Changle Liu

Engineering Studies on the EAST Tungsten Divertor

Experimental advanced superconducting tokamak device is a D-shaped full superconducting tokamak with actively water cooled plasma facing components. To achieve long pulse and high βH-mode plasma, new plasma position and shape are calculated and optimized during the campaign of 2013-2014. New divertors are designed and developed to fit the plasma and endure the heat flux up to 10 MW/m2. The divertor is International Thermonuclear Experimental Reactor-like. It bases on monoblock and cassette technology and is composed of plasma facing component (PFC) units, cassettes, and support systems. Monoblock structures are just employed on the PFC units of target plates of the divertors, thus W flat tiles are used on the baffles and dome. At the end of monoblocks, end boxes are applied. Cassettes act not only as the supports of the PFC units, but also as manifolds for the cooling channels of the units. The support systems consist of inner support rails, outer support rails, and auxiliary braces. The support systems are installed on the vacuum vessel before cassettes are put on. To dock the cassette with the support system, it is just lifted and pushed forward. To verify and optimize the structure, Research and Development work has carried out for the divertor. The prototypes of PFC units and cassettes are fabricated with different manufacture processes. The total assembly process is also simulated with these prototypes on mockup facility and the vacuum vessel. The results show that the divertor system can be manufactured successfully according to the requirements of drawings and installed conveniently and precisely on the tokamak. All these works certify that the design of the divertor is all right. The batch production of the divertors can be started.

Investigation on the Possibility of Tritium Self-Sufficiency for CFETR Using a PWR Water-Cooled Blanket

The neutron wall load (Pn) of Chinese fusion engineering testing reactor (CFETR) will be less than 1 MW/m2. To meet the net tritium breeding ratio (TBR) of the reactor, a new water-cooled blanket concept is considered. The blanket neutronics schemes are performed to explore the local TBR issues in the (Pn) range of 1-5 MW/m2, which aims at the effective design of the blanket concept considering the tritium self-sufficiency. As a result, the calculation results are compared with the local TBR values and the material fraction changes. It is found that the local TBR has the high value at low (Pn) while the blanket size in radial direction is determined. It is mainly because of the total breeding area increasing due to the pipe pitch increasing in the model. This leads to the possibility for CFETR using a simplified blanket interior. In addition, to match the pressurized water reactor (PWR) water-cooled condition, a reduced size of blanket module in toroidal direction is achievable. It can be concluded that a PWR water-cooled blanket has more benefits to CFETR engineering implementation in the future.

The Design and Manufacture of the Neutral Beam Injection Thermal Shield for the EAST Tokamak

Abstract Neutral beam injection (NBI) is a high-power auxiliary heating system for the EAST device. We present a thermal shield (TS) structure to protect the neck pipe of the EAST equatorial port to avoid damage from the NBI beam. Since the EAST port has a big trumpet structure, a straight section, and a small trumpet structure, to accommodate the port structure, a TS concept is put forward including its cooling system. The cooling loops and the sub-branches were designed with interfaces between the inner cooling branches. The heat removal capability is verified by a thermal hydraulics analysis based on ANSYS code. In particular, fabrication is addressed with technical processing technology, especially for the embedded cooling pipes in the heat sinks. The pipes are checked for leaks after bending and the embedding processing. The assembly activities are demonstrated in the spatial space zones of the port before the engineering installation. It is confirmed that the TS structure is safe and will run feasibly in the EAST discharge. It is indicated that the TS structure can provide thermal shielding and remove heat for the NBI device in the port region.

A multi-layer breeding blanket concept for CFETR based on PWR condition

A breeding blanket concept with the multi-layer structure based on the PWR water-cooled condition was presented for CFETR. To explore the feasibility of the blanket scheme, the neutronics and hydraulics programs were carried out. It was found when the Pn is less than 3MW/m2, the local TBR (tritium breeding ratio) would be in range of 1.46-1.7. Thus, the net TBR would be more than 1.05, which meets the tritium self-sustaining requirement for the fusion reactor. Especially, the local TBR is 1.66 at a neutron wall load (Pn) of 0.5 MW/m2 and the corresponding net TBR is 1.21. On the other hand, it also clarified a pipe bore with 7-8 mm at an inlet velocity of 3-4 m/s would be suitable for the heat removal of the blanket module. In addition, the total pressure drop would be under 0.2 MPa in the cooling system. It was concluded that the blanket concept would be more effective and benefit to CFETR in view of its neutron wall load level and the tritium self-sustaining efficiency.