Innovative Fusion Tokamak Powerplant and the Basic Engineering for Mhd Theory

All around the world an endeavour to develop the fusion process as a major alternative energy has been going on for about a half century. Aries-St is the spherical tokamak (St) a innovative fusion reactor engineering. This toroidal reactor is a type of system that facilitates the occurrence of the nuclear fusion and fission events together (Tillack et al. in Fusion Energ Des 65:215–261, 2003; El-Guebaly in Fusion Energ Des 65:263–284, 2003). The Aries-St power core consist of the components directly surrounding the burning plasma and serves important functions. In fusion applications, liquid metals are traditionally considered to be the best working fluids. Sufficient tritium breed amount must be TBR >1.1 for Aries-St fusion tokamak power plant (Tillack et al. in Fusion Energ Des 65:215–261, 2003; El-Guebaly in Fusion Energ Des 65:263–284, 2003). The Aries-St power core has designed for correlation with an optimized St plasma that develop through the investigation of extensive range of plasma magnetohydrodynamic (Mhd) equations. In this study, the engineering design plasma parameters are described with respect to Mhd equilibrium and nuclear analysis, stability, radiation heat transfer conditions, current drive, and safety. In addition, turbulence model extended to an incompressible Mhd flows and monte carlo simulation are used for modeling of low-conductivity fluid. In this study the modeling of aries-st tokamak reactor produced by using aries design technology, has performed by using the monte carlo code and Endf/b-V-VI nuclear data. Monte carlo method is the general name for the solution of experimental and statistical problems with a random approach.     

3D Neutronic Analysis in MHD Calculations at ARIES-ST Fusion Reactors Systems

In this study, we developed new models for liquid wall (FW) state at ARIES-ST fusion reactor systems. ARIES-ST is a 1,000 MWe fusion reactor system based on a low aspect ratio ST plasma. In this article, we analyzed the characteristic properties of magnetohydrodynamics (MHD) and heat transfer conditions by using Monte-Carlo simulation methods (ARIES Team et al. in Fusion Eng Des 49–50:689–695, 2000; Tillack et al. in Fusion Eng Des 65:215–261, 2003) . In fusion applications, liquid metals are traditionally considered to be the best working fluids. The working liquid must be a lithium-containing medium in order to provide adequate tritium that the plasma is self-sustained and that the fusion is a renewable energy source. As for Flibe free surface flows, the MHD effects caused by interaction with the mean flow is negligible, while a fairly uniform flow of thick can be maintained throughout the reactor based on 3-D MHD calculations. In this study, neutronic parameters, that is to say, energy multiplication factor radiation, heat flux and fissile fuel breeding were researched for fusion reactor with various thorium and uranium molten salts. Sufficient tritium amount is needed for the reactor to work itself. In the tritium breeding ratio (TBR) >1.05 ARIES-ST fusion model TBR is >1.1 so that tritium self-sufficiency is maintained for DT fusion systems (Starke et al. in Fusion Energ Des 84:1794–1798, 2009; Najmabadi et al. in Fusion Energ Des 80:3–23, 2006).     

Thermodynamics of Sc-H and Sc-D Systems: Experimental and Theoretical Studies

The pressure composition isotherms (p-c-T) of Sc-H and Sc-D systems have been experimentally measured using the PVT method in the same work. The enthalpies and entropies are extracted from van’t Hoff plots and compared with the literature values. The results show that the enthalpy and entropy for hydrogen absorption are in good agreement with the data reported by Manchester et al. (J Phase Equilib 18(2):194–205, 1997), while those for deuterium absorption are in variance with the data reported by Wu et al. (J Fusion Energ, 2012). First principles calculations further prove that the thermodynamic data of Sc-D system reported by us are more reasonable.     

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.     

Second Comment on: ‘‘Characterization of Micro-roughness Parameters in Titanium Nitride Thin Films Grown by DC Magnetron Sputtering’’ [J Fusion Energ DOI 10.1007/s10894-012-9534-4]

At recent comment written by [A. Gelali, A. Shafiekhani, A. Ghorbani, A. Ahmadpourian. J. Fusion Energ. DOI: 10.1007/s10894-012-9542-4] that is a comment on [S. Solaymani, A. Ghaderi, N. B. Nezafat. J. Fusion Energ. DOI: 10.1007/s10894-012-9534-4] some clear and unavoidable errors can be seen. These issues will be discussed by the present brief communication.     

Comment on: “Characterization of Microroughness Parameters in Titanium Nitride Thin Films Grown by DC Magnetron Sputtering” [J Fusion Energ DOI 10.1007/s10894-012-9510-z]

In recent article [Ali Gelali. Azin Ahmadpourian. Reza Bavadi. M. R. Hantehzadeh. Arman Ahmadpourian. J Fusion Energ DOI 10.1007/s10894-012-9510-z], Ali Geleli et al. studied the PSD and RMS Roughness parameters in Titanium Nitride thin films by AFM data and used the computed fractal dimension value of micrographs to describe the surface morphology of thin films. Here, the correct form of equations and relationship between PSD and RMS will be discussed.     

(p,α) Reaction Cross Sections Calculations of Fe and Ni Target Nuclei Using New Developed Semi-empirical Formula

Iron (Fe) and nickel (Ni) are important fusion structural materials in reactor technology. The gas production in the metallic structure arising from many different types of nuclear reactions has been a significant damage mechanism in structural components of fusion reactors. The hydrogen and its isotopes at high temperatures leave out of the metallic lattice but the alpha (α) particles that remain in the lattice generate helium (He) gas bubbles. In other words, the α particles can cause serious changes in the physical and mechanical properties of the fusion structural materials. In this study, the excitation functions of ^54,57Fe(p,α) and ^58,60,61,64Ni(p,α) reactions have been investigated in the incident proton energy range of 10–40 MeV to estimate the radiation damage effects on fusion structural materials used in the construction of the first walls and core of the reactor. The calculations of (p,α) reaction cross sections on ^54,57Fe and ^58,60,61,64Ni have been made by using PCROSS code and CEM95 code. The full exciton and cascade exciton model (CEM95) for pre-equilibrium calculations and Weisskopf-Ewing model for equilibrium calculations are used. Besides, the semi-empirical cross section formula with new coefficient obtained by Tel et al. (Pramana J Phys 74:931–943, 2010) has been applied for (p,α) reactions at 17.9 MeV proton incident energy.     

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.     

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.     

Soft X-ray Emission Optimization Studies with Krypton and Xenon Gases in Plasma Focus Using Lee Model

The X-ray emission properties of krypton and xenon plasmas are numerically investigated using corona plasma equilibrium model. Numerical experiments have been investigated on various low energy plasma focus devices with Kr and Xe filling gases using Lee model. The Lee model was applied to characterize and to find the optimum combination of soft X-ray yields (Y_sxr) for krypton (~4 Å) and xenon (~3 Å) plasma focus. These combinations give Y_sxr = 0.018 J for krypton, and Y_sxr = 0.5 J for xenon. Scaling laws on Kr and Xe soft X-ray yields, in terms of storage energies E₀, peak discharge current I_peak and focus pinch current I_pinch were found over the range from 2.8 to 900 kJ. Soft X-ray yields scaling laws in terms of storage energies were found to be as $ {\text{Y}}_{{{\text{sxr}},{\text{Kr}}}} = 0.0003 \times {\text{E}}_{0}^{1.43} $ Y sxr , Kr = 0.0003 × E 0 1.43 and $ {\text{Y}}_{{{\text{sxr}},{\text{Xe}}}} = 0.0064 \times {\text{E}}_{0}^{1.41} $ Y sxr , Xe = 0.0064 × E 0 1.41 for Kr and Xe, respectively, (E₀ in kJ and Y_sxr in J) with the scaling showing gradual deterioration as E₀ rises over the range. The maximum soft X-ray yields are found to be about 0.5 and 27 J from krypton and xenon, respectively, for storage energy of 900 kJ. The optimum efficiencies for soft X-ray yields (0.0002 % for Kr) and (0.0047 % for Xe) are with capacitor bank energies of 67.5 and 225 kJ, respectively.     

Calculation of Internal Inductance and Poloidal Beta Using Solovev’s Assumption in a Circular Cross Section Tokamak

We calculated the internal inductance and the poloidal beta in a circular cross section Tokamak using Solov’ev assumption in the solution of Grad–Shafranov equation (GSE). GSE is solved by considering linear source functions and fixed boundary conditions. This solution has the three quantities (plasma current I _p , plasma minor radius r _p and $ \beta_{p} + l_{i} /2 $ ) that they are as input data. In two different discharges on IR-T1 Tokamak with this solution, we have shown that the internal inductance at a low value is required to extend the duration of Tokamak plasma discharge that it is consistence with other method.     

Scaling Laws of Nitrogen Soft X-ray Yields from 1 to 200 kJ Plasma Focus

Numerical experiments are carried out systematically to determine the nitrogen soft X-ray yield for optimized nitrogen plasma focus with storage energy E₀ from 1 to 200 kJ. Scaling laws on nitrogen soft X-ray yield, in terms of storage energies E₀, peak discharge current I_peak and focus pinch current I_pinch were found. It was found that the nitrogen X-ray yields scales on average with $ {\text{Y}}_{\text{sxr,N}} = 1.93 \times {\text{E}}_{0}^{1.21} {\text{J}} $ (E₀ in kJ) with the scaling showing gradual deterioration as E₀ rises over the range. A more robust scaling is $ {\text{Y}}_{\text{sxr}} = 8 \times 10^{ - 8} {\text{I}}_{\text{pinch}}^{3.38} $ . The optimum nitrogen soft X-ray yield emitted from plasma focus is found to be about 1 kJ for storage energy of 200 kJ. This indicates that nitrogen plasma focus is a good water-window soft X-ray source when properly designed.     

Numerical Experiments on Oxygen Plasma Focus: Scaling Laws of Soft X-Ray Yields

Numerical experiments have been investigated on UNU/ICTP PFF low energy plasma focus device with oxygen filling gas. In these numerical experiments, the temperature window of 119–260 eV has been used as a suitable temperature range for generating oxygen soft X-rays. The Lee model was applied to characterize the UNU/ICTP PFF plasma focus. The optimum soft X-ray yield (Y_sxr) was found to be 0.75 J, with the corresponding efficiency of about 0.03 % at pressure of 2.36 Torr and the end axial speed was v_a = 5 cm/μs. The practical optimum combination of p₀, z₀ and ‘a’ for oxygen Y_sxr was found to be 0.69 Torr, 4.8 cm and 2.366 cm respectively, with the outer radius b = 3.2 cm. This combination gives Y_sxr ~ 5 J, with the corresponding efficiency of about 0.16 %. Thus we expect to increase the oxygen Y_sxr of UNU/ICTP PFF, without changing the capacitor bank, merely by changing the electrode configuration and operating pressure. Scaling laws on oxygen soft X-ray yield, in terms of storage energies E₀, peak discharge current I_peak and focus pinch current I_pinch were found over the range from 1 kJ to 1 MJ. It was found that the oxygen soft X-ray yields scale well with $ {\text{Y}}_{\text{sxr}} = 2 \times 10^{ - 7} {\text{I}}_{\text{pinch}}^{3.45} $ and $ {\text{Y}}_{\text{sxr}} = 6 \times 10^{ - 7} {\text{I}}_{\text{peak}}^{ 2. 9 2} $ for the low inductance (L₀ = 30 nH) (where yields are in J and currents in kA). While the soft X-ray yield scaling laws in terms of storage energies were found to be as $ {\text{Y}}_{\text{sxr,O}} = 5.354 \times {\text{E}}_{0}^{1.12} $ (E₀ in kJ and Y_sxr in J) with the scaling showing gradual deterioration as E₀ rises over the range. The oxygen soft X-ray yield emitted from plasma focus is found to be about 8.7 kJ for storage energy of 1 MJ. The optimum efficiency for soft X-ray yield (1.1 %) is with capacitor bank energy of 120 kJ. This indicates that oxygen plasma focus is a good soft X-ray source when properly designed.     

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.     

Alpha Induced Reaction Cross Section Calculations of Tantalum Nucleus

The fusion energy is attractive as an energy source because the fusion will not produce CO₂ or SO₂ and so fusion will not contribute to environmental problems, such as particulate pollution and excessive CO₂ in the atmosphere. The fusion reaction does not produce radioactive nuclides and it is not self-sustaining, as is a fission reaction when a critical mass of fissionable material is assembled. Since the fusion reaction is easily and quickly quenched the primary sources of heat to drive such an accident are heat from radioactive decay and heat from chemical reactions. Both the magnitude and time dependence of the generation of heat from radioactive decay can be controlled by proper selection and design of materials. Tantalum is one of the candidate materials for the first wall of fusion reactors and for component parts of irradiation chambers. Accurate experimental cross-section data of alpha induced reactions on Tantalum are also of great importance for thermonuclear reaction rate determinations since the models used in the study of stellar nucleosynthesis are strongly dependent on these rates (Santos et al. in J Phys G 26:301, 2000). In this study, neutron-production cross sections for target nuclei ¹⁸¹Ta have been investigated up to 100 MeV alpha energy. The excitation functions for (α, xn) reactions (x = 1, 2, 3) have been calculated by pre-equilibrium reaction mechanism. And also neutron emission spectra for ¹⁸¹Ta (α, xn) reactions at 26.8 and 45.2 MeV have been calculated. The mean free path multiplier parameters has been investigated. The pre-equilibrium results have been calculated by using the hybrid model, the geometry dependent hybrid (GDH) model. Calculation results have been also compared with the available measurements in literature.     

Design a 10 kJ IS Mather Type Plasma Focus for Solid Target Activation to Produce Short-Lived Radioisotopes 12C(d,n)13N

A 10 kJ (15 kV, 88 μF) IS (Iranian Sun) Mather type plasma focus device has been studied to determine the activity of a compound exogenous carbon solid target through ¹²C(d,n)¹³N nuclear reaction. The produced ¹³N is a short-lived radioisotope with a half-life of 9.97 min and threshold energy of 0.28 MeV. The results indicate that energetic deuterons impinging on the solid target can produce yield of $ \langle y\rangle $  = 6.7 × 10^?5 with an activity of A = 6.8 × 10⁴ Bq for one plasma focus shut and A _ν  = 4 × 10⁵ Bq for 6 shut per mint when the projectile maximum deuterons energy is E _max = 3 MeV.     

Feasibility of the electrochemical way in molten fluorides for separating thorium and lanthanides and extracting lanthanides from the solvent

An alternative way of reprocessing nuclear fuel by hydrometallurgy could be using treatment with molten salts, particularly fluoride melts. Moreover, one of the six concepts chosen for GEN IV nuclear reactors (Technology Roadmap - http://gif.inel.gov/roadmap/) is the molten salt reactor (MSR). The originality of the concept is the use of molten salts as liquid fuel and coolant. During the running of the reactor, fission products, particularly lanthanides, accumulate in the melt and have to be eliminated to optimise reactor operation. This study concerns the feasibility of the separation actinides-lanthanides-solvent by selectively electrodepositing the elements to be separated on an inert (Mo, Ta) or a reactive (Ni) cathodic substrate in molten fluoride media. The main results of this work lead to the conclusions that:

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The solvents to be used for efficient separation must be fluoride media containing lithium as cation.

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Inert substrates are suitable for actinide/lanthanide separation; nickel substrate is more suitable for the extraction of lanthanides from the solvent, owing to the depolarisation occurring in the cathodic process through alloy formation.