High-Speed Optical Signal Processing System Based on DSP for Neutral Beam Injector on EAST

Neutral Beam Injection (NBI) is one of the main plasma heating methods in nuclear fusion devices. To prevent the device from irreversible damage, all the status signal acquisition and transmission must meet high voltage isolation, fast-reaction and high reliability requirements. DSP performs well at complicated algorithms computing and meets the real-time requirements, but it is poor at parallel processing. FPGA is good at combination logic and sequential logic control. To meet the EAST-NBI control system experimental operational requirements, DSP and FPGA architecture is taken in this design, which converts the optical signal to high-precision analog signal to output. This paper analyzes the requirement of high-speed optical signal processing system based on DSP for EAST-NBI and introduces the detail modules designs, including seven modules: optical signal receiving and conversion interface circuit, FPGA decoding interface circuit, OMAP processing interface circuit, DAC interface circuit, Ethernet interface circuit, serial port interface circuit and amplification interface circuit.     

Design,Analysis and Optimization of the First Wall Cooling System of the CFETR Blanket

China Fusion Engineering Test Reactor (CFETR) is a superconducting tokamak which is designed by China National Integration design Group for Magnetic Confinement Fusion. CFETR Blanket, as a plasma-facing component withstand very high heat load, is very critical for fusion reactor operation. The first wall (FW) is one of the most significant components of the blanket. The cooling system of the FW has been designed. Meanwhile, thermal–dynamic calculations are performed to obtain the coolant feature and temperature distribution of the FW using ANSYS CFX code. Besides, thermo-mechanical coupling analysis is carried out using the temperature distribution from thermal–dynamic calculation as boundary condition. In addition, cooling channel optimization is proposed according to the analysis results. Analysis results of the optimization cooling channel indicate that the maximum temperature and thermal stress satisfy the design requirements of the FW.     

Thermodynamics Properties of Molten Salt Technology Assessment for New Generation Fusion Reactors

In this study, some important thermodynamic properties of the fusion reactor have been analyzed. The physical and chemical properties of molten salts have been extensively studied in the nuclear fusion program. In recent years, molten salts technology began to be used in some engineering areas, in the advanced nuclear field and especially in nuclear fusion reactor systems. Nowadays, Aries team has developed advanced designs by using the molten salts technology in order to get high thermodynamic and structural advantage on nuclear technology areas (Tillack et al. in Fusion Energ Des 65:215–261, 2003; Tillack et al. in Fusion Energ Des 49–50:689–695, 2000; El-Guebaly et al. in Fusion Energ Des 65:263–284, 2003). The Aries-St reactors are a 1000 MW fusion reactor system that based on a low aspect ratio ST plasma (Tillack et al. in Fusion Energy Des 65:215–261, 2003; Tillack et al. in Fusion Energy Des 49–50:689–695, 2000). The Aries team studies especially on liquid walls concepts and this liquid are used to increase neutronic performance of various structures of Aries-St reactors. In this research, candidate molten salts have been studied neutron effects on reactor performance which are the first wall (FW) and blanket. There are various candidate liquids that meet all the criteria such as Li₁₇Pb₈₃, flibe(Li₂BeF₄) and LiNaBeF₄, LiSn that are able to breed enough tritium. In this research, we used Li₁₇Pb₈₃, pure lithium and flibe as candidates that are in the Aries design. Montecarlo n-particle 4b-code is used for neutronics analysis and thermodynamic features. The value of tritium breeding ratio of the Aries-St reactors must be (TBR ≥ 1.1). This can be achieved in the region of LiPb/FW blanket of reactors. Aries-St spherical reactor has high heat flux (0.8 MW/m²) and NWL (6–8 MW/m²) in this region.     

Flow Distribution Analysis of Water-Cooled Solid Breeder Test Blanket Module

Blanket system is one of the most important systems in a fusion reactor, which plays an important role in heat removing, radiation shielding and tritium breeding. Water-cooled ceramic breeder (WCCB) blanket module (BM) is one of tritium breeding blanket module concepts for Chinese Fusion Engineering Testing Reactor. According to the preliminary design of WCCB BM, there are complicated cascade and parallel cooling channels, in which maybe exists the nonuniform distribution of flow rate, and resulting adverse effect on heat transfer and safety. In this paper, the whole model of one BM is built by thermal hydraulic analytical code named Relap 5 and the flow distribution issue of water-cooled solid breeder (WCSB) test blanket module (TBM) is analyzed. The systematic analysis results show that the flow rate differences of most parts of the WCSB TBM are less than 4 % in a steady state. Start-up, operational transient and loss of flow accident are also studied, and flow instability in these transient cases is found and needs for further analysis. Three dimensional local model of First Wall is also built by CFX, to investigate flow characteristics at partial WCSB TBM, which shows that flow distribution calculated by CFX is consistent with the results from the thermal hydraulic analytical code. Both of the results of the steady state and transient analysis show that the thermal hydraulic analytical code is appropriate in analyzing the flow distribution and transient issue from the systematic view.     

Influence of Different CH4/N2 Ratios on Structural and Mechanical Properties of a-CNx:H Film Synthesized Using Plasma Focus

Carbon nitride films were synthesized by operating the dense plasma focus device with different CH₄/N₂ admixture gas ratios and fixed 20 focus shots. The pressure and axial distance from anode tip were kept constant at 3 mbar and 8 cm respectively. Raman and X-ray photoelectron spectroscopy (XPS) techniques were used to observe the effect of CH₄/N₂ ratio on carbon nitride bonding. The XPS analysis showed that the terminating group C≡N is more dominant for the films synthesized using higher concentration of nitrogen which gives softer films. Field emission scanning electron microscopy results showed that the deposited films consist of nanoparticles and their agglomerates. The size of agglomerates increases with decreasing concentration of nitrogen in CH₄/N₂ admixture gas. Nanoindentation results showed the increase in hardness and elastic modulus values of films with decreasing concentration of nitrogen in CH₄/N₂ admixture gas. The hardness and elastic modulus values were found to be dependent on sp³ content in the film as well as the C≡N. The hardness and elastic modulus values of 10.7 and 229.8 GPa respectively were achieved for the films deposited with fixed 20 focus deposition shots and using CH₄/N₂ admixture gas ratio of 7:3.     

Data Server Application Software for Neutral Beam Injection Control System

The neutral beam injection (NBI) heating is one of the mainly auxiliary plasma heating methods for EAST. In this paper, a data server application software is presented for NBI control system. This application software is developed on Centos OS by C programming language with TCP and multithreading technology. It mainly focuses on storing and querying experimental pulse data and engineering data; experimental user’s operation records; experimental configuration information; the collected data after processing; as well as managing and releasing experimental discharge shot number and so on. By using the unified data header structure, the data server software provides universal information exchange interfaces for different service request terminals with different operating platform, such as WinCC, Labview. With the help of data server software, those problems related on the storage, querying and processing of information and data for NBI control system are well solved.     

The Development of Data Acquisition and Remote Real-Time Display System for EAST NBI

The data acquisition and remote real-time display system for the neutral beam injectors (NBI) on experimental advanced superconducting tokamak (EAST) are described in this paper. Distributed computer systems including local data acquisition (DAQ) facility, remote data server (DS), real-time display terminal are adopted with Linux and Windows operating system. Experimental signals are gathered by DAQ device at local working field. On the one hand, these gathered data will be sent to DS which runs on remote server main control layer on EAST NBI control network for saving and processing; on the other hand, these data will be sent to real-time display terminal which runs on remote monitoring layer on EAST NBI for displaying and monitoring experimental signals real-timely. Another point needs to be mentioned is that the real-time display software can call back historical data from DS for querying. The software of data acquisition and DS are programmed by C language while the real-time display software is programmed by Labview flow chart. The hardware mainly includes DAQ cards, server, industrial personal computer and others auxiliary hardware. Now the system proved to be performed well through experiments on NBI testing bed.     

Predictive Simulations of Plasma Heating and Current Drive by Neutral Beam Injection on EAST

Long pulse and high performance steady-state operation is the main scientific mission of experimental advanced superconducting tokamak (EAST). In order to achieve this objective, high-power auxiliary heating systems are essential. Radio frequency (RF) wave heating and neutral beam injection (NBI) are two principal methods. NBI is an effective method of plasma heating and current drive, and it has been used in many magnetic confinement fusion devices. Based on the plasma equilibrium of EAST (Li et al., Plasma Phys Control Fusion 55:125008, 2013) plus previous EAST experimental data used as initial conditions, the NBI module (Polevoi et al., JAERI-Data, 1997) employed in automated system for transport analysis (ASTRA) code (Pereverzev et al., IPP-Report, 2002) is applied to predict the effects of plasma heating and current drive with different neutral beam injection power levels. At certain levels of plasma densities and plasma current densities, the simulation results show that the NBI heats plasma effectively, also increases the proportions of NB current and bootstrap current among total current significantly.     

基于CSS的NBI射频离子源控制程序设计

离子源是中性束注入器(Neutral Beam Injector,NBI)的关键子系统之一,根据射频离子源的实验和运行需求,本文设计了一套基于物理实验与工业控制系统(Experimental Physics and Industrial Control System,EPICS)架构的控制程序,实现对射频离子源实验和调试过程的远程监控功能。NBI射频离子源控制程序通过集成开发平台(Control System Studio,CSS)的交互界面开发模块(Best OPI Yet,BOY)实现友好的人机交互界面,使用Jython实现界面逻辑,支持服务器/客户端和EPICS两种通信架构。程序已经上线近半年,实验表明,射频离子源控制程序具备了放电模式设置、时序幅值设置、设备状态实时监控、采集数据实时波形显示等功能,满足了射频离子源实验的远程控制和数据可视化的需求。     

Monitoring System of EAST’s Nuclear Radiation

Neutron and Gamma detectors have been used to monitor the nuclear radiation in the environment (Jianping in Nucl Electron Detect Technol 19(1), 1999; Chai et al. in Nucl Electron Detect Technol 25(1), 2005), during the experimental advanced superconducting tokamak (EAST) discharging. This paper is focus on the EAST’s nuclear radiation monitoring. Based on the full and real-time need for monitoring of radiation, the placement of radiation’s detectors around the EAST and the environment is studied. To get the radiation’s value, this paper gives the design of the monitoring system and presents the system’s acquisition module which transforms radiation to voltage signal. For the long distance and uneven distribution of each detector, transforming and processing module is designed, and the comparator’s principle and RS-485 transmission protocol are reviewed, the circuits of the comparator and RS-485 used in this system are designed. Then a conversion module is presented to have communication with personal computer, and framework of the whole monitoring system is introduced.     

Neutronic design of a sub-critical, source driven reactor at the RA-8 facility

We present a LEU-ADS design based on an existing Argentine experimental facility, the RA-8 pool type zero power reactor. The versatility of this reactor allows measurement of different core configurations using different fuel enrichment, burnable poison rods, water perturbations, different control rods types in critical or subcritical configurations with an external source.To assess the feasibility of the LEU-ADS, multiplication factors, kinetic parameters, spectra, and time flux evolution were computed. Two external sources were considered: an isotopic source, and a D-D pulsed neutron source.Parameters for different core configurations were calculated, and the feasibility of using continuous and pulsed neutron sources was verified.     

Development of a Helicon Plasma Source for Neutral Beam Injection System of the Alborz Tokamak

Amirkabir Helicon plasma source (AHPS) has been designed and developed at the fusion laboratory of Amirkabir University of Technology for the neutral beam injection (NBI) system of the Alborz tokamak. The design parameters of the Helicon plasma source are determined to get the \( n_{e} \cong 10^{19} \,{\text{m}}^{ - 3} \) plasma density. However, the Nagoya type III antenna was used to excite the Argon plasma by the applied magneto-static field of 750 G and RF power up to 2 kW at the frequency of 13.56 MHz. Preliminary experiments are performed to confirm the Helicon mode operation of the AHPS. Mode changing from the ICP to the Helicon mode was proved by using the optical emission spectroscopy and polaroid photography technique of the plasma “blue core”. Spectroscopy measurements results show that the emission line intensity of the \( {\text{Ar}}\;{\text{II}} \) at 434.8 nm increases as the RF power increases in the range of values of 300–1000 W at the neutral background pressure of 0.7 mTorr. Also, we found that, in Helicon mode, the increase of the magneto-static field from 350 to 750 G not only increases the plasma density besides the efficiency of the RF power absorption increases.     

Edge density profile variation during a magnetic axis shift in the CHS heliotron/torsatron

Edge density profiles were measured during natural and forced magnetic axis shifts in the lowaspect-ratio heliotron/torsatron CHS, using an 8 keV fast neutral lithium beam probe. The Shafranov shift of the magnetic axis (and hence the dislocation of the LCFS) that was negligible for a low density ECH discharge became substantial (4.5 cm) for a high density NBI discharge (n _e2.5× 10¹³cm^–3) and the corresponding radius of the LCFS increased about 1.2 cm in major radius compared to the vacuum case. For NBI discharges with different settings of the vacuum magnetic axisR _ax (fixed during the discharge), the measured edge density profiles indicated reasonable agreement between the theoretically and experimentally obtained LCFS radii for 90 cm <R _ax<101.6cm, while forR _ax<90 cm the measured radius was 10% larger than expected. When a change ofR _ax from 94.7 cm to 89.9 cm during a discharge was imposed externally, a well behaved plasma boundary moved inward smoothly by about 7.5 cm, while the steepness of the edge density profile changed for different values ofR _ax. The steepest profile was attained forR _ax=92.1 cm when the highest energy content and average density were achieved.     

Atomic data and spectral line intensities for Ca XVII

Electron impact collision strengths, energy levels, oscillator strengths, and spontaneous radiative decay rates are calculated for Ca XVII. The configurations used are 2s², 2s2p, 2p², 2l3l^′, 2l4l^′ and 2s5l^′, with l=s,p, and l^′=s,p,d giving rise to 92 fine-structure levels in intermediate coupling. Collision strengths are calculated at seven incident energies (15, 30, 75, 112.5, 150, 187.5 and 225 Ry) for the transitions within the three lowest configurations corresponding to the 10 lowest energy levels, and at five incident energies (75, 112.5, 150, 187.5 and 225 Ry) for transitions between the lowest five levels and the n=3,4,5 configurations. Calculations have been carried out using the distorted wave approximation. Excitation rate coefficients are calculated as a function of electron temperature by assuming a Maxwellian electron velocity distribution. Using the excitation rate coefficients and the radiative transition rates of the present work, and R-Matrix results for the 2s², 2s2p, 2p² configurations available in the literature, statistical equilibrium equations for level populations are solved at electron densities covering the range of at an electron temperature of logT_e(K)=6.7, corresponding to the maximum abundance of Ca XVII. Spectral line intensities are calculated, and their diagnostic relevance is discussed. This dataset will be made available in the next version of the CHIANTI database.     

For a better estimation of gamma heating in nuclear material-testing reactors and associated devices: status and work plan from calculation methods to nuclear data

The determination of gamma heating levels in material-testing reactors (MTRs) is of crucial importance as gamma heating affects both safety and performance parameters of MTRs [1 Amharrak H, Di Salvo J, Lyoussi A, Carette M, Reynard-Carette C. State of the art on nuclear heating in a mixed (n/gamma) field in research reactors. Nucl Instrum Methods A. 2014;749:57–67.[Crossref], [Web of Science ®] , [Google Scholar],2 Brun J, Reynard-Carette C, Lyoussi A, Merroun O, Carette M, Janulyte A, Zerega Y, Andre J, Bignan G, Chauvin J-P, Fourmentel D, Gonnier C, Guimbal P, Malo J-Y, Villard J-F. Numerical and experimental calibration of calorimetric sample cell dedicated to nuclear heating measurements. IEEE Trans Nucl Sci. 2012;59:3173–3179.[Crossref], [Web of Science ®] , [Google Scholar]]. The required accuracy (5% at one standard deviation) makes it necessary to calibrate bias and uncertainty associated with MTR gamma-heating calculations. Main steps of bias determination for gamma-heating calculations include, first, the development of a calculation methodology with the controlled use of physical approximations; second, the interpretation of gamma-heating measurements with reference calculations so as to determine bias supposed to be mainly due to nuclear data.     

Modeling of divertor power footprint widths on EAST by SOLPS5.0/B2.5-Eirene

The edge plasma code package SOLPS5.0 is employed to simulate the divertor power footprint widths of the experimental advanced superconducting tokamak(EAST)L-mode and ELM-free H-mode plasmas.The divertor power footprint widths,which consist of the scrape-off layer(SOL)widthλ_q and heat spreading 5,are important physical parameters for edge plasmas.In this work,a plasma current scan is implemented in the simulation to obtain the dependence of the divertor power footprint width on the plasma current I_p.Strong inverse scaling of the SOL width with I_p has been achieved for both L-mode and H-mode plasmas in the forms ofλ_(q,L-mode)=4.98×I_p~(-0.68)andλ_(q,H-mode)=1.86×I_p~(-1.08).Similar trends have also been demonstrated in the study of heat spreading with S_(L-mode)=1.95×I_p~(-0.542)and S_(H-mode)=0.756×I_p~(-0.872).In addition,studies on divertor peak heat load and the magnetic flux expansion factor show that both of them are proportional to plasma current.The simulation work here can act as a way to explore the power footprint widths of future tokamak fusion devices such as ITER and the China Fusion Engineering Test Reactor(CFETR).     

Determination of the optimum yield of enriched ore in radiometric enrichment of uranium ores

In this article a method is presented for determining the optimum conditions of operation of radiometric ore-sorting plants with allowance for the expenditure on geological prospecting S₁, ore extraction S₂, radiometric enrichment S₃, hydrometallurgical reduction S₄; also taken into account are the ore enrichment yield γ, the uranium content in the ore α, the coefficients of uranium recovery in the radiometric enrichment of the ore ε and hydrometallurgical reduction ε '. In order to determine the minimum cost of uranium salts S_m, an analytical method is used; S_m is represented in the form of a continuous function

$$S_m = F(S_1 ,S_2 ,S_3 ,S_4 ,\alpha ,\gamma , \in \in ').$$