Development Of Interfaces For Fusion Neutronics Codes And Their Application

China Fusion Engineering Test Reactor (CFETR) is proposed to be the next generation fusion device of China. The conceptual design of CFETR was completed and the engineering design is about to start. The main goals of CFETR (Phase I) are to realize the quasi steady-state operation of tokamak plasma, to verify the tritium self-sustaining ability of D-T fusion reactor and to explore the maintenance and remote handling scheme for fusion reactor, the accomplishment of which all will give CFETR with fusion power of 50 -200MW. One of the typical differences between CFETR, a fusion reactor, and the plasma experimental devices is the presence of high-energy fusion neutrons. Therefore, as one of the most important components in CFETR, blanket was designed to generate power and breed tritium by utilizing the high-energy fusion neutrons from plasma and to attenuate the neutrons at the same time. The main content of fusion neutronics is hereby to study the interaction between neutrons and CFETR components, of which blanket is the most concerned one. With the development of CFETR and supported by the National Magnetic Confinement Fusion Science Program Projects “Integration of CFETR Design Software and Optimization of Plasma Parameters",this work aims to focus on the interface issues in fusion neutronic evaluations by the transport code MCNP and inventory code FISPACT, on CFETR HECLIC(Helium Cooled LIthium Ceramic)blanket.The interface codes developed or tested are, the interface between fusion plasma physics and neutronic: SCG(Source Card Generator) that takes the 0-D plasma parameters or ONETWO code produced data as input and produces the neutron source definition card for MCNP; the interface between engineering and neutronic, in this interface, McCad and ANSYS DesignModeler are able to convert the CAD model into MCNP model, and NPST, developed in this work, is to implement parameterized MCNP modeling for CFETR; the interfaces with neutronic codes: the ADVANTG code for variance reduction and R2S code by coupling MCNP and FISPACT with R2S method to calculate the shutdown dose rate (SDR); the neutronics post-processing interface: MT2X and Mcmeshtran are adopted to convert the meshtal file from MCNP to .vtk file for Para View, ASCII files for ANSYS and CFX. What's more, the relatively new Unstructured Mesh capability of MCNP was test in CFETR neutronic calculation, that shows an well representation of high-order surfaces, which is useful in the neutronic calculation of CFETR engineering model.Meanwhile, efforts were made on the assessment of CFETR neutronic performances. Several MCNP models were set with different blanket concepts. And,the effects of different distribution of fusion neutron from plasma on neutronic performance of CFETR were studied. The focus of work in this part was to evaluate the neutronic performance of CFETR with HECLIC blanket. First, the shielding performance was obtained with 1-D model. And, then, the tritium breeding capability of blanket was obtained with 3-D model built manually. Based on the experience, the design and analyses of CFETR HECLIC with Breeding Unite (BU)structure was finally completed. The detailed neutronic performances are calculated.They are the tritium breeding ratio, the neutron wall loading distribution, the distribution of neutron flux, photon flux, nuclear heating, helium production and displacement damage accumulation in steel of structure; the integral dose to epoxy insulator, the peak fast neutron fluence (E > 0.1 MeV) to the superconductor, the peak displacement damage to copper stabilizer between TFC warm-ups and Peak nuclear heating in winding pack of TF coils; the activities, decay heat, contact dose rate and the shutdown dose rate. These are important results for CFETR HECLIC,out of which the heat may be transfer into other codes to act as heating source and the radiation related values are necessary data for radiation safety evaluation,maintenance and remote handling.