Z. W. Wang

Development of system code of CFETR design

China Fusion Engineering Test Reactor (CFETR) is now in the conceptual design phase. The main objective of CFETR is demonstration of fusion energy with 50 ~ 200 MW fusion power, fuel cycle of T self-sustained and steady-state operation with duty cycle of 0.3 ~ 0.5. The design of CFETR involves complex system structure, and there are complex constrains between physics and engineering. Between optimization of performance parameters and design of main structure and key components, numerous data exchange and iterative optimization are necessary for optimal design of sub-systems. To do the optimization design, a CFETR system code is under development. The main technical schemes for system code include: a physical design platform and various engineering design modules are developed, then a global framework integrates them by standard interfaces and communication technologies; and a standard material and design criterion database unifies the reference data for the system code. The detailed study will be presented in this conference.

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.

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.