Skip to content

Reactor Plasma Platform

Design in progress includes the following elements:
  • Aim for an ion temperature of 10+ keV
  • Comparison between super-conducting and resistive coils
  • Focus on hydrogen-only operation

Examining and Enhancing the Effects of Impurities and the Diverter Configuration

Our SR55 reactor design includes a proposal for a lower single null diverter configuration. In this design, the diverter will be situated in the lower section of the vacuum vessel. Its purpose is to eliminate particles (such as helium and impurities) from the plasma core and to discharge a significant portion of the plasma's thermal power. The ability of the diverter to effectively remove the plasma's thermal power is crucial for the successful and safe operation of a fusion power reactor.

The figure of merit, which is the fusion power per unit major radius (P/R), may be used to assess the heat load on the diverter. The greatest stationary heat flux that can currently be extracted by the diverter is 10 MW/m, despite the fact that P/R for the reactor Pf = 1GW mission may reach 15.69 or 19.11 MW/m depending on whether it is operating in entirely non-inductive or hybrid mode. As a result, a greater heat load on the diverter surface might stop the reactor from operating in a steady state or for lengthy pulses. The diverter geometry will still need to be improved; for instance, the field lines approaching the diverter target surfaces will be expanded, and the surfaces will be shaped to lessen the perpendicular heat flow.

The diverter region's plasma performance may be impacted by the plasma shape because of its tight link to the diverter geometry. For instance, the diverter volume on the high field side decreases as the plasma triangularity increases. It will limit the radiation in the diverter region, which will have an impact on the plasma detachment produced. Therefore, it is unquestionably necessary to optimize the location of the X-point and the diverter strike point as well as the plasma form. The equilibrium algorithm TEQ is used to build sequences of equilibria with varying triangularities (ranging from 0.3 to 0.8) based on the zero-dimensional parameters.

Through repeated adjustments to the diverter and TBM structures, it is possible to achieve a reference equilibrium for the reactor configuration. This equilibrium includes an acceptable position for the X-point and diverter strike point, with a dRsep value of 6cm. This equilibrium serves as the starting point for various design works. In addition to the standard diverter configuration, the snowflake+ (SF+) diverter configuration has also been taken into consideration. Calculations have shown that the SF+ configuration is compatible with the standard divertor and the current in the PF coils remains below the engineering limit. As previously discussed in the design process, incorporating two extra superconducting coils (DC1 and DC2) along with other PF coils allows for the realization of either an X-diverter or snowflake advance diverter configuration, with a field expansion ranging from 1.7 to 4.2 times.

L/R time vs. Toroidal mode number for (a)VV, (b) VV&BM (original resistance) and (c) VV&BM (original resistance times 10). Here X-axis denotes the toroidal mode number, while Y-axis the L/R time [s].

The reference equilibrium of the
configuration for the reactor

The heat would be partially radiated externally by impurities in the core that have a minimal detrimental effect on the plasma core's performance. Since introducing impurities would increase Zeff (the effective nuclear charge), a scan of Zeff using argon (Ar) gas injection in a steady-state burning plasma is conducted to find the highest attainable core impurity radiation without deteriorating or negatively influencing the performance of the core plasma. The fusion performance rises as Zeff rises before declining when it approaches the critical value of 2.86.

The highest values of the fusion performance are produced by an imbalance between the suppression of turbulence and rising radiation loss combined with rising Zeff. At the turning point, the PLOSS (Partial Loss) may only represent a small portion of the overall power or 26.3%. To ensure steady state functioning, more extra power is needed since impurities might reduce the efficiency of the present drive. It is important to note that the simulation's sensitivity to the presumptive impurity profile is rather low. As a component of the diverter heat load solution, this discovery is promising.

For our infrastructure design, there are currently several tasks, including layout design and system integration, superconducting magnet and cryogenics, vacuum vessel and Vacuum System, in-vessel components, standardization and design management, heating and current drive system, diagnostics, and remote maintenance system. Additionally, significant research has been conducted to utilize important technologies for the key elements as well as to test sophisticated designs.

Magnet System

To sustain a long burn duration as specified in our mission's duty time, the use of superconducting magnets is necessary. Several magnet systems are involved in generating the required magnetic field configuration and high plasma current (up to 15 MA) for confining the burning plasma. These systems include 18 toroidal field (TF) coils, 10 central solenoid (CS) modules, 8 poloidal field (PF) coils, and 20 correction coils (CC). Our main focus in this section is on the progress of the CS coils, TF coils, and the R&D (Research and Development) activities of the superconducting coils.

To fulfil our mission, we have designed 18 TF coils in a D-shaped configuration, consisting of six arcs and a straight leg. These coils have a plasma major radius of 6.3 m and a total of 162 turns. For the fabrication of the TF coils, we utilize high-performance superconducting magnets that incorporate advanced Jc Nb3Sn RRP superconducting strands. The TF coils are capable of operating at a maximum current of 91.6 kA, generating a toroidal magnetic field of up to 5.3 T on the plasma axis. In the inboard length, the maximum magnetic field reaches approximately 15 T.

The TF reactor has a significantly higher per turn and total storage energy (116.34 GJ) compared to other current reactors. To reduce costs, TF coils use Nb3Sn cable-in conduit conductors (CICC) with a rectangular cross-section. Stress analysis of each TF coil has been conducted, and the results are displayed in the figure. It is clear that each TF coil will experience a centripetal force towards the main machine in the TF inner leg, with a maximum force of approximately 821 MPa. The coil can deform up to 12.2 mm. To withstand this force, the coil case is designed to be nearly half a meter thick.

3D model Vacuum Vessel

Conceptual view of the reactor

Conceptual engineering design of the divertor structure.

Vacuum System

The cryostat, torus, and other lower-volume systems are only a few of the huge volume systems that make up the entire vacuum system. The vacuum vessel is envisioned as a toroidal container with an outside diameter of about 12.2 m and a height of approximately 4.5 m, which serves as both the initial confinement barrier for the tritium and other fuels as well as the high-quality vacuum needed for plasma operation. The whole tokamak-type reactor core is housed inside a cryostat with dimensions of 10 meters in diameter and 3.8 meters in height and a background pressure of 10-4 pa. Heat transmission to the superconducting magnet systems is minimized by a succession of stainless-steel thermal shields that are chilled by 1.8 MPa pressurized helium gas from the main cryoplant with a 72 K inlet temperature.

The VV is designed to have a torus shape with a D-shaped cross-section. It has 3 upper vertical ports, 6 lower ports, and 4 equatorial ports. The inner and outer shells, as well as the stiffening ribs, are connected by welding. To make manufacturing easier, the D-shaped cross-section is made up of three arcs and one straight line segment that are tangent to each other. The VV shells are 35 mm thick and made of 316L(N)-IG material. The 4 upper ports are used for blanket maintenance and disassembly. The 6 lower ports are for diverter maintenance and cryo-pumps. The 8 equatorial ports are used for NBI, diagnostics, and remote handling tools. To account for different neutron irradiation doses on the superconducting magnet coils, there is unequal spacing between the double-walled structure in the inboard and outboard regions.

The blanket modules, diverter parts, inner coils with their supports, and cooling system, are all kept within a radial dimension of at least 1000 mm from the VV inner shell to the plasma boundary. A maximum dimension of 15,820.5 mm in the vertical direction and 8,160 mm in the horizontal direction is allowed inside the VV. Additional research is being carried out on the gravity load, the frequency and mode of vibration, and the seismic load.

Remote Maintenance and Handling System

The operation of the burning plasma will produce high-energy neutrons that can activate the components and materials inside the reactor. Additionally, the plasma's interaction with stainless steel, tungsten, and other alloys used in the construction of these components will generate dust that is contaminated with tritium. As a result, it is crucial to have a comprehensive remote handling and maintenance capability within the facility. However, the design of the remote handling and maintenance system will greatly influence the overall layout and the designs of its components and systems.

The irradiation dose rate within the vacuum vessel will be in the range of hundreds of Svhr-1 throughout the nuclear phase of operations, as shown by the conservative estimate from the neutronic studies, hence all in-vessel maintenance must be carried out by robotic systems. If we wish to complete the required amount of duty time, two basic types of in-vessel components, the diverter and tritium breeding blanket, must be remotely maintained/replaced on a regular basis. However, the tritium breeding blanket and diverter’s bulk and volume will put a greater strain on the RHM system; as a result, consideration should be given to the RHM's capabilities while designing the tritium breeding blanket and diverter.

The engineering analysis of the RHM's capacity for in-vessel components has been accomplished, and the RHM system's strategy has been provided. This plan takes into account the tokamak layout with 18 TF coils and 3 vertical ports for maintenance. For maximum efficiency, the inboard and outboard blankets are combined into a single blanket sector and replaced from the top port.

The blanket is lifted out and moved to the heated cell using a crane mechanism along a fixed path. The parallel multipurpose platform will maintain the diverter through four lower ports. Except for blankets and the diverter, all other in-vessel components will be handled and maintained by the MPD subsystem. During the design phase, factors including security, dependability, accessibility, and compatibility were taken into account. The next stage of engineering design development will result in an exponential increase in RHM maturity.