Numerical simulation of tritium behavior under a postulated accident condition for CFETR TEP system

China Fusion Engineering Test Reactor(CFETR)is China's self-designed and ongoing next-generation fusion reactor project.Tritium confinement systems in CFETR guarantee that the radiation level remains below the safety limit during tritium handling and operation in the fuel cycle system.Our tritium technology team is responsible for studying tritium transport behavior in the CFETR tritium safety confinement systems of the National Key R&D Program of China launched in 2017,and we are conducting CFETR tritium plant safety analysis by using CFD software.In this paper,the tritium migration and removal behavior were studied under a postulated accident condition for the Tokamak Exhaust Processing system of CFETR.The quantitative results of the transport behavior of tritium in the process room and glove box during the whole accident sequence(e.g.,tritium release,alarm,isolation,and tritium removal)have been presented.The results support the detailed design and engineering demonstration-related research of CFETR tritium plant.

3D fluid model analysis on the generation of negative hydrogen ions for negative ion source of NBI

A radio-frequency(RF) inductively coupled negative hydrogen ion source(NHIS) has been adopted in the China Fusion Engineering Test Reactor(CFETR) to generate negative hydrogen ions.By incorporating the level-lumping method into a three-dimensional fluid model,the volume production and transportation of H^(-) in the NHIS,which consists of a cylindrical driver region and a rectangular expansion chamber,are investigated self-consistently at a large input power(40 k W) and different pressures(0.3–2.0 Pa).The results indicate that with the increase of pressure,the H^(-) density at the bottom of the expansion region first increases and then decreases.In addition,the effect of the magnetic filter is examined.It is noteworthy that a significant increase in the H^(-) density is observed when the magnetic filter is introduced.As the permanent magnets move towards the driver region,the H^(-) density decreases monotonically and the asymmetry is enhanced.This study contributes to the understanding of H-distribution under various conditions and facilitates the optimization of volume production of negative hydrogen ions in the NHIS.

Comparative studies for two different orientations of pebble bed in an HCCB blanket

The Indian Test Blanket Module（TBM） program in ITER is one of the major steps in its fusion reactor program towards DEMO and the future fusion power reactor vision. Research and development（RD） is focused on two types of breeding blanket concepts： lead–lithium ceramic breeder（LLCB） and helium-cooled ceramic breeder（HCCB） blanket systems for the DEMO reactor. As part of the ITER-TBM program, the LLCB concept will be tested in one-half of ITER port no. 2, whose materials and technologies will be tested during ITER operation. The HCCB concept is a variant of the solid breeder blanket, which is presently part of our domestic RD program for DEMO relevant technology development. In the HCCB concept Li_2TiO_3 and beryllium are used as the tritium breeder and neutron multiplier, respectively, in the form of a packed bed having edge-on configuration with reduced activation ferritic martensitic steel as the structural material. In this paper two design schemes, mainly two different orientations of pebble beds, are discussed. In the current concept（case-1）, the ceramic breeder beds are kept horizontal in the toroidal–radial direction. Due to gravity, the pebbles may settle down at the bottom and create a finite gap between the pebbles and the top cooling plate, which will affect the heat transfer between them. In the alternate design concept（case-2）, the pebble bed is vertically（poloidal–radial） orientated where the side plates act as cooling plates instead of top and bottom plates. These two design variants are analyzed analytically and 2 D thermal-hydraulic simulation studies are carried out with ANSYS, using the heat loads obtained from neutronic calculations.Based on the analysis the performance is compared and details of the thermal and radiative heat transfer studies are also discussed in this paper.

Application of Finite Difference Method to the Analysis of HTTR Reactor Cavity Natural Convection

An attempt is made to investigate the ability of the revised K FIX code to the solution to natural convection. The new experimental data of the benchmark problem of the IAEA coordinated research program (CRP 3) on heat transport and afterheat removal for GCRs under accident conditions provided by JAERI are used to calculate nitrogen natural convection in the pressurized vessel and air natural convection in the reactor cavity by using this revised code. Based on analysis, a refined mesh is used to solve the differential equations so as to get more detailed and more accurate result. The obtained velocity profiles are consistent with the result of TRIO EF code and the result of Bechtel laboratory. It can be drawn that the revised K FIX code can be used to solve this kind of problems.