Fault Current Tests for ITER External Bypass on DC Test Platform in ASIPP

Fault current tests of ITER external bypass are performed to verify its fault suppression capability. This paper describes the test requirements, test schemes and test procedures of fault current test for external bypass. The effectiveness of test schemes for fault current tests is verified by the simulation results and test results based on DC test platform.     

Preliminary Design and Analysis of the DC Reactor for the ITER Converter

The DC reactor is an important piece of equipment for restraining loop and ripple currents in the international thermonuclear experimental reactor （ITER） converter power supply system. As the reactor is operated at a steady state of 27.5 kA and needs to withstand a peak current of 175 kA, so the design of the DC reactor used in the ITER converter power supply system is necessary. A new water-cooling dry-type air-core reactor is designed in this work. The detailed structural parameters are calculated by theoretical formulas, and then the structure is optimized by electromagnetic simulation with ANSYS. Finally, thermal and dynamic stability analyses are performed to verify the temperature and stress at a rated current of 27.5 kA and pulsed current of 175 kA. The analysis results show that the temperature and stress meet the requirements of the ITER converter power supply system.     

Thermal Design of High-Power DC Reactor for ITER Poloidal Field Converter

This paper presents an alternative computationally efficient approach to the thermal design of high-power DC reactor which is applied to ITER poloidal field converter module. The proposed approach takes full advantages of the property of time-saving from heat transfer theory and high accuracy from 3-D finite-element analysis, focusing on the predictions of the temperature rise and pressure loss of circulating water. Thermal measurements from two prototypes test show a good agreement with the predictions, which proves the high efficiency and accuracy of the proposed approach in the design of hundred-kilowatt-class thermal power DC reactor.     

Harmonics analysis of the ITER poloidal field converter based on a piecewise method

Poloidal field (PF) converters provide controlled DC voltage and current to PF coils.The many harmonics generated by the PF converter flow into the power grid and seriously affect power systems and electric equipment.Due to the complexity of the system,the traditional integral operation in Fourier analysis is complicated and inaccurate.This paper presents a piecewise method to calculate the harmonics of the ITER PF converter.The relationship between the grid input current and the DC output current of the ITER PF converter is deduced.The grid current is decomposed into the sum of some simple functions.By calculating simple function harmonics based on the piecewise method,the harmonics of the PF converter under different operation modes are obtained.In order to examine the validity of the method,a simulation model is established based on Matlab/ Simulink and a relevant experiment is implemented in the ITER PF integration test platform.Comparative results are given.The calculated results are found to be consistent with simulation and experiment.The piecewise method is proved correct and valid for calculating the system harmonics.     

DC performance and AC loss of cable-in-conduit conductors for International Thermonuclear Experimental Reactor

A reliable prediction of AC loss is essential for the application of International Thermonuclear Experimental Reactor(ITER) cable-in-conduit conductors(CICCs);however,the calculation of AC loss of ITER CICCs is a cumbersome task due to the complicated geometry of the multistage cables and the extreme operating conditions in ITER.In this paper,we described the models developed for hysteresis and coupling loss calculation,which can be suitable for the construction of ITER magnetic system.Meanwhile,we compared the results of theoretical analysis with the SULTAN test result to evaluate the numerical model we used.In addition,we introduced the n-value and AC loss with transport current for CICCs based on the DC measurement results at SULTAN,which lays the foundation for the further study.     

Development of Data Acquisition Management System for ITER DC Testing Platform

To acquire multi-channel signals with 10 kHz sampling rate from various front-end sensors, a Data Acquisition Management System (DAMS) based on MDSplus was designed for the International Thermonuclear Experimental Reactor (ITER) Direct Current (DC) testing platform. Due to a large number of experimental data generated from long-pulse operation, it is very important to view and analyze experimental data online during operation. To meet the requirement of online data processing, slice storage and thumbnail technology were applied in DAMS. The long pulse data is gradually written in MDSplus database. The DAMS has been verified in the ITER DC power supply testing platform.     

核电厂蓄电池直流系统短路电流的计算方法及其危害分析

直流系统以其固有的供电连续性在核电厂运行中发挥着非常重要的作用,与交流系统相比,直流系统在短路电流计算上所应用的原理和方法上有着显著的不同.国内相关标准和文献中,对直流系统短路电流如何计算虽有描述,但并未对电流的变化曲线和故障的切除时间进行明确的描述,这对直流系统设备选型、故障影响和根本原因分析是不利的.文章参考IEE...     

Thermal Analysis of Power Thyristor Used in ITER Poloidal Field Converter

As a key component of ITER PF converter module, thyristor conducting large current produces a lot of power loss, therefore it is meaningful to study its heat transfer characteristic for improving the performance of PF converter. This paper presents the thermal analysis of the thyristor. A 3D finite-element model with multi material layers is built and simulated in steady state operation. A special temperature rise test scheme is designed and done to verify the analysis result. The simulation is well in compliance with the test result. The modeling method presented in this paper is proved to be practical in thyristor thermal analysis. The proposed measurement method in temperature rise test is also of valuable reference for thermal testing of power semiconductor devices.     

ITER氚增殖实验包层设计研究进展

国际热核实验反应堆(ITER)为人类开发聚变能提供重要的物理和工程技术实验平台,ITER氚增殖实验包层模块(TBM)技术是必须掌握的关键技术.参与ITER计划的成员国根据本国商用演示堆包层发展策略,分别提出了各自的实验包层概念,以便在ITER运行期间进行实验.本文对ITER-TBM目前已经开展和正在进行的主要设计研究工作进展进行总结,介绍了各方提出的设计方案、支撑设计的相关技术研究进展,以及合作实验窗口的分配现状.     

CENTER工程反应堆保护系统定期试验方案设计

CENTER工程反应堆保护系统采用了中国核动力研究设计院自主研发的"龙鳞"平台。根据GB/T5204和IEEE 338的设计要求,本文基于定期试验的设计准则,采用分段试验和相互交迭的设计思路,同时结合平台自身的特点对CENTER工程反应堆保护系统的定期试验总体方案进行了介绍,重点描述了输入通道试验(T1试验)、系统逻辑功能试验(T2试验)、输出通道及相关驱动器试验(T3试验)和响应时间试验的试验方案及其工作原理,可作为其他工程定期试验方案设计的参考。     

The first DC performance test and analysis of CC conductor short sample at ASIPP conductor test facility

The first DC performance experiments of ITER correction coil (CC) conductor short sample have been carried out in the conductor test facility of Institute of Plasma Physics, CAS (ASIPP) in January this year. Those experiments aim to investigate the DC performance of ITER CC conductor. The tested conductor short sample is bended as a half circle with the diameter of 270 mm to meet the background magnetic field shape. The half circle part of sample is longer than the final twist pitch. The current sharing temperature (T_cs) in the 3.86 T external magnetic field (B_ex), ≤12 kA could be measured including the critical current (I_c) run. There is no obvious impact of 1000 cycles on DC performance. Those measured T_cs results are in agreement with the expected results from strand scaling.     

R&D of High Current Balance Reactors

The design of high current balance reactors used in the ITER DC testing platform is presented,which is verified by simulations with finite element method software,and the reactors are fabricated and tested according to the design output.These reactors are chosen as multilayer multi-turn structure and cooled by water.The multilayer multi-turn structure is usually selected by some high voltage reactors,but is seldom used in high current situations.The analysis and testing results indicate that the multilayer multi-turn structure is also feasible for high current reactors with many advantages,and is of considerable significance to similar applications.     

Proposal of Integrated Test Facility for in-Vessel Thermofluid Safety of Fusion Reactors

A vacuum vessel (VV) of a tokamak fusion reactor like the International Thermonuclear Experimental Reactor (ITER) consists the first confinement barrier that includes the largest amount of radioactive materials such as tritium and activation products. The ingress of coolant event (ICE) is a design basis event in the ITER where water is used as the coolant. The loss of vacuum event (LOVA) is also considered as an independent design basis event. Based on the results of ICE and LOVA preliminary experiments, an integrated in-vessel thermofluid test is being planned and conceptual design of the facility is in progress. The main objectives of the integrated test are to investigate the consequences of possible interaction of the ICE and the LOVA and to validate the analytical model of thermofluid events in the VV of the fusion reactor. This paper introduces a conceptual design of the integrated test facility and a testing plan.     

Study on the Flexible Support Legs for ITER Transfer Cask System

In fields of remote handing i.e. robot technology for fusion engineering reactor, such as ITER or the China fusion engineering test reactor, the flexible support legs are key components for their transfer cask system to adjust its position, joining to hot cell or tokamak ports for maintaining the fusion device. For ITER machine, each support leg should withstand maximum 50 tons load and adjust its height in 150 mm. Defect in original ITER design was presented. A new concept for the support legs was configured and its feasibility was proven. Detailed design and simulation was done for the new support leg with virtual prototype technology. Simulation results show that new support leg could not only meet all required function but also has merits of constant load during the tuning process with linear relation of control variable parameters, which is intended to be used for Tokamak reactors.     

Preliminary design of high power magnet converter for CRAFT

In order to test the superconducting magnet performance for the Comprehensive Research Facility for Fusion Technology(CRAFT) project of China, a power supply converter needs to be designed. In this paper, short circuits upstream and downstream of the direct current(DC) reactor are analyzed, and the thyristor style and the parallel number are determined by the limit analysis of junction temperature and fault current I~2t. On this basis, the over current and voltage verification of fast fuse are finished to protect the thyristor at fault cases by considering the short circuit of the bridge arm. Then, the resistor and capacitor parameters of thyristor snubber are committed to decreasing the reverse over voltage. These analysis results will be used as the preliminary design of high power magnet converter for CRAFT.     

A Polywell Fusion Reactor Designed for Net Power Generation

A brief history of Polywell progress is recounted. The present PIC simulation explains why the most recent Polywell fusion reactor failed to produce fusion energy. Synchronized variations of multiple parameters would require DC power supplies, not available in historic model testing. Even with DC power, the simulation showed that the trapping of cold electrons would ruin plasma stability during start-up. A theoretical solution to this trapping problem was found in Russian literature describing diocotron-pumping of electrons out of a plasma trap at Kharkov Institute. In Polywell, diocotron-pumping required matching the depth of the potential-well to the electron-beam current falling on a special aperture installed in one of the electromagnets. With diocotron-pumping the reactor was simulated to reach steady-state, maximum-power operation in a few milliseconds of simulated time. These improvements, validated in simulating small-scale DD reactors, were scaled up by a factor of 30 to simulate a large, net-power reactor burning p + ¹¹B fuel. Power-balance was estimated from a textbook formula for fusion power density by numerically integrating the power density. Unity power-balance required the size of the p + ¹¹B reactor to be somewhat larger than ITER.     

Dynamic physical characteristics of DC arc on arcing horn for HVDC grounding electrode line

The dynamic physical characteristics of a DC arc on an arcing horn for a high voltage direct current(HVDC) grounding electrode line are significantly different from those of the switching device arc,secondary arc,AC fault arc and pantograph-catenary arc.In this work,an experimental platform for the DC arc on the arcing horn was built,and mechanisms of the arc column short circuit and arc root movement were studied.This work further analyzes the characteristics and mechanisms of the arc motion wh...     

Transient electrical behaviour of the ITER TF coils during fast discharge and two fault cases

Improvements of high voltage design criteria and quality assurance for ITER coils are indispensable taking into account the problems occurred during high voltage tests of the ITER TF model coil. One important aspect to consider is the transient electrical behaviour because fast changes of voltages may cause local overloading and destruction of the insulation system. This paper will present the calculation of the terminal voltages within the ITER TF coil system and the voltage stress of the insulation within an individual ITER TF coil for the fast discharge and two fault cases. Proposals for the high voltage tests are discussed based on the calculated voltage stress of the two fault cases and the experiences gained during the ITER TF Model Coil test to ensure appropriate dielectric quality of the ITER TF coils.     

On monitoring the tritium breeder in a lead–lithium cooled ceramic breeder (LLCB) module of the ITER

The operation of a tritium breeder is a most process among engineering problems of DEMO. In this study, a design for monitoring tritium-breeding in the reactor is discussed. Additionally, a system for the experimental estimation of the tritium-breeding ratio (TBR) and the tritium-breeding dynamics in a lead–lithium cooled ceramic breeder (LLCB) test module used in the ITER is proposed. The systems are based on tritium and neutron-flux measurements under the ITER plasma D–T experiments and the use of lithium ortho-silicate and lithium carbonate samples and neutron detectors. Different lithum-6 and lithium-7 isotope contents in the samples are used to measure neutron spectrum. The samples and detectors are delivered in containers to the test breeder module (TBM) on a monitor channel connecting the TBM to an operating zone of the ITER. The tritium content in the samples is measured in a laboratory by the liquid scintillation method.Pneumatic control is used to deliver the samples to the TBM and to extract the samples using the channel during plasma-operational pauses. Neutron calculation is performed to estimate the tritium content in the samples and the heat distribution in the materials of the channel under reactor irradiation. A measurement accuracy of the tritium content in the carbonate and orthosilicate samples can attain a level of 7% and 10%, respectively. The results of the channel-cooling calculation performed under the nominal operating conditions of the TBM (a plasma pulse) are presented in the paper.     

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.