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.     

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.     

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.     

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.     

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.     

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.     

(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.     

Adiabatic Compression of a Doublet Field Reversed Configuration (FRC)

The adiabatic compression of magnetized plasmas has come to the fore in recent times as an interesting hybrid of both inertial and magnetic fusion energy schemes, possibly allowing a means to reach fusion conditions in a compact pulsed system (R.P. Drake et al. Fusion Tech. 30, 310, (1996)). It is possible to compress a range of different magnetic configurations (D.D. Ryutov, R.E. Siemon, Com. Mod. Phys. 2, 185, (2001)), here we consider the compression of a FRC: a favorable target due to high ß. The literature relating to the adiabatic compression of magnetic concepts is reviewed. We present analytic modeling and MHD simulations of the reconnection and compression of a doublet FRC configuration that might serve as a target for compression.     

Proposal of an MFE-IFE cooperative system for a steady-state tritium balance

A new concept of a fusion reactor system, MFE-IFE cooperative system, is proposed. This concept combines the merits of a small-size MFE reactor and a dry-wall IFE reactor and aims at sufficient amount of tritium production and electricity generation without advanced technology. Design window analysis shows a NIF-scale (5 m chamber radius) dry-wall laser fusion reactor with a ～1 GWth fusion output and net tritium breeding ratio (TBR) of 1.74 can sustain an MFE power plant with a fusion power of 3 GWth and net TBR of 0.96. Although more detailed quantitative analyses are required, this concept can be a possible solution for a simultaneous achievement of tritium self-sufficiency and significant net electricity generation.     

Impact Analysis of Helium Cooled Solid Blanket Structures on the Tritium Breeding Performance in CFETR

Chinese Fusion Engineering Test Reactor (CFETR) is a test tokamak reactor to bridge the gap between ITER and future fusion power plant. As its objectives are to demonstrate generation of fusion power and to realize tritium self-sufficiency, the tritium breeding ratio (TBR) is a key design parameter. In the blanket design and optimization, the structures such as the first wall (FW), cooling plate (CP), stiffening plate (SP), cap and some other design parameters in detailed 3-D model have significant impacts on the tritium breeding performance. Based on a helium cooled solid breeder blanket option for CFETR, the impact analysis of the helium cooled solid blanket structures on tritium breeding performance was performed in this paper. Firstly, the detailed 3D neutronics model was built by using of a CAD to Monte Carlo Geometry conversion tool McCad. Then based on the detailed 3D neutronics model, the impact analyses of the blanket structures on tritium breeding performance were carried out, which include the FW, CP, SP, cap and side wall. By the sensitivity study of the blanket structures on the TBR, it gave the TBR variation trend and references for the blanket design and optimization.     

On the occurrence of burnout downstream of a flow obstacle in boiling two-phase upward flow within a vertical annular channel

If a flow obstacle, such as a spacer is placed in a boiling two-phase flow within a channel, the temperature on the surface of the heating tube is severely affected by the existence of the spacer. Under certain conditions, a spacer has a cooling effect, and under other conditions, the spacer causes dryout of the cooling water film on the heating surface. The burnout mechanism, which always occurs upstream of a spacer, however, remains unclear.In a previous paper [Fukano, T., Mori, S., Akamatsu, S., Baba, A., 2002. Relation between temperature fluctuation of a heating surface and generation of drypatch caused by a cylindrical spacer in a vertical boiling two-phase upward flow in a narrow annular channel. Nucl. Eng. Des. 217, 81–90], we reported that the disturbance wave has a significant effect on dryout and burnout occurrence and that a spacer greatly affects the behavior of the liquid film downstream of the spacer.In the present study, we examined in detail the influences of a spacer on the heat transfer and film thickness characteristics downstream of the spacer by considering the result in steam–water and air–water systems. The main results are summarized as follows:

(1) The spacer averages the liquid film in the disturbance wave flow. As a result, dryout tends not to occur downstream of the spacer. This means that large temperature increases do not occur there. However, traces of disturbance waves remain, even if the disturbance waves are averaged by the spacer.

(2) There is a high probability that the location at which burnout occurs is upstream of the downstream spacer, irrespective of the spacer spacing.

(3) The newly proposed burnout occurrence model can explain the phenomena that burnout does occur upstream of the downstream spacer, even if the liquid film thickness t_F m is approximately the same before and behind the spacer.

The JRC experimental program

JRC-ETHEL has chosen as the principle objective of its research program the improvement of protection measures in facilities handling large amounts of tritium. Technically, this involves investigating and assessing tritium propagation modes and transfer pathways in materials, components, equipment, and process plants. The experimental research work to be performed in ETHEL will basically aim at investigating:

?Loss mechanisms by identifying physico-chemical parameters such as adsorption/desorption rates, permeation rates, leakages of materials for fusion reactors and the effects of potential remedies like permeation barriers under process-like conditions.

?Multiple containment systems and fluid clean-up concepts under normal and accidental conditions.

?Methods for solid waste handling, treatment, conditioning, and final disposal.

?Techniques for tritium control, monitoring, and surveillance over the whole concentration range during both normal and accidental conditions and maintenance activities.

With the availability of two “climate chambers,” the small and large caissons of 5 and 350 m³ volume respectively, ETHEL is especially suited for benchmark and scale-up tests of many kinds of large gas volumes treatment system. This will help to close the gap between laboratory-scale results and plant-size design specifications and represents an important source of information for designers (NET, ITER) and regulatory authorities.     

Preliminary analysis of compatibility with low activation martensitic steel blanket for the fusion experimental breeder: Impact of ferromagnetic behavior on plasma burning cycle

The reduced activation ferritic martensitic steels is considered a candidate for the first wall (FW) blanket structural material because of its safety environmental advantages [R.L. Klueh, D.S. Geiles, et al., Ferritic/martensitic steels overview of recent results, J. Nucl. Mater. 307-312 (2002) 455-465; T. Muroga, M. Gasparotto, S.J. Zinkle, Overview of materials research for fusion reactors, Fusion Eng. Des. 61-62 (2002) 3-25]. An engineering design analysis concerning the electromagnetic issues is performed. Preliminary analysis results show that design effort of the fusion reactor can cope with the effect of the ferromagnetic FW blanket on the electromagnetic forces, which increases by 28-38% during a major plasma disruption and overcome the influence of the poloidal field, which reduces by 10-20%, comparing with the austenitic steel blanket. Both the effect and influence depend on the saturation magnetic susceptibility and blanket configurations.     

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.     

Effect of Internal Breeding of Tritium and Helium-3 on the Ignition of an ICF Fuel Pellet

Self-heating condition and following ignition in an Inertial Confinement Fusion (ICF) fuel pellet is evaluated by calculating the power equations, dynamically. In fact, the self-heating condition is a criterion that determines the minimum parameters of a fuel (such as temperature, density and areal density) that can be ignited. Deuterium is the main component of ICF fuels as large amounts of it are naturally available. In addition, the use of deuterium as a fuel in ICF causes the production of tritium and helium-3. However, pure deuterium has a high ignition temperature (\(\hbox {T}\ge 40\,\hbox {keV}\)) which makes it inefficient. In this paper, the power equations are solved, dynamically, and it has been indicated that internal tritium and helium-3 production at early evolution of compressed deuterium fuel causes ignition at lower predicted temperatures.     

Analysis of Time-Dependent Tritium Breeding Capability of Water Cooled Ceramic Breeder Blanket for CFETR

Attaining tritium self-sufficiency is an important mission for the Chinese Fusion Engineering Testing Reactor(CFETR) operating on a Deuterium-Tritium(D-T) fuel cycle. It is necessary to study the tritium breeding ratio(TBR) and breeding tritium inventory variation with operation time so as to provide an accurate data for dynamic modeling and analysis of the tritium fuel cycle. A water cooled ceramic breeder(WCCB) blanket is one candidate of blanket concepts for the CFETR. Based on the detailed 3D neutronics model of CFETR with the WCCB blanket,the time-dependent TBR and tritium surplus were evaluated by a coupling calculation of the Monte Carlo N-Particle Transport Code(MCNP) and the fusion activation code FISPACT-2007.The results indicated that the TBR and tritium surplus of the WCCB blanket were a function of operation time and fusion power due to the Li consumption in breeder and material activation.In addition, by comparison with the results calculated by using the 3D neutronics model and employing the transfer factor constant from 1D to 3D, it is noted that 1D analysis leads to an over-estimation for the time-dependent tritium breeding capability when fusion power is larger than 1000 MW.     

Power and Sizes of Tokamak Fusion Neutron Sources with NBI-Enhanced Reaction Rate

Steady state tokamak with deuterium–tritium plasma is considered as a basis for fusion neutron source for a hybrid fusion–fission reactor. Prototypes of such a system can be developed on the basis of the present day tokamaks as the plasma power gain factor Q ~ 1 is required for hybrid applications. Significant population of fast ions can be supported by a powerful neutral beam injection heating in regimes with Q ~ 1. The reaction rate for fast ions greatly exceeds the rate for thermal Maxwellian ions. The possible ranges of parameters are discussed for medium size tokamaks with minor plasma radius a = 0.5–1 m. Power and sizes of the neutron source are determined by the value of the injection energy. Power gain Q ≈ 1 can be achieved with injection energy of deuterium about 130 keV and tritium energy about 200 keV. Neutron power of 30–40 MW can be realized with a ≈ 1 m, and about of few megawatts with a ≈ 0.5 m.