scholarly journals Reconstruction of the eddy current profile on the vacuum vessel in a nuclear fusion device using only external magnetic sensor signals

2014 ◽  
Author(s):  
M. Itagaki ◽  
A. Sanpei ◽  
S. Masamune ◽  
K. Watanabe
Keyword(s):  
Eddy Current ◽  
Nuclear Fusion ◽  
Magnetic Sensor ◽  
Vacuum Vessel ◽  
Current Profile ◽  
Fusion Device ◽  
Sensor Signals

Mechanical design and error prediction of a flexible manipulator system applied in nuclear fusion environment

2017 ◽  
Vol 44 (6) ◽  
pp. 711-719 ◽  
Author(s):  
Shanshuang Shi ◽  
Huapeng Wu ◽  
Yuntao Song ◽  
Heikki Handroos
Keyword(s):  
Mechanical Design ◽  
Nuclear Fusion ◽  
Vacuum Condition ◽  
Flexible Manipulator ◽  
Error Prediction ◽  
Identification Algorithm ◽  
Vacuum Vessel ◽  
Redundant Manipulator ◽  
Content Type ◽  
Fusion Device

Purpose The purpose of this paper is to introduce a development and error modeling of a serial redundant manipulator system applied in nuclear fusion environment. Detailed mechanical design of vacuum-compatible arms and actuators are presented, and manipulator flexibility is studied through experiments and model identification algorithm to evaluate the deformation. Design/methodology/approach First, the manipulator is designed to be several modular segments to obtain enough and flexible workspace inside the fusion device with narrow and complex geometries. Joint actuators with “rotation-linear-rotation” chains are developed to provide both huge reduction ratios and vacuum sealing. The redundant manipulator system has 11 degree of freedoms in total with a storage cask system to dock with the device vacuum vessel. In addition, to improve the position accuracy, an error prediction model is built based on the experimental study and back-propagation neural network (BPNN) algorithm. Findings Currently, the implementation of the manipulator system has been successfully carried out in both atmosphere and vacuum condition. Excellent performance indicates that the mechanical design is suitable. The validation of BPNN model shows an acceptable prediction accuracy (94∼98 per cent) compared with the real measurement. Originality/value This is a special robot system which is practically used in a nuclear fusion device in China. It will allow remote inspection and maintenance of plasma facing components in the vacuum vessel of fusion device without breaking the ultra-high vacuum condition during physical experiments. Its design, mechanism and error prediction strategy have great reference values to the similar robots in vacuum and temperature applications.

Evaluation of 3D-Printed titanium alloy using eddy current testing with high-sensitivity magnetic sensor

2019 ◽  
Vol 102 ◽  
pp. 90-95 ◽  
Author(s):  
Dongfeng He ◽  
Zhi Wang ◽  
Masahiro Kusano ◽  
Satoshi Kishimoto ◽  
Makoto Watanabe
Keyword(s):  
Titanium Alloy ◽  
Eddy Current ◽  
High Sensitivity ◽  
Eddy Current Testing ◽  
Magnetic Sensor ◽  
Current Testing ◽  
3D Printed

scholarly journals MHD stability and disruptions in the SPARC tokamak

2020 ◽  
Vol 86 (5) ◽  
Author(s):  
R. Sweeney ◽  
A. J. Creely ◽  
J. Doody ◽  
T. Fülöp ◽  
D. T. Garnier ◽  
...  
Keyword(s):  
Safety Factor ◽  
Eddy Currents ◽  
Runaway Electron ◽  
Vacuum Vessel ◽  
Current Profile ◽  
Tearing Modes ◽  
Mhd Stability ◽  
Data Driven Approach ◽  
Operating Space ◽  
Stable Current

SPARC is being designed to operate with a normalized beta of $\beta _N=1.0$ , a normalized density of $n_G=0.37$ and a safety factor of $q_{95}\approx 3.4$ , providing a comfortable margin to their respective disruption limits. Further, a low beta poloidal $\beta _p=0.19$ at the safety factor $q=2$ surface reduces the drive for neoclassical tearing modes, which together with a frozen-in classically stable current profile might allow access to a robustly tearing-free operating space. Although the inherent stability is expected to reduce the frequency of disruptions, the disruption loading is comparable to and in some cases higher than that of ITER. The machine is being designed to withstand the predicted unmitigated axisymmetric halo current forces up to 50 MN and similarly large loads from eddy currents forced to flow poloidally in the vacuum vessel. Runaway electron (RE) simulations using GO+CODE show high flattop-to-RE current conversions in the absence of seed losses, although NIMROD modelling predicts losses of ${\sim }80$  %; self-consistent modelling is ongoing. A passive RE mitigation coil designed to drive stochastic RE losses is being considered and COMSOL modelling predicts peak normalized fields at the plasma of order $10^{-2}$ that rises linearly with a change in the plasma current. Massive material injection is planned to reduce the disruption loading. A data-driven approach to predict an oncoming disruption and trigger mitigation is discussed.

Modelling of Water Ingress in to Vacuum Vessel

2014 ◽  
Author(s):  
Tadas Kaliatka ◽  
Eugenijus Ušpuras ◽  
Algirdas Kaliatka
Keyword(s):  
Cooling System ◽  
Nuclear Fusion ◽  
Full Range ◽  
Water Evaporation ◽  
Experimental Investigations ◽  
Fusion Reactors ◽  
Vacuum Vessel ◽  
Temperature Plasma ◽  
Hydraulic Processes ◽  
Relap5 Code

An event of water coolant ingress into vacuum vessel is one of the most important events leading to severe consequences in nuclear fusion reactors. The ingress of coolant to the vacuum vessel could appear due to coolant pipe rupture of in-vessel components. Vacuum vessel could not withstand the high pressure inside. Pressure increase in vacuum vessel is due to water evaporation because of pressure difference and water contact with high temperature plasma facing components. If pressure in vacuum vessel is too high — safety valve opens and the steam containing activated dust will be transferred form the vacuum vessel to the environment. Therefore, it is important to understand the thermo hydraulic processes in vacuum vessel during the ingress of coolant event (ICE). There are few experimental investigations performed, modeling of ICE. In this article ingress of coolant event in vacuum vessel was modeled using RELAP5 code. RELAP5 is a “best estimate” system code suitable for the thermo-hydraulic analysis of all transients and postulated accidents in nuclear fission, light water reactor systems, including both large and small-break loss-of-coolant accidents as well as the full range of operational transients. The use of RELAP5 code for the accident analysis in nuclear fusion reactors allows to perform integral analysis of thermal-hydraulic processes in the cooling system and vacuum vessel. The comparisons of calculation results and experimental data showed that with some limitations the RELAP5 program could be used for the analysis of the thermal hydraulic processes in the vacuum vessel during ICE.

scholarly journals Heat load and deuterium plasma effects on SPS and WSP tungsten

Nukleonika ◽  
2015 ◽  
Vol 60 (2) ◽  
pp. 275-283 ◽  
Author(s):  
Monika Vilémová ◽  
Jiří Matějíček ◽  
Barbara Nevrlá ◽  
Maryna Chernyshova ◽  
Pawel Gasior ◽  
...  
Keyword(s):  
Nuclear Fusion ◽  
Edge Plasma ◽  
Deuterium Plasma ◽  
Thermally Induced ◽  
Fusion Device ◽  
Plasma Sintering ◽  
Water Stabilized Plasma ◽  
Spark Plasma ◽  
Tungsten Material ◽  
Induced Changes

Abstract Tungsten is a prime choice for armor material in future nuclear fusion devices. For the realization of fusion, it is necessary to address issues related to the plasma–armor interactions. In this work, several types of tungsten material were studied, i.e. tungsten prepared by spark plasma sintering (SPS) and by water stabilized plasma spraying (WSP) technique. An intended surface porosity was created in the samples to model hydrogen/helium bubbles. The samples were subjected to a laser heat loading and a radiation loading of deuterium plasma to simulate edge plasma conditions of a nuclear fusion device (power density of 108 W/cm2 and 107 W/cm2, respectively, in the pulse intervals up to 200 ns). Thermally induced changes in the morphology and the damage to the studied surfaces are described. Possible consequences for the fusion device operation are pointed out.

Dynamic Simulation of Blanket Module Maintenance inside Fusion Vacuum Vessel

2014 ◽  
Vol 556-562 ◽  
pp. 1220-1225
Author(s):  
Yan He Zhu ◽  
Lan Ming Guo ◽  
Jie Zhao
Keyword(s):  
Control Strategy ◽  
Hierarchical Control ◽  
Critical Issue ◽  
Vacuum Vessel ◽  
Positioning Accuracy ◽  
Experiment Platform ◽  
Fusion Device ◽  
Remote Maintenance ◽  
Control Scheme ◽  
Essential Problem

The most critical issue of the blanket module remote maintenance operation is to remove or replace the heavy module with high positioning accuracy of 1mm. Located in vacuum vessel (VV) of the nuclear fusion device, the blanket module is weight up to 500kg, thus the grasp and installation of blanket module come to be the essential problem during the maintenance operation. To meet the requirement, we propose a new hierarchical control strategy of rough and fine positioning technology based on combined sensors. The detail procedures and implementation of the control scheme has been carried out successfully on Virtual Robot Experiment Platform to demonstrate the feasibility of the control strategy.

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