Partial ionization centrifuge enabling differential pumping and lithium isotope separation for fusion energy

Abstract

A fuel processing system for a magnetic confinement fusion device, such as a fusion power reactor, is provided. The processing system includes a plasma centrifuge differential pump having a single outlet. When connected to an output of the fusion containment device, such as via a divertor, the plasma centrifuge differential pump maintains a pressure within the fusion device, thus increasing the tritium burn efficiency. The plasma centrifuge differential pump also allow for separation and removal of helium and other materials from the plasma used by the fusion device.

Description

Docket No. 165704-000310PCPARTIAL IONIZATION CENTRIFUGE ENABLING DIFFERENTIAL PUMPING AND LITHIUM ISOTOPE SEPARATION FOR FUSION ENERGYCROSS-REFERENCE TO RELATED APPLICATIONS

[0001] This application is a non-provisional of, and claims the priority benefit of U. S.Application No. 63 / 737,248, filed December 20, 2024, the disclosure of which is incorporated by reference in its entirety.BACKGROUND

[0002] Some of the most critical challenges to building and scaling pilot fusion power plants are driven by the low Tritium Burn Efficiency (TBE) expected in conventional fusion designs. Low TBE results in large recirculating flows of tritium that must be reprocessed and current technologies for this purpose are slow, resulting in large tritium inventories. This onsite inventory is expected to be on the scale of multiple kilograms for a 500 MWth reactor if existing fuel processing technologies are used. Currently, there are only about 20 kg of tritium worldwide, meaning large tritium inventories are inordinately expensive (~$50M / kg) and a major limitation on commercial fusion deployment. However, existing technologies exhibit a significant and costly decrease in fusion power density when operating at higher TBE due to limitations inherent in conventional designs.BRIEF DESCRITION OF THE DRAWINGS

[0003] The accompanying drawings, which are included to provide a further understanding of the disclosed subject matter, are incorporated in and constitute a part of this specification. The drawings also illustrate implementations of the disclosed subject matter and together with the detailed description explain the principles of implementations of the disclosed subject matter. No attempt is made to show structural details in more detail than can be necessary for a fundamental understanding of the disclosed subject matter and various ways in which it can be practiced.Docket No. 165704-000310PC

[0004] FIGS. 1 A and IB show a schematic representation of a plasma centrifuge with an outlet at the outer radius to achieve differential pumping according to embodiments disclosed herein. FIG. 1 A shows a side profile view; FIG. IB shows a top view.

[0005] FIGS. 2A and 2B show schematic representations of example systems including a magnetic containment fusion device and a fuel processing system as disclosed herein. FIG. 2A shows example helium / tritium densities for embodiments disclosed herein.

[0006] FIG. 3 shows the average pressure expressed as a multiple of the inlet pressure in a centrifuge differential pump as disclosed herein.

[0007] FIG. 4 shows an example plasma centrifuge configured for lithium separation according to embodiments disclosed herein.DETAILED DESCRIPTION

[0008] Embodiments disclosed herein provide a partial ionization centrifuge and methods of operating the same, to allow for separating various isotopes with relatively large separation factors and throughputs, specifically for light isotopes relevant to fusion applications. The embodiments disclosed herein have two implications for the technical and economic viability of fusion energy. First, as a differential fuel pumping system capable of accessing high Tritium Burn Efficiency (TBE) operation of 5%, 10%, 15%, 20% or more, in a fusion device at a fusion power density fraction of 90% or more, compared to 1-5% TBE in conventional fusion fuel system designs. For example, differential pumping of helium ash from the system provides a unique opportunity to improve TBE while maintaining high fusion power density. Second, embodiments disclosed herein may enable lithium isotope separation with high throughput and separation factor greater than 2, allowing for more effective breeding blankets in fusion devices, as well as providing a pathway for Li7enrichment for fission applications.

[0009] Achieving a sufficiently high tritium breeding ratio (TBR) in the blanket to achieve self-sufficiency while generating excess fuel for subsequent reactors has been a major challenge in the commercialization of conventional fusion power plants. Breeding blankets are responsible for generating the requisite tritium by using the neutrons from the fusion reaction to breed tritiumDocket No. 165704-000310PCfrom lithium. Though Li6and Li7both contribute, Li6provides the bulk of tritium generation due to its large (n,t) cross section for thermal neutrons. Because Li6has a natural abundance of just -7.5%, there is a large opportunity to increase enrichment of this isotope which would in turn relax requirements on blanket performance.

[0010] Two of the most significant challenges to scaling fusion from breakeven experiments to the first power plants are related to the fuel system. Gas-phase processing of deuterium and tritium fuel require rapid separation of these isotopes from impurities - most critically the helium generated by the D-T fusion reaction. In a typical device, it is expected that the TBE will be low, such that small fractions (-1-5%) of injected fuel atoms are burned, so it is generally expected that such reactors require large recirculating flows of tritium and. Therefore, a mechanism to preferentially pump helium at higher rates than deuterium and tritium may significantly increase the TBE, while also decreasing the level of required recirculating tritium flows and inventory, and mitigating any decreases in fusion power that would otherwise result from high helium concentration in the fusion plasma.

[0011] A second challenge relates to the breeding blanket of the fusion reactor, in which neutrons from the fusion reaction are used to generate tritium through reactions with Li nuclei. The cross section for this reaction is highest for Li6, with a natural abundance of -7.5%. While enrichment with Li6would improve blanket performance, large quantities are required, necessitating high throughput isotope separation. For example, it is expected that on the order of 50,000 kg of Li6are required for regular operation of a 500 MWh reactor.

[0012] Embodiments disclosed herein provide techniques to solve both of these critical light element separation challenges through novel implementations and arrangements of a partial ionization plasma centrifuge. Such a device may be uniquely suited to achieving large separation factors and high throughputs for the light isotopes that are most relevant to fusion energy.

[0013] As described in further detail below, a plasma centrifuge differential pump disclosed herein, such as shown in FIGS. 1A and IB, may be used to implement a differential pumping scheme in, or connected to, the divertor of a fusion reactor to preferentially extract heavy species.Docket No. 165704-000310PC

[0014] FIG. 2A shows an example of a a magnetic containment fusion device such as the torus of a fusion reactor, with an integrated fuel processing system as disclosed herein. The relative helium and tritium densities at various points in the system are indicated. As shown, a plasma centrifuge differential pump receives the spent or partially-spent fuel, such as from the divertor of a fusion reactor torus. The plasma output includes helium ash (i.e., helium as the byproduct of the hydrogen fusion process) and tritium. This is processed by the plasma centrifuge as disclosed in further detail herein, to separate the high-purity stream of hydrogen for reuse in the fusion reactor from helium-rich exhaust which can be further processed within the system. For example, a metal foil pump and / or other components may be used to process the exhaust. Other arrangements of fuel processing systems may be used, such as those disclosed in co-pending PCT Application No. PCT / US2025 / 046813, filed September 17, 2025, the disclosure of which is incorporated by reference in its entirety. As shown in FIG. 2B, a plasma centrifuge differential pump as disclosed herein may be used in conjunction with other processing components, such as a plasma centrifuge (“Marathon Plasma Centrifuge”) used to separate various fuel components as disclosed in PCT / US2025 / 046813, superpermeable pumps, such as the “metal foil pump” shown in FIG. 2B and as disclosed in further detail in co-pending PCT Application No. PCT / US2025 / 046796, the disclosure of which is incorporated by reference in its entirety, and other components as shown.

[0015] The operating principle of the plasma centrifuge differential pump in the fuel processing system as disclosed herein can be understood by examining the flow rates and densities for each species, as shown in FIG. 2A. An important high-level metric is the tritium burn efficiency (TBE), which can be expressed as the ratio of the tritium burn rate to the tritium exhaust rate. In terms of tritium and helium densities at the outlet where the gas is extracted from the system gives:where Z is the differential pumping factor, defined as ST / SHC, the ratio of volumetric pumping speeds for tritium and helium. As can be seen from this expression, because the tritium burn rate is fixed by the thermal power of the reactor, increasing the ratio of helium removal rate to tritiumDocket No. 165704-000310PCremoval rate in order to reduce the amount of tritium removed for each helium atom removed is the only way to increase tritium burn efficiency.

[0016] Observing that the ratio of tritium densities at the outlet in the case where the outlet is at the edge of a plasma centrifuge differential pump, and noting that the tritium burn efficiency can be expressed in terms of the compression of helium-4 relative to tritium,1TBE

[0017] Where a is the separation factor of the differential pump, as explained in further detail below. To first order and for purposes of the present disclosure, the ratio nr / nue can be assumed to be the same as that in the core of the plasma and fixed, though in practice, differential pumping, complex plasma physics processes, and other choices in the system may affect this.

[0018] The differential pump can then be understood to operate by effectively increasing the helium to tritium ratio at the outlet where gas is extracted. In this way, the volumetric pumping speed ratio can be maintained at one while still achieving the desired removal of helium at a higher relative rate than hydrogen.

[0019] Referring again to FIGS. 1A-1B, the principle of the plasma centrifuge is to make use of physical separation, similar to an ultracentrifuge approach used in uranium isotope separation but using a “plasma rotor” to drive an azimuthal flow through J x B forces as shown, with ions in a partially-ionized gas transferring momentum to a mostly neutral background. Centrifugal forces create a gradient in the radial pressure profile, generating a separation factor a between two gas species. Higher mass species, such as helium-4, achieve a higher compression ratio and can be preferentially extracted at the outer edge. As explained in further detail below, this effect can be used to remove the higher-mass species from the system while maintaining a desired pressure of other species, such as tritium, in the torus of the fusion device.

[0020] For an axisymmetric rotation profile, the separation factor a is given by:nr=a ve2dra=(n1 / n2)r=ar=0 kT r TDocket No. 165704-000310PCwhere n, is the density of species i, Mt is the atomic mass of the species, ve represents the azimuthal velocity of the rotating gas, T is the temperature, and r = 0, r = a correspond to the axis and outer radius of the device at the outlet of the centrifuge, respectively. Under simplified assumptions about the rotation profile, this expression reduces to( ni \o. ~ ™ e 2A T / ni \\ ns ) r—a

[0021] The key limitations on the separation factors achievable through this approach are greatly relaxed for lower mass species, such as those used in a fusion reactor system, due to much higher velocities being achievable before ionization occurs. This key observation may be used to improve the applications in fusion fuel processing and lithium isotope separation as disclosed herein, two areas where mass-based isotope separation of light species is critically important.

[0022] To model operation of the device, a rigid body rotation profile is assumed. The ratio of helium (He) to tritium (T) pumping is given by:Nr 2"He, out _ f Sjie (jnne-m^ve^a)NT, outJNeT,div ST1[ 2kT ]’ where N is the flow rate (number per second) of a given species, fneT, Avis the ratio of helium to tritium density in the divertor, ve(a) is the azimuthal velocity at the outer radius a, and S is the volumetric pumping rate of a given species.

[0023] Accordingly, the critical figure of merit for separation efficiency is the separation factor a for both the differential pumping and the Li separation applications disclosed herein.

[0024] To maximize a it becomes necessary to achieve high azimuthal velocity while minimizing the gas temperature. However, there is a limit at which the kinetic energy of an ion is equal to the ionization potential of the neutral gas, known as the Alfven critical ionization velocity. At this point, attempting to increase the rotational velocity of the gas in the system only results in higher ionization fraction as opposed to increased velocity and separation. While higher velocities can in principle be achieved in a fully ionized plasma, higher temperatures limitDocket No. 165704-000310PCseparation factors and introduce large engineering challenges. Accordingly, embodiments disclosed herein focus more on minimizing heating while driving engineering improvements.

[0025] According to embodiments disclosed herein, high separation factors and long-term device operation are achieved by optimizing the plasma driving the separation and operating at relatively high pressures (—0.1-1 torr upstream) for high throughput.

[0026] Two of the relevant parameters relating TBE and fusion power density are the differential pumping parameter S and the helium enrichment factor / / He, defined as S = A'lie / . S'o and / He = file, div If a, core, respectively. Here, Ni le is the effective pumping speed of He ash, SQ is the effective pumping speed of unburned hydrogenic species,. / He, div is the relative density of He ash to hydrogenic fuel in the divertor and / a, core is the relative density of He to fuel in the plasma core. S also plays an equivalent role in the relative pumping of He as a, so increases in S offer another way to achieve the same benefits as increases in a.

[0027] The performance characteristics required by a differential pumping system are related to the fusion power density in a fusion device by“2_ L— 3Sr / He ( 1 ~~ ) J

[0028] where Pt / f,max is the achieved fusion power relative to a plasma with zero He dilution and a 50:50 D-T mix. It has been shown in D. G. Whyte et al., “Tritium bum efficiency in deuterium-tritium magnetic fusion,” Nuclear Fusion, vol. 63, 2023, doi: 10.1088 / 1741-4326 / acf3fb that differential pumping can contribute to improvements in fusion power density and maximum TBE. As shown herein, a partial ionization centrifuge can provide such pumping and thus result in fusion reactor TBE rates of 5%, 10%, 15%, 20% or more and Pt! Pt, max of 75%, 80%, 85%, 90%, or more. This performance represents roughly an order of magnitude improvement in TBE over conventional systems and existing assumptions, and correspondingly about an order of magnitude decrease in tritium flow rates and inventories in the processing system - both significant improvements over conventional fusion fuel processing system designs.Docket No. 165704-000310PC

[0029] Assuming a typical value of / / He ~ 1, TBE = 20% corresponds to a fusion power density only 64% of the maximum possible for S = 1, but > 90% for S = 5, or equivalently a = 5. As a result, a = 5 may be used as a performance target for embodiments disclosed herein.

[0030] For the differential pumping application, flow rates of gas out the divertor can be estimated based on the ratio of the target bum rate mturn set by the fusion power output to the TBE: ihm. T = mbumlTBE. Assuming the TBE = 20% value and equal injection rates for deuterium and tritium, the total injection rate of gas into (and out of) the system in a 500 MWth reactor is 1.78X1021atoms / s, or 7.37X1 O'3g / s. In vacuum flow units, this corresponds to 3.65 Pa-m3 / s. Embodiments disclosed herein demonstrate throughputs of at least 7.4X10'4g / s, within one order of magnitude of a commercially-viable target.

[0031] To evaluate the performance of a plasma centrifuge used for lithium separation as disclosed herein, we can compare the needs to those of an ARC-like power plant using some of the design parameters from B. Sorbom et al., “ARC: A compact, high-field, fusion nuclear science facility and demonstration power plant with demountable magnets” (Fusion Engineering and Design, vol. 100, pp. 378-405, 2015, issn: 0920-3796, doi:doi.org / 10.1016 / j.fusengdes.2015.07.008) as a point of comparison. AFLiBe-based blanket has a volume of roughly 200 m3and a corresponding Li mass of -50,000 kg; to produce enough lithium for a single device in a year corresponds to a separated Li throughput of -140 kg / day or -1.6 g / s. Due to the additional challenges involved with working with Li metal, embodiments disclosed herein are evaluated at a more modest but still ambitious target of ~0.01g / s or ~lkg / day for production of -15% enriched Li6. This corresponds to a of about 2.

[0032] It is generally accepted in the art that existing fuel cycle technologies are insufficient to scale from current scientific experiments to fusion pilot plants and eventually economically viable reactors, in particular due to large tritium inventories. While high values of the TBE are in principle achievable in magnetic confinement devices, this parameter trades against the fusion power density achievable in the device. High values of TBE correspond to large rates of helium production, which leads to fuel dilution in the fusion plasma and a resulting decrease in fusion power density.Docket No. 165704-000310PC

[0033] When the tritium inventory in a reactor is dominated by the inventory in the gas processing loop, the inventory can be expressed asfHT, inventory fi'lin. T' p lh burn p / TBEwhere TPis the processing time. As such, the order of magnitude increase in TBE described here would correspond to roughly an order of magnitude decrease in inventory growth rate. It should be noted that this model is a simplification since it does not account for tritium inventories in the blanket, which vary depending on blanket fluid and the tritium extraction scheme. Here we consider only the gas processing loop which is agnostic to the blanket concept.

[0034] For either of the approaches described here, there are large benefits in engineering simplicity that arise from improvements in TBE or TBR. To assess these benefits we can assume that for fusion to be able to scale, a fixed doubling time must be achieved to enable a sufficient deployment rate for fusion. For a tritium-limited growth rate we can express the deployed fusion power P fusion at time t asPfusion^ =For a constant doubling time, increases in TBE greatly reduce the requirement on excess TBR, and increases in Li enrichment fraction mean that the TBR required is achievable with a substantially smaller blanket thickness. Increasing the rate of Li6enrichment from 10% to 50% reduces the expected blanket thickness from 40 to 30 cm while keeping a TBR > 1.1. Most of the benefit from Li6enrichment is obtained from -15-20% enrichment. Using common reactor dimensions gives a blanket volume reduction of -15%, and a reduction in the total torus volume of -10%. This is a substantial value, especially when it is noted that the size of magnets used in these systems scales with the stored magnetic energy, which is proportional to the system volume. Alternatively, the increase in baseline TBR achievable with Li6enrichment can be used to relax requirements on structural materials, which could for example help tokamaks handle disruption-driven forces on the first wall of the device.

[0035] Additionally, Li enrichment as disclosed herein benefits from limited effectiveness to high-mass separations which could pose proliferation risks, mitigating concerns present withDocket No. 165704-000310PCsome other approaches. Additional benefits for proliferation resistance in fusion systems are also gained from use of Li6in fusion blankets. Because of the very high cross section of Li6for thermal neutrons, thermal neutrons are more quickly consumed in blankets with higher fractions of Li6and fewer neutrons are available to fertile material that could in principle be added to fusion blankets.

[0036] Embodiments disclosed herein may bypass the need for rapid hydrogen recycling by solving the root of the problem: tritium bum efficiency. Because of the unique challenges in pumping the gas mixture present in the fusion context, there are no other known approaches that offer the ability to differentially pump helium at a higher rate than hydrogen isotopes. There is only a limited set of physical effects that could be used to achieve the desired gas separation in the differential pumping application. Because the D2 molecule has roughly the same mass as a He4atom and a deuterium nucleus has roughly the same charge to mass ratio as a He4nucleus, the partial ionization centrifuge approach disclosed herein represents a set of conditions uniquely suited to operating on an atomized feed supply of hydrogen atoms. As such, plasma centrifugebased differential pumping represents a fundamentally new technology in the field of fusion fuel processing.

[0037] Embodiments disclosed herein provide innovative and novel approaches to light isotope pumping and separation based on the unique capabilities of the partial ionization centrifuge concept. This work reimagines the partial ionization centrifuge concept through four innovations to enable the high throughput separation and pumping of light isotopes. The first two approaches describe systems and techniques that decrease the operating temperature of the plasma to enable high separation factors and long-term operation, while the latter two features describe specific implementations uniquely designed around the fusion application.

[0038] First, a high magnetic field separation region may be used to reduce or minimize resistive heating by the plasma current. The rotational velocity is one of the most critical operational parameters for increasing the separation factors achieved in the device. In a simplified model, the rotational velocity can be assumed to be proportional to the product of plasma current and background magnetic field, v~I*B, and a target separation factor is— ^^7 — | as shown above. While a particular velocity can be achievedDocket No. 165704-000310PCacross a range of currents and magnetic fields, increasing the current also increases the resistive heating of the plasma and so increases the temperature, which thereby decreases the separation factor. As such, it may be preferred to use a relatively high magnetic field, for example, a field of at least 0.1 T, 0.15 T, 0.2 T, or greater than 0.2 T to achieve high rotational velocities at relatively low currents. This also allows less power usage in the arc, which in some applications such as lithium separation as disclosed herein, improves the overall efficiency and economics of the system. In some embodiments superconducting magnets may be used to achieve these fields, or in other embodiments the magnetic fields already present near a magnetic confinement device may be used in order to achieve the desired effect.

[0039] Second, current profiles may be adjusted and optimized to mitigate viscous heating of the gas. For example, a plasma current of at least 100 A, 110 A, 120 A, 130 A, 140 A, 150 A, or more may be used. As previously described, at a fixed rotational velocity the neutral gas temperature is a key parameter that can affect the separation factor. In some regimes the viscous heating of the plasma due to rotation can become an important heating contribution. In these cases, the plasma current profile can be optimized through the use of a radially localized current, extending from an annular electrode, such as an annular hollow cathode, out to another annular anode spaced at a larger radius, but also spaced apart from the wall. An example of such an arrangement is shown in FIG. IB. In this arrangement, the velocity profile Ve of the gas at radii smaller than the electrode pair relaxes to that of a rigid body rotator, in which there is no viscous dissipation in steady state, thus reducing or minimizing the additional heating caused by rotation of the plasma.

[0040] For example, referring to FIG. IB, a hollow cathode and annular anode may be disposed around the central axis of the plasma centrifuge as shown. The hollow cathode may include resistive heating, inductive heating, or the like, so as to decouple arc initiation within the plasma centrifuge from arc heating of the cathode. The hollow cathode may be formed from any suitable material, such as lanthanum hexaboride, lanthanated tungsten, ceriated tungsten, barium oxide, or the like, or combinations thereof.Docket No. 165704-000310PC

[0041] Third, the partial ionization centrifuge may be configured to extract gas only at the large-radius boundary. This preferentially extracts heavy species and functions as a pumping system as previously disclosed.

[0042] Conventional partial ionization centrifuge systems use an inlet and two outlets to achieve separation of a heavy species from a light species, with each outlet being used to extract one of the species from the combined plasma. In embodiments disclosed herein, the plasma centrifuge has a single inlet and only a single outlet, because one of the primary effects of the centrifuge is preferential compression of the heavy species such as helium isotopes, as opposed to a clean separation between two distinct species. Specifically, a plasma centrifuge as disclosed herein preferentially compresses helium over hydrogen isotopes, specifically tritium. The absence of an outlet in the low-pressure central region mitigates or removes challenges around light product extraction, which must normally be achieved with specialized extractors located in regions where there is sufficiently high gas pressure to remove material. In this way, a plasma centrifuge differential pump as disclosed herein is distinct from an isotope separation system that produces two streams of separated materials.

[0043] The plasma centrifuge differential pump also may be located sufficiently close to the divertor of the confinement device that the concentration of species in the divertor should be substantially similar, if not identical to the distribution of species in the inlet of the plasma centrifuge. For example, the inlet of the plasma centrifuge may be disposed not more than about Im, 2m, 3m, 4m, or 5m downstream of an outlet of the divertor of a magnetic confinement fusion device. As used herein, a “downstream” distance such as the distance between the outlet of the divertor and the inlet of the centrifuge refers to the “duct length” or “duct distance” between the two, i.e., the total flow path length of a gas or plasma traveling between the two. The net desirable effect is to achieve local compression only of the species to be removed via the single outlet of the pump.

[0044] In some embodiments the size of the inlet to the plasma centrifuge is selected to maintain a desired upstream pressure in the divertor and a desired operating pressure in the centrifuge. For example, a smaller diameter inlet may be used on the centrifuge than the outlet ofDocket No. 165704-000310PCthe divertor. Such an arrangement may be desirable if the ideal operating pressure of the centrifuge is lower than the ideal operating pressure of the divertor.

[0045] The outlet of the centrifuge may also be throttled by an adjustable conductance element. An additional pump may be placed downstream of the centrifuge beyond this port in order to supply the exhaust mixture to downstream fuel processing systems. This can also be used to set the operating pressure in the centrifuge based on desired plasma parameters in the centrifuge itself.

[0046] Fourth, in embodiments where a partial ionization plasma centrifuge is used as, or as a component of an isotope separation system, additional surfaces may be used within the plasma centrifuge disclosed herein to prevent lithium condensation. In this arrangement, multiple temperature-controlled outlets may be used to allow for in-situ staging and improved separation. Such an arrangement may be used as an alternative, or within a separate differential pump component, to improve lithium separation within the fuel system. Specifically, in applications designed for lithium separation, a plasma centrifuge as shown in FIGS. 1A-1B may include a chamber or sub-chamber that is capable of operating above the temperature at which lithium vapor would condense on the walls. FIG. 4 shows an example of such an arrangement. For example, as shown in FIG. 4 a separate sub-vessel may be implemented inside the main vacuum vessel, and made of material capable of operating at an elevated temperature (for example, 700 C or higher) such that lithium vapor does not condense. The vacuum is used to insulate the inner chamber from the outer chamber, and layers of radiative insulation may be arranged around the sub-vessel to keep the outer vessel cool. The inner vessel can be made of materials that can operate at high temperature with lithium vapor, for example refractory materials such as tantalum, niobium, vanadium, tungsten, molybdenum, and the like. As indicated in FIG. 4, the centrifuge may be operated at temperatures under 300 C, while the sub-vessel is maintained at elevated temperatures of 700 C or more. A lithium separation plasma centrifuge as disclosed herein may be used separately from the fusion fuel processing system and combined systems shown in FIGS. 2A and 2B, for example to process lithium for use in the blanket of the fusion reactor.Docket No. 165704-000310PC

[0047] Embodiments disclosed herein achieve significant improvement in key metrics necessary for commercialization of fusion fuel processing systems, as summarized in Table 1.Table 1: Comparison of present technology performance for light isotope separation vs. example levels disclosed herein and commercially-preferred targets.Metric State-of-the-Art Disclosed Commercial Target Separation factor, He / T R~ 1 R > 5 R > 5 Throughput with N / A 7.4xl0-4g / s 7.4x1 O’3g / s differential pumpingfor He / TLi6Cost per kg $82k / kg (available <$5k / kg (15% <$5k / kg (20%supply) enriched) enriched) Throughput and Poor (very limited 0.01 Ig / s 1.6 g / s (-50,000 scalability for Li supply) (~kg / day) kg / yr, enough for one separation ARC)Separation factor in Li R~ 1.05 (exchange R> 2 R> 2processes, single pass)

[0048] As previously disclosed, to assess feasible separations for the isotopes of interest, the previous expression for the separation factor a may be modified to account for the Alfven Critical Ionization Velocity (CIV) of the gas being processed:(m1-ma2) eV=expion

[0049] where / w.min is the mass of the most easily ionized isotope and Eon is the ionization energy. The limiting species depends on the combination of the ionization energy and the mass of the species. In the present approach we assume that we will operate below the CIV, such that the ionization fraction is primarily set by temperature rather than rotation. With this constraint, the remaining free parameter for the separation is (under the assumption of rigid body rotation) the plasma temperature.

[0050] Under an assumed temperature range of 1-2 eV, a partial ionization centrifuge operating at the CIV would be expected to achieve R ~ 10-93 for the He4 / T separation, and R ~Docket No. 165704-000310PC1.5-2.2 for Li6 / Li7. Because of the much lower ionization energy for lithium, it may be possible to achieve operation at lower temperatures while still maintaining sufficient plasma conductivity.

[0051] Even in unoptimized systems, it is expected that embodiments disclosed herein can provide good performance for differential pumping, with R on the order of 5. To achieve the performance metrics disclosed herein, relatively low plasma currents are used, with increased magnetic fields to achieve the velocity targets. In the plasma centrifuge, the rotational velocity of the plasma scales as v ~ IB / jiz, where I is the plasma current, B is the externally applied magnetic field, and / z is the viscosity of the gas. Resistive dissipation in the plasma leads to deposited power density proportional to 12for fixed plasma parameters, implying that the resistive heating contribution scales like 1 / B2. Overall, the resistive heating represents a substantial fraction of the total heating contribution, in the range of about 8-41% depending on the discharge parameters. Hence, for some operational regimes, this will be an important mitigation to reduce temperatures. Representative plasma temperature estimates are in the 1-3 eV range. Throughput in the device can be estimated by noting that the mass flow rate through the system will be proportional to the density and the volume, and inversely proportional to the residence time in the system. This then requires a reasonable estimate for the operating pressure for the device. By integrating across the pressure profiles achieved through rotation, we can relate the average pressure p to the pressure along the axis po by the expressionp > exp(M')—lPo~ ~ M ’

[0052] where M = ErotlkT. This expression is plotted in FIG. 3, which shows the average pressure expressed as a multiple of the inlet pressure in the system.

[0053] Prior work has typically operated at ~1 torr pressures with good separation factors (R ~ 5 for H / D), and this happens to be in approximately the pressure range at the outlet of a tokamak divertor (~50 Pa or -0.38 torr). Therefore, it is a suitable range to assume for the operating pressure of the centrifuge. Assuming an upstream pressure of 50 Pa, the volume V of the plasma centrifuge required to achieve a desired mass flow rate can be estimated as:V= RT sthlpM,Docket No. 165704-000310PC

[0054] where R is the ideal gas constant, Tis the temperature of the gas, ires is the residence time, M is the molar mass of the species in g / mol, p is the average pressure in Pa, and rii is the mass flow rate in g / s. The residence time ires depends on the time required to equilibrate newly injected gas to the target velocity and density profiles. For simplicity, this can be estimated as ~1 s, noting that this is likely a highly conservative estimate, considering that large separation factors may be achieved in hydrogen after about 2.5 ms of operation.

[0055] For Emi / kT ~ 5, an upstream pressure of 50 Pa, and 1 eV plasma temperature assumed, the partial ionization plasma centrifuge pump as disclosed herein can achieve 10% of the mass flow rate needed by an eventual reactor in ~12 L of plasma volume.

[0056] For Erot / A - 5, an upstream pressure of 1 torr, and 0.3 eV plasma temperature, a lithium separator as disclosed herein can achieve a mass flow rate of -0.011 g / s in -10 L.

[0057] Various aspects or features of this disclosure are described with reference to the drawings, wherein like reference numerals are used to refer to like elements throughout. In this specification, numerous details are set forth in order to provide a thorough understanding of this disclosure. It should be understood, however, that certain aspects of disclosure can be practiced without these specific details, or with other methods, components, materials, or the like. In other instances, well-known structures and devices are shown in block diagram form to facilitate describing the subject disclosure.

[0058] The described embodiments are susceptible to various modifications and alternative forms, and specific examples thereof have been shown by way of example in the drawings and are herein described in detail. It should be understood, however, that the described embodiments are not to be limited to the particular forms or methods disclosed, but to the contrary, the present disclosure is to cover all modifications, equivalents, and alternatives. Additionally, elements of a given embodiment should not be construed to be applicable to only that example embodiment and therefore elements of one example embodiment can be applicable to other embodiments. Additionally, in some embodiments, elements that are specifically shown in some embodiments can be explicitly absent from further embodiments. Accordingly, the recitation of an element being present in one example should be construed to support some embodiments where such an element is explicitly absent.

Claims

Docket No. 165704-000310PC CLAIMSWhat is claimed is:

1. A fuel processing system for a fusion reactor, the fuel processing system comprising: a plasma centrifuge differential pump comprising:an inlet configured to receive a stream of material from a fusion reactor, the stream of material comprising helium and one or more hydrogen isotopes;a plasma centrifuge configured to apply a magnetic field to the stream of material to cause separation of at least some of the helium and at least some of the tritium; anda single outlet from which helium is extracted.

2. The fuel processing system of claim 1, wherein the plasma centrifuge differential pump compresses helium preferentially over the one or more hydrogen isotopes.

3. The fuel processing system of any previous claim, wherein the magnetic field has a strength of at least 0.2 T.

4. The fuel processing system of any previous claim, wherein the plasma centrifuge operates with a plasma current of at least 100 A.

5. The fuel processing system of any previous claim, wherein the plasma centrifuge comprises a central hollow cathode.Docket No. 165704-000310PC 6. The fuel processing system of claim 5, wherein the hollow cathode comprises one or more materials selected from a group consisting of lanthanum hexaboride, lanthanated tungsten, ceriated tungsten, and barium oxide.

7. The fuel processing system of claim 5, wherein the hollow cathode comprises a resistive heating element, an inductive heating element, or a combination thereof arranged and configured to heat the hollow cathode.

8. The fuel processing system of any previous claim, wherein the inlet is in fluid communication with an outlet of a divertor of a fusion reactor.

9. The fuel processing system of any previous claim, further comprising a secondary pump disposed downstream of the single outlet, the secondary pump arranged and configured to extract one or more isotope species from an outlet stream of the plasma centrifuge.

10. The fuel processing system of any previous claim, further comprising a variable conductance element disposed at the single outlet to control an operating pressure of the plasma centrifuge.

11. The fuel processing system of any previous claim, wherein the fuel processing system provides a separation factor between tritium and helium-4 of at least 5.

12. A system comprising:Docket No. 165704-000310PC a magnetic confinement fusion device arranged and configured to generate energy via fusion; anda fuel processing system as recited in claim 1.

13. The system of claim 12, wherein the inlet of the plasma centrifuge is connected to an outlet of a divertor of the magnetic confinement fusion device.

14. The system of any of claims 12-13, wherein the plasma centrifuge of the fuel processing system is arranged not more than 5 m downstream of a divertor of the magnetic confinement fusion device.

15. The system of any of claims 12-14, wherein the inlet of the plasma centrifuge is smaller than an outlet of the divertor.

16. A method of operating a plasma centrifuge as a differential pump within a fusion fuel processing system, the method comprising:receiving a stream of material from a magnetic confinement fusion device, the stream of material comprising tritium and helium;applying a magnetic field to generate an azimuthal flow of the material and cause radial separation of the tritium and helium; anddiverting the helium through a first outlet channel of the plasma centrifuge.

17. The method of claim 16, wherein the stream of material has a throughput of at least 7.4xl0'4g / s.Docket No. 165704-000310PC 18. The method of any of claims 16-17, wherein the fusion reactor operates at a tritium bum efficiency (TBE) of at least 5%.

19. The method of any of claims 16-18, wherein the magnetic field has a strength of at least 0.2 T.

20. The method of any of claims 16-19, wherein the stream of material comprises a fully-dissociated feedstock of hydrogen gas and one or more other fusion impurity gases.

21. The method of any of claims 16-20, wherein the plasma centrifuge is operated below a critical ionization velocity for hydrogen isotopes.

22. A lithium separation plasma centrifuge comprising:a main plasma centrifuge chamber; anda sub-vessel configured operate at a temperature sufficient to prevent condensation of lithium on one or more walls of the sub-chamber.

23. The lithium separation plasma centrifuge of claim 22, wherein the sub-vessel comprises one or more materials selected from a group consisting of: tantalum, niobium, vanadium, tungsten, and molybdenum.

24. The lithium separation plasma centrifuge of any of claims 22-23, wherein the sub-vessel is configured to operate at a temperature of at least 700 C.

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