma (problem #2) is also greatly reduced.

[0039] Such low temperatures also greatly reduce erosion of the divertor target due to sputtering (problem #3). We estimate how problematic it would be to operate a high temperature, low density SOL in a conventional divertor. For typical parameters, it would result in erosion of the divertor target to a depth of millimeters per day. The solid divertor target can only be about a centimeter thick at most to conduct the heat adequately. Therefore, the target would need to be replaced about every week, but such replacement would take several months, so this is unacceptable.

[0040] In addition to erosion of the divertor target, the eroded material will also be redeposited in very undesirable places (Problem #4). As is known in the art, much of this redeposition will likely be in the form of dust. This can contaminate the plasma by falling off the walls after accumulating in the plasma for a period of time. This process has been observed in operating equipment (such as large spiral equipment) where dust impacts on the MC plasma have resulted in high contamination, causing the plasma to terminate. In addition, dust can accumulate in locations including plasma diagnostic ports, RF antennas for heating and current drive, etc. This will ultimately lead to failure of these critical systems. It will also create safety hazards in fusion devices where the dust will be radioactive and fine dust can be dispersed through the atmosphere in an accident.

[0041] Dust is already recognized as a serious problem in conventional solutions in the art. Therefore, it would be unacceptable to make it worse by operating with high temperature, low density SOL. New aspects of enhancement are needed.

[0042] Finally, consider helium removal, SOL problem #5. Helium is a special impurity that is different from other impurities because it is continuously generated inside the MC plasma by fusion reactions. As is generally recognized in the art, SOL conditions are a key factor in how to effectively remove helium ash from the system. This removal is absolutely necessary: ​​if helium is removed effectively, it will accumulate in the MC plasma and stop the fusion reactions. Under typical conditions, this will occur in just a few minutes. This is obviously too short for a viable energy production device.

[0043] In the standard scheme, helium removal is achieved by pumping helium out of the plasma chamber as a gas. This pumping is difficult because helium is chemically inert and has an extremely low boiling point. It is well known in the art that there are relatively few types of pumps suitable for use in magnetic fusion environments. Only two types use mature technology. One type is called a cryogenic pump, which is used in current MC experimental equipment and will also be used in ITER to remove helium ash. Another type is a diffusion pump, which may use mercury as the working fluid. This pump also requires a quasi-cryogenic "cold trap". For simplicity, we can say that all these pumps need to operate in the low temperature region.

[0044] In the standard scheme, the steps for removing helium are as follows. Helium is generated in the MC plasma, enters the SOL in an ionized state, and collides with the remaining SOL particles into the divertor target. In the usual cycle, helium neutral gas is generated at the divertor target after the plasma particles hit the divertor target. The gas is transported through a tube to a pump with a low temperature area outside the neutron shield. Such pumps must be located in a very low neutron flux area because they have a low temperature area, making neutron heating unacceptable. Neutron heating is extremely intense near the MC plasma, so the pump must be located behind a fairly large neutron shield, and therefore, must be quite far away.

[0045] Of course, the pump can only effectively remove helium if it can be effectively transported through the tube from the area where the helium is generated (the divertor target) to the location of the pump. In the standard scheme, the operating scheme with a high SOL density greatly facilitates this key step by the following method. The high SOL density directly leads to the neutral gas also having a high particle density, both hydrogen and helium, and correspondingly, the mean free path is smaller than the typical size of the tube. In the terminology of the art, this neutral gas is in a low Knudsen number condition. (The Knudsen number is a standard dimensionless fluid parameter that is equal to the ratio of the mean free path of particle collisions to the length related to the system characteristics.) It is known that this greatly facilitates transportation through the tube. From a conceptual point of view, frequent collisions mean that all particles move together, as a large group of directed motions, and the tube wall is not a huge obstacle to transportation through the tube.

[0046] If the neutral gas density is low, it is recognized in the art that the transport of helium through the tube will be greatly reduced, making pumping insufficient. A low-density SOL will lead to exactly this unacceptable situation. Low density reduces the throughput of particles, both directly affected by density and indirectly affected by the greatly increased mean free path. As described in the art, this is a high Knudsen number flow. As is well known, the conductivity of a long tube is greatly reduced when the Knudsen number is high because the particles bounce off the wall and are randomized before reaching the end of the tube - the particles no longer move together in large numbers (the same as in the case of a small mean free path). In a fusion reactor, this situation cannot be compensated simply by increasing the cross-sectional area of ​​the tube, because this will also greatly increase the transport of neutrons down the tube, causing multiple unacceptable consequences: unacceptable heating of the cryogenic area required for the pump, radiation activation and neutron damage to the components of the pump and its surroundings, and reduced tritium breeding in a deuterium-tritium fusion reactor.

[0047] Thus, we see that for high temperature, low density SOL, all SOL problems become highly problematic in conventional techniques. We can easily understand why these difficult SOL problems lead conventional approaches to avoid high temperature, low density edges.

[0048] Avoiding high temperature, low density conditions, while reasonable in conventional fashion, is associated with high costs.

[0049] As emphasized before, the conventional approach (with cryogenic rim) places the entire burden of achieving high confinement on the MC core region; energy confinement and energy gain are thereby reduced. The equipment for achieving high fusion gain would then need to be very large in size, or high magnetic fields with expensive and huge coils. In any case, the cost of the equipment would be much higher than that of adopting our invention.

[0050] The present invention provides a solution to the SOL problem caused by plasma-wall interaction in high temperature, low density SOL.

[0051] This alternative (to the conventional approach) has been described in the prior art. Although one of the few attempts in this area, this one attempts to create conditions closest to those contemplated by this patent, using lithium to reduce the density of the SOL by chemically absorbing hydrogen-containing species in the plasma as they strike the divertor target (or other surface in contact with the SOL). By hydrogen, we shall mean any of the isotopes of hydrogen: protium, deuterium or tritium.

[0052] However, in the prior art, there is no satisfactory solution to the SOL problem that arises when lithium is used to reduce density. We will consider each of the five SOL problems in the context of lithium in the prior art:

[0053] Problem #1: Very high heat fluxes are generated. No satisfactory solution has been found in the prior art. The huge heat fluxes drive the surface temperature of the lithium beyond a tolerable limit, which is quite low. Some estimate this limit to be 400-450 degrees Celsius (Castro 2021), but others estimate it to be 300-380 degrees Celsius (Kessel 2019). Exceeding this temperature is accompanied by several undesirable effects: high evaporation, sputtering depending on the high temperature, and reduction of the critical chemical absorption of hydrogen. Many solutions have been proposed to maintain a tolerable temperature on the lithium surface, but none of them have proven satisfactory. The most discussed proposal is to flow the lithium through the high heat flux region at a very fast rate, so that the surface heating is limited by the instantaneous exposure. But lithium is a highly conductive metal; its rapid flow in a strong magnetic field is quite difficult. It is well known that it is extremely difficult to move a highly conductive fluid rapidly in a high magnetic field region due to rapid magnetic flow damping. Therefore, attempts to achieve suitable rapid flow with the high reliability required to continuously handle very high heat fluxes have not been successful. An additional problem is that the radioactive tritium deposited in the lithium must be removed. For fast flowing streams, very large amounts of lithium must be processed to remove the very weak concentration of tritium, which is not trivial.

[0054] Alternatively, it is very difficult to keep the surface of the lithium at an acceptable temperature by cooling it from behind. There are several reasons for this, including: a) the low temperature that the lithium surface can withstand b) water cooling, although relatively effective, cannot be used near lithium. Lithium burns violently in contact with water, and an accidental leak can result in a large release of radioactive tritium bound to the lithium, c) materials that are otherwise good for removing heat, such as copper or aluminum, cannot be used in a fusion environment. These materials are strongly degraded by neutrons or corroded by lithium in a fusion environment.

[0055] It is conceivable that some future flowing liquid surface approach could be used to remove heat from the lithium divertor target. We will refer to this hypothetical technology as the FLSM. It is important to recognize that even if the FLSM were somehow implemented, it would not solve all five SOL problems, but would only solve problems #1 and #3, and possibly problems #2 and #4 if other disadvantages were accepted. We will describe this below. But because of problem #5, the present invention would still be required for a working fusion reactor. And the problems imposed by problems #2 and #4 can be so onerous that the present invention would be a substantial improvement. The hypothetical FLSM would operate only as a special example of many possibilities on a divertor target, some of which have their surfaces covered by liquid. But as we describe below, and as indicated in the claims, even with the hypothetical FLSM, the present invention would still be beneficial, and in some cases absolutely necessary.

[0056] Problem #2: The low density edge in the MC plasma leads to a very strong tendency for unacceptable accumulation of impurities in the plasma. This is particularly well demonstrated in tokamak experiments with the highest energy confinement, which are known in the art as "H-modes". Extensive analysis shows that the physics known as neoclassical transport leads to a large increase in the impurity concentration in the MC core from the values ​​in the SOL. Detailed quantitative comparisons with experiments show that this physics works. The most careful analysis of the experimental data shows that "the results indicate that turbulent transport is of negligible importance for impurities over the edge transport barrier and, therefore, transport reaches the neoclassical level" [Putrich 2011]. Therefore, we can use neoclassical transport as a good guide to understand how impurities in the SOL enter the MC plasma. In the current experiments, this increased the high Z impurity concentration in the MC region by tens or even nearly a hundred times compared to the concentration in the SOL region. According to neoclassical dynamics, when the SOL density is reduced to the low values ​​envisioned using lithium, the tendency of impurities to concentrate in the core will become much stronger. For the very low SOL densities under consideration, even reasonably low levels of impurities within the SOL will accumulate in the MC region, albeit slowly, but eventually reaching unacceptable levels.

[0057] The most dangerous impurities are those with high Z. These impurities will come from the structural material of the MC plasma chamber walls (unless the walls are coated with Li or Be or some other low Z material) if there are high energy particles incident on the chamber walls of the MC plasma. But even low Z elements can accumulate to severely dilute the fuel in the MC plasma.

[0058] In addition to impurities passing through the SOL to reach the plasma, they can also reach the plasma in other ways. One important way is that neutral impurities sputtered from the divertor target can enter the MC region along a path that is not affected by the magnetic field and be ionized, thus becoming plasma impurities. When they are ionized near the edge, the neoclassical dynamics indicated above will strongly concentrate them in the MC plasma core.

[0059] Another way impurities can reach the plasma is through droplets or solid flakes dripping into the plasma from the walls. These occur due to large amounts of eroded material from the divertor targets, which will accumulate over time throughout the machine. As these accumulations become thicker, material eventually falls into the plasma.

[0060] Neutrons emitted from the SOL plasma near the divertor target (so-called charge exchange hydrogen neutrons) will travel in straight lines and sputter onto other materials inside the machine, which are usually higher Z. These impurities can enter the MC plasma, or add to the redeposited materials mentioned above.

[0061] Lithium has the lowest Z and is the least hazardous impurity, but can still become highly problematic as shown in the numerical example below. This example is for illustrative purposes only and is not intended to be limited to just this use case. Neoclassical theory and experimental experience with other low-Z elements indicate that the lithium concentration in a MC plasma cannot be lower than that in a SOL under steady-state conditions. Since sputtering is temperature dependent, measurements indicate that the lithium sputtering rate is about 20% of the hydrogen bombardment rate at lithium surface temperatures that can be as low as 300 degrees Celsius (see, e.g., Delner 2001, Figure 3 ). This would give a SOL lithium content of about 20%. A 20% lithium concentration in the MC plasma would reduce the fusion power in the core by several times due to dilution, and this is quite unacceptable. To prevent this, the surface temperature of the lithium in the divertor target must be limited to very low. This would directly translate into a limitation on the heat flux incident on the target.

[0062] To summarize the paragraphs above, in order to solve problem #2 according to the prior art, problem #1 becomes more difficult to solve. This is highly undesirable because problem #1 is already extremely challenging.

[0063] Problem #3: It is understood in the art that for divertor targets, even a slow flow of liquid can replenish eroded material at a sufficient rate. This requires a flow rate several orders of magnitude lower than that required to maintain the lithium surface within its temperature limits. Such small flow rates appear practical. This would apply to lithium. Therefore, having a slow flowing liquid lithium divertor target would solve this problem.

[0064] Problem #4: There is some improvement over solid divertor targets in that the liquid does not form dust. But there are still serious problems. The eroded material can quickly accumulate as liquid on the rest of the plasma facing components. This can disable RF antennas, critical diagnostics, and other equipment. It will accumulate and drip into undesirable places, including the MC plasma. These impurities injected into the MC region are likely to inhibit fusion.

[0065] These problems can be avoided if the walls of the fusion device are operated at a high enough temperature that lithium will evaporate faster than it can be deposited. This can eliminate problem #4, but problem #2 then becomes more problematic due to wall generated impurities. Or alternatively, the walls of the MC plasma chamber can be coated with mobile lithium. But then the temperature of the wall is limited by evaporation and temperature dependent sputtering, and this then limits the heat conversion efficiency of the blanket connected to it. Reduced heat conversion efficiency severely reduces the economics of a fusion reactor, although for many test devices low wall temperatures are acceptable.

[0066] So, to summarize the paragraphs above, problem #4 can be solved, but only at the expense of worsening the other problems.

[0067] Problem #5: Helium Expulsion. If the divertor target is near the plasma X-point, as is typical in conventional techniques, then a long tube must transport the helium from that vicinity to a pump, which in standard techniques must have a cryogenic region. As indicated above, for low SOL densities, the helium is in a high Knudsen number regime. This results in insufficient conduction through the tube to the pump. This cannot be remedied by increasing the cross-sectional area of ​​the tube, as this would greatly increase the transport of neutrons down the tube, causing multiple unacceptable consequences: unacceptable heating of the cryogenic region, reduced tritium breeding of the deuterium-tritium fusion reactor, and radiation activation and damage to the pump assembly and its surrounding components.

[0068] The inevitable conclusion from the preceding discussion is that the proposal to use lithium at the high temperature, low density edge has several serious problems, some of which are fatal if lithium is adopted in light of existing technology. This is the judgment of many in the field, as can be seen in their actions. Analyses in the literature claim that lithium can greatly improve energy confinement and, therefore, will enable energy gains in much lower cost equipment. A significant reduction in cost is a strong incentive to adopt a technology that is considered practical. Nevertheless, neither ITER nor SPARC, nor any of the fusion demonstration reactor designs proposed by major countries, including the Korean K-DEMO project or the Chinese CFETR project, have incorporated lithium in order to greatly enhance plasma energy confinement and reduce costs. Only Tokamak Energy has pursued a lithium-based divertor for enhanced confinement.

[0069] In the prior art, liquid metals that replace lithium have been considered for divertor targets. Examples of such metals include, but are not limited to, molten tin, molten gallium, and molten aluminum. They do not absorb hydrogen-containing species and therefore do not directly lead to low-density SOL plasmas.

[0070] But let's assume that the low-density plasma is arranged in some way by other means (such as applying lithium in another area, or using other pumping means). Examination of the advantages and disadvantages of liquid metals other than lithium in the context of the SOL problem reveals that the SOL problem is still not qualified despite the advantage in heat flux.

[0071] Advantages of alternative liquid metals in SOL problems:

[0072] Heat Flux (Issue #1): If the material chosen can operate at surface temperatures much higher than that of lithium, it will be easier for that material to withstand the huge heat fluxes on the divertor target (Issue #1). Alternative liquid metals have an advantage here. Recent work as part of the EU Fusion DEMO program has developed divertor designs in which liquid tin faces the plasma in a low-temperature, high-density SOL scenario known as high-cycle conditions [Lint 2021]. It appears that liquid metals other than lithium will be able to withstand considerable heat fluxes. But some additional improvement would still be desirable. The solution is part of our invention.

[0073] Erosion (Issue #3): Even in strong magnetic fields, operation with low flow rates of liquid metal is allowed. Then, a slow flow of metal is employed to replenish the eroded material on the divertor target, as in the case of lithium.

[0074] Moderate Disadvantages of Alternative Liquid Metals Compared to Lithium:

[0075] Redeposition (Issue #4): There is some improvement compared to solid divertor targets in that the liquids do not form dust. Also, the evaporation of these metals is likely to be much smaller than that of lithium, which will help with material redeposition. However, if the SOL has a high temperature and low density, the inevitable sputtering process will still result in a lot of erosion and redeposition. Much of the eroded material will still condense in inappropriate places on the first wall of the MC plasma. And after sufficient accumulation, these deposits can fall into the plasma. Since these deposits are higher Z substitution metals, the plasma tolerance to impurities is much lower than that of lithium.

[0076] Overall, lithium replacement liquid metals alone will not solve problem #4.

[0077] The problem is roughly the same:

[0078] Helium Expulsion (Problem #5): Lithium has a slight tendency to absorb helium, but for high temperature liquid lithium this tendency seems to be quite weak. One would not expect a replacement liquid metal to absorb helium. Therefore, the problem is roughly the same: there must still be a tube from the divertor region to the pump. For low SOL densities, the neutral gas in the tube is still in a high Knudsen number state, which does not allow the neutral gas to be adequately transported to the pump. Increasing the tube cross-sectional area does not lead to a satisfactory solution, because this would also result in an unacceptably large neutron flux down the tube to the pump, resulting in all the negative consequences described previously.

[0079] Serious disadvantages of replacing liquid metals:

[0080] Impurity Accumulation in MC Plasmas (Problem #2): These materials localize much higher atomic number Z than lithium. The neoclassical processes that produce high concentrations of impurities in the MC region are even more intense when Z is large (as is well known in the art). As a result, alternative metals with higher Z than lithium are more concentrated. Worse still, it is well known that the tolerable concentration of higher Z impurities in the MC region is several orders of magnitude smaller than the tolerable concentration of lower Z impurities.

[0081] Additional problems arise for higher Z alternative metals. Once their atoms are ionized in the SOL, they can strike the divertor target and create even more impurities. This is so-called self-sputtering, and each impurity atom can potentially lead to more than one new sputtered atom. Impurities will increase exponentially through this process, which is well known in the art and is called self-sputtering avalanche. This leads to large amounts of impurity generation. This obviously leads to unacceptably high impurity levels.

[0082] As before, sputtered impurities can reach the MC region via the SOL, or as sputtered neutral impurities entering the MC region.

[0083] Therefore, problem #2 is far more problematic than the case with lithium.

[0084] The invention disclosed herein provides a solution to this problem.

[0085] In summary, in the prior art, it is impossible to achieve high temperature, low density edge plasma to avoid all unqualified SOL problems.

[0086] Each of the following is incorporated herein in its entirety by reference: 1–Castro 2021 “Lithium, a path to make fusion energy affordable”, Physics Plasma 28, 050901 (2021) 2 – Derner 2001 “Measurements of erosion mechanisms from solid and liquid materials in PISCES-B”, Journal of Nuclear Materials 290-293 (2001) 3 – Kessel, 2019 “Critical Exploration of Liquid Metal Plasma-Facing Components in a Fusion Nuclear Science Facility”, Fusion Science and Technology, Vol. 75, November 2019 4. Putridge, J. 2011. “ELM flushing and impurity transport in the H-mode edgebarrier in ASDEX Upgrade”, Journal of Nuclear Materials, pp. 334–339 (2011) 5–Lint 2021 “Conceptual design of a liquid-metal divertor for the European DEMO”, Fusion Engineering and Design 173 (2021) 112812 6- Provisional patent application US2010 / 0046688 A1 filed on August 25, 2008 7-U.S. Patent 8,279,99 filed on October 10, 2008. Summary of the invention

[0087] While the present invention addresses the shortcomings of the prior art as discussed in more detail below, in general, the present invention provides a method of operating with a scraped layer (hereinafter referred to as SOL) having a high temperature and low density, which will result in substantially enhanced core energy confinement.

[0088] The present invention avoids the five unqualified SOL problems discussed in detail above; all of which are repeated here for convenience: 1) Unacceptable heat flux on the wall, resulting in unacceptable melting or evaporation of the wall 2) Unacceptable impurity accumulation in MC plasma due to sputtering and evaporation 3) Unacceptable erosion of the divertor target by sputtering or evaporation 4) Unacceptable redeposition of eroded material in Problem #3 5) Unacceptably low helium output

[0089] By having a high temperature and low density SOL while avoiding the SOL problems mentioned above, the present invention will increase the energy gain of MC plasma fusion, and it will allow high energy gain to be obtained from a much smaller MC plasma, so the equipment for operating MC plasma will be cheaper.

[0090] One aspect of the features of the present invention has the feature of a divertor target arranged in a region where the total magnetic field strength is lower than the total magnetic field strength of the region between the target and the SOL position of the X point. Figure 3 and Figure 4 As can be seen in Figure 1, the divertor target 105 is located at a larger major radius than point X, which generally corresponds to a lower total magnetic field strength. The total magnetic field strength includes all components of the magnetic field, not just those on the polar faces. Impurity accumulation in MC plasmas is greatly reduced by placing the divertor target in this way for physical reasons discussed below.

[0091] There are also surfaces in the divertor region that substantially block the line of sight from the divertor target to the MC plasma. Figure 3 and Figure 5 We refer to these as caps 107 and in Figure 3 , Figure 4 , Figure 5 In these figures we call it shield 109. In these figures we also see that for some embodiments, the side of cover 107 facing the SOL has liquid 108 on cover 107 at least some of the time, and the side of shield 109 facing the SOL has liquid 111 on shield 109 at least some of the time.

[0092] In various embodiments, the liquid on the surface of the cover facing the SOL or the liquid on the surface of the shield facing the SOL can be a molten metal, including but not limited to lithium, tin, gallium, aluminum, beryllium and alloys thereof. The lithium will chemically combine with the hydrogen neutrons emitted from the divertor target. All liquid metals can allow re-deposited material to be removed as long as the liquid metal flows at least at a slow rate. The liquid can be a metal that is always molten, or a metal that is molten only some of the time, to allow periodic removal of accumulated material, or periodic replenishment of eroded material.

[0093] In some embodiments, the strength of the magnetic field component in the angular plane on the divertor target (hereinafter referred to as the angular magnetic field) is greater than one-third of the angular magnetic field along the last closed flux surface (LCFS) of the annular confinement region. As described below, the width of the SOL produced in this way is not much longer than the mean free path of the neutrons emitted by the divertor target. This means that many neutrons can leave the SOL, which has advantages in solving the SOL problem, which advantages are described below. Examples of such neutrons are helium, higher Z impurities of the type that can cause self-sputtering avalanches if ionized, and in some embodiments, hydrogen neutrons.

[0094] In other embodiments, the divertor target has a large amount of neutron absorbing material between the divertor target and the magnetically confined plasma. For deuterium-tritium fusion, this may be a lithium (Li 6 ) of a tritium breeder blanket. Other neutron shielding materials may be used, either together or alone, including but not limited to B 10 This material, in addition to being located between the divertor region and the MC plasma, may also partially surround the divertor region.

[0095] In other embodiments, the surface of the divertor target facing the SOL is covered by a liquid, which includes but is not limited to molten metals such as tin, gallium, aluminum, beryllium, lithium, and alloys thereof. The liquid can be a metal that is always molten. Alternatively, for some metals, the metal may be molten only some of the time to allow it to be periodically replenished or redistributed, examples of such metals include but are not limited to beryllium, aluminum, or alloys of these metals.

[0096] In other embodiments, there is a tube for pumping helium or hydrogen to the pumping device that is much shorter than in standard technology. The shorter tube conducts the gas much better in high Knudsen number flow regimes. This allows for adequate transport of helium from the divertor target region to the pump. Figure 3 and Figure 4 In Figure 1, we see that there is neutron absorbing material between the pump tube inlet and the MC plasma, which allows fewer neutrons to enter the pump tube. Therefore, the area of ​​the tube can be increased to further improve the helium transport without causing unacceptable neutron leakage when passing through the tube to the cryogenic area or vacuum pump, and without seriously reducing tritium breeding due to neutron losses through the tube. Specifically, the distance from the divertor target to the pump is less than half the distance from the plasma point X to the pump, and there is neutron absorbing material between the divertor target and the MC plasma that produces the neutrons.

[0097] The purpose of the present invention is to create higher temperatures at the edge of the MC plasma than would otherwise be possible. Thus in other embodiments the electron temperature immediately adjacent to the divertor target is 25 eV. By temperature, in the following we mean 2 / 3 of the average energy. We note that much higher temperatures are possible, but such SOL temperatures in the divertor region are already beyond the reach of standard technology for fusion reactor relevant conditions due to erosion and redeposition. In other embodiments the LCFS temperature of the main plasma is 200 eV or higher. However, the latter is easier to measure. Therefore, we describe the requirement as either the electron temperature on the last closed flux surface of the MC plasma being above 200 eV or the electron temperature immediately adjacent to the divertor target reaching 25 eV.

[0098] In other embodiments, the density of the MC plasma in the LCFS is much smaller than the linear average density of the MC plasma. This is desirable for high energy confinement of the MC plasma and is the main reason for having a low density SOL because the SOL 103 is close to the LCFS 102. See Figure 3-Figure 4 We describe this by the most easily measurable quantity, namely, the ratio of the electron density on the last closed flux surface to the line-averaged electron density of the MC plasma, which is 0.2 or less.

[0099] When these elements are combined, they lead to embodiments that allow high temperature, low density SOL plasmas while avoiding the unqualified SOL problems described in the above section, and this will improve the energy gain of MC plasmas, and also allow much smaller and cheaper equipment with high energy gain. Furthermore, as will be explained below, these elements tend to work together.

[0100] In other embodiments, the plasma purity is maintained at a level such that fusion reactions can occur. It is known in the art that there are several important aspects of purity, which we divide as follows. In some embodiments, the purity can be maintained at a satisfactory level so that there is no accumulation of photon radiation over time until the final fusion reaction drops by 40% or the temperature drops by 40%. Experiments have found that excessive accumulation of impurities in MC plasmas causes this drop, which is one of the common characteristics of excessive impurities. In other embodiments, the purity is maintained high so that the radiation from the plasma is less than 70% of the heating power, where the heating power includes the sum of externally applied heating and heating from nuclear reactions in the plasma. Photon radiation in MC plasmas is mainly caused by impurities, and excessive energy losses of these photons prevent the MC plasma from remaining hot enough to provide fusion. In yet other embodiments, the purity is maintained high so that the amount of effective Z known in the art is maintained below 3, which is generally expressed in the art as Z. eff It is known in the art that too high Z effThis makes it difficult to drive sufficient current in the MC plasma to maintain the magnetic configuration for long enough. The effective Z used here is the average value for the MC plasma and is defined as s Z s 2 The ratio of the sum, where N s is the number of ions of species s in the MC plasma, and Z s is the charge state of each plasma species s (i.e., the number of electrons lost from the atoms), and the denominator is the number of electrons in the MC plasma N e In yet further embodiments, the purity of the MC plasma is such that there is little dilution of the fusion fuel, resulting in a high fusion reaction rate. The specification is that for particles comprising the MC plasma, the sum of the charges of each fusion fuel ion is greater than 60% of the sum of the charges of all electrons. The fusion fuel particles in the DT fuel cycle include deuterium and tritium, and D-He 3 Fusion fuel particles in the fuel cycle include deuterium and He 3 , and the fusion fuel particles in the P-boron fuel cycle include protons and boron.

[0101] The present invention can be applied to, but is not limited to, tokamaks, spherical tokamaks, stellarators, annular pinches, reverse field pinches (RFPs), spheromaks, reverse field configurations (FRCs), etc. Examples of geometries that can benefit from the application of the present invention include, but are not limited to, the ARC tokamak reactor being designed by Federal Fusion Systems, and spherical tokamak reactors such as the STEP device proposed by the UK Atomic Energy Agency and the device proposed by UK Tokamak Energy.

[0102] Examples of the types of FRC reactor configurations that may benefit from the application of the present invention include, but are not limited to, those disclosed in U.S. Patent No. 11,337,294, which describes the apparatus used by TAE Technologies, Inc. in experiments to develop FRC fusion reactors. The 294 patent describes how experiments used extremely thin layers of solid lithium in experimental discharges that lasted only a few milliseconds. Compared to the experiments in the 294 patent, our invention is valuable in discharges that: 1) last longer, so that the thin solid coating of lithium becomes saturated or eroded away 2) produce a considerable amount of neutrons or helium ash, and 3) have higher temperatures similar to those of fusion reactors, which result in higher sputtering and erosion. These conditions will occur as the experimental parameters become closer to those of energy-producing fusion reactors.

[0103] In some embodiments, an aspect of the invention features a toroidal confined plasma vessel comprising a toroidal plasma chamber and a magnetic confinement plasma region, wherein particles moving along the magnetic field substantially never strike a wall. The magnetic confinement plasma region is substantially symmetrical by rotation about a central axis. The plasma vessel further comprises a plurality of magnetic field coils and a divertor assembly with a divertor target. The plurality of magnetic field coils are configured to provide at least one X-point and to direct plasma particles from the magnetic confinement region to the divertor target. The divertor target has a cover, wherein a side of the cover substantially facing the divertor target comprises a material that is liquid on at least some surfaces of one side of the cover during at least some of the time the cover is in the toroidal confinement plasma vessel. The total magnetic field strength (including all components of the magnetic field) at the divertor target is lower than the total magnetic field strength (including all components of the magnetic field) between the divertor target and point X on the last closed flux surface closest to the divertor target; so that at least one of the following conditions exists: radiation from the magnetically confined plasma does not increase with time until the fusion rate in the magnetically confined plasma decreases by 40% or until the maximum plasma temperature in the magnetically confined plasma decreases by 40%, the power radiated from the magnetically confined plasma by photons does not exceed 70% of the heating power (where the heating power is the sum of externally applied heating plus heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is lower than 3 (where the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.6 times the sum of the charges of all electrons in the magnetically confined plasma.

[0104] In some embodiments, the divertor target has a shield that substantially blocks line of sight from the divertor target to the magnetically confined plasma region and to components important for maintaining operation of the apparatus.

[0105] In some embodiments, a shield that substantially blocks line of sight from the divertor target to the magnetic confinement region and to components important to maintaining device operation to maintain device operation is covered by liquid on a side substantially facing the divertor target during at least some of the time the cover is within the annular magnetic confinement plasma vessel.

[0106] In some embodiments, the component of the angular magnetic field is the component of the magnetic field on a plane perpendicular to the rotation direction of the central axis, and the amplitude of the component of the angular magnetic field at the divertor target is greater than one third of the maximum value of the angular magnetic field at the boundary of the magnetically confined plasma region.

[0107] In some embodiments, the neutron absorbing and moderating material is located substantially between the magnetically confined plasma region and the divertor target.

[0108] In some embodiments, the divertor target surface comprises a material that is liquid on at least some of the surface at least some of the time.

[0109] In some embodiments, the toroidal confined plasma vessel further comprises a pumping tube extending from a position near the divertor target to a pumping device to pump helium, hydrogen isotopes, other gases, or any combination thereof, and wherein a distance from the divertor target to the pumping device is less than one-half of a distance from point X to the pumping device.

[0110] In some embodiments, at least one of the following: the electron temperature at the boundary of the magnetic confinement region is greater than 200 eV or the electron temperature immediately adjacent to the divertor target is greater than 25 eV, and the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.2.

[0111] In some embodiments, at least one of the following: the electron temperature at the boundary of the magnetic confinement region is greater than 1000 eV, the electron temperature immediately adjacent to the divertor target is greater than 100 eV, and wherein the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.15.

[0112] In some embodiments, at least one of the following: radiation from the magnetically confined plasma does not increase over time until the fusion rate in the magnetically confined plasma decreases by 20% or until the maximum plasma temperature in the magnetically confined plasma decreases by 20%, the power radiated from the magnetically confined plasma by photons does not exceed 50% of the heating power (where the heating power is the sum of externally applied heating plus the heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is less than 2.5 (where the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.75 times the sum of the charges of all electrons in the magnetically confined plasma.

[0113] In some embodiments, another aspect of the invention features a toroidal confined plasma vessel comprising a toroidal plasma chamber and a magnetically confined plasma region, wherein particles moving along the magnetic field substantially never strike a wall, wherein the magnetically confined plasma region is substantially symmetrical by rotation about a central axis. The plasma vessel further comprises a plurality of magnetic field coils, a divertor assembly having a divertor target, and wherein the plurality of magnetic field coils are configured to provide at least one X-point and direct plasma particles from the magnetically confined region to the divertor target. At least one of the following: the electron temperature at the boundary of the magnetically confined region is greater than 200 eV, the electron temperature immediately adjacent to the divertor target is greater than 25 eV, and the ratio of the plasma electron density at the last closed flux surface to the line average electron density of a chord passing near the center of the magnetically confined plasma is less than 0.2. The divertor target comprises, on a surface facing the plasma, a material that is liquid at least some of the time and whose composition has a lithium content of less than 50% by atomic fraction. The total magnetic field strength (including all components of the magnetic field) at the divertor target is lower than the total magnetic field strength (including all components of the magnetic field) at a location between the divertor target and point X on the last closed flux surface closest to the divertor target in the SOL; so that at least one of the following conditions exists: radiation from the magnetically confined plasma does not increase with time until the fusion rate in the magnetically confined plasma decreases by 40% or until the maximum plasma temperature in the magnetically confined plasma decreases by 40%, the power radiated from the magnetically confined plasma by photons does not exceed 70% of the heating power (where the heating power is the sum of externally applied heating plus heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is lower than 3 (where the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.6 times the sum of the charges of all electrons in the magnetically confined plasma.

[0114] In some embodiments, the divertor target includes a cover, wherein a side of the cover substantially facing the divertor target includes a material that is liquid on at least some surface of the side of the cover.

[0115] In some embodiments, the neutron absorbing and moderating material is located substantially between the magnetically confined plasma region and the divertor target.

[0116] In some embodiments, the toroidal confined plasma vessel further comprises a pumping tube extending from a position near the divertor target to a pumping device to pump helium or hydrogen isotopes, other gases or any combination thereof, and wherein a distance from the divertor target to the pumping device is less than one-half of a distance from point X to the pumping device.

[0117] In some embodiments, the component of the angular magnetic field is the component of the magnetic field in a plane perpendicular to the rotation direction of the central axis, and the amplitude of the component of the angular magnetic field at the divertor target is greater than one third of the maximum value of the angular magnetic field around the boundary of the magnetically confined plasma region.

[0118] In some embodiments, a shield that substantially blocks line of sight from the divertor target to the magnetic confinement region and to components important for maintaining operation of the apparatus is covered by liquid on a side that substantially faces the divertor target.

[0119] In some embodiments, at least one of the following: the electron temperature at the boundary of the magnetic confinement region is greater than 1000 eV, the electron temperature immediately adjacent to the divertor target is greater than 50 eV, and the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.15.

[0120] In some embodiments, at least one of the following: radiation from the magnetically confined plasma does not increase over time until the fusion rate in the magnetically confined plasma decreases by 20% or until the maximum plasma temperature in the magnetically confined plasma decreases by 20%, the power radiated from the magnetically confined plasma by photons does not exceed 50% of the heating power (where the heating power is the sum of externally applied heating plus the heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is less than 2.5 (where the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.75 times the sum of the charges of all electrons in the magnetically confined plasma. BRIEF DESCRIPTION OF THE DRAWINGS

[0121] Figure 1A Examples from the prior art are given, and graphical examples of the terms used in the description are also given. The terminology is standard in the art. Surfaces tangential to the magnetic field are flux surfaces. Plasma particles travel primarily within these surfaces and cross them relatively slowly. The geometry is roughly a ring, and therefore has a rough axis of rotation. The distance of a point from this axis is called the major radius of the point. Such as Figure 1AThe plane in which the axis of rotation is shown is called the polar plane. The magnetic field component located in this plane is called the angular magnetic field. The last closed flux surface (LCFS) 102 has one or more X points 120, at which the angular magnetic field disappears. As is well known in the art, the magnet surface at the X point forms an "X", as shown in the figure. The area inside the LCFS 101 is a magnetically confined plasma, or MC plasma. In addition, as is well known in the art, the high temperature SOL plasma has most of the heat and particles that are essentially confined between two flux surfaces, one is an extension of the LCFS 130, referred to in the art as a separatrix, and the other is a boundary SOL surface 103.

[0122] The SOL plasma strikes the material surface at the strike points 104 and 140, and the material surfaces are referred to as divertor targets 105 and 150, also shown. Each X point has two divertor targets associated with it, one of which is farther from the axis of rotation 105, referred to as the outer divertor, and most of the particles and energy flow to this target. The other divertor is referred to as the inner divertor 150.

[0123] exist Figure 1A In the case of a plasma, point X is at the bottom of the plasma; this configuration is known in the art as the "lower null". The configuration in which the divertor is located near point X at the top of the plasma is known as the "upper null".

[0124] exist Figure 1B In the figure is another geometry of the prior art, called the Super-X divertor. Shown is the actual geometry of a device that has been built in the UK, called MAST-U. The figure shows the MC plasma 101, the LCFS 102, the lower X point 120, the boundary surface 103 of the large part of the SOL, the impact point 104 and the divertor target 105. The divertor target and the SOL near it are contained in the divertor chamber 190.

[0125] Figure 2A close-up of a conventional outer divertor region of the prior art and "lower zero point" configuration showing how neutral gas is pumped. It shows the MC plasma 101, the LCFS 102, the lower X point 120, the boundary surface 103 of most of the SOL, the impact point 104 and the divertor target 105. The wall of the region containing the plasma 160 is shown. Neutral gas particles are produced in the region of the divertor target and must be removed by pumping. This is accomplished by having a pumping tube 170 with an inlet 172 in the divertor region, which ends in a pumping device 171 for pumping neutral particles. Neutrons produced in the MC plasma 101 are absorbed by the neutron absorbing material 110 that generally surrounds the MC plasma 101, but the inlet of the divertor region has no significant neutron absorbing material between it and the MC plasma.

[0126] Figure 2B It is prior art and has a similar pumping pipe arrangement to the ITER equipment, which is similar to Figure 2 The arrangement in FIG. 1 is different. The figure shows the boundary surface of the boundary surface 103 of most of the SOL in the MC plasma 101, the LCFS 102, the lower X point 120, the boundary surface 103 and the divertor target 105. The wall of the region containing the plasma 160 is shown. Neutral gas particles are generated in the region of the divertor target and must be removed by pumping. This is done by having a pumping tube 170 with an inlet 172 in the divertor region, and at its end is a pumping device 171 that pumps the neutral particles. Neutrons generated in the MC plasma 101 are absorbed by the neutron absorbing material 110 that generally surrounds the MC plasma 101, but the inlet of the divertor region has no significant neutron absorbing material between it and the MC plasma.

[0127] When there is a lower zero point, Figure 3 An example of an embodiment of the invention of the present disclosure is given. There are many differences from the prior art, which will be explained in the detailed description section below. The figure shows MC plasma 101, LCFS 102, boundary SOL surface 103, divertor target 105, pumping tube 170, pumping tube inlet 172 and pumping device 171. The new aspects include material 106 on the divertor target that is liquid at least some of the time facing the SOL, cover 107, material 108 on the top of the cover that is liquid at least some of the time facing the SOL, shield 109, and material 111 on the top of the shield that is liquid at least some of the time facing the SOL. The figure also shows neutron absorbing material 110 located between the MC plasma and both the divertor target 105 and the pumping tube 172 inlet.

[0128] In other embodiments of the present invention, there will be an X point at the top of the upper half. This situation is an example of an "upper zero point", such as Figure 41. MC plasma 101, LCFS 102, boundary SOL surface 103, divertor target 105, pumping tube 170, pumping tube inlet 172, and pumping device 171 are shown. The new aspects include material 106 on the divertor target that is liquid facing the SOL at least some of the time, cover 107, material 108 on the top of the cover that is liquid facing the SOL at least some of the time, shield 109, and material 111 on the top of the shield that is liquid facing the SOL at least some of the time. Neutron absorbing material 110 located between the MC plasma and both the divertor target 105 and the pumping tube 172 inlet is also shown.

[0129] Some magnetic geometries have two X-points on (or very near) the LCFS, one near the top and one near the bottom. This is known in the art as a "double zero point." In this case, an exemplary embodiment of the present invention may have a lower divertor region (such as Figure 3 ) and the upper divertor region (such as Figure 4 ).

[0130] These images are for a magnetic configuration of a MC plasma, which may be a tokamak, a spherical tokamak, a toroidal pinch or a reverse field pinch (RFP). The invention may also be applied to other magnetic configurations, such as a reverse field configuration (FRC) or a spheromak. Examples of aspects of the invention applied to FRC are given in Figure 5 The magnetic field configuration is taken from U.S. Patent No. 11,337,294 (specifically, the Figure 2 ). For the sake of clarity, aspects of the drawings that are not relevant to the present patent considerations are omitted. Figure 5 MC plasma 101, LCFS 102, SOL boundary 103, divertor target 105, and lid 105 are shown. New aspects include material 106 on the divertor target that is liquid at least some of the time facing the SOL, and material 108 on the top of the lid that is liquid at least some of the time facing the SOL. Also shown is neutron absorbing material 110 between the MC plasma and the inlet of the pumping tube 172. DETAILED DESCRIPTION

[0131] Some embodiments of the present invention are for magnetic geometries with a lower single zero. Other embodiments are for magnetic geometries with an upper single zero. Still other embodiments of the present invention employ a dual zero geometry.

[0132] Note that in the specific case of double zero point geometry, the present invention differs in many respects from the US patent application of Buxton et al. (US 20210265068A1, published as US2021 / 0265068A1), which also applies double zero point geometry. The essence of our invention and its operation are quite different from the invention of Buxton et al. The object of the present invention is to find a solution to the SOL problem described above. The object of Buxton's invention is to find a method for applying a liquid metal divertor target to a double void geometry and, as implied in the claims, to use gravity to assist the liquid flow (because in Buxton, the flow rate at the inlet must be higher than the outlet). Most of the SOL problems that prompted the present invention are not even mentioned in Buxton's patent. Therefore, many aspects of the present invention are to solve the SOL problem, which are not present in Buxton's invention. And solving the SOL problem does not require double zero point geometry, which is the essence of Buxton's patent. Moreover, the use of gravity to assist the liquid metal flow is not critical to our invention in any respect. For example, heat can be removed by conducting the liquid metal to a heat sink behind it without using gravity assisted flow at all. It can therefore be easily appreciated that the present invention is very different from Buxton.

[0133] Some embodiments of the invention apply to both the lower single zero case and the upper single zero case. As another example, our invention addresses the problems of sputtering, erosion, and helium, and these problems are addressed in the description, and these problems motivate many of the claims. But in Buxton et al., the terms sputtering, erosion, and helium appear only once each in parenthetical notes, and certainly do not motivate any claims. Also, note that in Buxton et al., the inlet for the flow of liquid is above the outlet. But our invention can have a divertor target that can be substantially horizontal to promote a smooth surface. Our invention requires only a slow flow of liquid over the divertor target to replenish the eroded material, and this can be easily provided without gravity assistance, for example, by using capillary action, or by electromagnetic means, or by other means, so that our invention does not require a flow inlet above the outlet.

[0134] Buxton is largely irrelevant to the operation of our invention. The nature of the present invention is quite different from the invention of Buxton et al., and the claims of our invention apply to many situations in which Buxton does not apply. Nevertheless, it is possible that some embodiments of our invention may be double zero point geometries with the liquid flow inlet above the outlet.

[0135] There are several key insights that led to our invention. First, the goal is to produce high-temperature, low-density SOL in a practical way, i.e., so that all the unqualified SOL problems are solved. This will then lead to longer energy confinement times, higher fusion gains, and fusion gains in smaller devices.

[0136] The following insights, not in the prior art, show that there are conditions under which certain physical dynamics in SOL can take advantage. Various aspects of the present invention enable the favorable physical dynamics to actually manifest, and therefore, allow MC plasmas with high energy confinement by avoiding the SOL problem. Other additional aspects also help allow further resolution of the SOL problem.

[0137] Two key physical dynamics are described below, and aspects of the present invention will take advantage of these dynamics. These dynamics drive some of the key features of the present invention, so we discuss them below. We distinguish between the two by calling them Dynamic One and Dynamic Two.

[0138] 1) The following section describes the dynamics of No. 1 and its implications.

[0139] Our analysis shows that by placing the divertor target in an area with low magnetic field strength B, a strong electrostatic potential is generated which has the effect of preventing impurities generated at the divertor target from entering the MC plasma. The electrostatic potential actually protects the MC plasma from destructive impurities that are generated near the divertor target. Otherwise, these impurities would be ionized in other ways and travel along the magnetic field lines to reach the MC plasma. To the best of the authors' knowledge, until the present disclosure, this dynamic has not been fully exploited in the art.

[0140] Note that this dynamics only occurs in high temperature, low density SOLs, which are the physical conditions of concern to the present invention. Specifically, the electrostatic potential is strong when the mean free path of Coulomb collisions is longer than the characteristic distance that particles travel along the magnetic field lines from the divertor target to the MC plasma. This long mean free path is generated in a SOL with high temperature and low density. The electrostatic potential is much weaker under the conventional operating conditions of the divertor, which are the opposite: low temperature and high density under these conditions (as in MAST-U and many other cases in the art) lead to a short mean free path.

[0141] It seems true that, because the fusion community has focused primarily on the short Coulomb collision mean free path conditions, the ability of this electrostatic potential to help prevent contamination of magnetically confined plasmas in the long mean free path regime was not recognized until this disclosure. This patent arranges aspects of the invention to use this dynamics to advantage.

[0142] How this is done is explained below. Under the conditions of interest for the present invention, the magnitude of the potential difference along the magnetic field line between point 1 and point 2 is approximately (T e / e)ln(B1 / B2), where T e is the electron SOL temperature, e is the charge on the electron, B1 is the total magnetic field strength at point 1, B2 is the total magnetic field strength at point 2, and ln is the natural logarithm.

[0143] Under this condition, with T e The magnitude of the potential will have a very large effect on the path of the impurity. Impurities in SOL plasmas are generated by sputtering, evaporation or recycling, and in all these cases the energy of the impurities is in the range of a few eV or less. The present invention is applied to SOL in which T e The potential is about 200 eV or higher, which is much greater than the energy of the impurities. In this case, the potential can prevent the impurities from reaching the MC plasma. The impurities are positively charged, so the sign of the electrostatic potential will have this effect when the polarizer filter target is in a relatively small area compared to other locations between the polarizer filter target and the MC plasma.

[0144] Furthermore, if the mean free path of the Coulomb collisions is long, the time scale on which the electrostatic potential reflects the impurities back to the divertor is much shorter than the time it takes for the impurities to be heated by the SOL plasma to a high enough energy that the impurities can overcome the electrostatic potential. This heating is only strong enough under conditions of short mean free path, which is contrary to the conditions considered in the present invention.

[0145] To summarize the previous paragraphs: Under the density and temperature conditions of the embodiments of the present invention, if the divertor target is in a region of relatively low total magnetic field strength, the MC plasma is insulated from impurities generated in the SOL. This is critical to avoid contamination of the MC plasma, because, as we have described previously, experiments and neoclassical transport suggest that impurities in the SOL can be highly concentrated in the MC plasma.

[0146] It is therefore an aspect of the invention that the magnetic field strength at the divertor target is less than the strength at some location in the SOL between the target and the X point of the MC plasma.

[0147] Impurities for which such desirable kinetics apply include those generated by sputtering, evaporation, and recycling, but are not limited to elements and helium of the plasma facing material, and impurities from small amounts of air or other gases leaking into the plasma chamber.

[0148] As indicated in the previous paragraphs, one of the basic goals of the present invention is to prevent the MC plasma from being contaminated by unacceptable impurities. Impurities have several serious negative effects on plasmas, which are known in the prior art. The appropriate way to specify the purity of MC plasmas is to ensure that these negative effects are small. Therefore, we describe the purity of plasmas in terms of the following four adverse effects from impurities that must be small. 1) In experiments, an important manifestation of excessive contamination is that the radiation of the plasma increases over time until it becomes excessive and there is a collapse of the MC plasma temperature or fusion rate. Therefore, we specify low impurity conditions in the same way in the claims: the radiation of the magnetically confined plasma will not increase over time until the plasma temperature in the magnetic confinement region drops to a specified percentage or the fusion rate drops. The percentage drop beyond this condition will seriously affect the operation of the device. 2) Another important way that impurities degrade MC plasmas is by generating radiation power losses, which become a large part of the heating power, making the heating power insufficient to maintain a sufficiently high temperature. We describe this in the claims, the power radiated from the MC plasma by photons must not exceed the numerical portion of the heating power, where the heating power is the sum of the externally applied heating plus the heating generated by nuclear reactions in the MC plasma. If the radiation exceeds this fraction, the operation of the device will be severely affected. 3) Another way impurities can degrade the plasma is by increasing its resistance, making it difficult to drive the necessary current. It is known in the art that resistance is closely related to effective Z. Therefore, effective Z must be below a numerically specified value, exceeding which will severely affect the operation of the MC plasma. The effective Z used here is the average value of the MC plasma, and the effective Z is defined as a ratio, whose numerator is N s Multiply by Z s 2 The ratio of the sum, where N s is the number of ions of species s in the MC plasma, and Z s 2 is the square of the charge state of each plasma species s (the charge state is the number of electrons lost from an atom), and the denominator is the number of electrons in the MC plasma N e 4) A further way that impurities can reduce operation is by diluting the fusion fuel so that the fusion reaction rate is reduced. In the art, the dilution of fusion fuel is usually expressed as a relative value of the ratio of the number of fuel ions to the number of electrons. In a completely pure plasma, the charge on all fuel ions would be equal to the charge on all electrons. Impurities reduce the total charge carried by the fuel ions and the charge they carry. Therefore, we specify a low dilution of MC plasma by stating that the sum of the charge of each fusion fuel ion is greater than a certain fraction of the sum of the charges of all electrons. Dilutions greater than this low dilution will severely affect the Cao group of the device.

[0149] We now turn to the subject of how the electrostatic potential aspects also make solving SOL problems simpler to obtain benefits so that they work together advantageously.

[0150] In SOL, the magnetic field strength of the tokamak generally decreases with increasing major radius. Therefore, in order to advantageously apply the above dynamics, the divertor target 105 must generally be at a greater distance from the MC plasma 101. See Figure 3 and Figure 4 . And note that the inlet of the pumping tube 172 must be quite close to the divertor target 105. This brings another advantage for helium pumping: the inlet of the pumping tube 172 is quite far from the MC plasma, and therefore closer to the pump 171. Therefore, the tube to the pump 170 can be shorter, improving the transport of helium through the tube to the pump for removal. This solves the substantial problem described earlier in the SOL low density conditions: the relatively poor transport of helium through the tube.

[0151] Another aspect of the geometry as mentioned above can further facilitate helium pumping: if the divertor target 105 is at a distance from the MC plasma, it is possible to place the target in an area of ​​low neutron flux, because there is enough space between the divertor target 105 and the MC plasma 101 generating the neutrons to place the neutron absorbing material 110. Therefore, fewer neutrons enter the pumping tube inlet 172. Therefore, it is possible to have a larger cross-sectional area of ​​the pumping tube 170 without overheating the low temperature areas of the pump or degrading the vacuum pump itself due to an unacceptable number of neutrons transmitted through the tube. This further improves the transport of helium through the tube. Figure 2 As shown, a divertor configuration standard in the art has the inlet of the tube 172 at a much higher neutron flux regime, and therefore the tube cannot have too large an area to prevent allowing too many neutrons to pass through the tube.

[0152] Thus, to summarize the above paragraphs, several important advantages can work together in the invention disclosed herein.

[0153] With regard to the advantages of shorter tubes described above, we indicate in the claims that the distance between the divertor target and the pump is less than half the distance between point X and the pump.

[0154] This summarizes the section describing Kinetic One, as defined above, and all of the ways in which it can contribute beneficially to device operation.

[0155] 2) The following paragraphs describe the kinetics No. 2 that the claim will result in advantageous operation of the device as defined above:

[0156] For the parameters of the energy gain device and the parameters of the SOL temperature and density, there is a magnetic field geometry that has another important property: neutral atoms and molecules have a relatively long mean free path in the SOL, so that many of these atoms and molecules pass through the SOL without being ionized. Specifically, when this is applied to the divertor region, then many of the neutral particles generated at the divertor target tend to pass through the SOL and remain as neutral particles. These neutral particles include recycled hydrogen, helium, and sputtered impurities. This is very beneficial for reasons described below. Therefore, one aspect of the device is to make the width of the SOL smaller than the mean free path of the neutrons generated at the divertor plates, so that many of the neutral particles pass through the SOL.

[0157] One reason this is beneficial is that in order to achieve a low density SOL, it is not necessary to place a chemical absorbent material like lithium at the divertor target 105 where the heat flux is very high. Instead, the absorbent material can be placed in a different location than the divertor target where the heat flux is much lower. For example, in some embodiments, the cover 107 can have a liquid lithium surface facing the SOL 108. Or in some embodiments, the shield 109 can have a liquid lithium surface facing the SOL 111. Or in other embodiments, the neutral particles can enter the tube at 172, travel through the tube 170, and be removed by the pump 171. In fact, in some embodiments, there may be no chemical absorbent material at all, and other reliable pumping devices known in the art (such as a diffusion pump or a cryogenic pump) can remove the neutral particles.

[0158] Recall that a central feature of the prior art for obtaining low density, high temperature SOL is the use of lithium on the divertor target. And recall that this leads to a major problem in the prior art in that due to the high heat flux of the divertor target, it is difficult to maintain a temperature at the surface that is low enough to be acceptable for lithium (some estimate it to be 400-450 degrees Celsius, others estimate it to be 300-380 degrees Celsius). As described in the paragraphs above, these problems can be avoided in the present invention because the lithium does not need to be on the divertor target.

[0159] In summary, not having the SOL be too wide allows for the following important advantages: We can separate the two key functions of a) withstanding high heat fluxes, and b) pumping plasma particles to keep density low. Separating these functions allows the use of appropriate materials and devices for the best effect of each key function, without having to use lithium in both. Materials other than lithium can be used for the divertor target, which can withstand higher temperatures. This is a major benefit, because under the conditions of a practical fusion device, there is no material or device that is simultaneously optimal in both functions. And again, separation of these functions becomes possible when the width of the SOL is not much longer than the mean free path of the neutral particles.

[0160] An additional important advantage of the SOL being not much larger than the mean free path is as follows: this avoids self-sputtering avalanches, even for materials with self-sputtering coefficients greater than one at high SQL temperatures. This allows a much wider class of materials to be considered for the divertor plates, some of which have more favorable operating temperature ranges or other advantages. Recall what an avalanche is: ionized impurities are accelerated to high energies in the SOL, and when they hit the plate, they produce more than one neutral impurity atom. Obviously, this can lead to exponential growth: an avalanche. However, in the current geometry, many neutral sputtered atoms pass through the SOL without being ionized. Therefore, they will not be accelerated to energies that cause high self-sputtering. Thus, self-sputtering avalanches are avoided because many sputtered impurities avoid the critical step of being ionized in the SOL and therefore accelerated to high energies.

[0161] A further additional important advantage of operating in a regime where the SOL width is less than the mean free path is that pumping helium ash becomes much easier. In order to pump helium efficiently, the pressure of the neutral helium in the divertor region must be high enough to create a high enough pressure in the pump region. Some of this neutral helium will be ionized in the SOL and tend to cause plasma contamination. By making the SOL very thin, the amount of ionization is lower. This means that a higher density of neutral helium can be present in the divertor region, which makes it easier to pump helium in the pump region.

[0162] To summarize the paragraphs above, having the SOL width not much longer than the mean free path of the particles brings several benefits. A narrow SOL near the divertor target is clearly beneficial for this. Importantly, this situation is not the usual case in conventional technology: in general, the plasma SOL width near the divertor target is quite wide. This is because in conventional technology, the divertor target is usually not far from the MC plasma X point, and the SOL is very wide in this region.

[0163] The present invention uses a magnetic field configuration other than the conventional magnetic field configuration in the prior art in order to have this characteristic. To understand how this is done, a standard for a relatively narrow SOL width is first derived using methods well known in the art.

[0164] The particles in the SOL follow the magnetic field, especially at high temperatures and low densities. Therefore, the width w of the SOL is consistent with the width of the region between the flux surfaces that bind it. This means that w varies as 1 / (B_pol R), where R is the principal radius of the site and B_pol is the angular magnetic component of the magnetic field, i.e. the component of the magnetic field in the plane perpendicular to the direction of rotation around the axis.

[0165] Embodiments of the present invention ( Figure 3 and Figure 4) compared to the conventional divertor geometry, the divertor target 105 is closer to the vicinity of the plasma X-point 120 (see Figure 2 ). Point X is the position where the azimuthal magnetic field vanishes on the LCFS. For conventional geometry, where each divertor target is near point X, the B_pol on the divertor target is quite small. Therefore, such divertors will have a large SOL width and will not be able to obtain the advantage of a low SOL width.

[0166] To avoid the large SOL width of conventional divertor geometry, at least one divertor target must be placed in a different position than the conventional geometry. It is known in the art that each X point has two divertor targets associated with it, one having a larger major radius R than the other divertor target. Most of the heat and particles flow to the outer divertor target with the larger R, referred to as the outer divertor target.

[0167] Therefore, we give the angular magnetic field specification of the outer divertor target so that the angular magnetic field must not be too small. The specification is as follows, and the specification is based on the B_pol of the outer divertor target: For the present invention, at the location of at least one outer divertor target, the angular magnetic field must be at least as large as the coefficient of the maximum value of the angular magnetic field at any location on the LCFS of the MC plasma. The preferred value of this coefficient is 1 / 3.

[0168] The requirement for the divertor target to be located away from point X also has the consequence that it is easier to position the target in a region where the above described kinetics No. 1 is obtained, i.e. the divertor target is in a region where the total magnetic field strength is low, because if the target is close to point X it will be difficult to position the target in a region where the total magnetic field strength at the target is different from the value at point X or any other point therebetween. Thus, the requirements for kinetics No. 1 and kinetics No. 2 work together to a large extent.

[0169] In addition, the divertor target is positioned away from the LCFS X point, and there is space between the MC plasma and the divertor target to place the neutron absorbing material, so that there is a lower neutron flux on the divertor target. This is the same benefit of the low neutron flux in part 1) above. This is another method in which the present invention works together advantageously.

[0170] Some additional key aspects of the present invention assist in allowing all SOL problems to be solved simultaneously:

[0171] 3) As described above, if the SOL near the divertor target is narrow, this tends to concentrate heat, making it more difficult to solve problem #1. It is well known in the art that to spread the heat flux, the magnetic field can be made nearly tangential to the plate. It is recognized in the art that heat is spread by an amount inversely proportional to sin(θ), where θ is the angle of the magnetic field to the plate. In addition, it is well known in the art that the inhomogeneity of the target plate limits how small the angle can be. Therefore, the inhomogeneity of the plate limits the degree to which the heat flux can be reduced. This inhomogeneity arises in the mechanical design of conventional high heat flux divertors, such as on ITER. The divertor surfaces of these designs are composed of many so-called "monoblocks" that are not perfectly aligned. For ITER, due to this inhomogeneity, the minimum allowed angle θ is about 2 to 3 degrees. Improved mechanical designs (such as those proposed by General Atomics for fusion facilities) allow θ to be lower: about one degree. This angle will provide a lower heat flux. However, even such a small angle may not be enough to reduce the required heat flux.

[0172] It is therefore desirable to have a divertor plate that avoids inhomogeneities, ie that is smoother, so that the heat can be dispersed even more.

[0173] To this end, the invention uses the basic fact that liquid surfaces are smooth, as long as they do not flow too fast and become turbulent. Figure 3 , Figure 4 and Figure 5 The liquid surface 106 on the divertor target 105 in the divertor. This helps solve SQL problem 1. A smooth liquid surface appears on a very slowly moving liquid sheet that is nearly horizontal. Additionally, it is known in the art that magnetic fields make liquid metal flows more laminar, which can also enhance smoothness. Liquid metal flowing sufficiently slowly in a magnetic field can maintain high smoothness, even to some extent or in a direction that deviates from pure horizontal. Thus, in some embodiments, the present invention can have a liquid surface flowing on a divertor target.

[0174] As is known in the art, a capillary pore system (hereinafter referred to as CPS), if made of a fine enough mesh, can also have a very flat liquid surface. Therefore, in other embodiments, the present invention can also use liquid surfaces in this way, i.e., using liquids in CPS.

[0175] In summary therefore, the present invention uses a liquid surface which has the advantage of allowing the surface to be smoother to dissipate more heat. This advantageously works with another advantage of liquids mentioned previously: liquids allow replenishing rapidly eroded surfaces, such as divertor targets.

[0176] Another way to make a smooth surface is to have molten metal covering the divertor target and slowly cool it to solidify. In other words, the divertor target does not need to be liquid all the time. It can be liquid only some of the time to allow for the creation of a very smooth solidified surface.

[0177] In some embodiments, the heat that strikes the divertor target can be removed by cooling from the back of the divertor target facing the SOL surface. This is well known in the art for solid divertor targets. In other embodiments, if practical means become available to create a fast flow of liquid metal in the divertor to remove heat, this can also be used for that surface. In yet another embodiment, a combination of these two approaches can also be used.

[0178] 4) Many of the sputtered or evaporated impurities are completely neutral and travel in essentially straight lines (they do not respond to magnetic fields). These impurities originate primarily from the divertor target 105, or from the material covering the divertor target 106. These impurities can enter the MC plasma 101 and contaminate it. Therefore, an additional aspect of the present invention is to avoid line of sight from the divertor target to the MC plasma by using a material surface to block them. We call this the shield 109, which can be seen in Figure 3 and Figure 4 .

[0179] In some embodiments, such as Figure 3 , shield 109 is an extension of lid 107. The lid already blocks many lines of sight from the divertor target to the MC plasma, and the shield blocks additional lines of sight from the divertor target to the MC plasma. Figure 4 ), the cover 107 does not block the view of the MC plasma, but the shield 109 can do this.

[0180] By blocking the sputtered and evaporated material, the cover 107 and shield 109 also prevent the undesirable accumulation of such redeposited material in areas around the plasma chamber, thereby preventing material from accumulating on the walls around the MC plasma or in other areas that could cause problems.

[0181] The cover must also have an opening that allows particles to escape and be pumped. This opening is the entrance to a tube 172 and is connected to a tube 170, which has a pumping device 171 at the end. These particles that must be pumped may include helium, and may also include hydrogen particles, and may also include other particles.

[0182] 5) The cover 107 and shield 109 described in 4) may have other functions in addition to blocking the sputtering and evaporation of material by the MC plasma. These functions may be achieved by a further aspect of the invention, namely that the liquid surface 108 on the cover 107 faces the SOL, and the liquid surface 111 on the shield 109 faces the SOL. The material facing the SOL must be liquid at least some of the time and at least on some surfaces. Liquids may have several benefits, which we now describe.

[0183] If the liquid is lithium, one benefit is gained by having it chemically bond with the recovered hydrogen. This can produce a low density, high temperature SOL. Since neither the cover nor the shield is subject to the huge heat flux on the divertor target, it is much easier to maintain the lithium in a temperature range where chemical binding is very strong and evaporation and temperature-dependent sputtering are low. Therefore, placing the liquid lithium surface 108 or 111 at a location other than the divertor target has important advantages.

[0184] Liquids can also bring other benefits. These functions can be obtained by liquids (such as molten metals, including but not limited to lithium tin, gallium, lead and alloys containing these metals). To understand the function of these liquids, it is important to realize that the covers and shields will be subject to the redeposited sputtered material from the divertor target. Therefore, they can accumulate the redeposited material to a problematic thickness over time. The liquids 108 and 111 will allow the accumulated material to be removed, for example, by pumping the liquid through a tube without replacing any solid part of the divertor. The flow rate required for this is quite low. As mentioned earlier, replacing the solid part is usually very time consuming and is estimated to take several months. Removing material through the tube is faster and simpler, and does not require shutdown for maintenance.

[0185] An additional benefit that liquids 108 and 111 can impart is that, even if flowing slowly, the liquids can be easily replenished. This is important because when a high temperature SOL is nearby, there can be considerable erosion due to the generation of some high energy neutrals by charge exchange between recycled neutrals and the SOL plasma. The high energy neutrals will strike and erode the material of the cover 107 and shield 109. By having liquid 108 on the surface of the cover 107 or liquid 111 on the surface of the shield 109, the eroded material can be easily replenished.

[0186] Additionally, transient events in MC plasmas can lead to erosion due to deposition of high energy flux on divertor targets, shields. By covering these components with liquid, the material can be replenished with a slow liquid flow.

[0187] The liquid surface may flow continuously. Redeposited material will then dissolve in the liquid and be removed continuously. Or the surface may be solid most of the time, and the liquid will flow only in relatively short periods to periodically dissolve and remove accumulated material.

[0188] 6) Aspect 3 above indicates the advantages of having a liquid divertor target. However, a solid target is still possible. However, the life of the target will be limited due to erosion. Therefore, the divertor target can be solid most of the time, but will only be liquid when the liquid solidifies to allow replacement of eroded material and smoothing the surface again. This can happen during brief shutdowns of MC plasma operations. For example, supplementary material can be introduced through a tube while maintaining the divertor target in place.

[0189] This technique allows replenishing and smoothing the surface of the solid divertor target without removing it, and much faster than removing the solid target, which is estimated to take several months.

[0190] 7) All aspects described above describe physical arrangements that are suitable for placing neutron shielding around the divertor region. To understand this, note that the divertor target must be located quite far away because of the greater magnetic field strength than point X in the region. It is also true that the diagonal magnetic field requirements described above tend to place the divertor target far from the plasma point X. Either of these considerations means that the divertor target 105 is located so that there is space between the divertor target and the MC plasma 101, and the neutron absorbing material 110 can be placed in this space. See Figure 3 and Figure 4 .

[0191] For deuterium-tritium reactors, the neutron shield can be a neutron absorber Li 6 Other neutron shielding devices may also be used, either in conjunction with or in place of the lithium breeder blanket, including but not limited to materials containing B 10 In addition to being located between the divertor region and the MC plasma, the material may also partially surround the divertor region.

[0192] In addition, the neutron absorber 110 will reduce the neutron flux from the MC plasma 101 to the divertor target 105. Figure 3 and Figure 4 Several major benefits can be brought to the divertor target.

[0193] One of these benefits is that very high thermal conductivity materials can be used, which carry away heat more efficiently, but which are rapidly degraded by neutrons. Some examples of such materials known in the art include, but are not limited to, copper and copper alloys, carbon, graphite and carbon composites, and silicon carbide and silicon carbide composites. They have high thermal conductivity and other desirable properties, but degrade rapidly in fusion neutron flux. However, since the divertor target in the present invention can have a large amount of neutron shielding, it is possible to use these and other materials. These materials can remove high heat fluxes more efficiently and can have other advantages, such as corrosion resistance, etc.

[0194] Low neutron flux also makes it possible to use better coolants in the divertor region. As mentioned above, the water coolant used in ITER, due to its efficiency, will bring serious risks of lithium fire accidents if lithium is present in the divertor region. If the neutron bombardment is small, various low melting point liquids become available, which are not dangerous in contact with lithium. These liquids will be degraded by large neutron flux. These liquids include but are not limited to silicon-based liquids, mineral oils, synthetic oils, some molten salts or other liquids. These liquids have desirable properties, but when exposed to neutrons, they will either degrade rapidly or produce undesirable radioactive byproducts. Therefore, by placing the divertor target in a low neutron flux area, it becomes possible to use these coolants.

[0195] Due to these aspects in 7) and possibly 3), it may be easier to employ lithium divertor targets despite the low surface temperature limit.

[0196] 8) The object of the present invention is to create a SOL where the average electron energy is higher than is possible in conventional devices. We therefore summarize this as an independent claim.

[0197] In some embodiments, the electron temperature immediately adjacent to the divertor target must exceed 25 eV. Temperatures in this range will result in unacceptable erosion of the solid target operating at high duty cycles. We note that much higher temperatures are possible, but the SOL temperatures in the divertor region are already beyond the reach of standard technology for fusion-related conditions and cannot be operated for extended periods of time. In other embodiments, the LCFS temperature next to the main plasma is 200 eV or higher. Therefore, we describe this requirement as the electron temperature on the last closed flux surface of the MC plasma being above 200 eV, or the electron temperature immediately adjacent to the divertor target reaching 25 eV.

[0198] We note that it is sometimes technically difficult to measure the SOL electron temperature accurately. It is easier to measure the SOL temperature near the LCFS next to the MC plasma. This is another reason to describe this requirement as the electron temperature of the LCFS near the MC plasma, in addition to being the SOL temperature next to the divertor target.

[0199] In other embodiments, the density of the MC plasma in the LCFS is much smaller than the line average density of the MC plasma. This is also expected for high energy confinement. We describe this by the most easily measured quantity, that is, the ratio of the electron density on the last closed flux surface to the line average electron density of the MC plasma is 0.2 or less.

[0200] Before proceeding to the claims, we mention two important points. 1) It is hypothetically conceivable that some future flowing liquid surface device that could solve the surface heat flux problem becomes feasible. We will refer to this hypothetical future technology as the FLSM. It is important to recognize that even if the FLSM were somehow realized, it would not solve all five SOL problems, but perhaps only problems #1 and #3. Therefore, the present invention would still require a working fusion reactor or a neutron source for transmutation. Aspects of the present invention and the claims would still be necessary. The hypothetical FLSM would operate only on a divertor target as one specific example among many possibilities, referred to in the claims as a divertor target, with some of the divertor target surfaces being covered by liquid at least some of the time.

[0201] 2) Commercially, the applications of MC plasmas can be divided into two categories with different goals. The first type has a high duty cycle, that is, the MC device is operated for a large part of the time. Devices that produce useful energy or neutron sources for transmutation belong to the high duty cycle type. The second type of devices have a low duty cycle. They can be operated in a pulsed mode - the pulse lasts for a few seconds or minutes, followed by a long period of inactivity. Devices with a low duty cycle are usually research devices that are prototypes for the development of devices of the first type. Examples of commercial devices with a low duty cycle are the SPARC to be built by Commonwealth Fusion Systems and the ST40 currently in operation by Tokamak Energy Ltd. Government-operated research devices (such as ITER) also belong to the second type.

[0202] For low duty cycle equipment, target erosion (problem #3) and redeposition (problem #4) may not be as serious because these problems accumulate over days of continuous operation of the divertor target. However, the heat flux problem (problem #1) and the impurity accumulation problem (problem #2) are still critical. And depending on the research equipment, helium removal (problem #5) may or may not be important for the research equipment. It is definitely important for ITER, but not for ST40, and is somewhat important for SPARC, but not critically important. In addition, for low duty cycle equipment, some equipment produces a large neutron flux (e.g., ITER and SPARC), which can have an impact on the pumping device, while some equipment does not (e.g., ST40). Recall that as discussed above, neutrons can be a big problem for the pumping device, so it is very beneficial to place neutron absorbing material between the neutron source and the pumping tube inlet. These are not important for research equipment that does not produce neutrons.

[0203] We can apply the present invention to both low duty cycle and high duty cycle devices. The latter must have more aspects than the former in order to better solve problems #3 and #4. And while critical for high duty cycle devices, problem #5 is not important for some research devices.

[0204] The claims below are organized to reflect this, with some dependent claims being essential to type 1) devices but not to type 2) devices.

Claims

1. A toroidal confined plasma container, comprising: Toroidal plasma chamber; a region of magnetically confined plasma in which particles moving along the magnetic field essentially never strike the walls; wherein the magnetically confined plasma region is substantially symmetrical by rotation about a central axis; a plurality of magnetic field coils; a divertor assembly having a divertor target; wherein a plurality of magnetic field coils are configured to provide at least one X-point and direct plasma particles from the magnetic confinement region to the divertor target; wherein the divertor target has a cover, wherein a side of the cover substantially facing the divertor target includes a material that is liquid on at least some surfaces of the side of the cover during at least some of the time when the cover is in the toroidal confined plasma vessel; wherein the total magnetic field intensity (including all components of the magnetic field) at the divertor target is lower than the total magnetic field intensity (including all components of the magnetic field) at a position between the divertor target and point X on the last closed flux surface closest to the divertor target; Therefore, at least one of the following items: radiation from the magnetically confined plasma does not increase with time until the fusion rate in the magnetically confined plasma decreases by 40% or until the maximum plasma temperature in the magnetically confined plasma decreases by 40%, the power radiated from the magnetically confined plasma by photons does not exceed 70% of the heating power (wherein the heating power is the sum of externally applied heating plus the heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is lower than 3 (wherein the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.6 times the sum of the charges of all electrons in the magnetically confined plasma.

2. The toroidal confined plasma vessel of claim 1, wherein the divertor target has a shield that substantially blocks line of sight from the divertor target to the magnetically confined plasma region and to important components that maintain operation of the device.

3. The toroidal confined plasma container according to claim 2, wherein: The shield, which substantially blocks line of sight from the divertor target to the magnetic confinement region and to components important to maintaining device operation to maintain device operation, is covered by liquid on a side substantially facing the divertor target during at least some of the time the cover is within the annular magnetic confinement plasma vessel.

4. The toroidal confined plasma vessel of claim 1 , wherein the component of the angular magnetic field is the component of the magnetic field on a plane perpendicular to the rotation direction of the central axis, and the amplitude of the component of the angular magnetic field at the divertor target is greater than one third of the maximum value of the angular magnetic field at the boundary of the magnetic confinement plasma region.

5. The toroidal confined plasma vessel of claim 1, wherein the neutron absorbing and moderating material is located substantially between the magnetically confined plasma region and the divertor target.

6. The toroidal confined plasma vessel of claim 1 wherein said divertor target surface comprises a material that is liquid on at least some of the surface at least some of the time.

7. The toroidal confined plasma vessel of claim 1 , further comprising a pumping pipe extending from a position near the divertor target to a pumping device to pump helium, hydrogen isotopes, other gases, or any combination thereof, and wherein a distance from the divertor target to the pumping device is less than one-half of a distance from the X point to the pumping device.

8. The toroidal confined plasma vessel of claim 1 , wherein at least one of the following: the electron temperature at the boundary of the magnetic confinement region is higher than 200 eV or the electron temperature immediately adjacent to the divertor target is higher than 25 eV, and the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.

2.

9. The toroidal confined plasma vessel of claim 1 , wherein at least one of the following: the electron temperature at the boundary of the magnetic confinement region is greater than 1000 eV or the electron temperature immediately adjacent to the divertor target is greater than 100 eV, and wherein the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.

15.

10. The toroidal confined plasma vessel of claim 1 , wherein at least one of the following: radiation from the magnetically confined plasma does not increase over time until the fusion rate in the magnetically confined plasma decreases by 20% or until the maximum plasma temperature in the magnetically confined plasma decreases by 20%, the power radiated from the magnetically confined plasma by photons does not exceed 50% of the heating power (wherein the heating power is the sum of externally applied heating plus heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is less than 2.5 (wherein the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.75 times the sum of the charges of all the electrons in the magnetically confined plasma.

11. A toroidal confined plasma container, comprising: Toroidal plasma chamber; a region of magnetically confined plasma in which particles moving along the magnetic field essentially never strike the walls; wherein the magnetically confined plasma region is substantially symmetrical by rotation about a central axis; a plurality of magnetic field coils; a divertor assembly having a divertor target; wherein a plurality of magnetic field coils are configured to provide at least one X-point and direct plasma particles from the magnetic confinement region to the divertor target; wherein at least one of the following: the electron temperature at the boundary of the magnetic confinement region is higher than 200 eV, the electron temperature immediately adjacent to the divertor target is higher than 25 eV, and the ratio of the plasma electron density at the last closed flux surface to the linear average electron density of a chord passing through the vicinity of the center of the magnetically confined plasma is less than 0.2; wherein the divertor target comprises, on a surface facing the plasma, a material that is liquid at least some of the time and has a composition having a lithium content of less than 50% by atomic fraction; wherein the total magnetic field intensity (including all components of the magnetic field) at the divertor target is lower than the total magnetic field intensity (including all components of the magnetic field) at a location between the divertor target and point X on the last closed flux surface closest to the divertor target in SOL; Therefore, at least one of the following items: the radiation from the magnetically confined plasma does not increase with time until the fusion rate in the magnetically confined plasma decreases by 40% or until the maximum plasma temperature in the magnetically confined plasma decreases by 40%, the power radiated from the magnetically confined plasma by photons does not exceed 70% of the heating power (wherein the heating power is the sum of externally applied heating plus the heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is lower than 3 (wherein the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.6 times the sum of the charges of all electrons in the magnetically confined plasma.

12. The toroidal confined plasma vessel of claim 11, wherein the divertor target comprises a cover, wherein a side of the cover substantially facing the divertor target comprises a material that is liquid on at least some surface of the side of the cover.

13. The toroidal confined plasma vessel of claim 11, wherein the neutron absorbing and moderating material is located substantially between the magnetically confined plasma region and the divertor target.

14. The toroidal confined plasma vessel of claim 11 , further comprising a pumping pipe extending from a position near the divertor target to a pumping device to pump helium or hydrogen isotopes, other gases or any combination thereof, and wherein a distance from the divertor target to the pumping device is less than one-half of a distance from the X point to the pumping device.

15. The toroidal confined plasma vessel of claim 11, wherein the component of the angular magnetic field is the component of the magnetic field on a plane perpendicular to the rotation direction of the central axis, and the amplitude of the component of the angular magnetic field at the divertor target is greater than one third of the maximum value of the angular magnetic field at the boundary of the magnetic confinement plasma region.

16. The toroidal confined plasma vessel of claim 11, said shield substantially blocking line of sight from said divertor target to the magnetic confinement region and to components important for maintaining operation of the apparatus being covered by liquid on a side substantially facing said divertor target.

17. The toroidal confined plasma vessel of claim 11, wherein at least one of the following: the electron temperature at the boundary of the magnetic confinement region is greater than 1000 eV or the electron temperature immediately adjacent to the divertor target is greater than 50 eV, and wherein the ratio of the plasma electron density at the last closed flux surface to the line-averaged electron density of a chord passing near the center of the magnetically confined plasma is less than 0.

15.

18. The toroidal confined plasma vessel of claim 11, wherein at least one of the following: radiation from the magnetically confined plasma does not increase over time until the fusion rate in the magnetically confined plasma decreases by 20% or until the maximum plasma temperature in the magnetically confined plasma decreases by 20%, the power radiated from the magnetically confined plasma by photons does not exceed 50% of the heating power (wherein the heating power is the sum of externally applied heating plus heating produced by nuclear reactions in the magnetically confined plasma), the effective Z is less than 2.5 (wherein the effective Z is defined as the ratio whose numerator is the sum of all ions in the magnetically confined region multiplied by the square of the charge state of the ions and whose denominator is the total number of electrons in the magnetically confined plasma), and the sum of the charges of the fusion fuel ions in the magnetically confined plasma is greater than 0.75 times the sum of the charges of all the electrons in the magnetically confined plasma.

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