Spherical Tokamak Reactor Research Articles

Integrated Analysis on the Current Profile and the Operational Scenario of D-<sup>3</sup>He Spherical Tokamak Reactors

Integrated Analysis on the Current Profile and the Operational Scenario of D-<sup>3</sup>He Spherical Tokamak Reactors

Economical and Life-Cycle Energy Assessment of Magnetic Fusion Power Reactors

We analyzed several types of fusion reactors, tokamak (TR), spherical tokamak (ST), helical (HR), and inertial fusion reactor (IR) using physics, engineering and cost (PEC) code, which evaluates economic and lifecycle energy amount quantitatively. We compared the cost of electricity (COE) and the energy payback ratio (EPR) of each fusion reactors with those of fission power plant. Especially, we focus on the EPR of TR with several blanket and shield designs having scarce materials such as silicon carbide (SiC), vanadium alloy (V), and ferritic steel (FS). As the result, we found that the EPR of TR with SiC/LiPb blanket/shield model is the lowest. The COEs and the input energy of TR (βN = 4.0) and IR are lower than those of ST and HR. The COE of fusion reactor is two times higher than that of fission power plants. However the EPR of fusion reactor is as high as that of fission reactor. c

Life Cycle Assessment for Energy Payback of Spherical Tokamak Reactors

Life Cycle Assessment for Energy Payback of Spherical Tokamak Reactors

Assessment of Spherical Tokamak Reactors Comparing with Other Fusion Power Plants

Assessment of Spherical Tokamak Reactors Comparing with Other Fusion Power Plants

A Conceptual Design of Superconducting Spherical Tokamak Reactor

A Conceptual Design of Superconducting Spherical Tokamak Reactor

Conceptual design study of a superconducting spherical tokamak reactor with a self-consistent system analysis code

In a spherical tokamak (ST) reactor, the radial build of toroidal field coil and the shield play a key role in determining the size of the reactor. For self-consistent determination of the reactor components and physics parameters, a system analysis code is coupled with a one-dimensional radiation transport code. A conceptual design study of a compact superconducting ST reactor with an aspect ratio of up to 2.0 is conducted and the optimum radial build is identified. It is shown that the use of an improved shielding material and high-temperature superconducting magnets with high critical current density opens up the possibility of a fusion power plant with compact size and small re-circulating power simultaneously at a low aspect ratio, and that by using an inboard neutron reflector instead of a breeding blanket, tritium self-sufficiency is possible with an outboard blanket only and thus a compact-sized all superconducting coil ST reactor is viable.

Environmental and Economical Assessment of Various Fusion Reactors by the Calculation of CO2 Emission Amounts

We compared the CO2 emissions of several fusion reactors. The magnetic confinement systems evaluated here are the tokamak reactor (TR), helical reactor (HR), and spherical tokamak reactor (ST). These models are calculated by the Physics-Engineering-Cost (PEC) code. The inertial confinement fusion reactor (IR) is also evaluated, assuming its driver energy and driver efficiency. In addition, different blanket modules and fuels are considered in the TR designs. To calculate life-cycle CO2 emission from fusion reactors defined by plasma parameters and radial build, we used a basic unit for CO2 weights (kt-CO2/t-material). Calculation results indicate that CO2 is emitted mainly in the construction stage of superconducting magnet systems for magnetic confinement fusion reactors. For the IR design, the driver system construction and pellet fabrication stages involve considerable CO2 emission. By comparing fusion reactors with other electric power generation systems in terms of CO2 emission, we confirmed that fusion reactors emit less CO2. Therefore, introducing a carbon tax has little effect on the economics of fusion reactors, and the cost of electricity (COE) from fusion reactors might be lower than that of oil-fired electric power plants when a carbon tax of around several hundred yen/t-CO2 is introduced.

Burning plasma simulation and environmental assessment of tokamak, spherical tokamak and helical reactors

Reference 1-GWe DT reactors (tokamak TR-1, spherical tokamak ST-1 and helical HR-1 reactors) are designed using physics, engineering and cost (PEC) code, and their plasma behaviours with internal transport barrier operations are analysed using toroidal transport analysis linkage (TOTAL) code, which clarifies the requirement of deep penetration of pellet fuelling to realize steady-state advanced burning operation. In addition, economical and environmental assessments were performed using extended PEC code, which shows the advantage of high beta tokamak reactors in the cost of electricity (COE) and the advantage of compact spherical tokamak in life-cycle CO2 emission reduction. Comparing with other electric power generation systems, the COE of the fusion reactor is higher than that of the fission reactor, but on the same level as the oil thermal power system. CO2 reduction can be achieved in fusion reactors the same as in the fission reactor. The energy payback ratio of the high-beta tokamak reactor TR-1 could be higher than that of other systems including the fission reactor.

Liquid Lithium Divertor System in a Spherical Tokamak Reactor

The heat flux in the divertor target is one of the most crucial problems for a fusion reactor. This problem is more severe in a spherical tokamak (ST) reactor because it is more compact. This paper proposes a liquid lithium divertor system to solve this problem. The heat coming from the fusion plasma along the divertor leg is removed by evaporation of lithium. The lithium vapor is condensed on the wall, and the heat is transferred to the coolant that drives power generator. Narrow slit along the divertor leg and the differential evacuation chamber reduces leakage of lithium vapor to the plasma chamber.

二酸化炭素排出量の計算による核融合炉の環境負荷評価

Comparative studies of environmental burden have been made for fusion reactors based on different confinement systems and blanket modules by calculating CO2 emission amount. The confinement systems of fusion reactor considered in this paper are the Tokamak reactor (TR), helical reactor (HR), and spherical Tokamak reactor (ST). Several blanket modules such as Li/V, Flibe/FS, and LiPb/SiC blankets are evaluated under the condition of 1 GW electric power output and specific beta values. The calculated amounts of CO2 emission from fusion reactors are 9.2-11.3 g-CO2/kWh. This range is the same as that of emission from hydraulic power and atomic power plants which are regarded as clean energy sources now. A substantial amount of CO2 is emitted from superconducting magnet systems. TR and HR, which use large superconducting coil systems, emit much CO2 as a whole. If we adopt a higher beta design, the demand on coil systems is relaxed and better fusion reactors emitting less CO2 can be constructed.

Relativistic ray-tracing of electron Bernstein waves in a spherical tokamak reactor

Electron Bernstein waves (EBW) can propagate between electron cyclotron harmonics in hot, dense plasmas with no high-density cutoff (unlike electron cyclotron waves). ARIES-ST is a spherical tokamak (ST) reactor study. A promising method of heating and driving current in such an ST reactor would be with electron Bernstein waves. These could be mode converted from microwave vacuum modes (X-mode and O-mode) launched from the edge. In this paper, fully relativistic ray-tracing calculations (both damping and propagation) of EBWs in ARIES-ST are presented, using the relativistic dispersion solver R2D2 and the ray-tracing code GENRAY. The ray paths, damping locations, and polarizations are compared with the more commonly used non-relativistic EBW ray-tracing. A range of frequencies are shown for EBWs with small n∥ that could be produced by the X-B mode conversion process, and EBWs with large n∥ ≈ 0.6, produced by the O-X-B mode conversion process. With the density and temperature profile chosen for this paper, the greatest depth into the core that could be reached with mode-converted EBWs is a radial location of approximately ρ = 0.4. Although the radial location of the damping in most cases was not significantly different between the relativistic and non-relativistic cases, there are differences in the poloidal locations, as well as in the polarizations of the wave along the ray path, especially in the damping region.

Proposed Experiment to Study Relaxation Formation of a Spherical Tokamak with a Plasma Center Column

A spherical tokamak (ST) with a plasma center column (PCC) can be formed via driven magnetic relaxation of a screw pinch plasma. An ST-PCC could in principle eliminate many problems associated with a material center column, a key weakness of the ST reactor concept. This work summarizes the design space for an ST-PCC in terms of flux amplification, aspect ratio, and elongation, based on the zero-β Taylor-relaxed analysis of Tang & Boozer [Phys. Plasmas 13, 042514 (2006)]. The paper will discuss (1) equilibrium and stability properties of the ST-PCC, (2) issues for an engineering design, and (3) key differences between the proposed ST-PCC and the ongoing Proto-Sphera effort in Italy.

Possibility of utilizing the D-3He fuel cycle with 3He production in a spherical tokamak reactor

Possible operating regimes of a spherical tokamak reactor based on the D-3He fuel cycle with 3He production are considered. The parameters of the plasma and magnetic system are calculated for several versions corresponding to the high power efficiency (with a power gain factor in plasma of Q = 20) in a reactor with an aspect ratio of A = 1.5. According to calculations, for an axial magnetic field in vacuum of B 0 = 2 T, a plasma radius of a = 3 m, an average 〈β〉 value of 0.53, and a plasma temperature of 〈T〉 = 48 keV, the reactor power can reach P fus = 500 MW. In order to achieve a power of P fus = 1500 MW in a reactor with a = 2 m, 〈β〉 = 0.36, and 〈T〉 = 40 keV, the magnetic field should be increased to B 0 = 5 T.

Aspect ratio dependencies of D– 3He fueled tokamak reactors

Ignition capability is studied and compared for two types of D– 3He spherical tokamak (ST) and medium aspect ratio tokamak (MT) reactors. The required power fraction to ions for maintaining the hot ion mode, which is indispensable for D– 3He fusion, is smaller in a ST reactor than that in a MT reactor due to the lower synchrotron radiation loss. It is found that the nuclear elastic scattering capable of supplying 40% of the fusion energy to ions is enough to maintain the hot ion mode in the D– 3He ST reactor with the high wall reflectivity. For achieving the D– 3He MT reactor, the power fraction to ions should be additionally increased.

Transmutation of High-level Wastes in a Spherical-tokamak Reactor

We studied transmutation of high-level wastes in a spherical-tokamak reactor. We examined spent fuel, recovered fuel, natural UO2 fuel and reprocessed waste, though the natural UO2 fuel is not a high-level waste. We added a few percent of 239Pu to the former three fuels to increase the tritium breeding ratio and saw the change in the effective multiplication factor keff, number of fissions, multiplication of fusion energy and conversion ratio of 239Pu. We found that 2.5 and 5.5 fissions are needed to obtain the tritium breeding ratio of 1 and 2, respectively. High amplification of fusion energy occurs. Breeding of 239Pu is expected for the natural UO2 fuel even in a subcritical environment. For the reprocessed waste, 93Zr and 99Tc are transmuted efficiently by the (n,γ) reaction and so are 237Np and 241Am by the (n,f) reaction. Fusion neutrons of 100MW will transmute half of those nuclides produced from the generation of 1 GW(e) year in a large pressurized water reactor. However, tritium fuel may not be self-sustained in the reprocessed-waste burner.

Steady Burning Criteria for D-3He Spherical Tokamak Reactor with Internal Transport Barrier

Steady burning criteria for a D-3He fusion reactor have been investigated by using the 0-D power and particle balance equations and the IPB98(y,2) scaling law. A spherical tokamak (ST) configuration with the internal transport barrier (ITB) and the bootstrap current fraction of 100 % are assumed. As main results, the hot ion mode (Ti/Te > 1) and the high central beta are essential to realize a D-3He fusion reactor, and the confinement improvement does not reduce the requirement of the hot ion mode. The reactor size, the magnetic field, confinement improvement factor and the heat flux may be attainable if the hot ion mode is realized.

Plasma Current Rampup by the Outer Vertical Field Coils in a Spherical Tokamak Reactor

To find a solution of the plasma current rampup problem in a low aspect ratio spherical tokamak (ST) reactor, the effect of the outer vertical field coils on plasma current rampup is studied with noninductively driven current and bootstrap current but without the OH transformer during the fusion power rampup phase. As a lower elongation of [kappa] = 2 does not allow a large poloidal beta, a low density discharge with high heating/current drive power is necessary to increase the noninductive plasma current up to 30 MA, and then the vertical field can ramp the plasma current up to 48 MA just before reaching the steady state fusion burn phase. A higher elongation of [kappa] = 3 can reach a higher βp value, in which case the plasma current of 48 MA can be achieved by the vertical field without powerful heating/current drive. As the level of the vertical field current drive depends on the plasma energy, which is determined by the confinement time, a lower confinement factor can only produce a lower plasma current. The current rampup time can be chosen arbitrarily, shorter or longer, facilitating a flexible ST reactor operation.

Neutron radiation effects of the center conductor post in a spherical tokamak reactor

A Li metal center conductor post (Li metal CCP) is proposed for spherical tokamak reactors. The construction and materials of the core assembly for a Li metal CCP are described. 316Ti stainless steel is proposed for the can of the core and cooling tubes. Flowing water inside cooling tubes removes ohmic heat and nuclear heat in the CCP. Thermal hydraulic and neutronic calculations show that the average temperature of the Li conductor in operation is about 86 °C and the resistivity increases by about 20% after one year of operation. After operating one year the Li conductor can be melted by ohmic heating (no cooling water), and the liquid metal passed through a purification system to return the core assembly by an auxiliary liquid metal loop. The purified liquid metal can be cooled down to form the Li conductor again. The advantage of the proposed concept is discussed.

Conceptual study on liquid metal center conductor post in spherical tokamak reactors

A novel concept of liquid metal centre conductor post is proposed for spherical tokamak reactors. Flowing liquid metal Li inside the replaceable tube made of low activation materials (alloy, stainless steel etc.) coated by SiC is used as the medium carrying toroidal field coil current and removing Ohmic heat and nuclear heat in the center conductor post. The calculation and analysis have shown that the novel concept has potentially attractive advantages mainly including less transmutation wastes, no electric conductivity change, tolerable neutron structure damage, enhancement in tritium breeding ratio and longer service time in comparison with the conventional concept based on copper alloy conductor. The possible engineering and technological problems are specified for further study.

A new poloidal-bundle divertor for a spherical tokamak

In a spherical tokamak, power handling at the divertor region becomes more difficult than that of conventional tokamaks because of the narrow space at the divertor region. Here a new poloidal-bundle divertor is proposed for a spherical tokamak reactor. The bundle divertor is located at the bottom and/or top region of the torus, in addition to several poloidal guide coils. These coils are installed inside of the toroidal coil, because the copper conductor is adaptable in spherical tokamak devices. The magnetic field line outside the separatrix is guided to the lower region of the torus by the poloidal guide coils, and extracted outside of the torus through the adjacent toroidal coils by the bundle divertor coils. The bundle divertor configuration does not affect the shape of the core plasma as much, and the magnetic field ripple, due to these bundle coils, is negligible.