Huang Qunying

Interaction of CLAM Steel with Plasma in HT-7 Tokamak During High Parameter Operation

A Plasma Surface Interaction (PSI) experiment on China Low Activation Martensitic (CLAM) steel was done to check if CLAM steel could be used as a Plasma Facing Material (PFM). A specimen with a diameter of 45 mm was exposed to 897 shots of deuterium plasmas with a total duration of 712 sec at a minor radius of 30 cm in HT-7 tokamak. During the exposure experiment, no observable influence was found on plasma performance. After exposure, the surface of the specimen seemed as smooth as before but with some colour change at the margin of the specimen. Even though some micro-damage, such as dense blisters, melting, splashing, depositions, and dust, was found on local surfaces with Scanning Electron Microscopic (SEM) observation. The reflectivity of the specimen decreased only slightly. All of these shows CLAM steel has good stability and irradiation resistance. With further optimization, it could possibly be used as the first mirror material for plasma diagnostics in tokamaks.

Variation of Radiation Damage with Irradiation Temperature and Dose in CLAM Steel

The dependences of radiation induced defects on irradiation temperature up to 700°C at 15 dpa and on irradiation dose up to 85 dpa at room temperature have been investigated by the heavy ion irradiation and the positron annihilation lifetime spectroscopy for the CLAM. A void size peak is observed at ~500°C where the vacancy cluster contains 9 vacancies and has an average diameter of 0.59 nm. The size of the vacancy clusters increases with the increase of irradiation dose at room temperature, and the vacancy cluster at 85 dpa consists of 9 vacancies and reaches a size of 0.60 nm in diameter. The absolute values of the void size at the peak and the increase of void size with dose in the CLAM steel are negligible compared to those of the normal stainless steels, indicating that the CLAM steel has good radiation resistant property.

Activation Characteristics of Fuel Breeding Blanket Module in Fusion Driven Subcritical System

Shortage of energy resources and production of long-lived radioactivity wastes from fission reactors are among the main problems which will be faced in the world in the near future. The conceptual design of a fusion driven subcritical system (FDS) is underway in Institute of Plasma Physics, Chinese Academy of Sciences. There are alternative designs for multi-functional blanket modules of the FDS, such as fuel breeding blanket module (FBB) to produce fuels for fission reactors, tritium breeding blanket module to produce the fuel, i.e. tritium, for fusion reactor and waste transmutation blanket module to try to permanently dispose of long-lived radioactivity wastes from fission reactors, etc. Activation of the fuel breeding blanket of the fusion driven subcritical system (FDS-FBB) by D-T fusion neutrons from the plasma and fission neutrons from the hybrid blanket are calculated and analysed under the neutron wall loading 0.5 MW/m2 and neutron fluence 15 MW.yr/m2. The neutron spectrum is calculated with the worldwide-used transport code MCNP/4C and activation calculations are carried out with the well known European inventory code FISPACT/99 with the latest released IAEA Fusion Evaluated Nuclear Data Library FENDL-2.0 and the ENDF/B-V uranium evaluated data. Induced radioactivities, dose rates and afterheats, etc, for different components of the FDS-FBB are compared and analysed.

Synergistic Effect of Triple Ion Beams on Radiation Damage in CLAM Steel

The synergistic effect of triple ion beams is investigated by simultaneous and sequential irradiations of gold, hydrogen and helium ions on the low activation martensitic steel (CLAM) developed in China. The depth profile measurements of the positron annihilation Doppler broadening S Parameter are carried out as a function of slowpositron beam energy to examine the produced radiation damage. The synergistic effect of displacement damage and hydrogen and helium on the formation of radiation damage is clearly observed. In the preset case, this effect suppresses the radiation damage in the CLAM steel due to the helium and/or hydrogen filling of vacancy clusters.

Concise relation of substitution energy to macroscopic deformation in a deformed system

An ab initio study of the effect of macroscopic deformation on energetics of twelve alloying elements in bcc Fe has been performed under three specially designed strain modes. A concise relation of the macroscopic deformation effect on the substitution energy of alloying elements with linear dependences on defect formation volume and relative volume change was found. Based on this concise relationship, the following behaviors can be predicted by comparing defect formation volumes: the strain-induced solubility change of alloying atoms and then the degree or possibility of redistribution and segregation of alloying atoms, the stability transition between monovacancy and divacancy, and self-interstitial atom reorientation under heavy loading.

Effects of CT based Voxel Phantoms on Dose Distribution Calculated with Monte Carlo Method

A few CT-based voxel phantoms were produced to investigate the sensitivity of Monte Carlo simulations of x-ray beam and electron beam to the proportions of elements and the mass densities of the materials used to express the patients anatomical structure. The human body can be well outlined by air, lung, adipose, muscle, soft bone and hard bone to calculate the dose distribution with Monte Carlo method. The effects of the calibration curves established by using various CT scanners are not clinically significant based on our investigation. The deviation from the values of cumulative dose volume histogram derived from CT-based voxel phantoms is less than 1% for the given target.

Fusion-Driven Sub-Critical Dual-Cooled Waste Transmutation Blanket: Design and Analysis

The Fusion-Driven Sub-critical System (FDS) is one of the Chinese programs to be further developed for fusion application. Its Dual-cooled Waste Transmutation Blanket (DWTB), as one the most important part of the FDS is cooled by helium and liquid metal, and have the features of safety, tritium self-sustaining, high efficiency and feasibility. Its conceptual design has been finished. This paper is mainly involved with the basic structure design and thermal-hydraulics analysis of DWTB. On the basis of a three-dimensional (3-D) model of radial-toroidal sections of the segment box, thermal temperature gradients and structure analysis made with a comprehensive finite element method (FEM) have been performed with the computer code ANSYS5.7 and computational fluid dynamic finite element codes. The analysis refers to the steady-state operating condition of an outboard blanket segment. Furthermore, the mechanical loads due to coolant pressure in normal operating conditions have been also taken into account. All the above loads have been combined as an input for a FEM stress analysis and the resulting stress distribution has been evaluated. Finally, the structure design and Pb-17Li flow velocity has been optimized according to the calculations and analysis.

Thermal-Hydraulic Optimization of Water-Cooled Center Conductor Post for Spherical Tokamaks Reactor

This paper proposes a conceptual structure of segmental water-cooled Center Conductor Post (CCP) to be flexible in installment and replacement. Thermal-hydraulic optimization and sensitivity analysis of key parameters are performed based on a reference fusion transmutation system with 100 MW fusion power. Numerical simulation by using a commercial code PHOENICS has been carried out to be close to the thermal-hydraulic analytical results of the CCP mid-part.

Neutronics Optimization of Tritium Breeding Blanket for the FDS

Neutronics optimization calculations have been performed for the tritium breeding blankets with solid ceramic breeder Li2O and liquid eutectic breeder Li17Pb83, respectively, based on a 2-D geometrical configuration using the Monte Carlo neutron-photon transport code MCNP/4B. The effects of beryllium, 6Li enrichment and various structural materials on Tritium Breeding Ratio have been systematically analyzed.

Analysis of Plasma-Driven Tritium Permeation Through the First Wall of DFLL-TBM in ITER

Tritium permeation through the first wall (FW) from the plasma into helium coolant is evaluated for a dual-functional lithium-lead test blanket module (DFLL-TBM). The effect of the surface conditions on the plasma facing and coolant sides, both temperature gradient and beryllium layer clad on the plasma facing side, as well as trapping in defects on the tritium permeation is considered. The results show that most of the tritium implanted in FW re-entered the plasma. The plasma-driven tritium permeation is very sensitive to the surface conditions on the plasma facing side. With a higher sticking coefficient on the plasma-facing side, the tritium permeation into helium coolant is significantly reduced. The tritium permeation is strongly reduced with a beryllium layer clad on the front side of FW. The plasma driven tritium permeation will not seriously impact the tritium safety of DFLL-TBM. Based on tritium safety, it is reasonable to clothe the beryllium layer on FW and keep the surface clean to reduce the plasma driven tritium permeation.